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The query [[Category:ENS Notification]] [[Reactor type::Westinghouse PWR 2-Loop]] [[Scram::+]] was answered by the SMWSQLStore3 in 0.0368 seconds.


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 Entered dateSiteScramRegionReactor typeEvent description
ENS 5157028 November 2015 23:54:00Point BeachAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopUnit 1 automatic reactor trip actuated due to an automatic voltage regulator (AVR) malfunction which caused a generator lockout and turbine trip. The cause of the AVR malfunction is unknown at this time. All control rods fully inserted. The RCS is being cooled by forced flow (reactor coolant pumps). Secondary heat sink is being provided by the condenser steam dumps utilizing the main feed water system. The auxiliary feed water system actuated based on low steam generator level, but since has been secured. Off-site power remains available. No release is occurring and emergency core cooling systems did not actuate. Emergency plan entry was not required. The plant is in its normal shutdown electrical lineup at normal operating temperature and pressure. Unit 2 was not affected by the Unit 1 transient. The licensee has notified the NRC Resident Inspector
ENS 511367 June 2015 12:03:00Prairie IslandAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopThe following was received via phone and fax: On June 7, 2015, at 0735 CDT, the Unit 2 Reactor automatically tripped while operating at 100 percent power due to an automatic Turbine trip due to low bearing oil pressure. The crew entered the reactor trip emergency operating procedure and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The automatic Reactor trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the Auxiliary Feedwater Pumps as designed on low narrow range Steam Generator level and provided makeup flow to the Steam Generators. The Auxiliary Feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam Generator levels have been returned to normal. Auxiliary Feedwater has been secured. Steam Generators are being supplied by (the) 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of the Turbine trip remains under investigation. There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified.
ENS 511071 June 2015 02:25:00Prairie IslandManual ScramNRC Region 3Westinghouse PWR 2-LoopOn May 31, 2015 at 2220 CDT, the Unit 1 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 11 Condensate Pump followed by a lockout trip of 11 Main Feedwater Pump. Manual Reactor Trip is directed by the annunciator response procedure for the lockout alarm, C47010-0101, 11 Feedwater Pump Locked Out. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. Following the reactor trip, 15A Feedwater Heater relief lifted and failed to reseat. 12 Main Feedwater Pump was subsequently secured resulting in 15A Feedwater Heater relief valve reseating successfully. Steam generators are being supplied by 12 Motor Drive Auxiliary Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of 11 Condensate Pump trip remains under investigation. There was no effect on Unit 2 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified. Unit 2 is unaffected and remains at 100 percent power.
ENS 509503 April 2015 10:12:00Prairie IslandManual ScramNRC Region 3Westinghouse PWR 2-LoopOn April 3, 2015 at 0652 CDT, the Unit 2 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 21 Main Feedwater Pump as required by the annunciator response procedure for the lockout alarm. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. The auxiliary feedwater pumps have subsequently been secured and returned to automatic. Steam generators are being supplied by 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of 21 Main Feedwater Pump trip has been determined to be a failed suction pressure switch. There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified. The licensee plans to issue a press release.
ENS 506493 December 2014 01:10:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopInitiated a manual Unit 1 reactor trip due to imminent failure of 1P-25B Condensate Pump. Unit 1 had commenced a rapid down power due to the degradation of the pump. The event is reportable under 10CFR50.72(b)(2)(iv)(B) for a manual actuation of the reactor protection system when the reactor is critical and 10CFR50.72(b)(3)(iv)(A) for an actuation of specified system (6) PWR auxiliary feedwater system. Auxiliary feedwater system actuation was due to low steam generator water levels in both 'A' and 'B' Steam Generators, an expected system response during a reactor trip. Decay heat removal is by forced circulation and is being controlled by auxiliary feedwater system and condenser steam dumps. After the trip, both main steam generator feedwater pumps were secured due to feed pump suction pressure remaining low post trip. All other plant systems functioned as required. All control rods fully inserted in the core due to the manual trip. There was no ECCS actuation. Off-site power has been maintained throughout the event. No primary or secondary safety relief valves lifted during the reactor trip. Unit 1 is in a normal shutdown electrical lineup. There was no effect on Unit 2. The licensee notified the NRC Resident Inspector.
ENS 4921424 July 2013 17:46:00GinnaAutomatic ScramNRC Region 1Westinghouse PWR 2-LoopAt 1419 EDT on 7/24/2013, the reactor tripped due to a reactor protection system (RPS) actuation signal from a turbine trip, which was caused by a generator trip. All control rods inserted on the trip and reactor coolant system (RCS) pressure is currently 2235 psig and stable with RCS temperature stable at 547 degrees F. Decay heat is being removed by steam dumps (to the main condenser) and auxiliary feedwater which auto started as expected. The cause of the generator trip is under investigation. The plant will remain in Mode 3 until the cause of the trip is determined. The plant notified the NRC Resident Inspector.
ENS 4818915 August 2012 00:02:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopUnit 1 Manual Reactor Trip was initiated in anticipation of an auto turbine trip due to operators noticing the turbine governor valves closing in response to an Electro-Hydraulic Control System signal. All Control Rods are fully inserted. The RCS is being cooled by forced flow (reactor coolant pumps). Secondary heat sink is being provided by the condenser steam dumps utilizing the main feedwater system. The auxiliary feedwater system actuated based on low steam generator level, but has since been secured. There were no unexpected (inconsistent with nature of trip) pressure or level transients. Offsite power remains available. No release occurred nor is ongoing. Emergency Core Cooling did not actuate. No unexpected isolations occurred. Emergency Plan entry was not required. The licensee informed the NRC Resident Inspector.
ENS 4818614 August 2012 08:52:00Prairie IslandManual ScramNRC Region 3Westinghouse PWR 2-Loop

Prairie Island Unit 1 is currently being shutdown per Tech Spec 3.8.1.F due to both Diesel Generators inoperable for Unit 1. On August 13th at 0939 CDT, a planned entry to Tech Spec 3.8.1.B was performed for one Diesel Generator inoperable, due to the scheduled monthly surveillance run of D1 Emergency Diesel Generator. At 1048 CDT, a small candle sized flame was identified at the exhaust manifold and D1 was subsequently shutdown. Subsequent investigation by maintenance determined that there appeared to be a gasket leak on the turbocharger. D1 was tagged out of service and repairs are currently in progress. Tech Spec 3.8.1 required action B.3.1 requires a determination be made to verify the operable Diesel Generator is not inoperable due to a common cause failure. On August 14th at 0230 CDT, Unit 1 entered the Limiting Condition for Operation to perform the monthly surveillance run to verify no common cause failure existed. At 0312 CDT, the Shift Manager reported a small candle sized fire on the exhaust manifold for D2. Unit 1 entered an event or condition that could have prevented fulfillment of a safety function, a 10 CFR 50.72 (b)(3)(v)(D) report is required due to a loss of both D1 and D2. D2 was subsequently shutdown and declared inoperable. A Technical Specification shutdown was also required and a Unit 1 Shutdown was commenced at 0425 CDT and a 4 hour non-emergency notification is required per 10 CFR 50.72(b)(2)(i). With both Diesels inoperable at 0230 CDT, Tech Spec 3 8.1.E requires one diesel to be returned to operable status within 2 hours. However, as neither diesel generator could be returned to service in this time period, Tech Spec 3.8.1.E requires the plant to be in Mode 3 within 6 hours and Mode 5 within 36 hours. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 8/14/12 AT 1452 EDT FROM TERRY BACON TO DONG PARK * * *

A Technical Specification shutdown has been completed at 1025 CDT as planned for Unit 1. It was a normal manual reactor trip with no unexpected equipment issues. As expected due to plant electrical conditions, the Auxiliary Feedwater System auto started. This is reportable per 10 CFR 50.72(b)(3)(iv)(A) as a valid System Actuation, The Auxiliary Feedwater System operated as expected. Unit 1 is currently in Mode 3. The NRC Resident Inspector has been notified. Notified R3DO (Giessner).

ENS 4805328 June 2012 01:06:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-Loop

Unit 2 Manual Reactor Trip was actuated due to indications of a 100% Load Rejection. The cause of the Load Rejection is not known at this time. All Control Rods are fully inserted. The RCS is being cooled by forced flow (reactor coolant pumps). Secondary heat sink is being provided by the condenser steam dumps utilizing the main feedwater system. The auxiliary feedwater system actuated based on low steam generator level, but has since been secured. There were no unexpected (inconsistent with nature of trip) pressure or level transients. Off site power remains available. No release occurred nor is ongoing. Emergency Core Cooling Systems did not actuate. No unexpected isolations occurred. Emergency Plan entry was not required. The plant is stable at normal temperature and pressure. The electrical system is in a normal offsite power alignment. The Unit 2 Reactor Trip had no effect on Unit 1 which continues to operate at 100% power. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM ERIC SONNENBERG TO HOWIE CROUCH AT 2337 EDT ON 7/6/12 * * *

On June 27, 2012 at 2046 CDT, a Unit 2 Manual Reactor Trip was initiated in anticipation of an automatic turbine trip due to operators noticing the turbine governor valves closing and turbine load reduction. It was originally reported that the cause was not known. This notification is updated to provide information to the cause of the load reduction. Troubleshooting has shown that the reduction of turbine load was due to a turbine speed channel card failing in the turbine control system. The card failure resulted in sending the auxiliary governor in the turbine control system an incorrect indicated overspeed condition and throttling the turbine governor valves to reduce the turbine speed. There was no actual turbine overspeed condition. The auxiliary governor is not part of the reactor protection system. No reactor protection setpoints were exceeded. All other plant systems functioned as required, including the Reactor Protection System. All control rods fully inserted into the core due to the manual reactor trip. There was no Emergency Core Cooling System actuation. No Emergency Diesel Generators were started and power continued to be supplied from off site. The reactor coolant system had forced circulation and the condenser steam dumps were used for decay heat removal from the steam generators. The licensee has notified the NRC Resident Inspector. Notified R3DO (Daley).

ENS 4768322 February 2012 01:55:00Prairie IslandManual ScramNRC Region 3Westinghouse PWR 2-LoopDuring a normal shutdown in preparation for refueling outage 2R27, with Unit 2 at approximately 11.42% power, Unit 2 was manually tripped on 2/21/2012 at 2342 CST. The manual reactor trip was in response to a 21/22/23 Feedwater Heater Hi Hi alarm and was directed by the alarm response. Procedure 2E-0, 'Reactor Trip or Safety Injection,' was completed at 2345 CST. No Safety Injection was required. 2ES-0.1, 'Reactor Trip Recovery,' is in progress. Offsite power remains on all safeguards buses for both units. Unit 2 decay heat is via forced circulation and condenser steam dump with main feedwater providing flow to 21/22 steam generators. Auxiliary Feedwater start was not required and Unit 2 AFW remains in its safeguards alignment. No emergency event was declared as a result of this trip. Unit 1 remains at 100% power in Mode 1. Reportable actuations are: Unit 2 reactor protection (scram). The NRC Resident Inspector was notified. State (State of Minnesota) / local (Goodhue county) / Press release will be made. Other government agencies will not be notified. Nothing unusual / not understood. Unit 2 will continue to mode 5.
ENS 4733812 October 2011 02:33:00GinnaAutomatic ScramNRC Region 1Westinghouse PWR 2-LoopAutomatic Reactor Trip due to Turbine Auto Stop Valve Closure and Actuation of Auxiliary Feedwater System. At 2328 on 10/11/2011, the reactor tripped due to a RPS actuation Signal from a turbine trip, which was caused by a Turbine Auto Stop signal. All control rods inserted on the trip, RCS pressure is currently 2235 psig and stable, and RCS average temperature is 547 degrees and stable. Decay heat removal is being controlled by auxiliary feedwater which auto started as expected and steam generator atmospheric relief valves. The licensee is investigating the cause of the Auto Stop Signal. The plant will be maintained in MODE 3 until the cause of trip is determined. The licensee has notified the NRC Resident Inspector. There is no primary to secondary leakage. Offsite power is normal and all EDG's are available.
ENS 470171 July 2011 18:07:00Prairie IslandManual ScramNRC Region 3Westinghouse PWR 2-LoopWith Unit 1 at 100% power Unit 1 was manually tripped at 1552. The manual reactor trip was in response to the right main turbine stop valve failing closed as the result of an electro-hydraulic oil leak located at the stop valve. Procedure 1E-0 'Reactor Trip or Safety Injection' was completed at 1600. No SI (safety injection) required. 1ES-0.l 'Reactor Trip Recovery' is in progress. Offsite power remains on all safeguards buses for both units. 11 and 12 AFW pumps auto started on SG (steam generator) low level and are supplying Unit 1 Steam Generators. After the trip, power was lost to non-safety related 4160 VAC buses 11 and 14 as expected due to the electrical lineup. The loss of power to 4160 VAC bus 11 upon the reactor trip resulted in a loss of power to 11 RCP. 12 RCP continues to operate on offsite power. Unit 2 remains at 100% power/Mode 1. Reportable actuations are: Unit 1 reactor protection (scram), and Unit 1 AFW pumps auto start. The NRC Resident Inspector has been notified.
ENS 4695713 June 2011 23:29:00Point BeachAutomatic ScramNRC Region 3Westinghouse PWR 2-Loop

On June 13, 2011 at 1924 CDT, an automatic reactor trip was actuated during Shutdown Bank Insertion for Beginning of Life Physics Testing. Neutron flux lowered from the intermediate range to the point of automatic energization of the Source Range Nuclear Instruments. Energization of the Source Range Nuclear Instruments during this step is anticipated and permissible per the test procedure. When it energized, the N-31 Source Range Nuclear Instrument (Red Channel) failed high which initiated an automatic reactor trip. All plant systems functioned as required, including the Reactor Protection System. All control rods fully inserted into the core. There was no Emergency Core Cooling System or Auxiliary Feedwater System actuation. No Emergency Diesel Generators were started and power continues to be supplied from off site. The reactor coolant system has forced circulation and the atmospheric steam dumps are currently being used for decay heat removal from the steam generators. There was no radiological release and emergency plan implementation was not required. An investigation of the failure of the N-31 Source Range Nuclear Instrument is in progress There is no known primary to secondary leakage. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM RUSS PARKER TO DONALD NORWOOD AT 2155 EDT ON 6/14/11 * * *

At the time of the trip, the reactor was subcritical. This notification is updated to reflect reporting in accordance with 10CFR50.72(b)(3)(iv)(A). An investigation of the failure of the N-31 source range nuclear instrument continues. This event was initially reported by the licensee as a four-hour report in accordance with 10CFR50.72(b)(2)(iv)(B). The licensee notified the NRC Resident Inspector. Notified R3DO (Kozak).

  • * * UPDATE FROM ALEX RIVAS TO HUFFMAN ON 10/12/11 AT 1502 EDT * * *

'On June 14, 2011 at 2055 EDT, EN 46957 was amended to state the reactor was subcritical at the time of the trip. Subsequent review of this situation by the licensee resulted in a determination that at the time the test sequence was initiated, the reactor was critical and in MODE 2. Accordingly, this non-emergency report is amended to reflect EN 46957 as originally submitted on 06/13/11 at 2229 CDT. LER 301/2011-004-01 (submitted on July 25, 2011) will be revised accordingly.' The licensee has notified the NRC Resident Inspector. R3DO (Peterson) notified.

ENS 468309 May 2011 08:44:00Prairie IslandAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopWith Unit 2 at 100% power and Unit 1 in Mode 6 and severe weather in the vicinity, (a) Unit 2 Main Generator Lockout trip occurred at 0722 (CDT). The reactor trip was 'Turbine Trip'. Procedure 2E-0, 'Reactor Trip or Safety Injection' was completed at 0725 hrs. (with) no Safety Injection required. (Procedure) 2ES-0.1, 'Reactor Trip Recovery' is in progress. Offsite power remains on all safeguards buses for both units. (The) 21 and 22 AFW (Auxiliary Feed Water) pumps automatically started on steam generator low level and are supplying Unit 2 steam generators. Unit 1 shutdown cooling was not affected. Reportable actuations are: Unit 2 Reactor Protection (scram), Unit 2 AFW pumps automatic start. The licensee has notified the State of Wisconsin, the State of Minnesota, the Prairie Island Indian Nation and the NRC Resident Inspector. They will be issuing a press release.
ENS 4648215 December 2010 05:40:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopOn 12/15/10 at 0148 (CST) Control room personnel initiated a manual reactor trip in order to abort a startup in progress with the Unit 2 Reactor subcritical. During the performance of OP-1B, Reactor Startup for Unit 2 both Rod Control System Urgent and Non-Urgent alarms were received. Shortly after receiving the alarms multiple groups of control rods fell into the core as indicated by Individual Rod Position Indicators (IRPl's) and rod bottom lights. Based on these indications a manual reactor trip was initiated. All systems functioned as expected, with all control rods fully inserting. The plant is currently in Mode 3 and operating in accordance with normal plant procedures. The cause of the control rod alarms is under investigation. The licensee informed the NRC Resident Inspector.
ENS 4612926 July 2010 22:45:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopOn 7/26/2010 at 2001 (hrs. CDT), Control Room personnel initiated a manual reactor trip from approximately 19% reactor power. Unit 1 was in the process of coming off-line to support Main Generator repair. The generator breaker had just been opened and load transferred to condenser steam dumps when a loss of condenser vacuum occurred. The reactor was manually tripped due to a loss of main condenser vacuum with reactor power above P-10 permissive. All systems functioned as expected. All control rods fully inserted. Main Steam Isolation valves were manually shut. All reactor coolant system parameters are as expected, with reactor coolant temperature being maintained by atmospheric steam dumps. Currently, the plant is at normal operating temperature and pressure with the steam generators being fed by the main feed pumps. Feedwater is being supplied via the condenser and condensate storage tank. There were no lifts of safeties or reliefs during the transient. The plant is in its normal shutdown electrical line-up with no effect on Unit 2. There is no known primary-to-secondary leakage. The cause of the loss of vacuum is under investigation. The licensee has notified the NRC Resident Inspector.
ENS 460809 July 2010 10:44:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopOn 07/09/10 at 0647 (CDT) hours, control room personnel initiated a manual reactor trip of Unit 2 from approximately 64% power as a result of a failure of the "A" feedwater regulating valve (FRV). All (other) systems functioned as expected. All rods fully inserted into the core. The unit is stable in MODE 3 at normal RCS (Reactor Coolant System) pressure and temperature. The cause of the FRV failure is being investigated. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The unit electrical power is lined up to offsite power in a normal configuration. Decay heat is being removed from the steam generator through the steam dumps to the main condenser. The FRV failed open and the valve controller was unable to place the FRV into the correct position. The steam generator HI-Hi level provided a feedwater isolation signal and a high level lockout of the FRV. Currently, the feedwater bypass valve is controlling steam generator level. There was no impact on Unit 1. The licensee has notified the NRC Resident Inspector.
ENS 4603219 June 2010 11:25:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopAt 0636 hours, control room personnel initiated a manual reactor trip of Unit 2 from MODE 2 at 0% power. Just prior to the initiation of the manual reactor trip, an automatic turbine trip had occurred as a result of the receipt of a generator lockout signal. Prior to the automatic turbine trip, an orderly power reduction to 44% power had been completed and the unit was being maintained in a stable condition in MODE 1. At the time of the turbine trip, two sets of condenser steam dump valves were isolated, in preparation for scheduled condenser waterbox tube cleaning. Following the manual reactor trip, all safety systems and equipment operated as expected. The cause of the turbine trip, including receipt of the generator lockout signal, is under investigation. The unit is stable in MODE 3 at normal RCS temperature and pressure. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) t and 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been notified. Just prior to the trip, the unit was critical in Mode 2. All rods are fully inserted into the core.
ENS 4595225 May 2010 06:50:00Prairie IslandAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopDuring a normal plant power increase following a refueling outage on Unit 2, a reactor trip occurred at approximately 32% power. This reactor trip was the result of a turbine trip. The cause of the turbine trip is unknown at this time, however, a lock out trip occurred on the only running main feed water pump (21 main feedwater pump) at the time of the turbine and reactor trip. An investigation is ongoing. The reactor trip first actuated indication was a turbine trip. An automatic start of both Auxiliary Feed Water pumps occurred following the trip. The operating crew responded to the reactor trip utilizing emergency operating procedures for reactor trip and reactor trip recovery and transitioned into a normal shutdown procedure. All rods inserted as expected and all other systems operated as expected with the exception of a positive displacement charging pump that lifted a relief that failed to reclose. The positive displacement pump relief valve stuck open and the pump was shut down which isolated the relief valve. Decay heat was initially being removed to the main condenser however, steam leak by was causing a plant cooldown therefore the Main Steam Isolation Valves were shut. Decay heat is being removed using the steam generator atmospheric relief valves. There is no known primary to secondary leakage. The plant is in its normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector.
ENS 4507718 May 2009 14:41:00Prairie IslandAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopPrairie Island Unit 1 experienced an automatic turbine and reactor trip following a lockout trip of (the) 12 Circulating Water Pump. Lockout of the circulating water pump resulted in a condenser A/B differential pressure trip of the main turbine which in turn caused an automatic reactor trip. Auxiliary Feedwater Pumps automatically started on low steam generator level. All control rods fully inserted. Decay heat removal is via auxiliary feedwater and condenser steam dump. Offsite power was maintained to safeguards and non-safeguards AC buses. Operations are in progress per Reactor Trip Response emergency procedures to stabilize plant conditions, restore main feedwater flow to the steam generators, and then shut down auxiliary feedwater pumps. The plant will then be maintained per normal shutdown procedures until the cause of the trip is corrected. No safety or relief valves lifted during the transient. There was no impact on Unit 2. The licensee notified the NRC Resident Inspector.
ENS 4461530 October 2008 16:23:00Prairie IslandManual ScramNRC Region 3Westinghouse PWR 2-LoopDuring the performance of 030 (post refueling start-up testing), control rods were being inserted for dynamic rod worth measurement. An urgent failure occurred in the rod control system which caused Group 1 rods in Control Bank A to stop inserting while Group 2 rods continued to insert. Reactor was manually tripped following the receipt of rod control alarms due to rod misalignment within Control Bank A. All rods inserted as expected. The licensee notified the NRC Resident Inspector.
ENS 434075 June 2007 20:03:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopOn 06/05/07 at 1512 hours CDT, operators observed that the Unit 1 main feedwater regulating valve (1 FD-476B) was going full open to full shut and entered abnormal operating procedure (AOP) 2B for a feedwater system malfunction. An immediate inspection of the valve determined that the valve positioner arm was disconnected, with the positioner arm locknut found on the floor adjacent to the valve. Operators manually tripped the Unit 1 reactor at 1517 hours CDT in response to the loss of 'B' train main feedwater control. During the trip, the auxiliary feedwater system actuated due to low level in the 'B' steam generator and an actuation of the ATWS mitigating system (AMSAC). Plant systems and equipment functioned properly following the manual trip with the following exceptions: A switchyard bus section 2 lockout occurred, resulting in loss of 345 Kv Line 121; 1FD-2603 bleeder trip valve (1 HX-22A moisture reheater drain) stuck open; and 1 P-129A turbine bearing oil lift pump did not automatically start, but was successfully started manually. Troubleshooting and additional investigations are continuing. The affected equipment has been quarantined. Unit 1 is in MODE 3 and is stable. Unit 2 was unaffected by the Unit 1 manual trip. All control rods fully inserted on the trip. No safety or relief valves lifted from the trip. Reactor pressure and temperature are being maintained with main feedwater and steaming to the main condenser. Emergency Diesel Generator GO-1 (two EDGs per train) and one Service Water Pump are out of service for replacement. The licensee notified the NRC Resident Inspector.
ENS 432805 April 2007 14:01:00Prairie IslandAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopAt 09:08 am on 4/5/2007, during surveillance testing of Unit 2 Train A safeguards logic at power, a spurious Train A safety Injection (SI) actuation occurred resulting in reactor protection system (RPS) actuation. Train A SI was in "Test" at the time and should not have caused the RPS trip. The operating crew manually actuated Train B SI as required by emergency operating procedures. All automatic actions for a reactor trip and safety Injection occurred as required. Reactor Coolant System (RCS) pressure decreased below the shutoff head of the high head Emergency Core Cooling System (ECCS) pumps during the transient, resulting in momentary ECCS discharge to the RCS. SI has been terminated per emergency operating procedures. Prairie Island Unit 2 has been stabilized in mode 3, at about 2235 psig and 547 degrees average RCS temperature. Decay heat Is currently being removed by auxiliary feedwater and secondary steam dump to the main condenser. The cause of the actuation signal is under investigation. All control rods fully inserted. No primary power operated relief valves or safety valves lifted. No steam generator safeties lifted. Safeguards buses are powered by offsite power. The Unit 2 Emergency Diesel Generators (EDG) started but did not load. Unit 1 Control Rod Drive Mechanism cooling isolated as designed in response to the actuation and has since been restored. Otherwise, Unit 1 was unaffected and remains in mode 1 at 100% power. The licensee notified the NRC Resident Inspector. The licensee will also be notifying the State, local and other Government agencies and will be issuing a press release.
ENS 4324317 March 2007 02:13:00GinnaAutomatic ScramNRC Region 1Westinghouse PWR 2-LoopAt 2209 on 3/16/07, the plant tripped on a Safety Injection (SI) signal initiated because of low main steam line pressure in the 'A' main steam loop. The licensee is currently conducting a post trip review but believes that the low main steam line pressure in 'A' loop was caused by a spurious isolation of the 'B' Main Steam Line Isolation Valve (MSIV). The isolation in the 'B' main steam loop lead to high main steam flow and low main steam line pressure in the 'A' main steam loop. The 'A' MSIV then also auto-closed on the SI signal. The plant is currently stable in Mode 3 at about 2235 psig pressure and 547 degrees average Reactor Coolant System (RCS) temperature. All control rods fully inserted on the trip. Decay heat is currently being removed by auxiliary feedwater feeding the steam generators and steaming out the plant atmospheric steam valves. Since plant pressure did not decrease below 1500 psig, SI did not actually inject into the RCS. The licensee secured SI. No primary PORV's or safety valves lifted. No main steam safeties lifted according to plant closure indicators. There are no primary to secondary steam generator tube leaks. All electrical safeguards buses are powered by offsite power. The Emergency Diesel Generators (EDG) started but did not load and were shut down. The EDGs are operable and available if needed. The licensee notified the NRC Resident Inspector.
ENS 4312828 January 2007 00:02:00GinnaAutomatic ScramNRC Region 1Westinghouse PWR 2-LoopOn January 27, 2007 at approximately 2040 hours an automatic reactor trip occurred. The cause of the trip was Over Temperature Delta T (2/4). All systems functioned as designed. All control rods inserted on the trip. Decay heat removal is via condenser steam dump and Auxiliary Feedwater. The initial cause of the trip appears to be from a loss of load due to a turbine electro-hydraulic system issue. This is still under investigation. RCS Temperature is 547 Degrees F and stable RCS Pressure is 2235 psig and stable Both pressurizer PORVs momentarily opened and then closed during the transient. For 8Hr Non Emergency 10 CFR 50.72(b)(3) RPS actuation occurred. Auxiliary Feed Water actuation occurred. There was no testing or maintenance in progress at the time of the transient. The licensee informed the NRC Resident Inspector.
ENS 4298310 November 2006 17:33:00KewauneeAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopOn 11/10/2006 with a shutdown in progress to repair a degraded bearing on the turbine generator, an automatic reactor trip occurred due to a power range nuclear instrumentation (NI) low range - high flux trip. Reactor power had just been lowered to below 10% power (P-10) where the power range (NI) low range trips become active. The bistable for power range NI N-42 had been tripped due to an unrelated failure on 11/09/2006. When P-10 automatically unblocked, a power range NI low range high flux reactor trip was generated. At the time of the trip, reactor power was well below the trip setpoint of 24.5% power. Following the trip, Main Feedwater Regulating Valve, FW-7A, did not automatically close as required on the reactor trip coincident with Low Tave (554F). The Reactor Operator reported FW-7A was mid-position and attempted to manually close FW-7A. It did not respond. As a result, levels in steam generator A rose to greater than 67%, which initiated feedwater isolation. The feedwater isolation signal tripped the running feedwater pump. With no feedwater pumps running, both Auxiliary Feedwater Pump A and Auxiliary Feedwater Pump B automatically started as required. The High-High steam generator level also resulted in a second reactor trip initiation signal. The Reactor Operator manually controlled Auxiliary Feedwater flow to steam generator A to restore normal level. Following the feedwater isolation, FW-7A fully closed. Following the trip, MS-201B1, the steam supply to main steam reheater B1 was locally isolated to limit the RCS cool down. This was a previously discussed contingency action. Main steam isolation valves remained open and normal condenser heat sink remained available. Further investigation as to the cause of the trip is in progress. Recovery actions per normal operating procedures are in progress. The plant was being shut down at a rate of a half percent power per minute at the time of the trip. All control rods fully inserted on the reactor trip and no safety or relief valves lifted. The plant was aligned for the normal shutdown electrical lineup prior to the trip. The temperature on the generator bearing reached a maximum of 190F with trip guidance set at 225F. The licensee notified the NRC Resident Inspector.
ENS 4294730 October 2006 11:56:00KewauneeAutomatic ScramNRC Region 3Westinghouse PWR 2-Loop

During power operation, at 92% rated power, an automatic reactor trip occurred. The reactor protection signal that caused the reactor trip was steam generator 'B' steam flow greater than feedwater flow coincident with low water level on steam generator 'B.' The cause of the plant transient that led to the reactor trip was a loss of Instrument Bus 1 (Red Channel). Instrument Bus 1 unexpectedly deenergized during the performance of maintenance on the inverter (BRA-111) that feeds Instrument Bus 1. Following the reactor trip, the auxiliary feedwater pumps automatically started, as designed, due to a low level in the steam generators. After the trip, non-safety related 4160 Volt AC Bus 4 de-energized and secondary plant feedwater heater 15B relief valve lifted. The cause of the loss of Bus 4 and 158 feedwater heater relief lifting Is under Investigation. The plant is currently stable and in the hot shutdown (HSD) mode. Power was restored to Instrument Bus 1 at 1018. This event is reportable under 10 CFR 50.72(b)(2)(iv)(B) as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical because of the automatic reactor trip and under 10 CFR 50.72(b)(3)(iv)(A), any event or condition that results in valid actuation of any of the systems listed below because of the actions of the RPS and the automatic start of the AFW pumps. All control rods fully inserted. Decay heat is being removed by feeding the steam generators with AFW and steaming to the Condenser Dump. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM T. BUNKELMAN TO W. GOTT AT 1332 EST ON 10/30/06 * * *

Due to the loss of Bus 4, the running Circulating Water Pump was lost resulting in a loss of normal heat sink to the condenser. The standby Circulating Water Pump was started at 1002 CST and the condenser heat sink was restored. Until the condenser steam dump was restored, the plant was steaming through the steam generator PORVs (atmospheric steam dumps). There is no steam generator tube leakage. The licensee notified the NRC Resident Inspector. Notified R3DO (M. Phillips)

ENS 4253026 April 2006 22:39:00KewauneeManual ScramNRC Region 3Westinghouse PWR 2-Loop

At 2043 (CDT) on 4-26-06, during a plant shutdown with the reactor at approximately 35% power, the operating crew manually initiated a reactor trip. The operating crew had just stopped one of the two condensate pumps and then the remaining feedwater pump tripped unexpectedly. The operating crew recognized the turbine did not trip, as it is expected to automatically trip when no feedwater pumps are running. The automatic turbine trip would have automatically tripped the reactor. Therefore, the operating crew manually initiated a reactor trip. Because the reactor did not automatically trip (i.e., failure of RPS to initiate and complete a reactor trip), the Shift Manager declared an Alert, at 2049, based on Chart F of Table 2-1 EPIP-AD-02. Therefore, this is a one-hour notification in accordance with 10CFR50.72(a)(1)(i) 'The declaration of any of the emergency classes specified in the licensee's approved Emergency Plan.' The manual reactor trip is reportable (4-hour) in accordance with 10CFR50.72(b)(2)(iv)(B) 'Any event or condition that results in an 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.' All systems functioned as expected following the manual reactor trip. Service Water Train B is inoperable because of a one-gallon per minute leak. All rods inserted fully. Decay heat is being removed with the steam dump and secondary PORVs. The condenser is losing vacuum due to the turbine trip. Auxiliary Feedwater Pumps started as required. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM JERRY RISTE TO JOHN KNOKE AT 01:45 EDT ON 04/27/06 * * *

At 2049 on April 26, 2006, Kewaunee Power Station staff declared an Alert emergency classification (reference EN# 42530). The Kewaunee Power Station staff has assessed this event. There was no affect on the health and safety of the general public and no release of radiation. No plant personnel were injured and the only plant equipment problem was with the failure of a trip of both feedwater pumps to cause the main turbine to trip. The Kewaunee Power Plant staff has conducted a preliminary investigation of the control room indications and sequential events recorder, which indicates that before the manual reactor trip there was no automatic reactor trip signal present and a failure of the reactor trip breakers did not occur. The Alert was terminated at 0024 CDT (on 04/27/06). The unit is currently in the Hot Shutdown Mode with plans to cool the plant to less than 350 degrees Fahrenheit. The licensee notified the NRC Resident Inspector and will be notifying State and local government and issuing a press release. Notified NRR EO (MJ Ross-Lee), IRD Mgr (P. Wilson), R3DO (H. Peterson), DHS (Holz ), FEMA (Steindurf), NRC/EPA (Crews), DOE (Wyatt), USDA (Timmons), HHS (Peagler).

  • * * UPDATE ON 04/27/06 AT 1718 EDT FROM JERRY RISTE TO ARLON COSTA * * *

When performing a review of the event reported on April 26, 2006 (EN# 42530), the Kewaunee Power Station staff determined another reporting criterion was met. An eight-hour report is required to be made 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 10 CFR 50.72 except when the actuation results from and is apart of a pre-planned sequence during testing or reactor operation.' 10 CFR 50.72(b)(3)(iv)(B)(6) is PWR auxiliary or emergency feedwater system. As described in EN# 42530, a manual reactor trip of the Kewaunee Power Station was initiated at 2043 on April 26, 2006. The manual reactor trip was initiated when the plant experienced a loss of both feedwater pumps. With a loss of both feedwater pumps and a manual reactor trip, the narrow range water level in both steam generators decreased to the actuation setpoint value for starting the Auxiliary Feedwater Pumps, causing all three Auxiliary Feedwater Pumps to start as designed. Because the steam generator water level was below the actuation setpoint, this was a valid actuation of the auxiliary feedwater system. As a valid actuation of the auxiliary feedwater system, this condition is reportable under 10 CFR 50.72(b)(3)(iv). The untimeliness of the report has been entered into the Kewaunee Power Station's corrective action program. The licensee notified the NRC Resident Inspector. Notified R3DO (H. Peterson).

ENS 4250414 April 2006 16:07:00Prairie IslandManual ScramNRC Region 3Westinghouse PWR 2-LoopAt 1425 on April 14, 2006, a lockout trip of 11 Condensate Pump occurred. The condensate pump trip caused an expected lockout trip of 11 Main Feedwater Pump trip. With the loss of 50% of feedwater pump capacity, the Shift Supervisor directed a manual Unit 1 reactor trip. The manual reactor trip was successful and all systems responded as expected. The reactor protection system actuation is reportable under 10CFR 50.72(b)(2). A reactor trip from full power results in an expected steam generator narrow range level shrink to 0%. This resultant narrow range steam generator level caused an expected Auxiliary Feedwater System Actuation. Both 11 and 12 Auxiliary Feedwater Pumps started as expected. Auxiliary feedwater actuation is reportable under10CFR 50.72(b)(3). Investigation is underway to determine the cause of 11 Condensate Pump lockout. Plant operations are underway per emergency procedure 1ES-0.1, Reactor Trip Recovery, and 1C1.3, Unit 1 Shutdown, to stabilize the plant in Mode 3, Hot Standby. All control rods fully inserted. Steam generators are discharging steam to the condenser steam dump system. The Auxiliary Feedwater Pumps are maintaining Steam Generator level. The electrical grid is stable. The licensee will notify the NRC Resident Inspector.
ENS 4219913 December 2005 06:42:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopPoint Beach Unit 1 was manually tripped at 0339 CST on 12/13/05 due to a loss of condenser vacuum caused by a mechanical failure of the running circulating water pump. All plant systems responded normally, including an auxiliary feedwater actuation. The trip was uncomplicated. All rods fully inserted. MSIVs were isolated due to the loss of condenser vacuum so decay heat is being removed by the atmospheric dump valves. The licensee indicated that there are no known steam generator tube leak issues. All systems functioned as required. The licensee was not in any significant LCO at the time of the trip. The trip had no impact on the electrical lineup or on Unit 2 operations. The cause of the circ water pump failure is still under investigation but there is evidence of sheared bolts on the pump coupling. The licensee noted the turbine condenser rupture disks blew due to high pressure. The licensee notified the NRC Resident Inspector.
ENS 4217329 November 2005 01:24:00KewauneeAutomatic ScramNRC Region 3Westinghouse PWR 2-LoopAt 22:19 CST, Main Feedwater Pump B tripped on over current. A secondary plant runback from 100% power was automatically initiated. During the secondary plant runback, the reactor automatically tripped on Steam Generator B low-low level at 22:20 CST. All three Auxiliary Feedwater pumps automatically started due to low-low Steam Generator level. The plant has been stabilized at Hot Shutdown (RCS temperature approximately 547 degrees F, RCS pressure approximately 2235 psig). Investigation into the cause of the trip is on-going. This event is being reported under 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system (RPS) when the reactor is critical and 10CFR50.72(b)(3)(iv)(A) for valid actuation of the Auxiliary Feedwater System. All control rods fully inserted on the automatic trip. Steam generator water levels have recovered to indicate in the narrow range. The current decay heat removal path is auxiliary feedwater to the steam generators steaming through the power operated relief valves. There are no known primary to secondary leaks. All safety related buses are powered from offsite power. Emergency diesel generators are available and in standby. The licensee notified the NRC Resident Inspector.
ENS 4142520 February 2005 16:37:00KewauneeAutomatic ScramNRC Region 3Westinghouse PWR 2-Loop

The following information was reported by fax:

Following the initiation of RCS cooldown, Steam Generator levels began to decrease. The operator began to throttle the operating AFW pump's outlet valve open, however at 1158 on 2/20/05, the Control Room received the Annunciator for S/G B Low-Low Water Reactor Trip, which has a setpoint of 2/3 S/G level channels <17% narrow range level. This caused a Reactor Trip signal to be initiated by the Reactor Protection System. The trip signal was generated, however the plant was already in a shutdown condition with the reactor trip breakers open." All systems were functioning normally. The operator started an additional AFW pump and, at 1206, normal level was restored to S/G B and the automatic Reactor Trip signal cleared. The lowest level during this transient in Steam Generator B was 14.5% narrow range level. The licensee notified the NRC Resident Inspector.

ENS 4075415 May 2004 15:12:00Point BeachManual ScramNRC Region 3Westinghouse PWR 2-LoopA Unit 2 manual reactor trip was initiated when the control room was notified that a diver was entangled in the intake crib. Divers were being used to inspect the intake crib, install buoys, and the fish deterrent system. The diver's umbilical cord became snagged and attempts to free it were unsuccessful. The Unit 2 circulating water system was secured to aid in removing the diver from the water. The diver still had breathing air available during the transient. The diver was subsequently removed from the water unhurt. Plant systems functioned as required, including the Reactor Protection and Auxiliary Feedwater Systems. There was no Emergency Core Cooling System actuation. Note: The condenser was unavailable because circulating water was secured. This caused a loss of condenser vacuum and its use as a heat sink. The atmospheric steam dumps are currently being used for heat removal from the steam generators. The circulating water system was subsequently restored to service. This event is reportable pursuant to 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 and 50.72(b)(3)(iv)(A), PWR auxiliary feedwater system. All control rods inserted into the core. The electrical busses are in a normal shutdown line up. The licensee notified the NRC Resident inspector.
ENS 4024815 October 2003 13:25:00GinnaManual ScramNRC Region 1Westinghouse PWR 2-LoopThe plant was in mode 2 (1% power) making preparations to place the unit on-line per procedures. At 0921, the plant had entered ER-SC.1, Adverse Weather Plan, due to sustained winds greater than 55 mph. At 1024, offsite power circuit "751" was lost due to offsite storm damage(tree down on powerline). The loss of circuit "751" led to the loss of the "B" reactor coolant pump on undervotltage. Procedure AP-RCS.2, loss of Reactor Coolant flow was entered. At 1026 a manual Reactor Trip was initiated as directed by procedures. All rods fully inserted, no ECCS injection or safety valves lifted and all ESF systems functioned as designed. Due to the partial loss of offsite power, the "B" emergency diesel auto started and loaded to supply safeguards equipment. The "A" and "B" aux feedwater pumps also auto started. Offsite power was restored to safeguards equipment being supplied power by the "B" emergency diesel within 15 minutes from the second offsite power line. The NRC Resident inspector was notified.