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The query [[Category:ENS Notification]] [[Site.Company::Tennessee Valley Authority]] [[Scram::+]] was answered by the SMWSQLStore3 in 1.8478 seconds.


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 Entered dateSiteScramRegionReactor typeEvent description
ENS 542521 September 2019 00:09:00Watts BarManual ScramNRC Region 2While operating at 100 percent power, the Watts Bar Unit 1 reactor was manually tripped at 2055 EDT on August 31, 2019 due to loss of steam generator #2 level control. The trip was not complex. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems, including Auxiliary Feedwater, actuated as designed. Operations responded and stabilized the plant. Decay heat is being removed by the Auxiliary Feedwater and the Steam Dump System. Unit 2 is not affected. The cause of the loss of steam generator water level control is being investigated. The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The automatic actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was notified."
ENS 5407722 May 2019 05:45:00Watts BarManual ScramNRC Region 2On May 22, 2019, at 0233 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 reactor was manually tripped due to a failure of the #2 Main Feedwater Regulating Valve during power ascension following a refueling outage. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Dumps. Unit 2 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10 CFR 50.72(b)(3)(iv)(A) and 10 CFR 50.72(b)(2)(iv)(B). There was no impact to WBN Unit 1. The NRC Senior Resident has been notified."
ENS 5399914 April 2019 06:44:00SequoyahAutomatic ScramNRC Region 2

EN Revision Text: AUTOMATIC REACTOR TRIP DUE TO MAIN FEEDWATER PUMP TRIP At 0320 EDT, April 14, 2019, Sequoyah Unit 1 experienced an automatic reactor trip. The event was initiated by the trip of the 1A main feedwater pump. During the automatic unit runback, an automatic reactor trip was initiated due to low-low level in Steam Generator number 3. The Auxiliary Feedwater System (AFWS) automatically actuated as required when the expected post-trip feedwater isolation actuated. Reactor Coolant System temperature is being maintained by the AFWS and the steam dump system. During this operational cycle, one control Rod Position Indicator (RPI) for core position E-5 in shutdown bank 'A' has been inoperable, and the appropriate Condition and Required Actions of (Technical Specification Limiting Condition of Operation) 3.1.7 were complied with. Due to this inoperable RPI, the associated shutdown rod is conservatively assumed to be full out and untrippable. Consequently, boration was required to establish adequate shutdown margin. All other Control and Shutdown rods fully inserted. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the reactor trip. The unit is currently stable in Mode 3. Unit 1 is in a normal shutdown electrical alignment. There was no impact on Unit 2. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified.

  • * * UPDATE ON 8/6/19 AT 12:20 EDT FROM KEVIN MICHAEL TO KERBY SCALES * * *

The licensee provided an update to paragraph 2. The Auxiliary Feedwater System (AFWS) automatically actuated as required when the expected post-trip feedwater isolation actuated. Reactor Coolant System temperature is being maintained by the AFWS and the steam dump system. All Control and Shutdown rods fully inserted, except E-5 which was previously identified and conservatively assumed to be in a full out position. Applicable TS actions were performed to maintain shutdown margin. All safety systems responded as designed. No primary or secondary safety valves actuated during or after the reactor trip. The unit is currently stable in Mode 3. Unit 1 is in a normal shutdown electrical alignment. Notified the R2DO (Gerald McCoy)

ENS 5392310 March 2019 04:38:00Browns FerryAutomatic ScramNRC Region 2At 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 5369727 October 2018 16:52:00Watts BarManual ScramNRC Region 2On October 27, 2018, at 1533 EDT, Watts Bar Nuclear (WBN) Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. (Main Steam Isolation Valves) MSIVs were required to be isolated due to cooldown. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Generator Atmospheric Dump Valves. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident has been notified."
ENS 5355722 August 2018 11:23:00Watts BarAutomatic ScramNRC Region 2At 0943 EDT on August 22, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The reactor automatically tripped due to a main turbine trip signal. An investigation is in progress. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event."
ENS 5346722 June 2018 12:14:00Watts BarAutomatic ScramNRC Region 2At 0841 EDT on June 22, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 95% power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The reactor automatically tripped due to a main turbine trip. The turbine trip was caused by main generator electrical trip. An investigation is in progress. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip.
ENS 5332712 April 2018 12:14:00Watts BarAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 0920 EDT on April 12, 2018, the Watts Bar Unit 2 reactor automatically tripped while operating at 100 percent power. All control and shutdown bank rods inserted properly in response to the automatic reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and Steam Dump Systems. The cause of the automatic reactor trip is being investigated. The automatic actuation of the Reactor Protection System (RPS) is being reported as a four-hour report under 10 CFR 50.72 (b)(2)(iv)(B). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight-hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. The plant is currently stable at normal operating temperature and pressure. The grid is stable and the plant is in its normal shutdown electrical lineup. Unit 1 was unaffected by the Unit 2 trip.
ENS 5326918 March 2018 16:16:00Browns FerryAutomatic ScramNRC Region 2GE-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 FerryAutomatic ScramNRC Region 2GE-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 5311211 December 2017 11:06:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopWhile operating at 97% power, the Watts Bar Unit 2 reactor was manually tripped at 0857 EST on December 11, 2017 due to multiple dropped control rods. All control and shutdown bank rods inserted properly in response to the manual reactor trip. All safety systems including Auxiliary Feedwater actuated as designed. The plant is stable with decay heat removal through Auxiliary Feedwater and the Steam Dump System. The cause of the dropped rods is being investigated. The manual actuation of the Reactor Protection System (RPS) is being reported as a four hour report under 10 CFR 50.72 (b)(2)(iv)(B). The actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified for this event. No safety or relief valves lifted during this event.
ENS 5287225 July 2017 11:07:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopOn July 25, 2017, at 0428 Eastern Daylight Time (EDT) Watts Bar Nuclear Plant (WBN) Unit 2 was in Mode 3, beginning a Reactor Startup. While in the initial phase of withdrawing the first of four Control Rod banks, the two associated group demand position indicators deviated greater than 2 steps from each other. In accordance with Technical Requirement 3.1.7, Position Indication System, Shutdown, with one or more group demand position indicators inoperable, the reactor trip breakers are to be opened immediately. Operations personnel opened the reactor trip breakers immediately by initiating a manual trip of the Reactor Protection System (RPS). The Auxiliary Feedwater system was in service and controlling Steam Generator water levels at the time of the event and did not receive any valid actuation signals. No other system actuations occurred as a result of this reactor trip and all systems operated as designed. The cause of the position indication system inoperability is currently under investigation. NRC Resident Inspector has been notified.
ENS 527324 May 2017 19:53:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopOn May 4th, 2017, at 1709 EDT, Watts Bar Nuclear Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and Steam Generator Atmospheric Dump Valves. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72(b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident (Inspector) has been notified.
ENS 527252 May 2017 22:33:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopOn May 2nd, 2017, at 1945 EDT, Watts Bar Nuclear (WBN) Plant Unit 1 reactor was manually tripped due to a failure of the #3 Reactor Coolant Pump normal feeder breaker to close during the planned power transfer to unit power following startup. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All control and shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via auxiliary feedwater and main steam dump systems. Unit 1 is in a normal shutdown electrical alignment. This reactor trip and system actuation is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). There was no effect on WBN Unit 2. The NRC Senior Resident (Inspector) has been notified.
ENS 5264829 March 2017 23:36:00Browns FerryManual ScramNRC Region 2GE-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 5262520 March 2017 10:17:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopOn March 20, 2017 at 0813 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 operations personnel manually tripped the plant from approximately 91 percent power based on lowering steam generator levels. Prior to the plant trip, the 2A Hotwell pump tripped at 0758 EDT and the 2C Condensate Booster Pump subsequently tripped at 0802 EDT. Operations personnel commenced to lower plant power after the 2A Hotwell pump trip in an attempt to maintain steam generator levels, but were unable to recover level and manually tripped the unit. All control rods fully inserted and all automatically actuated safety related equipment operated as designed. At 0905 EDT, operations personnel exited the emergency operating instructions after the plant was stabilized. The cause of the event is under investigation. This event is reportable to the NRC within four hours under 10 CFR 50.72(b)(2)(iv)(B) as a result of the actuation of the Reactor Protection System and in eight hours under 10 CFR 50.72(b)(3)(iv)(A) as a result of actuation of the Auxiliary Feedwater system. The licensee notified the NRC Resident Inspector.
ENS 5246930 December 2016 16:37:00SequoyahManual ScramNRC Region 2Westinghouse PWR 4-LoopOn 12/30/16 at 1302 EST, Unit 1 began withdrawing control bank rods for an approach to criticality following a refueling outage. At 1305 EST, operators observed that control rod H-6 did not withdraw. Operators entered the applicable Abnormal Operating Procedure. Operators tripped the reactor as required by plant procedures and entered applicable Emergency Operating Procedures. The act of manually tripping the reactor generated a valid trip signal in the plant Reactor Protection System. Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater was already supplying the Steam Generators; a Feedwater Isolation occurred due to plant conditions. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators, and decay heat removal via the steam dumps. Method of decay heat removal is Steam Generators via the steam dumps. Current reactor coolant system conditions: Temperature at 548 degrees F and stable, pressure 2235 psig and stable. All control and shutdown banks are inserted. Electrical alignment is normal, supplied by offsite power. No impact to Unit 2, Unit 2 is in Mode 1 at 100 percent power. The licensee notified the NRC Resident Inspector.
ENS 5219423 August 2016 15:30:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopOn August 23, 2016, at 1356 EDT, Watts Bar Nuclear Plant (WBN) Unit 2 reactor was manually tripped due to a loss of main feedwater. Concurrent with the reactor trip, the Auxiliary Feedwater system actuated as designed. All control and shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and main steam dump systems. Unit 2 is in a normal shutdown electrical alignment. The cause is currently under investigation. This is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72(b)(2)(iv)(B). There was no effect on WBN Unit 1. The NRC Senior Resident Inspector has been notified.
ENS 5155923 November 2015 11:52:00SequoyahManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 0820 EST on 11/23/2015, Sequoyah Unit 1 was at 100% power when operators identified the Loop #3 Main Steam Isolation Valve (MSIV) came off its full open seat. This was evidenced by no OPEN indication on the main control board, dual indication on the post accident monitoring panel, and a change in both Tavg and steam pressure. Operators were dispatched locally to the MSIV and to the battery board room to investigate if a cause could be identified for the MSIV movement. The field investigation identified no issues. The operating crew manually tripped the reactor at 0844 EST due to an increasing Tavg-Tref mismatch. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the feedwater isolation signal. Unit 1 is currently being maintained in Mode 3 at normal operating temperature and pressure, approximately 547 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. There is no indication of any primary to secondary leakage. All control rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2 as it continues through the refueling outage with the core off-loaded. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. All control rods fully inserted during the reactor trip. The atmospheric steam dumps did operate during the transient and then shut. After the trip, the MSIV re-opened.
ENS 5139214 September 2015 08:12:00SequoyahManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 0426 EDT on 9/14/2015, Sequoyah Unit 1 was at 100% power when the Vital Instrument Power Board (VIPB) 1-II deenergized. A manual reactor trip was initiated in accordance with the Abnormal Operating Procedure for the loss of VIPB 1-II. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP (Normal Operating Temperature/Normal Operating Pressure), approximately 547 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. Current Temperature and Pressure - Reactor Coolant System (RCS) temperature is 547 degrees F and stable and RCS pressure is 2235 psig and stable. There is no indication of any primary to secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100 (percent). There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector.
ENS 5125924 July 2015 17:04:00SequoyahAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 1351 EDT on 7/24/2015, Sequoyah Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F and 2235 psig, with auxiliary feedwater supplying the steam generators and decay heat removal via the condenser steam dumps. The immediate cause of the trip was an electrically-induced turbine trip. There was no associated work in progress related to this and all systems were normally aligned. There is no indication of any primary/secondary leakage. All rods fully inserted on the reactor trip and remain inserted. The electrical alignment is normal with shutdown power supplied from off-site power. The 2B-B Emergency Diesel Generator is currently in service for the performance of an unrelated surveillance. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. The cause of the main generator lockout is under investigation.
ENS 5087811 March 2015 09:30:00SequoyahAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt (0621 EDT) on 3/11/2015, Sequoyah Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying decay heat removal via the steam generators and condenser steam dumps. The immediate cause of the trip was a Power Range Nuclear Instrumentation negative rate signal, caused by a malfunction in the rod control system. There was no associated work in progress related to this and all systems were normally aligned. Current Temperature and Pressure - temperature is 547 degrees F and stable, pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from Off-Site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector.
ENS 5083921 February 2015 12:46:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopOn February 21, 2015 at 10:32 EST, Watts Bar Nuclear Plant Unit 1 reactor was manually tripped due to rapidly dropping main condenser vacuum. Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater and S/G PORVs. Main Steam Isolation Valves are closed. The Station is in a normal shutdown electrical alignment. The cause is currently under investigation. This is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). The NRC Senior Resident has been notified. This event had no affect on Unit 2 (Under Construction)
ENS 5040426 August 2014 21:24:00Browns FerryAutomatic ScramNRC Region 2GE-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 5027813 July 2014 23:13:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopOn July 13, 2014 at 1937 (EDT), Watts Bar Nuclear Plant Unit 1 reactor was manually tripped due to automatic isolation of all low pressure feedwater heaters. Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater, steam dumps and the main condenser. The station is in a normal shutdown electrical alignment. The cause is currently under investigation. This is being reported under 10CFR 50.72(b)(3)(iv)(A) and 10CFR 50.72 (b)(2)(iv)(B). The NRC Senior Resident (Inspector) has been notified. Initial indications are that a level switch failed on the #7 heater drain tank which caused the heater drain tank pumps to trip. When the high level tank setpoint was reached, the feedwater system isolated.
ENS 500906 May 2014 13:27:00Browns FerryAutomatic ScramNRC Region 2GE-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 4915428 June 2013 16:34:00Watts BarAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopOn June 28, 2013 at 1330 (EDT), Watts Bar Nuclear Plant Unit 1 reactor automatically tripped due to an electrical fault causing a main generator lockout and subsequent turbine trip. The electrical fault generated an 'A' Main Bank Transformer Differential Relay actuation. Suspected cause is due to an offsite fault. This is being reported under 10CFR50.72(b)(2)(iv)(B). Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. This is being reported under 10CFR50.72(b)(3)(iv)(A). All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3, with decay heat removal via Auxiliary Feedwater, steam dumps and the main condenser. The station is in a normal shutdown electrical alignment. The NRC Senior Resident has been notified. There were no primary or secondary relief valve actuations during the plant transient. The cause of the electrical fault is being investigated.
ENS 4882919 March 2013 08:37:00Browns FerryManual ScramNRC Region 2GE-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 FerryAutomatic ScramNRC Region 2GE-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 4877824 February 2013 15:20:00SequoyahManual ScramNRC Region 2Westinghouse PWR 4-LoopOn February 24, 2013 at 1205 (EST) with reactor power at 25% and the turbine offline, a manual reactor trip for Sequoyah Unit 2 was initiated due to loss of condenser vacuum indication causing closure of condenser steam dumps, opening of the Steam Generator Atmospheric Relief Valves, and lowering hotwell level resulting in imminent loss of hotwell pumps. The cause of the event was determined to be a faulty test connection on B Condenser vacuum pressure switch. During the event, steam pressure rose to the setpoint for the first Steam Generator code safety valve (1064 psig). (The safety valve opened, then reseated). Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater actuated as expected on loss of the operating main feedwater pumps. The reactor trip was uncomplicated. Unit 2 is currently being maintained in Mode 3 at NOP/NOT (Normal Pressure and Temperature), with auxiliary feedwater supplying the steam generators and maintaining level at approximately 33% narrow range. Method of decay heat removal is via atmospheric reliefs to the atmosphere. Current RCS conditions: temperature (is) 547 degrees F and stable. Pressure (is) 2235 psig and stable. (There is) no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal and supplied from offsite power. (There is) no impact to Unit 1. Unit 1 is operating at 100% power / Mode 1. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart is 02/25/2013. (The licensee plans a press release.) The licensee notified the NRC Resident Inspector.
ENS 4862322 December 2012 16:39:00Browns FerryAutomatic ScramNRC Region 2GE-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 4823828 August 2012 06:13:00Watts BarAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopOn August 28, 2012, Watts Bar Nuclear Plant Unit 1 reactor automatically tripped due to low level in steam generator (SG) #2. The low level resulted when the Main Feedwater Control Valve for SG#2 (1-FCV-3-48) failed closed. This is being reported under 10CFR 50.72 (b)(2) (iv) (B). Concurrent with the reactor trip the Auxiliary Feedwater system actuated as designed. This is being reported under 10CFR 50.72(b)(3) (iv)(A). All Control and Shutdown rods fully inserted. All safety systems responded as designed. The unit is currently stable in Mode 3. The NRC Senior Resident has been notified. Decay heat is being removed to the main condenser via condenser steam dumps. The plant is in its normal shutdown electrical lineup. No steam safety or relief valves lifted during the event.
ENS 4797229 May 2012 07:22:00Browns FerryAutomatic ScramNRC Region 2GE-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 FerryManual ScramNRC Region 2GE-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 FerryAutomatic ScramNRC Region 2GE-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 FerryManual ScramNRC Region 2GE-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 FerryManual ScramNRC Region 2GE-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 FerryAutomatic ScramNRC Region 2GE-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 4716919 August 2011 01:10:00SequoyahAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 2250 EDT on 8/18/2011, Unit 1 Reactor/Turbine automatically tripped on RCP (Reactor Coolant Pump) Busses UV (Under-Voltage) trip. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected from the Feedwater Isolation Signal. No primary PORVs and/or Safety Valves opened during or after this trip. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F (degrees) and 2233 psig, with Auxiliary Feedwater supplying the Steam Generators. At the time of the trip, a 50G (instantaneous overcurrent ground) relay flag was found dropped on the '1A' 6.9 KV unit board. Subsequently, the '1A' 6.9 KV start bus was found to have transferred to its alternate supply, 'B' CSST (Common Station Service Transformer). 1A condenser circulating water pump motor trip out was also received in the MCR (Main Control Room). The method of decay heat removal is via steam dumps to the condenser with MSIVs open. The current temperature and pressure is stable. There is no indications of any primary/secondary leakage. All control rods inserted. The electrical alignment is normal with the exception of the above mentioned items, supplied from off-site power. There is no impact to Unit 2. Unit 2 is operating at 100% power/ Mode 1. The licensee notified the NRC Resident Inspector.
ENS 4708121 July 2011 00:53:00SequoyahAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 2129 EST on 7/20/2011, Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected from the Feedwater Isolation Signal. Primary PORVs and/or safety valves lifted and reseated as indicated by tailpipe temperatures and PRT pressure. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F and 2235 psig, with Auxiliary Feedwater supplying the steam generators. At the time of the trip, maintenance was in progress on Preferred Inverter #1. AOP-P.09 'Loss of 120VAC Preferred Power' was used to restore power to #1 Preferred board after the trip. Method of decay heat removal is via steam dumps to the condenser with MSIVs open. Current temperature and pressure: Temperature - 548 degrees Fahrenheit and stable, Pressure 2235 - psig and stable No indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from off-site power. No impact on Unit 2. Unit 2 is operating at 100% power/Mode 1 The NRC Resident Inspector has been informed. The licensee notified the State of Tennessee.
ENS 4699126 June 2011 19:39:00SequoyahAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 1614 EDT on 6/26/2011, Unit 1 reactor automatically tripped following a turbine trip from greater than 50% rated thermal power (P-9 interlock). Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected on loss of the operating main feedwater pumps. Initially, the steam dump system functioned as expected (all valves opened). Subsequently, the steam dump system was manually turned off when 3 of the valves did not close when expected. Consequently, decay heat removal is via the Steam Generators' atmospheric relief valves. No primary or secondary safety valves opened during or after this trip. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 degrees F and 2235 psig, with auxiliary feedwater supplying the steam generators. There is no indications of any primary to secondary leakage. All control and shutdown rods are inserted. The electrical alignment is normal, supplied from off-site power. There was no impact to Unit 2. Unit 2 is operating at 100% power / Mode 1 The NRC Resident Inspector has been notified.
ENS 4690229 May 2011 04:25:00Watts BarAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 0155 EDT, the Watts Bar Unit 1 reactor tripped from 100% power due to a turbine trip above P-9 (reactor trip on turbine trip permissive). The cause of the turbine trip is under investigation at this time. All systems functioned as designed with the exception of Pressurizer Backup Heaters which failed to energize on lowering Pressurizer pressure. The unit is stable in Mode 3 with Auxiliary Feedwater supplying the Steam Generators. The electrical system is in normal shutdown alignment with all Emergency Diesel Generators available in standby. There are no abnormal radiological conditions at this time. The reason the Pressurizer Backup Heaters failed to energize is unknown at this time. All control rods fully inserted. No relief valves or safety valves lifted. The licensee notified the NRC Resident Inspector.
ENS 4651126 December 2010 20:53:00Browns FerryManual ScramNRC Region 2GE-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 4649220 December 2010 03:47:00SequoyahManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 0050 EST on 12/20/2010, Unit 1 reactor was manually tripped due to a reported fire inside the Unit 1 main generator bus duct. A fire was reported at 0045 EST in the Unit 1 bus duct which is located inside the turbine building. The Unit 1 reactor was tripped to remove power from the generator bus. After the reactor trip, the fire was extinguished by the application of water to the bus duct by the fire brigade. The fire was reported extinguished at 0100 EST. Method of decay heat removal is via steam dumps. Current temperature and pressure: Temp. 548 degrees and stable; Pressure 2239 psig and stable. No indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from off-site power. No impact to Unit 2, (and) Unit 2 is operating at 100% power (in) Mode 1. The NRC Resident Inspector has been notified.
ENS 4642416 November 2010 23:43:00SequoyahManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 2210 EST on 11/16/2010, the Unit 1 reactor was manually tripped based on decreasing S/G level in Loop 4. Prior to the reactor trip power was at 26% and ascending following completion of a scheduled refueling outage. Moisture Separator Reheater 1C1 safety valve lifted and would not reseat. Turbine was tripped at 2206 EST to isolate steam leak. Following the Turbine trip, automatic S/G level control did not maintain S/G level. Manual control was taken however, S/G level could not be recovered. A manual reactor trip was initiated due to low narrow range S/G level. All other plant systems responded as expected. All rods fully inserted into the core during the trip. The reactor is at normal operating pressure and temperature and operators are removing decay heat via the steam dumps to condenser. As expected, the auxiliary feedwater system actuated during the transient. The grid is stable and the plant is in its normal shutdown electrical lineup. There is no known primary-to-secondary leakage. Plant response to the trip was considered normal and uncomplicated. Unit 2 was not affected by this event. The licensee has notified the NRC Resident Inspector.
ENS 4641814 November 2010 09:05:00Watts BarManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 0617 (hrs. EST), Watts Bar experienced a failure of the cooling system to the 'A' phase Main Bank Transformer. Due to rising oil temperatures on the 'A' phase Main Bank Transformer, the reactor was manually tripped at 0652 hrs. All systems responded as designed with no issues. All rods inserted during the trip. There were no primary or secondary relief valves that lifted during the transient. The grid is stable and the plant is in a normal shutdown electrical line-up. The reactor is at normal pressure and temperature with decay heat being removed via the steam dumps to condenser with auxiliary feedwater providing steam generator make-up. The cause of the loss of cooling to the 'A' phase Main Bank Transformer was the failure of a control power transformer that supplies the Main Bank Transformer cooling system components. The licensee has notified the NRC Resident Inspector.
ENS 4594421 May 2010 22:57:00Watts BarAutomatic ScramNRC Region 2Westinghouse PWR 4-LoopAt 1937 Eastern Daylight Saving Time (EDT), Watts Bar Nuclear Power Plant Unit 1 experienced a reactor trip due to a turbine trip. This caused an automatic AFW Pump start from P-4 coincident with Lo Tave signal. The cause of the turbine trip has not yet been identified, and is under investigation. The plant is stable and is being maintained in Mode 3, at normal operating pressure and temperature, with steam generator and pressurizer levels normal. Plant systems responded to return the plant to a stable condition without complication, and all systems performed as expected with one exception: The 'B' Motor Driven Auxiliary Feedwater Backpressure Control Valve failed closed, but the Steam Driven Auxiliary Feedwater Pump provided sufficient feedwater so that all Steam Generators were provided sufficient feedwater to maintain cooling and normal steam generator level. Plans for plant restart are pending awaiting the cause investigation. All control rods inserted into the core. Plant decay heat removal is through the steam dumps to the main condenser. Offsite power is available and lined up to plant system loads. Watts Bar (NRC) Resident Inspector has been notified of this event.
ENS 4552026 November 2009 04:00:00SequoyahManual ScramNRC Region 2Westinghouse PWR 4-LoopAt 0242 EST, 11/26/2009 Unit 2 reactor was manually tripped based upon indications that the 2A Main Feedwater Pump Turbine was losing vacuum. Prior to the trip, the reactor was at 30% RTP and ascending following completion of a scheduled refueling outage. 2A Main Feedwater Pump was in service; preparations were in progress to start 2B Main Feedwater Pump to support continued power ascension. Subsequent to the trip, the Auxiliary Feedwater Pumps (motor-driven and turbine-driven) started as expected in response to the trip of both main feedwater pumps following receipt of a normal feedwater isolation signal. No primary or secondary plant safety valves operated during the transient. All plant system responses to the trip were as expected. An investigation will be conducted to identify the cause of the indicated loss of vacuum and the required corrective actions. Expected restart date to be determined. All control rods fully inserted. The licensee has notified the NRC Resident Inspector.
ENS 4539130 September 2009 04:09:00Browns FerryAutomatic Scram
Manual Scram
NRC Region 2GE-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 FerryAutomatic Scram
Manual Scram
NRC Region 2GE-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.