Alert

Due to a lapse in appropriations, the NRC has ceased normal operations. However, excepted and exempted activities necessary to maintain critical health and safety functions—as well as essential progress on designated critical activities, including those specified in Executive Order 14300—will continue, consistent with the OMB-Approved NRC Lapse Plan.

Event Notification Report for March 30, 2011

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
03/29/2011 - 03/30/2011

** EVENT NUMBERS **

!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor	Event Number: 46699
Facility: PRAIRIE ISLAND Region: 3 State: MN Unit: [1] [ ] [ ] RX Type: [1] W-2-LP,[2] W-2-LP NRC Notified By: TERRY BACON HQ OPS Officer: BILL HUFFMAN	Notification Date: 03/25/2011 Notification Time: 17:27 [ET] Event Date: 03/25/2011 Event Time: 10:53 [CDT] Last Update Date: 03/29/2011
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(B) - UNANALYZED CONDITION	Person (Organization): JAMNES CAMERON (R3DO)

Unit	SCRAM Code	RX CRIT	Initial PWR	Initial RX Mode	Current PWR	Current RX Mode
1	N	Y	99	Power Operation	98	Power Operation

Event Text

POTENTIAL UNANALYZED CONDITION DUE TO POWER LEVEL GREATER THAN LIMIT

"During performance of maintenance to troubleshoot the B feedwater regulating bypass valve, the Thermal Power Monitor (TPM) indication exceeded the maximum thermal power assumed in the Safety Analysis Report. Operators were maintaining 12 Steam Generator (SG) water level in a band from 40 to 48 percent by controlling the B Feed Regulating Valve (FRV) in manual from the Control Room. Operators noted a power increase; adjustments were made via the FRV to reduce SG water level, however the valve response was sluggish and thermal power exceeded 100%. Immediate steps were taken to reduce power to below 100% by reducing 1st stage turbine pressure and inserting Bank D control rods 7 steps.

"The TPM indication was above the maximum thermal power limit of 100.36% for 1.68 minutes. The TPM indication peak was 100.39%. No concurrent increase in power was observed by the nuclear indication system.

"NRC Resident had been informed."

* * * RETRACTION FROM JOHN KEMPKES TO JOHN SHOEMAKER AT 1350 EDT ON 03/29/11 * * *

"An eight hour report (EN #46699) per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on March 25, 2011 for Thermal Power Monitor indication above the maximum thermal power limit of 100.36% for 1.68 minutes.

"Subsequent engineering investigation has determined that this specific transient had been previously analyzed. The transient was within the bounds of the safety analysis.

"The 10 CFR 50.72(b)(3)(ii)(8) report (EN #46699) is retracted.

"NRC Resident has been informed."

Notified R3DO (Peterson).

Power Reactor	Event Number: 46704
Facility: NINE MILE POINT Region: 1 State: NY Unit: [1] [ ] [ ] RX Type: [1] GE-2,[2] GE-5 NRC Notified By: BETHANY HINCKLEY HQ OPS Officer: JOE O'HARA	Notification Date: 03/29/2011 Notification Time: 02:48 [ET] Event Date: 03/29/2011 Event Time: 01:55 [EDT] Last Update Date: 03/29/2011
Emergency Class: UNUSUAL EVENT 10 CFR Section: 50.72(a) (1) (i) - EMERGENCY DECLARED	Person (Organization): RAY POWELL (R1DO) WILLIAM DEAN (RA) JACK GROBE (NRR) SCOTT MORRIS (IRD) MIKE CHEOK (NRR) HILL (DHS) BARDEN (FEMA)

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

Event Text

ELECTRICAL FIRE IN THE DRYWELL ON A PIECE OF LIFTING EQUIPMENT

"At approximately 0145 EDT, the Unit 1 control room was notified of elevated carbon monoxide (CO) levels in the Unit 1 Drywell. The cause of the elevated CO was a small fire on a 'Lift-A-Loft'. The fire was immediately extinguished followed by an evacuation of all personnel from the Drywell.

"Follow up atmosphere samples indicated carbon monoxide levels above the OSHA Threshold of 50 PPM. Readings were as high as 79 PPM and have been slowly lowering since the initial response. These values are considered to affect the health of plant personnel or safe plant operation and an Unusual Event was declared at 0155 EDT.

"As of 0225 EDT, all values within the Unit 1 Drywell have been confirmed to be below the 50 PPM threshold.

"As of 0226 EDT, the Unusual Event has been terminated."

There were no injuries, no offsite assistance required, and the NRC Senior Resident Inspector responded to the site.

The licensee notified the NRC Senior Resident Inspector, the State Emergency Communication Center, and the Oswego County Warning Point. No press release is planned.

Power Reactor	Event Number: 46707
Facility: BRAIDWOOD Region: 3 State: IL Unit: [1] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: JOE KLEVORN HQ OPS Officer: HOWIE CROUCH	Notification Date: 03/30/2011 Notification Time: 01:34 [ET] Event Date: 03/29/2011 Event Time: 20:00 [CDT] Last Update Date: 03/30/2011
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(B) - UNANALYZED CONDITION 50.72(b)(3)(v)(B) - POT RHR INOP	Person (Organization): HIRONORI PETERSON (R3DO)

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

POTENTIAL VOIDING IN AUXILIARY FEEDWATER ALTERNATE SUCTION LINE

"The design of the Auxiliary Feedwater (AF) system is for the AF pumps to normally take suction from the condensate storage tank. If the condensate storage tank is not available, the essential service water system provides the alternate supply. Due to the AF system suction piping and valve configuration, a voided section of pipe could exist in the portion that isolates the condensate storage tank supply from the essential service water supply. A preliminary vendor analysis has determined that the void fraction to reach the pump in a dynamic scenario exceeds the acceptance criteria for AF pump operability. Based on past operation in this configuration, the event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). Further review of the void model and pump performance characteristics are planned.

"In 2011, prior to the completion of this analysis, the void was refilled and verified full for the 'B' trains at Braidwood U1 and U2. Filling the voided piping of both 'A' trains at Braidwood U1 and U2 is in progress. Once filled, the AF systems are operable."

The licensee has notified the NRC Resident Inspector.

Power Reactor	Event Number: 46708
Facility: BYRON Region: 3 State: IL Unit: [1] [2] [ ] RX Type: [1] W-4-LP,[2] W-4-LP NRC Notified By: ALAN GUSTAFSON HQ OPS Officer: JOE O'HARA	Notification Date: 03/30/2011 Notification Time: 01:39 [ET] Event Date: 03/29/2011 Event Time: 20:00 [CDT] Last Update Date: 03/30/2011
Emergency Class: NON EMERGENCY 10 CFR Section: 50.72(b)(3)(ii)(B) - UNANALYZED CONDITION 50.72(b)(3)(v)(B) - POT RHR INOP	Person (Organization): HIRONORI PETERSON (R3DO)

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

Event Text

POTENTIAL VOIDING IN AUXILIARY FEEDWATER ALTERNATE SUCTION LINE

"The design of the Auxiliary Feedwater (AF) system is for the AF pumps to normally take suction from the condensate storage tank. If the condensate storage tank is not available, the essential service water system provides the alternate supply. Due to the AF system suction piping and valve configuration, a voided section of pipe could exist in the portion that isolates the condensate storage tank supply from the essential service water supply. A preliminary vendor analysis has determined that the void fraction to reach the pump in a dynamic scenario exceeds the acceptance criteria for AF pump operability. Based on past operation in this configuration, the event is being reported as a unanalyzed condition that significantly degrades plant safety and a condition that could have prevented the fulfillment of a safety function (i.e., remove residual heat) under 10CFR50.72(b)(3)(ii) and 10CFR50.72(b)(3)(v). Further review of the void model and pump performance characteristics are planned.

"In 2011, prior to the completion of this analysis. The void was refilled and verified full for both trains at Byron U1 and U2."

Unit 1 is defueled. This condition affects both 'A' and 'B' trains of auxiliary feedwater for both Unit 1 and Unit 2.

The NRC Resident Inspector has been notified.

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