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The following information was provided by the licensee via fax or email: At 2207 (EST) on 2/21/2022 with Unit 2 in Mode 1 at 68 percent power, the reactor was manually tripped due to lowering water level in the 2A Steam Generator. The trip was not complex with all systems responding normally post-trip. Operators responded and stabilized the plant. Decay heat is being removed by discharging steam to the main condenser using the turbine bypass valves. Units 1 and 3 were not affected. 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."

• * * UPDATE ON 3/23/22 AT 1643 EDT FROM CHRIS MCDUFFIE TO TOM KENDZIA * * *

The following information was provided by the licensee via phone and email: On 2/21/2022, Unit 2 was in Mode 1 increasing reactor power following startup from a forced outage. At 2205 (EST) with Unit 2 at 68 percent power, a feedwater control valve failed to properly control feedwater flow to the 2A Steam Generator and the Integrated Control System initiated an automatic runback. At 2207 (EST), the reactor was manually tripped from 39 percent power due to lowering water level in the 2A Steam Generator. Immediately following the manual reactor trip, an actuation of the Emergency Feedwater System (EFW) occurred. The 2A and 2B Motor Driven Emergency Feedwater (MDEFW) pumps automatically started as designed when the 'low steam generator level' signal was received for the 2A Steam Generator. The trip was not complex with all systems responding normally post-trip. Operators responded and stabilized the plant. Decay heat was removed by discharging steam to the main condenser using the turbine bypass valves. Units 1 and 3 were not affected. Unit 2 was restarted on 2/27/2022 following repairs. Due to the Reactor Protection System actuation while critical, this event was reported on 2/22/2022 as a four-hour, non-emergency notification per 10 CFR 50.72(b)(2)(iv)(B). Following further evaluation, it was determined that a valid EFW actuation occurred, therefore this event is now also being reported as a late 8-hour non-emergency notification of a valid actuation of the EFW system in accordance with 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. Notified the R2DO (Miller).

The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At 0357 EST on 2/5/22, Oconee Unit 2 declared an Unusual Event due to a multiple fire alarms and visual observance of a smoke filled room in the West Penetration Room (EAL HU 4.1). Unit 2 automatically tripped and entered Mode 3. There was an indication that a release to the environment potentially occurred. Units 1 and 3 remained at 100% power. The Licensee notified the NRC Resident Inspector, the State, and local authorities. Notified DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), and DHS Nuclear SSA (email).

• * * UPDATE ON 2/5/22 AT 0748 EST FROM CHUCK CLEMONS TO LLOYD DESOTELL * * *

The following information was provided by the licensee via email: At 0343 EST on February 5th, 2022, with Unit 2 in Mode 1 at 100% power, the reactor automatically tripped due to Reactor Coolant Pump (RCP) Flux/Flow Imbalance caused by the simultaneous trip of all 4 RCPs. The cause of the loss of all RCPs is under investigation. The trip was not complex. There was no inoperable equipment prior to the event that contributed to it. Operations responded and stabilized the plant. Decay heat is being removed by the Reactor Coolant System (RCS) in natural circulation with normal feedwater flow. Due to the Reactor Protection System (RPS) actuation while critical, this event is being reported as a four hour, nonemergency 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: The Licensee notified the State and local authorities. Notified R2DO (Miller)

• * * UPDATE ON 2/5/22 AT 0815 EST FROM CHUCK CLEMONS TO LLOYD DESOTELL * * *

The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: At 0811 EST on 2/5/22, Oconee Unit 2 terminated the notification of unusual event because EAL HU 4.1 criteria were no longer met. Unit 2 remains shutdown in Mode 3. The licensee has notified the State and local authorities and will notify the NRC Resident Inspector. Notified R2DO (Miller), IRD MOC (Gott), NRR EO (Regan), DHS SWO, FEMA Operations Center, CISA Central, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

• * * UPDATE ON 2/7/22 AT 2342 EST FROM PATRICK GADSBY TO OSSY FONT * * *

The following information was provided by the licensee via email: The follow-up investigation determine that the cause of the trip was determined to be a failed fuse in an undervoltage monitoring circuit for the electrical bus powering the reactor coolant pumps. The trip of the reactor coolant pumps resulted in an automatic reactor trip as designed. The plant responded as expected. There was no fire. The response of the main feedwater system caused an expected increase in temperature in a standby portion of the system which caused the breakdown of the pipe coating producing smoke. There was no release to the environment. The reading of a single instrument was determined to be invalid and other indications supported the conclusion that no release occurred. The following information was obtained from the licensee in accordance with Headquarters Operations Officers Report Guidance: The Licensee notified the NRC Resident Inspector, State and local authorities. A media release was issued on 2/5/22. Notified R2DO (Miller), IRD MOC (Gott), NRR EO (Regan).

At 2307 EDT on April 21, 2019, in Mode 1 at approximately 100 percent reactor power, Unit 1 automatically tripped due to a Main Turbine Trip. The Main Turbine Trip was a result of two out of three level instruments sensing a false high reactor water level. All control rods inserted as expected during the scram. Safety Relief Valves G and K lifted per design. The same level instruments that failed also tripped both Reactor Feed Pumps. As a result, reactor water level dropped below the Low Level 1 and 2 actuation setpoints. Per design, the Low Level 1 signal resulted in Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The Low Level 2 signals resulted in Group 3 (i.e. Reactor Water Cleanup) isolation, a secondary containment isolation signal, and an auto start of Standby Gas Treatment and Control Room Emergency Ventilation. Also, the Low Level 2 resulted in (high pressure coolant injection) HPCI and (reactor core isolation cooling system) RCIC automatically starting and injecting into the vessel. All systems responded as designed. This event is being reported in accordance with 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation and 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Decay heat is currently being removed via the turbine bypass valves. Condensate and feed water are maintaining water level. The reactor is still at saturation temperature and 475 psi, lowering slowly. The reactor is still in a normal electrical lineup. There was no impact to Unit 2 as a result of this event.

• * * UPDATE ON 04/22/19 AT 0220 EDT FROM ALAN SCHULTZ TO JEFFREY WHITED * * *

The licensee updated the event report to include a 4-Hr Non-Emergency Notification in accordance with 10 CFR 50.72(b)(2)(iv)(A) for Emergency Core Cooling System, HPCI, Discharge to the Reactor Coolant System. Notified R2DO (Dickson), NRR EO (Miller) and IR MOC (Gott).

On April 7, 2018, at 0836 EDT, with Unit 1 in Mode 1 at approximately 100 percent power, the reactor automatically tripped during testing of the stator cooling system. The trip was uncomplicated with all systems responding normally. No safety-related equipment was inoperable at the time of the event. 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).

Operations responded using Emergency Operating Procedures and stabilized the plant in Mode 3. Reactor water level being maintained via normal feedwater system. Decay heat is being removed through the bypass valves.

Reactor water level reached low level 1 (LL1) as a result of the reactor trip. The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sampling isolation valves) and Group 8 (i.e., shutdown cooling isolation valves) isolations. The LL1 isolations occurred as designed; the Group 8 valves were closed at the time of the event. Due to the Primary Containment Isolation System (PCIS) actuation, this event is also being reported as an eight-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the PCIS. Unit 2 was not affected. There was no impact on the health and safety of the public or plant personnel. The safety significance of this event is minimal. The automatic reactor trip was not complicated and all safety-related systems operated as designed. Investigation of the cause of the Reactor Protection System actuation is in progress. The licensee notified the NRC Resident Inspector.

At 1346 EST the licensee reported that at 1326, Brunswick Unit 1 declared an Alert under EAL HA 2.1 due to an explosion/fire in the Balance of Plant 4 kV switchgear bus area. Prior to the Alert declaration, the operators initiated a manual SCRAM due to an unexpected power decrease from 88% to 40%. The licensee has visually verified that there is no ongoing fire and is investigating the initial cause of the event. Offsite power is available to the site, but EDGs 1 and 2 are running and supplying Unit 1 loads. The MSIVs shut and HPCI/RCIC are being used to maintain vessel level. At 1412 EST, NRC decided to remain in Normal Mode. At 1704 EST the licensee reported the following: At 1313 hours Eastern Standard Time (EST) a manual reactor scram was initiated due to loss of both recirculation system variable speed drives as a result of an electrical fault. At this time, a Startup Auxiliary Transformer (SAT) experienced a lockout fault; interrupting offsite power to emergency buses 1 and 2. Emergency Diesel Generators (EDGs) 1, 2, 3, and 4 automatically started and EDGs 1 and 2 synchronized to emergency buses 1 and 2 per design. The power interruption resulted in closure of the Main Steam Isolation Valves, per design. The manual scram also resulted in closure of Group 2, 6, and 6 Containment Isolation Valves. The Reactor Core Isolation Cooling (RCIC) system was manually started and is being used to control reactor water level. The High Pressure Coolant Injection (HPCI) system was manually started and is being used for pressure control. The Plant response to the event was per design. Unit 2 is not directly affected by the event, however, due to the shared electrical distribution system is in a Technical Specification Action statement due to the Inoperable Unit 1 SAT. The public health and safety is not impacted by this event. At 1751 EST, the licensee reported that the emergency declaration had been downgraded to an Unusual Event at 1730 because the plant no longer meets the criteria for an Alert, but does meet the criteria for an Unusual Event due to a "loss of all offsite power to Emergency 4 kV buses E1 (E3) and E2 (E4) for greater than or equal to 15 minutes." The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).

• * * UPDATE FROM MARTY IRWIN TO DANIEL MILLS AT 1825 ON 2/07/16 * * *

At 1814 EST the emergency declaration was terminated because offsite power was restored. The NRC Resident Inspector has been notified. The licensee has notified the State and Local governments. Notified R2DO (Musser), NRR EO (Morris), IRD MOC (Stapleton), R2RA (Haney), NRR ET (Lubinski), NRR ET (Dean), DHS, FEMA, USDA, HHS, DOE, DHS NICC, EPA EOC, FEMA NWC (via email), FDA EOC (via email) and Nuclear SSA (via email).

(At 1016 EST, an) Alert (was declared) based on EAL # HA 2.1 Fire or explosion resulting in either: visible damage to any table H-1 structure or system/component required for safe shutdown of the plant, or control room indication of degraded performance of any safe shutdown structure, system, or component within any table H-1 area. Fire in 480V bus 1D2. Reactor was manually tripped 480 VAC safety related transformer fire in switchgear room. Plant reduced power and tripped the reactor manually. Reactor trip was uncomplicated. Fire was extinguished when the 480 VAC bus was de-energized. The licensee has notified the NRC Resident Inspector, the State of North Carolina, and other local authorities. Notified DHS SWO, DOE Ops Center, FEMA Operations Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and NuclearSSA via e-mail.

• * * UPDATE FROM JOEL DUHON TO JOHN SHOEMAKER AT 1602 EST ON 1/18/14 * * *

Harris Nuclear Plant secured from the Alert at 1551 EST, on 1/18/14. The plant is stable, the fire is out, the TSC and EOF have been secured and plant recovery has been transferred to the outage control center. There were no personnel injuries or radiological releases. Radiation monitor RM-*1TS-3653C (Technical Support Center Radiation Monitor) is out of service. The licensee has notified the NRC Resident Inspector. Notified the R2DO (King), R2RA (McCree), NRR (Leeds), IRD MOC (Grant), OPA (Brenner), NRR EO (Lee) Notified DHS SWO, DOE Ops Center, FEMA Operations Center, HHS Ops Center, NICC Watch Officer, USDA Ops Center, EPA EOC, and NuclearSSA via e-mail.

On November 14, 2013, at 1313 Eastern Standard Time, Unit 1 was manually tripped from 100% power due to indications of (four) dropped control rods. This manual reactor protection system actuation is reportable per 10 CFR 50.72(b)(2)(iv)(B). The cause of the dropped rods is not confirmed at this time, but may be related to maintenance in a Rod Control Power Cabinet ongoing at the time of the event. The Operations crew entered the reactor trip procedure and stabilized Unit 1 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip and all plant systems operated as designed. The Auxiliary Feedwater (AFW) system (1A and 1B motor-driven pumps) was manually started for steam generator level control following reactor trip. The start of the AFW system is reportable per 10 CFR 50.72 (b)(3)(iv)(A) for a valid system actuation. Decay heat is being removed via the steam generators (via steam dumps to the main condenser). This event does not impact public health and safety. Unit 2 was not affected by this event. The licensee notified the NRC Resident Inspector.

• * * UPDATE FROM WARREN MOORE TO DANIEL MILLS ON 11/18/13 AT 0950 EST * * *

A subsequent licensee evaluation determined that there were ten dropped control rods. 

Notified the R1DO (Desai).

At 0553 EDT on 10/24/2013, Oconee Unit 3 was manually tripped due to oscillations in the feedwater system in anticipation of an automatic reactor trip. At 0549 EDT, Unit 3 began experiencing small feedwater oscillations. The feedwater control system was placed in manual in an attempt to stabilize feedwater flows. Feedwater oscillations continued to grow in magnitude and at 0553 EDT, a manual trip was directed to prevent an automatic reactor trip. Due to an RPS actuation, this event is being reported as a 4 and 8 hour Non-Emergency per 10 CFR 50.72 (b)(2)(iv)(B) and 10 CFR 50.72 (b)(3) Following the reactor trip, four main steam relief valves failed to reseat. Procedure guidance was utilized to reduce main steam system pressure by approximately 30 psig to reseat the main steam relief valves. All main stream relief valves are now reseated. All other post trip conditions were normal and all other systems performed as expected. Unit 3 is currently in Mode 3 and stable. Operations have been stabilized on Unit 3. A post-trip investigation is in progress, per site procedures and directives. The licensee has notified the NRC Resident Inspector.

• * * UPDATE FROM BOB MEIXELL TO DONALD NORWOOD AT 1439 ON 9/10/14 * * *

Duke Energy reviewed NRC Event Number 49471 against NUREG 1022, Rev 3, section 3.2.6, "System Actuation" and determined this event should have been reported only per 10 CFR 50.72 (b)(2)(iv)(B), RPS Actuation (while critical). Thus, Duke Energy is revising NRC Event Number 49471 to remove the 8-hour report criteria 10 CFR 50.72 (b)(2)(iv)(A). The NRC Resident Inspector was notified of this revised report. This update has no effect on safety significance. Notified R2DO (Shaeffer).

On 11/16/11 at 0208 EST, Brunswick Nuclear Plant, Unit 2 calculated a drywall floor drain 42 minute leak rate of 5.88 gpm, following several hours of gradually rising floor drain leakage during a plant startup. Tech Spec 3.4.4 A was entered, requiring floor drain leakage to be restored below 5 gpm within 8 hours. At 0253 EST, a 45 minute leak rate of 10.11 gpm was calculated. At 0301 EST, Unusual Event SU 6.1 was declared for unidentified leakage exceeding 10 gpm, and at 0309 EST, a manual reactor scram was inserted from approximately 7% power (10 CFR 50.72(b)(2)(iv)(B)). Following the scram, the reactor was depressurized at a maximum cooldown rate of 92.5 deg F/hr, and the unidentified leak rate fell less than 10 gpm within 1 hour and less than 5 gpm within 2 hours. Leak rate at 0614 EST on 11/16/11 is 3.82 gpm with reactor pressure at 228 psig. The exact nature of the leak is unknown at this time. The current plan is to continue to depressurize and cool down the reactor to Mode 4, such that a full drywall inspection can commence. At present, Brunswick has not terminated the Unusual Event. Level control is currently being maintained with control rod drives (CRD). The MSIVs were manually closed to control cooldown. The maximum cooldown was observed to be 92.5 F/hour. The plant plans to reopen MISIVs and depressurized to condensate booster pump injection pressure of 350 psig. The plan is to achieve Mode 4 for a leak inspection. The licensee has notified the NRC Resident Inspector.

• * * UPDATE FROM DAVID FASHCHER TO CHARLES TEAL AT 0550 EST ON 11/16/11 * * *

The leakage rate is currently 3.73 gpm. The decrease is due to lower pressure which is currently at 258 psig. There are no additional changes. The leakage source is not identified at this time. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

• * * UPDATE FROM DUSTIN LUPTON TO JOHN SHOEMAKER AT 0648 EST ON 11/16/11 * * *

The leakage rate is currently below the T.S. limit due to lower pressure which is currently at 210 psig. There are no additional changes. The plant will remain in an Unusual Event (UE) until further notice. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

• * * UPDATE FROM DUSTIN LUPTON TO CHARLES TEAL AT 0749 EST ON 11/16/11 * * *

The leakage rate is stable. The leak rate is calculated at 3.04 gpm at 183 psig at 0708 EST. The current reactor pressure is 162 psig. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).

• * * UPDATE FROM DUSTIN LUPTON TO JOHN KNOKE AT 0832 EST ON 11/16/11 * * *

The licensee terminated from their Unusual Event at 0815 EST. The leakage is still unidentified. The licensee has notified the NRC Resident Inspector. Notified R2DO (Vias).