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Event Notification Report for February 8, 2005

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
02/07/2005 - 02/08/2005

** EVENT NUMBERS **


41322 41367 41379 41380 41381 41382 41383

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 41322
Facility: NORTH ANNA
Region: 2 State: VA
Unit: [1] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: MARK SCHRY
HQ OPS Officer: JEFF ROTTON
Notification Date: 01/10/2005
Notification Time: 14:57 [ET]
Event Date: 01/10/2005
Event Time: 13:45 [EST]
Last Update Date: 02/07/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
Person (Organization):
MARK LESSER (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

OFFSITE NOTIFICATION TO FEDERAL ENERGY REGULATORY COMMISSION

"At 1345 on 1-10-2005, the Federal Energy Regulatory Commission (FERC) Local Engineer was notified by voice mail message of a failure of the North Anna Spillway Emergency Diesel (1-EE-EG-4) during a surveillance PT. 1-EE-EG-4's voltage regulator would not control voltage. The cause of the failure is unknown at this time. This failure only affects the hydro project and there is no impact on station operation. The FERC notification was made per 18CFR12.10(a)."

The licensee has notified the NRC Resident Inspector.

* * * RETRACTION FROM MICHAEL WHALEN TO JEFF ROTTON AT 1200 EST ON 2/7/05 * * *

The following information was provided by licensee via facsimile:

"At 1457 hours on January 10, 2005, a 4 hour non-emergency notification was made to the NRC Operation Center as a result of an Offsite Notification to other government agency (i.e- FERC due to an inoperable spillway diesel) in accordance with 10 CFR 50.72 (b)(2)(xi). NRC guidance in NUREG -1022, Event Reporting Guidelines 10 CFR 50.72 and 50.73, provides discussion and notification examples intended to set a threshold that ensures reporting when an issue is safety significant. Some notifications to other government agencies may have little or no significance and may not be reportable to the NRC.

"Although the project safety device was inoperable and therefore reportable to FERC there was no safety significance since normal power was available throughout the time the device was out of service.

"Based on the discussions in NUREG - 1022, Section 3.2.12 for the notification in question, the North Anna Station Nuclear Safety Operating Committee has determined a NRC Notification was not required because there was no safety significance and is therefore being retracted.

The licensee notified the NRC Resident Inspector.

Notified R2RDO (Charlie Payne).

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General Information or Other Event Number: 41367
Rep Org: CALIFORNIA RADIATION CONTROL PRGM
Licensee: TRC LOWNEY ASSOCIATES
Region: 4
City: YORBA LINDA State: CA
County:
License #: 6773-30
Agreement: Y
Docket:
NRC Notified By: ROBERT GREGOR
HQ OPS Officer: MIKE RIPLEY
Notification Date: 02/03/2005
Notification Time: 15:28 [ET]
Event Date: 02/02/2005
Event Time: 12:05 [PST]
Last Update Date: 02/03/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
GARY SANBORN (R4)
M. WAYNE HODGES (NMSS)

Event Text

CALIFORNIA AGREEMENT STATE REPORT - DAMAGED MOISTURE DENSITY GUAGE

At 1205 on February 2, 2005, the licensee called the Brea office to report that one of their moisture density gauges had been run over by construction equipment at 1145 that morning. The gauge was damaged to the extent that they were unable to retract the source rod back into the shielded position (CPN, model 3, # M22054410 - 10 mCi Cs-137 and 50 mCi Am:Be). The incident happened at a construction site located at 3208 Quarter Horse Drive in Yorba Linda at the water reservoir.

The outer gauge casing was in pieces, the rod handle was broken and the source rod was extended into the ground. The maximum dose rate with the source rod extended into the ground was 2 mrem/hr at contact measured with a Bicron microrem instrument. The dose rate when the source rod was extracted from the ground was approximately 30 mrem/hr at one foot from the rod tip. The Cs-137 end of the gauge source rod was placed in a lead pig, which was taped into place to keep the Cs-137 source shielded. The unit remains were placed in a plastic bag which was then placed into the transportation case. The device was transported to the local CPN service representative - Maurer Technical Services. Maurer Technical Services stated the sources appeared to be intact, and they took a wipe for a leak test. The wipe was sent out for counting. Alpha and beta surveys at the accident scene did not detect any contamination. Two wipes of the source housings were collected, which were counted in the Brea office on a Ludlum 3030 alpha-beta counter the next day. No contamination was detected on these two wipes.

The incident apparently occurred when the construction equipment, a front end loader, changed direction to avoid a truck, and there wasn't time for the gauge operator, who was in a safe location about 15 feet from the gauge, to signal the front end loader to stop.

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Power Reactor Event Number: 41379
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [1] [ ] [ ]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: STEVE SMITH
HQ OPS Officer: JEFF ROTTON
Notification Date: 02/07/2005
Notification Time: 04:41 [ET]
Event Date: 02/06/2005
Event Time: 22:19 [MST]
Last Update Date: 02/07/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
GARY SANBORN (R4)
CORNELIUS HOLDEN (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

EMERGENCY DIESEL START DUE TO BUS DEENERGIZATION

The following information was provided by the licensee via facsimile:

"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 February 6, 2005, at approximately 22:19 Mountain Standard Time (MST) a valid actuation of the Palo Verde Nuclear Generating Station Unit 1 Train `B' Emergency Diesel Generator (EDG) occurred as a result of undervoltage on its respective safety bus (PBB-S04). EDG 'B' started and loaded as designed to energize PBB-S04.

"The loss of power to the safety bus was the result of a fault associated with 13.8KV breaker NAN-SO6J which caused breakers NAN-SO6H (normal power supply), NAN-S06K (alternate power supply), and NAN-SO6J (EOF & TSC Bldg power supply) to all trip open on Overcurrent. This action resulted in the deenergization of NAN-S06, NAN-S04, and PBB-S04. The PVNGS Fire Department and Auxiliary Operators responded to a report of smoke and upon arrival found no fire. The Fire Department verified the fire was completely extinguished and there were no extensions (secondary fires).

"Unit 1 entered Technical Specification LCO 3.8.1, Condition 'A', for one (of two) required offsite circuits inoperable. Various other Technical Specifications LCO's were momentarily entered and exited for PBB-S04 being deenergized for approximately 7 seconds. No Emergency Plan declaration was made and none was required.

"Unit 1 was at approximately 100% power, at normal operating temperature and pressure prior to and following the EDG actuation. No other ESF actuations occurred and none were required. No major equipment was inoperable prior to the event that contributed to the event. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.

"The NRC Resident Inspector has been notified of the ESF actuation and this ENS notification."

The B EDG is providing power to PBB-S04. Due to loss of power to the TSC, the TSC Diesel started and is providing power to the TSC. The Backup EOF located in Buckeye, AZ will be used in the case of an emergency event. There were no reported injuries. There was damage to the NAN-S06J breaker and it has been isolated. The event has been entered into the site's corrective action program for determining the cause of the breaker trip and damage. This event has no effect on the operation of the other site units.

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Power Reactor Event Number: 41380
Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: PATRICK FALLON
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 02/07/2005
Notification Time: 19:46 [ET]
Event Date: 02/07/2005
Event Time: 17:34 [EST]
Last Update Date: 02/07/2005
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
Person (Organization):
ERIC DUNCAN (R3)
WILLIAM BECKNER (NRR)
PETER WILSON (IRD)
KAREN CARTER (DHS)
MIKE EACHES (FEMA)

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

PLANT RECEIVED AN AREA RADIATION MONITOR ALARM

At 1734 on 2/7/2005, received an Area Radiation Monitor alarm in the reactor building basement airlock area at approximately 75Mr/hr. The valid alarm caused entry into EOP flowcharts for High Radiation in the Secondary Containment. Check of the relay room monitor showed 100 Mr/hr. Investigation showed that a failed open main steam line drain valve combined with placing hydrogen water chemistry in service caused the increasing radiation levels. A downstream steam line drain isolation valve was closed to isolate the steam flow path past the monitor. Hydrogen Water Chemistry injection rate was lowered. The rad levels returned to about 4 Mr/hr (normal levels) following valve closure (at 1745). The EOPs were exited at 1755. After review of the event it was determined that an unusual event should have been entered at the time of the EOP entry (EAL AU2, Unexpected Increase of Plant Radiation Levels), 1734 and exited at 1745 when area radiation levels returned to normal values. This is an after the fact notification of a missed emergency classification.

The NRC Resident Inspector was notified.

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Power Reactor Event Number: 41381
Facility: HATCH
Region: 2 State: GA
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: FRANK GORLEY
HQ OPS Officer: CHAUNCEY GOULD
Notification Date: 02/07/2005
Notification Time: 21:42 [ET]
Event Date: 02/07/2005
Event Time: 20:55 [EST]
Last Update Date: 02/08/2005
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
Person (Organization):
CHARLIE PAYNE (R2)
WILLIAM BECKNER (NRR)
PETER WILSON (IRD)
CHRIS LIGGETT (FEMA)
ROBERT BOZZO (DHS)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N N 0 Refueling 0 Refueling

Event Text

PLANT HAD A FREON LEAK IN THE DRYWELL CHILLER ROOM

The plant reported that there was a freon leak in the Drywell Chiller Room located in the Unit 2 Reactor Building. Maintenance was being performed on the refrigeration equipment relief valve when the gas began to escape into the room. There were six men working in the room and two were affected by the freon gas. They were treated on site and released. Oxygen content in the room was 20.5% and LEL for hydrocarbons was alarming at 19%. The drywell chiller room is secured and being cleared of the 1,200 pounds of freon gas. At this time an investigation is being made to determine the cause which may have been due to the relief valve opening or the workers removing the valve. The plant entered the NOUE due to a "toxic gas release". The NOUE will be terminated when the room is habitable.

The NRC Resident Inspector was notified along with State and Local agencies.

* * * UPDATE PROVIDED BY FRANK GORLEY TO JEFF ROTTON AT 0303 EST ON 02/08/05 * * *

Licensee reported that the NOUE was terminated at 0245 EST after the freon had been successfully cleared from the Drywell Chiller Room. Licensee will notify the NRC Resident Inspector.

Notified FEMA (Liggett), DHS (Knox), IRD (Wilson), R2DO (Payne), and NRR EO (Beckner).

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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: 02/07/2005
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.

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Power Reactor Event Number: 41383
Facility: SURRY
Region: 2 State: VA
Unit: [1] [ ] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DAVID HERRING
HQ OPS Officer: JEFF ROTTON
Notification Date: 02/08/2005
Notification Time: 00:14 [ET]
Event Date: 02/07/2005
Event Time: 20:23 [EST]
Last Update Date: 02/08/2005
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
CHARLIE PAYNE (R2)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 M/R N 0 Startup 0 Hot Shutdown

Event Text

MANUAL REACTOR TRIP DURING STARTUP DUE TO INDICATION OF MISALIGNED ROD

The following information was provided by the licensee via facsimile:

"While withdrawing Control Bank 'A' during the Reactor Startup, Rod B-10 indicated a rapid drop from approximately 42 steps to 17 steps on the CERPI (Computer Enhanced Rod Position Indication) panel. The reactor operator stopped withdrawal of 'A' control bank and the CERPI indication for rod B-10 remained at 17 steps. The remaining CERPIs in 'A' control bank varied from 40 to 45 steps.

"The startup was terminated and the reactor was manually tripped in accordance with AP-1 and 1-E-0 [was] initiated. All systems functioned as required on the trip. Initial investigation by I & C and Engineering found no problems with the CERPI indication. Rod Drop time data from the CERPI program shows all rods in Control Bank 'A' had a drop time of 0.32 to 0.38 seconds with the exception of B-10, which had a drop time of 0.18 seconds.

" An investigation is ongoing as to the cause of rod B-10 misalignment.

"This notification is being made pursuant to 10 CFR 50.72(b)(3)(iv)(A). The NRC resident was notified of this event."

The reactor was subcritical in the Source Range at the initiation of the event. All rods inserted fully during the manual reactor trip. S/G level is being maintained by main feedwater and decay heat removal is via the S/G PORV.

Page Last Reviewed/Updated Friday, March 30, 2012
Friday, March 30, 2012