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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.

• * * UPDATE ON 03/08/24 FROM PAUL BOKUS TO TOM HERRITY * * *

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)

• * * UPDATE ON 3/13/2024 AT 1538 FROM TROY RALSTON TO SAM COLVARD * * *

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)

• * * UPDATE ON 3/21/2024 AT 1525 FROM PAUL BOKUS TO IAN HOWARD * * *

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)

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

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

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

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

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

While Unit 3 was shutting down for 3R22 refueling outage, the mode switch was taken to shutdown position which is a manual scram signal. The manual scram signal was not received from the mode switch. A subsequent manual scram was inserted with the use of the manual scram push buttons. The Unit 3 reactor is shutdown with all rods inserted. Unit 2 was unaffected by the event and remains in Mode 1 at 100 percent power. The licensee notified the NRC Resident Inspector, Pennsylvania and Maryland State Agencies, local government. A media press release is planned. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email), FEMA Region 3 Watch Office (email).

• * * UPDATE AT 0316 EDT ON 10/21/19 FROM KEVIN GROMANN TO BETHANY CECERE * * *

Conditions no longer meet an Emergency Actuation Level and will not deteriorate. Unit 3 reactor is shutdown with all control rods fully inserted. The NOUE was terminated at 0230 EDT." The licensee notified the NRC Resident Inspector, Pennsylvania and Maryland State Agencies, local government. Notified the R1DO (Jackson), NRR EO (Miller), IRDMOC (Gott), R1RA (Lew via email), NRR (Nieh via email), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email), FEMA Region 3 Watch Office (email).

On 9/21/18, at 1755 EDT, Peach Bottom Atomic Power Station Unit 3 declared the High Pressure Coolant Injection system (HPCI) inoperable due to an inoperable differential pressure indicating switch (DPIS). The DPIS is used to isolate the HPCI system when there is a high steam line flow condition. Operations declared the HPCI system inoperable and entered Technical Specification 3.5.1 Condition C for HPCI being inoperable. Technical Specification 3.3.6.1 was also entered for HPCI instrumentation being inoperable. Other standby systems (Reactor Core Isolation Cooling and Low Pressure Emergency Core Cooling Systems) are OPERABLE. HPCI is a single train system. Therefore, per NUREG-1022, this condition is being reported pursuant to 10CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of the safety function of a system required to mitigate the consequences of a design event. This condition has been entered into the corrective action program (IR 4175355). Investigation of the exact cause of the indication issue is in progress. The NRC Resident has been informed of this notification.

• * * UPDATE AT 1317 EDT ON 09/22/2018 FROM CRAIG TAULMAN TO JEFF HERRERA * * *

On 09/22/18 at 0955 EDT, RCS (Reactor Coolant System) pressure boundary leakage was identified as the cause of the HPCI high steam flow indication issue. Technical Specification 3.4.4 was entered which will require the initiation of a nuclear plant shutdown. This indicates a degradation of a principal safety barrier. Current Unit 3 reactor power is 35%. This condition is being reported pursuant to 10 CFR 50.72(b)(2)(i) and 50.72(b)(3)(ii). This condition is being tracked in the corrective action program (IR 4175355). The NRC Resident has been informed". Peach Bottom will be notifying State and local agencies regarding the event. Notified the R1DO (Greives).

On 12/08/16 at approximately 2237 (EST), the Unit 2 HPCI (High Pressure Coolant Injection) system failed to meet surveillance testing requirements for achieving rated flow at greater than or equal to a minimum test pressure established per the surveillance. Operations declared the HPCI system inoperable and entered Technical Specification 3.5.1 Condition C for HPCI being inoperable. Other standby systems (Reactor Core Isolation Cooling and low pressure emergency core cooling systems) are operable. HPCI is a single train system. Therefore, per NUREG-1022, this condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that could have prevented the fulfillment of the safety function of a system required to mitigate the consequences of a design event. This condition has been entered into the Corrective Action program (IR 3951006). Investigation of the exact failure condition is in progress so that repairs can be made. At the surveillance flow of 5,000 gpm, the system was approximately 80 psi below the required pressure of 1,278 psi. Technical Specification 3.5.1, Condition C, is a 14-day Limiting Condition of Operation. The NRC Resident Inspector will be notified.

• * * RETRACTION AT 1440 EST ON 01/19/17 FROM ELMER KAUFFMAN TO S. SANDIN * * *

The licensee provided the following information as the basis for retracting this report: This is a retraction of an event notification made on 12/09/16 at 0529 EST (EN #52419). This event was initially reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as a condition that, at the time of discovery, was believed to have prevented the fulfillment of the High Pressure Coolant Injection (HPCI) system safety function. On 12/08/16 at 2237 EST, the Unit 2 HPCI system was declared inoperable due to failing to meet surveillance testing requirements for achieving rated flow at greater than or equal to a minimum test pressure established per the surveillance. Prompt troubleshooting was performed and it was determined that an adjustment to the HPCI turbine governor control system was required. This adjustment was performed and HPCI was returned to an operable status on 12/09/16. Subsequent to this occurrence, Engineering has completed an evaluation that concluded that HPCI was capable of fulfilling its safety function and that the associated Technical Specification (TS) Surveillance Requirement (SR) 3.5.1.8 was met. The evaluation concluded that HPCI was degraded, but met the threshold for TS operability. The NRC Senior Resident has been informed of this retraction." Notified R1DO (Kennedy).

This 60-day report, as allowed by 10 CFR 50.73(a)(1), is being made pursuant to 10 CFR 50.73(a)(2)(iv)(A) to describe an unplanned, invalid actuation of containment isolation valves in more than one system. On 3/28/16, at approximately 1320 (EDT), a loss of power occurred on the Unit 2 E124 480 volt load center due to an equipment operator inadvertently opening the main feed breaker during the process of applying a clearance to de-energize the E124-P-A motor control center for planned maintenance. Loss of the E124 load center resulted in Group II and Group III primary containment isolations due to an invalid ESF actuation signal. Systems impacted by the containment isolations included containment instrument nitrogen, containment atmospheric monitoring, reactor water cleanup, and secondary containment. Balance of plant impacts included partial loss of feedwater heating and a reduced condenser vacuum. Reactor power lowered to 86% as a result of the event and further decreased to approximately 80 percent when re-establishing the 3A, 4A and 5A feedwater heaters.

Following direction from the control room, the E124 main feed breaker was promptly re-closed by equipment operators. Affected equipment was restored to its normal or planned configuration and containment isolations were reset at 1406. The containment isolation signal was generated as a result of the loss of power to the E124 load center and was not due to actual plant conditions or parameters meeting design criteria for containment isolation. Therefore, this is considered to be an invalid actuation. The licensee has notified the NRC Resident Inspector.

This 60-day report, as allowed by 10CFR 50.73(a)(1), is being made pursuant to 10CFR 50.73(a)(2)(iv)(A) to describe an unplanned, invalid actuation of an emergency service water system. On 3/9/14, at approximately 1536 EST, the E2 Emergency Diesel Generator (EDG) was started for a post maintenance operability test. The 'A' safety-related Emergency Service Water (ESW) pump was being used to provide cooling to the EDG. At approximately 1613 EST and 1622 EST, unexpected automatic starts of the 'B' ESW pump and the non-safety related single train Emergency Cooling Water (ECW) pump occurred. The EDG run was terminated at 1622 EST to investigate the cause of the unexpected starts.

It was determined that a failed angle drive adaptor resulted in invalid input to the speed switch, which resulted in the unplanned starts of the 'B' ESW pump and the ECW pump. Except for the failure of the angle drive adaptor, all systems and equipment operated as expected. The NRC resident has been informed of this notification.

On January 09, 2012 the Technical Support Center(TSC) ventilation was discovered to be non-functional during system testing. At 1722 hours (EST), it was identified that control power was lost to four motor operated dampers. Investigation revealed that a control power fuse was blown. In addition, the TSC ventilation Supply Fan (AHU-1) was identified to have one of two fan belts failed. The fuses were replaced at 19:45 (EST) and the system was restored to service. However, the system has been removed from service this evening to replace the fan belts. An update will be provided when the system is restored to full functionality. The licensee has notified the NRC Resident Inspector.

• * * UPDATE AT 1744 EST ON 1/10/12 FROM JEFF MOORE TO HUFFMAN * * *

Maintenance personnel installed two new belts on AHU-1 and aligned the motor/fan sheaves to ensure correct operation. TSC ventilation has been restored to the normal condition. The licensee will notify the NRC Resident Inspector. R1DO (Trapp) notified.