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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 57003 | 3 March 2024 17:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Loss of Main Feedwater Pump Suction | 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. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 56803 | 19 October 2023 16:10:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip | The following information was provided by the licensee via email: On 10/19/2023, at approximately 1110 (CST), with Unit 1 in mode 1 at 100 percent power, the reactor automatically tripped. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The cause of the trip is being investigated. Operations responded and stabilized the plant. Auxiliary feedwater actuated as expected. Decay heat is being removed by the steam generator through the steam generator power operated relief valve. The trip was complex as non-safety related power was lost to both Unit 1 and Unit 2. Unit 1 is currently in mode 3 and on natural recirculation as both reactor coolant pumps are without power. Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was lost for approximately 70 minutes. No impacts to the SFP temperature were observed. 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). Due to the actuation of the auxiliary feedwater system following the reactor trip, this event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The second paragraph of the original report is amended as follows to correct information regarding the spent fuel pool for Unit 2: Unit 2 is currently in a refueling outage with all fuel in the spent fuel pool (SFP). SFP cooling was maintained at all times with one train of SFP cooling. The second train lost power and was restarted approximately 70 minutes (after power was lost). No impacts to the SFP temperature were observed. Notified R3DO (Orth) and IR MOC (Crouch) and NRR EO (Felts) via email | Steam Generator Reactor Protection System Auxiliary Feedwater Control Rod | |
| ENS 56543 | 27 May 2023 23:34:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Notification of Unusual Event Due to Multiple Fire Alarms in Containment Not Verified within 15 Minutes | The following information was provided by the licensee via email: Notification of Unusual Event, HU4.1 declared based on multiple fire alarms in the containment building not verified within 15 minutes. Turbine trip causing reactor trip due to fault on 2GT transformer. At 1845 CDT, verification of no fire in the containment building. Notified DHS Senior Watch Officer, FEMA Operations Center, CISA Central watch officer, DOE Operations Center (email), HHS Operations Center (email), EPA Emergency Operations Center (email), USDA Operations Center (email), FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email), FEMA NRCC (email) and CWMD watch desk (email).
The following information was provided by the licensee via email: This update is being made to report the actuation of the auxiliary feedwater system following the reactor trip at 1819 CDT. This event is being reported as a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). This update is also being made for the termination of the notification of unusual event at 2304 CDT on 5/27/2023. The basis for the termination was that there was no indication of a fire. Upon lockout of 2GT transformer, main to reserve power transfer did not occur on 3 of 4 non-safeguards buses. Subsequently, operator action successfully restored power to all non-safeguards buses at 1925 CDT. There was no impact to the health and safety of the public or plant personnel. The NRC resident inspector has been notified of the update. Notified R3DO (Benjam¡n), NRR EO (Walker), IRMOC (Grant), DHS Senior Watch Officer, FEMA Operations Center, CISA Central watch officer, DOE Operations Center (email), HHS Operations Center (email), EPA Emergency Operations Center (email), USDA Operations Center (email), FDA EOC (email), FEMA NWC (email) and DHS Nuclear SSA (email), FEMA NRCC (email) and CWMD watch desk (email). | Auxiliary Feedwater | |
| ENS 55503 | 3 October 2021 20:25:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Specified System Actuation | At 1525 CDT, 10/3/2021, with Unit 2 in Mode 5 at 0 percent power for a refueling outage, the 22 Turbine-Driven Auxiliary Feedwater (AFW) pump received an actuation signal during preparations for an Integrated Safety Injection test. The reason for the actuation signal is under investigation. The AFW steam admission valve opened and then, due to plant conditions, received a trip signal due to low discharge pressure. The steam supplies to the TD AFW pump were isolated. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. Unit 1 was not affected by this issue. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Auxiliary Feedwater | |
| ENS 54859 | 26 August 2020 18:19:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Flux Rate | 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. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater | |
| ENS 53408 | 17 May 2018 16:15:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Emergency Diesel Generator | At 1115 (CDT) on May 17, 2018 with Unit 2 in Mode 1 at 100% power, the station experienced an auto-start of Emergency Diesel Generator, D5. Preliminary information indicates that the Bus 25 Potential Transformer (PT) fuse drawer was inadvertently opened, causing Breaker 25-16 to open and de-energize bus 25. Operators were able to manually close the D5 EDG output breaker to re-energize bus 25. 22 Component Cooling (CC) Pump auto-started on the loss of 21 CC Pump due to low pressure in the CC system as designed. All equipment functioned as designed. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the EDG. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 51609 | 17 December 2015 19:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Declared Due to Fire Alarm in Containment Not Verified within 15 Minutes | Unusual Event HU2.1 declared at 1318 (CST). A fire alarm was received in unit 2 containment at 1307 (CST). Due to the location of the alarm, personnel were unable to verify the status within 15 minutes. At 1343 (CST), the fire alarm in containment cleared. This alarm came in shortly after a unit 2 reactor trip. The reactor trip was due to a turbine trip. Decay heat removal is via forced circulation with aux feed and steam dumps providing secondary cooling. Offsite power remains available. The reactor trip was uncomplicated and all control rods inserted. 25B feedwater heater relief valve lifted and has reseated. No offsite assistance was requested. The licensee has notified the NRC Resident Inspector. State and local authorities were notified.
The licensee terminated the NOUE (Notification of Unusual Event) at 1450 CST. The basis for the termination was determination that there was no smoke or fire in the Unit 2 containment observed during containment entry. NRC Resident Inspectors were notified. State and local governments were notified. The health and safety of the public was not at risk. Notified the R3DO (Valos), NRR EO (Morris), IRD (Grant), DHS SWO, FEMA Ops enter, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA. | Feedwater Decay Heat Removal Control Rod | |
| ENS 51136 | 7 June 2015 12:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Resulting from Turbine Trip on Low Oil Pressure | 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. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 51107 | 1 June 2015 03:20:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Trip of Condensate and Main Feedwater Pump | 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. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 50950 | 3 April 2015 11:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Loss of Main Feedwater Pump | 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. | Steam Generator Feedwater Auxiliary Feedwater Control Rod | |
| ENS 50221 | 23 June 2014 16:07:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator Auto Start Due to Degraded Bus Voltage Signal | At 1107 (CDT) on 6/23/14, Safeguards Bus 15 received a degraded voltage signal that resulted in the bus load shed and the automatic start of the D1 Emergency Diesel Generator (EDG). Bus 15 re-energized to normal voltage after the source breaker from D1 (EDG) automatically closed. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) for valid actuation of the D1 EDG. The initiating condition appears to be a failure of the automatic voltage tap changer on the 10-Bank transformer which affected voltage to the normal offsite power supply to the 1R transformer and Safeguards Bus 15. Currently Bus 15 is powered from offsite via the CT-11 transformer and D1 (EDG) has been shutdown. When Bus 15 power supply was transferred from D1 (EDG) to CT-11 transformer, a lockout was received on D1 (EDG) due to reverse power. Unit 1 is in TS (Technical Specification) 3.8.1 Condition D for one path inoperable and one EDG inoperable. At no time during the event were the health and safety of the public challenged. The redundant AC power supply train (Bus 16 and D2 EDG) remained operable and Bus 16 remained powered from offsite power. The Senior Resident Inspector has been notified. TS 3.8.1 Condition D requires that the one inoperable offsite power path or the D1 EDG be returned to operable status by 2334 CDT or be in Mode 3 in 6 hours. Unit 2 is not affected by this issue. | Emergency Diesel Generator | |
| ENS 48186 | 14 August 2012 08:12:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Based on Both Emergency Diesels Being Declared Inoperable | 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). | Emergency Diesel Generator Auxiliary Feedwater | |
| ENS 47683 | 22 February 2012 05:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Trip Due to Feedwater Heater Hi Hi Alarm | 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. | Steam Generator Feedwater Auxiliary Feedwater | |
| ENS 47017 | 1 July 2011 20:52:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unit 1 Manual Trip Due to Electro-Hydraulic Control Fluid Leak on Turbine Stop Valve | 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. | Steam Generator Main Turbine | |
| ENS 46830 | 9 May 2011 12:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Upon Turbine Trip Due to Main Generator Lockout | 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. | Steam Generator Shutdown Cooling | |
| ENS 45952 | 25 May 2010 08:14:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to Turbine Trip | 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. | Steam Generator Feedwater Main Steam Isolation Valve Auxiliary Feedwater Main Condenser | |
| ENS 45851 | 17 April 2010 03:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Trip Due to Turbine Trip | During a normal reactor shutdown for a refueling outage on Unit 2, a reactor trip occurred at approximately 13% power. This reactor trip was the result of a turbine trip due to high differential pressure between A & B condensers (greater than 2.5 inches). The cause for the vacuum difference between condensers is unknown at this time. The reactor trip first actuated indication was a high flux rate trip and the turbine trip first out indication was not received. The reason for this difference is unknown at this time. The operating crew responded to the reactor trip utilizing emergency operating procedures for reactor trip and reactor trip recovery and transitioned into normal shutdown procedures. All rods inserted as expected and all other systems operated as expected. The licensee notified the NRC Resident Inspector. According to the licensee: The plant is in a normal post-trip electrical lineup. No automatic relief valve operations occurred. The motor driven auxiliary feed pump was manually started. The main steam isolation valves are open and decay heat is being removed by steam through the turbine bypass valves to the main condenser. | Main Steam Isolation Valve Main Condenser | |
| ENS 45077 | 18 May 2009 18:05:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip/Reactor Trip Due to Condenser Vacuum | 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. | Steam Generator Feedwater Auxiliary Feedwater Main Turbine Decay Heat Removal Control Rod | |
| ENS 44920 | 19 March 2009 20:32:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Start of a Cooling Water Pump During Testing | In preparation for planned maintenance of 12 Diesel Driven Cooling Water Pump (Train A), 121 Cooling Water Pump was aligned as a safeguards replacement pump per plant procedures. Maintenance steps were then completed to a point allowing testing and 12 DDCLP was restored. The Operations procedure steps to enter the Cooling Water LCO 3.7.8 Condition A and realign 121 CL Pump away from being a safeguards replacement were missed and permission was granted to perform testing. 12 DDCLP was locally started per the PM and then tripped as directed from rated speed at 1513. The pump trip resulted in a cooling water pressure transient that automatically started 121 CLP and is reportable under 10CFR 50.72(b)(3) as an unplanned safety related system actuation. 121 CLP operated normally and there were no adverse plant effects from the transient. During the investigation of the automatic start, it was recognized that with 121 CLP still aligned as a safeguards replacement and 12 DDCLP running locally, a Safety Injection signal and start of 22 Diesel Driven Cooling Water Pump (Train B) would result in 121 CLP trip. LCO 3.7.8 Condition A was entered for one safeguards pump OOS and 121 CLP was returned to OPERABLE status at 1613. 121 Cooling Water Pump was shut down and returned to standby at 1936. Maintenance activities and testing for 12 DDCLP have been suspended pending investigation and corrective actions. 121 CLP remains aligned as a safeguards replacement and both cooling water headers have operable safeguards pumps. The licensee notified the NRC Resident Inspector.
The licensee has completed its initial investigation and corrective actions. Maintenance and testing activities for 12 Diesel Driven Cooling Water Pump has resumed. There was an error in the text of the initial notification, the time of the autostart of 121 Motor Driven Cooing Water Pump was 1532 not 1513. The licensee informed the NRC Resident Inspector. Notified R3DO (Ring). | ||
| ENS 44615 | 30 October 2008 19:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Manually Tripped Due to Failure in Rod Control System | 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. | Control Rod | |
| ENS 44377 | 31 July 2008 13:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Trip During Testing | Unit one (1) experienced a reactor trip during SP-1003 Analog Protection Functional Test. The yellow Tave channel was in test when the red Tave channel bistable failed causing an OTDT reactor trip. Applicable emergency operating procedures were entered and completed. The plant is now implementing 1C1.3, the normal plant shutdown procedure. All systems performed as expected with exception of 11 turbine driven auxiliary feed (AFW) pump auto started and tripped 50 seconds later on low suction / discharge pressure which the plant is continuing to investigate. All rods inserted and all other AFW system components are operating as expected. Decay heat removal is from Main and Auxiliary Feedwater to the Steam Dump system. No safety or relief valves actuated. The plant is in a normal electrical lineup. The licensee notified the NRC Resident inspector. | Auxiliary Feedwater Decay Heat Removal | |
| ENS 43280 | 5 April 2007 14:08:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Trip Caused by Spurious Safety Injection During Testing | 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. | Steam Generator Reactor Coolant System Reactor Protection System Emergency Diesel Generator Auxiliary Feedwater Emergency Core Cooling System Main Condenser Control Rod | |
| ENS 42504 | 14 April 2006 19:25:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Trip Due to Condensate Pump and Feedwater Pump Trip | 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. | Steam Generator Feedwater Reactor Protection System Auxiliary Feedwater Control Rod |