Semantic search
Jump to navigation
Jump to search
| Entered date | Site | Scram | Region | Reactor type | Event description | |
|---|---|---|---|---|---|---|
| ENS 57003 | 3 March 2024 15:51:00 | Prairie Island | Automatic Scram | NRC Region 3 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via email: At 1142 CST on 3/3/2024, with Unit 2 in Mode 1 at 29 percent power, the reactor automatically tripped due to a turbine trip caused by a loss of suction to the 22 main feedwater pump. All systems responded normally post trip. Decay heat is being removed via the auxiliary feedwater water system. Secondary steam control mechanism is the steam generator PORVs (power operated relief valves). Unit 1 remains at 100 percent power and is unaffected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The resident NRC inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The trip occurred while the licensee was returning to power operations after a refueling outage. During the trip, all rods inserted into the core. The plant is in a normal shutdown electrical lineup with offsite power available. The plant will be maintained at normal operating temperature and pressure. There is no known primary to secondary leakage. The cause of the loss of 22 main feedwater pump suction is under investigation. |
| ENS 54859 | 26 August 2020 18:10:00 | Prairie Island | Automatic Scram | NRC Region 3 | At 1319 CDT, on August 26, 2020, with Unit 1 in Mode 1 at 95.2 percent power in coast down for the 1R32 refueling outage, the reactor automatically tripped due to flux rate. All systems responded normally to these conditions with auxiliary feedwater initiating as expected. Operations stabilized the plant without complication. Decay heat is being removed via a main feedwater pump to the steam generators. Unit 2 is not affected and remains at 100 percent power. 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 an eight-hour report per 10 CFR 50.72(b)(3)(iv)(A), specified system actuation. The cause of the scram is under investigation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 51136 | 7 June 2015 12:03:00 | Prairie Island | Automatic Scram | NRC Region 3 | Westinghouse PWR 2-Loop | The 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 51107 | 1 June 2015 02:25:00 | Prairie Island | Manual Scram | NRC Region 3 | Westinghouse PWR 2-Loop | On 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 50950 | 3 April 2015 10:12:00 | Prairie Island | Manual Scram | NRC Region 3 | Westinghouse PWR 2-Loop | On 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 48341 | 25 September 2012 15:11:00 | Monticello | Automatic Scram | NRC Region 3 | GE-3 | During maintenance on 4160V Bus 12 ammeter, a Bus 12 lockout occurred. The station power was from 1R Reserve transformer for work on the 2R Auxiliary transformer. Net effect was Bus 12 locked out, removing power from 12 Reactor Feed Pump and 12 Reactor Recirculation pump. Reactor level lowered to +23 inches then began to rise. With both Main Feed Reg Valves in AUTO, the level transient reached +48 inches, the Reactor Water Level Hi Hi setpoint. The Main Turbine and 11 Reactor Feed Pump tripped as designed, and a Reactor SCRAM occurred. Reactor water level began to drop, and C.4.A Abnormal Procedure for SCRAM was used to restart 11 Reactor Feed Pump and recover water level. Minimum water level reached was -26 inches. Reactor Low Level SCRAM signal and Group 2 Primary Containment isolation occurred at +9 inches as designed, No Safety Relief valves lifted during this transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) did not receive an initiation signal due to not reaching their setpoints. There were no Emergency Core Cooling Systems initiation setpoints reached. Prior to the event, both divisions of Standby Liquid Control were inoperable as part of planned maintenance. All control rods fully inserted. Decay heat is being removed through the turbine bypass to the main condenser. The plant is in a normal shutdown electrical lineup and stable in Mode 3. The licensee has notified the NRC Resident Inspector and will notify the State and local governments. |
| ENS 48186 | 14 August 2012 08:52:00 | Prairie Island | Manual Scram | NRC Region 3 | Westinghouse 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.
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 47683 | 22 February 2012 01:55:00 | Prairie Island | Manual Scram | NRC Region 3 | Westinghouse PWR 2-Loop | During 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 47460 | 20 November 2011 02:10:00 | Monticello | Automatic Scram | NRC Region 3 | GE-3 | While performing a regularly scheduled Turbine Bypass Valve surveillance, prior to Turbine Bypass Valve movement, a 'B' half scram (signal) was received. Operators immediately suspended testing. Approximately 10 seconds later, a full Reactor Protection System actuation occurred. Following the reactor scram, reactor water level lowered below the Group II isolation initiation setpoint of +9 inches, (resulting in containment valve isolations). There were no radioactive releases associated with this event. No other alarms were received prior to the RPS actuation. The cause of the reactor scram is under investigation at this time. Also, due to the reactor scram, discharge canal temperature rate of change exceeded plant requirements. As a result, the State of Minnesota, and appropriate local agencies will be notified. All control rods inserted and the scram is considered uncomplicated. The plant is in a normal shutdown electrical configuration. The licensee notified the NRC Resident Inspector. |
| ENS 47364 | 21 October 2011 16:42:00 | Monticello | Automatic Scram | NRC Region 3 | GE-3 | The station experienced a lockout of the 2R Auxiliary Power Transformer. The resulting transient caused an automatic actuation of the RPS system. All control rods fully inserted. A Group 2 Primary Containment isolation occurred. Both 11 and 12 Emergency Diesel Generators started on a loss of voltage signal. Equipment response was that the 11 ESW (Emergency Service Water) pump (cooling for the #11 Emergency Diesel) failed to develop required pressure. The #13-4160V non-safety related bus failed to restore after the transient (and feed the Division 1 Essential Bus). Additionally, the #15 bus transferred to the 1AR transformer (and is feeding the Essential Bus). The #11 Emergency Diesel Generator is currently tagged out of service. Electrical supply is being provided by offsite power. Reactor heat is being removed through the main steam line to the main condenser and reactor water inventory is being provided by the feedwater system. The SRVs lifted and reseated. The HPCI system was manually place into a pressure control mode. The Minnesota Pollution Control Agency is being notified due to the licensee violating the site discharge canal temperature rate of change limit. The licensee notified the NRC Resident Inspector.
Prior to this event the 'B' Control Room Emergency Filtration (CREF) and 'B' Control Room Ventilation (CRV) Systems were inoperable for planned maintenance. On 10-21-11 at 1325 CDT, the #11 EDG ESW Pump was declared inoperable due to low cooling water pump flow, resulting in the #11 EDG being inoperable, which in turn resulted in the 'A' CREF and 'A' CRV being inoperable. Contrary to reporting requirements this condition was not identified and reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as required within 8-hours in the previous event notification. This condition resulted in a loss of safety function for both divisions of CREF and CRV. This update amends the 10-21-11 event notification to include this as an 8-hour non-emergency event pursuant to 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified the R3DO (Nick Valos) |
| ENS 47017 | 1 July 2011 18:07:00 | Prairie Island | Manual Scram | NRC Region 3 | Westinghouse PWR 2-Loop | With 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 46830 | 9 May 2011 08:44:00 | Prairie Island | Automatic Scram | NRC Region 3 | Westinghouse PWR 2-Loop | With 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 45952 | 25 May 2010 06:50:00 | Prairie Island | Automatic Scram | NRC Region 3 | Westinghouse PWR 2-Loop | During 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 45077 | 18 May 2009 14:41:00 | Prairie Island | Automatic Scram | NRC Region 3 | Westinghouse PWR 2-Loop | Prairie 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 44615 | 30 October 2008 16:23:00 | Prairie Island | Manual Scram | NRC Region 3 | Westinghouse PWR 2-Loop | During 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 43280 | 5 April 2007 14:01:00 | Prairie Island | Automatic Scram | NRC Region 3 | Westinghouse PWR 2-Loop | At 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 43114 | 23 January 2007 14:27:00 | Monticello | Automatic Scram | NRC Region 3 | GE-3 | The purpose of this notification is to inform the NRC that Nuclear Management Company (NMC) will be issuing a press release approximately two hours (CST) after this notification to the NRC on January 23, 2007, concerning an event previously reported to the NRC on January 10, 2007, via EN# 43088. The event in question involved an automatic reactor scram at 3:26 PM on January 10, 2007. As reported in that notification, all safety systems operated correctly. The scram occurred following the unexpected opening of the main turbine control valves. There was no release of radioactivity during the event. The purpose of the press release is to provide information to the media and the public regarding the results of NMC's investigation as to the cause of the January 10 event and the status of remedial actions. The licensee notified the NRC Resident Inspector. The licensee will notify State and Local authorities. |
| ENS 42504 | 14 April 2006 16:07:00 | Prairie Island | Manual Scram | NRC Region 3 | Westinghouse PWR 2-Loop | At 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 42301 | 2 February 2006 04:51:00 | Monticello | Automatic Scram | NRC Region 3 | GE-3 | Train "A" of the Emergency Filtration Train (EFT) Unit, which services the control room ventilation system, tripped off line due to a low flow condition. The cause was determined to be a rip in the rubber boot at the suction of the fan, thus causing an automatic trip of the EFT system from a low flow condition through the filter where flow is sensed. Both the "A" and "B" trains were declared inoperable due to the amount of leakage the "B" EFT was having through the ripped boot in the "A" EFT, and the condition found on "B" EFT rubber boot. Upon further evaluation of the "B" EFT boot condition, the "B" train was declared operable at 03:02 CST on 02/02/06. The "A" EFT will remain in a 7 day LCO until the rubber boot is replaced. The 8 hour notification was issued due to both EFT Units being declared inoperable. The licensee notified the NRC Resident Inspector. |