Semantic search
| Entered date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 56914 | 5 January 2024 15:56:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via phone and email: At 1552 (EST) on 01/05/2024, Perry Nuclear Power Plant reported elevated levels of tritium in the underdrain system to the state of Ohio as a non-voluntary reporting of tritium. An investigation is currently ongoing to identify the cause of the elevated tritium levels. The tritium levels in this location do 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. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56673 | 10 August 2023 04:03:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via email: At 0039 (EDT) on 8/10/23, with Unit 1 in Mode 1 at 100 percent power, the reactor automatically tripped during a reactor protection system (RPS) bus shift. All systems responding normally post-trip. There was no equipment inoperable at the time of the trip. Operations responded and stabilized the plant. Reactor water level being maintained via feedwater. Decay heat is being removed by cycling safety relief valves. An actuation of high-pressure core spray, division 3 diesel generator, and reactor core isolation cooling occurred during the scram and main steam line isolation closure. The reason for the auto-start was reaching Level 2 (130 inches in the reactor pressure vessel) during the transient. The systems automatically started as designed and injected to the reactor vessel when the Level 2 signal was received. The RPS actuation is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). The emergency core cooling system (ECCS) injection is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). The ECCS actuation is being reported as a eight-hour, non-emergency notification per 10 CFR 50.72(b)(2)(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 56588 | 23 June 2023 17:04:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via email: At 1521 EDT on 6/23/2023, Perry Nuclear Power Plant reported elevated levels of tritium in the underdrain system to the State of Ohio as a non-voluntary reporting of tritium. An investigation is currently ongoing to identify the cause of the elevated tritium levels. The tritium levels did not exceed any NRC regulations or reporting criteria. Tritium has not been detected in any other locations and is not expected to impact groundwater or exceed any limits in the Off Site Dose Calculation Manual (ODCM). This notification is being made solely as a four-hour, non-emergency notification for a Notification of Other Government Agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The following agencies were notified by licensee: Lake County Emergency Management Agency (EMA) Ashtabula County EMA Geauga County Department of Emergency Services Ohio EMA Radiological Branch |
| ENS 56569 | 11 June 2023 09:02:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 0130 EDT on June 11, 2023, it was discovered that the Beaver Valley Power Station, Unit No. 2 auxiliary building door A-35-5A, credited for tornado missile protection of the primary component cooling water system, was open and unlatched. Upon discovery, the door was shut and latched. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56527 | 19 May 2023 12:33:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 0852 (EDT) on May 19, 2023, with Unit 2 in Mode 3 at zero percent power, an actuation of the auxiliary feedwater system (AFW) occurred. The reason for the AFW auto-start was a failed start attempt of the 'B' main feedwater pump. The 'A' and 'B' motor driven auxiliary feedwater (MDAFW) pumps automatically started as designed when the 'Loss of Both Main Feedwater Pumps' 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 AFW system. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56519 | 16 May 2023 08:09:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone: A licensed operator failed a test specified by the FFD testing program. The individual's site access has been terminated. The NRC Resident Inspector has been notified. |
| ENS 56485 | 22 April 2023 21:35:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 1942 EDT on April 22, 2023, during the Beaver Valley Power Station, Unit No. 2 refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per ASME Code Section XI. The reactor vessel vent line penetration will require repair prior to returning the vessel head to service. The indication was not through-wall as there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. 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 56455 | 6 April 2023 16:46:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via phone and email: On March 4, 2023, it was determined that the main steam line (MSL) local leak rate test results for MSL 'B' were in exceedance of technical specification (TS) surveillance requirement (SR) 3.6.1.3.10 limits. Additionally, the leakage at the outboard main steam isolation valve (MSIV) 'B', was indeterminate due to a gross packing gland leak. An engineering calculation dated April 6, 2023, showed that this leakage, in conjunction with a design basis loss of coolant accident, would result in the radiological dose exceeding Updated Safety Analysis Report limits to the exclusion area boundary, the low population zone, and the control room. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the power plant being in an unanalyzed condition that degrades plant safety. Both inboard and outboard 'B' MSIVs have been reworked and are within the TS SR limits. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56435 | 28 March 2023 08:59:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | A non-licensed supervisor had a confirmed positive for a controlled substance 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 56417 | 17 March 2023 10:15:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via email: On March 16, 2023, at 2226 EDT, a site supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's authorization for access to the plant has been terminated. The Resident Inspector has been notified. |
| ENS 56350 | 12 February 2023 14:41:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone call and email: At 0800 on February 12, 2023, it was discovered that both trains of control room emergency ventilation system were simultaneously inoperable due to a safety injection relief valve discharging to a Unit 1 sump. This leakage in conjunction with design basis loss of coolant accident may result in radiological dose exceeding limits to the exclusion area boundary and to the control room, which is common to both Unit 1 and Unit 2. Therefore, this condition is being reported as an eight-hour, nonemergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(D) as an 'Unanalyzed Condition and a Condition that Could Have Prevented Fulfillment of a Safety Function.' There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
Retraction of EN56350, Control Room Emergency Ventilation System Inoperable: Based on subsequent evaluation, it was determined that the control room emergency ventilation system remained operable due to the maximum measured leak rate being within the bounds of the analysis. The maximum measured leak rate of 32,594 cc/hr from the safety injection system did not challenge the calculated maximum engineered safety features leak rate of 45,600 cc/hr and remained within the current dose analysis limits. As such, this was not an unanalyzed condition and did not prevent the fulfillment of a safety function to mitigate the consequences of an accident. The NRC Resident Inspector has been notified. Notified R1DO (Bickett). |
| ENS 56298 | 5 January 2023 15:39:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via phone and email: At 1242 (EST) on 05 January 2023, with the Unit in Mode 1 at 99 percent power, the reactor automatically tripped on low Reactor Pressure Vessel level while restoring power to Digital Feedwater Control Stations when there was a perturbation to the level controls. The reason for perturbation is unknown at this time. The trip was not complex, with all systems responding normally post trip. Operations responded and stabilized the plant. High pressure core spray was manually initiated in accordance with site procedures. Reactor water level is being maintained via the Feedwater System. Decay heat is being removed by the Main Condenser. 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)(A) and 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55962 | 24 June 2022 16:28:00 | Perry | NRC Region 3 | GE-6 | The following information was provided by the licensee via telephone: At 1257 EDT on June 24, 2022, it was discovered the Low Pressure Core Spray System (LPCS) was INOPERABLE. At Perry, the Low Pressure Core Spray System is considered a single train system in Modes 1, 2, and 3; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). Inoperability of the Low Pressure Core Spray system was caused by a loss of power to the LPCS Minimum Flow Valve during surveillance activities. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55953 | 21 June 2022 16:52:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via fax or email: At 1547 EDT on June 21, 2022, it was determined that Beaver Valley Power Station Unit No. 1 experienced a reportable chemical leak. Approximately 261 gallons of a Sodium Hypochlorite/Sodium Bromine mixture reached the ground and approximately 130.5 gallons (of the 261 gallons) progressed to the Ohio River (via storm drain). The source of the leakage has been isolated and absorbent material has been placed to contain the leakage. Following confirmation of this leakage, notifications were made to the following offsite agencies starting at 1615 EDT: National Response Center (Incident Report # 1339391) Pennsylvania Department Of Environmental Protection Beaver County Emergency Management This condition is being reported as a four-hour, non-emergency notification per 10CFR50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55943 | 15 June 2022 09:47:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 0724 EDT on 6/15/2022, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to lowering Steam Generator levels due to a secondary plant perturbation in the Heater Drain System. All control rods fully inserted into the core and the Auxiliary Feedwater System automatically started as designed in response to the full power reactor trip. The trip was not complex, with all systems responding normally post-trip. There was no equipment inoperable prior to the event that contributed to the reactor trip or adversely impacted plant response. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the condenser steam dump valves. Unit 2 is not affected and remains at 100 percent power and stable. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, this event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Auxiliary Feedwater System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55602 | 23 November 2021 09:10:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | 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 specific system actuation. At 0907 (EDT) on September 30, 2021, with Unit 1 in Mode 1, at 100 percent power, an actuation of the 1-1 emergency diesel generator (EDG) occurred during loss of voltage relay functional testing. The 1-1 EDG auto-start was due to human error during performance of the test procedure when the bus 1AE undervoltage signal was improperly defeated and a simulated undervoltage signal was applied. No actual undervoltage condition was present during this event. The 1-1 EDG automatically started as designed when the bus undervoltage signal was received. This was a complete actuation of an EDG to start and come to rated speed, and all affected systems functioned as expected in response to the actuation. Following the actuation, the relays were restored and the 1-1 EDG was shut down in accordance with plant procedures. This event is considered an invalid system actuation reportable under 10 CFR 50.73(a)(2)(iv)(A). The actuation was not initiated in response to actual plant conditions or parameters and was not a manual initiation. Therefore, in accordance with 10 CFR 50.73(a)(1), this telephone notification is provided within 60 days after discovery of the event instead of submitting a written Licensee Event Report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55586 | 17 November 2021 16:24:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | At 1313 hours on November 17, 2021, with Unit 2 in Mode 1 at 100 percent power, the reactor was manually tripped due to a loss of the 21B Main Feedwater Pump (due to low suction pressure). The Auxiliary Feedwater System automatically started as designed in response to the full power reactor trip. Additionally, the Main Steam Isolation Valves were manually closed to prevent excessive reactor coolant system cooldown. The trip was not complex, with all systems responding normally post-trip. There was no equipment inoperable prior to the event that contributed to the reactor trip or adversely impacted plant response. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the atmosphere using the Atmospheric Dump Valves. Beaver Valley Power Station Unit 1 is unaffected and remains at 100 percent power in Mode 1. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the automatic actuation of the Auxiliary Feedwater System is being reported as an eight-hour, non-emergency Specific System Actuation per 10 CFR 50.72(b)(3)(vi)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident has been notified. |
| ENS 55572 | 12 November 2021 13:52:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | 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 an oscillating Main Feedwater Pump Recirculation Valve. Additionally, the Main Steam Isolation Valves were manually closed to prevent excessive reactor coolant system cooldown. Decay heat is being removed by discharging steam to the atmosphere using the Atmospheric Dump Valves. The trip was not complex, with all systems responding normally post-trip. There was no equipment inoperable prior to the event that contributed to the reactor trip or adversely impacted plant response. Operations responded and stabilized the plant. Beaver Valley Power Station Unit 1 is unaffected and remains at 100 percent power in Mode 1. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification, per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 55540 | 22 October 2021 02:15:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | At 0104 EDT on October 22, 2021, during the Beaver Valley Power Station, Unit 2 refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that two penetrations could not be dispositioned as acceptable per ASME Code Section XI. Penetrations 28 and 40 will require repair prior to returning the vessel head to service. The indications were not through wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-6 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. 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 55507 | 5 October 2021 10:07:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | At 0632 EDT on October 5, 2021, with Unit 2 in Mode 1 at approximately 90 percent power for an end of cycle coastdown, the reactor automatically tripped due to an unexpected unblocking of the low power trip logic. The trip was not complex, with all systems responding normally post-trip. There was no equipment inoperable prior to the event that contributed to the reactor trip or adversely impacted plant response. The Auxiliary Feedwater System automatically started as designed in response to the reactor trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the condenser steam dump valves. Beaver Valley Power Station Unit 1 is unaffected and remains at 100 percent power in Mode 1. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Additionally, the automatic actuation of the Auxiliary Feedwater System is being reported as an eight hour, non-emergency Specific System Actuation 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 55413 | 18 August 2021 01:07:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | At 0024 EDT on 8/18/21, an unusual event was declared (EAL HU 4.1) due to receipt of multiple fire alarms and halon discharge in the cable tunnel. At 0036, the fire brigade verified no signs of fire. Unit 1 remained at 100 percent power and stable. The area is currently being ventilated. Unit 2 was not affected by this event. The licensee has notified State and local authorities and the NRC Resident Inspector. R1 Public Affairs (Screnci) was notified. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
At 0401 EDT, Beaver Valley terminated their notification of unusual event. The basis for termination was that there was no indication of fire. The licensee will be notifying the NRC Resident Inspector and has notified State and local authorities. Notified R1DO (Jackson), IRD MOC (Gott), NRR EO (Felts), DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). |
| ENS 55263 | 18 May 2021 09:04:00 | Perry | NRC Region 3 | GE-6 | 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 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 55172 | 6 April 2021 04:32:00 | Perry | NRC Region 3 | GE-6 | 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.
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 54848 | 21 August 2020 12:42:00 | Perry | NRC Region 3 | At 0953 EDT on 8/21/20, it was discovered that both trains of the standby liquid control system were simultaneously inoperable; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). All control rods remained operable during this time period. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Both trains of the standby liquid control system were declared inoperable due to an inadvertent addition of water to the storage tank which caused the boron concentration in the tank to go low out of specification. | |
| ENS 54749 | 18 June 2020 20:37:00 | Beaver Valley | NRC Region 1 | On June 18, 2020, Beaver Valley Power Station, Unit 1 determined that leakage from an outside out-of-service liquid waste pipe that is within a radiologically controlled area contains several isotopes. Analysis indicates greater than 2,000 picoCuries per liter of tritium, and isotopes of Mn-54, Co-58, Co-60, and Cs-137 are above the lower limit of detection. At 1725 EDT, in accordance with site procedures and NEI 07-07, 'Industry Ground Water Protection Initiative,' notification to the Commonwealth of Pennsylvania was planned. At 1930 EDT notification to the Commonwealth of Pennsylvania was completed. The leak is currently contained. The leakage did not exceed any NRC regulations or reporting criteria. This notification is being made solely as a four-hour, non-emergency notification for a planned notification of other government agency. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 54680 | 24 April 2020 07:00:00 | Beaver Valley | NRC Region 1 | At 0130 (EDT) on April 24, 2020, during the Beaver Valley Power Station, Unit 2 refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per ASME Code Section XI. Penetration 37 will require repair prior to returning the vessel head to service. The indication was not through wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-4 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. 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 54507 | 1 February 2020 16:29:00 | Perry | NRC Region 3 | EN Revision Imported Date : 3/17/2020 LOW PRESSURE CORE SPRAY INOPERABLE At 1150 EST on 2/1/2020, it was discovered that the Low Pressure Core Spray System was inoperable due to a divisional battery voltage being out-of-specification. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). The Low Pressure Core Spray is a single train safety system in Modes 1, 2, and 3. Low Pressure Core Spray was restored to operable, restoring function at 1230 on 2/1/2020. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
On February 01, 2020, event notification (EN-54507) was made to the NRC for Low Pressure Core Spray (LPCS) inoperability. This notification was made due to high DC bus voltages resulting in LPCS being declared inoperable which resulted in a loss of safety function. An Engineering Evaluation Request (EER) was performed to determine an upper limit for DC bus voltage. This EER determined that LPCS could perform its required functions with a voltage increase of up to 150V DC if the duration was not greater than two hours. The elevated voltage was experienced for approximately 59 minutes. The maximum voltage experienced was 147.07V DC. Therefore, LPCS remained operable and no loss of safety function existed. The NRC Resident Inspector has been briefed on the evaluation results and informed of this retraction. Notified R3DO (Hanna). | |
| ENS 54369 | 4 November 2019 09:06:00 | Beaver Valley | NRC Region 1 | At 0535 (EST) on November 4, 2019, with Unit 1 in mode 1 at 15 (percent) power, the reactor was manually tripped due to the lifting of an 'A' Main Steamline Safety Valve following a Condenser Steam Dump transient. The trip response was not complex, with all systems responding normally post-trip, and all control rods fully inserted into the core. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the Main Condenser by using the Condenser Steam Dump Valves. Offsite power is available and is currently supplying Normal and Emergency busses. The plant is currently stable in mode 3. Unit 2 is unaffected by this event and remains at 100 (percent) power in mode 1. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 54366 | 2 November 2019 23:03:00 | Beaver Valley | NRC Region 1 | At 1515 on November 2, 2019, the Refueling Water Storage Tank (RWST) was declared inoperable due to a Low Head Safety Injection relief valve discharging to the Safeguards Sump during routine surveillance testing. The leakage from the Low Head Safety Injection system in conjunction with a postulated Design Basis Accident (DBA) Loss of Coolant Accident (LOCA) with transfer to Safety Injection Recirculation may result in dose exceeding the Dose Analysis of the Exclusion Area Boundary (EAB) and the Control Room, which is common to both Unit 1 and Unit 2. This condition may not be bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. The Low Head Safety Injection relief valve has been isolated to prevent further leakage, and makeup to the RWST completed. At 1602 on November 2, 2019 the RWST was declared Operable. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B) and 10 CFR 50.72(b)(3)(v)(B), (C), (D) as an Unanalyzed Condition and a condition that could have prevented the Fulfillment of a Safety Function." The licensee notified the NRC resident inspector. | |
| ENS 54337 | 18 October 2019 02:28:00 | Beaver Valley | NRC Region 1 | At 1951 (EDT) on October 17, 2019, fretting indications on the reactor coolant system pressure boundary piping (pressurizer spray line) were identified. This condition does not appear to meet original construction code, ANSI B31.1, 1967 Edition thru summer 1971 Addenda. The condition will be resolved prior to plant startup. 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. This condition will be corrected prior to the plant entering Mode 4.
An engineering evaluation has determined that the subject fretting is not considered a flaw, but instead is considered wear. Appendix F of Section III of the ASME Boiler and Pressure Vessel Code was applied and it was determined that the pressurizer spray line piping maintained its required design safety functions in the as-found condition. The wear has been repaired during the current refueling outage in accordance with the original construction code (ANSI B3l.l, 1967 Edition through summer 1971 Addenda) as well as Owner's Requirements. The NRC Resident Inspector has been notified. Notified R1DO (Young). | |
| ENS 54203 | 6 August 2019 17:29:00 | Perry | NRC Region 3 | At 1335 EDT on 08/06/2019, it was discovered that reactor protection system (RPS) instrumentation functions for turbine stop valve closure and turbine control valve fast closure, end of cycle recirculation pump trip (EOC-RPT) instrumentation, and control rod block instrumentation were simultaneously inoperable due to a loss of feedwater heating; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). RPS instrumentation, EOC-RPT instrumentation, and control rod block instrumentation functions were restored at 1422 EDT on 08/06/2019. There was no impact on the health and safety of the public or plant personnel. The NRC Senior Resident Inspector has been notified. | |
| ENS 54185 | 27 July 2019 22:31:00 | Perry | NRC Region 3 | At 1929 EDT on 7/27/2019, with the Unit in Mode 1 at 98 percent power, the reactor automatically scrammed due to a Main Turbine Trip. The trip was not complex, with all systems responding normally post-trip. Main Steam Isolation Valves (MSIVs) were manually closed to prevent exceeding Reactor Pressure Vessel Cooldown Rate. Rector Core Isolation Cooling (RCIC) was manually initiated to stabilize Reactor Vessel Water Level and Pressure following MSIV closure. The Main Condenser and Feedwater are available. Operations responded and stabilized the plant. Reactor water level is being maintained via RCIC. Decay heat is being removed by discharging steam to the Main Condenser and RCIC. The cause of the Main Turbine Trip is currently under investigation. The site is in a normal electrical lineup. The licensee notified the NRC Resident Inspector. | |
| ENS 54183 | 26 July 2019 13:27:00 | Beaver Valley | NRC Region 1 | A 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 54085 | 24 May 2019 13:09:00 | Perry | NRC Region 3 | At 0730 (EDT) on May 24, 2019, it was discovered that the Low-Pressure Core Spray System was inoperable. At Perry, the Low-Pressure Core Spray system is considered a single train system in Modes 1, 2, and 3; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). Inoperability of the Low-Pressure Core Spray system was caused by Emergency Service Water Pump A inoperability. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | |
| ENS 53924 | 11 March 2019 10:00:00 | Beaver Valley | NRC Region 1 | On 03/11/2019 planned maintenance activities will be performed on the Beaver Valley Power Station (BVPS) Unit 2 Digital Radiation Monitoring System's (DRMS) Communications System. The work includes upgrades to the DRMS hardware, software and computer peripherals. Components to be upgraded under this planned maintenance include: Redundant Servers, Operator Console, Health Physics Office Console/Workstation, Printers and Portable Mass Storage/Backup and Computer peripherals necessary to interface with the computer system. This planned upgrade on the Unit 2 DRMS Communications System will result in the loss of Unit 2 radiological monitoring capability in the Control Room and on the Plant Computer System (PCS). Neither the Emergency Response Facilities nor the Emergency Response Data System (ERDS) will receive radiological data. No actual radiation monitors are affected. The scheduled work duration is approximately three weeks. When the DRMS is out of service for the upgrade, compensatory actions will be in place. Radiation monitors will be continuously monitored for any increases in radiation levels. The Unit 2 Shift Manager will be notified of any increase in radiation monitor readings, including exceeding Emergency Action Levels (EAL). This is an eight-hour, non-emergency notification for a Loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) because the upgrade to the DRMS Communications System will result in the loss of the Unit 2 radiological monitoring capability in the Control Room and to the Unit 2 Plant Computer System that affects the functionality of an Emergency Response Facility. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee is currently in a Technical Specification Action Statement, Limiting Condition for Operation for spent fuel pool radiation monitor and leakage rate monitor. | |
| ENS 53896 | 25 February 2019 03:44:00 | Perry | NRC Region 3 | At 0024 EST on 2/25/19, with Unit 1 in Mode 1 at 74 percent power, the reactor automatically tripped due to a generator trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Reactor water level is being maintained via the feed system. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The generator trip is under investigation, but is believed to be due to grid perturbations. | |
| ENS 53711 | 1 November 2018 15:38:00 | Beaver Valley | NRC Region 1 | On 11/01/2018, during the Beaver Valley Power Station Unit No. 2 (BVPS-2) refueling outage, while performing examinations of the 66 reactor vessel head penetrations, it was determined that one penetration could not be dispositioned as acceptable per ASME Code Section XI. Penetration 27 will require repair prior to returning the vessel head to service. The indication was not through wall and there was no evidence of leakage based on inspections performed on the top of the reactor vessel head. The examinations were being performed to meet the requirements of 10 CFR 50.55a(g)(6)(ii)(D) and ASME Code Case N-729-4 to find potential flaws/indications before they grow to a size that could potentially jeopardize the structural integrity of the reactor vessel head pressure boundary. The other 65 penetrations will be examined during the 2R20 (current) refueling outage. The plant is currently shutdown and in Undefined Mode. The reactor vessel head is not currently installed. Repairs are currently being planned and will be completed prior to startup. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A) since the as-found indications did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-4 to remain in-service without repair. The NRC Resident Inspector has been notified. | |
| ENS 53675 | 19 October 2018 12:04:00 | Perry | NRC Region 3 | During extent of condition review of a previously identified fire induced hot-short (Ref. EN#53644) an unfused circuit associated with the 0M23C0002A, Miscellaneous Switchgear Recirculation Fan was discovered. This condition is not bounded by existing design and licensing documents. This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit. Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed, challenging the ability to achieve and maintain safe shutdown. The postulated event would affect multiple fire zones in the control complex. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector.
Further extent of condition reviews have discovered another unfused circuit. The circuitry is related to 0M24C001A, Battery Room Exhaust Fan. This condition is not bounded by existing design and licensing documents. This results in an unanalyzed condition due to the possibility for a postulated fire induced hot-short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in the Appendix R Evaluation, Safe Shutdown Capabilities Report, due to an unfused circuit. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Senior Resident Inspector. Notified R3DO (Hills). | |
| ENS 53644 | 4 October 2018 12:29:00 | Perry | NRC Region 3 | Degraded or unanalyzed condition due to the possibility for a postulated fire induced hot short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in calculation SSC-001 due to an unfused circuit associated with the 1M43C0001A, Diesel Generator Building Ventilation Fan. This condition is not bounded by existing design and licensing documents. Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit without protection that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed challenging the ability to achieve and maintain safe shutdown. The postulated event would affect the following fire zones: 1CC-3c (Unit 1, Division 1 4160V and 480V Switchgear Room, 620 feet 6 inch elevation), 1CC-3e (Unit 1 West Corridor North of Elevator, 620 feet 6 inch elevation), DG-1d (Hallway Diesel Generator Building 620 feet 6 inch elevation), and 1DG-1c (Unit 1, Division 1 Diesel Generator Building 620 feet 6 inch elevation). This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Senior Resident Inspector. | |
| ENS 53548 | 12 August 2018 02:58:00 | Beaver Valley | NRC Region 1 | EN Revision Text: TECHNICAL SPECIFICATION REQUIRED SHUTDOWN - LOSS OF 480 VOLTAGE EMERGENCY BUS On 8-12-18 at 0158 EDT, Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. This resulted in a Loss of Safety Function due to the 2-2 Emergency Diesel Generator (EDG) being Inoperable coincident with the Residual Heat Release Valve (2SVS-HCV104). A Technical Specification shutdown is required per LCO 3.0.3. The Licensee also stated they were in an unanalyzed condition due to the EDG and Residual Heat Release Valve being inoperable at the same time. The Licensee is shutting down to Mode 5 (Cold Shutdown). The Licensee is notifying the Resident Inspector. The Licensee will be making a Press Release about the unplanned shutdown.
On 8-12-18 at 0158 (EDT) Beaver Valley Unit 2 experienced a loss of 480 Volt 2P Emergency Bus. Per operational guidance, this was determined to be a Loss of Safety Function due to the Unit 2 Emergency Diesel Generator (EDG) being INOPERABLE coincident with the Residual Heat Release Valve (2SVS-HCV104) 10 CFR 50.72(b)(3)(v)(B) and (D). This was also reported as an Unanalyzed Condition 10 CFR 50.72(b)(3)(ii)(b). No Press Release was performed for this event. The NRC Resident Inspector was notified. At 0410 (EDT) a Technical Specification Shutdown was commenced 10 CFR 50.72(b)(2)(i). At 2011 (EDT) the 480 Volt 2P Emergency Bus was restored and energized. Further evaluation of the event has determined that this event was not an Unanalyzed Condition and did not result in a Loss of Safety Function. The classifications of Unanalyzed Condition and Loss of Safety Function are being retracted. The accuracy of the existing guidance relative to Safety Function has been entered in the Corrective Action Program and interim actions have been taken to provide accurate guidance. Notified R1DO (Young) via email. | |
| ENS 53481 | 1 July 2018 08:24:00 | Perry | NRC Region 3 | On July 1st, 2018 at 0100 (EDT), a portion of the Division 1 Emergency Core Cooling System (ECCS) Loss Of Coolant Accident (LOCA) initiation logic was declared inoperable due to the discovery of a blown fuse. The fuse was replaced at 0215 on July 1st, 2018 and the Division 1 ECCS LOCA initiation logic was declared operable at 0230 on July 1st, 2018. The blown fuse caused the loss of a portion of the Division 1 ECCS LOCA initiation logic which would have prevented the initiation of the Emergency Closed Cooling (ECC) A system. ECC A and supported systems were declared inoperable. Low Pressure Core Spray (LPCS) was one of the supported systems that were declared inoperable. LPCS is considered a single train safety system. Inoperability of LPCS is considered an 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 mitigate the consequences of an accident. The blown fuse also caused the loss of a portion of the Division 1 ECCS LOCA initiation logic which would have prevented the automatic isolation of Nuclear Closed Cooling and Instrument Air to the Containment. The loss of Containment isolation capability is considered an 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 mitigate the consequences of an accident. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) The NRC Senior Resident Inspector has been notified. | |
| ENS 53419 | 22 May 2018 18:38:00 | Beaver Valley | NRC Region 1 | EN Revision Text: GAS VOIDS DISCOVERED IN BOTH TRAINS OF LOW HEAD SAFETY INJECTION On 5/22/2018, while operating at approximately 100 percent power, Ultrasonic Testing of the Beaver Valley Power Station (BVPS) Unit 1 Low Head Safety Injection (LHSI) pump suction piping identified gas voids in excess of the acceptable limit for void volume. Both trains of LHSI were declared inoperable. Technical Specification (TS) 3.5.2 for both trains of the LHSI system was entered along with TS 3.0.3 which requires the initiation of a plant shutdown. Time of TS entry was 12:56 (EDT). Plant shutdown was commenced at 15:56 (EDT) in accordance with plant procedures. At 15:59 (EDT) Train 'A' LHSI was restored to operable status, TS 3.0.3 Action was exited and the power reduction was stopped at approximately 99 percent. At 17:43 (EDT) Train 'B' LHSI was restored to operable status, TS 3.5.2 Actions were exited. This is reportable per 10 CFR 50.72(b)(3)(ii) Unanalyzed Condition, 10 CFR 50.72(b)(3)(v) Event or Condition that Could Have Prevented the Fulfillment of a Safety Function and 10 CFR 50. 72(b )(2)(i) TS Required Shutdown. The NRC Resident Inspector has been notified.
Further engineering evaluation has determined that the gas voids that existed at the time of discovery would not have rendered the LHSI (Low Head Safety Injection) system inoperable if it were required to actuate. The engineering evaluation concluded that filling of the containment sump during a Design Basis Accident would result in a void volume reduction such that the void in the LHSI suction piping would not be large enough to significantly impact the operability of the system. Therefore, the system remained operable but degraded. No TSs (Technical Specifications) were required to be entered and no shutdown was required. As such, all three reporting criteria do not apply and are being retracted. The NRC Resident Inspector has been notified. Notified R1DO (Burritt). | |
| ENS 53228 | 23 February 2018 20:29:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | At 1750 (EST) on 2/23/2018 credible information was obtained that a non-licensed supervisory person had intentionally misused a controlled substance. Unescorted access has been withdrawn. Per 10 CFR 26.719(b)(2)(i) this is a 24 hour reportable event. The NRC Resident Inspector has been notified. |
| ENS 53137 | 23 December 2017 05:00:00 | Perry | NRC Region 3 | GE-6 | High Pressure Core Spray System was declared inoperable due to the discovery of a through-wall leak on the Minimum Flow line. Leak rate is 60 drops per minute from ASME Class 2 Piping. The leak has been isolated and the High Pressure (Core Spray) System has been placed in Secured Status. High Pressure Core Spray is considered a single train safety system. Inoperability of (the) High Pressure Core Spray System is considered an 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 mitigate the consequences of an accident. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Senior Resident Inspector was notified. Technical Specifications Limiting Condition for Operation 3.5.1 Condition B was entered, requiring restoration of the High Pressure Core Spray System in 14 days. The licensee plans to notify State and Local Governments (Lake, Geauga, and Ashtabula Counties). |
| ENS 53058 | 7 November 2017 14:19:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | It was determined that a licensed employee was not placed in an additional Fitness-For-Duty (FFD) testing program as directed by the MRO (Medical Review Officer). Upon discovery, the individual was subsequently tested . . . and determined Fit-For-Duty per 10 CFR 26 requirements. This event is being reported pursuant to 10 CFR 26.719(b)(4). The NRC Resident Inspector has been notified. |
| ENS 53056 | 7 November 2017 08:29:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | On November 7, at 0504 (EST), BVPS (Beaver Valley Power Station) Unit 1 experienced an automatic reactor trip due to Main Unit Generator over current. The Auxiliary Feedwater system activated and remains in service. Offsite power supply is available. Normal and Emergency busses are being supplied by Offsite power. One Source Range channel failed to energize due to its corresponding Intermediate Range instrument being under compensated. It was manually energized and is not indicating as expected. The second Source Range instrument energized but is reading erratically. Both Source Range instruments have been declared inoperable and the appropriate Technical Specification has been complied with by making the Control Rods not capable of withdrawal and isolating all dilution flow paths. Plant trip response was as expected without complications, and all control rods fully inserted in the core. The plant is currently stable in Mode 3. This event is being reported as an actuation of the Reactor Protection system 10 CFR 50.72(b)(2)(iv)(B) and a Specified System Actuation (Auxiliary Feedwater System) 10 CFR 50.72(b)(3)(iv)(A). BVPS Unit 2 is unaffected by this event and remains at 100% power in Mode 1. The NRC Resident Inspector has been notified. |
| ENS 53000 | 4 October 2017 05:53:00 | Perry | NRC Region 3 | GE-6 | On October 4, 2017, at 0250 hours (EDT), the Perry Nuclear Power Plant commenced a Technical Specification (TS) shutdown by lowering reactor power from 100 percent rated thermal power to 98 percent to comply with TS LCO 3.0.3. Reactor power was further reduced to 82 percent rated thermal power at 0430 hours (EDT). The plant had entered TS 3.0.3 at 0155 hours (EDT) upon loss of MCC (Motor Control Center), Switchgear, and Miscellaneous Electrical Equipment Areas HVAC System train A while train B was removed from service for maintenance. MCC switchgear ventilation train A was declared inoperable based on excessive belt noise and a dropped belt on MCC switchgear supply fan A. This also constitutes a loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(2)(i) and 10 CFR 50.72(b)(3)(v)(D). The NRC Resident Inspector was notified.
Due to the loss of both trains of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC, actions were taken in LCO 3.8.7 for AC and DC Distribution Systems, LCO 3.8.4 for DC Sources, LCO 3.8.1 for AC Sources, and the associated support systems, the High Pressure Core Spray system was also declared inoperable, which is a single train safety system and therefore, an additional loss of safety function. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(B), 10 CFR 50.72(b)(3)(v)(C) and 10 CFR 50.72 (b)(3)(v)(D). At 0620 hours (EDT) the A train of MCC, Switchgear, and Miscellaneous Electrical Equipment Areas HVAC and High Pressure Core Spray was declared operable and LCO 3.0.3 was exited. The plant was restored to 100% (percent) power at 0804 (EDT). The NRC Resident Inspector was notified. Notified R3DO(Hills). |
| ENS 52903 | 15 August 2017 04:58:00 | Perry | NRC Region 3 | GE-6 | On August 14th, 2017 at 2257 (EDT), while shutting down the Annulus Exhaust Gas Treatment System (AEGTS) Train B, secondary containment pressure momentarily lowered. This resulted in the Technical Specification (TS) for Secondary Containment to not be met for 15 seconds. The minimum Secondary Containment vacuum observed during that time was 0.52 inch of vacuum water gauge. Secondary Containment pressure was returned to within the TS operability limit of 0.66 inch of vacuum water gauge (TS SR 3.6.4.1.1) by the AEGTS Train A that remained in operation. There were no radiological releases associated with this event. Declaring Secondary Containment inoperable is reportable under (10 CFR) 50.72(b)(3)(v)(C) & (D) for an event or condition that could have prevented the fulfillment of a safety function of a system that is needed to:(C) Control the release of radioactive material; or (D) Mitigate the consequences of an accident. The licensee has notified the NRC Resident Inspector. |
| ENS 52891 | 8 August 2017 20:22:00 | Perry | NRC Region 3 | GE-6 | On August 8, 2017, at 1554 hours (EDT), during restoration from testing of the High Pressure Core Spray (HPCS) Suppression Pool Level High Instrumentation, unexpected as-left indications were found that impacted both of the required channels of instrumentation. Subsequent venting of the instrumentation lines was completed and both channels of instrumentation are reading consistent with previously taken as-found data. The instrumentation was declared OPERABLE at 1635. The initial cause of the unexpected as-left indications appears to be the introduction of air into the instrumentation lines during the calibration activities. This is considered a loss of safety function based on both of the HPCS Suppression Pool Level High Instrumentation channels being declared INOPERABLE and the loss of the automatic HPCS suction swap to the Suppression Pool on a high level. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The (NRC Resident Inspector) has been notified. |
| ENS 52877 | 28 July 2017 16:40:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | On July 20, 2017, a fire protection supply line at BVPS (Beaver Valley Power Station) Unit 1 experienced a mechanical joint separation. When the mechanical joint separated there was an unplanned flow of water from the fire protection line to the ground, which eventually re-entered the Ohio River via storm drains. The source of the water to the BVPS Unit 1 Fire Protection System is the Ohio River. The fire protection water was not treated with chemical additives. The water flow from this fire protection line was terminated on the same day. Due to the situation described above, FENOC (FirstEnergy Nuclear Operating Company) provided written documentation describing this occurrence to the Pennsylvania Department of Environmental Protection, (PADEP) on July 28, 2017. This offsite notification to PADEP requires a 4-hour report per 10 CFR 50.72 (b)(2)(xi). No press release is planned. The NRC Resident Inspector has been notified. |