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 Entered dateSiteRegionScramReactor typeEvent description
ENS 5399914 April 2019 06:44:00SequoyahNRC Region 2Automatic Scram

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 5246930 December 2016 16:37:00SequoyahNRC Region 2Manual ScramWestinghouse 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 5155923 November 2015 11:52:00SequoyahNRC Region 2Manual ScramWestinghouse 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 08:12:00SequoyahNRC Region 2Manual ScramWestinghouse 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 17:04:00SequoyahNRC Region 2Automatic ScramWestinghouse 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 09:30:00SequoyahNRC Region 2Automatic ScramWestinghouse 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 4877824 February 2013 15:20:00SequoyahNRC Region 2Manual ScramWestinghouse 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 4716919 August 2011 01:10:00SequoyahNRC Region 2Automatic ScramWestinghouse 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 00:53:00SequoyahNRC Region 2Automatic ScramWestinghouse 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 19:39:00SequoyahNRC Region 2Automatic ScramWestinghouse 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 4649220 December 2010 03:47:00SequoyahNRC Region 2Manual ScramWestinghouse 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 23:43:00SequoyahNRC Region 2Manual ScramWestinghouse 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 4552026 November 2009 04:00:00SequoyahNRC Region 2Manual ScramWestinghouse 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 4509727 May 2009 22:12:00SequoyahNRC Region 2Automatic ScramWestinghouse PWR 4-LoopAt approximately 1904 on 05/27/09, Sequoyah Unit 2 received an Automatic Reactor trip on Power Range High Neutron Rate Reactor Trip. The automatic trip occurred during severe weather and lightning onsite. ESF functions initiated as designed including Aux Feed Water auto-start and automatic Feedwater isolation. All control rods fully inserted. The plant is currently being maintained in Mode 3 at approximately 547 degrees F 2235 PSIG. Decay heat is being removed by the Auxiliary Feedwater system and Steam Dump valves. The cause of the Power Range Neutron Flux High Rate Reactor Trip is not known at this time and investigation is ongoing. Unit 1 remains at 100% power". No relief or safety valves lifted. The electrical lineup is normal and the 2B Diesel Generator is tagged out for maintenance. The licensee has contacted the NRC Resident Inspector.
ENS 450457 May 2009 00:29:00SequoyahNRC Region 2Manual ScramWestinghouse PWR 4-LoopOn 5/6/09 at 2256 Sequoyah Unit-1 was manually tripped from 100% reactor power. The manual trip was in response to the failure of Loop 1 Feed Water Regulating Valve (FRV). Manual control was attempted to control level in Loop 1 Steam Generator however Loop 1 FRV failed to respond. A manual reactor trip was initiated as a result of this failure. In addition, Auxiliary Feedwater (AFW) initiated as required due to a Feedwater Isolation signal. The Loop 1 FRV did not isolate from the Feedwater Isolation signal, however the Loop 1 Feedwater Isolation Valve closed as designed. The Plant is currently being maintained in Mode 3 at NOT/NOP, approximately 547 (degrees) F and 2235 psig, with Auxiliary Feedwater supplying the steam generators and Steam Dumps to Main Condenser removing decay heat. Maintenance activities have been initiated to repair the Loop 1 FRV. All rods inserted on the trip. No safety or relief valves lifted as a result of the transient. The plant is in its normal shutdown electrical lineup. No grid instabilities exist and there was no effect on Unit 2. The licensee has notified the NRC Resident Inspector.
ENS 4502928 April 2009 23:50:00SequoyahNRC Region 2Manual ScramWestinghouse PWR 4-LoopAt 2159 Sequoyah Unit 1 was manually tripped from 18% reactor power. The manual trip was in response to automatic isolation of all intermediate pressure feedwater heater strings resulting in a loss of all condensate flow. The feedwater heater strings isolated following a manual trip of the main turbine. The main turbine was manually tripped in response to a failed open moisture separator reheater relief valve. At the time of the event Sequoyah Unit 1 was raising power following a refueling outage. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary feedwater automatically actuated as expected on loss of the operating main feedwater pump. The plant is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying the steam generators and steam dumps to main condenser removing decay heat. Maintenance activities have been initiated to repair the MSR relief valve. Electrical lineup is normal with all safety related equipment operable. Licensee is investigation the cause for the relief valve failure. The NRC Resident Inspector has been notified.
ENS 4493426 March 2009 08:29:00SequoyahNRC Region 2Automatic ScramWestinghouse PWR 4-LoopAt approximately 0452 on 03/26/09, Sequoyah Unit 1 received an Automatic Reactor trip on Reactor Coolant Pump Busses Undervoltage. A loss of Common Service Station Transformer C caused a loss of power to the 1B and 1D Unit Boards. The 1B and 1D unit boards are the 6.9Kv electrical feeds to the 1-2 and 1-4 RCPs, respectively. RCPs 1-1 and 1-3 are running. ESF functions initiated as designed including Aux Feed Water auto-start, automatic Feedwater isolation, and auto-start of all four EDGs. The 1A Shutdown Board is being powered from the 1A EDG. The plant is currently being maintained in Mode 3 at approximately 547 degrees F /2235 PSIG. Decay heat is being removed by the auxiliary feedwater system and Steam Generator Atmospheric Relief valves. The cause of the loss of Common Service Station Transformer C is not known at this time and investigation is ongoing. All rods inserted as expected. No safety related equipment is out of service. Unit 1 has no known Steam Generator Tube leaks. The licensee notified the NRC Resident Inspector.
ENS 4493526 March 2009 08:29:00SequoyahNRC Region 2Automatic ScramWestinghouse PWR 4-LoopAt approximately 0452 on 03/26/09, Sequoyah Unit 2 received an Automatic Reactor trip on Reactor Coolant Pump Busses Undervoltage. A loss of Common Service Station Transformer C caused a loss of power to the 2B and 2D Unit Boards. The 2B and 2D unit boards are the 6.9Kv electrical feeds to the 2-2 and 2-4 RCPs, respectively. RCPs 2-1 and 2-3 are running. ESF functions initiated as designed including Aux Feed Water auto-start, automatic Feedwater isolation, and auto-start of all four EDGs. The 2A Shutdown Board is being powered from the 2A EDG. The plant is currently being maintained in Mode 3 at approximately 547 degrees F /2235 PSIG. Decay heat is being removed by the auxiliary feedwater system and Steam Generator Atmospheric Relief valves. The cause of the loss of Common Service Station Transformer C is not known at this time and investigation is ongoing. All rods inserted as expected. No safety related equipment is out of service. Unit 2 has no known Steam Generator Tube leaks. The licensee notified the NRC Resident Inspector.
ENS 446499 November 2008 23:29:00SequoyahNRC Region 2Manual ScramWestinghouse PWR 4-LoopOn 11/19/08 at 1821 with unit 2 in Mode 3, the unit 2 reactor trip breakers were opened from the Main Control Room (MCR) due to indications of a Shutdown Bank "A" Rod E-11 dropping into the Reactor Core. At the time the reactor trip breakers were opened, the MCR Operators were in the process of withdrawing Shutdown Banks in preparations for entry into Mode 2. All other Shutdown Banks and Control Banks were inserted at the time the reactor trip breakers were opened. In addition, a Feedwater Isolation Signal was generated as designed. All safety related equipment operated as designed. The Plant is being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with AFW supplying the S/G's and Steam Dumps to Main Condenser removing decay heat. No primary system or steam generator safety valves opened due to this trip. An investigation has been initiated to determine the cause of the indications of Shutdown Bank 'A' Rod E-11 dropping into the Reactor Core. The licensee notified the NRC Resident Inspector.
ENS 446274 November 2008 02:58:00SequoyahNRC Region 2Manual ScramWestinghouse PWR 4-Loop

On 11/3/08 at 2322, Unit 2 was manually tripped due to the failure of Loop 4 Feed Water Reg. Valve (FRV Controller). Manual control was attempted to control level in Loop 4 Steam Generator (S/G); however, Loop 4 FRV failed to respond. A manual Reactor Trip was initiated as a result of this failure. In addition, Auxiliary Feedwater (AFW) initiated as required due to a Feedwater isolation signal The Loop 4 FRV did isolate from the Feedwater isolation signal. The Plant is being maintained in Mode 3 at NOT/NOP, 547 F and 2235 psig, with AFW supplying the S/G's and Steam Dumps removing decay heat. Additionally, Unit 2 has an indication of a primary leak inside lower containment. The leak rate is calculated to be approximately 2.0 gallons per minute. Based on current indications, the leak is suspected to be from a Pressurizer level transmitter No primary system or steam generator safety valves opened due to this trip. An investigation has been initiated to determine the cause of the Loop 4 Feed Water Reg. Valve (FRV Controller) failure and the source of the approximately 2.0 gallons per minute primary leak. A recovery plan will be developed. Pressurizer (PRZ) level is stable. All safety related systems are available and OPERABLE for safe plant shutdown. There is no impact on Unit 1. The loop 2 S/G blowdown sample line valve did not go closed as expected on the Feedwater isolation signal. The licensee is taking action to isolate it.

The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM LARRY PRUETT TO JOE O'HARA AT 0001 ON 11/5/08 * * *

The Sequoyah Nuclear Plant Unit - 2 reactor coolant system unidentified leakage was terminated on Nov. 4, 2008 @ 2205 hours when a manual valve was closed isolating the reactor coolant system from the leak location." The licensee closed the root valve to the PZR pressure channel 2-VLV-68-446A. The licensee will notify the NRC Resident Inspector. Notified R2DO(Desai)

ENS 4390916 January 2008 21:15:00SequoyahNRC Region 2Manual ScramWestinghouse PWR 4-LoopDuring a surveillance test on the Loop 3 Steam Generator pressure channel, the loop 3 Steam Generator main feedwater regulating valve went closed. The unit was manually tripped on lowering steam generator level. All control rods fully inserted on the reactor trip. Both motor driven and the turbine driven AFW pumps started successfully. The turbine driven AFW pump was secured and the motor driven AFW pumps are providing feedwater to the steam generators. Decay heat is being removed to the main condenser via the turbine steam dump valves. No PORV, safety relief valves, or atmospheric dump valves opened on the reactor trip. Unit 1 is in a normal shutdown electrical lineup. There was no effect on Unit 2 during this event. The licensee is investigating the unexpected closure of the Loop 3 Steam Generator main feedwater regulating valve. The licensee notified the NRC Resident Inspector.
ENS 4323313 March 2007 16:47:00SequoyahNRC Region 2Manual ScramWestinghouse PWR 4-LoopAt 1527 EST on 3/13/07 a manual reactor trip was actuated on Unit 2 due to 'A' MFPT (main feedwater pump turbine) malfunction. All systems responded as expected following the manual trip. The plant is currently stable in Mode 3 (Hot Standby) at 547 (degrees) F. No primary system or steam generator safety valves opened due to this trip. AFW (auxiliary feedwater pumps) system started and operated as designed. All emergency core cooling systems, emergency diesel generators are fully operable if needed and the electrical grid is stable. The NRC Resident Inspector was notified of this event by the licensee.
ENS 4311523 January 2007 14:44:00SequoyahNRC Region 2Automatic ScramWestinghouse PWR 4-LoopAn automatic reactor scram occurred resulting from loop #2 steam generator low level. The main feedwater regulating valve 2-FCV-3-48 failed closed when the air supply line to the valve severed. A significant leak had been identified and efforts to repair had been initiated. All systems responded as designed. The plant is stable on Auxiliary Feed Water (AFW) and steam generator cooling (to the condenser via steam dump valve) while repairs to the failed air line are planned. All control rods fully inserted into the core. No steam generator or pressurizer relief valves lifted during the transient. The licensee notified the NRC Resident Inspector.
ENS 415839 April 2005 13:20:00SequoyahNRC Region 2Automatic ScramWestinghouse PWR 4-Loop

At 11:11 Unit 1 Turbine Tripped initiating Reactor Trip. Initial Indications are Turbine Trip due to Lo Auto Stop Oil Pressure on 2 pressure switches. Investigating Switch Actuations.

  1. 2 and #4 S/G (Steam Generator) Atmospheric Relief Valves did not control properly and RxOp (Reactor Operator) closed valves with Control Switch.

Current Conditions: U1 (Unit 1) Mode 3, Stm Dumps In Service Normal, RCS temp at 547.4F, Pressure 2238 psig. 2 AFW (Auxiliary Feedwater) trains In Service at Normal Operation. All Rods Fully Inserted. Yes. The licensee indicated there was not excessive cooldown or depressurization in the steam generators whose atmospheric relief valves had to be manually closed. The NRC Resident Inspector has been notified.

  • * * UPDATE PROVIDED BY LICENSEE (FIELDS) TO NRC (HELD) ON 4/9/05 ON 1901 * * *

AFW Actuation - due to FW (Feedwater) Isol(ation) tripping both MFPs (Main Feed Pumps) (trips start AFW) AFW includes 2 motor-driven and 1 turbine-driven pump. R2DO (Rogers) notified.

ENS 4143724 February 2005 00:35:00SequoyahNRC Region 2Automatic ScramWestinghouse PWR 4-LoopThe following information, in addition to the phone report, was obtained from the licensee via facsimile: While performing maintenance, (Maintenance) inadvertently opened a 125 VDC battery board (four) breaker which resulted in a reactor trip from steam generator low-low level. The plant is being maintained in Mode 3 at NOP/NOT (Normal Operating Pressure/Normal Operating Temperature), 547 (degrees) and 2235 psig, with auxiliary feedwater supplying the steam generators and steam dumps removing the decay heat. All rods inserted on the trip. No relief valves lifted during the transient. The electric plant is stable with 125 VDC restored to a normal configuration. Steam generator level has been restored to normal levels. Unit 1 was not affected by the transient. The licensee has notified the NRC Resident Inspector.
ENS 4058915 March 2004 17:13:00SequoyahNRC Region 2Automatic ScramWestinghouse PWR 4-LoopAt 1518 on 3/15/04, Unit 1 Reactor tripped due to a Main Turbine trip. The Main Turbine tripped due to a Main Generator electrical fault, Auxiliary Feedwater (AFW) started when both Main Feedwater Pumps tripped on Low Tave Feedwater isolation. The AFW start was expected on the Reactor Trip. All safety systems performed as required. Investigation is in progress to determine and correct the cause of this trip. All control rods fully inserted; the electrical grid is stable; ECCS systems remain operable; decay heat is being removed via AFW and steam dumps. The licensee notified the NRC Resident Inspector.