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 Entered dateSiteRegionReactor typeEvent description
ENS 5411111 June 2019 17:55:00MonticelloNRC Region 3At 1132 CDT on 6/11/2019, both manual primary containment isolation valves in a one-inch service air line were found open. This resulted in an open primary containment penetration. Both valves are required to be closed for Primary Containment Isolation Valve Operability. Both valves were closed and independently verified closed at 1149 CDT on 6/11/2019. This is being reported under 10 CFR 50.72(b)(3)(v)(C) and (D), and 10 CFR 50.72(b)(3)(ii)(B). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee also notified the State of Minnesota State Duty Officer.
ENS 5400015 April 2019 11:36:00MonticelloNRC Region 3At 0511 CDT on 4/15/2019, transport of a potentially radiologically contaminated person from the Monticello Nuclear Plant to a local hospital was performed prior to conducting a radiological survey as a prudent measure to ensure timely medical support. At 0658 CDT a radiological survey determined that the individual and their clothing were not contaminated. This is reportable under 10 CFR 50.72(b)(3)(xii). The NRC Resident Inspector has been notified."
ENS 5399713 April 2019 02:04:00MonticelloNRC Region 3

EN Revision Text: HIGH ENERGY LINE BREAK DOOR FOUND IN INCORRECT POSITION RESULTING IN LPCI AND CORE SPRAY BEING INOPERABLE At approximately 1815 CDT on April 12, 2019, High Energy Line Break (HELB) Door-410A in the Reactor Building was discovered in the closed position. HELB Door-410B was previously closed for maintenance. Either Door-410A or Door-410B must be open to support the current HELB analyses. With both doors closed, this is considered an unanalyzed condition resulting in the loss of a post-HELB safe shutdown path. With Door-410A and Door-410B closed, LPCI (Low Pressure Coolant Injection) and Core Spray injection valves in both divisions are no longer considered available. This condition is being reported under 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function. The condition was resolved at approximately 1845 CDT on April 12, 2019 when Door-410A was blocked open. The health and safety of the public was not affected by this condition. The NRC Resident has been notified.

  • * * RETRACTION FROM JESSE TYGUM TO HOWIE CROUCH AT 1330 EDT ON 5/24/19 * * *

Event Notification (EN) #53997, made on 4/13/2019, is being retracted. An engineering evaluation completed subsequent to this event analyzed the discovered condition with both Door-410A and Door-410B being closed. The engineering evaluation determined that the environmental conditions present with both Door-410A and Door-410B closed would not have impacted the availability of both divisions of the LPCI (Low Pressure Coolant Injection) and Core Spray injection valves nor would it have resulted in the loss of a post-HELB safe shutdown path. Therefore, this condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii) as an unanalyzed condition that significantly degrades plant safety or per 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified. The licensee also notified the Minnesota State Duty Officer. Notified R3DO (Cameron).

ENS 5385331 January 2019 10:48:00Prairie IslandNRC Region 3

EN Revision Text: BOTH EMERGENCY DIESEL GENERATORS INOPERABLE DUE TO LOW AIR TEMPERATURE At 0743 (CST) on 1/31/2019, both trains of Unit 2 Diesel Generators were declared INOPERABLE due to outside air temperature exceeding the low temperature design limit for the diesel engines; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v) for an event or condition that could have prevented the fulfillment of a safety function. The Unit 2 Diesel Generators are still able to start if necessary to provide power. Additionally, multiple layers of defense in depth measures are in place to ensure safety. Prairie Island has five sources of offsite power; all of which are currently available. The Unit 1 Diesel Generators are OPERABLE and capable of being cross-connected to Unit 2. Additional equipment capable of responding to beyond design basis events is available on site providing another layer of defense in depth. Both Unit 2 Diesel Generators were returned to an OPERABLE status at 0810 on 1/31/2019 based on outside air temperature rising above the low temperature design limit with forecasted temperatures to remain above the low temperature design limit. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The air temperature limit was -30 degrees Fahrenheit. Unit 1 was not affected. The EDGs were supplied by a different manufacturer with different air temperature limits.

  • * * RETRACTION AT 1340 EDT ON 03/22/2019 FROM BRIAN JOHNSON TO JEFFREY WHITED * * *

Engineering analysis performed subsequent to the event notification has determined that both Unit 2 Diesel Generators would have been able to fulfill their safety function during the period of time when the outside air temperature had exceeded the low temperature design limit. Therefore, EN# 53853 is being retracted. The NRC Resident Inspector has been notified of the event notification retraction. Notified R3DO (McCraw).

ENS 5375728 November 2018 12:30:00Prairie IslandNRC Region 3At 0752 CST, on November 28, 2018, Dakota County inadvertently actuated their sirens while performing a scheduled weekly (Emergency Planning Fixed Siren Test). All seven (7) Dakota County sirens actuated for approximately 9 seconds before Dakota County Dispatch canceled the activation. This 4-hour non-emergency report is being made per 10 CFR 50.72(b)(2)(xi), Offsite Notification (which was made to Dakota County Dispatch). Capability to notify the public was never degraded during this time. All Emergency Notification sirens remain in service. No press release is planned at this time. The licensee has notified the NRC Resident Inspector.
ENS 5370028 October 2018 21:44:00Prairie IslandNRC Region 3

This event is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) for a major loss of emergency assessment capability at the Prairie Island Nuclear Generating Plant. At 1435 CDT on October 28, 2018, troubleshooting of the Seismic Monitoring Panel resulting from the receipt of Control Room annunciator 47023-0603 (Seismic Monitor Panel) determined that the '(Operational Basis Earthquake) OBE Exceedance' alarm on the Seismic Monitoring Panel will not alarm and determined the panel is non-functional. The Seismic Monitoring Panel system functions to provide indication that the OBE threshold has been exceeded following a seismic event and is used in the Prairie Island Nuclear Generating Plant Emergency Plan to perform classification of Initiating Condition 'Seismic event greater than OBE levels' and Emergency Action Level HU2.1. Station personnel are monitoring the seismic recorders for event alarms on a 15 minute frequency due to alarm function failure. The station is developing repair plans for restoration of the alarm function. This event does not adversely affect the safe operation of the plant or health and safety of the public.

The licensee has notified the NRC Resident Inspector."

ENS 5359711 September 2018 18:42:00MonticelloNRC Region 3On 9/10/2018, the 11 Core Spray (CSP) loop was placed in service to support quarterly surveillance testing. With the 11 CSP pump in service it was identified that the check valves isolating the 11 CSP system from the keep fill supply were leaking by. At 1129 CDT on 9/11/2018, it was identified that this leakage may have exceeded the leakage rate assumptions made in the dose analysis calculation for emergency core cooling system (ECCS) leakage outside containment following a loss of coolant accident (LOCA). Therefore, this is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades plant safety and 10 CFR 50.72(b)(3)(v) for any 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: (C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The potential ECCS leak pathway has been isolated. There is no impact to health and safety of the public. The NRC Resident Inspector has been notified."
ENS 5354610 August 2018 23:13:00Prairie IslandNRC Region 3

On 8/10/2018 at 1445 (CDT) both trains of Cooling Water (Cooling Water Pumps for Emergency Diesel Generators) were declared INOPERABLE and both units entered (Technical Specification) (TS) 3.0.3 due to corroded jacket cooling water plugs for (the) 12 and 22 cooling water pump motors; therefore this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). At 1543 (CDT), 08/10/2018 the 121 Cooling Water pump was aligned to the "A" Cooling Water train and the TS 3.0.3 condition was exited for both units. (After restoring train A cooling water the site entered a seven day limiting condition for operations, TS 3.7.8 for one inoperable cooling water pump.) There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 09/29/2018 AT 2128 EDT FROM BRIAN JOHNSON TO OSSY FONT * * *

Testing and forensic analysis performed subsequent to the notification has determined the as-found condition would not have impacted either diesel-driven pumps' ability to start, run, and meet flow/pressure requirements to perform their required safety function. Therefore, EN# 53546 is being retracted. The NRC Resident Inspector has been notified of the Event Notification retraction. Notified R3DO (Kozak).

ENS 5328624 March 2018 01:09:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopThis report is made for a loss of Emergency Assessment Capability associated with Emergency Action Levels for Toxic and Flammable Gas and is reportable under 10 CFR 50.72 (b)(3)(xiii). During an emergency equipment inventory it was identified that methods were not available to detect levels of toxic or flammable gas at the IDLH (Immediately Dangerous to Life and Health) level for a number of substances. The IDLH is used to assess the Alert Emergency Action Level. The ability of the Control Room Staff to detect and respond to the presence of toxic or flammable gas is unaffected. Because there have been no chemical spills or releases that would require sampling to be performed, the health and safety of the public was not affected. The NRC Resident Inspector has been notified. The State of Minnesota will be notified.
ENS 5328523 March 2018 21:07:00MonticelloNRC Region 3GE-3This report is made for a loss of Emergency Assessment Capability associated with Emergency Action Levels for Toxic and Flammable Gas and is reportable under 10 CFR 50.72(b)(3)(xiii). During an emergency equipment inventory, it was identified that methods were not available to detect levels of toxic or flammable gas at the IDLH (Immediately Dangerous to Life and Health) level for a number of substances due to the detector having an unsuitable range. The IDLH is used to assess the Emergency Action Level Alert Range. The ability of the Control Room Staff to detect and respond to the presence of toxic or flammable gas is unaffected. Because there have been no chemical spills or releases that would require sampling to be performed, the health and safety of the public was not affected. The resident NRC Inspector has been notified. The licensee will be notifying the state of Minnesota.
ENS 5313522 December 2017 11:09:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 2050 (CDT) on October 25, 2017, during the Unit 2 refueling outage, with no fuel in the Reactor Vessel, control room operators found that Unit 1 Train B Containment Fan Coil Units (FCUs) had swapped from chilled water to cooling water (CL). Construction Electricians were installing a new relay 2Sl-22X when the plunger on adjacent relay 2Sl-23X was bumped, which caused the swap of the Unit 1 Containment Fan Coil Units (CFCUs) from chilled water to cooling water. Relay 2SI-23X is a slave relay that starts 22 Turbine Driven Auxiliary Feed Water Pump, illuminates blue lights on various control switches, closes MV-32159 Loop A/B CLG WTR HDR XOVR MV B, closes chilled water Isolation Valves to Unit 1 Train B, and closes chilled water Isolation Valves to Unit 2 Train B. This actuation was as expected. CL is a shared system and, upon a Safety Injection (SI) signal on either unit, the CL header splits into two trains and, as a result, the CL supply is isolated to the chillers that supply chilled water to both units' CFCUs. By design, CL is the safety related source of cooling to the CFCUs. 22 Turbine Driven Auxiliary Feed Water Pump did not start as the unit was in 'No Mode' with the control switch for the pump in manual. MV-32159 did automatically close per design. Unit 2 Chilled Water was already isolated due to work in progress with the unit in 'No Mode.' There was no impact on public health and safety. This event is being reported as a 60-day telephone notification in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an invalid actuation of a containment heat removal system (the FCUs were running, but were swapped to their safeguards source due to an invalid actuation of a relay). The licensee notified the NRC Resident Inspector.
ENS 5313422 December 2017 11:09:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 0856 (CDT) on October 23, 2017, during the Unit 2 refueling outage, with no fuel in the Reactor Vessel, an unexpected auto start of the Unit 2 Train B Emergency Diesel Generator (D6) occurred when construction electricians inadvertently bumped the plunger for relay 2Sl-20X while working in the relay rack. Relay 2Sl-20X is a slave relay that actuates a light on the control board, starts D6, and starts 22 Residual Heat Removal (RHR) pump on a Safety Injection signal. In this instance, the RHR pump did not start as its control switch was in pull-out. It is expected that the control board light lit for the brief time the relay plunger was depressed, but this could not be confirmed. The D6 actuation resulted in an unexpected annunciator for D6 EMERGENCY GENERATOR SI SIGNAL EMERGENCY START. Operators responded per the alarm response procedure, performed a walk down of running D6 and then performed a shutdown of D6. D6 started and functioned as expected. There was no impact on public health and safety. This event is being reported as a 60-day telephone notification in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an invalid actuation of an emergency diesel generator. The NRC Resident Inspector has been notified.
ENS 5306713 November 2017 03:57:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 2119 (CST) on 11/12/2017 a Control Room board walk down discovered that both of the Unit 2 Containment Spray Pump control switches were in pull-out. With the control switches in pull-out, the pumps would not automatically start as required. Unplanned TS (Technical Specifications) 3.0.3 was entered at 2119 as a result of not complying with TS 3.6.5, Containment Spray and Cooling Systems, which requires both trains of Containment Spray to be Operable while in Mode 4. Unit 2 had entered Mode 4 at 0303 on 11/12/2017. TS 3.0.3 was exited at 2127 on 11/12/2017 when both Containment Spray Pump control switches were placed in Automatic restoring Operability. Preliminary investigation determined that while Unit 2 was in Mode 5, Surveillance SP 2099, Main Steam Isolation Valve Logic Test, had taken the Containment Spray Pump control switches to pull-out but did not re-align the control switches to automatic after the test was complete. This 8-hour Non-Emergency report is being made per 10 CFR 50.72(b)(3)(v)(D), Accident Mitigation. The NRC Senior Resident Inspector has been informed.
ENS 5301916 October 2017 22:09:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 1425 CDT on 10/16/17, investigation into a boric acid indication was determined to be a through wall leak at the socket weld that joins the 3/4 inch line 2RC-92 to valve 2RC-8-37. Unit 2 is currently in Mode 5 with Reactor Coolant system (RCS) Operational Leakage limits not applicable. The leak is downstream of two first off RCS isolation valves that are normally closed. The leak is not quantifiable as it only consists of a small amount of dry boric acid at the location. This failure constitutes welding or material defects in the primary coolant system that cannot be found acceptable under ASME Section Xl. Therefore, this is a degraded condition reportable under 10 CFR 50.72(b)(3)(ii)(A). At the time of this notification, the Prairie Island Nuclear Generating Plant Unit 2 is in Mode 5 for a planned refueling outage. The identified defect will be repaired prior to entering Mode 4. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified.
ENS 5281420 June 2017 08:18:00MonticelloNRC Region 3GE-3At 2353 CDT on 6/19/2017, while performing the High Pressure Coolant Injection (HPCI) quarterly surveillance following planned maintenance, the HPCI turbine did not start as expected due to the HPCI turbine stop valve failing to open. This issue is being reported under 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. Investigation into the failure of the HPCI system to start is in progress. The unit remains at 100% power. The health and safety of the public was not affected. The NRC Resident Inspector has been notified.
ENS 527811 June 2017 23:41:00MonticelloNRC Region 3GE-3Planned 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 5274911 May 2017 18:11:00MonticelloNRC Region 3GE-3A can of alcohol (16.9 ounce foreign beer) was discovered unopened in an administration building refrigerator. Site security took possession of the can of alcohol. The owner of the can of alcohol is unknown. This licensee is making this 24 hour notification in accordance with 10CFR26.719(b)(1). The licensee notified the NRC Resident Inspector.
ENS 5274210 May 2017 11:46:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt approximately 0755 CDT, on May 10, 2017, Pierce County inadvertently actuated their sirens while performing a scheduled weekly cancel test. All fifty two (52) Pierce County sirens actuated county wide for approximately 11 seconds before Pierce County Dispatch canceled the activation. This 4-hour non-emergency report is being made per 10 CFR 50.72(b)(2)(xi), Offsite Notification. Capability to notify the public was never degraded during this time. All Emergency Notification sirens remain in service. No press release is planned at this time. The license has notified the NRC Senior Resident Inspector.
ENS 5271528 April 2017 21:21:00MonticelloNRC Region 3GE-3This report is being made pursuant to 10 CFR 50.72(b)(2)(xi), as an event where notification to other government agencies has been made. On April 28, 2017, notification to the Minnesota State Duty Office was made due to a non-compliance with release of wastewater requirements in the Monticello Nuclear Generating Plant's National Pollutant Discharge Elimination System permit. There were no consequences to the health and safety of the public. The licensee notified the NRC Resident Inspector.
ENS 5268215 April 2017 13:03:00MonticelloNRC Region 3GE-3During 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.
ENS 5263824 March 2017 17:15:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

On February 3, 2017, Prairie Island staff performed maintenance on the transom above Battery Room Door 225. This activity resulted in the transom being unlatched for approximately five minutes. On February 6, 2017, a question from the NRC Resident Inspector resulted in an evaluation of this condition for past operability. On March 20, 2017, the past operability evaluation of Door 225 concluded that, in the event of a postulated HELB (High Energy Line Break), the transom being unlatched during the five minute maintenance period resulted in the inoperability of multiple systems in the Unit 1 and Unit 2 battery, auxiliary feedwater, and Unit 1 safeguards bus rooms that would be required to mitigate the postulated HELB. The loss of safety functions required to mitigate the postulated HELB make the condition reportable under 50.72(b)(3)(ii) for an unanalyzed condition that significantly degrades plant safety. Unlatching the transom above the Battery Room Door creates an opening not accounted for in design bases documents. This occurred due to an improperly prepared work permit. Corrective actions are in place to preclude recurrence. The licensee informed the NRC Resident Inspector.

  • * * RETRACTION FROM MARK LOOSBROCK TO JEFF ROTTON AT 1559 EDT ON 04/10/2017 * * *

Further analysis determined that an unlatched transom would result in a relative humidity of 100 percent in 11 Battery Room for about 10 minutes following a postulated HELB. Since the equipment in the Battery Rooms is not qualified for a harsh environment, the components in 11 Battery Room would have been inoperable. Temperature and relative humidity in the other Battery Rooms, Auxiliary Feedwater Rooms, and the Unit 1 Safeguards Bus Rooms would have remained within the allowable limits. Therefore, for the five minutes the strike was removed from the transom, only equipment in 11 Battery Room and supported A Train components would have been inoperable. This event was not an Unanalyzed Condition that significantly degraded plant safety, under 10 CFR 50.72(b)(3)(ii), as no safety function would have been lost. The licensee notified the NRC Resident Inspector. Notified R3DO (Skokowski).

ENS 524744 January 2017 20:07:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopA non-licensed employee supervisor had a confirmed positive for a prohibited substance during a random fitness-for-duty test. The individual's unescorted access to the plant has been denied. The licensee notified the NRC Resident Inspector.
ENS 5245421 December 2016 18:00:00MonticelloNRC Region 3GE-3

At 0935 (CST) on 12/21/2016, while performing the High Pressure Coolant Injection (HPCI) Comprehensive Pump and Valve Tests for post-maintenance testing following scheduled maintenance, the HPCI turbine did not start as expected due to the HPCI turbine stop valve failing to open. This issue is being reported under 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. Investigation into the failure of the HPCI system to start is in progress. The plant remains at 100% power with no challenges to the health and safety of the public. The NRC Resident Inspector has been notified. The plant is in a 14-day action statement under LCO 3.5.1, 'ECCS - Operating' due to the HPCI turbine stop valve failure. The licensee notified the Minnesota State Duty Officer.

  • * * RETRACTION FROM KIM HOFFMAN TO JOHN SHOEMAKER AT 1303 EST ON 1/17/18/17 * * *

On December 21, 2016, the NRC Operations Center was notified of Event Number 52454 that described a failure of the High Pressure Coolant Injection (HPCI) turbine stop valve to open during post maintenance testing prior to being declared operable. The condition was reported in accordance with 10 CFR 50.72 (b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. At the time, it was not readily apparent that the failure was due to the maintenance activities. Subsequent return-to-service testing showed the oil system vent and fill had been inadequate following the maintenance. This event occurred as a result of the maintenance process and would not have occurred during normal operation of the system. NUREG-1022, Revision 3 states, 'reports are not required when systems are declared inoperable as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that would have resulted in the system being declared inoperable).' There was no discovered condition that would have resulted in the safety function of the system being declared inoperable under normal, non-maintenance conditions. Based on the above additional information, Monticello Nuclear Generating Plant is retracting this report. The plant was in a planned evolution and did not discover a condition that could have prevented performing a safety function. The licensee has notified the NRC Resident Inspector. Notified R3DO (McCraw).

ENS 5245121 December 2016 16:38:00MonticelloNRC Region 3GE-3This 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to report an invalid actuation of the 12 Emergency Diesel Generator Emergency Service Water pump (12 ESW pump). At 1745 (CST) on October 24, 2016, an unexpected auto-start of the 12 ESW pump occurred. The 12 Emergency Diesel Generator (12 EDG), was previously properly removed from service and isolated for scheduled maintenance. Upon investigation, is was determined that no valid start signal was present and actuation occurred during relay replacement activities on the 12 EDG in C-92 (12 EDG (G-38) electrical control panel) cabinet when electricians inadvertently bumped a 12 EDG start relay. During this period, the Control Room received annunciators indicating the 12 EDG engine was running/cranking and the 12 ESW pump started. Due to being isolated, the 12 EDG did not actually start. The licensee notified the NRC Resident Inspector.
ENS 5239627 November 2016 21:22:00MonticelloNRC Region 3GE-3At 1447 (CST) on 11/27/2016 while troubleshooting a minor leak on the High Pressure Coolant Injection (HPCI) turbine, it was discovered that the HPCI turbine exhaust drain pot high level bypass switch was not functioning per design to support removal of condensate from the HPCI turbine casing. This resulted in some water accumulation within the HPCI turbine casing. Subsequently, HPCI was declared INOPERABLE and this issue is being reported under 10CFR50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of a safety function at the time of discovery. The plant remains at 100 percent power with no challenges to the health and safety of the public. The NRC Resident Inspector has been notified. Technical Specification limiting condition for operation requires HPCI to be Operable within 14 days. The licensee will be notifying the State of Minnesota regarding the event.
ENS 5217814 August 2016 01:59:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

Prairie Island Unit 1 declared an Unusual Event at 2359 CDT on 8/13/2016 based on Reactor Coolant System (RCS) identified leakage being greater than 25 gpm. The RCS leakage was 40 gpm for three (3) minutes. The RCS leakage was stopped when letdown flow was isolated. Minimum charging flow has been established and Excess Letdown was placed in service. Prairie Island Unit 1 is currently stable and continues to operate at 100 percent power. There was no impact on Prairie Island Unit 2. CV-31339 (Letdown Line Containment Isolation Valve) failed closed. VC-26-1 (Regenerative Heat Exchanger Letdown Line Outlet Relief to Pressurizer Relief Tank (PRT)) lifted with 40 gallons per minute to the PRT for three (3) Minutes. Operators entered procedure 1C12.1 AOP3, Loss of Letdown Flow to VCT. Letdown was isolated per 1C12.1 AOP3, relief valve VC-26-1 reseated and leakage to the PRT stopped. Charging flow was reduced to one (1) charging pump at minimum speed (16 GPM). Excess letdown was placed in service to maintain pressurizer level between 32 - 34 percent. The cause for CV-31339 closing has not yet been determined. The NRC Resident Inspector has been notified. Notified DHS SWO, FEMA Ops Center, DHS NICC. Notified FEMA National Watch and Nuclear SSA via E-mail.

  • * * UPDATE FROM PAUL FINHOLM TO DONALD NORWOOD AT 0525 EDT ON 8/14/2016 * * *

At 0329 CDT the Notice of Unusual Event was terminated based on confirmation that conditions meet all termination criteria. RCS conditions are stable. RCS leakage is less than Technical Specification limits. The current value (of RCS identified leakage) is 0.038 gpm. No classification criteria is currently met. The NRC Resident Inspector has been notified. Notified R3DO (Kozak), NRR EO (Miller), and IRD (Stapleton). Notified DHS SWO, FEMA Ops Center, DHS NICC. Notified FEMA National Watch and Nuclear SSA via E-mail.

ENS 521565 August 2016 13:58:00MonticelloNRC Region 3GE-3On 8/5/2016 at 1014 (CDT), the Monticello Nuclear Generating Plant (MNGP) was notified by the Minnesota Department of Health (MDH) of a notice of violation for exceeding the drinking water limit for carbon tetrachloride in the drinking water well that supplies the Security Access Facility. Additionally the MDH will be notifying the Minnesota Pollution Control Agency regarding the violation. As a result, this issue is being reported under 10CFR50.72(b)(2)(xi) for notifications to other offsite government agencies. There was no impact to the health and safety of the general public as a result of this issue. The drinking fountains in the Security Access Facility have been isolated. The NRC Resident Inspector has been notified.
ENS 521545 August 2016 06:26:00MonticelloNRC Region 3GE-3At 2240 CDT on August 4, 2016, it was discovered that the floor between the cable spreading room and the plant administration building (PAB) basement is not a credited Appendix R fire barrier. Because the cable spreading room and the plant administration building are located in the same fire area, a fire in the PAB could spread to the cable spreading room requiring evacuation of the control room. The travel path used to access the Alternate Shutdown Panel following control room evacuation traverses the same fire area in the PAB. Therefore, this event is being reported under 10 CFR 50.72(b)(3)(ii) for Degraded or Unanalyzed Condition as a fire in the PAB could have the potential to impact Division 1 equipment as well as impede the Operators ability to access Division 2 safe shutdown equipment. Fire watches have been established. There is no impact to the health and safety of the public. The NRC Resident Inspector has been notified. The licensee will notify the State of Minnesota.
ENS 521534 August 2016 22:04:00MonticelloNRC Region 3GE-3

At 1415 CDT on August 4, 2016, while performing a scheduled fire protection surveillance, it was discovered that a component within fire panel FZCP-7, BATTERY ROOM FIRE DETECTION had failed resulting in the inability of the installed fire detectors to detect a fire within the Division 1 and Division 2, 125 VDC battery rooms as well as the Division 2, 250 VDC battery room. This is being reported under 10 CFR 50.72(b)(3)(xiii) for a Loss Of Emergency Assessment Capability as the Control Room would not receive automatic notification of a fire in these areas for evaluation of HU2.1 and HA2.1 for fire within impacted battery rooms which are located within the Protected Area. There is no impact to the health and safety of the public. A 15 minute fire watch has been established for the affected fire zones. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM MARTIN RAJKOWSKI TO DANIEL MILLS AT 1050 EDT ON 08/05/2016 * * *

Event Notification 52153 completed at 2204 EDT on 8/4/2016 shown above contains an error. The failure of FZCP-7, BATTERY ROOM FIRE DETECTION, resulted in the inability to detect a fire within the Division 1 and Division 2 125 VDC battery rooms as well as the Division 1 250 VDC battery room. The Division 2 250 VDC battery room was not affected by this issue. Additionally, the State of Minnesota was notified of this issue. The NRC Resident Inspector has been notified of this update. Notified R3DO (Skokowski)

ENS 5202018 June 2016 06:25:00MonticelloNRC Region 3GE-3On 6/18/16 at approximately 0259 CDT, the Monticello Nuclear Generating Plant was notified by the Wright County Sheriffs Office of a spurious actuation of one emergency siren in the city of Monticello. As a result, this issue is being reported under 10CFR50.72(b)(2)(xi) for notifications to other off site government agencies as the licensee was notified by the Wright County Sheriff's Office. The source of the siren activation has not been determined. Wright County Sheriff's Office successfully deactivated the siren at 0322 CDT. There was no impact to the health and safety of the public as a result of this event as the offsite response capabilities remain functional with a single siren failure. The site is operating normally with no emergency conditions present. The NRC Resident Inspector has been notified. The licensee will notify the State of Minnesota concerning this event.
ENS 5187722 April 2016 00:03:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopMissing fire barrier between Fire Area (FA) 59 and 85. During a walk down of fire barriers for the NFPA 805 project, it was determined that the fire barrier between Fire Area 59 (Unit 1) and 85 (common) is not a rated barrier due to unsealed penetrations in the barrier. Evaluation FPEE 12-006 evaluated the acceptability of the barrier being unrated based on separation of safe shutdown equipment however a review of equipment credited for Appendix R safe shutdown identified that the redundant credited Appendix R equipment is on either side of the fire barrier which is not 3 hour rated. The conclusion of the FPEE is therefore no longer valid. Fire Hazard Analysis Drawings Do Not Match Boundary Description. The plant layout in F5 Appendix F, Rev. 28, Fire Hazard Analysis (FHA), does not agree with the boundary description in the FHA for the Unit 1 and 2 Containment Annulus fire areas, Fire Area (FA) 68 and 72. The layout should but does not show the fire area boundary between the annulus and adjacent fire areas, FA 60 and 75 on 735 (foot) and 61A on 755 (foot), as an Appendix R boundary. The annulus airlock doors are 3-hour fire rated and the airlock is constructed of concrete thick enough to qualify as a 3 hour fire barrier however, there are penetrations in the barrier that are not sealed with fire rated materials or inspected as required by the Fire Protection Program. Therefore, this is an unanalyzed condition reportable under 10 CFR 50.72(b)(3)(ii)(B). This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified.
ENS 5184031 March 2016 16:12:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopOn 3/31/2016 at approximately 0342 CDT, a worker within the Protected Area self-reported a can of beer had been packed in the worker's lunchbox. The worker reported after opening the can and taking a sip it was discovered to be a beer. This event is reportable under 10 CFR 26.719(b)(1). The worker notified Security who immediately escorted the worker from the Protected Area and disposed of the beer. The worker is not an Operator or a Supervisor. The investigation of this event is in progress. The public health and safety are not impacted. The NRC Resident Inspector was notified.
ENS 5181222 March 2016 06:38:00MonticelloNRC Region 3GE-3On 3/22/2016 during performance of HPCI FLOW CONTROL SYSTEM DYNAMIC TEST PROCEDURE, an oil leak was discovered on the hydraulic control oil piping. HPCI had previously been declared INOPERABLE due to planned maintenance, however as a result of the oil leak HPCI remains INOPERABLE. This oil leak would have cause HPCI to be declared INOPERABLE had it been found outside of the planned maintenance. The plant remains at 100% power with no challenges to the health and safety of the public. HPCI is in a 14 day technical specification to repair the oil leak. The licensee notified the NRC Resident Inspector.
ENS 516428 January 2016 00:03:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopPrairie Island's Appendix R calculations credit a procedurally established repair instruction to the Train B Pressurizer Vent valves for a postulated fire in Fire Area 59 (Unit 1) and Fire Area 74 (Unit 2) to obtain Mode 5 during a postulated fire in the affected areas. At 1900 (CST) on 1/7/2016, during a review of corrective actions associated with Prairie Island's NFPA 805 transition, it was identified that the required procedures are not in place to make the analyzed repairs. It has been determined that this condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. As compensatory measures, hourly fire watches are in place in the affected areas of the Auxiliary Building. The operating crew and Fire Brigade have been briefed on the impact of a fire in the affected area. This brief will continue to future operating shifts via a standing instruction. Fire detection equipment for the affected zones has been protected to ensure availability and operating crews are walking down the affected areas to verify any required transient combustibles in the affected areas are controlled in accordance with plant procedure. These compensatory measures, in addition to automatic fire detection and suppression capability in these fire areas, ensure protection of the potentially affected equipment until corrective actions can be completed. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified.
ENS 5161621 December 2015 22:12:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

As part of the License Amendment development to transition to NFPA 805, PINGP (Prairie Island Nuclear Generating Plant) Calculation ENG-ME-353, Mechanical MOV (Motor Operated Valve) Analysis to support IN-92-18 Response, revision 1, issued in 1998, was reviewed for applicability for the transition to NFPA 805. Recent consultation with an MOV engineer regarding the scope of the revision indicated ENG-ME-353 is out of date. On 12/21/2015, during technical review for a new weak link calculation, several MOVs were identified from the list of MOVs that are credited to be manually operated from outside the control room in the event of a fire in the control room or relay room per PINGP Procedure F5 Appendix B, Control Room Evacuation (Fire), that could be damaged if hot shorts were to bypass the torque and limit switches. There are also four other motor valves associated with the Gland Steam system of both Unit 1 and Unit 2 that were added to the procedure F5 Appendix B, Control Room Evacuation (Fire), that have not been analyzed for a weak link. This unanalyzed condition could impact the ability of plant operators to implement procedure F5 Appendix B, Control Room Evacuation (Fire). New hourly fire watch impairments were created for Fire Area 13 (Control Room) and Fire Area 18 (Relay and Cable Spreading Room) as compensatory measures. Therefore, this is an unanalyzed condition reportable under 10 CFR 50.72(b)(3)(ii)(B). The public health and safety is not impacted. The NRC Resident Inspector has been notified.

  • * * UPDATE AT 0107 EST ON 01/14/16 FROM NATHAN BIBUS TO DANIEL MILLS * * *

Reviews of the list of MOVs susceptible to hot shorts bypassing the torque and limit switches credited to be manually operated from outside the control room in the event of a fire have continued. Additional valves have been noted to be affected by this failure mechanism in areas outside of the Control Room or Relay Room. The additional MOVs affected by this unanalyzed condition could impact the ability of plant operators to implement PINGP Procedure F5 Appendix D, Impact of Fire Outside Control/Relay Room. As a compensatory measure, additional hourly fire watch impairments were created for the following fire areas: Fire Area 031 ( A Train Hot Shutdown Panel & Air Compressor/Aux 695 Feedwater Room) Fire Area 032 ( B Train Hot Shutdown Panel & Air Compressor/Aux 695 Feedwater Room) Fire Area 058 (Aux Building Ground Floor Unit 1) Fire Area 073 (Auxiliary Building Ground Floor Unit 2) The public health and safety is not impacted. The (NRC) Resident Inspector has been notified. Notified R3DO (Duncan).

ENS 5160917 December 2015 14:33:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

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.

  • * * UPDATE ON 12/17/2015 AT 1734 EST FROM TOM HOLT TO DONG PARK * * *

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.

ENS 5156624 November 2015 17:02:00MonticelloNRC Region 3GE-3At 1253 CST on 11/24/2015, XCEL Energy Environmental Services made a report to the Minnesota Pollution Control Agency (MPCA) and the Minnesota Department of Natural Resources (DNR) due to the recorded total fish loss (59) as a result of the Monticello Plant's discharge canal temperature change following the reactor scram on 11/23/2015. This notification is being made under 10 CFR 50.72(b)(2)(xi) based on a notification to another government agency. This issue has no safety significance and no impact on the health and safety of the general public. NRC Resident Inspector has been notified.
ENS 5156424 November 2015 14:19:00MonticelloNRC Region 3GE-3On 11/24/2015 at 0534 hours (CST) the plant was in MODE 3 (hot standby) for a forced outage. While initially placing Shutdown Cooling (SDC) in service, the 12 Residual Heat Removal (RHR) pump tripped approximately 8-10 seconds after start due to the closure of the RHR SDC suction isolation valves. This was determined to be from an invalid signal on the Reactor Pressure Vessel (RPV) suction interlock. This is being reported under 10 CFR 50.72 (b)(3)(v)(B) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. This event did not challenge the ability to maintain safe shutdown conditions, remove residual heat, control the release of radioactive material, or mitigate the consequences of an accident as other methods of decay heat removal were being utilized successfully to establish plant conditions. The NRC Resident Inspector has been notified. This incident places the RHR system in a 24-hour technical specification limiting condition for operations.
ENS 5156023 November 2015 15:03:00MonticelloNRC Region 3GE-3At 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.
ENS 5152811 November 2015 11:48:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

At 0826 CST on 11/11/2015, 1R-22, Shield Building Vent Gas Radiation Monitor, was removed from service for planned maintenance. This monitor has no compensatory measure that will allow timely classification of two Emergency Action Levels (EALs) - NUE (Notification of Unusual Event) and Alert classifications - when out of service. It is also used for offsite dose projection calculations. This results in a Loss of Emergency Assessment Capability while 1R-22 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii). Unit 1 Shield Building Ventilation Stack is also monitored by high range monitor, 1R-50, which is used for the same purpose in Site Area or General Emergency classifications. 1R-50 is being monitored and is indicating normal values. There are no radioactive leaks that will impact the Shield Building as evidenced by normal readings on 1R-22 prior to its removal from service. The duration of this maintenance is scheduled for 24 hours and will continue until the monitor is returned to service. Maintenance will not result in the unplanned release of radioactivity to the environment and will not adversely affect the safe operation of the plant or health and safety of the public.

The licensee has notified the NRC Resident Inspector.

  • * * UPDATE AT 1547 EST ON 11/12/15 FROM PAUL FINHOLM TO JEFF HERRERA * * *

The licensee indicated that the duration of maintenance was extended for approximately 24 hours to allow continued repair of the monitor. The NRC Resident Inspector was notified. Notified the R3DO (Kozak).

ENS 5149826 October 2015 15:20:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

At 0703 CDT on 10/26/2015, Prairie Island Nuclear Generating Plant (PINGP) identified Door 62, '11/21 Auxiliary Feedwater Pump (AFW) Room to 12/22 AFW Pump Room' to be in a closed position. Door 62 functions as a fire door, and closes in the event of a fire in either A or B Train AFW Pump Rooms. When Unit 1 or Unit 2 is in modes 1-4, Door 62 is required to remain open in the event of an internal flood in AFW Pump Room or a Turbine Building High Energy Line Break (HELB) Flood into the AFW Pump. Currently, Unit 2 is in MODE 6, so only the Unit 1 Turbine Building HELB applies. Door 62 is normally open with a fusible link to allow closure during a fire. Door 62 was closed during maintenance. With the door closed the Unit 1 side of the Auxiliary Feedwater Pump Room lost its ability to adequately drain water in a Unit 1 HELB event and was in an unanalyzed condition. Upon discovery, Door 62 was immediately repositioned to be open per the analyzed condition and a fire watch established per plant procedure. This notification is being conducted in accordance with 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition.

The plant remains safe, and this condition does not pose any additional risk to the public. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION FROM NATHAN BIBUS TO DONALD NORWOOD AT 1703 EST ON 11/30/2015 * * *

Prairie Island Nuclear Generating Plant is retracting this event notification, EN# 51498. Further analysis determined that the closure of Door 62 would not have prevented the structures, systems and components (SSC) located in the AFW Pump rooms, or SSCs powered from Motor Control Centers (MCCs) located in the AFW Pump rooms, from performing their safety functions. This is because the door closure would not have caused water level to rise above the maximum tolerable water height during any design basis flooding event. The acceptance criteria of the area flooding calculations were still met with Door 62 closed. Therefore, the Unit 1 side of the AFW Pump room did not lose its ability to adequately drain water in a Unit 1 HELB event, this event was not an 8-hour notification for an Unanalyzed Condition that significantly degrades plant safety, under 10CFR50.72(b)(3)(ii). The licensee has notified the NRC Resident Inspector. Notified R3DO (Lipa).

ENS 5149724 October 2015 11:45:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopPrairie Island Nuclear Generating Plant (PINGP) notified Local Law Enforcement Agency and the Prairie Island Community Tribal Council due to an incident onsite of local interest. This notification is being conducted in accordance with 10 CFR 50.72(b)(2)(xi) for notification to an outside government agency. The plant remains safe, and this condition does not pose any additional risk to the public. The licensee has notified the NRC Resident Inspector.
ENS 5142928 September 2015 21:54:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

At approximately 1327 CDT on September 28, 2015, both D1 and D2 Diesel Generators (EDG) were inoperable simultaneously until corrected at 1345 CDT. The D2 Diesel Generator had been declared inoperable for the planned performance of SP1307, D2 Diesel Generator 6 Month Fast Start Test. Tech Spec LCO 3.8.1 Condition B had been entered for D2 Diesel Generator. Subsequently, D1 Diesel Generator was determined to be inoperable but available due to Train A Cooling Water Header being inoperable during post maintenance testing of SV-33133, Backwash Water Supply to the 121 Safeguards Traveling Screen. Tech Spec LCO 3.7.8 Condition B was entered for the Cooling Water Header inoperability, which forced a cascade to Tech Spec 3.8.1 Condition B for D1 Diesel Generator. With both Emergency Diesel Generators inoperable, Tech Spec 3.8.1 Condition E was entered, which required the restoration of one Emergency Diesel Generator to operable status within 2 hours. D2 was returned to operable status through completion of SP 1307, and Tech Spec 3.8.1 Condition E was exited at 1345 CDT. With both Emergency Diesel Generators inoperable, this condition is reportable under 10 CFR 50.72 (b)(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. The plant remains safe, and this condition does not pose any additional risk to the public. Additionally, our defense in depth strategies are relied upon to take actions to protect the health and safety of the public. D2 Diesel Generator remained available with full cooling water flow during this time. The safety significance of this event is low, as engineering hydraulic analysis has demonstrated that with the safeguards traveling screen backwash water supply valve fully opened, the Cooling Water System would have continued to provide full cooling flow to the D1 Diesel Generator. The licensee has notified the NRC Resident Inspector.

  • * * RETRACTION ON 11/4/15 AT 1510 EST FROM NATHAN BIBUS TO DONG PARK * * *

An evaluation has been performed and it has been determined that SV-33133 and SV-33134 do not have an active close safety function. The Cooling Water System analysis of record, calculation ENG-ME-820, Rev 0B shows that the Cooling Water System continues to have flow margin with screen wash control valves SV-33133 and SV-33134 open. Therefore, there is no need for the valves to close to ensure the Cooling Water System's safety function. Because the valves do not have a safety function to close, this event was not an event or condition which could have prevented the fulfillment of a safety function of an SSC (structures, systems and components) required to mitigate the consequences of an accident and, therefore, did not require an 8 hour notification in accordance with 10 CFR 50.72(b)(3)(v)(D) for an event or condition that could have prevented fulfillment of a safety function (i.e., accident mitigation) under 10 CFR 50.72(b)(3)(v)(D). The notification is hereby retracted. The licensee has notified the NRC Resident Inspector." Notified R3DO (Orlikowski).

ENS 513798 September 2015 18:15:00MonticelloNRC Region 3GE-3A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the site has been terminated. The licensee notified the NRC Resident Inspector.
ENS 5130610 August 2015 19:49:00MonticelloNRC Region 3GE-3On August 10, 2015, at approximately 1555 CDT, the licensee was notified that emergency siren S-07 in Sherburne County, MN had inadvertently activated from approximately 1525-1538 CDT. The cause of the activation is under investigation. The siren vendor (NELCOM) was also contacted and made the notification to the licensee. As a result, this issue is being reported under 10CFR50.72(b)(2)(xi) for notifications to other offsite government agencies as the licensee was notified by the Sherburne County Sherriff's Office. The source of the activation signal has not been determined. The vendor is investigating. The siren is no longer actuating. There was no impact to the health and safety of the public as a result of this event as the offsite response capabilities remain functional with a failure of only 1 out of 106 total sirens. The site is operating normally with no emergency present. (The) NRC Resident Inspector has been notified. The licensee will notify the State.
ENS 5126627 July 2015 14:28:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopOn July 27, 2015 at 0902 (CDT), (the site commenced) a planned outage of the Emergency Response Data System (ERDS) and Safety Parameter Display System (SPDS), referred to as 'plant computer'. The unavailability of ERDS and SPDS could significantly affect the site's ability to respond to an emergency if one were to occur. During this time, Operations will be utilizing the site's procedures 1C1.5 and 2C1.5, 'OPERATION WITHOUT COMPUTER', which requires additional operators for monitoring of equipment affected by the loss of the plant computer. Additionally, as this is a planned outage, the work week schedule has been modified to ensure limited interactions required by Operations during this time frame. The site expects ERDS and SPDS to be operational 1200 July 28, 2015. This event is reportable under 10 CFR 50.72(b)(3)(xiii), Any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability (e.g., significant portion of Control Room indication, Emergency Notification System (ENS), or Offsite Notification System). The ENS and Offsite Notification System are not affected by this planned outage. The health and safety of the public are not impacted by this planned outage. The NRC Resident Inspector has been informed.
ENS 5124120 July 2015 13:37:00MonticelloNRC Region 3GE-3On 7/20/2015 at approximately 0931 CDT, the Monticello nuclear generating plant was notified by Wright County Sheriffs Office of a spurious actuation of one emergency response siren in the city of Monticello that occurred at approximately 0855 CDT (lasted for approximately three minutes). This actuation was confirmed by vendor system monitoring. As a result, this issue is being reported under 10 CFR 50.72(b)(2)(xi) for notifications to other offsite government agencies as the licensee was notified by the Wright County Sheriffs Office. The source of the activation signal has not been determined. The vendor is investigating. The siren is no longer actuating. There was no impact to the health and safety of the public as a result of this event as the offsite response capabilities remain functional with a single siren failure. The site is operating normally with no emergency conditions present. The NRC Resident Inspector has been notified. The State of Minnesota will be notified.
ENS 511367 June 2015 12:03:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopThe following was received via phone and fax: On June 7, 2015, at 0735 CDT, the Unit 2 Reactor automatically tripped while operating at 100 percent power due to an automatic Turbine trip due to low bearing oil pressure. The crew entered the reactor trip emergency operating procedure and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. All safety functions operated as designed. The automatic Reactor trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the Auxiliary Feedwater Pumps as designed on low narrow range Steam Generator level and provided makeup flow to the Steam Generators. The Auxiliary Feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam Generator levels have been returned to normal. Auxiliary Feedwater has been secured. Steam Generators are being supplied by (the) 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of the Turbine trip remains under investigation. There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified.
ENS 511071 June 2015 02:25:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopOn May 31, 2015 at 2220 CDT, the Unit 1 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 11 Condensate Pump followed by a lockout trip of 11 Main Feedwater Pump. Manual Reactor Trip is directed by the annunciator response procedure for the lockout alarm, C47010-0101, 11 Feedwater Pump Locked Out. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. Following the reactor trip, 15A Feedwater Heater relief lifted and failed to reseat. 12 Main Feedwater Pump was subsequently secured resulting in 15A Feedwater Heater relief valve reseating successfully. Steam generators are being supplied by 12 Motor Drive Auxiliary Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of 11 Condensate Pump trip remains under investigation. There was no effect on Unit 2 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified. Unit 2 is unaffected and remains at 100 percent power.
ENS 510342 May 2015 20:57:00MonticelloNRC Region 3GE-3On 5/2/2015 at 1247 CDT, while the plant was in MODE 5 for a refueling outage with the vessel cavity flooded, during logic testing of the non-credited essential 4kV bus, MNGP (Monticello Nuclear Generating Plant) experienced a human performance error that caused a loss of the 4kV bus and essential load center. Loss of the load center de-energized valve position indication on one shutdown cooling isolation valve. This tripped the shutdown cooling pump on a pump suction interlock, resulting in the loss of shutdown cooling. This is being reported under 10 CFR 50.72(b)(3)(v)(B) as an event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. Following the event, the safety related load centers were cross- This event did not result in the release of any radioactive material and did not challenge the health and safety of the public. The NRC Resident Inspector has been notified.