| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
04/20/2004 - 04/21/2004 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 40576 | Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [2] [ ] [ ]
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: DANIEL FORRY
HQ OPS Officer: MIKE RIPLEY | Notification Date: 03/08/2004
Notification Time: 17:04 [ET]
Event Date: 03/08/2004
Event Time: 11:30 [EST]
Last Update Date: 04/20/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
EUGENE COBEY (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text HPCI SYSTEM INOPERABLE DUE TO FAILURE OF TORUS SUCTION VALVE TO FULLY OPEN
"U2 HPCI was declared inoperable to fulfill its safety function to mitigate the consequences of an accident.
"U2 HPCI torus suction valve MO-2-23-058 failed to fully open, during performance of ST-O-023-301-2 'HPCI Pump, Valve, Flow, and Unit Cooler Functional and In-Service Test.' The valve stopped in a mid-position. HPCI flow path from Condensate Storage Tank remains available for HPCI injection. Investigation into cause is continuing."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION PROVIDED AT 1509 ET ON 4/20/04 BY D. FOSS TO J. ROTTON * * *
"The purpose of this notification is to retract a previous report made on 3/8/04 at 1130 hours (EN# 40576). Notification of the event to the NRC was initially made as a result of declaring the Unit 2 High Pressure Coolant Injection (HPCI) system inoperable when unexpected conditions were found during performance of routine surveillance testing of HPCI. Specifically, it was noted that a motor-operated Suppression Pool suction valve for HPCI (MO-58) did not complete its stroke in the open direction during testing. The HPCI system was not operating at the time of the discovery.
"Since the initial report, Engineering has determined that HPCI was capable of performing its safety function. The evaluation has determined that the MO-58 valve was operable for continued operations. The MO-58 stopped in the mid-stroke position due to motor operator torque switch operation. During design events, the torque switch is bypassed and would not have interrupted valve operation. The torque switch is only in the valve logic for remote manual valve operations (e.g. testing). The torque switch was adjusted and HPCI was returned to service on 3/10/04 by approximately 1500 hours [ET].
"During adjustments to the HPCI MO-58 motor operator, the suction source from the Suppression Pool was isolated in accordance with Technical Specifications since MO-58 is considered as a Primary Containment Isolation Valve. This occurred on 3/8/04 by approximately 1340 hours. This action was performed in accordance with station procedures and is considered planned maintenance. Throughout the time period of repairs to the MO-58, HPCI was available for operations with its suction source from the normally aligned Condensate Storage Tank.
"The NRC resident has been informed of the retraction."
Notified R1 DO (D. Silk). |
| General Information or Other | Event Number: 40679 | Rep Org: NV DIV OF RAD HEALTH
Licensee: Nortech
Region: 4
City: North Reno State: NV
County: Washoe
License #: 00-11-0309-01
Agreement: Y
Docket:
NRC Notified By: Stan Marshall
HQ OPS Officer: JAMIE HEISSERER | Notification Date: 04/16/2004
Notification Time: 12:23 [ET]
Event Date: 04/15/2004
Event Time: 16:15 [PDT]
Last Update Date: 04/16/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
DALE POWERS (R4)
ROBERTO TORRES (NMSS)
| Event Text NEVADA AGREEMENT STATE REPORT - LOST PORTABLE NUCLEAR GAUGE
The following information was received on 4/16/04 via facsimile:
At approximately 4:15 p.m. on 4/15/04 the State received a telephone call from the RSO for Nortech - Reno. The RSO reported the loss of a Pacific Nuclear Model MC-1 portable moisture density gauge, S/N is M12044864, which contains 10 millicuries of Cs-137 and 50 millicuries of Am-241/Be.
The operator had the device sitting on the tailgate of his pickup, out of the shipping box, while talking on the telephone. He got in the truck and proceeded to drive to a convenience store several miles away.
Upon getting out of the truck at the convenience store he noticed that the gauge was missing. He retraced his route looking for the device but did not find it. It is not known if the device was locked or not. The individual had been conducting testing, so there is the possibility that it was not locked.
The RSO has had four different individuals retracing the route looking for the device. They reportedly have been looking in alleys, dumpsters etc. but have not located the device.
The licensee has notified the Reno Police and the Washoe County Sheriff. |
| Power Reactor | Event Number: 40686 | 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: ARLON COSTA | Notification Date: 04/20/2004
Notification Time: 10:55 [ET]
Event Date: 03/20/2004
Event Time: 15:59 [EDT]
Last Update Date: 04/20/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
| Person (Organization):
CAUDLE JULIAN (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text INVALID ACTUATION OF THE MOTOR DRIVEN EMERGENCY FEEDWATER PUMPS
"By design, the Steam Generator (SG) Dry-out Protection Circuit actuates whenever two level indications (two of two actuation logic) in either SG are lower than the setpoint for more than 30 seconds. This start circuit is not credited in any ONS safety analysis, therefore this signal is considered an INVALID signal with respect to 50.73 (a)(2)(iv)(A). Upon actuation, the circuit starts both Motor Driven Emergency Feedwater Pumps (MDEFWPs) on the affected unit. It does not send a start signal to the Turbine Driven Emergency Feedwater Pump.
"Shortly after entering Mode 4, a control operator stopped one of two operating Reactor Coolant Pumps per the Unit shutdown procedure. A second control operator was manually controlling Feedwater and reduced flow to control the system cooldown rate, without recognizing that the 2B SG had reached the low level setpoint for SG Dry-out Protection Circuit actuation.
"In this event, the 2B SG level trains A and B indicated lower than the setpoint value (-21 inches) for more than 30 seconds, so the SG Dry-out Protection Circuit started the 2A and 2B MDEFWPs. The actuation was considered complete and the system started and functioned successfully. Both MDEFWPs started. The 2B SG control valve, 2FDW-316, opened and fed the 2B SG from the 2B MDEFWP for approximately one minute until the SG level reached the EFDW control setpoint (30 inches). The 2A pump did not feed the 2A SG because 2A SG level remained well above the EFDW control setpoint so there was no demand for the 2A SG control valve, 2FDW-315, to open."
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 40688 | Facility: ROBINSON
Region: 2 State: SC
Unit: [2] [ ] [ ]
RX Type: [2] W-3-LP
NRC Notified By: CHUCK BAUCOM
HQ OPS Officer: BILL GOTT | Notification Date: 04/20/2004
Notification Time: 13:56 [ET]
Event Date: 04/20/2004
Event Time: 13:10 [EDT]
Last Update Date: 04/20/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
| Person (Organization):
CAUDLE JULIAN (R2)
HO NIEH (IRO)
BOB DENNIG (NRR)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | N | N | 0 | Hot Shutdown | 0 | Hot Shutdown | Event Text OFFSITE NOTIFICATION - NON-WORK RELATED FATALITY
On April 20, 2004, a non-work related fatality occurred at H. B. Robinson. At 0730 EDT, a Progress Energy employee assigned to the main turbine maintenance crew for the current maintenance outage that began on April 20, suffered a condition that required immediate medical attention. Medical assistance was provided by onsite plant employees trained as first responders. The employee was transported by ambulance to a nearby medical facility where additional medical treatment was rendered. At 1310, site personnel were notified that attempts to revive the employee were not successful and that the employee had died. The employee was working in a non-radiological controlled area of the plant and no radioactive material or contamination was involved. Notification of the South Carolina Occupational Safety and Health Agency is planned.
The Licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 40689 | Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JEFF KNISLEY
HQ OPS Officer: JEFF ROTTON | Notification Date: 04/20/2004
Notification Time: 23:26 [ET]
Event Date: 04/20/2004
Event Time: 18:10 [PDT]
Last Update Date: 04/20/2004 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
RUSSELL BYWATER (R4)
WILLIAM BECKNER (NRR)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Cold Shutdown | 0 | Cold Shutdown | Event Text DEGRADED CONDITION - GREATER THAN ONE PERCENT OF STEAM GENERATOR TUBES DEFECTIVE
"On April 08, 2004, during the Unit 1 twelfth refueling outage (1R12), analysis of eddy current data on [Steam Generator] SG 1-4 indicated that greater than one percent of the total tubes inspected in SG 1-4 were defective. Ninety-two (92) defective tubes in SG 1-4 were detected and are being plugged. Most of the [pluggable] indications are due to circumferential primary water stress corrosion cracking in the Rows 5 to 8 U-bend region. Results of the SG tube inspection fall into Category C-3, which requires a four-hour non-emergency report in accordance with Technical Specification (TS) Table 5.5.9-2 and 10 CFR 50.72(b)(3)(ii) (an 8-hour requirement replacing the former (b)(2)(iii)(C), 4-hour requirement). [This was reported in EN#40659 dated 04/08/04.]
"An NRC teleconference in accordance with TS 5.6.10.d to report the results of the voltage-based repair criteria implemented for the tube support plate intersections was conducted on Wednesday April 14, 2004, at 1300 EDT.
"On April 20, 2004, following completion of SG eddy current testing, PG&E identified SG 1-1 had greater than one percent of the total tubes that were defective. Forty (40) defective tubes in SG 1-1 were detected and plugged. Most of the [pluggable] indications are due to outside diameter stress corrosion cracking at the SG tube to support plate intersections. SG 1-2 had twenty-six (26) and SG 1-3 had twenty-seven (27) defective tubes that required plugging.
"A Special Report in accordance with TS 5.6.10.a and e will be submitted prior to returning Unit 1 to power operation.
"A Licensee Event Report in accordance with 10 CFR 50.73(a)(2)(v)(C) will be submitted within 60 days."
The Licensee notified the NRC Resident Inspector. | |