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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 41436 | 23 February 2005 18:00:00 | 10 CFR 50.72(b)(3)(v)(A), Loss of Safety Function - Shutdown the Reactor 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Potential Vulnerability with Alternate Shutdown System (Asds) Isolation Design. | The licensee provided via facsimile the following report: During an extent of condition review of the corrective actions associated (with) Event Notification #41374, the Monticello Nuclear Generating Plant (MNGP) engineering staff made the following discovery. On February 22, 2005 at 12:00 hours, MNGP discovered a potential vulnerability with Alternate Shutdown System (ASDS) isolation design which could result in Bus 16 being locked out in the event of a Control Room or Cable Spreading Room fire. The Monticello Appendix R Safe Shutdown Analysis for Control Room/Cable Spreading Room fire assumes a loss of control of Division I and II equipment from the Control Room, however, safe shutdown is achieved remotely from the ASDS panel. ASDS design is such that a Control Room/Cable Spreading Room fire would not impede the ability to safely shutdown and maintain the plant in a shutdown condition. Contrary to the ASDS design, it was discovered that an un-isolated metering circuit from the 1AR transformer could result in Bus 16 being locked out in the event of a Control Room or Cable Spreading Room fire. The bus lockout relay from the 1AR transformer is not isolated by the ASDS transfer switches, therefore, this condition could result in failure of Bus 16 to re-energize during the implementation of the Shutdown Outside Control Room procedure. Since the Bus 16 feeder breaker from the 1AR transformer is not required at this time, it has been isolated from the safeguards bus to preclude occurrences of this event. The event is being reported as a potential loss of safety function (10CFR50.72(b)(v)(A,B and D) and as a degraded or unanalyzed condition (10CFR50.72(b)(3)(ii)(B)). The licensee informed NRC Resident Inspector. | 05000263/LER-2005-001 | |
| ENS 41441 | 25 February 2005 03:11:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Esf Actuation Following Trip of Reactor Protection Motor-Generator Set | A trip of the "A" RPS M-G set resulted in an "A" Group 2 isolation and startup of the standby gas treatment system. There was a preliminary report of a possible fire/smoke smell in the vicinity of the M-G set. However, when an operator reported to the location there was no observed fire. The fire brigade was also dispatched but found no fire or smoke in the M-G set area. The licensee is preparing to place the RPS on its alternate power supply. This will allow the Group 2 isolations and standby gas treatment system actions to be reset. The cause of the M-G set trip is still under investigation. The licensee will be notifying the NRC Resident Inspector as well as state and local authorities. | Standby Gas Treatment System | 05000263/LER-2005-002 |
| ENS 43088 | 10 January 2007 21:28:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram with Stuck Open Main Turbine Control Valves | At 1528 hours Central Time, a RPS trip occurred, all rods fully inserted. Turbine valve testing was in progress. A group 1 isolation occurred with all MSIVs closing. A partial group 2 occurred due to low reactor water level. All safety systems (valves) operated correctly. Currently reactor pressure/cooldown in progress using HPCI in pressure control, normal condensate and feedwater in service for reactor water level control. Investigation of cause is in progress. Group isolations resulted from low pressure and low reactor water level. During the transient and subsequent cooldown, Operators manually opened the main steam relief valves to maintain pressure control in accordance with the plant EOPs. HPCI was manually started to maintain reactor level. The group 2 isolation has been reset. The group 1 isolation (MSIVs) will not be reset until after the main turbine control valves are shut. The plant is being cooled down to initiate RHR cooling and is in the normal shutdown electrical lineup. The licensee notified the NRC Resident Inspector. | Feedwater Main Steam | |
| ENS 43246 | 18 March 2007 04:46:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Rps Trip Signal and Containment Isolations Received During Maintenance | With the plant in refueling outage with all rods in and in Mode 5, a RPS trip and a containment isolation was received at 2347, 03/17/07. This event occurred when Div #2 Bus 16 Bus Pot Drawer was opened during isolation of 1AR Transformer Metering under C/O-17605. The opening of this Pot Drawer caused Div #2 4KV Bus 16, Div #2 480V LC-104 and Div #2 480V MCC-141, 142, 143, and 144 to open. 'B' RPS tripped causing full RPS trip due to SRM shorting removed. The loss of power caused RBV (Reactor Building Ventilation) and Spent Fuel Pool Radiation Monitors to trip and cause containment isolation to occur. Investigation into C/O-17605, restoration of power, resetting of RPS and containment isolation in progress. The containment valves that received an isolation signal were: Primary Containment Atmosphere Control, Post Accident H2/O2 Control system, Post Accident Sampling system, O2 analyzing and SCTMT isolation (H&V). The licensee notified the NRC Resident Inspector. | Primary containment | |
| ENS 44484 | 12 September 2008 03:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram and Isolation Due to Transformer Lockout | While the 1R transformer was out of service for maintenance, the 2R transformer experienced a lockout resulting in a loss of normal offsite power, a reactor scram, and a Group 1, 2 and 3 isolation. All rods fully inserted as expected. The cause of the 2R transformer lockout is unknown at this time. After initiation, the high pressure coolant injection (HPCI) system would not trip at the high reactor water level set point +48", as required. The operators then manually isolated the HPCI steam lines. Plant decay heat removal is with the reactor core isolation cooling (RCIC) system and the safety relief valves. Torus cooling is in service. The vital electrical busses are being supplied by the 1AR transformer. Efforts are underway to restore the 1R transformer to service, and subsequently the non-vital busses. The licensee notified the NRC Resident Inspector.
The HPCI system was declared inoperable and isolated due to failure to automatically trip at +48" Reactor Water Level. The HPCI steam supply valves were automatically closed to remove HPCI from service. In addition, the HPCI turbine trip failed to trip with the turbine trip push button. The cause of the trip failure is unknown at this time. The licensee is continuing to investigate. HPCI did automatically start as designed and injected to the reactor vessel as designed. However, HPCI failed to trip at High Reactor Water Level as required. Additionally, the Automatic Depressurization System (ADS) timer showed erratic indication following the event. The ADS timer was inhibited to prevent automatic action. ADS is inoperable, but manual steam relief valve operation remains available. The licensee will notify the NRC Resident Inspector. Notified R3DO (Passehl), NRR EO (Ross-Lee), and IRD (McMurtray).
A (second) Group 2 isolation signal was received when reactor water level lowered below +9 (inches) (while pumping drywell sumps). All Group 2 valves except the drywell sump isolation valves were closed due to a previously reported Group 2 signal. The drywell sump valves had been opened to allow manual pumpdown of the sumps. The sump valves closed as expected. Licensee notified the NRC Resident Inspector. Notified R3DO (Passehl). | High Pressure Coolant Injection Reactor Core Isolation Cooling Automatic Depressurization System Decay Heat Removal Safety Relief Valve | 05000263/LER-2008-007 |
| ENS 44498 | 17 September 2008 15:17:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unusual Event Due to Loss of Shutdown Cooling | Monticello lost the 1R offsite power transformer due to an industrial accident that grounded the bus and resulted in a fatality of an employee onsite. At the time of this event, the 2R offsite transformer was already out of service due to a previous event (see EN #44484). This effectively resulted in a loss of all normal offsite power. The 1AR safety related offsite power source remained available and the safety related buses are energized and diesels are available. However, shutdown cooling was lost during the event due to a Group 2 isolation and cannot be restored. The shutdown cooling suction isolation valves #2020 and #2030 have power but until the Group 2 logic can be reset the valves cannot be reopened. The licensee needs to repower the RPS bus to reset the logic. Based on the uncertainty in reestablishing shutdown cooling an Unusual Event was declared based on Emergency Director judgment (EAL HU5.1). Current reactor coolant temperature is approximately 110 degrees with a heatup rate of about 20 degrees an hour. State and local authorities and the NRC resident inspector have been notified. The licensee will likely be making a press release.
On Wednesday, September 17, 2008 Monticello Nuclear Generating Plant (MNGP) experienced a loss of power to the station transformer resulting in a valid actuation of the following systems: Reactor Protection System (with the reactor shutdown), Containment Isolation, and Emergency Diesel Generators. The cause of the loss of power was due to contact of a 115kV transmission line by a manlift. A vendor employee was electrocuted. On-site Medical Emergency Response personnel responded until the individual was transported to North Memorial Medical Center. The individual was pronounced dead at North Memorial Medical Center. Notifications of offsite agencies and a media press release are in progress. 'This notification is being made in accordance 10CFR50.72(b)(2)(xi) and 10CFR50.72(b)(3)(iv)(A).' Licensee notified the NRC Resident Inspector." Notified R3IRC (Garza)
This is a follow-up to Event notification #44498. The station has completed notifications to off-site agencies and to the media. The station continues to troubleshoot and restore plant equipment and respond to media inquiries. NUE update - As of 1113 CDT shutdown cooling was restored." Notified R3 IRC.
Monticello Nuclear Generating Plant exited their Notification of Unusual Event at 1100 CDT on 9/21/08. The exit criteria is supported on the following information: (1) This event did not result in the loss or potential loss of a fission product barrier and did not change the status of the current fuel condition. All three fission barriers are intact and were maintained throughout the event. (2) There were no radiation releases as a result of the event. (3) Restoration of plant loads including shutdown cooling onto a normal offsite power source (the 1R transformer) has been completed. (4) The 1R transformer offsite power source was recovered and is being protected. Protection of the 1R transformer will continue until availability of the other offsite power source supplied from the 2RS/2R transformers has been restored. (5) Shutdown cooling is in service. (6) The site organization challenges in response to the injury event and loss of decay heat removal are no longer present. The licensee has notified the NRC Resident Inspector, as well as state and local agencies. A media release will be issued. Notified R3DO (Stone), NRR EO (Galloway), IRD MOC (Clark), OPA (Burnell), DHS (S. Moore), and FEMA (D. Fuller). | Reactor Protection System Emergency Diesel Generator Shutdown Cooling Decay Heat Removal | 05000263/LER-2008-007 |
| ENS 44508 | 21 September 2008 02:35:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Unplanned Loss of Shutdown Cooling Due to Containment Isolation | On Saturday, September 20, 2008, Monticello Nuclear Generating Plant (MNGP) experienced an actuation of the following systems: Reactor Protection System (with the reactor shutdown), Containment Isolation, and Emergency Diesel Generators. The apparent cause of the actuation was a pressure pulse in the reference leg of a Reactor level instrument that resulted when a CRD (Control Rod Drive) pump was started without the reference leg backfill system isolated from the CRD system. This notification is being made in accordance with 10CFR50.72(b)(3)(iv)(A). Due to the containment isolation, shutdown cooling was lost for approximately 90 minutes. Initial reactor temperature was ~95 degrees when the isolation occurred. When shutdown cooling was restored, reactor temperature had increased to ~120 degrees. The Emergency Diesel Generators started but did not load. The diesels have been restored to normal standby status. The licensee has notified the NRC Resident Inspector. | Reactor Protection System Emergency Diesel Generator Shutdown Cooling | |
| ENS 46496 | 20 December 2010 09:57:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Fuel Pool/Reactor Building Exhaust Plenum Primary Power Supply Failed | At 0357 December 20, 2010 the 'A' division Fuel Pool/Reactor Building Exhaust Plenum Primary Power Supply failed, resulting in upscale readings on both the Fuel Pool and Reactor Building Ventilation Plenum radiation monitors. This condition resulted in closure of the Group II Primary Containment Isolation Valves (PCIV), isolation of Secondary Containment (SCT), initiation of the Standby Gas Treatment System (SBGT), and a transfer of the Control Room Ventilation (CRV) and Control Room Emergency Filtration (CREF) systems to the High Radiation Mode. Conditions and Required Actions were entered for Technical Specification 3.3.6.2 (SCT Instrumentation), 3.3.7.1 (CREF Instrumentation), and 3.4.5 (RCS Leakage Detection - CAM). Radiation levels were verified to be normal in the affected areas. Isolations signals were reset and Secondary Containment ventilation systems were restored to a normal lineup. Repairs are currently in progress to replace the high voltage power supply to the affected radiation monitors and are expected to complete within the required action time limits of the applicable technical specifications. The licensee is in a 24 hour LCO. The licensee has notified the NRC Resident Inspector. | Secondary containment Primary containment Reactor Building Ventilation Standby Gas Treatment System | |
| ENS 47364 | 21 October 2011 17:50:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Lockout of Auxiliary Power Transformer | 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) | Feedwater Emergency Diesel Generator Primary containment Main Steam Line Main Condenser Control Rod | |
| ENS 47460 | 20 November 2011 05:12:00 | 10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram While Performing Turbine Bypass Valve Testing | 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. | Reactor Protection System Control Rod | |
| ENS 48341 | 25 September 2012 15:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram During Maintenance on 4160V Bus 12 Ammeter | 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. | High Pressure Coolant Injection Reactor Core Isolation Cooling Primary containment Main Turbine Reactor Recirculation Pump Standby Liquid Control Emergency Core Cooling System Safety Relief Valve Main Condenser Control Rod | |
| ENS 48503 | 12 November 2012 21:51:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Building Isolation with Standby Gas Treatment System Actuation During Radioactive Material Move | At 1551 EST on 11/12/12, the 'A' Refuel Floor Process Radiation Monitor reached 62 mR/hr during movement of the old steam dryer in the plant reactor building. This resulted in the isolation of the drywell containment air monitor and the oxygen analyzer primary containment isolation valves. The signal also resulted in a reactor building isolation (Secondary Containment), start of 'A' Standby Gas Treatment, and transfer of the control room ventilation to the High Radiation Mode. All automatic isolation valves have been reset. Reactor building and control room ventilation have been reset. Standby gas treatment has been secured. There were no challenges to the health and safety of the general public. The NRC Resident Inspector has been notified. | Secondary containment Primary containment Standby Gas Treatment System | |
| ENS 49066 | 24 May 2013 08:34:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of Emergency Diesel Generator Due to Bus Undervoltage | At 0334 (CDT) on 5/24/2013 MNGP (Monticello Nuclear Generating Plant) experienced a loss of power to Bus 15 (Division 1 4kV Essential Bus) during performance of preoperational testing on the 2R reserve transformer which initiated an Essential Bus Transfer of Bus 15 and automatic start of 12 Emergency Diesel Generator. MNGP was in Mode 5 operations with water level >21 feet 11 inches above the top of the RPV flange and all credited safety systems were lined up to Bus 16 (Division 2 4kV Essential Bus) which was unaffected by this event. Bus 15 was automatically repowered from the 1AR reserve transformer as designed. During this evolution all critical safety functions remained green and all systems responded as expected to the Essential Bus transfer. The cause of the sequence of events that led to the Bus 15 loss of power is being investigated. This event is reportable under 10CFR50.72(b)(3)(iv) as an event that results in a valid actuation of 12 Emergency Diesel Generator. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 49113 | 13 June 2013 19:30:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 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)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material | Emergency Diesel Generators Start | While preparing for an equipment test Thursday afternoon, Monticello Nuclear Generating Plant lost off-site power on its normal off-site power feed. Power for safety related loads was automatically transferred to the alternate off-site power source. The Emergency Diesel Generators started as designed but did not load onto the safety related busses due to the availability of off-site power. Operators stabilized the plant, which is shutdown for a refueling and maintenance outage, in less than an hour and are investigating the cause of the event. The current plant focus is on restoring the normal off-site power feed. The event posed no danger to the public or plant workers, and no one was injured. There was no release of radiation. Plant safety systems continue to be powered by the backup off-site power feed, with the emergency diesel generators available if needed. Event Specifics: At approximately 1430 CDT, during a refueling outage with the plant in Mode 4, reactor level at approximately 200 inches, and a full Scram already inserted, a loss of normal off-site power occurred due to a fault in a non-safety related bus supply breaker. The fault was in the 13.8 KV supply breaker to the #11 bus. This caused the Station 2R transformer to lockout, resulting in a loss of the normal off-site power to Essential Busses 15 and 16. Shutdown Cooling (SDC) was lost for approximately 1 hour due to loss of supply power and isolation of the common suction valves. Both 11 and 12 Emergency Diesel Generators (EDGs) automatically started but did not load onto their respective busses (as designed) due to the 1AR emergency off-site transformer re-energizing both 15 and 16 bus. This essential bus transfer is being reported as a 'Valid actuation of emergency AC electrical power systems' under 10CFR50.72(b)(3)(iv). During the event the decision was made to shut down the EDGs which rendered them inoperable for a short period of time until the Fast Start capability was reset. The period of time that the EDGs were inoperable is being reported as a 'Condition that could have prevented the fulfillment of the safety functions to remove residual heat, control the release of radioactive material, and mitigate the consequences of an accident under 10CFR50.72(b)(3)(v)(B), (C), and (D). Both EDGs have been restored to Automatic Standby Status and are operable. The loss of power resulted in a Group II Containment Isolation signal causing secondary containment to isolate and Standby Gas Treatment and Control Room Emergency Filtration to initiate as well as associated Group II Containment Isolation Valves to close. This is being reported as a 'General containment isolation signal ESF actuation' under 10CFR50.72(b)(3)(iv). The containment isolation has been reset, and SDC and SFPC have been restored. Reactor temperature rose approximately 4 degrees F during the event from 161 degrees to 165 degrees which remained in the prescribed operating band. Reactor level did not change. The licensee has notified the NRC Resident Inspector. | Secondary containment Emergency Diesel Generator Shutdown Cooling | |
| ENS 51560 | 23 November 2015 16:40:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to a Reactor Recirculation Pump Lockout | At 1040 CST, with the plant at 100% power, a lockout of the 11 recirculation pump occurred. Following the 11 recirculation pump lockout, at 1041 CST, a reactor scram and a Group 1 isolation occurred. All Main Steam Isolation Valves closed as a result of the Group 1 isolation signal. HPCI (High Pressure Core Injection) has been placed in service to control RPV (Reactor Pressure Vessel) pressure. HPCI did not inject into the RPV and was not needed to control RPV level. At 1104 CST, a Group 2 containment isolation signal was received due to RPV level less than +9 inches. The Group 2 isolation signal has been reset. The cause(s) of the 11 recirculation pump lockout, the reactor scram, and the Group 1 isolation are currently not known and are under investigation. This event is being reported under 50.72(b)(2)(iv)(B) due to the actuation of the Reactor Protection System when the reactor is critical. For the following reasons, this event is also being reported under 50.72(b)(3)(iv)(A): 1) This event resulted in a valid Group 2 containment isolation signal, 2) Since the cause of the Reactor Protection System actuation is not known, the event is being reported as a valid actuation of the Reactor Protection System, and 3) Since the cause of the Group 1 isolation is not known, the event is being reported as a valid primary containment isolation signal affecting multiple Main Steam Isolation Valves. All systems have responded as expected, all control rods fully inserted following the Reactor Protection System actuation. The plant is currently shutdown in mode 3, RPV pressure and RPV level are stable. This event did not result in any radiological release from the plant. This event did not challenge the health and safety of the public. The NRC Resident Inspector has been notified. The plant is in its normal shutdown electrical lineup. HPCI is in pressure suppression mode with RHR cooling the suppression pool. | Reactor Protection System Main Steam Isolation Valve Primary containment Reactor Recirculation Pump Control Rod | |
| ENS 52682 | 15 April 2017 09:41:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Reactor Protection System and Partial Primary Containment Isolation System Actuations on Low Water Level | During shutdown activities with the reactor subcritical, actions were being taken to remove 11 Reactor Feed Pump from service in support of a scheduled refueling outage. Reactor Water Level on Safeguards level instrumentation dropped below +9 inches, which resulted in a valid Reactor Protection System (RPS) Scram signal and Partial Group 2 Primary Containment Isolation System (PCIS) signal. All systems functioned as required. Reactor Water Level on Safeguards instrumentation was restored to greater than +9 inches immediately. RPS and PCIS logic was reset. There was no impact to the health and safety of the public as a result of this event. This actuation of these systems is being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of any of the systems listed in 10 CFR 50.72(b)(3)(iv)(B). The NRC Resident Inspector has been notified. | Reactor Protection System Primary Containment Isolation System | |
| ENS 56759 | 27 September 2023 15:41:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Actuation of Reactor Protection and Containment Isolation Systems | The following information was provided by the licensee via fax: (On 09/27/2023) at 1041 CDT, with the plant at 75 percent power and main turbine control valve testing in progress, a reactor pressure transient resulted in a reactor steam dome high pressure scram and subsequent group 1 primary containment isolation of the main steam lines (MSL). All main steam isolation valves closed as a result of the group 1 isolation signal. Additionally, a group 2 containment isolation signal was received due to reactor pressure vessel (RPV) level less than plus 9 inches during the transient. Operations personnel responded and stabilized the plant. The high-pressure coolant injection (HPCI) system was placed in service to control RPV pressure. HPCI did not inject into the RPV and was not needed to control RPV water level. The cause of the initial pressure transient is under investigation. The NRC Resident Inspector has been notified.
The utility notified the State of Minnesota and Wright and Sherburne counties. Notified R3DO (Orlikowski) | Main Steam Isolation Valve Primary containment Reactor Pressure Vessel Main Steam Line | |
| ENS 56995 | 28 February 2024 14:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram | The following information was provided by the licensee via fax and email: At approximately 0839 (CST) with Unit 1 in Mode 1 at 100 percent power, the reactor automatically scrammed due to the depressurization of the SCRAM air header caused by an invalid signal that (occurred) during system testing. The SCRAM was uncomplicated with all systems responding as expected. The cause and details of the event are under investigation. Containment isolation valves actuated and closed on a valid Group 2 signal. 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, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group 2 isolation signal. Operations responded using the emergency operating procedure and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. State as well as Wright and Sherburne Counties will be notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Anticipated Transient Without Scram (ATWS) circuit was being tested when an invalid signal was sent to depressurize the SCRAM air header. | Reactor Protection System Main Condenser |