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	Discovered date	Reporting criterion	Title	Event description
ENS 56987	2 January 2024 04:33:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Containment Isolation Valves	The following information was provided by the licensee via phone and email: This 60-day optional telephone notification is being made in lieu of a Licensee Event Report (LER) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 2333 EST on January 1, 2024, an invalid actuation of group 6 primary containment isolation valves (PCIVs) (i.e., containment atmospheric control/monitoring (CAC/CAM) and post-accident sampling system (PASS) isolation valves) occurred. Reactor building ventilation isolated and standby gas treatment started per design. No manipulations associated with the isolation or reset logic were ongoing at the time. Troubleshooting determined that the group 6 isolation signal resulted from spurious relay contact actuation in the main stack radiation high-high isolation logic due to relay contact oxidation. The main stack radiation monitor is a shared component that sends isolation signals to Unit 1 and Unit 2. There were no Unit 1 actuations. Only the relay contacts associated with Unit 2 actuated. The relay has been replaced. The actuation was not initiated in response to actual plant conditions. It was not an intentional manual initiation and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. During this event the PCIVs functioned successfully, and the actuations were complete. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector had been notified.
ENS 56980	19 February 2024 15:45:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Loss of Reactor Building Ventilation	The following information was provided by the licensee via email: At 1045 EST, on 2/19/2024, during a maintenance activity, a loss of all reactor building ventilation occurred on Unit 2. With no flow past the ventilation radiation monitors, the radiation monitors were inoperable to support their ability to perform primary and secondary containment isolation functions or start the standby gas treatment system. Reactor building ventilation was restored within 15 minutes. Due to this inoperability, the radiation monitor system was in a condition that could have prevented fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector will be notified. * * RETRACTION ON 3/15/24 AT 1315 EDT FROM BILL LINNELL TO ADAM KOZIOL * * * Upon further investigation, it was verified that the reactor building and the refueling floor radiation monitors are not needed to control the release of radiation for events described in chapter 14 of the updated Final Safety Analysis Report. For the analyzed loss of coolant accident (LOCA), the primary and secondary signals for this purpose were available and unaffected by this event. The radiation monitors provide a tertiary redundant method that is not credited within the station analysis. For all other analyzed accidents, the signal provided by the radiation monitors is not needed, as the secondary containment isolation function and start of the standby gas treatment system are not credited. Additionally, the fuel handling accident was not credible during the time of the event because no activities were in progress on the refueling floor. Therefore, the threshold for reporting the issue as an event or condition that could have prevented the fulfillment of a safety function was not met. The NRC Resident Inspector has been notified. Notified R1DO (Jackson)
ENS 56811	22 October 2023 16:49:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Momentarily Inoperable	The following information was provided by the licensee via fax and phone: On October 22, 2023, at 1149 CDT, with the reactor at 100 percent core thermal power and steady state conditions, the Cooper Nuclear Station secondary containment differential pressure exceeded the Technical Specification (TS) Surveillance Requirement (SR) 3.6.4.1.1 limit of -0.25 inches water gauge. The condition existed for approximately 80 seconds until the reactor building ventilation system responded to restore differential pressure to normal. Investigations identified a hinged duct access hatch found open. The hatch was closed and latched, and ventilation system parameters were returned to normal. There were no radiological releases associated with this event. Declaring secondary containment inoperable as a result of not meeting TS SR 3.6.4.1.1 is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material and mitigate the consequences of an accident. The NRC Senior Resident Inspector has been informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At the time the licensee notified the NRC Headquarters Operations Officer, the cause of the hinged access duct being open had not been determined. This event has been added to the licensee's corrective action program.
ENS 56287	9 November 2022 14:06:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Containment Isolation Valves	The following information was provided by the licensee via email: This 60-day optional telephone notification is being made in lieu of an LER (Licensee Event Report) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0906 Eastern Time (EST) on November 9, 2022, an invalid actuation of Group 6 Primary Containment Isolation Valves (PCIVs) (i.e., Containment Atmospheric Control/Monitoring and Post Accident Sampling isolation valves) occurred. In addition, per design, Reactor Building Ventilation isolated and Standby Gas Treatment started. It was determined that this condition was caused by faulty test equipment that was being used during preparation for the Main Stack Radiation Monitor High Radiation Response Time test. This test requires connecting a recording device to monitor for the test start signal on a Unit 2 relay associated with the Main Stack High Radiation signal. The recorder faulted which caused the associated fuse to blow and resulted in Unit 2 receiving a Main Stack High Radiation signal and Group 6 PCIV actuation. It was verified that the radiation monitor was not in trip electrically (i.e., there was no high radiation condition). The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. During this event the PCIVs functioned successfully, and the actuations were complete. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified.
ENS 56280	12 November 2022 04:33:00	10 CFR 50.73(a)(1), Submit an LER	60-DAY Telephonic Notification	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 50.73(a)(2)(iv)(A) for an invalid actuation of a primary containment isolation signal affecting more than one system. On November 11, 2022, at 2333 hours EST, Peach Bottom experienced an unplanned loss of the #343 Off-Site Startup Source. Due to the temporary loss of power during automatic bus transfers, several systems experienced Primary Containment Isolation System (PCIS) Group II and Group III (GP II/III) isolation signals. Plant Systems impacted by isolation valve closure included: Reactor Water Clean Up (RWCU), Containment Atmospheric Control (CAC), Traversing In-Core Probe (TIP) Purge, Primary Containment Floor and Equipment Drains, and the Instrument Nitrogen system. All equipment responded as designed. Plant conditions which initiate PCIS GP II isolation signals are Reactor Vessel Low Water Level, High Drywell Pressure, RWCU system High Flow or RWCU Non-Regenerative Heat Exchanger High Outlet Temperature. The PCIS GP III actuations are initiated by the Reactor Vessel Low Water Level, Primary Containment High Pressure, Reactor Building Ventilation High Radiation or Refuel Floor Ventilation High Radiation. At the time of the event, none of these actual plant conditions existed; therefore, the actuation of the PCIS was invalid. The loss of the #343 Off-Site Startup Source was caused by a failed printed circuit card in the programable logic controller (PLC) for the 3435 breaker. There is no time-based maintenance strategy for PLC replacement. The PLC circuit card was replaced, and the breaker restored to full qualification and service. Preventive maintenance strategy will be enhanced to address the identified vulnerability. The licensee has notified the NRC Resident Inspector.
ENS 56138	6 August 2022 10:28:00	10 CFR 50.73(a)(1), Submit an LER	60-DAY Optional Telephonic Notification of Invalid Actuation of Containment Isolation Valves	The following information was provided by the licensee via email: This 60-day optional telephone notification is being made in lieu of an LER submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for invalid actuations of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0628 Eastern Daylight Time (EDT) on August 6, 2022, an invalid actuation of group 6 Primary Containment Isolation Valves (PCIVs) (i.e., containment atmospheric control/monitoring and post accident sampling isolation valves) occurred. The group 6 isolation signal resulted from the reactor building ventilation radiation monitor `A' channel exceeding the setpoint value. This condition recurred at approximately 1305 EDT on August 12, 2022. In both instances, the `B' channel, located in the same plenum, remained steady and below the setpoint value through the entire event. This, along with readings made by radiation protection technicians, confirmed that there were no actual high radiation conditions in the reactor building exhaust in either instance. Following each invalid actuation, upon returning unit 2 reactor building ventilation to service, the `A' channel readings returned to be consistent with the `B' channel. It was determined that these invalid actuations likely resulted from degradation of circuit components associated with the radiation monitor. The `A' channel radiation monitor was replaced on September 22, 2022. During these two events, the PCIVs functioned successfully and the actuations were complete. The actuations were not initiated in response to actual plant conditions, they were not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, these events have been determined to be invalid actuations. These events did not result in any adverse impact to the health and safety of the public.
ENS 55756	4 January 2022 18:16:00	10 CFR 50.73(a)(1), Submit an LER	60-DAY Optional Telephonic Notification for Invalid Actuation of Containment Isolation Valves	The following information was provided by the licensee via email: This 60-day optional telephone notification is being made in lieu of an LER (Licensee Event Report) submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 1316 Eastern Standard Time (EST) on January 4, 2022, during performance of isolation logic periodic testing associated with Primary Containment Isolation System Groups 2 and 6, an invalid actuation of Group 6 Primary Containment Isolation Valves (PCIVs) (i.e., Containment Atmospheric Control/Monitoring (CAC/CAM) and Post Accident Sampling (PASS) isolation valves) occurred. This resulted in a Division I CAC isolation signal, a full CAM isolation, and a full PASS isolation. Reactor Building Ventilation isolated and Standby Gas Treatment started per design. No manipulations associated with the isolation or reset logic were ongoing at the time. Troubleshooting determined that the Group 6 isolation signal resulted from a high resistance contact on a relay associated with the main stack radiation high-high isolation logic. This condition interrupted electrical continuity and prevented the Group 6 logic from resetting. Following cleaning of the relay contacts, the isolation logic remained in the reset state. The main stack radiation monitor is a shared component that sends isolation signals to Unit 1 and Unit 2. It was verified that the radiation monitor was not in trip electrically and there were no Unit 2 actuations. Therefore, the actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. As a result, this event has been determined to be an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified.
ENS 55708	29 November 2021 17:28:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Secondary Containment Relays	The following information was provided by the licensee via email: This telephone notification is provided in accordance with 10 CFR 50.73(a)(1) to report an invalid actuation of secondary containment relays in accordance with 10 CFR 50.73(a)(2)(iv)(A). On November 29, 2021, the `B' Fuel Pool radiation monitor spiked high during restoration following the performance of the 0068 procedure `Spent Fuel Pool & Reactor Building Exhaust Plenum Monitor Calibration' due to cable to radiation monitor connector degradation from handling. This resulted in a Partial Primary Containment Group II isolation (gas systems), initiation of Standby Gas Treatment system, and isolation of the Reactor Building Ventilation system. All systems responded as designed to the actuation signal. Operations reset the Partial Primary Containment Group II isolation signal, shutdown Standby Gas Treatment System, and restored Reactor Building Ventilation system per procedures. At the time of the occurrence, the `A' Fuel Pool radiation monitor was reading normal at approximately 1.5 mr/hr. The `B' Fuel Pool radiation monitor spiked above the 50 mr/hr setpoint and continued to read erratically. Work was performed to clean and reconnect the connector and testing per 0068 procedure verified the condition was corrected. The `B' Fuel Pool radiation monitor returned to service. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 55660	20 October 2021 13:05:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of a General Containment Isolation Signal Affecting More than One 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 actuation of the Reactor Protection System (RPS). On October 20, 2021, at approximately 0705 hours Central Daylight Time (CDT), Browns Ferry, Unit 1, 1B RPS bus unexpectedly lost power. The loss of the bus resulted in a half scram, automatic Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolations, and Trains A, B, and C SBGT (Stand-By Gas Treatment) and A CREV (Control Room Emergency Ventilation system) started. All systems responded as expected. At 0720 hours CDT, the bus was placed on the alternate power supply and the half scram and PCIS isolations were reset. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the RPS bus loss was a trip of the underfrequency relay due to drift of the relay setpoint. The relay was replaced and 1B RPS bus was returned to the normal power supply on October 21, 2021, at 0510 hours CDT. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1729592. The NRC Resident Inspector has been notified of this event.
ENS 55627	6 December 2021 16:25:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Safety System Actuation	On December 6, 2021, at 1125 hours Eastern Standard Time (EST), during planned maintenance activities, electrical power was lost to the 4160V emergency bus E-3. The power loss to emergency bus E-3 affected both Unit 1 and 2. Emergency Diesel Generator #3 received an automatic start signal but was under clearance for planned maintenance. Emergency bus E-3 was re-energized at 1315 EST hours via offsite power. The loss of power to E3 resulted in Unit 2 Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 3 (i.e., Reactor Water Cleanup), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., air isolation to the drywell) isolations. The actuations of PCIVs were completed and the affected equipment responded as designed. Per design, no Unit 1 safety system group isolations or actuations occurred. Other Unit 2 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start signal to the Standby Gas Treatment (SGT) System trains A and B and the Control Room Emergency Ventilation System (CREV). Systems functioned as designed. Safety systems functioned as designed following the de-energization of bus E-3. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of PCIS on Unit 2 and an automatic start signal to Emergency Diesel Generator #3. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Except for the Emergency Diesel Generator, which is out of service for planned maintenance, all equipment has been returned to its normal alignment. * * UPDATE FROM JJ STRNAD TO THOMAS KENDZIA AT 2028 EST ON DECEMBER 6, 2021 * * * The loss of power to E3 resulted in Unit 1 Primary Containment Isolation System (PCIS) Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems). Other Unit 1 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start signal to the Standby Gas Treatment (SGT) System trains A and B and the Control Room Emergency Ventilation System (CREV). Systems functioned as designed. Safety systems functioned as designed following the de-energization of bus E-3. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of PCIS on Unit 1. All Unit 1 equipment was returned to its normal alignment. The NRC Resident will be notified. Notified R2DO (Miller).
ENS 55287	1 April 2021 18:02:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Telephonic Notification of Invalid Actuation of a General Containment Isolation Signal Affecting More than One System	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of the 2A Reactor Protection System (RPS). On April 1, 2021, at 1302 (CDT), Browns Ferry Unit 2, 2A RPS (Motor Generator) MG set tripped causing a half scram. Unit 2 experienced an unexpected trip of the 2A RPS MG Set that resulted in automatic Primary Containment Isolation System (PCIS) Group 2, 3, 6, and 8 isolations and Trains A, B, and C Standby Gas Treatment (SGT) and Train A Control Room Emergency Ventilation (CREV) starts. At the time of the event, Unit 2 was in a refueling outage and the rods were already fully inserted. All systems responded as expected. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. Based on the troubleshooting conducted, the cause was determined to be a loose wiring connection in the motor circuit. The lugs were replaced with ring lugs. Operations reset the 2A RPS Half Scram and PCIS in accordance with 2-AOI-99-1 on April 1, 2021, at 1324 CDT thus correcting the condition and returning RPS to service. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1683358. The NRC Resident Inspector has been notified of this event.
ENS 55240	10 March 2021 13:15:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of an Invalid Specified System Actuation	This telephone notification is provided in accordance with 10 CFR 50.73(a)(1) to report an invalid actuation of containment isolation valves in more than one system required by 10 CFR 50.73(a)(2)(iv)(A). On March 10, 2021, at 0815 (CST), during the Unit 2 Refueling Outage (L2R18), while performing a test to verify functionality of an isolation relay following replacement of the relay, a Group 4 isolation signal was actuated. The Group 4 isolation logic affects both the Reactor Building Ventilation (VR) and Containment Vent and Purge (VQ) system (for both units). All equipment responded as designed to the Group 4 isolation, including startup of Standby Gas Treatment (SBGT) to maintain secondary containment pressure (for both units). Investigation determined that the cause of the isolation was an inadvertent contact of the self-retracting grip jumper between two adjacent terminals that caused a short to ground and a blown fuse during the test performance. The fuse was replaced and systems restored as needed for the plant condition. The containment isolation was not due to actual plant conditions or parameters meeting design criteria for containment isolation. Therefore, this is considered an invalid actuation. The NRC Resident Inspector has been informed of this notification.
ENS 55191	17 February 2021 19:07:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Containment Isolation Valves	This 60-day optional telephone notification is being made in lieu of an LER submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 1507 EDT on February 17, 2021, during performance of isolation logic periodic testing associated with Primary Containment Isolation System Groups 2 and 6, an invalid actuation of Group 6 Primary Containment Isolation Valves (PCIVs) (i.e., Containment Atmospheric Control/Monitoring and Post Accident Sampling isolation valves) occurred. The Group 6 isolation signal resulted from the reactor building ventilation radiation monitor `B' Channel exceeding the setpoint value. This condition likely resulted from the radiation monitor electronics being impacted by humidity levels, which exceeded the instrument design requirements that developed in the area over time as a result of the Unit 2 reactor building ventilation being secured per the test procedure. The `A' Channel, located in the same plenum, remained steady and below the setpoint value through the entire event. This, along with readings made by a Radiation Protection Technician, confirmed that there was no actual high radiation condition in the reactor building exhaust. Upon returning Unit 2 reactor building ventilation to service, the `B' Channel readings returned to be consistent with the `A' Channel. The PCIVs functioned successfully and the actuation was complete. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified.
ENS 54932	6 August 2020 22:49:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of an Invalid Actuation of a General Containment Isolation Signal Affecting More than One System	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On August 6, 2020, at approximately 1749 CDT, Browns Ferry Nuclear Plant (BFN), Unit 2 experienced a loss of Reactor Protection System (RPS) Bus 2A. Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolated in response to this event. The PCIS isolations caused the initiation of Standby Gas Treatment (SBGT) trains A, B, and C, and Control Room Emergency Ventilation (CREV) subsystem A. Unit 2 declared RCS leakage detection instrumentation inoperable and entered TS LCO 3.4.5 condition A, B, and D with required action D.1 to enter LCO 3.0.3 immediately. Unit 2 entered TS LCO 3.0.3 with required actions to be in Mode 2 within 10 hours, Mode 3 within 13 hours, and Mode 4 within 37 hours. Upon investigation, it was discovered that an age-related overheating condition resulted in the failure of the 2A RPS Motor Generator (MG) set, causing the feeder beaker from the 2A 480v Remote Motor-Operated Valve distribution board to trip. On August 6, 2020, at approximately 1808 CDT, Operations personnel commenced restoration of Unit 2 to normal after transferring 2A RPS to its alternate power supply. The 2A RPS MG Set drive motor was replaced on August 24, 2020. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel (RV) Low Water Level or Drywell High Pressure. Plant conditions which initiate PCIS Group 3 actuations are RV Low Water Level or Reactor Water Cleanup Area High Temperature. Plant conditions which initiate PCIS Group 6 actuations are RV Low Water Level, High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation. Plant conditions which initiate PCIS Group 8 actuations are Reactor Vessel (RV) Low Water Level or Drywell High Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. All affected safety systems responded as expected. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1628707. The NRC Resident Inspector has been notified of this event.
ENS 54697	9 March 2020 01:21:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of a General Containment Isolation Signal Affecting More than One System	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On March 8, 2020, at approximately 2021 CDT, Browns Ferry Nuclear Plant Unit 2 experienced an unexpected loss of the 2A Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and the initiation of Standby Gas Treatment Trains A and B, and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The RPS MG Set trip was believed to have been caused by an intermittent short across a spike suppressor, which led to a loss of generator output signal to a voltage regulator. The affected components have been replaced. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1593265. The NRC Resident Inspector has been notified of this event.
ENS 54341	29 December 2018 07:20:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of a Signal Affecting More than One System	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On December 29, 2018, at approximately 0220 Central Standard Time (CST), Browns Ferry Nuclear Plant (BFN), Unit 3 experienced an unexpected loss of power to the 3A Reactor Protection System (RPS) Bus due to the trip of the 3A RPS motor generator (MG) set. This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected. This event is being reported as a late 60 day non-emergency notification. This missed notification was identified on August 23, 2019. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the trip of the RPS MG Set was a failure of the motor winding insulation of all three phases. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Reports 1478564 and 1543534. The NRC Resident Inspector has been notified of this event.
ENS 54332	20 August 2019 16:33:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of Invalid Specified System Actuation	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On August 20, 2019, at approximately 1133 hours Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN), Unit 2 experienced an unexpected loss of the 2A Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the RPS MG Set trip was dirty potentiometer windings on an Over Voltage Relay. The dirt prevented the potentiometer's wiper from contacting its windings, resulting in erratic setpoint values. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Reports 1542603, 1542608, and 1542569. The NRC Resident Inspector has been notified of this event.
ENS 54300	31 July 2019 21:50:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification Due to an Invalid Actuation of a Containment Isolation Signal	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On July 31, 2019, at approximately 1650 hours Central Daylight Time (CDT), Browns Ferry Nuclear Plant (BFN), Unit 1 experienced a Primary Containment Isolation System (PCIS) Group 6 isolation during performance of surveillance procedure 1-SR-3.3.6.2.3(A), Reactor/Refueling Zone Ventilation Radiation Monitor 1-RM-90-140/142 Calibration and Functional Test. The Group 6 isolation caused the initiation of Standby Gas Treatment (SBGT) Trains A, B, and C, and Control Room Emergency Ventilation (CREV) subsystem B. Unit 1 H2O2 Analyzer and Drywell Radiation Monitor CAM, 1-RM-90-256, were declared Inoperable and Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.5 Condition B was entered. All affected safety systems responded as expected. Plant conditions which initiate PCIS Group 6 actuations are Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. This condition was the result of two cleared fuses in the alarm logic. The apparent cause is a ground fault on the A6 Open Drain Input/Output Module. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Acton Program as Condition Report 1537358. The NRC Resident Inspector has been notified of this event.
ENS 54239	25 August 2019 16:02:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Scram Due to Main Generator Ground Fault Relay	On August 25, 2019, at 1102 (CDT), Quad Cities Unit 1 experienced an automatic scram from 100 percent power. All rods fully inserted and there were no complications. The trip was initiated from a main generator ground fault relay. Troubleshooting of the fault is in progress. All systems responded as designed. There were no systems inoperable and no TS (Technical Specification) action statements were in progress prior to the Reactor Scram. Reactor water level dropped below the Group 2 and Group 3 Reactor Water Level Isolation set-points as expected, and recovered via the Feedwater system. Standby Gas Treatment System auto started and Reactor Building Ventilation Isolation occurred as expected. Unit 1 remains in Mode 3. Decay heat is being removed using the steam bypass valves to the condenser and the safety relief valves did not lift as a result of the trip. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. Unit 2 was not affected.
ENS 54121	30 April 2019 11:50:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification Due to Invalid Actuation of a General Containment Isolation Signal	This 60-day telephone notification is being made in accordance with the reporting requirements specified by 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of a general containment isolation signal affecting multiple systems. On April 30, 2019, at approximately 0650 CDT, a level 2 containment isolation signal was introduced when a fuse for the Nuclear Steam Supply Shutoff System was removed for a maintenance clearance. The level 2 containment isolation signal caused a trip of the Division I DC bus back-up charger, leaving only the Division I battery to carry the DC bus. At 0707 CDT the bus was de-energized when another unrelated clearance opened the battery supply breaker to the DC bus causing another containment isolation signal. This event did not affect Shutdown Cooling or any other protected Safety Related Equipment. The containment isolation signals caused an isolation of the systems listed below. All components that were not removed from service, gagged in position, already in the expected position due to plant conditions, or de-energized due to plant condition performed as designed. Containment Isolation valves for the following systems isolated as expected: Drywell and Containment Floor Drains, Drywell and Containment Equipment Drains, Condensate Makeup, Fire Protection Water, Service Air, Instrument Air, Reactor Water Cleanup, Spent Fuel Cooling and Cleanup, Reactor Plant Component Cooling Water, Chilled Water, Reactor Recirculation, Main Steam Drains, Reactor Building Ventilation, and Fuel Building Ventilation. The licensee notified the NRC Resident Inspector.
ENS 53828	16 January 2019 05:00:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Differential Pressure Exceeded Technical Specification Allowed Value	On January 16, 2019, with James A. Fitzpatrick Nuclear Power Plant operating at 100 percent power, the Emergency and Plant Information Computer (EPIC) indicated that Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge while isolating Reactor Building Ventilation. The Secondary Containment differential pressure was less than 0.25 inches of vacuum water gauge for approximately ten (10) seconds, and then immediately returned to greater than or equal to 0.25 inches of vacuum water gauge. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the NRC Resident Inspector.
ENS 53778	5 December 2018 05:00:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Three Minute Loss of Secondary Containment Vacuum	At 1010 (EST) on December 5, 2018, Secondary Containment differential pressure exceeded the Technical Specification Surveillance Requirement of greater than or equal to 0.25 inches of vacuum water gauge. This condition existed for approximately 3 minutes before the differential pressure was restored to normal when the Standby Gas Treatment system was manually initiated. This event was caused by a trip of the service air compressor 39AC-2A. The loss of instrument air pressure caused Reactor Building ventilation to isolate and raise Secondary Containment differential pressure. The instrument air pressure was restored when 39AC-2A was isolated and the two backup air compressors started. This condition did not impact the leak tightness of Secondary Containment or the ability of the Standby Gas Treatment system to establish and maintain the required differential pressure. When Secondary Containment did not meet the Technical Specification Surveillance Requirement 3.6.4.1.1 for differential pressure, the Limiting Condition of Operation (LCO) was not met. Therefore, Secondary Containment was inoperable. This event is being reported under 10 CFR 50.72(b)(3)(v)(C). The licensee notified the NRC Resident Inspector.
ENS 53661	16 August 2018 05:00:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of a General Containment Isolation Signal Affecting More than One System	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On August 16, 2018, at approximately 1736 CDT, Browns Ferry Nuclear Plant (BFN), Unit 2 experienced an unexpected loss of the 2B Reactor Protection System (RPS). This resulted in Primary Containment Isolation System (PCIS) groups 2, 3, 6, and 8 isolations, and initiation of Standby Gas Treatment Trains A, B, and C and Control Room Emergency Ventilation System Train A. All affected safety systems responded as expected with the exception of the Unit 1 Refuel Zone Supply Fan Outboard Isolation Damper, 1-FCO-64-5, that failed to indicate closed position. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The cause of the RPS MG (Motor Generator) Set trip was a failed (shorted) operating coil associated with the 480 VAC motor starter inside the control box. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Reports 1440047 and 1440050. The NRC Resident Inspector has been notified of this event.
ENS 53189	31 January 2018 19:10:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Momentary Loss of Secondary Containment	At 1310 hours (CST) on January 31, 2018, the Unit 2B fuel pool radiation monitor spiked high due to an invalid actuation which caused the U1 and U2 reactor building ventilation system to isolate, B train standby gas treatment system (SBGTS) started, and the control room ventilation system also isolated as designed. Secondary containment vacuum was lost for approximately one minute, and then subsequently returned to an acceptable level in accordance with Technical Specification 3.6.4.1, 'Secondary Containment.' As a result of this transient, secondary containment was inoperable for approximately one minute. No emergency conditions were determined to exist. Troubleshooting of the radiation monitor spike is underway. Given the temporary loss of secondary containment vacuum, this event is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified.
ENS 53070	10 January 2017 09:00:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of a General Containment Isolation Signal Affecting More than One System	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On September 15, 2017, during a TVA (Tennessee Valley Authority) review of Operations logs, it was determined that a reportable condition occurred in January 2017 but no NRC report had been made. On January 10, 2017, at 0300 Central Standard Time (CST), Browns Ferry Nuclear Plant, Unit 3, received Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals. The Group 2, 3, 6, and 8 isolations caused the initiation of all three trains of the Standby Gas Treatment (SBGT) system and Control Room Emergency Ventilation (CREV) subsystem 'A.' At 0311 CST, Operations personnel discovered that the 3A1 RPS circuit protector had tripped on undervoltage. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywall Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywall Pressure. At the time of the event, these conditions did not exist; therefore the actuation of the PCIS was invalid. All affected equipment responded as designed. This condition was the result of an undervoltage condition on the 3A1 circuit protector. During trouble shooting, the undervoltage setpoints were found to be 116 VAC and 115 VAC, when the normal as left acceptance band is 109.7 VAC to 111.3 VAC. The 3A RPS protective relays had been previously replaced in September 2016. The most likely cause of the undervoltage condition in these relays is infant mortality. The NRC Resident Inspector has been notified of this event.
ENS 52999	3 October 2017 15:00:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable Due to Unexpected Isolation of Exhaust Valve	On October 3, 2017, at 0800 PDT, Reactor Building (Secondary Containment) pressure momentarily rose above the Technical Specification (TS) limit. Secondary Containment was declared inoperable and TS Action Statement 3.6.4.1.A was entered. The pressure rise was due to unexpected isolation of an exhaust valve in the Reactor Building ventilation system during electrical switchgear inspections. The cause of the closure is still under investigation. The Control Room operators reopened the Reactor Building exhaust valve and pressure returned to within limits automatically. Secondary Containment was declared operable at 0802 PDT and TS Action Statement 3.6.4.1.A was exited. This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and accident mitigation. The NRC Resident Inspector has been notified.
ENS 52974	17 September 2017 13:38:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Emergency Diesel Generator and Primary Containment Isolation System Actuations	On September 17, 2017, during planned surveillance activities involving Emergency Diesel Generator (EDG) 4, unexpected voltage and frequency indications were noted when EDG 4 was synchronized to Emergency Bus E4. With EDG 4 in manual mode, the Operator responded by lowering load to reopen the EDG 4 output breaker. Opening of the EDG 4 output breaker with the breakers from Balance of Plant (BOP) Bus 2C, which normally feeds the Emergency Bus E4, opened; resulted in de-energizing Emergency Bus E4. The EDG 4 voltage regulator and governor automatically reverted to auto control, and EDG 4 reconnected to Emergency Bus E4. Normal frequency and voltage were restored with EDG 4 in auto control. The momentary power interruption to Emergency Bus E4 resulted in Unit 2 Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., air isolation to the drywell) isolations. The actuations of Primary Containment Isolation Valves (PCIVS) were completed and the affected equipment responded as designed. Per design, no Unit 1 safety system group isolations or actuations occurred. These actuations are being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Additional Unit 2 actuations included PCIS Group 3 (i.e., Reactor Water Cleanup), Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start of Standby Gas Treatment (SGT) System subsystems A and B. These systems functioned as designed. This event did not impact public health and safety. The NRC Resident Inspector has been notified. The safety significance of this event is minimal. Safety systems functioned as designed following the power perturbation on E4. Plant systems responded as designed. The cause of the event is under investigation.
ENS 52966	12 September 2017 19:28:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Pressure Momentarily Above Technical Specification Limit	On September 12, 2017, at 1228 PDT, Reactor Building (Secondary Containment) pressure momentarily rose above the Technical Specification (TS) limit. Secondary Containment was declared inoperable and TS Action Statement 3.6.4.1.A was entered. The pressure rise was due to unexpected isolation of the supply and exhaust valves in the Reactor Building ventilation system due to an electrical transient on the power panel feeding the valve operators' solenoid pilot valves during maintenance. The cause of the electrical transient is under investigation. The Reactor Building differential pressure controller restored the building pressure to within limits. The Control Room operators reopened the Reactor Building ventilation supply and exhaust valves. Secondary Containment was declared operable at 1228 PDT and TS Action Statement 3.6.4.1.A was exited. This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and accident mitigation. The licensee notified the NRC Resident Inspector.
ENS 52888	4 August 2017 19:11:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable Due to Opening in Service Water Piping	On August 4, 2017, at 1511 EDT, Unit 1 Secondary Containment was declared inoperable due to a small (i.e., approximately 0.75 inch diameter) hole in Service Water system piping which was found during ultrasonic testing activities. The affected portion of piping penetrates Secondary Containment and flow in the piping creates a vacuum condition; thus bypassing Secondary Containment. The identified hole is being evaluated with respect to its impact on operability of the Service Water system. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(v)(C), as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. This event did not result in any adverse impact to the health and safety of the public. Initial Safety Significance Evaluation: The initial safety significance of this event is minimal. At the time of discovery, Unit 1 was at 100% steady state conditions. Reactor Building Ventilation was in service in a normal alignment. No abnormal radioactivity conditions existed within Secondary Containment. Corrective Actions: Temporary repair of the affected Unit 1 Service Water piping has been completed. This repair was evaluated by Engineering and it has been determined that the repair meets the requirements to maintain Secondary Containment operable. Unit 1 Secondary Containment operability was restored at 1704 EDT on August 4, 2017. The licensee notified the NRC Resident Inspector. * * RETRACTION FROM MIKE BRADEN TO RICHARD SMITH AT 1447 EDT ON 9/27/17 * * * Based upon further evaluation, Duke Energy is retracting Event Notification 52888. The safety objective of Secondary Containment is to limit the release of radioactivity to the environment after an accident so that the resulting exposures are kept to a practical minimum and are within regulatory limits. A bounding engineering evaluation was performed which demonstrates that potential releases from Secondary Containment could not have resulted in offsite or control room doses exceeding regulatory limits. Furthermore, the condition did not impact Technical Specification operability of Secondary Containment in that the ability of Secondary Containment to maintain the required vacuum was not impacted. Therefore, this condition does not represent an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and is not reportable in accordance with 10 CFR 50.72(b)(3)(v)(C), and the event notification is being retracted. The NRC Senior Resident was notified of this retraction. Notified R2DO (A. Masters).
ENS 52795	8 June 2017 19:27:00	10 CFR 50.72(b)(2)(xi), Notification to Government Agency or News Release 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 After Main Turbine Control Logic Loss of Power	At 1527 hrs (EDT) on June 8, 2017, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a loss of Main Turbine Electro-Hydraulic Control (EHC) logic power causing a High Flux Reactor Power RPS (Reactor Protection System) trip. All control rods (fully) inserted and both reactor recirculation pumps tripped due to reaching reactor water level 2. Reactor water level lowered to -49 inches causing Level 3 (+13 inches) and Level 2 (-38 inches) isolations. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) automatically initiated and were overridden by control room operators after RPV (Reactor Pressure Vessel) water level was restored to the normal band with feedwater. HPCI and RCIC injected to the Reactor Coolant System during reactor level stabilization. All isolations and initiations occurred as expected. No main steam relief valves opened. Pressure was controlled via main turbine bypass valve operation. All safety systems operated as expected. Secondary Containment Zone 1, 2, and 3 differential pressure lowered to 0 inch WG (Water Gauge) due to a trip of the Reactor Building Ventilation system that resulted from Unit 1 Level 2 isolation. Differential pressure was restored to Zones 1, 2, and 3 by the initiation of Standby Gas Treatment System on the Unit 1 Level 2 initiation. Unit 1 reactor is currently stable in Mode 3. Investigation into the loss of Main Turbine EHC logic power is underway. The NRC Resident Inspector has been notified. A voluntary notification to PEMA and press release will occur. The suspected cause of the loss of power to the EHC logic circuit is ongoing maintenance on the system.
ENS 52382	20 November 2016 22:02:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Differential Pressure Less than Technical Specification Requirement	On November 20, 2016 at 1402 PST, Reactor Building Exhaust Air Fan 1B, REA-FN-1B, failed to start in manual which caused the Technical Specification (TS) for secondary containment pressure boundary to not be met. The duration of the time that the secondary containment TS was not met was approximately less than one minute. REA-FN-1B was being started in manual during a shift of Reactor Building Ventilation to support a post-maintenance support task on REA-FN-1B. Secondary containment differential pressure was restored within the TS requirement of greater than or equal to 0.25 inch of vacuum water gauge by restarting Reactor Building HVAC Train A. The cause of REA-FN-1B failing to start is currently under investigation. This condition is being reported under 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72(b)(3)(v)(D) for an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and accident mitigation. The licensee has notified the NRC Resident Inspector.
ENS 52355	8 November 2016 11:10:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Reactor Building to Atmosphere Differential Pressure Out of Specifications	On 11/8/16, Operators were performing Division I Undervoltage Testing Surveillance on Unit 3, when a 2 (psi) drywell signal was inserted, Reactor Building Ventilation tripped and SBGT initiated as expected. At 0510 (CST), reactor building to atmosphere differential pressure dropped below the (negative) 0.25 inches water. This condition represents a failure to meet Surveillance Requirement 3.6.4.1.1. As a result, entry into Technical Specification 3.6.4.1 condition A was made due to Secondary Containment becoming inoperable. This event is being reported in accordance with 10CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the fulfillment of a safety function. At 0532, the 2/3 Reactor Building material interlock inner door was closed and Reactor Building (differential pressure) was restored to greater than (negative).25 inches of water column. An issue report has been initiated. An investigation will be conducted and a 60 day Licensee Event report will be submitted in accordance with 10 CFR 50.73(a)(2)(v)(C). The NRC Resident Inspector has been notified.
ENS 52042	24 June 2016 16:15:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram Due to Reactor Recirculation Pumps Degradation	At 1215 (EDT) on 6/24/2016, James A. FitzPatrick (JAF) was at 100% power when Breaker 710340 tripped and power was lost to L-gears L13, L23, L33, and L43. These provide non-vital power to Reactor Building Ventilation (RBV), portions of Reactor Building Closed Loop Cooling (RBCLC), and 'A' Recirculation pump lube oil systems. Off-site AC power remains available to vital systems and Emergency Diesel Generators (EDG) are available. Due to the loss of RBV, Secondary Containment differential pressure increased. At 1215 (EDT), Secondary Containment differential pressure exceeded the Technical Specifications (TS) Surveillance Requirement SR-3.6.4.1.1 of greater than or equal to 0.25 inches of vacuum water gauge. The Standby Gas Treatment (SBGT) system was manually initiated and Secondary Containment differential pressure was restored by 1219 (EDT). The 'A' Recirculation pump tripped at 1215 (EDT) and reactor power decreased to approximately 50%. 'B' Recirculation pump temperature began to rise due to the degraded RBCLC system. At 1236 (EDT), a manual scram was initiated. Reactor Pressure Vessel (RPV) water level shrink during the scram resulted in a successful Group 2 isolation. All control rods have been inserted. The RPV water level is being maintained with the Feedwater System and pressure is being maintained by main steam line bypass valves. A cooldown is in progress and JAF will proceed to cold shutdown (Mode 4). Due to complete loss of RBCLC system, the Spent Fuel Pool (SFP) cooling capability is degraded but the Decay Heat Removal system remains available. SFP temperature is slowly rising and it is being monitored. The time (duration) to 200 degrees is approximately 117 hours. The initiation of reactor protection systems (RPS) due to the manual scram at critical power is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The general containment Group 2 isolations are reportable per 10 CFR 50.72(b)(3)(iv)(A). In addition, the temporary differential pressure change in Secondary Containment is reportable per 10 CFR 50.72(b)(3)(v)(C), as an event that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector and the State of New York.
ENS 51725	11 February 2016 04:07:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable	This report is being made pursuant to 10CFR50.72(b)(3)(v)(C), event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material and 10CFR50.72(b)(3)(v)(D), event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. LaSalle Station's Unit 1 and Unit 2 were in Mode 1. At 2207 (CST) (on 2/10/16), Secondary Containment Differential Pressure dropped below the Technical Specification (TS) 3.6.4.1 minimum of 0.25 inches water vacuum. The initial indications are a failure of one Unit 1 Reactor Building Exhaust Isolation Damper, which resulted in a trip of the Unit 1 Reactor Building Exhaust Fans. At 2245, Secondary Containment Differential Pressure was restored to within the TS 3.6.4.1 limits by securing and isolating the Unit 1 Reactor Building Ventilation System. Troubleshooting plans are being developed to determine cause of the damper failure and to correct the deficient condition. The licensee has notified the NRC Resident Inspector.
ENS 51721	10 February 2016 03:42:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable	At 2142 (CST) on February 9, 2016, Reactor Building differential pressure did not meet the required 0.25 inches of vacuum water gauge due to failure of the control system. At 2205, the Unit 3 Reactor Building Ventilation System was secured and manually isolated. The Reactor Building differential pressure returned to (greater than or equal to) 0.25 inches of vacuum water gauge at 2207. This condition represents a failure to meet Surveillance Requirement 3.6.4.1.1. As a result, entry into Technical Specifications 3.6.4.1 condition A was made due to Secondary Containment being inoperable. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified.
ENS 51579	2 December 2015 01:36:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Inoperable Due to Secondary Containment Vacuum Being Less than Required Ts Value	On December 1, 2015 at 2036 EST, with James A. FitzPatrick Nuclear Power Plant (JAF) operating at 100 percent power, Secondary Containment differential pressure exceeded the Technical Specification (TS) Surveillance Requirement (SR) of greater than or equal to 0.25 inches of vacuum water gauge for approximately one (1) minute and twenty (20) seconds. Secondary Containment (SC) had been declared inoperable prior to this event, to facilitate a planned evolution related to a previous failure that occurred on September 18, 2015 (reference EN #51409). Operators attempted to restore the Reactor Building Ventilation System (RBVS) to the normal system lineup upon completion of the planned evolution. The Secondary Containment differential pressure trended positive, and exceeded the TS SR differential pressure requirement during this transition. Preliminary investigations indicate that the cause of this event is associated with the Above Refuel Floor Exhaust Fan (66FN-13B). The design of the Above Refuel Floor Exhaust portion of the RBVS includes an interlock between the exhaust fan and a downstream damper position switch, which starts the fan when the damper is in the full open position. During the approximate one (1) minute and twenty (20) second duration that the TS SR was not met, 66FN-13B was not running with the associated discharge damper in the open position. Secondary Containment was operable after the SC differential pressure was restored upon start of 66FN-13B, and remains operable. This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(c), as an event or condition that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector.
ENS 51480	23 August 2015 16:42:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of Invalid Primary Containment Isolation Signal	On 8/23/2015 at 1242 (EDT), with the reactor at 100% power, an invalid RPS MG (Reactor Protection System Motor-Generator) set 'A' trip resulting in a loss of RPS bus 'A'; this occurred during testing of the RPS instrument channels. All equipment operated as designed as a result of the loss of power to the 'A' RPS bus. The invalid trip was determined to be a result of the overvoltage relay being set too low. The above event meets the reporting criteria of 10CFR50.73(a)(2)(iv)(A) since the loss of RPS bus resulted in primary containment isolation signals affecting containment valves in more than one system. The following systems isolated as a result of the loss of 'A' RPS bus: Reactor Water Cleanup, Reactor Building ventilation, 'A' Containment Atmosphere Dilution, Torus Vent and Purge, Drywell Equipment and Floor Drain Sumps, 'A' Drywell Containment Atmospheric Monitors, Recirculation System Sample Line, Main Steam Line Drains and Residual Heat Removal drain valve to radwaste. 'A' Standby Gas Treatment System started as designed. This notification is being made in accordance with 10CFR50.73(a)(2)(iv)(A) to provide information pertaining to an invalid 'A' Reactor Protection System actuation. Completed actions were the replacement of overvoltage relay and voltage setpoint change, completed on 9/11/2015. In accordance with 10CFR50.73(a)(i) a telephone notification is being made instead of submitting a written Licensee Event Report.
ENS 51274	28 July 2015 17:58:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Discovered One Inch Diameter Hole Between Reactor Building and Auxiliary Building	At 1358 (EDT) on July 28, 2015, a 1 inch diameter hole was discovered in the secondary containment wall, between the Reactor Building and the Auxiliary Building, causing the Secondary Containment to become inoperable under Technical Specification 3.6.5.1. Reactor Building pressure was maintained at a negative pressure as required by Technical Specification 3.6.5.1 with the Reactor Building ventilation system in service before and after discovery of the hole. In addition, the Filtration, Recirculation and Ventilation system remained fully operable and remained in standby. The hole was sealed at 1600 and technical specification 3.6.5.1 was exited. Plant operation was not impacted by the event and was operating at 100% power. No personnel injuries resulted from this event. The hole was discovered by plant personnel that were walking past the wall. Due to the discovery of the hole, the plant is performing an extent of condition inspection. The licensee notified the NRC Resident Inspector and the Lower Alloways Creek Dispatch. * * * RETRACTION FROM MARIAZ DAVIS TO STEVEN VITTO AT 1232 EDT ON 08/ 12/15 * * * * This event is being retracted. Hope Creek Generating Station Unit 1, is retracting the 8-hour non-emergency notification (EN# 51274) made on July 28, 2015, at 1855 EDT. The notification on July 28, 2015, reported that secondary containment was declared inoperable when a 1 inch hole was discovered in the secondary containment wall, between the Reactor Building and the Auxiliary Building. Secondary containment was declared inoperable based on the initial interpretation of the definition of secondary containment. The hole did not impact the ability to maintain the Tech Spec required negative pressure. Subsequent evaluation determined that secondary containment was always operable. Based on the engineering evaluation, the condition reported in EN# 51274 did not result in an inoperability of the secondary containment. Therefore, there is no reportable condition and this event report is being retracted. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. The licensee also notified the Lower Alloways Creek Dispatch. Notified R1DO (Powell).
ENS 51228	14 July 2015 06:39:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Pressure Increase Above Technical Specification Limit	Reactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0.25 inches vacuum water gauge for approximately 2 minutes during a planned surveillance test due to a subsequent failure of REA-FN-1A (Exhaust Fan) to manually start during restoration from the surveillance test. This event is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. Prior to taking test data the surveillance test directs declaring Secondary Containment inoperable in anticipation of potentially exceeding 0.25 inches vacuum water gauge reactor building pressure during the conduct of the surveillance. Consequently Technical Specification LCO 3.6.4.1.A was entered with a 4 hour completion time to restore Secondary Containment to an operable state. Upon failure of REA-FN-1A to start immediate actions were taken to close reactor building ventilation dampers and secure ROA-FN-1A (Supply Fan). Following closure of ventilation dampers and stopping ROA-FN-1A reactor building pressure was quickly restored to less than 0.25 inches vacuum water gauge with Standby Gas Treatment that was already in operation as part of the surveillance test. There were no radiological releases associated with the event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector. Maximum Secondary Containment pressure noted was 0.1 inches positive water gage. * * RETRACTION AT 1351 EDT ON 8/25/2015 FROM MATT HUMMER TO MARK ABRAMOVITZ * * * Subsequent to the initial report, Columbia has since determined that per NUREG-1022 3.2.7 the event was not reportable as Secondary Containment was 'declared inoperable as a part of a planned evolution ... in accordance with an approved procedure and (Columbia's) TS (Technical Specifications).' No condition has been discovered that would have resulted in the system being declared inoperable prior to the surveillance. Therefore, this event is not considered to be a condition that could have prevented fulfillment of a safety function or a condition prohibited by TS and is not reportable to the NRC as a Licensee Event Report (LEA) per 10 CFR 50.73. The NRC Senior Resident Inspector will be notified. Notified the R4DO (Campbell).
ENS 51085	5 April 2015 09:35:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification for an Invalid Primary Containment Isolation System Actuation	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of general containment isolation signals affecting containment isolation valves in more than one system. On April 5, 2015 at 0435 CDT, during replacement of a failed fuse (2-FU1-64-16A-K33A), Unit 2 Primary Containment Isolation System (PCIS) logic received the B half of the Unit 2 Group 6 isolation signal. This caused initiation of the B and C Standby Gas Treatment, B Control Room Emergency Ventilation, isolation of the Unit 2 reactor zone and all three refueling zone ventilations. This was not a valid initiation of PCIS. Operations personal responded to the PCIS initiation, ensured that all equipment operated as designed, and returned the affected systems back to service. Plant conditions which initiate PCIS Group 6 actuations are Reactor Vessel Low Water (Level 3), High Drywell Pressure, and Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as PER 1010651. The NRC Resident Inspector was notified of this event.
ENS 50831	19 February 2015 08:04:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Building Declared Inoperable Due to Ventilation System Trip	At 0304 EST on February 19, 2015, Fermi 2 experienced a trip of the Reactor Building Ventilation (RB) (HVAC) during plant operations associated with very cold temperatures outside. At the time of the trip, outside air temperature was -1 degrees Fahrenheit and RB HVAC tripped due to a Freeze-Stat actuation (a freeze protection feature). The plant Technical Specifications require that Secondary Containment pressure be maintained greater than or equal to -0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1). This specification was not maintained and the highest pressure observed was -0.11 inches of vacuum water gauge. Subsequently, at 0450, during restoration activities, RB pressure degraded again to higher than -0.125 inches of vacuum water gauge for 38 seconds. The lowest observed pressure was -0.11 inches of vacuum water gauge. RB HVAC has been restored by resetting the Freeze-Stat and the Standby Gas Treatment System (SGTS) has been placed back in a standby condition. The technical specification requirement is to maintain secondary containment at -0.125 inches of vacuum water gauge for secondary containment operability. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector. * * RETRACTION FROM WARREN PAUL TO DANIEL MILLS AT 1035 ON 4/8/2015 * * * After reviewing the events that occurred on February 19, 2015 against the accident analyses in Chapter 15 of the UFSAR and design functions of the Standby Gas Treatment System and Secondary Containment structure, it is concluded that a condition that could have prevented the fulfillment of a safety function to control the release of radioactive material did not occur as a result of momentarily exceeding the Technical Specification for Secondary Containment vacuum after a loss of the normal Reactor Building Ventilation System. The Fermi 2 accident analysis for a LOCA does not assume that secondary containment is under vacuum throughout the duration of an accident and contains conservative leakage assumptions to bound the effects of a postulated ground level release. The accident analysis credits the operation of the Standby Gas Treatment System (SGTS); both divisions of SGTS were operable at the time of the event. Although secondary containment was declared inoperable due to exceeding the Technical Specification value for secondary containment vacuum, the structural integrity of the secondary containment was not degraded at the time. Upon receipt of an accident signal, SGTS would have automatically started and restored secondary containment vacuum to within the bounding analyses of Chapter 15 of the UFSAR. Secondary containment was capable of performing its design function of minimizing any ground level release of radioactive material by maintaining boundary integrity so that the SGTS may draw a vacuum in the Reactor Building and filter radioactive material at all times. The event reported in EN # 50831 did not result in a condition that could have prevented the fulfillment of a safety function to control the release of radioactive material. This event report is being retracted. The licensee informed the NRC Resident Inspector. Notified R3DO (Skokowski). * * UPDATE FROM WARREN PAUL TO CHARLES TEAL ON 4/15/15 AT 1348 EDT * * * Upon further review of NUREG-1022 section 3.2.7, the original Non-Emergency Event Notification, 50831, remains valid. The NRC Resident Inspect has been informed. Notified R3DO (McCraw).
ENS 50754	26 November 2014 20:27:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of a General Containment Isolation Signal Affecting More than One System	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of a general containment isolation signal affecting more than one system. On November 26, 2014, at approximately 1427 hours Central Standard Time (CST), the Browns Ferry Nuclear Plant (BFN), 1A Reactor Protection System (RPS) Motor-Generator (MG) Set Power Supply unexpectedly de-energized resulting in a BFN Unit 1 half scram and Primary Containment Isolation System (PCIS) Groups 1, 2, 3, 6, and 8 isolation signals. The PCIS Groups 1, 2, 3, 6, and 8 isolations caused the initiation of all three trains of the Standby Gas Treatment (SBGT) system and Control Room Emergency Ventilation (CREV) subsystem 'A', and isolations of the BFN, Unit 1, Reactor Zone ventilation and BFN, Units 1 and 2, Refuel Zone ventilation (Unit 3 Refuel Zone ventilation was tagged out under 3-TO-2014-0001 at the time of this event). Operations personnel responded to the PCIS initiation, ensured all equipment operated as designed, placed the BFN 1A RPS on alternate power, and reset the RPS logic and PCIS isolations. Plant conditions which initiate PCIS Group 1 actuations are Reactor Pressure Vessel (RPV) Low Low Low Water Level (Level 1), Main Steam Line (MSL) High Flow, MSL Area High Temperature, or MSL Low Pressure. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. The apparent cause for this condition was an intermittent problem with the BFN 1A RPS MG Set voltage adjust potentiometer. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Problem Evaluation Report 961518. The NRC Resident Inspector has been notified of this event.
ENS 50730	12 January 2015 11:00:00	10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness	Planned Maintenance on Radiological Monitor	At 0500 (CST) today (January 12, 2015), planned maintenance activities commenced on the Reactor Building Separate Particulate Iodine and Noble Gas (SPING) monitor. Work on the SPING affects classification and assessment of the Emergency Plan for a radiological release through the Reactor Building Vents. In the case of a radiological event, the Reactor Building Ventilation will be isolated via Area Radiation Monitors (ARM) in the ducting and the Main Chimney SPING will be used for classification and assessment. This would be the normal response for a radiological event. This maintenance activity is expected to be longer than 72 hours. The Reactor Building SPING is scheduled to be restored on Friday, January 16. This event is reportable per 10CFR50.72(b)(3)(xiii) since the maintenance activity affects classification and assessment. The NRC Resident Inspector has been notified.
ENS 50658	7 October 2014 15:35:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Specified System Actuation	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of general containment isolation signals affecting containment isolation valves in more than one system. On October 7, 2014, at 2135 (CDT), while in a refueling outage with the reactor non-critical (Mode 5), work activities were in progress that included replacement of an excess flow check valve and execution of a Technical Specification Surveillance Procedure on the Automatic Depressurization System. Subsequent to valving in a level transmitter (LT), water levels in both the variable and reference legs of the LT were disturbed resulting in a Unit 1 full scram and Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals due to receipt of an invalid low reactor water level signal. The PCIS Groups 2, 3, 6, and 8 isolations caused the initiation of Trains A, B, and C of the Standby Gas Treatment System and Control Room Emergency Ventilation Subsystem 'A'. The Reactor and Refuel Zone ventilation fans tripped and the secondary containment dampers isolated. Operations personnel responded to the PCIS initiation, ensured all equipment operated as designed, and placed affected systems back in service. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Problem Evaluation Report 943038. The NRC Resident Inspector has been notified of this event.
ENS 50579	28 October 2014 21:08:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Reactor Building Vacuum Below Technical Specification Limit	On the evening of October 28, 2014 at 1708 EDT, with James A. FitzPatrick (JAF) Nuclear Power Plant operating at 100 percent power, the Reactor Building differential pressure decreased below the JAF Technical Specification (TS) Surveillance Requirement (SR) value of at least 0.25 inches water vacuum for a period of thirty-four (34) seconds. This occurred during restoration of the Reactor Building Ventilation System (RBVS) following planned maintenance. The Reactor Building differential pressure was 0.50 inches water vacuum with the 'A' RBVS fans in-service in conjunction with the Standby Gas Treatment System (SGTS). The Reactor Building differential pressure decreased to 0.19 inches water vacuum when the SGTS was secured. The Reactor Building Vent was subsequently isolated, and the alternate 'B' RBVS fans were placed in-service; the differential pressure increased to within the required 0.25 inches water vacuum value. The JAF TS bases associated with Secondary Containment state that, 'for Secondary Containment to be considered OPERABLE, it must have adequate leak tightness to ensure that the required vacuum can be established and maintained.' Troubleshooting activities indicated that the transient was due to a non-safety related, non-TS damper downstream of one of the 'A' RBVS fans that did not fully stroke open. The subject damper is not part of Secondary Containment, and has no safety related function. This condition did not impact the leak tightness of Secondary Containment or the ability of the associated equipment to establish and maintain the required differential pressure. Secondary Containment would have fulfilled its safety function. However, because the JAF TS SR value of 0.25 inches water vacuum was not met, Secondary Containment was considered Technical Specification INOPERABLE for a period of thirty-four (34) seconds. The Secondary Containment is considered a single-train system; therefore, this condition is reportable pursuant to 10 CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector.
ENS 50565	27 August 2014 16:09:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of General Containment Isolation Signals	This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of general containment isolation signals affecting containment isolation valves in more than one system. On August 27, 2014, at 1109 hours Central Daylight Savings Time (CDT), while in a forced unit outage with the reactor noncritical (Mode 3) and with all control rods fully inserted, instrument mechanics were attempting to backfill reactor water level transmitter (LT) 3-53 sensing lines following performance of LT replacement. During this effort, water levels in both the variable and reference legs of the LT were disturbed resulting in a Browns Ferry Nuclear Plant (BFN) Unit 1 full scram and Primary Containment Isolation System (PCIS) Groups 2, 3, 6, and 8 isolation signals due to receipt of an invalid low reactor water level signal. The PCIS Groups 2, 3, 6, and 8 isolations caused the initiation of Trains B and C of the Standby Gas Treatment (SBGT) System and Control Room Emergency Ventilation (CREV) Subsystem 'A'. The Reactor and Refuel Zone ventilation fans tripped and the secondary containment dampers isolated. Train A of the SBGT System was tagged out of service during the event. Operations personnel responded to the PCIS initiation, ensured all equipment operated as designed, and placed affected systems back in service. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Problem Evaluation Report 928777. The NRC Resident Inspector has been notified of this event.
ENS 50427	5 September 2014 00:05:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	Control Room Emergency Ventilation (Crev) System Inoperable	On September 04, 2014, at 1905 hours (CDT), the Control Room Emergency Ventilation (CREV) system was declared inoperable due to the Air Handling Unit (AHU) tripping upon restoration of Control Room Ventilation following testing of Reactor Building Ventilation instrumentation. Troubleshooting is in progress at this time. Technical Specification 3.7.4, Condition A, was entered which requires the CREV system to be restored to an operable status in seven (7) days. Additionally, Technical Specification 3.7.5, Condition A, was entered which requires CREV AC to be restored to an operable status in 30 days. This notification is being made in accordance with 10CFR50.72(b)(3)(v)(D), '(any) event or condition that could have prevented fulfillment of a safety function,' because the CREV system is a single train system required to mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector. * * RETRACTION PROVIDED BY MARK BRIDGES TO JOHN SHOEMAKER AT 1721 EDT ON 10/23/2014 * * * The purpose of this notification is to retract the ENS notification made on September 4, 2014 (ENS 50427). Upon further investigation it was verified that the function of Control Room Emergency Ventilation System was not affected as discussed in Chapters 6 and 15 of the Updated Final Safety Analysis Report. Therefore, the threshold for reporting the issue as an event or condition that could have prevented the fulfillment of a safety function was not met (NUREG 1022 Revision 3 - Event Report Guidelines Section 3.2.7). The licensee has notified the NRC Resident Inspector and applicable State authorities. Notified R3DO (Pelke)
ENS 50187	11 June 2014 00:32:00	10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Secondary Containment Declared Inoperable	Nine Mile Point Unit 2 Secondary Containment was declared inoperable on 6/10/14 from 2032 EDT until 2036 EDT during the restoration of Reactor Building Ventilation System to a normal lineup. While performing actions to restore ventilation fans the Above Refuel Floor Exhaust Fan, 2HVR-FN5A, tripped when started resulting in building differential pressure becoming less negative than -0.25 inches. Tech Spec 3.6.4.1 Secondary Containment, Action A.1 to restore within 4 hours was entered at 2032 EDT on 6/10/14. The standby Above Refuel Floor Exhaust Fan, 2HVR-FN5B, was started per procedure and building differential pressure was restored. Secondary Containment was declared operable at 2036 EDT on 6/10/14 and Tech Spec 3.6.4.1 action was exited. Secondary containment being inoperable is a 8-hour report for 10 CFR 50.72(b)(3)(v)(c), Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector and will notify the State of New York.
ENS 50164	5 April 2014 11:00:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Primary Containment Isolation System Actuation	This notification is being made in accordance with 10 CFR 50.73 (a)(2)(iv)(A) to provide information pertaining to an invalid Primary Containment Isolation System (PCIS) Group 3 actuation signal that affected containment valves in more than one system. On April 5, 2014, with the reactor at 100% power, an invalid PCIS Group 3 actuation occurred from a momentary spike of the 'A' Refuel Floor radiation monitor which reached the instrument's high radiation trip set point. A radiation protection technician was dispatched to the refuel floor and dose rates in the vicinity of the 'A' radiation monitor detector were verified to be normal and below the alarm set points. The radiation monitor was verified to be indicating normal expected radiation levels. The detector was replaced, a functional check and calibration of the radiation monitor was completed satisfactory and the instrument channel was returned to service. The issue has been entered into the station's corrective action program. Both trains of Standby Gas Treatment System started as designed and Reactor Building ventilation isolated as a result of the invalid PCIS actuation. The PCIS functioned successfully, providing a complete Group 3 isolation. PCIS Group 3 involves the following system isolation valves: Drywell and Suppression Chamber Air and Vent: V16-19-6, 6A, 6B, 7, 7A, 7B, 8, 9, 10, 23 Containment Makeup: V-16-20-20, 22A, 22B Containment Air Sampling: VG-23, 26, V109-76A, 76B Containment Air Compressor Suction: V72-38A, 38B Containment Air Dilution: VG-9A, 9B, 22A, 22B, NG-11A, 11B, 12A, 12B, 13A, 13B Since no actual high radiation condition existed which required the PCIS Group 3 isolation, and the actuation was not in response to actual plant conditions satisfying the requirements for isolation, this event has been classified as an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. In accordance with 10 CFR 50.73(a)(1) a telephone notification is being made instead of submitting a written Licensee Event Report. The licensee has notified the NRC resident inspector.
ENS 50130	22 May 2014 06:10:00	10 CFR 50.72(b)(3)(xiii), Loss of Emergency Preparedness	Unplanned Isolation of Reactor Building Ventilation Radiation Monitor	At 0210 (EDT) on May 22, 2014, Nine Mile Point Unit 2, the reactor building vent radiation monitor (Vent WRGMS) was removed from service due to a problem with the check source. The unplanned isolation of Vent WRGMS is a 8-hour report for 10 CFR 50.72(b)(3)(xiii), any event that results in a major loss of emergency assessment capability. Until the equipment is restored, Chemistry will perform sampling requirements per the ODCM. The NRC Resident Inspector has been notified. The licensee notified the State of New York Public Service Commission.

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