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Event Notification Report for April 3, 2006

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
03/31/2006 - 04/03/2006

** EVENT NUMBERS **


42455 42457 42459 42461 42462 42463 42465 42466 42467 42468

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Power Reactor Event Number: 42455
Facility: QUAD CITIES
Region: 3 State: IL
Unit: [ ] [2] [ ]
RX Type: [1] GE-3,[2] GE-3
NRC Notified By: WALT COOMBS
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 03/31/2006
Notification Time: 02:26 [ET]
Event Date: 03/30/2006
Event Time: 21:00 [CST]
Last Update Date: 03/31/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
SONIA BURGESS (R3)

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

Event Text

FEEDWATER CHECK VALVE LEAKAGE ABOVE ALLOWABLE LIMITS

"At 21:00 hours on 3/30/06, Unit Two feed water header check valves 2-0220-58A and 2-0220-62A LLRT [local leak rate test] results were both indeterminate, as a result both were greater than the allowable La containment leakage rate allowed by Tech Spec 5.5.12." La is defined as the maximum allowable leak rate at a specified pressure.

"This is reportable under 10CFR50.72(b)(3)(ii)."

The "indeterminate" leak rate was higher than the test equipment could read.

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42457
Facility: BYRON
Region: 3 State: IL
Unit: [1] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: JEFF ATKINS
HQ OPS Officer: ARLON COSTA
Notification Date: 03/31/2006
Notification Time: 14:07 [ET]
Event Date: 03/31/2006
Event Time: 13:07 [CST]
Last Update Date: 03/31/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
OTHER UNSPEC REQMNT
Person (Organization):
SONIA BURGESS (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
2 N Y 100 Power Operation 100 Power Operation

Event Text

LOW TRITIUM CONCENTRATION LEVELS FOUND NEAR UNDERGROUND PIPING

"Followup information to the February 15, 2006, notification (ENS #42339 [at Braidwood]) involving the discovery of elevated levels of tritium in several vacuum breaker vaults located along the discharge piping to the Rock River. At that time we indicated we planned to install monitoring wells along this pipeline to determine if tritium has migrated from these vaults.

"Fifteen shallow test wells and eight deeper, more permanent wells were drilled on company property to obtain water samples. Of the 23 wells, two had measurable levels of tritium, however they were well below the Environmental Protection Agency's standard for drinking water (20,000 picocuries per liter).

"These elevated levels pose no health or safety hazard to the employees or public. Investigation into the source of the elevated tritium levels continues.

"A press release is planned for the afternoon of March 31, 2006 regarding the results obtained from these monitoring wells."

Incident reported according to 10 CFR 50.72 (c)(2) and the licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42459
Facility: DIABLO CANYON
Region: 4 State: CA
Unit: [ ] [2] [ ]
RX Type: [1] W-4-LP,[2] W-4-LP
NRC Notified By: DAVE TAGGERT
HQ OPS Officer: JEFF ROTTON
Notification Date: 03/31/2006
Notification Time: 16:01 [ET]
Event Date: 03/31/2006
Event Time: 11:50 [PST]
Last Update Date: 03/31/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
DALE POWERS (R4)
OMID TABATABAI (NRR)

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

DEFECTIVE RHR CHECK VALVE

"In accordance with 10 CFR 21.21(d)(1), Pacific Gas and Electric (PG&E) is hereby notifying the NRC of a defective component received from Flowserve, Flow Control Division, in Raleigh, NC, but not installed at Diablo Canyon Power Plant (DCPP). The component is an 8-inch tilting disk check valve that was procured for installation in the Residual Heat Removal (RHR) System during the Unit 2 refueling outage (2R13) scheduled to begin on 04/17/06.

"On 03/02/06, the defect was identified at DCPP during post-receipt bench testing and involved incorrect disc dimensions that caused the disc to stick in the valve bonnet (i.e., in the open position). This would have prevented the valve from performing its intended safety function of closing to prevent pump-to-pump interaction when both RHR pumps are running. (These check valves were installed in response to NRC Bulletin 88-04, 'Potential Safety Related Pump Loss.') Failure of this check valve, had it been installed, could have resulted in the loss of one RHR train on Unit 2, which could impact the ability to shut down the reactor and maintain it in a safe shutdown condition.

" On 03/08/06, PG&E notified Flowserve of the defect via Supplier Audit Finding Report #060670010 and requested corrective actions be taken.

"On 03/13/06, Flowserve concluded that the defect was caused by disc design error and test procedure error.

"On 03/16/06, Flowserve initiated Quality Problem Corrective Action Plan #169, in which they concluded a Part 21 evaluation was not required.

"On 03/31/06, PG&E Vice President, Diablo Canyon Operations and Station Director, [deleted], determined that the defect met 10 CFR 21.21 reporting requirements.

"PG&E initiated purchase of the 600 lb, stainless steel check valve on 06/02/05, and does not know whether any others have been manufactured by Flowserve. The valve was manufactured in accordance with Vendor Assembly Drawing W9023267 and ASME Section III, Subsection NC, 1989 Edition.

"PG&E subsequently repaired the check valve in accordance with instructions provided in a Flowserve letter to PG&E, dated 03/16/06. The valve has passed inspection and bench testing and will be installed during 2R13."

The licensee notified the NRC Resident Inspector.

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General Information or Other Event Number: 42461
Rep Org: WHITING CORPORATION
Licensee: WHITING CORPORATION
Region: 3
City: MONEE State: IL
County:
License #:
Agreement: Y
Docket:
NRC Notified By: MARK KWASNY
HQ OPS Officer: JEFF ROTTON
Notification Date: 03/31/2006
Notification Time: 17:56 [ET]
Event Date: 03/30/2006
Event Time: [CST]
Last Update Date: 03/31/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
SONIA BURGESS (R3)
WILLIAM COOK (R1)
THOMAS DECKER (R2)
DALE POWERS (R4)
FRANK GILLESPIE (NRR)
OMID TABATABAI (NRR)

Event Text

PART 21 - CRANE OVERSTRESS CONDITIONS

The following is a summary of information provided by the manufacturer via email:

On March 30, 2006, as part of a crane design analyses and study, Whiting Corporation identified two overstress conditions. One is on the Girder End connection of a low head room bridge and one on a single fastener in a main hoist gear case application. It is recommended that the affected cranes at the facilities listed below be restricted to 50% capacity for the over stressed end connection and 50% or 85 % capacity for the suspected over-stressed bolt until such time as the fastener or end connection can be inspected and evaluated or replaced. The following list identifies the facilities where the bridges and trolleys are suspected to be affected by the Whiting Corporation.

Oyster Creek, Three Mile Island, Seabrook

Oconee, McGuire, Shearon Harris, St. Lucie, Catawba

D. C. Cook

Cooper, Waterford, Grand Gulf, San Onofre, Palo Verde, Columbia Generating Station

Whiting Corporation is attempting to contact the customers directly to notify them of this circumstance.

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Power Reactor Event Number: 42462
Facility: PALISADES
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: CHRISTER DAHLGREN
HQ OPS Officer: JEFF ROTTON
Notification Date: 03/31/2006
Notification Time: 20:14 [ET]
Event Date: 03/28/2006
Event Time: 12:15 [EST]
Last Update Date: 03/31/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
SONIA BURGESS (R3)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 99 Power Operation 0 Refueling

Event Text

TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO HPSI INOPERABILITY

"On 3/28/06 at 1215, the left train High Pressure Safety Injection [HPSI] Pump P-66B subcooling valve CV-3070 was declared inoperable due to valve failing to stroke. Technical Specification Limiting Condition for Operation 3.5.2 ECCS - Operating, Required Action B.1 requires HPSI train to be restored to operable status within 72 hours. This required action expires on 04/01/06 at 1215. It has been determined that it will not be possible to restore operability prior to the expiration of this action statement.

"A Technical Specification required shutdown to Primary Coolant system temperature < 325 degrees F will be initiated on March 31, 2006 at 2100. The site will then commence a planned refueling outage that was scheduled to begin on April 01, 2006."

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42463
Facility: MILLSTONE
Region: 1 State: CT
Unit: [ ] [2] [ ]
RX Type: [1] GE-3,[2] CE,[3] W-4-LP
NRC Notified By: BARRETT NICHOLS
HQ OPS Officer: JEFF ROTTON
Notification Date: 03/31/2006
Notification Time: 22:36 [ET]
Event Date: 03/31/2006
Event Time: 22:20 [EST]
Last Update Date: 03/31/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(2)(i) - PLANT S/D REQD BY TS
Person (Organization):
WILLIAM COOK (R1)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 N Y 100 Power Operation 0 Hot Shutdown

Event Text

TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO INOPERABLE AFW TURBINE

During post maintenance testing of the Auxiliary Feedwater Terry Turbine driven pump, pump problems were discovered related to the turbine governor and pump bearings. The turbine driven pump is inoperable and can not be made operable within the Technical Specification LCO Action time limit of 72 hours. The unit will be commencing a reactor shutdown to Mode 4 and perform maintenance to correct the problems.

The licensee notified the NRC Resident Inspector, the Connecticut Department of Environmental Protection, and the Local Waterford Dispatch.

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Power Reactor Event Number: 42465
Facility: FERMI
Region: 3 State: MI
Unit: [2] [ ] [ ]
RX Type: [2] GE-4
NRC Notified By: PATRICK FALLON
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 04/01/2006
Notification Time: 02:45 [ET]
Event Date: 04/01/2006
Event Time: 00:39 [EST]
Last Update Date: 04/01/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(A) - DEGRADED CONDITION
Person (Organization):
SONIA BURGESS (R3)

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

Event Text

EXCESSIVE CHECK VALVE LEAKAGE AT A CONTAINMENT PENETRATION

"At 0039 hours on 4/01/2006, Fermi 2 feed water line check valves, B2100-F010A and B2100-F076A, failed their LLRT test. The leakage rate of the inboard check valve B2100-F010A was 324.21 SCFH and the leakage rate of the outboard check valve, B2100-F076A was above the measurement capability of the leak rate monitor. The combined penetration (X-9A) leakage value was thus 324.21 SCFH which is greater than the allowable containment leakage rate (La) value of 296.3 SCFH per Tech Spec 5.5.12."

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42466
Facility: COOPER
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: DUANE ALVENS
HQ OPS Officer: STEVE SANDIN
Notification Date: 04/02/2006
Notification Time: 15:21 [ET]
Event Date: 04/02/2006
Event Time: 14:04 [CDT]
Last Update Date: 04/02/2006
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
Person (Organization):
DALE POWERS (R4)
FRANK GILLESPIE (NRR)
PETER WILSON (IRD)
JOHN FROST (DHS)
ERWIN CASTO (FEMA)
BRUCE MALLETT (R4)
ART HOWELL (R4)
KRISS KENNEDY (R4)
SCOTT SCHWIND (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

Event Text

UNUSUAL EVENT DECLARED DUE TO ADVERSE WEATHER CONDITIONS

An Unusual Event was declared per EAL 7.1.4 at 1404 CDT due to strong winds exceeding 74 mph onsite for 5 minutes between 1355-1400 CDT (peak windspeed measured at 83 mph). The Unit remained at 100% power during the incident. Current plant conditions are stable with windspeed approximately 30 mph. Offsite power is stable with a surveillance of the #1 EDG in progress. After completion of walkdowns the licensee will assess exiting the Unusual Event. A preliminary damage report indicated loose guy wire(s) on the met tower.

The licensee informed state/local agencies and the NRC Resident Inspector who is onsite.

* * * UPDATE AT 1621 EDT ON 04/02/06 FROM GENE MACE TO S. SANDIN * * *

The licensee terminated the Unusual Event at 1500 CDT based on current sustained windspeeds less than 74 mph and passage of the storm front. The licensee is continuing their damage assessment walkdowns.

The licensee informed state/local agencies and the NRC Resident Inspector. Notified R4DO (Powers), EO (Gillespie), IRD (Wilson), DHS (Mitchell), and FEMA (Mick Roland).

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Power Reactor Event Number: 42467
Facility: BEAVER VALLEY
Region: 1 State: PA
Unit: [ ] [2] [ ]
RX Type: [1] W-3-LP,[2] W-3-LP
NRC Notified By: DANIEL SCHWER
HQ OPS Officer: ARLON COSTA
Notification Date: 04/02/2006
Notification Time: 15:46 [ET]
Event Date: 04/02/2006
Event Time: 14:02 [EDT]
Last Update Date: 04/02/2006
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):
WILLIAM COOK (R1)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
2 A/R Y 100 Power Operation 0 Hot Standby

Event Text

AUTOMATIC REACTOR TRIP DUE TO A MAIN GENERATOR TRIP

"At 1402 Beaver Valley Unit 2 experienced a reactor trip due to a Main Generator trip. Unit 2 was operating at 100% power at the time of the trip [control rods inserted fully]. All systems functioned as expected. Emergency busses [are] being supplied by offsite power. Auxiliary Feedwater started as expected. Both Emergency Diesel Generators started on momentary undervoltage and remained operating unloaded.

"The Control Room Crew entered Emergency Operating procedure E-0, 'Response to Reactor Trip and Safety Injection' and have transitioned as expected to ES-0.1 'Reactor Trip Response.' The plant is stable in Mode 3 with steam generator levels and Reactor Coolant System temperature restored to normal values with Auxiliary Feed System operation.

"The cause of the Generator / Reactor trip is being investigated.

"No radiological releases ongoing or caused by this event.

"Beaver Valley Unit 1 is in Mode 5 for a Steam Generator Replacement outage and was unaffected."

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42468
Facility: PALO VERDE
Region: 4 State: AZ
Unit: [ ] [ ] [3]
RX Type: [1] CE,[2] CE,[3] CE
NRC Notified By: STEVE SMITH
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 04/03/2006
Notification Time: 03:56 [ET]
Event Date: 04/02/2006
Event Time: 20:54 [MST]
Last Update Date: 04/03/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
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
3 N N 0 Cold Shutdown 0 Cold Shutdown

Event Text

EMERGENCY DIESEL TRIP DURING TESTING

"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 April 2, 2006 at approximately 20:54 Mountain Standard Time, Palo Verde Nuclear Generating Station Unit 3 experienced a valid Loss of Power (LOP) actuation on the Train 'B' 4.16 kV safety bus. The event occurred due to personnel error during performance of surveillance test 73ST-9DG02 (Class 1 E Diesel Generator and Integrated Safeguards Test Train B). EDG 'B' had been started In Emergency Mode per the surveillance test by opening the normal supply breaker to the associated 4.16 kV bus and initiating simulated Safety Injection Actuation System (SIAS) and Containment Isolation Actuation System (CIAS) signals. A subsequent portion of the surveillance test was in progress which demonstrates that the EDG 'test mode' trips are bypassed with the EDG operating in Emergency Mode. The step being performed was intended to simulate an Overcurrent (test mode) trip by installing a jumper at the Overcurrent relay. However the jumper was inadvertently installed at the Differential Current relay, which generated an 'Emergency Mode' trip of EDG 'B'. This resulted in the deenergization of the 4.16 kV bus. The operations staff entered Abnormal Operating Procedure (AOP) 40AO-9ZZ12 (Degraded Electrical) and reset EDG 'B'. Upon reset at approximately 21:26 MST, EDG 'B' automatically started in response to a valid Loss of Power (LOP) signal from the deenergized 4.16 kV bus. The EDG 'B' output breaker automatically closed to restore power to the Train 'B' 4.16 kV and equipment sequenced onto the 4.16 kV bus.

"Due to the loss of power on the Train 'B' 4.16 kV bus, the Train 'B' Control Room Essential Filtration System (CREFS) and Control Room Emergency Air Temperature Control System (CREATCS) were rendered inoperable and LCOs 3.7.11 Condition 'A' and 3.7.12 Condition 'A' were entered. Operability of Train 'B' CREFS and CREATCS was restored when the Train 'B' 4.16 kV bus was reenergized and these LCO Conditions were exited.

"Offsite power remained available to the Train 'A' 4.16 kV bus and EDG 'A' remained operable throughout the event. Shutdown Cooling was unaffected since it was being provided by the Train 'A' safety train, which was supplied by offsite power. The offsite electrical grid is stable.

"Palo Verde Unit 3 Is shutdown and in Mode 5 for its 12th refueling outage. No other ESF actuations occurred and none were required. There were no structures, systems, or components that were inoperable at the time of discovery that contributed to this condition. The event did not result in the release of radioactivity to the environment and 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 the ESF actuation and this ENS notification."

The site performed a work stand down to discuss this event and prevent recurrence.

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