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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56464 | 16 February 2023 12:05:00 | Other Unspec Reqmnt 10 CFR 50.73(a)(1), Submit an LER | 60 Day Notification for an Invalid Actuation of the Emergency Service Water System | The following information was provided by the licensee via email: 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 specific system actuation of the Emergency Service Water (ESW) System. On 2/16/2023, while performing a calibration planned maintenance (PM) for a jacket water pressure indicator during a D13 diesel generator system outage window, the 'C' ESW pump unexpectedly auto-started. Subsequent investigation identified that the affected jacket water pressure indicator shares a common sensing line with a jacket water pressure switch that provides a back-up to the engine speed switch for the engine running signal. At the time the jacket water pressure indicator calibration PM was being performed, the power circuits for D13 diesel generator instrumentation were energized. Pressurization of the energized jacket water pressure switch during the pressure indicator calibration activity resulted in initiation of a false engine running signal to the `C' ESW pump start logic. This event is considered an invalid system actuation because the 'C' ESW pump started in response to a false signal that the D13 EDG was running when the D13 EDG did not start. The actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. The ESW system functioned as expected in response to the actuation. The affected ESW pump was shut down in accordance with plant procedures. There was no impact on the health and safety of the public or plant personnel. The licensee notified the NRC Resident Inspector. | Service water | |
| ENS 56283 | 2 November 2022 23:29:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-DAY Telephonic Notification - Invalid Specific System Actuation | The following information was provided by the licensee via email: This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid specific system actuation of the Emergency Service Water System (ESW). On 11/2/2022, during normal reactor operations, multiple main control room alarms were received for D12 Emergency Diesel Generator (EDG) running and Unit 1 Division 2 Safeguard Battery Ground. The D12 EDG did not start; however, the 'B' ESW Pump auto started. Subsequent troubleshooting determined that the cause of the D12 EDG running alarms and the inadvertent auto start of the 'B' ESW Pump was a malfunction on the D12 EDG speed switch. This event is considered an invalid system actuation because the 'B' ESW Pump started in response to a false signal that the D12 EDG was running when D12 EDG did not start. This was a complete actuation of the ESW System and the system functioned as expected in response to the actuation. The affected ESW Pump was shut down in accordance with plant procedures and the degraded D12 EDG speed switch was replaced. There was no impact on the health and safety of the public or plant personnel. The licensee notified the NRC Resident Inspector. | Service water Emergency Diesel Generator | |
| ENS 55345 | 13 May 2021 11:00:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Containment Isolation Signal | This 60-Day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of containment isolation signal affecting more than one system. On May 13, 2021, during the restoration of the Unit 2 Refuel Floor High Radiation Isolation Logic an invalid isolation signal was received. The condition requiring an isolation signal was verified not to be present prior to restoring the logic; however, it was not recognized that a previous isolation signal was latched in and had not been reset. When the isolation logic was restored, the Primary Containment Isolation System (PCIS) isolated on the invalid signal. The systems successfully completed the isolation per the plant design and plant configuration. The following systems actuated due to the Unit 2 PCIS Group 6C Isolation: - Isolation of Containment Hydrogen and Oxygen Sampling Valves, - Start of the 2A Reactor Enclosure Recirculation System, - Trip of the Units 1 and 2 Refuel Floor HVAC, - Start of the A and B Trains of Standby Gas Treatment Systems. The NRC Resident Inspector was notified. | Primary Containment Isolation System HVAC Standby Gas Treatment System Reactor Enclosure Recirculation System | |
| ENS 54083 | 18 April 2019 06:49:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of a Unit 2 Containment Isolation Logic Due to a Blown Fuse | This 60-Day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of Limerick Generating Station Unit 2 containment isolation logic. On April 18, 2019, while performing a relay replacement on the Division 2/4 Main Steam Line logic, a partial containment isolation occurred due to a blown fuse. The following systems had components that actuated due to the partial isolation: Reactor Water Clean-Up System Primary Containment Instrument Gas System Drywell Chilled Water System Reactor Enclosure Cooling Water System Core Spray System The Residual Heat Removal System received an isolation signal; however, the system remained in service because the isolation was defeated in accordance with plant procedures. This event resulted in partial Group 2A, 3, 7A, 8A, and 8B isolations. The systems successfully functioned per the plant design and plant configuration. The licensee notified the NRC Resident Inspector. | Primary containment Core Spray Residual Heat Removal Main Steam Line | |
| ENS 54014 | 22 February 2019 15:30:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of a Unit 1 Containment Isolation Logic Due to a Blown Fuse | This 60-Day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of Limerick Generating Station Unit 1 containment isolation logic. On February 22, 2019, while performing work on the 1C Main Seam Line Rad Monitor a partial containment isolation occurred due to a blown fuse. The blown fuse caused a single channel 'C' isolation signal for the Refueling Area Ventilation Exhaust High Radiation and the Reactor Enclosure Ventilation Exhaust-High Radiation logic. The following systems had components that actuated due to the partial isolation: - Plant Process Radiation Monitoring System - Nuclear Boiler System - Control Rod Drive Hydraulic System - Containment Atmospheric Control System - Primary Containment Instrument Gas System This event resulted in partial Group VIC and partial Group VIIIB isolations. All the components that would actuate on a single 'C' isolation signal responded as designed. The licensee notified the NRC Resident Inspector. | Primary containment Control Rod | |
| ENS 51952 | 28 March 2016 05:50:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of Inboard Isolation Logic | On Monday, March 28, 2016, Unit 1 was in OPCON 5 (Refueling) conducting a refueling outage. A modification was being installed for an NSSSS (Nuclear Steam Supply Shutoff System) Test Box on Division 1A Group 1 NSSSS logic. At 0150 hours, a logic jumper was removed as directed by the work order and a logic fuse failed. The fuse failure caused an unplanned invalid actuation of the inboard isolation logic. The isolations were reset and the valves were restored to initial conditions at 0246 hours. On Sunday, April 3, 2016, at 0134 hours, one additional logic fuse opening event occurred during the testing which also caused an invalid actuation which was reset at 0405 hours. The fuse openings occurred during jumper manipulations as the modification was tested on the Division 1A and 1D logic during the refuel outage. The investigation determined the fuse openings were due to the testing process. The suspected devices that caused the condition are not permanent plant equipment and there is no degradation of the actual circuit. They were part of a temporary configuration that was installed to support modification installation and acceptance testing. The temporary devices have been removed. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The isolation was a partial actuation. This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report invalid automatic actuations of systems listed in paragraph (a)(2)(iv)(B). The listed system that actuated was general containment isolation signals affecting containment isolation valves in more than one system. Primary containment isolation valves (PCIVs) closed on reactor water cleanup (RWCU), drywell chilled water (DWCW), primary containment instrument gas (PCIG), drywell sumps and the suppression pool cleanup systems. The licensee has notified the NRC Resident Inspector. | Primary Containment Isolation System Primary containment Reactor Water Cleanup | |
| ENS 48575 | 13 October 2012 23:41:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Telephonic Notification of Invalid Multiple System Actuations | On Saturday, October 13, 2012, Unit 1 was operating at 100% power. At 1841 (EDT), the 1A RPS/UPS inverter tripped and the automatic transfer of the RPS and UPS 120 VAC distribution panel (1A-Y160) loads to the primary alternate AC power source was delayed. The delay in automatic load transfer caused the RPS series breakers to trip on undervoltage. The failure caused a loss of power to Division IA and IIA RPS relays and Division IA and IIA NS4 relays. This caused primary containment isolation valves (PClVs) to automatically close on more than one system. The IB and IIB channels were unaffected. The most probable cause for the delayed load transfer was a failed logic power supply with a momentary loss of synchronization. Troubleshooting continues (in order) to confirm the specific cause of the component failure. The distribution panel loads are currently supplied by an installed alternate AC power source. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The isolation was a partial actuation. This 60-day ENS report is being made per 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report invalid automatic actuations of systems listed in paragraph (a)(2)(iv)(B). The listed system that actuated was general containment isolation signals affecting containment isolation valves in more than one system. Primary containment isolation valves (PCIVs) closed on drywell chilled water (DWCW), reactor enclosure cooling water (RECW), primary containment instrument gas (PCIG), Unit 1 containment leak detector, and Unit 2 containment leak detector. The licensee notified the NRC Resident Inspector. | Primary Containment Isolation System Primary containment | |
| ENS 46883 | 2 April 2011 16:18:00 | 10 CFR 50.73(a)(1), Submit an LER | Loss of Rps/Ups Power Supply | On Saturday, April 2, 2011, Unit 2 refueling outage activities were in progress. The 2A RPS/UPS Static Inverter was out of service and bypassed with loads transferred to the primary alternate power supply. At 1218 hours, a post maintenance test was performed on the secondary alternate power supply. The inverter alternate power manual transfer switch was transferred from the 'primary alternate' to 'secondary alternate' position to support the post maintenance test. Since the transfer switch is 'break before make' the alternate power supply was interrupted momentarily. This deenergized the 2A RPS/UPS power distribution panel loads including the Division IA and IIA RPS relays and Division IA and IIA NSSSS (Nuclear Steam Supply Shutoff System) relays. Primary containment isolation valves (PCIVs) automatically closed on more than one system. The IB and IIB channels were unaffected. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The isolation was a partial actuation of the isolation actuation instrumentation. This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report invalid automatic actuations of systems listed in paragraph (a)(2)(iv)(B). The listed system that actuated was general containment isolation signals affecting containment isolation valves in more than one system. Primary containment isolation valves (PCIVs) closed on drywell chilled water (DWCW), reactor enclosure cooling water (RECW), primary containment instrument gas (PCIG), and suppression pool cleanup. The licensee has notified the NRC Resident Inspector. | Primary Containment Isolation System Primary containment | |
| ENS 45962 | 3 April 2010 01:56:00 | 10 CFR 50.73(a)(1), Submit an LER | Overvoltage Condition Causes Reactor Protective System Actuation | On Friday April 2, 2010, Unit 1 refueling outage activities were in progress and the 1A RPS/UPS Static Inverter was being removed from service. At 2156 hours, the inverter static transfer switch was placed in 'Bypass' which transferred the load from the inverter to the secondary alternate source. The manual bypass switch was then placed in 'Bypass' which was followed by the RPS/UPS series breakers tripping on an overvoltage condition. The actuation caused a loss of power to the IA RPS/UPS power distribution panel loads which provides power to the Division 1A and IIA RPS relays and Division IA and IIA NS4 relays. This caused primary containment isolation valves (PCIVs) to automatically close on more than one system. The IB and IIB channels were unaffected. Troubleshooting determined that the secondary, alternate source voltage was 132 VAC which exceeded the overvoltage setpoint of 126 VAC. During outages, the 13kV bus voltages are higher than on-line voltages due to low operating equipment loading on the buses. The secondary alternate source to the inverter is not regulated which can result in greater than normal voltage to the RPS/UPS loads. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The isolation was a partial actuation. This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) and 10CFR 50.73(a)(1) to report invalid automatic actuation of systems listed in paragraph (a)(2)(iv)(B). The listed system that actuated was general containment isolation signals affecting containment isolation valves in more than one system. Primary containment isolation valves (PCIVs) closed on drywell chilled water (DWCW), reactor enclosure cooling water (RECW), primary containment instrument gas (PCIG), and suppression pool cleanup. The licensee notified the NRC Resident Inspector. | Primary Containment Isolation System Primary containment | |
| ENS 45518 | 19 October 2009 16:05:00 | 10 CFR 50.73(a)(1), Submit an LER | Automatic Closure of Primary Containment Isolation Valves Due to Invalid Signal | On Monday, October 19, 2009, Limerick Unit 1 was operating at 100% power. At 1105 hours, an invalid actuation of the 1B Reactor Enclosure Ventilation Exhaust Radiation Monitor occurred. The actuation caused a Division 2 Group 6C isolation signal, which caused primary containment isolation valves (PCIVs) to automatically close on the Containment Leak Detector Radiation Monitor (10-S182) and the Drywell Hydrogen/Oxygen Analyzer (10-S205). The 1A, 1C, and 1D channels were unaffected and indicated normal ventilation exhaust radiation levels during the event. The cause of the event was a failure of a fuse holder in the 1B Reactor Enclosure Ventilation Exhaust Radiation Monitor. The radiation monitor is designed to fail-safe on a loss of power. The automatic closure of the PCIVs placed them in their fail-safe position. The failed fuse holder has been replaced and the radiation monitor was declared operable on Tuesday, October 20, 2009 at 1348 hours. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The only equipment malfunction during the event was the failed fuse holder. The Division 2 Group 6C isolation was a partial actuation. This event is reportable per 10CFR50.73(a)(2)(iv)(A) since isolation valves for the Containment Leak Detector Radiation Monitor and Drywell Hydrogen/Oxygen Analyzer automatically closed due to an invalid signal. The licensee notified the NRC Resident Inspector. | Primary Containment Isolation System Primary containment | |
| ENS 44260 | 9 April 2008 08:16:00 | 10 CFR 50.73(a)(1), Submit an LER | 60-Day Optional Notification for Invalid Specified System Actuation | On Wednesday April 9, 2008 at 0416 hours, an invalid actuation of the 1B Refueling Floor Ventilation Exhaust Radiation Monitor occurred. The actuation caused a Division 2 Group 6C isolation signal, which caused primary containment isolation valves (PCIVs) to automatically close on the Containment Leak Detector Radiation Monitor (10-S182) and the Drywell Hydrogen/Oxygen Analyzer (10-S205). The 1A, 1C, and 1D channels were unaffected and indicated normal ventilation exhaust radiation levels during the event. The cause of the event was a failure of the K2 relay in the 1B Refueling Floor Ventilation Exhaust Radiation Monitor. The failed relay has been replaced and the radiation monitor was declared operable on Wednesday April 9, 2008 at 2024 hours. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The only equipment malfunction during the event was the failed K2 relay. The Division 2 Group 6C isolation was a partial actuation. This event is reportable per 10CFR50.73(a)(2)(iv)(A) since isolation valves for the Containment Leak Detector Radiation Monitor and Drywell Hydrogen/Oxygen Analyzer automatically closed due to an invalid signal. The licensee has notified the NRC Resident Inspector. | Primary Containment Isolation System Primary containment | |
| ENS 43366 | 19 March 2007 06:20:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Specified System Actuations During Testing | This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) and 10CFR 50.73(a)(1) to report invalid automatic actuations of systems listed in paragraph (a)(2)(iv)(B), namely core spray (CS) and residual heat removal (RHR). Unit 2 commenced refueling outage activities on Saturday March 10, 2007. Relay replacements for the 4 kv safeguard bus LOCA auxiliary control time delay relays were planned for all four buses due to a relay failure on Unit 1 that was identified during the prior refueling outage testing. On Monday March 19, 2007, at 02:20 hours, during emergency diesel generator (EDG) surveillance testing a Unit 2 Division 3 LOCA signal was inadvertently initiated during the planned replacement of the D23 bus LOCA auxiliary control relay (162-117). The relay was being replaced at a pre-determined step in the test. The relay was in an energized state when removed. When the new relay was installed an unplanned actuation of LOCA load shedding and sequential loading occurred. The relay was subsequently replaced and tested successfully. D23 EDG had been secured just prior to the event. The following loads were tripped and automatically restored: 2C CS pump and D234 load center breaker. The C emergency service water (ESW) pump tripped and did not restart since the EDG was not running. The 2C RHR pump continued to run. The 2A CS loop received a partial actuation in that the Division 3 signal was initiated but the Division 1 signal was not initiated. The 2C CS pump was operating in full flow test mode; therefore, it tripped and re-started as designed and 2A CS pump did not start which was expected. The 2A CS loop automatic valve alignment is initiated by the Division 1 signal; therefore, no automatic 2A CS loop valve alignment occurred. On Wednesday March 21, 2007 at 14:08 hours during EDG surveillance testing a Unit 2 Division 2 LOCA signal was initiated during the test which started the D22 EDG and tripped the D224 load center breaker as expected. However, the load center breaker did not re-close which was not expected and other expected actions did not occur. At 14:41 hours, 33 minutes later, the remaining LOCA actions occurred when 2B RHR pump, 2B CS pump, and 2B reactor enclosure recirculation system (RERS) fan automatically started, and D224 load center breaker automatically closed due to a late actuation of the D22 bus LOCA auxiliary control relay (162-116). At 15:48 an additional unexpected relay actuation caused the 2B CS pump and D224 load center breaker to trip. The relay had been replaced earlier in the day and the ongoing testing was intended to satisfy the post maintenance test (PMT). However, the relay did not actuate at the point in the test designated as the PMT; the relay actuated unexpectedly 33 minutes later. The cause of the first event was a less than adequate technical review of a test revision that added a step to replace the bus LOCA auxiliary control relay. The affected tests have been revised to replace the relay at a point in the test when it is de-energized. The cause of the second event was an equipment failure due to an intermittent connection between the relay pin connector and the relay base. The affected relay and base have been replaced and tested successfully. All of the systems that received start signals functioned successfully. The only equipment malfunction was associated with the degraded relay. The RHR and CS starts were partial actuations. The D22 EDG train start was an expected actuation. The C ESW train was tripped but was not automatically started. This event is reportable per 10CFR50.73(a)(2)(iv)(A) since 2B RHR pump, 2B CS pump, and 2C CS pump automatically actuated on an invalid signal. Component data: Equipment name: D22 Bus LOCA Aux Control Time Delay Relay Equipment number: 162-116 Manufacturer: A348 Amerace Corp Model number: ETR14D3A002 Serial number: 83330224 The licensee notified the NRC Resident Inspector. | Service water Emergency Diesel Generator Core Spray Residual Heat Removal Reactor Enclosure Recirculation System | |
| ENS 42830 | 9 July 2006 22:49:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Actuation of the B Esw Pump | This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report an invalid automatic actuation of systems listed in paragraph (a)(2)(iv)(B), namely emergency service water (ESW). On Sunday July 9, 2006, at 18:49 hours, an invalid actuation of the B ESW pump occurred. The B ESW loop start was a partial actuation and the loop functioned successfully following the invalid actuation. The A ESW loop was not affected. An investigation identified that corrosion on the D12 emergency diesel generator (EDG) jacket water pump discharge pressure switch caused the invalid partial actuation. The cause of the pressure switch failure was due to a cracked supply tube, which allowed moisture to enter the cabinet causing the corrosion. EDG jacket water pump discharge pressure is utilized to initiate logic that starts the ESW pump when the EDG is running. ESW is the cooling medium for the EDG. The switch failure also caused inoperability of D12 EDG since it defeated the capability for the EDG to start. The failed pressure switch and associated tubing were replaced and successfully tested. D12 EDG was declared operable on Monday July 10, 2006 at 12:46 hours. An inspection of the other seven EDGs identified that the D11 EDG pressure switch was also corroded. The D11 EDG pressure switch was replaced. The inspection determined that the jacket water pressure switches on the other six EDGs were not degraded. This event is reportable per 10CFR50.73(a)(2)(iv)(A) since B ESW pump automatically actuated on an invalid signal. Component data: Equipment number: PSH-GA-110B Manufacturer: A160 Allen-Bradley Co. Model number: 636-C3 The licensee notified the NRC Resident Inspector. | Service water Emergency Diesel Generator | |
| ENS 42572 | 14 March 2006 21:40:00 | 10 CFR 50.73(a)(1), Submit an LER | Inadvertant Core Spray Pump Actuation | The notification is being made pursuant to 10CFR50.73(a)(2) (iv) (A) and is reported per 10CFR50.73(a)(1) An invalid actuation of the 1 C Core Spray Pump start during D13 LOCA/LOOP Testing. a) The specific train(s) and system(s) that were actuated. The 1C Core Spray Pump actuated automatically on an inadvertent Division 3 LOCA signal during LOCA/LOOP testing. No other ECCS train/system actuated. b) Whether each train actuation was complete or partial. The 1C Core Spray Pump actuated in the minimum flow protection mode. c) Whether or not the system started and functioned successfully. The 1 C Core Spray System started and functioned successfully but did not inject into the vessel. On March 14, 2006 at 5:40 PM during performance of the D13 LOCA/LOOP Test, ST-6-092-117-1, the 1 C Core Spray pump was inadvertently started. The pump started (but did not inject) when an I&C Technician attempted to demonstrate operation of the test switch. He inadvertently picked up the energized test switch which was next to 2 spare test switches and manipulated it which resulted in a Division 3 LOCA signal to the 1C Core Spray Pump. The 1C Core Spray Pump never injected and was running with min-flow protection and functioned as expected. The cause of the event was lack of attention to detail and self check. The licensee notified the NRC Resident Inspector. | Core Spray | |
| ENS 41667 | 6 March 2005 19:07:00 | 10 CFR 50.73(a)(1), Submit an LER | Invalid Safety System Actuation During an Outage | This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) to report an invalid automatic actuation of systems listed in paragraph (a)(2)(iv)(B), namely ECCS, emergency diesel generator (EDG), and emergency service water (ESW). On March 6, 2005 at 15:07 hours a Unit 2 Division 3 LOCA signal was inadvertently initiated during replacement of a power supply on ECCS. Continuity on the 24 VDC power distribution signal common was lost while lifting leads on the power supply. One terminal on the power supply was part of a daisy chain circuit connection. This resulted in an unplanned actuation of the high drywell pressure relays. Two 24 VDC power supplies are connected in parallel to supply power to Division 3 ECCS trip units and relays. It was planned to replace one of these power supplies while the power distribution network remained energized by the redundant power supply. However, the continuity of the power distribution network relied on the connection of two leads on one terminal on the out-of-service power supply. When the two leads were separated several relays lost their normal connection to the signal common. Reverse currents caused the false actuation of the Division 3 high drywell pressure relays. D23 EDG automatically started and ran unloaded. The C ESW pump automatically started due to the EDG start. The 2A Core Spray (CS) loop received a partial actuation in that the Division 3 signal was initiated but the Division 1 signal was not initiated. 2C CS pump started as designed and 2A CS pump did not start which was expected. The 2A CS loop automatic valve alignment is initiated by the Division 1 signal; therefore, no automatic 2A CS loop valve alignment occurred. All systems functioned as designed during the event. The cause of the event was that the work package planner for the power supply replacement failed to recognize the impact on plant systems and as a result applied less than adequate technical rigor. The maintenance planning procedure will be revised to ensure the appropriate level of technical rigor is applied for work packages that lift or manipulate a signal common lead. This event Is reportable per 10CFR50.73(a)(2)(iv)(A) since 2C Core Spray pump, D23 EDG and C ESW pump automatically actuated on an invalid signal. The NRC Resident Inspector has been notified. | Service water Emergency Diesel Generator Core Spray | |
| ENS 40895 | 26 May 2004 20:22:00 | 10 CFR 50.73(a)(1), Submit an LER | 60 Day Invalid Automatic Actuation Report: Containment Isoaltion Valves in More than One System | On May 26, 2004 at 16:22 hours and inadvertent primary containment isolation signal was initiated following restoration from a surveillance test on the 2A Reactor Enclosure ventilation exhaust duct radiation monitor. The isolation was due to a fuse failure that caused isolation signals on Group 6A, 6B, 6C, and 7A primary containment isolation valves (PCIVs). The Unit 2 reactor enclosure ventilation isolated. The 2A standby gas treatment system (SGTS) and 2A reactor enclosure recirculation system (RERS) trains initiated. The instrument gas compressor suction valve and the 2A instrument gas header supply valve closed. The containment nitrogen inerting valve and the 2A instrument gas header supply valve closed. The containment nitrogen inerting block valve closed. The containment atmospheric sample valves received an isolation signal but were in a closed position prior to the event. All system functioned as designed during the event. The laboratory analysis of the fuse determined that the most probable cause of the event was an age related degradation of the fuse that reduced its current carrying capacity. The fuse was operating under steady state conditions for approximately 30 seconds prior to the failure. A fuse replacement plan was in progress prior to the event due to prior age related failures of this type of fuse. The fuse replacement plan will be evaluated to ensure the current schedule is adequate. Component data: Type: Fuse Manufacturer: Bussmann Model number: MIN-5 The NRC Resident Inspector was notified of this event by the licensee. | Primary containment Standby Gas Treatment System Reactor Enclosure Recirculation System |