Event Desc|En No|Site Name|Licensee Name|Region No|City Name|State Cd|County Name|License No|Agreement State Ind|Docket No|Unit Ind1|Unit Ind2|Unit Ind3|Reactor Type|Nrc Notified By|Ops Officer|Notification Dt|Notification Time|Event Dt|Event Time|Time Zone|Last Updated Dt|Emergency Class|Cfr Cd1|Cfr Descr1|Cfr Cd2|Cfr Descr2|Cfr Cd3|Cfr Descr3|Cfr Cd4|Cfr Descr4|Staff Name1|Org Abbrev1|Staff Name2|Org Abbrev2|Staff Name3|Org Abbrev3|Staff Name4|Org Abbrev4|Staff Name5|Org Abbrev5|Staff Name6|Org Abbrev6|Staff Name7|Org Abbrev7|Staff Name8|Org Abbrev8|Staff Name9|Org Abbrev9|Staff Name10|Org Abbrev10|Scram Code 1|RX CRIT 1|Initial PWR 1|Initial RX Mode1|Current PWR 1|Current RX Mode 1|Scram Code 2|RX CRIT 2|Initial PWR 2|Initial RX Mode 2|Current PWR 2|Current RX Mode 2|Scram Code 3|RX CRIT 3|Initial PWR 3|Initial RX Mode 3|Current PWR 3|Current RX Mode 3|Event Text|
Power Reactor|46521|PILGRIM|ENTERGY NUCLEAR|1|PLYMOUTH|MA|PLYMOUTH||Y|05000293|1|||[1] GE-3|MERT PROBASCO|MARK ABRAMOVITZ|1/5/2011 00:00:00|09:03|1/5/2011 00:00:00|01:20|EST|3/4/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||NEIL PERRY|R1DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||REACTOR CORE ISOLATION COOLING DECLARED INOPERABLE  "On January 5, 2011, at 0120 hours, with the reactor at 100% thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNSP) declared the Reactor Core Isolation Cooling (RCIC) system inoperable due to the RCIC suction isolation valve from the Torus/Suppression Pool (RCIC-26) failing to go fully closed during planned surveillance testing.  The RCIC-26 is a motor-operated valve (MOV) and its normal position is closed.  The RClC-26 valve is redundant to the RCIC-25 valve, and is not the credited containment isolation valve.  The RCIC-26 valve has a safety function to be (manually) opened during certain event mitigation scenarios requiring a transfer of suction sources from the Condensate Storage Tank (CST) to the Torus.  "Based on the valve failing to fully close during MOV stroke time testing per PNPS Procedure 8.5.5.4, the RCIC system was declared inoperable at 0120 hours and the appropriate LCO was entered.  The RCIC-26 was subsequently returned to a full open position, caution tagged and the RCIC system was declared operable.  The LCO was exited at 0200 hours.  An investigation of the event is underway and continuing.  "This event had no impact on the health and/or safety of the public.  "The NRC Resident Inspector is on-site and has been notified.  "This is an 8-hour notification made in accordance with 50.72(b)(3)(v)(D)."  The licensee will notify the State of Massachusetts.   * * * RETRACTION FROM JOSEPH LYNCH TO JOHN KNOKE AT 1946 EST ON 3/4/11 * * *  "Event Notification 46521 was conservatively made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were satisfied pending the evaluation of RCIC System operability.  "On 01/05/11, at 0120 hours the RCIC System was declared inoperable due to uncertainty of RCIC System Operability when the Torus/Suppression Pool Suction Valve (RCIC-26) failed to go fully closed during planned surveillance testing. The valve was restored to the full open position and the valve was declared operable based on capability to meet the required safety function to fully open when RCIC pump suction from the suppression pool is required.  "The apparent cause evaluation concluded that valve failure was the result of high relay contact resistance in the closing control circuit components of the valve breaker. This failure prevented the valve from fully closing but had no affect on capability to open the valve. Surveillance testing verified that capability to open the valve was not affected.  "Corrective action was completed to clean or replace the control circuit relay contacts. Post work testing confirmed capability to open and close the valve. An extent of condition for similar breaker control circuit components was also performed. All relevant technical information is documented in the corrective action system.  "The failure observed did not affect the valve's required safety function and did not impact RCIC System operability. Thus there was no impact on nuclear safety. This event is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(D) .  "Event Number 46521, made on 01/05/2011, is being retracted."  The licensee has notified the NRC Resident Inspector.  Notified R1DO (Anthony Dimitriadis)|
Agreement State|46528|WISCONSIN RADIATION PROTECTION|SAINT NICHOLAS HOSPITAL|3|SHEBOYGEN|WI||117-1302-01|Y||||||CHRIS TIMMERMAN|JOE O'HARA|1/10/2011 00:00:00|14:29|1/10/2011 00:00:00||CST|3/1/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||MICHAEL KUNOWSKI|R3DO|JIM LUEHMAN|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - NUMEROUS MEDICAL EVENTS FROM PROSTATE BRACHYTHERAPY  The following was received from the state via fax;  "In July 2010, the Wisconsin Department of Health Services (DHS) sent out an Information Notice to all licensees who perform prostate brachytherapy and asked them to perform a comprehensive review of all prostate brachytherapy cases to determine whether any medical events had occurred.  On January 10, 2011, the licensee's Radiation Safely Officer reported the identification of five medical events involving permanent implants of I-125 for prostate brachytherapy where the total dose delivered differs from the prescribed dose by 20% or more. The licensee is identifying a medical event of any case where D90<135 Gy or D90>195 Gy for patients who receive seed implants only. [D90 is a recognized value in the regulatory guidelines and means a dose of 90% to the prostate.  Anything outside of the D90 value is considered to be a medical event.] The licensee performed a comprehensive review of all 44 prostate implants performed since August 2003. The licensee's five medical events include one overdose to the prostate and four underdoses to the prostate.  All were patients who received seed implants only. No medical events were identified involving doses to other organs or tissue above 0.50 Sv and 50% more than the expected dose. The licensee has notified the referring physicians and will not be notifying the affected patients per DHS 157.72(1)(e).  "Overdoses (medical event criteria used: D90>195 Gy): 11/13/2008: 199.15 Gy  "Underdoses (medical event criteria used: D90<135 Gy): 2/9/2007: 100.20 Gy;  11/12/2007: 127.34 Gy;  6/16/2008: 130.12 Gy; and  7/13/2010: 116.16 Gy"   * * * UPDATE FROM CHRIS TIMMERMAN TO JOHN KNOKE AT 1212 EST ON 2/1/11 * * *  "This is an update to Event Notification 46528. The licensee recently performed post-implant dosimetry on seven prostate brachytherapy patients whose post-implant dosimetry had never been performed. Evaluation of these seven implants prompted the licensee to report two additional medical events. The medical events involved permanent implants of l-125 for prostate brachytherapy where the total dose delivered to the prostate differs from the prescribed dose by 20% or more. The licensee is in the process of notifying the affected patients and referring physicians.  "Underdoses (medical event criteria used: D90<135 Gy):  8/22/2005: 102.89 Gy; and 5/8/2006: 126.24 Gy;  "DHS will send a special inspection team to determine the root cause(s) of these medical events."  WI Event Report ID No.: WI 110001 Update    Notified FSME(Angela McIntire) and R3DO (Richard Skokowski)   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.  * RETRACTION FROM MEGAN SHOBER TO JOHN SHOEMAKER VIA FACSIMLE AT 1355 EST ON 3/1/11 *  "This is a second update to Event Notification 46528. The licensee is retracting an overdose previously reported for a patient who received a permanent prostate implant on November 13, 2008.  During a reactive inspection conducted on February 18, 2011, DHS inspectors identified that post-implant dosimetry of prostate brachytherapy implants had not been performed appropriately.  Specifically, the licensee's former physics consultant generated post-plans that were not based on the number of I-125 seeds actually implanted in the patients.  For the patient in question, the original post-plan was based on an implant of 98 seeds; however, only 76 seeds were actually implanted. The licensee's current physicist generated a new post-plan using the correct number of I-125 seeds and observed a corresponding reduction in dose delivered.  The new D90 value for this patient does not meet the licensee's medical event criteria."  WI Event Report ID No.: WI 110002,  2nd Update    Notified FSME(McIntosh) and R3DO (Dickson)|
Power Reactor|46548|SAINT LUCIE|FLORIDA POWER & LIGHT CO.|2|FT. PIERCE|FL|ST LUCIE||Y|05000335|1|2||[1] CE,[2] CE|BRAD BISHOP|HOWIE CROUCH|1/18/2011 00:00:00|10:48|1/19/2011 00:00:00|02:00|EST|3/24/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||MARVIN SYKES|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0|Refueling|0|Refueling|N|N|0||0||EMERGENCY RESPONSE DATA ACQUISITION AND DISPLAY SYSTEM (ERDADS) REMOVED FROM SERVICE FOR MAINTENANCE  "On 1/19/11, St Lucie Unit 1 and Unit 2 will lose the computer trains associated with Emergency Response Data Acquisition and Display System (ERDADS). Unit 1 will be removed for corrective maintenance, and Unit 2 will be removed for system modification. It is expected that Unit 1 will be restored by 1/21/11, and Unit 2 will be returned to a functional status prior to core reload and fully operational by March 20, 2011. Further, neither Unit 1 nor Unit 2 ERDADS will be removed from service until Unit 2 has defueled (currently scheduled for 1/19/2011 at 0200). This is an advance notification of a planned loss of emergency assessment capability, which will be reportable under 10CFR50.72(b)(3)(xiii). Other means to monitor critical data exists. Notification will be made when each unit is restored to available status."  The licensee has notified the NRC Resident Inspector.  * * * UPDATE FROM REESE KILIAN TO HOWIE CROUCH @ 0952 EST ON 1/19/11 * * *  "Unit 1 and 2 ERDADS have been removed from service at 0955 [EST]."  The licensee has notified the NRC Resident Inspector.  * * * UPDATE FROM TIMOTHY KUDO TO JOHN SHOEMAKER @ 1527 EST ON 01/20/11 * * *   Unit 1 ERDADS has been returned to available status as of 1520 EST on 01/20/11.  Unit 2 ERDADS remains out of service.  The licensee has notified the NRC Resident Inspector.  * * * UPDATE FROM BISHOP TO HUFFMAN AT 2013 EDT ON 3/24/11 * * *  Unit 1 and Unit 2 ERDADS have both been restored to service as of 1600 EDT on 3/24/11.    The licensee has notified the NRC Resident Inspector.  R2DO (Rich) notified.|
Power Reactor|46562|KEWAUNEE|NUCLEAR MANAGEMENT COMPANY|3|KEWAUNEE|WI|KEWAUNEE||Y|05000305|1|||[1] W-2-LP|MIKE TERRY|JOHN KNOKE|1/21/2011 00:00:00|22:25|1/21/2011 00:00:00|15:39|CST|3/22/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|50.72(b)(3)(v)(B)|POT RHR INOP|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||TAMARA BLOOMER|R3DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||STEAM EXCLUSION DOOR DECLARED INOPERABLE  "On 1/21/2011 at 1539 CST, the NRC Resident Inspector informed the Control Room that the lower Cane bolt was disengaged on Steam Exclusion Door 3, between Emergency Diesel Generator Room B and the Cardox Room. While the Cane bolt was not engaged, the barrier was Non-Functional and, in accordance with TRM 3.0.9, all equipment supported by that steam exclusion barrier was immediately declared inoperable. This included both Emergency Diesel Generators A & 8, safety-related 4160 V Busses 5 & 6, Service Water Trains A & B, and safety-related 480 V Busses 51, 52, 61. & 62. In addition, with Service Water inoperable, the following equipment was also inoperable in accordance with TRM 3.3.1: Component Cooling Trains A & B, Safety Injection Trains A & B, Residual Heat Removal Trains A & B, Containment Spray and Cooling Trains A & B, Auxiliary Feedwater Pumps A & B, and the Turbine Driven Auxiliary Feedwater Pump. With all three AFW pumps inoperable. TS 3 A.b.2 was entered to immediately initiate action to restore one AFW Train to operable status and suspend all LCOs requiring mode changes until one AFW Train is restored to operable status.  "Steam Exclusion Door 3 was properly secured at 1545 CST on 1/21/2011, and LCO 3.0.c and TS 3 A.b.2 were exited at that time. All equipment affected by the steam exclusion barrier is operable.  "This is reportable under 10 CFR 50.72 (b)(3)(v)(B), 'Any event or condition that at the time of discovery could have prevented the fulfillment of a safety function,' and under 10 CFR 50.72(b)(3)(ii)(B) 'any event or condition that results in the nuclear plant being in an unanalyzed condition that significantly degrades plant safety.'"  The licensee notified the NRC Resident Inspector.   * * * RETRACTION FROM CRAIG J. NEUSER TO DONALD NORWOOD AT 1427 EDT ON 3/22/2011 * * *  "Retraction of EN #46562 Non-Functional Steam Exclusion Door.  "On January 21, 2011, EN #46562 provided notification that both trains of ESF equipment (e.g., SI, RHR, ICS, etc ) were inoperable following discovery that the lower cane bolt was disengaged on steam exclusion Door 3, between emergency diesel generator Room B and an adjacent equipment room in the turbine building.  With the lower cane bolt disengaged, the steam exclusion barrier was considered non-functional.  "A subsequent engineering evaluation determined that the Door 3 lower cane bolt was not required for Door 3 to fulfill its function as a steam exclusion barrier.  The previously reported condition would not have resulted in an environment that would have adversely impacted the equipment protected by Door 3.  Therefore, the door remained functional and the supported ESF equipment remained operable.  Consequently, this condition did not meet the reportability criteria in 10CFR50.72.  "As a result, the notification made on January 21, 2011, in EN #46562 is hereby retracted.    "The NRC Senior Resident Inspector has been notified."  Notified R3DO(Cameron).|
Agreement State|46577|MA RADIATION CONTROL PROGRAM|MASSACHUSETTS GENERAL HOSPITAL|1|BOSTON|MA|SUFFOLK|RCN01762|Y||||||ANTHONY CARPENITO|JOHN SHOEMAKER|1/31/2011 00:00:00|13:59|1/13/2011 00:00:00||EST|3/4/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||JAMES DWYER|R1DO|RICHARD TURTIL|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - POTENTIAL DOSE EXCEEDING ANNUAL OCCUPATIONAL DOSE LIMIT FOR ADULTS  The following was received via email:  "On 1/13/11, the licensee reported to the agency [Massachusetts Radiation Control Program] a potential dose exceeding the adult occupational total effective dose equivalent limit of 5 rem.  The situation causing this event occurred during late December 2010, when a number of emergency repairs within a cyclotron were conducted over several days.  The potential for overexposure was suspected on 1/5/11.  A dosimeter exposure readings report of the [Optically Stimulated Luminescence OSL] was received by the licensee from the dosimeter service, after quick read, on 1/13/11.  Affected individual's annual TEDE reported at 5457 mrem.  Exposure (1596 mrem) obtained by official [OSL] dosimeters worn during cyclotron repair operations differed significantly from exposure (620 mrem) obtained by electronic dosimeter worn simultaneously during cyclotron repair operations.  Electronic dosimeters are used by individuals for real-time readings during the repair operations.  The licensee removed affected individual from any potentially high exposure operations.  "Investigation ongoing.  Intermediate and draft reports have been received.  Awaiting final written report by licensee.  "The Agency considers this event to still be OPEN."  Massachusetts Event # 01-9454.  Notified R1DO (Dwyer) and FSME (McIntosh).  * * * UPDATE FROM TONY CARPENITO TO JOHN SHOEMAKER AT 1441 EST ON 03/03/11 VIA EMAIL * * *   "Subsequent on-site agency inspection performed.  Licensee submitted follow-up report [on] 2/28/11.  "Cause Description: Misinterpretation of licensee's pre-existing policy restricting workers when YTD [year-to-date] annual exposures approach in-house limits and over-reliance on real-time electronic dosimeters worn specifically during  potentially high exposure operations.  "Precipitating factor:  Over-reliance on real-time electronic dosimeters worn specifically during potentially high exposure operations.   "Corrective Action:  Licensee to implement policy re-write to minimize subjective misinterpretations, change full-time dosimeter exchange frequency to obtain more current year-to-date exposure totals, replace current job-specific dosimeters with different type of dosimeter better suited to monitor type of work involved, apply administrative correction factors to readings of job-specific dosimeters to obtain more conservative real-time results.   "The individual [involved in this event was] removed from potentially high exposure operations during investigation and re-instated several weeks later on 3/3/11.  "Although the Agency considers this specific situation to be closed, it will be revisited during future inspections."  The report did not state whether an over exposure actually occurred.  Notified R1DO (Dimitriadis) and FSME (McIntosh).  * * * UPDATE FROM TONY CARPENITO TO CHARLES TEAL AT 0832 EST ON 03/03/11 VIA TELEPHONE * * *   The Massachusetts Radiation Control Program determined the individual received a dose of 5457 mrem.  Notified R1DO (Dimitriadis) and FSME (McIntosh).|
Non-Agreement State|46601|CRITTENTON HOSPITAL MEDICAL CENTER|CRITTENTON HOSPITAL MEDICAL CENTER|3|ROCHESTER|MI||21-13562-01|N||||||V. ARTERBERY|MARK ABRAMOVITZ|2/8/2011 00:00:00|12:05|2/7/2011 00:00:00|16:00|EST|3/7/2011 00:00:00|NON EMERGENCY|35.3045(a)(1)|DOSE <> PRESCRIBED DOSAGE|||||||ERIC DUNCAN|R3DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MEDICAL EVENTS - ACTUAL DOSE LESS THAN PRESCRIBED DOSE  Nine breast cancer patients were treated with a multi-channel high dose rate afterloader with the incorrect dwell positions input into the afterloader controller.  The intended prescription was for 5 mm between dwell positions.  The actual input was 2.5 mm between dwell positions.  The prescribed dose was for 3400 cGy and the actual dose delivered was approximately 25% less.  No adverse clinical effects are expected.    The doctors and patients will be notified.  * * * UPDATE FROM DR. VIVIAN ARTERBERY TO JOHN SHOEMAKER AT 1507 EST ON 03/07/11 * * *   There were a total of eleven breast cancer patients involved.  Seven of the eleven patients had small areas of unintended target that received unplanned dose.    The doctors and patients have been notified.  Notified R3DO (Riemer) and FSME (McIntosh)    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.|
Power Reactor|46607|MILLSTONE|DOMINION GENERATION|1|WATERFORD|CT|NEW LONDON||N||||3|[1] GE-3,[2] CE,[3] W-4-LP|TODD FISHER|PETE SNYDER|2/11/2011 00:00:00|12:35|2/11/2011 00:00:00|08:45|EST|3/21/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(C)|POT UNCNTRL RAD REL|||||||RONALD BELLAMY|R1DO|||||||||||||||||||N|N|0||0||N|N|0||0||N|Y|100|Power Operation|100|Power Operation|TEARS DISCOVERED IN FLEXIBLE PIPE SEALS  "On February 11, 2011, during a plant walk down, it was discovered that two flexible boots sealing two piping penetrations in the secondary containment boundary had a total of three small tears rendering them inoperable.  "Technical Specification 3.6.6.2, 'Secondary Containment,' is applicable in Modes 1, 2, 3, and 4 and was entered.  Since Secondary Containment was rendered inoperable, Dominion is reporting that this condition could have prevented the fulfillment of the safety function to control the release of radioactive material.  Further engineering review will be conducted to more fully evaluate the impact on radiological controls."  No actual release occurred as a result of this condition.  The licensee notified the NRC Resident Inspector, the State of Connecticut and the Town of Waterford.   * * * RETRACTION FROM ROBERT ACQUARO TO DONALD NORWOOD AT 1503 EDT ON 3/21/2011 * * *  "On February 11, 2011, small tears were discovered on flexible boots sealing two piping penetrations in the secondary containment boundary on Millstone Power Station Unit 3.  Operators made a report in accordance with 10CFR50.72(b)(3)(v)(C).  "Subsequently, an engineering evaluation has been completed that concludes that the discovered tears would not have prevented the fulfillment of the safety function of the secondary containment boundary.  Therefore, the condition reported in event report 46607 is being retracted.  "The NRC Senior Resident Inspector has been notified."  Notified R1DO (Rogge).|
Agreement State|46643|TEXAS DEPARTMENT OF HEALTH|TRICAN WELL SERVICE|4|SPRINGTOWN|TX||GLA G02259|Y||||||KAREN BLANCHARD|JOE O'HARA|2/25/2011 00:00:00|19:03|2/25/2011 00:00:00|16:15|CST|2/25/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||VIVIAN CAMPBELL|R4DO|ROBERT LEWIS|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MISSING SHUTTER HANDLE ON BERTHOLD MODEL LB 8010 PORTABLE DENSITY GAUGE   On 2/22/11, Trican Well Service was preparing to use the portable gauge at a well jobsite in De Soto Parish, Louisiana when they discovered the gauge was missing its shutter handle.  The crew stopped work, took surveys to ensure the shutter was fully closed, and placed the gauge out of service until a repair can be performed. The device is a Berthold Model LB 8010 serial number 10074 portable density gauge used in well servicing operations.  The device contains 20 milliCuries of Cs-137, serial number 014808.  The crew does not know when the handle became detached.  They searched the jobsite for the handle but were unsuccessful in recovering it.  The device is in the custody of the licensee.  Berthold has been contacted and plans to repair the gauge on March 1, 2011.   Texas Incident No. I-8825|
Agreement State|46648|IOWA DEPARTMENT OF PUBLIC HEALTH|UNIVERSITY OF IOWA|3|IOWA CITY|IA||0037152AAB|Y||||||RANDAL DAHLIN|CHARLES TEAL|2/28/2011 00:00:00|12:14|2/22/2011 00:00:00||CST|2/28/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||BILLY DICKSON|R3DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - STATIC ELIMINATOR SOURCE SEAL INTEGRITY LOST  "The following report was received by the Iowa Department of Public Health (IDPH) on February 23, 2011. On February 22, 2011, the University of Iowa Environmental Health and Safety (EHS) staff discovered evidence that a Nickel-63 foil in a custom made static eliminator had lost its seal integrity. The source in question consists of two 8.951 mCi Ni-63 foils that are housed in a custom built static eliminator (2 inch diameter steel pipe with the foils glued to the walls of the pipe). The static eliminator is attached to a chamber apparatus located in a fume hood within a principal investigator's lab. The results of the leak test indicated approximately 40,203 dpm (0.0181 uCi's) of activity on a wipe taken of several areas in the apparatus housing the Ni-63 foils. EHS personnel bagged the static eliminator and returned it to EHS for disposal. The two Ni-63 sources were purchased from DuPont/Merck in March of 1995. The Principle Investigator (PI) had been conducting research using these sources since that time. The University RSO reports that the PI does not have any more of these custom devices and will be pursuing other options for research. A previous leaking Ni-63 foil was reported as NMED Item Number 100592."  Iowa Incident No. 110002|
Power Reactor|46649|INDIAN POINT|ENTERGY NUCLEAR|1|BUCHANAN|NY|WESTCHESTER||Y|05000247|2|||[2] W-4-LP,[3] W-4-LP|PHIL SANTINI|JOHN KNOKE|3/1/2011 00:00:00|12:54|3/1/2011 00:00:00|11:00|EST|3/1/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||ANTHONY DIMITRIADIS|R1DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||EMERGENCY DIESEL GENERATORS START ON A LOSS OF POWER FROM 138 KV CIRCUIT  "At 1100 EST, Indian Point Unit 2 experienced a loss of offsite power from the 138 Kv circuit.  All three Emergency Diesel Generators automatically started as required.  All other plant systems functioned as required.  Restoration of offsite power from the 13.8 Kv offsite circuit is in progress.  Investigation into the loss of the 138 Kv circuit is ongoing.  Indian Point, Unit 2 continues in Mode 1 at 100 % power."  Indian Point, Unit 2 is in a 72 hour LCO due to a loss of 1 of 2 offsite circuits.  Unit 3 was not affected.  The licensee has notified the NRC Resident Inspector and will be notifying the Public Service Commission of the State of New York.|
Fuel Cycle Facility|46650|GLOBAL NUCLEAR FUEL - AMERICAS|GLOBAL NUCLEAR FUEL - AMERICAS|2|WILMINGTON|NC|NEW HANOVER|SNM-1097|Y|07001113||||URANIUM FUEL FABRICATION|PHILLIP OLLIS|JOHN SHOEMAKER|3/2/2011 00:00:00|11:10|3/1/2011 00:00:00|15:00|EST|3/2/2011 00:00:00|NON EMERGENCY||OTHER UNSPEC REQMNT|||||||MALCOLM WIDMANN|R2DO|THOMAS HILTZ|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||FAILURE TO MAINTAIN MASS CONTROL OF UO2 POWDER  "At approximately 3:00 pm [EST] on Tuesday, March 1st, it was reported that 15.3kg of UO2 powder was removed from the UO2 Sinter Test Grinding Station [High Efficiency Particulate Air] HEPA filter housing transition. This clean-out was performed as a response to routine radiological surveys that indicated the presence of uranium.  It was determined that this material was present during prior filter replacement in early February in which 30.9 kg of UO2 powder was removed.  The total amount of UO2 powder present in the housing was therefore approximately 46 kg, which is greater than a safe mass.  "The Sinter Test Grinding Station and associated equipment was already shut down because of the HEPA filter housing clean-out and remained down pending investigation and implementation of corrective actions.    "Failure to maintain mass control resulted in a loss of double contingency for the filter housing.  The double contingency controls required include (1) mass control and (2) moderation control.  Moderation control, the 2nd leg of double contingency remained in place, was effective, and was not challenged.  As a result, no unsafe condition existed.    "The UO2 in the HEPA housing was transferred into [a] favorable geometry [of] 3-gallon cans per procedure.  An investigation is ongoing.  "At no time did an unsafe condition exist as the moderation control was in place, was effective, and was not challenged.  Immediate corrective actions [are] complete [transfer of material into 3-gallons cans].  Investigation of [the] event and implementation of long term corrective actions [are] pending.  The licensee will be notifying the NRC Region II, State, and Local Authorities.|
Non-Agreement State|46651|DEFENSE LOGISTICS AGENCY|DEFENSE LOGISTICS AGENCY|4|RED RIVER|TX|RED RIVER|37-30062-01|Y||||||DAVID COLLINS|JOE O'HARA|3/2/2011 00:00:00|15:20|3/1/2011 00:00:00||CST|3/7/2011 00:00:00|NON EMERGENCY|20.2201(a)(1)(ii)|LOST/STOLEN LNM>10X|||||||ANTHONY DIMITRIADIS|R1DO|BOB HAGAR|R4DO|ROBERT LEWIS|FSME|ILTAB EMAIL||MEXICO||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||LOST CHEMICAL DETECTION EQUIPMENT CONTAINING NI-63 SOURCES  "As part of a mass turn in of Chemical Detection Equipment (CDE) containing radioactive sources, an Air Force (AF) unit located in Fort Leonard Wood, MO. reported that it had shipped 52 CDE systems, each containing Ni-63 with 20 milli Curie sources, to the Defense Logistics Agency (DLA) Distribution Center at Red River Texas. The CDE systems were shipped in 7 separate packages and confirmation of 7 separate packages was provided which showed the receipt by DLA on 21 May 2010.  An inventory, conducted on 2 February 2011, of the 7 packages at the DLA Distribution Center could only account for 40 of the 52 CDE systems.  The AF Radiation Safety Office contacted the DLA Radiation Safety Office indicating that there was a discrepancy in the number of items receipted compared to the number of items identified as shipped by the unit.  The AF Radiation Safety Office provided the list of serial numbers for the 52 items reportedly shipped which were compared to the current inventory on hand.  It was confirmed that there were 12 serial numbers identified on the shipping list that were not in the current inventory.   An initial search of the designated Radioactive Material Storage area did not locate the items.  "It has been identified that a major issue with the turn-in process was the lack of prior notification of transfer of the CDE from the AF units to the DLA Distribution Site at Red River. A global instruction was provided by the AF PM to ship the material, without the PM providing the turn-in documents. Had the AF PM directed all of the shipment, a document would have been generated and the number of items to be shipped would have been provided to the depot. In the current process, the depot could only act reactively based on available material and paperwork.  "The transfer of CDE systems to the DLA Distribution Centers has been suspended until a more accountable process can be determined.  Additionally the AF will, in the future, direct all shipments directly to Wright Patterson AFB for demilitarization and disposal.  A continued effort to search the hazardous material warehouses for these items will continue."   * * * RETRACTION FROM DAVID COLLINS TO JOE O'HARA AT 1642 ON 03/07/11 * * *   "DLA Distribution is retracting the initial report of potential loss (Event # 46651) as the 12 Chemical Agent Detectors have been accounted for at Pine Bluff Arsenal, another Department of Defense facility and the ultimate destruction site.   The materiel was shipped from DLA Distribution Red River to Pine Bluff Arsenal on October 8, 2010, under a different NSN [National Stock Number] than the one originally sent to DLA.  The Chemical Agent Detection (CDE) Equipment can be in multiple configurations, each with their own NSN.  In this case they shipped from the AF unit under one NSN (entire detection system) and we subsequently shipped under a corrected NSN (detector component only)."  Notified R1DO(Hansell), R4DO(Farnholtz), FSME(Reis), Mexico, and ILTAB via e-mail.     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. For additional information go to  http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf  This source is not amongst those sources or devices identified by the IAEA Code of Conduct for the Safety & Security of Radioactive Sources to be of concern from a radiological standpoint.  Therefore is it being categorized as a less than Category 3 source|
Power Reactor|46653|DAVIS BESSE|FIRSTENERGY NUCLEAR OPERATING COMPANY|3|OAK HARBOR|OH|OTTAWA||Y|05000346|1|||[1] B&W-R-LP|TOM COBBLEDICK|JOHN KNOKE|3/3/2011 00:00:00|20:11|3/3/2011 00:00:00|13:53|EST|3/3/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(A)|POT UNABLE TO SAFE SD|50.72(b)(3)(v)(B)|POT RHR INOP|||||KENNETH RIEMER|R3DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||TEMPORARY LOSS OF EMERGENCY FEEDWATER TRAINS  "While testing fire detection systems, a radio was keyed in the vicinity of the Auxiliary Shutdown Panel. Control Room alarms that occurred at the same time led to a review of plant data. This review revealed two momentary events (approximately 8 and 19 seconds) over an approximate two minute period that caused momentary reductions in the control signals to the Auxiliary Feedwater Pump and Motor-Driven Feedwater Pump discharge control valves. These momentary signal reductions resulted in all trains of Emergency Feedwater being inoperable for approximately two minutes, pending further evaluation.  "With all trains of Emergency Feedwater inoperable, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v) as a momentary loss of safety function for equipment needed to (A) shut down the reactor and maintain it in a safe shutdown condition and to (B) remove residual heat.   "Fire detection testing has been completed, and a sign placed on the Auxiliary Shutdown Panel Room door stating that no radio usage is permitted inside the room."  All trains of Emergency Feedwater are now operable.  The licensee has notified the NRC Resident Inspector.|
Power Reactor|46654|SUSQUEHANNA|PPL SUSQUEHANNA LLC|1|ALLENTOWN|PA|LUZERNE||N|05000387|1|||[1] GE-4,[2] GE-4|RON FRY|BILL HUFFMAN|3/3/2011 00:00:00|23:27|3/3/2011 00:00:00|22:59|EST|3/3/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(i)|PLANT S/D REQD BY TS|||||||ANTHONY DIMITRIADIS|R1DO|||||||||||||||||||N|Y|99|Power Operation|86|Power Operation|N|N|0||0||N|N|0||0||TECHNCIAL SPECIFICATION REQUIRED SHUTDOWN DUE TO INOPERABLE HPCI  At 2259 EST, Susquehanna Unit 1 commenced a TS Required shutdown for High Pressure Coolant Injection (HPCI) T.S. 3.5.1.  Due to a suspected steam leak, the HPCI Inboard Steam Supply Valve HV155F002 was closed to attempt to identify and isolate an unknown drywell leakage condition.  After closing the HV155F002, a detectable change in drywell leak rate occurred, therefore, HV155F002 was left closed. Closing HV155F002 makes HPCI INOP and UNAVAILABLE.  TS 3.5.1 was entered for this condition on 2/25/2011 at 2136 EST. LCO completion time for T.S. 3.5.1 entry is 3/11/2011 at 2136 EST.  The licensee has notified the NRC Resident Inspector and State authorities.  The licensee also anticipates issuing a press release.|
Part 21|46655|FISHER CONTROLS INTERNATIONAL|FISHER CONTROLS INTERNATIONAL|3|MARSHALLTOWN|IA|||Y||||||DENNIS SWANSON|JOHN KNOKE|3/4/2011 00:00:00|12:53|2/25/2011 00:00:00||CST|3/4/2011 00:00:00|NON EMERGENCY|21.21|UNSPECIFIED PARAGRAPH|||||||KENNETH RIEMER|R3DO|MALCOLM WIDMANN|R2DO|PART 21 GROUP||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||PART 21 - ERROR IN VALVE BODY DRAWING   The purpose of this Fisher Information Notice (FIN) is to alert Duke Energy that as of February 25, 2011, Fisher Controls International LLC became aware of the possibility of a situation which may affect the performance of the applicable equipment provided to McGuire Nuclear Station. Specifically, an error was discovered on valve body drawing V112298, when, during a revision process on Revision B of the drawing, a dimension was omitted that set the depth of the valve shaft bearing bore. This error resulted in a greater possible variation of bearing position in the valve shaft bore. If the error was large, the valve could not be assembled which was not the case for these valves. The valves assembled without incident and passed the operational testing, including a seat leakage test, with no anomalies. In the case that the bearing position error was slight, it is possible that the seal and disc could experience more wear than normal and increased leakage would result. Because these valves are equipped with manual operators, Fisher expects that these valves will not be cycled enough to experience any of the potential problems described above.  This equipment included NPS 4, Class 150, Fisher Type A11 Butterfly Valve Assemblies equipped with Fisher Leverlock Manual Actuators. The NPS 4, A11 is a butterfly valve that uses internal bearings (located on either side of the disc) to provide a radial wear surface for shaft rotation and also serve as a centering system for the disc in the waterway. Centering of the disc is accomplished with a wear surface on the end of the bearings adjacent to the side of the disc. Lateral positioning of the disc is accomplished by controlling the length of the bearings and the depth of the bored holes in the body that accept the bearings.  Fisher has revised the drawings to ensure that this issue is corrected.|
Power Reactor|46656|BROWNS FERRY|TENNESSEE VALLEY AUTHORITY|2|DECATUR|AL|LIMESTONE||Y||||3|[1] GE-4,[2] GE-4,[3] GE-4|TIMOTHY SCOTT|BILL HUFFMAN|3/4/2011 00:00:00|16:24|1/3/2011 00:00:00|15:15|CST|3/4/2011 00:00:00|NON EMERGENCY|50.73(a)(1)|INVALID SPECIF SYSTEM ACTUATION|||||||MALCOLM WIDMANN|R2DO|||||||||||||||||||N|N|0||0||N|N|0||0||N|N|0|Refueling|0|Refueling|60 DAY TELEPHONE NOTIFICATON CONCERNING INVALID CONTAINMENT ISOLATION SYSTEM ACTUATIONS  "This 60-day telephone notification is being made under the reporting requirements specified by 10 CFR 50.73(a)(2)(iv) and 10 CFR 50. 73(a)(1) to report four closely-spaced invalid actuations of general containment isolation signals affecting more than one system.  "Four events of unplanned actuations of general containment isolation signals affecting containment Isolation valves in more than one system occurred during planned transfers of power between the normal and alternate power supply for the 3A 480V Shutdown Board.  The first event occurred on January 3, 2011, at 1515 hours Central Standard Time (CST), with Unit 3 in a forced outage and at 0 percent power (0 MWT).  The electrical power to the 3A Reactor Protection System (RPS) was interrupted during the planned transfer of the 3A 480V Shutdown Board from its normal supply to its alternate power supply.  During the transfer, the alternate feeder breaker did not close.  An attempt was made to return to the normal power supply; however, the normal feeder breaker did not initially close but did close on re-attempt. This resulted in the interruption of power to the 3A 480V Shutdown Board, which caused the 3A RPS to de-energize, resulting in a half scram and the actuation of Primary Containment Isolation System (PCIS) logic Groups 2, 3, 6, and 8, and the initiation of Trains A, B, and C Standby Gas Treatment and Train A Control Room Emergency Ventilation.  "Plant conditions, which require PCIS actuations and the associated system initiations (e.g., low reactor water level, high drywell pressure, abnormal area radiation level, or high area temperature), did not exist; therefore, the actuation was invalid.  The affected equipment responded as designed.  "On January 3, 2011, at approximately 1620 hours CST, Unit 3 Operations personnel restored 3A RPS power and re-aligned affected equipment, as appropriate.  This event was entered in the Corrective Action Program as Problem Evaluation Report (PER) 305070.  "Subsequent related failure to transfer events of the 3A 480V Shutdown Board occurred on January 4, 2011, at 2321 hours, on January 5, 2011, at 0448 hours, and on January 5, 2011, at 0841 hours.  In each event the plant conditions which require PCIS actuations and the associated system initiation did not exist; therefore, the actuations were invalid.  In each of the three subsequent events the affected equipment started and functioned successfully with one exception. During the last event, Main Control Room indication of a Secondary Containment outboard isolation damper closure was indeterminate (i.e., double-lit).  The damper was declared inoperable, and the associated Technical Specification LCO 3.6.4.2 Action was taken.  [A} work order [was] issued to investigate the problem repaired a damper limit switch.  This problem was entered in the Corrective Action Program as PER 305062.  "PERs were generated for each of these events (PER 305865, PER 305421, and PER 305893).  All of these events were consolidated into PER 305070.  "The Browns Ferry NRC Senior Resident Inspector has been notified."|
Power Reactor|46657|MONTICELLO|NUCLEAR MANAGEMENT COMPANY|3|MONTICELLO|MN|WRIGHT||N|05000263|1|||[1] GE-3|TODD LYNCH|BILL HUFFMAN|3/5/2011 00:00:00|13:04|3/5/2011 00:00:00|10:45|CST|3/5/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||KENNETH RIEMER|R3DO|||||||||||||||||||N|N|0|Hot Shutdown|0|Hot Shutdown|N|N|0||0||N|N|0||0||OFFSITE NOTIFICATION RELATED TO FISH KILL  "At 1045 the Monticello Nuclear Generating Plant (MNGP) control room was notified by an Xcel Energy environmental specialist that a fish kill count was conducted on the morning of 3/5/11 following reactor shutdown.  In accordance with the MNGP water appropriations permit for fish kill in the Mississippi river, the environmental specialist will be notifying the State of Minnesota Department of Natural Resources and Minnesota Pollution Control Agency.  "Notifications made to above government agencies meet the reporting criteria established in 10CFR50.72(b)(2)(xi) as an event or situation related to the protection of the environment for which a notification to government agencies has been or will be made.  "Total fish kill was determined to be approximately 100 fish downstream of the plant's discharge canal.  Fish kill was the result of cooldown of water being discharged to the river from the plant's discharge canal.  All temperature limits specified in the plant's water appropriation permit were met throughout the shutdown.  There was no release of any chemical or radioactive materials to the environment."  The licensee has notified the NRC Resident Inspector and will be notifying appropriate state and local authorities.|
Power Reactor|46658|NINE MILE POINT|CONSTELLATION NUCLEAR|1|SYRACUSE|NY|OSWEGO||Y|05000220|1|2||[1] GE-2,[2] GE-5|PATRICK RYAN|CHARLES TEAL|3/6/2011 00:00:00|01:37|3/5/2011 00:00:00|23:15|EST|3/6/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||ANTHONY DIMITRIADIS|R1DO|||||||||||||||||||N|Y|94|Power Operation|94|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||SPURIOUS SIREN ACTIVATION  "One of the 37 Prompt Notification System sirens surrounding the James A. Fitzpatrick (JAF)/Nine Mile Point (NMP) sites spuriously activated at 2315 EST.  "The Oswego County 911 Center notified the Nine Mile Point Emergency Preparedness Department of the inadvertent siren activation.  "Repair technicians have de-activated and silenced the faulty siren as of 0107 EST.  "The cause of the inadvertent siren activation is not understood at this time. The issue has been entered into the site's Corrective Action Program."  The licensee has notified the NRC Resident Inspector, the Oswego County 911 Center, and the Public Service Commission.|
Power Reactor|46659|FITZPATRICK|ENTERGY NUCLEAR|1|LYCOMING|NY|OSWEGO||Y|05000333|1|||[1] GE-4|TOM RESTUCCIO|CHARLES TEAL|3/6/2011 00:00:00|01:51|3/5/2011 00:00:00|23:15|EST|3/6/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||ANTHONY DIMITRIADIS|R1DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||SPURIOUS SIREN ACTIVIATION  "The purpose of this report is to provide a telephone notification under 10 CFR 50.72(b)(2)(xi) to notify the NRC of the inadvertent actuation of one Oswego County emergency notification siren at approximately 2315 EST on 3/5/11.  Initial notification to the JFA [James A. Fitzpatrick] Control Room of the siren activation was via on-site security personnel and verified with Oswego County 911 Center.  The faulted siren was alarming intermittently and repair personnel were dispatched to correct the problem.  At 0107 EST, power to the affected siren was de-energized by off-site repair personnel.  "Sirens affected provide coverage to Oswego County.  The sirens are utility owned and shared with the Nine Mile Point site.  "In the event the sirens are needed the county has it's Hyper-Reach (911 call back system) on standby.  "The NRC Senior Resident Inspector has been notified."|
Power Reactor|46660|TURKEY POINT|FLORIDA POWER & LIGHT CO.|2|MIAMI|FL|DADE||Y|05000250|3|||[3] W-3-LP,[4] W-3-LP|PAUL REIMERS|JOHN KNOKE|3/6/2011 00:00:00|19:38|3/6/2011 00:00:00|16:44|EST|3/6/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(iv)(B)|RPS ACTUATION - CRITICAL|||||||MALCOLM WIDMANN|R2DO|||||||||||||||||||M/R|Y|23|Power Operation|0|Hot Standby|N|N|0||0||N|N|0||0||MANUAL REACTOR TRIP DUE TO SECONDARY SODIUM CONCENTRATIONS EXCEEDING CHEMISTRY LIMITS  "This is a 4-hr Non-Emergency notification to the NRCOC [Nuclear Regulatory Commission Operations Center] for an event that results in actuation of the Reactor Protection System (RPS) when the reactor is critical in accordance with 10CFR50.72(b)(2)(iv)(B).  "On 3/6/11 at approx 16:20 [EST], Steam Generator sodium concentrations started to rise and exceeded 3-ONOP-071.1 (Secondary Chemistry Deviation from limits) Action Level 3 criteria (250 ppb Sodium). The plant power was reduced to 25% per 3-ONOP-100, Fast Load Reduction, and a manual plant trip [was] initiated per procedure at 16:44 [EST]. Unit [3] is stabilized in Mode 3, and [the licensee] is performing secondary clean-up."   All rods fully inserted. All safety systems functioned as required. The reactor trip was uncomplicated. Unit 4 was unaffected by this event.  The licensee has notified the NRC Resident Inspector.|
Power Reactor|46661|FARLEY|SOUTHERN NUCLEAR OPERATING COMPANY|2|ASHFORD|AL|HOUSTON||Y|05000348|1|||[1] W-3-LP,[2] W-3-LP|ALTON DEWEESE|JOHN SHOEMAKER|3/7/2011 00:00:00|10:31|3/7/2011 00:00:00|01:40|CST|3/9/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||MALCOLM WIDMANN|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||INADVERTANT AUTO-START SIGNAL TO THE 1B DIESEL GENERATOR DURING TESTING  "Farley Unit One was conducting FNP-1-STP-80.8, [test procedure for] 1B DG [Diesel Generator] 1000 KW load rejection.  After successfully completing the load rejection portion of the procedure, the control room staff was restoring the 1B diesel to a normal auto start alignment.  With the 1B diesel running, the plant operator was required to reset the 1B DG loading sequencer.  He incorrectly pressed the Emergency Start reset push-button instead of the Sequencer reset push-button. As a result, the Emergency Diesel generator stop light illuminated for a brief few seconds and then extinguished. Subsequently due to the test configuration, the 1B diesel received an auto-start signal and returned to the running condition prior to the Emergency Start reset. Although further investigation is continuing, this report is being made due to an apparent valid actuation of ESF equipment."  This event had no impact on other equipment or the plant electrical alignment.  The Sequencer reset push-button and the Emergency Start reset push-button are not in close proximity to each other.  The plant operator was assessed for fatigue and it was determined that fatigue was not a factor.  The plant operator was removed from duties pending remedial training and assessment.  The licensee has notified the NRC Resident Inspector.  * * * RETRACTION FROM STEVE GATES TO JOE O'HARA AT 1342 ON 03/09/11 * * *   "The 8-hour non-emergency report (EN #46661) per 10CFR50.72(b)(3)(iv)(A) was conservatively reported based on the potential for a valid actuation of an Emergency Diesel Generator (EDG) during a 1000 KW load rejection surveillance test.  "During the restoration phase of the load rejection test to align the 1B EDG to a normal shutdown configuration, a plant operator incorrectly pressed the Emergency Start Reset (ESR) push-button instead of the Sequencer Reset push-button. As a result, the 1B EDG stop light illuminated momentarily and then extinguished. The 1B EDG received a momentary shutdown signal, but remained in a running condition.  "Upon completion of the 1B EDG circuit analysis, It was determined that the 1B EDG did not receive a valid actuation of the EDG safety function. Depressing the ESR push-button caused the emergency start relays to deenergize and remain de-energized. The emergency start relays energize on receipt of valid signals in response to actual plant conditions or parameters satisfying the requirements for the initiation of the safety function of the EDG. Therefore per section 3.2.6 of NUREG-1022, the 1B EDG did not receive a valid actuation signal."  The NRC Resident Inspector has been notified.  Notified R2DO(Musser)|
Power Reactor|46662|MCGUIRE|DUKE POWER|2|CORNELIUS|NC|MECKLENBURG||Y|05000369|1|2||[1] W-4-LP,[2] W-4-LP|JIM DAIN|JOHN SHOEMAKER|3/8/2011 00:00:00|07:28|3/8/2011 00:00:00|03:57|EST|3/8/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||MALCOLM WIDMANN|R2DO|JOHN THORP|NRR|JEFFERY GRANT|IRD|||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0|Refueling|0|Refueling|N|N|0||0||ONSITE FATALITY  "At 0357 [EST on 3/8/11] the licensee received a call for medical assistance.  The caller reported an individual [contract employee] had passed out but was still breathing.  Our [McGuire] onsite medical response team reported the person was not breathing and in cardiac arrest.  The person was transported offsite via ambulance [Gileade Volunteer Fire and Rescue] to the local hospital.  At 0602 EST, the licensee received word that the individual had passed away at the hospital."  The licensee will notify the NRC Resident Inspector|
Fuel Cycle Facility|46663|GLOBAL NUCLEAR FUEL - AMERICAS|GLOBAL NUCLEAR FUEL - AMERICAS|2|WILMINGTON|NC|NEW HANOVER|SNM-1097|Y|07001113||||URANIUM FUEL FABRICATION|SCOTT MURRAY|CHARLES TEAL|3/8/2011 00:00:00|10:18|3/8/2011 00:00:00||EST|3/8/2011 00:00:00|NON EMERGENCY||OTHER UNSPEC REQMNT|||||||MALCOLM WIDMANN|R2DO|BRIAN SMITH|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||GNF PROCEDURE 40-32 LOSS OF A SINGLE MATERIAL CONTROL TRANSACTION  "At approximately 7:00 a.m. on Monday, March 7th, it was reported that a can of powder was present on a conveyor in the UO2 press feed area without the required material control transaction.  At approximately 10:10 a.m. it was discovered that the can contained three vacuum bags of powder.  The transaction is one criticality control for the conveyor to ensure only authorized dry materials are stored.  "The second controlled parameter (mass of uranium in each can) was maintained at all times.  As a result, no unsafe condition existed.  The total amount of UO2 powder in the improperly stored can was approximately 13.6 kg.  The material control transactions have been properly performed and the can has been transferred to an approved storage location.  "As a result, SNM movements have been ceased pending investigation and implementation of additional corrective actions."  The licensee has notified the NRC Resident (Thomas), New Hanover County Emergency Management, and North Carolina of Environment and Natural Resources.|
Power Reactor|46664|MCGUIRE|DUKE POWER|2|CORNELIUS|NC|MECKLENBURG||Y|05000369|1|2||[1] W-4-LP,[2] W-4-LP|MARK MCNEELY|JOHN SHOEMAKER|3/9/2011 00:00:00|11:20|3/9/2011 00:00:00|10:55|EST|3/14/2011 00:00:00|NON EMERGENCY|26.719|FITNESS FOR DUTY|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0|Refueling|0|Refueling|N|N|0||0||FITNESS FOR DUTY  A non-licensed employee failed to meet fitness-for-duty criteria.  The employee's access to the plant has been terminated.  Contact the NRC Headquarters Operations Officer for additional details.  * * * UPDATE FROM MARK S. MCNEELY TO JOHN SHOEMAKER AT 1025 EST ON 03/14/11 * * *   An update to this report has been provided by the licensee.  Contact the NRC Headquarters Operations Officer for additional details.  Notified R2DO Nease)|
Hospital|46665|UNIVERSITY OF MICHIGAN HOSPITAL|UNIVERSITY OF MICHIGAN HOSPITAL|3|Ann Arbor|MI||21-00215-04|N||||||MARK DRISCOLL|JOHN SHOEMAKER|3/10/2011 00:00:00|11:15|3/9/2011 00:00:00|10:00|EST|3/10/2011 00:00:00|NON EMERGENCY|35.3045(a)(1)|DOSE <> PRESCRIBED DOSAGE|||||||BILLY DICKSON|R3DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MEDICAL EVENT INVOLVING ADMINISTRATION OF GREATER THAN ORDERED DOSE  "An Authorized User physician from the University of Michigan Department of Radiation Oncology planned two liver infusion treatments for a patient with unresectable hepatocellular carcinoma using Y-90 Theraspheres.  As part of treatment preparation, an MRI was performed on the patient to determine the patient's liver segment volumes.  Liver segment volumes are used in calculating the Y-90 activity needed to deliver the prescribed radiation dose to particular segments of the liver.  The first treatment was to the right lobe and medial segment of the patient's liver and was performed on 12/15/2010.  It proceeded without incident and in accordance with the written directive.    "The Authorized User physician scheduled a second separate treatment to the patient's left lobe to be performed on 3/9/2011.  The Authorized User ordered a 74.4 Gy dose to the left lobe of the liver.  The medical physicist calculated a corresponding dosage of 60.5 mCi of Y-90 to be infused into the left lobe of the liver.  "However, in arriving at the Y-90 activity needed, a medical physicist used the liver segment volumes for the right lobe and medial segment combined instead of that for the left lobe.  The volume of the right lobe and medial segment is much larger than that for the left lobe.  As a result, the Y-90 dosage of 60.5 mCi exceeded what was actually needed to deliver the prescribed dose of 74.4 Gy.  The Y-90 was infused into to the left lobe on 3/9/2011 at approximately 10 [EST].  Based on the Authorized User's reassessment of the left lobe volume, the dose to the left lobe is calculated, post-administration, to be 159.4 Gy.  "The patient was notified of the event on 3/9/2011.  The referring physician was also notified on the morning of 3/10/2011.  The Authorized User physician has concluded that the elevated radiation dose to the patient's liver will not result in permanent medical damage or loss of function.  Upon initial investigation this event appears to possibly be due to a problem in transcription however causes and corrective actions are still being evaluated."  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.|
Power Reactor|46666|KEWAUNEE|NUCLEAR MANAGEMENT COMPANY|3|KEWAUNEE|WI|KEWAUNEE||Y|05000305|1|||[1] W-2-LP|SCOTT CIESLEWICZ|MARK ABRAMOVITZ|3/10/2011 00:00:00|18:29|3/10/2011 00:00:00|15:49|CST|3/10/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||BILLY DICKSON|R3DO|||||||||||||||||||N|N|0|Refueling Shutdown|0|Refueling Shutdown|N|N|0||0||N|N|0||0||INADVERTANT OPENING A SUBSTATION BREAKER CAUSING LOSS OF STATION POWER  "At 1549 on 03/10/2011 with the plant shut down and the reactor defueled, power was lost to Safeguards 4160 Volt Bus 6.  Diesel Generator B started and re-energized Bus 6.  "At the time of the event, Bus 6 was energized from the Main Auxiliary Transformer (MAT) on backfeed.  The event was caused by opening of substation breaker TA2066 as the result of an error by technicians working In the substation.  "All equipment operated as expected for the voltage restoration to Safeguards Bus 6.  Safeguards Bus 5 remained energized from offsite power through the Tertiary Auxiliary Transformer during the event.  "Spent Fuel Pool Cooling Train A remained In operation during the event.  Spent Fuel Cooling Train B was restarted following restoration of power to Bus 6.  "The loss of the MAT also resulted in the loss of non-safeguards 4160 V Buses 1-4.  In response to the loss of power to Bus 4, the Technical Support Center (TSC) / Station Blackout (SBO) Diesel started and failed to load onto 480 Volt Bus 46.  This resulted in a loss of power to the Technical Support Center.  The loss of power to the TSC is being reported as a loss of Emergency Assessment Capability.  At 1632, the TSC/SBO Diesel Generator tripped due to the failure of the output breaker to close and provide power to its support equipment.  The cause of the failure of the TSC/SBO output breaker to close is unknown at this time."  The licensee is investigating the cause of the breaker being opened and failure of the TSC/SBO diesel to load.  One Spent Fuel Pool cooling train was in service throughout the event and no pool heatup occurred.  The licensee notified the NRC Resident Inspector.|
Power Reactor|46667|NINE MILE POINT|CONSTELLATION NUCLEAR|1|SYRACUSE|NY|OSWEGO||Y|05000220|1|2||[1] GE-2,[2] GE-5|DAN COLEMAN|JOE O'HARA|3/11/2011 00:00:00|02:32|3/10/2011 00:00:00|20:15|EST|3/11/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||SAM HANSELL|R1DO|||||||||||||||||||N|Y|91|Power Operation|91|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||LOSS OF PRIMARY AND BACKUP METEROLOGICAL MONITORING EQUIPMENT ON BOTH TOWERS  "Loss of power to radiological monitoring equipment (primary and backup).  This constitutes a major loss of emergency assessment capability per Nine Mile Point procedures. The cause of the loss of monitoring equipment is a downed 13.2 kV power line.  Monitoring capability was restored at 0100."  The loss of power was caused by high winds, and the grid operator was able to restore power at 0100 EST.  The NRC Resident Inspector has been notified.|
Power Reactor|46668|DIABLO CANYON|PACIFIC GAS & ELECTRIC CO.|4|SAN LUIS OBISPO COUNTY|CA|SAN LUIS OBISPO||Y|05000275|1|2||[1] W-4-LP,[2] W-4-LP|K.R.THOMPSON|JOE O'HARA|3/11/2011 00:00:00|04:40|3/11/2011 00:00:00|01:23|PST|3/11/2011 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||THOMAS FARNHOLTZ|R4DO|JANE MARSHALL|IRD|ELMO COLLINS|RA|JACK GROBE|NRR|JONES|DHS|BISCOE|FEMA|||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||NOTICE OF UNUSUAL EVENT AS A RESULT OF A TSUNAMI WARNING IN THE AREA  The licensee declared a notice of unusual event as a result of a tsunami warning issued for the coastal areas of California.  The tsunami warning is a result of a 8.9 magnitude earthquake off the coast of Japan.  The licensee is in EAL HU1.5, 'Tsunami Affecting the Protected Area.'  The NRC remains in the normal response mode.   The NRC Resident Inspector has been notified.  * * * UPDATE AT 1134 EST ON 3/11/2011 FROM MIKE QUITTER TO JOE O'HARA * * *   "A classification of unusual event was declared at 0123 PST on March 11, 2011 due to a tsunami warning issued by the NOAA for the California West Coast.  Diablo Canyon Power Plant [DCPP] has implemented the requirements of Casualty Procedure M-5, 'Response to Tsunami Warning.'  Plant personnel were evacuated from the DCPP intake structure at 0742 PST.  Evacuation of personnel from the intake structure constitutes a deviation from DCPP license condition '2.E' and authorized pursuant to 10CFR50.54(x).  "No damage or injuries has been observed as a result of this tsunami event and there is no impact on the health and safety of the general public."  Notified the R4DO (Farnholtz), R4RA (Collins), IRD (Marshall), and NRR (Grobe).  * * * UPDATE AT 1843 EST ON 3/11/2011 FROM JOHN BECERRA TO MARK ABRAMOVITZ * * *  Diablo Canyon has terminated their Unusual Event at 1528 PST because the tsunami warning has been reduced to a tsunami advisory.  No damage occurred during this event.  Notified the R4DO (Farnholtz), IRD (Gott),  NRR (Galloway), DHS (Gates), and FEMA (O'Connell)|
Fuel Cycle Facility|46669|GLOBAL NUCLEAR FUEL - AMERICAS|GLOBAL NUCLEAR FUEL - AMERICAS|2|WILMINGTON|NC|NEW HANOVER|SNM-1097|Y|07001113||||URANIUM FUEL FABRICATION|JULIE OLIVIER|PETE SNYDER|3/11/2011 00:00:00|08:15|3/10/2011 00:00:00|09:15|EST|3/11/2011 00:00:00|NON EMERGENCY|PART 70 APP A (b)(2)|LOSS OR DEGRADED SAFETY ITEMS|||||||RANDY MUSSER|R2DO|BRIAN SMITH|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MALFUNCTIONING VALVE LEADS TO A LOSS OF DEFENSE IN DEPTH  "During a performance of temporary operating procedure on a laser optical device, it was identified that one of two valves used to isolate the device failed to operate.   The valves are operated as a pair and the valves are redundant to provide defense in depth.  One valve shut as expected.  The second valve did not shut.  "The valves are identified as an Item Relied on For Safety (IROFS).   The system was not operating and one of the valves operated as designed. No unsafe condition existed.  "Operability of both valves is required to meet the performance requirements of 10CFR70.61.    "This event is being reported pursuant to the requirements of 10CFR70 Appendix A(b)(2) within 24 hours. The affected device will remain shutdown pending further investigation and implementation of associated corrective actions.  "[This event is of] low safety significance - the discovery did not result in an unsafe condition."|
Hospital|46670|STATE OF FLORIDA|ANAZAOHEALTH CORPORATION|1|Tampa|FL||2975-1|Y||||||STEVE FURNACE|JOHN SHOEMAKER|3/14/2011 00:00:00|10:53|3/9/2011 00:00:00||EST|3/14/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||GLENN DENTEL|R1DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE - DOSE MORE THAN PRESCRIBED  "This information was provided by Danny Ray Jisha, Incident Investigator, Texas Department of State Health Services (DSHS).  Two patients identified to have been over dosed with Phosporus-32 (P-32).  The first patient received 565 Gy where 300 Gy was prescribed, and the second patient received 507 Gy where 200 Gy was prescribed.  It is suspected that Anazaohealth Corporation [Florida] supplied incorrect concentrations/doses of P-32.  The Tampa, Florida Inspection Office will investigate."  The over dose took place at the University of Texas Southwestern Medical Center, Dallas, Texas, however the cause appears to be due to improper labeling of the dose material at Anazaohealth Corporation in Tampa, Florida.  This event is being investigated by the State of Florida, Florida Incident # FL11-020, and the State of Texas.  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.|
Agreement State|46672|NORTH DAKOTA DEPARTMENT OF HEALTH|SANJEL CORP|4||ND|MCKENZIE||Y||||||DAVID STRADINGER|JOHN KNOKE|3/14/2011 00:00:00|23:01|3/13/2011 00:00:00||MST|3/15/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||CHUCK CAIN|R4DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - NUCLEAR GAUGES RECOVERED FROM OIL RIG FIRE  North Dakota Department of Health (NDDH) notified the NRC of two nuclear gauges affected by a fire at an oil rig in Mckenzie County, North Dakota.  The fire is out and one gauge has been retrieved with the shield intact with no unusual radiation readings.  The second gauge has also been retrieved. There is no damage apparent to the second gauge except for some shifting of the lead shielding around the gauge. There was no unusual radiation levels, exposure or contamination impact to personnel.  The gauges are Thermo-Fisher Scientific, model 5192.   Arrangements are being made to ship both gauges to the manufacture for inspection and repair.|
Power Reactor|46673|CRYSTAL RIVER|FLORIDA POWER CORP.|2|CRYSTAL RIVER|FL|CITRUS||Y|05000302|3|||[3] B&W-L-LP|WILLIAM KISNER|JOHN SHOEMAKER|3/15/2011 00:00:00|15:16|3/14/2011 00:00:00|13:11|EST|3/15/2011 00:00:00|NON EMERGENCY||OTHER UNSPEC REQMNT|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||ADDITIONAL DELAMINATED CONTAINMENT CONCRETE DISCOVERED  "Crystal River Unit 3 is currently shutdown.  Final tendon re-tensioning activities were being conducted following repair of delaminated (separated) concrete at the periphery of the containment wall.  At approximately 1311 [EDT] on March 14, 2011, an indication was received from acoustic monitoring instrumentation located on the containment wall outside the previously repaired area.  Re-tensioning activities were stopped and Plant Operations verified that there were no changes in plant parameters.  Non-destructive examinations were initiated and preliminary indications are that there is delaminated concrete in the area identified by the acoustic monitoring.  An assessment concluded that the containment continues to maintain the closure-pressure retaining capability required for Mode 5.  There continues to be no threat to the public heath and safety.  "Delaminated concrete at the periphery of the containment wall was created in late 2009 during the process of creating an opening in the structure to remove and replace the steam generators inside (Reference Event Report 45416).  The unit was already shut down for refueling and maintenance at the time the damage was found and has remained shut down."  The extent of newly discovered delaminated has not yet been determined.   The licensee plans to make a press release.  The licensee has notified the NRC Resident Inspector and the Florida Public Service Commission.|
Power Reactor|46674|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|ERICK MATZKE|JOHN KNOKE|3/15/2011 00:00:00|16:25|3/15/2011 00:00:00|14:56|CST|3/15/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||CHUCK CAIN|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||FLOOD BARRIER PENETRATION NOT SEALED  "During investigations of flood barrier penetrations, a 4 inch conduit has been identified that is not sealed. This conduit penetrates the South wall of the auxiliary building near the transformers into room 19. Flooding through the penetrations could have impacted the ability of the station's auxiliary feedwater (AFW) pumps to perform their design accident mitigation functions.  "This eight-hour notification is being made pursuant to 10 CFR 50.72 (b)(3)(v).  "The penetration is at an elevation of 1007'-8". The river level has been less than 995 feet Mean Sea Level (MSL) since prior to December 1, 2010. The AFW pumps are operable. There are not any indications of conditions that might result in a flood. Actions are in progress to plug the penetration."  The licensee has notified the NRC Resident Inspector.|
Agreement State|46675|MISSISSIPPI DIV OF RAD HEALTH|LEAF RIVER CELLULOSE|4|NEW AUGUSTA|MS||MS-565-02|Y||||||JULIA RALSTON|MARK ABRAMOVITZ|3/15/2011 00:00:00|16:43|1/10/2011 00:00:00||CDT|3/15/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||CHUCK CAIN|R4DO|CHRISTEPHER MCKENNEY|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||DAMAGED FIXED GAUGE SAMPLE WHEEL  The following report was received via fax.  "Licensee's RSO contacted DRH [Mississippi Department of Radiological Health] by letter to report a malfunction of the sample wheel mechanism in their basis weight fixed gauge (Model No. BWM-H Beta Gauge, Gauge Serial No. 25630138, Source Model No. Amersham Corporation Model SIC.L5, Source Serial No. HW338).  The device manufacturer was contacted and a representative removed the original faulty basis weight source head and replaced it with a refurbished unit.  The replacement did not function, so the original source head was repaired and reattached.  A leak test, survey and inspection were all conducted.  As a precaution, the licensee ordered a refurbished source head and the manufacturer's representative came out to replace the original faulty source head.  A leak test, survey and inspection were conducted.  Leak tests revealed no contamination and surveys were normal."  Source:  Sr-90, 15 mCi  Mississippi Tracking Number:  MS-11002|
Power Reactor|46676|NORTH ANNA|DOMINION GENERATION|2|RICHMOND|VA|LOUISA||N|05000338|1|2||[1] W-3-LP,[2] W-3-LP|DON TAYLOR|JOHN KNOKE|3/15/2011 00:00:00|17:38|3/15/2011 00:00:00|11:31|EDT|3/15/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||LOSS OF POWER TO TECHNICAL SUPPORT CENTER VENTILATION SYSTEM  "On 03/15/2011 at 1131 EDT, the Technical Support Center (TSC) ventilation system was rendered nonfunctional as a result of loss of power to the system. System power was lost approximately one hour after returning the normal power supply to service following maintenance. This condition has the potential to render the TSC unavailable due to the inability of the ventilation and filtration system to maintain a habitable atmosphere. Compensatory measures exist to relocate the TSC to alternate locations.  "On 03/15/2011 at 1545 EDT, after verifying the bus supplying the TSC ventilation was satisfactory for return to service, power was restored using an alternate feed. This notification is being made in accordance with 10CFR50.72(b)(3)(xiii) due to the potential loss of an emergency response facility.  "The NRC Senior Resident Inspector was notified."|
Power Reactor|46677|PEACH BOTTOM|EXELON NUCLEAR CO.|1|PHILADELPHIA|PA|YORK & LANCASTER||N|05000277|2|||[2] GE-4,[3] GE-4|CHRIS WEICHLER|HOWIE CROUCH|3/17/2011 00:00:00|00:01|3/16/2011 00:00:00|16:47|EDT|3/17/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||DON JACKSON|R1DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||HIGH PRESSURE COOLANT INJECTION DECLARED INOPERABLE  "On 03/16/11, at 1647 [EST], Peach Bottom Atomic Power Station Unit 2 declared the High Pressure Coolant Injection system inoperable for a condition found during testing which could cause the system to malfunction when swapping suction sources. While lined up to the suppression pool suction flow path, unsatisfactory results were obtained while venting for system fill verification, indicating potential voiding of a portion of the pump discharge piping. Unit 2 HPCI remains available while aligned to its normal suction, the condensate storage tank.  "This report is being submitted pursuant to 10CFR 50.72(b)(3)(v)(D).  "The NRC resident has been informed of this notification."|
Power Reactor|46678|CATAWBA|DUKE ENERGY NUCLEAR LLC|2|YORK|SC|YORK||Y|05000413|1|2||[1] W-4-LP,[2] W-4-LP|MATTHEW PARKER|DONALD NORWOOD|3/17/2011 00:00:00|13:58|3/17/2011 00:00:00|12:50|EDT|3/17/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||OFFSITE NOTIFICATION DUE TO VEHICLE FIRE  "Vehicle fire outside the protected area in the owner controlled area.  Offsite fire department contacted and responded to site.  Fire has been extinguished.  Fire was limited to vehicle and no plant equipment was affected.  Event had no affect on plant operation."  The licensee notified the State of North Carolina, the State of South Carolina, York county, Gaston county, and Mecklenburg county.  The licensee notified the NRC Resident Inspector.|
Hospital|46679|LIBERTY HOSPITAL|LIBERTY HOSPITAL|3|LIBERTY|MO||24-16178-01|N||||||CHRIS MOORE|STEVE SANDIN|3/17/2011 00:00:00|16:21|7/1/2007 00:00:00||CDT|3/17/2011 00:00:00|NON EMERGENCY|20.2201(a)(1)(i)|LOST/STOLEN LNM>1000X|||||||KENNETH RIEMER|R3DO|CHRISTEPHER MCKENNEY|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||TEN TRITIUM EXIT SIGNS DISCOVERED MISSING  On 6/5/2007 during an NRC Inspection, ten (10) tritium exit signs were verified in storage.  Approximately one month later in July 2007, the RSO noticed that the signs were gone and assumed that his Director had dispositioned them, i.e., returned them to the manufacturer for disposal.  On 3/17/2011 at 0800 CDT, in response to a followup question from an NRC Inspector onsite 3/16/2011, it was determined there was no documentation associated with the transfer or disposal of these signs.  The licensee contacted the NRC Inspector who was still in the area and informed her of this situation.  The licensee has no information on the manufacturer, activity or serial numbers of the missing exit signs.  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. For additional information go to  http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf  This source is not amongst those sources or devices identified by the IAEA Code of Conduct for the Safety & Security of Radioactive Sources to be of concern from a radiological standpoint.  Therefore is it being categorized as a less than Category 3 source|
Power Reactor|46680|MILLSTONE|DOMINION GENERATION|1|WATERFORD|CT|NEW LONDON||N|||2||[1] GE-3,[2] CE,[3] W-4-LP|WAYNE HARRELSON|JOE O'HARA|3/18/2011 00:00:00|05:32|3/18/2011 00:00:00|01:45|EDT|3/30/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||DON JACKSON|R1DO|||||||||||||||||||N|N|0||0||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||CONTROL ROOM BOUNDARY RENDERED INOPERABLE  "On March 17, 2011 during a control room walk-down, it was discovered that a halon bottle located in the control room was removed from its associated piping for scheduled work.  The open piping penetrates the control room boundary rendering it inoperable.    "Technical Specification 3.7.6.1 'Control Room Emergency Ventilation System' is applicable in Modes 1, 2, 3, 4, 5 and 6 was entered.  Since control room boundary was rendered inoperable, Dominion is reporting that this condition could have prevented the fulfillment of the safety function to mitigate the consequences of an accident.  Upon discovery the piping was capped re-establishing the control room boundary.    "Further engineering review will be conducted to more fully evaluate the impact on the control room boundary.    "This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(D)."  Offsite power is normal and all emergency diesel generators are operable.  There was no increase in plant risk.     The NRC Senior Resident Inspector has been notified.  * * * UPDATE AT 1002 EDT ON 03/30/11 FROM WAYNE WOOLERY TO S. SANDIN * * *   The Licensee is retracting this report based on the following:  "On March 17, 2011 during a control room walk down at Millstone Power Station Unit 2, it was discovered that a halon bottle located in the control room was removed from its associated piping for scheduled work. Since the associated piping penetrates the control room boundary, operators declared the control room boundary inoperable. Upon discovery, the piping was capped re-establishing the control room boundary.   "Operators made a report in accordance with 10CFR50.72(b)(3)(v)(D).  "Subsequently, an engineering evaluation has been completed that concludes that the piping opening created by the removal of the halon bottle would not have prevented the fulfillment of the safety function to mitigate the consequences of an accident. Therefore, the condition reported in event report 46680 is being retracted.   "The NRC Resident Inspector has been notified."  Notified R1DO (Powell).|
Power Reactor|46681|MCGUIRE|DUKE POWER|2|CORNELIUS|NC|MECKLENBURG||Y|05000369|1|2||[1] W-4-LP,[2] W-4-LP|JIM DAIN|CHARLES TEAL|3/18/2011 00:00:00|14:40|3/18/2011 00:00:00||EDT|3/18/2011 00:00:00|NON EMERGENCY|21.21|UNSPECIFIED PARAGRAPH|||||||REBECCA NEASE|R2DO|PART 21 GROUP||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0|Refueling|0|Refueling|N|N|0||0||PAINT CHIPS DISCOVERED IN WOODWARD GOVERNORS  Woodward governors purchased as nuclear safety related items for use in turbine driven auxiliary feedwater pumps and emergency diesel generators, were found to have paint chips on internal surfaces.  These governors were manufactured by Woodward Governor Company, Loveland, CO for use at the McGuire Nuclear Station.  The NRC Resident Inspector has been informed.|
Fuel Cycle Facility|46682|HONEYWELL INTERNATIONAL, INC.|HONEYWELL INTERNATIONAL, INC.|2|METROPOLIS|IL|MASSAC|SUB-526|Y|04003392||||URANIUM HEXAFLUORIDE PRODUCTION|BOB STOKER|CHARLES TEAL|3/18/2011 00:00:00|16:30|3/8/2011 00:00:00|12:30|CDT|3/18/2011 00:00:00|NON EMERGENCY|40.60(b)(3)|MED TREAT INVOLVING CONTAM|||||||REBECCA NEASE|R2DO|MICHAEL TSCHILTZ|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||UNPLANNED MEDICAL TREATMENT OF POTENTIALLY CONTAMINATED INDIVIDUAL  "A contract employee entered the onsite medical facility with a laceration on his right arm.  First aid was administered in preparation for transferring to an offsite medical facility.  The employee had spreadable contamination on his clothing (60dpm/100cm2).  The employee was decontaminated before leaving the site.  The employee received 8 stitches at the offsite medical facility."|
Agreement State|46683|MINNESOTA DEPARTMENT OF HEALTH|ABBOTT NORTHWESTERN HOSPITAL|3|MINNEAPOLIS|MN||1007-209-27|Y||||||BRYCE ARMSTRONG|DONALD NORWOOD|3/18/2011 00:00:00|16:13|3/17/2011 00:00:00||CDT|3/18/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||KENNETH RIEMER|R3DO|CHRISTEPHER MCKENNEY|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - YTTRIUM-90 MICROSPSHERES ADMINISTERED AT 150% OF PRESCRIBED DOSE  "A medical event took place at Abbott-Northwestern Hospital involving a Yttrium-90 (Y-90) SIR microsphere therapy patient treated on 3/17/2011.  It was discovered on 3/18/11, by the radiation oncologist covering the SIRS procedure from the day before, that the delivered amount of Y-90 SIRS wasn't 105% above the prescribed dose as intended, but actually 150% above the prescribed dose.  She then brought this error to the attention of the lead medical physicist who was the attending medical physicist responsible for this treatment delivery, for further clarification.  Upon investigation, it was deduced that the medical physicist had not read the patient's SIRS therapy (utilizing Y-90 radioactive isotope) written directive prescription correctly.  A higher than intended dosage was administered to the patient (1.66 GBq).  The correct dosage that was intended to be administered per the written directive was 1.11 GBq.  After calculation was made after the incident it was determined that the intended dose to the liver was 30.72 Gy and the actual dose to the liver was 45.93 Gy.   "Contributing factors to the above error identified by the licensee are as follows: "1. The amount of information presented in the SIRS written directive and the prescribed amount of isotope is hard to discern and is not set apart from all the other numbers presented. "2. The prescribed activity is manually transferred to a secondary worksheet used in Nuclear Medicine to draw the dose to be administered and this secondary activity worksheet is not verified by a secondary party.     "The licensee stated that to prevent such an event from occurring in the future, the SIRS written directive document will be modified to display the prescribed activity more predominantly on the form as well as a space for initializing by a secondary party that the prescribed dose has been transferred/entered properly on the secondary activity worksheet that is used in Nuclear Medicine to draw the dose to be administered.  "The referring physicians as well as the patient have been or are in the process of being notified of this event.  "According to the licensee's Radiation Oncologist and Interventional Radiologist that were asked to consult, this higher dose would slightly increase the patient's risk of radiation-induced liver disease.  The patient, as is standard for all SIRS (Y-90) patients, will receive liver function follow-up testing to track her status."   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.|
Agreement State|46684|MINNESOTA DEPARTMENT OF HEALTH|BOISE CASCADE PAPER CORPORATION|3|INTERNATIONAL FALLS|MN||5011-100-36|Y||||||BRANDON JURAN|DONALD NORWOOD|3/18/2011 00:00:00|17:34|3/18/2011 00:00:00||CDT|3/18/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||KENNETH RIEMER|R3DO|CHRISTEPHER MCKENNEY|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - DENSITY GAUGE STUCK SHUTTER  "The licensee Radiation Safety Officer (RSO) was performing a routine shutter check on a density gauge (Berthold model LB7440); the shutter was difficult to turn so the RSO used pliers to move the shutter to the indicated closed position. The shutter did not close.  The detector on the gauge was still reading the same as when the shutter was indicated open. The RSO contacted the manufacturer and is in the process of scheduling them to service the device. The RSO tagged the gauge with a do not operate tag warning people of the problem."|
Power Reactor|46685|WOLF CREEK|WOLF CREEK NUCLEAR OPERATING CORP.|4|BURLINGTON|KS|COFFEY||Y|05000482|1|||[1] W-4-LP|MARCY BLOW|VINCE KLCO|3/19/2011 00:00:00|06:54|3/19/2011 00:00:00|04:04|CDT|3/19/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(iv)(A)|ECCS INJECTION|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||CHUCK CAIN|R4DO|||||||||||||||||||N|N|0|Hot Standby|0|Hot Standby|N|N|0||0||N|N|0||0||SAFETY INJECTION DISCHARGE TO THE REACTOR  "Following a scheduled plant shutdown for refueling the operators were forced to close the Main Steam Isolation Valves (MSIV's) to limit plant cooldown.  While opening MSIV's to restore steam to the secondary, a Reactor Trip and Safety Injection (SI) occurred.  The MSIV bypass valves were opened to equalize pressure across the MSIV's.  Steam header pressure dipped when the MSIV for 'C' Steam Generator (S/G) was opened.  The low steamline pressure bistables are rate sensitive and actuated to cause the SI when steam pressure dipped.  Lowest steamline pressure was 1040 psig, the low steam line pressure SI actuates at 615 psig.    "During the SI the PZR [Pressurizer] PORV's cycled approximately 10 times to limit RCS pressure.  When the PORV's opened the 'B' PZR Code Safety Main Control Board (MCB) and plant computer alarm actuated but the actual MCB indication did not change nor does plant response indicate that a PZR Code Safety opened.  This appears to be an indication problem related to the PORV's cycling."  "All equipment functioned as required."  The station electric buses are aligned to normal offsite power.  Decay heat removal is being discharged to the atmospheric relief valves with no indication of primary to secondary leakage.    The licensee notified the NRC Resident Inspector.   * * * UPDATE FROM MARCY BLOW TO DONALD NORWOOD AT 1221 EDT ON 3/19/2011 * * *  "1.  The expected system actuations that occurred when the plant experienced a Safety Injection (SI) 03/19/11 at 04:04 CDT, previously reported on EN 46685 for 10CFR50.72(b)(2)(iv)(A), is also reportable under 10CFR50.72(b)(3)(iv)(A) for Specified System Actuation.    "2.  During the recovery of the Safety Injection (SI) actuation that occurred 03/19/11 at 04:04 CDT and previously reported on EN 46685, the Safety Injection Signal was reset which blocked any further automatic actuation.  This was directed per the appropriate procedure step.  There is no Technical Specifications allowed condition for both trains of ECCS to be inoperable, therefore the unit entered Tech. Spec. 3.0.3 due to the Auto SI feature being blocked.  LCO 3.5.2 action C.1. directs immediate entry into LCO 3.0.3.  The entry into TS 3.0.3 was made at 0411 CDT and exited at 0639 CDT when the Reactor Trip Breakers were reclosed which re-enabled the automatic SI signal.  This is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation.  "NRC Resident was notified of the update."  Notified R4DO(Cain).|
Power Reactor|46686|BYRON|EXELON NUCLEAR CO.|3|BYRON|IL|OGLE||Y|05000454|1|||[1] W-4-LP,[2] W-4-LP|MIKE LINDEMANN|DONALD NORWOOD|3/19/2011 00:00:00|16:21|3/19/2011 00:00:00|08:00|CDT|3/30/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(A)|DEGRADED CONDITION|||||||KENNETH RIEMER|R3DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||ULTRASONIC EXAMINATION RESULTS IN INDICATIONS ON TWO REACTOR HEAD PENETRATIONS  "On March 19, 2011, during the Byron Station Unit 1 refueling outage, it was determined that the results of planned ultrasonic (UT) examinations performed on two penetrations of the reactor vessel head would not meet the applicable acceptance criteria.  Both require repair prior to returning the vessel head to service.  These indications are not in the reactor coolant pressure boundary; however they are very near the toe of the J-groove weld.  The examinations were being performed to meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1, to ensure the structural integrity of the reactor vessel head pressure boundary.  The UT examinations continue for the remaining head penetrations.  All of the penetrations will be examined during the current refueling outage.  Repairs are currently being planned and will be competed prior to startup.  "This is reportable pursuant to 10CFR50.72(b)(3)(ii)(A) since the as found indications did not meet the applicable acceptance criteria referenced in ASME Code Case N-729-1 to remain in-service without repair.  "The NRC Resident Inspector has been notified."  * * * UPDATE FROM BLAINE PETERS TO JOHN SHOEMAKER AT 0910 EDT ON 03/30/11 * * *  "Ultrasonic examination [made on March 19, 2011] resulted in indications on two reactor head penetrations.  "As mentioned in Event Notification 46686, reactor vessel head penetrations [In-service inspection] ISI examinations were still in progress.  On Wednesday, March 30, 2011, two additional Unit 1 reactor head penetrations were found to contain indications that will require repair prior to returning the reactor head to service.  The indications on these two penetrations are within the reactor vessel head pressure boundary."  The NRC Resident Inspector has been notified by the licensee.  Notified the R3DO (Peterson).|
Power Reactor|46687|GRAND GULF|ENTERGY NUCLEAR|4|PORT GIBSON|MS|CLAIBORNE||Y|05000416|1|||[1] GE-6|RICKY LIDDELL|VINCE KLCO|3/20/2011 00:00:00|07:08|3/19/2011 00:00:00|22:36|CDT|3/20/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||CHUCK CAIN|R4DO|||||||||||||||||||N|Y|96|Power Operation|96|Power Operation|N|N|0||0||N|N|0||0||HIGH PRESSURE CORE SPRAY VALVE BREAKER OVERCURRENT TRIP SETPOINT OUT OF TOLERANCE  "During testing of the High Pressure Core Spray [HPCS] system minimum flow valve breaker, it was discovered that the overcurrent trip setpoint was out of tolerance. This testing was being performed as a result of a breaker trip that occurred during a surveillance. The discovery occurred while the system was inoperable for maintenance and no TS [Technical Specification] limits or action times were exceeded."  The overcurrent trip setpoint was placed within tolerance and HPCS is in operable status.  The licensee notified the NRC Resident Inspector.|
Power Reactor|46688|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|DONNA GUINN|CHARLES TEAL|3/21/2011 00:00:00|13:19|3/18/2011 00:00:00||CDT|3/21/2011 00:00:00|NON EMERGENCY|21.21|UNSPECIFIED PARAGRAPH|||||||GREG PICK|R4DO|PART 21 GROUP||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||DEFECTIVE KPR-14DG-125 RELAYS DISCOVERED DURING BENCH TESTING  "The following event description is based on information currently available.  The condition is reported under 10 CFR 21.21(d)(3)(i).  "Part 21 Report - Potter Brumfield KRP-14DG-125 Relays, supplied by Southern Testing Services (STS) division of Argo Turboserve Corporation (ATC).  "On March 17, 2011, during bench testing of the KRP-14DG-125 Relays, Fort Calhoun Station (FCS) discovered that some of these relays were defective in that one of the contacts would not close properly after energizing and de-energizing the relay coils.  If installed in the plant, the improper closure of this contact could defeat the safety function of the relays that provide a signal to the component supported by that contact.  This deviation from the design specifications is reportable per 10 CFR 21.  FCS does not have any of the affected relays installed in the plant.  "FCS returned the batch of relays to the vendor for further failure modes and effects evaluation and reporting.    "FCS has not provided any of these relays from our stock to any other licensee.  "The vendor and the [NRC] Resident Inspector have been notified."|
Power Reactor|46689|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|ROBERT KROS|MARK ABRAMOVITZ|3/22/2011 00:00:00|14:22|3/22/2011 00:00:00|10:58|CDT|3/22/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||GREG PICK|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||CONTAINMENT COOLERS DECLARED INOPERABLE  "At 10:58 CDT, today during the performance of IC-ST-IA-3010B, I&C found NG-HCV-400A-A3, CCW INLET VALVE HCV-400A NITROGEN SUPPLY ISOLATION VALVE, closed which is required to remain open for VA-3A to remain operable.  This valve supplies backup nitrogen to VA-3A CCW cooler isolation valve HCV-400A on loss of instrument air to maintain cooling flow to the ventilation during an accident condition.  While the nitrogen valve NG-HCV-400A-A3 was closed, performance of IC-ST-IA-3010B on VA-3B, CONTAINMENT AIR RECIR FAN, placed the containment cooler in an inoperable status.  This led to Technical Specification 2.0.1 entry due to both trains of cooling being inoperable.  The cause for the mispositioning of NG-HCV-400A-A3 is unknown at this time.  The inoperability of VA-3A along with VA-3B rendered the containment cooling trains unavailable to perform their safety function during an accident condition.  This condition is being reported pursuant to 10 CFR 50.72(b)(3)(V)(D) for mitigating the consequences of an accident.  FCS entered into Technical Specification at 10:58 CDT and exited Technical Specification at 11:14 CDT."  The licensee notified the NRC Resident Inspector.|
Power Reactor|46690|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|ERICK MATZKE|MARK ABRAMOVITZ|3/22/2011 00:00:00|17:11|3/22/2011 00:00:00|13:08|CDT|3/22/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||GREG PICK|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||POTENTIAL FLOODING OF RAW WATER PUMP  "During ongoing investigations of flood barrier penetrations at the station, a weakness in the flood protection strategy that would prevent protection of the raw water pumps for floods above 1007'-6" Mean Sea Level (MSL) was discovered.  Cell in-leakage through penetrations at 997' 10" MSL would be beyond the capacity of the raw water pumps.  During the preparation of a calculation to demonstrate the validity of this method it was determined that the grid backwash pipe for each grid and the surface sluice penetrate the east wall of the intake structure through an unsealed penetration (a total of 7 penetrations).  The grid backwash line is an 18" pipe passing through a 24" sleeve.  Flooding through the penetrations could have impacted the ability of the station's raw water pumps to perform their design accident mitigation functions.    "This eight-hour notification is being made pursuant to 10 CFR 50.72 (b)(3)(v).  "The penetration is at an elevation of 997'10" MSL.  The design flood for the station is at 1014' MSL.  The raw water pumps would not be affected until a river level of 1007'6" MSL was reached.  The river level is currently approximately 993.5' MSL and has been less than 995'MSL since prior to December 1, 2010.  The raw water pumps are currently operable.  The National Weather Service Weather Forecast Office is predicting a rise in river level of 2 feet over the next 5 days. Actions are in progress to seal the penetration."  The licensee has instituted a temporary plug contingency plan dependant on river level.  The licensee will notify the NRC Resident Inspector.|
Agreement State|46691|UTAH DIVISION OF RADIATION CONTROL|APPLIED GEOTECHNICAL ENGINEERING CONSULTANTS INC|4|LINDON|UT||UT1800298|Y||||||GWYN GALLOWAY|MARK ABRAMOVITZ|3/22/2011 00:00:00|20:45|3/22/2011 00:00:00|17:35|MDT|3/22/2011 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||GREG PICK|R4DO|JIM WHITNEY|ILTA|ANGELA MCINTOSH|FSME|||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MOISTURE DENSITY GAUGE STOLEN AND RECOVERED  "A Troxler Electronic Laboratories, Inc. Model 3430, portable gauging device [serial number 22936, containing approximately 8.0 millicuries of cesium-137, and approximately 40 millicuries of americium-241/beryllium] was stolen from the licensee's vehicle while parked at the Home Depot in Lindon, Utah. The Cs-137 source was in the safe shielded position when it was stolen and the transportation case was also secured. The device had been secured by two independent physical barriers, but both barriers were breached. The device was recovered at approximately 5:55 p.m. MST by licensee personnel. The transportation case had been opened, but the source rod was still secured in the shielded position.  "The licensee's vehicle was an open bed pickup truck with a mechanism to secure the device as required."  Utah Report: UT110001  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. For additional information go to  http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf|
Part 21|46692|AUTOMATIC VALVE|AUTOMATIC VALVE|3|NOVI|MI|||N||||||KEVIN ARMSTRONG|PETE SNYDER|3/23/2011 00:00:00|16:24|3/22/2011 00:00:00||EDT|3/23/2011 00:00:00|NON EMERGENCY|21.21|UNSPECIFIED PARAGRAPH|||||||JOHN ROGGE|R1DO|DANIEL RICH|R2DO|PART 21 GROUP|EMAI|||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||VALVE PLUNGER GUIDE MALFUNCTION  "Initial Concern: Valve, serial number 73386, does not consistently return to the closed position."  "Nature of the Defect: A dent in the plunger guide may prevent the valve from changing state.  "Number and Location of Components: "Model: U0204GBBR-AA      Quantity: 32  Customer: EXELON LIMERICK, "Model: U0204GBBR-DEEL  Quantity: 30  Customer: ALABAMA POWER FARLEY "Model: U0204GBBR-DE      Quantity: 4    Customer: ALABAMA POWER FARLEY "Model: U0204FBBR-DE      Quantity: 8    Customer: RALPH A. HILLER "Model: U0204GBBR-DEL   Quantity: 4    Customer: DRESSER MASONEILAN "Model: U0204GBBR-DEP   Quantity: 3   Customer: DRESSER MASONEILAN  "Advice to Purchasers:  "Any of the valves identified above may be inspected by removing the coil and checking the plunger guide for any defects. A dent which will prevent plunger movement is noticeable without magnification.  It typically occurs approximately 0.10 inches from the base of the valve.  Any valves thought to contain defects will be rebuilt or replaced by Automatic Valve."|
Power Reactor|46693|CALLAWAY|AMEREN UE|4|FULTON|MO|CALLAWAY||N|05000483|1|||[1] W-4-LP|DAVID HURT|JOHN SHOEMAKER|3/24/2011 00:00:00|06:02|3/23/2011 00:00:00|23:54|CDT|3/24/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||GREG PICK|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||PRESSURE TRANSMITTERS NEEDED FOR AUXILIARY FEEDWATER SUCTION PATH NOT ANALYZED FOR POTENTIAL HIGH ENERGY LINE BREAK  "While performing an extent of condition review of high energy line break (HELB) analyses, a detailed review of the auxiliary steam system was being performed.  During this review, sections of pipe that run through rooms 1206/1207 in the Auxiliary Building were identified that have design ratings indicating that they could possibly be classified as high energy lines.   "The pipes were verified to have not been considered in the current HELB analyses. This condition affects pressure transmitters ALPT0037, 38, & 39 which are not qualified for operation in a harsh environment.  These pressure transmitters provide the Auxiliary Feedwater Pump [AFW] Suction Transfer signal on low suction pressure from the non safety Condensate Storage Tank to the Safety Related supply (Essential Service Water).  "Technical Specification [TS] 3.3.2-6.h bases state: "since these detectors are in an area not affected by HELBs or high radiation, they will not experience any adverse environmental conditions and the Trip Setpoint reflects only steady state instrument uncertainties."   "Based upon the above bases, with the identified aux steam lines in service, the pressure transmitter's operability could not be assured.  This represented an unanalyzed condition and had the potential to affect equipment used for accident mitigation.  TS 3.0.3 was entered at time 2354 [CST] on 3/23/2011.  At 0009 [CST] on 3/24/2011, Aux Steam valves FBV0158, FBV0I48, FAV0002, and FAV0003 were isolated, removing the HELB concern [TS 3.0.3 was exited at this time].  These are the active feed [isolation valves] to the lines passing through the Aux Building Rooms 1206/1207."  The licensee notified the NRC Resident Inspector.|
Power Reactor|46694|BRAIDWOOD|EXELON NUCLEAR CO.|3|BRACEVILLE|IL|WILL||Y|||2||[1] W-4-LP,[2] W-4-LP|SCOTT BUTLER|JOHN SHOEMAKER|3/24/2011 00:00:00|11:39|3/24/2011 00:00:00|10:18|CDT|3/24/2011 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||JAMNES CAMERON|R3DO|WILLIAM GOTT|IRD|ERIC LEEDS|NRR|CYNTHIA PEDERSON|R3|JOHN KNOX|DHS|LORI BURCKART|FEMA|||||||||N|N|0||0||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||UNEXPECTED LOSS OF ANNUCIATORS DURING PLANNED MAINTENANCE  "During a planned maintenance activity on the Unit 2 main control room alarm cabinets, it was identified that all Unit 2 safety system annunciators were lost.  This was identified at 1006 [CDT].  Main Control Board indicators remained functional.  At 1018 [CDT], the Shift Manager declared an Unusual Event under Emergency Action Level MU6.  This was due to an unplanned loss of most (approximately 75%) safety system annunciators for > 15 minutes.  The planned maintenance activity was not expected to affect the amount of annunciators that were lost.  "At 1030 [CST], the [planned maintenance] clearance order was cleared and power was restored to the Unit 2 annunciators."  There was not transient on other plant equipment and the plant remained stable before and after this event.  The cause for the unexpected loss of annunciators is not clearly understood and is still under investigation.  The licensee notified the NRC Resident Inspector.    * * * UPDATE FROM SCOTT BUTLER TO JOHN SHOEMAKER AT 1216 EDT ON 03/24/11 * * *  The Unusual Event was terminated at 1047 CDT on 03/24/11.  All annunciators have been restored and an investigation will be conducted to determine the cause.  Notified R3DO (Cameron)|
Non-Agreement State|46695|GRADY MEMORIAL HOSPITAL|VARIAN MEDICAL SYSTEMS|1|CHARLOTSVILLE|VA||45-30957-01|Y||||||RICHARD PICCOLO|CHARLES TEAL|3/24/2011 00:00:00|16:20|3/23/2011 00:00:00|03:00|EDT|3/24/2011 00:00:00|NON EMERGENCY|30.50(b)(2)|SAFETY EQUIPMENT FAILURE|||||||JOHN ROGGE|R1DO|KEVIN HSUEH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||EMERGENCY RETRACT MECHANISM FAILURE  The Varian equipment representative provided notification of the following event that occurred at the Grady Memorial Hospital in Atlanta, GA.  A technician was installing a Varisource IX high-dose afterloader when the active wire composed of a 10 Ci Ir-192 source failed to extend.  After troubleshooting it was discovered that the wire was stuck on the wedge block which is part of the emergency retract mechanism.  The active wire was removed and the emergency retract mechanism was replaced.    The technician received 0.2 mrem during the repair work.|
Fuel Cycle Facility|46696|LOUISIANA ENERGY SERVICES|LOUISIANA ENERGY SERVICES, L.P.|2|EUNICE|NM|LEA|SNM-2010|Y|70-3103|||||JACK ROLLINS|PETE SNYDER|3/24/2011 00:00:00|19:02|1/28/2011 00:00:00||MDT|3/24/2011 00:00:00|NON EMERGENCY|21.21|UNSPECIFIED PARAGRAPH|||||||DANIEL RICH|R2DO|PART 21 MATERIALS GR|EMAI|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||TURNBUCKLE COLLAR THREADS OUT OF TOLERANCE  "Name of firm constructing or supplying the basic component which fails to comply or contains a defect: OFI Fabrication, Richmond, VA.  "Safety Hazard: Failure of turnbuckles could ultimately cause a failure of IROFS41 (i.e. breach of UF6 piping) during a seismic event resulting in release of licensed material, but this release would not exceed 10 CFR 70.61 requirements for the public or workers.  "The date on which the information or such defect or failure to comply was obtained: January 28, 2011.  "In a case of a basic component which contains a defect or a failure to comply, the number and location of all such components in use at or supplied for the URENCO USA Facility: Currently there are 164 non-defective components installed and 96 non-defective components in storage.  "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: Replacement NQA-1 turnbuckles were installed; and the Receipt Inspection Plan (RIP) for turnbuckles (collar as well as paddle assembly) was revised to include 100 percent verification of thread tolerances. All out of tolerance turnbuckle collars were returned to the vendor.  "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: Implement 100 percent verification of turnbuckle collar thread tolerances using appropriate 'Go-No Go' thread gauges."|
Power Reactor|46697|NORTH ANNA|DOMINION GENERATION|2|RICHMOND|VA|LOUISA||N|05000338|1|2||[1] W-3-LP,[2] W-3-LP|DON TAYLOR|CHARLES TEAL|3/25/2011 00:00:00|11:14|3/24/2011 00:00:00|14:00|EDT|3/25/2011 00:00:00|NON EMERGENCY||OTHER UNSPEC REQMNT|||||||DANIEL RICH|R2DO|MARISSA BAILEY|NMSS|||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||NON-COMPLIANCE WITH NUHOMS TECHNICAL SPECIFICATION  "This 24-hour report is being issued in accordance with the requirements of NRC Certificate of Compliance 1030, Amendment 0, for the NUHOMS Storage System, Technical Specification (TS) 2.2, Functional Operating Limit Violations.  "During a review of historical North Anna NUHOMS dry storage canister (DSC) loading certification documents, a discrepancy was identified.  The NUHOMS Certificate of Compliance 1030 Amendment 0 Technical Specifications include a Figure 2, "Heat Load Zones" which specifies the maximum decay heat load for each of the 32 assembly locations in a DSC.  The figure includes limits for two zone '1b' locations and two zone '1a' locations in the four center locations of the DSC.  The zone '1b' decay heat limit of 0.8 kw is specified for the two 'upper compartments' and zone '1a' decay heat limit of 1.05 kw is specified for the two 'lower compartments' on the figure.  Contrary to this, the loading certifications for 7 of 10 DSCs already loaded at NAPS (North Anna Power Station) were not developed to maintain this orientation when loaded in the horizontal storage module (HSM).  As a result, the DSC zone '1b' heat load limits were exceeded in some cases for these 7 DSCs.  "The heat load limit for all other zones in the DSCs are symmetric, and those assemblies were verified to the correct limit and are unaffected by this error.  In addition the total heat load limit for the sum of the center assemblies was met for all DSCs.  The maximum heat load of any zone '1b' assembly at the time of loading was 0.859 kw, which is slightly higher than the 0.8 kw limit.  The lower heat load of assemblies in the other compartments offset the slightly higher heat load effects, and it is expected that the thermal analysis acceptance criteria would still have been met at the time of loading.  "The decay heat of the assemblies has continued to decrease since their initial loading and it was confirmed that 12 of the 13 assemblies that initially exceeded the 0.8 kw limit currently meet the zone '1b' heat load limits.  The current decay heat of the remaining assembly is slightly above the 0.8 kw limit.  Based on the offsetting margins identified above all of the affected DSCs are currently in a safe condition as loaded in the HSMs."  The NRC Resident Inspector has been informed.|
Power Reactor|46698|SURRY|DOMINION GENERATION|2|SURRY|VA|SURRY||N|05000280|1|2||[1] W-3-LP,[2] W-3-LP|RETT GARNER|CHARLES TEAL|3/25/2011 00:00:00|11:40|3/25/2011 00:00:00||EDT|3/25/2011 00:00:00|NON EMERGENCY||OTHER UNSPEC REQMNT|||||||DANIEL RICH|R2DO|MARISSA BAILEY|NMSS|||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|98|Power Operation|98|Power Operation|N|N|0||0||NON-COMPLIANCE WITH NUHOMS TECHNICAL SPECIFICATION  "This 24-hour report is being issued in accordance with the requirements of NRC Certificate of Compliance 1030, Amendment 0, for the NUHOMS Storage System, Technical Specification (TS) 2.2, Functional Operating Limit Violations.  "During a review of historical Surry NUHOMS Dry Storage Canister (DSC) loading certification documents, a discrepancy was identified.  The NUHOMS Certificate of Compliance 1030 Amendment 0 Technical Specifications include a Figure 2, 'Heat Load Zones' which specifies the maximum decay heat load for each of the 32 assembly locations in a DSC.  The figure includes limits for two zone '1b' locations and two zone '1a' locations in the four center locations of the DSC.  The zone '1b' decay heat limit of 0.800 kw is specified for the two 'upper compartments' and zone '1a' decay heat limit of 1.05 kw is specified for the two 'lower compartments' on the figure.  Contrary to this, the loading certifications for 6 of 12 DSCs already loaded at Surry were not developed to maintain this orientation when loaded in the horizontal storage module (HSM).  As a result, the DSC zone '1b' heat load limits were exceeded in some cases for these 4 DSCs.  "The heat load limit for all other zones in the DSCs are symmetric, and those assemblies were verified to the correct limit and are unaffected by this error.  In addition the total heat load limit for the sum of the center assemblies was met for all DSCs.  The maximum heat load of any zone '1b' assembly at the time of loading was 0.806 kw, which is slightly higher than the 0.800 kw limit.  The lower heat load of assemblies in the other compartments offset the slightly higher heat load effects, and it is expected that the thermal analysis acceptance criteria would still have been met at the time of loading.  "The decay heat of the assemblies has continued to decrease since their initial loading and all assemblies currently meet the upper central compartment limits.  The affected fuel assemblies are in a safe condition as required by NUHOMS TS 2.2.1.  "The NRC Resident Inspector has been notified of this event."|
Power Reactor|46699|PRAIRIE ISLAND|NUCLEAR MANAGEMENT COMPANY|3|WELCH|MN|GOODHUE||N|05000282|1|||[1] W-2-LP,[2] W-2-LP|TERRY BACON|BILL HUFFMAN|3/25/2011 00:00:00|17:27|3/25/2011 00:00:00|10:53|CDT|3/29/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|||||||JAMNES CAMERON|R3DO|||||||||||||||||||N|Y|99|Power Operation|98|Power Operation|N|N|0||0||N|N|0||0||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|46700|BEAVER VALLEY|FIRSTENERGY NUCLEAR OPERATING COMPANY|1|SHIPPINGPORT|PA|BEAVER||N|||2||[1] W-3-LP,[2] W-3-LP|DANIEL SCHWER|BILL HUFFMAN|3/25/2011 00:00:00|21:35|3/25/2011 00:00:00|14:00|EDT|3/27/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||JOHN ROGGE|R1DO|||||||||||||||||||N|N|0||0||N|N|0|Refueling|0|Refueling|N|N|0||0||BOTH TRAINS OF EMERGENCY DIESEL GENERATORS ARE INOPERABLE  "On March 25, 2011, the Train B emergency diesel generator (2EGS-EG2-2) was inoperable and unavailable due to being out of service for scheduled maintenance. At 1400 hours, the Train A emergency diesel generator (2EGS-EG2-1) was declared inoperable, but available, after questions were raised about the adequacy of the assembly method for fuel injection line compression fittings by the manufacturer.  Without assurance that the fittings meet full qualification requirements, the Train A emergency diesel generator was declared inoperable.  "The Unit is currently in Mode 6 with fuel loaded and the upper internals installed in the reactor vessel. The reactor vessel head is removed with 23 feet of water in the cavity, two Operable Residual Heat Removal trains one of which is in operation. With both emergency diesels inoperable, the safety functions needed for accident mitigation could be impaired in the event of a loss of off-site power. Actions are currently in progress to restore one emergency diesel generator to an Operable status.  "This event is reportable pursuant to 10 CFR 50.72(b)(3)(v)(D) due to both emergency diesel generators being inoperable. This event will be evaluated for 10 CFR Part 21 applicability."  The licensee is in Technical Specification 3.8.2.  With both diesels declared inoperable, they have to suspend all core alterations and possible reactivity additions and return an EDG to service.  The licensee plans to return the B train EDG an available status (but not Operable per Technical Specification) by tomorrow.  The licensee will then replace the discrepant fuel injector line compression fittings on the A train and return it to Technical Specification Operable status.  The licensee has notified the NRC Resident Inspector.  * * * UPDATE AT 0226 ON 3/27/11 FROM DANIEL SCHWER TO MARK ABRAMOVITZ * * *  "Following replacement of the questionable fuel injection line compression fittings and successful surveillance and post maintenance testing, the Train A emergency Diesel Generator (2EGS-EG2-1) was declared OPERABLE at 0058 hours on 3/27/2011.  One Diesel Generator was maintained available at all times while the issue was being addressed."  The licensee notified the NRC Resident Inspector.  Notified the R1DO (Rogge).|
Power Reactor|46701|DIABLO CANYON|PACIFIC GAS & ELECTRIC CO.|4|SAN LUIS OBISPO COUNTY|CA|SAN LUIS OBISPO||Y|||2||[1] W-4-LP,[2] W-4-LP|DOUGLAS DYE|HOWIE CROUCH|3/26/2011 00:00:00|21:10|3/26/2011 00:00:00|14:49|PDT|3/26/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(iv)(B)|RPS ACTUATION - CRITICAL|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||GREG PICK|R4DO|||||||||||||||||||N|N|0||0||M/R|Y|88|Power Operation|0|Hot Standby|N|N|0||0||MANUAL REACTOR TRIP WHEN A LOSS OF THE 2-1 MAIN FEEDWATER PUMP OCCURRED  "This notification provides the 4-hour non-emergency event report for the manual reactor trip of Diablo Canyon Power Plant Unit 2 in accordance with 10 CFR 50.72(b)(2)(iv)(B) 'RPS Actuation (scram)'. Additionally, this notification provides the 8-hour non-emergency event report of the automatic actuation of the auxiliary feedwater system as a result of the reactor trip in accordance with 10 CFR 50.72(b)(3)(iv)(A) 'Specified System Actuation'.  "On March 26, 2011 at 1449 PDT operators at Diablo Canyon Power Plant Unit 2 manually initiated a reactor trip in response to loss of main feedwater pump 2-1. The main feedwater pump is believed to have tripped due to non-radioactive water spray on its control console. The water spray was caused by leakage from the flange of the relief valve on the feedwater heater 2-1A condenser dump valve line. Emergency plan activation was not required.  The unit is stable in mode 3 (Hot Standby) with offsite power being supplied to all buses via the 230 kV startup circuit. Diesel generators 2-1 and 2-2 remain OPERABLE in standby. Diesel generator 2-3 remains unavailable due to pre-planned maintenance. All rods fully inserted on the reactor trip. The reactor is being cooled by the auxiliary feedwater system with the condenser in service. All systems performed as designed with no unexpected pressure or level transients. ECCS actuation was not required. Automatic main feedwater isolation, auxiliary feedwater actuation, and steam generator blowdown isolation occurred as expected.  "Unit 1 was unaffected by this event and remains at 100% power."  The licensee has notified the NRC Resident Inspector, San Luis Obispo County, and the State of California.|
Power Reactor|46702|SEABROOK|FPL ENERGY SEABROOK|1|MANCHESTER|NH|ROCKINGHAM||Y|05000443|1|||[1] W-4-LP|JEFF MCNALLY|JOHN SHOEMAKER|3/28/2011 00:00:00|12:38|3/28/2011 00:00:00|11:43|EDT|3/28/2011 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||RAY POWELL|R1DO|SCOTT MORRIS|IRD|MIKE CHEOK|NRR|ERIC LEEDS|NRR|DAVID LEW|R1|DENNIS VIN|FEMA|ROD BRADSHAW|DHS|||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||UNUSUAL EVENT DECLARED DUE TO SMOKE COMING FROM AN ELEVATOR CONTROL CABINET  At 1129 EDT on 03/28/11, smoke was detected coming from an elevator power supply transformer in the "B" Residual Heat Removal (RHR) vault.  The licensee declared a Unusual Event at 1143 EDT.  At 1146 EDT the fire brigade responded and found smoke only and no flame.  The smoke stopped when the cabinet was de-energized.  At 1158 EDT, the fire was proven to be out when the cabinet was opened.  No fire fighting extinguishing agents were used.  There was no damage to other plant equipment and no personnel injuries.  At 1151 EDT, the licensee notified the States of New Hampshire and Massachusetts and at 1209 EDT, activated their Technical Support Center (TSC).  The plant was operating at 100% power and remained stable during and after the event.    The licensee notified the NRC Resident Inspector.  * * * UPDATE FROM JEFF MCNALLY TO JOHN SHOEMAKER AT 1341 EDT ON 03/28/11 * * *  "At 1243 EDT on 03/28/11, the Unusual Event was terminated.  Subsequent investigation has revealed that no plant equipment damage beyond a power transformer that supplies an equipment elevator occurred.  There was no indication of [visible] flame (smoke only) and no personnel injuries as a result of the event."  The licensee notified the NRC Resident Inspector.  Notified R1DO (Powell), NRR EO (Cheok), IRD (Morris).|
Power Reactor|46704|NINE MILE POINT|CONSTELLATION NUCLEAR|1|SYRACUSE|NY|OSWEGO||Y|05000220|1|||[1] GE-2,[2] GE-5|BETHANY HINCKLEY|JOE O'HARA|3/29/2011 00:00:00|02:48|3/29/2011 00:00:00|01:55|EDT|3/29/2011 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||RAY POWELL|R1DO|WILLIAM DEAN|RA|JACK GROBE|NRR|SCOTT MORRIS|IRD|MIKE CHEOK|NRR|HILL|DHS|BARDEN|FEMA|||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||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|46707|BRAIDWOOD|EXELON NUCLEAR CO.|3|BRACEVILLE|IL|WILL||Y|05000456|1|2||[1] W-4-LP,[2] W-4-LP|JOE KLEVORN|HOWIE CROUCH|3/30/2011 00:00:00|01:34|3/29/2011 00:00:00|20:00|CDT|3/30/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|50.72(b)(3)(v)(B)|POT RHR INOP|||||HIRONORI PETERSON|R3DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||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|46708|BYRON|EXELON NUCLEAR CO.|3|BYRON|IL|OGLE||Y|05000454|1|2||[1] W-4-LP,[2] W-4-LP|ALAN GUSTAFSON|JOE O'HARA|3/30/2011 00:00:00|01:39|3/29/2011 00:00:00|20:00|CDT|3/30/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|50.72(b)(3)(v)(B)|POT RHR INOP|||||HIRONORI PETERSON|R3DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||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.|
Power Reactor|46709|MCGUIRE|DUKE POWER|2|CORNELIUS|NC|MECKLENBURG||Y|||2||[1] W-4-LP,[2] W-4-LP|TONY COOK|HOWIE CROUCH|3/30/2011 00:00:00|07:01|3/30/2011 00:00:00|00:10|EDT|3/30/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||MARVIN SYKES|R2DO|||||||||||||||||||N|N|0||0||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||CONTROL ROD MALFUNCTION DURING ROD TESTING RESULTS IN OPERATORS MANUALLY OPENING TRIP BREAKERS  "Rod L-13 did not function as expected during control rod movement test.  This rod is in Shutdown Bank C. When withdrawing this bank, rod L-13 did not withdraw and when the bank was manually inserted, rod L-13 began to withdraw.  The [operating] crew went to Enclosure 13.2 of the procedure to deal with the misaligned rods.  This enclosure has procedural guidance to open the reactor trip breakers, if desired.  The reactor trip breakers were opened and all 211 rods are fully inserted. The reactor was not critical. This activity was performed twice [at the request of reactor engineering]."  The licensee will remain in Mode 5 (Cold Shutdown) until troubleshooting and repair is completed.  The licensee will be notifying the NRC Resident Inspector.|
Fuel Cycle Facility|46710|GLOBAL NUCLEAR FUEL - AMERICAS|GLOBAL NUCLEAR FUEL - AMERICAS|2|WILMINGTON|NC|NEW HANOVER|SNM-1097|Y|07001113||||URANIUM FUEL FABRICATION|SCOTT MURRAY|JOE O'HARA|3/30/2011 00:00:00|09:07|3/29/2011 00:00:00|10:00|EDT|3/30/2011 00:00:00|NON EMERGENCY|PART 70 APP A (b)(1)|UNANALYZED CONDITION|||||||MARVIN SYKES|R2DO|TIM MCCARTIN|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||COMPLETION OF ISA ACTION PLAN CONVERSION MILESTONE  "In response to a Notice of Violation (NOV), Global Nuclear Fuels - America (GNFA) committed to perform a review of the existing Integrated Safety Analysis (ISA). An ISA Action Plan and schedule for performing the ISA review was described in GNF-A's response to the NOV and the first milestone (conversion) was scheduled for completion by January 31, 2011. This milestone was subsequently extended by approximately 60 days.  "On 3/29/11, GNF-A completed the ISA review for the conversion area and has identified 87 existing safety controls that are now being designated as items relied on for safety (IROFS). Implementation of the revised safety basis, IROFS and application of management measures to the new IROFS will be completed within 90 days per the ISA Action Plan.  "Because the revised ISA has designated existing safety controls as additional IROFS, GNF-A is making a report of this completion pursuant to the reporting requirements of 10CFR70 Appendix A(b)(1) within 24 hours.  "Safety Significance of Events: There was no event or plant condition that resulted in a degraded safety condition.    "Safety Equipment Status: Existing conversion area safety controls have now been designated as IROFS per ISA Action Plan. Controls are available to perform their safety function.  "Status of Corrective Actions: Conversion Area milestone complete. ISA Action Plan continues."  The licensee intends to discuss this issue further with the State of North Carolina, New Hanover County, and the NRC Region 2 office.|
Power Reactor|46711|SAN ONOFRE|SOUTHERN CALIFORNIA EDISON  COMPANY|4|SAN CLEMENTE|CA|SAN DIEGO||Y|||2|3|[1] W-3-LP,[2] CE,[3] CE|EDGAR DEGIOVANNI|VINCE KLCO|3/30/2011 00:00:00|16:24|3/30/2011 00:00:00|07:00|PDT|3/30/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||RYAN LANTZ|R4DO|||||||||||||||||||N|N|0||0||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|EMERGENCY SIRENS OUT OF SERVICE DUE TO INADVERTENT SECURING OF POWER  "City of San Clemente inadvertently secured power to 19 Community Alert Sirens. Trouble alarms were received at 0700 [PDT] but it is not known at what time specifically power was removed. All sirens were functional the day before when an inspection was performed on 3/29/2011 at 1213 [PDT]. Power was restored to the sirens at 1145 [PDT] 3/30/2011. This event resulted in 20 community sirens being non-functional for greater than 1 hour.  "Siren State Parks 4 (SP-4) was also non-functional for unrelated reasons and remains out of service."  Power was restored and all but the state park siren capabilities were declared operable.  The licensee notified the NRC Resident Inspector.|
Power Reactor|46712|BRAIDWOOD|EXELON NUCLEAR CO.|3|BRACEVILLE|IL|WILL||Y|||2||[1] W-4-LP,[2] W-4-LP|MIKE DEBOARD|VINCE KLCO|3/30/2011 00:00:00|17:00|3/30/2011 00:00:00|15:38|CDT|3/30/2011 00:00:00|UNUSUAL EVENT||OTHER UNSPEC REQMNT|||||||HIRONORI PETERSON|R3DO|WILLIAM GOTT|IRD|THOMAS BLOUNT|NRR|||||||||||||||N|N|0||0||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||DISCOVERY OF AFTER-THE-FACT EMERGENCY CONDITION (UNUSUAL EVENT)  "An extent of condition review of Braidwood Unit 2 unplanned loss of safety system annunciators Emergency Plan Unusual Event on March 24, 2011 (ENS number 46694) was performed for both Units of Braidwood Station. During this review it was identified that a previous unknown loss of annunciators had also occurred on August 10, 2010 from 1024 to 1136 CT on Unit 2. This condition occurred during planned maintenance on annunciator cabinet 2PA19J power supply capacitors.  "The maintenance performed on August 10, 2010 would normally not cause a loss of all Unit 2 annunciators. During the work, it was expected to lose approximately one third of the annunciators. Latent annunciator system problems identified from the March 24, 2011 event caused a loss of all Unit 2 annunciators and contributed to this condition being unknown to Main Control Room operators. All Unit 2 indications and computer points to the sequence of events recorder remained available and Unit 2 was stable during this timeframe.  "At 1538 CT on 3/30/11, it was determined that the August 10, 2010 condition met the threshold for Emergency Action Level MU6, UNPLANNED loss of most or all safety system annunciation or indication in the control room for greater than 15 minutes. This notification is being made as an undeclared Unusual Event Emergency Plan Classification per 10 CFR 50.72(a)(1)(ii).  "Per NUREG 1022, a 1- hour notification is required when a condition existed which met the emergency plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of the discovery."  The licensee notified the NRC Resident Inspector.|
Power Reactor|46713|BROWNS FERRY|TENNESSEE VALLEY AUTHORITY|2|DECATUR|AL|LIMESTONE||Y|05000259|1|||[1] GE-4,[2] GE-4,[3] GE-4|RAY SWAFFORD|DONALD NORWOOD|3/30/2011 00:00:00|19:57|3/30/2011 00:00:00|15:42|CDT|3/30/2011 00:00:00|NON EMERGENCY|50.72(b)(2)(i)|PLANT S/D REQD BY TS|||||||MARVIN SYKES|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||TECHNICAL SPECIFICATION REQUIRED SHUTDOWN INITIATED  "On March 30, 2011 at 1443 CDT, during a refueling outage, Browns Ferry Unit 2 received an invalid Common Accident Signal (CAS) as a result of maintenance activities.  All four Unit 1/2 diesel [generators] auto started and all four Unit 3 diesel [generators] auto started.  Unit 2 received a full Reactor SCRAM and Core Spray pumps A, B, C, and D auto started and injected into the reactor.  Unit 2 Division I RHR [Residual Heat Removal] system was in Shutdown Cooling with only the C pump in service.  The A RHR pump auto started and Shutdown Cooling flow increased, as expected.  Unit 2 Division II RHR system was tagged out for maintenance.  HPCI [High Pressure Coolant Injection] and RCIC [Reactor Core Isolation Cooling] received auto initiation signals, however the steam isolation valves were tagged closed and the systems did not start.  MSIVs [Main Steam Isolation Valves] isolated as a result of the CAS signal.  "Unit 1 was at 100% power when the CAS on Unit 2 occurred.  This caused initiation of the Preferred Pump Logic which designates Division I CS [Core Spray] and LPCI [Low Pressure Coolant Injection] pumps on Unit 1 (1A and 1C) and the Division II pumps on Unit 2 (2B and 2D) and prevents Unit 1 Division II pumps from auto starting.  This resulted in the Division II RHR and CS pumps [being] inoperable for Unit 1.  Unit 1 subsequently entered Technical Specification 3.5.1, Condition H (two or more low pressure ECCS [Emergency Core Cooling System] injection/spray subsystems inoperable) which requires entering LCO 3.0.3 immediately.  "Unit 1 entered LCO 3.0.3 at 1443, which requires that actions shall be initiated within one hour to place the unit in Mode 2 within 10 hours; Mode 3 within 13 hours; and Mode 4 within 37 hours.  At 1542 Unit 1 began lowering power in order to comply with LCO 3.0.3.  "CAS logic was reset at 1812 and Unit 1 exited LCOs 3.5.1.H and 3.0.3.  "This condition requires a four hour report in accordance with 50.72(b)(2)(i) - 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.'  "The NRC Resident Inspector was notified.  Service Request #346544 was initiated in the Corrective Action Program."|
Power Reactor|46714|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|ERICK MATZKE|JOE O'HARA|3/31/2011 00:00:00|11:17|3/30/2011 00:00:00|14:05|CDT|3/31/2011 00:00:00|NON EMERGENCY|26.719|FITNESS FOR DUTY|||||||RYAN LANTZ|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||CONTRABAND FOUND INSIDE THE PROTECTED AREA  "While cleaning out portable toilets being used at the site, the waste disposal vendor employee discovered a two ounce (shot sized) liquor bottle in one of the units. The unit had been on-site since 2/25/11 and was staged in the Protected Area. The waste disposal vendor indicated that it is unlikely the bottle was in the portable toilet prior to delivery to the site."  The NRC Resident Inspector has been notified.|
Power Reactor|46715|CALLAWAY|AMEREN UE|4|FULTON|MO|CALLAWAY||N|05000483|1|||[1] W-4-LP|MICAH BENNINGFIELD|DONG HWA PARK|3/31/2011 00:00:00|20:50|3/31/2011 00:00:00|14:32|CDT|3/31/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||RYAN LANTZ|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||UNANALYZED CONDITION IDENTIFIED FOR INOPERABLE RWST LEVEL  "In response to a condition identified in late 2010 concerning the control and removal of hazard barriers in the plant, a review of the basis and analysis for high energy line breaks (HELBs) and the barriers for protecting against such events has been underway at Callaway in accordance with the plant's corrective action program. While following up on a question from the NRC Resident Inspector, and as a result of an additional question from the Nuclear Oversight organization at Callaway, it was identified that non-safety piping located in the valve room associated with the Refueling Water Storage Tank (RWST) could potentially [make] all four RWST low water level pressure transmitters inoperable in the event of a malfunction of the non-safety piping concurrent with a design-basis loss-of-coolant accident (LOCA) and/or following a seismic event. The RWST water level transmitters (which are located in the RWST valve room) perform a safety-related function for the emergency core cooling system (ECCS) by automatically swapping suction sources for the ECCS during a LOCA from the RWST to the containment sumps when a low water level condition is reached in the RWST. These instrument channels are required to be OPERABLE in Modes 1, 2, 3 and 4 per Callaway Technical Specification 3.3.2, 'Engineered Safety Feature Actuation System (ESFAS) Instrumentation.'   "The subject non-safety piping delivers steam supplied by the Auxiliary Steam system to (and from) heaters surrounding the RWST for maintaining RWST contents above the minimum required temperature during winter conditions. The piping passes through the RWST valve room containing the noted RWST water level transmitters which were designed only for a mild environment. It has been identified, however, that the non-safety Auxiliary Steam piping constitutes a high energy line and that its failure could create harsh (hot and wet) conditions in the valve room to which the RWST water level instrumentation was not designed.  "Per the Callaway FSAR, where non-safety piping interfaces with safety-related piping or systems, the design must be such that failure of the non-safety piping does not adversely affect the safety function(s) of the interfacing safety-related piping or system (since non-safety piping may be assumed to malfunction in conjunction with a design-basis accident). In this case, and based on a conservative interpretation of the FSAR, if the non-safety piping in the RWST valve room is assumed to malfunction (i.e., break), a failure of the RWST instrumentation could occur, thereby preventing the ECCS suction swap over from occurring as required or assumed for LOCA mitigation.  "This condition required declaring all four RWST water level channels inoperable. In light of recognizing that the RWST water level instruments could be subject to a harsh environment when they were only designed for a mild environment, and could thus fail as a result, this condition represents an unanalyzed condition that significantly degrades plant safety. With regard to the impact on the required ECCS suction swap over function that requires the RWST water level channels to be operable, the inoperability of all four instrument channels is a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to shutdown the reactor and maintain it in a safe condition, remove residual heat, control the release of radioactive material, and mitigate the consequences of an accident.  "Upon declaring the RWST water level instrument channels inoperable, TS Limiting Condition for Operation (LCO) 3.0.3 was entered at time 1432 CDT on 3/31/2011. At 1634 CDT, the Auxiliary Steam system was isolated and depressurized. This removed the energy that could be released from a break in the non-safety piping, thereby restoring Operability for the RWST water level instruments.  "The NRC Senior Resident Inspector was notified."|
Power Reactor|46716|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|AMY BURKHART|DONG HWA PARK|3/31/2011 00:00:00|23:32|3/31/2011 00:00:00|21:26|CDT|3/31/2011 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||RYAN LANTZ|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||IDENTIFIED UNSEALED FLOOD BARRIER PENETRATIONS  "During investigations of flood barrier penetrations, two approximately 4 inch conduits have been identified that are not sealed.  These conduits penetrate the south wall of the auxiliary building near the transformers into room 19.  Flooding through the penetrations could have impacted the ability of the station's auxiliary feedwater (AFW) pumps to perform their design accident mitigation functions.    "This eight-hour notification is being made pursuant to 10 CFR 50.72 (b)(3)(v).  "The penetrations are at an approximate elevation of 1007 [feet].  The river level has been less than 997 feet Mean Sea Level (MSL) since prior to December 1, 2010.  The AFW pumps are operable. There are not any indications of conditions that might result in a flood. Actions are in progress to plug the penetrations."  The NRC Resident Inspector has been notified.|
