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The query [[Category:ENS Notification]] [[Reactor type::Westinghouse PWR 2-Loop]] was answered by the SMWSQLStore3 in 0.0401 seconds.


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 Entered dateSiteRegionReactor typeEvent description
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 5327219 March 2018 13:09:00GinnaNRC Region 1Westinghouse PWR 2-LoopEmergency Assessment Capability cannot be performed in the Technical Support Center due to an equipment deficiency in the HVAC system which could impact facility habitability. An Alternate Technical Support Center is in place at the Emergency Offsite Facility. Priority maintenance is in progress to correct the deficiency. The licensee notified the NRC Resident Inspector.
ENS 532391 March 2018 20:05:00Point BeachNRC Region 3Westinghouse PWR 2-LoopDuring review of protection of equipment from damaging effects of tornados, Point Beach Nuclear Plant identified a potential vulnerability for the turbine driven auxiliary feedwater pumps due to steam supply piping that is not routed through a Class 1 structure. Immediate compensatory measures were taken to mitigate the potential consequences of a tornado generated missile impact. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) and per 10 CFR 50.72(b)(3)(v)(A) and (D). The identified vulnerability is being addressed in accordance with EGM 15-002 and DSS-ISG-2016-01, enforcement discretion memorandum and interim guidance document for resolution of noncompliance with tornado-generated missile protection. The NRC Resident Inspector has been notified.
ENS 5318526 January 2018 13:06:00GinnaNRC Region 1Westinghouse PWR 2-LoopOn January 26, 2018, a containment entry was made to identify the source of elevated Unidentified Reactor Coolant System (RCS) operational leakage. A through-wall leak was identified on a Class 1 piping weld on the letdown line at 0853 EST. It was determined that the leak was RCS pressure boundary leakage. Ginna entered Technical Specification (TS) LCO (Limiting Condition for Operation) 3.4.13, RCS Operational Leakage, Condition B. for the existence of pressure boundary leakage. This condition requires the plant to be in MODE 3 within 6 hours and MODE 5 within 36 hours. The leak was isolated and TS LCO 3.4.13 exited at 1015 EST. This event is reportable within 8 hours in accordance with 10CFR50.72(b)(3)(ii)(A) for 'Any event or condition that results in: (A) The condition of the nuclear power plant, including its principal safety barriers, being seriously degraded'. The Station (Ginna) is developing an evaluation and a repair plan at this time. This condition has no impact on public health and safety. The licensee has informed 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 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 5312417 December 2017 11:56:00GinnaNRC Region 1Westinghouse PWR 2-LoopGinna notified New York State Department of Environmental Conservation of a sulfuric acid spill of approximately 270 gallons in the AVT, All Volatile Treatment, building. Ginna is currently contacting offsite support for hazardous chemical cleanup. The spilled sulfuric acid is currently contained within the secondary containment structure associated with the sulfuric acid tank. There is no release to the environment. There is no impact to habitability in the AVT building at this time. The licensee notified the NRC Resident Inspector.
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 5304230 October 2017 12:48:00Point BeachNRC Region 3Westinghouse PWR 2-LoopDuring a scheduled refueling outage, an inspection of components inside containment revealed a suspected weld defect on 1CV-309B, 1P-1B RCP Labyrinth Seal 1PT-124 Upper Root. 10 CFR 50.2 (2)(i) defines the reactor coolant pressure boundary as being connected to the reactor coolant system, up to and including the outermost containment isolation valve in system piping which penetrates primary reactor containment. The weld defect is located on the transmitter side of 1CV-309B. This can be isolated from the RCS by shutting 1CV-309B and 1CV-308B, 1P-1B RCP Labyrinth Seal 1PT-124 Lower Root. Based on the definition provided in 10 CFR 50.2, the condition is considered reportable under 50.72(b)(3)(ii). Unit 1 is currently in mode 3. Repairs for the condition are being determined. The NRC Resident Inspector has been notified.
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 530049 October 2017 09:15:00Point BeachNRC Region 3Westinghouse PWR 2-Loop

At 0737 CDT on 10/9/17, Point Beach declared an Unusual Event with Emergency Action Level HU 3.1 due to report of toxic gas from a spill in a service building within the protected area. The spill is contained and cleanup operations are in progress. The spill was not in a contaminated area or vital area. The janitorial worker injured while mixing cleaning chemicals in a closet was taken to the hospital. The licensee notified the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 10/9/17 AT 1111 EDT FROM DENNY SMITH TO BETHANY CECERE * * *

Point Beach has terminated the Unusual Event at 0944 (CDT) on 10/9/2017. The Unusual Event condition is no longer warranted. The NRC Resident Inspector has been notified. Notified R3DO (Hills), NRR EO (King), IRD (Stapleton), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

ENS 5297619 September 2017 00:01:00Point BeachNRC Region 3Westinghouse PWR 2-LoopAt 1724 (CDT) on 9/18/17 during Control Room Ventilation testing Door-61, South Control Room Door, became wedged against its door stop and stuck open. Door-61 is a credited High Energy Line Break (HELB) / Fire / Flood Barrier in addition to its function to maintain the Control Room envelope. The door stop was subsequently unbolted from the floor and the door was free to close. Door-61, South Control Room Door, has since been inspected, and at 1750 (CDT), was declared functional as a HELB / Fire / Flood Barrier and Operational for purposes of maintaining the Control Room Envelope. During the 26 minutes the door was stuck open, the Control Room was in an unanalyzed condition with regards to protection from a High Energy Line Break. The licensee notified the NRC Resident Inspector.
ENS 529462 September 2017 19:36:00GinnaNRC Region 1Westinghouse PWR 2-LoopMCR (Main Control Room) area radiation monitor R-1 failed at 1148 (EDT on) 9/2/2017. This caused a loss of capability to classify EAL (Emergency Action Level) RA3.1, Dose Rates greater than 15 mrem/hr in either of the following areas requiring continuous occupancy to maintain plant safety functions: Control Room (R-1) or CAS (Central Alarm Station). Compensatory measures are currently in place with a portable radiation monitor in the MCR with alarm setpoints consistent with R-1. Priority maintenance is being planned to restore R-1 to service. The licensee will notify 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 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 5263623 March 2017 21:06:00Point BeachNRC Region 3Westinghouse PWR 2-LoopOn 3/23/17, at 0325 hours CDT, it was discovered that a prohibited item was present in the protected area from 0508-1718 hours on 3/22/17, which resulted in a reportable condition pursuant to 10 CFR 26.719(b)(1). The licensee has notified the NRC Resident Inspector.
ENS 5262720 March 2017 17:53:00Point BeachNRC Region 3Westinghouse PWR 2-Loop

At 1620 (CDT), an unusual event was declared due to a smoke detector alarm in Unit 1 containment. (There were) no indications of any other detector alarms, no abnormal equipment indications, and containment parameters are normal (temperature, humidity). At 1631 (CDT), visual inspection (of the) 66 ft. hatch indicated no smoke or abnormal smell. At 1640 (CDT), local inspection of Unit 1 containment verified no fire or hot spots. The licensee has notified the NRC Resident Inspector. Notified DHS SWO, FEMA Ops Center, DHS NICC. Notified FEMA National Watch and Nuclear SSA via E-mail.

  • * * UPDATE FROM RYAN RODE TO DONG PARK AT 2208 EDT ON 3/20/2017 * * *

Event transmitted under ENS # 52627 is terminated at 2022 (CDT on) 3/20/17." The NRC Resident Inspector has been notified. Notified R3DO (Orlikowski), 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 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 519085 May 2016 11:36:00Point BeachNRC Region 3Westinghouse PWR 2-LoopAn individual failed to comply with the NextEra Energy fitness-for-duty policy during a follow-up fitness-for-duty test. The individual's access to the plant has been terminated. This is reportable under 10 CFR 26.719(b)(2)(ii). The licensee has notified the NRC Resident Inspector.
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 5179215 March 2016 15:14:00Point BeachNRC Region 3Westinghouse PWR 2-LoopDuring a scheduled refueling outage, an inspection of containment components revealed a suspected through wall leak on 1CV-200B, Letdown Orifice 'B' Outlet Control. Non-destructive engineering inspection has been completed and determined that an indication exists. 10 CFR 50.2(2)(i) defines the reactor coolant pressure boundary as being connected to the reactor coolant system, up to and including the outermost containment isolation valve in system piping which penetrates primary reactor containment. 1CV-200B is isolable from the Reactor Coolant System (RCS) by a single motor operated valve, 1RC-427, Reactor Coolant Loop 'B' Leg to CVCS Letdown Isolation valve. 1CV-200B is located inside of containment between 1RC-427 and the two containment isolation valves for the letdown line, 1CV-371 and 1CV-371A. Based on the definition provided in 10 CFR 50.2, the condition is considered pressure boundary leakage and is considered reportable under 10 CFR 50.72(b)(3)(ii). Unit 1 is currently in Mode 6. Repairs for the condition are being determined. The NRC Resident Inspector has been notified.
ENS 5173012 February 2016 03:51:00GinnaNRC Region 1Westinghouse PWR 2-LoopOn 02/11/2016, at 2305 (EST), Ginna Station experienced a loss of Station Service Transformer 12A causing Emergency Diesel Generator 1A to automatically start due to under-voltage signals to safeguards buses 14 and 18. All plant systems responded as designed. Control room operators stabilized the plant per abnormal operating procedures. The plant is currently in a 100/0 electrical lineup (supplied by the 12B Service Station Transformer) on the off site circuit 767 with the 1A Emergency Diesel Generator secured. The loss of the station service transformer is currently under investigation. This is reportable as a valid system actuation that was not part of a pre-planned sequence during testing. The NRC Resident Inspector has been 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 5160414 December 2015 14:32:00GinnaNRC Region 1Westinghouse PWR 2-LoopThis report is being made per paragraphs 50.73(a)(1) and 50.73(a)(2)(iv)(A) to address an actuation of Emergency Diesel Generator 'A' on October 21, 2015, during the performance of a Diesel Generator Load and Safeguard Sequence Test. The Emergency AC Electrical Power system, including Diesel Generators is a system named in 50.73(a)(2)(iv)(B). During the performance of the Diesel Generator Load and Safeguard Sequence Test restoration steps a human performance error, while taking resistance readings on a relay, resulted in the unintentional start of the 'A' Emergency Diesel Generator. The testing was aborted and the affected Diesel Generator was restored to standby service in accordance with plant procedures. This is defined as an 'invalid signal' in that the 'A' Diesel Generator did not start as the result of an actual initiating condition. The start signal is considered an INVALID signal with respect to 50.73(a)(2)(iv)(A), however the system was not fully removed from service. The 'B' train was not affected by this event. The actuation signal was considered complete since all necessary components responded as would have been expected if there had been a valid signal. The NRC Resident Inspector was notified. The licensee has determined that shorter test probe tips, when taking resistance readings, are necessary to prevent reoccurrence of this event.
ENS 5157028 November 2015 23:54:00Point BeachNRC Region 3Westinghouse PWR 2-LoopUnit 1 automatic reactor trip actuated due to an automatic voltage regulator (AVR) malfunction which caused a generator lockout and turbine trip. The cause of the AVR malfunction is unknown at this time. All control rods fully inserted. The RCS is being cooled by forced flow (reactor coolant pumps). Secondary heat sink is being provided by the condenser steam dumps utilizing the main feed water system. The auxiliary feed water system actuated based on low steam generator level, but since has been secured. Off-site power remains available. No release is occurring and emergency core cooling systems did not actuate. Emergency plan entry was not required. The plant is in its normal shutdown electrical lineup at normal operating temperature and pressure. Unit 2 was not affected by the Unit 1 transient. The licensee has notified the NRC Resident Inspector
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 5150629 October 2015 10:38:00Point BeachNRC Region 3Westinghouse PWR 2-LoopAt 0348 CDT, while Point Beach Unit 2 was performing outage activities, it experienced a Main Power Transformer lockout and associated loss of busses (2A-01, 2A-02, 2B-01 and 2B-02). The loss of the two non-vital 4160 V buses resulted in actuation of the Unit 2 undervoltage logic which resulted in actuation of the Auxiliary Feedwater System. The Auxiliary Feedwater System functioned normally upon actuation. This condition was determined to be reportable per 10CFR50.72(b)(3)(iv)(A)(6), PWR auxiliary or emergency feedwater system actuation. This event did not affect the operating Unit 1. The NRC Senior Resident Inspector has been notified.
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 5130711 August 2015 02:01:00GinnaNRC Region 1Westinghouse PWR 2-LoopAt approximately 2332 EDT on August 10, 2015, the Ginna Control Room was notified of an inadvertent siren activation by the Monroe County Emergency Center. It is unclear at this time why the siren inadvertently activated. Company personnel are addressing the issue with the siren. The licensee notified the NRC Resident Inspector. The siren activated for approximately 1 minute. The licensee will remove the siren from service until the cause of the inadvertent actuation can be corrected. The licensee has a sufficient number of sirens to allow this siren to be removed from service.
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 5122513 July 2015 15:45:00GinnaNRC Region 1Westinghouse PWR 2-LoopAt approximately 1210 (EDT) on July 13, 2015 during conduct of vendor maintenance, a contract maintenance worker inadvertently activated siren 71. The licensee was notified of the siren activation by the vendor at 1211 (EDT). Wayne County was notified of the siren activation by the vendor at 1212 (EDT). One of the 96 sirens in the 10-mile Emergency Planning Zone (EPZ) were activated for less than one minute. No press release is planned by Exelon. The NRC Resident Inspector has been notified.
ENS 5119430 June 2015 20:05:00Point BeachNRC Region 3Westinghouse PWR 2-LoopThis notification is being made in accordance with NUREG-1022, Event Report Guidelines 10 CFR 50.72 and 50.73 Section 3.2.12, News Release or Notification of Other Government Agency. On June 30, 2015 at 1256 (CDT) an employee was injured while conducting a work activity and a non-contaminated hospital transport was completed. 29 CFR 1904.39(a) requires a report to the Occupational Safety and Health Administration (OSHA), US Department of Labor within twenty four (24) hours after the in-patient hospitalization of one or more employees as a result of a work-related incident. At 1635 CDT, it was determined that this is a 24-hour OSHA reportable occurrence. The licensee has notified the NRC Resident Inspector.
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).