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The following information was provided by the licensee via phone and email: At 0209 EST on 02/23/2026, with Hope Creek in Mode 1 at 100 percent power, the reactor automatically tripped due to a loss of offsite power. All emergency diesel generators (EDGs) started and loaded as expected to power vital buses. Operations responded and stabilized the plant in Mode 3. Reactor water level is being maintained via high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC). Reactor pressure is being maintained using safety relief valves (SRVs). 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). An Unusual Event (SU1.1) was declared at 0229 on 02/23/2026, due to a loss of offsite power to all vital buses for greater than 15 minutes. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified, and all offsite notifications have been made. Notified DHS SWO, FEMA Operations Center, FEMA NWC, CISA Central Watch Officer, CWMD Watch Desk, DHS Nuclear SSA (email), DHS NRCC THD Desk (email).

• * * UPDATE ON 02/23/2026 AT 1352 EST FROM HOPE CREEK TO NESTOR MAKRIS * * *

On 02/23/2026, at 1340 EST, Hope Creek terminated their Unusual Event (SU1.1), which was declared due to a loss of offsite power.

The following information was provided by the licensee via phone and fax: Entry into shutdown technical specification action statement due to an identified breach in ventilation ductwork. At 2250 EDT on September 9, 2025, it was discovered that there was degraded manway sealant on the manway to fire damper HV-298A. This degraded sealant results in a direct path from the enclosure building to the atmosphere, challenging the enclosure building boundary. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(C). There is no impact on the health and safety of the public or plant personnel. The plant is currently in a 24-hour technical specification action statement (3.6.5.2) for Unit 2. Unit 3 is not impacted and continues to operate at 100 percent power. The Resident Inspector has been notified.

• * * RETRACTION ON SEPTEMBER 12, 2025, AT 0908 EDT FROM JARED FARLEY TO ERIC SIMPSON * * *

The following information was provided by the licensee via phone and fax: Millstone Unit 2 is retracting NRC Event Notification (EN) 57912, made on September 10, 2025, at 0523 EDT, regarding a condition identified at Millstone Power Station Unit 2. The condition involved degraded sealant on manways to fire dampers HV-298A/B/G, which resulted in a direct path from the enclosure building to the atmosphere, challenging the integrity of the enclosure building boundary. This condition was initially reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(C) for an event or condition that could have prevented the fulfillment of a safety function (control of release of radioactive material). A subsequent review using additional information on hatch design and actual seating surface determined that there is reasonable assurance the enclosure building boundary remained operable and retained its safety function to control the release of radioactive material and mitigate the consequences of an accident. Based on this assessment, Unit 2 exited Technical Specification action statement 3.6.5.2, and the condition is not reportable under 10 CFR 50.72(b)(3)(v)(C). Therefore, NRC EN 57912 is being retracted. The Resident Inspector has been notified. Notified R1DO (Ford).

The following information was provided by the licensee via phone and email: On April 7, 2025, at 2000 EDT, 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 flow control valve for the `D' steam generator remains functional, as well as the capability of the auxiliary feedwater system to supply all steam generators. 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 technical specifications (TS). 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: Troubleshooting is ongoing for the cause of the loss of control power. The licensee is currently in technical specification limiting condition for operation 3.7.1.2 action statement 'D' to restore at least one auxiliary feedwater pump to operable status as soon possible.

• * * RETRACTION ON 04/09/2025 AT 1049 EDT FROM BOB MURRELL TO JORDAN WINGATE * * *

The following information was provided by the licensee via phone and email: The purpose of this notification is to retract a previous report made on 04/08/2025 at 1344 EDT, EN 57652. Notification of the event to the Nuclear Regulatory Commision was initially made because of declaring all pumps in the emergency feedwater system inoperable in response to discovering 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 flow control valve for the `D' steam generator remained functional, as well as the capability of the EFW system to supply all steam generators. Subsequent to the initial report, NextEra has concluded that the loss of control power to the `B' train EFW flow control valve did not prevent EFW from fulfilling its safety function. Specifically, the `A' train of EFW was fully capable of fulfilling all EFW safety functions. Therefore, this event is not considered a safety system functional failure and is not reportable pursuant 10 CFR 50.72(b)(3)(v)(B). The NRC Resident Inspector has been notified. Notified R1DO (Schussler)

The following information was provided by the licensee via phone and email: At approximately 0230 EST on 2/18/2025, it was determined that the primary and backup methods of activating the site Emergency Response Organization (ERO) were not available. External network access was unavailable, to include access to the Dominion Energy Emergency Notification System (DEENS) rendering the system non-functional. Subsequently, the backup method for activating the station ERO was attempted and was unable to be verified. On 2/18/2025, at approximately 0400, an additional backup ERO activation phone number was provided and verified to be functional. At approximately 0740 on 2/18/2025, external network access was restored and DEENS was restored to functional. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(xiii). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

• * * UPDATE ON 2/19/25 AT 1030 EST FROM JOSHUA LINDSEY TO JORDAN WINGATE * * *

The following information was provided by the licensee via phone and email: It has since been determined that the (backup) phone number provided at 0400 would not have worked to activate the ERO. The NRC Resident Inspector has been notified. Notified R1DO (Deboer)

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.

• * * RETRACTION ON 12/30/2024 AT 1025 EST FROM JAMIE SMITH TO JORDAN WINGATE * * *

The following information was provided by the licensee via phone and email: On 11/12/2024 an 8-hour notification was made describing the failure of one main steam isolation valve (MSIV) to close during testing while the plant was in MODE 3. This notification was made pursuant to 10 CFR 50.72(b)(3)(v) as a condition that could have prevented the fulfillment of a safety function. Further engineering evaluation has determined that this condition was not reasonably expected to prevent the fulfillment of a safety function based on a review of the accident analyses and the redundant equipment which is known to have been capable of performing the safety functions. Therefore, this does not result in a reportable condition under this criterion. At the same time, a 4-hour notification was made pursuant to 10 CFR 50.72(b)(2)(i) as an initiation of a shutdown required by plant technical specification due to initiating a transition to MODE 4 to exit TS applicability. The plant was in hot standby (MODE 3) at the time of the event and negative reactivity was not added in order to move to MODE 4, therefore, this event is not reportable under this criterion. The NRC Resident Inspector has been notified. Notified R1DO (Lilliendahl)

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.

• * * UPDATE ON 9/23/2024 AT 1550 EDT FROM RYAN LOOMIS TO IAN HOWARD * * *

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).

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.

• * * RETRACTION ON 06/20/2024 AT 1423 EDT FROM CAMERON KELLER TO ROBERT THOMPSON * * *

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).