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 Entered dateSiteRegionReactor typeEvent description
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 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 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 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 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 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 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 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 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 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 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 5101325 April 2015 04:10:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 2125 on 4/24/15, 1R-2, Containment Vessel Area Radiation Detector failed. Previously, 1R-7, Incore Seal Table Area Radiation Detector, had failed on 4/20/15. The compensatory measure for 1R-2 out-of-service is to verify 1R-7 operating properly and the compensatory measure for 1R-7 out-of-service is to verify 1R-2 operating properly. With both monitors out-of-service and Unit 1 operating in Mode 5, no compensatory measure is available that will allow timely classification of two Emergency Action Levels (EALs) - Notification of Unusual Event (NUE) classification (RU2.2) and Alert classification (RA3.2). This results in a Loss of Emergency Assessment Capability while 1R-2 and 1R-7 are concurrently out-of-service. This is a reportable condition per 10 CFR 50.72(b)(3)(xiii). Monitoring of radiological conditions in Unit 1 Containment showed no indication of RCS leakage or elevated radiation levels prior to the failure of 1R-2. Unit 1 Containment also remains monitored by 1R-48, Containment Hi Range Area Radiation Detector A and 1R-49, Containment Hi Range Area Radiation Detector B, which currently indicate normal radiation levels. Unit 1 Shield Building Stack is also monitored by 1R-50, Shield Building High Range Vent Gas Radiation Detector, which also currently indicates normal radiation levels. Additionally, a temporary portable radiation monitor has been placed near the location of 1R-2 and is being continuously monitored. The plant remains in a safe condition and there was no effect to the health and safety of the public. The licensee has notified the NRC Resident Inspector.
ENS 5099419 April 2015 23:04:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopOn April 19, 2015, it was determined that in the event of an Appendix-R fire in the Control Room/Relay Room, fire induced circuit damage can potentially result in Reactor Coolant Pump's (RCP) restarting. Procedure F5 APP B does not take actions to open RCP DC knife switches as required per Appendix R calculation GEN-PI-026, GEN-PI-054, and GEN-PI-055. Neither unit is currently susceptible to this condition due to the installation of improved RCP seals which allow adequate time to restore seal cooling prior to seal failure. However, the condition has occurred within three years of the time of discovery and is reportable under 10 CFR 50.72(b)(3)(ii)(B). The Appendix R analysis requires that manual actions are taken to ensure that the RCPs are tripped and actions are taken to prevent restarting of the RCPs. The RCPs breaker must be verified open at the associated bus and DC knife switches located in the breaker cubicles are to be opened. These actions are required due to fire induced, loss of remote trip (spurious breaker closure) and loss of RCP seal cooling water and loss of Component Cooling (CC) water to the thermal barrier heat exchanger. These actions are required to achieve and maintain Mode 3, Hot Standby in the event of a catastrophic fire that results in the functional loss and/or evacuation of the Control Room/Relay Room. Current procedures verify the RCP breakers are open; however, they do not open the DC knife switch and therefore a hot short could result in a RCP restarting. This procedure omission meets the reporting criteria for 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Senior Resident Inspector.
ENS 509667 April 2015 22:33:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt approximately 1833 CDT on April 07, 2015, the licensee was notified by the siren vendor (NELCOM) that Dakota County Dispatch reported an inadvertent activation of an emergency siren (D-1), in Dakota County, MN. The cause of the siren activation is unknown. The siren was deactivated at 1810 CDT after sounding for approximately 25 minutes. The siren vendor (NELCOM) has been contacted to repair the siren. The siren remains out of service. This is the only siren out of service within the 10 mile Emergency Planning Zone (EPZ). There are 123 Emergency Notification sirens. Of the seven (7) credited sirens in Dakota County, six (6) remain in service. Thus, emergency notification capabilities remain in effect. NRC Resident has been informed.
ENS 509503 April 2015 10:12:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopOn April 3, 2015 at 0652 CDT, the Unit 2 reactor was manually tripped while operating at 100 percent power due to a lockout trip of 21 Main Feedwater Pump as required by the annunciator response procedure for the lockout alarm. This also resulted in a turbine trip. The crew entered the reactor trip emergency operating procedures and stabilized the unit in Mode 3 at normal operating pressure and temperature. All control rods fully inserted into the core following the trip. The manual trip is reportable per 10 CFR 50.72(b)(2)(iv)(B). The Auxiliary Feedwater System actuated to start the auxiliary feedwater pumps as designed on low narrow range steam generator level and provided makeup flow to the steam generators. The auxiliary feedwater actuation is reportable per 10 CFR 50.72(b)(3)(iv)(A). Steam generator levels have been returned to normal. The auxiliary feedwater pumps have subsequently been secured and returned to automatic. Steam generators are being supplied by 22 Main Feedwater Pump and decay heat is being removed by the condenser steam dump system. The cause of 21 Main Feedwater Pump trip has been determined to be a failed suction pressure switch. There was no effect on Unit 1 as a result of this trip. The health and safety of the public and site personnel were not at risk at any time during this event. The NRC Senior Resident Inspector has been notified. The licensee plans to issue a press release.
ENS 509482 April 2015 17:29:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopNorthern States Power Company - Minnesota (NSPM) has completed a review of seismic monitor performance at the Prairie Island Nuclear Generating Plant (PINGP) over the past 3 years. The emergency preparedness plan requires seismic monitoring instruments to diagnose an earthquake for emergency action levels (EAL) HA1.1 (Seismic Event Greater Than Operating Basis Earthquake (OBE) as indicated by 'OBE Exceedance' alarm on Seismic Monitoring Panel) or HU1.1 (Earthquake felt in plant as indicated by Valid 'Event' alarm on Seismic Monitoring Panel). Contrary to that requirement, this review identified 6 unplanned instances where the seismic monitor was non-functional that were not previously reported, and 3 planned instances where the seismic monitor was non-functional for greater than 24 hours that were not previously reported. Since there was no compensatory measure that could be credited when the seismic monitor was non-functional, an emergency classification at the ALERT or UNUSUAL EVENT level could not be obtained with site instrumentation for a seismic event. The seismic monitor is currently functional, however it was determined to be non-functional on the following dates: Unplanned out of service: 1. August 14, 2012 2. November 16, 2012 3. November 18 2012 4. November 21, 2012 5. December 5, 2012 6. January 16, 2013 Planned greater than 24 hour out of service: 1. December 14, 2012 2. September 3, 2014 3. September 30, 2014 The unplanned non-functional conditions of the seismic monitor have been corrected and were entered into the NSPM Corrective Action Program. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). This report is required per 10 CFR 50.72(a)(1)(ii) as an event that occurred within 3 years of the date of discovery. Corrective actions are in progress to address the missed reporting of seismic monitor unavailability. The licensee notified the NRC Resident Inspector.
ENS 508707 March 2015 15:05:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 1155 CST on March 7, 2015, a small cooling water leak was identified on the 21 Containment Fan Coil Unit east face u-bend on the north east corner bottom bundle. Unit 2 Containment was declared inoperable, which required entry into Technical Specifications (TS) LCO 3.6.1, Condition A, Containment inoperable, applicable in MODES 1, 2, 3, and 4. Immediate actions were taken to isolate the Fan Coil Unit within 1 hour from the initial identification of the leak. 21 Containment Fan Coil Unit was isolated, Containment was declared operable and TS 3.6.1 Condition A was exited at 1220 CST on 3/7/15. This condition is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. The plant remains in a safe condition and there was no effect to the health and safety of the public. The licensee has notified the NRC Resident Inspector.
ENS 508665 March 2015 05:35:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

The Instrument Air Containment Isolation failed closed on Unit 2. This isolated normal letdown / excess letdown and required Pressurizer level to be maintained by diverting Pressurizer level to the Pressurizer Relief Tank. The Pressurizer Relief Tank rupture disc ruptured which resulted in a fire alarm in Unit 2 Containment. The fire alarm could not be validated within 15 minutes which resulted in a declaration of an Unusual Event based on EAL HU2.1. The loss of Instrument Air to Unit 2 Containment resulted in a loss of cooling to the reactor vessel gap and support cooling systems. Due to the loss of reactor vessel ventilation systems a plant shutdown to Mode 3 has been initiated. No radioactive releases to the environment are in progress or expected to occur. The public health and safety has not been jeopardized. The licensee informed state/local agencies and the NRC Resident Inspector and does plan to issue a press release. Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA).

  • * * UPDATE AT 0130 EST ON 03/06/15 FROM TERRY BACON TO DANIEL MILLS * * *

At 1725 (CST) on 3/5/15 Instrument Air to Unit 2 containment was established. Letdown was then reestablished. These actions stabilized the plant and stopped the 30 GPM (gallons per minute) identified leakage out of the PZR Relief Tank rupture disc into containment. This condition is what caused the fire detection alarm in containment, and is also Unusual Event criteria. Unit 2 containment was entered and it was confirmed that NO fire existed in containment. The Unusual Event was terminated at 0018 CST on 3/6/15 based on no fire and no identified RCS leakage into containment. The plant is currently in Mode 3. The health and safety of the public was not jeopardized. The licensee informed state/local agencies and the NRC Resident Inspector and does plan to issue a press release. Notified R3DO (Stone), EO (Ross-Lee) and IRD (Stapleton). Notified other Federal Agencies (DHS SWO, FEMA Ops, FEMA NWC, NICC Watch Officer and NuclearSSA).

ENS 5082617 February 2015 00:13:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

On February 16, 2015, at 1644 (CST), Prairie Island Nuclear Generating Plant (PINGP) Unit 1 and Unit 2 Control Room Envelope Boundary was declared inoperable when it was discovered that Door 158, Auxiliary Building to 122 Control Room Chiller Room, would not latch. Both trains of Control Room Special Ventilation were declared inoperable and Tech Spec LCO 3.7.10 Condition B was entered. In addition, with the Control Room Envelope Boundary inoperable, Tech Spec 3.7.11 Condition E was required to be entered due to both Control Room Chillers inoperable. The required actions of Tech Spec 3.7.11 Condition E required both Units to enter Tech Spec LCO 3.0.3. As a mitigating action, station personnel were dispatched to secure Door 158. This condition was corrected on February 16, 2015, at 1709 (CST) when the deadbolt was engaged to maintain Door 158 closed. Tech Specs 3.7.10 Condition B, 3.7.11 Condition E, and 3.0.3 were all exited at 1709 (CST) February 16, 2015. 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 to Mitigate the Consequences of an Accident. Based on immediate implementation of mitigating actions and restoration of the Control Room Envelope, the protection of the health and safety of the public was not affected by this issue. This event has been entered into the sites Corrective Action Program. The licensee has notified the NRC Senior Resident Inspector.

  • * * UPDATE AT 1220 EST ON 02/24/15 FROM NATHAN BIBUS TO S. SANDIN * * *

The licensee is retracting this report based on the following: After discussions and interviews with personnel involved, it was determined that the Aux Building Operators found Door 158 latched. After passing through the door, the Operators checked to ensure the door was latched and discovered it was sticking and required assistance to latch by agitating the latch operating mechanism. The door was checked additional times and the door would latch with assistance. The Operators ensured the door was latched and notified the Control Room at 1644 (CST). When the Operators installed the dead bolt and padlock at 1709 (CST), the door was still latched. As the door was never left unlatched and was always able to latch, the door was operable and there was no loss of safety function. The licensee has notified the NRC Resident Inspector. Notified R3DO (Orlikowski).

ENS 5080811 February 2015 05:12:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 2205 CST on February 10, 2015, a cooling water leak of approximately 60 to 90 drops per minute was identified on the 14 Containment Fan Coil Unit Cooling Water face gasket. As a result, Unit 1 Containment was declared inoperable. This required entry into Technical Specifications (TS) LCO 3.6.1 Condition A, Containment inoperable, applicable in MODES 1, 2, 3, and 4. Immediate action was taken to isolate the fan coil unit within 1 hour from the initial identification of the leak. After isolating the cooling water leak to 14 Containment Fan Coil Unit, containment was declared operable and TS 3.6.1 Condition A was exited at 2232 CST. A Work Request (WR) has been initiated to restore 14 Containment Fan Coil Unit to an operable condition. This condition is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. The plant remains in a safe condition and there was no effect to the health and safety of the public. The licensee has notified the NRC Resident Inspector.
ENS 5080110 February 2015 10:41:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

On February 10, 2015, Prairie Island Unit 1 was shutdown in Mode 3 during a planned outage. Ultrasonic testing in support of Unit 1 Emergency Core Cooling System (ECCS) void verifications identified existing voids with calculated volumes in excess of the OPERABILITY limits specified by the procedure. This rendered both trains of Residual Heat Removal (RHR) systems inoperable requiring entry into Technical Specification 3.0.3 at 0250 (CST). The station took prompt actions to vent the identified voids. The void at 1 RH-12 was vented to within acceptable limits allowing LCO 3.0.3 to be exited at 0538 on February 10. Venting at 1RH-11 is in progress. Voiding was identified at location 1-RH-11 with a calculated volume of 62.21 cubic inches with an OPERABILITY limit of 11.62 cubic inches. Voiding was identified at location 1-RH-12 with a calculated volume of 350 cubic inches with an OPERABILITY limit of 22.84 cubic inches. There was no impact to the health and safety of the public as Safety Injection was available and the time both trains of RHR were INOPERABLE was limited. This event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). The licensee notified the NRC Resident Inspector.

  • * * RETRACTION FROM TOM HOLT TO JEFF HERRERA AT 1444 EDT ON 4/11/15 * * *

Further analysis was performed on the two void locations, 1RH-11 and 1RH-12. Based on this additional analysis from AREVA, it was determined that the void located at 1RH-11 (RHR Train A) was operable. The calculations for past operability at inspection location 1RH-11 provides reasonable assurance that a void of 65 cubic inches will not generate forces that will fault any piping and supports. The void location at 1RH-12 (RHR Train B) was considered inoperable due to exceeding current procedural operability limits. The void located at 1RH-11 was determined to be nonconforming due to exceeding procedural design basis limits. Therefore, with RHR Train A determined to be operable, this event was not an 8-hour notification for an unanalyzed condition that significantly degrades plant safety, nor a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). The licensee has notified the NRC Resident Inspector. Notified the R3DO (Skokowski).

ENS 5070223 December 2014 13:00:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopThis 60-day telephone notification is being made under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of the Unit 1 11 RHR Pump while Unit 1 was at NO MODE. It was a complete actuation. This was not the result of a preplanned sequence during testing and the system was not removed from service. At 0340 CDT on October 26, 2014, the Unit 1 11 RHR Pump was noted to have started while investigating an NSSS (nuclear steam supply system) annunciator system ground. The operator verified that the pump had adequate suction path via the RCS Cold Leg and stable pump discharge pressure. The 11 RHR Pump was stopped per the plant's abnormal operating procedure. Evaluation of this event has shown that the auto-start of 11 RHR pump was due to a human performance error while performing relay replacements. Following troubleshooting of the auto-start and completion of the relay maintenance, the 11 RHR pump was run; the pump and system operated properly. The health and safety of the public was not affected by this issue. The licensee has notified the NRC Resident Inspector.
ENS 5067212 December 2014 11:34:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 0248 CST on December 12, 2014, 2R-30, Auxiliary Building Normal Vent Radiation Monitor, failed. This monitor was the redundant monitor for 2R-37, Auxiliary Building Normal Vent Radiation Monitor, which was previously taken out of service for planned maintenance. With 2R-30 and 2R-37 out of service, there are no monitors that will allow for timely classification of two Emergency Action Levels (EALS) - NUE (Notification of Unusual Event) and Alert Classifications. This results in a loss of emergency assessment capability while 2R-30 and 2R-37 are out of service. This is a reportable condition in accordance with 10CFR50.72(b)(3)(xiii). The Auxiliary Building ventilation effluent monitor readings were normal and there were no elevated readings in Auxiliary Building area monitors prior to the unavailability of these monitors. The health and safety of the public was not affected by this issue. The activity to repair and return one of the monitors to service is continuous until restored. The licensee has notified the NRC Resident Inspector.
ENS 506523 December 2014 16:02:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

At 0950 CST on December 3, 2014, R-1 Control Room Area Radiation Monitor failed low and was taken out of service. This monitor has no redundant monitor that will allow timely classification of two Emergency Action Levels (EALs) - NUE (Notification of Unusual Event) and Alert classifications - when out of service. This results in a Loss of Emergency Assessment Capability while R-1 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii). Control Room readings were normal as evidenced on R-1 prior to its failure. Compensatory measures have been established with a Portable Area Monitor installed in the Control Room adjacent to R-1. Corrective maintenance is in progress and will continue until the monitor is returned to service. The health and safety of the public was not affected by this issue.

The licensee has notified the NRC Resident Inspector.

ENS 5063020 November 2014 16:29:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 1022 CST on November 20, 2014, a cooling water leak of approximately 20 to 30 drops per minute was identified on the 14 Containment Fan Coil Unit NW side lower face gasket. Unit 1 Containment was declared inoperable, which required entry into Technical Specifications (TS) LCO 3.6.1 Condition A, containment inoperable, applicable in MODES 1, 2, 3, and 4. Immediate actions were taken to isolate the Fan Coil Unit within 1 hour from the initial identification of the leak. 14 Containment Fan Coil Unit was isolated and Containment was declared operable and TS 3.6.1 Condition A was exited at 1109 CST. A Work Request (WR) was initiated and 14 Containment Fan Coil Unit has been restored to an operable condition. This condition is reportable under 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. The plant remains in a safe condition and there was no effect to the health and safety of the public. The licensee has notified the NRC Resident Inspector. The licensee verified the other Unit 1 Containment Fan Coil Units did not have the same problem.
ENS 505986 November 2014 13:41:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopA licensed operator has been found in violation of the Northern States Power Minnesota Fitness for Duty Policy. The individual's access to the plant has been suspended and the operator has been removed from duty. The NRC Resident Inspector has been informed.
ENS 5053113 October 2014 20:48:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAs part of a fire modeling analysis evaluating a missing fire barrier reported in ENS notification 50475 (a gap where the ventilation duct passes through the wall from the Bus 150/160 room to the Auxiliary Feedwater (AFW) pump room), an additional missing fire barrier was identified. The barrier is related to separation of redundant pressurizer heater banks credited for safe shutdown. For Fire Area 32 (AFW pump room), Group E Pressurizer Heaters are credited for safe shutdown because Group A and Group B Pressurizer Heater cables could be affected by a fire in this area. It was determined that a cable associated with the Group C, D, and E Pressurizer Heaters is routed in Fire Area 32. Therefore, a fire in Fire Area 32 could affect all five Pressurizer Heater Groups. An evaluation has previously demonstrated that Mode 3 Hot Standby could be maintained with no charging pumps or pressurizer heaters available, but it has not been determined if Mode 5 could be achieved. Therefore, this missing fire barrier meets the reporting criteria for 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition that significantly degrades plant safety. The protection of the health and safety of the public was not affected by this issue. Unit 1 is in mode 6 and no fire has occurred. Compensatory measures (fire watches) are in place for Fire Area 32. The licensee has notified the NRC Senior Resident Inspector. Unit 2 has a different cable routing and is not affected.
ENS 5052311 October 2014 00:03:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

During a Fire Penetration walkdown, an opening less than 1 (inch) tall under a duct was identified between the Train B Aux Feedwater Pump (AFWP) Room (Unit 1 side) and the Bus150/160 room. This constitutes a missing fire barrier between Fire Area (FA) 32 and FA 37 such that the required degree of separation for redundant safe shutdown trains is lacking. A firewatch has been established on the Unit 1 side of the AFWP Room and Bus 150/160 room. The compensatory fire watch will remain in place per F5 Appendix K until the fire barrier is returned to full functional status. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Resident Inspector of this event.

  • * * RETRACTION FROM DOUG LARSON TO DANIEL MILLS AT 2118 EST ON 11/18/2014 * * *

Northern States Power Company (NSPM) has determined that the opening under the duct is not separating redundant equipment and no longer meets the regulatory definition of an unanalyzed condition. This was identified when it was determined that the redundant pressurizer heater cables credited for safe shutdown were routed through the same fire area. This was reported under EN 50531 (Unanalyzed Condition Due To Missing Fire Barrier). Since the missing fire barrier does not separate redundant equipment, the condition does not degrade plant safety, and is no longer reportable. The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Resident Inspector. Notified R3DO (Peterson).

ENS 505116 October 2014 15:35:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 1059 CDT on October 6, 2014, the State of Minnesota and Environmental Services were notified that approximately 100 fish were found dead against the trash racks of the plant screen house. The licensee had completed zebra mussel treatment several days prior to this event and the fish deaths are attributed to this treatment. This is reportable under 10 CFR 50.72 (b)(2)(xi), News Release or Notification of Other Government Agency. The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Senior Resident Inspector.
ENS 5049225 September 2014 00:04:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

At 1533 CDT on September 24, during a Past Operability review for the Unit 1 Train B Inadequate Core Cooling Monitor (ICCM), it was discovered that due to data errors, the as left VDC (voltage DC) value for the Remote Display Power Supply was outside of acceptance criteria during the last performance of SP 1213B 'Inadequate Core Cooling Monitor Calibration,' completed on 8/25/2014. The Reactor Vessel Level Instrumentation System (RVLIS) and the Core Exit Thermocouples (CETs) are subsystems of the ICCM System and were determined to be inoperable during this timeframe. Therefore Tech Spec 3.3.3 Condition A for Train B RVLIS and Tech Spec 3.3.3 Condition B for CETs were not met. Also during this time, Unit 1 Train A ICCM was out of service from 1110 (CDT) on 9/8/2014 to 1530 (CDT) on 9/12/2014 for calibration. Thus, both trains of Core Exit Thermocouples and RVLIS were inoperable during this 4 day, 4 hour and 20 minute time period. ICCM System indication is used for Safety Injection Termination and Reinitiation criteria in the Emergency Operating Procedures (EOPs). During the time when both trains of ICCM were inoperable, B Train remained in service and was thought to be operable. This provided the potential for operators to perform an untimely SI Termination or Reinitiation during an accident based on inaccurate information. The Emergency Response Computer System was available to provide accurate backup information to the operators and for timely Emergency Action Limit classification This is reportable under 10 CFR 50.72 (b)(3)(v)(D), Event or Condition that Could Have Prevented Fulfillment of a Safety Function to Mitigate the Consequences of an Accident. Based on new information currently being received from Westinghouse, the ICCM Remote Display Power Supply only functions to communicate data to the display and has no effect on the data values. Thus the ICCM System may have been functioning properly. An update to this information will be provided once it is available The protection of the health and safety of the public was not affected by this issue. The Unit 1 Train B ICCM System was returned to operable status at 2300 (CDT) on 9/24/2014. The license has notified the NRC Senior Resident Inspector.

  • * * RETRACTION AT 1408 EDT ON 10/6/2014 FROM NATHAN BIBUS TO DONG PARK * * *

Based on further analysis and input from Westinghouse, the remote display 12V power supply is operable as long as data is still updating on the display. It was determined that the tolerance acceptance criteria within our procedure was conservative. For this application, the power supply can function at much less than 12V. The acceptance criteria within surveillance procedures were incorrect and procedure changes have been submitted to correct the tolerance. Therefore, there was no loss of safety function. The licensee has notified the NRC Resident Inspector. Notified R3DO (Orlikowski).

ENS 5048724 September 2014 08:41:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

At 0728 CDT on 9/24/2014, 2R-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 2R-22 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii). Unit 2 Shield Building Ventilation Stack is also monitored by high range monitor, 2R-50, which is used for the same purpose in Site Area or General Emergency classifications. 2R-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 2R-22 prior to its removal from service. The duration of this maintenance is scheduled for 1 hour 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.

ENS 5047519 September 2014 17:54:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

As part of an extent of condition review related to ENS report 50362 (Unanalyzed Condition Due to Lack of Appropriate Fuse Protection), it was identified that, within the past three years, three additional examples existed of missing fire barriers that could have impacted separation of redundant safe shutdown trains. For the affected areas associated with each of the three missing barriers, compensatory measures (hourly fire watches) had been in place for three years prior to the discovery of the condition. However, it was determined the missing barriers met the reporting criteria for 10 CFR 50.72(b)(3)(ii)(B) as unanalyzed conditions that significantly degraded plant safety. The first condition was the use of rodofoam material in Auxiliary Building seismic joint seals. Rodofoam is a combustible foam material and does not provide a fire rating equivalent to the barriers in which it was used. This issue was corrected in a plant modification. The second condition is the lack of required fire dampers in ventilation supply ducts to the Aux. Feedwater (AFW) Pump Rooms (Fire areas 31 and 32). An engineering change package has been approved to implement a plant modification to install the required dampers as part of the NFPA-805 transition. The third condition is a gap where the ventilation duct passes through the wall from the Bus 150/160 room into the AFW pump room. An engineering change package has been approved to implement a plant modification to correct this condition as part of the NFPA-805 transition. The protection of the health and safety of the public was not affected by this issue. In all cases, compensatory measures existed to ensure the protection of safe shutdown equipment. The licensee has notified the NRC Senior Resident Inspector.

  • * * UPDATE FROM DOUG LARSON TO DANIEL MILLS AT 2122 EST ON 11/18/14 * * *

Northern States Power Company (NSPM) has determined that the missing fire damper and the wall gap (the second and third identified conditions) are not separating redundant equipment and no longer meet the regulatory definition of an unanalyzed condition. This was identified when it was determined that the redundant pressurizer heater cables credited for safe shutdown were routed through the same fire area. This was reported under EN 50531 (Unanalyzed Condition Due To Missing Fire Barrier). Since the missing fire barrier does not separate redundant equipment, the condition does not degrade plant safety, and is no longer reportable. The protection of the health and safety of the public was not affected by this issue. The licensee has notified the NRC Resident Inspector. Notified R3DO (Peterson).

ENS 5044611 September 2014 11:32:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt 1009 CDT on September 11, 2014, 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 8 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 affect the safe operation of the plant or health and safety of the public. The licensee has notified the NRC Resident Inspector.
ENS 504244 September 2014 09:50:00Prairie IslandNRC Region 3Westinghouse PWR 2-Loop

At 0843 CDT on September 4th, 2014, 1R-50 High Range Shield Building Vent Gas Monitor was removed from service for planned maintenance. This monitor provides indication of release of gaseous radioactivity to the environment. There is not a compensatory measure that will allow timely classification of two Emergency Action Levels when out of service: General Emergency Event RG1.1 and Site Area Emergency Event RS1.1. This results in a loss of emergency assessment Capability while 1R-50 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 radiation monitor 1R-22 that is used for the same purpose in Alert or Unusual Event Emergency classifications. 1R-22 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-50 prior to its removal from service. The maintenance is scheduled for ten hours or 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 affect the safe operation of the plant or health and safety of the public.

The licensee has notified the NRC Resident Inspector.

ENS 5036213 August 2014 17:40:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopBased on industry operating experience, a drawing review was performed and it was determined that the control circuits for the DC Emergency Turbine Oil Pumps for both Unit 1 and Unit 2 are not fused properly. Therefore an overload within the control circuit could result in a fire that could propagate to multiple fire areas affecting safe shutdown equipment that could be compromised, which affects the Appendix R safe shutdown analysis. Based on this information, the determination was made that this condition meets the reporting criteria for 10 CFR 50.72(b)(3)(ii)(B). Compensatory measures were already in place in accordance with F5 Appendix K, Fire Protection system minimum requirements, for impaired fire protection equipment. Hourly fire watches are in place in the affected locations as a result of previous conditions. The presence of the compensatory measures in addition to automatic fire detection in these fire areas ensures the protection of the equipment. The protection of the health and safety of the public is not affected by this issue. The license has notified the NRC Senior Resident Inspector.
ENS 5035812 August 2014 16:43:00Prairie IslandNRC Region 3Westinghouse PWR 2-LoopAt approximately 1248 CDT, on August 12, 2014, Emergency Planning was notified by the siren vendor that Pierce County inadvertently actuated sirens at 1239 (CDT) while performing a cancel test. Fifty two (52) of the 123 sirens were actuated county wide for approximately 15 seconds before Pierce County could cancel activation. Per F3-5.2, Response to a False Siren Activation, this requires a 10 CFR 50.72(b)(2) 4 hour non-emergency report. Capability to notify the public was never degraded during this time. The sirens remain in service. No press release is planned at this time. The license has notified the NRC Senior Resident Inspector.