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The query [[Category:ENS Notification]] [[Site::Prairie Island]] [[Scram::+]] was answered by the SMWSQLStore3 in 0.3595 seconds.


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 Entered dateSiteRegionScramReactor typeEvent description
ENS 511367 June 2015 12:03:00Prairie IslandNRC Region 3Automatic ScramWestinghouse 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 IslandNRC Region 3Manual ScramWestinghouse 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 IslandNRC Region 3Manual ScramWestinghouse 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 4818614 August 2012 08:52:00Prairie IslandNRC Region 3Manual ScramWestinghouse 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 4768322 February 2012 01:55:00Prairie IslandNRC Region 3Manual ScramWestinghouse 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 470171 July 2011 18:07:00Prairie IslandNRC Region 3Manual ScramWestinghouse 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 468309 May 2011 08:44:00Prairie IslandNRC Region 3Automatic ScramWestinghouse 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 4595225 May 2010 06:50:00Prairie IslandNRC Region 3Automatic ScramWestinghouse 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 IslandNRC Region 3Automatic ScramWestinghouse 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 IslandNRC Region 3Manual ScramWestinghouse 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 432805 April 2007 14:01:00Prairie IslandNRC Region 3Automatic ScramWestinghouse 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 4250414 April 2006 16:07:00Prairie IslandNRC Region 3Manual ScramWestinghouse 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.