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| Event date | Site | Region | Reactor type | Event description | |
|---|---|---|---|---|---|
| ENS 57431 | 19 November 2024 18:50:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1350 (EST) on 11/19/2024, with Unit 1 in mode 1 at 100% power, the reactor was manually tripped due to an automatic trip of the `B' main feedwater pump turbine. The reactor trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in mode 3. Decay heat removal is being accomplished by the steam dumps to the condenser. Emergency feedwater actuated due to low-low steam generator (water) level, as expected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the 'B' main feedwater pump trip is under investigation. There was maintenance involving the 'B' main feedwater pump at the time of the scram. |
| ENS 57432 | 19 November 2024 18:50:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 1350 (EST) on 11/19/2024, with Unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to an automatic trip of the `B' main feedwater pump turbine. The reactor trip was uncomplicated with all systems responding normally post trip. Operations stabilized the plant in mode 3. Decay heat removal is being accomplished by the steam dumps to the condenser. Emergency feedwater actuated due to low-low steam generator level, as expected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The cause of the 'B' main feedwater pump turbine trip is under investigation. |
| ENS 57420 | 11 November 2024 22:31:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 2250 EST on November 11, 2024, a technical specification required shutdown was initiated at Beaver Valley Power Station Unit 2. The following technical specification limiting conditions of operation (LCOs) were entered at 1939 EST on November 11, 2024: LCO 3.6.3, containment isolation valves, condition C, one or more penetration flow paths with one containment isolation valve inoperable; required action C.1, isolate the affected penetration flow path by use of at least one closed and de-activated automatic valve, closed manual valve, or blind flange. LCO 3.7.2, main steam isolation valves (MSIVs), condition C, one or more MSIVs inoperable in mode 2 or 3; required action C.1, close MSIV within 8 hours. These technical specification required actions will not be completed within the completion time; therefore, a technical specification required shutdown was initiated, and this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(i). With one main steam isolation valve inoperable, this condition is also being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). 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 failure occurred during planned surveillance testing in preparation for reactor startup. |
| ENS 57419 | 11 November 2024 20:10:00 | Pilgrim | NRC Region 1 | GE-3 | The following is summary of information provided by the licensee via phone and email: On November 11, 2024, at 1510 EST, site personnel identified what appeared to be water bubbling up from the pavement adjacent to the sanitary lift station 'C' outside of the facility industrial area. Less than 100 gallons of non-radiological sanitary water ran to a catch basin connected to permitted outfall number 007. Visual inspection did not identify any odor or indication of flow at outfall number 007 discharge. By 1530, the lift station pumps had been secured, sources of influent to the lift station were removed from service, and efforts were underway to pump the tank. At 1611, an offsite notification was made to the Environmental Protection Agency's Enforcement and Compliance Assurance Division in accordance with Section B of the station's National Pollutant Discharge Elimination System (NPDES) Permit No. 0003557. The event was associated with leakage from underground sewage system piping from a non-radiological underground tank and lift station. The NRC Resident Inspector will be notified. |
| ENS 57422 | 10 October 2024 14:02:00 | Millstone | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: At 0902 EST, on 10/10/2024, with Millstone Unit 3 in mode 1 at 100 percent power, it was discovered that the secondary containment boundary was inoperable when the latch that secured a hatch that was part of the secondary containment boundary was not functional. The latch was repaired by 1115, on 10/10/2024, and the secondary containment boundary was declared operable at 1200, on 10/10/2024. The initial assessment of reportability concluded that an immediate report was not required. However, upon additional review, it has been determined that because the secondary containment boundary is a single-train system that performs a safety function, an 8-hour report was required in accordance with 10 CFR 50. 72 (b)(3)(v)(C) and (D). This report should have been made on 10/10/2024 and is late. There has been no impact to Unit 2, and Unit 3 continues to operate in mode 1 at 100 percent power. There is no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57372 | 10 October 2024 09:57:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via phone and email: At 0557 EDT on 10/10/2024, with Unit 2 in mode 1 at 100 percent power, the reactor automatically tripped due to turbine generator loss of field. The trip was not complex with all systems responding normally post-trip. 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). Operations responded using EOP-0, post trip immediate actions, and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. 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 exciter is suspected to being the cause and is under investigation. All control rods fully inserted. |
| ENS 57424 | 1 October 2024 12:38:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via phone and email: NextEra Energy Seabrook LLC. makes the following notification under 10 CFR 21.21(d)(3)(i) of a defect found in a GE - Hitachi Relay, CR120B (Model #DD945E118P0060) during pre-installation bench testing. During bench testing, the relay failed to energize and transfer all associated contacts. The relay was purchased from GE - Hitachi (GEH) as safety-related, GE CR-120B relays. All GE CR-120B relays that were purchased in the same batch as the failed relay were located and quarantined in order to be returned to GEH for forensic testing. NextEra Energy Seabrook, LLC has concluded that this defect constitutes a substantial safety hazard (SSH). A SSH exists because the nature of the defect was such that, if installed in certain safety-related applications and failed, it would have prevented the fulfillment of a safety function. On November 12, 2024, the Seabrook site Vice President was notified of the requirement to report this event under 10 CFR 21.21. This is a non-emergency notification required by 10 CFR 21.21(d)(3)(i). A written notification will be provided in accordance with 10 CFR 21.21(d)(3)(ii). Because the defect was discovered prior to installation, there was no impact to safety-related equipment. The NRC Senior Resident Inspector has been informed. |
| ENS 57437 | 26 September 2024 22:01:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: This notification is a 10 CFR 21.21(a)(2) interim report for General Electric thermal overload relay, model CF124G011, part number DD317A7861P003. A sample of overload relays were sent to PowerLabs for parts quality initiative testing. The results were reviewed by James A. FitzPatrick Nuclear Power Plant (JAF) and a deviation in one relay component was discovered. Testing identified a failure to latch on trip, which is a deviation from the performance characteristics of the relay. Under normal operation, the relay would latch in the tripped state requiring a manual reset of the relay. If the relay with the deviation were installed, the relay would trip when required; however, it would automatically reset. The unexpected reset could result in unintended cycling of associated equipment including repeated exposure to inrush current and potential damage. Bench testing would be expected to identify this condition prior to installation. Based on a review, this potential condition does not affect installed equipment. The affected relay was stored at JAF since July 1998. The cause of the deviation cannot be investigated because the part is not available; however, the evaluation of the potential effect of the condition on equipment where the relay could have been used at JAF is ongoing, and it is expected to be completed by February 28, 2025. This notification is being submitted as an interim report per 10CFR21.21(a)(2). The NRC resident inspector has been notified. |
| ENS 57333 | 23 September 2024 11:20:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: At 0720 EDT on September 23, 2024, James A. FitzPatrick was at 100 percent power when an automatic scram occurred as a result of a main turbine trip due to an automatic trip of the generator output breakers; the cause is still under investigation. The scram was not complex. The automatic scram inserted all control rods. A subsequent reactor pressure vessel (RPV) low water level resulted in a group 2 isolation and initiation of high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems. RCIC did inject, but HPCI did not inject, as expected, based on RPV water level recovery with the feedwater system. Reactor pressure is being maintained by main steam line bypass valves. The plant is stable in Mode 3 with the 'A' reactor feed pump maintaining RPV water level. The initiation of the reactor protection system (RPS) due to the automatic scram signal while critical is reportable per 10 CFR 50.72(b)(2)(iv)(B). The general containment Group 2 isolations and HPCI and RCIC system actuations are reportable per 10 CFR 50.72(b)(3)(iv)(A). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The group 2 containment isolation affects multiple systems.
On 9/23/2024 at 1156 EDT Constellation communications provided a media statement to Oswego area news media contacts summarizing the events that had occurred at Nine Mile Point Unit 2 and FitzPatrick Unit 1. This is a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified. Notified R1DO (Dimitriadis) |
| ENS 57334 | 23 September 2024 11:20:00 | Nine Mile Point | NRC Region 1 | GE-5 | The following information was provided by the licensee via phone and email: On 9/23/2024 at 0720 EDT, with Unit 2 in mode 1 at 100 percent power, the reactor automatically scrammed due to turbine stop valve closure on a turbine trip. The scram was not complex. Due to the reactor protection system (RPS) actuation while critical, this event is being reported as a four hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Following the scram, reactor water level dropped below level 2 (108.8 inches), starting high pressure core spray (HPCS) and reactor core isolation cooling (RCIC); both injected into the reactor. RCIC is being used with turbine bypass valves to remove decay heat. Due to the emergency core cooling systems HPCS and RCIC discharging into the reactor coolant system, this event is being reported a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(A), and an eight hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). In addition, with reactor water level below level 2 (108.8 inches), primary containment isolation signals actuated resulting in group 2 recirculation sample system isolation, group 3 traveling in-core probe (TIP) isolation valve isolation, group 6 and 7 reactor water cleanup isolation, group 8 containment isolations, and group 9 containment purge isolations. This event is being reported as an eight hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A). Operations responded using procedure N2-EOP-RPV (1-3) and stabilized the plant in mode 3. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector was informed. There was no impact on Unit 1.
On 9/23/2024 at 1156 EDT Constellation communications provided a media statement to Oswego area news media contacts summarizing the events that had occurred at Nine Mile Point Unit 2 and FitzPatrick Unit 1. This is a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector has been notified. Notified R1DO (Dimitriadis), NRR EO (Felts), and IR MOC (Grant). |
| ENS 57282 | 21 August 2024 16:00:00 | Millstone | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via fax and phone: At 1200 EDT on 8/21/2024, with Millstone unit 3 in mode 1 at 100 percent power, it was discovered that the secondary containment boundary was inoperable while maintenance activities on the system were in progress. Therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(C) and (D). There is no impact on the health and safety of the public and plant personnel. The NRC Resident Inspector has been notified. Unit 3 continues to operate in mode 1 at 100 percent power with actions in progress to restore the system to operable within the technical specification allowed outage time. There has been no impact to unit 2, which remains at 100 percent power. The state of Connecticut and local towns were notified. |
| ENS 57233 | 18 July 2024 19:24:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via phone and email: At 1524 (EDT) on 07/18/2024, with Unit 2 in Mode 1 at 100 percent power, the reactor automatically tripped due to a turbine trip. The trip was not complex, with all systems responding normally post-trip. Operations responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 implemented AOP-7K (abnormal operating procedure), overcooling event, due to a grid transient. Operations responded and stabilized Unit 1 in Mode 1 at 100 percent power. Due to the reactor protection system actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. There were no other specified system actuations. |
| ENS 57302 | 10 July 2024 13:02:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: A 10 CFR 50.73(a)(1) invalid specified system actuation reported under 10 CFR 50.73(a)(2)(iv)(a) invalid actuation of residual heat removal (RHR). This 60-day telephone notification is being made per 10 CFR 50.73 (a)(2)(iv)(a) under the provision of 10 CFR 50.73 (a)(1) as an invalid actuation of the RHR. On July 10, 2024, while at 100 percent power, a partial train actuation of RHR was initiated by an invalid actuation signal while performing RHR valve logic testing. The cause for the RHR system logic actuation was due to improper configuration of an emergency core cooling system (ECCS) logic tester. The RHR system started and functioned as designed for the actuation signals it received from the ECCS logic tester. There was no impact on the health and safety of the public or plant personnel. The NRC resident inspector was notified. |
| ENS 57221 | 10 July 2024 11:28:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: At 0728 EDT on July 10, 2024, with Unit 2 in Mode 1 at 24 percent power, the reactor automatically scrammed due to a manual turbine trip. The (reactor) scram was not complex with all systems responding normally. Reactor vessel level reached the low-level set-point following the scram, resulting in valid Group 2 and Group 3 containment isolation signals. Due to the reactor protection system actuation while critical, this event is being reported as a four hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the Group 2 and Group 3 isolations. Operations responded using emergency operating procedures and stabilized the plant in Mode 3. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 3 was not affected. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 57172 | 13 June 2024 17:31:00 | FitzPatrick | NRC Region 1 | GE-4 | At 1331 EDT on 6/13/2024, it was determined that a non-active licensed operator supervisor tested positive in accordance with the fitness for duty testing program. The individual's authorization for site access has been denied. The NRC Resident Inspector has been notified. |
| ENS 57145 | 25 May 2024 08:00:00 | Millstone | NRC Region 1 | CE Westinghouse PWR 4-Loop | The following information was provided by the licensee by phone and email: A 50 ml bottle of vodka was found in the Unit 3 debris basket on the exterior of the intake structure. The bottle likely came from the ultimate heat sink (Niantic Bay) during normal backwash operations by the system that collects debris. Security has discarded the contraband. 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 bottle was found unsealed. |
| ENS 57132 | 19 May 2024 04:30:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via email: At 0030 (EDT) on 5/19/24, with Beaver Valley Unit 1 in mode 1 at 14 percent power, the reactor was manually tripped due to inability to control the A steam generator water level. The trip was not complex, with all systems responding normally post-trip. The turbine driven auxiliary feedwater pump automatically started on a valid actuation signal. All control rods inserted into the core. Operations responded and stabilized the plant. Decay heat is being removed by the feedwater system and the main condenser. Beaver Valley Unit 2 is unaffected. Due to the reactor protection system 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). Due to the emergency safety feature system actuation (automatic start of the turbine driven auxiliary feedwater pump) while critical, this event is being reported as an eight-hour, non-emergency notification per 10CFR 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 verbally notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 is stable on off-site power, normal configuration. All emergency systems are available. |
| ENS 57120 | 9 May 2024 20:29:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: At 1629 EDT on 05/09/2024, the high pressure coolant injection (HPCI) system was declared inoperable due to a pinhole through-wall leak identified on the seal drain line for 23HOV-1 (HPCI trip throttle valve) downstream of the restricting orifice 23RO-137A. The location of the defect is in the class 2 safety related piping. HPCI is a single train safety system and this notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(D). The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: This pinhole leak was discovered during normal operator rounds. Although HPCI is declared inoperable and in a 14-day limited condition of operation, the system function remains available. In addition, all other ECCS systems are currently operable. Compensatory measures (walkdowns) have been implemented to ensure the leak rate does not significantly increase.
FitzPatrick performed an additional technical evaluation of the steam leak identified on May 9, 2024. The evaluation concluded that the HPCI system would have remained operable and performed its specified safety function with a postulated complete failure of this pipe, considering its size, location, and impact of the leak. Additionally, all components in the vicinity would have retained their required safety functions. Based on this conclusion, EN 57120 is being retracted. The NRC Senior Resident Inspector has been notified. Notified R1DO (Elkhiamy). |
| ENS 57115 | 9 May 2024 12:00:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 0800 EDT on May 9, 2024, it was identified during leak rate testing that through-wall flaws existed on reactor plant river water piping inside the containment building. This determination resulted in a containment bypass condition such that a gaseous release could have occurred at a location not analyzed for a release in the loss of coolant accident dose consequence analysis. This condition is not bounded by existing design and licensing documents. Evaluation of the condition of the piping is ongoing to support repair prior to startup. With the plant currently in cold shutdown, the containment, as specified in Technical Specification 3.6.1, is not required to be operable. There was no impact on the health and safety of the public or plant personnel. This event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(A), 10 CFR 50.72(b)(3)(ii)(B), and 10 CFR 50.72(b)(3)(v)(C). The NRC Resident Inspector has been notified. |
| ENS 57103 | 3 May 2024 08:11:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0411 EDT on 5/03/2024, it was determined that primary containment did not meet TS (Technical Specification) 4.6.1.2 (surveillance) requirement due to a primary containment leak rate test exceeding `La (allowable leakage rate). 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. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The final observed leak rate is still being calculated as the test is still within the stabilization period. Testing is allowed within the stabilization period for an unspecified amount of time. Short term corrective actions are to identify and repair any leak paths. No mode changes are required due to this event. |
| ENS 57075 | 13 April 2024 04:35:00 | Beaver Valley | NRC Region 1 | Westinghouse PWR 3-Loop | The following information was provided by the licensee via phone and email: At 0035 EDT on April 13, 2024, with Unit 1 at 97 percent power, the reactor automatically tripped due to 1 of 3 reactor coolant pump (RCP) low flow reactor trip (signal) associated with a loss of the A and B 4160 volt normal buses. Auxiliary feedwater and the 1-1 emergency diesel generator (EDG) automatically started on valid actuation signals. The 1-1 EDG sequenced on to supply all required loads per plant design. All control rods fully inserted and the trip was not complex with all systems responding normally post-trip. Operators have responded and stabilized the unit in Mode 3 (Hot Standby). Decay heat is being removed by discharging steam to the main condenser via the condenser steam dump system with steam generators being supplied by the main feedwater system. Unit 2 is not affected by the event. 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). Due to the valid actuations of auxiliary feedwater and the 1-1 EDG, this event is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). There was no impact to the health and safety of the public or plant personnel. The NRC senior resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Power for the A-E Bus is on the 1-1 EDG. The D-F Bus is on offsite power. One electrical train of offsite power is down. |
| ENS 57004 | 4 March 2024 00:42:00 | Nine Mile Point | NRC Region 1 | GE-5 | The following information was provided by the licensee via email: On 3/3/24 at 1942 EST, while performing a plant shutdown in preparation for a refuel outage, Nine Mile Point Unit 2 experienced a reactor scram due to a main turbine trip on low condenser vacuum. The plant was at approximately 55 percent power at the time of the reactor scram. Additionally, following the scram a low RPV (reactor pressure vessel) level scram and containment isolation signal on level 3 was received, as expected. The containment isolation signal impacted RHR (residual heat removal) shutdown cooling, RHR letdown to radwaste, and RHR sampling. All impacted valves were closed at the time the isolation occurred. All control rods were fully inserted. Plant response was as expected. Post scram, the main turbine bypass valves are being used to control decay heat, and normal post scram level control is via the feed / condensate system. This is being report under 10 CFR 50.72(b)(2)(iv)(B), 'RPS Actuation', and 10 CFR 50.72(b)(3)(iv)(A), 'Specified System Actuation'. Unit 1 is not affected. 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 cause of the low condenser vacuum was a momentary loss of sealing steam. The condenser remained viable for decay heat removal. All safety equipment is available. The grid is stable with the plant in its normal shutdown electrical configuration. |
| ENS 56997 | 28 February 2024 18:50:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via phone and email: At 1350 EST on 2/28/2024, with Calvert Cliffs Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 65 percent power, an actuation of the '1A' and '2A' emergency diesel generators' auto-start occurred due to an undervoltage condition on the number 11 and number 21 4kV buses which are fed from the number 11 13kV bus. The '1A' and '2A' emergency diesel generators automatically started as designed when the 4kV buses' undervoltage signals were received. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of the '1A' and '2A' emergency diesel generators. 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 undervoltage condition was caused by the feeder breaker to the number 11 13 kV bus opening during electrical maintenance. |
| ENS 56991 | 24 February 2024 20:46:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via email: At 1546 EST, with unit 2 at 100 percent power, the reactor was manually tripped due to the '22' steam generator feed pump tripping. The trip was uncomplicated with all systems responding normally post-trip. 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). Operations responded using emergency operation procedure EOP-0, Post Trip Immediate Actions and EOP-1, Uncomplicated Reactor Trip and stabilized the plant in mode 3. Decay heat is removed by discharging steam to the main condenser using the turbine bypass valves. Unit 1 is not affected. ESFAS (engineered safety features actuation systems) actuation (auxiliary feedwater manual actuation) is reportable under 10 CFR 50.72(b)(3)(iv)(A) 8-hour report. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56980 | 19 February 2024 15:45:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 1045 EST, on 2/19/2024, during a maintenance activity, a loss of all reactor building ventilation occurred on Unit 2. With no flow past the ventilation radiation monitors, the radiation monitors were inoperable to support their ability to perform primary and secondary containment isolation functions or start the standby gas treatment system. Reactor building ventilation was restored within 15 minutes. Due to this inoperability, the radiation monitor system was in a condition that could have prevented fulfillment of a safety function; therefore, this condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector will be notified.
Upon further investigation, it was verified that the reactor building and the refueling floor radiation monitors are not needed to control the release of radiation for events described in chapter 14 of the updated Final Safety Analysis Report. For the analyzed loss of coolant accident (LOCA), the primary and secondary signals for this purpose were available and unaffected by this event. The radiation monitors provide a tertiary redundant method that is not credited within the station analysis. For all other analyzed accidents, the signal provided by the radiation monitors is not needed, as the secondary containment isolation function and start of the standby gas treatment system are not credited. Additionally, the fuel handling accident was not credible during the time of the event because no activities were in progress on the refueling floor. Therefore, the threshold for reporting the issue as an event or condition that could have prevented the fulfillment of a safety function was not met. The NRC Resident Inspector has been notified. Notified R1DO (Jackson) |
| ENS 56957 | 9 February 2024 18:22:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: On 2/9/24 at 1322 EST, it was determined that the unit was in an unanalyzed condition. A review of DC feeder circuit protection schemes identified a circuit for the fuel pool cooling system is uncoordinated due to inadequate fuse sizing. This results in a concern that postulated fire damage in one area could cause a short circuit without adequate protection, leading to the unavailability of equipment credited for in 10 CFR 50 Appendix R, Fire Safe Shutdown. This condition is not bounded by existing design and licensing documents; however, it poses no impact to the health and safety of the public or plant personnel. Therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). The postulated event affects the following fire zones: fire areas 6S and 6N (within the Unit 2 reactor building). Compensatory actions for affected fire areas have been implemented. An extent of condition review is being performed. The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Fire watches have been established in the affected areas. These will be maintained until the protection scheme is revised.
The following updated information was provided by the licensee via email and phone call: On 03/08/24 at 1418, extent of condition reviews identified circuit(s) in the Units 2 and 3 Reactor Protection Systems (RPS) which are also uncoordinated due to improper fuse sizing. These circuits are not bounded by existing design and licensing documents for 10 CFR 50 Appendix R Fire Safe Shutdown and, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel. The postulated event affects the following fire areas: 32, 33, 38 and 39 (Units 2 and 3 Switchgear Rooms). In accordance with procedural requirements, compensatory actions for the affected fire areas have been implemented and will remain until the condition is resolved. The NRC Senior Resident Inspector has been notified. Notified R1DO (Arner)
On March 13, 2024, at 1350 EDT, extent of condition reviews identified a circuit in the Unit 2 reactor protection system (RPS) which is also uncoordinated due to improper fuse sizing. This circuit is not bounded by existing design and licensing documents for 10 CFR 50, Appendix R, Fire Safe Shutdown, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel. The postulated event affects fire area 57 (Switchgear Corridor, common to Units 2 and 3). In accordance with procedural requirements, compensatory actions for the affected fire areas have been implemented and will remain until the condition is resolved. Additionally, it was previously reported that fire area 6N contained a circuit which was not bounded by the Fire Safe Shutdown analysis; however, after further review it has been determined that compliance is maintained in this fire area and is therefore retracted from the scope of this report. The NRC Senior Resident Inspector has been notified. Notified R1DO (Jackson)
The following information was provided by the licensee via email: On 03/21/24 at 1211, extent of condition reviews identified an annunciator circuit for the Unit 3 emergency service water (ESW) and high pressure service water (HPSW) pump structure heating and ventilation panel that is also uncoordinated due to improper fuse sizing. This circuit is not bounded by existing design and licensing documents for 10 CFR 50 Appendix R Fire Safe Shutdown and, therefore, this event is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(ii)(B). This event poses no impact to the health and safety of the public or plant personnel. The postulated event affects fire area 47 (Unit 3 pump structure for `B' ESW and `3A'-`3D' HPSW pumps) and the yard fire area (Manhole 026D). In order to restore immediate compliance, the cable has been de-energized to eliminate the possibility of the event of concern. This circuit will remain de-energized or other measures will be implemented until the condition is permanently resolved. The NRC Senior Resident Inspector has been notified. Notified R1DO (Ford) |
| ENS 56952 | 7 February 2024 21:00:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: A programmatic vulnerability, failure, or degradation was discovered within the fitness for duty (FFD) program that may permit undetected drug or alcohol use or abuse by individuals within the protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program. Public and plant safety have not been affected. The NRC Resident Inspector was notified. |
| ENS 56939 | 30 January 2024 14:37:00 | Limerick | NRC Region 1 | GE-4 | The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: On January 30, 2024, a non-licensed employee supervisor, after investigation, was determined to be in involved with a controlled substance. The employee's access to the site has been placed on administrative hold, pending further investigation. The NRC Resident Inspector has been notified. |
| ENS 56936 | 29 January 2024 17:02:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram caused by a main turbine trip. Investigation is still ongoing. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All control rods were fully inserted. The licensee indicated that the turbine trip may have been caused by a power load imbalance, however the cause of the incident is under investigation. The scram was not complex. Decay heat is currently being removed thru bypass valves dumping to the main condenser. Initially unit 2 lost the use of the bypass valves due to lack of condenser vacuum. Unit 2 used the high pressure coolant injection (HPCI) system in the condenser storage tank (CST) to CST mode to remove decay heat. Residual heat removal was used to keep the torus cool. Condenser vacuum was regained and unit 2 is back to removing decay heat with the turbine bypass valves. There was no impact to unit 3. The licensee confirmed there was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
The following information was provided by the licensee via email: Licensee adds 8-hour non-emergency 10 CFR 50.72(b)(3)(iv)(A) specified system actuation report to original 4-hour non-emergency 10 CFR 50.72(b)(2)(iv)(B) RPS Actuation report. At approximately 1202 EST on 01/29/24, unit 2 experienced a reactor scram by a main turbine trip. All control rods inserted. Reactor core isolation cooling system (RCIC) was manually initiated for level control. HPCI was manually initiated for pressure control. Primary containment isolation system (PCIS) Group II and III isolations occurred (specified system actuation). Investigation is ongoing. The NRC Resident Inspector has been notified. |
| ENS 56889 | 15 December 2023 00:39:00 | Hope Creek | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone call and email: On December 14, 2023, at 1939 EST, Hope Creek reactor scrammed following closure of turbine control valve number 4. All control rods fully inserted into the core. All safety systems responded as designed and expected. There was no radiological release. The unit is stable in mode 3 with decay heat being removed via the turbine bypass valves rejecting steam to the main condenser. Normal feedwater level control is providing makeup to the reactor vessel. No personnel injuries resulted from the event. The outage control center has been staffed to determine the cause of the reactor scram. The Hope Creek NRC Resident Inspector has been notified. |
| ENS 56875 | 29 November 2023 20:00:00 | Indian Point | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: This notification is being made per 10 CFR 50.72(b)(2)(xi), as a result of notifications made to State and local government agencies for the discovery of an oil sheen in the discharge canal outside Unit 3. The New York State Department of Environmental Conservation and Westchester County Department of Health were notified. No sheen was observed in the river or at the southern end of the discharge canal near the outfall gates. Clean up efforts are underway. The licensee will notify the NRC Project Manager. |
| ENS 56856 | 16 November 2023 07:27:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via email: At 0227 EST on 11/16/23, Calvert Cliffs Unit 2 experienced an automatic trip from the reactor protection system (RPS) based on reactor trip bus undervoltage (UV). At that time, a loss of U-4000-22 (13 kV to 4 kV transformer) caused a loss of 22, 23, and 24 4 kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV. The loss of 22 and 23 4 kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4 kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4 hour report ESFAS (engineering safety features actuation system) actuation (2B DG start on UV) is reportable under 10 CFR 50.72(b)(3)(iv)(A) - 8 hour report AFW operation is reportable under 10 CFR 50.73(a)(2)(iv)(A) - 60 day report The NRC Senior Resident Inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: All rods fully inserted. There was no impact on Unit 1 operations. Unit 2 is stable in mode 3.
ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8 hour report Notified R1DO (Defrancisco). |
| ENS 56846 | 10 November 2023 08:14:00 | Susquehanna | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: At 0118 EST, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually scrammed due to degrading main condenser vacuum. The scram was not complex, with all systems responding normally post-scram. The main turbine bypass valves opened automatically to maintain reactor pressure. Operations responded and stabilized the plant. Reactor water level is being maintained via feedwater pumps. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Unit 2 is not impacted. Due to Reactor Protection System actuation while critical, this event is being reported as a four-hour and eight-hour non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). Unit 1 reactor is currently stable in mode 3. An investigation is in progress into the cause of the degrading condenser vacuum. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56841 | 8 November 2023 11:45:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via phone and email: At 0645 EST, on November 8, 2023, with Unit 2 in Mode 3 at zero percent power, a manual actuation of the auxiliary feedwater system (AFW) occurred during a planned plant cooldown. The reason for the AFW manual-start was a trip of the 22 steam generator feed pump due to a high casing level. The 23 AFW motor driven pump was manually started in accordance with implementation of AOP-3G, Malfunction of Main Feedwater System to restore steam generator levels. There was no impact to Unit 1. 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. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: No other systems were affected. No other compensatory or mitigation strategies implemented. Plant cooldown was the only significant evolution in progress. No impact to other technical specifications or limiting conditions for operation. All systems functioned as required. The electric plant is being supplied by offsite power with all diesel generators available. No significant increase in plant risk. There was nothing unusual or not understood. |
| ENS 56839 | 7 November 2023 21:17:00 | Calvert Cliffs | NRC Region 1 | CE | The following information was provided by the licensee via email: At 1617 on 11/7/2023, Calvert Cliffs Unit 2 experienced an automatic trip from a Reactor Protection System (RPS) based on reactor trip bus under voltage (UV). At that time a loss of U-4000-22 caused a loss of 22, 23, and 24 4kV busses. This resulted in a loss of both motor generator (MG) sets causing the reactor trip bus UV condition. The loss of 22 and 23 4kV non-safety related busses resulted in a loss of main feedwater. Auxiliary feedwater (AFW) was manually initiated and is feeding both steam generators. The 2B diesel generator (DG) started and restored the 24 4kV safety related bus. Heat removal is via the normal turbine bypass valves to the main condenser. RPS actuation is reportable under 10 CFR 50.72(b)(2)(iv)(B) - 4-hour report. ESFAS actuation (2B DG start on UV) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. ESFAS actuation (AFW manual initiation) is reportable under 10CFR50.72(b)(3)(iv)(A) - 8-hour report. Site Senior NRC resident inspector has been notified. The following additional information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: Unit 1 was unaffected. Estimation of duration of shutdown is 24 hours. |
| ENS 56838 | 7 November 2023 17:00:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On November 07, 2023 at 1200 EST, it was discovered that all pumps in the Auxiliary Feedwater system were inoperable due to the loss of control power to the 'B' train Emergency Feedwater (EFW) flow control valve which supplies the 'D' steam generator. The redundant 'A' train EFW control valve for the 'D' steam generator remains functional, as well as the capability of the Auxiliary Feedwater system to supply all steam generators. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(B). 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 'A' and 'B' EFW Flow Control Valves are arranged in a series configuration for each Steam Generator. Failure of any of the 8 EFW Flow Control Valves to meet its Surveillance Requirements will render all EFW Pumps inoperable per tech specs. |
| ENS 56822 | 30 October 2023 16:00:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone call and email: A non-licensed supervisory employee had a confirmed positive test 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 56810 | 23 October 2023 00:48:00 | Nine Mile Point | NRC Region 1 | GE-2 | The following information was provided by the licensee via phone and email: On October 21, 2023, at 2048 EDT, reactor recirculation pump (RRP) 12 tripped. The cause for the trip is under investigation. Following the RRP trip, the average power range monitors (APRMs) flow bias trips were inoperable due to reverse flow through RRP 12. The APRMs were restored to operable on October 21, 2023, at 2058 EDT, when the RRP 12 discharge blocking valve was closed. This 8-hour non-emergency report is being made based upon requirements of 10CFR50.72(b)(3)(v)(A) which states: "Licensee shall notify the NRC of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) Shut down the reactor and maintain it in a safe shutdown condition." The NRC Resident Inspector has been notified. |
| ENS 56790 | 13 October 2023 01:27:00 | Ginna | NRC Region 1 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via email: On 10/12/23 at 2127 EDT, with the Unit 1 in Mode 1 at 100% Power, operators identified degrading condenser vacuum and manually tripped the reactor. All control rods inserted as expected. The trip was not complex, and all systems responded normally post-trip. The cause of the degraded condenser vacuum was an unexpected closure of the condenser air ejector regulator. The cause of the air ejector regulator going closed is not fully understood and is being investigated. Following the SCRAM, Operators responded and stabilized the plant. Decay heat is being removed by the Main Steam System through the Atmospheric Relief Valves (ARVs) and Auxiliary Feed Water (AFW) systems. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56753 | 20 September 2023 19:15:00 | Peach Bottom | NRC Region 1 | GE-4 | The following information was provided by the licensee via phone and email: A licensed operator had a confirmed positive test for alcohol during another entity's pre-access fitness-for-duty screening for unescorted access authorization. The individual's unescorted access at Peach Bottom Atomic Power Station has been denied. The NRC Resident Inspector has been notified. |
| ENS 56731 | 9 September 2023 15:43:00 | Ginna | NRC Region 1 | Westinghouse PWR 2-Loop | The following information was provided by the licensee via email: On 9/9/23 at 1143 EDT, with the Unit 1 in Mode 1 at 100 percent power, all 4 turbine control valves closed resulting in a reactor protection system (RPS) automatic reactor trip on over temperature differential temperature. All control rods inserted as expected. The trip was not complex and all systems responded normally post-trip. The cause of the control valve closure has not been determined. Following the SCRAM, operators responded and stabilized the plant. Decay heat is being removed by the main steam system through the atmospheric relief valves and auxiliary feed water systems. Due to the RPS actuation while critical, this event is being reported as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B) and an eight-hour non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for a valid specified system actuation. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. |
| ENS 56710 | 2 September 2023 10:32:00 | Nine Mile Point | NRC Region 1 | GE-5 | The following information was provided by the licensee via email: On 9/2/2023 at 0632 EDT, a feedwater transient occurred resulting in an reactor protection system (RPS) automatic reactor scram on low level (Level 3, 159.3 inches). Following the scram, reactor water level dropped below Level 2 (108.8 inches) resulting in a Group 2 recirculation sample system isolation, Group 3 traveling in-core probe (TIP) isolation valve isolation, Group 6 and 7 reactor water cleanup isolation, and Group 9 containment purge isolations. All control rods inserted as expected. High pressure core spray and reactor core isolation cooling initiated and injected as expected. ECCS systems have been secured and normal reactor pressure and level control has been established for hot shutdown. Nine Mile Point Unit 2 is stable and in Mode 3. These 4 hour and 8 hour non-emergency reports are being made in accordance with 10 CFR 50.72(b)(2) (iv)(A), 10 CFR 50.72(b)(2)(iv)(B), and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident was informed. There was no impact on Unit 1. |
| ENS 56667 | 8 August 2023 04:00:00 | FitzPatrick | NRC Region 1 | GE-4 | The following information was provided by the licensee via email: A licensed (non-active) individual failed to comply with fitness for duty testing policies. The individual's unescorted access was terminated. |
| ENS 56652 | 3 August 2023 14:03:00 | Nine Mile Point | NRC Region 1 | GE-2 | The following information was provided by the licensee via email: A licensed employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The NRC Senior Resident Inspector was notified. |
| ENS 56645 | 30 July 2023 19:26:00 | Seabrook | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: On July 30, 2023 at 1526 EDT, with unit 1 in mode 1 at 100 percent power, the reactor was manually tripped due to low main turbine electro-hydraulic control oil level. The trip was uncomplicated with all systems responding normally post-trip. Operations stabilized the plant in mode 3. Decay heat removal is being accomplished using the steam dumps in steam pressure mode to the main condenser. Emergency Feedwater actuated due to low-low steam generator level as expected. This event is being reported pursuant to 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. |
| ENS 56637 | 25 July 2023 13:24:00 | Millstone | NRC Region 1 | CE | The following information was provided by the licensee via email: At 0924 (EDT) on July 25, 2023, it was discovered that both trains of control room air conditioning system were simultaneously inoperable due to an inoperable control room envelope boundary. The boundary was restored at 0925 (EDT) on July 25, 2023. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. There has been no impact to Unit 3 which remains at 100% power.
The licensee determined in a subsequent engineering evaluation of the conditions that existed at the time, that the access hatch being open did not have an adverse impact upon the control room emergency ventilation system and the control room envelopes boundary's ability to perform their safety function including: Radiation dose to the occupants did not exceed the licensing basis, design basis accident calculated value. Protection of control room occupants from hazardous chemicals and smoke. Therefore, this condition is not reportable and NRC Event EN56637 is being retracted. The basis for this conclusion has been provided to the NRC Resident Inspector. Notified R1DO (Lally) |
| ENS 56620 | 12 July 2023 08:49:00 | Millstone | NRC Region 1 | Westinghouse PWR 4-Loop | The following information was provided by the licensee via email: At 0449 (EDT) on 7/12/2023, Millstone Unit 3 declared the 'B' train of the emergency core cooling system (ECCS) inoperable due to a degraded damper associated with the ventilation support system for the 'B' charging pump. At the time of this event, the 'A' train of service water was already inoperable due to planned maintenance on a breaker that would have prevented an 'A' service water valve powered from this breaker from closing on a safety signal. This configuration resulted in the possibility that the 'A' train of ECCS would not have been available to fulfill its design function under all postulated accident conditions. This event is being reported under 10 CFR 50.72(b)(3)(v)(B), '(any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) remove residual heat).' Subsequently, the 'A' train of service water was restored to operable at 0548 on 7/12/2023. Repairs and investigation continue on the 'B' train ECCS damper. The NRC resident has been notified. This event did not impact Millstone Unit 2. There was no impact to the public.
The following information was provided by the licensee via email: The condition was reported to the NRC pursuant to 10 CFR 50.72(b)(3)(v)(B), via an 8-hour report as 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 remove residual heat. A subsequent engineering review of the conditions that existed at the time determined that, based on area temperature response, any impact on ventilation flows into and out of the `B' charging pump cubicle did not generate an observable change in the temperature trend. Based on this, it is concluded with reasonable assurance that the functional requirement of the support system was maintained and the `B' charging pump would have continued to perform its safety function until the `A' train of service water was restored to operable and as a result safety function was not lost. Therefore, this condition is not reportable and NRC Event Number 56620 is being retracted. The basis for this conclusion has been provided to the NRC Resident Inspector." Notified R1DO (Bicket). |
| ENS 56610 | 6 July 2023 16:32:00 | Millstone | NRC Region 1 | CE | The following information was provided by the licensee via email: On July 6, 2023, at 1232 EDT, while operating in Mode 1 at 100 percent power, the supply check valve from the number 2 steam generator to the turbine driven auxiliary feedwater pump was determined during troubleshooting that it is not able to perform its isolation function. This failure would have resulted in the blowdown of both steam generators during a main steam line break in the number 2 steam generator main steam line upstream of the main steam isolation valves until the operators could isolate the faulted steam generator. Previous evaluation has determined that this condition constituted an unanalyzed condition that could impact containment pressure. There was no radioactive release to the environment. The steam line from the steam generator to the turbine driven auxiliary feedwater pump was isolated by use of a motor operated valve in the discharge line of the number 2 steam generator. There was no impact to Unit 3 which remains at 100 percent power. The NRC Senior Resident Inspector was notified. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B) as a condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. |
| ENS 56569 | 11 June 2023 05:30:00 | 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 56560 | 6 June 2023 13:37:00 | Three Mile Island | NRC Region 1 | B&W-L-LP | The following information was provided by the licensee via email: At 0937 EDT on June 6, 2023, it was discovered that a site employee suffered a non-work-related fatality. The individual was found non-responsive outside the Radiological Controlled Area. This is a four-hour, non-emergency notification for which a notification to other government agencies has been made. This event is reportable in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Region I inspector has been notified. |