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| Entered date | Event description | |
|---|---|---|
| ENS 45584 | 21 December 2009 20:17:00 | High Pressure Coolant Injection (HPCI) system was declared inoperable due to failure of the turbine governor valve to open with the Auxiliary Oil Pump (AOP) running. The AOP had been started to obtain oil samples for routine maintenance. The turbine was not expected to roll since there was no initiation signal, but the HPCI governor valve was expected to open to try to raise HPCI speed. Failure of the governor valve to open would prevent HPCI from achieving rated system flow and pressure. HPCI is also considered unavailable. All remaining ECCS systems and RCIC are operable. Cause of the condition is currently unknown. Troubleshooting activities are being planned. HPCI is a single train system. This report is submitted in accordance with 10 CFR 50.72(b)(3)(v) as a condition that at time of discovery could prevent the fulfillment of the safety function of an SSC (Structure, System and Components) needed to mitigate the consequence of an accident. The NRC Resident Inspector has been notified of this event. |
| ENS 42711 | 17 July 2006 21:38:00 | During review of instrumentation calibration, it was determined that a faulty flow instrument was used for determination of CREFS (Control Room Emergency Filtration System) flow. Due to the error of the flow instrumentation previous test results were deemed to be high outside acceptance criteria. These tests were performed in October of 2005. The CREFS was declared inoperable and preparation was made to rerun the surveillance. At 1900 the surveillance was completed. The results of the surveillance determined that CREFS flow was below the allowable limit. The minimum flow for CREFS is a requirement for post DBA dose calculation for Control Room personnel. With minimum flow not maintained, the Control Room post accident dose calculations may be violated. CREFS remains inoperable pending adjustments of ventilation dampers and successful performance of surveillance testing. Resident inspector has been notified. The licensee has entered Technical Specification LCO 3.7.4, Condition A, which is a 7 day LCO.
The licensee is retracting this report based on the following: This notification is being made to retract Event Notification #42711 which reported a loss of safety function due to the unplanned inoperability of the Control Room Emergency Filtration System (CREFS). CREFS was declared inoperable due to flow below the acceptance criteria required per surveillance testing. Upon further evaluation, CNS has determined that CREFS satisfies the Technical Specifications (TS) surveillance requirements and that the safety function was not lost. The surveillance procedure acceptance criteria for CREFS minimum flow is not required by the TS and is not a direct indicator of system operability. The procedure has been revised to remove the minimum flow acceptance criteria. It has been determined that CREFS, at the flow conditions documented on July 17, 2006 and during a past surveillance in 2005, satisfied TS and USAR requirements and was capable of performing its safety function. The licensee will inform the NRC Resident Inspector. Notified R4DO (Farnholtz). |
| ENS 42375 | 26 February 2006 04:32:00 | This report is being made pursuant to 10CFR50.72 (b) (2) (iv) (B) actuation of RPS when the reactor is critical (4 hour report) and 10CFR50.72 (b) (3) (iv) (A) actuation of PCIS group 2 (8 hour report). At 0250 on 2/26/06, Central Standard Time, Cooper Nuclear Station was manually scrammed due to main turbine reheat valve remaining closed following testing concurrent with a high level in the moisture separator. Alarm card procedure for this condition required removing the turbine from service. Subsequent to the scram, reactor vessel level lowered to minus 20 inches wide range indication. This corresponds to approximately 140 inches above the top of the fuel. A primary containment isolation system (PCIS) group 2 isolation occurred as expected due to level transient. All automatic actions occurred as expected. The NRC Senior Resident Inspector has been informed of the event. The reactor scram was uncomplicated, all control rods fully inserted, and no relief or safety valves lifted. The electrical lineup is normal, and the decay heat path is to the main condenser through the turbine bypass valves. |
| ENS 41588 | 11 April 2005 19:11:00 | At approximately 1604 CDT on 4/11/05, Cooper Nuclear Station (CNS) was informed by Emergency Preparedness officials of Atchison County that the radio transmission tower that provides signals to activate tone alert radios within the 10-mile Emergency Planning Zone (EPZ) was not functioning. Tone alert radio is relied on by approximately 650 households in the EPZ who are not in audible range of sirens for notification of an emergency at CNS. Based on a total EPZ population currently estimated at 4600 persons (who are alerted principally by sirens and the tone alert radio), this is considered to be a major loss of the Public Prompt Notification system capability, and is reportable under 10CFR50.72(b)(3)(xiii). Investigations are ongoing, a repair technician has been dispatched to the tower. During the interim, compensatory measures have been verified to be in place via state and local emergency planning officials for backup route alerting for personnel within the 10-mile EPZ. NRC Resident Inspector was notified.
22:53 on 4/11/05- CNS was notified by the NWS (National Weather Service) that the transmitter had been returned to high power service at 19:35. The cause of the high power transmitter loss was a lightning strike to the transmitter. The strike did not result in any equipment damage, however the transmitter had to be manually reset. 22:55 on 4/11/05- CNS contacted Nemaha and Richardson County Sheriff's Departments and the Atchison County 911 center to notify them that the transmitter/tower had been returned to service. Operation of equipment was verified. Transmitter tower and tone alert radios have been returned to service. NRC Resident Inspector was notified by the licensee. R4DO (Nease) was notified. |
| ENS 41382 | 7 February 2005 22:11:00 | 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 hours. 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 hour 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 hours) 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 hours for the pool temperature to reach 150 degrees Fahrenheit and 94 hours 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. |
| ENS 41356 | 25 January 2005 23:36:00 | This report is being made pursuant to 10 CFR 50.72(b)(3)(xii), 'Any event requiring the transport of a radioactively contaminated person to an offsite medical facility for treatment.' At 20:58, Central Standard Time, Cooper Control Room was notified of an individual that was overheated and not feeling well. Cooper is currently in a refueling outage and this individual was working inside the drywell. Initial radiological surveys were started at the scene but were not completed until transportation to medical facilities was commenced. Individual was transported in Cooper Station ambulance. At the time of transport the individual was listed as potentially contaminated. Radiological surveys were completed prior to arrival at the medical facility. These radiological surveys determined that the individual was not contaminated. NRC Resident has been notified. |
| ENS 40435 | 10 January 2004 13:48:00 | This Notification is being made pursuant to 10 CFR 50.72 (b)(3)(v)(D). On January 8, 2004 at 17:35 CST with the Control Room Emergency Filtration System (CREFS) inoperable for planned maintenance, CREFS failed post work testing due to system flow being higher than that allowed by technical specifications. It was originally believed that the failure was connected with the planned maintenance because the system flow rate was initially within the specifications during surveillance testing the previous day. Following system troubleshooting, at 05:00 CST on January 10, 2004, it was determined that the cause of the high flow rates was indeterminate. Based on this evaluation, it could not be determined whether the CREFS failure was independent of planned maintenance or not. Additionally, it is not immediately clear if the high flow condition would result in a loss of safety function. It was therefore conservatively determined that the condition is reportable per 10 CFR 50.72 (b)(3)(v)(D) as a failure of a single train system which could have prevented the fulfillment of a safety function that is needed to mitigate the consequences of an accident. CREFS flow rates are currently within specifications, but the system will remain inoperable until the proper surveillances have been completed. The senior NRC Resident has been notified.
The licensee retracted this event and provided the following information: On January 10, 2004, Cooper Nuclear Station (CNS) made an eight hour event notification report to the NRC pursuant to 10CFR50.72.(b)(3)(v)(D), failure of a single train system which could have prevented the fulfillment of a safety function that is needed to mitigate the consequences of an accident. In particular, testing to restore the Control Room Emergency Filter System (CREFS) to operable status following planned maintenance indicated the flow rate was greater than that allowed by Technical Specifications. It was not known whether the high flow condition would result in a loss of safety function. Flow rate was restored to within Technical Specification Limits and CREFS was returned to operable status on January 11, 2004 at 1803 hours Subsequent evaluation determined: 1. the high flow condition did not impact the system safety functions to maintain the Control Room at a positive pressure with respect to adjoining areas or to isolate the outside air intake on relevant Group isolation signals. 2. the CREFS safety function to limit the radiation exposure to Control Room personnel during any one of the postulated design basis events to within regulatory limits (10CFR50, Appendix A, GDC 19) was maintained during the high flow condition. 3. the design functions of CREFS for the CNS toxic hazards assessment and Fire Protection Program requirements were not affected by the CREFS high flow. The evaluation concluded the safety function of the CREFS during the high flow condition was maintained. Therefore, CNS is retracting this 10CFR50.72 Event Notification. The licensee will notify the NRC Resident Inspector. Notified R4DO (Jeffery Clark) |