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 Entered dateSiteRegionScramReactor typeEvent description
ENS 5418527 July 2019 22:31:00PerryNRC Region 3Automatic ScramAt 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 5389625 February 2019 03:44:00PerryNRC Region 3Automatic ScramAt 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 527273 May 2017 12:59:00PerryNRC Region 3Manual ScramGE-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 517168 February 2016 17:50:00PerryNRC Region 3Manual ScramGE-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 3Manual ScramGE-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 506017 November 2014 11:13:00PerryNRC Region 3Automatic ScramGE-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 3Automatic ScramGE-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 477101 March 2012 05:51:00PerryNRC Region 3Manual ScramGE-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 4591812 May 2010 03:12:00PerryNRC Region 3Manual ScramGE-6On May 11, 2010, at approximately 2318 hours, a manual Reactor Protection System (RPS) actuation was initiated as required by Technical Specification (TS) Limiting Condition for Operation (LCO) 3.1.5 'Control Rod Scram Accumulators.' Control Rod Drive (CRD) charging water header pressure was less than 1520 psig (i.e., no CRD pumps operating) and there were multiple accumulator faults on withdrawn control rods. 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 10 CFR 50.72 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 event initiator, an invalid Division 2 Loss of Coolant Accident (LOCA), i.e., High Drywell Pressure/Low Reactor Vessel Water Level, signal, is currently under investigation. Prior to the manual RPS Actuation, the invalid LOCA signal resulted in invalid actuations of Division 2 equipment and systems including, the Division 2 Emergency Diesel Generator (EDG), (which started but did not load onto the bus), Low Pressure Coolant Injection B and C subsystems (which started the pumps but did not inject into the vessel), discharge of the Suppression Pool Makeup subsystem B into the suppression pool, startup of the Control Room Emergency Recirculation subsystem B, and isolation of Group 2B Containment isolation valves which included the Nuclear Closed Cooling System Containment Return Isolation valve that was not already closed. The affected equipment is being restored in accordance with plant procedure. The NRC Resident Inspector has been notified. The licensee experienced an instrumentation rack loss of power which appears to have resulted in the inadvertent Division 2 initiation. The initiator of this event also and led to a loss of power to both control rod drive charging water header pumps resulting in charging water header pressure less than required and related accumulator faults which placed the licensee in a technical specification required shutdown condition. The action statement allows only 20 minutes to restore the condition which was insufficient time for the licensee to correct the condition so a manual scram was initiated from 100% power. The scram was characterized as an uncomplicated scram and all system responses (not related to the initial instrument fault) functioned as required.
ENS 4544016 October 2009 01:48:00PerryNRC Region 3Manual ScramGE-6On October 16, 2009 at 0048 a manual reactor scram was inserted at the Perry Nuclear Plant. The plant was conducting a planned shutdown due to the Division 2 Emergency Service Water inoperability. While shifting reactor recirculation pumps to slow speed the 'A' pump failed to transfer and tripped off. Following stabilization from this event a manual reactor scram was inserted from approximately 30% power. This was different from the initial planned shutdown sequence. Following the scram all systems operated as expected. The plant is stable in Mode 3. The plant will transition to Mode 4 in accordance with Technical Specification 3.7.1 (Emergency Service Water Inoperability) required actions. All control rods fully inserted and the plant electrical power is in a normal line-up. The licensee notified the NRC Resident Inspector.
ENS 4514721 June 2009 20:58:00PerryNRC Region 3Automatic ScramGE-6On June 21, 2009, at approximately 1750 hours, 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 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 reactor scram is currently under investigation. Preliminary indications are that the cause of the RPS actuation is related to a main turbine trip. The NRC Resident Inspector has been notified. No safety relief valves lifted during the event and a reactor cooldown is in progress.
ENS 4336315 May 2007 01:53:00PerryNRC Region 3Automatic ScramGE-6

An automatic reactor scram occurred due to lowering reactor water level. Digital feedwater tuning activities were in progress at the time of the scram. All systems functioned as designed. The reactor water level has been restored to normal level band. The plant is stable in Hot Shutdown. There were no ECCS injections. At the time of the scram, the feedwater pump was in manual control for feedwater tuning. When water level started going down quickly, the operator was not able to restore sufficient feedwater flow before the level 3 (water level low) actuation. All control rods fully inserted on the scram. No valves repositioned and no safety or relief valves lifted after the scram. Reactor water level is being maintained with the motor feed pump and decay heat is being removed to the main condenser. The plant is in the normal shutdown electrical lineup. Reactor pressure is 509 psi and stable. The licensee notified the NRC Resident Inspector.

  • * * UPDATE FROM FREDERICK SMITH TO FANGIE JONES ON 05/17/07 AT 1301 * * *

This call is to clarify information provided in notification Event Number 43363 made by the Perry Nuclear Power Plant on May 15, 2007. The second paragraph, fourth sentence of the notification states: 'No valves repositioned and no safety or relief valves lifted after the scram.' The intent of the portion of the sentence 'No valves repositioned' was supposed to be in reference to the reporting requirement in 10 CFR 50.72(b)(3)(iv) to report any event or condition that results in valid actuation of general containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).

The event on May 15, 2007, did not meet the 'containment isolation valves in more than one system / MSIVs criterion', however, subsequent review of the event identified that Residual Heat Removal 'B' Heat Exchanger Second Vent to Suppression Pool Containment Isolation Valve 1E12-F073B, closed, as designed, in response to the reactor coolant level 3 (water level low) condition that was present during the event. No additional reporting criteria have been identified and this update has been provided to clarify the earlier statement. The NRC Resident Inspector has been notified." Notified R3 RDO, (Ring), and NRR EO, (Ross Lee).

ENS 4304913 December 2006 05:49:00PerryNRC Region 3Manual ScramGE-6

A loss of instrument air caused a transient in feed system causing a lowering Hot Surge Tank Level. The Reactor was manually scrammed when Reactor Feed Booster Pumps were cavitating. Control Rod 42-55 did not insert on the initial scram. Control Rod 42-55 did insert when ARI was manually initiated. The loss of instrument air was the result of an air line rupture, and the licensee is steaming to the condenser while maintaining vessel level with feedwater. There was no ECCS injection, and the licensee is investigating the cause of the line rupture. The licensee notified the NRC Resident Inspector.

  • * * UPDATE AT 0814 ON 12/13/2006 FROM MICHAEL BROGAN TO MARK ABRAMOVITZ * * *

This update is being made in accordance with 10CFR50.72(c)(2) to immediately report the results of ensuing evaluations or assessments of plant conditions, the effectiveness of response or protective measures taken, and information related to plant behavior that is not understood. On December 13, 2006, at 0549 hours, the Perry Nuclear Power Plant notified the NRC Operations Center (Event Number 43049) of a manual Reactor Protection System actuation associated with a loss of instrument air. In the initial report, it was stated that control rod 42-55 did not insert on the initial scram, but did insert when Alternate Rod Insertion was manually initiated. Subsequently, reactor engineering review of control rod performance during the event determined that the control rod scram time for control rod 4255 was satisfactory and that the problem was with the control rod indication. The plant remains stable at this time. Control rod indication has two channels. Control rod 42-55 displayed one fully inserted indication and one blinking green indication (i.e. not fully inserted). When the system was reset, all rods indicated fully inserted with a "00" position indication. The licensee notified the NRC Resident Inspector. Notified the R3DO (Lara).

ENS 413106 January 2005 04:01:00PerryNRC Region 3Manual ScramGE-6At approximately 0106 on 1-6-05, Reactor Recirculation Pumps A and B down-shifted from fast to slow speed which resulted in reactor power decreasing from 100% to approximately 46%. As operators started to reduce power using control rod insertion, Reactor Recirculation Pump A tripped to 'off.' At 0112 on 1-6-05, a manual reactor scram was inserted due to operating under undesirable power to flow conditions. At approximately 0119, the operators were unable to start the Motor Feedwater Pump, Reactor Core Isolation Cooling system was manually started. Level control was established using a Reactor Feed Pump Turbine and Reactor Core Isolation Cooling. The Main Steam Isolation Valves were closed to limit cooldown. Reactor level is being controlled with the Reactor Core Isolation Cooling System and safety relief valves are available for reactor pressure control. The cause of the Reactor Recirculation Pumps down-shifting and the subsequent trip of Reactor Recirculation Pump A is still under investigation. The cause of the Motor Feedwater Pump failure to start is likewise under investigation. All control rods fully inserted. The lowest reactor level reached was 154 inches above TAF. The electrical grid is stable and ESF systems remain available. Reactor pressure and level are being maintained by the Reactor Core Isolation Cooling system. The licensee notified the NRC Resident Inspector.
ENS 4129024 December 2004 03:07:00PerryNRC Region 3Automatic ScramGE-6At 2345 on 12/23/04 both reactor recirc pumps down shifted from fast to slow speed which resulted in reactor power reducing from 100% to 44%. At 2354 an automatic scram occurred from the activation of the Oscillation Power Range Monitor (OPRM) instrumentation. After the scram all safety systems responded as designed. Cause of the reactor recirc pump downshift is still under discovery. All rods fully inserted. No safety relief valves lifted during the transient. The licensee was in no major LCO at the time. Pressure control is via the steam bypass valves. The licensee has notified the NRC Resident Inspector.
ENS 4059819 March 2004 16:17:00PerryNRC Region 3Manual ScramGE-6This report is being made in accordance with 10CFR50.73 (a)(1), which states, in part, "in the case of an, invalid actuation reported under 10 CFR50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER." These invalid actuations are being reported under 10CFR50.73(a)(2)(iv)(A). On February 22, 2004, at 1016 hours, while operating at 100 percent power, the reactor protection system manual scram channel functional test was performed. When a division 1 manual scram pushbutton was depressed, in addition to the expected half scram, the reactor protection system (RPS) A motor-generator electrical protection assembly circuit breakers tripped open. Loss of power from the RPS A motor-generator caused the loss of electrical power to the RPS instrumentation. This resulted in the invalid actuation of the logic that is powered by RPS A. The actuations resulted in isolation signals that closed one or more valves in each of the following division 1 subsystems: main steam line drains, containment and drywell radiation monitors, reactor water cleanup, fuel pool cooling, suppression pool cleanup, containment and drywell radwaste sumps and containment chilled water. All systems and components responded as designed for the signals that resulted from the loss of RPS A buss. This is considered a partial actuation since only division 1 components were effected. Following restoration of power to the RIPS A buss, the actuation logic was reset, and the equipment/systems were returned to the status required by plant conditions. Discussion of the causes and corrective actions associated with this event are documented in the corrective action program in condition report 04-00901. The resident inspector has been notified.
ENS 4006314 August 2003 21:33:00PerryNRC Region 3Automatic ScramGE-6

Automatic reactor scram due to a loss of offsite power. All rods fully inserted. Supplying power to vital buses via emergency diesel generators. All system operating properly. NRC Resident Inspector was notified of the event by the licensee.

  • * * UPDATE ON 08/15/03 @ 2007 BY STRACH TO GOULD * * *

The NOUE was terminated at 1952. Off site power was restored and the emergency diesels secured. Notified FEMA (Austin), Reg 3 RDO (Burgess) and EO (Richards)