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Event Notification Report for November 13, 2006

U.S. Nuclear Regulatory Commission
Operations Center

Event Reports For
11/10/2006 - 11/13/2006

** EVENT NUMBERS **


42846 42880 42929 42959 42965 42975 42983 42985 42986

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 42846
Facility: GINNA
Region: 1 State: NY
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: KYLE GARNISH
HQ OPS Officer: BILL GOTT
Notification Date: 09/18/2006
Notification Time: 22:23 [ET]
Event Date: 09/18/2006
Event Time: 15:00 [EDT]
Last Update Date: 11/10/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
EUGENE COBEY (R1)

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

LOSS OF SPENT FUEL POOL (SFP) COOLING

"At 1500 on 9/18/06, the 'A' SFP pump was declared inoperable due to abnormal noise. The 'B' SFP pump had previously been declared inoperable at 2330 on 9/12/06 [due to a vibration problem]. These events have resulted in a loss of SFP cooling. Current SFP temperature if 89F. Per the Technical Requirements Manual, SFP temperature is being verified within limits of less than or equal to 150 F once per 4 hours, and actions have been initiated to restore one SFP cooling system to OPERABLE status prior to the SFP exceeding 120 F."

The licensee is attempting to repair both SFP pumps in parallel efforts. In addition, a skid mounted SFP cooling system is being prepared for service. The spent fuel pool temperature was at 80 F when cooling was lost and has increased to 89 F at the time this report was made. The licensee has calculated that the SFP cooling will have to be out of service for 48 hours to reach the 150 F limit.

The licensee notified the NRC Resident Inspector.

* * *RETRACTION FROM D. DUTTON TO J. KOZAL ON 11/10/2006 AT 1355 * * *

"The purpose of this notification is to retract the ENS report made on 9/18/2006 at 2223 EDT (Event # 42846). The initial report was conservatively made when both trains of spent fuel pool cooling were lost. Ginna reported the situation as an event or condition that could have prevented the fulfillment of a safety function per 10 CFR 50.72 (b)(3)(v)(D).

"The spent fuel pool cooling system is not credited to mitigate any design basis event, and is not required by the Technical Specifications. The Ginna UFSAR credits the volume of water in the spent fuel pool for providing cooling should forced spent fuel pool cooling be lost."

"Based on this, ENS report (# 42846) is being retracted."

The licensee notified the NRC Resident Inspector.

Notified R1DO (Perry).

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General Information or Other Event Number: 42880
Rep Org: SOUTHERN TESTING SERVICES
Licensee: OTEK CORPORATION
Region: 1
City: KNOXVILLE State: TN
County:
License #:
Agreement: Y
Docket:
NRC Notified By: WILLIAM R. WILLIS
HQ OPS Officer: PETE SNYDER
Notification Date: 10/05/2006
Notification Time: 19:20 [ET]
Event Date: 10/05/2006
Event Time: [EDT]
Last Update Date: 11/10/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
21.21 - UNSPECIFIED PARAGRAPH
Person (Organization):
PAUL KROHN (R1)
THOMAS DECKER (R2)
VERN HODGE (Email) ()

Event Text

POTENTIAL DEFECT IN OTEK PANEL METERS

Southern Testing Services (STS) provided information on a potential defect in OTEK Panel Meters. The information applies to 90-265 Volt AC / 10- 32 Volt DC, OTEK Part Numbers: HI-Q114, HI-Q116, HI-Q117, HI-Q118, and HI-Q119.

"STS has identified a potential failure cause for the HI-Q class meters that would result in a frozen indicator with no indication that the failure has occurred. The specific failure mode is a failure of the main processor (only) that freezes the display processor. The frozen display inhibits the detection of the main processor failure except by cycling power to the Panel Meter. This failure mode was previously addressed in the Failure Mode Effects Analysis (FMEA) conducted by STS in Test Report S4000-RP-03; however, no credible causes for the failure beyond infant mortality were postulated or had been experienced at the time. Infant mortality failures are minimized by the dedication process that verifies operation of the units prior to shipment to the client.

"STS has just learned from OTEK that of one (1) OTEK Panel Meter had experienced a main processor failure which resulted in a frozen display indicator. This Panel Meter was reportedly in a non-nuclear application. OTEK determined that the likely cause of this failure was a high frequency spike on the power lines caused by running the meter off of a DC generator.

"The STS EMI/RFI qualification program for the OTEK meters qualified them with an anomaly, limiting the surge protection to 500 [Volts peak to peak] for DC powered units. Additionally, high frequency susceptibility testing was successfully completed with a continuous signal of 3.5 [Volts RMS] on the power lines (both AC and DC powered units) from 10 kHz to 200 MHz. Additional testing was done on AC powered units at 7 [Volts RMS]. Testing was conducted in accordance with EPRI TR-102323 to the methods specified in EN 61000-46. The units passed the susceptibility testing at the limits specified, as reported in Test Report S4000-RP-03.

"Based on the successful surge and high frequency susceptibility testing conducted by STS, it is concluded that the single known failure noted by OTEK was related to operating outside of the parameters tested during qualification of these OTEK meters. The likelihood of such a failure in a nuclear safety related application is considered to be remote."

STS determined that 294 units were provided to nuclear plants in potentially safety related applications where EMI/RFI requirements were imposed. The following plants were effected purchasers: Vermont Yankee, Pilgrim, St. Lucie, and Brown's Ferry.

"OTEK is currently in the process of updating the display board processor programming to detect a failure of the main processor, and provide an indication on the display that a main processor failure has occurred. STS will coordinate with the above utilities when this update becomes available."

* * * UPDATE FROM SOUTHERN TESTING SERVICES (W. WILLIS) TO M. RIPLEY 0840 EST 11/10/06 * * *

The following information was provided via facsimile:

"The date on which the information of such defect or failure to comply was obtained: OTEK notified STS via email on September 18, 2006 following in-person conversations on September 12, 2006 with Southern Testing Services and TVA personnel.

"In case of a basic component which contains a defect or fails to comply, the number and location of all such components in use at, supplied for, or being supplied for one or more facilities or activities subject to the regulations in this part.

"A total of 294 units have been provided to Nuclear Plants in potentially safety-related applications where EMI/RFI requirements were imposed. The following Effected Purchasers have been identified: Vermont Yankee, St. Lucie, Pilgrim, Brown's Ferry

"The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action, and the length of time that has been or will be taken to complete the action:
OTEK has provided a prototype meter with the display processor code modified to detect failure of the main processor and provide an indication on the meter by generating a full scale bargraph colored amber. STS has successfully performed testing on the unit to confirm proper operation of the modified display processor code. STS will coordinate with the above utilities when testing of this update is complete to upgrade or replace the existing units if required by the utilities. OTEK will include the modified software in the HI-Q series meters in all new orders, starting immediately.

"Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees:
STS and Otek acknowledge the definition of 'Defect' within 10 CFR, Part 21 but believe the issue is more accurately described as an absence of a specific failure mode indication that was not previously required."

Notified R1 DO (N. Perry), R2 DO (P. Fredrickson), and NRR Part 21 (EMAIL & FAX)

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 42929
Facility: CLINTON
Region: 3 State: IL
Unit: [1] [ ] [ ]
RX Type: [1] GE-6
NRC Notified By: PAT RYAN
HQ OPS Officer: JOHN MacKINNON
Notification Date: 10/24/2006
Notification Time: 09:08 [ET]
Event Date: 10/24/2006
Event Time: 01:42 [CDT]
Last Update Date: 11/10/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
MARK RING (R3)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 N Y 96 Power Operation 96 Power Operation

Event Text

HIGH PRESSURE CORE SPRAY (HPCS) DECLARED INOPERABLE FOR APPROX. 3.4 HOURS

"At 0142 on October 24, 2006, while aligning the High Pressure Core Spray system for surveillance testing of the Reactor Core Isolation Cooling System Storage Tank Level instrumentation, 1E22-F015, the Suppression Pool suction valve for the High Pressure Core Spray pump, failed to stroke fully open. High Pressure Core Spray was declared inoperable as a result. This event is considered a loss of a single train system needed to mitigate the consequences of an accident.

The High Pressure Core Spray system was restored to an operable condition at 0506 on October 24, 2006 after the suction valve was successfully stroked open and the HPCS suction source was aligned to the Suppression Pool in accordance with Technical Specification Limiting Condition for Operation 3.5.1. The cause of the event is currently under investigation."

All other Emergency Core Cooling systems were fully operable during the time period HPCS was inoperable.

The Senior Resident Inspector has been notified by the licensee.

* * * RETRACTION FROM SIMPSON TO HUFFMAN AT 1534 EST ON 11/10/06 * * *

"Upon further review of this event, the High Pressure Core Spray (HPCS) system remained operable. Based upon valve motor operator thrust verification testing data and troubleshooting, the cause of the suppression pool suction valve for the HPCS pump stopping in mid-position was determined to be tripping of the open-direction torque switch prior to the open limit switch setpoint.

"Normally, the condition of the open-direction torque switch has no safety-related consequence since the torque switch is bypassed during design basis events and the valve's motor gearing capability is sufficient to open the valve when the torque switch is bypassed.

"During this event, as directed by the surveillance test procedure, operators placed the HPCS Motor Operated Valve (MOV) test switch to the test position which resulted in the open-direction torque switch not being bypassed (i.e., was in the circuit) during repositioning of the HPCS suppression pool suction valve. Due to placing the HPCS MOV test switch to test, operators entered the action of Operational Requirements Manual section 2.5.2 (Motor Operated Valves Thermal Overload Protection). The action requires operators to return the MOV test switch to normal (removing the torque switch from the circuit) if an emergency condition occurs requiring valve repositioning.

"As operators were opening the HPCS suppression pool suction valve for testing, suction for the HPCS pump was aligned from the RCIC storage tank. When the HPCS suction valve from suppression pool stopped in mid-position, the HPCS suction valve from the RCIC storage tank was still fully open (per design, stays full open until the HPCS suppression pool suction valve is full open). Therefore, if an accident occurred requiring HPCS to initiate and inject water into the reactor pressure vessel during this event suction would have initiated from the RCIC storage tank. The HPCS system can take suction from either the RCIC storage tank or the suppression pool, and a HPCS initiation signal does not automatically swap HPCS pump suction from the RCIC storage tank to the suppression pool or vice versa.

"The operators immediately recognized the HPCS suppression pool suction valve did not fully open. If an accident condition occurred, operators would reposition the HPCS MOV test switch to Normal (to bypass the open torque switch). In the event a condition requiring a HPCS suction transfer to the suppression pool occurred, the suppression pool suction valve would fully open and the RCIC storage tank suction valve would fully close, completing the required suction shift.

"On this basis, the HPCS system was capable of performing its function to mitigate the consequences of an accident and this issue is not reportable under 10 CFR 50.72(b)(3)(v)(D)."

The NRC Resident was notified of this retraction. R3DO(Cameron) notified.

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!!!!! THIS EVENT HAS BEEN RETRACTED. THIS EVENT HAS BEEN RETRACTED !!!!!
Power Reactor Event Number: 42959
Facility: PALISADES
Region: 3 State: MI
Unit: [1] [ ] [ ]
RX Type: [1] CE
NRC Notified By: DAN MALONE
HQ OPS Officer: JOHN MacKINNON
Notification Date: 11/02/2006
Notification Time: 19:10 [ET]
Event Date: 11/02/2006
Event Time: 14:36 [EST]
Last Update Date: 11/10/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
50.72(b)(3)(ii)(B) - UNANALYZED CONDITION
50.72(b)(3)(v)(D) - ACCIDENT MITIGATION
Person (Organization):
MONTE PHILLIPS (R3)

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

HIGH PRESSURE SAFETY INJECTION (HPSI) PUMPS ALIGNMENT BLOCKS IMPROPERLY INSTALLED

"At 1436 hours on November 2, 2006, with the plant in Mode 3, it was determined that less than 100% of the required Emergency Core Cooling System (ECCS) flow was available per Technical Specification (TS) 3.5.2.D. Therefore, TS LCO 3.0.3 was entered.

"Each High Pressure Safety Injection (HPSI) pump (one in each ECCS train) is designed with alignment blocks in its mounting, which ensures pump and motor alignment for the thermal expansion experienced by the pump upon initiation of sump recirculation flow. These alignment blocks ( 2 per pump) were discovered to be improperly installed, or missing altogether. In this condition, the HPSI pump could potentially be rendered inoperable upon initiation of sump recirculation.

"This condition is reportable in accordance with 10CFR 50.72(b)(3)(ii)(B) and (b)(3)(v)(D) as an unanalyzed condition, and a condition that could have prevented the fulfillment of the safety function of the HPSI pumps to mitigate the consequences of an accident, respectively."

The NRC Resident Inspector was notified of this event by the licensee.

* * * RETRACTION FROM MALONE TO HUFFMAN AT 1335 EST ON 11/10/06 * * *

"EN # 42959 reported on November 2, 2006 that both Emergency Core Cooling System Trains were inoperable. The reason for that determination involved the observation that the alignment blocks [keys] associated with the mounting of the high pressure safety injection (HPSI) pumps were either incorrectly installed or were missing. The alignment keys were believed to be necessary to ensure appropriate HPSI pump and motor alignment for the thermal expansion experienced upon initiation of sump recirculation flow. The condition was reported as an unanalyzed condition and a condition that could have prevented the fulfillment of the safety function of the HPSI pumps to mitigate the consequences of an accident.

"Subsequently, further evaluation of the HPSI pump mounting configuration determined that the alignment keys are not required for pump operability. Therefore, there was no unanalyzed condition and no condition that could have prevented the fulfillment of the safety function of the HPSI pumps to mitigate the consequences of an accident."

The licensee notified the NRC Resident Inspector. R3DO(Cameron) notified.

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General Information or Other Event Number: 42965
Rep Org: TEXAS DEPARTMENT OF HEALTH
Licensee: BEN TAUB GENERAL HOSPITAL
Region: 4
City: HOUSTON State: TX
County:
License #: L01303
Agreement: Y
Docket:
NRC Notified By: RAY JISHA
HQ OPS Officer: JOHN MacKINNON
Notification Date: 11/05/2006
Notification Time: 18:07 [ET]
Event Date: 11/05/2006
Event Time: 16:00 [CST]
Last Update Date: 11/06/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
AGREEMENT STATE
Person (Organization):
MICHAEL SHANNON (R4)
ED HACKETT (NMSS)

This material event contains a "Less than Cat 3" level of radioactive material.

Event Text

TEXAS AGREEMENT STATE REPORT - MISSING SOURCE

Two Radiation Oncologist Residents removed an LDR applicator from a cervical cancer patient. The applicator was supposed to have contained four Cesium-137 sources, 15 millicuries each, but only 3 of the sources were found. The patient's bed sheets had been changed and taken out of the room. Search for missing source is in progress.

* * * UPDATE FROM R. JISHA TO P. SNYDER AT 0843 ON 11/6/06 * * *

The lost source material was found in the laundry. The state continues to investigate. Further information will be provided later.

Notified R4DO (J. Clark) and NMSS EO (M. Burgess).

A "Medical Event" may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient.

THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL

Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks.

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Hospital Event Number: 42975
Rep Org: WILLIAM BEAUMONT HOSPITAL
Licensee: WILLIAM BEAUMONT HOSPITAL
Region: 3
City: Royal Oak State: MI
County:
License #: 21-01333-01
Agreement: N
Docket:
NRC Notified By: CHERYL SCHULTZ
HQ OPS Officer: BILL HUFFMAN
Notification Date: 11/08/2006
Notification Time: 17:39 [ET]
Event Date: 11/07/2006
Event Time: 13:45 [EST]
Last Update Date: 11/08/2006
Emergency Class: NON EMERGENCY
10 CFR Section:
35.3045(a)(1) - DOSE <> PRESCRIBED DOSAGE
Person (Organization):
JAMNES CAMERON (R3)
KEITH McCONNELL (NMSS)

Event Text

MEDICAL EVENT - LESS THAN PRESCRIBED DOSE OF YTTRIUM - 90

The Radiation Safety Office for the licensee reported an event where a patient received less than the prescribed dose during a treatment for liver cancer using Yttrium - 90 microspheres. Specifically, the patient was prescribed 9.8 millicuries to the liver using Yttrium - 90 SirTex Sirspheres using a intrahepatic catheter. The patient only received 6.5 millicuries due to problems in the administration of the dose.

After administering about half of the treatment dose, the physician started to encounter injection resistance which is not uncommon with this treatment due to vasculature loading. The physician stopped the treatment and was trying to view the microsphere placement in the liver using fluoroscopy when he noted some unusual "clumping" of the microspheres between the delivery vial and a 3-way stop cock that connects to the catheter. Because of the clumping and the resistance, the physician elected to discontinue the administration of the remainder of the dose.

The licensee has contacted SirTex and plans to send the delivery device with the clumped microspheres to SirTex when the Yttrium - 90 has decayed away (in a couple weeks) for further evaluation of the product.

The licensee has also contacted the Region 3 NRC inspector (Piskura) about this event.

A "Medical Event" may indicate potential problems in a medical facility's use of radioactive materials. It does not necessarily result in harm to the patient.

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Power Reactor Event Number: 42983
Facility: KEWAUNEE
Region: 3 State: WI
Unit: [1] [ ] [ ]
RX Type: [1] W-2-LP
NRC Notified By: SCOTT LIESLEWICZ
HQ OPS Officer: MARK ABRAMOVITZ
Notification Date: 11/10/2006
Notification Time: 17:33 [ET]
Event Date: 11/10/2006
Event Time: 14:20 [CST]
Last Update Date: 11/10/2006
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):
JAMNES CAMERON (R3)

Unit SCRAM Code RX CRIT Initial PWR Initial RX Mode Current PWR Current RX Mode
1 A/R Y 10 Power Operation 0 Hot Shutdown

Event Text

AUTOMATIC REACTOR TRIP AT LOW POWER

"On 11/10/2006 with a shutdown in progress to repair a degraded bearing on the turbine generator, an automatic reactor trip occurred due to a power range nuclear instrumentation (NI) low range - high flux trip. Reactor power had just been lowered to below 10% power (P-10) where the power range (NI) low range trips become active. The bistable for power range NI N-42 had been tripped due to an unrelated failure on 11/09/2006. When P-10 automatically unblocked, a power range NI low range high flux reactor trip was generated. At the time of the trip, reactor power was well below the trip setpoint of 24.5% power.

"Following the trip, Main Feedwater Regulating Valve, FW-7A, did not automatically close as required on the reactor trip coincident with Low Tave (554F). The Reactor Operator reported FW-7A was mid-position and attempted to manually close FW-7A. It did not respond. As a result, levels in steam generator A rose to greater than 67%, which initiated feedwater isolation. The feedwater isolation signal tripped the running feedwater pump. With no feedwater pumps running, both Auxiliary Feedwater Pump A and Auxiliary Feedwater Pump B automatically started as required. The High-High steam generator level also resulted in a second reactor trip initiation signal. The Reactor Operator manually controlled Auxiliary Feedwater flow to steam generator A to restore normal level. Following the feedwater isolation, FW-7A fully closed.

"Following the trip, MS-201B1, the steam supply to main steam reheater B1 was locally isolated to limit the RCS cool down. This was a previously discussed contingency action. Main steam isolation valves remained open and normal condenser heat sink remained available.

"Further investigation as to the cause of the trip is in progress. Recovery actions per normal operating procedures are in progress."

The plant was being shut down at a rate of a half percent power per minute at the time of the trip. All control rods fully inserted on the reactor trip and no safety or relief valves lifted. The plant was aligned for the normal shutdown electrical lineup prior to the trip. The temperature on the generator bearing reached a maximum of 190F with trip guidance set at 225F.

The licensee notified the NRC Resident Inspector.

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Power Reactor Event Number: 42985
Facility: COOPER
Region: 4 State: NE
Unit: [1] [ ] [ ]
RX Type: [1] GE-4
NRC Notified By: SCOTT COON
HQ OPS Officer: STEVE SANDIN
Notification Date: 11/11/2006
Notification Time: 06:46 [ET]
Event Date: 11/11/2006
Event Time: 05:30 [CST]
Last Update Date: 11/11/2006
Emergency Class: UNUSUAL EVENT
10 CFR Section:
50.72(a) (1) (i) - EMERGENCY DECLARED
50.72(b)(2)(xi) - OFFSITE NOTIFICATION
Person (Organization):
JEFFREY CLARK (R4)
BILL BATEMAN (NRR)
PETER WILSON (IRD)
GOMEZ (SWO) (DHS)
KEN SWEETSER (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

UNUSUAL EVENT DECLARED DUE TO AN ELECTRICAL FIRE INVOLVING HEAT TRACE

At 0530 CST, CNS declared an Unusual Event per EAL 5.1.1 "Any fire within the Protected Area which takes longer than 10 minutes to extinguish," due to a fire involving the Standby Liquid Control (SLC) heat tracing. Operators in the control room received a report from the Refueling Floor that heat trace installed on piping of the SLC System in Reactor Building 976 East was "arcing & sparking." Responders on-scene extinguished the fire after de-energizing the heat trace by opening the associated electrical feeder breaker. All personnel are accounted for and there are no reported injuries.

The licensee will conduct a comprehensive review to determine the extent of damage. The SLC System is a TS required safety system but is not required for current plant conditions. No other safety-related equipment was affected. The licensee is not augmenting staff in response to this event and no offsite assistance was requested.

The licensee informed State and local agencies and the NRC Resident Inspector who is enroute to the site.

* * * UPDATE AT 0733 EST ON 11/11/06 FROM STEVEN JOBE TO S. SANDIN * * *

The licensee terminated the Unusual Event at 0558 CST on 11/11/06 based on the judgment of Station Management since the extent of condition is well understood, the affected electrical circuit is isolated and that no reflash watch is necessary. The licensee informed State and local agencies and the NRC Resident Inspector.

Notified R4DO (Clark), EO (Bateman), IRD (Wilson), DHS (Gomez) and FEMA (Sweetser).

* * * UPDATE AT 0942 EST ON 11/11/06 FROM ED McCUTCHEN TO S. SANDIN * * *

The licensee will be issuing a press release regarding this incident and has informed the NRC Resident Inspector. Notified R4DO (Clark).

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Power Reactor Event Number: 42986
Facility: BRUNSWICK
Region: 2 State: NC
Unit: [ ] [2] [ ]
RX Type: [1] GE-4,[2] GE-4
NRC Notified By: MARK SCHALL
HQ OPS Officer: JASON KOZAL
Notification Date: 11/11/2006
Notification Time: 16:31 [ET]
Event Date: 11/11/2006
Event Time: 12:43 [EST]
Last Update Date: 11/11/2006
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):
PAUL FREDRICKSON (R2)

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

Event Text

REACTOR SCRAM DUE TO CONDUCTIVITY EXCURSION

"At 1243 EST, during startup activities, a manual reactor scram was inserted as a result of high conductivity in the condenser. It is believed that the high conductivity was the result of a condenser tube leak. Upon receipt of the conductivity excursion alarm, abnormal operating procedures were consulted and the manual scram was inserted. Unit 2 was at approximately 1 percent of rated thermal power and reactor pressure was approximately 100 psi. At the time of the conductivity excursion, the condensate system was not in service and, as such, reactor water chemistry was not adversely affected. All safety systems operated per design. No emergency core cooling systems (ECCS) actuated.

"Unit 2 will be taken to mode 4 and the necessary repairs will be completed."

All control rods inserted as expected. The licensee believes there is no spread of high conductivity to adjacent systems (e.g. CRD and the CST). Confirmatory samples are in progress. Decay heat is being removed by RCIC in the pressure control mode with the intention of placing shutdown cooling in-service. The electrical system is in a normal shutdown lineup.

The licensee notified the NRC Resident Inspector.

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