| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
03/01/2004 - 03/02/2004 ** EVENT NUMBERS ** |
 | General Information or Other | Event Number: 40551 | Rep Org: CALIFORNIA RADIATION CONTROL PRGM
Licensee: DEL MONTE FOODS
Region: 4
City: TERMINAL ISLAND State: CA
County:
License #: GL
Agreement: Y
Docket:
NRC Notified By: ROBERT GREGER
HQ OPS Officer: BILL GOTT | Notification Date: 02/26/2004
Notification Time: 21:02 [ET]
Event Date: 02/26/2004
Event Time: [PST]
Last Update Date: 02/26/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
GARY SANBORN (R4)
E. WILLIAM BRACH (NMSS)
| Event Text AGREEMENT STATE REPORT OF A LOST GAUGE
"Del Monte Foods, a general licensee has been unable to locate a fixed gauge containing 100 milliCuries of Americium-241. The unit was one of 4 generally licensed gauges used as fill/level detectors at the Del Monte facility at Terminal Island. Del Monte had the 4 lines removed by contractors and after the completion of the project Del Monte realized one gauge remains unaccounted for. They have searched their facility and checked with all contractors and no information was obtained regarding the whereabouts of the gauge. The licensee has been requested to provide [the California Radiation Control Program Office] additional information regarding the dates/times involved, the names of the contractors who removed the lines, and the names of individuals who may have purchased the equipment." |
| Power Reactor | Event Number: 40557 | Facility: OCONEE
Region: 2 State: SC
Unit: [ ] [2] [ ]
RX Type: [1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP
NRC Notified By: RANDY TODD
HQ OPS Officer: CHAUNCEY GOULD | Notification Date: 03/01/2004
Notification Time: 04:17 [ET]
Event Date: 02/29/2004
Event Time: 20:11 [EST]
Last Update Date: 03/01/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
CHARLES R. OGLE (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 96 | Power Operation | 96 | Power Operation | Event Text UNIT 2 AUTOMATIC FEEDWATER ISOLATION SYSTEM (AFIS) DECLARED INOPERABLE
The licensee reported that both digital trains of the automatic feedwater isolation system were declared inoperable when one channel's input failed to zero and about the same time the second channel showed a shift in readouts which was fairly significant that they also declared it inoperable. These two channels have a shared neutral bus and it is believed that there may be a loose connection somewhere in the circuit. They are currently troubleshooting the problem. Plant will initiate a unit shutdown if the problem cannot be corrected by 0600 on 03/01/04 (12-hour LCO A/S).
The NRC Resident Inspector will be notified. |
| Power Reactor | Event Number: 40558 | Facility: FARLEY
Region: 2 State: AL
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: WAYNE VAN LANDINGHAM
HQ OPS Officer: STEVE SANDIN | Notification Date: 03/01/2004
Notification Time: 07:23 [ET]
Event Date: 03/01/2004
Event Time: 05:22 [CST]
Last Update Date: 03/01/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
CHARLES R. OGLE (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | A/R | Y | 100 | Power Operation | 0 | Hot Standby | Event Text UNIT 1 EXPERIENCED A REACTOR TRIP ON TURBINE TRIP DUE TO A FEEDWATER SYSTEM PROBLEM
"Reactor trip Unit 1 as a result of Turbine Trip from Hi-Hi Steam Generator level '1C' Steam Generator. Steam Generator Feed pump suction pressure initially dropped, deviation alarms were received on 'A' and 'C' Steam Generator levels. Third condensate [pump] was started, increasing feed to S/Gs. '1C' S/G went high and caused turbine trip/reactor trip. Autostart of 'A' and 'B' motor driven feed pumps occurred following the trip."
All rods fully inserted following the reactor trip. Both motor-driven auxiliary feedwater pumps are currently supplying the steam generators with the steam dump system in-service to remove decay heat via the main condenser. Offsite power is stable with the EDGs in standby, if needed. All systems functioned as required. The licensee is conducting an investigation to determine the root cause.
The licensee will inform the NRC Resident Inspector. |
| Power Reactor | Event Number: 40559 | Facility: BRAIDWOOD
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: LAWRENCE BROOKS
HQ OPS Officer: ARLON COSTA | Notification Date: 03/01/2004
Notification Time: 19:05 [ET]
Event Date: 03/01/2004
Event Time: 16:00 [CST]
Last Update Date: 03/01/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
| Person (Organization):
JAMES CREED (R3)
BILL BATEMAN (NRR)
|
| 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 24-HOUR CONDITION OF LICENSE REPORT INVOLVING POTENTIAL VIOLATION OF MAXIMUM POWER LEVEL
"This 24-hour report is being made as required by Braidwood Unit 1 License Condition 2.G as a potential violation of the maximum power level 3586.6 MWt as stated in Unit 1 License Condition 2.C(1).
"Braidwood is conservatively reporting an overpower condition on Unit 1 due to the implementation of an ultrasonic flow measurement system. Unit 1 was potentially overpowered a maximum of 1.07%. This value is based upon the maximum ultrasonic flow meter correction factor used for Unit 1 during the period between June 1999 and September 2003.
"During post installation testing of the permanent ultrasonic flow measurement system, discrepancies in the ultrasonic flow measurement system were identified that could have resulted in an overpower condition of Unit 2 during the previous use of the ultrasonic system (i.e., during the period between June 1999 and September 2003). The ultrasonic flow measurement system was used to correct venturi feedwater flow measurements. The corrected feedwater flow measurements were then used is the calorimetric calculation for reactor power.
"Currently Braidwood Unit 1 and Unit 2 are controlling power level based only on the venturi feedwater flow indication."
The licensee notified the NRC Resident Inspector. | |