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 Entered dateSiteRegionReactor typeEvent description
ENS 542036 August 2019 17:29:00PerryNRC Region 3At 1335 EDT on 08/06/2019, it was discovered that reactor protection system (RPS) instrumentation functions for turbine stop valve closure and turbine control valve fast closure, end of cycle recirculation pump trip (EOC-RPT) instrumentation, and control rod block instrumentation were simultaneously inoperable due to a loss of feedwater heating; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). RPS instrumentation, EOC-RPT instrumentation, and control rod block instrumentation functions were restored at 1422 EDT on 08/06/2019. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified."
ENS 5418527 July 2019 22:31:00PerryNRC Region 3At 1929 EDT on 7/27/2019, with the Unit in Mode 1 at 98 percent power, the reactor automatically scrammed due to a Main Turbine Trip. The trip was not complex, with all systems responding normally post-trip. Main Steam Isolation Valves (MSIVs) were manually closed to prevent exceeding Reactor Pressure Vessel Cooldown Rate. Rector Core Isolation Cooling (RCIC) was manually initiated to stabilize Reactor Vessel Water Level and Pressure following MSIV closure. The Main Condenser and Feedwater are available. Operations responded and stabilized the plant. Reactor water level is being maintained via RCIC. Decay heat is being removed by discharging steam to the Main Condenser and RCIC. The cause of the Main Turbine Trip is currently under investigation. The site is in a normal electrical lineup. The licensee notified the NRC Resident Inspector.
ENS 5418326 July 2019 13:27:00Beaver ValleyNRC Region 1A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector.
ENS 5408524 May 2019 13:09:00PerryNRC Region 3At 0730 (EDT) on May 24, 2019, it was discovered that the Low-Pressure Core Spray System was inoperable. At Perry, the Low-Pressure Core Spray system is considered a single train system in Modes 1, 2, and 3; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). Inoperability of the Low-Pressure Core Spray system was caused by Emergency Service Water Pump A inoperability. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."
ENS 5392411 March 2019 10:00:00Beaver ValleyNRC Region 1On 03/11/2019 planned maintenance activities will be performed on the Beaver Valley Power Station (BVPS) Unit 2 Digital Radiation Monitoring System's (DRMS) Communications System. The work includes upgrades to the DRMS hardware, software and computer peripherals. Components to be upgraded under this planned maintenance include: Redundant Servers, Operator Console, Health Physics Office Console/Workstation, Printers and Portable Mass Storage/Backup and Computer peripherals necessary to interface with the computer system. This planned upgrade on the Unit 2 DRMS Communications System will result in the loss of Unit 2 radiological monitoring capability in the Control Room and on the Plant Computer System (PCS). Neither the Emergency Response Facilities nor the Emergency Response Data System (ERDS) will receive radiological data. No actual radiation monitors are affected. The scheduled work duration is approximately three weeks. When the DRMS is out of service for the upgrade, compensatory actions will be in place. Radiation monitors will be continuously monitored for any increases in radiation levels. The Unit 2 Shift Manager will be notified of any increase in radiation monitor readings, including exceeding Emergency Action Levels (EAL). This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the upgrade to the DRMS Communications System will result in the loss of the Unit 2 radiological monitoring capability in the Control Room and to the Unit 2 Plant Computer System that affects the functionality of an Emergency Response Facility. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee is currently in a Technical Specification Action Statement, Limiting Condition for Operation for spent fuel pool radiation monitor and leakage rate monitor.
ENS 5389625 February 2019 03:44:00PerryNRC Region 3At 0024 EST on 2/25/19, with Unit 1 in Mode 1 at 74 percent power, the reactor automatically tripped due to a generator trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via the feed system. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The generator trip is under investigation, but is believed to be due to grid perturbations.
ENS 537111 November 2018 15:38:00Beaver ValleyNRC Region 1On 11/01/2018, during the Beaver Valley Power Station Unit No. 2 (BVPS-2) refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per ASME Code Section XI. Penetration 27 will require repair prior to returning the vessel head to service. The indication was not through wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-4 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. The other 65 penetrations will be examined during the 2R20 (current) refueling outage. The plant is currently shutdown and in Undefined Mode. The reactor vessel head is not currently installed. Repairs are currently being planned and will be completed prior to startup. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as-found indications did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-4 to remain in-service without repair. The NRC Resident Inspector has been notified."
ENS 5367519 October 2018 12:04:00PerryNRC Region 3

During extent of condition review of a previously identified fire induced hot-short (Ref. EN#53644) an unfused circuit associated with the 0M23C0002A, Miscellaneous Switchgear Recirculation Fan was discovered. This condition is not bounded by existing design and licensing documents. This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit. Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed, challenging the ability to achieve and maintain safe shutdown. The postulated event would affect multiple fire zones in the control complex. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector.

  • * * UPDATE ON 10/19/2018 AT 1454 EDT FROM EDWARD CONDO TO ANDREW WAUGH * * *

Further extent of condition reviews have discovered another unfused circuit. The circuitry is related to 0M24C001A, Battery Room Exhaust Fan. This condition is not bounded by existing design and licensing documents. This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Senior Resident Inspector. Notified R3DO (Hills).

ENS 536444 October 2018 12:29:00PerryNRC Region 3Degraded or unanalyzed condition due to the possibility for a postulated fire induced hot short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in calculation SSC-001 due to an unfused circuit associated with the 1M43C0001A, Diesel Generator Building Ventilation Fan. This condition is not bounded by existing design and licensing documents. Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit without protection that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed challenging the ability to achieve and maintain safe shutdown. The postulated event would affect the following fire zones: 1CC-3c (Unit 1, Division 1 4160V and 480V Switchgear Room, 620 feet 6 inch elevation), 1CC-3e (Unit 1 West Corridor North of Elevator, 620 feet 6 inch elevation), DG-1d (Hallway Diesel Generator Building 620 feet 6 inch elevation), and 1DG-1c (Unit 1, Division 1 Diesel Generator Building 620 feet 6 inch elevation). This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Senior Resident Inspector."
ENS 5354812 August 2018 02:58:00Beaver ValleyNRC Region 1

EN Revision Text: TECHNICAL SPECIFICATION REQUIRED SHUTDOWN - LOSS OF 480 VOLTAGE EMERGENCY BUS On 8-12-18 at 0158 EDT, Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. This resulted in a Loss of Safety Function due to the 2-2 Emergency Diesel Generator (EDG) being Inoperable coincident with the Residual Heat Release Valve (2SVS-HCV104). A Technical Specification shutdown is required per LCO 3.0.3. The Licensee also stated they were in an unanalyzed condition due to the EDG and Residual Heat Release Valve being inoperable at the same time. The Licensee is shutting down to Mode 5 (Cold Shutdown). The Licensee is notifying the Resident Inspector. The Licensee will be making a Press Release about the unplanned shutdown.

  • * * UPDATE ON 08/16/2018 AT 1424 EDT FROM BLASE BARTKO TO KEN MOTT * * *

On 8-12-18 at 0158 (EDT) Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. Per operational guidance, this was determined to be a Loss of Safety Function due to the Unit 2 Emergency Diesel Generator (EDG) being INOPERABLE coincident with the Residual Heat Release Valve (2SVS-HCV104) 10 CFR 50.72(b)(3)(v)(B) and (D). This was also reported as an Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(b). No Press Release was performed for this event. The NRC Resident Inspector was notified. At 0410 (EDT) a Technical Specification Shutdown was commenced 10 CFR 50.72(b)(2)(i). At 2011 (EDT) the 480 Volt 2P Emergency Bus was restored and energized. Further evaluation of the event has determined that this event was not an Unanalyzed Condition and did not result in a Loss of Safety Function. The classifications of Unanalyzed Condition and Loss of Safety Function are being retracted. The accuracy of the existing guidance relative to Safety Function has been entered in the Corrective Action Program and interim actions have been taken to provide accurate guidance. Notified R1DO (Young) via email.

ENS 534811 July 2018 08:24:00PerryNRC Region 3On July 1st, 2018 at 0100 (EDT), a portion of the Division 1 Emergency Core Cooling System (ECCS) Loss Of Coolant Accident (LOCA) initiation logic was declared inoperable due to the discovery of a blown fuse. The fuse was replaced at 0215 on July 1st, 2018 and the Division 1 ECCS LOCA initiation logic was declared operable at 0230 on July 1st, 2018. The blown fuse caused the loss of a portion of the Division 1 ECCS LOCA initiation logic which would have prevented the initiation of the Emergency Closed Cooling (ECC) A system. ECC A and supported systems were declared inoperable. Low Pressure Core Spray (LPCS) was one of the supported systems that were declared inoperable. LPCS is considered a single train safety system. Inoperability of LPCS is considered an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. The blown fuse also caused the loss of a portion of the Division 1 ECCS LOCA initiation logic which would have prevented the automatic isolation of Nuclear Closed Cooling and Instrument Air to the Containment. The loss of Containment isolation capability is considered an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) The NRC Senior Resident Inspector has been notified."
ENS 5341922 May 2018 18:38:00Beaver ValleyNRC Region 1

EN Revision Text: GAS VOIDS DISCOVERED IN BOTH TRAINS OF LOW HEAD SAFETY INJECTION On 5/22/2018, while operating at approximately 100 percent power, Ultrasonic Testing of the Beaver Valley Power Station (BVPS) Unit 1 Low Head Safety Injection (LHSI) pump suction piping identified gas voids in excess of the acceptable limit for void volume. Both trains of LHSI were declared inoperable. Technical Specification (TS) 3.5.2 for both trains of the LHSI system was entered along with TS 3.0.3 which requires the initiation of a plant shutdown. Time of TS entry was 12:56 (EDT). Plant shutdown was commenced at 15:56 (EDT) in accordance with plant procedures. At 15:59 (EDT) Train 'A' LHSI was restored to operable status, TS 3.0.3 Action was exited and the power reduction was stopped at approximately 99 percent. At 17:43 (EDT) Train 'B' LHSI was restored to operable status, TS 3.5.2 Actions were exited. This is reportable per 10 CFR 50.72(b)(3)(ii) Unanalyzed Condition, 10 CFR 50.72(b)(3)(v) Event or Condition that Could Have Prevented the Fulfillment of a Safety Function and 10 CFR 50. 72(b )(2)(i) TS Required Shutdown. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 6/21/18 AT 1535 EDT FROM SHAWN KEENER TO RICHARD SMITH * * *

Further engineering evaluation has determined that the gas voids that existed at the time of discovery would not have rendered the LHSI (Low Head Safety Injection) system inoperable if it were required to actuate. The engineering evaluation concluded that filling of the containment sump during a Design Basis Accident would result in a void volume reduction such that the void in the LHSI suction piping would not be large enough to significantly impact the operability of the system. Therefore, the system remained operable but degraded. No TSs (Technical Specifications) were required to be entered and no shutdown was required. As such, all three reporting criteria do not apply and are being retracted. The NRC Resident Inspector has been notified. Notified R1DO (Burritt).

ENS 5322823 February 2018 20:29:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 1750 (EST) on 2/23/2018 credible information was obtained that a non-licensed supervisory person had intentionally misused a controlled substance. Unescorted access has been withdrawn. Per 10 CFR 26.719(b)(2)(i) this is a 24 hour reportable event. The NRC Resident Inspector has been notified.
ENS 5313723 December 2017 05:00:00PerryNRC Region 3GE-6High Pressure Core Spray System was declared inoperable due to the discovery of a through-wall leak on the Minimum Flow line. Leak rate is 60 drops per minute from ASME Class 2 Piping. The leak has been isolated and the High Pressure (Core Spray) System has been placed in Secured Status. High Pressure Core Spray is considered a single train safety system. Inoperability of (the) High Pressure Core Spray System is considered an event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Senior Resident Inspector was notified. Technical Specifications Limiting Condition for Operation 3.5.1 Condition B was entered, requiring restoration of the High Pressure Core Spray System in 14 days. The licensee plans to notify State and Local Governments (Lake, Geauga, and Ashtabula Counties).
ENS 530587 November 2017 14:19:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopIt was determined that a licensed employee was not placed in an additional Fitness-For-Duty (FFD) testing program as directed by the MRO (Medical Review Officer). Upon discovery, the individual was subsequently tested . . . and determined Fit-For-Duty per 10 CFR 26 requirements. This event is being reported pursuant to 10 CFR 26.719(b)(4). The NRC Resident Inspector has been notified.
ENS 530567 November 2017 08:29:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopOn November 7, at 0504 (EST), BVPS (Beaver Valley Power Station) Unit 1 experienced an automatic reactor trip due to Main Unit Generator over current. The Auxiliary Feedwater system activated and remains in service. Offsite power supply is available. Normal and Emergency busses are being supplied by Offsite power. One Source Range channel failed to energize due to its corresponding Intermediate Range instrument being under compensated. It was manually energized and is not indicating as expected. The second Source Range instrument energized but is reading erratically. Both Source Range instruments have been declared inoperable and the appropriate Technical Specification has been complied with by making the Control Rods not capable of withdrawal and isolating all dilution flow paths. Plant trip response was as expected without complications, and all control rods fully inserted in the core. The plant is currently stable in Mode 3. This event is being reported as an actuation of the Reactor Protection system 10 CFR 50.72(b)(2)(iv)(B) and a Specified System Actuation (Auxiliary Feedwater System) 10 CFR 50.72(b)(3)(iv)(A). BVPS Unit 2 is unaffected by this event and remains at 100% power in Mode 1. The NRC Resident Inspector has been notified.
ENS 530004 October 2017 05:53:00PerryNRC Region 3GE-6

On October 4, 2017, at 0250 hours (EDT), the Perry Nuclear Power Plant commenced a Technical Specification (TS) shutdown by lowering reactor power from 100 percent rated thermal power to 98 percent to comply with TS LCO 3.0.3. Reactor power was further reduced to 82 percent rated thermal power at 0430 hours (EDT). The plant had entered TS 3.0.3 at 0155 hours (EDT) upon loss of MCC (Motor Control Center), Switchgear, and Miscellaneous Electrical Equipment Areas HVAC System train A while train B was removed from service for maintenance. MCC switchgear ventilation train A was declared inoperable based on excessive belt noise and a dropped belt on MCC switchgear supply fan A. This also constitutes a loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(2)(i) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector was notified.

  • * * UPDATE ON 10/04/17 AT 0926 EDT FROM DAN HARTIGAN TO STEVEN VITTO * * *

Due to the loss of both trains of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC, actions were taken in LCO 3.8.7 for AC and DC Distribution Systems, LCO 3.8.4 for DC Sources, LCO 3.8.1 for AC Sources, and the associated support systems, the High Pressure Core Spray system was also declared inoperable, which is a single train safety system and therefore, an additional loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B), 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D). At 0620 hours (EDT) the A train of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC and High Pressure Core Spray was declared operable and LCO 3.0.3 was exited. The plant was restored to 100% (percent) power at 0804 (EDT). The NRC Resident Inspector was notified. Notified R3DO(Hills).

ENS 5290315 August 2017 04:58:00PerryNRC Region 3GE-6On August 14th, 2017 at 2257 (EDT), while shutting down the Annulus Exhaust Gas Treatment System (AEGTS) Train B, secondary containment pressure momentarily lowered. This resulted in the Technical Specification (TS) for Secondary Containment to not be met for 15 seconds. The minimum Secondary Containment vacuum observed during that time was 0.52 inch of vacuum water gauge. Secondary Containment pressure was returned to within the TS operability limit of 0.66 inch of vacuum water gauge (TS SR 3.6.4.1.1) by the AEGTS Train A that remained in operation. There were no radiological releases associated with this event. Declaring Secondary Containment inoperable is reportable under (10 CFR) 50.72(b)(3)(v)(C) & (D) for an event or condition that could have prevented the fulfillment of a safety function of a system that is needed to:(C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector.
ENS 528918 August 2017 20:22:00PerryNRC Region 3GE-6On August 8, 2017, at 1554 hours (EDT), during restoration from testing of the High Pressure Core Spray (HPCS) Suppression Pool Level High Instrumentation, unexpected as-left indications were found that impacted both of the required channels of instrumentation. Subsequent venting of the instrumentation lines was completed and both channels of instrumentation are reading consistent with previously taken as-found data. The instrumentation was declared OPERABLE at 1635. The initial cause of the unexpected as-left indications appears to be the introduction of air into the instrumentation lines during the calibration activities. This is considered a loss of safety function based on both of the HPCS Suppression Pool Level High Instrumentation channels being declared INOPERABLE and the loss of the automatic HPCS suction swap to the Suppression Pool on a high level. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The (NRC Resident Inspector) has been notified.
ENS 5287728 July 2017 16:40:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopOn July 20, 2017, a fire protection supply line at BVPS (Beaver Valley Power Station) Unit 1 experienced a mechanical joint separation. When the mechanical joint separated there was an unplanned flow of water from the fire protection line to the ground, which eventually re-entered the Ohio River via storm drains. The source of the water to the BVPS Unit 1 Fire Protection System is the Ohio River. The fire protection water was not treated with chemical additives. The water flow from this fire protection line was terminated on the same day. Due to the situation described above, FENOC (FirstEnergy Nuclear Operating Company) provided written documentation describing this occurrence to the Pennsylvania Department of Environmental Protection, (PADEP) on July 28, 2017. This offsite notification to PADEP requires a 4-hour report per 10 CFR 50.72 (b)(2)(xi). No press release is planned. The NRC Resident Inspector has been notified.
ENS 527273 May 2017 12:59:00PerryNRC Region 3GE-6On April 30, 2017, at 1818 (EDT), the main turbine steam bypass valve #1 partially opened. Power was incrementally lowered. While lowering power the bypass valve would shut and then reopen and power would again be lowered. When power was lowered to approximately 74 percent the bypass valve remained closed. During the transient the reactor protection system (RPS) Turbine Stop Valve Closure and Control Valve Fast Closure trip functions were declared inoperable due to the opening of the bypass valve which affects the bypass setpoint for those RPS trip functions. With the loss of these RPS trip functions a loss of safety function existed intermittently for approximately 37 minutes. The manual reactor trip function and other RPS functions remained operable. Both channels of the rod withdrawal limiter (RWL) and the end of cycle reactor recirculation pump trip (EOC-RPT) function were also declared inoperable. These functions are credited in accident analysis, this also resulted in a loss of safety function. Currently the bypass valve is closed and the RWL, EOC-RPT and RPS function are operable. Troubleshooting continues to determine the issue with the main turbine that caused the bypass valve to open. NRC Resident Inspector has been notified.
ENS 5257123 February 2017 13:02:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopIn order to address the concerns outlined in RIS (Regulatory Information Summary) 2015-06 'TORNADO MISSILE PROTECTION', an evaluation of tornado missile vulnerabilities and their potential impact on Technical Specification (TS) plant equipment was conducted. This evaluation concluded that the following Structures, Systems, and Components (SSCs) are potentially vulnerable to tornado generated missiles: The BVPS Unit 1 (BV-1) and BVPS Unit 2 (BV-2) Main Steam Safety Valve (MSSVs) discharge flow paths to atmosphere (reference TS 3.7.1) are potentially vulnerable to tornado generated missiles. A tornado could generate missiles capable of striking the exhaust piping of the MSSVs potentially crimping the piping and resulting in reduced flow capacity. In the worst case, all MSSV's could be rendered inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(A). The BV-1 and BV-2 Atmospheric Steam Dumps (ADVs) discharge flow paths to atmosphere (reference TS 3.7.4) are potentially vulnerable to tornado generated missiles. A tornado could generate missiles capable of striking the exhaust piping of the ADVs potentially crimping the piping and resulting in reduced flow capacity. In the worst case, all ADVs could be rendered inoperable. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(A). The BV-2 Auxiliary Building tornado missile shield door (A-35-5A), credited for tornado missile protection of the Primary Component Cooling Water (PCCW) system, was found to not be fully closed and latched. A tornado could generate missiles capable of striking the PCCW system with the missile door open rendering both trains of the PCCW system inoperable. This door is now maintained closed and latched except when opened under administrative controls. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B). The BV-2 Spent Fuel Building tornado missile shield door (F-66-3), credited for tornado missile protection of the irradiated fuel assemblies in the Spent Fuel Pool including the Cask Pit, was found to not be fully closed and latched. This door is now maintained closed and latched except when opened under administrative controls. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B). The potential tornado missile vulnerabilities for the MSSVs and ADVs (discussed above) are being addressed in accordance with EGM-15-002 Rev 1 and DSS-ISG-2016-01 (NRC enforcement discretion and interim guidance documents). Immediate compensatory measures were taken to reduce the likelihood and mitigate the potential consequences of an onsite tornado generated missiles. The NRC Resident inspector has been notified.
ENS 5251327 January 2017 07:25:00Beaver ValleyNRC Region 1Westinghouse PWR 3-Loop

An Unusual Event (HU.4) was declared at BVPS (Beaver Valley Power Station) Unit 1 at 0641 EST due to a fire alarm in Containment which was unable to be verified extinguished within 15 minutes. Current Containment parameters do not indicate an actual fire. The Fire Alarm has subsequently been reset (at 0648 EST). The Licensee notified the NRC Resident Inspector." The Licensee has notified State and local government agencies. A containment entry is planned for 1000 EST to verify that there is no fire. Unit 1 continues at 100 percent power. Unit 2 was unaffected. Notified DHS SWO, FEMA Ops Center, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA.

  • * * UPDATE FROM MICHAEL SMITH TO STEVEN VITTO ON 1/27/2017 AT 0952 EST * * *

The time of the fire alarm reset was at 0648 EST on 1/27/2017. The Unusual Event has been terminated at 0934 EST on 1/27/2017 after a containment entry and visual inspection determined no fire or indications of a fire exist. The following notifications were made: Commonwealth of Pennsylvania Emergency Management Agency, Beaver County in Pennsylvania, The State of Ohio Emergency Management Agency, Columbia County in Ohio, The State of West Virginia Emergency Management Agency and Hancock County in West Virginia. The Licensee notified the NRC Resident Inspector. Notified R1DO (Welling), NRR EO (Miller), and IRD MOC (Gott). Notified DHS SWO, FEMA Ops Center, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA.

ENS 5246830 December 2016 14:40:00PerryNRC Region 3GE-6On December 28, 2016 at 2119 EST, the Standby Liquid Control system (SLC) subsystem A was declared inoperable in accordance with the surveillance instruction for performance of a routine surveillance test. At 2229 EST, control room operators received an out-of-service alarm for the explosive-actuated injection valve for SLC subsystem B and declared subsystem B inoperable, thereby rendering both subsystems inoperable. With both subsystems inoperable, the SLC system was unable to fulfill its safety function. At 2335, the surveillance was completed and subsystem A was restored to operable status, which restored the ability for the system to fulfill its safety function. Troubleshooting determined that the cause for subsystem B inoperability was an intermittent electrical connection for the explosive-actuated injection valve. Repairs were conducted and the subsystem was restored to operable status on December 29, 2016 at 1708 EST. This issue was entered into the Corrective Action Program and during post reportability review, it was determined that this was a reportable event under 10 CFR 50.72(b)(3)(v)(A) for an event or condition that could have prevented the fulfillment of a safety function of a system that is needed to shut down the reactor and maintain it in a safe shutdown condition and under 10 CFR 50.72(b)(3)(v)(D) for a system that was unavailable for accident mitigation. The NRC Resident Inspector has been notified.
ENS 5239022 November 2016 18:44:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopDuring routine floor cleaning activities inside of the BVPS (Beaver Valley Power Station) Main River Water Intake Structure, approximately one and one half cups (approximately 12 oz.) of Simple Green cleaning solution diluted in three gallons of mop cleaning water was inadvertently released into the Ohio River. Subsequent to confirmation of the inadvertent release to the Ohio river, a notification was made to the Pennsylvania Department of Environmental Protection, (PADEP). This offsite notification to PADEP requires a 4-hour report per 10 CFR 50.72(b)(2)(xi). Actions were taken to preclude any further liquid cleaning solution to be released, and no detectable impact to the Ohio River was observed. The amount of cleaning solution does not exceed any Reportable Quantity limit that would be hazardous to the environment or personal health. No press release is planned. The NRC Resident Inspector has been notified.
ENS 5213229 July 2016 01:51:00Beaver ValleyNRC Region 1Westinghouse PWR 3-Loop

A review of the Beaver Valley Unit 2 Fire Protection Safe Shutdown Report (FPSSR) found that a postulated fire had the potential to spuriously open all three individual steam generator atmospheric dump valves in addition to a common residual heat release valve. Previous analysis did not consider all of the valves spuriously opening from a fire. The potential impact of these valves spuriously opening is a cooldown that could adversely affect shutdown margin. Hourly fire tours have been put in place for those fire areas that have the potential to initiate this condition. This condition is reportable as an 8 hour report in accordance with 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector was notified. This condition is not applicable to Unit 1.

  • * * RETRACTION FROM DAN SCHWER TO VINCE KLCO ON 9/26/16 AT 1428 EDT * * *

Retraction of EN 52132 'Postulated Fire Event That Could Adversely Impact Safe Shutdown Equipment'. On 07/29/2016, an 8-hour notification (EN 52132) was made describing the discovery of a postulated fire event that could adversely impact safe shutdown equipment, specifically the spurious opening of the three Atmospheric Steam Dump Valves (ASDVs) and the Residual Heat Release (RHR) Valve, simultaneously. This notification was made pursuant to 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition because of the potential effect on shutdown margin. Further engineering evaluation has determined that the required shutdown margin is not challenged by the event, as bounded by previous analysis of a similar scenario as well as validated operator actions. Therefore, this does not result in a reportable condition. The Resident NRC Inspector has been notified. Notified the R1DO (Cook).

ENS 5194820 May 2016 18:40:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt approximately 1531 (EDT) hours, a tornado signal was inadvertently activated from Hancock County, West Virginia which sounded 118 out of 120 Emergency sirens. The remaining 2 sirens do not receive a tornado signal. The sirens alarmed for approximately 3 minutes. A silent test was subsequently performed for all 120 sirens and validated the emergency siren system remains functional. In accordance with 10 CFR 50.72(b)(2)(xi), News Release or Notification of Other Government Agency, a 4 hour notification is required. The (NRC) Senior Resident Inspector has been notified. No press releases are planned. The licensee notified state and local agencies.
ENS 5172911 February 2016 19:38:00PerryNRC Region 3GE-6At 1504 EST on February 11, 2016, with the plant shutdown in a forced outage, the Division 1, 4.16 Kv Safety Bus (EH11) lost power. Division 1 Shutdown Cooling was in service at the time and the Division 1 Shutdown Cooling pump A tripped. The Division 1 Emergency Diesel Generator (EDG) started and loaded EH11 as designed. However, the Emergency Service Water (ESW) A pump, which supplies cooling water to the EDG did not start. Due to the absence of cooling water to the EDG, operators took manual action to secure the Division 1 EDG. Division 2 Shutdown Cooling was operable during this transient and was subsequently started. The Division 1 Shutdown Cooling common suction isolation valve (1E12F0008) had previously been de-energized in the open position to support planned maintenance. The Division 2 Shutdown Cooling isolation valve was not affected by the loss of bus EH11. Shutdown Cooling was re-established at 1544 EST using the Division 2 Shutdown Cooling pump. Reactor coolant temperature rose from approximately 89 degrees Fahrenheit to 115 degrees Fahrenheit during the event. The cause of the loss of EH11 and subsequent failure of ESW A pump to start are currently under investigation. This event is being reported under 10 CFR 50.72(b)(3)(iv)(A) as a specific system actuation due to the auto start of the Division 1 EDG on a valid signal. The plant remains shutdown with Division 2 Shutdown Cooling in operation. The plant is in a normal electrical line up with the exception of bus EH11 being de-energized. The licensee notified the NRC Resident Inspector.
ENS 517168 February 2016 17:50:00PerryNRC Region 3GE-6At 1500 EST on February 8, 2016, two safety relief valves (SRV) opened upon a spurious Division 2 initiation signal. This caused suppression pool temperature to increase. At 1503 EST, plant operators took action to manually SCRAM the reactor at 95 degrees Fahrenheit in the suppression pool per plant procedures. The SRVs closed immediately following the scram at 1503 EST. The cause of the SRVs opening is currently under investigation. During the scram, all rods fully inserted into the core. Reactor Pressure is stable with decay heat being removed via turbine bypass valves to the main condenser. Reactor level control is currently being maintained via feedwater. Main Steam Isolation Valves are open. Cool down and depressurization to Mode 4 to follow. The plant is in a normal post SCRAM electrical line-up. The licensee notified the NRC Resident Inspector.
ENS 517125 February 2016 03:49:00Beaver ValleyNRC Region 1Westinghouse PWR 3-Loop

On February 5, 2015 at 0109 EST, the Control Room Emergency Ventilation System (CREVS) was declared inoperable due to a higher than allowed identified in-leakage rate for the Control Room Envelope (CRE) when in the Normal Operating Mode. Unit 1 remains at 100 percent power and Unit 2 remains in Mode 3 for an unrelated planned maintenance outage. Unit 1 and Unit 2 share a common CRE. This in-leakage was detected during additional testing following the event documented in EN #51584. At the time of discovery, there is a reasonable expectation this condition could prevent the fulfillment of the safety function of a system that is required to mitigate the consequences of an accident, thus satisfying the reporting criteria for 10CFR50.72(b)(3)(v)(D). Actions to implement mitigating actions were immediately initiated in accordance with Technical Specification 3.7.10. CREVS has been placed in Recirculation Ventilation Mode, isolating the control room from outside air. The NRC Senior Resident Inspector has been notified of the condition.

  • * * RETRACTION FROM DAVID HELD TO VINCENT KLCO AT 1411 EDT ON 4/4/16 * * *

Following the 8-hour 10 CFR 50.72 notification made on 02/05/2016 (EN 51712) regarding the Control Room Emergency Ventilation System (CREVS) inoperability, further engineering evaluation has determined the identified in-leakage does not result in exceeding the design criteria for dose to the control room personnel. Therefore the degraded Control Room Emergency Ventilation System remained operable with the identified air in-leakage as determined by the Control Room Envelope Habitability Program. As such, the safety function was never lost and the event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(D). The NRC Resident has been notified. Notified the R1DO (Jackson).

ENS 5168124 January 2016 11:23:00PerryNRC Region 3GE-6At 0357 hours (EST) on January 24, 2016, during a shut down required by plant Technical Specifications (see EN#51679), the 3A Feedwater heater isolated while performing a Reactor Recirculation Pump downshift. A consequence of this Feedwater heater isolation was that all 8 of the Average Power Range Monitors (APRM) became inoperable due to a calibration set point being out of tolerance. The APRM's are relied upon for the reactors high neutron flux trips. The inoperable APRM's resulted in a loss of RPS Trip Capability and a loss of safety function. The manual reactor trip function and other RPS functions remained available. RPS Trip capability for the APRM's was restored at 0445 hours on January 24, 2016. This notification is being made under (10 CFR) 50.72(b)(3)(v)(A) for an event or condition that could have prevented the fulfillment of a safety function affecting the ability to shut down the reactor. The licensee has notified the NRC Resident Inspector.
ENS 5167924 January 2016 00:56:00PerryNRC Region 3GE-6

At 2100 hours (EST), on January 23, 2016, the Perry Nuclear Power Plant commenced a reactor shutdown due to unidentified leakage in the drywell. At 2122 hours, drywell unidentified leakage exceeded the Technical Specification 3.4.5.d limit of 'less than or equal to 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in Mode 1.' The unidentified leakage increased to approximately 3.8 gpm at 2122 hours. Current unidentified leakage is 3.02 gpm. Technical Specification 3.4.5 actions allow 4 hours to reduce the leakage within limits or be in Mode 3 within 12 hours and Mode 4 within 36 hours. The plant is required to be in Mode 3 by 1322 hours on January 24, 2016 and Mode 4 by 1322 hours on January 25, 2016. A drywell entry will be made in Mode 3 to identify the leak source.

This notification is being made due to an expected inability to restore the leakage within limits prior to exceeding the LCO action time. Follow up question from NRC: Event times do not match (2100 versus 2122) - explained downpower was commenced at 2100 with leakage less than TS limit. When Reactor Core flow was reduced, un-identified leakage increased above the TS limit. The Licensee has notified the NRC Resident Inspector.

  • * * UPDATE FROM MIKE DOTY TO DANIEL MILLS AT 1123 EST ON 1/24/16 * * *

At 1007 hours, on January 24, 2016 with the plant at 8% power during a feedwater shift to place the motor feed pump in service, reactor level rose to the level 8 scram set point and the Reactor Protection System (RPS) initiated, scramming the reactor. During the scram, all rods fully inserted into the core. Decay heat is being removed via turbine bypass valves to the main condenser. Reactor level control is currently being maintained via feedwater. The plant is stable with cool down and depressurization to Mode 4 to follow. The cause of the rise in feedwater level is under investigation. This notification is being made under 50.72(b)(2)(iv)(B) for a RPS initiation while critical. All safety shutdown systems are available. The electric plant is in its normal shutdown alignment being supplied by offsite power. The licensee has notified the NRC Resident Inspector. Notified R3DO (Cameron). NRR (Morris) and IRD (Gott) were notified via email.

  • * * UPDATE FROM DAVID O'DONNELL TO HOWIE CROUCH AT 1915 EST ON 1/24/16 * * *

Following a shutdown required by plant Technical Specifications a small leak was identified coming from the Reactor Recirculation Loop A Pump Discharge Valve vent line. The Recirculation Loop is part of the reactor coolant system making this reportable under 50.72(b)(3)(ii)(A) as a degraded condition. It was subsequently determined to require a plant cool down in accordance with Technical Specification 3.4.5, Action C which requires the plant to be in MODE 4 within 36 hours. Technical Specification 3.4.5 was previously entered for increased unidentified leakage in the drywell. The plant is required to be in Mode 4 by 1322 hours on January 25, 2016. The licensee has notified the NRC Resident Inspector. Notified R3DO (Cameron). NRR (Morris) and IRD (Gott) were notified via email.

ENS 516311 January 2016 17:19:00Beaver ValleyNRC Region 1Westinghouse PWR 3-Loop

Unusual Event declared at BVPS (Beaver Valley Power Station) Unit 1 at 1634 EST due to a fire alarm in Containment (which was) unable to be verified extinguished within 15 minutes. Current Containment parameters do not indicate an actual fire. The licensee notified the NRC Resident Inspector and State and local government agencies. Notified DHS SWO, FEMA Ops Center, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA.

  • * * UPDATE FROM SAMANTHA BALDWIN TO DONALD NORWOOD AT 2057 EST ON 1/1/2015 * * *

At 1655 the fire alarm was reset. The Unusual Event has been terminated at 2035 EST after a containment entry and visual inspection determined no fire or indications of fire exist. The licensee notified the NRC Resident Inspector and State and local government agencies. Notified R1DO (McKinley), NRR EO (Morris), and IRD MOC (Gott). Notified DHS SWO, FEMA Ops Center, and NICC Watch Officer. E-mailed FEMA NWC and Nuclear SSA.

ENS 515844 December 2015 01:33:00Beaver ValleyNRC Region 1Westinghouse PWR 3-Loop

On December 3, 2015 at 2107 (EST), the Control Room Emergency Ventilation System (CREVS) was declared inoperable due to a higher than allowed identified in-leakage rate for the Control Room Envelope (CRE) when in the Normal Operating Mode. Both Unit 1 & Unit 2 remain at 100 percent power and they share a common CRE. At the time of discovery, there is a reasonable expectation this condition could prevent the fulfillment of the safety function of a system that is required to mitigate the consequences of an accident, thus satisfying the reporting criteria for 10CFR50.72(b)(3)(v)(D). Actions to implement mitigating actions were immediately initiated in accordance with Technical Specification 3.7.10. CREVS has been placed in Recirculation Ventilation Mode, isolating the control room from outside air. The NRC Senior Resident Inspector has been notified of the condition.

  • * * RETRACTION ON 1/7/16 AT 1110 EST FROM SHAWN SNOOK TO DONG PARK * * *

Following the 8-hour 10CFR50.72 notification made on 12/4/2015 (EN 51584) regarding the Control Room Emergency Ventilation System (CREVS) inoperability, an engineering evaluation determined inleakage did not exceed limits described in the Beaver Valley licensing basis. Therefore, the degraded Control Room Emergency Ventilation System remained Operable with the identified air inleakage as determined by the Control Room Envelope Habitability Program. As such, the safety function was never lost and the event notification is being retracted as it is not reportable pursuant to 10CFR50.72(b)(3)(v)(D). FENOC is planning to repair the degraded components of the system. The NRC Resident (Inspector) has been notified." Notified R1DO (Dimitriadis).

ENS 5147917 October 2015 19:30:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 1230 EDT, it was discovered that sodium hypochlorite had leaked into the ground soil at the bottom of a leak sensing pit, which is located near the BVPS-1 Cooling Tower Pump House, at the Beaver Valley Power Station Unit Number 1 (BVPS-1). The source of the sodium hypochlorite, is suspected to be from a piping leak, and has been terminated. The cause of the leak has not yet been conclusively determined, and there is no evidence that the sodium hypochlorite reached the surface water of the Ohio River. At 1613 EDT on October 17, 2015, it was determined that there was CERCLA (Comprehensive Environmental Response Compensation & Liability Act) reportable leak of a minimum 294 gallons of sodium hypochlorite to ground soil, and the following offsite notifications were made starting at 1625 EDT: National Response Center (Incident Report #1131084); Pennsylvania (PA) Department of Environmental Protection; Beaver County Emergency Management; PA Emergency Management Agency. This condition is being reported pursuant to 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified.
ENS 5147816 October 2015 15:41:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 1115 EDT there was a non-oil chemical spill at the main intake structure during delivery of corrosion inhibitor. The spill occurred due to a failed tank-fill connection. Approximately 10 gallons of the inhibitor reached the Ohio River. The spill has been stopped and contained, and recovery is in progress. This is not a CERCLA (Comprehensive Environmental Response Compensation and Liability Act), Reportable Quantity for the Environmental Protection Agency. However, at 1245 EDT, it was determined to be reportable to the Pennsylvania Department of Environmental Protection (PADEP), in accordance with the Pennsylvania Clean Streams Law and NPDS (National Pollutant Discharge Elimination), permit; thus, a notification was made, to the PADEP, at 1250 hours EDT . This condition is being reported pursuant to 10 CFR 50.72 (b)(2)(xi). The NRC Resident Inspector has been notified.
ENS 514535 October 2015 15:01:00Beaver ValleyNRC Region 1Westinghouse PWR 3-Loop

On 10/5/2015, during the Beaver Valley Power Station Unit No. 2 (BVPS-2) refueling outage, while performing planned ultrasonic examinations (UT) on the 66 reactor vessel head penetrations, it was determined, that two penetrations could not be dispositioned as acceptable per ASME (American Society of Mechanical Engineers) Code Section XI in a Reactor Coolant System pressure boundary. The indications of a degraded condition, on these two penetrations, are not through wall, as no leak path was identified. The examinations are being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D), and ASME Code Case N-729-1, to find potential flaws/indications well before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. All 66 reactor vessel head penetrations are scheduled to be examined during the current refueling outage. The plant is currently shutdown and in Mode 6. The reactor vessel head is not currently installed. Repairs are currently being planned and will be completed prior to startup. This is reportable, pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as found indications did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair. The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM DAN SCHWER TO STEVEN VITTO ON 10/22/2015 AT 1039 EDT * * *

The Beaver Valley Power Station Unit 2, FirstEnergy Nuclear Operating Company (FENOC) is retracting the 8-hour non-emergency notification made on 10/05/2015 (EN 51453). This retraction is based on additional examinations and subsequent engineering assessments completed that concluded the penetrations meet the requirements of ASME Code Case N-729-1 as amended by 10CFR50.55a(g)(6)(ii)(D) and no repairs are required. The two reactor head penetration indications, are therefore not reportable, pursuant to 10CFR50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. R1DO(Gray) has been notified.

ENS 5134526 August 2015 09:45:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 05:48 (EDT) BVPS (Beaver Valley Power Station) received notification that siren #6, Potter Township Municipal Building, was sounding intermittently. The fire department activation cable to the siren was severed by a motor vehicle. The ENS activation function remains functional. This event is reportable as a 4-hour Non-Emergency Notification 10 CFR 50.72(b)(2)(xi) as 'a News Release or Notification of Other Government Agency.' The Resident Inspector has been notified.
ENS 5132317 August 2015 12:59:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 0837 EDT on August 17, 2015, it was determined that Beaver Valley Unit 2 had experienced a small oil leak of approximately 1 liter from equipment located inside the Alternate Intake structure, some of which progressed to the Ohio River. Subsequent to confirmation of this discovery, notification was made to the following offsite agencies starting at 1030 EDT: National Response Center (Incident Report# 1125865), Pennsylvania (PA) Department of Environmental Protection, Beaver County Emergency Management, PA Emergency Management Agency, and downstream water authorities (Midland Water Authority, Allegheny Ludlum, East Liverpool Water Co.) This notification is a required 4-hour report per 10 CFR 50.72 (b)(2)(xi). The source of the oil leak has been stopped. Absorbent material has been placed to contain the oil that had leaked. No press release is planned. The NRC Resident Inspector has been notified.
ENS 511982 July 2015 10:54:00PerryNRC Region 3GE-6This event is being reported in accordance with 10CFR50.72(b)(2)(xi). On July 2, 2015, at approximately 0830 EDT, an inadvertent actuation of the Perry Nuclear Power Plant's (PNPP) alert notification system occurred. Seventy-six of seventy-six sirens sounded for three minutes affecting the emergency planning zone in Ashtabula, Geauga, and Lake Counties. Following the actuation, county agencies received calls from members of the public. PNPP's capability to notify the public in an emergency was not affected. The siren actuation was not related to any condition or event at the PNPP. An investigation is in progress to determine the cause of the inadvertent actuation. Preliminarily, it appears that the wrong test was initiated from a county agency; an audible test was initiated instead of a silent test. At the time of the event, the plant was in Mode 1 at 100 percent rated thermal power. The public was informed of the inadvertent actuation by way of the Emergency Alert System (EAS). Additionally, social media was used to respond to social media inquiries. The NRC Resident Inspector has been notified. The licensee notified the Ohio Emergency Management Branch Chief, and the County Emergency Managers of Ashtabula, Geauga, and Lake Counties.
ENS 5115916 June 2015 11:34:00PerryNRC Region 3GE-6At 0452 EDT hours on June 16, 2015, during performance of a surveillance test for the Division 3 4160 Volt Bus Undervoltage/Degraded Voltage Channel Calibration and Logic System Functional Test, the K36 degraded voltage time delay relay was found outside of the Technical Specification 3.3.8.1 allowable value, resulting in an inoperable condition of the Division 3 Emergency Diesel Generator (EDG). The Division 3 EDG had previously been declared inoperable for performance of the surveillance testing. The K36 degraded voltage time delay relay initiates load shedding, isolates the Division 3 bus, and starts the Division 3 EDG. The Technical Specification allowable value is 180 to 270 seconds. The as-found time was 272.66 seconds. The K36 relay was calibrated in accordance with plant procedures and returned to service at 0730 hours on June 16, 2015. The Division 3 EDG is the on-site power source for the High Pressure Core Spray system which is a single train system. Therefore, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been informed.
ENS 5108422 May 2015 09:24:00PerryNRC Region 3GE-6

This 60-day telephone notification is being made per the 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 a containment isolation signal affecting more than one system. At 1629 EDT on March 26, 2015, the plant received a Division 1 balance of plant outboard containment and drywell isolation signal. The isolation signal was received while removing fuses to establish a clearance for the replacement of an average power range monitor bypass switch. The removal of two fuses removed power to the manual initiation logic resulting in an isolation signal. The following component actuations occurred: valves 1P51F0150 and 1P51F0652, isolating service air to the containment and drywell; valves 1G61F0155 and 1G61F0170, isolating the containment and drywell floor drain sumps; valves 1D17F0071A and 1D17F0079A, isolating the drywell radiation monitor; valves 1D17F0081A and 1D17F0089A, isolating the containment radiation monitor; valve 1P11F0080, isolating the containment pools drain; valves 1P50F0060 and 1P50F0150, isolating the containment vessel chilled water system; valves 1P53F0070 and 1P53F0075, isolating the upper and lower airlock local leak rate air supply; valves 1P52F0160 and 1P52F0170, isolating the upper and lower airlock air supply; valve 1P22F0015, isolating mixed bed water to the drywell; and valve 1P54F0395, isolating fire protection carbon dioxide to the drywell. This event is considered an invalid system actuation reportable under 10 CFR 50.73(a)(2)(iv)(A).

The isolation was not initiated in response to actual plant conditions or parameters, and was not a manual initiation. Therefore, this notification is provided via a 60-day optional phone call in accordance with 10 CFR 50.73(a)(1) instead of submitting a written Licensee Event Report. The event meets reporting criteria specified in 10 CFR 50.73(a)(2)(iv)(B)(2) as a general containment isolation valve signal affecting containment isolation valves in more than one system. All affected systems functioned as expected in response to the loss of power to the manual initiation logic. The valves were reopened in accordance with plant procedures. The inadvertent isolation signal was the result of a human performance error. The NRC Resident Inspector has been notified.

ENS 5106411 May 2015 20:38:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 1629 EDT on May 11, 2015, Beaver Valley Power Station (BVPS) Unit 1 (Mode 6 Refueling) opened circuit breaker PCB-92 to de-energize system service transformer 1A due to an emergent issue with a transformer bushing. This resulted in a loss of power to the ERF (Emergency Response Facility) substation which powers various plant support equipment at both units. The BVPS Unit 2 Digital Radiation Monitoring System (DRMS) was non-functional for approximately 21 minutes. This was a loss of radiation monitor capability to alarm and indicate in the control room. Following power restoration to the ERF substation, the radiation monitor system was restored at 1650 EDT. Both Unit 1 and Unit 2 remain stable. Since the BVPS Unit 2 DRMS was non-functional, this event resulted in a loss of emergency assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified.
ENS 5098515 April 2015 07:32:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 0411 EDT on April 15, 2015, Beaver Valley Power Station (BVPS) Unit 1 manually tripped the reactor from approximately 85% power due to the trip of a condensate pump. The unit was performing an emergent power reduction due to a degraded condensate pump prior to the manual reactor trip. An end of cycle Tave coastdown was in progress at the time of the event. All control rods fully inserted into the core. All three auxiliary feed water pumps started as expected and were subsequently secured in accordance (with) station procedures. The main feedwater system remains available and in service. The unit is currently stable in Mode 3. Unit 2 was unaffected and remains at full power. Decay heat removal is via main feedwater system with steam discharge to the main condenser via the steam bypass valves. Unit 1 is in a normal shutdown electrical lineup. No primary or secondary reliefs or safeties lifted during the transient. The licensee informed the NRC Resident Inspector.
ENS 5091823 March 2015 16:28:00PerryNRC Region 3GE-6

At approximately 1526 EDT, the control room received a report of an individual experiencing chest pains. An ambulance was called to transport the individual to an offsite medical facility. The initial Radiation Protection survey did not detect any contamination, however the protective clothing the individual wore could not be removed. The individual is considered 'potentially contaminated' due to not being able to perform a complete frisk. Radiation Protection personnel escorted the individual offsite. The individual will be frisked at the medical facility." The licensee has notified the NRC Resident Inspector and will notify the state and local government.

  • * * UPDATE AT 1722 EDT ON 3/23/2015 FROM ED CONDO TO MARK ABRAMOVITZ * * *

At approximately 1652 (EDT), Perry Radiation Protection confirmed the individual was not contaminated. Additionally no contamination was found in the ambulance or at the hospital. Perry Radiation Protection is in possession of and returning all protective clothing worn by the individual to the plant. The licensee notified the NRC Resident Inspector. Notified the R3DO (Cameron).

ENS 5089514 March 2015 23:32:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopAt 1708 EDT on March 14, 2015, Beaver Valley Power Station (BVPS) Unit 2 determined that the Digital Radiation Monitoring System (DRMS) was not operating. This resulted in a loss of radiation monitor capability to alarm and indicate in the control room. BVPS Unit 2 DRMS was declared non-functional. Repair efforts were initiated and compensatory measures were initiated. The radiation monitoring computer system is a stand-alone system. No cyber interference is suspected. At 2056 EDT on March 14, 2015, following initial investigation and repairs, BVPS Unit 2 emergency assessment capability was restored in conjunction with DRMS restoration. Since the BVPS Unit 2 DRMS was non-functional, this event resulted in a loss of emergency assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified.
ENS 5085227 February 2015 16:10:00PerryNRC Region 3GE-6At 1518 EST on Feb 27, 2015, the Perry Shift Manager received notice from the Radiation Protection group that an Exclusive Use closed transport vehicle arrived on site exceeding the 10 CFR 71.47 radiation levels on contact with a box on the vehicle. The truck that arrived had two boxes containing four rebuilt control rod drive mechanisms to be used during the Perry refueling outage. One of the boxes had a contact dose reading of 1290 MR/HR. This is above the 1000 MR/HR limit as noted in 10 CFR 71.47. No other limits were exceeded on the exterior of the vehicle. Specifically, the cab of the truck was reading 0.1 MR/HR which is less than the 2 MR/HR limit. Also at 2 meters around the truck, the highest level reading was 1.2 MR/HR which is below the 10 MR/HR (limit). Also on direct contact with the outside of the vehicle, the highest reading was 30 MR/HR, which is below the 200 MR/HR limit. The Site Radiation Protection Shipping Coordinator contacted the shipping organization of this finding at Perry. This was the Director of Operations of Energy Solutions in Memphis, Tennessee. The box was taken into the Perry Fuel Handling Building and is posted per the Perry Radiation Control Program. The vehicle is parked outside the Fuel Handling Building and is being controlled. The NRC Resident Inspector has been informed.
ENS 507186 January 2015 17:18:00PerryNRC Region 3GE-6In accordance with 29 CFR 1904.39(a)(2), notification was made to the Occupational Safety and Health Administration regarding the in-patient hospitalization of an individual while in the owner controlled area. The licensee has notified the NRC Resident Inspector. The licensee notified the State of Ohio and local authorities The individual employee is currently under medical treatment and is not contaminated.
ENS 5062719 November 2014 09:57:00Beaver ValleyNRC Region 1Westinghouse PWR 3-Loop

The Beaver Valley Power Station (BVPS) Unit 1 Special Particulate, Iodine and Noble Gas (SPING 4) monitors were removed from service for a planned equipment upgrade/replacement. During the replacement process certain Emergency Action Level (EAL) required monitors will not be functional, therefore, alternate methods of monitoring have been placed in effect. The replacement activities are expected to take approximately 60 days. A follow-up notification will be made after the required monitors are returned to service and declared functional. This is an 8 hour notification per 10 CFR 50.72(b)(3)(xiii). No BVPS Unit 2 systems will be adversely affected by the Unit 1 replacement activity. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM PATRICK HARTIG TO CHARLES TEAL AT 1728 EST ON 2/27/15 * * *

The planned equipment upgrade was completed and the Beaver Valley Power Station (BVPS) Unit 1 Particulate, Iodine and Noble Gas (PING) monitors were returned to service and declared functional at 1714 EST on 2/27/2015. The NRC Resident Inspector has been notified. Notified R1DO (Gray).

ENS 506017 November 2014 11:13:00PerryNRC Region 3GE-6The Perry Nuclear Power Plant experienced an automatic reactor scram due to a loss of feedwater, which resulted in receiving valid reactor vessel water Level 3 and Level 2 initiation signals. The High Pressure Core Spray system and the Reactor Core Isolation Cooling system started and injected. Reactor water level and pressure have been stabilized in the required bands. The motor feed pump automatically started and is being used to control reactor vessel water level. The High Pressure Core Spray and Reactor Core Isolation Cooling systems have been returned to the standby mode. As a result of receiving a reactor vessel water Level 2 signal a Balance of Plant containment isolation signal was received. All systems isolated as required and the plant is restoring isolated systems in accordance with plant procedures. During the scram, all rods fully inserted into the core. Decay heat is being removed via turbine bypass valves to the main condenser. The electrical grid is stable and is supplying plant loads. An emergency diesel generator (Division 3 High Pressure Core Spray) started, as designed, as a result of the reactor vessel water Level 2 signal. No safety relief valves lifted as a result of the transient. The plant is stable with cooldown and depressurization to Mode 4 in progress. The cause of the loss of feedwater is under investigation. The NRC Resident Inspector has been notified. The State of Ohio and local officials will be notified.