ENS 41382
ENS Event | |
|---|---|
21:58 Feb 7, 2005 | |
| Title | |
| Event Description | This report is being made pursuant to 10CFR50.72(b)(3)(v)(B) 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat;'
This report is being made due to a trip of the Emergency Diesel Generator during testing that resulted in the RHR loops potentially becoming depressurized. This has the potential to render all RHR Shutdown Cooling unavailable and prevent the removal of decay heat. Sequence of events (all times CST): At 12:00 [02/07/05], Shutdown Cooling was removed from service to prepare for Sequential Load testing of DG #1. This was a planned evolution. At this time decay heat was being removed by the fuel pool cooling system with 2 fuel pool cooling pumps and 2 fuel pool cooling heat exchangers. Time to boil was calculated to be 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br />. At 15:58, the Sequential Load Test commenced on the inoperable DG. The DG came up to speed and sequenced on the initial loads (RHR pumps, a CS pump and a SW pump). Shortly into the sequencing of the DG, the DG tripped due to a blown fuse in the DG control circuit. Sequential loading was not completed. The trip occurred between 13 seconds and 20 seconds of the sequential load. This resulted in the initial loads losing power. Procedurally, the minimum flow valves for the RHR and CS pumps were being remotely opened from the Control Room at the time the DG tripped. This resulted in low-pressure alarms on both RHR systems and one CS system. One fuel pool cooling pump was deenergized, per design, during the sequential load test. Both fuel pool cooling heat exchangers remained in service. With these conditions, the fuel pool cooling lineup does not qualify as an alternate decay heat removal method. At 16:04, both RHR loops were declared inoperable due to depressurizing the RHR loops. At 16:02, the tripped fuel pool cooling pump was restored to operation and previous decay heat removal was restored. No unexpected rise in temperature occurred during the time that only 1 fuel pool cooling pump was in operation. This reestablished the fuel pool cooling system as an alternate decay heat removal method. At 19:11, the B loop of RHR was returned to a standby lineup and declared operable. At this time investigation into why the DG fuse blew is ongoing. All indications are that other equipment performed as designed. The licensee notified the NRC Resident Inspector.
The following is a correction to the original report received via facsimile (licensee text in quotes): Instead of the minimum flow valves for RHR and CS being opened, the suppression pool inboard cooling valve for RHR and the test line recirculation valve for CS were being opened. The licensee will notify the NRC Resident Inspector. Notified R4 DO (T. Pruett)
The following information was provided by the licensee (licensee text in quotes): On 2/7/2005 at 1558 CST, Cooper Nuclear Station made an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 50.72 non-emergency notification to the NRC. The report was made pursuant to 10 CFR 50.72(b)(3)(v), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat.' A control power failure during Emergency Diesel Generator #1 (DG) surveillance testing resulted in the loss of the Residual Heat Removal (RHR) pressure maintenance pump. This resulted in the potential de-pressurization and unavailability of all RHR Shutdown Cooling (SDC) and the ability to remove decay heat using RHR. NUREG 1022 Revision 2 defines the safety functions to be considered for Reportability under this section of the rule as being those that are listed in the regulation itself. Thus, the lost safety function being reported was 'remove decay heat'. Plant conditions prior to the testing were: Mode 5 (Refueling) with the Reactor Vessel and Drywell heads removed and reactor water level flooded up and Spent Fuel Pool transfer gates removed. Division II RHR was in service providing SDC for decay heat removal. In preparation for the DG testing and in accordance with Technical Specifications, all RHR SDC was removed from service. With RHR SDC out of service, reactor coolant circulation was verified to be by natural circulation with operators monitoring reactor coolant temperatures once per hour. Alternate decay heat removal was provided by the credited lineup of two Fuel Pool Cooling (FPC) pumps and two FPC heat exchangers. FPC receives cooling water from the Reactor Equipment Cooling System (REC), which in turn is cooled by the Service Water System (SW). During the preparation period (approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) for the DG #1 testing, reactor coolant temperature was allowed to slowly go from 85 degrees Fahrenheit to 90 degrees Fahrenheit. During load sequencing testing of DG #1, the DG tripped due to a control system failure and de-energized the Division I 4160 Volt (V) critical bus. (Note: The bus was previously de-energized for a short period of time as part of the test.) This caused the pump providing pressure maintenance for the RHR to trip potentially depressurizing the RHR loop (Division II) that had been lined up to provide SDC. A conservative decision was made to declare Division II SDC inoperable during the DG trip recovery. If the test had proceeded as planned one RHR pump would have been running in Division I in the test mode (pumping water to the suppression pool). No RHR pumps would have been running in Division II (lined up to allow the Division I test to be conducted). DG #2 remained in normal standby lineup. Division II 4160 V bus was energized supplying power to connected loads. Due to the DG #1 trip the Division I 4160 V bus was deenergized. Shutdown Cooling using RHR could not be placed in service as a result of the test lineup established for DG #1 testing. Reactor coolant circulation was by natural circulation and reactor decay heat removal was by one FPC pump and two FPC heat exchangers. The trip of one FPC pump is expected and verified during this surveillance test. REC was operating with cooling supplied by Division II SW. During the period of time after the DG trip and prior to the restoration of electrical power to the Division I 4160 V bus, coolant circulation continued by natural circulation with one FPC pump and two FPC heat exchangers providing decay heat removal. At approximately the time of the DG trip coolant temperature was 90 degrees Fahrenheit. Just after power was restored coolant temperature was 89 degrees Fahrenheit. Operators had adjusted REC temperatures and flows to provide additional cooling to Fuel Pool Cooling. An additional FPC pump was started to provide a two FPC pump and two FPC heat exchanger lineup for reactor decay heat removal. The small variation in coolant temperature demonstrates that the FPC lineup was adequate to provide decay heat removal. Engineering performed an evaluation to investigate bulk water temperature response to the event with one FPC pump and two heat exchangers supplying cooling with the fuel pool gates removed. The results show extended periods of time for pool heat-up and are considered bounding. It takes 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> for the pool temperature to reach 150 degrees Fahrenheit and 94 hours0.00109 days <br />0.0261 hours <br />1.554233e-4 weeks <br />3.5767e-5 months <br /> for the bulk temperature to reach a maximum value of 182 degrees Fahrenheit. Based on this evaluation CNS concludes the maximum bulk temperature would not exceed 182 degrees Fahrenheit. As discussed above, RHR SDC was removed from service to support Emergency Diesel Generator surveillance testing. While RHR SDC was out of service, reactor coolant circulation was provided by natural circulation. At the same time, the safety function of decay heat removal was provided by Fuel Pool Cooling. Since the decay heat removal safety function was never lost this is not a reportable event. The licensee notified the NRC Resident Inspector. Notified the R4DO (Graves).
The following is a change to paragraphs 3 and 4 of the above retraction statement During sequential load testing of DGI, the normal expected response after loads are sequenced on, is to have an RHR pump in each division recirculating back to the suppression pool via the suppression pool cooling line. This path is established when the respective RHR pump automatically starts. During load sequencing testing of DG # 1, the DG tripped due to a control system failure and de-energized the Division I 4160 Volt (V) critical bus. (Note: The bus was previously de-energized for a short period of time as part of the test.). Due to the timing of the DG failure, both RHR pumps started and both suppression pool cooling valves were opened. Subsequently the DG tripped and the RHR pumps stopped due to no power available. The suppression pool cooling valves were unable to be closed prior to depressurizing both RHR loops. A conservative decision was made to declare Division II SDC inoperable during the DG trip recovery. DG #2 remained in normal standby lineup. Division II 4160 V bus was energized supplying power to connected loads. Reactor coolant circulation was by natural circulation and reactor decay heat removal was by one FPC pump and two FPC heat exchangers. The trip of one FPC pump is expected and verified during this surveillance y test. REC was operating with cooling supplied by Division II SW. The NRC Resident Inspector will be informed. Reg 4 RDO(Linda Howell) was notified. |
| Where | |
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Cooper  Nebraska (NRC Region 4) | |
| Reporting | |
| 10 CFR 50.72(b)(3)(v)(B), Loss of Safety Function - Remove Residual Heat | |
| Time - Person (Reporting Time:+0.22 h0.00917 days <br />0.00131 weeks <br />3.01356e-4 months <br />) | |
| Opened: | Andrew Ohrablo 22:11 Feb 7, 2005 |
| NRC Officer: | Mike Ripley |
| Last Updated: | Apr 7, 2005 |
| 41382 - NRC Website
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