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	Discovered date	Reporting criterion	Title	Event description
ENS 56977	19 February 2024 04:25:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Start of Emergency Diesel Generator	The following information was provided by the licensee via phone and email: At approximately 2325 EST on February 18, 2024, with Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 100 percent power, emergency diesel generator 2 automatically started due to the unexpected loss of AC power to emergency bus E2 during a planned transfer of E2 DC control power from normal to alternate for the 1B-1 battery. In addition, the unexpected loss of AC power to E2 resulted in Unit 1 primary containment isolation system (PCIS) partial Group 2 (i.e., drywell equipment and floor drain, residual heat removal (RHR), discharge to radioactive waste, and RHR process sample), Group 6 (i.e., containment atmosphere control/dilution, containment atmosphere monitoring, and post accident sampling systems), and partial Group 10 (i.e., air isolation to the drywell) isolations. Emergency diesel generator 2 automatically started and re-energized the E2 bus as designed when the loss of E2 signal was received. The PCIS actuations were as expected for the outage plant line up on Unit 1 at the time. The cause of the loss of electrical power to emergency bus E2 is under investigation at this time. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of emergency diesel generator 2 and PCIS. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: This event will be entered into the plant's corrective action program.
ENS 56936	29 January 2024 17:02:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram	The following information was provided by the licensee via email: At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram caused by a main turbine trip. Investigation is still ongoing. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods were fully inserted. The licensee indicated that the turbine trip may have been caused by a power load imbalance, however the cause of the incident is under investigation. The scram was not complex. Decay heat is currently being removed thru bypass valves dumping to the main condenser. Initially unit 2 lost the use of the bypass valves due to lack of condenser vacuum. Unit 2 used the high pressure coolant injection (HPCI) system in the condenser storage tank (CST) to CST mode to remove decay heat. Residual heat removal was used to keep the torus cool. Condenser vacuum was regained and unit 2 is back to removing decay heat with the turbine bypass valves. There was no impact to unit 3. The licensee confirmed there was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. * UPDATE ON 01/29/24 AT 1935 EST FROM PAUL BOKUS TO NATALIE STARFISH * * The following information was provided by the licensee via email: Licensee adds 8-hour non-emergency 10 CFR 50.72(b)(3)(iv)(A) specified system actuation report to original 4-hour non-emergency 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation report. At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram by a main turbine trip. All control rods inserted. Reactor core isolation cooling system (RCIC) was manually initiated for level control. HPCI was manually initiated for pressure control. Primary containment isolation system (PCIS) Group II and III isolations occurred (specified system actuation). Investigation is ongoing. The NRC Resident Inspector has been notified.
ENS 56478	20 April 2023 05:48:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Trip Due to a Turbine Trip	The following information was provided by the licensee via phone and email: At 0148 Eastern Daylight Time (EDT) on April 20, 2023, with Unit 1 in Mode 1 at 100% power, the reactor automatically tripped due to a turbine trip. Turbine Bypass valves did not open on the trip due to Turbine Protection system power supply failure; the Safety Relief Valves (SRVs) opened automatically to control reactor pressure. Reactor Pressure reached approximately 1095 psig on the trip; exceeding the 1060 psig RPS trip setpoint. Operations responded and stabilized the plant. Operations was able to transition from SRVs to main steam line drains to the condenser. Reactor water level is being maintained via the Condensate / Feedwater system. Decay heat is being removed by discharging steam to the main condenser using the main steam line drains. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Reactor water level reached low level 1 (LL1) following the reactor trip. The LL1 signal causes Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves), and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the valid Primary Containment Isolation System (PCIS) actuation and RPS actuation from the reactor pressure signal, this event is also being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 2 is not affected by this event. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been 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 56220	13 November 2022 04:19:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Scram	The following information was provided by the licensee email: On November 12, 2022, at 2319 CST, an actuation of the reactor protection system (RPS) initiated a full scram. The plant was in Mode 2, reactor pressure was 149 pounds. The high pressure coolant injection (HPCI) injection valve, HPCI-MOV-MO19, opened and injected cold water into the reactor vessel while HPCI system testing was in progress. The cause is still under investigation. All control rods inserted. Plant is currently in Mode 3 and stable. All systems operated as designed with no Primary Containment Isolation System group isolations. This event is being reported under two event classifications: 50. 72(b)(2)(iv)(B) -- "Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. 50. 72(b)(3)(iv)(A) -- "Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. The NRC Resident has been informed.
ENS 55859	7 March 2022 04:40:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation 60-DAY Telephone Notification	The following information was provided by the licensee via fax or email: This 60-day telephone notification is being made in lieu of an LER submittal per 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 systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0040 Eastern Standard Time (EST) on March 7, 2022, Unit 1 received inadvertent High-Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiation signals. Subsequently, at approximately 0148 EST on March 7, 2022, Unit 1 received inadvertent Low-Pressure Coolant Injection (LPCI) and Core Spray initiation signals. In addition, all four Emergency Diesel Generators auto started, Group 10 (Instrument Air) Primary Containment Isolation System actuations occurred, and the Residual Heat Removal (RHR) Service Water Booster pumps tripped resulting in a brief interruption (approximately 9 minutes) to the Shutdown Cooling (SDC) heatsink. Jumpers, installed per planned refueling outage activities, prevented discharge of Emergency Core Cooling Systems into the reactor. HPCI, RCIC, and RHR Loop `A' were removed from service and under clearance. RHR SDC remained operable via RHR Loop `B' and forced circulation was maintained in the reactor. At the time of these events, Unit 1 was shutdown for refueling and the `A' and `C' reactor water level transmitters had been isolated in preparation for planned replacement. Leak-by of the instrument isolation valves occurred on both transmitters. Leak-by on the `C' instrument occurred at a faster rate with the `A' instrument providing the confirmatory signals resulting in Low Level 2 (LL2) and Low Level 3 (LL3) indication at approximately 0040 EST and 0148 EST, respectively. All actuations occurred as designed for LL2 and LL3 signals. During these events, reactor water level remained stable at the Reactor Vessel Head Flange and the `B' and `D' reactor water level transmitters remained off-scale-high, as expected under these conditions. Therefore, the actuations were 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 (i.e., there was no low reactor water level condition). Considering the above, these actuations were invalid. There was no impact on the health and safety of the public or plant personnel.
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 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 55575	14 November 2021 10:25:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Manual Trip Due to Lowering Main Condenser Vacuum	At 0525 EST, November 14, 2021, "Unit 2 was manually scammed by operations due to lowering main condenser vacuum. This resulted in PCIS (primary containment Isolation system) Group II/III isolation signals. All control rods inserted, and all systems operated as designed. Unit 3 is unaffected and remains at 100 percent power in Mode 1. The Resident Inspector was notified.
ENS 55345	13 May 2021 11:00:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Containment Isolation Signal	This 60-Day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of containment isolation signal affecting more than one system. On May 13, 2021, during the restoration of the Unit 2 Refuel Floor High Radiation Isolation Logic an invalid isolation signal was received. The condition requiring an isolation signal was verified not to be present prior to restoring the logic; however, it was not recognized that a previous isolation signal was latched in and had not been reset. When the isolation logic was restored, the Primary Containment Isolation System (PCIS) isolated on the invalid signal. The systems successfully completed the isolation per the plant design and plant configuration. The following systems actuated due to the Unit 2 PCIS Group 6C Isolation: - Isolation of Containment Hydrogen and Oxygen Sampling Valves, - Start of the 2A Reactor Enclosure Recirculation System, - Trip of the Units 1 and 2 Refuel Floor HVAC, - Start of the A and B Trains of Standby Gas Treatment Systems. The NRC Resident Inspector was notified.
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 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 54826	10 August 2020 17:58:00	10 CFR 50.72(a)(1)(i), Emergency Class Declaration	Notice of Unusual Event Due to Loss of Offsite Power Due to High Winds	At 1258 CDT on August 10, 2020, Duane Arnold Energy Center declared an Unusual Event due to a loss of offsite power due to high winds. The event at the single unit plant resulted in an automatic scram from 82 percent power (Mode-1) to zero percent power (Mode-3). They are headed to Mode-4. There is damage on site, but the Reactor Building is intact. All rods inserted and cooling is being addressed via Reactor Core Isolation Cooling (RCIC) for level control and Safety Relief Valves are removing decay heat to the torus. Both Standby Diesel Generator are running. The licensee notified the NRC Resident Inspector, the Iowa Department of Emergency Management, and the Linn County and Benton County Emergency Management agencies. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). * * UPDATE ON 08/10/2020 AT 1554 EDT FROM CURTIS HANSEN TO OSSY FONT * * * This report is being made under CFR 50.72 (b)(2)(iv)(B) for an automatic reactor scram due to loss of offsite power due to high winds. In addition, this report is being made under CFR 50.72 (b)(3)(iv)(A) and (B) due to PCIS ((Primary Containment Isolation System)) Groups 1, 2, 3, 4 and 5 (activating) due to loss of offsite power. All isolations went to completion. RCIC injecting for level control. All rods fully inserted during the scram. The plant electrical line up is both SBDGs (Standby Diesel Generators) are running. Decay heat is being removed via SRVs (Safety Relief Valves) to the torus. Progress towards shutdown cooling. NRC Senior Resident (Inspector) notified at 1448. Notified R3DO (Pelke).
ENS 54820	8 June 2020 08:24:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Specified System Actuation	This report is being made as required by 10 CFR 50.73(a)(2)(iv)(A) to describe an automatic actuation of containment isolation valves in more than one system. Because the actuation was invalid, this 60-day telephone notification is being made instead of a written LER (licensee event report), in accordance with 10 CFR 50.73(a)(1). On 06/08/2020, at approximately 0424 EDT, a trip of the Unit 3 'A' reactor protection system (RPS) MG-Set resulted in a partial activation of the primary containment isolation system and inboard containment isolation valves closed in multiple systems. All affected Group III containment isolation valves were verified to be closed. It was determined that the normal power supply for the Unit 3 'A' RPS had failed. Power was transferred from the normal to the alternate source and the RPS 'A' channel was reset. Investigation determined that the 3A RPS MG Set motor contactor coil winding had failed due to an internal short circuit. The motor contactor has been replaced. The containment isolation was not due to actual plant conditions or parameters meeting design criteria for containment isolation. Therefore, this is considered to be an invalid actuation. The NRC Resident Inspector has been informed of this notification.
ENS 54812	4 August 2020 03:12:00	10 CFR 50.72(a)(1)(i), Emergency Class Declaration	Unusual Event Declared Due to a Loss of Offsite Power	At 2312 EDT, on August 3, 2020, Brunswick Unit 1 declared an Unusual Event due to a loss of offsite power. The unit was at approximately 20 percent power and was not synced to the grid when the unit automatically scrammed. All control rods fully inserted. Emergency Diesel Generators started and began powering the safety buses. Safety systems actuated as expected. The Unit also experienced a loss of Fuel Pool Cooling and Cleanup System, but one pump was returned to service. Unit 2 remains at 100 percent power and is unaffected. The licensee notified State and local governments, as well as the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, CISA IOCC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). * * UPDATE FROM MARK TURKAL TO DONALD NORWOOD AT 0120 EDT ON 8/4/2020 * * * At approximately 2302 EDT, a loss of offsite power occurred on Unit 1. This resulted in a Reactor Protection System (RPS) actuation. Per design, emergency diesel generators 1 and 2 properly started and loaded to their respective emergency buses. The Reactor Core Isolation Cooling (RCIC) system was manually started and is being used to control reactor water level. The High Pressure Coolant Injection (HPCI) system was manually started and is being used for pressure control. As previously reported, an Unusual Event was declared at 2312 EDT due to the loss of offsite power. At the time of the event, Unit 1 was in the process of shutting down for maintenance associated with a ground on the main generator. Due to the RPS actuation while critical, this event is being reported as a four-hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B). As a result of the reactor trip, reactor water level reached low level 1 (LL1). The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sample isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Per design, the loss of offsite power also caused a Group 1 (i.e., main steam isolation valve) isolations. Due to the Emergency Diesel Generator and Primary Containment Isolation System (PCIS) actuations, this event is also being reported as an eight-hour, nonemergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The safety significance of the event is minimal. All safety related systems operated as designed. Investigation of the cause of the loss of offsite power is in progress. The licensee notified the NRC Resident Inspector. Notified R2DO (Inverso). * * UPDATE ON 8/4/2020 AT 1534 EDT FROM JOSEPH ELKINS TO ANDREW WAUGH * * * At 1454 EDT on August 4, 2020, the Unusual Event was exited when offsite power was restored to Unit 1. Per design, when the loss of offsite power to Unit 1 occurred, all four emergency diesel generators (EDGs) started and EDGs 1 and 2 properly suppled emergency buses 1 and 2. Since Unit 2 was not affected by the loss of power, EDGs 3 and 4 ran unloaded. With restoration of offsite power to Unit 1, EDG 2 has been secured. EDGs 1, 3, and 4 are being secured as required by plant operating procedure. Notified R2DO (Inverso), NRR EO (Miller), IRD MOC (Grant), DHS SWO, FEMA OC, DHS NICC WO, CISA IOCC (email), DHS SWO (email), FEMA NWC (email), FEMA Ops Center (email), FEMA-NRCC-sasc (email), NRCC THD Desk (email), NuclearSSA (email). ********************************************************************************************************************************
ENS 54795	21 July 2020 09:35:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation	Biofouling of Intake Structure Leads to Manual Reactor Scram of Unit 1	The following was received from TVA - Brown's Ferry at 0858, 21 July 20. On July 21, 2020, at 0435 hours Central Daylight Time, Browns Ferry Unit 1 inserted a manual reactor scram due to degrading main condenser vacuum from marine biofouling at the intake structure. Browns Ferry Unit 2 is in Mode 4 and Browns Ferry Unit 3 is at approximately 76% rated thermal power and stable. Primary Containment Isolation Systems received an actuation signal for groups 2, 3, 6, 8 on reactor water level below +2". All Primary Containment Isolation System groups that received an actuation signal performed as designed. Additionally, all other systems functioned as designed. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B) - Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This event also requires an 8 hour report per 10CFR50.72(b )(3)(iv)(A), Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation and (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs). The NRC Resident Inspector has been notified. The plant is stable in Mode 3 and will remain shutdown until marine growth clogging the intake structure abates.
ENS 54709	16 March 2020 06:02:00	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 a general containment isolation signal affecting more than one system. On March 16, 2020, at approximately 0102 CDT, Browns Ferry Nuclear Plant (BFN), Unit 3 received motor trip-out alarms and diagnosed Group 2 and 3 Primary Containment Isolation System (PCIS) Isolations, 3C Residual Heat Removal (RHR) Pump tripping and Reactor Water Cleanup (RWCU) system isolating. All affected safety systems responded as expected. BFN, Unit 3, was nearing the end of the U3R19 refueling outage at the time of the event, and was still dependent on the Shutdown Cooling (SDC) system. 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. At the time of the event, these conditions did not exist: therefore, the PCIS actuation was invalid. The event was determined to have been caused by clearance restoration activities in an unprotected control panel. A fuse re-installation inadvertently created a fault condition between two different plant 120 VAC power sources when the fuse holder's lower spring clip contacted a different fuse. This was a result of age-related degradation of the fuse holder, its close proximity to other fuses, and the lack of insulating isolation barriers between fuses. 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 1594925. 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 54603	24 March 2020 16:05:00	10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown	Technical Specification-Required Shutdown Due to Unidentified Leakage	At 1205 Eastern Daylight Time (EDT) on March 24, 2020, a Technical Specification-required shutdown was initiated on Unit 1 due to indication of a leak in the drywell. Technical Specification Action 3.4.4.A, Unidentified Reactor Coolant System (RCS) leakage increase not within limit, requires RCS leakage to be reduced to within limits within 8 hours. It was expected that the leakage would not have been reduced to within limits within the required Technical Specification completion time; therefore, this event is being reported in accordance with 10 CFR 50.72(b)(2)(i). Reactor water level reached low level 1 (LL1) following the reactor shutdown. The LL1 signal causes Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves), and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the valid Primary Containment Isolation System (PCIS) actuation, this event is also being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 2 is not affected by this event. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
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 54137	27 June 2019 13:37:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Loss of Power Resulted in a Valid Isolation Signal	At 0937 EDT on 6/27/19, Susquehanna Unit 2 had a control power fuse fail that caused a loss of power to one of two power distribution systems of the Reactor Protection System. This loss of power resulted in a valid isolation signal and actuation of the Primary Containment Isolation System that affected multiple systems, and is reportable per 10 CFR 50.72(b)(3)(iv)(A). The situation has been resolved and the plant is currently using an alternate power source. Offsite power and the Emergency Diesel Generators are still available. The licensee notified the NRC Resident Inspector.
ENS 54016	22 April 2019 03:07:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Reactor Trip and Specified System Actuation	At 2307 EDT on April 21, 2019, in Mode 1 at approximately 100 percent reactor power, Unit 1 automatically tripped due to a Main Turbine Trip. The Main Turbine Trip was a result of two out of three level instruments sensing a false high reactor water level. All control rods inserted as expected during the scram. Safety Relief Valves G and K lifted per design. The same level instruments that failed also tripped both Reactor Feed Pumps. As a result, reactor water level dropped below the Low Level 1 and 2 actuation setpoints. Per design, the Low Level 1 signal resulted in Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The Low Level 2 signals resulted in Group 3 (i.e. Reactor Water Cleanup) isolation, a secondary containment isolation signal, and an auto start of Standby Gas Treatment and Control Room Emergency Ventilation. Also, the Low Level 2 resulted in (high pressure coolant injection) HPCI and (reactor core isolation cooling system) RCIC automatically starting and injecting into the vessel. All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Decay heat is currently being removed via the turbine bypass valves. Condensate and feed water are maintaining water level. The reactor is still at saturation temperature and 475 psi, lowering slowly. The reactor is still in a normal electrical lineup. There was no impact to Unit 2 as a result of this event. * * UPDATE ON 04/22/19 AT 0220 EDT FROM ALAN SCHULTZ TO JEFFREY WHITED * * * The licensee updated the event report to include a 4-Hr Non-Emergency Notification in accordance with 10 CFR 50.72(b)(2)(iv)(A) for Emergency Core Cooling System, HPCI, Discharge to the Reactor Coolant System. Notified R2DO (Dickson), NRR EO (Miller) and IR MOC (Gott).
ENS 53966	30 March 2019 21:47:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Manual Reactor Scram and Specified System Actuation	At 17:47 Eastern Daylight Time (EDT) on March 30, 2019, with Unit 2 in Mode 1 at approximately 23 percent reactor power and main turbine startup in progress coming out of a refuel outage, a high temperature was sensed at main turbine bearing #9. As a result of and to arrest the high temperature condition, the main control room inserted a manual reactor scram. All control rods inserted as expected during the scram. When the scram was inserted, reactor water level dropped below the Low Level 1 actuation setpoint. Per design, the Low Level 1 signal resulted in Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The main control room manually closed all Main Steam Isolation Valves (MSIVs), in anticipation of a low vacuum prior to the Group 1 automatic closure signal being received. High Pressure Coolant Injection (HPCI) was aligned for pressure control and Reactor Coolant Isolation System (RCIC) was aligned for level control. The Reactor Coolant Sample Line Isolation valves closed as expected on low main condenser vacuum. All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. At the time of notification, decay heat was being removed by the condenser through one open MSIV and a feedwater pump running.
ENS 53962	28 March 2019 20:54:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Actuation of the Primary Containment Isolation System and the Reactor Protecton System	At 1654 EDT on March 28, 2019, with Unit 1 in Mode 3 at 0 percent power, an actuation of the Primary Containment Isolation System occurred, closing the outboard Main Steam Isolation Valves (MSIVs) due to a low condenser vacuum signal. The MSIVs had been manually closed, per procedure, during the shutdown evolution to address drywell leakage. The inboard MSIVs had not been reopened when the isolation occurred. Subsequently, at 1658 EDT a Reactor Protection System (RPS) actuation occurred due to reactor water level dropping below the actuation setpoint. All control rods were inserted at the time of the actuation. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System and the Reactor Protection System. There was no impact on the health and safety of the public or plant personnel. The safety function of both the MSIVs and the RPS had already been completed at the time of the event. The NRC Resident Inspector has been notified.
ENS 53953	24 March 2019 05:59:00	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 Degrading Condenser Vacuum	At 0159 (EDT), with Unit 2 in Mode 1 at 25 percent power, the reactor was manually tripped due to degrading condenser vacuum. After the turbine was tripped, the station service electrical buses did not transfer to alternate supply resulting in loss of the condensate feedwater system and level being controlled by the RCIC system. Operators responded and stabilized the plant. Reactor water level is being maintained via the RCIC system. Pressure is being controlled and decay heat is being removed by the HPCI system in pressure control mode. Unit 1 is not affected. Additionally, an actuation of the primary containment isolation system occurred during the reactor scram. The reason for the actuation was a group II isolation signal was received on reactor water level and a group I isolation was received on decreasing vacuum. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non- emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the primary containment isolation system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 53923	10 March 2019 04:59:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Scram Resulting in Rps and Eccs Actuation	At 2259 CST on 3/9/2019, Browns Ferry Unit-3 received an automatic SCRAM on Main Generator Breaker Failure and Turbine Load Reject. Unit-3 declared a Notification of Unusual Event SU1 for loss of offsite AC power to Unit-3 specific 4kV Shutdown Boards for greater than 15 minutes. Primary Containment Isolation Systems (PCIS) Groups 1, 2, 3, 6, and 8 isolation signals were received. Upon receipt of these signals, all required components actuated as required. Main steam relief valves lifted on the initial transient. High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) initiated on low reactor water level. HPCI remains in service for reactor level and pressure control. RCIC is not in service at this time, the station is investigating low flow from the pump. All four Unit-3 Diesel Generators started and loaded as expected. Residual Heat Removal System is in service for suppression pool cooling. 4kV Station Unit Boards have been restored from the 161kV system. Actions are in progress to restore 4kV Shutdown Boards to offsite power. This event is reportable within 1 hour in accordance with 10 CFR 50.72(a)(1)(i) for declaration of the Licensees Emergency Plan. Complete as documented on EN 53922. This event requires a 4 hour report per 10 CFR 50.72(b)(2)(iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' This event also requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A). 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (1) Reactor protection system (RPS) including: reactor scram or reactor trip, (2) General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs), (4) ECCS (Emergency Core Cooling System) for boiling water reactors (BWRs) including: core spray systems; high-pressure coolant injection system; low pressure injection function of the residual heat removal system, (5) BWR reactor core isolation cooling system; isolation condenser system; and feedwater coolant injection system, and (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. As of the event report, the MSIVs were opened and decay heat was being removed via the bypass valves to the condenser.
ENS 53911	5 March 2019 10:35:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Actuation of the Primary Containment Isolation System	At 05:35 Eastern Standard Time (EST) on March 5, 2019, with Unit 2 in Mode 5 at 0% power, an actuation of the Primary Containment Isolation System occurred during hydrolazing of the reactor water level variable leg instrumentation line nozzle N011B in the reactor cavity. The hydrolazing activity caused low reactor water level to be sensed on Division II of the shutdown range level instrumentation. Per design, the low level 1 signal resulted in Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The Group 8 was reset and shutdown cooling was restored at approximately 05:45 EST. The safety significance of this event was minimal. Although there was a brief interruption of the shutdown cooling, the Residual Heat Removal (RHR) shutdown cooling system operation was restored in approximately 10 minutes without extensive troubleshooting or maintenance, and remained operable. The RHR shutdown cooling system is not credited in any Updated Final Safety Analysis Report Chapter 6 or 15 accidents or transients. The NRC Resident Inspector has been notified.
ENS 53776	13 October 2018 05:00:00	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 Primary Containment Isolation System (PCIS) Group 1 for Main Steam Isolation Valves (MSIVs), Group 3 for Reactor Water Cleanup (RWCU), Group 6 for Secondary Containment isolation, Group 7 for Reactor Water Sampling, Diesel Generator, Reactor Core Isolation Cooling (RCIC) System logic, and Residual Heat Removal (RHR) logic. Group 1, Group 6, Diesel Generator actuation, RCIC actuation and RHR actuation are within scope of 10 CFR 50.73(a)(2)(iv). Group 3 and Group 7 are not within scope as they affect only one system. Cooper Nuclear Station (CNS) was shut down in Mode 5 at the time of the event with the reactor cavity flooded. On October 13, 2018, at 0028 Central Daylight Time, CNS received full PCIS Groups 1, 3, and 6, and a half Group 7 on the Division 1 side. The MSIVs and RWCU isolation valves were already closed for maintenance. The Secondary Containment isolated. Control Room Emergency Filter and the Standby Gas Treatment Systems initiated. The inboard Reactor Water Sample valve isolated. Diesel Generator #1 started but was not required to connect to the critical bus. Reactor Core Isolation Cooling System logic actuated with no expected response due to being isolated for shutdown conditions. Division 1 RHR pump logic actuated. Division 1 RHR system was operating in shutdown cooling mode. The actuation caused the Division 1 RHR outboard injection and heat exchanger bypass valves to open. Shutdown cooling was unaffected and remained in service throughout the event. The plant systems responded as expected with no Emergency Core Cooling System injection. At the time of the event, an in-service inspection of welds inside the reactor vessel was taking place using a robot scanner that uses two vortex thrusters to hold the robot to the vessel wall. The robot inadvertently passed over an instrument penetration, drawing suction on the process leg, resulting in low reactor water level indications and the subsequent invalid Level 1 and 2 system actuations. Actual reactor vessel water level remained steady at cavity flooded conditions. The NRC Resident Inspector has been notified of this event.
ENS 53676	19 October 2018 05:00:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge	Automatic Reactor Trip Due to Feedwater Regulating Valve Failing Closed	At 1725 CDT, a Feedwater Regulating valve failed closed, resulting in a reactor level transient, which initiated a reactor trip, Primary Containment Isolation System signals to valves in Groups 2, 3, and 4 and initiation of High Pressure Coolant Injection and Reactor Core Isolation Cooling. All control rods inserted and level has been restored to normal. The cause of the feedwater valve failure is under investigation. All other systems responded as expected. This report is being made under 10 CFR 50.72 (b)(2)(iv)(B), (b)(3)(iv)(A) and (b)(2)(iv)(A). The Senior Resident Inspector has been informed. Decay heat is being removed via the main condenser and reactor vessel water level is being maintained by the condensate and feedwater systems.
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 53497	9 July 2018 05:00:00	10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident	High Pressure Coolant Injection System Declared Inoperable	On 07/09/2018 at 1111 CDT, Browns Ferry Unit 1 Operators identified U1 High Pressure Cooling Injection system steam supply valves were isolated. After reviewing ICS (Integrated Computer System), Operations determined isolation occurred at 0958 CDT during performance of surveillance testing. The Browns Ferry Nuclear Plant Unit 1 High Pressure Coolant Injection (HPCI) system was declared inoperable at 0958 CDT due to an inadvertent isolation that occurred during testing. During performance of surveillance procedure 1-SR-3.3.6.1.2(3B) HPCI System Steam Supply Low Pressure Functional test, an erroneous signal was induced causing actuation of primary containment isolation system group IV (i.e., HPCI Isolation). Technical Specification 3.5.1, ECCS-Operating, Condition C was entered as a result of the inoperable HPCI system. This constitutes an unplanned HPCI system inoperability and requires an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(v)(D). The erroneous signal was cleared and the HPCI isolation was reset. Upon reset of the isolation signal, the HPCI system was returned to available status. The HPCI system was unavailable for 2 hours and 55 minutes, however the HPCI system remains inoperable. There was no impact to the health and safety of the public or plant personnel as a result of this condition. The NRC Resident Inspector has been notified. A condition report has been entered into the Licensee's Corrective Action Program to capture this event.
ENS 53319	7 April 2018 12:36:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation	Automatic Reactor Trip and Pcis Actuation During Stator Cooling System Testing	On April 7, 2018, at 0836 EDT, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped during testing of the stator cooling system. The trip was uncomplicated with all systems responding normally. No safety-related equipment was inoperable at the time of the event. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Operations responded using Emergency Operating Procedures and stabilized the plant in Mode 3. Reactor water level being maintained via normal feedwater system. Decay heat is being removed through the bypass valves. Reactor water level reached low level 1 (LL1) as a result of the reactor trip. The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the Primary Containment Isolation System (PCIS) actuation, this event is also being reported as an eight-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PCIS. Unit 2 was not affected. There was no impact on the health and safety of the public or plant personnel. The safety significance of this event is minimal. The automatic reactor trip was not complicated and all safety-related systems operated as designed. Investigation of the cause of the Reactor Protection System actuation is in progress. The licensee notified the NRC Resident Inspector.
ENS 53167	1 December 2017 04:30:00	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 under reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) to describe an invalid actuation of a containment isolation signal affecting more than one system. At 2230 CST on November 30, 2017, with the Duane Arnold Energy Center (DAEC) operating at 100 percent power, an invalid Group 3 isolation on the 'B' side of the Primary Containment Isolation System (PCIS) occurred. Group 3 isolation signals were generated for Primary Containment Isolation Valves for Drywell and Torus Ventilation and Purge, Containment Nitrogen Compressor Suction and Discharge, Recirculation Pump Seals, and Post Accident Sample System. This event was caused by a fault on the 1D25 Instrument AC Inverter. The fault was caused by an insufficient design clearance to ground and was corrected by increasing the clearance. All equipment responded in accordance with the plant design. Specifically, all actuations were complete and successful. There were no safety consequences or impacts on the health and safety of the public. The event was entered into DAEC's corrective action program for resolution. The NRC Resident Inspector has been notified.
ENS 53147	4 January 2018 19:10:00	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 Partial Loss of Offsite Power During Winter Storm	On January 4, 2018, at 1410 hours EST, with the reactor at approximately 100 percent power and steady state conditions, the winter storm across the Northeast caused the loss of offsite 345 kV Line 342. Reactor power was reduced to approximately 81 percent and a procedurally required manual reactor scram was initiated. All control rods fully inserted. As a result of the reactor scram, indicated reactor water level decreased, as expected, to less than +12 inches resulting in automatic actuation of the Primary Containment Isolation Systems for Group II - Primary Containment Isolation and Reactor Building Isolation System, and Group VI - Reactor Water Cleanup System. Reactor Water Level was restored to the normal operating band. The Primary Containment Isolation Systems have been reset. The Reactor Protection System signal has been reset. Following the reactor scram, the non-safety related Control Rod Drive Pump "B" tripped on low suction pressure. Control Rod Drive Pump "A" was placed in service. All other systems operated as expected, in accordance with design. This event is reportable per the requirements of Title 10, Code of Federal Regulations (CFR) 50.72 (b)(2)(iv)(B) - "RPS Actuation" and 10 CFR 50.72 (b)(3)(iv)(A) - "Specified System Actuation. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The main steam isolation valves are open with decay heat being removed via steam to the main condenser. Offsite power is still available from 345kV line 355. As a contingency, emergency diesel generators are running and powering safety busses per licensee procedure. The licensee notified the Commonwealth of Massachusetts. The licensee will be notifying the town of Plymouth as part of their local notifications. The licensee will be issuing a press release.
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 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 52685	17 February 2017 15:21:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephonic Notification of Invalid Specified System Actuation	On February 17, 2017 at 1021 EST, secondary containment isolated and Standby Gas Treatment (SBGT) systems started on Unit 1 and Unit 2 during a maintenance activity to replace a relay in the Unit 2 primary containment isolation system. The work was being conducted as part of planned maintenance during the Unit 2 refueling outage. Poor work instruction led to a jumper not being installed as required, thus causing relays to de-energize, resulting in an invalid actuation of the Unit 1 and Unit 2 outboard primary and secondary containment isolation valves and auto-start of the Unit 1 and Unit 2 SBGT system. This event is reportable per 10 CFR 50.73(a)(2)(iv)(A) since the containment isolation and auto-start of SBGT on both units was not part of a pre-planned sequence and the event resulted in the invalid actuation of general containment isolation valves in more than one system. All primary and secondary containment isolation valves and SBGT systems functioned successfully. The associated wires were re-landed and secondary containment was returned to normal service. The licensee notified the NRC Resident Inspector.
ENS 52682	15 April 2017 09:41:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Reactor Protection System and Partial Primary Containment Isolation System Actuations on Low Water Level	During shutdown activities with the reactor subcritical, actions were being taken to remove 11 Reactor Feed Pump from service in support of a scheduled refueling outage. Reactor Water Level on Safeguards level instrumentation dropped below +9 inches, which resulted in a valid Reactor Protection System (RPS) Scram signal and Partial Group 2 Primary Containment Isolation System (PCIS) signal. All systems functioned as required. Reactor Water Level on Safeguards instrumentation was restored to greater than +9 inches immediately. RPS and PCIS logic was reset. There was no impact to the health and safety of the public as a result of this event. This actuation of these systems is being reported per 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of any of the systems listed in 10 CFR 50.72(b)(3)(iv)(B). The NRC Resident Inspector has been notified.
ENS 52645	31 January 2017 19:25:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Report Due to Invalid Eccs Actuation Signal	The following report is made pursuant to 10 CFR 50.73(a)(2)(iv)(A) due to an unintended initiation signal that occurred on January 31, 2017 with James A. FitzPatrick Nuclear Power Plant (JAF) in Mode 5 at zero (0) percent power. On January 31, 2017 at 1425 (EST) the control room received multiple annunciations associated with the following Systems / Trains: Primary Containment Isolation System (PCIS) / Trains A and B Residual Heat Removal System (RHR) / Trains A and B Core Spray (CS) / Trains A and B Reactor Core Isolation Cooling (RCIC) All four (4) Emergency Diesel Generators (EDG) auto-started with their associated Emergency Service Water pumps operating. RHR and CS both received initiation signals but were defeated per procedure. The HPCI (High Pressure Coolant Injection) auxiliary oil pump was taken to Pull-to-Lock per procedure, and the RCIC steam isolation valve cycled until the breaker was opened to close the valve. An evaluation concluded that the (Emergency Core Cooling System - ECCS) initiation signals were caused by the opening of a portable job box that was stored near sensitive equipment. Upon opening the job box, the lid bumped a reference leg resulting in the initiation signals. All initiation signals were reset and systems restored to normal shutdown lineups. The licensee notified the NRC Resident Inspector. * * UPDATE ON 3/30/17 AT 0840 EDT FROM DUSTIN SCURLOCK TO DONG PARK * * * To the original report, the licensee added, "This condition recurred at 1624 (EDT on 1/31/17). The licensee notified the NRC Resident Inspector. Notified R1DO (Cook).
ENS 52436	28 October 2016 07:39:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Containment Isolation Valves	This telephone notification is being made in lieu of a Licensee Event Report in accordance with 10CFR50.73(a)(1) to notify the NRC of on invalid actuation of a group isolation, reportable under 10CFR50.73(a)(2)(iv)(A). On October 28, at 0139 Central Daylight Time, electricians were removing a test jack from a terminal on a primary containment isolation system relay due to the test jack being installed on the wrong relay. During removal of the test jack, a neutral wire came out of the terminal causing multiple relay actuations resulting in a half Group 2 isolation signal on the Division 2 side. Because of the plant being in Mode 5, refueling, the only isolation valves that closed because of the invalid isolation signal were the Division 2 isolation valves associated with the drywell ventilation monitor and the drywell floor and equipment sumps. The specific train and system actuated was the Primary Containment Isolation System (PCIS) Division 2 portion of the Group 2 isolation circuit (i.e. half group isolation). This was a partial actuation affecting only half of the logic and causing isolation of two systems. Based on the inadvertent actuation of the affected PCIS relays, the system isolated and functioned as expected. The isolation signal was reset at 0230 and the affected systems restored to service. The licensee notified the NRC Resident Inspector.
ENS 51952	28 March 2016 05:50:00	10 CFR 50.73(a)(1), Submit an LER	Invalid Actuation of Inboard Isolation Logic	On Monday, March 28, 2016, Unit 1 was in OPCON 5 (Refueling) conducting a refueling outage. A modification was being installed for an NSSSS (Nuclear Steam Supply Shutoff System) Test Box on Division 1A Group 1 NSSSS logic. At 0150 hours, a logic jumper was removed as directed by the work order and a logic fuse failed. The fuse failure caused an unplanned invalid actuation of the inboard isolation logic. The isolations were reset and the valves were restored to initial conditions at 0246 hours. On Sunday, April 3, 2016, at 0134 hours, one additional logic fuse opening event occurred during the testing which also caused an invalid actuation which was reset at 0405 hours. The fuse openings occurred during jumper manipulations as the modification was tested on the Division 1A and 1D logic during the refuel outage. The investigation determined the fuse openings were due to the testing process. The suspected devices that caused the condition are not permanent plant equipment and there is no degradation of the actual circuit. They were part of a temporary configuration that was installed to support modification installation and acceptance testing. The temporary devices have been removed. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The isolation was a partial actuation. This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report invalid automatic actuations of systems listed in paragraph (a)(2)(iv)(B). The listed system that actuated was general containment isolation signals affecting containment isolation valves in more than one system. Primary containment isolation valves (PCIVs) closed on reactor water cleanup (RWCU), drywell chilled water (DWCW), primary containment instrument gas (PCIG), drywell sumps and the suppression pool cleanup systems. The licensee has notified the NRC Resident Inspector.
ENS 51838	31 March 2016 10:03:00	10 CFR 50.72(b)(3)(iv)(A), System Actuation	Valid Isolation Signal Due to a Human Performance Error	On March 31, 2016 at 0603 (EDT), with Susquehanna Unit 1 in its 19th Refueling and Inspection Outage, Unit 1 received a valid isolation signal. Preliminary investigation indicates the isolation signal was the result of a human performance error. The systems affected by the isolation signal responded as designed for the current shutdown plant conditions. This isolation of multiple primary containment isolation systems is being reported under 10CFR50.72(b)(3)(iv)(A) and per the guidance of NUREG 1022, Rev. 3, section 3.2.6 as a system actuation. The licensee notified the NRC Resident Inspector.
ENS 51505	10 September 2015 01:03:00	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 invalid actuation containment isolation signals affecting containment isolation valves in more than one system. On September 9, 2015 at 2103 hours Eastern Daylight Time (EDT), Unit 1 experienced a loss of electrical power to motor control center 1CB when the substation E6 feeder breaker tripped. The loss of power resulted in closure of primary containment isolation valves (PCIVs) in Unit 1 Primary Containment Isolation System (PCIS) Group 2 (i.e. Drywell Equipment and Floor Drains, Residual Heat Removal (RHR) Discharge to Radwaste, RHR Process Sample, and Traversing lncore Probe), Group 3 (i.e., Reactor Water Cleanup), and Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems). It has been determined that affected PCIVs appropriately closed. However, the limit switch within the motor operator of the inboard RWCU PCIV (i.e. 1-G31-F001) malfunctioned; resulting in an inaccurate remote position indication. Testing has confirmed that 1-G31-F001 properly closed and can perform its intended safety function. These PCIV isolations were the result of a substation E6 feeder breaker trip to motor control center 1CB and not in response to actual plant conditions (i.e., to mitigate the consequences of an event) and, therefore, were invalid. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified.
ENS 51391	14 September 2015 03:05:00	10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(3)(v)(C), Loss of Safety Function - Release of Radioactive Material	Manual Scram Due to Loss of Turbine Building Closed Cooling Water	At 2305 EDT on September 13, 2015, a manual scram was initiated in response to a loss of all Turbine Building Closed Cooling Water (TBCCW). All control rods fully inserted. The lowest Reactor Water Level (RWL) reached was 137 inches. All isolations and actuations for RWL 3 occurred as expected. Decay heat was initially being removed through the Main Turbine Bypass System to the Main Condenser, however, as a result of the loss of TBCCW, the Main Feed Pumps lost cooling and had to be secured. At 2310, Standby Feedwater was initiated and Main Feedwater was secured. The loss of TBCCW also caused all Station Air Compressors (SACs) to trip on loss of cooling. The loss of SACs caused the Instrument Air header pressure to degrade to the point at which the Secondary Containment isolation dampers drifted closed. This resulted in the Reactor Building vacuum exceeding the Technical Specification limit. At 2325, operators started the Standby Gas Treatment system and manually initiated a Secondary Containment isolation signal. Secondary Containment vacuum was promptly restored to within Technical Specification limits. Additionally, Operators were monitoring for expected MSIV drift due to the degraded Instrument Air header pressure. When outboard MSIVs were observed to be drifting, Operators closed the outboard and inboard MSIVs at 2345. At 2352, Safety Relief Valves (SRVs) reached the Low-Low Setpoint and began cycling to control reactor pressure. RWL is currently being maintained in the normal level band with the Standby Feedwater and Control Rod Drive systems. Reactor Pressure is being controlled with Safety Relief Valves. Operators are currently in the Emergency Operating Procedure for Reactor Pressure Vessel control. Investigation into the loss of TBCCW continues. No safety-related equipment was out of service at the time of the event. All offsite power sources were adequate and available throughout the duration of the event. The NRC resident inspector has been notified. * * UPDATE AT 0555 EDT AT 09/14/15 FROM CHRIS ROBINSON TO JEFF HERRERA * * * At 0409 EDT the Reactor Core Isolation Cooling (RCIC) system was placed in service due to identification of an unisolable leak in the Standby Feedwater System. Reactor water level and pressure is now being controlled though the RCIC system and Safety Relief Valves. This event update is reportable as a valid manual initiation of a specified safety system under 10CFR50.72(b)(3)(iv)(A). The NRC resident inspector has been notified. The leak rate was reported as approximately 5-10 gallons per minute from a weld on the standby feedwater pump header drain valve F326. The licensee reported the leak stopped once RCIC was placed into service. The licensee is still investigating the issue. Notified the R3DO (Pelke), IRD Manager (Grant), NRR EO (Morris). * * UPDATE PROVIDED BY CHRIS ROBINSON TO JEFF ROTTON AT 2135 EDT ON 09/14/2015 * * * At 1847 EDT on September 14, 2015, a valid automatic Reactor Protection System (RPS) actuation occurred due to Reactor Water Level 3 while shutdown in MODE 3. Operators were manually controlling Reactor Pressure Vessel (RPV) level and pressure with Reactor Core Isolation Cooling (RCIC) and Safety Relief Valves (SRV). While operators were cycling SRVs, the RPV level went below the Level 3 setpoint. Operators promptly restored RPV level by manual operation of RCIC. The Level 3 actuation and associated isolations were verified to operate properly. The scram signal has been reset. Fermi 2 remains in MODE 3 controlling RPV Level and Pressure through manual operation of RCIC and SRVs. This is the second occurrence of a valid specified safety system actuation reportable under 10CFR50.72(b)(3)(iv)(A) for this ongoing event. The NRC Resident Inspector has been notified. Notified R3DO (Riemer), IRD Manager (Grant), and NRR EO (Morris) * * UPDATE FROM BRETT JEBBIA TO JOHN SHOEMAKER AT 1446 EST ON 2/27/16 * * * This update provides clarification of the applicable reporting criteria for this Event associated with primary containment isolation actuations. Upon the manual reactor scram at 2305 EDT on September 13, 2015, Reactor Protection System (RPS) Level 3 actuated and Primary Containment Isolation System (PCIS) Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for these actuations is 10 CFR 50.72(b)(3)(iv)(A). The applicable reporting criterion for the manual closure of the inboard and outboard main steam isolation valves at 2345 EDT on September 13, 2015, is also 10 CFR 50.72(b)(3)(iv)(A). In addition, the manual closures of all MSIV lead to a loss of condenser vacuum which resulted in the actuation of PCIS Group 1 at 0001 EDT on September 14, 2015, as expected. The applicable reporting criterion for this actuation is also 10 CFR 50.72(b)(3)(iv)(A). Upon reaching Level 3 at 1847 EDT on September 14, 2015, PCIS Groups 4, 13 and 15 actuated as expected. The applicable reporting criterion for this actuation is 10 CFR 50.72(b)(3)(iv)(A). The licensee informed the NRC Resident Inspector. Notified the R3DO (Stone).
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 51003	24 February 2015 22:02:00	10 CFR 50.73(a)(1), Submit an LER	60-Day Optional Telephone Notification of Invalid System Actuation Due to Motor Generator Output Breaker Trip	On February 24, 2015, at approximately 1702 CDT, while the plant was in cold shutdown, power was lost on the Division 1 reactor protection system (RPS) bus. This event resulted in the automatic closure of the Division 1 primary containment isolation valves in the residual heat removal (RHR) and reactor water cleanup systems. Additionally, the primary containment atmospheric monitoring system automatically actuated, and ventilation systems in the fuel building, auxiliary building, and control building shifted to emergency mode. The closure of the isolation valves in the residual heat removal system caused an automatic trip of the 'A' RHR pump, which had been in the shutdown cooling alignment. The equipment response to the isolation signal was as expected. This event is being reported in accordance with 10 CFR 50.73(a)(1) as an invalid actuation of the Division 1 primary containment isolation system. The isolation was promptly diagnosed as having resulted from a trip of the output breaker of the RPS motor generator (MG) set 'A,' and not from a valid signal. Operators implemented the appropriate response procedures to align power to the bus via the alternate source, and began restoring the affected systems. The 'A' RHR pump was re-started within twelve minutes, during which time coolant temperature increased approximately seven degrees to a maximum of approximately 100F. Other affected systems were restored over the next few hours. The causal analysis concluded that the MG set output breaker tripped due to an overly conservative setpoint on the overvoltage trip relay. The low trip setpoint was a latent condition that had existed since the output voltage was raised in 1988 at the recommendation of the vendor, but at which time the trip setpoint was not changed. To correct this condition, the MG overvoltage trip setpoint was raised to restore adequate operating margin to the normal MG output voltage. At the time of the event, the plant was in MODE 5 with the reactor cavity flooded to greater than 23 feet above the vessel flange. The shutdown cooling system was promptly restored to service. This event was of minimal safety significance to the health and safety of employees and the public. The licensee has notified the NRC Resident Inspector.

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