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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56673 | 10 August 2023 04:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Trip | The 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. | Feedwater Reactor Protection System Reactor Core Isolation Cooling Reactor Pressure Vessel Core Spray Emergency Core Cooling System Main Steam Line Safety Relief Valve | |
| ENS 56298 | 5 January 2023 17:42:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Scram | The 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. | Feedwater Reactor Protection System Reactor Pressure Vessel High Pressure Core Spray Main Condenser | |
| ENS 54185 | 27 July 2019 23:29:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Scram Due to Main Turbine Trip | At 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. | Feedwater Main Steam Isolation Valve Reactor Pressure Vessel Main Condenser | |
| ENS 53896 | 25 February 2019 05:24:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Due to Generator Trip | At 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. | Reactor Protection System Main Condenser | |
| ENS 51716 | 8 February 2016 20:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Following Spurious Opening of Two Safety Relief Valves | At 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. | Feedwater Main Steam Isolation Valve Safety Relief Valve Main Condenser | 05000440/LER-2016-002 |
| ENS 51679 | 24 January 2016 02:22:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Ts Required Shutdown Due to Unidentified Leakage in Drywell | 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.
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.
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. | Reactor Coolant System Feedwater Reactor Protection System Main Condenser | 05000440/LER-2016-001 |
| ENS 50601 | 7 November 2014 13:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Scram Due to Loss of Feedwater | The 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. | Feedwater Emergency Diesel Generator Reactor Core Isolation Cooling High Pressure Core Spray Safety Relief Valve Main Condenser | |
| ENS 50551 | 20 October 2014 06:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Scram on Loss of Feedwater | 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.
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). | Feedwater Emergency Diesel Generator Reactor Core Isolation Cooling High Pressure Core Spray | |
| ENS 48688 | 22 January 2013 08:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Protection System Actuation | On January 22, 2013, at approximately 0332 hours (EDT), an automatic Reactor Protection System (RPS) actuation occurred at the Perry Nuclear Power Plant, Unit 1. At the time of the event, the plant was in Mode 1 at 100% power. All control rods are inserted into the reactor core and the plant is currently stable in Mode 3 (Hot Shutdown) with reactor pressure and level being maintained in the normal shutdown range. The RPS actuation was initiated by a low reactor water level (Level 3 - 178") signal. In response to the RPS actuation and subsequent reactor Level 2 (130") signal, the High Pressure Core Spray (HPCS) system and Reactor Core Isolation Cooling (RCIC) system both actuated and injected to maintain reactor coolant level. The reactor level is currently being maintained in its normal band by the feedwater system and decay heat is being removed by (turbine bypass valves to) the condenser (both HPCS and RCIC have been returned to standby). The plant is in a normal electrical line-up with all three Emergency Diesel Generators operable and available, if needed. The Containment Isolation Valves (responded to the Level 2 and 3) isolation signals as designed. The cause of the RPS actuation is under investigation. The NRC Resident Inspector has been notified. | Feedwater Reactor Protection System Emergency Diesel Generator Reactor Core Isolation Cooling High Pressure Core Spray Control Rod | 05000440/LER-2013-001 |
| ENS 47710 | 1 March 2012 05:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Protection System Actuation Due to Automatic Turbine Runback | On 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. | Feedwater Reactor Protection System Emergency Diesel Generator Reactor Water Cleanup Stator Water Cooling Emergency Core Cooling System Control Rod | 05000440/LER-2012-001 |
| ENS 45918 | 12 May 2010 03:18:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Manual Reactor Scram Due to Loss of Control Rod Drive Charging Water Header Pumps | On 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. | Feedwater Reactor Protection System Emergency Diesel Generator Emergency Core Cooling System Control Rod | 05000440/LER-2010-003 |
| ENS 45440 | 16 October 2009 04:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Reactor Recirculation Pump Trip | On 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. | Service water Reactor Recirculation Pump Control Rod | 05000440/LER-2009-003 |
| ENS 45147 | 21 June 2009 21:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Automatic Reactor Scram Related to a Main Turbine Trip | On 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. | Feedwater Reactor Protection System Emergency Diesel Generator Emergency Core Cooling System Safety Relief Valve Control Rod | 05000440/LER-2009-001 |
| ENS 43808 | 28 November 2007 12:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Scram and Eccs Injection | A reactor scram occurred at full power due to either a main turbine trip or loss of feedwater (cause is still under investigation). All rods fully inserted. RCIC started as expected but tripped shortly thereafter on a preliminary indication of low suction pressure. The Digital Feedwater System backup motor driven feedwater pump did not function as required and reactor water level decreased to level 2 ( 130 inches). High Pressure Core Spray (HPCS) started automatically at level 2 and restored water level. Currently reactor water level is at 188 inches and reactor pressure is at 927 PSI. Decay heat is being removed via the turbine bypass valves. No other significant equipment was out of service at the time of the scram. The scram had no impact on offsite or onsite power availability. The licensee attempted to restore RCIC a second time and experienced another trip. In addition, the licensee attempted to restore the digital feedwater and was unsuccessful. Feedwater continues to be supplied as needed via the HPCS while the licensee attempts to restore RCIC and Digital Feedwater System. The licensee notified the NRC Resident Inspector. | Feedwater High Pressure Core Spray | 05000440/LER-2007-004 |
| ENS 43363 | 15 May 2007 04:58:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram During Testing | 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.
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). | Feedwater Main Steam Isolation Valve Residual Heat Removal Main Condenser Control Rod | 05000440/LER-2007-001 |
| ENS 43049 | 13 December 2006 09:35:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram - Loss of Instrument Air Resulting in Feedwater Transient and Manual Reactor Scram | 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.
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). | Feedwater Reactor Protection System Control Rod | 05000440/LER-2006-005 |
| ENS 41310 | 6 January 2005 06:12:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Manual Reactor Scram Due to Reactor Recirculation Pump Trip | At 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. | Feedwater Main Steam Isolation Valve Reactor Core Isolation Cooling Reactor Recirculation Pump Safety Relief Valve Control Rod | 05000440/LER-2005-001 |
| ENS 41290 | 24 December 2004 04:54:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation | Reactor Scram Following Down Shift of Both Recirc Pumps from Fast to Slow Speed | At 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. | Safety Relief Valve | 05000440/LER-2004-002 |