| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
05/11/2006 - 05/12/2006 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42478 | Facility: PEACH BOTTOM
Region: 1 State: PA
Unit: [ ] [3] [ ]
RX Type: [2] GE-4,[3] GE-4
NRC Notified By: TODD STRAYER
HQ OPS Officer: JOHN MacKINNON | Notification Date: 04/05/2006
Notification Time: 16:43 [ET]
Event Date: 04/05/2006
Event Time: 16:15 [EDT]
Last Update Date: 05/11/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
| Person (Organization):
PAT FINNEY (R1)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text HIGH PRESSURE COOLANT INJECTION DECLARED INOPERABLE (HPCI)
"During performance of a Unit 3 High Pressure Coolant Injection System (HPCI) Logic System Functional Test, the High Pressure Coolant Injection (HPCI) system was found to be inoperable. The inoperability is due to a logic failure that would prevent the automatic opening of MO-3-23-015, 'HPCI TURBINE STEAM LINE INBOARD ISOLATION VAVLE'. MO-3-23-015 is a normally open valve located inside Primary Containment. The HPCI system initiates upon receipt of a reactor low water level (level 3) signal or a high drywell pressure signal. Upon a HPCI system initiation MO-3-23-015 is required to automatically open, if closed, with no isolation signal present. The automatic opening of MO-3-23-015 is required to ensure the design function of HPCI is fulfilled. The design function of HPCI is to assure that the reactor is adequately cooled to limit fuel-clad temperature in the event of a small break in the nuclear system and loss of coolant, which does not result in rapid depressurization of the reactor vessel. HPCI permits the nuclear plant to be shut down while maintaining sufficient reactor vessel water inventory until the reactor vessel is depressurized.
"The HPCI system inoperability has no immediate impact on plant operations. The inoperability places Unit 3 in a 14-day shutdown Tech Spec Action Statement. The 14-day shutdown Tech Spec Action Statement commenced on 04/05/06 @ 0824 when the Logic System Function Test was started.
"A troubleshooting and repair team has been initiated to determine a cause and repair of the deficiency."
The licensee notified the NRC Resident Inspector.
* * * RETRACTION AT 09:10 ON 5/11/2006 FROM DAVE FOSS TO ABRAMOVITZ * * *
"The purpose of this notification is to retract a previous report made on 4/5/06 at 1643 hours (EN# 42478). On 4/5/06, the High Pressure Coolant Injection (HPCI) system was removed from service and declared inoperable for the performance of a Logic System Functional Test (LSFT). During the LSFT, a condition was discovered in the open control logic circuit of the MO-3-23-015 (HPCI Turbine Steam Line Inboard Isolation Valve) that would have inhibited automatic opening of the valve on a HPCI initiation signal. Notification of this issue to the NRC on 4/5/06 was initially made as a result of the belief that the MO-3-23-015 may not have been capable of being opened for design events. It was subsequently determined that a relay contact exhibited high resistance during the LSFT. The relay was replaced on 4/6/06 (IR 475307).
"Since the initial report, Engineering has performed an operability review that evaluated the impact that the relay contact high resistance condition had on the ability of the HPCI system to perform its safety function. It was determined that the relay contact degradation did not cause the HPCI system to be inoperable. The MO-3-23-015 is a normally open Primary Containment Isolation Valve. MO-3-23-015 would only isolate for conditions where HPCI would be rendered inoperable. Automatically re-opening the MO-3-23-015 valve is not a safety related function. Therefore, HPCI was capable of performing its safety function with the degraded relay contact."
Notified the R1DO (Dimitriadis). The licensee notified the NRC Resident Inspector. |
| General Information or Other | Event Number: 42562 | Rep Org: MA RADIATION CONTROL PROGRAM
Licensee: HERBERT HENKEN
Region: 1
City: MILTON State: MA
County: SUFFOLK
License #:
Agreement: Y
Docket: 03-6234
NRC Notified By: MIKE WHALEN
HQ OPS Officer: JOE O'HARA | Notification Date: 05/09/2006
Notification Time: 08:48 [ET]
Event Date: 11/01/2005
Event Time: [EDT]
Last Update Date: 05/09/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
| Person (Organization):
ANTHONY DIMITRIADIS (R1)
GREG MORELL (NMSS)
| Event Text MASSACHUSETTS AGREEMENT STATE - ABANDONED RA-226 QUANTITY 1000 TIMES GREATER THAN LABELING REQUIREMENTS
The following information was received from the state via fax:
"Cause Description: During Clean-Out of Deceased Dermatologist, the family discovered a box with Ra-226.
"About 11/1/2005: [Deleted] contacted the Boston Regional (Office) of the EPA to dispose of Ra-226 discovered at family home. EPA recommended to [Deleted] that it would be cheaper if he disposed of the Ra-226 directly via a waste broker. EPA suggested Radiac Environmental Services of Brooklyn, NY.
"11/17/2005: Radiac (Environmental Services of Brooklyn, NY) visited the Milton, MA residence and estimated that the amount of Ra-226 was about 2.9mCi. In addition, Radiac placed the box of Ra-226 sources into a 5-Gallon DOT Type A steel drum with a ring bolt closure, placed a security seal was placed over the ring, and labeled the drum with yellow/magenta radioactive label. [Deleted] was going to Florida for the winter.
"4/27/06: The EPA informed the Mass. Radiation Control Program about the Ra-226 in Milton, MA.
"Corrective Actions: Material to be disposed of by licensed broker - Radiac Environmental Services of Brooklyn, NY - during the week of May 8, 2006." |
| Power Reactor | Event Number: 42569 | Facility: PALISADES
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: PATRICK PITCHER
HQ OPS Officer: BILL HUFFMAN | Notification Date: 05/11/2006
Notification Time: 17:35 [ET]
Event Date: 05/11/2006
Event Time: 15:14 [EDT]
Last Update Date: 05/11/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
RICHARD SKOKOWSKI (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 8 | Power Operation | 0 | Hot Standby | Event Text MANUAL ACTUATION OF REACTOR PROTECTION SYSTEM DUE TO FAILURE OF CONTROL ROD TO WITHDRAW
"Following a startup from a refueling outage with the plant at approximately 24% power, it was determined that one control rod appeared to be fully inserted in the core as determined by both incore and excore flux tilts. The Off Normal Procedure for a dropped control rod was entered. The decision was made to take the plant offline to troubleshoot and correct the apparent condition. The Off Normal Procedure does not allow operation in Mode 2 with a dropped control rod. At 1514 hours EDT the reactor was manually tripped from 8% power. The reactor trip was uncomplicated.
"This is reportable under 10 CFR 50.72(b)(2)(iv)(B) as a manual Reactor Protection System actuation while critical."
Decay heat is being discharged to the condenser via the turbine bypass valves and cooling is being supplied from AFW. All systems besides the control rod in question functioned as required.
It appears that the control rod in question may have been inserted since startup. The rod position indicators were showing a normal rod position. The licensee believes that the rod may not be coupled. This rod has had a previous history of being difficult to couple. The licensee stated that normal rod testing was conducted prior to startup along with low power physics testing which did not indicate anything significantly out of spec.
The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42571 | Facility: CALLAWAY
Region: 4 State: MO
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: JOHN DAMPF
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 05/12/2006
Notification Time: 05:11 [ET]
Event Date: 05/12/2006
Event Time: 00:53 [CDT]
Last Update Date: 05/12/2006 | 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):
DALE POWERS (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | M/R | Y | 12 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP AFTER HIGH TURBINE VIBRATION
"While lowering turbine load to 45% for planned maintenance to replace an RCS Loop flow transmitter, Callaway Plant experienced high vibration on two main turbine bearings. The main turbine was manually tripped at 0047 in accordance with off-normal procedures. At 0052, received a Steam Generator High-High Level (P-14) on the "A" S/G resulting in a Feed Water Isolation Signal (FWIS) and Auxiliary Feed Water Actuation (AFAS). The reactor was manually tripped at 0053. Emergency Operating Procedures were completed and exited at 0115."
After receiving the main turbine high vibration alarm, the plant reduced power to below the reactor trip/turbine trip setpoint and manually tripped the main turbine. Steam generator level rose to the P-14 feedwater isolation setpoint at which time, the reactor was manually tripped.
All rods fully inserted on the trip. Decay heat is being removed by condenser steam dumps. Steam generator level is being maintained with the startup feed pump. The electrical grid is stable. No relief valves or safety valves lifted. The cause of the high vibration is being investigated.
The licensee notified the NRC Resident Inspector. | |