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| Entered date | Event description | |
|---|---|---|
| ENS 52781 | 1 June 2017 23:41:00 | Planned maintenance to restore normal power to Plant Computer Systems resulted in an unexpected loss of all Meteorological (MET) Tower Data (at 1645 CDT). As a result, this represents a Loss of Emergency Assessment Capability and is reportable under 10CFR 50.72 (b)(3)(xiii). The isolation was restored and MET Tower Data was restored at 1845. The health and safety of the public was not affected as the plant is operating in a normal condition with no severe weather or storms in the area. Additionally meteorological data was available from the National Weather Service should this data had been necessary. The NRC Resident Inspector has been notified." The licensee will be notifying the State of Minnesota. |
| ENS 50806 | 10 February 2015 20:39:00 | On February 10, 2015, at 1240 EST, Northern States Power-Minnesota (NSPM) determined that the Station Blackout (SBO) implementation at Monticello Nuclear Generating Plant (MNGP) was not consistent with the NRC Safety Evaluation (SE). Specifically, the High Pressure Coolant Injection (HPCI) system was not being utilized in a manner consistent with the NRC SE for SBO. Current battery calculations do not reflect a full complement of HPCI system equipment running for the duration (coping requirements) of the SBO event. The calculation assumed a manual action to remove the HPCI auxiliary oil pump from operating during an SBO event in order to preserve the station battery. NSPM is reporting this as an Unanalyzed Condition pursuant to the requirements of 10 CFR 50.72(b)(3)(ii)(B). The health and safety of the public was not affected since no SBO event occurred. All station batteries and the HPCI system remain operable in accordance with the plant Technical Specifications. The NRC Resident Inspector was notified of the event.
An engineering analysis was performed updating the battery calculations for Station Blackout (SBO) implementation demonstrating the ability of the safety related station batteries to provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of the SBO for the specified four hours. Therefore, the battery calculation is analyzed and specifically the High Pressure Cooling Injection (HPCI) System is analyzed to run in automatic for the entire duration of the SBO event meeting the site licensing basis for SBO. The SBO procedure has been revised to incorporate HPCI running in automatic for the entire duration of the SBO event. The NRC Resident Inspector has been notified. Notified R3DO (Duncan) |
| ENS 50496 | 26 September 2014 02:53:00 | At 2200 CDT on September 25, 2014, the Duty Shift Manager was notified that Agastat relays associated with Primary Containment Isolation valves on the Hydrogen-Oxygen Analyzing System are beyond the analyzed shelf life for relays that are in the normally energized state and are considered INOPERABLE. This affected both primary containment isolation valves for a containment penetration on multiple flow paths. This issue was determined to be reportable under (10 CFR) 50.72 (b)(3)(v)(C) & (D) for an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material and mitigate the consequences of an accident. Additionally, the required actions involved isolating six flow paths via manual isolation valves. This action rendered the Hydrogen-Oxygen Analyzers non-functional for both trains and constitutes a loss of Emergency Preparedness and Accident Assessment Capability. This is reportable under (10 CFR) 50. 72(b)(3)(xiii). The Primary Containment Isolation Valves have been, and remain, in their closed position to satisfy their Primary Containment Function and protect the health and safety of the public. The NRC Senior Resident Inspector has been notified. The licensee will notify the State of Minnesota. The relays of concern were manufactured 19 years ago and have been in operation for 11 years, versus a manufacturer assumption of a 10 year operational lifespan.
Partial retraction for EN 50496. This is an update of Emergency Notification System (ENS) report 50496 that was submitted at 0253 EDT on Friday, September 26, 2014. ENS notification was made due to four relays associated with the sampling valves on the Hydrogen-Oxygen Analyzing (HOA) system that perform Primary Containment Isolation Valve (PCIV) functions. These relays were discovered installed beyond their manufacturer qualified service life, which called operability into question. The portions 10 CFR 50.72 (b)(3)(v)(C) & (D) are being retracted after subsequent bench testing and investigation of system operability. Based on the past operability evaluation, all four relays associated with PCIV functions on the HOA system would have performed their specific safety function of primary containment isolation, as required by the facility's technical specifications. Therefore, this event does not meet the threshold of an event or condition that would prevent fulfilment of a safety function. The loss of emergency preparedness and accident assessment capability previously reported under 10 CFR 50.72 (b)(3)(xiii) remains unchanged. The NRC Resident Inspector has been notified. Notified R3DO (Peterson). |
| ENS 50026 | 10 April 2014 22:40:00 | Monticello Nuclear Generating Plant personnel discovered the remains of what appeared to be a deceased duck on plant property. The cause of death was not immediately apparent, no work was ongoing within the vicinity at the time. Notifications to the Minnesota Department of Natural Resources and the Division of Fish and Wildlife will be made for this discovery. This event is reported per 10CFR50.72(b)(2)(xi). The licensee has notified the NRC Senior Resident Inspector. Plant personnel could not determine if the duck was an endangered species. |
| ENS 49113 | 13 June 2013 22:02:00 | 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. |
| ENS 48190 | 15 August 2012 04:37:00 | At 2045 (CDT) on 8/14/12, MNGP (Monticello Nuclear Generating Plant) Operations determined that valves RHR-82 and RHR-84 had been inappropriately closed as part of an isolation clearance order for work on shutdown cooling suction piping. These valves are required to be open to provide overpressure protection for RHR piping passing through primary containment penetration X-12. Upon discovery of the condition, Primary Containment was declared Inoperable and the Required Actions of Tech Spec 3.6.1.1 were entered. Following discovery, the isolation was restored and the valves opened. At 0001 (CDT) on 8/15/12, Primary Containment was declared Operable. This issue is being reported in accordance with 10CFR50.72(b)(3)(v)(C) and 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety functions of a system needed to control the release of radioactive material or to mitigate the consequences of an accident. The MNGP Senior NRC Resident Inspector has been notified of this issue. The licensee will contact the Minnesota State Duty Officer.
This notification is a retraction of ENS 48190 based on further engineering evaluation. Monticello had previously evaluated penetration X-12 for thermally induced over pressurization. The evaluation qualified the piping components in the penetration for a maximum pressure of 3,306 psig using ASME Section III Appendix F operability criteria. The peak pressure calculated for the penetration was 2,743 psig based on Reactor pressure of 1000 psig with Reactor in Mode 1, and at worse case LOCA conditions for the Drywell. These assumptions and parameters envelop those that were present when valves RHR-82 and RHR-84 were closed on August 14, 2012. Therefore, this event would not have prevented the fulfillment of the safety function reported. The NRC Resident Inspector has been notified. Notified R3DO (Duncan). |
| ENS 47479 | 28 November 2011 01:04:00 | After transitioning to Mode 2 from Mode 4, while performing the Rod Worth Minimizer (RWM) operability test, it was discovered that the RWM control switch was in the BYPASS position. The RWM enforces predetermined control rod withdrawal and insertion sequences. Complying with these predetermined sequences ensures a Control Rod Drop Accident does not exceed analytical limits. With the control switch in the BYPASS position, the RWM was inoperable and would not have enforced the predetermined control rod withdrawal sequence. The RWM control switch was restored to the OPERATE position and the RWM was verified to be operable. This issue is being reported under 50.72(b)(3)(v)(D) as an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of the RWM, which is a system needed to mitigate the consequences of the Control Rod Drop Accident. The licensee will be notifying the NRC Resident Inspector. |