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| Start date | Reporting criterion | Title | Event description | System | LER | |
|---|---|---|---|---|---|---|
| ENS 56928 | 18 January 2024 20:04:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Residual Heat Removal Degraded Due to Service Water Leakage | The following information was provided by the licensee via email: On January 18, 2024, at 0030 PST, diesel generator 2 (DG2) was shut down following a monthly surveillance run. Subsequently, a leak was discovered in the DG2 building. Service water pump '1B' was secured at 0117, effectively stopping the leak. The leak was determined to be service water coming from a diesel generator mixed air cooling coil. Service water system 'B' and DG2 were subsequently declared inoperable at 0135. After discussion with engineering, it was identified that the amount of service water leakage from the cooling coil was assumed to be greater than the leakage allowed by the calculation to assure adequate water in the ultimate heat sink to meet the required mission time of 30 days. At 1204, it was determined that entry into Technical Specification 3.7.1 condition D was warranted since the assumed leakage from the cooling coil could exceed the calculated allowed value. At 1238, the control power fuses for service water pump '1B' were removed. DG2 and service water system 'B' were declared unavailable, and the technical specification condition for the inoperable ultimate heat sink was exited. With the control power fuses removed, the pump is kept from auto starting, effectively preventing the leak and ensuring the safety function of the ultimate heat sink is maintained while the cooling coil is repaired or replaced. Due to the leakage assumed greater than the calculated allowable value this condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition and per 10 CFR 50.72(b)(3)(v)(B) as an event or condition that could have prevented fulfillment of the safety function of structures or systems that are needed to remove residual heat. There was no impact to the health and safety of the public. The NRC Resident has been notified.
The following information was provided by the licensee via email: On January 18, 2024 at 2138 EST, Columbia Generating Station notified the NRC under 10 CFR 50.72(b)(3)(ii)(B) of an unanalyzed condition on the available capacity of the ultimate heat sink (UHS) and under 10 CFR 50.72(b)(3)(v)(B) of an event or condition that could have prevented fulfillment of the safety function of structures or systems needed to remove residual heat. On January 18, 2024, following monthly surveillance of the diesel generator DG2, a DG2 room cooler flow alarm was received at 0115. A leak was discovered in the diesel mixed air (DMA) air handler unit. Service Water Pump '1B' was secured and the leakage was stopped at 0117. The service water system 'B' and diesel generator system 'B' were declared inoperable at 0135. The leak was assumed to be greater than that allowed to ensure adequate water in the UHS required to meet the 30-day mission time, and the UHS was declared inoperable at 1204. Control power fuses for the service water pump '1B' were removed to fully eliminate the leakage path from the cooler, and the UHS was declared operable at 1238. Following the event, engineering performed an analysis based on the size and location of the leak, and concluded it would have taken 1.4 days to deplete the available excess water in the UHS to below the minimum technical specification required water level of the spray pond. Operations were able to secure the service water subsystem of the UHS prior to exceeding the volumetric margins in the spray ponds to ensure the 30-day mission time was met. The condition did not represent a safety significant unanalyzed condition nor a loss of safety function. The NRC Resident Inspector has been notified. Notified R4DO (Gepford). | Service water Residual Heat Removal Spray Pond | |
| ENS 51201 | 6 July 2015 21:20:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Multiple Spurious Operations Scenario That Could Adversely Impact Post-Fire Safe Shutdown | A recent review of Fire Protection and Post Fire Safe Shutdown (PFSS) Programs at Columbia Generating Station (CGS) identified a potential unanalyzed condition with Multiple Spurious Operation (MSO) Scenario 2x. Review of the circuit design for High Pressure Core Spray (HPCS) HPCS-V-10, HPCS-V-11 and HPCS-V-15 identified that fire-induced circuit failure (hot shorts) on the OPEN function control circuits for each valve would create the flow path to potentially drain inventory from the suppression pool (SP). The normal operation of HPCS-P-3 (keep-fill pump) would allow additional inventory from the SP to be transferred to the CSTs (Condensate Storage Tank). If a fourth hot short is postulated, HPCS-P-1 would transfer inventory from the SP to the CST at a much faster rate. HPCS-V-11 was deactivated on 6/12/2015 due to a maintenance repair issue and will be left in the fully closed position. This plant alignment resolves current concern for MSO scenario 2x as fire-induced circuit damage cannot cause spurious opening of HPCS-V-11. However, with an incomplete analysis for MSO scenario 2x, compliance with PFSS MSO requirements would have been challenged from the completion of the MSO project (October 2012) up to June 2015. CGS is reporting this event as an unanalyzed condition in conformance with 10 CFR 50.72(b)(3)(ii)(B). Further analyses are being implemented to confirm the condition and to develop appropriate remedial actions. The licensee will notify the NRC Resident Inspector. | High Pressure Core Spray | 05000397/LER-2015-006 |
| ENS 49898 | 11 March 2014 17:01:00 | 10 CFR 50.72(b)(3)(ii)(B), Unanalyzed Condition | Postulated Hot Short Fire Event That Could Adversely Impact Safe Shutdown Equipment | An extent of condition review of all unfused ammeters circuits in Direct Current (DC) distribution systems at Columbia Generating Station (Columbia) identified areas in the plant which may be susceptible to secondary fires due to hot shorts from these unfused ammeters. It is postulated that a fire in one fire area can damage these circuits and cause short circuits without protection that would overheat the cables and possibly result in secondary fires in other fire areas where the cables are routed. The secondary fires could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. This condition is reportable as an 8-hour ENS report in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector. | 05000397/LER-2014-002 | |
| ENS 47209 | 28 August 2011 03:21:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Shutdown Cooling Temporarily Isolated | In accordance with 10 CFR 50.72 (b)(3)(v)(B) this event notification is being made to report an event that could have prevented the fulfillment of the safety function to remove residual heat. On August 27, 2011, following refueling, the plant was shutdown in Mode 4 with reactor coolant temperature being maintained 100-120 degrees F. Reactor coolant temperature was being controlled with RHR B operating in the Shutdown Cooling mode. At 2021 (EDT), an equipment failure caused a loss of Reactor Protection System (RPS) B resulting in a closure of RHR-V-9 (RHR Shutdown Cooling Inboard Isolation Valve), interrupting shutdown cooling flow. Operators transferred RPS B to the alternate power supply and restored RHR B to shutdown cooling. Shutdown cooling was out of service for 34 minutes. During the 34 minute interruption of shutdown cooling, reactor coolant temperature rose from 107 degrees F to 111 degrees F. Following restoration of shutdown cooling, reactor coolant temperature was restored. All containment isolations resulting from the loss of RPS B occurred as expected. Further investigation is ongoing to determine the cause of the equipment failure that caused the loss of RPS B. The licensee notified the NRC Resident Inspector. | Reactor Protection System Shutdown Cooling | 05000397/LER-2011-002 |
| ENS 42961 | 3 November 2006 11:09:00 | 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | Procedural Error Caused Residual Heat Removal Shutdown Cooling Inboard Containment Isolation Valve to Close. | This event notification is being made to report an event that could have prevented the fulfillment of the safety function to remove residual heat. On November 3, 2006, the plant was shutdown in Mode 4 (Cold Shutdown) with the reactor vessel level being maintained between 60-80 'inches' and temperature was being maintained 110-120F. Operators were performing a procedure to transfer RPS B to its alternate power supply. During the procedure, RHR-V-9 (RHR shutdown cooling inboard containment isolation valve) closed and shutdown cooling was interrupted. Reactor Recirculation Pump A remained in service providing forced flow through the core. Operators reopened RHR-V-9 and restored shutdown cooling. Reactor temperature reached 148F and vessel level reached 95 'inches' (bottom of main steam lines is 116 inches) while shutdown cooling was not in service. Shutdown cooling was restored in 45 minutes (120 minutes time to boil from 120F) after RHR-V-9 went closed. Reactor temperature and level were restored to their previous operating bands by 0451 PST. Initial indications are that RHR-V-9 closed due to an error in the procedure which failed to remove control power from RHR-V-9 during the RPS transfer. Without removing control power, a motor-closed demand would have been created in the circuit during the transfer and upon restoration of power to the valve motor, the valve would motor close. Further investigation is ongoing. The NRC Resident Inspector was notified of this event by the licensee. | Shutdown Cooling Reactor Recirculation Pump Residual Heat Removal Main Steam Line |