| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
04/07/2005 - 04/08/2005 ** 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 THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!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: 04/07/2005 | 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).
* * * UPDATE ON 04/07/05 @ 0725 BY COY BLAIR TO CHAUNCEY GOULD * * *
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. |
| Power Reactor | Event Number: 41533 | Facility: SURRY
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: JAMES SHELL
HQ OPS Officer: JOHN KNOKE | Notification Date: 03/28/2005
Notification Time: 14:21 [ET]
Event Date: 03/28/2005
Event Time: 14:21 [EST]
Last Update Date: 04/07/2005 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
DAVID AYRES (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | Y | 100 | Power Operation | 100 | Power Operation | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text PLANNED OUTAGE OF SPDS DUE TO SYSTEM UPGRADE
The licensee faxed the following information:
"On Tuesday, March 29, 2005, Surry Power Station will remove a portion of the Emergency Response Facilities Computer System (ERFCS) in the Main Control Room (MCR) for planned upgrades to a new system for increased reliability and improved human interface design. For approximately 1 week, there will not be a qualified Safety Parameter Display System (SPDS) available in the MCR.
"During the time period that SPDS is unavailable in the MCR while the upgrade is occurring, the replacement SPDS system is available in the MCR on the Plant Computer System (PCS), but final testing will not qualify the software until completion of this work. There will also remain one ERFCS terminal operable in both the Technical Support Center (TSC) and Local Emergency Operations Facility (LEOF) with operable SPDS throughout the duration of this work. The Emergency Response Data System (ERDS) will remain available from those terminals for the duration of this planned work and normal data transmission capability will remain in the event of an emergency.
"Since SPDS will be out of service for greater than 1 hour, this report is being submitted in accordance with the guidance in 10CFR50.72(b)(3)(xiii) and NUREG-1022."
The licensee has notified the NRC Resident Inspector.
* * * UPDATE PROVIDED TO NRC (HUFFMAN) BY DEANE ON 1827 EDT ON 4/7/05 * * *
The following information was provided by the licensee via facsimile:
"This report is an update to previous report EN #41533, dated 03/28/05, regarding the availability of the Emergency Response Facilities Computer System (ERFCS) and Safety Parameter Display System (SPDS) in the main control room for Surry Power Station.
"The ERFCS and SPDS modifications and testing were completed at 1630 hours on 04/07/05. These functions have been transferred to the Surry Plant Computer System.
"Senior NRC resident, Norm Garrett was notified via alphanumeric page. Notification was made to NRC duty officer Bill Huffman."
R2DO (Rogers) notified. |
| Hospital | Event Number: 41561 | Rep Org: LANCASTER GENERAL HOSPITAL
Licensee: LANCASTER GENERAL HOSPITAL
Region: 1
City: LANCASTER State: PA
County: LANCASTER
License #: 37-11866-04
Agreement: N
Docket:
NRC Notified By: TONY MONTAGNESE
HQ OPS Officer: HOWIE CROUCH | Notification Date: 04/04/2005
Notification Time: 15:01 [ET]
Event Date: 09/30/2003
Event Time: [EST]
Last Update Date: 04/07/2005 | Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(b) - PATIENT INTERVENTION DAMAGE
21.21 - UNSPECIFIED PARAGRAPH
| Person (Organization):
JAMES NOGGLE (R1)
JOHN HICKEY (NMSS)
| Event Text MEDICAL EVENT NOTIFICATION
In September, 2003, a patient at Lancaster General Hospital's Health Campus was undergoing Co-60 gamma knife treatment. In preparation for the treatment, the patient was immobilized to ensure that the gamma knife was accurately aimed at the treatment site. During the treatment, the patient became uncomfortable and asked to move. He was told to move only his legs, but he shifted his body. As a result, the patient shifted 7cm away from the gamma knife. Licensee was unable to calculate dose to unintended site. There were no adverse effects to the patient.
* * * UPDATE FROM LICENSEE (MONTAGNES) TO NRC (HUFFMAN) AT 1048 EDT ON 4/07/05 * * *
The following additional details were provided on the event. The licensee also is reporting this as a 10 CRF Part 21 report.
The Licensee reported and incident involving component of a Leksell Model 23004 Type B Gamma Knife unit (Elekta AB, Stockholm, Sweden). The specific components involved are the "z-bars". During the treatment with this device, a patient's head is secured into a head frame in preparation for irradiation treatment of deep head disease. This head frame is then also secured to a helmet and couch assembly to ensure immobilization during treatment. In, order to localize the therapeutic radiation beam to a precise location, the frame is able to be adjusted in the X, Y, or Z directions. In the "Z" direction, there are metal bars that are adjusted to a desired position, then locked into place by the treatment staff with a number of screens.
During the case which lead to the incident, a large patient is believed to have made an exaggerated movement during treatment, against staff instructions, as a means of alleviating some discomfort. At the conclusion of the treatment, the staff noted that the patient's position - as indicated by the z-bars - had slipped caudally by approximately 7 cm.
The incident occurred on September 30, 2003 in the Lancaster General Gamma Knife Center, 2102 Harrisburg Pike, Lancaster, Pennsylvania, 17603. Although this original incident occurred on September 30, 2003, this report is only being filed now as a result of an investigation by the NRC that concluded on April 5, 2005.
R1DO (Noggle) and NMSS (Essig) notified. |
| Hospital | Event Number: 41572 | Rep Org: UNIVERSITY OF VIRGINIA
Licensee: UNIVERSITY OF VIRGINIA
Region: 1
City: CHARLOTTESVILLE State: VA
County:
License #: 45-00034-26
Agreement: N
Docket: 03000329
NRC Notified By: RALPH ALLEN
HQ OPS Officer: STEVE SANDIN | Notification Date: 04/06/2005
Notification Time: 16:59 [ET]
Event Date: 04/06/2005
Event Time: [EST]
Last Update Date: 04/07/2005 | Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(a)(2) - DOSE > SPECIFIED EFF LIMITS
| Person (Organization):
JAMES NOGGLE (R1)
MELVYN LEACH (NMSS)
| Event Text MEDICAL EVENT INVOLVING ADMINISTRATION OF WRONG DIAGNOSTIC TEST
On the morning of 4/6, the Resident Physician reviewed the prescribing Physician's order for administration of a brain scan diagnostic test to image a tumor and instructed the technician to perform a "standard" brain scan which images blood flow. The Technician administered 30 mCi Tc-99m as instructed rather than the 3 mCi Thallium prescribed. The RSO noted that the test performed would result in a total dose of 3.22 mGy and a urinary bladder wall dose of 81 mGy (information from package insert). The RSO does not believe there will be any adverse consequences to the patient in that this was a diagnostic test. The error was identified by the Director of Nuclear Medicine during review. The patient had not been informed as of the time of this report. The patient will be rescheduled for the appropriate diagnostic test after elimination and decay of the Tc-99m.
* * * UPDATE FROM LICENSEE (STEVA) TO NRC (HUFFMAN) AT 1251 EDT ON 4/07/05 * * *
The license stated that the patient involved was a 66 year old female. The licensee does not consider that this event meets the criteria of a medical event.
R1DO (Noggle) and NMSS (Essig) notified. | |