Semantic search

Jump to navigation Jump to search
 Entered dateSiteRegionReactor typeSystemScramEvent description
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).

ENS 552999 June 2021 15:02:00SurryNRC Region 2Westinghouse PWR 3-Loop

At 1115 EDT on June 9, 2021, during a siren activation test, a loss of the capability to activate the sirens from both Surry local activation sites was identified. The Virginia EOC was participating in the activation test and is aware of the issue and notified the local government authorities in the Surry EPZ of the situation. The NRC Resident has been notified of this issue. The station telecommunications department has been contacted and is aware of the issue. In the event that a radiological emergency should occur at the Surry Power Station, Primary Route Alerting procedures will be put in use by the local jurisdictions. This report is being made in accordance with 10 CFR 50.72(b)(2)(xi) and 10 CFR 50.72(b)(3)(xiii) due to notification of other state and local government agencies of the failure of the Alert & Notification system for Surry.

  • * * PARTIAL RETRACTION ON 6/18/2021 AT 0959 FROM STEPHEN MITCHELL TO THOMAS KENDZIA * * *

Surry Power Station Event Notification 55299 is being retracted based upon further evaluation. Surry has three localities (State SAU, James City, and Surry County) with access to the redundant activation trains (primary and backup systems). The actuation tests only one primary and one backup activation panel at two localities, only primary at the State SAU (Situational Awareness Unit) and back up at James City County were tested. Follow-up telecom and vendor investigation revealed that the primary server was functional from James City County that would have actuated all 71 sirens; Consequently, it was concluded that all of the sirens were fully functional from the James City primary system and there was no loss of all sirens as originally reported on 6/9/2021 (EN 55299). EN 55299 also contained a 4-hour Offsite Notification per 10 CFR 50.72(b)(2)(xi) that is unaffected. The NRC Resident Inspector has been notified. Notified R2DO (Miller).

ENS 552978 June 2021 08:37:00SummerNRC Region 2Westinghouse PWR 3-LoopA contract employee supervisor 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 552957 June 2021 18:31:00Comanche PeakNRC Region 4Westinghouse PWR 4-LoopSteam Generator
Feedwater
Main Transformer
Main Condenser
Automatic ScramAt 1527 (Central Standard Time) Unit 2 Reactor tripped caused by a turbine trip due to a fault and fire on Unit 2 Main Transformer #1. All Aux Feedwater Pumps started due to steam generator Lo-Lo levels. Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007B. The Emergency Response Guideline Network has been exited. Decay heat is being rejected to the Main Condenser via the steam dump valves. Fire was extinguished at 1546 without offsite assistance. No major injuries reported and no personnel transported offsite for medical attention. Cause of the fault and fire are under investigation. NRC Resident Inspector has been notified. All rods inserted into the core during the trip. There were no relief valves or safety valves lifted during the transient. The plant is stable in its normal shutdown electrical lineup via the auxiliary transformer with all safety equipment available. Unit 1 was not affected by the transient.
ENS 552871 June 2021 17:46:00Browns FerryNRC Region 2GE-4Reactor Protection System
Reactor Water Cleanup
Control Room Emergency Ventilation
This 60-day telephone notification is being made per the reporting requirements specified by 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation of the 2A Reactor Protection System (RPS). On April 1, 2021, at 1302 (CDT), Browns Ferry Unit 2, 2A RPS (Motor Generator) MG set tripped causing a half scram. Unit 2 experienced an unexpected trip of the 2A RPS MG Set that resulted in automatic Primary Containment Isolation System (PCIS) Group 2, 3, 6, and 8 isolations and Trains A, B, and C Standby Gas Treatment (SGT) and Train A Control Room Emergency Ventilation (CREV) starts. At the time of the event, Unit 2 was in a refueling outage and the rods were already fully inserted. All systems responded as expected. Plant conditions which initiate PCIS Group 2 actuations are Reactor Vessel Low Water Level (Level 3) or High Drywell Pressure. The PCIS Group 3 actuations are initiated by Reactor Vessel Low Water Level (Level 3) or Reactor Water Cleanup Area High Temperature. The PCIS Group 6 actuations are initiated by Reactor Vessel Low Water Level (Level 3), High Drywell Pressure, or Reactor Building Ventilation Exhaust High Radiation (Reactor Zone or Refuel Zone). The PCIS Group 8 actuations are initiated by Low Reactor Vessel Water Level (Level 3) or High Drywell Pressure. At the time of the event, these conditions did not exist; therefore, the actuation of the PCIS was invalid. Based on the troubleshooting conducted, the cause was determined to be a loose wiring connection in the motor circuit. The lugs were replaced with ring lugs. Operations reset the 2A RPS Half Scram and PCIS in accordance with 2-AOI-99-1 on April 1, 2021, at 1324 CDT thus correcting the condition and returning RPS to service. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program as Condition Report 1683358. The NRC Resident Inspector has been notified of this event.
ENS 5528531 May 2021 10:50:00Arkansas NuclearNRC Region 4CEA licensed operator had a confirmed positive 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 5528128 May 2021 09:04:00WaterfordNRC Region 4CEFeedwaterThis 60-day telephone notification is being made in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to provide information pertaining to an invalid Engineered Safety Feature actuation signal. On April 1, 2021, at Waterford 3, while performing a replacement of power supplies on the Plant Protection System, a spurious signal caused a partial actuation of the Emergency Feedwater Actuation Signal. A partial Emergency Feedwater (EFW) logic trip path was met causing the opening of valves EFW-228A (EFW to SG 1 Primary Isolation), EFW-229A (EFW to SG 1 backup isolation), EFW-228B (EFW to SG 2 Primary Isolation), and EFW-229B (EFW to SG2 Backup Isolation). This inadvertent actuation was spurious and was not a valid signal resulting from parameter inputs. The 1992 Statements of Consideration (57 FR 41378) define an invalid signal to include spurious signals. Therefore, this actuation is considered invalid. This event was entered into the Waterford 3 corrective action program for resolution. This event did not result in any adverse impact to the health and safety of the public. The plant responded as expected. In accordance with 10 CFR 50.73(a)(1) a telephone notification is being made in lieu of submitting a written Licensee Event Report. The NRC Senior Resident Inspector has been notified.
ENS 5527525 May 2021 21:38:00CatawbaNRC Region 2Westinghouse PWR 4-LoopAt 1751 EDT on May 25, 2021, it was determined the local leak rate test (LLRT) for the 2EMF-IN containment penetration did not meet 10 CFR 50 Appendix J requirements for both the inboard and outboard containment isolation valves (2MISV5230 and 2MISV5231). The LLRT was performed during the previous refueling outage at which time primary containment was not required to be operable. The leakage assigned to the penetration also resulted in total leakage exceeding the allowed overall leakage. The valves were repaired and retested satisfactory prior to entering the mode of applicability, This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A), There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5527224 May 2021 12:01:00SequoyahNRC Region 2Westinghouse PWR 4-LoopReactor Protection System
Auxiliary Feedwater
Automatic ScramUnit 2 is not impacted and remains stable in Mode 1 at 100 percent power. 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) and in accordance with 10 CFR 50.72 (b)(3)(iv)(A) as an event that results in a valid actuation of the AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. No relief valves opened. All Rods fully inserted. Decay heat is being removed by Auxiliary Feedwater via the steam dumps. The plant is in a normal post-trip electrical line-up.
ENS 5526519 May 2021 08:35:00Palo VerdeNRC Region 4CESteam Generator
Reactor Coolant System
Reactor Protection System
Emergency Diesel Generator
Auxiliary Feedwater
Automatic Scram

At 0315 MST on May 19, 2021, Unit 2 reactor automatically tripped during testing of the Plant Protection System. The Reactor Protection System actuated to trip the reactor on High Pressurizer Pressure, although no plant protection setpoints were exceeded. Main Steam Isolation Signal (MSIS), Safety Injection Actuation Signal (SIAS), and Containment Isolation Actuation Signal (CIAS) were received. No injection of water into the Reactor Coolant System occurred. Auxiliary Feedwater Actuation Signals (AFAS) 1 and 2 actuated on low Steam Generator water level post trip as designed. This event is being reported as a reactor protection system and a specified system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Following the reactor trip, all (Control Element Assemblies) CEAs inserted fully into the core. All systems operated as expected. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. Unit 2 is stable and in Mode 3. Steam Generator heat removal is via the class 1 E powered motor driven auxiliary feedwater pump and Atmospheric Dump Valves. The NRC Senior Resident Inspector has been informed.

  • * * UPDATE ON 5/19/21 AT 1351 EDT FROM JASON HILL TO BRIAN P. SMITH * * *

The Unit 2 reactor tripped because of actual High Pressurizer Pressure that occurred as a result of a Main Steam Isolation Signal actuation. At 0337 MST, both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) were made inoperable when the injection valves were overridden and closed in accordance with station procedures. At 0346 MST, in accordance with station procedures, both trains of Containment Spray, LPSI, and HPSI pumps were overridden and stopped, rendering Containment Spray inoperable as well. This represents a condition that would have prevented the fulfillment of a safety function required to mitigate the consequences of an accident per 10 CFR 50.72(b)(3)(v)(D). Additionally, at the time of the Safety Injection Actuation Signal (0315 MST), both trains of Emergency Diesel Generators actuated as required and both 4160 VAC busses remained energized from off-site power. The NRC Senior Resident Inspector has been informed. Notified R4DO (Young)

  • * * UPDATE ON 7/02/21 AT 1943 EDT FROM YOLANDA GOOD TO JEFFREY WHITED * * *

The inoperability of both trains of Low Pressure and High Pressure Safety Injection (LPSI and HPSI) and both trains of Containment Spray (CS) following the Unit 2 reactor trip has been determined to be an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B). Additionally, inoperability of both trains of HPSI resulted in a reportable condition that could prevent fulfillment of its credited safety function to maintain the reactor in a safe shutdown condition per 10 CFR 50. 72(b)(3)(v)(A). The additional reporting criteria were discovered during review of the event and corresponding safety analyses. The NRC Senior Resident Inspector has been informed. Notified R4DO (Werner)

ENS 5526318 May 2021 09:04:00PerryNRC Region 3GE-6This 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. On March 23, 2021, during the performance of the Division 1 ECCS ((Emergency Core Cooling System)) Integrated Test, the Division 1 Diesel Generator (DG) unexpectedly started. While performing the local lockout testing, per the procedure, a step was performed that initiated the unexpected DG start. The following step was to verify the diesel did NOT start. It was later determined that this was a procedural deficiency. The DG started and ran as designed. The DG did not tie to the safety bus as no undervoltage condition was detected. 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, this notification is provided via a 60-day optional phone call in accordance with 10 CFR 50.73(a)(1) instead of submitting a written Licensee Event Report. All affected systems functioned as expected in response to the actuation. The DG was shut down in accordance with plant procedures and the testing procedure corrected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5526117 May 2021 13:12:00Peach BottomNRC Region 1GE-4

(Peach Bottom Atomic Power Station declared an unusual event due to a) "receipt of a single fire alarm in the Unit 2 drywell and the existence of the fire not verified in less than 30 minutes of alarm receipt." The NRC Resident Inspector and State and Local Authorities were notified. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

  • * * UPDATE ON 5/17/21 AT 1423 EDT FROM BRETT HENRY TO HOWIE CROUCH * * *

At 1355 EDT, the licensee terminated the notification of unusual event. The basis for termination was that the smoke has dissipated and there were no signs of fire. The licensee notified State and Local Authorities and the NRC Resident Inspector. Notified R1DO (Grieves), NRR EO (Miller), and IRD MOC (Grant). Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), FEMA NRCC THD (email) and FEMA NRCC SASC (email).

  • * * RETRACTION ON 6/8/2021 AT 1249 EDT FROM JAMES BROWN TO DONALD NORWOOD * * *

Peach Bottom Atomic Power Station is retracting notification EN 55261, 'Peach Bottom - Unusual Event,' based on the following additional information not available at the time of the notification: Following a Unit 2 drywell inspection, analysis of temperature data, and evaluation of equipment in operation; it was concluded that a fire did not exist. The smoke's most likely apparent cause was the result of heating residual oil/grease in the drywell. Peach Bottom reported the condition and entry into the UE initially based on the available information at the time and to ensure timeliness with emergency declaration and reporting notification requirements. The licensee has notified the NRC Resident Inspector. Notified R1DO (Ferdas).

ENS 5525915 May 2021 00:55:00CallawayNRC Region 4Westinghouse PWR 4-LoopAt approximately 1300 CDT on 05/14/2021, a contract worker, who was using a scaffold ladder to access their work area on the iso-phase bus duct system for the main transformers at the Callaway plant, fell approximately 27 feet to the ground. An ambulance was dispatched to transport the individual to a local hospital. Union Electric (Ameren Missouri) subsequently learned that the event caused the individual to have a serious injury that required an overnight hospital stay. This event is reportable to OSHA per 29 CFR 1904.39(a)(2) by the contract worker's employer and is reportable to the Missouri Public Service Commission in accordance with Missouri regulation 20 CSR 4240-3.190(3)(A). This notification is being made to the NRC pursuant to 10 CFR 50.72(b)(2)(xi) due to other government notifications that will occur as a result of a situation related to the health and safety of onsite personnel. The NRC Senior Resident Inspector has been notified of this event. The individual was not working in a contamination area.
ENS 5525613 May 2021 20:35:00South TexasNRC Region 4Westinghouse PWR 4-LoopA non-licensed temporary supervisor had a confirmed positive 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 5525212 May 2021 15:41:00Wolf CreekNRC Region 4Westinghouse PWR 4-LoopSteam Generator
Feedwater
Auxiliary Feedwater
Automatic ScramWith Reactor power at approximately 8 percent following a refueling outage, Steam Generator levels began to oscillate while in automatic control. Manual control of Main Feedwater Regulating Valves was unable to stabilize steam generator levels prior to reaching the "C" Steam Generator Low Level Reactor Trip setpoint. Reactor Trip, Main Feedwater Isolation and Auxiliary Feedwater Actuation automatically actuated. The plant is stable in Mode 3 at Hot Standby. All equipment has responded as expected. The Resident has been contacted.
ENS 5524710 May 2021 21:59:00CallawayNRC Region 4Westinghouse PWR 4-LoopOn May 10, 2021, Callaway determined that a violation of 10 CFR 26.4(c) occurred. A licensee employee was assigned to perform Emergency Response Organization (ERO) duties that required that employee to be subject to the Fitness for Duty (FFD) program. However, the individual had been removed from the FFD program. The individual's unescorted access to the plant had been temporarily removed, but the individual was still required to report to the Emergency Operations Facility in accordance with the emergency plan procedures. The individual's ERO qualification has been deactivated. A review determined that this condition did not apply to any other ERO responders. This discovery is reported pursuant to 10 CFR 26.719(b)(4). The NRC Senior Resident Inspector has been notified of the event.
ENS 552447 May 2021 17:21:00MonticelloNRC Region 3GE-3The following is a summary of information received via email from Monticello Nuclear Generating Plant (MNGP): On March 9, 2021, GE Hitachi (GEH) issued Safety Communication SC 21-02 for PRC 21-02, Transfer of Information, Revision 0, pursuant to 10 CFR 21.21(b). Monticello Nuclear Generating Plant (MNGP) was listed as a potentially affected plant. This is a deviation from the Power Range Nuclear Monitor (PRNM) Licensing Topical Report, NEDC 32410P-A. On May 5, 2021, MNGP completed an evaluation of this deviation and concluded this condition represents a substantial safety hazard in that the condition could result in an Average Power Range Monitoring (APRM) flux reading either below or above the Technical Specification (plus/minus) 2% band. The condition is a defect that is reportable pursuant to 10 CFR 21.21(d)(4). The GEH NUMAC PRNMS was installed in 2009 with System Part No. 299X739NF. Name and Address of the Individual or Individuals Informing the Commission: Thomas A. Conboy, Site Vice President Monticello Nuclear Generating Plant Northern States Power Company - Minnesota 2807 West County Road 75 Monticello, MN 55362 The licensee notified the NRC Resident Inspector.
ENS 552406 May 2021 17:05:00LaSalleNRC Region 3GE-5Secondary containment
Reactor Building Ventilation
This telephone notification is provided in accordance with 10 CFR 50.73(a)(1) to report an invalid actuation of containment isolation valves in more than one system required by 10 CFR 50.73(a)(2)(iv)(A). On March 10, 2021, at 0815 (CST), during the Unit 2 Refueling Outage (L2R18), while performing a test to verify functionality of an isolation relay following replacement of the relay, a Group 4 isolation signal was actuated. The Group 4 isolation logic affects both the Reactor Building Ventilation (VR) and Containment Vent and Purge (VQ) system (for both units). All equipment responded as designed to the Group 4 isolation, including startup of Standby Gas Treatment (SBGT) to maintain secondary containment pressure (for both units). Investigation determined that the cause of the isolation was an inadvertent contact of the self-retracting grip jumper between two adjacent terminals that caused a short to ground and a blown fuse during the test performance. The fuse was replaced and systems restored as needed for the plant condition. The containment isolation was not due to actual plant conditions or parameters meeting design criteria for containment isolation. Therefore, this is considered an invalid actuation. The NRC Resident Inspector has been informed of this notification.
ENS 552396 May 2021 14:00:00North AnnaNRC Region 2Westinghouse PWR 3-LoopAuxiliary Feedwater
Main Condenser
Manual ScramOn May 6, 2021 at 1223 (EDT), Unit 1 was manually tripped from 60 percent power due to degrading main condenser vacuum. Unit 1 was in the process of decreasing power due to increased secondary sodium levels identified earlier in the day. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. This reactor protection actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). Auxiliary Feedwater pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the Auxiliary Feedwater pumps is reportable per 10 CFR 50,72(b)(3)(iv)(A) for a valid actuation of an ESF (Engineered Safeguards Features) system. Decay heat is being removed by the condenser steam dump system. The electrical system is in normal lineup for shutdown conditions. There was no effect on Unit 2 operation. The NRC resident inspector has been notified.
ENS 552355 May 2021 13:22:00Diablo CanyonNRC Region 4Westinghouse PWR 4-LoopA non-licensed employee supervisor had a confirmed positive 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 552313 May 2021 15:39:00FermiNRC Region 3GE-4At 0930 EDT on 5/3/2021, it was determined that during entries into the Fermi 2 Reactor Building Steam Tunnel (RBST) on 4/17/2021, 4/18/2021, and 4/21/2021 that the door was not controlled according to site procedures. The RBST door is credited as a hazard barrier for various high-energy line break (HELB) scenarios. On the identified dates, the RBST door was left open for brief periods during maintenance related activities in the RBST. This condition is not bounded by existing analyses as the door is assumed to be closed throughout a HELB event. The time period that the door was open was less than one hour in each case, as stay times in the room are inherently limited by industrial and radiological conditions. Individuals remained in the area to close the door if needed, but existing analyses do not address the ability to perform those actions under all HELB scenarios. There is no impact to the health and safety of the public or plant personnel as the door is currently closed and latched and access into the area has been restricted to normal ingress and egress per site procedures, which ensures consistency with existing analyses. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). Investigation into the cause is ongoing. Preliminary review of the extent of this condition identified entries into the RBST on other occasions during the past three years where the conditions may also have not been bounded by existing analyses. The additional occasions where the door may have been held open were on 9/22/2018 (MODE 3), 10/26/2018 (MODE 1 ), 11/2/2018 (MODE 1), and 3/21/2020 (MODE 3). Each of these instances was also less than one hour with the exception of the occurrence beginning on 10/26/2018 which lasted approximately 10 hours to support packing leak repairs on a HPCI (High Pressure Coolant Injection) Outboard Isolation Valve. The licensee notified the NRC Resident Inspector.
ENS 552291 May 2021 15:39:00CatawbaNRC Region 2Westinghouse PWR 4-LoopSteam Generator
Feedwater
Reactor Protection System
Auxiliary Feedwater
Automatic ScramAt 0755 EDT, on May 1, 2021, with Unit 2 in Mode 3 at 0 percent (not critical) power, the reactor trip breakers opened during heat-up activities. The trip was not complex, with all systems responding normally post-trip. At 1013 EDT, on May 1, 2021, with Unit 2 in Mode 3 at 0 percent power, an actuation of the Auxiliary Feedwater (AFW) System occurred. The loss of both main feedwater pump turbines caused an AFW auto-start. The 2A and 2B motor driven auxiliary feedwater (MDAFW) pumps automatically started as designed when the loss of both main feedwater pumps signal was received. The cause of the actuation is still being evaluated. Operations responded and stabilized the plant. Decay heat is being removed by the steam generators and discharging steam to the condenser. Unit 1 is not affected. Due to the Reactor Protection System (RPS) actuation while not critical and the actuation of the AFW system, this event is being reported as an eight-hour, non-emergency 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.
ENS 5522430 April 2021 07:38:00Peach BottomNRC Region 1GE-4High Pressure Coolant Injection
Reactor Core Isolation Cooling
On 4/29/21 at 2354 (EDT), an alarm was received for U2 HPCI Inverter Power Failure. (It was) identified that the High Pressure Coolant Injection (HPCI) flow controller had lost power due to a failure of an inverter. Without the flow controller, HPCI would not auto start to mitigate the consequences of an accident; thus, HPCI was declared inoperable. All other emergency core cooling systems and reactor core isolation cooling (RCIC) system remain operable. HPCI is a single train system with no redundant equipment in the same system; therefore, this failure is reportable as an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident per 10CFR50.72(b)(3)(v)(d). The NRC Resident has been informed of this notification.
ENS 5521526 April 2021 19:30:00ColumbiaNRC Region 4GE-5A non-licensed employee supervisor had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident has been notified.
ENS 5521325 April 2021 11:50:00HarrisNRC Region 2Westinghouse PWR 3-LoopAt 1200 EDT on April 25, 2021, planned maintenance activities on the Harris Nuclear Plant Seismic Monitoring System will be performed. The work includes performance of preventive maintenance and system upgrades. The work duration is approximately 10 days and compensatory measures will be in place for seismic monitoring. This is an eight-hour, non-emergency notification for a planned loss of emergency assessment capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii). There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5520522 April 2021 15:41:00Palo VerdeNRC Region 4CEAt 0925 Mountain Standard Time (MST) on April 22, 2021, Palo Verde Nuclear Generating Station staff received reports that Emergency Notification sirens were activated. Current information indicates that the inadvertent activation of the sirens was caused by an offsite agency during performance of a planned silent test that occurred at approximately 0916 MST. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). All sirens remain functional, and the NRC Resident Inspectors have been notified of the issue. Additional notifications will be made as needed.
ENS 5520221 April 2021 09:45:00PilgrimNRC Region 1GE-3On April 21, 2021, at 0752 hours (EDT), an offsite notification was made to the Commonwealth of Massachusetts Department of Environmental Protection (MADEP) in accordance with Regulation 310 CMR 40.0000: Massachusetts Contingency Plan (MCP). The notification documents non-radiological contaminants found above reportable concentrations in select samples collected during site characterization efforts. The reported reportable concentrations were slightly above reporting limits in a soil sample for Per and Polyfluoroalkyl Substances (PFAS), two groundwater sampling locations for PFAS, and isolated instances of metals in groundwater including Arsenic, Vanadium, Lead, Antimony, Beryllium, Cadmium, Chromium, Nickel, and Thallium. This report is being submitted in accordance with 10CFR50.72(b)(2)(xi) based on notification being made to another government agency. Concentrations above reporting limits have been entered into the site's corrective action program. As per MCP, the site will proceed with requirements to implement the phased MCP process. This condition does not represent a threat to station personnel or to members of the general public.
ENS 5520121 April 2021 02:37:00CatawbaNRC Region 2Westinghouse PWR 4-LoopReactor Coolant SystemDuring the performance of reactor vessel closure head (RVCH) examinations, at 2230 EDT on April 20, 2021, it was determined that the Unit 2 RVCH penetration nozzle number 74 did not meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME code case N-729-6 . All other RVCH penetration examinations have been completed per 10CFR50.55a(g)(6)(ii)(D) and ASME code case N-729-6 with no other relevant indications identified. The condition of the Unit 2 reactor vessel head penetration nozzle number 74 will be resolved prior to re-installation of the Unit 2 RVCH. This event is being reported as an eight-hour, non-emergency notification per 10CFR50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5520020 April 2021 15:36:00Browns FerryNRC Region 2GE-4A non-licensed, employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector.
ENS 5519114 April 2021 13:00:00BrunswickNRC Region 2GE-4Primary Containment Isolation System
Reactor Building Ventilation
This 60-day optional telephone notification is being made in lieu of an LER submittal as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). At approximately 1507 EDT on February 17, 2021, during performance of isolation logic periodic testing associated with Primary Containment Isolation System Groups 2 and 6, an invalid actuation of Group 6 Primary Containment Isolation Valves (PCIVs) (i.e., Containment Atmospheric Control/Monitoring and Post Accident Sampling isolation valves) occurred. The Group 6 isolation signal resulted from the reactor building ventilation radiation monitor `B' Channel exceeding the setpoint value. This condition likely resulted from the radiation monitor electronics being impacted by humidity levels, which exceeded the instrument design requirements that developed in the area over time as a result of the Unit 2 reactor building ventilation being secured per the test procedure. The `A' Channel, located in the same plenum, remained steady and below the setpoint value through the entire event. This, along with readings made by a Radiation Protection Technician, confirmed that there was no actual high radiation condition in the reactor building exhaust. Upon returning Unit 2 reactor building ventilation to service, the `B' Channel readings returned to be consistent with the `A' Channel. The PCIVs functioned successfully and the actuation was complete. The actuation was not initiated in response to actual plant conditions, it was not an intentional manual initiation, and there were no parameters satisfying the requirements for initiation of the system. Therefore, this event has been determined to be an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident Inspector was notified.
ENS 5518712 April 2021 09:17:00HatchNRC Region 2GE-4At 2323 EST on 02/12/2021, with Unit 2 in Mode 5 at zero percent power, an actuation of the Group I containment isolation logic occurred during fluid flushing of turbine stop valves. The reason for the actuation was due to a maintenance activity resulting in turbine stop valve movement with no condenser vacuum which is a Group I isolation signal. Two Group I isolation valves, 2B31F019 and 2B31F020, reactor water sample valves, automatically isolated as designed when the system actuation signal was received. The other Group I valves had already been removed from service as part of the refueling outage schedule. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that results in an invalid actuation of the Group I containment isolation system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5518812 April 2021 09:17:00HatchNRC Region 2GE-4At 2320 EST on 02/17/2021, with Unit 2 in Mode 5 at zero percent power, an actuation of the Group 2 containment isolation logic occurred on the inboard valves. The reason for the actuation was most likely due to air entrapment in reactor water level sensing lines following maintenance. Group 2 inboard isolation valves in the drywell floor and equipment drain system and the fission product monitor system automatically isolated as designed. As a corrective action, the variable leg and reference leg of the instrumentation were backfilled with water to ensure all air was removed from the line. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that results in an invalid actuation of the Group 2 containment isolation system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 551859 April 2021 16:08:00VogtleNRC Region 2W-AP1000A contract employee supervisor had a confirmed positive for illegal drugs during a random fitness-for-duty test. The employee's access to the plant had been terminated.
ENS 551726 April 2021 04:32:00PerryNRC Region 3GE-6Reactor Protection System
Control Rod
Manual Scram

At 2149 EDT on April 5, 2021, with the power plant in Mode 2 at zero percent power, an actuation of the RPS system occurred following the decision to abort plant start-up. The reason for the RPS actuation was to align the plant to Mode 3, from Mode 2, following manually inserting all control rods using the Rod Control System. The RPS system initiated as designed when the mode switch was taken from 'Start-up' to 'Shutdown' to align the plant to Mode 3 from Mode 2. 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 RPS system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

  • * * RETRACTION ON 5/12/21 AT 1345 EDT FROM JOHN NAKEL TO KERBY SCALES * * *

This is a retraction of an event notification made on 4/6/2021 at 0432 EST (EN#55172). This event was initially reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the RPS System. This event was later determined to be pre-planned, in accordance with Technical Specifications, and not the result of a significant event, therefore not meeting the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). On the evening of April 4, 2021, while commencing reactor start up, it was determined that control rod withdrawal to add positive reactivity for the start-up would not overcome the negative reactivity of plant heat up. The control room team determined that the proper course of action would be to insert all control rods . The control room briefed and notified the Outage Control Center about its decision, then proceeded to insert all control rods. The control room manually inserted all control rods using the control rod hydraulic system. Following insertion of all control rods, the mode switch was taken to the shutdown position to meet the prerequisites of the procedure for maintaining hot shutdown. This action establishes Mode 3 in accordance with Technical Specifications and aligns the plant to perform the necessary work prior to a plant restart. By placing the mode switch in the shutdown position, a scram signal is generated for 10 seconds. NUREG-1022 offers guidance that states 'Actuations that need not be reported are those initiated for reasons other than to mitigate the consequences of an event.' The actions the operating crew took that night are accurately described by this statement in NUREG-1022 'shifting alignment of makeup pumps or closing a containment isolation valve for normal operational purposes would not be reportable.' In this situation, the Mode switch was taken to shutdown to align the plant to mode 3 for normal operational purposes, and not to mitigate a significant event. When the mode switch was taken to shut-down, RPS initiated as designed, there was no mis-operation or unnecessary actuation. This actuation was determined to be pre-planned, in accordance with Tech Specs, and not the result of a significant event, therefore not meeting the reporting criteria of 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident has been notified. Notified R3DO (McGraw).

ENS 551692 April 2021 14:29:00River BendNRC Region 4GE-6Automatic ScramAt 1017 CDT on April 2, 2021, while operating at 85 percent power, River Bend Station experienced an automatic reactor scram caused by a turbine trip signal. The cause of the turbine trip signal is not known at this time and is being investigated. Reactor water level is being maintained by feedwater pumps and reactor pressure is being maintained by turbine bypass valves. The scram was uncomplicated and all plant systems responded as designed. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72(b)(3)(iv)(A) Specified System Actuation as result of expected post scram level 3 isolations. No radiological releases have occurred due to this event from the unit. The NRC Resident Inspector has been notified of this event.
ENS 5516030 March 2021 14:00:00FermiNRC Region 3GE-4At 1058 EDT on 3/30/2021, during routine pump down activities from the sites Equalization Basin, open to the environment (consisting of groundwater and runnoff), to a sanitary system manhole, there was a backflow from the sanitary system to the environment (nearby grassy area). The total amount of overflow is estimated to be 150 gallons. Fermi 2 Environment is currently investigating and clean-up is in progress and the backflow has stopped. The cause of the backflow is under investigation. As a result of the backflow reaching the environment, reports are being made to the Michigan Department of Environment, Great Lakes, and Energy (EGLE), the Monroe County Health Department, and the local news media. Since these reports are in the process of being made, this is considered a News Release or Notification to Other Government Agencies, therefore this event is reportable under 10 CFR 50.72(b)(2)(xi). The licensee has notified the NRC Resident Inspector.
ENS 5515425 March 2021 13:37:00River BendNRC Region 4GE-6Main Steam Isolation Valve
Steam Jet Air Ejector
Manual ScramOn March 25, 2021 at 0901 CDT, River Bend Station Unit 1 (RBS) was operating at 93 (percent) reactor power (limited by 100 (percent) recirculation flow) when condenser vacuum began to lower due to ARC-AOV1A, Steam Jet Air Ejector Suction Valve, going closed. At 0918 CDT, a manual reactor SCRAM was inserted at approximately 80 (percent) reactor power due to condenser vacuum continuing to lower. After the SCRAM, all systems responded as designed and condenser vacuum was restored by starting a mechanical vacuum pump. The cause of the Steam Jet Air Ejector Suction Valve closure is unknown at this time and being investigated. Currently RBS is stable, and pressure is being maintained using Turbine Bypass Valves. The Main Steam Isolation Valves remained opened throughout the event. This event is being reported under 10 CFR 50.72(b)(2)(iv)(B), as any event or condition that results in actuation of the Reactor Protection System (RPS) when the reactor is critical and 10 CFR 50.72 (b)(3)(iv)(A) Specified System Actuation as result of expected post SCRAM level 3 isolations. No radiological releases have occurred due to this event from the unit. NRC Resident Inspector has been notified of this event.
ENS 5515324 March 2021 22:06:00Turkey PointNRC Region 2Westinghouse PWR 3-LoopAn Unusual Event was declared at Turkey Point Unit 4 Nuclear Generating Station at 2129 EDT on 03/24/2021 due to a Fire Alarm in Containment. The licensee was not able to validate the alarm within 15 minutes. Following containment entry there was not smoke or fire present. At 2214 EDT, Turkey Point Unit 4 Nuclear Generating Station terminated the Unusual Event. The cause of the spurious fire alarm is under investigation. The licensee notified the NRC Resident Inspector. Notified IRD MOC (Gott), NRR EO (Felts), R2DO (Miller), DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
ENS 5514822 March 2021 13:16:00SusquehannaNRC Region 1GE-4At 1005 EDT on 3/22/2021, the control room was notified of a personal medical event in the Radiologically Controlled Area. An ambulance entered Susquehanna plant property at 1019 and exited at 1028 to transport the individual to a local hospital. Ambulance did not enter the Protected Area. The individual was considered potentially contaminated since a complete frisk could not be performed prior to transport. Following transportation to a local hospital, Radiation Protection (RP) technicians confirmed the individual and ambulance were not contaminated. This event is reportable under 10CFR50.72(b)(3)(xii). An Event of Potential Public Interest (EPPI) was made to the Pennsylvania Emergency Management Agency (PEMA) due to an emergency vehicle accessing plant property. The NRC Resident Inspector was notified.
ENS 5514721 March 2021 23:57:00Calvert CliffsNRC Region 1CESteam Generator
Auxiliary Feedwater
Manual ScramAt 2216 EDT on 3/21/2021, Calvert Cliffs Unit 2 was manually tripped from 37 percent power due to lowering level in the 21 Steam Generator. All systems responded per design. Main Feedwater was secured and Auxiliary Feedwater was manually initiated. The Site Senior Resident has been notified. The cause of the lowering level in the 21 Steam Generator is under investigation.
ENS 5514620 March 2021 04:55:00Quad CitiesNRC Region 3GE-3Emergency Diesel GeneratorAt 20:30 CDT on March 19, 2021, with the Unit 1 in Mode 5 at 0 percent power, an actuation of the Unit 1 Emergency Diesel Generator (EDG) occurred during outage activities on Transformer 12 (T-12) resulting in a trip. The cause of the Unit 1 EDG auto-start was bus undervoltage as a result of the T-12 trip. The Unit 1 EDG automatically started as designed when the Bus 14-1 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 Unit 1 EDG. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 5514317 March 2021 12:59:00Watts BarNRC Region 2Westinghouse PWR 4-LoopReactor Protection System
Auxiliary Feedwater
At 1004 EDT on March 17, 2021, with Unit 2 in Mode 1 at 90 percent power, the reactor automatically tripped due to a main turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by the Auxiliary Feedwater and Steam Dump Systems. Unit 1 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). The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 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. All controls rods fully inserted and the electrical system is in normal shutdown alignment. The cause of the turbine trip is being investigated.
ENS 5513814 March 2021 18:05:00Arkansas NuclearNRC Region 4B&W-L-LPFeedwater
Emergency Diesel Generator
Manual ScramOn March 14, 2021, at 1315 CDT, Arkansas Nuclear One, Unit 1(ANO-1) was manually tripped due to degraded voltage and momentary loss of the A-2, non-vital 4160 V Bus in accordance with Abnormal Operating Procedure. All control rods fully inserted. Degraded voltage of the A-2 non-vital 4160 V Bus resulted in de-energizing the A-4 vital 4160 V Bus. Emergency Diesel Generator No. 2, K-4B, started automatically and is currently powering the A-4 vital 4160 V Bus. All other Vital and Non-Vital Buses transferred power automatically to the Startup Transformer No. 1. Offsite power remains energized and available for ANO-1. All other systems responded as designed. The loss of the A-2 Non-Vital Bus is still under investigation. ANO-1 is currently stable in MODE 3 (Hot Standby), maintaining pressure and temperature with Main Feedwater pumps 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 both 10 CFR 50.72(b)(2)(iv)(B) for the Reactor Protection System actuation and 10 CFR 50.72(b)(3)(iv)(A) for the actuation of the Emergency Diesel Generator. The Licensee has notified the NRC Senior Resident Inspector.
ENS 5513713 March 2021 01:11:00Nine Mile PointNRC Region 1GE-2Reactor Recirculation PumpOn March 12, 2021, at 2102 (EST), Reactor Recirculation Pump (RRP) 13 tripped. The cause for the trip is under investigation. Following the RRP trip, the Average Power Ranger Monitors (APRMs) flow bias trips are inoperable due to reverse flow through RRP 13. The APRMs were restored to operable on March 12, 2021, at 2110 (EST) when the RRP 13 Discharge Blocking Valve was closed. This 8-hour non-emergency report is being made based upon requirements of 10 CFR 50.72(b)(3)(v)(A) which states: 'Licensee shall notify the NRC of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition.' The licensee has notified the NRC Resident Inspector.
ENS 5513412 March 2021 12:12:00Calvert CliffsNRC Region 1CEA licensed operator had a confirmed positive alcohol test during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Resident Inspectors have been notified.
ENS 5513210 March 2021 18:26:00Quad CitiesNRC Region 3GE-3On March 10, 2021, Exelon reported an unpermitted release of a radionuclide (i.e., tritium) into the groundwater within the site boundary at Quad Cities Nuclear Power Station to the Illinois Environmental Protection Agency (IEPA) and the Illinois Emergency Management Agency (IEMA) in accordance with Illinois state regulations. There has been no detection of the release beyond the site boundary. The suspected source for the increased groundwater tritium levels is an onsite water storage tank or pipe; however, an investigation is in progress and the exact cause and source is not yet known. The increase in groundwater tritium concentration does not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
ENS 551309 March 2021 11:58:00SequoyahNRC Region 2Westinghouse PWR 4-LoopThis 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 50.73(a)(2)(iv)(A). The event was an invalid actuation of the Unit 1 Containment Ventilation Isolation (CVI) system. On January 11, 2021 at 1152 Eastern Standard Time (EST) with Unit 1 at 100% power, Train 'A' of the CVI System actuated due to an invalid high radiation signal from 1-RM-90-130, Containment Purge Air Exhaust Monitor. The cause of the signal was determined to be a failed sample pump associated with the radiation monitor. 1-RM-90-130 was in service at the time of the invalid signal. The Train 'A' Containment Ventilation Isolation signal was a full actuation of that train and the system functioned as designed. Prior to and following the invalid high radiation alarms, all radiation monitors except 1-RM-90-130 were stable at their normal values; therefore, the CVI was invalid. Control room operators performed appropriate checks and confirmed that all required automatic actuations occurred as designed. The failed pump was replaced and returned to service. This event was entered into the corrective action program as CR 1663398. The NRC Resident Inspector was notified.
ENS 551289 March 2021 08:08:00SusquehannaNRC Region 1GE-4High Pressure Coolant Injection
Primary containment
At 0313 EST on March 9th, 2021, during performance of Unit 1 High Pressure Coolant Injection (HPCI) valve exercising, the inboard vacuum breaker isolation valve did not stroke closed as expected, but remained mid-position. The affected penetration of primary containment was isolated by closing the outboard HPCI vacuum breaker isolation valve. This results in an unplanned inoperability of the Unit 1 HPCI system. This is being reported as a loss of an entire safety function condition in accordance with 10CFR50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Unit 1 is in a 14-day LCO for Tech Spec 3.5.1(d), HPCI inoperability. Tech Spec 3.6.1.3(a), Containment Penetration Valve, was completed with closing the outboard HPCI vacuum breaker isolation valve. The Units are in a normal offsite power line-up.
ENS 551224 March 2021 04:00:00Calvert CliffsNRC Region 1CEReactor Coolant SystemAt time 0323 (EST) on March 04, 2021, it was determined that the Reactor Coolant System (RCS) pressure boundary did not meet the acceptance criteria under ASME, Section XI IWB-3600, "Analytical Evaluation of Flaws." This condition will be resolved prior to plant start up. This event is being reported as an eight hour non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident has been notified.
ENS 551171 March 2021 14:07:00Turkey PointNRC Region 2Westinghouse PWR 3-LoopSteam Generator
Feedwater
Reactor Protection System
Automatic ScramOn 3/1/21 at 1112 EST, with Unit 3 in Mode 1 at approximately 100 percent Rated Thermal Power (RTP), the reactor automatically tripped. Auxiliary Feedwater initiated as designed to provide Steam Generator (S/G) water level control. Emergency Operating Procedures (EOPs) have been exited and General Operating Procedures (GOPs) were entered. Unit 3 is stable in Mode 3 at normal operating temperature and pressure. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. All rods are inserted, decay heat is being removed via S/G through normal secondary systems. The plant is in a normal electrical line-up. The cause of the automatic reactor trip is (unknown at this time and is) being investigated. There was no effect on Unit 4.