| U.S. Nuclear Regulatory Commission
Operations Center Event Reports For
06/16/2006 - 06/19/2006 ** EVENT NUMBERS ** |
 | !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42534 | Facility: KEWAUNEE
Region: 3 State: WI
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: JOEL STODOLA
HQ OPS Officer: JOHN KNOKE | Notification Date: 04/28/2006
Notification Time: 03:25 [ET]
Event Date: 04/27/2006
Event Time: 19:29 [CDT]
Last Update Date: 06/16/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
HIRONORI PETERSON (R3)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Intermediate Shutdown | 0 | Intermediate Shutdown | Event Text TWO TRAINS OF SHIELD BUILDING VENTILATION INOPERABLE
At 19:29 on 04/27/2006 the Kewaunee Power Station declared two trains of Shield Building Ventilation (SBV) inoperable. Train B SBV was declared inoperable on 4/25/2006 at 17:25 when SW Train B was declared out of service for a leak that developed on the branch header to Diesel Generator B and TS 3.3.e.2 was entered. On 4/27/2006 at 19:29 QA typing discrepancies in RR-119 were discovered. QA-2 components were used in a QA-1 system, which could potentially cause a failure of some safety related equipment powered from this relay rack. As a result, all safety related equipment powered from RR-119 were declared inoperable and the appropriate technical specifications entered. SBV Train A Damper Control is powered from RR-119 and was declared inoperable at 19:29 on 4/27/2006 resulting in two trains of SBV being inoperable and TS 3.6.c.1 entered. The QA-2 components that may have an adverse affect on QA-1 safety related components in RR-119 have been removed and both trains of SBV were declared operable at 00:29 on 4/28/2006.
The licensee notified the NRC Resident Inspector.
* * * RETRACTION FROM JERRY RISTE TO W. GOTT 1524 EDT ON 6/16/06 * * *
"This event was reported on April 27, 2006 (Event Number 42534) for two trains of shield building ventilation (SBV) being declared inoperable. Train B SBV was declared inoperable due to loss of the service water spray system for the SBV charcoal filters and Train A SBV was declared inoperable due to relay rack RR-119 Quality Assurance typing discrepancies (relay rack provides power for SBV Damper Control). Subsequent review of analysis determined that the SBV spray system is not required for post-LOCA operation to control the release of radioactive material. Therefore, Train B of SBV was not required to be declared inoperable when Train B service water was declared out of service. With only one Train of SBV being inoperable, this event is not reportable and is being retracted.
"The licensee notified the NRC Resident Inspector."
Notified R3DO (P. Louden). |
| Power Reactor | Event Number: 42640 | Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: KEN JOHNSTON
HQ OPS Officer: BILL GOTT | Notification Date: 06/14/2006
Notification Time: 17:30 [ET]
Event Date: 06/14/2006
Event Time: 09:41 [PDT]
Last Update Date: 06/16/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(xiii) - LOSS COMM/ASMT/RESPONSE
| Person (Organization):
RUSSELL BYWATER (R4)
|
| 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 ERDS INOPERABLE
"On June 14, 2006, at 0941 PDT, plant operators declared both Units 1 and 2 Emergency Response Data System (ERDS) inoperable due to the failure to complete scheduled surveillance testing. The in-plant emergency response facility data system (ERFDS) and safety parameter display system (SPDS) are operating as expected, but data transfer could not be accomplished via the digital data transfer to the NRC Event Response Center.
"Plant personnel are actively investigating the cause of the data transfer failure, and will update this notice upon completion of troubleshooting, repair, and successful performance of the surveillance test.
"Plant management, emergency personnel, and the NRC Resident Inspector(s) will be informed of the condition, planned maintenance actions, and resolution of this condition."
* * * UPDATE FROM DAVE BAHNER TO W. GOTT AT 1841 EDT ON 06/16/06
"PG&E restored the ERDS data system to service and satisfactorily completed communication testing with the NRC Incident Response Center." The system was declared operable at 1433 PT.
The licensee will notify the NRC Resident Inspector.
Notified R4DO (A. Gody) |
| !!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!! | | Power Reactor | Event Number: 42641 | Facility: OCONEE
Region: 2 State: SC
Unit: [1] [ ] [ ]
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: PETE SNYDER | Notification Date: 06/14/2006
Notification Time: 19:08 [ET]
Event Date: 06/14/2006
Event Time: 14:00 [EDT]
Last Update Date: 06/16/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
| Person (Organization):
BRIAN BONSER (R2)
|
| 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 LEAKING DECAY HEAT REMOVAL ISOLATION VALVE BYPASS LINE
"On 2-21-06, during a tour of containment during normal operation at 100% power, a small leak (one (1) to three (3) drops per second) was noted from a 1/2 inch line connected to the decay heat removal (DHR) drop line. It was identified as being a body-bonnet leak on valve 1LP-167 subject to a TS limit of 10 gpm.
"At approximately 1400 hours on 6-14-06 following a shutdown for an unrelated issue, the source was identified as a leak at a weld in a "tee" joint adjacent to 1LP-167. This is considered RCS pressure boundary leakage, subject to a TS limit of zero leakage. The leak was isolated by closing a normally open valve in the 1/2 inch line and the leakage stopped.
"Initial Safety Significance: The leak is in a 1/2 inch line which provides over pressure protection from thermal expansion in the volume between 1LP-1 and 1LP-2 (the main pressure boundary isolation valves between the high pressure RCS and the LPI (DHR) system). The leak rate (1 to 3 drops per second) was not significant, except that it was RCS pressure boundary leakage. 1LP-1 is normally closed, but must be opened to establish a DHR path. Valve 1LP-167 is a 1/2 inch check valve which would have limited RCS leakage. Thus, if the leak had grown, it would have been limited to the amount of seat leakage past either 1LP-167 or 1LP-1. It would also have been limited by the 1/2 inch size of the line containing the leak."
Technical Specification LCO 3.4.13 applies to RCS leakage in modes 1 to 4. The licensee plans to fix the leak prior to entry into mode 4.
The licensee notified the NRC Resident Inspector.
* * * RETRACTION AT 00:15 ON 6/16/2006 FROM SAM LARK TO ABRAMOVITZ * * *
"On 6-14-06 at 1908 hours Oconee reported an RCS pressure boundary leak in a 1/2 inch line connected to the decay heat removal (DHR) line near valve 1LP-1 inside containment. Oconee has reviewed the event in greater detail and has concluded that the event is not reportable. The Basis for TS 3.4.13 states that RCS LEAKAGE includes leakage from connected systems up to and including the second normally closed valve (or outermost isolation valve for systems penetrating containment). However TS 1.1 contains a definition of LEAKAGE which includes 'Pressure Boundary LEAKAGE: LEAKAGE (except SG LEAKAGE) through a nonisolable fault in an RCS component body, pipe wall, or vessel wall.' The leakage in this event was isolable, and therefore does not meet the definition of Pressure Boundary LEAKAGE. Therefore the zero leakage criterion of TS 3.4.13 does not apply to this leak. The applicable criterion is 10 gpm identified LEAKAGE. Since the leak does not meet the criterion as Pressure Boundary LEAKAGE, the leak was isolable, and the applicable TS LEAKAGE limit was not exceeded, this event does not meet the reportability criteria for 10 CFR 50.72 or 50.73 and event notification 42641 is hereby RETRACTED.
"Additional information and clarification: "During normal operation the leak was isolated by one barrier (valves 1LP-167 and 1LP-1, closed in parallel). The leakage observed on 2-21-06 during a containment tour at Mode 1 was recorded as 1 drop per second. As stated in the initial notification, at that time the leak was believed to be a body-bonnet leak. It was observed at Mode 1 again on 5-25-06 and recorded as 3 drops/second. On 6-14-06, the leakage was recorded as one drop/second while at reduced pressure in Mode 4, before the DHR systems was placed in service. At that point, the leak was isolated by closing an additional valve (1LP-166, normally open), and the leak stopped. The Low Pressure Injection system was placed in service for DHR, which opened 1 LP-1. Later, with system pressure at approximately 285 psig in Mode 5 (outside the applicability of TS 3.4.13), 1LP-166 was reopened to allow additional verification of the leak location. At that time the leak was described as a 'spray' but no leak rate was measured before 1LP-166 was reclosed. The leak rate at that time was estimated as well less than 10 GPM.
"Corrective Action: The affective section of 1/2 inch pipe and associated fittings have been removed for transfer to a Duke laboratory for analysis. Repairs will be completed prior to return to mode 4."
The licensee notified the NRC Resident Inspector. Notified the R2DO (Bonser). |
| Power Reactor | Event Number: 42647 | Facility: SAINT LUCIE
Region: 2 State: FL
Unit: [ ] [2] [ ]
RX Type: [1] CE,[2] CE
NRC Notified By: SHAWN ELLIOTT
HQ OPS Officer: MARK ABRAMOVITZ | Notification Date: 06/16/2006
Notification Time: 02:18 [ET]
Event Date: 06/15/2006
Event Time: 22:23 [EDT]
Last Update Date: 06/16/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(iv)(B) - RPS ACTUATION - CRITICAL
| Person (Organization):
BRIAN BONSER (R2)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 2 | M/R | Y | 45 | Power Operation | 0 | Hot Standby | Event Text MANUAL REACTOR TRIP DUE TO DIGITAL ELECTRO-HYDRAULIC LEAK
"On 6/15/06 at 2223 hrs an unplanned manual reactor trip was initiated on St. Lucie Unit 2 from 45% power due to severe DEH Leak on the #1 Throttle Valve. DEH Leak ceased upon Turbine Trip. Following the reactor trip, EOP-1, Standard Post Trip Actions, and EOP-2, Reactor Trip Recovery procedures were completed without contingencies and Unit 2 was stabilized in Mode 3. All control rods fully inserted and no S/G Safety Valves Lifted. Feedwater to the S/G was supplied by the main FW Pumps. All safe shutdown equipment operated as expected. There were no major equipment failures."
Decay heat is being removed with main feedwater and dumping steam to the condenser. The grid is stable. The fire brigade was activated following the trip and set a fire watch (no fire).
The NRC Resident Inspector was notified of this event by the licensee. |
| Other Nuclear Material | Event Number: 42650 | Rep Org: MICHIGAN DEPT OF TRANSPORTATION
Licensee: MICHIGAN DEPT OF TRANSPORTATION
Region: 3
City: State: MI
County: WAYNE
License #:
Agreement: N
Docket:
NRC Notified By: TOM KILLINGSWORTH
HQ OPS Officer: MIKE RIPLEY | Notification Date: 06/18/2006
Notification Time: 19:02 [ET]
Event Date: 06/18/2006
Event Time: 01:00 [EDT]
Last Update Date: 06/18/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
INFORMATION ONLY
| Person (Organization):
PATRICK LOUDEN (R3)
ELMO COLLINS (NMSS)
| Event Text DAMAGED MOISTURE DENSITY GAUGE
While performing back-scatter measurements at a construction site on I-75 between Davison Freeway and I-94 in Wayne County, MI, a Model 3440 Troxler moisture density gauge (8 millicuries Cs-137, 40 millicuries Am-241:Be) belonging to the department was damaged when it was run over by a truck. The gauge was returned to the department office and survey results, pending a leak test on Monday 06/19, were negative indicating the sources remained intact. The licensee notified the NRC Resident Inspector. |
| Power Reactor | Event Number: 42651 | Facility: PALO VERDE
Region: 4 State: AZ
Unit: [1] [2] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: JIM BLASEK
HQ OPS Officer: JOE O'HARA | Notification Date: 06/19/2006
Notification Time: 00:29 [ET]
Event Date: 06/18/2006
Event Time: 14:55 [MST]
Last Update Date: 06/19/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
50.72(b)(3)(v)(B) - POT RHR INOP
| Person (Organization):
ANTHONY GODY (R4)
|
| Unit | SCRAM Code | RX CRIT | Initial PWR | Initial RX Mode | Current PWR | Current RX Mode | | 1 | N | N | 0 | Refueling | 0 | Refueling | | 2 | N | Y | 100 | Power Operation | 100 | Power Operation | | 3 | N | Y | 100 | Power Operation | 100 | Power Operation | Event Text LOSS OF ONSITE EMERGENCY SAFETY FUNCTION - "B" EDG FAILED TO START DURING TEST RUN
"The following event description is based on information currently available. If through subsequent reviews of this
event, additional information is identified that is pertinent to this event or alters the information being provided at this
time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.
"On June 18, 2006, at approximately 1455 Mountain Standard Time (MST), Palo Verde Nuclear Generating Station
Unit 1 was operating at 0% power in Mode 6 with the 'A' Emergency Diesel Generator (EDG) inoperable for troubleshooting of the train 'A' load sequencer, The 'B' EDG failed to start during a test run resulting in a loss of the onsite emergency power safety function required by General Design Criterion (GDC) 17. The offsite power source safety function remains available to the plant.
"The unit entered Technical Specification 3.8.2, AC Sources - Shutdown Condition B for one required DG inoperable.
There was no movement of irradiated fuel assemblies, therefore the unit remained in compliance with the Required
Actions. The offsite electrical grid is stable.
"At 1810 MST the 'A' EDG was declared operable, exiting Technical Specification 3.8.2. With the 'A' EDG operable,
the safety function for the onsite emergency power was also restored.
"There were no structures, systems or components that were inoperable at the start of event that contributed to the
event. This condition did not result in any challenges to the fission product barrier or result in any releases of
radioactive materials. There were no adverse safety consequences or implications as a result of this event. This
condition did not adversely affect the safe operation of the plant or health and safety of the public.
"The NRC Resident Inspector has been notified of this condition and this ENS notification." |
| Power Reactor | Event Number: 42652 | Facility: MONTICELLO
Region: 3 State: MN
Unit: [1] [ ] [ ]
RX Type: [1] GE-3
NRC Notified By: RYAN RICHARDS
HQ OPS Officer: JOHN MacKINNON | Notification Date: 06/19/2006
Notification Time: 04:06 [ET]
Event Date: 06/18/2006
Event Time: 23:19 [CDT]
Last Update Date: 06/19/2006 | Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
| Person (Organization):
PATRICK LOUDEN (R3)
|
| 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 RADIATION MONITOR SPIKED THREE TIMES
"At 2319 on 06/18/2006, the 'A' Plenum Radiation Monitor spiked high which resulted in the closure of the Drywell CAM (Continuous Air Monitor) and the Oxygen Analyzer Primary Containment Isolation Valves. The spike's (3) of approx. 30 mR/HR occurred during this event. The valves isolated once, the Reactor Building Ventilation trip was reset once and re-isolation occurred several minutes later from another spike on the Rad Monitor. Trip setpoint is 26 mR/hr. The Plenum high Rad signal also resulted in Reactor Building isolation (twice), start of 'A' Standby Gas Treatment, and transfer of the Control Room Ventilation to the high Rad Mode. The 'R' Plenum Rad Monitor remained constant at 1.3 mR/hr. The Reactor Building Ventilation and Control Room Ventilation have been reset and Standby Gas Treatment has been secured. The 'A' Plenum Radiation Monitor has been declared inoperable."
Instrument & Control Technicians are currently troubleshooting the problem associated with the 'A' Plenum Radiation Monitor.
The NRC Resident Inspector's have been left messages by the licensee. | |