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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).

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).

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).

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.

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).

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).

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)

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).

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.