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 Entered dateSiteRegionReactor typeEvent description
ENS 5624128 November 2022 08:38:00FermiNRC Region 3GE-4The following information was provided by the licensee via email: At 0400 EST on November 28, 2022, during the performance of Division 2 Residual Heat Removal (RHR) cooling tower fan operability and RHR Service Water valve lineup verification, it was reported that the Mechanical Draft Cooling Tower (MDCT) Fan 'B' was making a loud metallic noise. The cause of the metallic noise is unknown at this time. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the Division 2 Emergency Equipment Cooling Water (EECW) system. The EECW system cools various safety related components including the High Pressure Coolant Injection (HPCI) system room cooler. An unplanned HPCI inoperability occurred based on inoperable cooling water to the HPCI room cooler, per LCO 3.0.6. Investigation into the Division 2 MDCT Fan 'B' abnormal noise is in progress. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. The NRC Resident Inspector has been notified.
ENS 5622013 November 2022 04:01:00CooperNRC Region 4GE-4The 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 5617422 October 2022 17:36:00CooperNRC Region 4GE-4The following information was provided by the licensee via fax: During Mode 5 Refueling operations, while attempting to establish flow through the Fuel Pool Cooling system filter demineralizers, an air operated valve to a radioactive waste tank failed to close automatically. This caused the Fuel Pool Cooling system to pump water from the Skimmer Surge Tanks (SST) to the radioactive liquid waste system. In response to the loss of inventory from the SSTs, the Control Room operating crew started Core Spray Pump A to restore normal operating level In the SST. This prevented the loss of the Fuel Pool Cooling/Alternate Decay Heat Removal system which was the only in service system meeting the safety function of decay heat removal. Core Spray Pump A was used for Injection for less than 3 minutes. This is reportable as a discharge of ECCS into the RCS in response to an event, but not part of a pre-planned sequence under 10 CFR 50.72(b)(2)(iv)(A) and actuation of a specified system under 10 CFR 50.72(b)(3)(Iv)(A). The resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Licensee reported approximately 6000-7000 gallons of water was injected into the RCS. The stuck open air operated valve was closed. Proceeding with refueling outage operations.
ENS 5616818 October 2022 22:08:00Browns FerryNRC Region 2GE-4The following information was provided by the licensee via email: On 10/18/2022 at 1440 CDT, Browns Ferry Unit 3 declared both trains of standby liquid control (SLC) inoperable due to acceptance criteria failure of 3-SI-3.1.7.6, 'Standby Liquid Control System ATWS Equivalency Calculation for Newly Established Pump Flow Rate.' The purpose of this surveillance is to ensure the anticipated transient without scram (ATWS) calculation criteria is met after each pump flow test. Chemistry performed the surveillance following pump flow testing and the requirement for equivalency calculation failed low with a result of less than 1.0. CR 1810303 documents this condition in the corrective action program. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(A), 10 CFR 50.72(b)(3)(v)(C), and 10 CFR 50.72(b)(3)(v)(D). This condition is also reportable as a 60-day written report in accordance with 10 CFR 50.73(a)(2)(v)(A),10 CFR 50.73(a)(2)(v)(C), and 10 CFR 50.73(a)(2)(v)(D). The NRC Resident Inspector has been notified of this event. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officer's report guidance: The plant entered an 8 hour limiting condition for operation based on the above. The condition was resolved at 2053 CDT when the system was restored to normal operation.
ENS 5616314 October 2022 18:03:00LimerickNRC Region 1GE-4The following information was provided by the licensee via email: Unit 1 High Pressure Coolant Injection (HPCI) was declared inoperable due to an inadvertent division 2 isolation signal and subsequent valve closure. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. The NRC Resident will be notified.
ENS 5615812 October 2022 09:21:00FitzPatrickNRC Region 1GE-4The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: On 10/08/22, a non-supervisory employee violated the station's FFD policy. The individual's site access has been terminated.
ENS 561527 October 2022 05:41:00CooperNRC Region 4GE-4The following information was provided by the licensee via fax: On 10/7/2022 at 0050 CDT, a potentially contaminated individual was transported off-site via ambulance to a local hospital. Due to the nature of the medical condition, an initial on-site survey for radioactive contamination was not performed prior to transport. Prior to arrival at the hospital, it was confirmed the individual and (the individual's) clothing were not radiologically contaminated. Follow-up surveys performed by radiation technicians identified no radiological contamination of the ambulance and response personnel. This event is being reported per 50.72(b)(3)(xii) - 'Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.' The NRC Resident Inspector has been notified.
ENS 561384 October 2022 10:27:00BrunswickNRC Region 2GE-4The following information was provided by the licensee via email: This 60-day optional telephone notification is being made in lieu of an LER submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for invalid actuations of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 0628 Eastern Daylight Time (EDT) on August 6, 2022, an invalid actuation of group 6 Primary Containment Isolation Valves (PCIVs) (i.e., containment atmospheric control/monitoring and post accident sampling isolation valves) occurred. The group 6 isolation signal resulted from the reactor building ventilation radiation monitor `A' channel exceeding the setpoint value. This condition recurred at approximately 1305 EDT on August 12, 2022. In both instances, the `B' channel, located in the same plenum, remained steady and below the setpoint value through the entire event. This, along with readings made by radiation protection technicians, confirmed that there were no actual high radiation conditions in the reactor building exhaust in either instance. Following each invalid actuation, upon returning unit 2 reactor building ventilation to service, the `A' channel readings returned to be consistent with the `B' channel. It was determined that these invalid actuations likely resulted from degradation of circuit components associated with the radiation monitor. The `A' channel radiation monitor was replaced on September 22, 2022. During these two events, the PCIVs functioned successfully and the actuations were complete. The actuations were not initiated in response to actual plant conditions, they were not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, these events have been determined to be invalid actuations. These events did not result in any adverse impact to the health and safety of the public.
ENS 5611926 September 2022 05:41:00FitzPatrickNRC Region 1GE-4

The following information was provided by the licensee via email: At 0001 EDT on September 26, 2022, James A. FitzPatrick (JAF) removed the generator from service as part of a planned shutdown for refueling. At 0306 EDT, with the mode switch in Startup/Hot Standby and inserting rods, JAF experienced a spurious Scram and closure of seven out of eight main steam isolation valves (MSIV's). The reactor protection system (RPS) actuated during the event, resulting in all control rods being fully inserted. The cause of the closure of MSIV's and the Scram is being investigated. This condition is being reported as a four-hour NRC report per 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation, and as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the safety system actuation based on the multiple main steam isolation valves closing on an isolation signal. There was no impact to the health and safety of the public. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 10/4/22 AT 2047 EDT FROM ANDREW WEAVER TO KERBY SCALES * * *

The following update was provided by the licensee via email: This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A) for Reactor Protection System (RPS) actuation along with Main Steam Isolation Valves (MSIV) system actuation. An analysis of reactor criticality was performed for the period of time prior to the RPS actuation event. Operators were inserting control rods per the shutdown Reactivity Management Plan. The Intermediate Radiation Monitoring (IRM) readings preceding the scram signal demonstrate a negative reactivity direction without control rod movement. The analysis concluded that the reactor was subcritical when RPS was actuated. The NRC Resident Inspector has been notified. Notified R1DO (Young).

ENS 5611317 September 2022 13:06:00HatchNRC Region 2GE-4The following information was provided by the Southern Nuclear Company via email: At 2257 EDT on 09/16/2022, it was determined that there was a programmatic vulnerability of the Fleet FFD program. Specifically, it was determined that some individuals were not placed into the follow-up pool for additional screening when required by the program. All identified personnel were in the random FFD pool, and were subject to the behavioral observation program. This is reportable in accordance with 10CFR26.719(b)(4) for all Units and 10CFR26.417(b)(1) for Vogtle Units 3 and 4. The NRC Resident Inspectors have been notified. See EN#s 56112, 56114, and 56115.
ENS 5605418 August 2022 01:20:00FermiNRC Region 3GE-4

The following information was provided by the licensee via email: At 2108 EDT on August 17, 2022 the Division 2 Mechanical Draft Cooling Tower (MDCT) fans were declared inoperable due to failure of the over speed fan brake inverter. The brakes prevent fan over speed from a design basis tornado. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the Division 2 Emergency Equipment Cooling Water (EECW) system. The Division 2 EECW system cools various safety related components, including the High Pressure Coolant Injection (HPCI) room cooler and Division 2 Control Center HVAC (CCHVAC) chiller. An unplanned HPCI inoperability occurred based on a loss of the HPCI Room Cooler. At the time of the event, Division I CCHVAC was inoperable for maintenance (but was running for a maintenance run) and the event caused an inoperability of Division 2 CCHVAC. This resulted in an inoperability of both divisions of CCHVAC. Failure of the Division 2 MDCT Fan brake inverter occurred due to a trip of the DC input breaker. The breaker was reset at 2128 EDT restoring Division 2 UHS Operability. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfilment of the safety function of structures or systems that are needed to mitigate the consequences of an accident based on a loss of a single train safety system and loss of both divisions of a safety system. The Senior NRC Resident Inspector has been notified

  • * * RETRACTION ON 09/08/2022 AT 0856 EDT FROM JEFF MYERS TO MIKE STAFFORD * * *

The following information was provided by the licensee via email: On 8/17/22 at 2108 EDT the Division 2 (Div. 2) mechanical draft cooling tower (MDCT) brake inverter input breaker tripped for an unknown cause. The result of the loss of power was the inoperability of the MDCT fan brakes which impacts the ultimate heat sink (UHS) (TS 3.7.2). The UHS cascades to the EECW (emergency equipment cooling water) (TS 3.7.2) which is a support system for Div. 2 CCHVAC (Control Cell) Chiller A/C system (TS 3.7.4). This resulted in the inoperability of the Div. 2 CCHVAC Chiller. The cause for the breaker to trip is an intermittent electrical transient. Immediate corrective action was to reset the breaker, and the long-term action is to implement a modification to mitigate susceptibility to voltage variations. Div. 1 has implemented this long-term mod and no unexpected trips have occurred to date. Div. 1 CCHVAC Chiller was previously inoperable from equipment issues which was repaired, and the unit was in service for a 24-hour confidence run. Although licensed personnel had not completed the administrative actions for documenting operability during the 24-hour confidence run to monitor parameters, the (post maintenance test) PMT related to the maintenance was already completed, which included a 4-hour run in accordance with surveillance 24.413.01, Div. 1 and Div. 2 Chilled Water Pump and Valve, to verify normal operation and motor current. These PMT's were completed prior to the identified inoperability of the Div. 2 UHS due to the tripped breaker on the brake power supply. At the time of the MDCT brake inverter trip, the Operations' Senior License and the Night Shift Manager were aligned that, although still operating as part of the 24-hour confidence run, the unit was in service and capable of performing its safety function, but the administrative tasks were not completed, the Limited Condition of Operation (LCO) sheet had not been cleared, and no log entries were made. Since the Div. 1 Chiller was, in fact, operable at the time of the trip of the breaker on the inverter, this would allow the use of Technical Specification (TS) 3.0.9 'Barriers'. Per Operations Department Expectation (ODE)-12 `LCOs' (standard guidance and expectations for preparing and implementing an LCO), Operations determined that the MDCT brakes are barriers to a tornado event and TS 3.0.9 could be utilized. By invoking TS 3.0.9, as long as all other supported systems in the other division are operable, Div. 2 supported systems relying upon the UHS can remain operable and the Automatic Depressurization System (ADS) and Reactor Core Isolation Cooling (RCIC) system can be used as backup to the High Pressure Coolant Injection (HPCI) system. Based on this information, there was no loss of safety function with CCHVAC A/C system or HPCI. Therefore, the NRC non-emergency 10CFR50.72(b)(3)(v)(D) report was not required and the NRC report 56054 can be retracted. The NRC Resident Inspector has been notified. Notified R3DO (Orlikowski)

ENS 5601625 July 2022 13:30:00BrunswickNRC Region 2GE-4The following information was provided by the licensee via email: At 1058 Eastern Daylight Time (EDT) on July 25, 2022, it was determined that a non-licensed supervisor failed a test specified by the FFD testing program for the substance alcohol. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified.
ENS 5599715 July 2022 23:41:00BrunswickNRC Region 2GE-4The following information was provided by the licensee via email: At 2020 Eastern Daylight Time (EDT) on July 15, 2022, the HPCI System was declared inoperable. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The Reactor Core Isolation Cooling (RCIC) System and Automatic Depressurization System (ADS) were operable during this time. HPCI availability was restored at 2023. Additional investigation is in-progress. There was no impact on the health and safety of the public or plant personnel. Unit 2 is not affected by this event. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: HPCI is considered inoperable but available at this time, resulting in a 14-day Shutdown LCO (Limiting Condition for Operation), due to the HPCI inoperability.
ENS 5599212 July 2022 17:25:00Browns FerryNRC Region 2GE-4The following information was provided by the licensee via fax or email: At 0917 CDT on 7/12/2022, during the performance of U1 (Unit 1) High Pressure Coolant Injection (HPCI) rated flow test, the 1-FCV-73-19 (HPCI governor valve) failed to operate as expected. This condition results in U1 HPCI being inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The Automatic Depressurization System (ADS) and Reactor Core Isolation Cooling (RCIC) system remain operable. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. U1 entered TS LCO 3.5.1 Condition C, 14-day Shutdown LCO (Limiting Condition for Operation), due to the HPCI inoperability.
ENS 559817 July 2022 22:49:00CooperNRC Region 4GE-4The following information was provided by the licensee via phone and fax: On 7/7/2022, at 0740 CDT, the National Weather Service reported to Cooper Nuclear Station that the NAWAS ((National Warning System)) radio tower near Shubert, Nebraska would neither transmit nor receive. The Shubert Tower transmitter activates the EAS ((Emergency Alert System))/Tone Alert Radios used for public notification. Additional information from the National Weather Service received 7/7/2022 at 1601 (CDT) determined that the Shubert Tower transmitter is non-functional and would not likely be repaired within 24 hours. The backup notification system has been verified to be available throughout this period. This is considered to be a major loss of the Public Prompt Notification System capability. The primary notification system is not expected to be restored to service within 24 hours, and therefore this condition is reportable under 10 CFR 50.72(b)(3)(xiii), since the backup alerting methods do not meet the primary system design objective. The backup notification system is available to use for notifications if needed. The NRC Senior Resident has been informed. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The backup notification system has to be manually activated.
ENS 5596425 June 2022 01:00:00FermiNRC Region 3GE-4The following information was provided by the licensee via email: At 2338 EDT, on June 24, 2022, with the unit in Mode 1 at 100 percent power, the reactor automatically scrammed due to an RPS actuation following a Main Turbine Trip. The cause of the turbine trip is not known at this time. The scram was not complex, with systems responding normally post-scram. Operations responded and stabilized the plant. Reactor water level has been recovered and maintained at the normal level. Decay Heat is being removed by the Main Steam system to the main condenser using the Turbine Bypass Valves. All Control Rods inserted into the core. The transient occurred with no surveillances or activities in progress. Investigation into the cause of the Turbine Trip is in progress. 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). The low reactor water level caused an isolation of Primary Containment (Groups 4/13/15) as expected. The Primary Containment Isolation Event is being reported under 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified.
ENS 5590923 May 2022 21:01:00SusquehannaNRC Region 1GE-4The following information was provided by the licensee via email: At 1716 hours EDT on May 23, 2022, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed. Unit 1 reactor was being operated at approximately 100 percent (Rated Thermal Power) RTP. The Control Room received indication that both divisions of (Reactor Protection System) RPS actuated from (Reactor Pressure Vessel) RPV high pressure signals and all control rods fully inserted. The Main Turbine bypass valves opened automatically to control reactor pressure. Reactor water level lowered to -42 inches causing Level 3 and Level 2 isolations. (High Pressure Coolant Injection) HPCI (Emergency Core Cooling System) ECCS actuation occurred as designed at -38 inches and injected to the Reactor Vessel. No other ECCS system actuations occurred. (Reactor Core Isolation Cooling) RCIC automatically initiated as designed at -30 inches. The Operations crew subsequently maintained reactor water level at the normal operating band using Feedwater pumps. The reactor is currently stable in Mode 3. An investigation is in progress into the cause of the Automatic SCRAM. The NRC Senior Resident Inspector was notified. A voluntary notification to (Pennsylvania Emergency Management Agency) PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10 CFR 50.72(b)(2)(iv)(A) & 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A).
ENS 5590723 May 2022 11:23:00CooperNRC Region 4GE-4The following information was provided by the licensee via email: On May 23, 2022, at 0455 CST, Cooper Nuclear Station experienced a spike in Secondary Containment differential pressure which exceeded the Technical Specifications Surveillance Requirements 3.6.4.1.1 limit of -0.25 inches of water gauge. Secondary Containment differential pressure restored to Technical Specification limits within two minutes and further investigation is ongoing. This unplanned Secondary Containment inoperability constitutes a condition reportable under 10CFR50.72(b)(3)(v)(C) and (D). The NRC Senior Resident Inspector has been informed.
ENS 5589916 May 2022 19:51:00Peach BottomNRC Region 1GE-4The following information was provided by the licensee via fax: Unit 2 experienced multiple electrical transients resulting in a Group I Primary Containment Isolation Signal (PCIS) isolation and subsequent unit reactor scram. Low reactor water level during the automatic scram caused PCIS Group II and III isolation signals. Following the PCIS Group I isolation, all main steam lines isolated. All control rods inserted and all systems operated as designed. The following additional information was obtained from the licensee via phone in accordance with Headquarters Operations Officers Report Guidance: Peach Bottom Unit 2 automatically scrammed from 100 percent power due to an electrical transient and subsequent PCIS Group I isolation (Main Steam Isolation Valve closure). Unit 2 lost main feedwater due to the PCIS Group I isolation, however, all other systems responded as expected following the scram. High Pressure Coolant Injection is maintaining pressure control while Condensate Pumps are maintaining inventory. The unit is currently stable and in Mode 3. Peach Bottom Unit 3's Adjustable Speed Drives were impacted by the electrical transients and the unit reduced power to 98 percent power. The NRC Resident Inspector was notified.
ENS 5589411 May 2022 22:25:00FermiNRC Region 3GE-4The following information was provided by the licensee via email: During performance of High Pressure Coolant Injection (HPCI) Pump and Valve Operability surveillance in accordance with procedure 24.202.01, the turbine tripped without operator action. The plant was operating in Mode 1 at 10 percent power with the HPCI turbine running in a test mode at 5100 gpm with all surveillance criteria met. The surveillance was near completion at the point where the HPCI turbine is manually tripped. Before the manual trip was performed, the HPCI turbine tripped without operator action. Prior to performance of the surveillance, HPCI was provisionally operable with only satisfactory completion of Post Maintenance Testing (PMT) surveillance remaining to declare HPCI operable. HPCI surveillance testing was performed at low reactor pressure (165 psig) in Mode 2 satisfactorily. Investigation into the cause of this trip is in progress. HPCI has been declared inoperable from the time of release of the surveillance. Reactor Coolant Isolation Cooling (RCIC) was verified to be operable prior to and after the surveillance in accordance with Technical Specifications 3.5.1 condition E.1. This report is being made pursuant to 10 CFR 50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5587129 April 2022 20:44:00FitzPatrickNRC Region 1GE-4

The following information was provided by the licensee via email: At 1251 EDT on April 29, 2022, while troubleshooting the failure of the High Pressure Coolant Injection (HPCI) Exhaust Drain Pot High Level Alarm to clear, it was discovered that the High Pressure Coolant Injection exhaust line condensate drain system was not functioning as designed to support removal of condensate from the turbine exhaust. This resulted in some water accumulation in the turbine casing. Subsequently, the High Pressure Coolant Injection System was declared inoperable. As a result, this condition is being reported under 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented fulfillment of the safety function at the time of discovery.

  • * * RETRACTION ON 07/15/22 AT 1943 EDT FROM EVAN THOMPSON TO LLOYD DESOTELL * * *

A technical evaluation of this event was performed and concluded that the HPCI system would have been operable with this condition. If HPCI turbine actuated with the estimated amount of condensate accumulated in the casing and connecting piping, it would have performed its safety function; the HPCI Turbine Exhaust Rupture Disc would not have been challenged by calculated peak pressures; and calculated water hammer loads were within specified load capacities of the turbine flange, downstream piping, struts, snubber, and spring hanger. Based on this, the condition reported in EN 55871 is being retracted. Notified R1DO (Bickett)

ENS 5585926 April 2022 13:13:00BrunswickNRC Region 2GE-4The 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 558182 April 2022 15:10:00Browns FerryNRC Region 2GE-4

The following information was provided by the licensee via fax: At 1345 CDT, Browns Ferry declared a Notification of Unusual Event due to a fire at the 3B Reactor Feedwater Pump within the Turbine Building which was not extinguished within 15 minutes. Subsequently, the fire was extinguished at 1402 CDT. Unit 3 remains in Mode 1 at approximately 9.5 percent rated thermal power (RTP). Unit 1 and 2 remain at 100 percent RTP and unaffected. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The fire began at 1332 CDT. It is believed that the fire was in the oil system of the Feedwater Pump. The fire was extinguished by the on-site fire brigade. No off-site assistance was requested. The Unusual Event was declared under Emergency Action Level HU-4. The licensee notified the NRC Resident Inspector and required State and local government agencies. Unit 3 is currently stable. Notified DHS-SWO, FEMA Ops Center, and CISA Central Watch Officer, and FEMA NWC, DHS NRCC THD Desk, and DHS NuclearSSA via email.

  • * * UPDATE FROM CHASE HENSLEY TO DONALD NORWOOD AT 1650 EDT ON 4/2/2022 * * *

The Notification of Unusual Event was exited at 1544 CDT. Notified R2DO (Miller), IR-MOC (Kennedy), NRR-EO (Felts), DHS-SWO, FEMA Ops Center, and CISA Central Watch Officer.

ENS 557809 March 2022 23:20:00BrunswickNRC Region 2GE-4

The following information was provided by the licensee: At 2013 EST on March 9, 2022, the HPCI System was declared inoperable following evaluation of routine HPCI surveillance testing data indicating that the required response time for reaching rated conditions was not met. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The Reactor Core Isolation Cooling (RCIC) System and Automatic Depressurization System (ADS) are operable. There was no impact on the health and safety of the public or plant personnel. Investigation is in-progress to determine the cause. Unit 1 is not affected by this event. Unit 1 is in a refueling outage. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 05/04/22 AT 1135 EDT FROM CHARLIE BROOKSHIRE TO DAN LIVERMORE * * *

The following information was provided by the licensee via email: At 20:13 EST on March 9, 2022, the HPCI System was declared inoperable following evaluation of routine HPCI surveillance testing data indicating that the required response time for reaching rated flow and pressure was not met. Subsequent to this, it was determined that the required response time was overly conservative for assuring the safety function of the system could be fulfilled. The required response time was revised. The operability determination for this event has been updated indicating that system operability was never lost for this event. There was not a condition that could have prevented the system from fulfilling the safety function. The NRC Resident Inspector has been notified. Notified R2DO (Miller).

ENS 5575624 February 2022 14:35:00BrunswickNRC Region 2GE-4The 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 557324 February 2022 20:30:00FermiNRC Region 3GE-4The following information was provided by the licensee via email: At 1700 EST, on February 4, 2022 with the unit in Mode 1 at 58 percent power, the reactor automatically scrammed due to low Reactor water level due to a transient on the Feedwater System while preparing to shutdown for a refueling outage. The scram was not complex, with systems responding normally post-scram. Operations responded and stabilized the plant. Reactor water level has been recovered and maintained at normal level. Decay Heat is being removed by the Main Steam system to the main condenser using the Turbine Bypass Valves. All Control Rods inserted into the core. The transient occurred while in the process of removing the South Reactor Feed Pump from service. While reducing speed on the South, the North Reactor Feed Pump increased in speed and tripped on low suction. The plant was preparing to shut down for a refueling outage when the trip occurred. 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, in preparation of plant shutdown, Primary Containment De-Inerting was in progress. The low Reactor water level caused an isolation of Primary Containment (Groups 4/13/15). The Primary Containment Isolation Event is being reported under 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC resident has been notified.
ENS 5571527 January 2022 15:07:00CooperNRC Region 4GE-4

The Licensee provided the following information via email: On January 27, 2022 at 1038 CST, with Cooper Nuclear Station in Mode 1, 100 percent power, the meteorological tower primary and backup data acquisition system failed, which resulted in a loss of meteorological data to the plant. Information technology personnel investigated and restored the primary system to service. Meteorological data to the plant was restored at 1105 CST on January 27, 2022. This notification Is being made due to a loss of emergency assessment capability In accordance with 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been Informed.

  • * * RETRACTION ON FEBRUARY 23, 2022 AT 1658 EST FROM LINDA DEWHIRST TO LLOYD DESOTELL * * *

The following information was provided by the licensee via fax: This notification is being made to retract event EN 55715 that was reported on January 27, 2022. Based on further investigation, the Emergency Plan and Emergency Plan Implementing Procedures provide acceptable alternative methods for performing emergency assessments that are in addition to the data obtained from the primary and backup meteorological tower information. It was determined that no actual or potential major loss of emergency assessment capability existed per 10 CFR 50.72(b)(3)(xiii). This is consistent with NUREG 1022, Revision 3, Supplement 1 and NEI 13-01, Revision 0. The NRC Resident Inspector has been notified of the retraction. Notified R4DO (O'Keefe)

ENS 5570616 January 2022 06:41:00Browns FerryNRC Region 2GE-4The following information was provided by the licensee via email: On 1/15/2022 at 2320 (EST) during the planned F108 outage on Browns Ferry Nuclear Plant Unit 1, personnel entered the Unit 1 Drywell for leak identification. Personnel discovered a through-wall piping leak on a 3/4 inch test line upstream of the test valve. This 3/4 inch test line is located on the RHR Shutdown Cooling & RHR Return Line to the reactor vessel and is classified as ASME Code Class 1 Piping. This constitutes an 8-hour NRC notification in accordance with 10 CFR 50.72(b)(3)(ii)(A) - Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The NRC resident has been notified.
ENS 5570313 January 2022 12:03:00CooperNRC Region 4GE-4The following information was provided by the licensee via fax: On 1/13/2022 at 0806 CST, Nebraska Public Power District was notified by Atchison County Missouri of a spurious actuation of (Cooper Nuclear Station) (CNS) Emergency Siren 2113 near Rockport, Missouri from approximately 0800 to 0805 CST. Nebraska Public Power District will issue a press release for this event. The CNS Emergency Alert System (EAS) was not activated. This condition is reportable under 10 CFR 50.72(b)(2)(xi) for any event or situation for which a news release is planned or notification to other government agencies has been or will be made which is related to heightened public or government concern. The NRC Senior Resident Inspector has been informed.
ENS 556853 January 2022 17:01:00LimerickNRC Region 1GE-4The following information was provided by the licensee via email: On January 3, 2022, a Licensed Reactor Operator violated the station's Fitness for Duty policy. The employee's unescorted access to Limerick Generating Station has been terminated in accordance with station procedures. The event was determined to be reportable under 10 CFR 26.719(b)(2)(ii). The NRC Resident Inspector has been notified.
ENS 5567929 December 2021 19:16:00HatchNRC Region 2GE-4This following information was conveyed by the licensee via phone and email: At 1552 EST on 12/29/21, with Unit 1 in Mode 1 at 90 percent power, the reactor was manually tripped due to reactor pressure perturbations. The cause of the reactor pressure perturbations is under investigation. Additionally, closure of (containment isolation valves) CIVs in multiple systems occurred during the trip as a result of reaching the actuation setpoint on reactor water level. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via condensate / feedwater. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not affected. 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). It is also reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of CIVs. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5566016 December 2021 14:57:00Browns FerryNRC Region 2GE-4
  • 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 556276 December 2021 18:14:00BrunswickNRC Region 2GE-4

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 5561630 November 2021 16:22:00SusquehannaNRC Region 1GE-4At 1254 EST on November 30, 2021, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed during Turbine Valve Cycling surveillance activities. Unit 1 reactor was being operated at approximately 80 percent rated thermal power with turbine valve cycling surveillance activities in progress. The Control Room received indication that both divisions of RPS (reactor protection system) actuated from turbine valve closure signals and all control rods fully inserted. The Main Turbine was manually tripped, and turbine bypass valves opened automatically to control reactor pressure. Reactor water level lowered to -35 inches causing Level 3 and Level 2 isolations. No ECCS (emergency core cooling systems) actuations occurred. RCIC (reactor core isolation cooling) automatically initiated as designed at -30 inches. The Operations crew subsequently maintained reactor water level at the normal operating band using Feedwater pumps and RCIC was placed in a standby lineup. The reactor is currently stable in Mode 3. An investigation is in progress into the cause of the turbine valve closure signals. The NRC Senior Resident Inspector was notified. A voluntary notification to PEMA (Pennsylvania Emergency Management Agency) will be made. This event requires a 4-hour ENS notification in accordance with 10CFR50.72(b)(2)(iv)(B) and an 8-hour ENS notification in accordance with 10CFR50.72(b)(3)(iv)(A). Unit 2 was not affected and remains at 100 percent power, Mode 1.
ENS 5559319 November 2021 00:50:00FitzPatrickNRC Region 1GE-4On November 18, 2021, during the performance of High Pressure Coolant Injection (HPCI) surveillance testing, 23MOV-19 (HPCI PUMP DISCH TO REACTOR INBD ISOL VALVE) did not go open as expected while performing the sensed low water level portion of the test. The ability to manually open 23MOV-19 from the control room was unaffected as such, the HPCI system remained available for use. Failure of 23MOV-19 to open automatically prevents the HPCI system from performing its safety function as such this condition renders HPCI inoperable but available and is being reported as a condition that could have prevented the fulfillment of the safety function of a system needed to mitigate the consequences of an accident per 10CFR50.72(b)(3)(v)(D). HPCI inoperable placed the licensee in a 14-day limiting condition for operation for Tech Spec 3.5.1.c. The NRC Resident Inspector was notified.
ENS 5557514 November 2021 08:50:00Peach BottomNRC Region 1GE-4At 0525 EST, November 14, 2021, "Unit 2 was manually scammed by operations due to lowering main condenser vacuum. This resulted in PCIS (process control and instrumentation 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 555583 November 2021 13:38:00FermiNRC Region 3GE-4At 0241 (EDT) on June 3, 2021, during performance of a High Pressure Coolant Injection (HPCI) Condensate Storage Tank (CST) level functional surveillance, the HPCI torus suction inboard isolation valve was slow to open during swap of suction from the CST to the Torus. On June 9, 2021, it was determined that as a result of the June 3, 2021, slow swap condition, TS 3.3.5.1 Required Action D.1 to declare HPCI inoperable within 1 hour was applicable due to inoperable CST low level instrumentation channels. At 1817 (EDT) on June 3, 2021, HPCI suction was swapped to the torus, making TS Required Action D.1 no longer applicable. Reactor Core Isolation Cooling (RCIC) was available throughout this condition. At 0900 (EDT) on November 3, 2021, it was determined that an NRC event report due to HPCI inoperability should have been made. This event is being reported as a late 8-hour non-emergency notification pursuant to 10CFR50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. The cause of the slow valve opening was later determined to be corrosion products on contacts of a relay in the CST low level instrumentation logic. On June 4, 2021 at 1451 (EDT), the HPCI CST Level Functional Test was completed Satisfactorily, restoring HPCI Instrumentation to Operable. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5552214 October 2021 19:27:00FermiNRC Region 3GE-4

At 1320 EDT, during a Traversing In-Core Probe (TIP) run for a scheduled Local Power Range Monitors (LPRM) calibration, it was reported to the Main Control Room that TIP A would not fully retract to the In-Shield position. With TIP A unable to fully retract to the In-Shield position the TIP A Ball Valve was declared Inoperable due to not being able to close and meet its safety function in that configuration. Furthermore the TIP A Shear Valve was previously declared Inoperable due to the Firing Fuses being removed. With the two valves Inoperable the penetration could not be isolated and Primary Containment boundary isolation could not be established. TIP A was subsequently manually hand cranked and placed back into its In-Shield position at 1333 EDT restoring TIP A Ball Valve Operable. This report is being made pursuant to 10CFR50.72(b)(3)(v)(C) based on control the release of radioactive material. The Senior NRC Resident Inspector has been notified.

  • * * RETRACTION ON NOVEMBER 24, 2021 AT 1232 EST FROM LEVI SMITH TO BRIAN P. SMITH * * *

The purpose of this notification is to retract a previous report made on October 14, 2021 (EN 55522). At 1320 EDT on October 14, 2021 while performing Traversing In-Core Probe (TIP) Machine Gain Adjustment in support of Local Power Range Monitor (LPRM) calibration, an unplanned inoperability of the TIP 'A' Primary Containment Isolation Valve (PCIV) was reported pursuant to 10CFR50.72(b)(3)(v)(C) by EN 55522. On October 14, it was reported to the Main Control Room that TIP 'A' would not fully retract to the In-Shield position. With TIP 'A' unable to fully retract to the In-Shield position, the TIP 'A' Ball Valve PCIV was declared Inoperable due to not being able to close and meet its safety function in that configuration. The TIP 'A' Shear Valve PCIV was previously declared inoperable due to firing fuses being removed. Further investigation determined that a "FAULT: MOVEMENT LIMITED" error was received. This TIP error condition did not present a primary containment isolation issue in the event of a primary containment isolation signal. The Automatic TIP Control Unit (ATCU) is designed to command the TIP drive mechanism to continuously retract a TIP probe to the in-shield position in the event of a containment isolation signal with this condition. In the event of a containment isolation signal, the TIP machine would withdraw the TIP detector back to the in-shield position and the TIP A ball valve PCIV would have closed to perform its safety function. Therefore, the inoperability of TIP 'A' ball valve reported under criterion 10CFR50.72(b)(3)(v)(C) was not met, and EN 55522 is hereby retracted. The NRC Resident Inspector has been notified. Notified R3DO (Peterson)

ENS 5551411 October 2021 17:05:00SusquehannaNRC Region 1GE-4At 1321 EDT on October 11, 2021, Susquehanna Steam Electric Station Unit 2 reactor automatically scrammed due to a trip of the Main Turbine. Unit 2 reactor was being operated at approximately 95 percent RTP (rated thermal power) with no evolutions in progress. The Control Room received indication of a Main Turbine trip with both divisions of RPS (Reactor Protection System) actuated and all control rods inserted. Turbine bypass valves opened automatically to control reactor pressure and subsequently failed open causing the reactor to depressurize. When reactor pressure reached approximately 560 psig, the operations crew manually closed the Main Steam Isolation Valves (MISVs) to stop the depressurization. Reactor water level lowered to -31 inches causing Level 3 (+13 inches) isolations. No (automatic) ECCS (Emergency Core Cooling System) actuations occurred. HPCI (High Pressure Coolant Injection) and RCIC (Reactor Core Isolation Cooling) were manually initiated to control reactor water level. The Operations crew subsequently maintained reactor water level at the normal operating band using RCIC and reactor pressure was controlled with HPCI in pressure control mode and main steam line drains. The Reactor Recirculation Pumps tripped as designed on EOC-RPT (end of cycle recirculation pump trip). The reactor is currently stable in Mode 3. An investigation into the cause of the turbine trip is underway. The NRC Resident Inspector was notified. A voluntary notification to PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10 CFR 50.72(b)(3)(iv)(A).
ENS 5548423 September 2021 18:46:00LimerickNRC Region 1GE-4During planned testing of the Unit 1 HPCI (high pressure coolant injection) system, flow controller oscillations occurred which prevented successful completion of the surveillance test. Operators secured Unit 1 HPCI and declared the system inoperable. HPCI inoperable placed the licensee in a 14-day limiting condition for operation that was extended to 30 days after their risk-informed completion time evaluation was done. The licensee has notified the NRC Resident Inspector.
ENS 554488 September 2021 08:40:00HatchNRC Region 2GE-4At 0159 EDT on 09/08/2021, the HPCI pump discharge valve failed to reopen during a valve surveillance, resulting in the HPCI system being declared INOPERABLE. HPCI does not have a redundant system; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). Reactor Core Isolation Cooling system and low pressure Emergency Core Cooling Systems were OPERABLE during this time. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5542724 August 2021 16:51:00FitzPatrickNRC Region 1GE-4During an extent of condition review of DC control circuits, it was identified there are additional unprotected DC control circuits which are routed between separate Appendix R fire areas. A postulated fire in one area can cause a short circuit and potentially result in secondary fires or cable fires in other fire areas where the cables are routed. The secondary fires or cable failures degrade the degree of separation for redundant safe shutdown trains and are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B). Compensatory actions for affected fire areas have been implemented. Design modifications in the affected control circuits are being developed and will be scheduled to correct this condition.
ENS 5542322 August 2021 12:10:00FermiNRC Region 3GE-4At 0529 EDT on August 22, 2021, HPCI ((High Pressure Coolant Injection System)) was declared inoperable due to receiving the HPCI Inverter Circuit Failure annunciator. The cause of the annunciator was a fuse failure. The cause of the fuse failure is unknown at this time and is under investigation. Concurrent with the HPCI fuse failure was a similar fuse failure within the Division 2 EDG ((emergency diesel generators)) Load Sequencer which renders the Division 2 EDGs inoperable. Relation to the HPCI issue is unknown and is part of the investigation. The RCIC ((Reactor Core Isolation Cooling System)) was verified operable per Tech Spec 3.5.1 E.1. In addition, offsite circuits were verified operable per Tech Spec 3.8.1.B. Division 1 EDGs remain operable. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. There was no impact on the health and safety of the public or plant personnel. The Senior NRC Resident Inspector has been notified.
ENS 5542020 August 2021 12:53:00HatchNRC Region 2GE-4A licensed operator failed a pre-access authorization test specified by the FFD testing program test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 5540311 August 2021 11:32:00FermiNRC Region 3GE-4

At 0634 EDT on August 11, 2021 (high pressure coolant injection) HPCI was declared inoperable due to a pump flow controller problem. The cause of the controller problem is unknown at this time and is under investigation. (Reactor core isolation cooling) RCIC was verified operable per Tech Spec 3.5.1 E.1. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM WHITNEY HEMMINGWAY TO KAREN COTTON ON 10/6/2021 AT 1036 EDT * * *

The purpose of this notification is to retract a previous report made on August 11, 2021 (EN 55403). At 0634 EDT on August 11, 2021, an unplanned inoperability of the High Pressure Coolant Injection system (HPCI) was reported pursuant to 10 CFR 50.72(b)(3)(v)(D) by EN 55403. HPCI was declared inoperable due to receipt of an alarm associated with the pump flow controller. The HPCI system operating procedure states that HPCI should be declared inoperable when this alarm is received. The cause of the alarm, a loose transmitter connection, was identified and corrected. Following clearance of the alarm, HPCI was declared operable at approximately 1930 EDT on August 11, 2021. This alarm indicated a fault in the signal from the transmitter to the HPCI flow controller; in this case, the HPCI flow controller would have continuously called for maximum HPCI flow. The controller is configured with a high limiter to prevent an overspeed trip. An engineering evaluation of the event identified that HPCI was capable of performing its required safety functions while this alarm was present. The condition was that the HPCI flow controller would have continuously called for maximum HPCI flow upon HPCI initiation, however operators would be able to manually control HPCI flow upon HPCI initiation. Additionally HPCI would have run until Reactor Pressure Vessel (RPV) level reached Level 8 where it would trip until RPV level decreased to Level 2 then automatically restart. The licensee notified the NRC Resident Inspector. Notified R3DO (Peterson).

ENS 553943 August 2021 13:18:00HatchNRC Region 2GE-4At 1026 EDT on 8/3/21, with Unit 1 in MODE 1 at 100 percent power, the reactor automatically tripped due to low reactor water level. The low reactor water level condition was due to a loss of both reactor feed pumps. The cause of the loss of feed pumps is under investigation. Additionally, the low reactor water level resulted in the automatic actuation of High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systems, and Containment Isolation Valves (CIVs) in multiple systems. All safety systems responded normally. Operations responded and stabilized the plant. Reactor water level is being maintained via RCIC system. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not affected. 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). It is also reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the HPCI and RCIC systems and CIVs. There was no impact on the health and safety of the public or plant. The Licensee notified the NRC Resident Inspector. The Unit will proceed to Mode 4 while the cause of the loss of feed pumps is under investigation.
ENS 5537021 July 2021 20:50:00SusquehannaNRC Region 1GE-4At 1826 EDT on July 21, 2021, Susquehanna Steam Electric Station Unit 1 reactor automatically scrammed due to a trip of the Main Turbine. Unit 1 reactor was operating at 100 percent reactor power with no evolutions in progress. The Control Room received indication of a Main Turbine trip with both divisions of RPS (Reactor Protection System) actuated and all control rods inserted. The Reactor Recirculation Pumps tripped on EOC-RPT (end of cycle recirculation pump trip). Reactor water level lowered to +8 inches causing Level 3 (+13 inches) isolations. No ECCS (Emergency Core Cooling Systems) or RCIC (Reactor Core Isolation Cooling system) actuations occurred. The Operations crew subsequently maintained reactor water level at the normal operating band using Reactor Feed Water. The reactor is currently stable in Mode 3 with main condenser available. Investigation into the trip of the Main Turbine is in progress. The NRC Resident Inspector was notified. A voluntary notification to PEMA will be made. This event requires a 4 hour ENS notification in accordance with 10CFR50.72(b)(2)(iv)(B) and an 8 hour ENS notification in accordance with 10CFR50.72(b)(3)(iv)(A) and 10CFR50.72(b)(3)(iv)(B).
ENS 5535715 July 2021 21:36:00FermiNRC Region 3GE-4While preparing for the June 2021 Discharge Monitoring Report (DMR), Environmental was entering data per the lab results that were sent from Pace Analytical for the June DMRs. On June 1, 2021, a National Pollutant Discharge Eliminating System (NPDES) sample was collected at outfall 001A to test for copper, there is a NPDES permit condition to monitor for copper on a quarterly basis. The lab report was returned to Fermi Environmental on June 15, 2021. The results came back at 41.2 micrograms/liter. Fermi's NPDES permit maximum limit is 40 micrograms/liter for outfall 001A. Due to the June 1, 2021 sample exceeding the permit limit, a second sample was collected on June 21, 2021 as a verification sample and the copper results came back July 13, 2021. Those results came back at 5.9 micrograms/liter which is within the permit limit. Environmental was aware of the June 1, 2021 copper exceedance limit but failed to recognize the reporting requirement at the time of the discovery because it was thought that the exceedance would be reported through the DMR submittal. The June DMRs are due on July 20, 2021. At approximately 1740 EDT on July 15, 2021, a Fermi environmental engineer was preparing and reviewing the Discharge Monitoring Report and identified that a recent sample result for outfall 001A was outside of the NPDES permit limit for Copper. The Copper sample result was 41.2 micrograms/liter with a limit of 40 micrograms/liter. Subsequent discussions with Environmental personnel determined that this issue should be reported to the state of Michigan Department of Environment, Great Lakes and Energy (EGLE). A discussion is planned with EGLE on July 16, 2021. This notification is being made pursuant to 10 CFR 50.72(b)(2)(xi) based on the planned notification to EGLE. The licensee notified the NRC Resident Inspector.
ENS 553458 July 2021 20:07:00LimerickNRC Region 1GE-4This 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 5530311 June 2021 18:06:00HatchNRC Region 2GE-4

At 1710 EDT on June 11, 2021, a Technical Specification required shutdown was initiated at Plant Hatch Unit 1. Technical Specification Condition 3.4.4.B unidentified LEAKAGE increase not within limits, was entered due to a greater than 2 gpm increase in unidentified LEAKAGE within the previous 24 hour period in MODE 1. This specification was entered on June 11, 2021, at 1615 EDT with a REQUIRED ACTION to restore leakage increase within limits within 4 hours. This REQUIRED ACTION could not be completed within the COMPLETION TIME; therefore, a Technical Specification required shutdown was initiated, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 6/17/2021 AT 1309 FROM JASON BUTLER TO JEFFREY WHITED * * *

Upon further review of the leakage rates, it was determined that at 1900 EDT on 6/11/2021 the drywell floor drain unidentified leakage increased greater than 2 gpm within the previous 24 hours while in MODE 1. Technical Specification (TS) 3.4.4.B was entered to reduce leakage increase to within limits within 4 hours. At 2000 EDT on 6/11/2021 unidentified leakage was reduced below the 2 gpm increase within the previous 24 hours due to actions taken to lower reactor power and pressure. Therefore, the TS required shutdown per TS 3.4.4.C was not applicable. Thus Event Report 55303 is being retracted. The NRC resident has been notified of the retraction. Notified R2DO (Miller).