Semantic search
| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56977 | 19 February 2024 04:25:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of Emergency Diesel Generator | The following information was provided by the licensee via phone and email: At approximately 2325 EST on February 18, 2024, with Unit 1 in Mode 5 at 0 percent power and Unit 2 in Mode 1 at 100 percent power, emergency diesel generator 2 automatically started due to the unexpected loss of AC power to emergency bus E2 during a planned transfer of E2 DC control power from normal to alternate for the 1B-1 battery. In addition, the unexpected loss of AC power to E2 resulted in Unit 1 primary containment isolation system (PCIS) partial Group 2 (i.e., drywell equipment and floor drain, residual heat removal (RHR), discharge to radioactive waste, and RHR process sample), Group 6 (i.e., containment atmosphere control/dilution, containment atmosphere monitoring, and post accident sampling systems), and partial Group 10 (i.e., air isolation to the drywell) isolations. Emergency diesel generator 2 automatically started and re-energized the E2 bus as designed when the loss of E2 signal was received. The PCIS actuations were as expected for the outage plant line up on Unit 1 at the time. The cause of the loss of electrical power to emergency bus E2 is under investigation at this time. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of emergency diesel generator 2 and PCIS. 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: This event will be entered into the plant's corrective action program. | Emergency Diesel Generator Primary Containment Isolation System Residual Heat Removal | |
| ENS 56478 | 20 April 2023 05:48:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip Due to a Turbine Trip | The following information was provided by the licensee via phone and email: At 0148 Eastern Daylight Time (EDT) on April 20, 2023, with Unit 1 in Mode 1 at 100% power, the reactor automatically tripped due to a turbine trip. Turbine Bypass valves did not open on the trip due to Turbine Protection system power supply failure; the Safety Relief Valves (SRVs) opened automatically to control reactor pressure. Reactor Pressure reached approximately 1095 psig on the trip; exceeding the 1060 psig RPS trip setpoint. Operations responded and stabilized the plant. Operations was able to transition from SRVs to main steam line drains to the condenser. Reactor water level is being maintained via the Condensate / Feedwater system. Decay heat is being removed by discharging steam to the main condenser using the main steam line drains. 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). Reactor water level reached low level 1 (LL1) following the reactor trip. 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 and RPS actuation from the reactor pressure signal, 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. | Feedwater Reactor Protection System Primary Containment Isolation System Shutdown Cooling Main Steam Line Safety Relief Valve Main Condenser | |
| ENS 55627 | 6 December 2021 16:25:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Safety System Actuation | On December 6, 2021, at 1125 hours Eastern Standard Time (EST), during planned maintenance activities, electrical power was lost to the 4160V emergency bus E-3. The power loss to emergency bus E-3 affected both Unit 1 and 2. Emergency Diesel Generator #3 received an automatic start signal but was under clearance for planned maintenance. Emergency bus E-3 was re-energized at 1315 EST hours via offsite power. The loss of power to E3 resulted in Unit 2 Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 3 (i.e., Reactor Water Cleanup), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., air isolation to the drywell) isolations. The actuations of PCIVs were completed and the affected equipment responded as designed. Per design, no Unit 1 safety system group isolations or actuations occurred. Other Unit 2 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start signal to the Standby Gas Treatment (SGT) System trains A and B and the Control Room Emergency Ventilation System (CREV). Systems functioned as designed. Safety systems functioned as designed following the de-energization of bus E-3. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of PCIS on Unit 2 and an automatic start signal to Emergency Diesel Generator #3. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. Except for the Emergency Diesel Generator, which is out of service for planned maintenance, all equipment has been returned to its normal alignment.
The loss of power to E3 resulted in Unit 1 Primary Containment Isolation System (PCIS) Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems). Other Unit 1 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start signal to the Standby Gas Treatment (SGT) System trains A and B and the Control Room Emergency Ventilation System (CREV). Systems functioned as designed. Safety systems functioned as designed following the de-energization of bus E-3. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that resulted in a valid actuation of PCIS on Unit 1. All Unit 1 equipment was returned to its normal alignment. The NRC Resident will be notified. Notified R2DO (Miller). | Secondary containment Emergency Diesel Generator Primary Containment Isolation System Reactor Building Ventilation Residual Heat Removal Reactor Water Cleanup Control Room Emergency Ventilation | |
| ENS 54016 | 22 April 2019 03:07:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Automatic Reactor Trip and Specified System Actuation | 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.
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). | Reactor Coolant System Secondary containment Primary Containment Isolation System Shutdown Cooling Reactor Water Cleanup Control Room Emergency Ventilation Emergency Core Cooling System Safety Relief Valve Control Rod | |
| ENS 53966 | 30 March 2019 21:47:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram and Specified System Actuation | At 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. | Feedwater High Pressure Coolant Injection Main Steam Isolation Valve Primary Containment Isolation System Main Turbine Shutdown Cooling Main Condenser Control Rod | |
| ENS 53962 | 28 March 2019 20:54:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of the Primary Containment Isolation System and the Reactor Protecton System | At 1654 EDT on March 28, 2019, with Unit 1 in Mode 3 at 0 percent power, an actuation of the Primary Containment Isolation System occurred, closing the outboard Main Steam Isolation Valves (MSIVs) due to a low condenser vacuum signal. The MSIVs had been manually closed, per procedure, during the shutdown evolution to address drywell leakage. The inboard MSIVs had not been reopened when the isolation occurred. Subsequently, at 1658 EDT a Reactor Protection System (RPS) actuation occurred due to reactor water level dropping below the actuation setpoint. All control rods were inserted at the time of the actuation. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuations of the Primary Containment Isolation System and the Reactor Protection System. There was no impact on the health and safety of the public or plant personnel. The safety function of both the MSIVs and the RPS had already been completed at the time of the event. The NRC Resident Inspector has been notified. | Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Control Rod | |
| ENS 53955 | 25 March 2019 08:02:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Start of Four Emergency Diesel Generators | At 0402 Eastern Daylight Time (EDT) on March 25, 2019, an actuation of the four Emergency Diesel Generators (EDGs) occurred. At the time of the event, Unit 1 was in Mode 1 at approximately 100% power and Unit 2 was in Mode 4 at 0% power. Unit 2 was in the process of aligning the electrical distribution system to power the emergency buses via the Unit Auxiliary Transformer (UAT) in accordance with plant procedures. It was determined that a fault occurred on the power path between the 230 KV switchyard and the UAT. This caused a main generator differential lockout relay to actuate; thereby starting the EDGs. All emergency buses remained energized from offsite power via the Startup Auxiliary Transformer and, therefore, the EDGs did not tie to their respective buses. The EDGs responded per design to this event. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in valid actuation of the EDGs. Due to the shared configuration of the Brunswick electrical system, both Unit 1 and Unit 2 are affected. The Unit 2 main generator lockout was reset and the EDGs have been restored to standby condition. Troubleshooting activities to determine the cause of the fault are in progress. There was no impact on the health and safety of the public or plant personnel. The NRC Resident Inspector has been notified. | Emergency Diesel Generator | |
| ENS 53911 | 5 March 2019 10:35:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of the Primary Containment Isolation System | At 05:35 Eastern Standard Time (EST) on March 5, 2019, with Unit 2 in Mode 5 at 0% power, an actuation of the Primary Containment Isolation System occurred during hydrolazing of the reactor water level variable leg instrumentation line nozzle N011B in the reactor cavity. The hydrolazing activity caused low reactor water level to be sensed on Division II of the shutdown range level instrumentation. 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. This event is being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A) as an event that results in a valid actuation of the Primary Containment Isolation System. There was no impact on the health and safety of the public or plant personnel. The Group 8 was reset and shutdown cooling was restored at approximately 05:45 EST. The safety significance of this event was minimal. Although there was a brief interruption of the shutdown cooling, the Residual Heat Removal (RHR) shutdown cooling system operation was restored in approximately 10 minutes without extensive troubleshooting or maintenance, and remained operable. The RHR shutdown cooling system is not credited in any Updated Final Safety Analysis Report Chapter 6 or 15 accidents or transients. The NRC Resident Inspector has been notified. | Primary Containment Isolation System Shutdown Cooling Residual Heat Removal | |
| ENS 53319 | 7 April 2018 12:36:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Trip and Pcis Actuation During Stator Cooling System Testing | 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. | Feedwater Reactor Protection System Primary Containment Isolation System Shutdown Cooling | |
| ENS 52974 | 17 September 2017 13:38:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Emergency Diesel Generator and Primary Containment Isolation System Actuations | On September 17, 2017, during planned surveillance activities involving Emergency Diesel Generator (EDG) 4, unexpected voltage and frequency indications were noted when EDG 4 was synchronized to Emergency Bus E4. With EDG 4 in manual mode, the Operator responded by lowering load to reopen the EDG 4 output breaker. Opening of the EDG 4 output breaker with the breakers from Balance of Plant (BOP) Bus 2C, which normally feeds the Emergency Bus E4, opened; resulted in de-energizing Emergency Bus E4. The EDG 4 voltage regulator and governor automatically reverted to auto control, and EDG 4 reconnected to Emergency Bus E4. Normal frequency and voltage were restored with EDG 4 in auto control. The momentary power interruption to Emergency Bus E4 resulted in Unit 2 Primary Containment Isolation System (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 10 (i.e., air isolation to the drywell) isolations. The actuations of Primary Containment Isolation Valves (PCIVS) were completed and the affected equipment responded as designed. Per design, no Unit 1 safety system group isolations or actuations occurred. These actuations are being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Additional Unit 2 actuations included PCIS Group 3 (i.e., Reactor Water Cleanup), Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start of Standby Gas Treatment (SGT) System subsystems A and B. These systems functioned as designed. This event did not impact public health and safety. The NRC Resident Inspector has been notified. The safety significance of this event is minimal. Safety systems functioned as designed following the power perturbation on E4. Plant systems responded as designed. The cause of the event is under investigation. | Secondary containment Emergency Diesel Generator Primary Containment Isolation System Primary containment Reactor Building Ventilation Residual Heat Removal Reactor Water Cleanup | |
| ENS 52683 | 17 April 2017 04:04:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Actuation of Emergency Diesel Generators | On April 17, 2017, at 0004 Eastern Daylight Time (EDT), an automatic actuation of the four Emergency Diesel Generators (EDGs) was received. At the time of the event, Unit 2 was in the process of starting the main turbine following a refueling outage. Operations personnel tripped the main turbine due to elevated bearing vibrations. When the main turbine was tripped, Power Circuit Breakers (PCBs) 29A and 29B failed to open. This caused a main generator primary lockout due to generator reverse power and the subsequent automatic actuation of all four EDGs. All emergency buses remained energized from offsite power and therefore, the EDGs did not tie to their respective buses. The protective relaying and EDGs responded per design to this event. This event is being reported in accordance with 10 CFR 50.73(b)(3)(iv)(A) as an event that results in a valid actuation of the EDGs. Due to the shared configuration of the Brunswick electrical system, both Unit 1 and Unit 2 are affected. This event did not impact public health and safety. The NRC Resident lnspector has been notified. | Emergency Diesel Generator Main Turbine | 05000325/LER-2017-002 |
| ENS 51715 | 7 February 2016 18:26:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Scram and Alert Declaration Due to Electrical Fault Resulting in Fire/Explosion | 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).
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). | High Pressure Coolant Injection Emergency Diesel Generator Main Steam Isolation Valve Reactor Core Isolation Cooling | 05000325/LER-2016-001 |
| ENS 47817 | 9 April 2012 09:29:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Emergency Diesel Generator Actuation | On April 9, 2012, at 0529 hours, electrical power was lost to the 4160v emergency bus E1. Activities to support testing of emergency bus E1 were in progress when technicians connected a recorder across the terminals of an under-voltage relay on emergency bus E1 and caused the normal supply breakers for emergency bus E1 to open. The power loss to emergency bus E1 affected both Unit 1 and Unit 2. Emergency diesel generator #1 automatically started and re-energized the E1 emergency bus. Unit 1 was in Mode 5 and electrical systems were aligned to support testing of emergency bus E1. As a result, no other safety system isolations or actuations occurred. Per design, no Unit 2 safety system isolations or actuations occurred. The safety significance of this event is minimal. Safety systems functioned as designed when emergency bus E1 de-energized. There was no interruption of Unit 1 shutdown cooling as a result of this event. Normal power supply was restored to emergency bus E1 and emergency diesel generator #1 was shutdown at 0701 hours. Reporting requirements met by this notification: 10CFR50.72(b)(3)(iv)(A) with specified system in 10CFR50.72(b)(3) (iv)(B)(8). The NRC Resident Inspector has been notified. | Emergency Diesel Generator Shutdown Cooling | |
| ENS 47697 | 24 February 2012 05:37:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Reactor Protection System Actuation During Neutron Instrument Testing | On 2/24/2012 at 0037 EST, Unit 1 was in Mode 4 when an unplanned Reactor Protection System (RPS) actuation occurred. The trip occurred while operators were returning the Mode Switch to the 'Shutdown' position during restoration from Neutron Instrumentation testing. Jumpers had not been installed to bypass this actuation signal at the time the Mode Switch was operated, resulting in a valid signal of the RPS. The RPS actuation is reportable in accordance with 10 CFR 50.72(b)(3)(iv)(A). The safety significance of this event was minimal. Plant equipment performed as expected. All control rods were inserted prior to the RPS actuation and remain inserted. The RPS actuation was reset and the plant remains in Mode 4. The licensee has notified the NRC Resident Inspector.
Based on a detailed review of NUREG-1022, Revision 2, 'Event Reporting Guidelines 10 CFR 50.72 and 50.73,' this event has been determined not to be reportable under 10 CFR 50.72(b)(3)(iv)(A). The RPS actuation was inadvertent and was caused by a human error (i.e., failure to install appropriate jumpers) that occurred during a surveillance test. This RPS actuation was not in response to actual plant conditions satisfying the requirements for initiation of the RPS. There was no plant condition present that either warranted a scram or would prompt manual operator action in anticipation of scram condition. Therefore, this RPS actuation is considered invalid and is not reportable per 10 CFR 50.72(b)(3)(iv)(A). Investigation of this condition is documented in the corrective action program in Condition Report (CR) 519432. The NRC Resident Inspector was notified of this retraction." Notified the R2DO (Mark Franke). | Reactor Protection System Control Rod | |
| ENS 47690 | 23 February 2012 04:19:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to High Delta-P Across Circulating Water Pump Traveling Screens | 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).
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). | High Pressure Coolant Injection Reactor Protection System Main Steam Isolation Valve Reactor Core Isolation Cooling Primary containment Main Condenser Control Rod | |
| ENS 45902 | 5 May 2010 15:44:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Reactor Feed Pump Trip | On 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. | High Pressure Coolant Injection Reactor Protection System Primary Containment Isolation System Reactor Core Isolation Cooling Primary containment Shutdown Cooling Reactor Recirculation Pump Reactor Pressure Vessel Residual Heat Removal Decay Heat Removal Control Rod | 05000325/LER-2010-003 |
| ENS 45732 | 27 February 2010 06:58:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Core Isolation Cooling (Rcic) System Manually Started to Maintain Reactor Pressure Vessel Level Following a Pre-Planned Reactor Scram | On February 27, 2010, at 0116 hours, as part of a pre-planned sequence of events, Unit 1 control room operators manually inserted a Reactor Protection system trip to shut down the reactor from approximately 21 percent of rated thermal power to begin a planned refuel outage. At 0158 hours, the Reactor Core Isolation Cooling (RCIC) system was manually started to maintain reactor pressure vessel (RPV) coolant level after the 1A Reactor Feedwater Pump (RFP) was shutdown due to high turbine casing drain level, and the 1B Reactor Feedwater Pump (RFP) had been removed from service and isolated to support maintenance activities. The Reactor Core Isolation Cooling (RCIC) system started and maintained reactor pressure vessel (RPV) coolant level until the 1B Reactor Feedwater Pump (RFP) could be returned to service at 0248 (hours). The Reactor Core Isolation Cooling (RCIC) system was shutdown at 0306 (hours). The safety significance of this occurrence is considered minimal. The RCIC system is designed to operate either automatically or manually following RPV isolation, accompanied by a loss of normal coolant flow from the Reactor Feedwater system, to provide adequate core cooling, and control of the RPV coolant level. The function of the RCIC system is to respond to transient events by providing makeup coolant to the reactor. For this event, the RPV was not isolated, but there was a loss of normal coolant flow from the Reactor Feedwater System. The RCIC system was manually started and successfully controlled RPV coolant level for approximately 68 minutes until Reactor Feedwater system flow could be restored to the RPV. All emergency core cooling systems were operable and ready for use, if needed, during this event. This transient is bounded by the analyses in the Updated Final Safety Analysis Report. RPV level remained in the normal post-shutdown band of 170-200 inches during the transient. The licensee informed the NRC Resident Inspector. | Feedwater Reactor Protection System Reactor Core Isolation Cooling Reactor Pressure Vessel Emergency Core Cooling System | 05000325/LER-2010-001 |
| ENS 45375 | 21 September 2009 18:46:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Protection System and Primary Containment Isolation System Actuation After Placing Hpci System Inservice | On September 21, 2009, at 14:46 EDT, Unit 1 received valid actuations of the Reactor Protection System (RPS) and the Primary Containment Isolation System. Unit 1 was non-critical, operating in Mode 3, when a RPS actuation occurred. Operators were placing the High Pressure Coolant Injection (HPCI) system in service for reactor pressure control, when a resulting water level shrink caused level in the Reactor Pressure Vessel to drop to Low Level 1 causing the actuation of RPS and the Primary Containment Isolation system. The HPCI system was secured, and level stabilized in the normal band. Primary Containment Isolation system 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. RHR was not in shutdown cooling at the time of the isolation signal. Actuations of the Primary Containment Isolation Valves (PCIVs) were completed and the affected equipment responded as designed, with the following exceptions: Two Group 2 valves (1-G16-F003 and 1-G16-F019) and two Group 6 valves (1-CAC-V6 and l-CAC-V9) did not automatically isolate and were manually isolated from the control panel. Investigation is under way to determine why these valves did not automatically close. Unit 1 was non-critical, in Mode 3, with all rods inserted at the time of the event. The four primary containment isolation valves that did not automatically close did not create an unisolated primary containment penetration. The Emergency Core Cooling System (ECCS), along with Reactor Core Isolation Cooling (RCIC), were operable and available. The licensee has notified the NRC Resident Inspector. | High Pressure Coolant Injection Reactor Protection System Primary Containment Isolation System Reactor Core Isolation Cooling Primary containment Shutdown Cooling Reactor Pressure Vessel Residual Heat Removal Emergency Core Cooling System | |
| ENS 45190 | 8 July 2009 14:13:00 | 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Valid Emergency Diesel Generator Actuation | On July 8, 2009, at 1013 hours, during planned maintenance activities, electrical power was lost to the 4160V emergency bus E-2. The power loss to emergency bus E-2 affected both Unit 1 and 2. Emergency diesel generator #2 automatically started and re-energized the E-2 bus. The loss of power to E2 resulted in Unit 1 Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 3 (i.e., Reactor Water Cleanup), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., air isolation to the drywell) isolations. The actuations of PCIVs were completed and the affected equipment responded as designed. Per design, no Unit 2 safety system group isolations or actuations occurred. Other Unit 1 actuations included the Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start of Standby Gas Treatment (SGT) System trains A and B and the Control Room Emergency Ventilation System (CREV). Systems functioned as designed. Safety systems functioned as designed following the de-energization of bus E-2. The safety significance of this event is minimal. Plant systems responded as designed. The cause of the event is under investigation. Reporting requirements met by this notification: 10 CFR 50.72(b)(3)(iv)(A). The NRC Resident Inspector has been notified. Licensee was conducting maintenance on the switchgear at the time of the loss of power. | Secondary containment Emergency Diesel Generator Primary Containment Isolation System Reactor Building Ventilation Residual Heat Removal Reactor Water Cleanup Control Room Emergency Ventilation | |
| ENS 44685 | 26 November 2008 17:00:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Automatic Reactor Scram Due to Ehc Malfunction | At 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. | Reactor Protection System Main Steam Isolation Valve Primary Containment Isolation System Main Steam Line Control Rod | 05000325/LER-2008-007 |
| ENS 44647 | 9 November 2008 16:08:00 | Other Unspec Reqmnt 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Manual Reactor Scram Due to Spurious Srv Opening | One 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. | Feedwater High Pressure Coolant Injection Primary Containment Isolation System Reactor Core Isolation Cooling Reactor Recirculation Pump Safety Relief Valve Control Rod | 05000324/LER-2008-002 |
| ENS 44453 | 30 August 2008 19:03:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Electro-Hydraulic Control Malfunction Leads to Reactor Scram | At 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. | Reactor Protection System Primary Containment Isolation System Control Rod | 05000324/LER-2008-001 |
| ENS 43062 | 25 December 2006 10:39:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Reactor Scram Due to Trip from Neutron Monitoring System | On 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. | Feedwater Secondary containment High Pressure Coolant Injection Reactor Protection System Reactor Core Isolation Cooling Shutdown Cooling Reactor Recirculation Pump Reactor Water Cleanup Control Room Emergency Ventilation Control Rod | 05000324/LER-2006-003 |
| ENS 42986 | 11 November 2006 17:43:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scram Due to Conductivity Excursion | At 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. | Shutdown Cooling Emergency Core Cooling System Control Rod | |
| ENS 42955 | 1 November 2006 23:37:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(a)(1)(i), Emergency Class Declaration 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Unit 2 Declared an Unusual Event Due to Loss of Offsite Power to the 4Kv Emergency Buses | 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.
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).
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). | Secondary containment High Pressure Coolant Injection Emergency Diesel Generator Main Steam Isolation Valve Reactor Core Isolation Cooling Shutdown Cooling Reactor Water Cleanup Control Room Emergency Ventilation Safety Relief Valve Control Rod | 05000324/LER-2006-001 |
| ENS 41837 | 13 July 2005 13:17:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Turbine Trip/Reactor Trip | On July 13, 2005, at approximately 0917 hours, Unit 1 received a Reactor Protection System (RPS) trip and a Main Turbine trip. All control rods fully inserted into the core. Plant response to the reactor shutdown resulted in a reactor coolant level transient that caused a Low Level 2 signal and subsequent High Pressure Coolant Injection (HPCI) system initiation. The HPCI system started but did not inject into the vessel because reactor coolant level was already recovered by the Reactor Feedwater system. Proper operation of the RCIC system has not been conclusively determined. Both Reactor Recirculation pumps tripped, as expected, from the low reactor coolant level and high pressure transient. Additionally, primary containment isolation system actuation signals for valve groups 2, 3, 6, and 8 were received and the valves, that were open, closed as required. The Reactor Building Ventilation System isolated and both trains of the Standby Gas Treatment (SBGT) System automatically started and operated successfully. The RPS trip was reset and the HPCI turbine secured. Both trains of the SBGT system were secured. At 1119 hours, the plant exited the scram recovery procedure and entered General Plant Operating procedure 0GP-05, "Unit Shutdown." Investigation into the cause of the RPS and Main Turbine trip is still in progress. The plant is currently in Mode 3 (i.e., Hot Shutdown) and activities are in progress to transition to Mode 4 (i.e., Cold Shutdown). The resident inspector has been notified. INITIAL SAFETY SIGNIFICANCE EVALUATION The initial safety significance of this condition is considered to be minimal. The plant responded as designed to the transient, with the exception of the verification of RCIC system performance, and the plant was safely shut down. CORRECTIVE ACTIONS An event investigation team has been assembled to determine the cause of the event. Plant response to the event is being evaluated and identified issues will be addressed prior to plant restart. During the transient, four safety relief valves lifted and reset. The plant is currently in a normal shutdown electrical lineup. Reactor vessel level is being maintained using normal feedwater. The bypass valves are available for cooldown. There was no effect on unit-2.
On July 13, 2005, at approximately 0917 hours, Unit 1 received a Reactor Protection System (RPS) trip and a Main Turbine trip. All control rods fully inserted into the core and the plant safely shut down. A non-emergency notification (Event Number 41837) was made to the NRC Operations Center at 1303 hours. This follow-up notification discusses plant recovery from the Unit 1 event. After investigation team review of plant data, it was determined that the RCIC system performed appropriately (i.e., was not required to start) in response to the reactor coolant level transient. Reactor coolant level decreased to near, but did not exceed, the RCIC System actuation instrumentation setpoint. The investigation team determined that the direct cause of the RPS/Main Turbine trip is the shorting to ground of one phase of the Unit 1 Main Generator No-Load Disconnect Switch, which electrically connects the generator to the Main Transformer. The switch grounding caused a generator ground fault that resulted in a backup generator lockout and fast closure of the turbine control valves. Fast closure of the turbine control valves provided the trip input to the RPS. Unit 1 is now making preparations to go to mode 4, cold shutdown. Unit 1 and Unit 2 are currently in Technical Specification (TS) Limiting Condition for Operation (LCO) 3.8.1 'AC Sources - Operating,' for one of two required Unit 1 offsite AC circuits inoperable. The resident inspector has been notified. CORRECTIVE ACTIONS Further testing of RCIC System Low Reactor Level actuation instrumentation will be performed to further verify satisfactory RCIC System performance during the event. Activities are in progress to address the damaged No-Load Disconnect Switch and to determine the cause of the switch failure. Notified the R2DO (Fredrickson). | Feedwater High Pressure Coolant Injection Reactor Protection System Primary Containment Isolation System Main Transformer Reactor Recirculation Pump Reactor Building Ventilation Safety Relief Valve Control Rod | 05000325/LER-2005-005 |
| ENS 41582 | 9 April 2005 05:50:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation | Reactor Scrammed from 61% Power Due to Low Reactor Water Level | On 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. | Secondary containment High Pressure Coolant Injection Reactor Protection System Reactor Core Isolation Cooling Shutdown Cooling Reactor Recirculation Pump Reactor Water Cleanup Control Rod | 05000324/LER-2005-002 |
| ENS 40297 | 4 November 2003 22:32:00 | 10 CFR 50.72(b)(2)(iv)(B), RPS System Actuation 10 CFR 50.72(b)(3)(iv)(A), System Actuation 10 CFR 50.72(b)(2)(iv)(A), System Actuation - ECCS Discharge | Brunswick Unit 2 Scram Due to Generator/Turbine Trip | On November 4, 2003, at approximately 1732 hours, Unit 2 received a generator/turbine trip due to loss of excitation, which resulted in a Reactor Protection System (RPS) trip. Plant response to the reactor shutdown resulted in High Pressure Coolant Injection and Reactor Core Isolation Cooling (RCIC) system actuations on low reactor coolant level. Additionally, primary containment isolation system actuation signals for valve groups 1, 2, 3 , 6, 8, and 10 were received and the valves closed as required. All four emergency diesel generators automatically started, but did not load because power was never lost to the emergency buses. The loss of power from the generator trip resulted in reactor building ventilation isolation and automatic start of both trains of the Standby Gas Treatment (SBGT) system. SBGT Train A immediately tripped, but was successfully placed in service. All control rods fully inserted into the core. At approximately 1857 hours, another RPS trip was received due to low reactor coolant level while cycling Safety Relief Valves; however, all control rods were already inserted. The safety significance of this event is considered to be minimal. The plant responded as designed to the transient with the exception of the SBGT Train A initial starting issue. An event investigation team has been assembled to determine the cause of the event. Plant response to the event is being evaluated and identified issues will be addressed prior to plant restart. The licensee reported that the station electrical grid is normal and the emergency diesel generators have been shutdown and returned to standby status. The current plant conditions are 550 psi, 496 degrees F with RCIC operating to maintain reactor water level. The main condenser is available and being used to dump steam from the reactor. Two safety relief valves opened during the transient and reclosed as expected. The reactor water level decreased to minimum of approximately 90 inches during the transient. The cause of main steam isolation valve (MSIV) closure is under investigation. The licensee also reported that this event caused a Group 6 Isolation, reactor building ventilation isolation, and standby gas treatment actuation signal on Brunswick Unit 1. These Unit 1 systems have been returned to a normal configuration. The licensee has notified the NRC Resident Inspector. | High Pressure Coolant Injection Reactor Protection System Emergency Diesel Generator Main Steam Isolation Valve Primary Containment Isolation System Reactor Core Isolation Cooling Reactor Building Ventilation Safety Relief Valve Main Condenser Control Rod | 05000324/LER-2003-004 |