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 Entered dateSiteRegionReactor typeEvent description
ENS 5744426 November 2024 14:14:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: At 1158 EST on 11/26/24, the Division 3 diesel generator was declared inoperable due to failure of the right bank air start motor during a planned monthly surveillance run. Troubleshooting of the issue is in progress. This condition could prevent the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). All other emergency core cooling systems were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Division 3 diesel generator supports high pressure core spray, a single train system.
ENS 5740729 October 2024 18:07:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: At 1730 (EDT), on 10/29/2024, Vistra authorized a report to the State of Ohio in accordance with NEI 07-07, 'Industry Groundwater Protection Initiative,' of a liquid spill to the plant's outdoor ground from an outdoor cask holding radioactive waste. This spill contained radionuclides cobalt-60 and manganese-54 with activities greater than the site procedural limits for state reporting. The outdoor cask has been removed from the onsite storage area; remediation and causal investigation activities are being performed. This exceedance did not exceed any NRC regulations or reporting criteria. There was no impact on the health and safety of the public or plant personnel. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No EPA limits were exceeded
ENS 5740629 October 2024 17:32:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: At 1152 EDT, on 10/29/2024, the division 3 diesel generator was declared inoperable due to variations in steady state voltage during a planned monthly surveillance run. Troubleshooting of the issue is in progress. This condition could prevent the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). All other emergency core cooling systems were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Limiting condition for operations (LCO) 3.8.1.b was entered as a result of this event.
ENS 5723824 July 2024 13:22:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: At 1046 EDT on 7/24/2024, fourteen Ashtabula County emergency response sirens were inadvertently activated during system maintenance. Ashtabula County Emergency Management Agency was contacted and made aware of the situation. This notification is being made solely as a four-hour, non-emergency notification for a notification of other government agencies in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact to the health and safety of the public or plant personnel. The NRC resident inspector has been notified of the issue.
ENS 5718119 June 2024 15:07:00PerryNRC Region 3GE-6

The following information was provided by the licensee via phone and email: At 1640 EDT on 06/18/2024, the division 3 diesel generator was declared inoperable. This condition could prevent the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). All other emergency core cooling systems were operable during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The division 3 diesel generator was declared inoperable due to potential water intrusion into the electrical generator. Inspection of the generator is in progress. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: This event resulted in Perry Unit 1 entering a 72 hour limiting condition for operation (LCO) in accordance with Technical Specification 3.8.1. condition 'B'.

  • * * RETRACTION ON 8/15/2024 AT 0943 EDT FROM TONY MCGOWAN TO ERNEST WEST * * *

Investigation and assessment of the division 3 diesel generator exciter and generator as-found condition revealed that no water intrusion occurred and, therefore, would not have impacted the diesel generator's ability to start, run, and produce the required electrical power to perform its required safety function. Therefore, EN 57181 is being retracted. The NRC resident inspector has been notified of the Event Notification retraction. Notified R3DO (Feliz-Adorno)

ENS 5713623 May 2024 04:25:00PerryNRC Region 3GE-6The following information was provided by the licensee via email: At 0400 EDT on May 23, 2024, a technical specification required shutdown was initiated at Perry. Technical specification action 3.4.5 condition B (unidentified reactor coolant system leakage exceeds 5 gallons per minute) was entered on May 23, 2024 at 0000 with a required action to reduce leakage to within limits within 4 hours, due by 0400 on May 23, 2024. This required action was not completed within the completion time, therefore, a technical specification required shutdown was initiated, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The reactor coolant leakage was stable at approximately 6 gallons per minute.
ENS 5713016 May 2024 20:53:00PerryNRC Region 3GE-6The following information was provided by the licensee via email: On May 16, 2024 at 0840 EDT, operations declared the reactor water cleanup (RWCU) leak detection instruments related to the high differential flow signal inoperable. Technical specification (TS) 3.3.6.1, primary containment and drywell isolation instrumentation, conditions `A and `B were entered as one required channel of instrumentation was inoperable, and an automatic function with isolation capability was not maintained. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D). All other RWCU primary containment isolation instrumentation functions remained operable. At 1210 EDT, the affected leak detection instruments were declared operable, and the TS limiting condition for operation 3.3.6.1 was declared met. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5709225 April 2024 20:24:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: On April 25, 2024, it was determined that between March 25, 2024, 2015 (EDT) and March 30, 2024, 2024 (EDT), the condensate transfer and storage system was employed as a method of alternate keepfill in place of the installed residual heat removal (RHR) systems waterleg pump for RHR system loops `B and `C. This condition is not bounded by existing design and licensing documents. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 569145 January 2024 15:56:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: At 1552 (EST) on 01/05/2024, Perry Nuclear Power Plant reported elevated levels of tritium in the underdrain system to the state of Ohio as a non-voluntary reporting of tritium. An investigation is currently ongoing to identify the cause of the elevated tritium levels. The tritium levels in this location do not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5667310 August 2023 04:03:00PerryNRC Region 3GE-6The following information was provided by the licensee via email: At 0039 (EDT) on 8/10/23, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped during a reactor protection system (RPS) bus shift. All systems responding normally post-trip. There was no equipment inoperable at the time of the trip. Operations responded and stabilized the plant. Reactor water level being maintained via feedwater. Decay heat is being removed by cycling safety relief valves. An actuation of high-pressure core spray, division 3 diesel generator, and reactor core isolation cooling occurred during the scram and main steam line isolation closure. The reason for the auto-start was reaching Level 2 (130 inches in the reactor pressure vessel) during the transient. The systems automatically started as designed and injected to the reactor vessel when the Level 2 signal was received. The RPS actuation is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The emergency core cooling system (ECCS) injection is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). The ECCS actuation is being reported as a eight-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5658823 June 2023 17:04:00PerryNRC Region 3GE-6The following information was provided by the licensee via email: At 1521 EDT on 6/23/2023, Perry Nuclear Power Plant reported elevated levels of tritium in the underdrain system to the State of Ohio as a non-voluntary reporting of tritium. An investigation is currently ongoing to identify the cause of the elevated tritium levels. The tritium levels did not exceed any NRC regulations or reporting criteria. Tritium has not been detected in any other locations and is not expected to impact groundwater or exceed any limits in the Off Site Dose Calculation Manual (ODCM). This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The following agencies were notified by licensee: Lake County Emergency Management Agency (EMA) Ashtabula County EMA Geauga County Department of Emergency Services Ohio EMA Radiological Branch
ENS 564556 April 2023 16:46:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: On March 4, 2023, it was determined that the main steam line (MSL) local leak rate test results for MSL 'B' were in exceedance of technical specification (TS) surveillance requirement (SR) 3.6.1.3.10 limits. Additionally, the leakage at the outboard main steam isolation valve (MSIV) 'B', was indeterminate due to a gross packing gland leak. An engineering calculation dated April 6, 2023, showed that this leakage, in conjunction with a design basis loss of coolant accident, would result in the radiological dose exceeding Updated Safety Analysis Report limits to the exclusion area boundary, the low population zone, and the control room. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the power plant being in an unanalyzed condition that degrades plant safety. Both inboard and outboard 'B' MSIVs have been reworked and are within the TS SR limits. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5641717 March 2023 10:15:00PerryNRC Region 3GE-6The following information was provided by the licensee via email: On March 16, 2023, at 2226 EDT, a site supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's authorization for access to the plant has been terminated. The Resident Inspector has been notified.
ENS 562985 January 2023 15:39:00PerryNRC Region 3GE-6The following information was provided by the licensee via phone and email: At 1242 (EST) on 05 January 2023, with the Unit in Mode 1 at 99 percent power, the reactor automatically tripped on low Reactor Pressure Vessel level while restoring power to Digital Feedwater Control Stations when there was a perturbation to the level controls. The reason for perturbation is unknown at this time. The trip was not complex, with all systems responding normally post trip. Operations responded and stabilized the plant. High pressure core spray was manually initiated in accordance with site procedures. Reactor water level is being maintained via the Feedwater System. Decay heat is being removed by the Main Condenser. 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)(A) and 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.
ENS 5596224 June 2022 16:28:00PerryNRC Region 3GE-6The following information was provided by the licensee via telephone: At 1257 EDT on June 24, 2022, it was discovered the Low Pressure Core Spray System (LPCS) 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)(D). Inoperability of the Low Pressure Core Spray system was caused by a loss of power to the LPCS Minimum Flow Valve during surveillance activities. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5526318 May 2021 09:04:00PerryNRC Region 3GE-6This 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 specific system actuation. On March 23, 2021, during the performance of the Division 1 ECCS ((Emergency Core Cooling System)) Integrated Test, the Division 1 Diesel Generator (DG) unexpectedly started. While performing the local lockout testing, per the procedure, a step was performed that initiated the unexpected DG start. The following step was to verify the diesel did NOT start. It was later determined that this was a procedural deficiency. The DG started and ran as designed. The DG did not tie to the safety bus as no undervoltage condition was detected. This event is considered an invalid system actuation reportable under 10 CFR 50.73(a)(2)(iv)(A). The actuation 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. All affected systems functioned as expected in response to the actuation. The DG was shut down in accordance with plant procedures and the testing procedure corrected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 551726 April 2021 04:32:00PerryNRC Region 3GE-6

At 2149 EDT on April 5, 2021, with the power plant in Mode 2 at zero percent power, an actuation of the RPS system occurred following the decision to abort plant start-up. The reason for the RPS actuation was to align the plant to Mode 3, from Mode 2, following manually inserting all control rods using the Rod Control System. The RPS system initiated as designed when the mode switch was taken from 'Start-up' to 'Shutdown' to align the plant to Mode 3 from Mode 2. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 5/12/21 AT 1345 EDT FROM JOHN NAKEL TO KERBY SCALES * * *

This is a retraction of an event notification made on 4/6/2021 at 0432 EST (EN#55172). This event was initially reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS System. This event was later determined to be pre-planned, in accordance with Technical Specifications, and not the result of a significant event, therefore not meeting the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). On the evening of April 4, 2021, while commencing reactor start up, it was determined that control rod withdrawal to add positive reactivity for the start-up would not overcome the negative reactivity of plant heat up. The control room team determined that the proper course of action would be to insert all control rods . The control room briefed and notified the Outage Control Center about its decision, then proceeded to insert all control rods. The control room manually inserted all control rods using the control rod hydraulic system. Following insertion of all control rods, the mode switch was taken to the shutdown position to meet the prerequisites of the procedure for maintaining hot shutdown. This action establishes Mode 3 in accordance with Technical Specifications and aligns the plant to perform the necessary work prior to a plant restart. By placing the mode switch in the shutdown position, a scram signal is generated for 10 seconds. NUREG-1022 offers guidance that states 'Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event.' The actions the operating crew took that night are accurately described by this statement in NUREG-1022 'shifting alignment of makeup pumps or closing a containment isolation valve for normal operational purposes would not be reportable.' In this situation, the Mode switch was taken to shutdown to align the plant to mode 3 for normal operational purposes, and not to mitigate a significant event. When the mode switch was taken to shut-down, RPS initiated as designed, there was no mis-operation or unnecessary actuation. This actuation was determined to be pre-planned, in accordance with Tech Specs, and not the result of a significant event, therefore not meeting the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident has been notified. Notified R3DO (McGraw).

ENS 5484821 August 2020 12:42:00PerryNRC Region 3At 0953 EDT on 8/21/20, it was discovered that both trains of the standby liquid control system were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). All control rods remained operable during this time period. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Both trains of the standby liquid control system were declared inoperable due to an inadvertent addition of water to the storage tank which caused the boron concentration in the tank to go low out of specification.
ENS 545071 February 2020 16:29:00PerryNRC Region 3

EN Revision Imported Date : 3/17/2020 LOW PRESSURE CORE SPRAY INOPERABLE At 1150 EST on 2/1/2020, it was discovered that the Low Pressure Core Spray System was inoperable due to a divisional battery voltage being out-of-specification. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The Low Pressure Core Spray is a single train safety system in Modes 1, 2, and 3. Low Pressure Core Spray was restored to operable, restoring function at 1230 on 2/1/2020. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 03/16/2020 AT 1156 EDT FROM JOHN NAKEL TO BETHANY CECERE * * *

On February 01, 2020, event notification (EN-54507) was made to the NRC for Low Pressure Core Spray (LPCS) inoperability. This notification was made due to high DC bus voltages resulting in LPCS being declared inoperable which resulted in a loss of safety function. An Engineering Evaluation Request (EER) was performed to determine an upper limit for DC bus voltage. This EER determined that LPCS could perform its required functions with a voltage increase of up to 150V DC if the duration was not greater than two hours. The elevated voltage was experienced for approximately 59 minutes. The maximum voltage experienced was 147.07V DC. Therefore, LPCS remained operable and no loss of safety function existed. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. Notified R3DO (Hanna).

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 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.