| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
03/31/2005 - 04/01/2005 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 41382 | Facility: COOPER
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: ANDREW OHRABLO
HQ OPS Officer: MIKE RIPLEY | Notification Date: 02/07/2005
Notification Time: 22:11 [ET]
Event Date: 02/07/2005
Event Time: 15:58 [CST]
Last Update Date: 03/31/2005 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
BLAIR SPITZBERG (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | Event Text RESIDUAL HEAT REMOVAL SYSTEM INOPERABLE DUE TO EMERGENCY DIESEL GENERATOR TRIP DURING TESTING
"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.
* * * UPDATE FROM C. BLAIR TO M. RIPLEY 1548 EST 03/08/05 * * *
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)
* * * RETRACTION FROM COY BLAIR TO MARK ABRAMOVITZ 3/31/2005 AT 14:40 * * *
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). |
| General Information or Other | Event Number: 41542 | Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: LOWELL GENERAL HOSPITAL
Region: 1
City: LOWELL State: MA
County:
License #: 44-0060
Agreement: Y
Docket: 02-5396
NRC Notified By: TONY CARPENITO
HQ OPS Officer: JOHN KNOKE | Notification Date: 03/29/2005
Notification Time: 12:46 [ET]
Event Date: 01/17/2005
Event Time: 12:00 [EST]
Last Update Date: 03/29/2005 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
JAMES TRAPP (R1)
TOM ESSIG (NMSS)
| Event Text MEDICAL MISADMINISTRATION - WRONG TREATMENT SITE
The State provided the following NMED information:
"On 1/18/05, the licensee reported to the Agency a teletherapy misadministration that occurred on 1/17/05. The situation was described as "the treatment field was displaced by six cm when the therapist set the central axis of the field at the tattoo marking the bottom of the field. An area of 15 cm x 6 cm (treatment site) received dose when the plan called for that tissue to receive no dose. The dose was one day's treatment of 180 cGy. The treatment monitor units and energy were correct, the error was entirely an issue of field placement."
The Regulation Code cited is 105 CMR 120.502 (4) (a). The intended dose to the patient was 180 cGy, however, only 90 cGy was given. Per the treating physican, no corrective action is needed. The hospital retrained staff to ensure confirming proper positioning of therapy prior to treatment. |
| Power Reactor | Event Number: 41552 | Facility: TURKEY POINT
Region: 2 State: FL
Unit: [3] [4] [ ]
RX Type: [3] W-3-LP,[4] W-3-LP
NRC Notified By: DANIEL EDDINGER
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 03/31/2005
Notification Time: 13:00 [ET]
Event Date: 03/30/2005
Event Time: 15:30 [EST]
Last Update Date: 03/31/2005 | Emergency Class: NON EMERGENCY
10 CFR Section:
26.73 - FITNESS FOR DUTY
| Person (Organization):
MALCOLM WIDMANN (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | | 4 | N | Y | 60 | Power Operation | 60 | Power Operation | Event Text TWO CONTRACT SUPERVISORS TESTED POSITIVE DURING A RANDOM DRUG TEST
Two contract supervisors had a positive test during a random drug sampling. Both supervisors have had their plant access terminated and were escorted offsite. A review of the work performed by both individuals is in progress. Contact the Headquarters Operations Officer for additional details.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 41555 | Facility: FITZPATRICK
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: ANDY HALLIDAY
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 03/31/2005
Notification Time: 14:29 [ET]
Event Date: 02/14/2005
Event Time: 07:18 [EST]
Last Update Date: 03/31/2005 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
JAMES TRAPP (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text INADVERTANT SAFETY SYSTEM ACTUATION DURING MAINTENANCE
The following information was provided by the licensee via fax (licensee text in quotes):
"At 0718 on February 14, 2005, with the James A. FitzPatrick Nuclear Power Plant operating at 100% reactor power, the High Voltage Power Supply (HVPS) for Reactor Building (RB) Radiation Monitor 17RM-452A was inadvertently de-energized.
"At 0724 the HVPS was re-energized and the technician began to adjust the output of the high voltage power supply. This adjustment resulted in an unanticipated voltage spike in the radiation monitor circuit. The voltage spike caused the radiation monitor to see a false Hi-Hi signal resulting in an actuation of the RB ventilation system isolation logic, start-up of the 'A' Standby Gas Treatment (SBGT) System, and a Half PCIS Group II Isolation. AOP-15, Isolation Verification and Recovery, was entered and at 0728 an 'A' train Group II PCIS Isolation was verified for that Channel.
"At 0750 AOP-15 was exited and plant systems were restored from the Half Group II PCIS Isolation.
"The reactor remained at 100% power throughout the event, and short-term LCOs were entered for equipment isolated as a result of the isolation, as applicable. While the Hi-Hi signal was falsely generated by the voltage spike encountered when re-energizing the HVPS the actuation logic functioned properly and the plant equipment responded as designed. There were no equipment failures associated with this event and neither plant operation nor the health and safety of the public were affected by this event.
"The condition meets the reporting criteria of 10 CFR 50.73 (a)(2)(iv)(a) because the invalid RB Radiation Monitor Hi - Hi signal resulted in a general containment isolation signal affecting containment isolation valves in more than one system ('A' SBGT, 'A' H2/02 Exosensors, 'A' Drywell Cam, 'A' Containment Atmosphere Dilution (CAD), 'A' PCP Vent and Purge). Since the signal was invalid this event meets the criteria in 10 CFR 50.73 (a)(1) for being reported as a 60-day phone call rather than as an LER.
"The event has been entered into the corrective action program and the resident inspector has been briefed.
"The apparent cause evaluation identified 1) a failure on the part of the technicians to plan for or provide a barrier to prevent the inadvertent actuation of the HVPS switch resulting in the initial deenergization of the radiation monitor circuit and then 2) failure to recognize that the restoration of the HVPS switch after inadvertent actuation should have been accomplished by entering the site work process and obtaining a work order which would have allowed for proper review and incorporation of barriers to minimize the probability of inadvertent isolations."
The licensee notified the NRC Resident Inspector and the State. | |