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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 54603 | 24 March 2020 16:05:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification-Required Shutdown Due to Unidentified Leakage | At 1205 Eastern Daylight Time (EDT) on March 24, 2020, a Technical Specification-required shutdown was initiated on Unit 1 due to indication of a leak in the drywell. Technical Specification Action 3.4.4.A, Unidentified Reactor Coolant System (RCS) leakage increase not within limit, requires RCS leakage to be reduced to within limits within 8 hours. It was expected that the leakage would not have been reduced to within limits within the required Technical Specification completion time; therefore, this event is being reported in accordance with 10 CFR 50.72(b)(2)(i). Reactor water level reached low level 1 (LL1) following the reactor shutdown. The LL1 signal causes 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 valid Primary Containment Isolation System (PCIS) actuation, this event is also being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Unit 2 is not affected by this event. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Reactor Coolant System Primary Containment Isolation System Shutdown Cooling | |
| ENS 53961 | 28 March 2019 18:50:00 | 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Unusual Event Declared Due to Rcs Unidentified Leakage | At 1450 EDT on March 28, 2019, the licensee observed that the Unit 1 unidentified Reactor Coolant System (RCS) leakage was greater than 10 gallons per minute (gpm) for greater than or equal to 15 minutes. The licensee declared an Unusual Event in accordance with their EAL SU 5.1. The licensee initiated a unit shutdown in accordance with their procedures and the unit was approximately 58 percent reactor power at 1507 EDT, with unit shutdown in progress. The licensee also received an alarm due to increasing Drywell Pressure at 1.7 pounds drywell pressure. At 1600 EDT the licensee called with an update. Unit 1 was still in an Unusual Event with the unit at 37 percent power with the shutdown continuing. Drywell Pressure had decreased to 0.8 pounds. At 1603 the licensee scrammed Unit 1. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
At 1437 EDT on March 28, 2019, with Unit 1 in Mode 1 at approximately 100 percent power, a Technical Specification-required shutdown was initiated due to indication of a leak in the drywell. Technical Specification Action 3.4.4.A, Unidentified Reactor Coolant System (RCS) leakage increase not within limit, requires RCS leakage to be reduced to within limits within 8 hours. It is expected that the leakage would not have been reduced to within limits within the required Technical Specification completion time; therefore, this event is being reported in accordance with 10 CFR 50.72(b)(2)(i). There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified.
At 0259 EDT on March 29, 2019, the Unusual Event was terminated because RCS leakage was reduced to less than 10 gallons per minute. The most recent leakage rate measured at 0225 EDT was 3.9 gpm. The source of the leak will be identified when plant conditions allow containment entry. No elevated radiation levels were observed during this event. Drywell pressure is currently 0.0 psig. Unit 1 is in Mode 4. The licensee notified the NRC Resident Inspector. Notified R2DO (Bonser), NRR EO (Miller), IRD MOC (Grant), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). | Reactor Coolant System | |
| ENS 47687 | 23 February 2012 01:14:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Inoperability of Emergency Core Cooling Systems | At 1859 hours EST, the Brunswick site experienced a loss of balance of plant (BOP) bus Common C. As a result, makeup pumps to the ECCS discharge line keepfill systems lost power. At 1905 on Unit 1, 'A' loop of the Core Spray (CS) system received a low discharge pressure alarm and was declared inoperable. At 1916 hours, 'B' loop of the Residual Heat Removal (RHR) system received a low discharge pressure alarm and was declared inoperable. With the loss of the second low pressure ECCS system, Condition J of Technical Specification 3.5.1, 'ECCS Operating,' was entered, which requires the Unit 1 to enter LCO 3.0.3 immediately. At 1931 hours, 'A' loop of RHR was declared inoperable due to low discharge pressure. Power reduction of Unit 1 was initiated at 2014 hours. At 2055 hours on Unit 2, 'A' loop of the Residual Heat Removal (RHR) system received a low discharge pressure alarm and was declared inoperable. At 2128 hours, 'B' loop of the Core Spray (CS) system received a low discharge pressure alarm and was declared inoperable. With the loss of the second low pressure ECCS system, Condition J of Technical Specification 3.5.1, 'ECCS Operating,' was entered, which requires the Unit 2 to enter LCO 3.0.3 immediately. Power reduction of Unit 2 was initiated at 2219 hours. This event reportability is in accordance with 10CRF50.72(b)(2)(i), Technical Specification Required Shutdown, due to inoperability of ECCS systems. The initial safety significance of this event is minimal. Offsite power and the Emergency Diesel Generators are operable. The High Pressure Coolant Injection (HPCI) system remains operable on both Unit 1 and Unit 2. The Reactor Core Isolation Cooling (RCIC) system remains operable on Unit 1 and is being restored following maintenance on Unit 2. Troubleshooting activities to determine the loss of the BOP Common C bus are in progress. Efforts are in progress to install temporary power to the keepfill makeup pumps. The licensee will notify the NRC Resident Inspector.
Unit 1 - At 2315 hours, temporary power was provided to the ECCS keepfill makeup pump and the ECCS systems were restored. LCO 3.0.3 was exited on Unit 1 at 0041 hours with restoration of the 'A' and 'B' loops of the RHR systems. The 'A' loop of the Core Spray system was restored at 0058 hours on 2/23/2012. During the shutdown, Unit 1 was manually scrammed due to high delta-pressure across the Circulating Water Pump traveling screens. See EN #47690 for details. Unit 2 - At 2315 hours, temporary power was provided to the ECCS keepfill makeup pump and the ECCS systems were restored. LCO 3.0.3 was exited on Unit 2 at 2354 hours with restoration of 'B' loop of the RHR system. The 'A' loop of the Core Spray system was restored at 0039 hours. Unit 2 was at 96% of Rated Thermal Power when the shutdown was terminated. The licensee notified the NRC Resident Inspector. Notified R2DO (Ernstes). | High Pressure Coolant Injection Emergency Diesel Generator Reactor Core Isolation Cooling Core Spray Residual Heat Removal Emergency Core Cooling System | 05000325/LER-2012-001 |
| ENS 45368 | 20 September 2009 18:00:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Inoperability of Diesel Generator | Event reportability, is in accordance with 10 CFR 50.72(b)(2)(i), Technical Specification Required Shutdown, due to inoperability of Diesel Generator #4 extending from planned maintenance. Brunswick Nuclear Plant Units 1 and 2 are initiating unit shutdowns in anticipation of Technical Specification Required Shutdown as required by Technical Specification 3.8.1, Condition H due to the inoperability of Diesel Generator #4 lasting longer than seven (7) days. Power reduction commenced at 1400 on Unit 1, and is scheduled to commence at 2200 09/20/2009 on Unit 2 in accordance with site procedures. Both units will continue the shutdown to Mode 4 or until the emergency diesel generator is declared operable following appropriate repairs and testing. The licensee has notified the NRC Resident Inspector. | Emergency Diesel Generator | 05000325/LER-2009-004 |
| ENS 43269 | 1 April 2007 14:45:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Tech Spec Required Shutdown Due to Edg Inoperability | During post-maintenance operability testing, Diesel Generator 4 exhibited unstable power output while synchronized to the grid. This testing was being performed at the conclusion of a planned Diesel Generator 4 maintenance outage. Preventive maintenance activities included in this scheduled outage included replacement of various relays in the Diesel Generator control logic and replacement of the engine governor. Brunswick Nuclear Plant Unit 1 is commencing a unit shutdown as required by Technical Specification 3.8.1 Condition H due to the inoperability of Diesel Generator 4 lasting longer than seven (7) days. TS 3.8.1 Required Action H.1 requires Unit 1 to be in MODE 3 by 1615 on 4/1/2007 and in MODE 4 by 1615 on 4/2/2007. This event is reportable per 10CFR50.72(b)(2)(i). Power reduction was commenced at 1045 in accordance with station procedures for unit shutdown. Power level has been reduced from 100% to 53%. Unit 1 shutdown will continue until the unit is in MODE 4 or until the engine is declared operable following appropriate repairs and testing. Unit 2 is in MODE 4, Cold Shutdown. The inoperable Diesel Generator 4 is not needed to meet the requirements of Unit 2 Technical Specification 3.8.2. The inoperability of Diesel Generator 4 is of minimal safety significance for Unit 1. Unit 1 has two operable off-site circuits and three (3) Diesel Generators are operable. The remaining three (3) operable Diesel Generators and off-site circuits are adequate to supply electrical power to the on-site Class 1 E Distribution System. This maintains the safety function of the AC sources. Unit 1 is being shutdown to place the unit in MODE 4 where the inoperable Diesel Generator 4 is not required. Unit 2 is currently in MODE 4 and does not require Diesel Generator 4 to be operable. As such, this event is of zero safety significance for Unit 2. During post-maintenance operability testing, Diesel Generator 4 exhibited unstable power output while synchronized to the grid. Station engineering and a vendor representative continue to evaluate the engine/generator performance. Repair plans will be developed and implemented as appropriate following completion of the engineering evaluation. The inoperability of Diesel Generator 4 will prevent Unit 2 from changing mode for returning the plant to service. The licensee notified the NRC Resident Inspector. | ||
| ENS 42960 | 2 November 2006 23:53:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Shutdown | On 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. | Emergency Diesel Generator | |
| ENS 41895 | 5 August 2005 22:40:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Technical Specification Required Shutdown Due to Inoperability of All Emergency Diesel Generators | Event reportability is in accordance with 10 CFR 50.72(b)(3)(v)(D), 'Accident Mitigation,' and 10 CFR 50.72(b)(2)(i), 'TS Required Shutdown,' due to potential COMMON CAUSE FAILURE of all four Emergency Diesel Generators. On 05 August 2005, at 14:31 hrs, DG2 was started for data gathering purposes, and it experienced a trip and lockout on differential current. Investigation revealed that set points for the installed DG differential current protective devices may be set too conservatively. Since each DG has the same relay with the same set point, all DGs were declared inoperable. Preparations are in progress to place both units in Cold Shutdown. The site has insufficient operable backup AC sources. The current effect on the plant is minimal due to all normal AC sources being available and operable. However, due to the potential for losing a normal AC source resulting in insufficient power being available to some emergency systems, both units will be placed in cold shutdown until all emergency diesel generators can be restored to operable. Corrective actions are in progress to correct differential current protective device problem in order to restore availability of DG2 and operability of all four DGs. The licensee plans on commencing a down power on both units (Unit 2 first) within several hours. Technical Specifications require both units to be in Hot Shutdown within 12 hours (approximately 0700 on 08/06/05) and Cold Shutdown within 36 hours. Neither unit has any other significant systems LCOs. The licensee notified the NRC Resident Inspector.
This is a follow up message to the initial event reported in accordance with 10 CFR 50.72(b)(3)(v)(D), 'Accident Mitigation' and 10 CFR 50.72(b)(2)(i), 'TS Required Shutdown' due to the potential COMMON CAUSE FAILURE of all four Emergency Diesel Generators. Unit 2 entered Mode 3 (Hot Shutdown) at 04:46. Unit 1 entered Mode 3 (Hot Shutdown) at 05:31. Both units are continuing to cool down in preparation for entering Mode 4 (Cold Shutdown)." The licensee notified the NRC Resident Inspector. Notified R2DO(Landis), NRR EO (Haney), and IRD (McGinty). | Emergency Diesel Generator | 05000325/LER-2005-006 |
| ENS 41692 | 12 May 2005 08:11:00 | 10 CFR 50.72(b)(3)(v)(D), Loss of Safety Function - Mitigate the Consequences of an Accident 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown 10 CFR 50.73(a)(1), Submit an LER | Technical Specification Required Shutdown Due to Loss of Emergency Bus E1 | 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.
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. | Reactor Coolant System Secondary containment Emergency Diesel Generator Primary Containment Isolation System Primary containment Reactor Building Ventilation Residual Heat Removal Reactor Water Cleanup Control Room Emergency Ventilation | 05000325/LER-2005-004 |
| ENS 40905 | 29 July 2004 13:42:00 | 10 CFR 50.72(b)(2)(i), Tech Spec Required Shutdown | Plant Entered Lco Action Statement Due to Malfunction of a Suppression Chamber-To-Drywell Vacuum Breaker. | 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.
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). | Primary containment Shutdown Cooling Residual Heat Removal Control Rod |