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 Entered dateSiteRegionReactor typeEvent description
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 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 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 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 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 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 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 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 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 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 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 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 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 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.
ENS 5055120 October 2014 03:55:00PerryNRC Region 3GE-6

The Perry Power Plant experienced a reactor scram during a shift of non-essential vital power supply to the alternate source. Feedwater was lost resulting in receiving a valid level 3 and level 2 signal. High Pressure Core Spray and Reactor Core Isolation Cooling started and injected. Reactor level and pressure have been stabilized to required bands. The motor feed pump has been started and is controlling level. High Pressure Core Spray and Reactor Core Isolation Cooling have been returned to standby. During the scram, all rods fully inserted into the core. Decay heat is being removed via the steam dumps to the condenser. The electrical grid is stable and supplying plant loads. An emergency diesel generator started, as designed, as a result of the level 2 signal but did not load. No safety valves lifted as a result of the transient. The cause of the loss of feedwater is under investigation. The licensee will be notifying the State of Ohio and Perry Township and has notified the NRC Resident Inspector.

  • * * UPDATE FROM DOUG SHORTER TO HOWIE CROUCH AT 0933 EDT ON 10/20/14 * * *

The plant is currently in Mode 3, stable with cooldown and depressurization to Mode 4 in progress. Level control is being provided by the motor feedwater pump. Troubleshooting of the cause of the scram and loss of feed water is on-going. The initial notification identified 10CFR50.72(b)(3)(iv)(A), 'Specified System Actuation', as a reporting criteria. The specific system that actuated was not provided. As a result of receiving a reactor vessel water level 2 signal a containment/BOP isolation signal was received. All systems isolated as required and the plant is restoring isolated systems in accordance with procedure. The licensee will be notifying the State of Ohio and Perry Township and has notified the NRC Resident Inspector. Notified R3DO (Pelke).

ENS 5045012 September 2014 10:12:00PerryNRC Region 3GE-6

At 1313 (EDT) on 7/26/14, the plant experienced an electrical transient on bus EK-1-B1 (safety-related 120 volt AC distribution panel) that resulted in partial Balance of Plant Division 2 isolation signals and alarms received in the Control Room. The following component actuations occurred: valve 1P50F140 closed, resulting in a trip of Containment Vessel Chilled Water C; valve 1G41F140 closed, isolating the Fuel Pool Cooling and Clean-up return from the containment building upper pools; valve 1B33F019 closed, isolating Reactor Water sampling; valve 1D17F071B closed, isolating the Drywell Atmosphere Radiation Monitor; valve 1D17F081B closed, isolating the Containment Atmosphere Radiation Monitor; valves 1G61-F030, 1G61-F150, 1G61-F075, and 1G61-F165 closed, isolating the Containment and Drywell Floor and Equipment drain sumps; valve 1G50-F272 closed isolating the Reactor Water Cleanup Backwash Receiving Tank: 1M25F020B, Control Room HVAC Inboard supply damper, closed and Division 2 indicated an auto initiation (M25-S12, Auto Initiate Active amber light was on). 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 electrical transient on bus EK-1-B1 that resulted in the partial Balance of Plant Division 2 isolation signals. The valves were reopened in accordance with plant procedures. The failure mechanism that caused the electrical transient was a failed capacitor in regulating transformer EFB1B2. The capacitor was replaced and tested with satisfactory results. The NRC Resident Inspector has been notified.

ENS 5021319 June 2014 17:06:00PerryNRC Region 3GE-6A review of industry operating experience regarding the impact of unfused Direct Current (DC) circuits has determined the described condition to be applicable to the Perry Nuclear Power Plant (PNPP) resulting in an unanalyzed condition with respect to fire safe shutdown requirements. In the postulated event, a fire induced hot short could adversely impact safe shutdown equipment. The potential exists for a secondary fire to occur due to unfused DC control circuits associated with the Turbine Emergency Bearing Oil Pump, Reactor Feed Pump Turbine 'A' Emergency Lube Oil Pump, Turbine Emergency Seal Oil Pump, and Reactor Feed Pump Turbine 'B' Emergency Lube Oil Pump. These circuits are routed from the respective equipment to other plant areas including the Unit 1 Control Room, Division 1 Cable Spreading, and Division 1 Cable Chase. Without overcurrent protection for these circuits, the potential exists that an initial fire event affecting these circuits could cause short circuits 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 these circuits are routed challenging the ability to achieve and maintain safe shutdown. 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.
ENS 499872 April 2014 14:42:00PerryNRC Region 3GE-6

Release of toxic or flammable gas affecting the Protected Area boundary deemed detrimental to the safe operation of the plant. Emergency Action Level entered: MU-1. The leak is Trichloroethylene (TCE) gas used in the Off-Gas building. The Off-Gas building ground and basement levels were evacuated due to the leak. There is no safe-shutdown equipment located in the Off-Gas building. The licensee is working to isolate the leak. The licensee informed the NRC Resident Inspector. The licensee notified the State of Ohio and the local counties. Notified DHS SWO, FEMA Ops Center, NICC Watch Officer, DOE Ops Center, USDA Ops Center, HHS Ops Center, and Nuclear SSA via email.

  • * * UPDATE AT 1630 EDT ON 4/2/14 FROM DON ROGERS TO S. SANDIN * * *

The licensee notified the following outside agencies: U.S. EPA National Response Center, Ohio EPA, Perry Township Fire Department, Lake County Emergency Planning Committee, and the U.S. Coast Guard. Notified R3DO (Passehl).

  • * * UPDATE FROM MICHAEL ADLER TO DANIEL MILLS AT 0115 EDT ON 04/05/2014 * * *

Unusual Event has been terminated on 4/5/2014 at 0059 EDT. The trichloroethylene leak has been stopped. Access has been restored to all normally accessible areas. Unit 1 remains in Mode 1 at 100% power. The licensee notified the NRC Resident Inspector and the Local and State emergency agencies. Notified the IRD MOC (Gott), R3DO (Passehl), and NRR EO (McGinty). Notified DHS SWO, DOE Ops Center, FEMA Ops Center, HHS Ops Center, NICC Watch Officer, USDA OPS Center, EPA EOC, FDA EOC, and Nuclear SSA via email.

ENS 498047 February 2014 08:19:00PerryNRC Region 3GE-6

This event is being reported in accordance with 10CFR 50.72(b)(2)(i), 'Initiation of a Shutdown Required by Technical Specifications.' At 2043 hours (EST) on February 06, 2014, the Perry Nuclear Power Plant entered Technical Specification 3.6.1.3 Primary Containment Isolation Valves (PCIVs), action C.1, due to leakage identified during local leak rate testing of the containment penetration for the Containment and Drywell Purge system. Leakage was identified on the outboard containment isolation valve resulting in the plant exceeding the limit for secondary containment bypass leakage. The Containment and Drywell Purge system penetration is normally isolated and remains isolated in accordance with Technical Specifications. Action C.1 requires restoration of the leakage rate within four hours. At 0043 hours on February 7, 2014, the plant entered Technical Specification 3.6.1.3, 'Primary Containment Isolation Valves (PCIVs)', action E as the leakage rate was not restored. Action E requires the plant be in Mode 3 in 12 hours and Mode 4 in 36 hours. At 0600 hours on February 07, 2014, the Perry Nuclear Power Plant initiated a shutdown in accordance with Technical Specification 3.6.1.3, action E. Repairs to restore the penetration leakage to within allowable limits are in progress. The NRC Resident Inspector has been notified.

  • * * UPDATE PROVIDED BY DAVE ODONNELL TO JEFF ROTTON AT 1220 EST ON 02/07/2014 * * *

At 0943 hours (EST) the reactor shutdown to comply with Technical Specification 3.6.1.3 action E was terminated (with the reactor at 42% power). A blind flange was installed downstream of the outboard containment isolation valve. Local leak rate testing of the containment penetration for the Containment and Drywell Purge system verified that leakage was within the limits for secondary containment bypass leakage. The NRC Resident Inspector has been notified. The licensee has commenced increasing reactor power. Notified R3DO (Orlikowski)

ENS 4974621 January 2014 02:19:00PerryNRC Region 3GE-6This notification is being made pursuant to (10 CFR) 50.72(b)(2)(xi), notification of other government agency. Notification to other government agency, State of Ohio, was made at 0140 (EST) on 1/21/14. At 1310 on 1/20/2014, a leak was identified on a feed water Venturi. In response to the water leak, samples were taken to check for the spread of tritium. A positive result for tritium was identified in the under drain system in the Auxiliary Building which requires communications as part of the NEI ground water protection initiative. The positive sample results were obtained at 2330 on 1/20/14. Actions are in progress stop the leak (perform leak injection). The EPA limit for groundwater is 20,000 pCi/l. The samples taken by the licensee indicated 46,000 pCi/l. The licensee has notified the NRC Resident Inspector and will notify local counties.
ENS 4944617 October 2013 16:25:00PerryNRC Region 3GE-6A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined the described condition to be applicable to the Perry Nuclear Power Plant resulting in a potentially unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1 E batteries control room ampere indications do not include overcurrent protection features to limit the fault current. In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane. Simultaneously, it is postulated that the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This could cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (i.e., heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. 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. See the following related Event Numbers: 49411, 49419, 49422, and 49444.
ENS 4912116 June 2013 02:42:00PerryNRC Region 3GE-6This event is being reported in accordance with 10 CFR 50.72(b)(2)(i) and 50.72(b)(3)(ii)(A). On June 16, 2013 at 0200 EDT, the Perry Nuclear Power Plant commenced a controlled plant shutdown. The shutdown was due to a small leak through the base of a vent line on the 'B' Reactor Recirculation Flow Control Valve. On June 15, 2013 at 2250 EDT, the leak was identified and was subsequently determined to require a plant shutdown in accordance with Technical Specification 3.4.5, Action (C) which requires the plant to be in Mode 3 within 12 hours. The NRC Resident Inspector has been notified." The licensee will also be notifying state and local authorities. The licensee had come down in power to make a drywell entry and investigate drywell leakage indications. Steam was observed to be coming from a vent line that comes off the top of the recirc flow control valve. The licensee was unable to characterize the leak rate other than a small leak. The licensee stated that the steam appeared be coming from a weld location where the vent line comes out of the flow control valve which would classify it as pressure boundary leakage.
ENS 4893717 April 2013 05:20:00PerryNRC Region 3GE-6

The Perry Nuclear Power Plant is reporting an event or condition pursuant to 10 CFR 50.72(b)(3)(v)(D). On April 16, 2013, at 2323 EDT, it was identified that Emergency Service Water (ESW) pump 'A' was inoperable due to an inability to maintain minimum flow requirements. As a result, ESW 'A' and the supported Division 1 Emergency Diesel Generator (EDG) were declared inoperable. Coincident with this discovery, a test of the Division 2 emergency systems was in progress with the associated ESW 'B' pump and Division 2 EDG inoperable. Division 2 EDG was available to support the Shutdown Defense In-Depth Strategy. Division 3 EDG was operable and could supply High Pressure Core Spray system injection, if needed. Both EDGs were inoperable simultaneously and Technical Specification 3.8.2 'AC Sources-Shutdown' was entered and required actions taken. These actions included immediately suspending core alterations and immediately initiating actions to restore the required EDG. The test of Division 2 emergency systems was suspended and ESW 'B' and the Division 2 EDG were restored to operable status at 0135 EDT on April 17, 2013. The failure of ESW 'A' minimum flow is currently under investigation. The Resident Inspector has been notified.

  • * * RETRACTION FROM JOHN PELCIC TO CHARLES TEAL ON 4/20/13 AT 1355 EDT * * *

Engineering personnel performed an immediate investigation of the ESW 'A' minimum flow condition. The investigation results showed that the ESW 'A' pump flow exceeded the minimum flow requirement to protect the ESW 'A' system. Therefore, continued operation of ESW 'A' was acceptable and the minimum flow condition originally reported did not cause the Division 1 Emergency Diesel Generator to be inoperable. The condition would not have prevented the fulfillment of a safety function to mitigate the consequences of an accident. Reporting is not required under 10 CFR 50.72(b)(3)(v)(D) and this notification is retracted. The NRC Resident Inspector has been notified. Notified R3DO (Orth).

ENS 488947 April 2013 17:47:00PerryNRC Region 3GE-6During an extent of condition review of past radiological events, it was identified that an event on November 17, 2010 met the E-Plan entry criteria for GU1, 'Unexpected Increase In Plant Radiation Levels'. Due to an equipment deficiency, dose rates in one section of the Radwaste building rose from 0.08 mrem/hr to 80 mrem/hr. This satisfied the E-Plan criteria of a 1000 times change over normal radiation levels. This was initially identified in (Perry) Condition Report 2010-85937. The licensee notified the NRC Resident Inspector and will notify State and local authorities.
ENS 4876919 February 2013 16:39:00PerryNRC Region 3GE-6

On February 19, 2013, at approximately 1303 EST, the control room was notified that a supplemental worker (i.e., a contract individual) had fallen and was injured. The worker was in a contaminated area. Due to the individual's condition, the individual was not surveyed by a Health Physics technician prior to being transported in their anti-contamination clothing. The individual was transported by ambulance accompanied by Health Physics personnel to the local hospital for medical treatment (i.e., TriPoint Medical Center). Subsequently, the worker was declared deceased at the hospital. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xii) and 50.72(b)(2)(xi). Additionally, OSHA was notified pursuant to the requirements of 29 CFR 1904.39. The Lake County Coroner was also notified. Subsequent surveys found no contamination on the worker, hospital, medical personnel, or ambulance. No press release is planned. The NRC Resident Inspector has been notified.

* * * UPDATE ON 3/21/13 AT 2032 EDT FROM LLOYD ZERR TO PETE SNYDER * * *

The Lake County Coroner has determined that the individual died of natural causes. The NRC Resident Inspector has been notified. Notified R3DO (Passehl).

ENS 4870831 January 2013 09:28:00PerryNRC Region 3GE-6On January 31, 2013, at approximately 0210 hours (EST), the ability to transfer plant parameter data via the Emergency Response Data System (ERDS) was lost. ERDS capability was restored at 0701 hours (EST). The cause is under investigation. In the event of an emergency while ERDS was unavailable, contingency plans were in place to transmit plant parameter data, This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii), The NRC Resident Inspector has been notified.
ENS 4868822 January 2013 06:57:00PerryNRC Region 3GE-6On January 22, 2013, at approximately 0332 hours (EDT), an automatic Reactor Protection System (RPS) actuation occurred at the Perry Nuclear Power Plant, Unit 1. At the time of the event, the plant was in Mode 1 at 100% power. All control rods are inserted into the reactor core and the plant is currently stable in Mode 3 (Hot Shutdown) with reactor pressure and level being maintained in the normal shutdown range. The RPS actuation was initiated by a low reactor water level (Level 3 - 178") signal. In response to the RPS actuation and subsequent reactor Level 2 (130") signal, the High Pressure Core Spray (HPCS) system and Reactor Core Isolation Cooling (RCIC) system both actuated and injected to maintain reactor coolant level. The reactor level is currently being maintained in its normal band by the feedwater system and decay heat is being removed by (turbine bypass valves to) the condenser (both HPCS and RCIC have been returned to standby). The plant is in a normal electrical line-up with all three Emergency Diesel Generators operable and available, if needed. The Containment Isolation Valves (responded to the Level 2 and 3) isolation signals as designed. The cause of the RPS actuation is under investigation. The NRC Resident Inspector has been notified.
ENS 4861020 December 2012 07:51:00PerryNRC Region 3GE-6

Computer engineering personnel will be taking the plant integrated computer system (ICS) out-of-service for planned maintenance. During the time ICS is out-of-service, the Safety Parameter Display System (SPDS) and the Emergency Response Data System (ERDS) will be unavailable. The computer outage is scheduled for six hours. Contingency plans have been established to transmit plant parameter data and perform the dose assessment function in the event of an emergency while ERDS is unavailable. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). A follow-up notification will be made when the maintenance activities are complete and the equipment is restored. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM JIM CASE TO HOWIE CROUCH AT 1309 EST ON 12/20/12 * * *

At 1300 EST, the plant integrated computer system was restored and SPDS and ERDS was returned to service. Notified R3DO (Cameron) and ERDS Group email.

ENS 4854428 November 2012 19:47:00PerryNRC Region 3GE-6At 1930 (EST) on November 28, 2012, it was determined that a Notification of Unusual Event (NOUE) was not declared for an event that occurred on June 3, 2012, when an equipment failure resulted in a deposit of ion exchange resin onto the floor of the Radioactive Waste building. Subsequent radiological surveys indicated that conditions met the requirements for a NOUE in accordance with the Perry Nuclear Power Plant Emergency Plan. This report is being provided within one hour of the recognition of the undeclared event. As discussed in NUREG 1022, Revision 2, an actual declaration of an Unusual Event is not necessary. The Initiating Conditions for the emergency classification no longer existed at the time of recognition. The NRC Resident Inspector has been notified.
ENS 4854228 November 2012 10:50:00PerryNRC Region 3GE-6

Entered an Unusual Event (under Emergency Action Level) MU1, toxic gas, carbon monoxide (CO), detected in the Radwaste Control Room. Levels rose to 34 ppm and the Radwaste Control Room was evacuated prior to reaching the First Energy exposure limit of 35 ppm. The source of the CO has not been determined. There is no radiation release from this event. There were no personnel injuries and offsite assistance was not requested. There was no effect on plant operations. The licensee has notified state and local authorities and the NRC Resident Inspector. Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email.

  • * * UPDATE AT 1504 EST ON 11/28/12 FROM MARIKIO BLOUNT TO HUFFMAN * * *
(At 1452 EST, the licensee) terminated the Unusual Event for MU1, due to toxic gas - carbon monoxide, detected in the Radwaste Control Room.  The source of the carbon monoxide readings was determined to be from a leaking acetylene bottle.  The acetylene bottle has been removed from the building.  Carbon monoxide readings have returned to normal.

The licensee noted that the acetylene is detected as carbon monoxide by the toxic gas monitoring devices. The licensee has notified State and local authorities and the NRC Resident Inspector. Notified R3DO (Stone), NRR (EO) Lubinski) , IRD (Marshall), DHS SWO, FEMA, DHS NICC and NuclearSSA via email.

ENS 4844928 October 2012 16:27:00PerryNRC Region 3GE-6At approximately 1008 EDT on October 28th, 2012, a failure between the Plant Computer and the MMI (Man Machine Interface) occurred. The cause is due to a failure of the data diode. The Plant Computer is still working however the MMI is not, therefore Safety Parameter Display System (SPDS) outside of the Control Room and the Emergency Response Data System (ERDS) is unavailable. In the event of an emergency, plant parameter data will be communicated to the facilities through the status board ring down circuit with back-up by the Private Branch Exchange (PBX), Off Premise Exchange (OPX), and various redundant intra-facility circuits throughout the emergency facilities. The dose assessment function is maintained during this out of service time period by manual input of data into the Meteorological Information and Dose Assessment System (MIDAS). The ability to open and maintain an 'open line' using the Emergency Notification System is not affected and will be the primary means for transferring plant data to the NRC as a contingency until the ERDS can be returned to service. At 1548 EDT on October 28th, 2012, a re-start of the data diode was successful in restoring the connection between the Plant Computer and the MMI. SPDS and the ERDS are functioning as designed. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified. The licensee has notified the State and local agencies.
ENS 4841517 October 2012 07:40:00PerryNRC Region 3GE-6

At approximately 0800 hours EDT on October 17, 2012, computer engineering personnel will be taking the plant integrated computer system (ICS) out-of-service for planned maintenance. During the time ICS is out-of-service, the Safety Parameter Display System (SPDS) and the Emergency Response Data System (ERDS) will be unavailable. The computer outage is scheduled for twelve hours. In the event of an emergency, plant parameter data will be communicated to the facilities through the Status Board Ring Down circuit with back-up by the Private Branch Exchange, Off Premise Exchange, and various redundant intra-facility circuits throughout the emergency facilities. The dose assessment function will be maintained during the out-of-service time period by manual input of data into the Meteorological Information and Dose Assessment System (MIDAS). The ability to open and maintain an 'open line' using the Emergency Notification System will not be affected and will be the primary means of transferring plant data to the NRC as a contingency until the ERDS can be returned to service. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). A follow-up notification will be made when the maintenance activities are complete and the equipment is restored. The (NRC) Resident Inspector has been notified.

  • * * UPDATE FROM THOMAS MORSE TO VINCE KLCO ON 10/17/12 AT 2146 EDT * * *

As of 2140 EDT on 10/17/12, the ERDS system was tested and restored to service. The licensee notified the NRC Resident Inspector. Notified the R3DO (Orth).

ENS 4831918 September 2012 12:58:00PerryNRC Region 3GE-6On July 23, 2012, at 2057 hours, the Perry Nuclear Power Plant experienced a loss of the normal power supply to the Reactor Protection System (RPS) A electrical bus. The loss of RPS bus A caused an actuation of several Division 1 containment outboard isolation valves. The actuation signal caused full closure of one or more valves in each of the following Division 1 subsystems: Main Steam line drains, Containment Radiation Monitor, Drywell Radiation Monitor, Reactor Water Cleanup, Fuel Pool Cooling and Cleanup, Liquid Radwaste Sumps, Containment Vessel Chilled Water, Containment Vacuum Relief, Condensate Transfer and Storage, Mixed Bed Demineralizer and Distribution, Containment Personnel Airlocks, Service Air, and Instrument Air. Division 2 components and valves were not affected. 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 an outboard isolation signal. The valves were reopened in accordance with plant procedures. The failure mechanism that caused the loss of RPS bus A was a degraded voltage regulator. The voltage regulator was replaced and retested with satisfactory results. The NRC Resident Inspector has been notified.
ENS 480693 July 2012 20:35:00PerryNRC Region 3GE-6This event is being reported in accordance with 10CFR50.72(b)(2)(xi). On July 3, 2012, at approximately 1738 hours, an inadvertent actuation of the Perry Nuclear Power Plant's (PNPP) alert notification system occurred. Seventy-five of seventy-six sirens sounded for three minutes, affecting 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 has been retained. 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. 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). The NRC Resident Inspector has been notified. The licensee has notified State and Local agencies.
ENS 4801311 June 2012 14:41:00PerryNRC Region 3GE-6On June 11, 2012, at 0845 hours, an unexpected Division 3 battery DC system trouble alarm was received in the control room along with indication of lowering battery voltage. As a result of this condition, the plant operators declared the Division 3 DC electrical power subsystem inoperable at 0852 hours and entered the applicable Technical Specifications which require the High Pressure Core Spray System (HPCS) be declared inoperable. HPCS is a single-train safety system and its inoperable status is considered a loss of safety function. The cause of the trouble alarm was failure of the normal battery charger. Following a walkdown inspection of the Division 3 DC electrical bus with no abnormalities noted, the reserve charger was placed in service at 0858 hours to supply the bus. At 1245 hours, the HPCS system was declared operable following restoration of the Division 3 DC electrical power subsystem to operable status. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector has been notified.
ENS 4791310 May 2012 10:29:00PerryNRC Region 3GE-6

Beginning at approximately 1100 hours EDT on May 10, 2012, plant personnel will be taking the plant integrated computer system (ICS) out-of-service for planned maintenance. During the time ICS is out-of-service, the Safety Parameter Display System (SPDS), the Emergency Response Data System (ERDS), and the automatic mode calculation of the Computer Aided Dose Assessment Program (CADAP) will be unavailable. The computer outage is scheduled for two hours. In the event of an emergency, plant parameter data will be orally transmitted to the facilities through the Status Board Ring Down circuit with back-up by the Private Branch Exchange, Off Premise Exchange, and various redundant intrafacility circuits throughout the emergency facilities. The dose assessment function will be maintained during the out-of-service time period by manual input of data into CADAP and, if required, by manual calculation. The ability to open and maintain an 'open line' using the Emergency Notification System will not be affected and will be the primary means of transferring plant data to the NRC as a contingency until the ERDS can be returned to service. This event is being reported in accordance with 10 CFR 50.72(b)(3)(xiii). A follow-up notification will be made when the maintenance activities are completed and the equipment is restored. The NRC Resident Inspector has been notified.

  • * * UPDATE AT 1245 EDT ON 5/10/12 FROM MORSE TO HUFFMAN * * *

The maintenance activities were completed as scheduled and the integrated computer system and associated systems SPDS, ERDS and CADAP has been returned to service as of 1238 EDT. The licensee will notify the NRC Resident Inspector. R3DO (Giessner) notified.

ENS 479027 May 2012 16:16:00PerryNRC Region 3GE-6This event is being reported in accordance with 10 CFR 50.72(b)(2)(xi). On May 7, 2012, at approximately 1233 hours (EDT), an inadvertent actuation of the Perry Nuclear Power Plant's alert notification system occurred. Twenty of the seventy-six total sirens sounded for three minutes affecting Ashtabula, Geauga, and Lake Counties. Following the actuation the county agencies received calls from members of the public. A successful quiet test of the sirens had been conducted earlier in the day (at approximately 0830 hours (EDT)). At this time, all sirens are functioning correctly. The siren actuation was not related to any condition or event at the Perry Nuclear Power Plant. The actuation signal originated from the Lake County Emergency Operations Center while thunderstorms were passing through the area. Additional investigation is in progress to determine the cause of the inadvertent actuation. At the time of the event, the plant was in Mode 1 at 100 percent rated thermal power. Lake County officials plant to issue a press release. The NRC Resident Inspector has been notified.
ENS 479016 May 2012 20:53:00PerryNRC Region 3GE-6This event is being reported in accordance with 10 CFR 50.72(b)(2)(xi). On May 6, 2012, during daily chlorination activities, it was identified that the National Pollutant Discharge Elimination System (NPDES) permit limit for Total Residual Chlorine was exceeded between approximately 0935 hours (EDT) and 0947 hours (EDT) when the noncompliance was corrected. The maximum measured value was 0.29 mg/L, which exceeded the NPDES Maximum Concentration Limit of 0.2 mg/L. On May 6, 2012, at approximately 1930 hours (EDT), a 'Noncompliance Notification for Exceedance of a Daily Maximum Discharge Limit' was made to the Ohio Environmental Protection Agency. The cause of the NPDES permit noncompliance is under investigation. Chlorination evolutions have been suspended pending investigation results. At the time of the event, the plant was in Mode 1 at 100 percent rated thermal power. This event is also being reported in accordance with the plant's operating license, appendix B, Environmental Protection Plan, which states, in part, that any occurrence of an unusual or important event that indicates or could result in environmental impact causally related to plant operation shall be recorded and reported to the NRC within 24 hours followed by a written report. The licensee notified the NRC Resident Inspector.
ENS 477101 March 2012 05:51:00PerryNRC Region 3GE-6On March 1,2012, at approximately 0224 (EST), a manual Reactor Protection System (RPS) actuation was initiated due to 3 turbine bypass valves going open as a result of an automatic turbine runback signal. At the time of the event, the plant was in Mode 1 at 100% power. All control rods are inserted into the core and the plant is currently stable in Mode 3 (Hot Shutdown) with reactor pressure at approximately 930 psig. No Emergency Core Cooling Systems were required or utilized to respond to the event and there were no other reportable actuations. Reactor coolant level is being maintained in its normal band by the feedwater system and decay heat is being removed by the condenser. The plant is in a normal electrical line-up with all three Emergency Diesel Generators operable and available if needed. The cause of the automatic turbine runback has not been determined and is being investigated. During the transient, Reactor Water Cleanup System (RWCU) tripped. No automatic isolation signal was received. At the time of the event, restoration of a Stator Water Cooling pressure gauge was being performed (following maintenance). The NRC Resident Inspector has been notified.
ENS 4754521 December 2011 21:37:00PerryNRC Region 3GE-6

On December 21, 2011, at 1359 hours EST, it was determined that the test instruction for the weekly groundwater level readings contained non-conservative acceptance criteria. The test instruction acceptance criteria for groundwater level exceeds the initial assumptions used in the Updated Safety Analysis Report (USAR) Chapter 15 accident analysis for 'Postulated Radioactive Releases due to Liquid Containing Tank Failures.' The issue was identified during the conduct of a prompt functionality assessment evaluating the plant underdrain system performance and to ensure all USAR described functions were being met. The test instruction acceptance criteria is less than 575 feet. The calculation supporting the Chapter 15 accident analysis assumes an initial groundwater elevation of 568 feet in order to accumulate a sufficient volume of groundwater to dilute the tank inventory prior to exiting the underdrain system. The accumulation volume results in hold up time allowing for mixing and radioactive decay. With the degraded performance of the underdrain system pumps and high precipitation, the groundwater level had risen above 568 feet but was still less than 575 feet. A preliminary engineering evaluation has determined that significant margin exists in the calculation. A calculation revision is being pursued in parallel with this notification. Compensatory actions involving the repair of 1 permanent non-safety pump and installation of temporary pumps were previously initiated and the groundwater levels are decreasing. Restoration of the remaining permanent plant pumps continues. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 2/15/12 AT 1315 EST FROM LLOYD ZERR TO ERIC SIMPSON * * *

The calculation supporting the USAR Chapter 15 accident analysis was reviewed and subsequently revised. The revised calculation verifies and initial groundwater elevation of 575 feet is consistent with the preliminary assessment that substantial margin existed in the calculation for the underdrain system. Because the condition reported in Event Number 47545 was not an event or condition that results in the nuclear power plant being in an unanalyzed condition that degrades safety, the condition is not reportable, and this notification is retracted. The evaluation for this condition is documented in condition report 2011-07169. The NRC Resident Inspector and R3DO (Passehl) were notified.

ENS 475159 December 2011 22:41:00PerryNRC Region 3GE-6

On December 7, 2011, a 10 CFR 21 report (reference NRC EN No. 47498) was received from a vendor for a defect with NUS Controllers. The defect involves spring clips that form part of the seismic restraints for the controllers. The controllers referenced in the report are installed for the Reactor Core Isolation Cooling (RCIC) system in the control room and remote shutdown panel. Based on initial information provided by the vendor, it was determined that the RCIC system remained operable. On December 9, 2011, additional information provided by the vendor did not support the immediate operability determination and the RCIC system was declared inoperable for Technical Specification (TS) Limiting Condition for Operation (LCO) 3.5.3 Condition A at 1835 hours (EST). At 1932 hours (EST), the High Pressure Core Spray system was verified operable per TS LCO 3.5.3 Required Action A.1. TS LCO 3.5.3 Required Action A2 requires restoration of the RCIC system to operable status within 14 days. Qualified spring clips have been obtained and will be installed on the controllers. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system needed to remove residual heat. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM CHARLES ELBERFELD TO JOHN KNOKE AT 1415 EST ON 12/10/11 * * *

As a follow-up to the condition reported above, we have replaced the affected seismic clips on the controllers and the Reactor Core Isolation Cooling system is now operable as of 0734 on December 10, 2011. The NRC Resident Inspector has been notified." R3DO (Skokowski) notified.

  • * * RETRACTION FROM LLOYD ZERR TO CHARLES TEAL ON 2/6/12 AT 1504 EST * * *

The vendor provided a seismic report to the station. This report showed that the seismic clips holding the Reactor Core Isolation Cooling (RCIC) controller meet the Operating Basis Earthquake (OBE) test requirements and design requirements for a Safe Shutdown Earthquake (SSE) for Perry. Based on this review, it was determined that the spring clips would function properly during and OBE and SSE. Because the condition reported in Event Number 47515 would not have prevented the fulfillment of the safety function of a system needed to remove residual heat, the condition is not reportable, and this notification is being retracted. The evaluation for this condition is documented in condition report 2011-06531. The NRC Resident Inspector has been informed." Notified R3DO (Giessner) and Part 21 Group via email.

ENS 475087 December 2011 18:29:00PerryNRC Region 3GE-6

On November 16, 2011, at 2000 hours (EST), control room operators accepted the results of an immediate investigation related to the adequacy of a calculation related to the flooding analysis for service water piping in the control complex. The conclusions of the original analysis assumed operator actions to mitigate the flooding. The existing procedural guidance at that time lacked specificity for the required operator actions. This was identified as a non-conforming condition with respect to a USAR flooding analysis. The immediate investigation determined that significant margin exists for mitigation of the service water leakage crack compared to the 30 minute actions required in the original analysis. A preliminary strategy for flood mitigation was identified in the immediate investigation. Guidance was provided to the operators for a leak mitigation strategy in a night order, and a prompt functionality assessment was requested for the control complex building with respect to the flooding analysis. On November 22, 2011, at 1957 hours (EST), the prompt functionality assessment was accepted by the control room operators. The assessment included compensatory measures that simplified guidance for mitigating the flooding from the service water system. The compensatory measures were implemented. On December 2, 2011, the NRC Component Design Basis Inspection team debriefed that the condition identified should have been called in to the NRC Operations Center within eight hours and that missing the call was a violation of 10 CFR 50.72(b)(3)(ii)(B). On December 7, 2011, at 1320 hours (EST), a call was received from the NRC Region III informing the compliance supervisor that the eight-hour notification should still be made. The NRC Resident Inspector has been notified.

  • * * UPDATE FROM CHARLES ELBERFELD TO JOHN KNOKE AT 1412 EST ON 12/10/11 * * *

The licensee added further clarification to the event reported above as follows: Given that an unanalyzed condition existed until the appropriate measures were implemented, an eight-hour call in accordance with the aforementioned section of 10 CFR 50.72 should have been made. After further consideration, station management decided the eight-hour call was missed and it is being reported as required. The NRC Resident Inspector has been notified." R3DO (Skokowski) notified.