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|
Non-Agreement State|47599|AVERA MCKENNAN HOSPITAL|AVERA MCKENNAN HOSPITAL|4|SIOUX FALLS|SD|MINNEHAHA|4016571-01|N||||||RICHARD MASSOTH|VINCE KLCO|1/17/2012 00:00:00|13:04|1/16/2012 00:00:00|16:00|MST|4/12/2012 00:00:00|NON EMERGENCY|35.3045(a)(1)|DOSE <> PRESCRIBED DOSAGE|||||||GREG PICK|R4DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MEDICAL EVENT DUE TO POTENTIAL DIFFERENT FRACTIONAL DOSE DELIVERED THAN PRESCRIBED  The licensee provided notification that a patient received 2 occurrences of a dose less than prescribed when delivering ten fractions of a treatment.  Each of the underdoses were approximately 50% of the 340 Gray prescribed fractional dose.  The patient will receive additional dose fractions in order to achieve the written directive total dose.  The Radiation Oncologist has notified the patient and attending physician.  * * UPDATE FROM RICHARD MASSOTH TO JOHN KNOKE AT 1826 EST ON 01/31/12 * *   "On January 17, 2012 the NRC Operations Center was verbally notified of two Therapeutic Underdose Occurrences discovered by the licensee on January 16 and 17, 2012. These occurrences involved a fractionated Breast High Dose Rate Afterloader (HDR) treatment with a SenoRx Contura multicatheter breast applicator. The first and third delivered treatment fractions were found to be less than 50% of the intended fractional dose. The entire course of the treatment in the written directive included ten equal-dose fractions of 3.4 Gray per fraction for a total dose of 34 Gray to the prescribed treatment site. To correct for the underdose occurrences, two additional treatment fractions were added and the treatment plan was modified to achieve the total dose specified in the written directive.  "The licensee now believes that this medical event has also caused an unintended dose to skin outside of the prescribed treatment site, requiring notification under 10CFR35.3045(a)(3). The licensee has performed computer simulation, calculations and physical measurements using TLDs simulating the treatment geometry to model the unintended skin dose. The event delivered an unintended skin dose exceeding at least the skin erythema threshold (2 Gy). The licensee is continuing to monitor the patient response to the skin dose and is working to refine the unintended skin dose estimates. An NRC reactive inspection team is on-site."  Notified R4DO (Jeff Clark) and FSME (Greg Suber)  * * * UPDATE FROM TRACI HOLLINGSHEAD TO HOWIE CROUCH AT 1044 EDT ON 4/12/12 * * *   The licensee confirmed that they agree with their medical consultants' findings that the patient received approximately 2720 rads of unintended skin dose.  Notified R4DO (Gaddy) 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.|
Part 21|47693|ROSEMOUNT NUCLEAR|ROSEMOUNT NUCLEAR|3|CHANHASSEN|MN|||Y||||||DUYEN PHAM|VINCE KLCO|2/23/2012 00:00:00|11:08|2/16/2012 00:00:00||CST|4/2/2012 00:00:00|NON EMERGENCY|21.21|UNSPECIFIED PARAGRAPH|||||||MEL GRAY|R1DO|MIKE ERNSTES|R2DO|PATTY PELKE|R3DO|RICK DEESE|R4DO|PART 21 GRP - EMAIL||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||PART 21 REPORT - ROSEMOUNT PRESSURE TRANSMITTERS WITH NONZERO BASED CALIBRATIONS  The following information was received by facsimile:  "During the course of qualification testing to replace certain diodes identified for obsolescence, RNII [Rosemount Nuclear Instruments, Inc.] has determined that Model 1154 Series H range code 4-8 pressure transmitters with a significantly elevated or suppressed 4 mA point may not meet the published steam pressure/temperature accuracy specification.  "The out of tolerance condition observed during steam pressure/temperature qualification testing is not related to the replacement diode changes. It is an inherent performance characteristic related to large zero elevation or suppression. The steam pressure/temperature accuracy specification will be revised to account for nonzero based calibrations with a significantly elevated or suppressed 4 mA point.  "This revised specification supersedes the published steam pressure/temperature accuracy specification for all Model 1154 Series H pressure transmitters affected by this notification.  "RNII recommends that users review the application where 1154 Series H range code 4-8 pressure transmitters are used to determine if there are safety considerations related to the revised steam pressure/temperature specification."  Rosemount Nuclear has provided instruments to the following list of domestic U.S. customers: Alabama Power; American Electric Power; Arizona Public Service/Pinnacle West; Bechtel Power; Constellation Energy; Dominion Nuclear Connecticut/Dominion Generation; Duke Energy; ECFS MCS; Edison Material Supply; Electro Mechanics; Entergy; Exelon Generation; Florida Power and Light; FPL Energy; Georgia Power; Northern States Power-Minnesota DBA XCEL Energy; Pacific Gas and Electric; Progress Energy Florida; Progress Energy Carolinas; PSEG Nuclear; South Carolina Electric and Gas; Southern California Edison; Southern Nuclear Operating Company; STP Nuclear Operating; Tennessee Valley Authority; TXU/Luminant; Westinghouse Electric.   * * * UPDATE FROM DUYEN PHAM TO PETE SNYDER ON 4/2/12 AT 1224 EDT * * *   The following information was received by facsimile:  "This revision only affects Section 4.0 of the 23 February 2012 notification letter. The pressure values listed in Section 4.0 at 8 hours and 56 hours have been corrected. No other changes have been made."   Notified R1DO(Caruso), R2DO(Lesser), R3DO(Dickson), R4DO(Haire) and Part 21 Group via email.|
Power Reactor|47696|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|DAVID SPARGO|CHARLES TEAL|2/24/2012 00:00:00|04:24|2/23/2012 00:00:00|22:00|CST|4/5/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||RICK DEESE|R4DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||LOSS OF 21 OUT OF 101 EMERGENCY SIRENS  "Communications has been lost to 21 sirens out of 101.  The loss of communications does not allow the activation of the sirens.  Almost all of Harrison County and Pottawattamie County in Iowa are without communications to the sirens.  There are compensatory measures in place to ensure notification by local law enforcement in case of an actual emergency to inform the public in these areas.    "This is being reported per 10CFR50.72(b)(3)(xiii) for 'Any event that results in a major loss of emergency assessment capability, off site response capability, or communications capability'.  "An attempt is being made to reboot the siren's communication system in order to restore the sirens.  During the reboot, sirens in Washington County, Nebraska, will also lose communications and therefore will not be functional.  Local Law Enforcement has been notified in Washington County to perform compensatory measures in case of an emergency."  The NRC Resident Inspector has been informed.   * * * UPDATE FROM DAVID SPARGO TO DONALD NORWOOD AT 0621 EST ON 2/24/2012 * * *  "As of 0518 CST, communications has been reestablished and all sirens were returned to service."  The licensee notified the NRC Resident inspector.  Notified R4DO (Deese).  * * * UPDATE FROM ERICK MATZKE TO HOWIE CROUCH AT 1602 EDT ON 04/05/12 * * *  "Following an investigation of the siren failure, it was determined that all sirens were lost for a period of time from approximately 1809 CST February 23, 2012 until 0518 CST February 24, 2012. The control room was notified at 0215 CST February 24, 2012 and compensatory measures were established for all affected counties in Iowa and Nebraska. No new compensatory measures or actions are or were required."  The licensee has notified the NRC Resident Inspector. Notified R4DO (Haire).|
Part 21|47716|ROSEMOUNT NUCLEAR|ROSEMOUNT NUCLEAR|3|CHANHASSEN|MN|||Y||||||DUYEN PHAM|JOHN KNOKE|3/2/2012 00:00:00|16:51|2/16/2012 00:00:00||CST|4/3/2012 00:00:00|NON EMERGENCY|21.21|UNSPECIFIED PARAGRAPH|||||||LAWRENCE DOERFLEIN|R1DO|MARK FRANKE|R2DO|MARK RING|R3DO|JACK WHITTEN|R4DO|PART 21 GROUP-EMAIL||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||PART 21 REPORT -  ROSEMONT PRESSURE TRANSMITTERS OUT OF TOLERANCE CONDITION  The following information was received by facsimile:  "During the course of qualification testing to replace certain diodes identified for obsolescence, an out of tolerance condition during steam pressure/temperature testing was observed on a test pressure transmitter in the qualification program. Through investigation into the out of tolerance condition, it was determined that the condition was not related to the replacement diode, but rather the result of the magnitude of a resistance change made to the transmitter temperature compensation circuitry during a final factory acceptance test. A brief description of the final acceptance test follows.  "During pressure transmitter final processing, a test is conducted on Model 1154 Series H pressure transmitters to assess the effects of ambient temperature changes on the 4mA to 20mA analog output. The temperature test is performed on each transmitter to verify the transmitter accuracy performance over normal operating temperatures from 40 F to 200 F. If the transmitter does not meet the acceptance criteria for the test, changes can be made to the compensating resistor values to optimize performance in the normal operating temperature range. If the resistance adjustment is too large, it has been determined that the accuracy of the pressure transmitter relative to the published steam pressure/temperature profile and accuracy specification may be exceeded during a steam pressure/temperature event.  "The magnitude of output shift during steam pressure/temperature conditions, that is attributable to the resistance changes made in final processing, is predictable based on the specific amount of resistance added or subtracted during the compensation process for each transmitter. Accordingly, RNII [Rosemount Nuclear Instruments, Inc] has reviewed the production records for the potentially affected transmitters and a revised steam pressure/temperature accuracy specification has been established for each shipped pressure transmitter. The revised steam pressure/temperature accuracy specification has been listed by serial number.  "This revised specification supersedes the published steam pressure/temperature accuracy specification for all Model 1154 Series H pressure transmitters affected by this notification. Also note that for certain calibrations with large zero elevation and suppressions as defined in RNII's Part 21 Notification, dated February 23, 2012, an alternate steam pressure/temperature accuracy specification is also provided for pressure transmitters that are affected by both notifications. All other 1154 Series H published specifications remain unchanged.  "On February 23, 2012, it was concluded that a substantial safety hazard may exist. RNII does not have sufficient information to determine the potential safety impact in plant applications. As a result, a notification about the potential substantial safety hazard is being made in accordance with 10 CFR Part 21 to customers who purchased affected 1154 Series H pressure transmitters.  "The manufacturing rework procedure 01153-3000 has been updated to limit the amount of acceptable resistance that may be added or subtracted during final testing. A revised steam pressure/temperature accuracy specification for each pressure transmitter impacted by this notification has been determined.  "RNII recommends that users review the application where 1154 Series H pressure transmitters are used to determine if there are safety considerations related to the revised steam pressure/temperature accuracy specification."  The United States nuclear sites associated with this Part 21 are: Farley, D.C. Cook, Palo Verde, ANO-1, Calvert Cliffs, Point Beach, Oconee, Catawba, Waterford 3, Braidwood, La Salle, Davis Besse, Crystal River 3, Vogtle, Millstone, Diablo Canyon, Salem, Hope Creek, San Onofre, South Texas, Watts Bar, Comanche Peak, Wolf Creek, St. Lucie, Dominion Energy - Unspecified, and Northeast Nuclear Energy - Unspecified.    * * * UPDATE FROM DUYEN PHAM TO PETE SNYDER ON 4/3/12 AT 1531 EDT * * *   The following information was received by facsimile:  This revision potentially affects the following commercial nuclear power plant sites in the United States: 1. Calvert Cliffs, 2. Millstone, 3. Hope Creek, 4. Salem, 5. Catawba, 6. Crystal River, 7. Farley,  8. Oconee, 9. St. Lucie, 10. Vogtle, 11. Watts bar, 12. Braidwood, 13. Davis Besse, 14. D.C. Cook, 15. LaSalle, 16. Point Beach, 17. Arkansas Nuclear One, 18. Comanche Peak, 19. Diablo Canyon, 20. Palo Verde, 21. San Onofre, 22. South Texas Project, 23. Waterford 3, and 24. Wolf Creek.  "Note: Notification #1 [EN #47693], dated 23 February 2012 (and revised on 2 April 2012) impacts ALL 1154 Series H pressure transmitters range codes 4 to 8 shipped since 1988. This does NOT impact the 1154SH9 since the 1154SH9 has a wider steam pressure/temperature specification. The revised specification is the same for all affected transmitters.  "Note: Notification #2 [this report - EN #47716], dated 2 March 2012 (and revised as of today [3 April 2012]) impacts only a subset of 1154 Series H pressure transmitters range codes 4 to 9 shipped from March 1994 to February 2012. This is a smaller affected population and does include the range 9 in the scope. The revised specification for this second issue is serial number dependent and only affects the transmitters listed in Appendix A [of the full part 21 report]."  Notified R1DO(Caruso), R2DO(McCoy), R3DO(Bloomer), R4DO(Haire), and Part 21 Group via email.|
Non-Agreement State|47722|INDIANA UNIVERSITY MEDICAL CENTER|INDIANA UNIVERSITY MEDICAL CENTER|3|INDIANAPOLIS|IN||13-02752-03|N||||||MACK RICHARD|JOHN KNOKE|3/7/2012 00:00:00|14:45|7/5/2011 00:00:00||EST|4/19/2012 00:00:00|NON EMERGENCY|35.3045(a)(1)|DOSE <> PRESCRIBED DOSAGE|||||||MARK RING|R3DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MEDICAL EVENT - DOSAGE DELIVERED TO PATIENT DIFFERED BY 20% FROM PRESCRIBED DOSAGE  On 7/5/11 a male patient receiving treatment for prostate cancer was supposed to receive 52  Iodine-125 seeds to the prostate.  The physician conducting the surgery injected 21 of the 52 seeds to an area outside the target volume.  Only 31 of the 52 seeds were injected into the targeted area.   The physician and patient will be advised of the disparity in dosage.  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 AT 1100 EST ON 4/19/12 FROM RICHARD TO SIMPSON * * *   The licensee is retracting this event report after a review and evaluation of the doses administered.  The licensee has determined that this event is not reportable based on the delivered dose not differing by 20% from the prescribed dosage and is therefore not reportable under 10 CFR 35.3045.  The licensee has discussed this conclusion with NRC Region 3 (Warren).  R3DO (Peterson) and FSME (McIntosh) have been notified.|
Agreement State|47771|PA BUREAU OF RADIATION PROTECTION|THOMAS JEFFERSON UNIVERSITY HOSPITAL|1|PHILADELPHIA|PA||PA-0130|Y||||||JOSEPH MELNIC|DONALD NORWOOD|3/26/2012 00:00:00|14:49|3/22/2012 00:00:00||EDT|3/26/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||CHRISTOPHER NEWPORT|R1DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - PATIENT RECEIVED LESS THAN INTENDED DOSE  The following information was received via facsimile:  "On March 23, 2012 the Southeast Regional Office [of the Pennsylvania Department of Environmental Protection - Bureau of Radiation Protection (Department)] received notice from the hospital of a medical event involving yittrium-90 TheraSpheres.  "The patient was being treated for disease of the liver and received 76% of the intended dose in the first delivery and 56% of the intended dose in the second delivery.  "The cause of the event is currently under investigation.  More information will be provided when received.  "No harm to the patient is expected.  The oncologist was satisfied with the treatment and no additional delivery is anticipated.  The Department will conduct a reactive inspection."  Pennsylvania Event Report ID No:  PA120010  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|47779|ILLINOIS EMERGENCY MGMT. AGENCY|WOOD RIVER REFINERY|3|ROXANA|IL||IL-01282-01|Y||||||DAREN PERRERO|HOWIE CROUCH|3/28/2012 00:00:00|10:57|3/26/2012 00:00:00||CDT|3/28/2012 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||ILLINOIS AGREEMENT STATE REPORT - SHUTTER MECHANISM FAILURE  The following information was obtained from the State of Illinois via email:  "The Radiation Safety Office for the licensee reported a second device at their facility had a shutter mechanism failure.  The shutter was found stuck in the open position on March 26, 2012 during continuation of their routine maintenance check and survey.  The gauge is near the top floor of a coker drum away from any routinely occupied work station. The device/area has been posted with appropriate warning notices. Measured radiation levels appear normal for this type of device with the shutter open.  There are two other gauges on this vessel which are operating normally.  Operators in the area have been informed of the situation and vessel access is prohibited.  The area is subject to high humidity and rust build up is the suspected cause of the shutter failure.  In the past, the manufacturer has recommended covers for these devices but none are available. The licensee's consultant and the vendor have been called.  Both will be on-site again on March 30, 2012 as previously arranged to service a similar other device that failed last week. They will investigate this recent failure at that time."  The gauge is an Ohmart Corp. model number SHLG-2, serial number 2605CN and contains a 5 Ci Cs-137 source.  The previous failure was reported under NRC EN# 47759.  Illinois Item Number: IL12006|
Agreement State|47790|FLORIDA BUREAU OF RADIATION CONTROL|LARKIN COMMUNITY HOSPITAL|1|SOUTH MIAMI|FL||2825-2|Y||||||CHARLES ADAMS|MARK ABRAMOVITZ|3/30/2012 00:00:00|08:07|3/8/2012 00:00:00||EDT|3/30/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||CHRISTOPHER NEWPORT|R1DO|CHRISTEPHER MCKENNEY|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - GAMMA KNIFE MEDICAL OVEREXPOSURE  The following information was received from the State of Florida via email:  "Licensee reported a possible medical event on 8 March which was confirmed in a written report received on 22 March.  Patient was being treated by Gamma Knife with 8 shots at four sites when the fifth shot was interrupted for a bathroom break.  While in the bathroom the patient fell and dislodged the stereotactic frame.  The frame was reapplied and treatment plan recalculated.  However, the computer did not start the treatment at the correct site resulting in an increase in dose of 22.2% at site 2, 14.8% at site 3 & 15, and 7% at site 4.  The doses are well within clinical practice.  A conservative dose prescription had been elevated to a moderate prescription.  The referring physician and patient have been notified.  No medical consequences are expected.  The cause is human error.  Any further investigation referred to Licensing and Materials.  No further action will be taken on this incident by this office."  Device: Leksell Perfexion Serial Number: 6130  Florida Incident: FL-12-034  "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.|
Non-Agreement State|47791|US ARMY|US ARMY|3|WARREN|MI||21-32838-01|N||||||THOMAS GIZICKI|CHARLES TEAL|3/30/2012 00:00:00|15:31|12/16/2009 00:00:00||EDT|4/5/2012 00:00:00|NON EMERGENCY|20.2201(a)(1)(i)|LOST/STOLEN LNM>1000X|||||||CHRISTOPHER NEWPORT|R1DO|ANGELA MCINTOSH|FSME|BILLY DICKSON|R3DO|||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||MISSING CHEMICAL AGENT DETECTOR  When returning from deployment a Virginia National Guard unit discovered that they had lost a M43A1 Chemical Detector containing 250 uCi of Am-241.  The unit searched the facility as well as a storage warehouse located in Richmond, VA for the device, but they were unable to locate it.  * * * RETRACTION FROM THOMAS GIZICKI TO VINCE KLCO ON 4/5/2012 AT 1201 EDT * * *   Retraction due to the detector found in the same location in a different Army unit.  Notified the R1DO (Caruso), R3DO (Bloomer) and the FSME (McIntosh).  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|
Agreement State|47792|ARIZONA RADIATION REGULATORY AGENCY|UNKNOWN|4|TEMPE|AZ|||Y||||||AUBREY GODWIN|CHARLES TEAL|3/30/2012 00:00:00|16:36|3/28/2012 00:00:00|13:35|MST|3/30/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||THOMAS FARNHOLTZ|R4DO|CHRISTEPHER MCKENNEY|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - A MOISTURE DENSITY GAUGE WAS FOUND IN BUILDING  The following information was received from the State of Arizona via email:  "At approximately 10:00 AM on March 28, 2012, the Agency [Arizona Radiation Regulatory Agency (ARRA)] was informed that radioactive material had been detected by the Tempe, AZ Fire Department in an old flour mill.  The Agency [ARRA] dispatched a response team to the site to locate and confirm information.  The response team determined that the isotope involved is Cesium-137 and based on the label information approximately 25 millicuries remain in the device.  The device is labeled on the accessible surfaces as a Kay Ray Model 7107 gauge and also as a Model 7519B gauge assembly SN 13454.  Based on the site history and the labels, it appears that the device was transferred to Bay State Milling in the 4th quarter of 1980 or the 1st quarter of 1981.  The Agency [ARRA] did not issue a specific license for this site.  We assume that the device was transferred as a GL (Generally Licensed) device and remained in use into 1998.  Only about 10% of the site has been surveyed thus far.  The site security is being maintained by the Tempe FD (Fire Department).  "The Agency [ARRA] is now searching historical records for information, including any other devices on site.  "The Agency [ARRA] continues its investigation."  Arizona Report #:  12-008|
Power Reactor|47794|COOK|INDIANA/MICHIGAN POWER CO.|3|BRIDGMAN|MI|BERRIEN||N|05000315|1|2||[1] W-4-LP,[2] W-4-LP|BUD HINCKLEY|STEVE SANDIN|4/1/2012 00:00:00|04:26|4/1/2012 00:00:00|04:26|EDT|4/1/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||BILLY DICKSON|R3DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0|Refueling|0|Refueling|N|N|0||0||UNAVAILABILITY OF TSC VENTILATION SYSTEM DUE TO SCHEDULED MAINTENANCE  "At 0445 EDT on Sunday, April 1, 2012, the Cook Nuclear Plant (CNP) Technical Support Center (TSC) air conditioning and charcoal filtration systems will be removed from service for scheduled maintenance.  "Under certain accident conditions the TSC may become unavailable due to the inability of the air conditioning and charcoal filtration systems to maintain a habitable atmosphere. Compensatory measures exist to relocate TSC personnel to the unaffected unit's control room if necessary.  "TSC ventilation system maintenance and post maintenance testing is scheduled to be completed by 1900 EDT on Sunday, April 1, 2012.  "The licensee has notified the NRC Resident Inspector.  "This notification is being made in accordance with 10 CFR 50.72 (b)(3)(xiii) due to the loss of an emergency response facility."   * * * UPDATE FROM DEAN BRUCK TO PETE SNYDER ON 4/1/12 AT 1633 EDT * * *   "The TSC ventilation system maintenance was completed satisfactorily and the system was restored to service at 1620 EDT on 4/1/12."  The licensee will notify the NRC Resident Inspector. Notified R3DO (Dickson).|
Power Reactor|47795|CATAWBA|DUKE ENERGY NUCLEAR LLC|2|YORK|SC|YORK||Y|05000413|1|2||[1] W-4-LP,[2] W-4-LP|THOMAS GARRISON|JOHN KNOKE|4/2/2012 00:00:00|11:45|4/2/2012 00:00:00|06:24|EDT|4/2/2012 00:00:00|NON EMERGENCY||INFORMATION ONLY|||||||MARK LESSER|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0|Refueling|0|Refueling|N|N|0||0||INFORMATION ONLY - FIRE IN PLANT VITAL BATTERY ROOM AREA  "At 0624 EDT on 04/02/2012 the control room was notified by a plant operator that there was a fire in the plant vital battery room area on Unit 2. The fire was located in the lower compartment of inverter 2KUI. This equipment supplies power to 120V AC non-safety related components. The fire brigade captain was dispatched along with the site fire brigade. When the fire brigade captain arrived the operator was staged near the inverter with a CO2 [carbon dioxide] fire extinguisher. The fire brigade captain instructed the NLO [non-licensed operator] to extinguish the fire with the extinguisher. The fire was immediately extinguished once the CO2 was applied from the extinguisher. At 0633 EDT the fire brigade captain reported to the control room that the fire was extinguished. Upon further investigation it appears that transformer in the lower compartment of the inverter had caught fire. The fire was confined to the lower compartment of the inverter and did not affect any other components in the area. Due to the loss of the inverter the loads automatically swapped to the alternate AC source as designed and no indications or equipment were lost due to the loss of the inverter.  "At the time of this event Unit 1 was at 100% power and Unit 2 was shutdown in Mode 6 with reactor coolant system drained to 8% level with 2B Residual Heat Removal pump supplying core cooling. There was no impact to either unit as a result of this event. All equipment required to maintain core cooling, AC and DC power, and reactor coolant inventory on Unit 2 was maintained during this event."  The licensee notified the NRC Resident Inspector.|
Agreement State|47796|ILLINOIS EMERGENCY MGMT. AGENCY|MIDWEST REGIONAL MEDICAL CENTER|3|ZION|IL||IL-01104-01|Y||||||DAREN PERRERO|PETE SNYDER|4/2/2012 00:00:00|17:38|3/30/2012 00:00:00||CDT|4/2/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||BILLY DICKSON|R3DO|CHRISTEPHER MCKENNEY|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - THERASPHERES FOUND IN DELIVERY EQUIPMENT AFTER PROCEDURE  The following information was received from the State of Illinois via email:  "On April 2, 2012 the licensee contacted the [Illinois Emergency Management] Agency [IEMA] to provide a preliminary notification that they believed a medical event had occurred as a result of a Therasphere treatment the previous workday.  On March 30, 2012 a patient had been administered Y-90 for the treatment of liver cancer according to their standard protocol.  No unusual anomalies took place during the treatment and no shunting of the dose was observed.  The treatment was completed as expected including initial agitation of the vial, 'tapping' of the system during administration and repeated flushing of the delivery system upon completion of the five minute process.   "However, when post-treatment measurements were conducted of the delivery system and vial, a measurable amount of activity was determined to still be present.  Analysis of the associated microcatheter, tubing, vial and the rest of the delivery system resulted in an assessment that only 92 of the prescribed 130 Gy [71%] of dose had been delivered.  Additional measurements and imaging identified that an aggregate of microspheres remained in the vial and at the hub of the microcatheter where it connects to the delivery system.  No other activity or contamination was noted in associated equipment.   "The patient was advised the next day of the substantial lowering of the dose delivered.  No immediate effect on the patient is expected at this time and no determination has been made as to any corrective action or additional dose to supplement the partial treatment.  A follow up with the patient is pending.  Similarly, the licensee is unaware of any corrective action necessary for future treatments pending additional evaluations and assessments.  Consideration is being given to having the delivery system returned to the manufacturer for an engineering analysis once the radioactivity present has decayed.  An investigation by IEMA is pending."  IL Item Number: IL12007  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|47797|TURKEY POINT|FLORIDA POWER & LIGHT CO.|2|MIAMI|FL|DADE||Y|05000250|3|||[3] W-3-LP,[4] W-3-LP|JOSE VASQUEZ|PETE SNYDER|4/2/2012 00:00:00|20:40|4/2/2012 00:00:00|19:33|EDT|4/2/2012 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||MARK LESSER|R2DO|TIM McGINTY|NRR|VICTOR MCCREE|R2|DAN DORMAN|NRR|WILLIAM GOTT|IRD|||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||UNUSUAL EVENT - FIRE FROM HYDROGEN SUPPLY TO VOLUME CONTROL TANK  "At 1933 EDT on 4/2/12, a fire was discovered on top of the Auxiliary Building Roof coming from a one inch hydrogen supply header to the U-3 Volume Control Tank.  "At 1948 EDT [Turkey Point] declared an Unusual Event due to a fire greater than 15 minutes in the protected area (HU2).  "At 1949 EDT [operators] closed the hydrogen isolation valve and hydrogen header pressure began to lower.  "At 1956 EDT a second isolation valve was closed. Header pressure continued to lower.  "At 2018 EDT the fire is out.  No damage to plant equipment or adverse effect to safe shutdown systems. [Turkey Point] is currently working on a de-escalation process."  The licensee notified the NRC Resident Inspector, State and local agencies.  Notified DHS SWO, FEMA, NICC and Nuclear SSA.   * * * UPDATE FROM JOSE VASQUEZ TO PETE SNYDER T 2139 EDT ON 4/2/12 * * *   After the fire was out the pipe became cool to the touch.  At 2110 EDT Turkey Point terminated the Unusual Event for Unit 3.  The licensee notified the NRC Resident Inspector, State and local agencies.  Notified DHS SWO, FEMA, NICC and Nuclear SSA via email.|
Power Reactor|47798|GRAND GULF|ENTERGY NUCLEAR|4|PORT GIBSON|MS|CLAIBORNE||Y|05000416|1|||[1] GE-6|FRANK WEAVER|PETE SNYDER|4/2/2012 00:00:00|22:56|4/2/2012 00:00:00|15:11|CDT|4/2/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||MARK HAIRE|R4DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||AUTOSTART OF DIVISION 3 DIESEL GENERATOR FOLLOWING 4160 VOLT LINE OUTAGE  "On 4/2/12 at 1511 [CDT], GGNS [Grand Gulf Nuclear Generating Station] received a valid ESF actuation for emergency AC power to Division 3 4160V bus due to degraded voltage.      "One of the two 500KV offsite feeders (Tech Spec Offsite Power Source) tripped causing a drop in grid voltage which resulted in a trip of the ESF feeder breaker for 4160 Volt Division 3 bus. The HPCS (High Pressure Core Spray) Diesel Generator automatically started and energized the bus. The HPCS system was not running and no ECCS initiation occurred during this event. The plant was in Mode 5 with RHR A in shutdown cooling. Divisions 1 and 2 ESF power monitoring instrumentation responded to the grid voltage transient but no actuation setpoints were reached. Division 1 and 2 ESF 4160V buses remained energized and shutdown cooling remained in service. The 500KV offsite feeder (Tech Spec Offsite Power Source) and additional 115 KV feeder (Tech Spec Offsite Power Source) remained in service. The 500KV feeder that tripped was restored by the dispatcher at approximately 1515 CDT. The Division 3 bus was subsequently transferred back to offsite power and the HPCS Diesel Generator was secured. This event is reportable per 10 CFR 50.72(b)(3)(iv).  "A lightning strike resulted in a voltage transient on the GGNS electrical distribution system. Due to this transient, the 'A'  Control Room Air Conditioning Unit (CRAC 'A') tripped and had to be manually restarted. CRAC 'A' was not running for approximately two minutes. During this timeframe CRAC 'B' was tagged out of service.  This was evaluated and determined to not be a loss of safety function."  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47799|QUAD CITIES|EXELON NUCLEAR CO.|3|CORDOVA|IL|ROCK ISLAND||Y|05000254|1|||[1] GE-3,[2] GE-3|BRIAN MAGNUSON|CHARLES TEAL|4/2/2012 00:00:00|23:32|4/2/2012 00:00:00|14:52|CDT|4/2/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(i)|PLANT S/D REQD BY TS|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||BILLY DICKSON|R3DO|||||||||||||||||||N|Y|100|Power Operation|90|Power Operation|N|N|0||0||N|N|0||0||INITIATION OF REACTOR SHUTDOWN REQUIRED BY TECHNICAL SPECIFICATIONS (TS) 3.8.1.F  "On April 2, 2012, at 1452 hours, Unit 1 received Panel 901-8 A7 U1 Emergency Diesel Generator (EDG) Trouble alarm.  Equipment Operators were dispatched and the Unit 1 EDG was found running unloaded, without a generator field flash, and no auto start signal received.  Troubleshooting identified that a 125 VDC ground had caused the Unit 1 EDG to start.  As a result, the Unit 1 EDG was declared inoperable.  "At this time Unit 2 is in a refueling outage and the Unit 2 EDG is currently inoperable for repairs.  Due to the inoperability of the Unit 1 and Unit 2 EDGs, at 2000 hours a Reactor Shutdown was initiated on Unit 1 in accordance with TS 3.8.1.F.  "In addition, since the EDGs supply emergency power to both Unit's Standby Gas Treatment Systems (SBGTS), emergency power was unavailable to SBGTS; however, normal power supplies remained available.  "This notification is being made in accordance with 10 CFR 50.72(b)(2)(i), and in accordance with 10 CFR 50.72(b)(3)(v)(D).  "At 2151 hours, the Unit 1 EDG was declared operable following repairs and successful operability testing."  The NRC Resident Inspector has been informed.|
Power Reactor|47800|FERMI|DETROIT EDISON CO.|3|NEWPORT|MI|MONROE||N|05000341|2|||[2] GE-4|JEFF GROFF|JOHN KNOKE|4/3/2012 00:00:00|05:57|4/3/2012 00:00:00|05:24|EDT|4/6/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||BILLY DICKSON|R3DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||SPDS AND ERDS REMOVED FROM SERVICE FOR PLANNED MAINTENANCE  "On 4/03/12 at 0524 EDT, the Safety Parameter Display System (SPDS) and Emergency Response Data System (ERDS) were removed from service to support activities for a planned maintenance outage on the UPS vital bus power supply. The duration of work is expected to be approximately 80 hours. During this time, the majority of the control room indications remain available to the plant staff, and will be used for emergency response, if needed. Information will be communicated to the NRC using other available communication systems, if needed. The plant is currently in Mode 5, and will remain in Mode 5, for the duration of the SPDS and ERDS unavailability. Since the unavailability will last greater than 8 hours, this is considered a Loss of Emergency Assessment Capability, and reportable under 10 CFR 50.72(b)(3)(xiii).  "Follow up notification will be made when SPDS and ERDS have been restored."  The licensee has notified the NRC Resident Inspector.  * * * UPDATE FROM GEORGE PICCARD TO VINCE KLCO ON 4/6/2012 AT 1227 EDT * * *   "At 0702 EDT on April 6, 2012, Safety Parameter Display System (SPDS) and Emergency Response Data System (ERDS) were restored, following restoration of the associated power bus."  The licensee notified the NRC Resident Inspector.  Notified the R3DO (Bloomer)|
Power Reactor|47801|COOK|INDIANA/MICHIGAN POWER CO.|3|BRIDGMAN|MI|BERRIEN||N|||2||[1] W-4-LP,[2] W-4-LP|BRAD LEWIS|JOHN KNOKE|4/3/2012 00:00:00|07:36|4/3/2012 00:00:00|08:00|EDT|4/3/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||BILLY DICKSON|R3DO|||||||||||||||||||N|N|0||0||N|N|0|Refueling|0|Refueling|N|N|0||0||UNIT 2 PLANT PROCESS COMPUTER OUT-OF-SERVICE FOR PLANNED MAINTENANCE  "The Unit 2 DC Cook Nuclear Plant (CNP) Plant Process Computer (PPC) will be removed from service on Tuesday, April 3, 2012 at 0800 EDT to support scheduled maintenance. This will cause the entire Unit 2 PPC, including the ERDS [Emergency Response Data System] to be unavailable to the NRC Operations Center. The planned maintenance also affects the Safety Parameter Display System (SPDS), the Real Time Data Repository (RDR), and PPC data to Emergency Response Facilities at CNP.   "The scheduled maintenance, returning of equipment to service, and post maintenance testing is expected to be completed by 0000 EDT on Wednesday, April 4, 2012.    "Compensatory measures exist within the DC Cook Emergency Response procedures to provide plant data via the Emergency Notification System to the NRC Operations Center until the ERDS can be returned to service.  "The licensee has notified the NRC Senior Resident Inspector.   "This notification is being made in accordance with 10 CFR 50.72 (b)(3)(xiii) due to any event that results in a major loss of emergency assessment capability, offsite response capability, offsite communications capability (e.g. significant portion of control room indication, Emergency Notification System, or offsite notification system)."  * * * UPDATE FROM BRAD LEWIS TO CHARLES TEAL ON 4/3/12 AT 1536 EDT * * *  At time 1500 EDT, the Plant Process Computer and ERDS system has been restored to service.     The licensee has notified the NRC Resident Inspector.  Notified R3DO (Dickson).|
Fuel Cycle Facility|47802|HONEYWELL INTERNATIONAL, INC.|HONEYWELL INTERNATIONAL, INC.|2|METROPOLIS|IL|MASSAC|SUB-526|Y|04003392||||URANIUM HEXAFLUORIDE PRODUCTION|LIDIA LIBINSKI|JOHN KNOKE|4/3/2012 00:00:00|11:29|4/2/2012 00:00:00|15:20|CDT|4/3/2012 00:00:00|NON EMERGENCY|40.60(b)(3)|MED TREAT INVOLVING CONTAM|||||||GERALD MCCOY|R2DO|ERIC BENNER|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||CONTAMINATED INDIVIDUAL TREATED IN THE ON-SITE DISPENSARY FOR INJURED FINGER  "On 4/2/12 approximately at 1520 CDT a person working in the Feed Materials Building injured his finger between a drum lid and drum ring. The employee received first aid treatment in the on-site dispensary. The individual was found to have contamination on his boots. The employee was not transported to an off-site facility. The employee was released to go home at the end of his shift and appropriately exit monitored. There was no detectable contamination on the employee when he left the site."  The contamination was uranium ore concentrates.  The licensee informed NRC Region II.|
Power Reactor|47803|INDIAN POINT|ENTERGY NUCLEAR|1|BUCHANAN|NY|WESTCHESTER||Y|05000247|2|||[2] W-4-LP,[3] W-4-LP|BRIAN VANGOR|STEVE SANDIN|4/4/2012 00:00:00|05:15|4/4/2012 00:00:00|05:00|EDT|4/4/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(i)|PLANT S/D REQD BY TS|||||||JOHN CARUSO|R1DO|||||||||||||||||||N|Y|48|Power Operation|43|Power Operation|N|N|0||0||N|N|0||0||TECHNICAL SPECIFICATION REQUIRED SHUTDOWN DUE TO INOPERABLE STATIC INVERTER  "Indian Point 2 commenced a Technical Specification required shutdown in accordance with Limiting Condition of Operation 3.8.7, Condition B at 0500 hours on 04/04/2012 due to the inoperability of '24' Static Inverter - a component in the 118VAC Distribution System. Indian Point 2 is required to be in Mode 3 by 0802 hours. Repair of the '24' Static Inverter is in progress.   "The NRC Resident Inspector has been notified."    The licensee also notified the NY State Public Service Commission.   * * * UPDATE FROM P. SCHOAN TO DONALD NORWOOD AT 1053 EDT ON 4/4/2012 * * *  At 0538 hours, the Technical Specification shutdown was stopped following the repair and return to service of the '24' Static Inverter.  The licensee notified the NRC Resident Inspector.  Notified R1DO (Caruso).|
Power Reactor|47805|CATAWBA|DUKE ENERGY NUCLEAR LLC|2|YORK|SC|YORK||Y|05000413|1|2||[1] W-4-LP,[2] W-4-LP|DON WISNIEWSKI|PETE SNYDER|4/4/2012 00:00:00|20:25|4/4/2012 00:00:00|20:03|EDT|4/5/2012 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||GERALD MCCOY|R2DO|ROBERT NELSON|NRR|VICTOR MCCREE|R2|DAN DORMAN|NRR|WILLIAM GOTT|IRD|||||||||||A/R|Y|100|Power Operation|0|Hot Standby|N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||NOTICE OF UNUSUAL EVENT DUE TO DUAL UNIT LOSS OF OFFSITE POWER  At 2003 EDT on 4/4/12 Catawba Unit 1 automatically tripped due to a loss of offsite power. The licensee declared a Notice of Unusual Event due to entry into EAL 4.5.U.1, 'AC Electrical Power from All Offsite Sources Has Been Lost for Greater Than 15 Minutes.'  Power is still available from onsite sources. All four Emergency Diesel Generators [EDG] started and powered their respective safety buses.   "The EAL poses no threat to the safety of the public."  The Unit 1 control rods fully inserted, there is no known primary to secondary leakage, and the auxiliary feedwater water system automatically started.  The licensee notified the NRC Resident Inspector  Notified: DHS SWO, FEMA, DHS NICC, USDA, HHS, DOE, EPA, and Nuclear SSA via email.   * * * UPDATE AT 0150 EDT ON 04/05/12 FROM DARRYL HELTON TO S. SANDIN * * *  Catawba Nuclear Generating Station exited the Unusual Event at 0137 EDT after restoring offsite power at 0135 EDT to the Unit 1A and Unit 2B essential busses.  The EDGs powering these busses have been secured and returned to Standby.  The Unit 1B and Unit 2A essential busses remain energized by their respective EDGs.  The licensee does not have an estimate as to when offsite power will be restored for these busses.  Additional information provided as follow-up includes:  4-hour notification per 10CFR50.72(b)(2)(iv)(B) due to the Unit 1 Reactor Protection System [RPS] actuation (Rx Trip),   4-hour notification per 10CFR50.72(b)(2)(xi) for offsite notifications to the States of North and South Carolina and the counties of York, Gaston and Mecklenburg,  8-hour notification per 10CFR50.72(b)(3)iv)(A) due to the Unit 1 Rx Trip, Unit 1 Auxiliary Feedwater automatic start, Unit 1A and 1B and Unit 2A and 2B EDG auto starts.  8-hour notification per 10CFR50.72(b)(3)(v)(B) due to the loss of Residual Heat Removal [RHR] cooling on the initial loss of power.  Unit 2 RHR was restored in less than three (3) minutes when the essential busses were re-energized by the EDGs.  The licensee informed the NRC Resident Inspector.    Notified R2IRC, EO (Nelson), IRD (Gott) and FEDS (DHS SWO, DOE Ops Center, FEMA, HHS Ops Center, DHS NICC, USDA Ops Center, EPA EOC) and NuclearSSA via email.|
Power Reactor|47806|QUAD CITIES|EXELON NUCLEAR CO.|3|CORDOVA|IL|ROCK ISLAND||Y|||2||[1] GE-3,[2] GE-3|BRIAN MAGNUSON|VINCE KLCO|4/4/2012 00:00:00|21:32|4/4/2012 00:00:00|17:16|CDT|4/4/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(A)|DEGRADED CONDITION|||||||TAMARA BLOOMER|R3DO|||||||||||||||||||N|N|0||0||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||DEGRADED CONDITION DUE TO IDENTIFIED REACTOR PRESSURE VESSEL TEST LEAKAGE  "On April 4, 2012, at 1716 [CDT], with Unit 2 shutdown for refueling, leakage was identified from a 2-inch vessel nozzle during a Reactor Pressure Vessel (RPV) pressure test. The leakage amount was approximately one drop per second. The penetration (N-11B) is a reference leg used for reactor vessel instrumentation. The leakage originates from the area where the nozzle penetrates the vessel wall. The nozzle is welded on the inside of the vessel, so the actual attachment weld could not be examined at the time of this report. The RPV pressure test has been stopped and the reactor vessel depressurized. The cause and resolution are under evaluation. The condition is being reported under 50.72(b)(3)(ii)(A) given the defect was associated with the primary coolant system pressure boundary."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47807|SUSQUEHANNA|PPL SUSQUEHANNA LLC|1|ALLENTOWN|PA|LUZERNE||N|05000387|1|2||[1] GE-4,[2] GE-4|RON FRY|PETE SNYDER|4/4/2012 00:00:00|22:59|4/4/2012 00:00:00|18:35|EDT|4/4/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||JOHN CARUSO|R1DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||INOPERABLE ACCIDENT MITIGATION EQUIPMENT   "On 4/4/2012 at 1517 EDT, the 'A' Emergency Diesel Generator (EDG) was declared inoperable for performance of a maintenance surveillance. At 1835 EDT on 4/4/2012 the 'B' Control Structure Chiller was declared inoperable due to an unrelated problem. [With] the 'B' Control Structure Chiller inoperable coincident with the 'A' EDG inoperable, the 'A' Control Structure Chiller would not be available to perform its design function on a loss of offsite power. This is a condition that, at the time of discovery, could have prevented fulfillment of a Safety Function and is reportable under 50.73(a)(2)(v) as an 8 hour notification. Note that Technical Specifications allows four hours to correct the condition before further actions are required, i.e. declare the features ('A' Control Structure Chiller) supported by the inoperable diesel inoperable. The 'A' EDG was restored to operable at 2200 which restored safety function capability for the 'A' Control Structure Chiller."   The licensee notified the NRC Resident Inspector.|
General Information|47808|BUREAU OF INDIAN AFFAIRS|UNKNOWN|3|ASHLAND|WI|| 48-32735-01|Y||||||JEFF BRADLEY|STEVE SANDIN|4/5/2012 00:00:00|10:45|4/5/2012 00:00:00|09:00|CDT|4/5/2012 00:00:00|NON EMERGENCY|20.2201(a)(1)(i)|LOST/STOLEN LNM>1000X|||||||TAMARA BLOOMER|R3DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||TROXLER MOISTURE DENSITY GAUGE FOUND ON INDIAN RESERVATION  The Bureau of Indians Affairs Great Lakes Agency (License #48-32735-01) reported that a Troxler Moisture Density Gauge has been found on the Bad River Indian Reservation.  The gauge was delivered to the Radiation Safety Officer for the Bureau of Indian Affairs.  The gauge is currently stored in a safe location and the source rod is in the locked position. The licensee and owner of the gauge is not known at this time.  The Troxler Gauge is a model 3411B, S/N 8975, and contains an Am-241 40 mCi and a Cs-137 7.5 mCi source.  The RSO contacted the Troxler Corporation for more information.  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|
Power Reactor|47809|FARLEY|SOUTHERN NUCLEAR OPERATING COMPANY|2|ASHFORD|AL|HOUSTON||Y|05000348|1|||[1] W-3-LP,[2] W-3-LP|ALTON DEWEESE|VINCE KLCO|4/5/2012 00:00:00|18:21|4/5/2012 00:00:00|12:20|CDT|4/5/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||GERALD MCCOY|R2DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||UNPLANNED SYSTEMS ACTUATIONS WHILE PERFORMING A SURVEILLANCE TEST  "Unit 1 was performing a planned refueling outage surveillance test, FNP-1-STP-40.0, 'Safety Injection with Loss of Off-Site Power [LOSP].' The systems were being returned to normal following the actuation portion of the test. When the B1F Sequencer Test Trip Override Switch was taken to the 'ON' position, the 1-2A Diesel Generator output breaker opened, which caused a loss of power to the 'A' Train 4 kV busses. Prior to the event, the 1-2A Diesel Generator was running at normal speed and voltage carrying the 'A' Train 4kV busses. When the diesel generator output breaker opened, it then reclosed upon receipt of the LOSP signal causing the LOSP sequencer loads to automatically start. This included the 1C Component Cooling Water Pump, the 1A High Head Safety Injection Pump (discharge isolation was closed prior to the event), and the 1A and 1B SW pumps.  Therefore, during the test, the system actuated in a way that was not part of the planned surveillance testing.  The 1A RHR pump was in shutdown cooling mode at the time of the event and was load shed. RHR was restarted manually by the operating crew approximately 1 minute later (no auto start [signal] present due to a loss of site power - LOSP signal without a safety injection signal present).  "The investigation revealed that a step in the procedure sequence was not performed during the restoration portion of the test. The operator did not parallel the diesel with off-site power prior to operating the B1F Sequencer Test Trip Override Switch which opened the diesel output breaker without off-site power aligned to the 'A' Train 4kV busses. The 1-2A Diesel Generator was subsequently paralleled to the grid and properly shutdown per the test procedure restoration."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47810|OCONEE|DUKE ENERGY NUCLEAR LLC|2|SENECA|SC|OCONEE||Y|||2||[1] B&W-L-LP,[2] B&W-L-LP,[3] B&W-L-LP|STEPHEN NEWMAN|PETE SNYDER|4/6/2012 00:00:00|03:57|4/5/2012 00:00:00|22:38|EDT|4/13/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|||||||GERALD MCCOY|R2DO|||||||||||||||||||N|N|||0||N|Y|100|Power Operation|85|Power Operation|N|N|0||0||STANDBY SHUTDOWN FACILITY (SSF) NOT ANALYZED FOR ALL OPERATING CONDITIONS  "On 3/29/2012 Duke Energy identified that unanalyzed conditions exist for SSF mitigated events since associated thermal and hydraulic analyses do not consider all initial operating conditions, especially lower operating modes and lower decay heat.  Specifically there are four (4) conditions where the SSF is not currently analyzed:  "1. SSF operating at less than 525 degrees F and less than normal operating pressure (approximately 2155 psig), 2. SSF operation before four (4) Effective Full Power Days (EFPDs), 3. SSF reactor coolant make up at low Reactor Coolant System (RCS) pressure. 4. A reactor trip from less than 85 percent power and less than 579 degrees F.  "On 4/4/2012, an immediate determination of operability concluded that for the first three (3) conditions the SSF was operable but degraded/nonconforming (OBDN).  For the 4th issue, there was reasonable assurance that 1% delta k/k shutdown margin would be maintained if T average. remained above 500 degrees F. Based on a lack of analysis and an increased likelihood of reducing T average. below 525 degrees F during a 72 hour event, the SSF was declared OBDN with a separate operability determination required to validate the Unit 3 power coastdown and end of life T average. reduction analysis. Until additional analysis is performed, the SSF is inoperable on any unit where the power level is reduced below 85 percent.  "A second operability determination for Unit 3 concluded that the SSF will maintain greater than or equal to 525 degrees F with an initial power level of 70 percent and a 570 degree F T average. The SSF will be declared inoperable on Unit 3 if power is reduced to less than 70 percent. Seventy percent was chosen as a conservative value to ensure the unit stayed inside the bounds of existing analyses. Unit 3 is currently at approximately 85 percent power and reducing power at approximately 1 percent per day in preparation for the Unit 2 end of core 26 refueling outage. For Unit 3, the SSF is OBDN based on preliminary calculation results.  "On 4/5/2012, due to a worsening component cooling water system leak on Unit 2, it was necessary to bring the unit down to Mode 3 to implement repairs. Upon down power, when Unit 2 transitioned below 85 percent power, the ability of the SSF to perform its design function, in consideration of the information above, could not be confirmed and the SSF was declared inoperable for Unit 2.  "Currently, there is no conclusive information that would support SSF operability while Unit 2 is below 85 percent power. As such, this event is being conservatively reported under 50.72(b)(3)(ii)(B), 'The nuclear plant being in an unanalyzed condition that significantly degrades plant safety.' Due to their current power levels, this condition does not affect Units 1 and 3.  "Initial Safety Significance: Until confirmed by analysis, the lack of decay heat may result in an initial over cooling of the RCS and potentially an interruption of natural circulation or inadequate shutdown margin. Consequently, the SSF was declared inoperable.  "Corrective Action(s): Additional analyses are being completed to reestablish SSF operability to bound the unanalyzed entry conditions.  "The NRC Resident Inspector has been informed."  * * * UPDATE FROM DEAN PORTER TO JOHN KNOKE AT 1403 EDT ON 04/13/12 * * *  "Update to ENS Notification Number 47810: ENS Notification number 47810 identified an unanalyzed condition for Oconee Unit 2. This update includes Oconee Units 1, 2 and 3 because the condition of the SSF affects all three units during certain plant conditions.   "The Standby Shutdown Facility (SSF) serves as a backup for existing safety systems. Currently the events that rely upon the SSF for mitigation are unanalyzed on Units 1 and 2 when the power level is reduced below 85 percent and on Unit 3 when the power level is reduced below 70 percent, or for any unit operating with less than 4 Effective Full Power Days (EFPDs) at 100% full power since its most recent shutdown. Since the SSF events are unanalyzed for these conditions and until further analysis and evaluation can be completed, Duke Energy is conservatively calling the SSF inoperable when in these conditions.  "For these conditions, this event is being reported for Oconee Units I, 2, and 3 under 50. 72(b )(3)(ii)(B), that is, the nuclear plant being in an unanalyzed condition that significantly degrades plant safety. Based on current analyses, the SSF will be declared inoperable, and action statements entered on Units 1 or 2 if power is reduced below 85% power and on Unit 3 if power is reduced below 70% power, or for any unit operating with less than 4 EFPDs since its most recent shutdown.  "Initial Safety Significance: Until confirmed by analysis or evaluation, the lack of decay heat may result in an initial over cooling of the RCS and potentially fail the various acceptance criteria of the events required to be augmented by the SSF. Consequently, the SSF will be declared inoperable for the conditions stated above. Unit 3 is being shut down for a routine refueling outage. Units 1 and 2 continue to operate at 100% power with no problems.  "Corrective Action(s): Additional evaluations are being completed to establish whether the existing analyses are applicable to the conditions outside of which they were performed and, if not, there is a reasonable assurance that successful mitigation can be accomplished with the existing procedure. If not, further licensing action may be required. These evaluations were initiated on April 4, 2012, and are ongoing."  The licensee has notified the NRC Resident Inspector.  Notified the R2DO (Kathleen O'Donohue).|
Power Reactor|47811|HATCH|SOUTHERN NUCLEAR OPERATING COMPANY|2|BAXLEY|GA|APPLING||Y|05000321|1|2||[1] GE-4,[2] GE-4|GUY GRIFFIS|DONALD NORWOOD|4/6/2012 00:00:00|08:41|4/6/2012 00:00:00|04:43|EDT|4/6/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||GERALD MCCOY|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||LOSS OF TONE ALERT SYSTEM  "On Friday April 6, 2012, at 0443 EDT, the National Weather Service (NWS) Jacksonville Florida office notified Hatch Nuclear Plant (HNP) of a loss of the NWS Tone Alert Weather Radio System.  Site Emergency Procedures define a loss of Tone Alert System for greater than 15 minutes as a significant loss of Emergency Communications.  "The NWS Tone Alert Weather Radios are utilized as the Prompt Notification System (PNS) for HNP.  This loss impacts the ability to notify the Emergency Planning Zone (EPZ) population for the HNP.  This failure meets NRC 8 hour reporting criteria 10CFR50.72(b)(3)(xiii).  "The State of Georgia Department of Homeland Security / Georgia Emergency Management Agency (GEMA) 24-hour warning point and the '911' dispatch centers of Appling, Jeff Davis, Tattnall and Toombs counties were notified so compensatory measures would be available should the Prompt Notification System be needed.  This consists of utilizing route alerting and the 'Code Red' system, which is a reverse 911 feature available from the county 911 center.  This notification meets the 4-hour notification reporting criteria 10CFR50.72(b)(2)(xi).  "On Friday April 6, 2012, at 0503 EDT, the NWS Jacksonville Florida office notified HNP that the NWS Tone Alert Weather Radio System had been returned to service."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47812|SUSQUEHANNA|PPL SUSQUEHANNA LLC|1|ALLENTOWN|PA|LUZERNE||N|05000387|1|||[1] GE-4,[2] GE-4|RON FRY|VINCE KLCO|4/6/2012 00:00:00|20:33|4/6/2012 00:00:00|17:57|EDT|4/6/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(A)|DEGRADED CONDITION|||||||JOHN CARUSO|R1DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||LOCAL LEAK RATE TEST RESULTS EXCEED PATHWAY TECHNICAL SPECIFICATION LIMITS  "Appendix J local leak rate testing has determined that secondary containment bypass leakage (SCBL) has been exceeded for Unit 1. During performance of leak rate test SE-159-026 for X-9A penetration it was determined the combined SCBL limit of 15 scfh [standard cubic feet per hour] for as-found minimum pathway was exceeded as specified in Tech Spec SR 3.6.1.3.11.  "Acceptance criteria test results were within acceptance criteria for the 10CFR50 Appendix J limits of 0.6 La.  "This event is being reported as a degraded or unanalyzed condition pursuant to 10CFR50.72(b)(3)(ii)."  Licensee corrective actions are to repair the identified valve seats.  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47813|FARLEY|SOUTHERN NUCLEAR OPERATING COMPANY|2|ASHFORD|AL|HOUSTON||Y|05000348|1|||[1] W-3-LP,[2] W-3-LP|DARRIN GARD|VINCE KLCO|4/6/2012 00:00:00|22:18|4/6/2012 00:00:00|14:44|CDT|4/6/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||GERALD MCCOY|R2DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||VALID LOAD SHED SIGNAL OCCURRED DUE TO LOSS OF THE 1B STARTUP TRANSFORMER  "At 1444 [CDT] on April 6, 2012, during a planned refueling outage on Unit 1, maintenance activities in the high voltage switchyard caused feeder breaker 820 to inadvertently trip. With the second feeder breaker, 924, already out of service, power was lost to the 1B startup transformer. An undervoltage condition was then experienced on the 1G 4160 V emergency bus. As a result, the B1G Sequencer initiated a valid load shed of the 1G 4160 V emergency bus. Due to outage conditions, the B-Train, 1B Emergency Diesel Generator (EDG) was tagged out and did not automatically start but did receive a valid start signal. None of the ESF loads supplied by the 1G bus started automatically since the 1B EDG was out of service. With a B-Train equipment outage in progress, the 1A RHR pump (A-Train) remained in service for shutdown cooling throughout the event. Although the bus safety function was not needed for plant conditions a valid load shed signal occurred and therefore this event is considered reportable.  "The 1G 4160 V emergency bus was restored to service at 1542 on April 6, 2012. Investigation revealed a technical inaccuracy in the instructions used during the maintenance activity in the high voltage switchyard that caused feeder breaker 820 to trip."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47814|SUSQUEHANNA|PPL SUSQUEHANNA LLC|1|ALLENTOWN|PA|LUZERNE||N|05000387|1|2||[1] GE-4,[2] GE-4|DAVE WALSH|VINCE KLCO|4/7/2012 00:00:00|15:27|4/7/2012 00:00:00|13:54|EDT|4/18/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xii)|OFFSITE MEDICAL|||||||JOHN CARUSO|R1DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||TRANSPORT OF POTENTIALLY CONTAMINATED WORKER  "On 04/07/2012 at 1354 [EDT], Susquehanna Steam Electric Station requested an offsite ambulance via the 911 system for medical assistance. The individual was in the radiologically controlled area and was treated as contaminated. An offsite ambulance arrived on site at 1413 hrs. and the ambulance departed the site at 1424 hrs. enroute to the Berwick Hospital.  "This is considered a transport of a contaminated individual requiring an 8 hour ENS Notification per 10CFR50.72(b)(3)(xii)."  Licensee health physic technicians accompanied the individual to the hospital.  The licensee notified the NRC Resident Inspector and the Pennsylvania Emergency Management Agency.  * * * RETRACTION FROM DARVIN DUTTRY TO JOHN SHOEMAKER ON 04/18/2012 AT 1522 EDT * * *  "On 04/07/2012, PPL Susquehanna reported that a potentially contaminated individual was transported offsite via ambulance for medical assistance.  The individual had been in the radiologically controlled area when the event occurred, and for medical reasons could not be completely surveyed for radioactive contamination prior to transport to the hospital.  Therefore the event was considered transport of a contaminated individual.  "Health Physics personnel accompanied the individual to the hospital and conducted surveys of the individual, ambulance and hospital equipment and facilities.  The results of these surveys indicated that no contamination was detected and the individual, ambulance and all hospital facilities and equipment were non-contaminated.  "Based on the above information, reporting pursuant to 10CFR50.72(b)(3)(xii) described in the referenced Event Notification is retracted."  The licensee has notified the NRC Resident Inspector.  Notified R1DO (Joustra).|
Power Reactor|47815|PALISADES|NUCLEAR MANAGEMENT COMPANY|3|COVERT|MI|VANBUREN||N|05000255|1|||[1] CE|GREG SMITH|MARK ABRAMOVITZ|4/8/2012 00:00:00|05:01|4/8/2012 00:00:00|05:01|EDT|4/8/2012 00:00:00|NON EMERGENCY||OTHER UNSPEC REQMNT|||||||TAMARA BLOOMER|R3DO|||||||||||||||||||N|Y|60|Power Operation|0|Cold Shutdown|N|N|0||0||N|N|0||0||OFFSITE NOTIFICATION OF NOISE ASSOCIATED WITH PLANT SHUTDOWN  "Notified Van Buren County Sheriff of atmospheric steam dump valve usage for cooldown of Palisades Nuclear Plant for start of refueling outage."  The licensee will notify the NRC Resident Inspector.|
Power Reactor|47816|SUSQUEHANNA|PPL SUSQUEHANNA LLC|1|ALLENTOWN|PA|LUZERNE||N|05000387|1|2||[1] GE-4,[2] GE-4|ALEX McLELLAN|MARK ABRAMOVITZ|4/9/2012 00:00:00|02:07|4/9/2012 00:00:00|01:02|EDT|4/9/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||JOHN CARUSO|R1DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||BOTH CONTROL STRUCTURE CHILLERS DECLARED INOPERABLE WHILE SWITCHING POWER SUPPLIES  "On 4/9/2012, starting at 0102 EDT, the 'A' and 'C' Emergency Diesel Generators (EDG) were sequentially and briefly declared inoperable to switch their DC control power back to their normal supplies.  Switching power to the normal supply is required by Unit 2 technical specification 3.8.4 following maintenance work on the U1 power supplies.  Previously, at 18:35 EDT on 4/4/2012, the 'B' Control Structure Chiller was declared inoperable due to an unrelated problem.  With the 'B' Control Structure Chiller inoperable coincident with the 'A' EDG or 'C' EDG inoperable, neither Control Structure Chiller would be available to perform its design function on a loss of offsite power.  This is a condition that, at the time of discovery, could have prevented fulfillment of a Safety Function and is reportable under 50.72(b)(3)(v)(D) as an 8 hour notification.  "Switching the power supplies was a planned evolution. The duration of the loss of safety function was a total of eight minutes. As a mitigating action, operators were continuously available with communication to the control room. The associated diesel generator could have been returned to an operable condition promptly if required.  "Note that Technical Specifications allows four hours to correct the condition before further actions are required, i.e. declare the features ('A' Control Structure Chiller) supported by the inoperable diesel inoperable."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47817|BRUNSWICK|CAROLINA POWER AND LIGHT CO.|2|SOUTHPORT|NC|BRUNSWICK||Y|05000325|1|2||[1] GE-4,[2] GE-4|LEE GOLDSTEIN|MARK ABRAMOVITZ|4/9/2012 00:00:00|09:15|4/9/2012 00:00:00|05:29|EDT|4/9/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||GERALD MCCOY|R2DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||VALID EMERGENCY DIESEL GENERATOR ACTUATION  "On April 9, 2012, at 0529 hours, electrical power was lost to the 4160 V emergency bus E1.  Activities to support testing of emergency bus E1 were in progress when technicians connected a recorder across the terminals of an under-voltage relay on emergency bus E1 and caused the normal supply breakers for emergency bus E1 to open.    "The power loss to emergency bus E1 affected both Unit 1 and Unit 2.  Emergency diesel generator #1 automatically started and re-energized the E1 emergency bus.  "Unit 1 was in Mode 5 and electrical systems were aligned to support testing of emergency bus E1.   As a result, no other safety system isolations or actuations occurred.  Per design, no Unit 2 safety system isolations or actuations occurred.  "The safety significance of this event is minimal.  Safety systems functioned as designed when emergency bus E1 de-energized.  There was no interruption of Unit 1 shutdown cooling as a result of this event.  Normal power supply was restored to emergency bus E1 and emergency diesel generator #1 was shutdown at 0701 hours.  "Reporting requirements met by this notification: 10CFR50.72(b)(3)(iv)(A) with specified system in 10CFR50.72(b)(3) (iv)(B)(8).  The NRC Resident Inspector has been notified."|
Part 21|47818|ABB INC|ABB INC|1|FLORENCE|SC|||Y||||||DAVID BROWN|CHARLES TEAL|4/9/2012 00:00:00|17:06|4/9/2012 00:00:00||EDT|4/27/2012 00:00:00|NON EMERGENCY|21.21(d)(3)(i)|DEFECTS AND NONCOMPLIANCE|||||||BLAKE WELLING|R1DO|GERALD MCCOY|R2DO|DAVID HILLS|R3DO|VINCENT GADDY|R4DO|PART 21 GROUP|EMAI|||||||||||N|N|0||0||N|N|0||0||N|N|0||0||PART 21 REPORT - HK CIRCUIT BREAKER STUDS FAILED TO MEET SPECIFICATION  "This letter is submitted in accordance with 10 C.F.R. 21.21(d)(3)(ii) with respect to a failure to comply with the specifications associated with two studs P/N 163392A00 and 192247A00 used in medium voltage HK circuit breakers that may be subject to failure due to hydrogen embrittlement due to incorrect processing during plating.  These studs were manufactured at the ABB Medium Voltage Service facility in Florence, SC from steel rod, heat treated in-house, and then sent to Surtronics for zinc plating with chromate treatment, including hydrogen embrittlement relief baking immediately following plating.  A total of 51 pieces of P/N 163392A00 and 104 pieces of P/N 192247A00 were plated by Surtronics."  * * * UPDATE FROM DAVID BROWN VIA FAX AT 1309 EDT on 4/27/12 * * *  The vendor has notified the affected licensees, removed all remaining studs from inventory and will be auditing Surtronics established process during the next finishing production run.  The licensees affected include EFH/Luminant, Progress Energy and TVA.  Notified R1DO (Jackson), R2DO (Musser), R3DO (Lara) and R4DO (Proulx).|
Power Reactor|47819|TURKEY POINT|FLORIDA POWER & LIGHT CO.|2|MIAMI|FL|DADE||Y|05000250|3|4||[3] W-3-LP,[4] W-3-LP|DAVID BROOKINS|CHARLES TEAL|4/10/2012 00:00:00|12:42|12/4/2010 00:00:00||EDT|4/10/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||KATHLEEN O'DONOHUE|R2DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||HABITABILITY REQUIREMENT NOT MET DUE TO MAINTENANCE ON THE TSC    "Portions of the Turkey Point Technical Support Center (TSC) Heating Ventilation and Air Conditioning System (HVAC) were deenergized for maintenance for the periods of 12/4/10 to 7/13/11 and 10/10/11 to 10/28/11 (prior Event Notification 47387, made on 10/28/11, for the 10/10-28/11 period was retracted by FPL on 12/14/11).  "For the period 10/10/11 to 10/28/11, an Equipment Clearance Order (ECO) deenergized the TSC HVAC dampers in the non-emergency position while the ECO was in effect.  The ECO was put into effect on 10/10/11.  The ECO was released at 1630 [EDT] on 10/28/2011 and functionally tested restoring TSC ventilation recirculation capability.  "For the period 12/4/10 to 7/13/11, a similar ECO deenergized all but one of the TSC HVAC dampers in the non-emergency position while the ECO was in effect.  The normal air inlet damper was stuck in the closed (emergency) position during this period.  The ECO was put into effect on 12/4/10 and released on 7/13/11.  "Compensatory measures were not put into effect during either period that the TSC HVAC dampers were deenergized.  Accordingly, FPL [Florida Power and Light] cannot conclude that the TSC would have remained continuously habitable for 30 days under every postulated design basis accident scenario as required by NRC standards.  However, the alternate Turkey Point TSC remained available at all times."  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47820|SUSQUEHANNA|PPL SUSQUEHANNA LLC|1|ALLENTOWN|PA|LUZERNE||N|05000387|1|2||[1] GE-4,[2] GE-4|ALEX MCLELLAN|JOHN KNOKE|4/11/2012 00:00:00|01:18|4/10/2012 00:00:00|21:48|EDT|4/14/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||BLAKE WELLING|R1DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||SPDS AND ERDS REMOVED FROM SERVICE DUE TO PLANNED MAINTENANCE  "At 2148 EDT, on April 10, 2012 the Unit 1 and Unit 2 Safety Parameter Display System (SPDS) and Emergency Response Data System (ERDS) were removed from service to support a planned maintenance outage on the 1A Engineered Safeguards System (ESS) Bus as part of the U1 17th Refueling and Inspection outage. The Bus Outage is expected to have a duration greater than 8 hours, but less than 72 hours. During this time, required control room hardwire indications will be available from the unaffected ESS buses. An update will be provided when SPDS/ERDS becomes available.  "Since the Unit 1 and Unit 2 SPDS/ERDS computer system will be unavailable for greater than 8 hours, this is considered a Loss of Emergency Assessment Capability and reportable under 10 CFR50.72(b)(3)(xiii)."  The licensee has notified the NRC Resident Inspector.  * * * UPDATE FROM RON FRY TO JOHN KNOKE AT 2313 EDT ON 04/14/12 * * *  "As of 1800 hours EDT on 04/13/12, Unit 1 and Unit 2 ERDS and SPDS were restored to normal operation. Operation of these systems has been monitored since that time and operation has been determined to be reliable."  The licensee has not notified the NRC Resident Inspector.  Notified the R1DO (Blake Welling)|
Non-Agreement State|47821|SINAI-GRACE HOSPITAL|SINAI-GRACE HOSPITAL|3|DETROIT|MI||SNM-1991|N||||||TIM APPLEGATE|JOHN KNOKE|4/11/2012 00:00:00|11:15|1/31/2012 00:00:00||EDT|4/12/2012 00:00:00|NON EMERGENCY|20.2201(a)(1)(i)|LOST/STOLEN LNM>1000X|||||||DAVID HILLS|R3DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||PATIENT BURIED WITH PLUTONIUM-238 PACEMAKER  A former patient at St. John Macomb-Oakland Hospital expired on 1/31/12 and the body was released to the family for burial.  The patient had a Medtronic pacemaker with a radionuclide of Plutonium-238.  The RSO at Sinai-Grace Hospital notified the NRC of this event.  There is no planned action to recover the pacemaker from the buried patient.   THIS MATERIAL EVENT CONTAINS A "CATEGORY 3" LEVEL OF RADIOACTIVE MATERIAL  Category 3 sources, if not safely managed or securely protected, could cause permanent injury to a person who handled them, or were otherwise in contact with them, for some hours.  It could possibly - although it is unlikely - be fatal to be close to this amount of unshielded radioactive material for a period of days to weeks. These sources are typically used in practices such as fixed industrial gauges involving high activity sources (for example: level gauges, dredger gauges, conveyor gauges and spinning pipe gauges) and well logging.  For additional information go to  http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf|
Non-Agreement State|47822|SABIC INNOVATIVE PLASTICS|SABIC INNOVATIVE PLASTICS|3|MT. VERNON|IN||GL|N||||||RANDY BOYER|CHARLES TEAL|4/11/2012 00:00:00|16:02|3/22/2012 00:00:00|09:30|EDT|4/11/2012 00:00:00|NON EMERGENCY|20.2201(a)(1)(i)|LOST/STOLEN LNM>1000X|||||||DAVID HILLS|R3DO|ANDREW PERSINKO|FSME|ILTAB via EMAIL||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||SIX TRITIUM EXIT SIGNS UNACCOUNTED FOR AFTER SHIPPING TO MANUFACTURER  On 1/16/12, the licensee replaced the tritium exit signs in their facility.  These exit signs were installed in the late 60's.  They put the signs back into the original packaging to return them to the manufacturer.  Gexpro [exit sign distributors] picked up the signs with the understanding that they would be shipped to SRB Technologies Inc. (the sign manufacturer).  SRB technologies did not receive the signs.  A search of the Gexpro warehouse did not locate the signs.  Additionally, SRB Technologies had no record of receiving the signs.  A total of 6 tritium exit signs are unaccounted for.  The serial numbers of the signs are 600275, 600276, 600277, 600278, and 600279.  The remaining exit sign did not have a serial number attached any longer.  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|47823|LIMERICK|EXELON NUCLEAR CO.|1|PHILADELPHIA|PA|MONTGOMERY||N|05000352|1|2||[1] GE-4,[2] GE-4|PAUL MARVEL|STEVE SANDIN|4/11/2012 00:00:00|16:32|3/19/2012 00:00:00|03:00|EST|4/11/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||BLAKE WELLING|R1DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||OFFSITE NOTIFICATION TO STATE AND LOCAL AGENCIES DUE TO NPDES DISCHARGE  "The following ENS is being submitted late.  "On Monday, March 19, 2012 at 3 a.m. a manhole overflowed during a scheduled and permitted radiological release through the cooling tower blow down outfall. As a result, several thousand gallons of water overflowed briefly, formed puddles in the area, and was discharged through a different National Pollutant Discharge Elimination System (NPDES) permitted outfall.  "The overflow was terminated, water samples were collected from the impacted areas, and the area was remediated in keeping with the Station's environmental monitoring program. Several samples showed increased levels of tritium that were well below permitted Commonwealth and Federal effluent limits.  "There were no public health risks associated with this event. The NRC Resident Inspector was notified, and courtesy notifications were made to all appropriate government agencies and local stakeholders in accordance with NEI 07-07, Industry Groundwater Protection Initiative, on March 20, 2012. Due to notification of government agencies, this event is being reported under 10CFR50.72 (b)(2)(xi)."  The licensee informed the Commonwealth of Pennsylvania, and Berks, Montgomery and Chester counties.|
Power Reactor|47824|ARKANSAS NUCLEAR|ENTERGY NUCLEAR|4|RUSSELVILLE|AR|POPE||N|05000313|1|2||[1] B&W-L-LP,[2] CE|STEVEN COFFMAN|CHARLES TEAL|4/11/2012 00:00:00|17:02|4/11/2012 00:00:00|12:40|CDT|4/11/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||VINCENT GADDY|R4DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||DRILL MESSAGE INADVERTENTLY MISTAKEN FOR ACTUAL EVENT  "As part of a planned NRC Emergency Plan Exercise on 04/11/2012, local radio and television stations were given access to drill exercise messages to facilitate public awareness of the drill.  These messages were plainly marked as 'drill' activity.  At 12:40 p.m. CST, an email was sent to Entergy Executive Management showing the contents of a web page that stated in part 'A site area emergency was declared Wednesday at Arkansas Nuclear One, Unit 1, today at 10:18 a.m. by Entergy Operations Inc. officials...'  A Little Rock, Arkansas local television station had posted a story on their internet web page at 10:53 a.m., indicating a real event, not a drill activity.  The television station News Director stated that once the error was identified, the story was removed from their web page within a matter of minutes.  This unplanned media event is being reported in accordance with 10CFR 50.72.(b)(2)(xi)."  The NRC Resident Inspector has been informed.|
Power Reactor|47825|GRAND GULF|ENTERGY NUCLEAR|4|PORT GIBSON|MS|CLAIBORNE||Y|05000416|1|||[1] GE-6|SCOTTY BEACH|CHARLES TEAL|4/11/2012 00:00:00|20:03|4/11/2012 00:00:00|18:26|CDT|4/11/2012 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||VINCENT GADDY|R4DO|ELMO COLLINS|RA|BRUCE BOGER|NRR|JOSEPH GIITTER|NRR|WILLIAM GOTT|IRD|||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||UNUSUAL EVENT DUE TO A FIRE IN THE 'A' MAIN CONDENSER  "At 1811 [CDT] hours a fire was reported inside the 'A' Main Condenser.  The fire brigade was dispatched to combat the fire and the area was evacuated.  "An Unusual Event (HU4) was declared at 1826 [CDT] hours due to the fire being in the Protected Area boundary and not extinguished within 15 minutes of detection.  "The Claiborne County, Mississippi Fire Department was notified to provide assistance.  "At 1847 [CDT] hours the fire was reported to be extinguished.  "The Unusual Event was terminated at 1900 [CDT] hours."  The source of the fire is still under investigation.  There were no personnel injuries.  The licensee notified the NRC Resident Inspector, Mississippi Emergency Management, Governor Office of Homeland Security, Claiborne County Sheriffs Department, Tensas Sheriffs Office, Mississippi Highway Patrol, and the Louisiana Department of Environmental Quality.  Notified DHS SWO, FEMA, NICC and Nuclear SSA via email.|
Power Reactor|47826|FERMI|DETROIT EDISON CO.|3|NEWPORT|MI|MONROE||N|05000341|2|||[2] GE-4|JEFF GROFF|CHARLES TEAL|4/11/2012 00:00:00|22:38|4/11/2012 00:00:00|18:07|EDT|4/11/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(B)|POT RHR INOP|||||||DAVID HILLS|R3DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||BRIEF LOSS OF RHR PUMP DUE TO VOLTAGE TRANSIENT  "On 4/11/12 at 1807 EDT, with the plant shutdown in Mode 5 during Refueling Outage 15, the 'A' Residual Heat Removal pump tripped while operating in the Shutdown Cooling Mode.  The pump trip was due to an isolation of the E1150F009 'Division 1 RHR Shutdown Cooling Inboard Isolation Valve.'  This resulted in an interruption of primary decay heat removal for approximately 11 minutes.  Approximate calculated time to boil was 23.1 hours.  This is being reported under 10CFR50.72(b)(3)(v)(B) 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat.'  "At the time of the pump trip, the station was attempting to restore power to Division 2 Bus 65E from the 64T cross tie bus.  A voltage transient occurred due to a fault and caused a Group 4 (Shutdown Cooling/Head Spray) isolation signal.  Abnormal Operating Procedure 20.205.01 'Loss of Shutdown Cooling,' was entered, the Group 4 isolation was reset and Shutdown Cooling was restored at 1818 EDT.  "An investigation is in progress to determine the cause of the bus fault."  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47827|TURKEY POINT|FLORIDA POWER & LIGHT CO.|2|MIAMI|FL|DADE||Y|05000250|3|||[3] W-3-LP,[4] W-3-LP|DAVID BROOKINS|HOWIE CROUCH|4/12/2012 00:00:00|07:28|4/11/2012 00:00:00|10:48|EDT|4/12/2012 00:00:00|NON EMERGENCY|26.719|FITNESS FOR DUTY|||||||KATHLEEN O'DONOHUE|R2DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||FITNESS FOR DUTY - A CONTRACT SUPERVISOR TESTED POSITIVE ON A DRUG TEST  A non-licensed contract supervisor tested positive for illegal drugs on a random fitness-for-duty test.  The individual's access has been terminated. Contact the Headquarters Operations Officer for additional details.|
Power Reactor|47828|ROBINSON|CAROLINA POWER & LIGHT CO.|2|HARTSVILLE|SC|DARLINGTON||Y|05000261|2|||[2] W-3-LP|GEORGE CURTIS|HOWIE CROUCH|4/12/2012 00:00:00|09:42|4/12/2012 00:00:00|09:30|EDT|4/13/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||KATHLEEN O'DONOHUE|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||TSC  AND EOF OUT OF SERVICE DUE TO MAINTENANCE  "At approximately 0930 hours EDT on Thursday, April 12, 2012, the H. B. Robinson Steam Electric Plant, Unit No. 2, Technical Support Center (TSC)/Emergency Response Facility (EOF) air conditioning and charcoal filtration systems will be removed from service to facilitate the replacement of the charcoal filtration media.  The duration of work is expected to be approximately 11 hours.  Since the unavailability will last greater than 8 hours, this is considered a Loss of Emergency Assessment Capability, and reportable under 10 CFR 50.72(b)(3)(xiii).  "Due to the inability of the TSC/EOF ventilation system to maintain a habitable atmosphere, as a compensatory measure, Emergency Responders assigned to these facilities have been informed to report to the alternate facilities until such time that the TSC/EOF ventilation system has been returned to service.  "TSC/EOF ventilation system maintenance and post maintenance testing is scheduled to be completed by 2030 hours EDT on Thursday April 12, 2012.  "The NRC Resident Inspector has been informed."  * * * UPDATE FROM WARREN WONKA TO HOWIE CROUCH AT 0435 EDT ON 4/13/12 * * *   At 1814 EDT on 4/12/12, maintenance on the TSC/EOF ventilation system was completed and the TSC/EOF was returned to service.   Notified R2DO (O'Donohue).|
Fuel Cycle Facility|47829|GLOBAL NUCLEAR FUEL - AMERICAS|GLOBAL NUCLEAR FUEL - AMERICAS|2|WILMINGTON|NC|NEW HANOVER|SNM-1097|Y|07001113||||URANIUM FUEL FABRICATION|SCOTT MURRAY|JOHN KNOKE|4/12/2012 00:00:00|18:30|4/12/2012 00:00:00|08:00|EDT|4/12/2012 00:00:00|NON EMERGENCY||OTHER UNSPEC REQMNT|||||||KATHLEEN O'DONOHUE|R2DO|MICHELE SAMPSON|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||INTERNAL REPORTING REQUIREMENT FOR CRITICALITY SAFETY CONTROLS  "At approximately 8:00 am on April 12, 2012, a criticality safety engineer was notified that waste in the Dry Conversion Process (DCP) area was improperly placed into a designated storage location. Upon investigation, it was determined that waste in the designated location contained materials with a total of less than 5 kg of uranium. As a result, no unsafe condition existed.  "An operator placed a bag of waste adjacent to a partially filled receptacle instead of placing the bag into the receptacle. This resulted in a portion of one of the documented administrative controls for criticality safety, requiring 24 inches separation between waste storage locations, to be degraded. This event is being conservatively reported per internal procedural requirements.   "As an immediate corrective action, the material was removed and transferred to the waste processing area. In addition, a shop wide communication to Fuel Manufacturing Operations is underway to inform operators of the issue.  "Additional corrective actions and extent of condition are being evaluated.  "At no time was an unsafe condition present."  Notifications were sent to state and local agencies and NRC Region II.|
Agreement State|47830|VIRGINIA RAD MATERIALS PROGRAM|ECS MID-ATLANTIC, LLC|1||VA|ARLINGTON|107-314-1|Y||||||CHARLES COLEMAN|HOWIE CROUCH|4/13/2012 00:00:00|10:07|4/11/2012 00:00:00||EDT|4/13/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||BLAKE WELLING|R1DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - DAMAGED MOISTURE DENSITY GAUGE  "On April 11, 2012, a technician performing compaction tests using a CPN MC portable moisture density gauge (10 millicuries cesium-137; 50 millicuries americium-241), left the gauge during a test to prepare the next test location 20 to 30 feet away. A compaction roller ran over the gauge and shattered the gauge housing. The technician cordoned off the area and contacted the Radiation Safety Officer. The RSO contacted the Virginia Emergency Operations Center and returned the gauge to its storage area after ensuring the sources were inside their shields. Based on an onsite investigation by members of the Virginia Radioactive Materials Program, it was determined that no individual was likely to have received a radiation dose and that the gauge sources were secured in their shields. The licensee has contacted the gauge distributor to return the gauge."  Virginia Report No.: VA-12-001|
Agreement State|47831|CALIFORNIA RADIATION CONTROL PRGM|SAINT JOSEPH HOSPITAL|4|EUREKA|CA||1703|Y||||||GENE FORRER|JOHN KNOKE|4/13/2012 00:00:00|11:50|4/4/2012 00:00:00||PDT|4/13/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||VINCENT GADDY|R4DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - TC-99M GENERATOR INADVERTANTLY SENT TO NON-LICENSED RECIPIENT  The following information was received from the State of California by email:  "The hospital had sent a depleted Tc-99m generator to an incorrect recipient, a non-licensee.  They had intended to return it to Mallinckrodt in St. Louis, [Missouri].  He [RSO] thought the generator still contained about 100 mCi.  The package was shipped on 4/4/12, and was retrieved on 4/6/12 after the mistake was discovered."  The package was  returned to Saint Joseph Hospital and the following week the hospital resent the package to the correct recipient, Mallinckrodt.|
Agreement State|47832|TEXAS DEPARTMENT OF HEALTH|UNIVERSITY OF TEXAS  MD ANDERSON CANCER|4|HOUSTON|TX||00466|Y||||||ART TUCKER|JOHN KNOKE|4/13/2012 00:00:00|13:35|4/11/2012 00:00:00||CDT|4/13/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||VINCENT GADDY|R4DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - UNDERDOSE TO PATIENT RECEIVING YTTRIUM-90 MICROSPHERES TREATMENT  The following information was sent by the State of Texas via email:  "On April 13, 2012, the Agency [Texas Department of Health] was notified by the licensee that a therapy event had occurred on April 11, 2012. A patient was prescribed 105 grays using Yttrium-90 microspheres. The patient received only 77.4 grays. The licensee stated that the reduced exposure was because a flow clamp was not in the fully opened position allowing microspheres to plate out in the flow tube instead of being administered to the patient. The licensee stated that there were no adverse effects on the patient.   "Additional information will be provided as it is received in accordance with SA-300."  Texas Report No: I-8948  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.|
Part 21|47833|MITSUBISHI NUCLEAR ENERGY SYSTEMS|MITSUBISHI HEAVY INDUSTRIES, LTD|1|ARLINGTON|VA|||Y||||||EI KADOKAMI|JOHN KNOKE|4/13/2012 00:00:00|15:58|4/13/2012 00:00:00||EDT|4/19/2012 00:00:00|NON EMERGENCY|21.21(a)(2)|INTERIM EVAL OF DEVIATION|||||||BLAKE WELLING|R1DO|KATHLEEN O'DONOHUE|R2DO|DAVID HILLS|R3DO|VINCENT GADDY|R4DO|PART 21 GROUP|EMAI|||||||||||N|N|0||0||N|N|0||0||N|N|0||0||PART 21 INTERIM REPORT - STEAM GENERATOR TUBE WEAR  This interim Part 21 is in regard to San Onofre Nuclear Generating Station, Unit 2, Steam Generator replacement.  "During the first refueling outage following steam generator replacement, eddy current testing identified ten total tubes with depths of 90 to 28 percent of the tube wall thickness. Some of the affected tubes were located adjacent to retainer bars. The retainer bars are part of the floating anti-vibration bar (AVB) structure that stabilizes the u-bend region of the tubes.  "Other tubes in the two steam generators had detectable wear associated with support points elsewhere in the AVB structure. Each steam generator has 9727 tubes with an 8 percent (778 tubes) design margin for tube plugging.    "Discovery Date: February 13, 2012  "Evaluation completion schedule date: May 31, 2012"  "Those Mitsubishi Heavy Industries customers potentially affected by this issue have been notified and will receive a copy of this interim report."  Reference Document: UET-20120089 Interim Report No: U21-018-IR (0)  * * * UPDATE FROM JOSEPH TAPIA TO KARL DIEDERICH ON 4/19/12 AT 1306 EDT  * * *   "On January 31, 2012, San Onofre Unit 3 shut down due to indications of a steam generator [SG] tube leak. Steam generator tube inspections confirmed one small leak on one tube in one of the two steam generators. Continuing inspections of 100% of the steam generator tubes in both Unit 3 steam generators discovered unexpected wear, including tube to tube as well as tube to tube support structural wear. Inspection, testing, and analysis of SG tube integrity in both Unit 3 SGs is ongoing. In-situ pressure testing identified eight Unit 3 SG tubes that did not meet the target performance criteria in Technical Specification for tube integrity. One of the failed tubes was the leaking tube that required the Unit 3 shutdown.  "Discovery date: February 21, 2012  "Evaluation completion schedule date: August 31, 2012"  Notified R1DO (Joustra), R2DO (Nease), R3DO (Peterson), R4DO (O'Keefe), and Part 21 Group via email.|
Power Reactor|47834|BROWNS FERRY|TENNESSEE VALLEY AUTHORITY|2|DECATUR|AL|LIMESTONE||Y|05000259|1|2|3|[1] GE-4,[2] GE-4,[3] GE-4|JOHN RIDINGER|JOHN KNOKE|4/13/2012 00:00:00|19:31|4/13/2012 00:00:00|15:25|CDT|4/13/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|||||||KATHLEEN O'DONOHUE|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0|Refueling|0|Refueling|UNANALYZED CONDITION IMPACTING EMERGENCY DIESEL GENERATOR LOADING  "On March 14, 2012, it was determined that in the event of an Appendix R fire, fire damage to cables in certain fire areas could cause a Residual Heat Removal Service Water System (RHRSW) pump to spuriously start, overload EDG A and B, and render them inoperable during certain Appendix R fires. This was reported as an unanalyzed condition (Ref. EN #47764).  "An extent of condition analysis was completed on April 13, 2012. From this analysis it was determined that EDG A, D, 3EC, and 3ED could exceed the maximum rated loading due to the potential for an automatic or spurious start of RHRSW Pumps B3 and D3 that supply Emergency Equipment Cooling Water (EECW) to essential safety equipment.   "The following are the Fire Areas (FA) affected:   EDG A in FA 21 EDG D for FA 2-3 and 9 EDG 3EC in FA 1-1, 1-3, and 20, and  EDG 3ED in FA 1-1, 1-3, 1-4, and 20.  "This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). This is also reportable as a 60 day written report IAW 10 CFR 50.73(a)(2)(ii)(B). This event was entered into the licensee's Corrective Action Program as PER 536176.   "The NRC Resident Inspector has been notified of this event."|
Agreement State|47835|NV DIV OF RAD HEALTH|TOP DOLLAR RECYCLING|4|LAS VEGAS|NV|||Y||||||SNEHA RAVIKUMAR|JOHN KNOKE|4/13/2012 00:00:00|20:37|4/11/2012 00:00:00||PDT|4/13/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||VINCENT GADDY|R4DO|||||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - RADIOACTIVE MATERIAL DETECTED IN LOAD OF SCRAP METAL  The following information was provided by the State of Nevada via email:  On 4/11/12 a report was received [by Nevada State Health Division] that a truckload of radioactive scrap metal belonging to Top Dollar Recycling was refused in Southern California. This scrap steel originally came from a mine in Utah. On 4/12/12, the 40 foot semi-truck of scrap metal returned to Top Dollar Recycling and a site visit by the State Radiation Control Program (RCP) was conducted.   RCP staff took measurements of the semi-truck's scrap metal. Background reading was <10uR/hr, and the highest readings of 50uR/hr on contact were near the back of the truck on the driver's side. The truck was unloaded using a small bobcat loader, spread out and surveyed. One piece of scaffolding, about 4 foot long and bent roughly in half, had elevated readings of 180 uR/hr on contact and identified as Ra-226. The scaffold, which was tubular steel and hard packed with soil type debris, was the source of radiation.   The remainder of the semi-truck was unloaded, and all scaffolding metal was segregated and surveyed individually. No additional sources of radiation were identified. The other metal on the truck was also surveyed with no elevated readings. The outside of a very large pile (12' x 20') of metal from the same mine was scanned and no elevated reading were found.   Top Dollar Recycling took possession of the item, wrapped it in a plastic garbage bag before moving it to a sturdy container in the storage/control location.  Item Number: NV120011   .|
Power Reactor|47836|VOGTLE|SOUTHERN NUCLEAR OPERATING COMPANY|2|WAYNESBORO|GA|BURKE||Y|05000424|1|||[1] W-4-LP,[2] W-4-LP|JEFF TODD|JOHN KNOKE|4/14/2012 00:00:00|16:12|4/14/2012 00:00:00|13:46|EDT|4/14/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(iv)(B)|RPS ACTUATION - CRITICAL|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||KATHLEEN O'DONOHUE|R2DO|||||||||||||||||||M/R|Y|100|Power Operation|0|Hot Standby|N|N|0||0||N|N|0||0||MANUAL REACTOR TRIP DUE TO LOW MAIN FEEDWATER FLOW  "At 1346 EDT, Vogtle Unit 1 reactor was manually tripped from 100% power due to Main Feedwater Pump 'B' discharge flow lowering unexpectedly. All control rods fully inserted. AFW system automatically actuated as expected. System responses allowed for an uncomplicated reactor trip response. Plant is stable in Mode 3 during cause investigation."  The electrical lineup remained normal.  No safety valves lifted due to the trip. Decay heat is being removed via the steam dumps to the main condenser.  The licensee has notified the NRC Resident Inspectors.|
Power Reactor|47837|PALO VERDE|ARIZONA NUCLEAR POWER PROJECT|4|WINTERSBURG|AZ|MARICOPA||Y||||3|[1] CE,[2] CE,[3] CE|MICHAEL KOHRT|JOHN KNOKE|4/15/2012 00:00:00|17:13|4/15/2012 00:00:00|12:20|MST|4/15/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(iv)(B)|RPS ACTUATION - CRITICAL|||||||VINCENT GADDY|R4DO|||||||||||||||||||N|N|0||0||N|N|0||0||M/R|Y|0|Startup|0|Hot Standby|MANUAL REACTOR TRIP DUE TO CONTROL ROD DEVIATION DURING STARTUP  "The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73.  "On April 15, 2012 at approximately 1220 Mountain Standard Time (MST), Palo Verde Unit 3 was manually tripped during low power physics testing.  "While conducting low power physics testing following a refueling outage, Regulating Group 1 rods were being inserted while simultaneously diluting to maintain a constant power level below the Point of Adding Heat. While inserting rods one rod deviated from its subgroup when it stopped moving. The Reactor Operator immediately ceased rod motion and the dilution was stopped. The residual positive reactivity in the core caused a corresponding reactor power increase that approached procedural power limits set forth in the low power physics testing procedure. Based on these indications, operators initiated a manual reactor trip.  "Following the reactor trip, all CEAs inserted fully into the core. All systems operated as expected and this event was diagnosed as an uncomplicated reactor trip. No emergency plan classification was required per the Emergency Plan. Safety related busses remained powered during the event from offsite power and the offsite power grid is stable. No major equipment was inoperable prior to the event that contributed to the event or complicated operator response. Unit 3 is stable and in Mode 3 feeding Steam Generators with Auxiliary Feedwater Pump 'N'. The event did not result in any challenges to fission product barriers and there were no adverse safety consequences as a result of this event. The event did not adversely affect the safe operation of the plant or the health and safety of the public.  "The NRC Resident Inspector was informed of the Unit 3 reactor trip."  The electrical lineup remained normal. Decay heat is being removed via the steam bypass to the main condenser.|
Power Reactor|47838|HARRIS|CAROLINA POWER & LIGHT CO.|2|RALEIGH|NC|WAKE & CHATHAM||Y|05000400|1|||[1] W-3-LP|CURTIS BULLOCK|VINCE KLCO|4/16/2012 00:00:00|07:38|4/16/2012 00:00:00|08:00|EDT|4/16/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||KATHLEEN O'DONOHUE|R2DO|||||||||||||||||||N|Y|92|Power Operation|92|Power Operation|N|N|0||0||N|N|0||0||PLANNED MAINTENANCE ON THE TECHNICAL SUPPORT CENTER NORMAL POWER SUPPLY  "At approximately 0800 [EDT] on April 16, 2012, the Harris Nuclear Plant (HNP) Technical Support Center (TSC) normal power feed will be removed from service for scheduled maintenance.  "The maintenance will consist of first switching the TSC to the TSC backup power supply. The normal supply will be disconnected and replaced with another offsite power source which is independent of the Harris switchyard. This power arrangement will remain in place while maintenance is performed on the TSC normal power supply and is expected to last approximately two months. A backup diesel generator is stationed near the TSC which can be connected if necessary during an emergency.  "An update will be provided when the TSC normal power supply has been returned to its normal alignment  "This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2, since this work activity affects an emergency response facility for the duration of the maintenance.  "The [NRC] Senior Resident Inspector has been informed."|
Agreement State|47839|KENTUCKY DEPT OF RADIATION CONTROL|CLOSE THE LOOP, INC.|1|HEBRON|KY||401-865-430|Y||||||CHRISTOPHER KEFFER|JOHN SHOEMAKER|4/16/2012 00:00:00|14:22|4/13/2012 00:00:00|07:00|CDT|4/16/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||JUDY JOUSTRA|R1DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||LOST STATIC ELIMINATOR  The licensee notified the Kentucky Department of Public Health and Safety, Radiation Health Branch on April 16, 2012 at 1135 (CDT) of a missing static eliminator.  The static eliminator was discovered to be missing by an oncoming operator on April 13, 2012 at 0700 (CDT).  An operator on the previous shift reported the device to be in place.  The Director of Facility Operations was not in the office on April 13, 2012 and was not made aware of the missing device until April 16, 2012 at 0715 (CDT).  The licensee initiated a search for the device on April 16, 2012 at 0715 (CDT) but has been unsuccessful in locating the device.  The licensee plans to continue searching for the device.  The licensee does not suspect tampering or theft and no exposures to personnel have occurred or are expected at this time.  The device is a Model P-2021 Static Eliminator, Serial # A2HU670, with a 10 mCi Po-210 source.  Kentucky Radiation Health Branch Report number KY120007.  No other notifications have been made by the licensee.  THIS MATERIAL EVENT CONTAINS A "CATEGORY 3" LEVEL OF RADIOACTIVE MATERIAL  Category 3 sources, if not safely managed or securely protected, could cause permanent injury to a person who handled them, or were otherwise in contact with them, for some hours.  It could possibly - although it is unlikely - be fatal to be close to this amount of unshielded radioactive material for a period of days to weeks. These sources are typically used in practices such as fixed industrial gauges involving high activity sources (for example: level gauges, dredger gauges, conveyor gauges and spinning pipe gauges) and well logging.  For additional information go to  http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf  Note: This device is assigned an IAEA Category 3 value based on the actual radioactivity of the source, not on the device type. (Reference IAEA RG-G-1.9)|
Power Reactor|47840|HARRIS|CAROLINA POWER & LIGHT CO.|2|RALEIGH|NC|WAKE & CHATHAM||Y|05000400|1|||[1] W-3-LP|CURTIS BULLOCK|JOHN SHOEMAKER|4/16/2012 00:00:00|17:11|4/16/2012 00:00:00|09:32|EDT|4/16/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||KATHLEEN O'DONOHUE|R2DO|||||||||||||||||||N|Y|92|Power Operation|92|Power Operation|N|N|0||0||N|N|0||0||UNAVAILABILITY OF TECHNICAL SUPPORT CENTER RADIATION MONITOR  "At 0932 EDT on April 16, 2012, the Harris Nuclear Plant (HNP) Technical Support Center (TSC) normal power feed was removed from service for scheduled maintenance.  "Upon restoration of power to the TSC, it was identified that the TSC radiation monitor (RM-3653C) did not return to service as expected.  Following discovery, Radiation Protection technicians were dispatched to investigate the cause of RM-3653C not operating properly.  "At 1627 EDT, the TSC radiation monitor has been returned to service and is operable.  "This event is reportable per 10CFR50.72(b)(3)(xiii) as described in NUREGĀ1022, Rev. 2 due to the loss of the TSC radiation monitor.  "The NRC Senior Resident Inspector has been informed"  The cause of the radiation monitor failure was related to a dead back-up battery which did not allow the radiation monitor to properly re-boot after power was restored.|
Agreement State|47841|PA BUREAU OF RADIATION PROTECTION|PINNACLE HEALTH HOSPITALS|1|CAMP HILL|PA||PA-0037|Y||||||DAVID ALLARD|MARK ABRAMOVITZ|4/17/2012 00:00:00|10:53|4/13/2012 00:00:00||EDT|4/17/2012 00:00:00|NON EMERGENCY||AGREEMENT STATE|||||||JUDY JOUSTRA|R1DO|ANGELA MCINTOSH|FSME|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||AGREEMENT STATE REPORT - RADIOTHERAPY UNDERDOSE  The following report was received via fax:  "On April 13, 2012, a patient was to receive a prescribed dose of 100 mCi of iodine-131 (I-131) for thyroid cancer.  Two capsules of I-131 were received from the radiopharmacy in a single vial which was assayed to assure proper dose prior to administering to the patient.  The patient was given the contents of the vial for treatment and the treatment was considered complete.  The Nuclear Medicine staff prepared the remaining container for return to the radiopharmacy on Monday April 16, 2012.  The radiopharmacy picked up the I-131 container and returned it to their facility.  Upon inspection of the returned package, the radiopharmacy discovered that one of the two capsules intended for the treatment remained lodged in the vial.  The radiopharmacy informed Tristan and the RSO at 12:30 EDT on Monday April 16, 2012.  The patient received 50 mCi of the intended 100 mCi dose.  "Patient and referring physician are in the process of being notified. The [PA] Department [of Radiation Protection] plans to do a reactive inspection on April 17, 2012."  PA Event ID: PA120013  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.|
Part 21|47842|SWAGELOK COMPANY|SWAGELOK COMPANY||||||N||||||BRUCE FLUSCHE|JOHN SHOEMAKER|4/17/2012 00:00:00|13:51|4/17/2012 00:00:00||EST|4/17/2012 00:00:00|NON EMERGENCY|21.21(d)(3)(i)|DEFECTS AND NONCOMPLIANCE|||||||REBECCA NEASE|R2DO|NEIL OKEEFE|R4DO|PART 21 GROUP|EMAI|||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||PART 21 REPORT - SWAGELOK U SERIES BELLOWS VALVES LOOSENING OF STEM TIPS  "Swagelok received a return from Duke Energy (Oconee) for two 8U series bellows valves for investigation of stem tips that had loosened during performance testing of equipment.  "Our evaluation confirmed loosening of the stem tips and determined the root cause to be higher than normal torque being applied to the valve handle during closure. (Please note that we did not specify a minimum or maximum torque for our operating instructions).  This caused the stem and the stem insert interface to loosen, but not fully disengage. Our tests show that closure to catalog specification of 4.0 x 10-9 atm. cc/sec of helium can still be achieved with this condition.  "Therefore, it is the opinion of Swagelok that there is no inherent safety risk associated due to this condition, however lower than expected flow through the valve, or erratic flow, can occur if the stem tip loosens. If utilities consider full flow as a safety function, they should evaluate the valves currently in service for this condition.  "Applicability - There have been no other field returns for this condition. The possible condition extends to the Swagelok 4U, 6U and 8U series bellows valves. Only hand operated valves are susceptible; air operated valves are excluded, as are the 12U series valves supplied by Swagelok."  Callaway may also be affected and will be notified by the Supplier.|
Power Reactor|47843|HARRIS|CAROLINA POWER & LIGHT CO.|2|RALEIGH|NC|WAKE & CHATHAM||Y|05000400|1|||[1] W-3-LP|JUSTIN KELLY|JOHN SHOEMAKER|4/17/2012 00:00:00|16:45|4/17/2012 00:00:00|09:00|EDT|4/17/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|Y|91|Power Operation|91|Power Operation|N|N|0||0||N|N|0||0||TECHNICAL SUPPORT CENTER HAD ONLY ONE POWER SOURCE DURING PREPLANNED MAINTENANCE  "On April 16, 2012, at 0738 hours, the Harris Nuclear Plant notified the NRC Operations Center (i.e., Event Number 47838) of preplanned maintenance on the Technical Support Center (TSC) normal power supply.  "Following completion of the power transfer, it was discovered that in the current alignment, the TSC is only powered from one power source, which is the backup power supply.    "A backup diesel generator is stationed near the TSC which can be connected if necessary during an emergency.  "Activities are in progress to modify the existing procedure to allow the TSC to be connected to the offsite power source which will restore two sources of power to the TSC.  This normal power arrangement is expected to remain in place while maintenance is performed on the TSC normal power supply for approximately two months.  "The NRC Resident Inspector has been informed.  "This report was made in accordance with 10 CFR 50.72(b)(3)(xiii)."|
Power Reactor|47844|SUSQUEHANNA|PPL SUSQUEHANNA LLC|1|ALLENTOWN|PA|LUZERNE||N|||2||[1] GE-4,[2] GE-4|ALEX MCLELLAN|JOHN SHOEMAKER|4/17/2012 00:00:00|21:02|4/17/2012 00:00:00|15:40|EDT|4/17/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(C)|POT UNCNTRL RAD REL|||||||JUDY JOUSTRA|R1DO|||||||||||||||||||N|N|0||0||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||UNIT 2 SECONDARY CONTAINMENT AFFECTED BY VIOLATION OF UNIT 1 SECONDARY CONTAINMENT INTEGRITY DURING OUTAGE  "At 1540 (EDT) on 4/17/12, with Unit 1 in mode 5 and Unit 2 in mode 1, the Work Control Center was notified that the U1 #2 Main Stop Valve (MSV) was disassembled.  The U1 #2 MSV was required to be intact to maintain Unit 1 Secondary Containment.  Ongoing work on the D Main Steam Line Outboard Valve created a pathway that violated Unit 1 secondary containment integrity.  Unit 1 Secondary Containment is required to be operable for Unit 2 while Unit 1 Zone 1 is aligned to the Recirculation Plenum.  Unit 1 Zone 1 was isolated from the recirculation plenum and Unit 2 Secondary Containment was restored at 1643 (EDT) on 4/17/12.  Unit 2 Secondary Containment differential pressures were maintained throughout the event.  "This is considered a loss of an entire safety function and requires an 8 hour report per 10CFR50.72(b)(3)(v)(C)."  The licensee is still investigating the cause but it appears to be associated with recent administrative changes to the Reactor Vessel draining definition and work process procedures.  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47845|BRAIDWOOD|EXELON NUCLEAR CO.|3|BRACEVILLE|IL|WILL||Y|05000456|1|||[1] W-4-LP,[2] W-4-LP|DALE RUSH|JOHN SHOEMAKER|4/17/2012 00:00:00|23:12|4/17/2012 00:00:00|21:30|CDT|4/17/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||HIRONORI PETERSON|R3DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||PLANT PROCESS COMPUTER REMOVED FROM SERVICE FOR PLANNED REPLACEMENT  "At 2130 (CDT) on April 17, 2012, the Unit 1 Plant Process Computer (PPC) was removed from service for a planned replacement in the current Unit 1 Refueling Outage.  The Unit 1 PPC feeds the Safety Parameter Display System (SPDS) used in the Main Control Room (MCR) and the Technical Support Center (TSC).  The Unit 1 PPC also feeds the Emergency Response Data System (ERDS).  The Unit 1 and Unit 2 PPCs also feed the Plant Parameter Display System (PPDS) used in the MCR, TSC and Emergency Operations Facility (EOF).  Meteorological data will remain available in the MCR but not through ERDS for either Unit 1 or Unit 2.  The dose assessment program will remain functional as the Unit 2 Plant Process computer will be capable of providing the necessary data through PPDS to run the program.  The dose assessment program is not affected by the Unit 1 PPC being out of service.  As compensatory measures, a proceduralized backup method to fax or communicate via a phone circuit applicable data to the NRC, TSC, and EOF exists.  There is no impact to the Emergency Notification System (ENS) or Health Physics Network (HPN) communication systems.  "The new Unit 1 PPC is scheduled to be functional on April 21, 2012.  However, based on the mode Unit 1 will be in, this will limit the number of points that would provide usable data.  The Unit 1 PPC will be tested as mode changes occur.  The Unit 1 PPC is planned to be declared functional by Mode 2.  A follow-up ENS call will be made once the Unit 1 PPC is declared functional.    "The loss of SPDS and ERDS is a 'major loss of assessment capability' and is reportable under 10CFR50.72(b)(3)(xiii).  "The NRC Senior Resident Inspector and the State of Illinois (through the Illinois Emergency Management Agency Resident Inspector) have been notified of this ENS call."|
Part 21|47846|FISHER CONTROLS INTERNATIONAL|FISHER CONTROLS INTERNATIONAL, LLC|3|MARSHALLTOWN|IA|||Y||||||GEORGE BAITINGER|JOHN SHOEMAKER|4/18/2012 00:00:00|14:46|4/18/2012 00:00:00||CDT|4/18/2012 00:00:00|NON EMERGENCY|21.21(d)(3)(i)|DEFECTS AND NONCOMPLIANCE|||||||JUDY JOUSTRA|R1DO|REBECCA NEASE|R2DO|NEIL OKEEFE|R4DO|PART 21 GROUP|EMAI|||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||FISHER TYPE 9500 BUTTERFLY VALVES CONTAIN NON DEDICATED PARTS  "While reviewing an incoming order for a Type 9500 Butterfly Valve, it was discovered that two parts, namely the Thrust Plate and Thrust Plate Cap Screws, were not called out for Commercial Grade Dedication per Fisher Manufacturing Procedure (FMP) 2K27.  These parts are considered an essential-to-function part in order for the valve to perform an active safety function.  "Upon discovery of this possible failure to comply, an investigation was performed on all Safety-Related Type 9500 Butterfly Valves that have been shipped, are in active service or inventory, and are required contractually to be dedicated and it was determined that the Thrust Plate and Thrust Plate Cap Screws in these valves were supplied commercial grade and had not been dedicated per FMP 2K27.  "The Thrust Plates retain the Thrust Sleeve Assemblies which provide a bearing surface for the Valve Shaft.  The Thrust Sleeve Assemblies also provide adjustment to the Valve Body Liner.  The Thrust Plate Cap Screws retain the Thrust Plates.  Without the Thrust Plates or the Thrust Plate Cap Screws, there is a risk that the Thrust Sleeve Assemblies will disengage from the Valve Body, preventing the valve from performing its active safety-related function.  "It is Fisher's opinion that while the Thrust plates and Thrust Plate Cap Screws were not dedicated, the failure to dedicate does not pose an inherent safety risk given the materials used.    "Fisher will revise its procedures to ensure that this issue is corrected and does not happen in the future.  "Action Required: Fisher will supply properly dedicated Thrust Plates and Thrust Plate Cap Screws for the affected valves that are in inventory or service, at no cost to the affected customers.  Notification has been made to plants with the Type 9500 Butterfly Valves for; Cooper Plant (NPPD), Vogtle Plant (Georgia Power), Ft. Calhoun Plant (OPPD), and Millstone Plant (Dominion) and specific valve order numbers and serial numbers were provided.|
Power Reactor|47847|QUAD CITIES|EXELON NUCLEAR CO.|3|CORDOVA|IL|ROCK ISLAND||Y|||2||[1] GE-3,[2] GE-3|JAMES COX|JOHN SHOEMAKER|4/18/2012 00:00:00|18:07|4/18/2012 00:00:00|15:11|CDT|4/18/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(iv)(B)|RPS ACTUATION - CRITICAL|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||HIRONORI PETERSON|R3DO|||||||||||||||||||N|N|0||0||A/R|Y|20|Power Operation|0|Hot Shutdown|N|N|0||0||UNIT TWO AUTOMATIC REACTOR SCRAM ON HIGH REACTOR PRESSURE  "On April 18, 2012, at 1511 hours (CDT), an automatic scram occurred on high reactor pressure.  The pressure increase occurred during post-modification testing on the main generator automatic voltage regulator, which had been upgraded during the recent refueling outage.  "Following the reactor scram, reactor water level decreased to approximately zero inches, which resulted in automatic Group II and III isolations (Reactor Water Clean Up and Secondary Containment Isolation) as expected.  All systems responded properly to the event.  The cause of the event is still under investigation.  "Unit 1 was unaffected by the event and remains at 100% power.  "This report is being made in accordance with 10CFR50.72(b)(2)(iv)(B) and 10CFR50.72(b)(3)(iv)(A)."  All Control Rods fully inserted, decay heat is being removed through the bypass steam valves to the main condenser, and the plant remains in a normal shutdown electrical alignment.  The high reactor pressure appears to have been caused by a load rejection associated with the main generator voltage regulator testing.  The licensee has informed the NRC Resident Inspector.|
Power Reactor|47848|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|AMY BURKHART|JOHN SHOEMAKER|4/18/2012 00:00:00|18:53|10/6/2011 00:00:00|13:44|CDT|4/18/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|||||||NEIL OKEEFE|R4DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||WASTE DISPOSAL SYSTEM CLASS ONE SEISMIC SUPPORT INOPERABLE  "The Waste Disposal System [WDS] Class 1 piping requires operable seismic supports downstream of the isolation valve class break.  Currently, eight (8) INC [International Nuclear Safety, Corp.] snubbers have been degraded to [Non Nuclear System] NNS Class 4 ridged struts.  The snubbers original design function was to allow thermal motion but restrain seismic motion.   "The snubbers have been identified as potential to create an unanalyzed condition that over stresses the safety class 1 drain pipe upstream of the isolation valve if the snubbers on the drain pipe downstream of the isolation valve were in a locked condition (acting as a strut).  Per NRC bulletin 81-01, these snubbers are assumed to be frozen and do not allow movement of the pipe; thus, they have been degraded to rigid struts as they are not in the snubber program and are not tested. They still provide a seismic safety function for [class] II/I issues and act as a strut to provide horizontal restraint to the WDS piping.    "The snubbers were removed from the piping system and tested to determine their performance and if they would have moved to allow thermal growth.  Six snubbers failed the test and were either in a locked condition or their movement was dimensionally small relative to the required movement. The [Reactor Coolant System] RCS is within acceptable stress values with the snubbers removed.  "The 8-hour regulatory reporting time has been exceeded."  An initial Reportability Evaluation was completed on March 26, 2012 and had determined the supports were operable.  A second Reportability Evaluation later determined the supports have been inoperable since October 6, 2011.  The WDS is used to drain the RCS.  The licensee will notify the NRC Resident Inspector.|
Power Reactor|47849|KEWAUNEE|NUCLEAR MANAGEMENT COMPANY|3|KEWAUNEE|WI|KEWAUNEE||Y|05000305|1|||[1] W-2-LP|LOGAN MILLER|JOHN SHOEMAKER|4/18/2012 00:00:00|22:46|4/18/2012 00:00:00|19:55|CDT|4/18/2012 00:00:00|NON EMERGENCY|26.719|FITNESS FOR DUTY|||||||HIRONORI PETERSON|R3DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||FITNESS FOR DUTY  A non-supervisor licensed operator had a confirmed positive result for alcohol during a random fitness-for-duty test.  The operator's access to the plant has been suspended.  Contact the Headquarter Operations Officer for additional details.  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47850|LIMERICK|EXELON NUCLEAR CO.|1|PHILADELPHIA|PA|MONTGOMERY||N|05000352|1|||[1] GE-4,[2] GE-4|BRANDON SHULTZ|KARL DIEDERICH|4/19/2012 00:00:00|11:10|4/19/2012 00:00:00|07:53|EDT|4/20/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(iv)(B)|RPS ACTUATION - CRITICAL|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||JUDY JOUSTRA|R1DO|||||||||||||||||||M/R|Y|100|Power Operation|0|Hot Shutdown|N|N|0||0||N|N|0||0||MANUAL REACTOR SCRAM FOLLOWING LOSS OF RECIRCULATION PUMPS  "Limerick Unit 1 was manually scrammed from 100% power at 0753 hours on 4/19/12 in accordance with plant procedure OT-112 'Recirculation Pump Trip' when both 1A and 1B Recirculation Pump Adjustable Speed Drives (ASDs) tripped due to an electrical fault affecting the 144D and 114A non-safety related 480V Load Centers. The shutdown was normal and the plant is stable in Hot Shutdown with normal pressure control via the Main Steam Bypass valves to the main condenser and normal level control using feedwater. The manual RPS actuation is reportable under 10 CFR 50.72(b)(2).  "The Technical Support Center (TSC) Normal Air conditioning systems shut down due to loss of power from the 144D Load Center. The loss of power also affects the flow indication for the Emergency Ventilation system. This is considered a Loss of Emergency Assessment Capability, and reportable under 10 CFR 50.72(b)(3)(xiii). The Emergency TSC Ventilation system is available but flow cannot be verified. During a required activation the TSC, responders would report to the TSC. If conditions required use of the Emergency Ventilation system, the Station Emergency Director would assess habitability in accordance with Station procedures. TSC relocation of personnel would be directed as required until such time that the TSC ventilation system is returned to service"  The licensee notified the NRC Resident Inspector.  * * * UPDATE AT 1726 ON 4/20/2012 FROM BRANDON SHULTZ TO MARK ABRAMOVITZ * * *  "The Technical Support Center (TSC) 144D load center has been re-energized, restoring the emergency ventilation flow indication and emergency assessment capability to its normal stand-by condition."    The switchgear was inspected for any potential grounds and then reenergized at approximately 0800 EDT on 4/20/2023.  The licensee notified the NRC Resident Inspector.    Notified the R1DO (Joustra).|
Fuel Cycle Facility|47851|GLOBAL NUCLEAR FUEL - AMERICAS|GLOBAL NUCLEAR FUEL - AMERICAS|2|WILMINGTON|NC|NEW HANOVER|SNM-1097|Y|07001113||||URANIUM FUEL FABRICATION|SCOTT MURRAY|ERIC SIMPSON|4/19/2012 00:00:00|13:30|4/18/2012 00:00:00|13:45|EDT|4/19/2012 00:00:00|NON EMERGENCY|70.50(b)(2)|SAFETY EQUIPMENT FAILURE|||||||REBECCA NEASE|R2DO|DENNIS DAMON|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||LOCAL ALARM HORN WAS FOUND TO BE INAUDIBLE DURING ROUTINE TESTING  "At approximately 1345 EDT on 4/18/12, during routine testing on an outdoor Criticality Accident Alarm System (CARS) Data Acquisition Module (DAM #22), the local alarm horn in the Wilmington Field Services Center (WFSC) building #3 inspection records area was found to be inaudible. The cause and extent of the condition is under investigation.  "Personnel were removed from the inspection records area until compensatory measures were established. There are no active fissile material operations impacted by this discovery, thus no unsafe condition existed.  "This event is being reported pursuant to the requirements of 10CFR70.50 (b)(2)."  The licensee notified the NC Division of Radiation Protection and the New Hanover County Emergency Response Center.|
Power Reactor|47852|BEAVER VALLEY|FIRSTENERGY NUCLEAR OPERATING COMPANY|1|SHIPPINGPORT|PA|BEAVER||N|05000334|1|||[1] W-3-LP,[2] W-3-LP|DAN SCHWER|BILL HUFFMAN|4/19/2012 00:00:00|17:47|4/19/2012 00:00:00|11:58|EDT|4/19/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xii)|OFFSITE MEDICAL|||||||JUDY JOUSTRA|R1DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||OFFSITE TRANSPORT OF POTENTIALLY CONTAMINATED INJURED WORKER  "At approximately 1156 EDT on 4/19/2012, a craft contractor worker was transported to an off-site medical facility due to a work-related injury. The injured individual was partially surveyed by a Health Physics technician in their anti-contamination clothing prior to leaving the site and no radioactive contamination was detected. The injured individual was then transported by ambulance accompanied by a Health Physics technician to the local hospital for medical treatment.  "At the hospital, the individual and applicable areas and equipment were surveyed by the Health Physics technician and no radioactive contamination was detected. The individual's anti-contamination clothing was returned to the site.  "This notification is being made under the 10CFR50.72(b)(3)(xii) reporting requirements since a complete survey of the injured individual was unable to be made and he was considered to be potentially contaminated prior to being transported offsite.  "The site NRC Resident Inspector has been notified."  The worker was in containment on a polar crane platform when injured.  He was subsequently released from the hospital and cleared to work.|
Power Reactor|47853|BROWNS FERRY|TENNESSEE VALLEY AUTHORITY|2|DECATUR|AL|LIMESTONE||Y|05000259|1|||[1] GE-4,[2] GE-4,[3] GE-4|RODNEY NACOSTE|MARK ABRAMOVITZ|4/19/2012 00:00:00|20:58|4/19/2012 00:00:00|14:30|CDT|4/19/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||HPCI FAILED TRIP TESTING SURVEILLANCE  "On 04/19/12 at 1430 while performing 1-SR-3.5.1.7, HPCI [High Pressure Coolant Injection] Main & Booster Pump Set developed head & flow rate at rated reactor pressure.  The HPCI turbine failed to trip using the manual trip pushbutton.  This manual trip pushbutton should have caused the 1-FCV-73-18, HPCI TURBINE STOP VALVE, to go closed.  HPCI was secured by taking the 1-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to close.  The 1-FCV-73-18, HPCI TURBINE STOP VALVE, also failed to go closed locally using the 1-XCV-73-18, HPCI TURBINE MECHANICAL TRIP, nor did it go closed when the auxiliary oil pump was secured.  "With the 1-FCV-73-18, HPCI TURBINE STOP VALVE, open, the HPCI ramp generator is no longer in the circuit therefore, should an initiation occur and cause the 1-FCV-73-16, HPCI TURBINE STEAM SUPPLY VALVE, to open there is the potential for the HPCI turbine to over speed. Therefore, HPCI was isolated using 1-FCV-73-3, HPCI STEAM LINE OUTBD ISOL VALVE.  This incident is reportable as an 8-hour ENS notification under 10CFR 50,72 (b)(3)(v) as 'any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident.'  "It also requires a 60 day written report in accordance with 10CFR 50.73(a)(2)(vii).  "The NRC Resident Inspector has been notified."|
Power Reactor|47854|MCGUIRE|DUKE POWER|2|CORNELIUS|NC|MECKLENBURG||Y|05000369|1|||[1] W-4-LP,[2] W-4-LP|RICK JACKSON|MARK ABRAMOVITZ|4/19/2012 00:00:00|22:07|4/19/2012 00:00:00|17:45|EDT|4/19/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|Y|100|Power Operation|100|Power Operation|N|N|0||0||N|N|0||0||POTENTIAL FIRE INDUCED SPURIOUS OPERATION OF STEAM GENERATOR PORV BLOCK VALVES  "On March 22, 2012, McGuire Nuclear Station notified the NRC of a condition which could result in fire induced spurious operation of the block valves for the Unit 2 Steam Generator (SG) Power Operated Relief Valves (PORVs).  This potential condition could adversely affect the ability to achieve and maintain a cold shutdown of Unit 2.  This represented a condition reportable as per the requirements of 10 CFR 50.72(b)(3)(ii)(B). Reference Event #47764 [for Unit-2].  "In accordance with 10 CFR 50.72(c)(2)(i), McGuire is reporting results of ensuing cause and extent of condition evaluation of Event #47764, which has identified a similar condition on McGuire Unit 1.  This condition could result in a fire induced spurious operation of the block valves for the Unit 1 SG PORVs.  This spurious operation could potentially damage these valves and render them inoperable.  Since these valves are unisolable, they could not be repaired to allow the Unit 1 SGs to be used for Unit cool down if needed.  This potential condition could adversely affect the ability of the plant to achieve and maintain a cold shutdown of Unit 1. This represents a condition reportable as per the requirements of 10 CFR 50.72(b)(3)(ii)(B).  This potential condition is mitigated by existing compensatory measures (i.e. fire watches) which have been in place as part of McGuire's transition to NFPA 805."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47855|HARRIS|CAROLINA POWER & LIGHT CO.|2|RALEIGH|NC|WAKE & CHATHAM||Y|05000400|1|||[1] W-3-LP|ADAM HELSEL|MARK ABRAMOVITZ|4/19/2012 00:00:00|22:32|4/19/2012 00:00:00|17:24|EDT|4/19/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(xiii)|LOSS COMM/ASMT/RESPONSE|||||||REBECCA NEASE|R2DO|||||||||||||||||||N|Y|90|Power Operation|90|Power Operation|N|N|0||0||N|N|0||0||TSC VENTILATION INOPERABLE  "At approximately 1724 EDT on April 19, 2012 with the Harris Nuclear Plant operating at 90% power in mode 1, Air Handler 17 was discovered not operating.  Air Handler 17 provides ventilation supporting environmental habitability to the Technical Support Center (TSC) to keep ambient temperatures habitable for personnel and to ensure communications and assessment equipment remains functional.  This degrades the heating and cooling capacity of the TSC by approximately one third.  Due to the cool weather in the forecast, the remaining cooling systems will be operating more efficiently than they would in the hot summer months, which mitigates the impact of the degraded ventilation system.  The system has been repaired as of 1950 EDT on April 19, 2012.  The cause of the equipment failure was a broken belt and a condition report has been entered into the site's corrective action program.  This report is being made in accordance with 10 CFR 50.72, criterion (b)(3)(xiii) as a condition that may impair the functionality of an Emergency Response Facility.  The plant continues to operate at 90%.  Other equipment functioned as expected.  The on call Emergency Response Manager and Site Emergency Coordinator (TSC) have been notified and the alternate TSC is available."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47856|SAN ONOFRE|SOUTHERN CALIFORNIA EDISON  COMPANY|4|SAN CLEMENTE|CA|SAN DIEGO||Y|||2||[1] W-3-LP,[2] CE,[3] CE|BILL POIER|BILL HUFFMAN|4/20/2012 00:00:00|16:29|4/20/2012 00:00:00|12:49|PDT|4/20/2012 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||WILLIAM RULAND|NRR|BILL GOTT|IRD|ERIC LEEDS|NRR|ELMO COLLINS|R4 R|RYAN LANTZ|R4|||||||||||N|N|0||0||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||UNUSUAL EVENT DECLARED DUE TO ELECTRICAL PANEL FIRE IN TURBINE BUILDING  San Onofre Unit 2 declared an Unusual Event at 1249 PDT today due to an electrical fire within a fire detection panel (2L319) in the Unit 2 turbine building.  The fire was declared extinguished at 1314 PDT after the panel was de-energized.  The location of the fire had no impact on plant operation, equipment, or personnel safety.  No offsite assistance was necessary.   The licensee exited the Unusual Event at 1341 PDT after the licensee confirmed that the fire was terminated and there was no ongoing risk to the plant.  The licensee notified state and local authorities and the NRC Resident Inspector.    The NRC Operations Center notified other Federal Agencies (DHS SWO, FEMA Ops, DHS NICC, and NuclearSSA via e-mail).|
Power Reactor|47857|HARRIS|CAROLINA POWER & LIGHT CO.|2|RALEIGH|NC|WAKE & CHATHAM||Y|05000400|1|||[1] W-3-LP|AL BOSTIC|MARK ABRAMOVITZ|4/21/2012 00:00:00|12:45|4/21/2012 00:00:00|05:15|EDT|4/24/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(C)|POT UNCNTRL RAD REL|50.72(b)(3)(v)(D)|ACCIDENT MITIGATION|||||REBECCA NEASE|R2DO|||||||||||||||||||N|N|0|Hot Shutdown|0|Hot Shutdown|N|N|0||0||N|N|0||0||MAIN STEAM ISOLATION VALVES FAIL TO SHUT  "At approximately 0515 EDT on April 21, 2012, the Harris Nuclear Plant [HNP] was in the process of a normal plant shutdown for a refueling outage.  HNP was at 0% power in Mode 4.  During OST-1046, MSIV Operability Test, 'B' and 'C' Main Steam Isolation Valves (MSIV), failed to close from the main control board.  "At 0648 EDT an April 21, 2012 'B' MSIV shut immediately after the instrument air supply was isolated.  "At 0938 EDT an April 21, 2012 'C' MSIV shut after the instrument air supply was isolated.  "The cause of the equipment failure is not yet known but is currently being investigated.  This report is being made in accordance with 10 CFR 50.72 (b)(3)(v)(C), inability to isolate and mitigate a radioactive release, and 10 CFR 50.72 (b)(3)(v)(D), a condition which could have prevented the fulfillment of a safety function to mitigate the consequences of an accident. The plant continues to remain shutdown at 0% power.  Other equipment functioned as expected including the turbine isolation valves."  The licensee notified the NRC Resident Inspector.  * * * UPDATE FROM JOHN CAVES TO HOWIE CROUCH AT 1258 EDT ON 4/24/12 * * *  "Investigation into the condition revealed the instrument air supply to the MSIVs was isolated at 0530 EDT and 'B' MSIV indicated [drifted] shut at 0607 EDT. The plant is currently in Mode 6, Refueling, and the investigation is ongoing."  The licensee notified the NRC Resident Inspector. Notified R2DO (Musser).|
Power Reactor|47858|DIABLO CANYON|PACIFIC GAS & ELECTRIC CO.|4|AVILA BEACH|CA|SAN LUIS OBISPO||Y|05000275|1|2||[1] W-4-LP,[2] W-4-LP|DAN STERMER|PETE SNYDER|4/22/2012 00:00:00|02:50|4/21/2012 00:00:00|16:15|PDT|4/25/2012 00:00:00|NON EMERGENCY||INFORMATION ONLY|||||||NEIL OKEEFE|R4DO|||||||||||||||||||N|Y|82|Power Operation|20|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||APPARENT PHASE IMBALANCE BETWEEN NORMAL AND ALTERNATE POWER CIRCUITS  "On April 21, 2012, at 1615 PDT, Diablo Canyon Power Plant (DCPP) operators identified that an attempted manual transfer of the Unit 1 AC power sources from the normal offsite circuit (500 kv) to the alternate offsite circuit (230 kv) failed.  "The attempted manual transfer of the offsite power source from the normal 500 kv to the alternate 230 kv was a planned activity  to shut down Unit 1 for its 17th refueling outage.  The failure to manually transfer was concluded to be caused by an indicated out of phase condition between the 500 kv and 230 kv system.  DCPP control transfer systems prevent paralleling and transferring the power supplies from the normal to alternate offsite circuit with an out of phase condition.  The offsite AC power sources remain capable of being manually transferred from the normal to the alternate source by way of transferring to diesels as an intermediate step.  "The 230 kv system ability to perform its accident mitigation safety function following a reactor trip or safety injection was not affected by the inability to perform a direct manual transfer from the 500 kv offsite power system.  "DCPP staff are currently troubleshooting the out of phase indication and will provide an update when the cause is concluded.  "The licensee notified the NRC Resident Inspector."  * * * UPDATE FROM MIKE QUITTER TO HOWIE CROUCH AT 2013 EDT ON 4/25/12 * * *  "This is an update to EN #47858 reported on April 22, 2012, where Pacific Gas & Electric Company (PG&E) reported that an attempt to manually transfer the DCPP Unit 1 AC power sources from the normal offsite circuit (500 kV) to the alternate required offsite circuit (230 kV) failed.  "PG&E determined that the out-of-phase condition between the 500kV system and the 230 kV system resulted from the offsite power system configurations and conditions existing at the time of the attempted transfer.  PG&E established alternate operating guidance to allow operators to perform bus transfers between the 500kV source and the 230 kV source in the event that an out-of-phase condition exists when bus transfer is desired.   "PG&E personnel informed the NRC Resident Inspector."  Notified R4DO (Proulx).|
Power Reactor|47859|HATCH|SOUTHERN NUCLEAR OPERATING COMPANY|2|BAXLEY|GA|APPLING||Y|05000321|1|||[1] GE-4,[2] GE-4|DANIEL KOMM|PETE SNYDER|4/24/2012 00:00:00|07:52|2/25/2012 00:00:00|01:38|EDT|4/24/2012 00:00:00|NON EMERGENCY|50.73(a)(1)|INVALID SPECIF SYSTEM ACTUATION|||||||RANDY MUSSER|R2DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||INVALID ACTUATION OF STANDBY GAS TREATMENT SECONDARY CONTAINMENT ISOLATION VALVES (SCIVs)  "During restoration of Unit 1 reactor protection system (RPS) following preventive maintenance (PM), a link was closed on February 25, 2012 at 0138 EST that resulted in the automatic actuation of the standby gas treatment (SGT) trains for Units 1 and 2 and the automatic isolation of the associated SCIVs (Secondary Containment Isolation Valves). The instruments that input to the actuation logic were in the tripped condition during the performance of the PM with links opened to preclude the automatic response of the SCIVs. The PM procedure did not contain the necessary steps to ensure that the associated instruments that input into the actuation logic were reset prior to reclosing the links that were previously opened to disable their input to the logic. Reclosing the first link associated with a Unit 1 reactor building high radiation monitor trip resulted in an actuation signal that caused the Unit 1 and 2 SGT trains to automatically start as designed, Unit 1 and 2 reactor building normal ventilation to shut down and the Unit 1 and 2 SCIVs to automatically close. This actuation was therefore not the result of a valid signal. The automatic actuation of the SGT system and the isolation of Unit 1 and 2 SCIVs are considered an invalid actuation since the parameters that cause this actuation to occur had not been exceeded. For this reason the actuation is considered invalid and a report to the NRC is not required by 10CFR50.72(b)(3)(iv); however, because the secondary containment isolation signals affected containment isolation valves in more than one system (Unit 1 and 2 components affected) the event is reportable as required by 10CFR50.73(a)(2)(iv)(B)(2). A licensee event report (LER) is required, but a telephone notification is allowed by 10CFR50.73. In the case of an invalid actuation reported under 10CFR50.73(a)(2)(iv), other than actuation of the reactor protection system (RPS) when the reactor is critical, the licensee may, at its option, provide a telephone notification to the NRC Operations Center within 60 days after discovery of the event instead of submitting a written LER. The affected procedure(s) will be revised to ensure the affected instruments are reset prior to returning the system to service to preclude recurrence during the performance of future RPS PM activities.  "The four Standby Gas Treatment (SBGT) fans auto started and both Unit 1 and Unit 2 reactor building and refueling floor normal ventilation systems automatically shutdown and isolated as designed. The SBGT initiation and the ventilation system shutdown were both complete actuations."  The licensee notified the NRC Resident Inspector.|
Fuel Cycle Facility|47861|WESTINGHOUSE ELECTRIC CORPORATION|WESTINGHOUSE ELECTRIC CORPORATION|2|COLUMBIA|SC|RICHLAND|SNM-1107|Y|07001151||||URANIUM FUEL FABRICATION|GERARD COUTURE|HOWIE CROUCH|4/24/2012 00:00:00|13:48|4/23/2012 00:00:00|22:00|EDT|4/24/2012 00:00:00|NON EMERGENCY|70.50(b)(3)|MED TREAT INVOLVING CONTAM|||||||RANDY MUSSER|R2DO|GORDON BJORKMAN|NMSS|||||||||||||||||N|N|0||0||N|N|0||0||N|N|0||0||CONTAMINATED INDIVIDUAL REQUIRED MEDICAL TREATMENT  "Westinghouse Environmental Health and Safety (EH&S) staff responded to an operator who was exposed to dilute nitric acid (30%) on the left forearm and left foot.  Employee was cleaning scrubber piping in the conversion area of the plant when exposed to the nitric acid solution containing uranium.  The employee was treated within the onsite medical facility where decontamination was performed.  Medical, emergency response and health physics procedures were followed.  Upon completion of the decontamination efforts, the smearable alpha reading was < 50 dpm/100 cm2 and alpha direct reading on forearm was 818 dpm/100 cm2, on foot direct alpha reading was 280 dpm/100 cm2. Medical staff recommended transfer to the hospital for further treatment.  Westinghouse followed contaminated injury protocols and had the employee transported to the hospital emergency room via ambulance.  A Westinghouse health physics (HP) technician accompanied the employee to the hospital.  Surveys were conducted of the ambulance and all results were below established limits. Material Safety Data Sheets for the nitric acid and uranium were provided to the hospital in accordance with procedures. Hospital report describes injury as, 'irritation noted over the left forearm as well as over the left anterior part of the foot and dorsal part of the foot ...some mild orange discoloration noted to these areas.' Employee was monitored for a period of time, given treatment for pain, and then released from the hospital.  "Immediate Actions: Initial investigation into the event is ongoing. This event has been entered into the Facility Corrective Action Process Issue # 12-115-001. Local county authorities and state authorities are aware of this event."|
Power Reactor|47862|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|ROBERT KROS|HOWIE CROUCH|4/25/2012 00:00:00|17:22|4/25/2012 00:00:00|09:00|CDT|4/25/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(B)|UNANALYZED CONDITION|||||||DAVID PROULX|R4DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||UNANALYZED CONDITIONS RESULTING FROM A NON-CONSERVATIVE ERROR IN CALCULATIONS  "A non-conservative error was identified in the Proto-Flo input calculation FC06644 for LPSI (Low Pressure Safety Injection) flow post-RAS [recirculation actuation signal].  The calculation used an incorrect (non-conservative) input for LPSI pump performance.  Also, the associated procedure (EOP/AOP Attachment 11) as written does not provide adequate direction during the Alternate Hot Leg Injection mode of operation.      "EOP/AOP Attachment 11 (Alternate Hot Leg Injection) used 140 psia as the entry point.  The LPSI pumps may not be able to meet minimum flow requirements at this pressure, affecting core cooling and possibly resulting in pump damage.  Also the EOP/AOP attachment directs the operator to verify that flow is approximately 400 gpm as indicated on FIC-326. If 400 gpm cannot be achieved the contingency is to open any LPSI loop injection isolation valve.  This step would not depressurize the RCS low enough to allow the 400 gpm flow rate to be achieved which would cause insufficient flow.   "Therefore, it is reasonable to conclude that the referenced procedural guidance may not be able to complete the safety function of providing adequate core cooling during the Alternate Hot Leg Injection mode of operation under a worst case scenario.  "Therefore, this condition is an unanalyzed condition and reportable under 10CFR50.72(b)(3)(ii)(B)."  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47863|POINT BEACH|NUCLEAR MANAGEMENT COMPANY|3|TWO RIVERS|WI|MANITOWOC||N|05000266|1|2||[1] W-2-LP,[2] W-2-LP|ALEX RIVAS|VINCE KLCO|4/25/2012 00:00:00|22:42|4/25/2012 00:00:00|21:06|CDT|4/26/2012 00:00:00|ALERT|50.72(a) (1) (i)|EMERGENCY DECLARED|||||||JULIO LARA|R3DO|JOHN LUBINSKI|NRR|CYNTHIA PEDERSON|R3 R|BRUCE BOGER|NRR|JASON KOZAL|IRD|ELLIOT BRENNER|PAO|||||||||N|Y|100|Power Operation|100|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||ALERT DECLARED DUE TO TOXIC GAS BUILD UP FOLLOWING AN EMERGENCY DIESEL GENERATOR TEST RUN  On April 25, 2012 at 2106 CDT, Point Beach Nuclear Plant declared an Alert under EAL HA3.1 due to toxic gas in a vital area. During a maintenance run on the plant's emergency diesel generator, diesel exhaust caused the concentration of carbon monoxide in an adjacent vital area room (Plant Instrument Air Compressor Room) to exceed OSHA IDLH (Immediately Dangerous to Life and Health) levels. The diesel was immediately secured and the room was ventilated. There were no personnel injuries and no public health and safety issues associated with this event.   The licensee notified the NRC Resident Inspector.  Notified DHS SWO, FEMA, USDA, DOE, DHS NICC, HHS AND EPA.  * * * UPDATE FROM JOHN RODGERS TO JOHN SHOEMAKER AT 0022 EDT ON 04/26/12 * * *   The Alert at Point Beach was terminated at 2314 CDT on 04/25/12.    The licensee notified the NRC Resident Inspector.  Notified R3DO (Lara), NRR EO (Lubinski), IRD (Scott via email only).  Notified DHS SWO, FEMA, USDA, DOE, DHS NICC, HHS, EPA, and Nuclear SSA (email only).|
Power Reactor|47864|COOK|INDIANA/MICHIGAN POWER CO.|3|BRIDGMAN|MI|BERRIEN||N|||2||[1] W-4-LP,[2] W-4-LP|BUD HINCKLEY|JOHN SHOEMAKER|4/26/2012 00:00:00|01:21|4/25/2012 00:00:00|22:46|EDT|4/26/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||JULIO LARA|R3DO|||||||||||||||||||N|N|0||0||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||NOTIFICATION TO OFFSITE AGENCIES DUE TO OIL SPILL  "At 2246 EDT on April 25, 2012, D.C. Cook notified the State of Michigan and local authorities of an oil spill from the Unit 2 [Main Generator] Seal Oil system at 2110 EDT which resulted in a portion of this oil entering the absorption pond, which is in the owner controlled area.  An oil sheen is present on the absorption pond which has a surface area of approximately 2 acres (87,000 square feet) and it is estimated that five to ten gallons of oil are on the absorption pond.  "The NRC Resident Inspector was notified."  The licensee has contained the oil leak and clean up efforts are ongoing.|
Power Reactor|47865|DIABLO CANYON|PACIFIC GAS & ELECTRIC CO.|4|AVILA BEACH|CA|SAN LUIS OBISPO||Y|||2||[1] W-4-LP,[2] W-4-LP|DAVE GOUVEIA|CHARLES TEAL|4/26/2012 00:00:00|04:20|4/25/2012 00:00:00|22:27|PDT|4/26/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||DAVID PROULX|R4DO|||||||||||||||||||N|N|0||0||N|Y|25|Power Operation|0|Hot Standby|N|N|0||0||PRESS RELEASE ISSUED FOR SHUTDOWN OF UNIT 2 DUE TO SALP INFLUX  "On April 25, 2012, at 2227 PDT, Diablo Canyon Power Plant (DCPP) U2 was manually taken offline and the reactor shut down due to an influx of sea salp - a small, jellyfish-like organism - in the intake structure.  Operators manually started the auxiliary feedwater pumps in accordance with the plant operating procedures.   "PG&E will not restart Unit 2 until conditions improve at the intake structure.  As previously announced, Unit 1 was safely shut down for a planned refueling and maintenance outage.  "All systems operated as designed and no unexpected equipment performance issues were noted.  "The licensee notified the NRC Resident Inspector."  The licensee plans on issuing a press release.|
Power Reactor|47866|SURRY|DOMINION GENERATION|2|SURRY|VA|SURRY||N|05000280|1|2||[1] W-3-LP,[2] W-3-LP|BRET RICKERT|VINCE KLCO|4/26/2012 00:00:00|13:50|4/26/2012 00:00:00|09:55|EDT|4/26/2012 00:00:00|NON EMERGENCY|50.72(b)(2)(xi)|OFFSITE NOTIFICATION|||||||RANDY MUSSER|R2DO|||||||||||||||||||N|Y|94|Power Operation|94|Power Operation|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||OFFSITE NOTIFICATION DUE TO DEATH OF A BALD EAGLE  "At 0955 [EDT] on April 26, 2012, a bald eagle (bird of prey) caused a phase to phase fault on the Station's 428 power line (power to buildings outside the protected area). The fault temporarily de-energized the 428 line. The bald eagle was found dead beneath the power lines.  "Two outside agencies, The Virginia Department of Game and Inland Fisheries and The U.S. Fish and Wildlife Service will be notified.  "This is being reported pursuant to 10 CFR 50.72(b)(2)(xi) for an event that required notification of other government agencies.  "The NRC Resident has been informed."|
Power Reactor|47868|FERMI|DETROIT EDISON CO.|3|NEWPORT|MI|MONROE||N|05000341|2|||[2] GE-4|GEORGE PICCARD|HOWIE CROUCH|4/26/2012 00:00:00|14:55|4/26/2012 00:00:00|10:12|EDT|4/26/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(iv)(A)|VALID SPECIF SYS ACTUATION|||||||JULIO LARA|R3DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||VALID ACTUATION OF THE REACTOR PROTECTION SYSTEM DURING TESTING  "At 1012 EDT on April 26, 2012, during the Reactor Pressure Vessel Hydrostatic Test, a valid high pressure reactor scram occurred due to issues related to controlling pressure near rated values. This actuation of the Reactor Protection System was not part of the pre-planned testing sequence. All control rods were fully inserted at the time of the scram. This report is being made in accordance with 10CFR50.72 (b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' The reactor scram was reset after reactor pressure was lowered."  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47869|TURKEY POINT|FLORIDA POWER & LIGHT CO.|2|MIAMI|FL|DADE||Y|05000250|3|4||[3] W-3-LP,[4] W-3-LP|ADRIAN GONZALEZ|VINCE KLCO|4/27/2012 00:00:00|13:19|4/27/2012 00:00:00|08:00|EDT|4/27/2012 00:00:00|NON EMERGENCY|26.719|FITNESS FOR DUTY|||||||DAVID PROULX|R4DO|||||||||||||||||||N|N|0|Refueling|0|Refueling|N|Y|100|Power Operation|100|Power Operation|N|N|0||0||FITNESS FOR DUTY - NON-LICENSED CONTRACT SUPERVISOR TESTED POSITIVE FOR DRUGS  A non-licensee contract supervisor tested positive for illegal drugs during a random fitness-for-duty test.  The employee's access to the plant has been terminated.  Contact the Headquarters Operations Officer for additional details.  The licensee notified the NRC Resident Inspector.|
Power Reactor|47870|FORT CALHOUN|OMAHA PUBLIC POWER DISTRICT|4|FORT CALHOUN|NE|WASHINGTON||Y|05000285|1|||(1) CE|ROBERT KROS|HOWIE CROUCH|4/27/2012 00:00:00|16:02|4/27/2012 00:00:00|13:45|CDT|4/27/2012 00:00:00|NON EMERGENCY|26.719|FITNESS FOR DUTY|||||||DAVID PROULX|R4DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||FITNESS FOR DUTY - NON-LICENSED SUPERVISOR TESTED POSITIVE FOR ILLEGAL DRUGS  A non-licensed supervisory employee was determined to be under the influence of illegal drugs during a random test. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details.  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47871|KEWAUNEE|NUCLEAR MANAGEMENT COMPANY|3|KEWAUNEE|WI|KEWAUNEE||Y|05000305|1|||[1] W-2-LP|LOGAN MILLER|HOWIE CROUCH|4/27/2012 00:00:00|22:23|4/27/2012 00:00:00|17:42|CDT|4/27/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(B)|POT RHR INOP|||||||JULIO LARA|R3DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||BOTH TRAINS OF RESIDUAL HEAT REMOVAL DECLARED INOPERABLE  "At 1742 EDT on 04/27/2012, while in Mode 5, both trains of RHR (Residual Heat Removal) were declared inoperable due to a through wall leak on a socket welded connection of ASME code class piping. Currently, with both trains of RHR in service for decay heat removal, the leakage impacts redundant equipment required to fulfill a safety function. In the current condition, both trains are required to be operable to meet Technical Specification [TS] 3.4.7, 'RCS Loops - MODE 5, Loops Filled'.  "The leakage is isolable from the reactor coolant system and is therefore not considered RCS pressure boundary leakage per LCO 3.4.13, 'RCS Operational Leakage'.  "This event is being reported under 50.72(b)(3)(v)(B), 'Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) remove residual heat'.  "Action has been initiated in accordance with TS 3.4.7 required action C.2 to restore one RHR loop to operable status."  The licensee is performing engineering analysis to determine the most appropriate repair method.  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47872|DIABLO CANYON|PACIFIC GAS & ELECTRIC CO.|4|AVILA BEACH|CA|SAN LUIS OBISPO||Y|05000275|1|||[1] W-4-LP,[2] W-4-LP|DENNIS PETERSEN|DONALD NORWOOD|4/29/2012 00:00:00|17:35|4/29/2012 00:00:00||PDT|4/29/2012 00:00:00|NON EMERGENCY|21.21(d)(3)(i)|DEFECTS AND NONCOMPLIANCE|||||||DAVID PROULX|R4DO|PART 21 GRP BY EMAIL||||||||||||||||||N|N|0|Refueling|0|Refueling|N|N|0||0||N|N|0||0||PART 21 REPORT - CONTROLLERS DID NOT MEET EMI / RFI GUIDELINES  "This report constitutes a 10 CFR Part 21 notification.  On March 23, 2012, Scientech (a business unit of Curtis Wright Flow Control), sent a 10 CFR Part 21 report (Letter Number 12-09-BB) to Pacific Gas & Electric Company.  Diablo Canyon Nuclear Plant (DCPP) entered this report into its corrective action program on April 23, 2012.  This Part 21 report identified that eight AMS286 and AMS287 hand controllers supplied to DCPP had a deviation from the purchase specification.  DCPP specified the controllers to be qualified to meet electromagnetic interference (EMI) / radio frequency interference (RFI) emissions and susceptibility in accordance with NRC Regulatory Guide 1.180, Rev. 1.  Contrary to this requirement, Scientech did not perform the EMI / RFI testing on the AMS286 controllers shipped to DCPP with a Certificate of Conformance to DCPP's specification.  Scientech tested the AMS287 controllers, which failed some of the required testing, but were shipped to DCPP with a Certificate of Conformance nonetheless.  DCPP accepted the controllers based on the Certificates of Conformance, but did not install them.  Following acceptance of the controllers, DCPP identified the nonconformance with its specification, and had another vendor modify the controllers to subsequently pass required testing.   "DCPP determined the deviation could have potentially created a substantial safety hazard if the controllers were installed at DCPP.  The controllers were intended to control auxiliary feedwater from the Unit 1 control room and the hot shutdown panel as part of an upcoming process control system modification.  Following a postulated fire in the control room, plant shutdown would be managed from the hot shutdown panel.  In this instance, radios would be used to communicate with operators at the panel.  The controllers would be susceptible to EMI/RFI interference and could potentially affect the operators' ability the safely shut down the plant and maintain it in a safe shutdown condition.  DCPP notified Scientech of this conclusion on 4/27/12."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47873|KEWAUNEE|NUCLEAR MANAGEMENT COMPANY|3|KEWAUNEE|WI|KEWAUNEE||Y|05000305|1|||[1] W-2-LP|CRAIG NEUSER|JOHN KNOKE|4/30/2012 00:00:00|10:55|4/30/2012 00:00:00|06:00|CDT|4/30/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(v)(B)|POT RHR INOP|||||||JULIO LARA|R3DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||BOTH TRAINS OF RESIDUAL HEAT REMOVAL SYSTEM DECLARED INOPERABLE  "At 1742 CDT on 04/27/2012, while in Mode 5, both trains of Residual Heat Removal were declared inoperable due to a through wall leak in a 3/4-inch pipe socket weld connection. The leak developed in an ASME Section XI, Code Class 2 weld upstream of a sample isolation valve. The leak is not isolable from the common 10-inch Residual Heat Removal discharge piping. However, the leakage is isolable from the Reactor Coolant System and is therefore not considered RCS pressure boundary leakage per Technical Specification LCO 3.4.13, RCS Operational Leakage. Currently, with both trains of RHR in service for decay heat removal, the leakage impacts redundant equipment required to fulfill a safety function. In the current condition, both trains are required to be operable to meet Technical Specification LCO 3.4.7, RCS Loops - Mode 5, Loops Filled. This event was reported per EN #47871.  "At 0600 CDT on 04/30/2012 during the subsequent repair of the leaking weld a second leak of approximately 0.03 gallons per minute developed. The second leak occurred down stream of the original leak while welding a temporary clamp to the 3/4-inch sample line. The cause of the second leak is being directly attributed to the welding activity and not degradation of the pipe. The structural integrity of the pipe in the area of the second leak was verified to be acceptable. Both trains of Residual Heat Removal remain in service removing decay heat from the core. No other equipment is being affected by the leak. Repair options are currently being evaluated.   "The new leak is being reported under 50.72(b)(3)(v)(B), 'Any event that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (B) Remove residual heat.'  "Actions continue in accordance with Technical Specification LCO 3.4.7 Required Action C.2 to restores one RHR loop to operable status."  The licensee has notified the NRC Resident Inspector.|
Power Reactor|47874|SALEM|PSEG NUCLEAR LLC|1|HANCOCKS BRIDGE|NJ|SALEM||N|05000272|1|||[1] W-4-LP,[2] W-4-LP|SCOTT RIDDELL|CHARLES TEAL|4/30/2012 00:00:00|10:45|4/30/2012 00:00:00|10:18|EDT|4/30/2012 00:00:00|UNUSUAL EVENT|50.72(a) (1) (i)|EMERGENCY DECLARED|50.72(b)(2)(iv)(B)|RPS ACTUATION - CRITICAL|50.72(b)(2)(iv)(A)|ECCS INJECTION|50.72(b)(2)(i)|PLANT S/D REQD BY TS|RICHARD CONTE|R1DO|BILL DEAN|RA|DAN DORMAN|NRR|WILLIAM GOTT|IRD|||||||||||||N|Y|100|Power Operation|0|Hot Standby|N|N|0||0||N|N|0||0||UNUSUAL EVENT DUE TO A POTENTIAL FIRE IN CONTAINMENT  "Performing an I&C functional [test] caused an inadvertent Safety Injection signal resulting in a reactor trip/safety injection.  All safety systems responded as designed for a safety injection.  Electrical systems are aligned to normal offsite power sources.  All fire alarms have been validated by the Fire Protection Department as invalid alarms and confirmed that no fire event in the protected area.  "The reactor trip was successful and all rods [fully] inserted.  Decay heat removal is via auxiliary feedwater through the atmospheric [steam] dumps.  Unknown at this time is the cause of the inadvertent safety injection signal.  No injuries occurred as a result of this event."  The licensee believes that the trip/safety injection may have caused piping to shake resulting in dust near the fire detection equipment resulting in the invalid fire indication.  The instrument being tested was the high steam flow channel-1 bistable for PT505.  The maximum pressurizer level during this event was 95%.    The licensee notified the NRC Resident Inspector.  * * * UPDATE AT 1400 EDT ON 4/30/2012 FROM JOHN KOKOVALCHICK TO MARK ABRAMOVITZ * * *  "At 1003 hours on April 30, 2012, Salem Unit 1 experienced a reactor trip and safety injection (SI) signal due to a high steam flow coincident with a low steam pressure signal.  At the time of the safety injection signal, function testing of the 1PT505 turbine inlet pressure channel was in progress. This testing required the tripping of the high steam flow bistables.  "As a result of the reactor trip and safety injection signal, the Emergency Diesel Generators started but did not load, the ECCS system (high head safety injection pumps actuated and injected into the reactor vessel, intermediate head safety injection pumps and low head (RHR) safety injection pumps) actuated.  All 4 main steam isolation valves closed along with feedwater isolation and start of the auxiliary feedwater pumps.  All control rods fully inserted following the reactor trip.  Following the main steam line isolation, the atmospheric relief valves opened along with the lifting of several main steam safety valves.  "The unit is currently in Mode 3 and will be cooling down to Mode 4.  Train A SSPS [Solid State Protection System] is currently out of service and suspected of causing the safety injection signal.  Train B SSPS has not been reset due to the standing safety injection signal.  With Train A SSPS inoperable and Train B SSPS not reset, TS 3.0.3 was entered and a shutdown required by TS 3.0.3 was commenced at 1345 hours.  "This report is being made in accordance with 10CFR50.72(b )(2)(iv)(B), 50.72(b)(3)(iv)(A), 50.72(b)(2)(i) and 50.72(b)(2)(iv)(A)."  The licensee exited the Unusual Event at 1249 EDT.   The licensee notified the NRC Resident Inspector.  Notified the R1DO (Conte).  The NRC Operations Center notified other Federal Agencies (DHS SWO, FEMA Ops, DHS NICC, and NuclearSSA via e-mail).|
Power Reactor|47875|QUAD CITIES|EXELON NUCLEAR CO.|3|CORDOVA|IL|ROCK ISLAND||Y|05000254|1|||[1] GE-3,[2] GE-3|JAMES COX|MARK ABRAMOVITZ|4/30/2012 00:00:00|13:38|3/24/2012 00:00:00|19:36|CDT|4/30/2012 00:00:00|NON EMERGENCY|50.73(a)(1)|INVALID SPECIF SYSTEM ACTUATION|||||||MARK RING|R3DO|||||||||||||||||||N|Y|100|Power Operation|90|Power Operation|N|N|0||0||N|N|0||0||INVALID PROTECTION SYSTEM ACTUATION DURING STATION ELECTRICAL TRANSIENT  "The purpose of this report is to provide a telephone notification for an invalid actuation. On March 24, 2012, following the completion of switch yard work, the Control Room received switching orders to open Bus Tie 9-10 Bus 9 disconnect and close Bus Tie 9-10 Bus 10 disconnect. Operations was unaware that a grounding device, installed for personnel protection during the work activities, had not been removed. Consequently, when the Bus Tie 9-10 Bus 10 disconnect was closed the switchyard was grounded resulting in an electrical transient. Protective relaying operated as designed to clear the fault, and there were no injuries.  "During the electrical transient, the voltage depression tripped the Unit 1 "A" Reactor Protection System (RPS) bus which caused a 1/2 scram and certain protective logic systems to de-energize by design. The following invalid actuations occurred as a result of the loss of power to the RPS bus: partial Group 11 Isolation (Primary Containment); Group III Isolation (Reactor Water Cleanup); Reactor Building Ventilation Isolation; Control Room Ventilation Isolation; and Standby Gas Treatment Initiation.  "The electrical transient also tripped the Unit 1 ECCS keepĀfill pump, resulting in the Core Spray (CS) discharge pressure decreasing to the alarm setpoint. Both CS subsystems were conservatively declared inoperable and entry into Technical Specifications (TS) 3.0.3 occurred at 1936 hours. Subsequent fill and vent activities confirmed no air existed in the discharge headers of the CS subsystems (no loss of safety function) and both subsystems were declared operable with TS 3.0.3 being exited at 2017 hours. Following the electrical transient, the Unit 1 generator was temporarily limited to approximately 90% load due to elevated vibration on Turbine Bearing No. 10. In-plant walk-downs identified no other equipment concerns. Unit 1 returned to full power on April 2, 2012, following confirmation the bearing vibration is acceptable for long-term operation. Unit 2 was in a scheduled refueling outage during the event and was unaffected by the electrical transient. A Root Cause Investigation is ongoing."  The licensee notified the NRC Resident Inspector.|
Power Reactor|47880|GRAND GULF|ENTERGY NUCLEAR|4|PORT GIBSON|MS|CLAIBORNE||Y|05000416|1|||[1] GE-6|LEROY PURDY|DONALD NORWOOD|4/30/2012 00:00:00|22:34|4/30/2012 00:00:00|18:01|CDT|4/30/2012 00:00:00|NON EMERGENCY|50.72(b)(3)(ii)(A)|DEGRADED CONDITION|||||||GREG WERNER|R4DO|||||||||||||||||||N|N|0|Cold Shutdown|0|Cold Shutdown|N|N|0||0||N|N|0||0||WELD DEFECT INDICATION FOUND IN RESIDUAL HEAT REMOVAL SYSTEM TO REACTOR PRESSURE VESSEL NOZZLE  "Grand Gulf Nuclear Station is currently in Mode 4 (less than 200 degrees F) executing Refueling Outage 18 (RF-18) including in-service inspections.  "General Electric notified Entergy of a weld indication that was detected by automated ultrasonic testing.  The indication is in the weld root area of N06B-KB Reactor Coolant System Pressure Boundary weld.  The N06B nozzle connects Residual Heat Removal System 'C' to the Reactor Pressure Vessel.  The dimension of the indication is approximately 0.9 inches in length, approximately 0.5 inches in depth and with no discernible width.  Nominal wall thickness is 1.3 inches.  "The indication does not penetrate the entire thickness of the pipe wall and there is no leakage at the indication.  There has been no release of radioactive material due to the indication.  No systems were actuated due to this event.  There are currently no other systems affected.  The cause is under investigation and corrective action plans are being explored.  "The weld defect has been evaluated by Entergy Engineering and determined to meet the criteria for reporting identified in NUREG-1022: Welding or material defects in the primary coolant system that cannot be found acceptable under ASME Section XI, IWB-3600, 'Analytical Evaluation of Flaws,' or ASME Section XI, Table IWB-3410-1, 'Acceptable Standards'."  The NRC Resident Inspector has been informed.|