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| Entered date | Event description | |
|---|---|---|
| ENS 47690 | 23 February 2012 02:55:00 | 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). |
| ENS 47687 | 22 February 2012 23:36:00 | 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). |
| ENS 44453 | 30 August 2008 17:47:00 | 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. |
| ENS 44425 | 19 August 2008 17:27:00 | On August 18, 2008, during biennial (i.e., every two years) non-Technical Specification related testing, Emergency Diesel Generator (EDG) No. 4 failed to start from the EDG local control panel. This occurred with the EDG aligned for Alternate Safe Shutdown (ASSD) response versus its normal alignment. The purpose of the testing is to confirm that the EDG control logic is isolated from the control room and only operable via local controls. Based on troubleshooting activities, it was discovered that this failure to start was due to improper wiring in the circuitry associated with the EDG lockout relay. The improper wiring was introduced by a modification performed on all four EDGs in 2007. At 1110 hours Eastern Daylight Time (EDT) on August 19, 2008, it was concluded that this condition impacted the ability of EDG Nos. 2, 3, and 4 to perform their intended ASSD function. The 2007, modification affected the termination point for the power to the EDG lockout relay reset coil. Per the modification, this termination point was incorrectly connected to the downstream side of the alternate shutdown isolation switch which results in a loss of power to the lockout reset when the switch is placed in local control. Therefore, in the unlikely event of a fire, an induced failure could potentially initiate a lockout signal which could not be reset from the EDG local control panel. Although the modification was installed on all four EDGs, only the local control of EDGs 2, 3, and 4 is credited in the safe shutdown analysis. This condition does not affect the Technical Specification operability of the EDGs and they remain fully capable of performing their intended design basis accident response functions. The NRC Resident Inspector has been notified. Compensatory fire watches have been established and an engineering change package to correct the condition is being developed. |