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 Entered dateSiteRegionScramReactor typeEvent description
ENS 5401622 April 2019 01:51:00BrunswickNRC Region 2Automatic Scram

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).

ENS 5396630 March 2019 21:06:00BrunswickNRC Region 2Manual ScramAt 17:47 Eastern Daylight Time (EDT) on March 30, 2019, with Unit 2 in Mode 1 at approximately 23 percent reactor power and main turbine startup in progress coming out of a refuel outage, a high temperature was sensed at main turbine bearing #9. As a result of and to arrest the high temperature condition, the main control room inserted a manual reactor scram. All control rods inserted as expected during the scram. When the scram was inserted, reactor water level dropped below the Low Level 1 actuation setpoint. 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 main control room manually closed all Main Steam Isolation Valves (MSIVs), in anticipation of a low vacuum prior to the Group 1 automatic closure signal being received. High Pressure Coolant Injection (HPCI) was aligned for pressure control and Reactor Coolant Isolation System (RCIC) was aligned for level control. The Reactor Coolant Sample Line Isolation valves closed as expected on low main condenser vacuum. 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. At the time of notification, decay heat was being removed by the condenser through one open MSIV and a feedwater pump running.
ENS 533197 April 2018 12:10:00BrunswickNRC Region 2Automatic ScramGE-4

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.

ENS 517157 February 2016 13:46:00BrunswickNRC Region 2Manual ScramGE-4

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).

ENS 4769023 February 2012 02:55:00BrunswickNRC Region 2Manual ScramGE-4

At 2319 hours EST, a manual Reactor Protection System (RPS) actuation was inserted on Unit 1 in anticipation of a loss of condenser vacuum. Shortly before the manual RPS actuation, Circulating Water Intake Pump (CWIP) 1B tripped due to high delta-pressure across the intake traveling screen. This caused the trip of the remaining pumps. Previously, at 1859 hours, balance of plant (BOP) bus Common C unexpectedly de-energized. This caused loss of power to the CWIP traveling screen motors which, in turn, lead to the high delta-pressure across the traveling screen(s). All control rods inserted properly. As a result of the scram, reactor water level reached the Low Level 1 actuation set point and Primary Containment (i.e., Group 6) isolation occurred. All systems functioned as designed. The High Pressure Coolant Injection (HPCI) system is being used, as needed, for pressure control. The Reactor Core Isolation Cooling (RCIC) system is being used, as needed, for level control. No Safety/Relief Valves (SRVs) actuated as a result of the manual RPS actuation. The manual RPS actuation is reportable in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A). The actuation of the HPCI and RCIC systems and the Group 6 isolation are reportable in accordance with 10CFR50.72(b)(3)(iv)(A). The unit is currently in Mode 3 with a cooldown in progress. The licensee notified the NRC Resident Inspector. Notified R2DO (Ernstes).

  • * * UPDATE FROM STEWART BYRD TO CHARLES TEAL AT 0741 EST ON 2/23/12 * * *

At 2319 hours EST, a loss of all Circulating Water Intake Pumps caused a lowering vacuum on Unit 1. As previously reported (i.e. Event Notification 47690), a manual Reactor Protection System (RPS) actuation was inserted on Unit 1 at this time. In addition, a valid actuation of the RPS, High Pressure Coolant Injection (HPCI), Reactor Core Isolation Cooling (RCIC), and a Group 6 isolation was reported in accordance with 10CFR50.72(b)(3)(iv)(A). At 2342, Main Condenser vacuum was 15 in. Hg and lowering. All Main Steam Isolation Valves were slow closed in anticipation of Group 1 isolation at this time. This follow-up notification is being made to report the manual actuation of the Group 1 isolation valves in accordance with 10 CFR 50.72(b)(3)(iv)(A). The Group 1 isolation was discussed with the NRC during initial notification of EN 47690, and this follow-up is providing written notification of the MSIV closure. The NRC Resident Inspector has been notified. Notified R2DO (Ernstes).

ENS 4744416 November 2011 03:33:00BrunswickNRC Region 2Manual ScramGE-4

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).

ENS 459025 May 2010 15:11:00BrunswickNRC Region 2Automatic ScramGE-4On May 5, 2010, at 1144 hours Eastern Daylight Time (EDT), an automatic reactor scram occurred on Unit 1 following a trip of the 1B Reactor Feed Pump (RFP). Following the 1B RFP trip, the reactor recirculation pumps did not run back as expected. The resulting water level shrink caused level in the Reactor Pressure Vessel (RPV) to drop to Low Level 1, causing the activation of the Reactor Protection System (RPS) and the Primary Containment Isolation System (PCIS). All control rods properly inserted. PCIS Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 8 (i.e., RHR Shutdown Cooling) isolation signals were received on Low Level 1. Actuations of the Primary Containment Isolation Valves (PCIVs) were completed and the affected equipment responded as designed. Due to the expected RPV level reduction following a reactor scram, water level in the RPV momentarily reached Low Level 2. This initiated the High Pressure Coolant Injection (HPCI) System, the Reactor Core Isolation Cooling (RCIC) System, and a partial Group 3 PCIS (i.e., RWCU) isolation. The HPCI and RCIC systems did not inject. The 1-G31-F001 isolated (i.e., inboard isolation) but 1-G31-F004 (i.e., outboard isolation) did not automatically isolate. Based on a preliminary assessment, this response appears to be in accordance with plant design. Further assessments of plant response are on-going to validate plant response. The licensee has notified the NRC Resident Inspector. The scram was uncomplicated. No SRVs lifted. Decay heat removal is via the 'A' feed water pump via the turbine bypass valves to the condenser. The electrical line-up of Unit 1 is normal. Brunswick Unit 2 was not affected.
ENS 4468526 November 2008 15:12:00BrunswickNRC Region 2Automatic ScramGE-4At 12:00 hours EST, an apparent Electro-Hydraulic Control (EHC) system malfunction while synchronizing the Main Generator to the grid resulted in a Group 1 Main Steam Isolation Valve (MSIVs) closure on low reactor pressure and a subsequent automatic reactor scram. Preliminary investigation of the automatic scram signal indicates that Main Steam Line Low Pressure Instruments (B21-PT-N015 A thru D) sensed low steam line pressure after the Main Generator was paralleled to the grid. This resulted in the closure of all MSIVs. Closure of MSIVs in Mode 1 results in an automatic reactor scram. All control rods fully inserted. With the exception of the EHC system, all systems responded as designed. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to the Reactor Protection System (RPS) actuation and 10CFR 50,72(b)(3)(iv)(A) due to the Primary Containment Isolation System (PCIS) Groups 1, 2, and 6 actuations. Unit 2 was not affected by this event and remains at 100% power. Reactor Pressure is 844 psig, reactor temperature is 526 degrees Fahrenheit. MSIV's remain shut. Decay heat was removed via MSIV bypass lines. RCIC actuated for a period of time for level control but has been secured. CRD pumps are providing makeup water. No SRV's or relief valves lifted. The NRC Resident Inspector has been notified.
ENS 446479 November 2008 12:08:00BrunswickNRC Region 2Manual ScramGE-4One hour reportable event based on a safety relief valve (SRV) failure to close (NUREG 0626 and NUREG 0660). On 11/09/08 at approximately 11:08 with Unit 2 at 100% steady state power SRV 2-B21-F013H spuriously failed full open with no operator action or testing in progress. The valve's control switch was cycled as required by Abnormal Operating Procedure AOP-30 with no success. At 1113 the valve was successfully closed by pulling the associated fuses. At 1117, a manual reactor scram was inserted based on a Torus temperature of 109.8 degrees F (Technical Specifications require a scram to be inserted at 110 degrees F). All control rods (fully) inserted from the manual scram signal. Reactor water level lowered to Low Level 2 resulting in Primary Containment Isolation System (PCIS) isolations of Groups 2, 3, 6, and 8. In addition, this resulted in a Reactor Core Isolation Cooling (RCIC) system actuation and injection into the reactor. The High Pressure Coolant Injection (HPCI) system actuated but did not inject because reactor water level recovered. An Alternate Rod Insertion signal was received, the Standby Gas Treatment (SBGT) system initiated, and the Reactor Recirculation Pumps tripped as designed. Plant safety systems responded as designed to the transient. (Licensee) investigations are underway to determine the cause of the SRV failure. Reactor decay heat is being removed through the main turbine bypass valves to the condenser. Reactor make-up is being maintained by the normal feedwater system. The plant is in its normal shutdown electrical lineup supplied by offsite power. The diesel generators are available for service to the plant The Licensee notified the NRC Resident Inspector.
ENS 4445330 August 2008 17:47:00BrunswickNRC Region 2Manual Scram
Automatic Scram
GE-4At 1503 hours EDT, an Electro-Hydraulic Control (EHC) system malfunction caused the Unit 2 Main Turbine bypass valves (BPV) to start cycling. Initially, BPV 1 partially opened and closed followed shortly thereafter by four BPVs going full open. At that time the order was given to insert a manual scram. An automatic scram signal occurred just as the operator was beginning to insert the manual scram. Preliminary investigation of the automatic scram signal indicates that it was initiated by low Relay Emergency Trip Supply (RETS) pressure to the main turbine control valves due to the EHC malfunction. Reactor water level momentarily dropped below Low Level during the response. This resulted in Primary Containment Isolation System (PCIS) Group 2 and Group 6 isolations, as expected. All control rods fully inserted. All systems responded as designed. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) due to the Reactor Protection System actuation and 10 CFR 50.72(b)(3)(iv)(A) due to the PCIS Group 2 and Group 6 actuations. Unit 1 was not affected by this event and remains at 100% power. The NRC resident inspector has been notified.
ENS 4306225 December 2006 09:27:00BrunswickNRC Region 2Automatic ScramGE-4On 12/25/06 at approximately 05:39 an automatic reactor scram occurred on Brunswick Unit 2. The Reactor Protection System (RPS) actuated on Neutron Monitoring System (APRM/OPRM) trip for APRM 2 and 4. All control rods properly inserted when the scram occurred from the RPS signal. Reactor water level reached low level 1 (LL1) and low level 2 (LL2) as a result of the scram. The LL1 signal causes a Group 2 (floor and equipment drain isolation valves), Group 6 (monitoring and sampling isolation valves) and Group 8 (shutdown cooling isolation valves) isolation signal. The LL1 isolations occurred as designed. The LL2 (signal) causes a Reactor Core Isolation Cooling (RCIC) system actuation, High Pressure Coolant Injection (HPCI) system actuation, Group 3 (reactor water cleanup valves) isolation signal, a secondary containment isolation signal, a Standby Gas Treatment (SBGT) initiation signal, a Control Room Emergency Ventilation (CREV) initiation signal, Reactor Recirculation Pump trip and an Alternate Rod Insertion (ARI) actuation signal. The low level 2 condition was reached momentarily and did not affect all instruments due to calibration differences. Initial assessment concludes that the appropriate LL2 isolations and actuations occurred as designed. Further evaluation of LL2 isolation and actuations will be conducted. The RCIC system actuation resulted in injection into the reactor as designed. The HPCI system actuated but did not inject because reactor water level was recovered. The plant is in a stable condition. An investigation is in progress to determine the cause of the Neutron Monitoring System trip. RCIC started momentarily and then was secured. Reactor water level being maintained via normal feedwater system. Decay heat being removed through the bypass valves. Normal electrical lineup for shutdown. EDGs available. Unit 1 not affected by this transient. The licensee notified the NRC Resident Inspector.
ENS 4298611 November 2006 16:31:00BrunswickNRC Region 2Manual ScramGE-4At 1243 EST, during startup activities, a manual reactor scram was inserted as a result of high conductivity in the condenser. It is believed that the high conductivity was the result of a condenser tube leak. Upon receipt of the conductivity excursion alarm, abnormal operating procedures were consulted and the manual scram was inserted. Unit 2 was at approximately 1 percent of rated thermal power and reactor pressure was approximately 100 psi. At the time of the conductivity excursion, the condensate system was not in service and, as such, reactor water chemistry was not adversely affected. All safety systems operated per design. No emergency core cooling systems (ECCS) actuated. Unit 2 will be taken to mode 4 and the necessary repairs will be completed. All control rods inserted as expected. The licensee believes there is no spread of high conductivity to adjacent systems (e.g. CRD and the CST). Confirmatory samples are in progress. Decay heat is being removed by RCIC in the pressure control mode with the intention of placing shutdown cooling in-service. The electrical system is in a normal shutdown lineup. The licensee notified the NRC Resident Inspector.
ENS 429602 November 2006 22:13:00BrunswickNRC Region 2Manual ScramGE-4On 11/01/06 at approximately 18:23 (EST) Brunswick Unit 2 experienced a loss of the unit's Startup Auxiliary Transformer (SAT) and a loss of reactor forced circulation. Due to the loss of the SAT and subsequent manual reactor scram of Unit 2, a loss of Offsite Power resulted to the Unit 2 power buses. All four site Emergency Diesel Generators (EDGs) then started as designed. On 11/02/06 at approximately 0400 (EST) EDG no. 1 tripped on low lube oil pressure due to high differential pressure on the EDG lube oil strainer. The EDG was not loaded at the time of the trip. Due to the loss of one offsite qualified circuit (Unit 2 SAT) and the loss of one EDG (EDG 1), Unit 1 entered Technical Specification (TS) 3.8.1, Condition F, which requires restoration of the inoperable offsite circuit or restoration of the inoperable Diesel Generator within 12 hours. At 1600 on 11/02/06, Brunswick Unit 1 entered TS 3.8.1, Condition H, Required Action H.1 to be in MODE 3 in 12 hours and Required Action H.2 to be in MODE 4 within 36 hours. On 11/02/06 at 17:54 EDG 2 was declared inoperable due to being placed in manual for a required loaded run due to having been operated at no load for a period of time. Unit 1 entered Technical Specification 3.0.3 due to having EDG 1, EDG 2, and one offsite qualified circuit (Unit 2 SAT) inoperable. Per Technical Specification 3.0.3, action shall be initiated within one hour to place the unit in Mode 2 within 7 hours, Mode 3 within 13 hours, and Mode 4 within 37 hours. Unit 1 began a Technical Specification Required shutdown at 18:53 and is presently at 99% power. On 11/02/06 at 21:30, EDG 2 was declared operable following the loaded run and is presently in standby in the auto mode. Technical Specification 3.0.3 was exited at that time. Unit 1 remains in Technical Specification 3.8.1 Condition H at the present time. At 21:59, the NRC resident was notified of this event.
ENS 429551 November 2006 19:10:00BrunswickNRC Region 2Manual ScramGE-4

At 1823 EST, Unit 2 was manually scrammed due to a loss of offsite power from the Startup Auxiliary Transformer to both 4KV Emergency (E) buses. Both Emergency Diesel Generators (EDGs) 3&4 autostarted and re-energized the affected electrical buses. At 1823 EST, an Unusual Event was declared based on EAL 06.01.01, "Inability to power either 4KV E bus from offsite power." Unit 2 is currently stable in mode 3, Hot Shutdown, with MSIVs closed and HPCI controlling pressure and RPV Water Level. All control rods fully inserted following the manual reactor scram. The licensee determined that no emergency facilities will be activated and that no offsite assistance is needed at this time. The licensee informed both state and local agencies and will inform the NRC Resident Inspector.

  • * * UPDATE PROVIDED BY JOEL LEVINER TO JEFF ROTTON AT 2214 EST ON 11/01/06 * * *

On 11/01/06 at approximately 18:23 (EST) Brunswick Unit 2 experienced a loss of the unit's Startup Auxiliary Transformer and a loss of reactor forced circulation. A manual reactor scram was performed as required by station Abnormal Operating Procedures. Due to the loss of the Startup Auxiliary Transformer and subsequent manual reactor scram, a loss of Offsite Power resulted to the unit's power buses when unit shutdown was completed. All control rods properly inserted when the manual reactor scram was performed. All four site emergency diesels started and diesels 3 and 4 are supplying the Unit 2 emergency buses. Reactor water level reached low level 1 (LL1) and low level 2 (LL2) as result of the reactor scram and loss of offsite power. The LL1 signal resulted in Group 2 (floor and equipment drain isolation valves), Group 6 (monitoring and sampling isolation valves) and Group 8 (shutdown cooling isolation valves) isolation signals. All low level 1 isolations occurred as designed. The LL2 resulted in a Reactor Core Isolation Cooling (RCIC) system actuation, High Pressure Coolant Injection (HPCI) system actuation, Group 3 (reactor water cleanup valves) isolation signal, a secondary containment isolation signal, a Standby Gas Treatment (SBGT) initiation signal, a Control Room Emergency Ventilation (CREV) initiation signal, and an Alternate Rod Insertion (ARI) actuation signal. All isolation and actuations occurred as designed with the exception the CREV initiation and ARI actuation. CREV initiation and ARI actuations were performed by manual actions. The failure of the CREV and ARI initiation/actuations are under investigation. The RCIC and HPCI systems were used to restore reactor water level to the normal operation band. Reactor vessel pressure is being controlled in the normal band with manual operation of Safety Relief Valves (SRV), and HPCI/RCIC in pressure control mode. The Main Steam Isolation Valves (MSIVs) (Group 1) and the drywell pneumatic isolation valves (Group 10) closed on the loss of power. The plant is a stable condition. Troubleshooting activities are in progress to determine the cause of the event. At 1910, the NRC was previously notified of the Unusual Event declaration. Initial Safety Significance Evaluation: The safety significance of this event is minimal and Unit 2 is in a stable condition. All control rods properly inserted when the manual scram was performed. Plant safety systems responded as required with the exception of the CREV and ARI systems which did not automatically initiate but functioned properly when manually actuated. All four emergency diesels started and Unit 2 diesels 3 and 4 are supplying the Unit 2 emergency buses. Reactor pressure and level are being controlled per procedure, with HPCI and RCIC. Actions are in progress to re-establish off site power supply to emergency buses 3 and 4 via backfeed through the Unit Auxiliary Transformer (UAT). Corrective Actions: Actions are in progress to re-establish offsite power supply to emergency buses 3 and 4 via backfeed through the UAT. Investigations are in progress to determine the cause of the SAT failure and the failure of CREV and ARI to auto-initiate. The licensee has notified the NRC Resident Inspector and the State and local emergency agencies. Update provided also added the following reportable notifications due to the event: 10CFR50.72(b)(2) (iv)(A) and(iv)(B) and 10CFR50.72(b)(3)(iv)(A). Notified R2DO (Evans).

  • * * UPDATE PROVIDED BY MARK SCHALL TO JEFF ROTTON AT 1805 EST ON 11/02/06 * * *

Licensee reported that the Unusual Event was terminated at 1745 EST on 11/02/06 after Offsite power was restored to both 4 KV E Buses from the Unit Auxiliary Transformer (UAT) on Unit 2. The #3 and #4 EDGs have been secured and are in Standby. #1 EDG remains inoperable and #2 EDG is presently being Load Tested. The licensee will be notifying the NRC Resident Inspector and the State and local emergency agencies. Notified R2DO (Evans), NRREO (Richards), IRD Manager (Leach), DHS (Barnes), and FEMA (Kuzia).

ENS 4169212 May 2005 11:22:00BrunswickNRC Region 2Manual ScramGE-4

On May 12, 2005, at 0411 hours, electrical power was lost to the 4160 VAC Emergency Bus E1. Emergency Diesel Generator 1 was inoperable for maintenance at the time of the electrical power loss. This power loss to Emergency Bus E1 affected both Units 1 and 2. Unit 1 The loss of power to E1 resulted in Division 1 Primary Containment Isolation Valve (PCIV) actuations. The actuations included the Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Traversing In-core Probe, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 3 (i.e., Reactor Water Cleanup), and Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems) valves, as well as the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation) and the automatic start of Standby Gas Treatment (SGT) System train B. The actuations of PCIVs and Reactor Building Ventilation System isolation were complete and the affected equipment responded as designed to the invalid signal (i.e., the valves and dampers that were open, at the time of the event, closed). Additionally, SGT System train B started and functioned successfully. Loss of power to E1 also resulted in entry into LCO 3.0.3 (i.e., be in Mode 2 within 7 hours, Mode 3 within 13 hours, and Mode 4 within 37 hours) due to all required reactor coolant leakage detection instrumentation/systems being inoperable. At 0440 hours, it was discovered that all three Control Room Air Conditioning (AC) subsystems became inoperable due to failure of the control building air compressors and Technical Specification LCO 3.0.3 was entered. At 0515 hours, it was determined that both Control Room Emergency Ventilation (CREV) subsystems became inoperable when the dampers drifted shut. At 0546 hours, a control building air compressor was started and the control room air conditioning and CREV subsystems were returned to operable status. Operators initiated a plant shutdown for Unit 1, as required by Technical Specifications at 0948 hours. Unit 2 Conditions and activities associated with the Control Room AC and CREV systems apply to Unit 2 as well as Unit 1. Reporting Requirements Met by this Notification 10 CFR 50.72(b)(2)(1), the initiation of any nuclear plant shutdown required by the plant's Technical Specifications, applies to Unit 1. 10 CFR 50.72(b)(3)(v)(D), a condition that, at the time of discovery, could have prevented the fulfillment of the safety function of systems that are needed to mitigate the consequences of an accident (i.e., Control Room AC and CREV), applies to both Units 1 and 2. 10 CFR 50.73(a)(i), invalid actuation of general containment isolation signals affecting containment isolation valves in more than one system, applies to Unit 1. INITIAL SAFETY SIGNIFICANCE EVALUATION Currently Unit 2 is operating at steady state with Unit 1 being shut down. Specified systems actuated as designed. No adverse impact to the control room environment occurred during the period (i.e., one hour and 35 minutes) the affected ventilation system was inoperable. The other redundant emergency busses are operable. Prior to the event reactor coolant leakage level for Unit 1 was well within operating limits. The actions as required by the applicable Technical Specifications have been established. CORRECTIVE ACTIONS Activities are currently under way to determine the cause of the E1 power loss and restore electrical power to Emergency Bus E1. Causes and actions to preclude recurrence will be addressed in accordance with the corrective action program and provided to the NRC in the associated licensee event report. The licensee notified the NRC Resident Inspector.

  • * * UPDATE ON 05/13/05 @ 0952 BY LEONARD BELLER TO CHAUNCEY GOULD * * *

On May 12, 2005, at 0411 hours, electrical power was lost to the 4160 VAC Emergency Bus E1. Emergency Diesel Generator 1 was inoperable for maintenance at the time of the electrical power loss. This power loss to Emergency Bus E1 affected both Units 1 and 2. A non-emergency notification (Event Number 41692) was made to the NRC Operations Center at 112:2 hours. This follow-up notification discusses plant recovery from the Emergency Bus E1 power loss. Unit 1 Loss of power to E1 resulted in entry, into LCO 3.0.3 (i.e., be in Mode 2 within 7 hours, Mode 3 within 13 hours, and Mode 4 within 39 hours) due to all required reactor coolant system (RCS) leakage detection instrumentation being inoperable. Operators initiated a plant shutdown for Unit 1, as required by Technical Specifications at 0948 hours. A Notice of Enforcement Discretion (NOED) was requested from the NRC to waive compliance with the shutdown requirements associated with RCS leakage detection instrumentation in order to provide more time for an orderly plant shutdown. In lieu of the RCS leakage detection shutdown requirements (i.e., be in Mode 2 by 1111 hours), Unit 1 would adhere to the shutdown requirements associated with loss of Emergency Bus E1 (i.e.,, be in Mode 3 by May 13, 2005, at 0011 hours). The requested NOED was verbally granted by the NRC on May 12, 2005 at 1050 hours, so shutdown activities for Unit 1 continued versus the insertion of a manual reactor scram, with the unit at approximately 65 percent of rated thermal power. Power was restored to Emergency Bus E1 and the LCO associated with RCS leakage detection instrumentation was exited on May 12, 2005 at 1740 hours. The LCO associated with loss of power to Emergency Bus E1 was exited at 2015 hours. Unit 2 The LCO associated with loss of power to Emergency Bus E1 also applied to Unit 2, and was exited at 2015 hours, CORRECTIVE ACTIONS Emergency Diesel Generator 1 was made available, but not operable, on May 13, 2005, at 0117 hours. The licensee notified the NRC Resident Inspector. Reg 2 RDO (Moorman) was notified.

ENS 415829 April 2005 04:53:00BrunswickNRC Region 2Automatic ScramGE-4On 04/09/05 at approximately 00:50 an automatic reactor scram occurred on Brunswick Unit 2. The Reactor Protection System (RPS) actuated on low reactor water level (LL1). All control rods inserted from the RPS signal. The LL1 signal also provided a Group 2 (floor and equipment drain isolation valves), 6 (monitoring and sampling isolation valves) and 8 (shutdown cooling isolation valves) isolation signal for the respective containment Isolation valves. Reactor low level 2 (LL2) resulted in a Reactor Core Isolation Cooling (RCIC) system actuation and injection into the reactor. The High Pressure Coolant Injection (HPCI) system actuated but did not inject because reactor water level recovered. The Reactor Water Cleanup system (RWCU) isolated (Group 3 isolation). Secondary Containment isolated and the Standby Gas Treatment (SBGT) system initiated. An Alternate Rod Insertion signal was received and the Reactor Recirculation Pumps tripped as designed. An investigation is in progress to determine the cause of the reactor level transient. Safety systems and isolations functioned as designed. The NRC Resident Inspector was notified.
ENS 4090529 July 2004 12:14:00BrunswickNRC Region 2Manual ScramGE-4

On July 29, 2004, at approximately 0351 hours, testing of Suppression Chamber-to-Drywell vacuum breakers commenced in accordance with Periodic Test 0PT02.3.1, "Suppression Chamber-to-Drywell Vacuum Breaker Operability Test." At 0357 hours, it was determined that vacuum breaker 2-CAC-X18D had an open indication. Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.6.1.6, "Suppression Chamber-to-Drywell Vacuum Breakers," Condition B was entered for one suppression chamber-to-drywell vacuum breaker (2-CAC-X18D) not indicating closed. Required Action B.1 of TS LCO 3.6.1.6 provides 4 hours to verify the vacuum breaker closed. Attempts to confirm the vacuum breaker closed were unsuccessful and at 0757 hours, TS LCO 3.6.1.6 Condition C was entered. Required Action C.1 is to be in Mode 3 (Hot Shutdown) in 12 hours and Required Action C.2 is to be in Mode 4 (Cold Shutdown) in 36 hours. Initiation of the plant shutdown required by TS LCO 3.6.1.6 commenced at 0942 hours. This condition is being reported in accordance with 10 CFR 50.72(b)(2)(i) as the initiation of any nuclear plant shutdown required by the plant's TS. INITIAL SAFETY SIGNIFICANCE EVALUATION The initial safety significance of this condition is considered to be minimal. The actual position of the vacuum breaker cannot be confirmed, however if one vacuum breaker was not closed, communication between the drywell and suppression chamber air space could occur, and as a result, there is a potential, for primary containment overpressurization due to this bypass leakage if a loss of coolant accident were to occur. The plant is currently proceeding to hot shutdown with all Emergency Core Cooling systems operable. CORRECTIVE ACTIONS Determine the required corrective actions after entry into the drywell following shutdown. The safety significance of the event will be reviewed when the actual condition of the vacuum breaker is known. The NRC Resident Inspector has been notified.

  • * * UPDATE ON 07/29/04 @ 1654 FROM KEN CHISM TO CHAUNCEY GOULD * * *

On July 29, 2004, at 1517 hours, a manual scram was inserted to shut down the Unit 2 reactor from approximately 23 percent of rated thermal power. The reactor shutdown was implemented to meet the requirements of Technical Specification (TS) Limiting Condition for Operation (LCO) 3.6.1.6 Required Action C.1 to be in Mode 3 in 12 hours. All control rods fully inserted into the core and no ECCS actuation occurred or relief valves lifted. Following the scram, an expected reactor vessel coolant level shrink occurred causing coolant level to decrease below the Low Level 1 setpoint. This coolant level decrease resulted in a Primary Containment Isolation Valve isolation signal to Group 2 (Drywell Equipment and Floor Drain, Traversing In-core Probe, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 6 (Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling System), and Group 8 (RHR Shutdown Cooling Suction and RHR Inboard Injection) isolation valves. The isolation signal closed all of the valves that were open at the time of the expected initiation. The plant is currently in mode 3 and proceeding with a normal cooldown to Mode 4. The NRC Resident Inspection was informed. Notified Reg 2 RDO (Tom Decker).