LER Code: 0252023001R00 Title: Manual Actuation of Reactor Protection System During Mode 3 Due to an Inadequate Procedure Step Facility/Unit: Vogtle Electric Generating Plant / 3 Event Date: 1/14/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On January 14, 2023. at 0721 EST with Vogtle Electric Generating Plant (VEGP) Unit 3 in Mode 3 at 0 percent power, the Reactor Protection System (RPS) was manually actuated while conducting pre-criticality testing. The manual RPS actuation was in response to low gland steam pressure. The reactor trip breakers were in an open state prior to the RPS actuation. The cause of this event was an inadequate procedure step. This step was revised to require checking reactor trip breakers are not open prior to manually tripping the reactor in response to gland steam pressure below the procedural limit. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as a condition that resulted in manual actuation of the RPS. VEGP Units 1, 2, and 4 were unaffected by this event.
LER Code: 0252023004R00 Title: Manual Reactor Protection System Actuation Due to Condenser Flanges Not Removed During Construction Facility/Unit: Vogtle Electric Generating Plant / 3 Event Date: 5/2/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On May 2, 2023, at 0423 EDT with Vogtle Electric Generating Plant (VEGP) Unit 3 in Mode 1 at 14 percent power, during startup testing, the reactor was manually trippedmanually tripped the reactorloss of feedwater flow. The loss of feedwater was due to high differential pressure across the main feedwater pump suction strainers, which was caused by a secondary plant transient. The secondary plant transient was caused by shipping flanges which had remained installed in the condensate flowpath. The operators responded to ensure plant stability, with decay heat removal by discharging steam to the main condenser using the steam dumps and startup feedwater supplying the steam generators. The corrective actions for this event included removal of the shipping flanges, inspection and cleaning of Main Feedwater Pump suction strainers, and repair of the feedwater heater drain cooler bypass valve that failed due to an increased cycling frequency from blockage of the condensate normal drain path. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) as manual actuation of the Reactor Protection System (RPS)manual actuation of the Reactor Protection System (RPS)). VEGP Units 1, 2, and 4 were unaffected by this event.
LER Code: 2662021001R00 Title: Main feedwater pump trip results in manual reactor trip. Facility/Unit: Point Beach Nuclear Plant / 1 Event Date: 7/31/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
At 1646 on July 31, 2021, with Unit 1 operating in MODE 1 at full power, Operators removed Unit 1 from service by manually tripping the reactor when operators identified control board indications warranting prompt removal of the reactor from service. All control rods fully inserted into the core due to the manual trip. The auxiliary feedwater system started as expected when a valid system actuation occurred after the reactor trip. There was no emergency core cooling system actuation and offsite power was maintained throughout the event. After the reactor trip, equipment anomalies occurred for which the operators effectively responded. The cause of the condition requiring the manual reactor trip was a failure of the main steam generator feedwater pump B motor. Corrective actions included replacement of the main steam generator feedwater pump B motor and additional maintenance activities for anomalies identified during the reactor trip. This event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A)).
LER Code: 2802022003R00 Title: Loss of Emergency Switchgear Room Cooling Due to Use of Incorrect Air Handler Fan V-Belts Facility/Unit: Surry Power Station / 1 Event Date: 9/15/2022 Reported Basis: 50.73.(a)(2)(i)(B), 50.73.(a)(2)(v)(A), 50.73.(a)(2)(vii)
As part of planned maintenance beginning on 7/29/2022 to 9/15/2022, the air handling unit (AHU) fan V-belt failurebelts for both of the Unit 1 Emergency Switchgear Room (ESGR) cooling trains (1-VS-AC-6 and 1-VS-AC-7) were replaced. When the AHU was placed in-service, these V-belts eventually failed resulting in the inoperability of one or both of the AHU(s)). The cause of this event was that the wrong V-belt variant of the BX-99 V-belt ordered (Torq Titan vs. the Grip Notch V-belt expected) and these belts were inadvertently accepted and installed at the station. Contributing to this cause was that there were missed opportunities to recognize that the new V-belts were not the same design as those being replaced. There were no safety consequences because of this event since the maximum temperature in the room stayed well below the design basis upper limit. Planned corrective actions include procedural enhancements and coaching of personnel to ensure that the parts received at the station comply with those ordered.
LER Code: 2822021001R00 Title: 12 Diesel Driven Cooling Water Pump Auto Start Due to Low System Pressure Facility/Unit: Prairie Island Nuclear Generating Plant / 1 Event Date: 1/26/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
On January 26, 2021, the Prairie Island Nuclear Generating Plant (PINGP), 12 Diesel Driven Cooling Water Pump (DDCLP) auto started. Maintenance was in progress on 22 DDCLP and 121 Motor Driven Cooling Water Pump (MDCLP) was aligned as the replacement safety related pump for 22 DDCLP while non-safety related 21 Cooling Water (CL) pump was supporting the system loadssecuring the 12 DDCLP auto started on a sensed low header pressure following the isolation of 22 Cooling Water Strainer for maintenance and subsequent auto initiation of backwash of the other Cooling Water Strainers. This event is reportable under 10CFR 50.73(a)(2)(iv)(A) due to a valid Emergency Service Water system actuation. There were no nuclear safety impacts. The system operated as designed. All systems responded normally. The direct cause of this event was low pressure at the discharge of 12 DDCLP due to the isolation of the Cooling Water Strainer while 22 DDCLP was isolated during low CL system flow conditions. The corrective actions included securing the 12 DDCLP. In addition, procedure updates are planned.
LER Code: 2822021002R00 Title: Loss of Electrical Bus Results in 121 Motor Driven Cooling Water Pump Auto Start Facility/Unit: Prairie Island Nuclear Generating Plant / 1 Event Date: 10/17/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
On October 17, 2021, the Prairie Island Nuclear Generating Plant (PINGP) 2RY Transformer was de-energized when operations personnel opened the 2RSY Reserve Auxiliary Transformer 34.5KV B Disconnect Switch instead of closing the 2RSX Reserve Auxiliary Transformer 34.5KV B Disconnect Switch during restoration of the 2RX Transformer in the substation. The de-energization of 2RY caused a loss of power to the Unit 2 4.16 KV Bus 23. This led to an auto-start of 121 Motor Driven Cooling Water Pump on a sensed low header pressure. This event is reportable under 10CFR 50.73(a)(2)(iv)(A) due to a valid Emergency Service Water system actuation. Outage Shutdown Safety Assessment and Probability Risk Assessment remained green during the event and recovery actions. The cause of the de-energization of 2RY was individual errors during hard match and concurrent verification. The corrective action implemented required all substation switching to be identified as high risk with field supervisor oversight required.
LER Code: 3062021001R00 Title: 22 Turbine Driven Auxiliary Feedwater Pump Actuation Signal Due to Procedure Error Facility/Unit: Prairie Island Nuclear Generating Plant / 2 Event Date: 10/3/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
On October 3, 2021, with Prairie Island Nuclear Generating Plant (PINGP) Unit 2 in Cold Shutdown, the 22 Turbine Driven Auxiliary Feedwater (AFW) Pump received a valid actuation signal. While performing the prerequisite checklists for Surveillance Procedure (SP) 2083ASurveillance Procedure (SP) 2083A ?Å“Unit 2 Integrated SI Test with a Simulated Loss of Offsite Power Train A,??the Train B 22 Turbine Driven Auxiliary Feedwater (AFW) Pump selector switch in the Main Control Room was placed in Shutdown Auto from Manual. This, combined with having the Non-Safety Related 4160 Volt buses 21 and 22 isolated for maintenance, completed the automatic start signal. The turbine and pump did not turn because the equipment was out of service with the steam supply valves closed. This event is reportable under 10CFR 50.73(a)(2)(iv)(A) due to a valid Pressurized Water Reactor Auxiliary Feedwater actuation signal. There was no impact on the ability to maintain safe shutdown conditions. The system operated as designed. The cause of this event was a latent procedure error in SP 2083A. The corrective action was an update to the SP 2083A prerequisite checklist to place the 22 Turbine Driven AFW Pump selector switch to Manual.
LER Code: 3062023001R00 Title: 2GT/XFMR lockout due to failure of lightning arrester caused Unit 2 Reactor Trip Facility/Unit: Prairie Island Nuclear Generating Plant / 2 Event Date: 5/27/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
At 18:19 on May 27, 2023, with Prairie Island Nuclear Generating Plant Unit 2 operating at approximately 100% power and steady-state operation, Unit 2 Generator Transformer (2GT/XFMR) lockout occurred causing the turbine and subsequently reactor to trip. Operators responded to the event in accordance with approved procedures and safely placed the plant in Mode 3Mode 3. The direct cause of this event was the failure of the A-phase lightning arrester on 2GT/XFMR. The failed arrester was sent out for a failure analysis to be performed by a third-party laboratory. May 27, 2023 at 19:20 Event Notification (EN) # 56543 was reported to the NRC as a 4-hour notification under 10 CFR 50.72(b) (2)(iv)(B)actuation of the reactor protection system and a Notification of Unusual Event. On May 28, 2023 at 00:45, updated EN # 56543 reported to the NRC for 8-Hr Non-Emergency report IAW 10 CFR 50.72(b)(3)(iv)(A) for an AFW Actuation. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A)73(a)(2)(iv)(A) due to a Reactor Trip and a valid Pressurized Water Reactor Auxiliary Feedwater actuation signal.
LER Code: 3152022001R00 Title: Manual Reactor Trip Following Manual Turbine Trip due to High Vibrations on Main Turbine Facility/Unit: Donald C. Cook Nuclear Plant / 1 Event Date: 5/24/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On May 24, 2022, at 0414 EDT, Donald C. Cook Unit 1 was operating at approximately 12 percent power, while emerging from the Unit 1 Cycle 31 refueling outage (U1C31) which included significant maintenance on the Unit 1 high vibrations Pressure Turbine. While rolling the Unit 1 Main Turbine, it experienced high vibrations due to an apparent rub, and was manually tripped. Following the Main Turbine trip, the high vibrations persisted, and the Unit 1 Reactor was manually tripped, the Main Steam Stop Valves were closed, and main condenser vacuum was broken. Following the reactor trip and AFW actuation was initiated or required, and Decay Heat Removal was through the Unit 1 Steam Generator Power-Operated Relief Valves (SG PORVs). All required equipment operated as expected, and the trip was not complicated. The initiating event for the Unit 1 manual reactor trip was to allow the control room operators to break Main Condenser vacuum and prevent further potential damage to the Main Turbine due to the high vibrationshigh vibrations. The event is reportable in accordance with 10 CFR 50.73(a)(2)(iv)(A)), System Actuation, due to the valid actuation of the Reactor Protection System (RPS) and the Auxiliary Feedwater System, as a result of the manual reactor trip.
LER Code: 3232022001R00 Title: Unit 2 Reactor Coolant System Pressure Boundary Degradation Facility/Unit: Diablo Canyon Power Plant / 2 Event Date: 10/23/2022 Reported Basis: 50.73.(a)(2)(ii)(A)
At 0830 PDT on 10/23/2022, during routine outage inspections on Unit 2, it was determined that the reactor coolant pressure boundary did not meet American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section XI acceptance criteriaAmerican Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) Section XI acceptance criteria due to finding a through-wall indication at a 2-inch stainless steel socket weld (Weld No. WIB-975D) on the cold leg Loop 1 vacuum refill connecting piping (Line No. 1140), and was therefore reportable. This event is being reported per 10 CFR 50.73(a)(2)(ii)(A) as a degraded condition. The presumed cause of the degradation was vibration-induced fatigue propagation of a flaw initiated at a weld defect. Corrective actions to address the condition consisted of performing a weld repair in accordance with ASME BPVC Section XI Case N-666-1 during the refueling outage. There was no impact to the health and safety of the public or plant personnel.
LER Code: 3272021001R00 Title: Reactor Trip on High Neutron Flux Rate Due to Dropped Control Rods Facility/Unit: Sequoyah Nuclear Plant / 1 Event Date: 5/24/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
On May 24, 2021, at 0915 eastern daylight time (EDT), SQN Unit 1 experienced an automatic reactor trip. Approximately 0.5 seconds before the reactor trip, an unexpected Rod Control Urgent Failure alarm annunciated and Control Bank B, Group 2 rods began to lower. The reactor trip first out alarm indicated the trip was from a Power Range High Neutron Flux Rate detected by the Power Range Nuclear Instruments. During troubleshooting, it was discovered that there were bad pin connections on the backplane of the phase card associated with Control Bank B (the card connection between the Control Bank B, Group 2 stationary gripper phase control card and the backplane of its card cage). All plant safety systems responded as designed. All rods fully inserted as required. The direct cause of the failure has been attributed to the card connection between Control Bank B Group 2 stationary gripper phase control card and the backplane of the card cage. Periodic instruction procedure, PI-674 Periodic Calibration of the Full Length Rod Control Powerprocedure, PI-674 Periodic Calibration of the Full Length Rod Control Power, directs activities to remove phase control cards during calibration which can result in pin deformation. Therefore, the corrective action for this event is to revise the procedure to modify how calibration activities are performed based on learnings obtained from the equipment failure evaluation.
LER Code: 3272021003R00 Title: Exceeded Breach Margin Renders Both Trains of the Auxiliary Building Gas Treatment System Inoperable Facility/Unit: Sequoyah Nuclear Plant / 1 Event Date: 8/20/2021 Reported Basis: 50.73.(a)(2)(i)(B)
On August 20, 2021, at 0905 eastern daylight time main control room operators were notified that auxiliary building secondary containment enclosure (ABSCE) boundary Door A118 was open. This was discovered by Fix-it-now personnel that were in the process of establishing compensatory measures associated with an active breach permit for Door A118. A past operability evaluation determined that ABSCE Door A118 had been left openABSCE Door A118 had been left open without compensatory measures in place on four occasions between August 18 and August 20, 2021. The open door created a breach of the ABSCE boundary that exceeded the allowed breach margin and rendered both trains of the Auxiliary Building Gas Treatment System (ABGTS) inoperable. Additionally, the POE determined that due to the unknown inoperability of the ABGTS, the requirements of Limiting Condition for Operation 3.7.12 Condition E were not met. The cause of the event was miscommunication between the Work Control Center Senior Reactor Operator and Maintenance Services personnel on the requirements of the ABSCE breaching procedure, related to the work being performed, resulting in a breach permit not being issued when required. Corrective actions included revising the ABSCE breaching procedure to clarify exception requirements and require a breach permit if an ABSCE door will be left open greater than the allowable time and briefings with Operations personnel concerning the requirements of the ABSCE breaching procedure.
LER Code: 3282021001R00 Title: Ice Bed Inoperable Due to Exceeding Surveillance Requirement Frequency Facility/Unit: Sequoyah Nuclear Plant / 2 Event Date: 7/25/2021 Reported Basis: 50.73.(a)(2)(v)(C), 50.73.(a)(2)(v)(D)
On July 25, 2021, at 0544 eastern daylight time (EDT), the SQN Unit 2 ice bed temperature monitoring system stopped providing accurate data for the ice bed. The data is used to complete a surveillance instruction (SI) to verify the ice bed temperature does not exceed 27 degrees Fahrenheit (F) as required by Technical Specification (TS) Surveillance Requirement (SR) 3.6.12.1 at a Frequency of every 12 hours. Due to the failure, Main Control Room operators authorized performance of 2-SI-IXX-061-138.0, Backup Ice Condenser Temperature Monitoring2-SI-IXX-061-138.0, Backup Ice Condenser Temperature Monitoring. At 1258, SR 3.6.12.1 was successfully performed; however, the SQN Unit 2 ice bed temperature monitoring system stopped providing accurate data completion time exceeded the specified Frequency plus the 25 percent extension allowed by SR 3.0.2 by 20 minutes (SR 3.6.12.1 had last been completed at 2138 on July 24). Therefore, the Unit 2 ice bed inoperable for 20 minutes from 1238 until 1258. The cause of the event was the failure of the ice bed temperature monitoring system remote scanner power supply failure due to an internal power supply failure. Corrective actions for this event include developing procedural guidance for the response required for a nonfunctional ice bed temperature monitoring system and replacement of obsolete instrumentation in the ice bed temperature monitoring system.
LER Code: 3282021002R00 Title: Turbine Trip Function Inoperable Due to Slow to Close Turbine Throttle Valve Facility/Unit: Sequoyah Nuclear Plant / 2 Event Date: 11/8/2021 Reported Basis: 50.73.(a)(2)(i)(B), 50.73.(a)(2)(v)(D)
On October 28, 2021, Main Steam Throttle Valve 1 (TV-1), associated with the high-pressure turbine and turbine trip function, was replaced during the Unit 2 refueling outage. On November 5, Surveillance Requirement 3.3.2.9 was successfully completed for the Turbine Trip function. On November 8, at 2115 eastern standard time (EST), during turbine overspeed trip testing, it was identified that TV-1 took longer than expected to close on the overspeed trip signal. In response, station personnel developed a support/refute matrix that included validating installation of an orifice block vice a flushing block in TV-1. This led to the discovery that TV-1 was installed with the incorrect block. The flushing block was replaced with the correct orifice block. On November 9, at 0700, SR 3.3.2.9 was successfully completed restoring the valve to operable status. The cause of the event was the failure to follow procedural requirements regarding a quality critical maintenance (QCM) stepprocedural requirements regarding a quality critical maintenance (QCM) step. The maintenance procedure addressing quality critical maintenance requires that supervision/oversight must be present in the field during the execution of steps requiring additional oversight. Contrary to this requirement, The TVA QCM signee performed the step in an independent verification role vice providing oversight and witnessing the installation of the correct orifice plate. Corrective actions included providing a briefing for Outage Services personnel regarding oversight responsibilities associated with QCM steps and revising the vendor's procedure "Critical Step" with photos and guidance to visually confirm and photograph blocks when installing.
LER Code: 3342022001R00 Title: Manual Reactor Trip and Auxiliary Feedwater Actuation due to Heater Drain System Valve Failure Facility/Unit: Beaver Valley Power Station / 1 Event Date: 6/15/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
At 0724 on June 15, 2022 while at approximately 100 percent power, the Beaver Valley Power Station, Unit No. 1 (BVPS-1) reactor was manually tripped due to lowering steam generator water levels. This was due to reduced heater drain flow from the heater drain system (HDS) to the main feedwater pumps (MFPs) when the heater drain receiver normal level control valve, LCV-1SD-106B, experienced a valve plug to stem separation. The Operators manually tripped the reactor as required per the predefined trip criteria for low steam generator water level. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the Reactor Protection System, 10 CFR 50.73(a)(2)(iv)(B)(1), and the automatic actuation of the Auxiliary Feedwater System, 10 CFR 50.73(a)(2)(iv)(B)(6). Corrective actions include replacing the plug, stem, and pin (trim) of LCV-1SD-106B with an enhanced design during the next refueling outage (Fall 2022).
LER Code: 3382021001R00 Title: Manual Reactor Trip on Degrading Condenser Vacuum Due to Piping Failure Facility/Unit: North Anna Power Station / 1 Event Date: 5/6/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
On May 6, 2021 at 1223 hours, Unit 1 was manually tripped from 60% 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. At 1400 hours on May 6, 2021, a 4-hour and 8-hour non-emergency report was made per 10 CFR 50.72(b)(2)(iv)(B) for RPS Actuation (scram) and 10 CFR 50.72(b)(3)(iv)(A) for a valid actuation of an ESF system, respectively. The direct cause of the event was High Cycle Fatigue Piping Failure of the "A" High Pressure Heater Drain Receiver High Level Divert Line that caused damage to condenser tubes and degraded condenser vacuum. Unit 2 continued to operate at 100% power, Mode 1, during the event. The health and safety of the public were not affected by this event.
LER Code: 3392023001R00 Title: Reactor Coolant Pressure Boundary Leak Due to Poor Weld Workmanship Facility/Unit: North Anna Power Station / Unknown Unit Event Date: 10/3/2023 Reported Basis: 50.73.(a)(2)(ii)(A)
On September 11, 2023, at 1558 hours with Unit 2 in Mode 5 at 140 degrees F and 30 psig for a refueling outage, a boric acid leak was discovered on tubing associated with a Pressurizer level transmitter. The leak was not quantifiable as it consisted of a small amount of dry boric acid. Non-destructive examination (NDE) was performed on the leak to determine if it was a through wall leak. On October 3, 2023, at 1154 with Unit 2 in Mode 6 at 100 degrees F and atmospheric pressure, the NDE determined the leak was a through wall leak. This failure constitutes welding or material defects in the primary coolant system that cannot be found acceptable under ASME Section Xl. Therefore, an 8-hour report was made for a degraded condition under 10 CFR 50.72(b)(3)(ii)(A). The direct cause of the weld failure was due to inadequate welding process control by the welder. Unit 1 was not impacted by this event. The health and safety of the public were not affected by this event.
LER Code: 3482021001R00 Title: Automatic Reactor Trip due to Turbine Generator Trip Facility/Unit: Joseph M. Farley Nuclear Plant / 1 Event Date: 4/22/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
At 1450 CDT on April 22. 2021, with Farley Nuclear Plant (FNP) Unit 1 in Mode 1 at 48% power an automatic reactor trip occurred due to a main turbine trip. The main turbine / generator trip was due to data input error in the exciter switchgear software during exciter testing. The trip was not complex with all systems responding normally. Operations stabilized the plant in Mode 3Mode 3. Due to low decay heat load, the Main Steam Isolation valves were closed to minimize cooldown and the Atmospheric Relief Valves were used to remove decay heat. This event was initially reported as event notification 55206. Corrective Actions included correcting the data entry in the exciter switchgear control software and performing causal analysis on the human performance aspect of this event. This event is reportable under 10CFR50.73(a)(2)(iv)(A) due to the automatic actuation of a system listed in 10CFR50.73(a)(2)(iv) (B)). FNP Unit 2 was not affected during this event.
LER Code: 3482023002R00 Title: Residual Heat Removal Pump Inoperable for Longer than Allowed by Technical Specifications Facility/Unit: Joseph M. Farley Nuclear Plant / 1 Event Date: 10/2/2023 Reported Basis: 50.73.(a)(2)(i)(B)
At approximately 10:00 CDT on October 2, 2023, with Joseph M. Farley Nuclear Plant (FNP) Unit 1 in Mode 1 and 100 percent power, the 1B residual heat removal (RHR) pump failed to start on demand during the quarterly surveillance run and was subsequently declared inoperable. The immediate investigation determined that the 1B RHR pump circiut breaker racking release handle had been placed in the "Trip Free" position instead of "Operational". The same 1B RHR pump surveillance run was last completed satisfactorily on July 5, 2023, and there was no work on the 1B RHR pump between July 5th and October 2nd that would have required manipulation of the component's breaker. The most likely cause was an inadvertent misposition of the wrong breaker during the 1B Emergency Diesel (EDG) surveillance the morning of August 28, 2023. Corrective actions included placing the circuit breaker racking release handle in the "Operational" position and addressing human performance issues with the performance management process. The surveillance was completed satisfactorily and the 1B RHR pump returned to operable status. This event is reportable under 10 CFR 50.73(a)(2)(i)(B) for the 1 B RHR pump being inoperable for longer than allowed by Technical Specification 3.5.2. FNP Unit 2 was not affected during this event.
LER Code: 3482023003R00 Title: Unit 1 Emergency Diesel Generator (EDG) lube oil pump outlet coupling leak Facility/Unit: Joseph M. Farley Nuclear Plant / 1 Event Date: 2/26/2023 Reported Basis: 50.73.(a)(2)(i)(B)
On February 26, 2023 at 23:42, with Unit 1 in Mode 1 at 100 percent power, the 1B Emergency Diesel Generator (EDG) circulating lube oil pump outlet coupling connection developed an oil leak resulting in the inoperability of the 1B EDG. The cause of the leak was inadequate restraints on the piping, which allowed movement of the piping and resulted in a failure of the lube oil pressure boundary by the piping separating from the coupling. An additional cause was determined to be inadequate troubleshooting from a similar issue which occurred on November 4, 2022. To correct the issue, the existing lube oil pipe restraint for the 1B EDG was modified, and a new additional restraint was installed. The condition was later eliminated with the implementation of rigid piping. the 1B EDG being inoperabletroubleshooting process deficiencies and implementation weaknesses were corrected by revisions to the procedures that clearly define roles, responsibilities and required documentation for issue response, troubleshooting and corrective actions addressing deficient conditions. This event is reportable under 10 CFR 50.73(a)(2)(i)(B) due to the 1B EDG being inoperable, which failed to meet the LCO in accordance with TS 3.0.1 and 3.8.1 between December 7, 2022, and March 3, 2023.
LER Code: 3642023001R00 Title: Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits Facility/Unit: Joseph M. Farley Nuclear Plant / 2 Event Date: 8/1/2023 Reported Basis: 50.73.(a)(2)(i)(B)
On June 14, 2023, Farley Unit 2 entered Mode 5 to conduct planned maintenance on the 2A Pressurizer Code Safety Valve (PSV) which had indications of leaking. the 2A PSV failed to liftThe PSV was removed from the system and delivered to a vendor to test in accordance with Technical Specification (T/S) 3.4.10, Pressurizer Safety Valves. On August 1, 2023, while in Mode 1 at 100% power level, Farley Unit 2 was informed by the vendor that The 2A PSV was replacedset pressure results were outside of the T/S as-found acceptance criteriaT/S as-found acceptance criteria of 2423 - 2510 psig. During as-found testing the 2A PSV failed to lift. Additional testing by the vendor using in-situ device identified that 2A PSV lifted at approximately 2599 psig. Although this value is greater than T/S limits, the 110% ASME Code limitation and Safety Limit were not exceeded. The failure to lift within the acceptable band was determined to be the result of internal steam cutting of the disc insert and nozzle. The 2A PSV was replaced during the planned outage with a pre-tested spare.
LER Code: 3642023003R00 Title: Pressurizer Code Safety Valve Lift Pressure Outside of Technical Specifications Limits Facility/Unit: Joseph M. Farley Nuclear Plant / 2 Event Date: 10/24/2023 Reported Basis: 50.73.(a)(2)(i)(B)
On October 24, 2023, while at 0% power level and Mode 6 (refueling), it was discovered that a Unit 2 pressurizer code safety valve (PSV), which had been removed during the refueling outage (2R29) and shipped off-site for testing, failed its as-found lift pressure test. The PSV lifted above the Technical Specification (TS) 3.4.10 allowable lift setting value. Setpoint drift of the PSV is the most likely cause of the failure. It is likely that the PSV was outside of the TS limits longer than allowable by the Required Action Statement (15 minutes) during the previous operating cycle in all applicable modes of operation. Therefore, this condition is being reported in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by TS. The PSV was replaced during the October 2023 refueling outage.
LER Code: 3642024001R00 Title: Manual Reactor Trip due to Rising Steam Generator Levels Facility/Unit: Joseph M. Farley Nuclear Plant / 2 Event Date: 11/14/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On November 14, 2023, Farley Nuclear Plant (FNP) Unit 2 was performing an initial reactor startup following Refueling Outage 29 (2R29). With reactor power at approximately 10% power%, steam generator (SG) levels began to rise while in automatic control. Operators took manual control of the main feedwater regulating and main feedwater regulating bypass valves but were unable to control the rise in SG levels. At 1041 CST the operators initiated a manual reactor tripmanual reactor trip prior to reaching high SG level setpoint. All equipment responded as expected. NRC was notified in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to reactor scram and 10 CFR 50.72(b)(3)(iv)(A) for a specified system actuation via ENS 56852. The cause of the event was determined to be Ovation steam and feedwater flow signal errors resulting in the automatic feedwater controls transferring to 'high power mode' at a lower reactor power level than designed (10% versus 20%). This resulted in rising SG Levels and led the operators to manually trip the reactormanually trip the reactor. Following correction of Ovation signal errorsOvation signal errors post trip, operators were able to complete the reactor startup and reached full powerfull power on November 19, 2023.
LER Code: 3692023001R00 Title: Automatic Actuation of the 1A Motor Driven Auxiliary Feedwater Pump Due to Human Error Facility/Unit: McGuire Nuclear Station / Unknown Unit Event Date: 10/18/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On October 18, 2023, an actuation of the A Train Auxiliary Feedwater (AFW) pump occurred while Unit 1 was in Mode 5 during a planned refueling outage. Maintenance technicians were performing PT/0/A/4600/012 A (Train A Reactor Trip Breakers Actuating Device Operational Test for Manual Trip Function) and removed the incorrect fuse in the 1A Solid State Protection System (SSPS) cabinet, temporarily deenergizing Train A SSPS Slave Relays. This caused the 1A AFW Auto-Start Defeat circuit to reset, and in turn, start the 1A AFW pump. The Automatic Recirculation Valve (ARV) provided the required flow path for pump protection. Auxiliary Feedwater was not supplied to the respective Steam Generators due to the motor operated discharge isolation valves (MOVs) being closed as required for current plant conditions. At the time of the 1A AFW actuation, the Main Feedwater (MFW) pumps were in a tripped condition, and decay heat removal was being provided by the 1A Residual Heat Removal (RHR) System. Both MFW pumps being in tripped status provided the logic for the AFW actuation signal once the 1A AFW Auto-Start Defeat circuit was reset. Corrective actions included personnel accountability and reinforcement of human performance tools.
LER Code: 3702022001R00 Title: Valid Actuation of the Unit 2 Reactor Protection System and Auxiliary Feedwater System Facility/Unit: McGuire Nuclear Station,- / 2 Event Date: 2/18/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On 2/18/2022, at 0449 hours, McGuire Nuclear Station Unit 2 Digital Electro-Hydraulic Control System (DEH) experienced a component failurecomponent failure causing a spurious rapid turbine load rejection and un-demanded closure of Throttle Valves (TV) #2 and #4. During the event, the DEH system reverted to manual control, reactor power decreased, and plant systems responded as expected. Due to plant stability concerns, the reactor was manually tripped at 0459. Following the reactor was manually tripped trip, steam generator (SG) levels decreased to the Auxiliary Feedwater System (AFW) autostart setpoint and all AFW pumps started as expected and began providing feedwater to the SGs. This additional SG input lowered reactor coolant average temperature and resulted in a Feedwater Isolation as designed. The reactor trip was uncomplicated with all systems responding normally post-trip. Subsequent analyses determined that an electrical disturbance, originating from within the DEH control system cabinet caused DEH to react in an unexpected manner and produce un-demanded turbine valve movement. A contributing factor is that some DEH circuit cards had embedded batteries operating past their expected service life. These cards did not retain control logic and did not reconfigure as expected when reenergized. Corrective actions included hardening the DEH system by performing power-to-ground audits and replacing failed controller modules, cabinet power supplies, and a subset of circuit cards with embedded batteries. A DEH replacement project is currently in the design phase scheduled for implementation in Fall 2024 (Unit 2) and Spring 2025 (Unit 1).
LER Code: 3702023001R00 Title: Manual Actuation of the Unit 2 Motor Driven Auxiliary Feedwater Pumps Facility/Unit: McGuire Nuclear Station / 2 Event Date: 4/2/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On April 2, 2023, at 0341, Operations entered the Senior Reactor Operator Decision Making Process after receiving a report that called into question the functionality of the operating 2B Main Feedwater (MFW) Pump Recirculation Valve, 2CF-81. At the time, Unit 2 was in Mode 3, the 2B MFW pump was feeding the steam generators, and the 2A MFW Pump Recirculation Valve 2CF-76 was nonfunctional. At 0352 hours, Operations manually started the Auxiliary Feedwater (AFW) Motor Driven Pumps to feed the steam generators to allow corrective maintenance on the MFW System. The AFW Motor Driven Pumps started as designed. Flow to the steam generators was not adversely impacted during this sequence. Subsequent investigation determined that the 2A and 2B MFW recirculation piping experienced abnormal vibration when flow was introduced through valves 2CF-76 and 2CF-81, making the valves susceptible to failure. The recirculation piping had recently been modified to replace the carbon steel elbow piping immediately downstream of the valves with a stainless steel target tee configuration. To mitigate the issue, the target tee was replaced with the original carbon steel elbow arrangement. This event had no impact on the health and safety of the public.
LER Code: 3902021001R00 Title: Control Room Emergency Ventilation System Inoperable due to Main Control Room Door Being Left Open Facility/Unit: Watts Bar Nuclear Plant / 1 Event Date: 3/1/2021 Reported Basis: 50.73.(a)(2)(v)(D)
At 1511 Eastern Standard Time (EST) on March 1, 2021, a Main Control Room (MCR) alarm was received for low control room positive pressure. In response to the alarm, a Control Room Envelope (CRE) door was found ajar and immediately closed. Technical Specification Limiting Condition for Operation (LCO) 3.7.10, Control Room Emergency Ventilation System (CREVS), was declared not met for both trains and Condition B entered. At 1513 EST on March 1, 2021, the alarm cleared, CREVS was declared operable and LCO 3.7.10, Condition B was exited. The direct cause of this event was a human performance error when an individual traversing the control building complex did not fully challenge and ensure a MCR envelope boundary door was properly latched and secured. Contributing to this event, standards had not been properly established, communicated, and reinforced with plant staff related to operation of plant doors. Immediate corrective actions included closing the MCR door and coaching the individual. Actions to reduce the probability of recurrence include the development and communication of management expectations associated with operating plant doors. This condition is being reported as an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident in accordance with 10 CFR 50.73(a)(2)(v)(D).
On February 9, 2022, VCSNS identified kilowatt oscillations in the load of the 'B' Emergency Diesel Generator (EDG)kilowatt oscillations in the load of the 'B' Emergency Diesel Generator (EDG) during the performance of surveillance test procedure STP-125.002B. The 'B' EDG satisfied STP-125.002B acceptance criteria; however, since the kilowatt oscillations were not immediately explainable and based on engineering qualitative review for prompt operability, the 'B' EDG was declared inoperable. The unexpected condition of the 'B' EDG resulted in inoperability of the 'B' EDG, contrary to VCSNS Technical Specifications (TS) Limiting Condition of Operation (LCO) 3.8.1.1.b, from January 16 to February 9, 2022.
On April 29, 2021, while Shearon Harris Nuclear Power Plant, Unit 1 (HNP), was in Mode 6 for a refueling outage, surveillance testing identified that the lift setting of pressurizer safety valve (PSV) 1RC-123 was -1.3 percent (%) from its setpoint, which is outside the Technical Specification (TS) allowed tolerance of +/-1%. The cause of the PSV lift setting being out of tolerance is a personnel error that resulted in inadvertent installation of a valve that had previously demonstrated setpoint drift incompatible with TS criteria. The PSV was replaced with another PSV confirmed to have a lift setting within the TS allowed tolerance and no history of significant drift. The PSV with a history of significant setpoint drift will be retired from service. The +/-1% TS tolerance is more restrictive than the American Society of Mechanical Engineers code allowed tolerance of +/-3%. In addition, the safety analysis for the PSVs assumes a PSV tolerance of +/- 3%, so the as-found PSV lift setting of -1.3% is well within the safety analysis. Therefore, this event had no significance with respect to the health and safety of the public.
LER Code: 4002021005R00 Title: Past Inoperability of Effluent Accident Monitor Exceeded Time Allowed by Technical Specifications Facility/Unit: Shearon Harris Nuclear Power Plant / 1 Event Date: 6/22/2021 Reported Basis: 50.73.(a)(2)(i)(B)
On June 22, 2021, the Waste Processing Building (WPB) Vent Stack 5A High Range Noble Gas Radiation Monitor was found to have incorrect database values used for flow rate measurement, which rendered this monitor inoperable. The database values for this radiation monitor were immediately corrected upon discovery of this condition and operability was restored on June 22, 2021. Following investigation, it was determined that this radiation monitor was inoperable since October 6, 2020. During completion of a surveillance test on this radiation monitor on October 6th, the database values were restored incorrectly. Maintenance personnel did not recognize a note in the Radiation Monitor Data Sheet Library that requires substitute values to be entered when the flow transmitter for this radiation monitor is inoperable. Since this radiation monitor was not known to be inoperable from October 6, 2020, through June 22, 2021, no alternate method for monitoring was implemented and no special report was completed as required by Technical Specifications (TS). The WPB Vent Stack 5A High Range Noble Gas Radiation Monitor is credited as one of two WPB exhaust effluent monitors in TS Table 3.3-10, Accident Monitoring Instrumentation. This event had no significance with respect to the health and safety of the public. Actions to ensure tracking of the flow rate monitor status have been completed.
LER Code: 4002022003R00 Title: Manual Reactor Trip due to Degrading Condenser Vacuum Facility/Unit: Shearon Harris Nuclear Power Plant / 1 Event Date: 4/29/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On April 29, 2022, at 04:05 Eastern Daylight Time, with Shearon Harris Nuclear Power Plant, Unit 1 (HNP), in Mode 1 at 100 percent power, the reactor was manually trippedmanually tripped the reactordegrading condenser vacuum approaching the turbine trip setpoint. The trip was not complex, with all systems responding normally post-trip. The Reactor Protection System and Auxiliary Feedwater System actuated as designed. A monthly swap of the Main Condenser Air Removal System vacuum pumps (CVPs) was in progress at the time of this event. When the 'B' CVP was secured, the suction isolation valve, 1AE-16, and the 'B' CVP discharge check valve, 1AE-19, failed to close, resulting in a rapid decrease in condenser vacuum. The rate of vacuum degradation did not allow time for manual isolation prior to the manual reactor trip. The root cause of this event is the combination of two simultaneous equipment failures, resulting in a large volume of main condenser air-in-leakage. The 1AE-16 failure cause is indeterminate; however, The CVP operating procedure likely cause is an intermittent malfunction of the control relay associated with the solenoid operated valve (SOV) that controls 1AE-16 actuator operation. The 1AE-19 failure cause is attributed to poor valve construction with the disk arm not being centered, resulting in contact between the disc and valve body. Corrosion product buildup along this contact surface caused valve binding. The CVP operating procedure was revised to close the CVP manual suction isolation valve prior to securing the CVP. Control relays and a circuit breaker mechanism operated cell assembly for 1AE-16 were replaced. 1AE-19 was replaced. The 1AE-16 SOV and the ??A??CVP discharge check valve will be replaced.
LER Code: 4002022004R00 Title: Both Trains of High Head Safety Injection Inoperable Facility/Unit: Shearon Harris Nuclear Power Plant / 1 Event Date: 5/2/2022 Reported Basis: 50.73.(a)(2)(i)(B), 50.73.(a)(2)(v)(A), 50.73.(a)(2)(v)(D)
On May 2, 2022, at 02:26 Eastern Daylight Time, with Shearon Harris Nuclear Power Plant, Unit 1 (HNP), in Mode 1 at 100 percent power, testing was being performed in accordance site procedure OST-1093, “Chemical Volume Control System/ Safety Injection System Operability Train B.”.??When Charging Safety Injection Pump (CSIP) discharge cross-connect valve 1CS-220 was stroked closed, the Main Control Room (MCR) received alarm ALB-008/2-1, Reactor Coolant Pump (RCP) seal injection low flow, and seal injection flow was lowering to zero. Valve 1CS-220 was immediately restored to the open position and seal injection flows recovered to normal within approximately 23 seconds. Further investigation identified that the ??B??CSIP discharge valve, 1CS-197, was locked shut from prior post-maintenance testing instead of being in its required open position, the result of not validating assumptions related to a ??B??CSIP clearance. With 1CS-197 locked shut when 1CS-220 was stroked closed, both ??A??Train high head safety injection (HHSI) and ??B??Train HHSI were inoperableA??Train high head safety injection (HHSI) and ??B??Train HHSI were inoperable until 1CS-220 could be restored to open. This was a violation of Technical Specification (TS) requirementsTechnical Specification (TS) requirements to have at least one CSIP operable to meet TS 3.1.2.2, TS 3.1.2.4, and TS 3.5.23.1.2.2, TS 3.1.2.4, and TS 3.5.2 in Modes 1, 2, and 3. If a condition occurred that initiated a safety injection signal while 1CS-220 was closed in accordance with test procedure OST-1093test procedure OST-1093, an operator in the MCR would procedurally restore the valve to its open position, restoring the ??A??CSIP and its discharge path for the Emergency Core Cooling System to operable. This event did not impact plant safety and there was no actual safety consequence on the health and safety of the public as a result of this event.
LER Code: 4002022005R01 Title: Manual Reactor Trip due to ?Å“B??Condensate Pump Motor Failure Facility/Unit: Shearon Harris Nuclear Power Plant / 1 Event Date: 8/28/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On August 28, 2022, at 03:29 Eastern Daylight Time, with Shearon Harris Nuclear Power Plant, Unit 1 (HNP), in Mode 1 at 100 percent power, the reactor was manually tripped in response to an automatic trip of the ?Å“B??main feedwater pump (MFP) caused by a trip of the ?Å“B??condensate pump (CP). The ?Å“B??CP tripped following an electrical failure of its motor. The trip of the ?Å“B??CP resulted in subsequent trips of the ?Å“B??condensate booster pump and the ?Å“B??MFP as designed. In accordance with plant procedures, the operations crew manually tripped the reactor upon the trip of a MFP with initial reactor power greater than 90 percent. The ?Å“A??MFP tripped shortly after the manual reactor trip due to a low suction pressure condition created from the condensate and feedwater transient. The auxiliary feedwater pumps auto-started on low steam generator level as designed. Safety systems functioned as required. This event had no impact on the health and safety of the public. The ?Å“B??CP motor electrical failure was caused by a lightning strike. The design of CP motors and associated surge protection did not prevent motor damage and failure from a lightning strike. The CP motor was replaced ?Å“B??CP motor was replaced. The ?Å“A??CP and ?Å“B??CP motor surge protectors were inspected and tested, with no deficiencies identified. A study will be completed to determine changes needed to prevent future lightning induced failures of the CP motors.
At 20:50 Eastern Daylight Time on October 27, 2022, with Shearon Harris Nuclear Power Plant, Unit 1 (HNP), in Mode 3, conditions existed such that all auxiliary feedwater (AFW) pumps were declared inoperableall auxiliary feedwater (AFW) pumps were declared inoperable. The capability to throttle flow to the 'B' steam generator was not maintained due to improper operation of a flow control valve, 1AF-51improper operation of a flow control valve, 1AF-51, which is located in the discharge piping line to the `B' steam generator from the common header of the motor-driven AFW pumps. Since 1AF-51 operation can impact AFW flow control from the common header of the motor-driven AFW pumps, both motor-driven AFW pumps were declared inoperable. The turbine-driven AFW pump was inoperable at the time of this event due to incomplete post-maintenance testing following planned maintenance. The motor-driven AFW pumps were able to supply discharge flow to the steam generators during this event since the 1AF-51 failure never impacted the valve's ability to open. Based upon the declared inoperability of all three AFW pumps, this condition was reported on October 28, 2022, under 10 CFR 50.72(b)(3)(v) as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (D) mitigate the consequences of an accident, by event notification number 56186. This event had no impact on the health and safety of the public. The cause of the 1AF-51 inoperability was an actuator malfunction. The actuator for 1AF-51 was replaced to restore proper flow control capability.
LER Code: 4002022007R01 Title: Automatic Reactor Trip due to Loss of Power from the ??A??Auxiliary Bus Facility/Unit: Shearon Harris Nuclear Power Plant / 1 Event Date: 10/30/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
At 06:53 Eastern Daylight Time, with Shearon Harris Nuclear Power Plant, Unit 1, in Mode 1, at sixteen percent power following the completion of a refueling outage, an automatic reactor trip occurred due to an under-voltage condition on the 'A' reactor coolant pump (RCP) and the 'C' RCP that resulted from a loss of power from the 'A' auxiliary bus from the ??A??auxiliary bus. Power was lost from the 'A' auxiliary bus while operators were performing a procedure to transfer power from the 'A' start-up transformer to the 'A' unit auxiliary transformer (UAT)operators were performing a procedurea procedure to transfer power from the 'A' start-up transformer to the 'A' unit auxiliary transformer (UAT)the 'A' main feedwater pump (MFP) tripped loss of power from the ??A??auxiliary bus, the ??A??main feedwater pump (MFP) tripped. The ??B??MFP was not in service and with the loss of the last running MFP, the auxiliary feedwater system actuated as designed. Safety systems functioned as required. This event did not impact public health and safety. An investigation determined that the current transformers (CTs) in the ??A-3??cubicle were mis-wired, resulting in a differential current protective relay sensing the equivalent of a differential current in the ??C??phase on the ??A??auxiliary bus. When current was applied through the ??A??UAT to the ??A-3??cubicle, the differential current protective relay actuated, which actuated the lockout of the A auxiliary bus. The wiring error occurred during maintenance activitieswiring error occurred during maintenance activities on the CTs that were reinstalled during the refueling outage. Corrective actions involved rewiring of the CTs in accordance with design.
LER Code: 4002022008R00 Title: Automatic Actuation of Auxiliary Feedwater System Facility/Unit: Shearon Harris Nuclear Power Plant / 1 Event Date: 10/30/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On October 30, 2022, at 20:57 Eastern Daylight Time, Shearon Harris Nuclear Power Plant, Unit 1 (HNP), was in Mode 3 at 0% power. An automatic actuation of the auxiliary feedwater (AFW) system occurred during an attempt to start the ??B??main feedwater pump (MFP). Prior to the event, the ??A??MFP was removed from service due to its power supply, the ??A??electrical auxiliary bus, being under clearance. The ??B??MFP switch was taken to start and the ??B??MFP breaker did not close, which resulted in a valid AFW system actuation signal for loss of the last running MFP. The ??A??and ??B??motor-driven AFW (MDAFW) pumps were in service for steam generator inventory control prior to the AFW system actuation and remained in service following the event. The MDAFW flow control valves (FCVs) automatically opened as designed when the AFW actuation signal was received. Operations took action to control the AFW flow rate by throttling the FCVs. This event is being reported in accordance with 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in an automatic actuation of the AFW system. The AFW system responded to plant conditions as designed. The cause of the ??B??MFP failure to start was an open limit switch alignment deficiency on the ??B??MFP recirculation line condenser isolation valve. Misalignment of the limit switch prevented meeting the start permissive for the ??B??MFP. The limit switch on the ??B??MFP recirculation line condenser isolation valve was adjusted to obtain proper alignment.Procedural guidance will be revised to require start permissive verification checks prior to starting a MFP. This event did not impact public health and safety.
LER Code: 4122021002R00 Title: Unit 2 Automatic Reactor Trip and Automatic Actuation of the Auxiliary Feedwater System Facility/Unit: Beaver Valley Power Station / 2 Event Date: 10/5/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
At 0632 on October 5, 2021, Beaver Valley Power Station, Unit 2 (BVPS-2) automatically tripped due to a loss of the Reactor Trip System Interlock for Power Range Neutron Flux (P-10) and associated power range trip block signals. BVPS-2 was at approximately 90% power and in an end of cycle coast down. The Auxiliary Feedwater System automatically started as designed. The apparent cause was a premature failure of a Train A Solid State Protection System (SSPS) universal logic board (ULB) due to manufacturing defects with the solder joints which may have resulted in an intermittent loss of connection. This is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in automatic actuation of the Reactor Protection System per 10 CFR 50.73(a)(2)(iv)(B)(1)10 CFR 50.73(a)(2)(iv)(B)(6) and an automatic actuation of the Auxiliary Feedwater System per 10 CFR 50.73(a)(2) (iv)(B)(6). Corrective actions include replacement of the ULB, vendor testing of several ULBs within the SSPS, review of the final vendor test report for additional actions, and establishment of maintenance plans for periodic testing of the SSPS ULBs at both units.
LER Code: 4122021004R00 Title: Unit 2 Manual Reactor Trip Facility/Unit: Beaver Valley Power Station / 2 Event Date: 11/12/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
At 1007 EST on November 12, 2021, with Unit 2 in Mode 1 at approximately 17 percent power following a refueling outage, the reactor was manually tripped due to increasing steam generator water levels due to oscillating Main Feedwater Pump Recirculation Valves. The oscillation of the valves led to a steam generator water level transient that met predefined reactor trip criteria. The direct cause was the key lock switches for the recirculation valves were left in AUTO allowing the valves to modulate based on flow indication. The apparent cause was the unit supervisor did not properly read a procedure step from a startup procedure and the step was not performed. This is reportable under 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the Reactor Protection System per 10 CFR 50.73(a)(2)(iv)(B)(1). Corrective actions are procedural clarifications and cautions.
LER Code: 4122021005R00 Title: Manual Reactor Trip and Auxiliary Feedwater Actuation due to trip of a Main Feedwater Pump Facility/Unit: Beaver Valley Power Station / 2 Event Date: 11/17/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
At 1313 on November 17, 2021, the Beaver Valley Power Station, Unit No. 2 (BVPS-2) reactor was manually tripped while at approximately 100 percent power following a trip of a main feedwater pump (MFP) caused by a loss of adequate suction pressure. A transient in the Heater Drain System occurred from a level increase in a Second Point Heater Drain Receiver Tank. The level fluctuation resulted in an unexpected actuation of the tank's low-low level switch, which tripped the Train A Heater Drain and Separator Drain Receiver Drain Pumps. The trip of these pumps reduced flow and lowered MFP suction pressure. The standby condensate pump was not available to automatically start on low MFP pressure. When the MFP tripped, operators manually tripped the reactormanually tripped the reactor per procedure. With the spare condensate pump unavailable, MFP suction pressure was unable to be restored and the manual reactor trip was procedurally required. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in a manual actuation of the Reactor Protection Systemmanual actuation of the Reactor Protection System, 10 CFR 50.73(a)(2)(iv)(A)10 CFR 50.73(a)(2)(iv)(B)(1)), and an automatic actuation of the Auxiliary Feedwater System, 10 CFR 50.73(a)(2)(iv)(B)(6)). Corrective actions include changes to procedural controls for conditional single point vulnerabilitiesconditional single point vulnerabilities and condensate pump alignments above 40% power.
LER Code: 4122023001R00 Title: Indications Identified During Reactor Vessel Head Inspection Facility/Unit: Beaver Valley Power Station / 2 Event Date: 4/22/2023 Reported Basis: 50.73.(a)(2)(ii)(A)
On April 22, 2023, during the BVPS-2 twenty-third refueling outage (2R23), it was determined that two indications identified on the reactor pressure vessel head vent line did not meet the applicable acceptance criteria. Because the indications could not be dispositioned as acceptable per American Society of Mechanical Engineers (ASME) Code Section XI in a Reactor Coolant System pressure boundary, it was reported under Event Notification 56485 as a degraded condition per 10 CFR 50.72(b)(3)(ii)(A) as a degraded condition, and is being reported under 10 CFR 50.73(a)(2)(ii)(A)). Primary Water Stress Corrosion Cracking (PWSCC) of the Alloy 600 penetration tube material was determined to be the cause of the identified flaws. The vent line penetration was repaired in accordance with NB-3200 of ASME Section III, ASME Code Case N-638-10, and relief request 2-TYP-4-RV-06.
LER Code: 4122023002R00 Title: Automatic Actuation of Auxiliary Feedwater System Facility/Unit: Beaver Valley Power Station / 2 Event Date: 5/19/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
At 0852 on May 19, 2023, with Beaver Valley Power Station, Unit No. 2 (BVPS-2) in Mode 3 at 0 percent power, an actuation of the Auxiliary Feedwater (AFW) System occurred. Both Motor Driven Auxiliary Feedwater (MDAFW) Pumps automatically started as designed when a 'Loss of Both Main Feedwater Pumps (MFP)' signal was received upon the start failure of the 'B' MFP, 2FWS-P21B. The 2FWS-P216 failure to start was because the lube oil (LO) pressure start permissive was not met when the LO system relief valve, 2FWS-RV205B, setpoint was set too low during the refueling outage. Also, the MDAFW pump auto-start signal was able to be generated during a plant condition where it was not necessary. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) as an event that resulted in the automatic actuation of the AFW System, 10 CFR 50.73(a)(2)(iv)(B)(6). Planned corrective actions include adjusting 2FWS-RV2056 setpoint to restore LO system margin and start-up procedures will be revised to disable the MDAFW Pump auto-start circuit when plant conditions do not require the auto-start function. Also, maintenance documents will be enhanced to clarify the desired setpoint value.
LER Code: 4142021001R00 Title: Indication Identified During Reactor Pressure Vessel Head Nozzle Penetration Inspection Facility/Unit: Catawba Nuclear Station / 2 Event Date: 4/20/2021 Reported Basis: 50.73.(a)(2)(ii)(A)
On April 20, 2021, during the performance of the Catawba Nuclear Station Unit 2 reactor vessel closure head (RVCH) examinations, it was determined the Unit 2 RVCH nozzle penetration 74 did not meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME code case N-729-6. A relevant ultrasonic testing (UT) leak path indication was identified on Unit 2, nozzle penetration 74. The indication was confirmed in the J-groove weld with supplemental eddy current and liquid penetrant surface examinations. A bare metal visual examination of the RVCH was performed with no visual evidence of head penetration leakage detected. The cause of the Unit 2 RVCH nozzle penetration 74 relevant indication is due to Primary Water Stress Corrosion Cracking (PWSCC) from the Alloy 82/182 weld application in the primary system. Unit 2 RVCH nozzle penetration 74 was repaired in accordance with the NRC approved Westinghouse embedded flaw repair method (WCAP-15987-P-A) and Duke Energy relief request RA-21-0145. This event was reported to the NRC as an eight-hour, non-emergency Event Notification Number 55201 on April 21, 2021, per 10 CFR 50.72(b)(3)(ii)(A)). This report is being submitted in accordance with 10 CFR 50.73(a)(2)(ii)(A)), "any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degradedseriously degraded."
LER Code: 4142021002R00 Title: Valid Actuation of the Unit 2 Reactor Protection System and Auxiliary Feedwater System Facility/Unit: Catawba Nuclear Station / 2 Event Date: 5/1/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
At 0755 hours on May 1, 2021, during reactor coolant system temperature cross calibration testing activities, with Unit 2 in Mode 3 at 0 percent (not critical) power with all control rods fully inserted, Catawba Unit 2 received an automatic reactor trip signal on Overpower Delta Temperature. The trip was not complex, with all systems responding normally post-trip. As a result of the reactor trip signal, a feedwater isolation signal was generated. At 1013, on May 1, 2021, during feedwater recovery actions, with Unit 2 in Mode 3 at 0 percent power, the 2B Feedwater Pump Turbine tripped, which caused an actuation of the Auxiliary Feedwater (CA) system. The cause of the reactor trip signal was due to inadequacies in the reactor coolant system temperature cross calibration testing equipment and procedure. The cause of the CA system actuation was due to inadequate oversight and standards adherence of the Unit 2 feedwater isolation recovery actions. Corrective actions include procedural revisionsprocedural revisions and personnel accountability actions. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A)). There was no impact to the health and safety of the public.
LER Code: 4142021003R00 Title: Containment Penetration Failure Resulting in Integrated Leak Rate Test Failure Facility/Unit: Catawba Nuclear Station / 2 Event Date: 5/25/2021 Reported Basis: 50.73.(a)(2)(ii)(A)
At 1751 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. The LLRT was performed during the C2R24 refueling outage at which time primary containment was not required to be operable. The leakage assigned to the penetration also resulted in total containment leakage exceeding the allowed overall leakage (L ). The valves were repaired and retested a satisfactorily prior to entering the mode of applicability. The most likely cause of the leakages from 2MISV5230 and 2MISV5231 is the degradation of the elastomer seat aggravated by the presence of a buildup of debris in the seating area and overly conservative testing methodology. This event was reported to the NRC as an eight-hour, non-emergency Event Notification Number 55275 on May 25, 2021, per 10 CFR 50.72(b)(3)(ii)(A), ?Å“event or condition that the nuclear power plant, including its principal safety barriers, being seriously degraded??
LER Code: 4142022001R00 Title: Manual Reactor Trip and Auxiliary Feedwater Start due to Misaligned Control Rods Facility/Unit: Catawba Nuclear Station / 2 Event Date: 4/23/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On April 23, 2022, at 0224 hours, with Unit 2 in Mode 1 at 100 percent power, two control rods partially dropped during control rod testing resulting in misalignment, which required a manual reactor trip in accordance with plant procedure. All control rods fully inserted into the core following the manual reactor trip. All safety systems functioned as expected. The Auxiliary Feedwater system actuated as designed to provide makeup flow to the steam generators. This event did not impact public health and safety. The cause of the event was an intermittent high resistance connection in the rod control system moveable regulation circuitry resulting from insulation inappropriately inserted into a crimped lug connection. The connection was made during initial plant construction and degraded over time. This event was reported to the NRC as a four-hour, non-emergency Event Notification Number 55856 on April 23, 2022 per 10 CFR 50.72(b)2(iv)(B), Actuation of the Reactor Protection System and an eight- hour non-emergency notification under the same Event Notification per 10 CFR 50.72(b)(3)(iv)(A)), Specified Safety System Actuation.
LER Code: 4142022003R00 Title: Manual Reactor Trip and Auxiliary Feedwater System Actuation Facility/Unit: Catawba Nuclear Station / 2 Event Date: 10/24/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On 10/24/2022 at 0857 EDT, with Unit 2 in Mode 1 at approximately 9% power, the reactor was manually tripped the reactorprocedural guidance due to a 213 train main feedwater pump trip. The trip was not complicated with all systems responding normally post-trip. The auxiliary feedwater system started automatically as expected. The 213 main feedwater pump tripped on loss of condenser vacuum due to a deficiency in the unit startup procedureunit startup procedure. The procedure did not ensure the water/condensate in the main steam equalization header was properly drained prior to opening the main steam supply to the feedwater pump turbine (CFPT) valve. This introduced water into the 213 main feedwater pump turbine condenser causing a loss of vacuum which resulted in the 213 CFPT trip and manual reactor trip. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A)). There was no impact to the health and safety of the public.
LER Code: 4242023002R00 Title: Manual Reactor Trip Due to Failure of a Heater Drain Pump and Standby Condensate Pump Facility/Unit: Vogtle Electric Generating Plant / 1 Event Date: 8/22/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On August 22, 2023 at 1724 EDT, with Unit 1 in Mode 1 and 100 percent power, the Control Room Operators manually tripped the reactormanually tripped the reactor due to lowering Main Feed pump suction pressure and Steam Generator water levels. The initiating event was Heater Drain pump 'B' motor tripping on overcurrent. The cause of the reactor trip was the failure of Condensate pump 'A' to start automatically or manually. The trip was not complex, and all other systems responded normally post trip. At 1735, the Control Room was notified of the Heater Drain pump 'B' motor being on fire, and the fire was extinguished by the onsite fire brigade at 1807. Units 2, 3, and 4 were not affected. The operating crew performed as expected. The unit was stabilized in Mode 3 with decay heat removed through the Steam Dumps to the Condenser. The Heater Drain Pump 'B' motor was replaced and the Condensate Pump 'A' differential and lockout relays were replaced. Due to the actuation of the Reactor Protection System and the Auxiliary Feedwater System, this is being reported under 10 CFR 50.73 (a)(2)(iv)(A)).
LER Code: 4432023002R00 Title: Automatic Actuation of Emergency Service Water during Testing due to a Defective Relay Facility/Unit: Seabrook Station / Unknown Unit Event Date: 5/2/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On May 2, 2023, during planned Engineered Safety Features Actuation System Testing, a Tower Actuation (TA) signal occurred. The TA occurred due to the failure of the first level undervoltage load shedding scheme to shed loads from the 4KV bus prior to the Service Water (SW) Ocean Pumps coasting down following the opening of the 4KV supply breaker. With no power to the bus, and the SW Ocean Pumps breakers still closed, a low SW pump discharge pressure condition occurred, resulting in the TA logic being satisfied. Troubleshooting determined that a defective relay contact in the first level undervoltage load shedding scheme caused the failure. The relay was replaced, and the system was returned to service. There was minimal impact to the station due the unplanned actuation, therefore this event had no impact on the health and safety of the public. In addition, there were no other Structures, Systems, or Components (SSCs) that contributed to this event. This event is being reported pursuant to 10 CFR 50.73(a)(2)(iv)(A) for actuation of the emergency service water systemactuation of the emergency service water systemactuation of the emergency service water system that does not normally run that serves as the ultimate heat sink.
LER Code: 4452022001R00 Title: Reactor Trip Due To Turbine Trip Greater Than 50% Power Facility/Unit: Comanche Peak Nuclear Power Plant / Unknown Unit Event Date: 9/5/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On September 05, 2022, Unit 1 was operating at approximately 100% power. At 2345, a spurious failure of the channel 1 shaft displacement signal coupled with a previous failure of the channel 2 shaft displacement signal caused a trip of the Unit 1 turbine through the turbine trip logic. This led to a Unit 1 reactor trip due to a turbine trip with reactor power greater than 50%. Entry into the Emergency Plan was not required. The cause of this event was the spurious, momentary failure of the channel 1 main turbine shaft displacement signal coupled with the complete failure of the channel 2 main turbine shaft displacement signal. These excessive displacement signals satisfied the shaft displacement trip logic for the turbine. The cause of the signal failures was determined to be inadequate protection of the shaft displacement probe cabling against chafing. Corrective actions include replacement of all three channels of shaft probes, cables, and drivers, as well as defeating the shaft displacement trip through the next refueling outage. Additional corrective actions included revising the shaft displacement probe installation instructionsshaft displacement probe issueshaft displacement probe installation instructions and developing permanent modifications for the routing of the turbine shaft displacement probes for implementation during the next refueling outage. All times are in Central Daylight Time (CDT).
LER Code: 4452023001R01 Title: Automatic Reactor Trip Due to Lo-LoSteam Generator Level Facility/Unit: Comanche Peak Nuclear Power Plant / Unknown Unit Event Date: 6/16/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On June 16, 2023, unit 1 was operating at 100% power. At 18:22, the 1B Main Feedwater Pump (MFP) began indicating oscillating speed. The oscillations increased in severity and became unresponsive to manual control input. Due to the loss of control, the 1B MFP was manually tripped as it approached the overspeed setpoints. Following the manual trip of the 1B MFP the Unit responded with an automatic runback to 700MWe. Following the runback, the combined effects of the entire transient resulted in an automatic reactor trip due to Steam Generator #4 Lo-Lo water level. The direct cause of this event was a degraded servo on the 1B MFP was manually tripped. Vendor analysis of the component did not identify a specific failure mechanism and the cause is inconclusive. Based on engineering evaluation, the most probable cause is particulate accumulation in the servo valve due to historical water intrusion events into the 1B lube oil system. An additional supplement will be submitted if further analysis has more conclusive results. Corrective action included replacement of the servoservo. This LER is a supplement to the original LER 23-001-00 submitted on 8/15/2023. All times are in Central Daylight Time (CDT).
A pressurizer safety valve (PSV) was removed and tested during the Unit 1 fall 2021 refueling outage (B1 R24) under the Inservice Testing (1ST) program. The as-found lift setting was outside the Technical Specifications (TS) 3.4.10 and 1ST program limits. This required the removal and testing of the remaining two PSVs. The remaining two PSVs were also outside the TS and 1ST program limits. The three PSVs were replaced during the outage. An engineering analysis on the effects of these valves lifting at the as- found settings concluded that all acceptance criteria in the Updated Final Safety Analysis Report Chapter 15 analyses are still met. This condition of multiple pressurizer safety valves being outside of their required lift setting tolerance band is reportable in accordance with 10 CFR 50.73(a)(2)(i)(b), Any operation or condition which was prohibited by the plant's Technical Specifications..."
LER Code: 4542023001R01 Title: OA Control Room Ventilation Inoperable due to Jumpers left on OPRO31J and OPRO32J Facility/Unit: Byron Station / 1 Event Date: 3/16/2023 Reported Basis: 50.73.(a)(2)(i)(B)
On March 16, 2023, at 14:40 CDT, the OA Control Room Ventilation (VC) failed to actuate when performing the 1A Diesel Generator (DG) sequencer testing due to installed jumpers on OPRO31J and OPRO32J, Main Control Room Outside Air Intake A Monitors. Jumpers that were installedjumpers were not removed as expectednot removed as expected at the conclusion of the bus outage. These jumpers prevented OPRO31J and OPRO32J, Main Control Room Outside Air Intake A Monitors from causing OA Train VC actuations when required during 1A DG sequencer testing. OPRO31J and OPRO32J were declared inoperable. Once identified, the jumpers were removed from OPRO31J and OPRO32J and the monitors were restored to operable status. This event is reportable in accordance with 10 CFR 50.73(a)(2)(i)(B) for any operation or condition which was prohibited by the plant's Technical Specifications.
LER Code: 4562021001R00 Title: Train A and B Source Range Neutron Flux Trip Functions Bypassed During Plant Startup Facility/Unit: Braidwood Station / 1 Event Date: 4/23/2021 Reported Basis: 50.73.(a)(2)(i)(B)
On April 23, 2021 at 1216 hours, with Unit 1 in Mode 2 during reactor startup activities, control room personnel identified that both Train A and Train B source range neutron flux reactor trip functions were bypassed, and immediately restored by taking the bypass switches to normal. The condition existed since April 21, 2021 at 0738 hours with the unit in Mode 5. In accordance with Technical Specification (TS) Limiting Condition for Operations (LCO) 3.3.1, Reactor Trip System (RTS) Instrumentation, the source range neutron flux channels are required in Modes 2, 3, 4 and 5 in accordance with TS LCO 3.0.4 Table 3.3.1-1. A review determined periods where TS 3.3.1 Conditions were not met, and of mode changes not in accordance with TS LCO 3.0.4. The cause of the event was failure of the Operations crew to adequately track equipment statusfailure of the Operations crew to adequately track equipment status in accordance with operating procedures and processes. This event is being reported as any operation or condition which was prohibited by Technical Specificationsoperation or condition which was prohibited by Technical Specifications in accordance 10 CFR 50.73(a)(2)(i)(B)), and as any event or condition that could have prevented the fulfillment of the safety function needed to shut down the reactor and maintain it in a safe shutdown condition in accordance with 10 CFR 50.73(a)(2)(v)(A)).
LER Code: 4572023001R00 Title: Unit 2 Train B Auxiliary Feedwater Pump was inoperable due to Degraded Oil in the Crank Case Facility/Unit: Braidwood Station / 2 Event Date: 9/22/2023 Reported Basis: 50.73.(a)(2)(i)(B), 50.73.(a)(2)(v)(B)
On September 1, 2023, an oil sample from the 2B auxiliary feedwater (AF) pump diesel engine was obtained. On September 21, 2023, the oil sample results were received which indicated oil viscosity and fuel percent in the Fault range, and a confirmatory sample was drawn. On September 22, 2023, resample results were received confirming the original analysis, and at 1157 Operations declared the 2B AF train INOPERABLE2B AF train INOPERABLE and entered Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.5, "Auxiliary Feedwater (AF) System," Condition A, "One AF train inoperable." Maintenance was performed to repair fuel leaks and crankcase oil was changed. The 2B AF train was returned to service on September 23, 2023, at 1340, and at 1501 LCO 3.7.5 was exited. The cause of the event was vendor performance and quality control gap in refurbishing fuel injectors. The corrective action was repair of the 2B AF diesel engine lube oil fuel leaks. A supplement to this report will be provided as more information is available.
LER Code: 4822021003R00 Title: Automatic Reactor Trip Due to Low Steam Generator Level Facility/Unit: Wolf Creek Generating Station / Unknown Unit Event Date: 5/12/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
On 5/12/2021, Wolf Creek Generating Station was performing initial reactor startup following Refueling Outage 24 (RF24). With reactor power at approximately 8%%, steam generator levels began to oscillate while in automatic control. Operators took manual control of the main feedwater regulating bypass valves but were unable to stabilize steam generator levels prior to reaching the ?Å“C??steam generator low level reactor trip setpoint. At 1125 Central Daylight Time (CDT), the reactor tripped and an auxiliary feedwater actuation occurred. All equipment responded as expected. ENS notification #55252 was made at 1441 CDT in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to reactor scram, and 10 CFR 50.72(b)(3)(iv)(A) for a specified system actuation. The cause of the event was determined to be during the 7300 modification, improper utilization of main feedwater regulating bypass valves inherent valve curves did not take into consideration system flow characteristics. The result was an inaccurate correlation of feed flow with valve position within the system. This led to a mismatch in feed flow demand and actual feed flow to the steam generators. Due to improper gain settings, the mismatch began to diverge and led to a reactor trip on low steam generator water level. Following tuning of control parameters, operators were able to commence reactor startup and reached full power on 5/15/2021.
LER Code: 4822021004R00 Title: Low Steam Generator Level due to Main Feedwater Valve Failure Caused Automatic Reactor Trip Facility/Unit: Wolf Creek Generating Station / Unknown Unit Event Date: 8/18/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
At 1036 Central Daylight Time (CDT) on 8/18/2021, Wolf Creek Generating Station (WCGS) experienced a reactor trip due to low level in the 'B' steam generator. WCGS was operating in MODE 1 at 100% power when the trip occurred. ENS notification #55416 was made at 1251 CDT in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to reactor scram, and 10 CFR 50.72(b)(3)(iv) (A) for an auxiliary feedwater system actuation. All control rods dropped, all equipment functioned as designed, and offsite power remained available. The direct cause was a fracture of the valve stem for the 'B' steam generator main feedwater regulating valve, causing the valve to fail closed and resulting in a loss of feed flow control to the ??B??steam generator. The hardware failure anaylsis performed on the valve stem determined that the fracture was due to a fatigue crack which had propagated through the stem of the valve. Tool marks within the thread root, as well as the thread root being cut deeper and narrowerTool marks within the thread root, as well as the thread root being cut deeper and narrower were identified as stress risers which allowed the crack to propagate into the material of the stem. Dye penetrant testing of the replacement stem, and on the existing valve stems of the three other main feedwater regulating valves, showed no relevant indications.
LER Code: 4822022001R00 Title: Low Steam Generator Level due to Main Feedwater Valve Failure Caused Reactor Trip Facility/Unit: Wolf Creek Generating Station / Unknown Unit Event Date: 7/18/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
At 1803 Central Daylight Time (CDT) on July 18, 2022, Wolf Creek Generating Station (WCGS) experienced an automatic reactor trip due to low level in the 'B' steam generator. WCGS was operating in MODE 1 at 100% power when the trip occurred. ENS notification #56005 was made at 2017 CDT in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to reactor scram, and 10 CFR 50.72(b)(3)(iv)(A) for an auxiliary feedwater system actuation. All control rods dropped, all equipment functioned as designed, and offsite power remained available. The direct cause of the event was a failure of the valve stem for the 'B' steam generator (SG) main feedwater regulating valve (MFRV), causing the valve to fail closed and resulting in a loss of feedwater flow control to the 'B' SG. The hardware failure analysis performed on the valve stem determined that the failure of the valve stem was due to stress corrosion cracking (SCC) initiating from the external surface within the region contained within the packing area. The root cause investigation found that during the previous valve rebuild, an inappropriate chemical (anti-seize) was applied to the stem for lubrication to assist in the rebuild. It was the presence of this chemical which caused the SSC that led to the failure. All four MFRVs are scheduled to be rebuilt during the upcoming refueling outage in October 2022 with all four valve stems then being sent off to be examined to look for evidence of SCC.
LER Code: 4832021001R00 Title: Manual Actuation of Essential Service Water System Facility/Unit: Callaway Plant. / 1 Event Date: 2/9/2021 Reported Basis: 50.73.(a)(2)(iv)(A)
On February 9, 2021, at 0838 operators commenced placing both trains of Essential Service Water (ESW) in manual alignments in accordance with annunciator response procedure OTA-RK-00014, Addendum 12A, "Service Water Pump Lockout." The manual actuation of ESW was in response to a condition where two non-safety Service Water (SW) pumps tripped and locked out due to low lubrication (lube) water pressure. The cause of the trips and subsequent lockouts of the 'A' and 'B' SW pumps was attributed to low lube water pressure caused by fouling of the lube water Y-strainers. The fouling occurred since the normal weekly backflush of the Y-strainers could not be performed while a temporary alteration in support of maintenance (TASM) was installed in order to repair a leak in the permanent lube water piping. Preventive Maintenance (PM) tasks had not been initiated prior to the event to periodically clean the Y-strainers in order to compensate for the inability to backflush. the section of the common lube water piping where the leak occurred initial corrective action taken in response to the event was cleaning of the Y-strainers in the lube water piping. In addition, PM tasks were created to clean the Y-strainers on a weekly basis for the remaining time that the TASM was installed. The permanent corrective action was to complete a modification to replace the section of the common lube water piping where the leak occurred. Both trains of ESW were returned to standby alignments on February 9, 2021 at 1140.
With the plant at full power on January 7, 2022, an undetected relay failure caused the solid-state reactor protection system (SSPS) to actuate during restoration from a Reactor Trip Breaker 'B' Trip Actuating Device Operational Test, thus resulting in an unplanned reactor trip. The reactor trip was reported per Event Notification 55698 in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 50.72(b)(3)(iv)(A). The cause of the reactor trip was that procedure OSP-SB-0001B, "Reactor Trip Breaker 'B' Trip Actuating Device Operational Test," did not include adequate instructions for verifying the absence of a "B" train General Warning trip signal prior to operating the "A" train Multiplexer Test Switch during test restoration. This was due to inadequate guidance for development of the procedure. Corrective Actions include revising Operations surveillance procedures to include verification of green and amber test lights at panel SB032B during restoration and revising Operations and l&C surveillance procedures to include verification that contacts for relays that have actuated during the test have changed back to their normal state, prior to test restoration. Additional corrective actions include revising the procedure preparation process, improving the screening and incorporation of Operating Experience, improving lesson plans, conducting training, placing operator aids adjacent to the Multiplexer Test Switch, and requiring an inspection on other master relays.
LER Code: 4982023004R01 Title: Condition Prohibited by Technical Specifications Due to Inoperable Train of Essential Chilled Water Facility/Unit: South Texas / 1 Event Date: 11/27/2023 Reported Basis: 50.73.(a)(2)(i)(B)
On 11/20/23 at 1530 Essential Chiller 12C was declared OPERABLE following maintenance. The following day at 1245, Train 'C' Essential Chilled Water System (EChWS) was declared inoperable due to excessive chilled water leakage. A subsequent Engineering evaluation determined that due to the leakage, Train 'C' EChWS should be considered inoperable from the date and time the 12C Essential Chiller was restored from maintenance and declared OPERABLE. With only two EChWS loops OPERABLE, Technical Specification (TS) 3.7.14 Action a requires restoration of the inoperable loop within 7 days or application of the Configuration Risk Management Program (CRMP), or to be in at least HOT STANDBY within the next 6 hours. The Engineering evaluation determined that Train 'C' EChWS was inoperable without any of the required TS actions taken from 1530 on 11/20/23 to 1245 on 11/28/23 when the CRMP was entered. This is a total of 189 hours and 15 minutes which exceeds the 174 hours (7 days plus 6 hours) allowed by the TS. Therefore, this event is reportable under 10 CFR 50.73(a)(2)(i)(B) as a condition which is prohibited by the plant's TS.
LER Code: 4992021001R00 Title: Condition Prohibited by Technical Specifications Due to Inoperable Containment Isolation Valve Facility/Unit: South Texas / 2 Event Date: 7/1/2021 Reported Basis: 50.73.(a)(2)(i)(B)
On July 1, 2021, at 2344 hours, the STP Nuclear Operating Company discovered that Steam Generator 2D outside reactor containment isolation valve FV-4150 was indicating valve closure but only reduced the flow rate by about 8.33%%. On July 2, 2021, at 0350 hours, the valve was declared inoperable due to inadequate isolation. Site engineering determined on July 12, 2021, that FV-4150 was inoperable since March 29, 2021, due to preventative maintenance performed on that date. On July 3, 2021, FV-4150 was determined to be operable following a valve stem adjustment and a satisfactory performance of the surveillance test. The causes of the event were associated with inadequate written guidance. This event is reportable because the Technical Specification 3.6.3 required action statement for an inoperable containment isolation valve was not met. Specifically, the containment isolation valve stem adjustment was inoperable from March 29, 2021, to July 3, 2021, for a total of 96 days. The allowed outage time is 24 hours or apply the requirements of the configuration risk management program. Planned corrective actions include revising the work instructionsrevising the post-maintenance testing guidancework instructionswork instructions and the post-maintenance testing guidancepost-maintenance testing guidance.
LER Code: 4992022001R00 Title: ' Automatic Actuati . _of_Fmera- . ?? - Generator due to I or:ri it of Switchyard Electrical Bus_ Facility/Unit: South Texas / 2 Event Date: 1/6/2022 Reported Basis: 50.73.(a)(2)(iv)(A)
On January 06, 2022, with both units operating at 100% power, the South Texas Project 345 kV south switchyard electrical bus unexpectedly de-energized. The de-energization of the south bus resulted in a loss of power to the Unit 2 standby transformer which was supplying power to the Engineered Safety Features (ESF) 4160V busses for the Unit 2 B Train. The associated Emergency Diesel Generator 22 automatically started in response to the undervoltage condition, as designed. The cause of this event was an electrical flashover on a Transmission Distribution Service Provider owned insulator on the south bus. This event is reportable as an event which resulted in the automatic actuation of the Unit 2 emergency AC electrical power systems. Corrective actions included replacing the insulator, performing inspections for abnormal electrical Corona on the insulators (no adverse values were found), and cleaning of the other insulators on the North and South bus. There were no previous indications of issues with the insulator. The event did not result in any offsite release of radioactivity or increase of offsite dose rates and there were no personnel injuries or damage to any safety-related equipment associated with this event. Additionally, all ESF equipment operated as designed. Therefore, there was no adverse effect on the health and safety of the public.
LER Code: 4992023002R00 Title: Automatic Actuation of Emergency Diesel Generator due to Lockout of Switchyard Electrical Bus Facility/Unit: South Texas / 2 Event Date: 12/2/2023 Reported Basis: 50.73.(a)(2)(iv)(A)
On December 2, 2023, with both units operating at 100% power, the South Texas Project (STP) 345 kV south switchyard electrical bus unexpectedly de-energized. The de-energization of the south bus resulted in a loss of power to the Unit 2 standby transformer which was supplying power to the Engineered Safety Features (ESF) 4160V busses for the Unit 2 B Train. The associated Emergency Diesel Generator 22 automatically started in response to the under- voltage condition, as designed. The cause of this event was the failure of a stand-off insulator occurring on Phase ?Å“A??in 13.8kV Terminal Cabinet. A root cause could not be determined after Engineering evaluations. Corrective actions included replacing the insulator and performing visual inspections of both the transformer and bus for signs of arcing and/or water intrusion. There were no previous indications of issues with the insulator. This event is reportable as an event which resulted in the automatic actuation of the Unit 2 emergency AC electrical power systemsautomatic actuation of the Unit 2 emergency AC electrical power systems. Additionally, all ESF equipment operated as designed.
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Attributes Definition
Extracted items may include an attributes dictionary providing secondary classification or context.
Cause may include category (e.g., conflicting_procedure).
CorrectiveAction may include action_type (e.g., revision).