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Event Notification Report for July 27, 2011

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
07/26/2011 - 07/27/2011

** EVENT NUMBERS **


46909 46917 47082 47083 47094 47096

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 46909
Facility: WOLF CREEK
Region: 4 State: KS
Unit: [1] [ ] [ ]
RX Type: [1] W-4-LP
NRC Notified By: TIM DUNLOP
HQ OPS Officer: BILL HUFFMAN
Notification Date: 06/01/2011
Notification Time: 17:07 [ET]
Event Date: 06/01/2011
Event Time: 13:15 [CDT]
Last Update Date: 07/26/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
50.72(b)(3)(v)(B) - POT RHR INOP
Person (Organization):
BLAIR SPITZBERG (R4DO)

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

Event Text

BOTH TRAINS OF COMPONENT COOLING WATER DECLARED INOPERABLE DUE TO VOIDING

"At 1315 (CDT), while in Mode 3 at normal operating pressure and 552?F, both trains of Component Cooling Water (CCW) were inoperable due to indications of voiding. The 'A' Train CCW had been declared inoperable at 1000 (CDT) when review of pump test data indicated a potential void and Technical Specification LCO 3.7.7, 'Component Cooling Water (CCW) System,' Condition A was entered. At 1315 (CDT), indications of voiding were identified in the common service loop piping, which was aligned to the 'B' Train CCW. The 'B' Train CCW was declared inoperable and the plant entered LCO 3.0.3 due to both CCW Trains being inoperable.

"At 1410 (CDT) plant cool down to Mode 4 was commenced. The 'A' Train CCW has subsequently been vented and void volume is currently within allowable limits for operability. However, further evaluation of this voiding is underway prior to declaring the 'A' Train CCW operable."

The licensee has notified the NRC Resident Inspector.

* * * RETRACTION ON 7/26/11 AT 1305 EDT FROM ISCH TO HUFFMAN * * *

The licensee is retracting this event based on the following:

"Further engineering evaluation concluded the amount of gas ingested by the 'A' CCW pump would not cause any degradation to the pump. The remainder of the gas in the system was less than the acceptance criteria for the CCW system. The 'A' CCW train was capable of performing its specified safety function and therefore would have been considered operable. The condition would not have prevented the CCW System from fulfilling its safety function and would not be reportable under 10 CFR 50.72."

The licensee has notified the NRC Resident Inspector. R4DO (Gaddy) notified.

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 46917
Facility: GINNA
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: CRAIG JONES
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 06/03/2011
Notification Time: 05:54 [ET]
Event Date: 06/03/2011
Event Time: 00:39 [EDT]
Last Update Date: 07/26/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(iv)(A) - VALID SPECIF SYS ACTUATION
Person (Organization):
RICHARD CONTE (R1DO)

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

UNEXPECTED EMERGENCY DIESEL GENERATOR ACTUATION

"On 6/3/2011 at 0039 hours, during the performance of a work order to test components associated with Service Water Isolation, Emergency Diesel Generator (EDG) 'A' unexpectedly started automatically and its supply breaker to Safeguards Bus 14 closed. The Control Room staff observed normal voltage on Diesel Generator 'A'. Bus 14 voltage was never lost during this event, however, they also noted an associated Bus 14 undervoltage annunciator on the Main Control Board. Seconds later, Emergency Diesel Generator 'A' tripped on Reverse Power and its supply breaker to Bus 14 tripped open. The initiating action was the removal of the Bus 14 Normal Feed Breaker Control Power Fuses as part of the work order package.

"The Ginna EDG's have the following automatic start signals and logic: manual, safety injection signal (1/2 trains), undervoltage on respective safeguards bus, 'A' EDG Bus 14 or 18 (1 out of 2 degraded voltage + 1 out of 2 loss of voltage), 'B' EDG Bus 16 or 17 (1 out of 2 degraded voltage + 1 out of 2 loss of voltage).

"Investigation has commenced to determine the cause of the EDG start and undervoltage signal.

"The NRC Resident Inspector has been notified."

* * * RETRACTION ON 7/26/11 AT 1214 EDT FROM SLABY TO HUFFMAN * * *

"The purpose of this report is to retract the event discussed in Emergency Notification System report #46917 submitted on June 3rd, 2011. The ENS notification reported an unexpected start of Emergency Diesel Generator `A' during testing of a service water valve isolation circuit. As reported, Emergency Diesel Generator 'A' unexpectedly started and its supply breaker to Bus 14 closed. Seconds later, the Emergency Diesel Generator tripped on reverse power and its output breaker to Bus 14 opened. At the time of the event it was not understood why the diesel generator started. Subsequent troubleshooting and causal investigation identified that the signal was caused by a degraded control relay that unexpectedly changed state when control power was removed. This relay was expected to remain mechanically latched and would have remained in the desired position had control power not been removed as part of the test. Bus 14 voltage remained in the normal operating range throughout the event. Since this was not a valid undervoltage signal, the June 3rd, 2011 event is being retracted. A follow-up report will be made in accordance with 10CFR50.73(a)(1) and 10CFR50.73(a)(2)(iv)."

The NRC Resident Inspector has been notified. R1DO(Henderson) notified. See related EN #47094.

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Agreement State Event Number: 47082
Rep Org: TEXAS DEPARTMENT OF HEALTH
Licensee: BRAZOS VALLEY INSPECTION SERVICES, INC
Region: 4
City: BRYAN State: TX
County:
License #: 02859
Agreement: Y
Docket:
NRC Notified By: ART TUCKER
HQ OPS Officer: DONALD NORWOOD
Notification Date: 07/21/2011
Notification Time: 08:38 [ET]
Event Date: 07/20/2011
Event Time: [CDT]
Last Update Date: 07/21/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
VIVIAN CAMPBELL (R4DO)
RON ZELAC (FSME)

Event Text

AGREEMENT STATE REPORT - FAILURE OF RADIOGRAPHY CAMERA LOCKING MECHANISM

The following information was received via facsimile:

"On July 20, 2011, the Agency [Texas Department of Health] was notified by the licensee that they had completed radiography operations at a field site using a INC IR-100 radiography camera containing an 82 curie iridium-192 source and retracted the source to its fully shielded position. The radiographer surveyed the camera and found the readings to be normal. When he disconnected the drive cable from the source pigtail, he found that the pigtail was no longer protruding from the back of the camera, it was now flush with the rear of the device. The shipping plug and the front dust cover were placed on the camera. The camera is being returned to the manufacturer for repair. Additional information will be provided as it is received. The licensee believes that the locking device that holds the pigtail in place inside the camera has failed. The serial number for the camera is 4590."

Texas Incident Report Number: I-8873

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Agreement State Event Number: 47083
Rep Org: FLORIDA BUREAU OF RADIATION CONTROL
Licensee: UNIVERSAL ENGINEERING SCIENCES, INC.
Region: 1
City: FT. PIERCE State: FL
County:
License #: 1136-12
Agreement: Y
Docket:
NRC Notified By: CHARLES E. ADAMS
HQ OPS Officer: STEVE SANDIN
Notification Date: 07/21/2011
Notification Time: 12:42 [ET]
Event Date: 07/20/2011
Event Time: [EDT]
Last Update Date: 07/21/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
GLENN DENTEL (R1DO)
DUNCAN WHITE (FSME)

Event Text

AGREEMENT STATE REPORT - DAMAGED TROXLER MOISURE DENSITY GAUGE

The following report was received via email from the State of Florida:

"An accident caused damage to a moisture density gauge at a work site on the Kissimmee River Restoration/CSX Bridge project. The operator of a large earthmoving vehicle lost control and ran over a parked truck with the gauge sitting on the tailgate. The truck & contents were mangled and buried under the earthmoving vehicle. The tech was not injured. A 15 foot area was roped off around the accident site. When the gauge was recovered it appeared to have some damage to the top housing. Survey readings were elevated to around 40-50 mR/hr at one foot. Using shielding, the gauge was transported to the Troxler office in Orlando for repair. Licensee sent a written report to this office. This office will take no further action on this incident."

The device is a Troxler Model No. 3440 S/N 37720 with two sources; 9.0 mCi Cs-137 and 44 mCi Am-241/Be.

Florida Incident No.: FL-11-062

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Power Reactor Event Number: 47094
Facility: GINNA
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: MIKE SLABY
HQ OPS Officer: BILL HUFFMAN
Notification Date: 07/26/2011
Notification Time: 12:14 [ET]
Event Date: 06/03/2011
Event Time: 00:39 [EDT]
Last Update Date: 07/26/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.73(a)(1) - INVALID SPECIF SYSTEM ACTUATION
Person (Organization):
PAMELA HENDERSON (R1DO)

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

INVALID ACTUATION OF EMERGENCY DIESEL GENERATOR

"This report is being made per paragraphs 50.73(a)(1) and 50.73(a)(2)(iv)(A) to address an actuation of Emergency Diesel Generator 'A' on June 3rd, 2011 while performing service water valve isolation signal testing. Emergency AC Electrical power system, including emergency diesel generators, is a system named in 50.73(a)(2)(iv)(B).

"During a refueling outage, testing was in progress to verify that service water isolation valves received the proper close signal during a safety injection. The test configuration required pulling control power fuses for the Bus 14 normal supply breaker to prevent operation and allow for manual relay actuation. Within seconds of pulling these fuses, the control room received a Bus 14 undervoltage annunciator, Emergency Diesel Generator 'A' started, and the generator output breaker closed onto Bus 14. Upon further investigation, the cause of this signal was identified as a degraded control relay that failed to mechanically latch and unexpectedly changed state when control power was removed. This resulted in an invalid undervoltage signal. Bus voltage remained within normal operating range. Given that the diesel generator was in unit mode of operation and was not fully synchronized with the normal bus supply, the diesel generator tripped shortly after starting due to a valid reverse power signal. A field verification and technical review was performed to ensure that this condition did not cause significant stress on the generator or engine.

"This start signal is considered an INVALID signal with respect to 50.73(a)(2)(iv)(A), however the system was not fully removed from service. The 'B' train was not affected by this event. The actuation was considered complete since all necessary components responded to the undervoltage signal as expected under the actual field conditions. The control relay would have remained in the desired position and performed its required function under design conditions with normal control power available. Therefore the degradation was not determined to have an impact on the safety function."

The NRC Resident Inspector was notified. See related EN #46917.

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Power Reactor Event Number: 47096
Facility: DAVIS BESSE
Region: 3 State: OH
Unit: [1] [ ] [ ]
RX Type: [1] B&W-R-LP
NRC Notified By: TOM COBBLEDICK
HQ OPS Officer: BILL HUFFMAN
Notification Date: 07/26/2011
Notification Time: 16:47 [ET]
Event Date: 07/26/2011
Event Time: 16:00 [EDT]
Last Update Date: 07/26/2011
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(A) - POT UNABLE TO SAFE SD
50.72(b)(3)(v)(B) - POT RHR INOP
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
JOHN GIESSNER (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

Event Text

UNANALYZED CONDITIONS INVOLVING THE SAFETY RELATED DIRECT CURRENT (DC) SYSTEM

"Information was received in regards to an old design issue identified in a Component Design Basis Inspection Unresolved Item. Two issues were identified with the Safety-Related Direct Current (DC) System:

"1. The plant's licensing basis states that non-safety-related electrical equipment, whose failure under postulated environmental conditions could prevent satisfactory accomplishment of the specified safety-related electrical equipment required safety functions, is qualified as required. However, the Reactor Coolant Pump (RCP) backup lift oil pump motors and the Containment Emergency Lighting Panel L49E1 are located inside containment and are not environmentally qualified. This could challenge the adequacy of electrical separation between the potentially grounded non-safety related equipment and the safety related batteries.

"2. Automatic transfer switches are installed to automatically transfer non-safety related loads such as non-nuclear instrumentation, station annunciators, plant computer, and integrated control system between two non-safety related inverters, which receive power from the safety-related DC power system. If a ground fault existed on one of these switches, the fault could be transferred from one power source to the redundant source, potentially impacting the ability of both safety-related DC power sources to perform their required functions. This type of transfer is not permitted by the plant's licensing basis.

"The breakers for the 4 RCP backup lift oil pump motors and for the Containment Emergency Lighting were opened. One train of instrumentation power was placed on its alternate power source from the Alternating Current (AC) system, eliminating the potential to impact both trains of the DC power system.

"This condition is being reported per 10 CFR 50.72(b)(3)(ii)(B) as a condition that results in the plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v)(A-D) as an event or condition that could have prevented fulfillment of a safety function."

The licensee has notified state and local authorities and the NRC Resident Inspector.

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