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 Entered dateSiteRegionReactor typeSystemScramEvent description
ENS 5578512 March 2022 06:56:00Turkey PointNRC Region 2Westinghouse PWR 3-LoopMain Steam Isolation Valve

The following information was provided by the licensee via email: At 0050 EST on 3/12/22, while shutting down for entry into a scheduled refueling outage, the station discovered that a single Main Steam Isolation Valve (4A MSIV) did not fully close on demand. All other equipment operated as expected. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v). The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 04/26/22 AT 1422 EDT FROM DAVID STOIA TO BRIAN PARKS * * *

The following information was provided by the licensee via email: On 3/12/2022 at 0656 EDT Turkey Point Unit 4 notified the NRC pursuant to 10 CFR 50.72(b)(3)(v)(D) that a single Main Steam Isolation Valve (MSIV) did not fully close when manually demanded from the control room during shutdown of Unit 4 for a refueling outage. Following disassembly and inspection of the MSIV, Florida Power & Light Engineering identified the cause of the deficiency and determined that the valve would have fully seated under its design accident conditions. This notification is a retraction of EN# 55785. The NRC Resident Inspector has been notified of this retraction. Notified R2DO (Miller).

ENS 557809 March 2022 23:20:00BrunswickNRC Region 2GE-4High Pressure Coolant Injection
Reactor Core Isolation Cooling
Automatic Depressurization System

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 557759 March 2022 00:47:00Davis BesseNRC Region 3B&W-R-LP

The following information was provided by the licensee via phone and email: A non-licensed, contract employee supervisor had a confirmed positive for alcohol during a follow-up fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.

  • * * RETRACTION FROM GERALD WOLF TO DONALD NORWOOD AT 1448 EDT ON 3/16/2022 * * *

The following information was received from the licensee via E-mail: This is a retraction of EN55775. The measured Blood Alcohol Level (BAC) of the individual was below the Fitness-For-Duty program limits, so this event did not constitute a violation of the Fitness-For-Duty program. The NRC Resident Inspector has been notified. Notified R3DO (Hills) and the FFD E-mail group.

ENS 557696 March 2022 00:55:00ByronNRC Region 3Westinghouse PWR 4-LoopThe following information was provided by the licensee: At 2115 CST on March 5, 2022 Byron Station Technical Support Center (TSC) emergency ventilation system supply fan belt failed. This failure affected the ability of the TSC ventilation system to maintain adequate radiological habitability in the event of an emergency with an airborne radiological release. All other capabilities of the TSC are unaffected by this condition. If an emergency was declared requiring TSC activation during this period, the TSC would be staffed and activated using existing emergency planning procedures. If the TSC becomes uninhabitable, the Station Emergency Director would relocate the TSC staff to an alternate TSC location in accordance with applicable procedures. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the discovered condition affected the functionality of an emergency response facility. The licensee notified the NRC resident inspector.
ENS 557685 March 2022 15:12:00Turkey PointNRC Region 2Westinghouse PWR 3-LoopThe following information was provided by the licensee via telephone and email: A licensed employee had a confirmed positive for alcohol during a random fitness for duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 5575825 February 2022 16:10:00McGuireNRC Region 2Westinghouse PWR 4-LoopThe following information was provided by the licensee: On 2/25/22, at 1133 EDT, the Technical Support Center (TSC) high temperature alarm annunciated in the Control Room due to an equipment malfunction that resulted in an unplanned loss of the TSC for greater than seventy-five minutes. If an emergency had been declared requiring TSC activation during this period, the TSC would have been staffed and activated using existing emergency planning procedures. If relocation of the TSC had been necessary, the Emergency Coordinator would have relocated the TSC staff to an alternate location in accordance with applicable site procedures. This is an eight-hour, non-emergency notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the equipment malfunction affected the functionality of an emergency response facility. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5575725 February 2022 14:53:00Turkey PointNRC Region 2Westinghouse PWR 3-LoopThe following information was provided by the licensee via telephone: A non-licensee contractor supervisor had a confirmed positive for a controlled substance during a fitness for duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 5575624 February 2022 14:35:00BrunswickNRC Region 2GE-4Primary Containment Isolation System
Primary containment
Reactor Building Ventilation
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 5575022 February 2022 01:44:00OconeeNRC Region 2B&W-L-LPSteam Generator
Feedwater
Reactor Protection System
Main Condenser
Manual Scram

The following information was provided by the licensee via fax or email: At 2207 (EST) on 2/21/2022 with Unit 2 in Mode 1 at 68 percent power, the reactor was manually tripped due to lowering water level in the 2A Steam Generator. The trip was not complex with all systems responding normally post-trip. Operators responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Units 1 and 3 were 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). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified."

  • * * UPDATE ON 3/23/22 AT 1643 EDT FROM CHRIS MCDUFFIE TO TOM KENDZIA * * *

The following information was provided by the licensee via phone and email: On 2/21/2022, Unit 2 was in Mode 1 increasing reactor power following startup from a forced outage. At 2205 (EST) with Unit 2 at 68 percent power, a feedwater control valve failed to properly control feedwater flow to the 2A Steam Generator and the Integrated Control System initiated an automatic runback. At 2207 (EST), the reactor was manually tripped from 39 percent power due to lowering water level in the 2A Steam Generator. Immediately following the manual reactor trip, an actuation of the Emergency Feedwater System (EFW) occurred. The 2A and 2B Motor Driven Emergency Feedwater (MDEFW) pumps automatically started as designed when the 'low steam generator level' signal was received for the 2A Steam Generator. The trip was not complex with all systems responding normally post-trip. Operators responded and stabilized the plant. Decay heat was removed by discharging steam to the main condenser using the turbine bypass valves. Units 1 and 3 were not affected. Unit 2 was restarted on 2/27/2022 following repairs. Due to the Reactor Protection System actuation while critical, this event was reported on 2/22/2022 as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Following further evaluation, it was determined that a valid EFW actuation occurred, therefore this event is now also being reported as a late 8-hour non-emergency notification of a valid actuation of the EFW system in accordance with 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Notified the R2DO (Miller).

ENS 5574818 February 2022 08:35:00McGuireNRC Region 2Westinghouse PWR 4-LoopSteam Generator
Feedwater
Reactor Protection System
Auxiliary Feedwater
Control Rod
Manual ScramThe following information was provided by the licensee via telephone and email: On 2/18/2022, McGuire Nuclear Station Unit 2 experienced a turbine runback to 55 percent power. Based on concerns with unit stability, the reactor was manually tripped at 0459 (EST). All Auxiliary Feedwater pumps started on low steam generator level as required. The reactor trip was uncomplicated with all systems responding normally post trip. A feedwater isolation occurred as designed. Unit 1 was not affected. Due to the Reactor Protection System actuation while critical, actuation of the Turbine Driven Auxiliary Feedwater Pump and Motor Driven Auxiliary Feedwater pumps along with the Feedwater Isolation, this event is being reported as a four hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an 8 hour nonemergency notification per 10 CFR 50.72(b)(3)(iv)(A). 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: All control rods fully inserted. Decay heat is being removed via the condenser and normal feedwater. Unit 2 is in a normal shutdown electrical lineup.
ENS 5574216 February 2022 17:01:00SequoyahNRC Region 2Westinghouse PWR 4-LoopThe following information was provided by the licensee via fax or email: At 1128 EST on 2/16/2022, the SQN (Sequoyah Nuclear) Shift Manager was notified that TVA (Tennessee Valley Authority) attempted to notify Tennessee Emergency Management Agency (TEMA) regarding routine siren testing at 0750. TVA was unable to reach TEMA via telephone land line or the Emergency Communication and Notification System (ECNS). TEMA Watch Point staff were located at their back-up facility. TVA subsequently notified TEMA via cell phone that there were communication issues with the primary and backup notification methods. It was determined that the TEMA back-up facility was not able to receive incoming calls. At 0820, TEMA positioned personnel at their primary facility in order to respond to notifications. This restored primary and backup means of notifying the state because the primary facility was not affected by the communication issues. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Offsite Communications Capability because it affected TVA's ability to notify the State of TN. The licensee has notified the NRC Resident Inspector.
ENS 5574116 February 2022 16:42:00Watts BarNRC Region 2Westinghouse PWR 4-LoopThe following information was provided by the licensee via fax or email: At 1159 EST, on 2/16/2022, the Watts Bar Nuclear, Shift Manager was notified that Tennessee Valley Authority (TVA) attempted to notify Tennessee Emergency Management Agency (TEMA) regarding routine siren testing at 0750 EST. TVA was unable to reach TEMA via telephone land line or the Emergency Communication and Notification System (ECNS). TEMA Watch Point staff were located at their back-up facility. TVA subsequently notified TEMA via cell phone that there were communication issues with the primary and backup notification methods. It was determined that the TEMA back-up facility was not able to receive incoming calls. At 0820 EST, TEMA positioned personnel at their primary facility in order to respond to notifications. This restored primary and backup means of notifying the state because the primary facility was not affected by the communication issues. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Offsite Communications Capability because it affected TVA's ability to notify the State of TN. The licensee has notified the NRC Resident Inspector.
ENS 557357 February 2022 15:49:00VogtleNRC Region 2W-AP1000The following information was provided by the licensee via telefone: A non-licensed contractor superintendent had a confirmed positive for alcohol during a for-cause fitness for duty test. The employee's access to the plant has been terminated. The NRC Resident Inspector has been notified.
ENS 557345 February 2022 18:54:00Davis BesseNRC Region 3B&W-R-LPReactor Coolant SystemThe following information was provided by the licensee via email: At approximately 1402 EST on 2/5/2022, with the Unit in Mode 1 at approximately 98 percent power, Operations was performing a valve lineup and inadvertently isolated a portion of the Reactor Coolant System (RCS) Letdown System, resulting in the system relief valve lifting and entry into the Makeup and Purification System Malfunction Abnormal Procedure due to loss of letdown. Pressurizer level increased and Technical Specification (TS) Limiting Condition for Operation (LCO) 3.4.9 CONDITION A was entered at 1414 EST due to Pressurizer level not below the limit of 228 inches, which has a REQUIRED ACTION to restore Pressurizer level within one hour. A rapid plant down power was initiated at approximately 1430 EST to reduce Pressurizer level. At 1514 EST on 2/5/2022, TS LCO 3.4.9 CONDITION B was entered, which has a REQUIRED ACTION to place the Unit in MODE 3 in 6 hours and in MODE 4 in 12 hours. As the Unit was continuing to down power, this represents initiation of a Technical Specification required shutdown, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). At approximately 1542 EST the down power was stopped at 15 percent power. Pressurizer level was restored to less than 228 inches at approximately 1603 EST, and TS LCO 3.4.9 was exited. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 557335 February 2022 04:32:00OconeeNRC Region 2B&W-L-LPReactor Coolant System
Feedwater
Reactor Protection System
Automatic Scram

The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At 0357 EST on 2/5/22, Oconee Unit 2 declared an Unusual Event due to a multiple fire alarms and visual observance of a smoke filled room in the West Penetration Room (EAL HU 4.1). Unit 2 automatically tripped and entered Mode 3. There was an indication that a release to the environment potentially occurred. Units 1 and 3 remained at 100% power. The Licensee notified the NRC Resident Inspector, the State, and local authorities. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), and DHS Nuclear SSA (email).

  • * * UPDATE ON 2/5/22 AT 0748 EST FROM CHUCK CLEMONS TO LLOYD DESOTELL * * *

The following information was provided by the licensee via email: At 0343 EST on February 5th, 2022, with Unit 2 in Mode 1 at 100% power, the reactor automatically tripped due to Reactor Coolant Pump (RCP) Flux/Flow Imbalance caused by the simultaneous trip of all 4 RCPs. The cause of the loss of all RCPs is under investigation. The trip was not complex. There was no inoperable equipment prior to the event that contributed to it. Operations responded and stabilized the plant. Decay heat is being removed by the Reactor Coolant System (RCS) in natural circulation with normal feedwater flow. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Licensee notified the State and local authorities. Notified R2DO (Miller)

  • * * UPDATE ON 2/5/22 AT 0815 EST FROM CHUCK CLEMONS TO LLOYD DESOTELL * * *

The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At 0811 EST on 2/5/22, Oconee Unit 2 terminated the notification of unusual event because EAL HU 4.1 criteria were no longer met. Unit 2 remains shutdown in Mode 3. The licensee has notified the State and local authorities and will notify the NRC Resident Inspector. Notified R2DO (Miller), IRD MOC (Gott), NRR EO (Regan), DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 2/7/22 AT 2342 EST FROM PATRICK GADSBY TO OSSY FONT * * *

The following information was provided by the licensee via email: The follow-up investigation determine that the cause of the trip was determined to be a failed fuse in an undervoltage monitoring circuit for the electrical bus powering the reactor coolant pumps. The trip of the reactor coolant pumps resulted in an automatic reactor trip as designed. The plant responded as expected. There was no fire. The response of the main feedwater system caused an expected increase in temperature in a standby portion of the system which caused the breakdown of the pipe coating producing smoke. There was no release to the environment. The reading of a single instrument was determined to be invalid and other indications supported the conclusion that no release occurred. The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Licensee notified the NRC Resident Inspector, State and local authorities. A media release was issued on 2/5/22. Notified R2DO (Miller), IRD MOC (Gott), NRR EO (Regan).

ENS 557324 February 2022 20:30:00FermiNRC Region 3GE-4Feedwater
Reactor Protection System
Primary containment
Main Condenser
Control Rod
The 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 557231 February 2022 16:24:00LaSalleNRC Region 3GE-5The following information was provided by the licensee via fax or email: At approximately 1328 CST on 2/1/2022, LaSalle Generating Station was made aware of the following event that resulted in additional county emergency sirens sounding. The Grundy County monthly siren test had issues with siren activation from the County's primary controller. The buttons to activate were being pressed, but the intended sirens were not initiating. The Grundy County operator continued to attempt activation unknowingly activating the sirens in the Northeast quadrant several times between 1000-1015 CST. This event is reportable per 10 CFR 50.72(b)(2)(xi), News Release or Notification of Other Government Agencies. This is a 4-Hour Reporting requirement. The LaSalle NRC Resident has been notified.
ENS 557211 February 2022 15:23:00DresdenNRC Region 3GE-3The following information was provided by the licensee via fax or email: At approximately 1025 CST, the Dresden Main Control was notified of Grundy County warning sirens issues during the intended monthly test. The Grundy County scheduled monthly siren test had issues with siren activation from the County's primary controller. The buttons to activate were being pressed but the intended sirens were not initiating. The Grundy County operator continued to attempt activation unknowingly activating the sirens in the northeast quadrant between 1000-1015 CST. This event is reportable per 10 CFR 50.72(b)(2)(xi), News Release or Notification of Other Government Agencies. This is a 4-Hour Reporting requirement. The Dresden NRC Resident has been notified.
ENS 557221 February 2022 15:20:00BraidwoodNRC Region 3Westinghouse PWR 4-LoopThe following information was provided by the licensee via fax or email: At approximately 1025 CST on 2/1/22, the Braidwood Station Main Control Room was notified of a public notification of multiple inadvertent siren actuation affecting Braidwood Station in Will County, Illinois while testing sirens. This event is reportable per 10 CFR 50.72(b)(2)(xi), News release or notification of other Government Agencies. Braidwood NRC Resident has been notified.
ENS 557201 February 2022 11:42:00Diablo CanyonNRC Region 4Westinghouse PWR 4-LoopThe following information was provided by the licensee via fax or email: At 1350 PST on 01/31/2022, Pacific Gas and Electric determined that a non-licensed employee supervisor violated Diablo Canyon FFD policy and had a confirmed positive on a direct observed test. The employee's access to the plant has been terminated and permanent denial has been entered into PADS. The NRC Senior Resident Inspector 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 5571326 January 2022 18:49:00OconeeNRC Region 2B&W-L-LP

Licensee provided the following information via email: At 1050 EST on January 26, 2022, it was discovered that the required offsite and the overhead and underground paths of onsite emergency AC power were simultaneously inoperable; therefore this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(A). The overhead path and both required offsite paths were inoperable due to a lockout of the Unit 3 Startup Transformer CT-3. The underground path was made momentarily inoperable as part of aligning an additional offsite power source to Unit 3 to provide defense in depth to a loss of power. The safety function was restored at 1051 EST when the underground path was declared OPERABLE. The offsite power source was restored at 1651 EST. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 1/26/22 AT 2237 EST FROM GABE SLAUGHTER TO KAREN COTTON * * *

The Licensee updated the 3rd paragraph above as follows via email: The safety function was restored at 1051 EST when the underground path was declared OPERABLE. An offsite power source was restored through Transformer CT-5 within 15 minutes. Startup transformer CT-3 was restored and all offsite and the onsite overhead power source were returned to service by 1651 EST. Investigation of the cause of the CT-3 lockout is in progress. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified of this update.

ENS 5570821 January 2022 10:25:00MonticelloNRC Region 3GE-3Secondary containment
Primary containment
Reactor Building Ventilation
Standby Gas Treatment System
The following information was provided by the licensee via email: This telephone notification is provided in accordance with 10 CFR 50.73(a)(1) to report an invalid actuation of secondary containment relays in accordance with 10 CFR 50.73(a)(2)(iv)(A). On November 29, 2021, the `B' Fuel Pool radiation monitor spiked high during restoration following the performance of the 0068 procedure `Spent Fuel Pool & Reactor Building Exhaust Plenum Monitor Calibration' due to cable to radiation monitor connector degradation from handling. This resulted in a Partial Primary Containment Group II isolation (gas systems), initiation of Standby Gas Treatment system, and isolation of the Reactor Building Ventilation system. All systems responded as designed to the actuation signal. Operations reset the Partial Primary Containment Group II isolation signal, shutdown Standby Gas Treatment System, and restored Reactor Building Ventilation system per procedures. At the time of the occurrence, the `A' Fuel Pool radiation monitor was reading normal at approximately 1.5 mr/hr. The `B' Fuel Pool radiation monitor spiked above the 50 mr/hr setpoint and continued to read erratically. Work was performed to clean and reconnect the connector and testing per 0068 procedure verified the condition was corrected. The `B' Fuel Pool radiation monitor returned to service. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5570616 January 2022 06:41:00Browns FerryNRC Region 2GE-4Shutdown CoolingThe 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 556987 January 2022 16:29:00CallawayNRC Region 4Westinghouse PWR 4-LoopFeedwater
Auxiliary Feedwater
Control Rod
The following information was provided by the licensee via email: At 1223 CST on January 7, 2022, Callaway Plant was in Mode 1 at approximately 100 percent power when a turbine trip / reactor trip occurred. All safety systems responded as expected with the exception of an indication issue with the 'B' Feedwater Isolation Valve, which was confirmed closed. A valid Feedwater Isolation Signal and Auxiliary Feedwater Actuation Signal were also received as a result of the reactor trip. The plant is being maintained stable in Mode 3. All control rods fully inserted from the reactor trip signal, and decay heat is being removed via the Auxiliary Feedwater and Steam Dump Systems. The NRC Senior Resident Inspector was notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The plant is in a normal shutdown electrical lineup.
ENS 556947 January 2022 09:32:00Comanche PeakNRC Region 4Westinghouse PWR 4-LoopThe following information was provided by the licensee via fax or email: At 0120 (CST) on 01/07/2022, a partial loss of the 25KV Power Distribution System caused a loss of both the Primary and Backup Meteorological Towers at the Comanche Peak Nuclear Power Plant. This resulted in a loss of emergency assessment capability with regard to meteorological conditions. A backup diesel generator for the primary Meteorological Tower did not start due to a dead battery. After the battery issue was resolved, the diesel generator started but it subsequently tripped due to a loose fuse. The 25 KV Plant Support Power Loop feeds certain non-safety-related equipment and does not affect plant operation. Power was restored to both Meteorological Towers at 0305 (CST) on 01/07/2022 and proper operation was verified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The NRC Resident Inspector was notified.
ENS 556936 January 2022 20:51:00Saint LucieNRC Region 2CEOn January 6, 2022 at 1937 (EST), St Lucie Unit 2 commenced a reactor shutdown as required by Technical Specification 3.1.3.1 Action 'e', due to Control Element Assembly number 27 slipping from 133 inches to 120 inches withdrawn and unable to be recovered within the prescribed time limits. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 2 entered 6 hour LCO to shutdown to mode 3 at 1539 EST as required by Technical Specification 3.1.3.1 Action 'e'. There was no impact on Unit 1.
ENS 556926 January 2022 12:29:00South TexasNRC Region 4Westinghouse PWR 4-LoopEmergency Diesel GeneratorAt 0603 CST on 1/6/2022, with Unit 2 in Mode 1 at 100 percent power, the South Texas Project (STP) south switchyard electrical bus was de-energized momentarily and re-energized approximately 40 seconds later. Emergency Diesel Generators (EDG) 22 automatically started in response to loss of offsite power on Train B Engineered Safety Feature (ESF) Bus. This event is reportable under 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in the valid actuation of an emergency AC electrical power system (50.72(b)(3)(iv)(B)(8)). 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: Unit 2 is in a 72 hour LCO per TS 3.8.1.1.A for the loss of one offsite power supply. The plant is in a normal electrical lineup. There was no impact on Unit 1.
ENS 556916 January 2022 11:06:00CookNRC Region 3Westinghouse PWR 4-Loop

The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: On 01/06/22 at 1044 (EST), an Unusual Event was declared due to a Fire Detection Actuation in the Unit 1, auxiliary cable vault (EAL H.U 4.1). No fire was detected. Unit 1 and Unit 2 remain at 100 percent power. The Llcensee notified the NRC Resident Inspector, the state, and local authorities. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), and DHS Nuclear SSA (email).

  • * * UPDATE FROM DAN WALTER TO TOM KENDZIA AT 1452 (EST) ON 01/06/2022 * * *

At 1441 (EST), DC Cook Unit 1 terminated their notification of unusual event. The basis for termination was that the inspection identified no damage to cables or cable trays. The fire protection system is out of service for the auxiliary cable vault with compensatory measures in effect. The licensee has notified the state and local authorities and will notify the NRC Resident Inspector. Notified R3DO (Skokowski), IRD MOC (Grant), NRR EO (Felts), IR (Kennedy)(email), NRR (Veil)(email), R3 DRA (Shuaibi)(email), DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

ENS 556905 January 2022 13:36:00Fort CalhounNRC Region 4CEThe following information was provided by the licensee via email: At 0907 CST, a small fire was reported in the Intake Structure at Fort Calhoun Station. Offsite fire departments were notified at 0909 CST and responded at 0922 CST. Fire was confirmed extinguished at 0949 CST. Fire was extinguished using offsite resources per the Station Fire Plan. There were no injuries reported. The fire occurred in the Non-Radiological area of the plant and there was no release of radioactivity or hazardous materials.
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 556843 January 2022 15:58:00Calvert CliffsNRC Region 1CEAutomatic ScramThe following information was provided by the licensee via email: At 1223 (EST) on 01/03/2022, Calvert Cliffs Unit 2 automatically tripped from 100 percent power due to loss of electrical load. The cause is under investigation. The site Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods inserted and decay heat is being removed via the condenser. The plant is in a normal shutdown electrical lineup. There was no impact on Unit 1.
ENS 556821 January 2022 17:35:00ColumbiaNRC Region 4GE-5Service water
Reactor Core Isolation Cooling
High Pressure Core Spray

The Licensee provided the following information via fax: During performance of a surveillance of the High Pressure Core Spray (HPCS) service water system on January 1, 2022, the HPCS system was declared inoperable for performance of the surveillance. During the surveillance, pump discharge pressure and flow were above the action range curve specified in the surveillance. For the given flow rate, pump discharge pressure was too high. This condition prevents declaring the HPCS service water system and HPCS system operable. The HPCS service water and HPCS systems remain inoperable. The station entered Technical Specification (TS) 3.7.2.A and TS 3.5.1.B at 0910 (PST) on January 1, 2022. In accordance with TS 3.5.1.B, the Reactor Core Isolation Cooling (RCIC) system was verified to be operable. TS 3.5.1 Action B provides a 14-day completion time to restore HPCS to an operable status. All other Emergency Core Cooling systems (ECCS) are operable. This event is being reported as an event or condition that could have prevented the fulfillment of a safety function credited for mitigating the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). The HPCS system is a single train system at Columbia. The NRC resident has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The licensee is investigating the cause of the high pump discharge pressure and verifying instrumentation accuracy.

  • * * RETRACTION ON 1/6/22 AT 1715 EST FROM CHASE WILLIAMS TO TOM KENDZIA * * *

This Notification is to retract EN 55682, Unplanned High Pressure Core Spray (HPCS) Inoperability. On 1/1/2022 at (1735 EST), Columbia Generating Station notified the NRC under 10 CPR 50.72(b)(3)(v)(D) of the inoperability of a single train of safety system (HPCS) for performance of the surveillance. During the surveillance pump discharge pressure and flow were above the action range curve specified in the surveillance. Engineering performed an analysis of this event and concluded the HPCS was operable during the event and would have performed its required safety function. The results of initial IST testing of HPCS-P-2 via OSP-SW/IST-Q703 on 01/01/22 resulted in measured parameters falling outside of the acceptable range specified for this pump. Systematic error was suspected as the cause of the failure and the test was reperformed following taking actions to eliminate the suspected systematic errors. The second performance of the test on 01/01/22 resulted in acceptable pump performance. Evidence exists that the initial performance of the test failed due to imprecise averaging techniques due to difficulties in averaging continuously changing values on the test instrument. The second performance of OSP-SW/IST-Q703 should be considered a successful test and the test of record as the systematic error was eliminated and measured parameters are considered valid. The NRC Resident Inspector has been notified. The HOO notified R4DO (Rolando-Otero).

ENS 5567929 December 2021 19:16:00HatchNRC Region 2GE-4Feedwater
Reactor Protection System
Main Condenser
Manual ScramThis 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 5567427 December 2021 16:49:00HarrisNRC Region 2Westinghouse PWR 3-Loop

The following information was provided by the licensee via email: On December 27, 2021, at 1014 EST, a system error in the site's Alert and Notification Siren System was identified, indicating a loss of the siren system affecting a greater than 25% of the emergency planning zone population. Review of the system's data logger indicates the system error has been present within the system since December 22, 2021, at 1245 EST. The fleet's telecommunications department has been contacted and is aware of the issue. In the event that a radiological emergency should occur at the Shearon Harris Nuclear Power Plant, Primary Route Alerting procedures will be put in use by the local jurisdictions. This condition is reportable as a Loss of Emergency Preparedness Capabilities per 10 CFR 50.72(b)(3)(xiii). The NRC Resident, state and local agencies have been notified.

  • * * RETRACTION ON 12/29/21 AT 1630 EST FROM SARAH MCDANIEL TO KAREN COTTON * * *

The following information was provided by the Licensee via email: Further troubleshooting efforts identified that the Chatham County EOC Siren Activation Point remained capable of sending an alert signal to the sirens for the duration of the event described above. This ensures siren activation would be performed in a timely manner in the event of a radiological emergency. This Event Notification is therefore retracted, as no loss of emergency preparedness capabilities has occurred. The NRC Resident and local agencies have been notified. Notified R2DO (Miller)

ENS 5566016 December 2021 14:57:00Browns FerryNRC Region 2GE-4Reactor Protection System
Primary Containment Isolation System
Reactor Building Ventilation
Reactor Water Cleanup
  • 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 5564915 December 2021 16:45:00HarrisNRC Region 2Westinghouse PWR 3-Loop
  • The following information was provided by the licensee via email:

At 0927 EST on December 15, 2021, it was determined that a non-licensed employee supervisor failed a test specified by the fitness for duty (FFD) testing program. The individual's authorization for site access has been terminated. The NRC Resident Inspector has been notified.

ENS 5564110 December 2021 13:35:00Saint LucieNRC Region 2CESteam Generator
Feedwater
Main Condenser
Manual ScramOn 12/10/2021, at 1024 EST, with Unit 1 at 100 percent power, the reactor was manually tripped due to lowering level in the steam generators. All systems responded as expected following the trip. The reactor is currently stable in Mode 3 and operators restored steam generator level utilizing main feedwater. The cause of the reduction in feedwater flow is under investigation. St. Lucie Unit 2 was not affected and remains at 100 percent power. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) for the reactor trip. The NRC Resident Inspector has been notified. All rods inserted into the core during the trip. The plant is in its normal shutdown electrical lineup. Decay heat is being maintained by steam discharge to the main condenser using the turbine bypass valves.
ENS 5563810 December 2021 03:54:00OconeeNRC Region 2B&W-L-LPReactor Protection System
Main Condenser
At 0049 EST, on December 10, 2021, with Unit 2 in Mode 1 at 73 percent power, the reactor automatically tripped due to an unknown condition. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being maintained by discharge steam to the main condenser using the turbine bypass valves. Units 1 and 3 are not affected. The cause of the trip is under investigation. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)iv)(B). There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. All rods inserted into the core during the trip. The plant is in its normal shutdown electrical lineup maintaining normal operating pressure and temperature.
ENS 556276 December 2021 18:14:00BrunswickNRC Region 2GE-4Secondary containment
Emergency Diesel Generator
Primary Containment Isolation System
Reactor Building Ventilation
Residual Heat Removal
Reactor Water Cleanup
Control Room Emergency Ventilation

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 556266 December 2021 17:17:00Palo VerdeNRC Region 4CEFeedwater
Steam Bypass Control System
The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73. At 1203 MST on December 6, 2021, the Unit 3 reactor automatically tripped on low departure from nucleate boiling ratio. A part-strength control element assembly was being moved at the time of the trip. Unit 3 is stable and in Mode 3. In response to the reactor trip, all control element assemblies inserted fully into the core. Safety-related electrical power remains energized from off-site power sources and reactor coolant pumps continue to provide forced circulation through the reactor. Decay heat is being removed by the steam bypass control system and main feedwater system. Required systems operated as expected. No emergency classification was required per the Emergency Plan. The NRC Senior Resident Inspector has been informed. Units 1 and 2 were unaffected by this transient.
ENS 556192 December 2021 00:58:00Quad CitiesNRC Region 3GE-3At 1847 CST on December 1, 2021, it was discovered that the HPCI (high pressure coolant injection) system was inoperable. Due to this inoperability, the system was in a condition that could have prevented the fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The Unit 1 RCIC (reactor core isolation cooling) system was Operable during this time period. There was no impact on the health and safety of the public or plant personnel. The NRC Resident has been notified. Unit 1 HPCI operability was restored at 2110 CST.
ENS 5561630 November 2021 16:22:00SusquehannaNRC Region 1GE-4Feedwater
Main Turbine
Control Rod
At 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 5561429 November 2021 18:47:00Arkansas NuclearNRC Region 4B&W-L-LPReactor Coolant System
Feedwater
Reactor Protection System
Auxiliary Feedwater
On November 29, 2021 at 1458 CST, Arkansas Nuclear One, Unit 1, (ANO-1) automatically tripped due to high Reactor Coolant System pressure after the 'A' Main Feedwater Pump was manually tripped due to lowering speed. ANO-1 is currently stable in MODE 3 (Hot Standby) maintaining pressure and temperature with the P-75 Auxiliary Feedwater pump and steaming to the Condenser. There are no indications of a radiological release on either unit as a result of this event. This report satisfies the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A) for the Reactor Protection System actuation. The NRC Senior Resident Inspector has been notified. Unit 2 was not affected.
ENS 5561227 November 2021 13:16:00OconeeNRC Region 2B&W-L-LPDecay Heat RemovalAt 0519 EST on November 27, 2021, with Unit 2 in Mode 5 at zero percent power, an actuation of the Emergency AC Electrical Power System occurred. The reason for the Emergency AC Electrical Power System auto-start was a lockout of the CT-2 transformer; causing a temporary loss of AC power to the main feeder bus. The Keowee Hydroelectric Units 1 and 2 automatically started as designed when a main feeder bus undervoltage signal was received. 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 Emergency AC Electrical Power System. Additionally, the temporary loss of AC power resulted in a loss of Decay Heat Removal (DHR) that was restored upon power restoration to the main feeder bus. Therefore, this condition is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v) for an event or condition that could have prevented fulfillment of a safety function. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The loss of the CT-2 transformer is under investigation. Main feeder bus power was restored within a minute so no plant heat up occurred as a result of the loss of the decay heat removal system.
ENS 5561024 November 2021 21:42:00Comanche PeakNRC Region 4Westinghouse PWR 4-LoopA violation occurred concerning Comanche's Peak's Fitness-For-Duty Program. Two empty mini-bottles of alcohol were discovered in a trash can within the protected area. The event has been documented in the corrective action program. The resident inspector has been notified.
ENS 5560924 November 2021 20:24:00GinnaNRC Region 1Westinghouse PWR 2-LoopReactor Coolant SystemThis 60-day telephone notification is provided in accordance with 10 CFR 50.73(a)(1) to report one invalid actuation of the Unit 1 Containment Isolation System Train "A" in accordance with 10 CFR 50.73(a)(2)(iv)(A). On October 17, 2021 at approximately 1358 (EDT), a DC breaker was opened to perform an inspection of a Containment Isolation (CI) rack. A CI signal was produced and resulted in a loss of Letdown during filling and venting the Reactor Coolant System (RCS) with the RCS at 344 psig. RCS pressure began to rise, and prompt actions were taken by the Control Room to secure a Charging Pump within 20 seconds. The RCS pressure rise continued and both Pressure Operated Relief Valves cycled at 409.9 psig as designed, lowering RCS pressure. The CI Train "A" was not part of a pre-planned sequence and the event resulted in the invalid actuation of Train "A" Containment Isolation valves in more than one system. All valves functioned successfully. The DC breaker was closed, CI signal reset, and associated CI valves re-opened. All systems functioned as required and returned to normal service. The NRC Senior Resident Inspector has been notified.
ENS 5560223 November 2021 09:10:00Beaver ValleyNRC Region 1Westinghouse PWR 3-LoopEmergency Diesel GeneratorThis 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid specific system actuation. At 0907 (EDT) on September 30, 2021, with Unit 1 in Mode 1, at 100 percent power, an actuation of the 1-1 emergency diesel generator (EDG) occurred during loss of voltage relay functional testing. The 1-1 EDG auto-start was due to human error during performance of the test procedure when the bus 1AE undervoltage signal was improperly defeated and a simulated undervoltage signal was applied. No actual undervoltage condition was present during this event. The 1-1 EDG automatically started as designed when the bus undervoltage signal was received. This was a complete actuation of an EDG to start and come to rated speed, and all affected systems functioned as expected in response to the actuation. Following the actuation, the relays were restored and the 1-1 EDG was shut down in accordance with plant procedures. This event is considered an invalid system actuation reportable under 10 CFR 50.73(a)(2)(iv)(A). The actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. Therefore, in accordance with 10 CFR 50.73(a)(1), this telephone notification is provided within 60 days after discovery of the event instead of submitting a written Licensee Event Report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5559721 November 2021 14:28:00Calvert CliffsNRC Region 1CESteam Generator
Reactor Protection System
Auxiliary Feedwater
Manual ScramAt 1046 EST on November 21, 2021, with Calvert Cliffs Nuclear Power Plant Unit 2 in Mode 1 at 100 percent power, the reactor was manually tripped due to lowering levels in both steam generators following a loss of the 21 and 22 steam generator feed pumps. An Auxiliary Feedwater System actuation occurred to restore steam generator water levels. The trip was not complicated, with all systems responding normally. Decay heat is being removed by the Auxiliary Feedwater System. Calvert Cliffs Nuclear Power Plant Unit 1 is unaffected and remains in Mode 1 at 100 percent power. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification. RPS actuation, per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the automatic actuation of the Auxiliary Feedwater System is being reported as an eight-hour, non-emergency notification, Specific System Actuation, per 10 CFR 50.72(b)(3)(vi)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.