Semantic search

The following information was provided by the licensee via email: The following event description is based on information currently available. If through subsequent reviews of this event additional information is identified that is pertinent to this event or alters the information being provided at this time a follow-up notification will be made via the ENS or under the reporting requirements of 10 CFR 50.73. At 2144 MST on April 8, 2023, the Unit 1 reactor automatically tripped due to the loss of reactor coolant pumps stemming from the loss of 13.8 kV power to the pumps. Prior to the reactor trip, the main turbine tripped due to a loss of hydraulic pressure. The main generator output breakers did not automatically open on the turbine trip as expected so the control room operators opened the breakers per procedural guidance. Once the breakers were opened, the two 13.8 kV electrical distribution buses failed to complete a fast bus transfer, which resulted in the loss of power to the reactor coolant pumps, initiating the reactor trip. The control room operators manually actuated a main steam isolation signal per procedure, requiring use of the atmospheric dump valves. Following the reactor trip, all control element assemblies inserted fully into the core. No automatic specified system actuation was required or occurred. No emergency plan classification was required per the Emergency Plan. Safety related buses remained powered from offsite power during the event and the offsite power grid is stable. Unit 1 is stable and in Mode 3. Decay heat is being removed by the atmospheric dump valves and the class 1E powered motor driven auxiliary feedwater pump. The loss of hydraulic pressure, the main generator output breakers failing to automatically open and the fast bus transfer not actuating are being investigated. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B). The NRC Senior Resident Inspector has been informed. Unit 2 is in a refueling outage in Mode 5 and Unit 3 is in Mode 1 at 100 percent power.

• * * UPDATE ON 4/9/23 AT 0835 EDT FROM TANNER GOODMAN TO ADAM KOZIOL * * *

This update is being made to report the manual actuation of the B-train auxiliary feedwater pump and manual main steam isolation signal (MSIS) actuation affecting multiple main steam isolation valves (MSIVs) following the reactor trip. This event is being reported as a reactor protection system actuation in accordance with the reporting criteria of 10 CFR 50.72(b)(2)(iv)(B) and a specified system actuation in accordance with 10 CFR 50.72(b)(3)(iv)(A). The NRC Senior Resident Inspector has been informed of the update. Notified R4DO (Warnick)

• * * UPDATE ON 5/3/23 AT 1945 EDT FROM LORRAINE WEAVER TO JOHN RUSSELL * * *

This update is intended to clarify the initial description of the event that occurred on 4/8/2023. Prior to the reactor trip, the main turbine tripped due to a loss of hydraulic pressure. The main generator output breakers did not automatically open on the turbine trip. The control room operators manually opened the breakers per procedural guidance. Once the breakers were opened, the two 13.8 kV electrical distribution buses de-energized. A fast bus transfer did not occur per design, which resulted in the loss of power to the reactor coolant pumps, initiating the reactor trip. The control room operators manually actuated a main steam isolation signal per procedure, requiring use of the atmospheric dump valves. The NRC Senior Resident Inspector has been informed of the update. Notified R4DO (Gaddy)

The following information was provided by the licensee via email: At 0001 EDT on September 26, 2022, James A. FitzPatrick (JAF) removed the generator from service as part of a planned shutdown for refueling. At 0306 EDT, with the mode switch in Startup/Hot Standby and inserting rods, JAF experienced a spurious Scram and closure of seven out of eight main steam isolation valves (MSIV's). The reactor protection system (RPS) actuated during the event, resulting in all control rods being fully inserted. The cause of the closure of MSIV's and the Scram is being investigated. This condition is being reported as a four-hour NRC report per 10 CFR 50.72(b)(2)(iv)(B) for RPS actuation, and as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(iv)(A) for the safety system actuation based on the multiple main steam isolation valves closing on an isolation signal. There was no impact to the health and safety of the public. The NRC Resident Inspector has been notified.

• * * UPDATE ON 10/4/22 AT 2047 EDT FROM ANDREW WEAVER TO KERBY SCALES * * *

The following update was provided by the licensee via email: This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A) for Reactor Protection System (RPS) actuation along with Main Steam Isolation Valves (MSIV) system actuation. An analysis of reactor criticality was performed for the period of time prior to the RPS actuation event. Operators were inserting control rods per the shutdown Reactivity Management Plan. The Intermediate Radiation Monitoring (IRM) readings preceding the scram signal demonstrate a negative reactivity direction without control rod movement. The analysis concluded that the reactor was subcritical when RPS was actuated. The NRC Resident Inspector has been notified. Notified R1DO (Young).

At 2312 EDT, on August 3, 2020, Brunswick Unit 1 declared an Unusual Event due to a loss of offsite power. The unit was at approximately 20 percent power and was not synced to the grid when the unit automatically scrammed. All control rods fully inserted. Emergency Diesel Generators started and began powering the safety buses. Safety systems actuated as expected. The Unit also experienced a loss of Fuel Pool Cooling and Cleanup System, but one pump was returned to service. Unit 2 remains at 100 percent power and is unaffected. The licensee notified State and local governments, as well as the NRC Resident Inspector. Notified DHS SWO, FEMA Operations Center, CISA IOCC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).

• * * UPDATE FROM MARK TURKAL TO DONALD NORWOOD AT 0120 EDT ON 8/4/2020 * * *

At approximately 2302 EDT, a loss of offsite power occurred on Unit 1. This resulted in a Reactor Protection System (RPS) actuation. Per design, emergency diesel generators 1 and 2 properly started and loaded to their respective emergency buses. 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. As previously reported, an Unusual Event was declared at 2312 EDT due to the loss of offsite power. At the time of the event, Unit 1 was in the process of shutting down for maintenance associated with a ground on the main generator. Due to the RPS actuation while critical, this event is being reported as a four-hour, nonemergency notification per 10 CFR 50.72(b)(2)(iv)(B). As a result of the reactor trip, reactor water level reached low level 1 (LL1). The LL1 signal causes a Group 2 (i.e., floor and equipment drain isolation valves), Group 6 (i.e., monitoring and sample 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. Per design, the loss of offsite power also caused a Group 1 (i.e., main steam isolation valve) isolations. Due to the Emergency Diesel Generator and Primary Containment Isolation System (PCIS) actuations, this event is also being reported as an eight-hour, nonemergency notification in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The safety significance of the event is minimal. All safety related systems operated as designed. Investigation of the cause of the loss of offsite power is in progress. The licensee notified the NRC Resident Inspector. Notified R2DO (Inverso).

• * * UPDATE ON 8/4/2020 AT 1534 EDT FROM JOSEPH ELKINS TO ANDREW WAUGH * * *

At 1454 EDT on August 4, 2020, the Unusual Event was exited when offsite power was restored to Unit 1. Per design, when the loss of offsite power to Unit 1 occurred, all four emergency diesel generators (EDGs) started and EDGs 1 and 2 properly suppled emergency buses 1 and 2. Since Unit 2 was not affected by the loss of power, EDGs 3 and 4 ran unloaded. With restoration of offsite power to Unit 1, EDG 2 has been secured. EDGs 1, 3, and 4 are being secured as required by plant operating procedure. Notified R2DO (Inverso), NRR EO (Miller), IRD MOC (Grant), DHS SWO, FEMA OC, DHS NICC WO, CISA IOCC (email), DHS SWO (email), FEMA NWC (email), FEMA Ops Center (email), FEMA-NRCC-sasc (email), NRCC THD Desk (email), NuclearSSA (email). ********************************************************************************************************************************

On Friday, May 17, 2019 at 2303 (EDT), with the reactor at 70 (percent) core thermal power, Pilgrim Nuclear Power Station initiated a manual reactor scram due to degrading condenser vacuum as a result of the trip of Seawater Pump B. All control rods inserted as designed. The plant is in hot shutdown. Plant safety systems responded as designed. Pressure is being controlled using the Mechanical Hydraulic Control System and Main Condenser. Reactor water level is being maintained with the feedwater and condensate system. During the manual reactor scram, the plant experienced the following isolation signals as designed:

"Group 2 Isolation: Miscellaneous containment isolation valves
Group 6 Isolation: Reactor Water Clean-up
Reactor Building Isolation Actuation

Due to the Reactor Protection System actuation while critical, this event is being reported as a four-hour, non-emergency notification in accordance with 10 CFR 50.72(b)(2)(iv)(B), 'any event that results in actuation of the reactor protection system (RPS) when the reactor is critical...' This notification is also being made in accordance with 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) of this section...' (B)(2) 'General containment isolation signals affecting containment isolation valves in more than one system or multiple main steam isolation valves (MSIVs).' This event has no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. The licensee will notify the Massachusetts Emergency Management Agency.

EN Revision Text: POTENTIAL LOSS OF MSIV SCRAM FUNCTION DURING MAIN STEAM LINE ISOLATION VALVE TESTING At approximately 1040 EST on January 5, 2019, during evaluation of test results for the 'C' Main Steam Isolation Valve (MSIV), it was determined that closure of three of four Main Steam Lines would not necessarily have resulted in a full scram during testing due to failure of a limit switch (LS-6) associated with MSIV-1C while in the test configuration. This condition is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v), 'Any event or condition that 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 system was restored from the testing configuration at 1057 EST and the failed trip channel was placed in the tripped condition at 1326 EST thus restoring the design function. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.

• * * RETRACTION AT 1529 EST ON 02/11/19 FROM JOSEPH FRATTASIO TO JEFF HERRERA * * *

The purpose of the notification is to retract ENS Notification 53815 made on 01/05/19 for Pilgrim Nuclear Power Station. The previous notification reported that there was a potential loss of Main Steam Isolation Valve (MSIV) scram function during main steam line isolation valve testing, at the time of discovery, due to failure of a limit switch (LS-6) associated with MSIV-1C while in the test configuration. Subsequent evaluation has demonstrated that the scram function credited in the design basis was not lost. Specifically, after an Engineering Evaluation, it has been determined that the MSIV position RPS logic was not lost for those functions within the design basis and, as such, was capable of performing its intended safety function. The NRC Resident Inspector has been notified. Notified the R1DO (Cahill).

EN Revision Text: INOPERABILITY OF EQUIPMENT FOR CONTROL OF RADIOLOGICAL RELEASE At 2130 CST on 11/27/2018, Division 1 Main Steam Positive Leakage Control System (MS-PLCS) was declared inoperable because of a leaking check valve that caused excessive cycling of the associated air compressor. Division 2 MS-PLCS had been declared inoperable on 11/27/2018 at 1400 CST when a pressure control valve in the system exceeded the maximum allowable stroke time. Because MS-PLCS supplements the isolation function of the main steam isolation valves (MSIVs) by processing fission products that could leak through the closed MSIVs, both divisions of MS-PLCS inoperable at the same time represents a condition that could prevent the fulfillment of a safety function of an SSC (Structures, Systems and Components) that is needed to control the release of radioactive material. The station diesel air compressor is available to supply backup air to the safety relief valves as required by the Technical Requirements Manual." (This is associated with operability of the safety relief valves, due to the inoperable MS-PLCS air compressor.) The unit is in a 7 day shutdown Limiting Condition for Operation (LCO), 1-TS1-18-Div 1 & 2 MSPLCS-685, for the two divisions of MS-PLCS being inoperable. The licensee notified the NRC Resident Inspector.

• * * RETRACTION ON 12/03/18 AT 1551 EST FROM TIM GATES TO BETHANY CECERE * * *

This event was initially reported under 10 CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the Main Steam Positive Leakage Control System (MS-PLCS) from fulfilling its safety function to control the release of radioactive material. Division I was declared inoperable due to a failed component. Division II was declared inoperable due to a pressure control valve in the system exceeding the maximum allowable time to close by 0.50 seconds. An engineering evaluation has since been performed and concluded that the 2 second maximum allowable time to close was based on the pressure control valve being classified as a rapid closure valve and was established from the original baseline data of 0.50 seconds. This baseline data is an administrative target value per the In-Service Testing Program. There are no technical specification requirements associated with the 2 second closure time. The engineering evaluation also determined that the volume of air supplied through the pressure control valve during the extra 0.50 seconds of valve closure would have an inconsequential effect on the pressure within the volume of leakage barrier between the Main Steam Isolation Valves associated with the MS-PLCS pressure control valve or have any effect on containment over-pressurization. Based on the information provided by the engineering evaluation, the Division II MS-PLCS has been declared operable-degraded non-conforming since time of initial discovery. Consequently, this event is not reportable as a condition that could have prevented the Main Steam Positive Leakage Control System (MS-PLCS) from fulfilling its safety function. The (NRC) Resident Inspector has been notified via e-mail. Notified the R4DO (Gaddy).

On March 25, 2018 at 1833 CDT, while at 100 percent power, Farley Unit 1 (FNP-1) conservatively declared a single Main Steam Isolation Valve (MSIV) inoperable on the 1C Steam Generator line due to indication of Steam Generator pressure rise with a corresponding reduction in flow of that loop. FNP-1 began a reactor shutdown at 0400 CDT on March 26, 2018 to establish plant conditions to support testing the affected main steam line MSIVs while in the required action time of Technical Specification 3.7.2. At 1338 CDT on March 26, 2018, testing confirmed that the single MSIV was inoperable and that valve disassembly will be required. The duration of the valve repair would exceed the required action time of Technical Specification 3.7.2. This report is being made in accordance with 10 CFR 50.72(b)(2)(i), as a plant shutdown required by technical specifications. The licensee has notified the NRC Resident Inspector.

• * * UPDATE FROM DOUGLAS HOBSON TO KEN MOTT AT 0202 EDT ON 5/16/18 * * *

This EN (event notification) is being updated to clarify the reporting criteria as 'Voluntary'. Farley Technical Specification 3.7.2 allows continuous operation in MODE 2 with an INOPERABLE MSIV as long as the other MSIV in the affected Main Steam Line is closed. The initiation of the shutdown was performed as a prudent action to repair and restore OPERABILITY of the affected MSIV and was not a requirement of the Farley Technical Specifications. The licensee notified the NRC Resident Inspector. The R2DO (Masters) was notified.

During a rain and lightning storm, plant operators observed arcing from the main transformer bus duct and notified the control room. The decision was made to trip the main generator which resulted in an automatic reactor trip. The plant entered EAL SU.1 as a result of the loss of offsite power for greater than fifteen minutes. Plant safety busses are being supplied by both emergency diesel generators while the licensee inspects the electrical system to determine any damage prior to bringing offsite power back into the facility. Offsite power is available to the facility. No offsite assistance was requested by the licensee. During the trip, all rods inserted into the core. Decay heat is being removed via the atmospheric dump valves with emergency feedwater supplying the steam generators. The main steam isolation valves were manually closed to protect the main condenser. There were no safeties or relief valves that actuated during the plant transient. There is no known primary-to-secondary leakage. Reactor cooling is via natural circulation. All safety equipment is available for the safe shutdown of the plant. The licensee has notified the NRC Resident Inspector, Louisiana Department of Environmental Quality and the local Parish emergency management agencies. Notified DHS SWO, FEMA, DHS NICC, FEMA National Watch Center (email) and Nuclear SSA (email).

• * * UPDATE ON 7/17/17 AT 2007 EDT FROM MARIA ZAMBER TO DONG PARK * * *

This notification is also made under 10 CFR 50.72(b)(3)(v)(D). This is a non-emergency notification from Waterford 3. On July 17, 2017 at 1606 CDT, the reactor automatically tripped due to a loss of Forced Circulation, which was the result of Loss of Offsite Power (LOOP) to the electrical (safety and non-safety) buses. Both 'A' and 'B' trains of Emergency Diesel Generators (EDGs) started as designed to reenergize the 'A' and 'B' safety buses. The LOOP caused a loss of feedwater pumps, resulting in an automatic actuation of the Emergency Feedwater (EFW) system. Prior to the reactor trip, at 1600 CDT, personnel noticed the isophase bus duct to main transformer 'B' glowing orange due to an unknown reason. Due to this, the main turbine was manually tripped at 1606 CDT. Following the turbine trip, the electrical (safety and non-safety) buses did not transfer to the startup transformers as expected due to an unknown reason. The plant entered the Emergency Operating Procedure for LOOP/Loss of Forced Circulation Recovery. At 1617 CDT, an Unusual Event was declared due to Initiating Condition (IC) SU1 - Loss of all offsite AC power to safety buses (greater than) 15 minutes. All safety systems responded as expected. The plant is currently in mode 3 and stable with the EDGs supplying both safety buses and with EFW feeding and maintaining both steam generators. Offsite power is in the process of being restored. The licensee has notified the NRC Resident Inspector, Louisiana Department of Environmental Quality and the local Parish emergency management agencies.

• * * UPDATE FROM ADAM TAMPLAIN TO HOWIE CROUCH AT 2203 EDT ON 7/17/17 * * *

The licensee terminated the Notification of Unusual Event at 2056 CDT. The basis for terminating was that offsite power was restored to the safety busses. The licensee has notified Louisiana Department of Environmental Quality, St. John and St. Charles Parishes, Louisiana Homeland Security Emergency Preparedness, and will be notifying the NRC Resident Inspector. Notified IRD (Stapleton), NRR (King), R4DO (Hipschman), DHS SWO, FEMA, DHS NICC, FEMA National Watch Center (email) and Nuclear SSA (email).

• * * UPDATE FROM SCOTT MEIKLEJOHN TO HOWIE CROUCH AT 1724 EDT ON 7/19/17 * * *

This update is being reported under 10 CFR 50.72(b)(3)(v)(B). During the event discussed in EN# 52863, at 1642 CDT (on July 17, 2017), Condensate Storage Pool (CSP) level lowered to less than 92% resulting in entry to Technical Specification (TS) 3.7.1.3. Level in the CSP was lowered due to feeding from both Steam Generators with EFW. Normal makeup to the CSP was temporarily unavailable due to the LOOP. Filling the CSP commenced at 1815 CDT (on July 17, 2017), and TS 3.7.1.3 was exited on July 18, 2017 at 0039 CDT. The licensee notified the NRC Resident Inspector. Notified R4DO (Hipschman).

• * * UPDATE FROM SCOTT MEIKLEJOHN TO HOWIE CROUCH AT 1233 EDT ON 9/14/17 * * *

Waterford 3 is retracting a follow up notification made on July 19, 2017 for EN# 52863, concerning the loss of safety function associated with the Condensate Storage Pool (CSP) per 10 CFR 50.72(b)(3)(v)(B). The Condensate Storage Pool was performing its required safety function by providing inventory to the Emergency Feed Water pumps when the required Tech Spec level (T.S. 3.7.1.3) dropped below 92%. The Technical Specification was entered at 1624 (CDT) on July 17, 2017 and exited after filling at 0039 on July 18, 2017. The total allowed outage time allowed by Tech Spec 3.7.1.3 is 10 hours to be in Hot Shutdown if not restored. The Condensate Storage Pool level was restored to greater than 92% prior to exceeding the allowed outage time. Based on level being restored and the Condensate Storage Pool performing its required safety function, 10 CFR 50.72(b)(3)(v)(B) does not apply. Prior to the automatic reactor trip, Condensate Storage Pool level was greater than 92%. The NRC Resident Inspector has been notified of the retraction. Notified R4DO (Groom).