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ENS 53719 | 7 November 2018 17:28:00 | The following information was received from the Commonwealth of Kentucky by email: On 11/6/2018, a former licensee (formerly licensed as Wickliffe Paper Co.) reported discovery of a nuclear gauging device (TN (Texas Nuclear) model 5036 originally containing 200 mCi assayed 12/94) that it was unaware it possessed. The license was terminated on August 9, 2016 and at that time, the former licensee provided information related to the disposition of all devices the licensee was aware it possessed. License termination was due to plant closure. During engineering surveys to assess plant conditions for restart, personnel discovered the device still mounted on plant equipment. The former licensee is taking steps to have the device transferred to a licensed manufacturer for disposal. There is no reason to believe any individuals received any exposure at levels which would exceed the regulatory limits." Kentucky Event: KY180004 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 |
ENS 53716 | 5 November 2018 11:26:00 | The following information was received via email: When conducting the annual reconciliation, Karcher North America, INC. reported eleven lost static eliminators. Static Eliminators: Model: P-2021 8101. Isotope/units: PO-210, 10 mCi ea. Serial Numbers: A2JZ217, A2KH719, A2CP799, A2DM543, A2DT589, A2DU443, A2DU444, A2EZ668, A2GS233, A2JD061, A2JD062. Colorado Event Report ID No.: CO180027 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 |
ENS 53720 | 7 November 2018 15:44:00 | The following information was received by the State of Florida: At noon (on 11/7/18), (Akumin) called (the State of FL Bureau of Radiation Control) to report that both Akumin Hollywood and Akumin Aventura View ordered F-18 Fluciclovine, and received packages that were labeled as F-18 Fluciclovine, but were subsequently notified by their radiopharmaceutical vendor PET NET Solutions-Ft Lauderdale, on Thursday, November 1, 2018 that due to a 'batch error,' the packages actually contained F-18 FDG (Fludeoxyglucose). Three patients were reported as receiving the incorrect radiopharmaceutical. Activity reported as approximately 10 mCi. Florida Incident: FL18-137 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. |
ENS 53705 | 30 October 2018 16:33:00 | Replacement camera sources were properly delivered by the common carrier to St. Mary's Hospital located in Jefferson City, Missouri. The sources were received by the Biomed Hospital Imaging Specialist and placed in the biomed office. Currently, the package containing the sources is missing. The licensee investigation continues. Sources are two Gd-153 (10 mCi each) and two Co-57 (0.5 microCi each).
This event is being retracted. The sources were discovered to be delivered to SSM Hospital's biomed office instead of nuclear medicine. The sources were secured and in control of the SSM Hospital at all times. There were no exposures to personnel. The licensee notified NRC Region 3 (Warren). Notified R3DO (Stoedter) and NMSS Events Notification and ILTAB via email. 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 |
ENS 53700 | 28 October 2018 21:44:00 | This event is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) for a major loss of emergency assessment capability at the Prairie Island Nuclear Generating Plant. At 1435 CDT on October 28, 2018, troubleshooting of the Seismic Monitoring Panel resulting from the receipt of Control Room annunciator 47023-0603 (Seismic Monitor Panel) determined that the '(Operational Basis Earthquake) OBE Exceedance' alarm on the Seismic Monitoring Panel will not alarm and determined the panel is non-functional. The Seismic Monitoring Panel system functions to provide indication that the OBE threshold has been exceeded following a seismic event and is used in the Prairie Island Nuclear Generating Plant Emergency Plan to perform classification of Initiating Condition 'Seismic event greater than OBE levels' and Emergency Action Level HU2.1. Station personnel are monitoring the seismic recorders for event alarms on a 15 minute frequency due to alarm function failure. The station is developing repair plans for restoration of the alarm function. This event does not adversely affect the safe operation of the plant or health and safety of the public. The licensee has notified the NRC Resident Inspector. |
ENS 53695 | 26 October 2018 12:11:00 | A patient was prescribed 200 mCi of Lutetium-177. Due to dose administration issues, a delivered dose of 135 mCi was received by the patient. The licensee notified the NRC Region 3 contact (Gattone). 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. |
ENS 53646 | 5 October 2018 09:52:00 | EN Revision Text: MAIN STEAM ISOLATION VALVES EXCEEDED PRIMARY CONTAINMENT LOCAL LEAK RATE ACCEPTANCE CRITERIA At 0520 (CDT), on October 05, 2018, it was discovered that a Primary Containment local leak rate test performed on Main Steam Isolation Valves (MSIV) exceeded its acceptance criteria. During Mode 1, 2, and 3, Surveillance Requirement 3.6.1.3.10 requires MSIV leakage for a single MSIV line to be less than or equal to 106 standard cubic feet per hour (scfh) when tested at 29 psig and Surveillance Requirement 3.6.1.3.12 requires the combined leakage rate for all MSIV leakage paths to be less than or equal to 212 scfh when tested at 29 psig. As-found for the 'C' MSIV line leakage results were unquantifiable and gave a (minimum) path value greeter than 160 scfh. This leakage rate lead to Surveillance Requirement 3.6.1.3.10 and 3.6.1.3.12 limits to be exceeded. This event is being reported as a condition of the nuclear power plant, including its principal safety barriers, being seriously degraded per 10 CFR 50.72(b)(3)(ii)(A) since the Primary Containment Isolation Valves leakage limits for MSIVs were exceeded. The NRC Resident Inspector has been notified.
CNS (Cooper Nuclear Station) is retracting the 8-hour non-emergency notification made on October 5, 2018 at 0520 CDT (EN# 53646). Subsequent evaluation concluded that overall as-found 'C' MSIV leakage rate was not at a level that exceeded the surveillance requirement 3.6.1.3.10 and 3.6.1.3.12 limits and thus the Primary Containment Isolation Valve leakage rate limits for the MSIVs were not exceeded. The NRC Senior Resident Inspector has been notified. Notified the R4DO (Drake). |
ENS 53643 | 4 October 2018 07:57:00 | EN Revision Text: MANUAL REACTOR TRIP DURING LOW POWER PHYSICS TESTING At 0544 EDT on October 4, 2018, with Unit 1 in Mode 2 with reactor power in the intermediate range performing low power physics testing, the reactor was manually tripped due to a rod control urgent failure alarm. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam system. Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods inserted as expected. The cause of the rod control urgent failure is being investigated.
This Event Notification is being updated to clarify that the reactor was not critical when this event occurred. Therefore, the reporting requirement is changed from 10 CFR 50.72(b)(2)(iv)(B) to 10 CFR 50.72 (b)(3)(iv)(A). During Dynamic Rod Worth Measurement testing, Control Bank Charlie was inserted approximately 153 steps when the urgent failure occurred (CBC positioned at 75 steps out). Following the scram, additional analysis concluded that the reactor was subcritical when the Reactor Protection System was actuated." The licensee notified the NRC Resident Inspector. Notified the R2DO (McCoy). |
ENS 53623 | 26 September 2018 15:10:00 | At 0946 CDT on 9/26/2018, a disruption in power to the offsite 138 kV line and the subsequent trip of the Emergency Reserve Auxiliary Transformer (ERAT) Static VAR Compensator (SVC) resulted in a degraded voltage signal on the Division 1- 4.16 kV safety bus. The degraded voltage signal resulted in a trip of the ERAT feed to the bus, blocking closure of the 345 kV Reserve Auxiliary Transformer (RAT) feed to the bus and auto start of the Division 1 Emergency Diesel Generator (EDG). The Division 1 EDG successfully started and re-energized the Division 1- 4.16 kV bus as designed. The unit is stable with the Division 1 EDG carrying the Division 1- 4.16 kV bus. The Ameren Transmission System Operator in St. Louis, MO informed the station that they had received a report that a 138 kV to 13.8 kV transformer at Clinton Route 54 substation was on fire and the South feed to the Tabor substation cycled as a result of this fault. The NRC Resident Inspector and Illinois Emergency Management Agency Resident Inspector have been notified. |
ENS 53626 | 26 September 2018 23:25:00 | On September 26, 2018 at 1908 CDT. an automatic scram was received on U1 following main generator 345 kV output breaker 7-8 trip with 345 kV output breaker 6-7 already opened for maintenance on line 0401. Following the reactor scram, reactor water level decreased to approximately minus 15 inches, which resulted in automatic Group II and Group Ill isolations (expected response). Reactor pressure rose to approximately 1083 psig, and the 3B and 3C low set relief valves opened briefly to control reactor pressure. Reactor water level and reactor pressure have been restored to their normal bands. All systems responded properly to the event. Unit 1 remains in Mode 3, with reactor pressure being controlled on the turbine bypass valves. The cause and details of the event are under investigation. Unit 2 was unaffected by the event and remains at 100% power. This report is being made in accordance with 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A)." All control rods inserted. Decay heat is being removed via the main condenser. The licensee notified the NRC Resident Inspector. |
ENS 53625 | 26 September 2018 21:43:00 | On 9/26/2018 at 1530 EDT, it was discovered that the HPCI system was inoperable due to a blown fuse in the 10C617 Panel, E21-F15A. Therefore, this condition Is being reported as an eight-hour, non-emergency notification per 10 CFR 50.72(b)(3)(v)(D). The blown fuse also impacts 'A' channel Residual Heat Removal (RHR) subsystem and 'A' Core Spray (CS) subsystem. These Emergency Core Cooling subsystems have been declared inoperable. Remaining Emergency Core Cooling subsystems and the Reactor Core Isolation Cooling (RCIC) system remain OPERABLE. There was no impact on the health and safety of the public or plant personnel." The licensee notified the NRC Resident Inspector and will notify the local authorities. |
ENS 53619 | 24 September 2018 14:06:00 | On September 22, 2018, at approximately 0050 (CDT), Duane Arnold Energy Center (DAEC) Security was contacted by a site assigned contractor that they had located what appeared to be drug paraphernalia inside the Protected Area. Local Law Enforcement was contacted and responded to DAEC. The Linn County Sheriff's office took the items into evidence for testing to determine if there was any presence of a controlled substance. On September 24, 2018, at 1013, the Linn County Sheriff's office notified DAEC that the items tested positive for the presence of a controlled substance. Therefore, this is being reported in accordance with 10 CFR 26.719. DAEC Site security is working with NextEra Corporate security regarding the investigation into this incident. The Resident Inspector has been notified. |
ENS 53669 | 16 October 2018 10:12:00 | The Clorox Company discovered a missing fixed gauge containing radioactive material. The gauge was a Filtec, model FT-2 containing 100 microCuries of Americium-241. Gauge S/N: 105382; Source S/N: 1786. 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 |
ENS 53609 | 15 September 2018 15:45:00 | EN Revision Text: UNUSUAL EVENT DUE TO SITE CONDITIONS PREVENTING PLANT ACCESS A hazardous event has resulted in on site conditions sufficient to prohibit the plant staff from accessing the site via personal vehicles due to flooding of local roads by Tropical Storm Florence. Notified DHS SWO, FEMA OPS, and DHS NICC. Notified FEMA NWC, NuclearSSA, and FEMA NRCC via email.
On 9/18/2018 at 1400 EDT, the Unusual Event at Brunswick was terminated due to the ability to transport personnel to the site. The licensee will notify the NRC Resident Inspectors. Notified the R2DO (Guthrie), NRR EO (Miller) and the IRD MOC (Grant). Notified DHS SWO, FEMA OPS, and DHS NICC. Notified FEMA NWC, NuclearSSA, and FEMA NRCC via email. |
ENS 53660 | 11 October 2018 09:39:00 | The following information was received by from ABB INC by facsimile: 1. This letter provides a notification of a defect associated with dry type transformer serial # 24-26458. The failure was caused by the breakdown of layer to layer insulation within the 4160 volt winding due to dielectric stress. Deterioration of the insulation resulted in an internal fault within the bravo phase 4160 volt winding, triggering a ground fault trip shutdown of equipment. This failure was reported by Exelon's Clinton Nuclear Station and it is the only known reported occurrence of safety related transformer failure caused by the breakdown of layer to layer insulation. Information is provided as specified in 10 CFR 21 paragraph 21.21(d)(4). 2. Notifying individual: Joey Chandler, Plant Manager, ABB ((PGTR) Power Grids Transformer Division, US), 171 Industry Drive, Bland, VA 24315. 3. Identification of the Subject component: ABB P/N 24-26458 dry type transformer. This transformer is used for stepping down voltage and was intended for providing power to safety related electrical equipment. 4. Nature of the deviation: The Exelon Clinton Nuclear Generating Station shut down due to a ground fault alarm on the 4160 volt side of the stepdown transformer that provides power to numerous safety-related components at the plant. Subsequent troubleshooting of the problem revealed that the dry-type transformer supplying 480 volt power had dielectrically failed due to apparent internal fault within the Bravo phase. Further investigation of this failure revealed an operational voltage design stress on the Nomex 410 insulation between the 4160 volt winding's layers of conductor of greater than recommended by the manufacturer (DuPont) for a 40 year design life. At the time of failure, the subject transformer had been in operation for approximately 33.5 years and had progressed 37 years and two months into its intended 40 year life given the 10/1980 ship date. ABB has no knowledge of any adverse operational variances over the course of the approximate 33.5 year life of operation to be able to assess or comment on this potential impact in terms of life. 5. The function of this dry type transformer is to step voltage down from 4160 volts to 480 volts while providing transfer of power to safety related components. Exelon's Clinton Nuclear Power Station has identified this transformer's power transfer to feed safety related equipment. An interruption of this transfer in power would result in a loss of power to the safety related equipment downstream and could potentially result in a compromise in safety. 6. ABB was notified of this transformer failure on 12/9/2017. This notification was delayed while the failure was being investigated. This investigation is documented in report: Exelon Clinton Failure Analysis_26458_011218 rev5.doc.pdf dated 09/10/2018. 7. Corrective actions include: a. Reviewed and verified current electrical engineering safety related design standard for allowable design stress on insulation per DuPont's recommendation for 40 year life. (Complete.) b. Reviewed the material used for transformer 24-26458. Found only affected safety related product to be isolated to Clinton Nuclear Station, though records may be incomplete as these records have been archived for over 35 years. (Complete.) c. Re-trained all involved personnel of the 10 CFR 21 reporting requirements, and the need to provide an interim report within 60 days of discovery. d. ABB worked directly with Clinton Nuclear to ensure all transformers of respective design was replaced with new transformers following ABB's Technical Evaluation for Nuclear 1E Transformer, Rev. 18 which documents operational design stresses be less than or equal to 30 volts / mil of Nomex 410 insulation between layer to layer of conductor for 40 year life. 8. Recommendation: Because of the possible existence of additional affected transformers, ABB (PGTR) cannot determine the potential for a substantial safety hazard exists at any other licensee's facility. Licensees are requested to evaluate any Gould-Brown Boveri/ITE dry type transformer with the following nameplate identification below. Transformers associated with this identification are recommended to be replaced. kVA: 750AA/ 1000 FA HV: 4160 Delta Connected LV: 480 Wye Connected Class: AA/ FA Type: Vent Frequency: 60 Hz Temp Rise: 80 degrees C Date of Manufacture: 10/1988 and older models Questions concerning this notification should be directed to the Quality Manager (Rick Kinder) at the ABB transformer plant in Bland, VA at (276) 688 -3325. |
ENS 53577 | 1 September 2018 10:43:00 | While filling up a licensee company truck at a gas station located in Ripley, West Virginia, an individual stole the company truck and its associated radiography camera (QSA 880; S/N 677846;108 Ci; Iridium-192 source). The licensee notified LLEA (West Virginia State Police) and the vehicle was recovered. The licensee inspected the properly secured equipment and observed no impact to the radiography camera. The licensee stated there was no radiological impact to the public or employees. THIS MATERIAL EVENT CONTAINS A "CATEGORY 2" LEVEL OF RADIOACTIVE MATERIAL Category 2 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 a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioactive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachytherapy. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
ENS 53549 | 13 August 2018 12:05:00 | The following information was received from the State of Texas via email: On August 12, 2018 at approximately 1135 (CDT)., the licensee notified the Agency (Texas Department of State Health Services) that one of its radiography crews had experienced a source disconnect. The event occurred on August 11, 2018, at approximately 1200 (CDT) at a temporary job site near Whitsett, TX. The device involved was an INC IR-100 (SN: 6792) containing a 91 curie iridium-192 source (SN: ZH0109). The crew had set up the device and performed a procedure shot and everything functioned properly. They performed the first shot of the job and the source would not retract into the device--it felt as though it had stuck on something. After a second unsuccessful attempt, the source was cranked back out into the collimator, boundaries set, and an authorized person came to the site and performed the retrieval. The drive cable and source were both new. There was no observable cause for the failure. The device and associated equipment will be sent to the manufacturer for evaluation. Per readings from all three individuals' self-reading pocket dosimeters, there were no overexposures. The source retriever's dosimetry badge is being sent for processing. An investigation into this event is ongoing. More information will be provided as it is obtained in accordance with SA-300.
The following update information was received from the State of Texas via email: Clarification: The source assembly (never disconnected) from the drive cable. (The licensee was) unable to retract it back into the exposure device. Notified R4DO (Deese) and NMSS Events Notification Group via email. Texas Incident: I-9606 |
ENS 53537 | 3 August 2018 14:10:00 | At 0940 EDT on August 3, 2018, the Division 2 Mechanical Draft Cooling Tower (MDCT) fans were declared inoperable due to failure of the over speed fan brake inverter. The brakes prevent fan over speed from a design basis tornado. The MDCT fans are required to support operability of the Ultimate Heat Sink (UHS). The UHS is required to support operability of the Division 2 Emergency Equipment Cooling Water (EECW) system. The EECW system cools various safety related components, including the High Pressure Coolant Injection (HPCI) system room cooler. An unplanned HPCI inoperability occurred based on a loss of the HPCI Room Cooler. Investigation into why the Division 2 MDCT fan over speed brake inverter failed is in progress. This report is being made pursuant to 10CFR50.72(b)(3)(v)(D) based on an unplanned HPCI inoperability. The NRC Resident Inspector has been notified. |
ENS 53535 | 2 August 2018 16:46:00 | The following information was received from the State of Louisiana via email: On 07/26/2018, (the) Radiation Safety Officer (RSO) for ExxonMobil Chemical Co. (ExMCo) reported a multi-source gauge failure to the Department (Louisiana Department of Environmental Quality), LDEQ by e-mail. On 07/25/2018 during routine annual maintenance and pm (preventative maintenance) checks it was discovered the level/density gauge had several shutters stuck in the open position. Three sources would not retract into the shielded position. However, the remaining four sources are functioning properly. The gauge is a Berthold Technologies USA multi-source device, Model LB 300 IS, utilizing AEA Technologies, Model CKC.P4 sources. There are seven (nominal) 50 mCi Co-60 sources in the device. The sources involved in this malfunction are source #1 s/n 1369-08-02, source #2 s/n 1370-08-02, and source #6 s/n 1374-08-02. All three sources will not retract into the shielded position. The device has a SS&D Registration # TN-1031-D-801-S. Only one device was manufactured and is no longer being manufactured. The manufacturer is Berthold Technologies GmbH & Co. KG, D-75323 Bad Wildbad Germany. The Berthold Model LB300 IS level density gauge is installed on G-Line High Pressure Reactor Vessel, V5300 and G-Line high pressure separator production line. ExMCo engineers and Flowmaster/Berthold engineers & service company have been contacted to fix the problem by repairing the source holders or replace the device with other comparable technology. Event type: The gauge is installed on processes and does not pose a health and safety threat to the general public or the ExMCo employees. The gauge will remain on the operational process until the repair is made to the device. This is considered an equipment failure for reporting requirements. Event Location: ExxonMobil Chemical Co. Baton Rouge Plastics Plant 11675 Scotland Avenue, (Hwy 19) Baton Rouge, LA 70807, Event description: Shutters stuck in the open position or difficult to operate shutters were detected on a level/density gauge installed on processes at ExMCo. A service company has been contacted to make the repair or replace the device. The Department will be provided a final report with corrective actions. The Department was notified and the incident was reported to the NRC Operation Center. The report to the NRC as required by 10 CFR Part 30.50 (b) (2) and required by LAC 33:XV.341.B.2.b. Louisiana Event: LA 180015 |
ENS 53522 | 24 July 2018 00:57:00 | 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 (Emergency Notification System) or under the reporting requirements of 10CFR50.73. This event is being reported pursuant to 10CFR50.72(b)(3)(xiii) as a Loss of Emergency Preparedness Capabilities at Palo Verde Nuclear Generating Station (PVNGS) Unit 2. On July 23, 2018, at approximately 1631 Mountain Standard Time (MST), the Unit 2 control room experienced an unplanned loss of Steam Generator #1 steam line monitor (RU-139), channels A and B. This monitor is used to assess dose projections for Main Steam line exhaust while in Modes 1-4 and is used in the PVNGS Emergency Plan to perform classification of Initiating Conditions 'RS1' and' RG1' and Emergency Action Levels (EALs) 'RS1.2' and 'RG1.2'. The PVNGS Emergency Plan does have two additional EALs that can be assessed for each Initiating Condition. The loss of this monitor constitutes a reportable loss of emergency assessment capability. The NRC Resident Inspector has been informed of this condition. |
ENS 53603 | 13 September 2018 14:38:00 | The following information was received by the State of Texas: On 09/13/2018, the Agency (Texas Department of State Health Services) was notified by a licensee that an employee received a personnel dosimetry report of 16.0 rem for the second quarter of 2018. The employee had received a report of a 3.3 rem exposure for the previous quarter. The licensee suspects that someone tampered with the dosimeter. The employee works in the nuclear medicine department and always leaves the badge attached to the lab coat on the door to the hot lab. Other employees performing similar work only received minimal exposures. Due to the amount of the reported exposure, the agency will conduct an investigation on site. Texas Incident: I-9613 |
ENS 53500 | 11 July 2018 03:58:00 | On July 11, 2018, as part of pre-planned maintenance, the site meteorological tower will be removed from service. The tower will be out of service for approximately 11 days. As a result, this is reportable under 10CFR 50.72 (b)(3)(xiii). During the time the data is not available from the meteorological tower; compensatory measures will be in place to obtain the data from the National Weather Service if necessary. The (NRC) Resident Inspector has been notified. |
ENS 53485 | 3 July 2018 19:07:00 | EN Revision Text: DISCOVERY OF AN UNANALYZED CONDITION THAT SIGNIFICANTLY DEGRADES PLANT SAFETY On July 3, 2018, while performing a review of Emergency Operating Procedures, a concern was identified regarding the potential for excessive loss of ultimate heat sink inventory (UHS) through the auxiliary feedwater (AFW) system mini-flow recirculation pathway. This condition would have the potential to prevent the ultimate heat sink from providing an adequate inventory of water for a 30-day mission time. The normal water supply for the Callaway AFW system is the condensate storage tank (CST). The CST is a non-safety grade component. The safety-grade supply for AFW is the essential service water (ESW) system. The ESW system is supplied by the UHS. The UHS thermal performance analysis accounts for a loss of UHS inventory to the AFW system up until the point of the accident sequence that the AFW pumps would be secured. The analysis did not include an allowance for loss of UHS inventory through the AFW mini-flow recirculation pathway following the AFW pumps being secured. The EOP guidance that secures the AFW pumps does not isolate the mini-flow recirculation pathway. Initial estimates indicate that loss of UHS inventory through the mini-flow recirculation pathway, if not isolated, would preclude the UHS from completing its 30-day mission time. This potential for depletion of the UHS placed the plant in an unanalyzed condition that significantly degraded safety. Callaway has issued interim guidance to the on-shift personnel regarding this concern to ensure that the ultimate heat sink water level is maintained at a level that will be adequate to mitigate the potential loss of inventory. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspectors have been notified of this condition.
Event Notification (EN) 53485, made on July 3, 2018, is being retracted because re-evaluation performed subsequent to the notification has demonstrated, based on actual plant equipment and environmental conditions, that the unanalyzed inventory losses previously reported by EN 53485 would not have depleted the UHS inventory to an unacceptable level during its 30-day mission time. The re-evaluation has led to the conclusion that the previously unanalyzed losses of UHS inventory would not have prevented the UHS from performing its specified safety functions and meeting its 30-day mission time requirements. With the UHS capable of performing its specified safety functions and meeting its 30-day mission time requirements, the systems supported by the UHS would have remained capable of performing their specified safety functions. Based on these considerations, it has been determined that the condition reported in EN 53485 did not result in the plant being in an unanalyzed condition that significantly degraded safety. Consequently, the condition did not meet the criteria for an 8-hour notification per 10 CFR 50.72(b)(3)(ii)(B) for an unanalyzed condition that significantly degrades safety. The NRC Resident Inspector has been notified of the Event Notification retraction. Notified R4DO (Gaddy). |
ENS 53487 | 4 July 2018 23:19:00 | The following information was received from the State of Texas by email: On July 4, 2018, the Agency (Texas Department of State Health Services) was notified by the licensee's radiation safety officer (RSO) that one of their crews has experienced a source disconnect. The crew was using a QSA 880D exposure device containing a 70 Curie Iridium - 192 source. The licensee did not have a lot of details on the event, but stated the source had been recovered and that no over exposures had occurred. The licensee stated the connector ball on the drive cable was tested after the event and failed the test. The RSO stated they would provide additional information on July 5, 2018. Additional information will be provided as it is received in accordance with SA-300. Texas Incident- I-9591 |
ENS 53484 | 3 July 2018 12:00:00 | At 0954 (EDT) on July 3, 2018, with Unit 1 in Mode 1 at 100 percent power, the reactor was manually tripped due to high steam generator water level. The trip was not complex, with all systems responding normally. Operations stabilized the plant in Mode 3. Decay heat is being removed through the main steam lines through the steam dumps and into the condenser. The expected actuation of the Auxiliary Feedwater System (an engineered safety feature) is being reported as an eight hour report under 10 CFR 50.72 (b)(3)(iv)(A). Unit 2 was not affected. There was no impact to the health and safety of the public or plant personnel. The NRC Resident Inspectors have been notified. All control rods inserted and Unit 1 is in an electrical shutdown lineup. The cause of the high steam generator water level transient is being investigated. |
ENS 53478 | 28 June 2018 10:10:00 | The following information was received from the State of Texas via email: On June 27, 2018, the Agency (Texas Department of State Health Services) was notified by the licensee's corporate radiation safety officer (CRSO) that one of his radiography crew had experienced a source disconnect. The crew was using a QSA 880D exposure device containing a 113.1 Curie iridium - 192 source. After completing the first shot on a new location on the pipeline, the crew could not get the source to return to the exposure device. The crew contacted the CRSO and set up new barriers at 2 millirem. The CRSO and a second individual qualified for source retrieval arrived at the site at 1743 hours. The retrieval team (RT) surveyed the guide tube and determined the source was in the collimator. The RT removed the guide tube from the exposure device and removed the camera from the area. Using a set of long tongs, the guide tube was removed from the pipe and the source slid down the guide tube until the connector was exposed. They could see the drive cable had broken near the connector. The source was shielded with bags of lead shot. The CRSO disconnected the broken drive cable from the source pigtail and connected the pigtail to a new drive cable that had been installed on the camera. The source was retracted to the shielded position in the camera. The camera and crank out device and drive cable will be sent to the manufacturer for inspection. No overexposures occurred as a result of this event. Additional information will be provided as it is received in accordance with SA-300. Texas Incident # - 9590 |
ENS 53469 | 22 June 2018 15:12:00 | At 0900 (EDT) on June 22, 2018, a non-licensed supervisory contractor subverted a random Fitness for Duty test. The contractor's site access has been terminated. The NRC Resident Inspector was notified. |
ENS 53287 | 25 March 2018 23:43:00 | On March 25, 2018 at 1616 hours (EDT), with the reactor in cold shutdown condition, two control rod drive piping lines were determined to be potentially inoperable in the event of a design basis earthquake due to support defects. The control rod drive piping forms a portion of the reactor coolant pressure boundary and primary containment boundary. The supports will be repaired prior to plant startup. This event had no impact on the health and/or safety of the public. The NRC Resident Inspector has been notified. The licensee will notify the Commonwealth of Massachusetts.
The purpose of the notification is to retract ENS notification 53287 made on 03/25/18 for Pilgrim Nuclear Power Station. The previous notification reported that control rod drive (CRD) piping could be potentially inoperable in the event of a design basis earthquake, at the time of discovery, due to piping support defects. Subsequent evaluation has demonstrated that the piping was not inoperable. Specifically, after an engineering evaluation, it has been determined that the CRD Hydraulic System operability was never lost and the system was operable, although non-conforming, based on the support configuration not conforming to the pipe support drawings. The affected pipe supports have been restored or reworked to the proper design condition in accordance with the design drawings. The CRD System has subsequently been restored to a fully operable status. Notified R1DO (Jackson) and IRD MOC (Pham). |
ENS 53265 | 15 March 2018 22:08:00 | At 1524 (EDT) on Thursday, March 15, 2018, Operations was notified of a failure to meet Appendix R requirements for Peach Bottom Atomic Power Station (PBAPS) Unit 2 and Unit 3. Valves associated with the feedwater system for both units were not properly considered as Hi-Lo Pressure interface valves as required by the Appendix R program. This results in the susceptibility to a hot short condition that could open valves, diverting flow from the reactor, damage piping and prevent injection. U3 (Unit 3) Fire Safe Shutdown Credited Reactor Core Isolation Cooling (RCIC) System is affected. U2 (Unit 2) is affected by a potential leak path through the Reactor Water Cleanup system. This event is being reported as an occurrence of an event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety under 10 CFR 50.72(b)(3)(ii). The Station (PBAPS) is performing hourly fire watches for the impacted areas and is also evaluating this condition for corrective action. The licensee notified the NRC Resident Inspector. |
ENS 53242 | 3 March 2018 02:19:00 | At 2315 EST on March 2, 2018, Pilgrim Nuclear Power Station (PNPS) determined, based on information received from the Commonwealth of Massachusetts, that there may be a potential loss of offsite response capabilities due to ongoing severe natural hazard conditions (i.e., major winter storm) along the coast of Massachusetts. According to information received by PNPS, towns within the 10 Mile EP Radius could be hampered in implementing some protective actions specified in the emergency plan in the unlikely event an emergency were to occur. There is no condition at the Station that would warrant implementation of any emergency plan at this time. PNPS continues to operate safely and is monitoring the weather conditions closely. The Station maintains emergency assessment, response, and communication capability. This report is being made conservatively in accordance with 10 CFR 50.72(b)(3)(xiii) which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. As stated previously, the Station maintains emergency assessment, response, and communication capability. The licensee notified the NRC Resident Inspector. |
ENS 53202 | 11 February 2018 23:36:00 | On February 11, 2018 at 2203 (EST), the Susquehanna Control Room received indication that a loss of Secondary Containment Zone 2 differential pressure (DP) had occurred. Control Room operators noted a differential pressure of <.25" WC (inches Water Column) for several seconds. System DP was restored to normal in 1 minute. The cause of the pressure swings is under investigation. Zone 2 differential pressures being less than 0.25" WC constitutes a loss of Secondary Containment based on not meeting requirements of SR 3.6.4.1.1. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment system. The licensee notified the NRC Resident Inspector. |
ENS 53192 | 1 February 2018 14:23:00 | At 1057 CST on February 1, 2018 with the unit in Mode 1 at approximately 27% power, a manual actuation of the Reactor Protection System (RPS) was initiated due to an unexpected trip of the B Recirc Pump with A Recirc Pump in fast speed. B Recirc Pump tripped during transfer from slow to fast speed resulting in single loop operation. Operators were unable to reconcile differing indications of core flow. This resulted in a conservative decision to initiate a manual scram. The cause of the B Recirc Pump trip and the apparent issues with core flow indication are under investigation. The plant is currently stable in Mode 3. The plant response to the scram was as expected. All control rods (fully) inserted as expected; the feedwater system is maintaining reactor vessel water level in the normal control band and reactor pressure is being maintained with steam line drains and main turbine bypass valves. The NRC Senior Resident (Inspector) has been notified.
This event was initially reported under 10 CFR 72(b)(2)(iv)(B) as a manual actuation of the RPS due to an unexpected trip of the B Reactor Recirculation Pump with the A Reactor Recirculation Pump running in fast speed (Single Loop Operations). Operations was unable to reconcile differing indications of core flow and made the conservative decision to perform a planned shutdown in accordance with normal operating procedures. Therefore, this event 'resulted from and was part of a pre-planned sequence during testing or reactor operation' as specified in 10 CFR 50.72(b)(2)(iv)(B), 10 CFR 50.73(a)(2)(iv)(A) and NUREG-1022 Section 3.2.6. Consequently, this event is not reportable as an actuation of RPS. The NRC Resident Inspector has been notified. R4DO (Groom) has been notified. |
ENS 53191 | 1 February 2018 13:50:00 | A non-licensed (employee) supervisor had a confirmed positive test for alcohol during a random fitness-for-duty (FFD) test. The individual's unescorted access to the plant has been (terminated). The NRC Resident Inspector has been notified. |
ENS 53180 | 23 January 2018 05:02:00 | At 0400 (CST) on 1/23/2018 the Braidwood Technical Support Center (TSC) HVAC (Heating, Ventilation and Air Conditioning) Emergency Makeup Air Filter train was taken out of service to perform a planned Makeup Air Filter charcoal replacement. The TSC HVAC Makeup Air Filter train will be rendered nonfunctional during the charcoal replacement. Subsequent charcoal and HEPA filter testing will restore functionality of the TSC HVAC Makeup Air Filter train. The expected duration of the charcoal replacement and subsequent testing is 30 hours. If an emergency is declared requiring TSC activation during the time TSC HVAC is non-functional, the TSC will be staffed and activated using existing emergency planning procedure unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff to an alternate location in accordance with applicable site procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to a major loss of emergency preparedness capability. An update will be provided once the TSC HVAC Emergency Makeup Air Filter train functionality has been restored. The licensee has notified the NRC Resident Inspector.
On 1/26/18 at time 1539 EST, the TSC HVAC Emergency Makeup Air Filter train was returned to service following the planned Makeup Alr Filter charcoal replacement. Functionality was verified by charcoal and HEPA filter post maintenance testing. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Cameron). |
ENS 53153 | 6 January 2018 06:05:00 | Oyster Creek Declared an Unusual Event HU 6 Hazardous Event for an Abnormal Intake Structure Level Less than or equal to -3.0 feet MSL (Mean Sea Level) on points 23 and 24 in the Main Control Room at time 0524 (EST). The licensee notified the NRC Resident Inspector, State, and local authorities. Notified DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email).
Oyster Creek Terminated Unusual Event HU 6 Hazardous Event for an Abnormal Intake Structure Level at time 2308 (EST). The licensee notified the NRC Resident Inspector, State, and local authorities. Notified R1DO (Werkheiser), NRR EO (King), IRD MOC (Gott), DHS SWO, FEMA Operations Center, DHS NICC, FEMA NWC (email), DHS Nuclear SSA (email), and FEMA NRCC SASC (email). |
ENS 53132 | 20 December 2017 18:18:00 | On December 20, 2017, at 1040 Eastern Standard Time (EST), the Watts Bar Nuclear Plant (WBN) 1B-B 6.9kV Shutdown Board (SDBD) normal feeder breaker opened. The loss of voltage to the 1B-B SDBD resulted in the start of the 1B-B Motor Driven Auxiliary Feedwater (MDAFW) pump, the Unit 1 Turbine Driven Auxiliary Feedwater (TDAFW) pump, and the start of all four Emergency Diesel Generators (EDGs). Power was restored to the 1B-B 6.9 kV SDBD when it loaded on to its associated EDG. Following initial investigation, the 1B-B 6.9 kV SDBD was transferred to its alternate offsite power source, Common Station Service Transformer (CSST) C at 1217 EST. At 1230 EST, the 1B-B 6.9 kV SDBD alternate feeder breaker opened. The loss of voltage to the 1B-B SDBD did not result in the restart of the 1B MDAFW pump, the Unit 1 TDAFW pump, or EDGs; this equipment remained running from the earlier event. Power was restored to the 1B-B 6.9 kV SDBD when it loaded on to its associated EDG. Restoration of normal offsite power to the 1B-B SDBD was completed at 1654. Other than several common Unit Technical Specifications having not been met, Unit 2 was not operationally impacted by the transfer of the 1B-B Shutdown Board to onsite power and remains in Mode 1 at 100% power. This report is made per 10 CFR 50.72(b)(3)(iv)(A). NRC Resident Inspector has been notified. The licensee investigation continues for the cause of the event. |
ENS 53128 | 19 December 2017 17:17:00 | During regular power operations at 100% power, DG#1 and DG#2 were declared inoperable due to a common issue associated with indicating lights and the associated sockets installed in various control and auxiliary circuits for both DG's. The indicating lights in question are incandescent 120V AC style 120MB bulbs in a socket with a 550 ohm resistor. Style 120MB light bulbs have a failure mechanism where the bulb can cause a short circuit rather than the more common open circuit that is expected when an incandescent bulb filament fails. Cooper originally believed that the socket's integral resistor was sufficient to protect the circuit. In testing performed by an outside laboratory and confirmed on-site using warehouse stock, it was determined that the integral resistor may not have the power dissipation capability to protect the circuit ln which the light and socket are installed if a bulb fails in short circuit. This condition resulted in both DG's being declared inoperable at 1340 (CST) due to a loss of reasonable expectation that they would meet their safety function required action to start, load and run to support loads required to mitigate the consequences of an accident. This is a loss of safety function under 10CFR 50.72(b)(3)(v)(D) subject to an 8 hour report. As a result of both DG's being inoperable, the Control Room Emergency Filtration System is also inoperable. This is also a loss of safety function subject to an 8 hour report for the same criterion. The Senior Resident has been notified.
CNS is retracting the 8-hour non-emergency notification made on December 19, 2017 at 1340 CST (EN# 53128). Subsequent evaluation concluded a postulated lamp short circuit failure in any of the affected circuits would not impact the ability of the Diesel Generators to perform their safety function and therefore, were operable. With DG operability not affected, the Control Room Emergency Filtration System also remained operable. The NRC Resident Inspector has been notified. Notified the R4DO (Werner). |
ENS 53127 | 19 December 2017 17:05:00 | The following information was received from the State of Illinois: The University of Chicago Medical Center reported an underdose of Y90 Theraspheres today (12/19/17) to a patient. 53.4 % of dose was delivered with 46.6% stuck in catheter. 21 mCi was ordered and 11.21 was delivered. CT scan verified dose administered in correct location. On 12/18/17, it was still undetermined why remaining dose hung up in catheter. Additional dose (is) being ordered to complete the therapy as a fractionated dose. The licensee is investigating why the catheter became blocked. A 15 day written report will follow. Illinois Incident: IL177059 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. |
ENS 53114 | 11 December 2017 16:29:00 | The following information was received from the State of South Carolina by email: On Monday, December 11, 2017 at (1100 CST), the Department (South Carolina Department of Health and Environmental Control) was notified by (the) Corporate RSO (Radiation Safety Officer) of STERIS Isomedix Services that one of the source racks had been stuck in the 'up' position. The incident happened at (0328 CST) on Saturday, December 8, 2017. The worker saw that there was an unload fault on the system indicating that the rack was stuck so he called maintenance to try to correct the problem. At (0340 CST) the Radiation Safety Officer (and then the corporate RSO were notified) about the event. (The RSO) called and left a message on an employee voicemail rather than calling the 24 hour emergency phone number. The workers were able to go into the penthouse to correct the problem and lower the source rack back into the pool. The workers found that a carrier had a cracked hinge. They checked all of their other carriers and replaced a total of two carrier doors. The RSO informed the CRSO (Corporate Radiation Safety Officer) that the situation was resolved at (0724 CST). The licensee stated that a written report will be sent within 30 days of the event. |
ENS 53109 | 8 December 2017 17:26:00 | The following information was excerpted from an email received from the State of Kansas: The licensee is reporting that an ionizer containing a radioactive source (Model Number P-2063-1000) was lost. The licensee currently has 3 other ionizers of the same model. The device radioactive source was Polonium-210 (SN: A2KT674) with an activity of 31.5 mCi and was last leak tested on 9/20/2016. The device was checked out by the licensee and placed within the secure test floor while testing electrical devices at the Integra Technologies facility. The missing device use was last logged on 8/25/2017. The licensee believes that their maintenance department mistakenly threw the device away. Upon discovering that the device was missing, the licensee searched their facility several times over without finding the device. The prevention for further loss is that the remaining 3 units will be mounted in permanent locations using security screws so they cannot be removed by unauthorized personnel. 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 |
ENS 53108 | 8 December 2017 17:25:00 | U/2 HPCI (Unit 2/High Pressure Coolant Injection) was declared inoperable due to leak by of the pump discharge check valve after pump shutdown from flow testing. This resulted in cycling of the minimum flow valve. The discharge valve was closed to prevent the continued cycling of the minimum flow valve. This condition was identified during normal surveillance testing. The licensee notified the NRC Resident Inspector. |
ENS 53107 | 7 December 2017 15:03:00 | This non-emergency notification is being made in accordance with 10 CFR 50.72(b)(2)(xi), any event or situation related to the protection of the environment for which notification to other government agencies has been made. Dominion Energy is in the process of informing the Virginia Department of Health, Department of Environmental Quality, Department of Emergency Management, and the Surry County Administrator of recent groundwater monitoring results at Surry Power Station in accordance with NEI 07-07, Industry Groundwater Protection Initiative (GPI). On December 6, 2017 at 1138 EST, Surry Power Station received analysis results of recent samples from the on-site groundwater monitoring program. As part of the program, 10 new groundwater monitoring wells were recently placed in service within the Protected Area to provide early detection, to better define the site's hydrology, and if necessary, to mitigate any potential leaks. The analysis results from one of the new wells indicated tritium activity level above the GPI communication threshold. Samples were re-analyzed, resulting in different values, with the highest result of 59,300 picoCuries per liter. Since each result was above the voluntary reporting threshold, Surry stakeholder communication was implemented in accordance with the NEI GPI Voluntary Communication Protocol, Criterion 2.2. There are no known active leaks at this location; however, Dominion Energy is continuing to investigate the source of the tritium and the reason for the variability in the sample results. Tritium was not detected in the on-site monitoring locations outside of the Protected Area. No tritium has been detected in the on-site and off-site drinking water wells. Since the activity is contained within the site restricted area, the health and safety of on-site personnel and members of the public are not affected. A 30-day report will be submitted to the NRC in accordance with NEI 07-07. The NRC Senior Resident Inspector has been notified. |
ENS 53474 | 26 June 2018 17:30:00 | The following information was received from the State of Texas: During the review of an event, the Agency (Texas Department of State Health Services) found a letter from a licensee reporting the shutter on a Ohmart model SHD-45 containing a 50 millicurie cesium - 137 source had failed in the closed position. The report was dated November 29, 2017. The shutter did not pose an exposure risk to any individual. The licensee has worked with the manufacturer and the gauge was scheduled to be replaced on June 21, 2018. The Agency has not been able to confirm if the gauge was repaired/replaced. Additional information will be provided as it is received in accordance with SA-300. Texas Incident Number: I-9588 |
ENS 53126 | 19 December 2017 12:46:00 | The following information was excerpted from a facsimile received from Crane Nuclear: This letter provides notification of a defect in a Weak Link Analysis provided to the Tennessee Valley Authority (TVA) by Crane-Aloyco, Inc. (CAI), a Crane Nuclear, Inc. (CNI) predecessor business unit, for a Chapman Gate Valve, Figure L900, Item # 18, Drawing CC05307, Revision B for the Browns Ferry Nuclear (BFN) plant. The subject valve was originally procured from Crane Chapman in 1968. In 1988, TVA requested Crane to supply a Weak Leak Analysis for the original valve. A Weak Link Analysis (OTC-258 Rev.0) was developed by CAI, which identified a maximum thrust capacity of approximately 112,000 lbf. In November 2017, Crane Nuclear, Inc. developed a new Weak Link Analysis for the valve. Crane Nuclear, Inc. provided the new Weak Link Analysis (WL-103 Rev. 0) to TVA on November 17th, 2017. Crane Nuclear. Inc. identified in the new Weak Link Analysis a maximum thrust capacity of approximately 96,000 lbf. CNI is reviewing our records to determine if the maximum thrust rating in any other Weak Link Analyses provided by CNI for gate valve designs with an SMB-4T or SMB-5T actuator exceeds the rating for the thrust bearings. Should you have any questions regarding this matter, please contact me, Joyce Hamman, Director, Safety & Quality at (678) 451-2280, Burt Anderson, Site Leader, at (630) 226-4990, or Samson Kay, Engineering Manager at (630) 226-4983. |
ENS 53074 | 16 November 2017 08:17:00 | At 0008 CST on 11/16/2017, Cooper Nuclear Station (CNS) was notified by Omaha Weather that the NOAA broadcast and the Shubert radio tower for this area is off. This affects the tone alert radios used to notify the public in event of an emergency condition. This is considered to be a major loss of the Public Prompt Notification System capability, and is reportable under 10CFR50.72(b)(3)(xiii). The transmission outage actually began at 2007 (CST), 11/15/2017, but CNS was not notified until 0008 (CST), 11/16/2017. Backup notification methods remained available throughout the period. At time 0447 CST on 11/16/2017, Cooper Nuclear Station was notified that the NOAA broadcast and Shubert radio transmission tower was returned to service. Nemaha County, NE, Richardson County, NE, and Atchison County, MO authorities within the 10 mile EPZ were notified by Cooper Nuclear Station of the condition and the effect on the tone alert radios at 0642 (CST), 11/16/2017. This is reportable under 10CFR50.72(b)(2)(xi) as a 4 hour report. The NRC Senior Resident has been informed. |
ENS 53129 | 19 December 2017 18:28:00 | The following information was excerpted from an email received from the State of Florida: The State of Florida received a notice on 11/15/17 that a Troxler Gauge and a company vehicle was stolen from the licensee by an employee. A City of Orlando Police report was issued (2017-442672). On 12/7/2017, a notice was received from the licensee to inform State of Florida Bureau of Radiation Control that the gauge was found intact and undamaged. The Troxler moisture density gauge is a model number 3440; serial number 27931; Cs-137/AmBe; 8mCi/40mCi. Florida Incident Number: FL17-298. 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 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 |
ENS 53067 | 13 November 2017 03:57:00 | At 2119 (CST) on 11/12/2017 a Control Room board walk down discovered that both of the Unit 2 Containment Spray Pump control switches were in pull-out. With the control switches in pull-out, the pumps would not automatically start as required. Unplanned TS (Technical Specifications) 3.0.3 was entered at 2119 as a result of not complying with TS 3.6.5, Containment Spray and Cooling Systems, which requires both trains of Containment Spray to be Operable while in Mode 4. Unit 2 had entered Mode 4 at 0303 on 11/12/2017. TS 3.0.3 was exited at 2127 on 11/12/2017 when both Containment Spray Pump control switches were placed in Automatic restoring Operability. Preliminary investigation determined that while Unit 2 was in Mode 5, Surveillance SP 2099, Main Steam Isolation Valve Logic Test, had taken the Containment Spray Pump control switches to pull-out but did not re-align the control switches to automatic after the test was complete. This 8-hour Non-Emergency report is being made per 10 CFR 50.72(b)(3)(v)(D), Accident Mitigation. The NRC Senior Resident Inspector has been informed. |
ENS 53130 | 19 December 2017 17:40:00 | The following was excerpted from an email received from the State of Florida: The State of Florida received a notice of an over-exposure from the licensee. An employee received a whole body dose of 5019 mR read on her dosimeter on 10/10/2017. The investigation determined that the most likely cause of the over-exposure was due to an unusual number of equipment failures with the synthesis units requiring employee intervention to correct the issues. Dose rates and doses to the employee were not being monitored real time. The employee has been retrained on the standard operating procedures. Alarming personal electronic dosimeters have been purchased and are in use to alert personnel of the radiation fields. Florida Incident Number: FL17-299 |
ENS 53059 | 7 November 2017 22:09:00 | On November 7, 2017 at 1810 (CST), Unit 1 High Pressure Coolant Injection (HPCI), was manually isolated following failure of the remote turbine trip pushbutton to function. Unit 1 HPCI Operability Testing was in progress to the point of securing the HPCI turbine with the remote manual pushbutton. The pushbutton failed to trip the turbine resulting in operator action to lower the flow controller setpoint and isolating the HPCI steam line. HPCI remains isolated and is Inoperable pending resolution of the Turbine Trip circuitry. This event is being reported as a condition that could have prevented fulfillment of a safety function in accordance with 10CFR50.72(b)(3)(v)(D). The HPCI system is a single train system and the loss of HPCI could impact the plant ability to mitigate the consequences of an accident. The Reactor Core Isolation Cooling (RCIC) system was confirmed operable. The NRC Senior Resident Inspector has been notified. |
ENS 53055 | 6 November 2017 13:21:00 | The following information was received from the State of Texas by email: On November 6, 2017, the Agency (Texas Department of State Health Services) was notified by the licensee that a shutter was stuck in the closed position. The Ronan SA1 shutter was closed for maintenance on a hopper and failed to reopen. The gauge contains a 50 millicurie Cesium-137 source. The Licensee stated a service company has been contacted to repair the gauges in the next few days. No individual received significant exposure to radiation due to this event. Additional information will be provided as it is received in accordance with SA-300. Texas Incident Number: I-9519 |
ENS 53040 | 28 October 2017 13:29:00 | A patient receiving treatment for a liver disease was prescribed 60 milliCuries of Y-90 SIR-Spheres. The delivered dose was calculated to be 11 milliCuries and stasis was not achieved. The patient was notified of the misadministration and is scheduled to receive the fully prescribed dose. 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. |
ENS 53037 | 26 October 2017 17:00:00 | The following information was received by facsimile from the vendor: Component: Speed switch P/Ns ESl50267C, ESl50267E, ESl50267H, and ESl50267K. Summary: Engine Systems Inc. (ESI) began a 10CFR21 evaluation on September 12, 2017 upon notification of a potential issue with speed switch P/N ESl50267K supplied to Hope Creek Nuclear Generating Station. The speed switch had reportedly failed in service which resulted in a failure to start of the emergency diesel generator (EDG). An analysis performed by Exelon Powerlabs determined the failure was due to a shorted capacitor that is installed on the speed switch's relay output contacts to ground. The evaluation was concluded on October 25, 2017 and it was determined that this issue is a reportable defect as defined by 10CFR Part 21. The speed switch output contacts are utilized in the engine's start circuitry and failure to function properly could adversely affect the safety-related operation of the emergency diesel generator set. Impact on Operability: If the resistance path to ground were sufficiently low, the ability of the relay output contacts to pick-up and/or drop-out associated components would be compromised. The speed switch relays are used in safety-related EDG start circuitry to control various electrical relays. Failure to properly control any of these components could adversely affect the safety-related operation of the emergency diesel generator. Root cause evaluation: The root cause of the failure is a deficiency in the design and selection of the EMC (Electromagnetic compatibility) mitigating components. Consideration was not given for the impact of voltage transients imparted on the capacitors during coil de-energization. For customers without suppression from the inductive kick, the magnitude of voltage transients may be sufficient to damage and ultimately degrade the capacitors to the point of failure. Affected nuclear plants include Nine Mile Point, Quad Cities, Dresden, Davis Besse and Hope Creek. |
ENS 53016 | 16 October 2017 08:36:00 | The following information was received from the State of California: The Licensee discovered at approximately (1700 PDT) on 10/15/17 that two 3 Ci Am-241Be well logging sources had been stolen from their storage area at Weller Ranch in Kern County. Locks had been cut and the sources were removed from the approximate 12-foot storage pipe. Additionally, an approximate 2500 (pound) calibration water tank was also missing. The FBI was notified by CA Radiologic Health Branch (RHB) at approximately (2030 PDT) on 10/15/17. RHB will be onsite 10/16/17. California 5010 Number: 101517
The two well logging sources have been accounted for. An unauthorized individual had accessed the sources, removing them from their storage location on his father's ranch land and discarding them a short distance away without any knowledge of what they were. The unauthorized individual also took the water calibration tank for his personal use. The well logging sources are back in the possession of the licensee. Notified R4DO (Vasquez) , NMSS Events Notification, CNSNS (MEXICO) and ILTAB via email. 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 |
ENS 52991 | 25 September 2017 14:04:00 | The following information was received from the Commonwealth of Pennsylvania via email: Notifications: On September 25, 2017, the licensee informed the Department (PA DEP Bureau of Radiation Protection) of a failure of an electronic component of a fixed gauge. It is reportable per 10 CFR 30.50(b)(2)(i). Event Description: The electronic component of the automatic shutter on an IMS Model 5301-01 gauge containing approximately 20 curies of Cesium-137 failed to close on its own. The licensee immediately notified the RSO, as per their emergency procedure, who was able to remotely log in to the computer software system and bypass the automatic mode to close the shutter. The gauge is housed in a secure and entry restricted enclosure and instructions have been given to all operators to ensure that the shutter is closed while not in use. The manufacturer, IMS, was notified and is scheduled to make repairs on September 26, 2017. All regulatory precautions were taken and no overexposures have occurred. Cause of the Event: Equipment failure. ACTIONS: The Department will perform a reactive inspection. The manufacturer has already been scheduled to correct the problem. More information will be provided upon receipt. PA Event Report ID No.: PA170014 |
ENS 53017 | 16 October 2017 11:08:00 | The following information was received from the State of New York via facsimile: On September 21, 2017 the Department (New York State Department of Health) was notified that a Best Medical International, Inc., Model #2301 lodine-125 seed used for localization of non-palpable lesions and lymph nodes was lost. On September 18, a patient was implanted with a 125.2 microCurie lodine-125 seed. The seed was verified to be implanted by use of a survey meter. When the patient returned for explant three days later, the iodine-125 seed could not be detected. The licensee surveyed the patient's vehicle, house, laundry, and trash and no radioactivity was detected. The licensee reported placing the seed 'superficially' within the patient and the licensee speculates that the seed may have become dislodged from the patient at some point between the implant and explant. New York Report ID No.: NYDOH-NY-17-08 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 |
ENS 52974 | 17 September 2017 16:49:00 | On September 17, 2017, during planned surveillance activities involving Emergency Diesel Generator (EDG) 4, unexpected voltage and frequency indications were noted when EDG 4 was synchronized to Emergency Bus E4. With EDG 4 in manual mode, the Operator responded by lowering load to reopen the EDG 4 output breaker. Opening of the EDG 4 output breaker with the breakers from Balance of Plant (BOP) Bus 2C, which normally feeds the Emergency Bus E4, opened; resulted in de-energizing Emergency Bus E4. The EDG 4 voltage regulator and governor automatically reverted to auto control, and EDG 4 reconnected to Emergency Bus E4. Normal frequency and voltage were restored with EDG 4 in auto control. The momentary power interruption to Emergency Bus E4 resulted in Unit 2 Primary Containment Isolation System (PCIS) Group 2 (i.e., Drywell Equipment and Floor Drain, Residual Heat Removal (RHR) Discharge to Radwaste, and RHR Process Sample), Group 6 (i.e., Containment Atmosphere Control/Dilution, Containment Atmosphere Monitoring, and Post Accident Sampling Systems), and Group 10 (i.e., air isolation to the drywell) isolations. The actuations of Primary Containment Isolation Valves (PCIVS) were completed and the affected equipment responded as designed. Per design, no Unit 1 safety system group isolations or actuations occurred. These actuations are being reported in accordance with 10 CFR 50.72(b)(3)(iv)(A). Additional Unit 2 actuations included PCIS Group 3 (i.e., Reactor Water Cleanup), Reactor Building Ventilation System isolation (i.e., Secondary Containment isolation), and the automatic start of Standby Gas Treatment (SGT) System subsystems A and B. These systems functioned as designed. This event did not impact public health and safety. The NRC Resident Inspector has been notified. The safety significance of this event is minimal. Safety systems functioned as designed following the power perturbation on E4. Plant systems responded as designed. The cause of the event is under investigation. |
ENS 52969 | 14 September 2017 16:38:00 | The following information was received from the State of California via email: On September 14, 2017, (The RSO) of Southwest Calibration & Training notified the RHB Brea office that United Inspection & Testing, Inc., RML # 4788-33, had a Troxler, 3411B, serial #6644 radioactive gauge run over and damaged. On September 14, 2017, RHB Brea contacted (The RSO) of United Inspection & Testing, Inc. (The RSO of United Inspection & Testing, Inc.) informed our office that the radioactive gauge had been run over by a backhoe on the afternoon of September 13, 2017 at approximately 1400 (PDT), at the intersection of Banana Street and Daurin Street at a construction site in Fontana, CA. As a result of the accident the radioactive gauge had the handle broken off. The RSO was able to return the Cs-137 source to its shielded position, but it could not be locked in the shielded position due to the damage to the gauge. The authorized user of the gauge was also struck by the backhoe and died of his injuries. (The RSO of the United Inspection & Testing, Inc.) retrieved the gauge from the accident site and transported it to Southwest Calibration & Training to be inspected. (The RSO) of Southwest Calibration & Training reported that the Troxler radioactive gauge read 0.9 mR/hr at 1 foot. The gauge was extensively damaged and may not be repairable. California 5010 Number: 091417 |
ENS 52971 | 15 September 2017 15:17:00 | The following information was received by the State of Illinois by email: IEMA (Illinois Emergency Management Agency) was notified at (1353 CDT) on 9/8/17 that a load of ferrous metal was being rejected from a scrap metal recycling facility (Omnisource in Indiana) back to Gaby Iron in Chicago Heights. The max exposure rate was reported at 20 microR/hour (4 microR/hour background). The load is being returned under DOT SP IN-IL-17-010. The suspect load was inspected Monday, September 11th. An Alnor dew pointer device with an intact 7 microCurie Ra-226 source was recovered. No removable contamination was identified. The device was impounded by IEMA and is pending return to an appropriate entity. No additional radiation sources were discovered and the remainder of the load was released without further restriction. Pending appropriate disposal or return to the manufacturer, this matter is being considered closed. The device was an Alnor Instrument portable gauge; Model 7350; Serial Number 230667. The Amersham sealed source was a model RAM.X452; Ra-226; 7 microCurie activity. NMED Report: IL177030 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 |
ENS 53062 | 8 November 2017 15:06:00 | The following information was received from the State of Ohio via email: A local Health District employee had a Niton Xlp 300 XRF with a 50 mCi Cadmium-109 source stolen overnight on Saturday, September 2, 2017. It was in the car in their garage and someone came in and took it. The employee had worked late at a job site that day and brought the gauge home instead of returning to the office. Employee's garage door did not close for some reason that night and they were unaware that it was open when they went to bed. There were several other cars broken into that night in employee's neighborhood. A report was filed with local police department. Device has not yet been recovered. Source/Radioactive Material: Sealed Source; Radionuclide: Cd-109; Activity: 50mCi; Device Name: X-RAY Fluorescence (XRF); Model Number: Niton XLp 300; Manufacturer: Thermo Scientific Analytical; Serial Number: 98149. Ohio Item Number: OH170007
The following report was received via e-mail: Note: According to device owner, the manufacturer told them that this incident was NOT reportable to their regulatory agency. The owner reported the event on 11/6/17 as a result of more research on their part. UPDATE: The gauge was found by a member of the public in their yard, where it had apparently been abandoned. The local health district was notified based on contact information on case. The case was still locked when found. The device is now back in the possession of the local health district as of 11/27/17. Notified the R3DO (Duncan), NMSS Events Resources and CNSC (via e-mail) . THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIAL Sources that are "Less than IAEA Category 3 sources," are either sources that are very unlikely to cause permanent injury to individuals or contain a very small amount of radioactive material that would not cause any permanent injury. Some of these sources, such as moisture density gauges or thickness gauges that are Category 4, the amount of unshielded radioactive material, if not safely managed or securely protected, could possibly - although it is unlikely - temporarily injure someone who handled it or were otherwise in contact with it, or who were close to it for a period of many weeks. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
ENS 52909 | 16 August 2017 15:41:00 | On 8/16/2017, at 1039 (EDT), an un-planned trip of the Peach Bottom Station Blackout Transformer 34.5 kV feeder breaker 1005 and a loss of the 191-00 line occurred causing a loss of power to Unit 1 and the TSC. Power was not restored to the TSC or the ventilation system within 1 hour. Power was subsequently restored to the TSC at 1207 hours (EDT) and the ventilation system was restored to available. This report is being submitted pursuant to 10CFR50.72(b)(3)(xiii) as a Major Loss of Emergency Preparedness Capabilities due to a reduction in the effectiveness of the Onsite Technical Support Center (TSC). The NRC Resident Inspector has been informed of this notification. |
ENS 52905 | 15 August 2017 14:07:00 | On August 15, 2017, during evaluation of protection for Technical Specification (TS) equipment from the damaging effects of tornados, Callaway Plant identified a non-conforming condition in the plant design such that specific Technical Specification equipment is considered not to be adequately protected from tornado missiles. The recirculation lines for all three independent trains of Auxiliary Feedwater (AFW) connect to the Condensate Storage Tank (CST) inside the CST Valve House, which is not a tornado missile-resistant structure. The direct impact by a design basis missile could result in crimping of the recirculation lines, thereby creating the potential to cause damage to the Train A and B Motor-Driven Auxiliary Feedwater Pumps (MDAFPs) and the Turbine-Driven Auxiliary Feedwater Pump (TDAFP) by restricting recirculation flow to less than the design requirements. This condition is reportable per 10 CFR 50.72(b)(3)(ii)(B) for any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety, and per 10 CFR 50.72(b)(3)(v) for 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 (A) shut down the reactor and maintain it in a safe shutdown condition, (B) remove residual heat, or (D) mitigate the consequences of an accident. These conditions are being addressed in accordance with NRC's Enforcement Guidance Memorandum EGM 15-002 and Interim Staff Guidance DSS-ISG-2016-01 (enforcement discretion and interim guidance documents). The NRC Resident Inspector has been notified. |
ENS 52907 | 15 August 2017 16:52:00 | The following information was provided by the State of California via email: On 08/14/17, RHB (California Radiation Health Branch) received an incident report from Office of Emergency Services regarding a damaged moisture density gauge. A CPN gauge, Model MC-3, S/N M380108935, containing 10 mCi of Cs-137 and 50 mCi of Am-241 was run over by a heavy construction equipment at a job site in the city of Santa Clara, CA. The top of the gauge housing was damaged with a broken rod, however, the user managed to retrieve the source back into shielded position. The damaged gauge was placed in the transport case and taken to the licensee's facility for disposal. Fire department was at the incident site, performed surveys using a survey meter (no survey meter information available) and the readings did not indicate any contamination. According to the gauge user, Fire Department readings indicated 500 uR/hr at the damaged gauge and 6 uR/hr at 15 feet from the gauge. The gauge will be transported to CPN for leak testing and disposal on 08/15/17. RHB will be following up on this incident. California 5010 Number: 081417 |
ENS 52906 | 15 August 2017 14:24:00 | Following a panoramic irradiator two day shutdown, a restart with three source racks commenced. Air pressure was applied to raise the source racks. During the restart, two source racks (racks 1 and 3) did not descend into the irradiator pool as designed. During an investigation, two release valves associated with the two source racks did not operate properly. Operators manually released air pressure and all source racks descended into the irradiator pool. The deficient release valves were replaced and the source racks were satisfactorily retested. The source racks all properly descended into the pool. The time the source racks were inoperable for approximately 1.5 hours. |
ENS 52891 | 8 August 2017 20:22:00 | On August 8, 2017, at 1554 hours (EDT), during restoration from testing of the High Pressure Core Spray (HPCS) Suppression Pool Level High Instrumentation, unexpected as-left indications were found that impacted both of the required channels of instrumentation. Subsequent venting of the instrumentation lines was completed and both channels of instrumentation are reading consistent with previously taken as-found data. The instrumentation was declared OPERABLE at 1635. The initial cause of the unexpected as-left indications appears to be the introduction of air into the instrumentation lines during the calibration activities. This is considered a loss of safety function based on both of the HPCS Suppression Pool Level High Instrumentation channels being declared INOPERABLE and the loss of the automatic HPCS suction swap to the Suppression Pool on a high level. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D). The (NRC Resident Inspector) has been notified. |
ENS 52876 | 28 July 2017 15:22:00 | Containment atmosphere oxygen level was measured at 18.4 percent. This is below normal habitability level. The cause of the low oxygen level is a nitrogen leak inside containment Nitrogen has been isolated from containment and operators are preparing to purge containment. The licensee notified the State of California, local authorities and the NRC Resident Inspector. Notified the DHS SWO, FEMA OPS, USDA OPS, HHS OPS, DOE OPS, DHS NICC, EPA EOC. Notified FDA EOC, NuclearSSA, FEMA NWC and FEMA NRCC SASC via email.
The ALERT was terminated on 7/28/17 at 1819 PDT. The containment atmosphere was restored to normal conditions. The nitrogen source was isolated. The cause of the nitrogen leak into containment is under investigation. The licensee will notify the NRC Resident Inspector. The licensee has notified the State of California and the local authorities. The licensee plans to issue a press release. Notified the R4DO (Hay), NRR EO (Miller), IRD MOC (Grant). Notified the DHS SWO, FEMA OPS, USDA OPS, HHS OPS, DOE OPS, DHS NICC, EPA EOC. Notified FDA EOC, NuclearSSA, FEMA NWC and FEMA NRCC SASC via email. |
ENS 52874 | 27 July 2017 18:54:00 | (Unit 2) HPCI was declared inoperable due to improper valve alignment stemming from an incorrect sequence directed from a work order. (Unit 2) HPCI was inoperable for 20 minutes and was manually re-aligned to an operable status. The licensee notified the NRC Resident Inspector. |
ENS 52872 | 25 July 2017 11:07:00 | On July 25, 2017, at 0428 Eastern Daylight Time (EDT) Watts Bar Nuclear Plant (WBN) Unit 2 was in Mode 3, beginning a Reactor Startup. While in the initial phase of withdrawing the first of four Control Rod banks, the two associated group demand position indicators deviated greater than 2 steps from each other. In accordance with Technical Requirement 3.1.7, Position Indication System, Shutdown, with one or more group demand position indicators inoperable, the reactor trip breakers are to be opened immediately. Operations personnel opened the reactor trip breakers immediately by initiating a manual trip of the Reactor Protection System (RPS). The Auxiliary Feedwater system was in service and controlling Steam Generator water levels at the time of the event and did not receive any valid actuation signals. No other system actuations occurred as a result of this reactor trip and all systems operated as designed. The cause of the position indication system inoperability is currently under investigation. NRC Resident Inspector has been notified. |
ENS 52871 | 24 July 2017 23:50:00 | |
ENS 52867 | 21 July 2017 10:40:00 | The following information was received by the licensee via email: Pursuant to 10 CFR 21, this is a non-emergency notification by Susquehanna Nuclear, LLC concerning a defect in an Eaton/Cutler Hammer A200 series starter that failed while in service at Susquehanna Steam Electric Station. The failed starter was manufactured by Eaton Corporation in 2014 and purchased by Susquehanna from AZZ/NLI as part of an MCC bucket assembly. The starter failed with its contacts stuck in the energized state when it was de-energized. A failure analysis identified the contactor sticking to be due to the pole faces of the coil laminations and those of the armature laminations adhering to one another at normal operating temperatures. There was residue/material on the pole faces which closely matched Polydimethylsiloxane (PDMS) and silicone grease. One of the characteristics of PDMS is that at cooler temperatures it is more of a solid consistency, and at higher temperatures it becomes more viscous and tacky. A previous Part 21 report submitted by Curtiss-Wright QualTech NP (Event Notification Number 51611) in December 2015 provided notification of Eaton/Cutler Hammer A200 series starters failures due to silicon based mold release that remained on the molded parts and would come between the moving (magnet) and fixed armatures. The Part 21 stated that when heated for extended period of time, the material would become sticky causing anywhere from a minor delay in opening to a frozen closed condition. Eaton/Cutler Hammer determined that the silicone mold release was first introduced into the manufacturing facility in May 2008 and used periodically until October 2012. According to Eaton/Cutler Hammer, any starters manufactured after January 1, 2013 should be silicon mold release free. Following the failure of the 2014 starter at Susquehanna, Eaton Corporation performed an investigation and reconfirmed that silicon mold release was banned from molding production in October 2012 and has not been used since that time. Eaton concluded that the contamination does not appear to be systemic, but rather random and intermittent and that the contamination was most likely introduced either by operators and assemblers on the manufacturing lines, or by others who disassemble and inspect the product after shipment from their plant. Susquehanna does not take the components apart during receipt for testing or visual inspection. Eaton concluded that there is no evidence that the issue is systemic and considers it a random event. Susquehanna has evaluated the condition and has concluded that the condition could create a substantial safety hazard. The licensee notified the NRC Resident Inspector. |
ENS 52858 | 14 July 2017 11:13:00 | The following information was received from the State of Arizona via email: This First Notice constitutes EARLY notice of events of POSSIBLE safety or public interest significance. The information is as initially received WITHOUT verification or evaluation, and is basically all that is known by the Agency (Arizona Radiation Regulatory Agency) Staff at this time. During an inspection of the licensee on July 13, 2017, an inspector found one portable gauge where the radiation source exposure shutter would not close when moved to the closed position. The inspector's dose measurement at contact with the device was approximately 100 mR/hr. The gauge is a Troxler model 3430, Serial Number 30302, containing 8 milliCuries of Cesium-137 and 40 milliCuries of Americium-241. The licensee has contacted a repair company to fix the gauge as soon as possible. The Agency is investigating the event. The Governor's office and U.S. NRC are being notified of this event. Arizona First Notice: 17-009 |
ENS 52851 | 13 July 2017 10:33:00 | The following information was received from the State of Colorado via email: This is an initial report regarding a misadministration event in Colorado. University of Colorado Hospital (License Number: CO 828-01) had a misadministration of Y-90 microspheres (SIRTex SIRSpheres) on Wednesday, July 12, 2017. At approximately 11 (MDT), the post administration measurements of the waste from the SIRSpheres Administration indicated that the activity administered to segment 2/3 of the patient's liver was only 68.7 percent of the prescribed activity. The written directive called for an activity of 0.24 GBq and residual waste activity measurements indicated that 0.165 GBq was delivered. The physician indicated that stasis was not reached during the administration to this segment. There was a separate administration to segment (four) of the liver in which stasis was reached. Follow-up information will be provided after they are available. 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. |
ENS 52829 | 26 June 2017 18:39:00 | On June 26, 2017, at 1531 (EDT), Indian Point Unit 2 inserted a manual reactor trip prior to Steam Generator levels reaching the automatic reactor trip setpoint. Steam Generator water level perturbation resulted from a loss of 22 Main Boiler Feed Pump. All Control Rods verified inserted. The Auxiliary Feedwater System started as designed and supplied feedwater to the Steam Generators. Heat removal is via the Main Condenser through the High Pressure Steam Dumps. Offsite power is being supplied through the normal 138kV feeder 95332. The cause of the 22 Main Boiler Feed Pump loss is currently under investigation. Entergy is issuing a press release/news release on this issue. Unit 2 is stable and in Mode 3. There was no impact on Unit 3. The licensee notified the State of New York and the NRC Resident Inspector. |
ENS 52826 | 24 June 2017 15:42:00 | On June 24, 2017 at 1028 (EDT), a loss of secondary containment occurred due to trip of 2V217A Zone III Filtered Exhaust Fan causing a reduction in D/P (differential pressure) to less than the required 0.25 WC (water column). 2V217B Zone III Filtered Exhaust Fan started on low flow in AUTO as designed and secondary containment D/P was restored to greater than 0.25 WC by 1029 hours. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022, Rev. 3, section 3.2.7 as a loss of a safety function. There is no redundant Susquehanna secondary containment system. The licensee notified the NRC Resident Inspector. |
ENS 52822 | 23 June 2017 11:37:00 | The following information was received from the State of North Carolina via email: On June 22, 2017 at (1130 EDT), North Carolina Radiation Protection Section (RPS) was informed by the Radiation Safety Officer for Hospira, Inc. (Pfizer), Rocky Mount, NC (License 064-0969-1) that they were experiencing an issue involving their Wet Shielded Irradiator (Nordion Model JS-8900, Serial Number IR-183, approved for 4,800,000.00 Ci of Co-60). RSO stated that during routine maintenance checks the Source 1 Rack of the irradiator would not trip the down switch to confirm the source rack was in the down position on the control panel and that they were following emergency procedures. Nordion was then contacted by the licensee to obtain assistance. RPS inspectors were immediately dispatched to the licensee's site. Once on site, RSO informed RPS that visual confirmation was made of source position via hydraulic cylinders that were fully extended, comparison of cable tightness on roof was observed, and that no indication of radiation in the vault was detected; all leading to the unconfirmed indication that the source rack had moved to the down position. With the assistance from Nordion, Hospira staff were able to initiate bypass procedures and gain access to the vault where confirmation was made that the source racks were in the down position. Nordion advised that a faulty down position switch was the cause for the failure. Switch was repaired on site by Hospira engineers, same day. Following repair, Hospira personnel cycled the sources which were brought up into position for one sterilization cycle and then the sources were brought down to test the position sensor. The test was successful, as indicated by the down position indicator lamps and screen on the operator's panel. Nordion staff was informed of the successful test and Hospira staff continued procedural tests to confirm full functionality. After confirming cycling up and down of the source racks, Hospira personnel performed full monthly QA check before resuming operations. 30-Day report is pending to RPS.
The following information was received from the State of North Carolina via email: We have completed our investigation and have no further information to provide in this event report. We would like to request (NMED) Event 170315 be Closed & Complete. Notified R1DO (Lilliendahl) and NMSS Events Notification via email. |
ENS 52823 | 23 June 2017 16:13:00 | The following information was received from the Commonwealth of Massachusetts via email: The licensee reported on June 23, 2017 that licensee learned from its licensed leak test service provider on June 21, 2017 that one 6 millicurie, cobalt-57 sealed source out of 25 sources received in a package on June 16, 2017 from the source manufacturer, Eckert & Ziegler Isotope Products, tested positive for leakage. The leakage was reported as being 4.2 times the limit of 0.005 microCuries (0.021 microCuries). The other 24 sources showed no contamination. The leaking sealed source is an Eckert & Ziegler Isotope Products Model 3901-2 source, serial number P6-883. The licensee reported that the leaking source was contained and secured in an individual zip lock type plastic bag; that there is no facility contamination based on area surveys performed; that the external surfaces of the package received, that had contained all of the 25 source, had been wipe tested and that the package was not contaminated; and that the sources were not used pending leak test results. The licensee reported that it notified the source manufacturer on June 21, 2017, received a return authorization number from the manufacturer, and shipped the source back to the manufacturer on June 22, 2017. The Agency (Massachusetts Radiation Control Program) considers this event to be open. |
ENS 52820 | 22 June 2017 20:33:00 | On June 20, 2017, at 1444 hours (EDT), with the reactor at 100% core thermal power and steady state conditions, plant personnel identified that both doors in one of the secondary containment airlocks (Door #58 and Door #85) were open briefly as part of normal passage of personnel. The Technical Specification definition of SECONDARY CONTAINMENT INTEGRITY states 'At least one door in each access opening is closed.' Actions were taken to immediately close both doors and restore operability of secondary containment. PNPS (Pilgrim Nuclear Power Station) is providing an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(v)(C), an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The NRC Resident Inspector has been notified. The licensee notified the Commonwealth of Massachusetts. |
ENS 52792 | 7 June 2017 12:13:00 | The following information was provided by the State of Arkansas via email: During routine shutter checks performed by the licensee on June 6, 2017, the licensee noted that the shutter would not close. The gauge is identified as Berthold Model LB 300 L source holder containing 0.189 milliCuries of Cobalt-60. The gauge remains operational in the normal use location and the RSO will place additional signs in the area. No maintenance is planned in the area that would require closing of the shutter. The RSO has performed a radiation survey to ensure that radiation exposure is maintained at less than 2 mR/hr in the vicinity of the gauge. The licensee has contacted the technical representative who was expected to be at the facility on July 6, 2017, for other maintenance and will examine this gauge. In accordance with RH-1502.f.2 (10 CFR 30.50(b)(2)) the malfunctioning shutter is reportable within 24 hours. The State of Arkansas is awaiting a written report from the licensee and final disposition information for the gauge. The State's event number is AR-2017-003.
The following was received via e-mail: A report submitted on July 3, 2017, indicated that the cause of the shutter failure was the exposure to alkaline pulp material and the carbon steel construction of the source holder. The source holder was replaced on January 9, 2018 with a comparable source holder constructed of stainless steel. The Department (Arkansas Department of Health) considers this event to be closed. Notified R4DO (Groom) and NMSS Events Notification via email. |
ENS 52770 | 24 May 2017 12:20:00 | The following information was received from the State of Texas by email: On October 16, 1998, the Agency (Texas Department of State Health Services) was notified that a Humboldt model 5001 moisture/density gauge containing a 10 millicurie cesium - 137 and a 40 millicurie americium - 241 source was lost during transport from San Antonio to Laredo, Texas. The gauge was to be delivered to the Texas Department of Transportation (TXDOT). A search of the transportation companies warehouses and delivery locations along the transportation route did not find the gauge. The investigation was placed in "Inactive" status. On May 17, 2017, the Agency received an email string showing that a moisture/density gauge was for sale on the internet site 'eBay'. A search of the eBay site found that the gauge serial number matched the serial number of the gauge reported missing in 1998. The Federal Bureau of Investigation (FBI) was contacted and a request was made for assistance in gathering information on the seller. Using the information gathered by the Agency and the FBI, the Agency was able to contact the seller. The seller removed the posting off of eBay immediately. The seller stated they purchase materials from companies who are going out of business and resell them. The seller stated they did not remember when or where the gauge was purchased. The seller stated they had just moved all the materials they store in a large warehouse into two smaller warehouses and that is when they discovered the gauge. They did some research on the use for the gauge online and decided to sell it. The seller turned the gauge over to TXDOT on May 24, 2017. Dose rates taken on the gauge by TXDOT were normal. The gauge will be leak tested and returned to the manufacturer. Additional information will be provided as it is received in accordance with SA-300. Event #35040 initially reported the event on 11/16/1998 as a lost source while in transit. Texas Incident: I-7394 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 |
ENS 52760 | 16 May 2017 23:40:00 | The Midwest Inspection Services building has been severely damaged by a tornado storm. All radiography cameras have been accounted for and are stored in a safe location. The vault is intact. Tornados are forecasted to continue throughout the evening.
The following update was received from the Oklahoma Department of Environmental Quality via email: Shortly after 22:30 (CDT) on the evening of May 16th, Oklahoma DEQ (Department of Environmental Quality) verbally notified the HOO (NRC Headquarters Operations Officer) of an event affecting an Oklahoma radiography company. This is a follow-up report to confirm the verbal report and provide more details. Facility Name: Desert NDT, LLC dba Shawcor (note) the facility was historically known as Midwest Inspections, and was accidentally referred to by that name in the initial report. Facility license number: OK-32104-01 Because of concern generated by news reports, DEQ (Oklahoma Department of Environmental Quality) contacted the facility at about 22:00 (CDT) and over the following few minutes, we were able to reach the facility manager. (The facility manager) reported that their licensed facility at Elk City had been largely destroyed by the tornado reported in the media. He indicated that the vault was mostly intact, but had damage to the ceiling. All power at the facility was out. (The facility manager) indicated they (Desert NDT) had fifteen cameras in the vault, and others were out in trucks on jobs around the region. (The facility manager) reported that they (Desert NDT) had done an inventory on the fifteen cameras in the vault, and confirmed that they were accounted for. (The facility manager) had no reports of problems with any sources dispatched on jobs. (The facility manager) explained that they (Desert NDT) did not regard the damaged vault as suitable for secure storage, but they (Desert NDT) had one radiography truck that was largely intact, and they (Desert NDT) were storing the fifteen cameras in one truck, and keeping the truck under constant surveillance by an employee who was authorized unescorted access. Media reports indicated that another storm, weaker than the first, but still having potential tornadoes was headed for the area. In a second call, the facility manager reported that because of concerns about further storms, they had moved the cameras into a storm shelter in (a secure location). (The facility manager) indicated that the storm shelter was under surveillance, and was lockable, and would remain locked unless being directly accessed. In view of the remarkable circumstances, (Oklahoma) DEQ approved this arrangement as an interim measure. About 8:20 (CDT) on the morning of the 17th, (Oklahoma) DEQ contacted the manager again. (The facility manager) indicated that the fifteen cameras were still secured in the storm shelter. (The facility manager) reported that there was no known further damage during the night, and that the company would be conducting a confirmatory inventory of the fifteen cameras, and conducting an inventory to ensure that cameras out on jobs were safe and under control. (The facility manager) will report the results of this to (Oklahoma) DEQ when available. (The facility manager) explained that they were doing an assessment of undamaged trucks that were suitable for secure storage under Part 37, and that they planned to retain some sources at the Elk City facility using the trucks that were suitable. (The facility manager) indicated that excess sources would be moved to a licensed company facility out of state. (The facility manager) will follow up with (Oklahoma) DEQ later today. (Oklahoma) DEQ has used GIS (Geographic Information System) to identify seven other licensed facilities that are near the storm track, and are not considered as having as much concern. We (Oklahoma DEQ) contacted all of them by phone this morning and confirmed that all is well. Notified the R4DO (Miller) and NMSS via email. |
ENS 52753 | 14 May 2017 21:27:00 | On May 14, 2017 at time 1823 (CDT), Waterford 3 Steam Electric Station notified St. Charles Parish Emergency Services via 911 of a fire in the Generation Support Building (GSB), the Hahnville, Luling and Killona Fire Departments were dispatched. The GSB is an Administrative and Engineering Building outside the Protected Area and on the Owner Controlled Area. The fire was reported out at 1841. No personnel were injured due to the fire. The fire appeared to be from an external building exhaust fan. There was no internal or structural damage to the building. There was no radiological release. No Safety Related Systems were required to function. The licensee notified the NRC Resident Inspector. |
ENS 52752 | 14 May 2017 16:09:00 | At 0730 (CDT) on 5/14/2017, a visitor was working in the Protected Area (PA) on the turbine building roof and discovered a blue 12 ounce can of beer in their cooler. This was discovered when the visitor was removing items from their cooler into a larger community cooler. The visitor immediately notified their escort of the prohibited item. The escort then notified Security of the event. Security took possession of the item and the individual was escorted offsite. The individual stated when they packed their cooler at home they thought they had picked up a blue can of soda and did not notice it was a blue can of beer. This event is being reported per 10CFR26.719(b). The licensee notified the NRC Resident Inspector. |
ENS 52749 | 11 May 2017 18:11:00 | A can of alcohol (16.9 ounce foreign beer) was discovered unopened in an administration building refrigerator. Site security took possession of the can of alcohol. The owner of the can of alcohol is unknown. This licensee is making this 24 hour notification in accordance with 10CFR26.719(b)(1). The licensee notified the NRC Resident Inspector. |
ENS 52748 | 11 May 2017 15:24:00 | On Wednesday May 10, 2017 at approximately (1700 EDT), the Reactor Operator (RO) that was signed in on the reactor console logbook completed a ('key on') checklist in preparation for a routine reactor startup. The RO left the control room and brought the log book to the reactor bridge for the Designated Senior Reactor Operator (DSRO) to sign off for the ('key on') startup. The RO immediately realized his mistake concerning the procedural requirement for a reactor operator to be present in the control room at all times when the reactor is not secured (procedure OP-103), and returned to the control room. The DSRO followed the RO to the control room and observed that the reactor key was in the on position, the control rods were all fully inserted, and reactor power was at residual levels. The reactor was shutdown, but was not secured. The DSRO determined that this constituted a violation of procedure OP-103 and could be a Reportable Occurrence as defined under Technical Specification 1.2.24 h. The DSRO reviewed Technical Specification (TS) 6.6.2, Action to be Taken in the Event of a Reportable Occurrence. The DSRO determined that under TS 6.6.2a that reactor conditions had been returned to normal by the presence of the licensed operator in the control room. The DSRO then signed the Key On checklist authorization for reactor startup and the reactor was started. The DSRO spoke with the Manager of Engineering and Operations (MEO) by telephone about this matter at approximately 1800 on May 10, 2017. The MEO concurred that procedure OP-103 was violated and would be reportable to the Nuclear Regulatory Commission (NRC). The DSRO and MEO agreed to discuss this matter with the Director, Nuclear Reactor Program and the Reactor Health Physicist on May 11, 2017. The MEO stated on May 11, 2017 that TS 6.1.3a, the specification implemented by procedure OP-103, was not met. It was agreed that required notifications to NRC would be made by (1700) on May 11, 2017 to meet the 24 hour notification requirement. |
ENS 52993 | 26 September 2017 15:39:00 | An endocrinologist specified a therapy dose of 20 milliCuries of I-131. An authorized dose directive was incorrectly written for 30 milliCuries of I-131. The patient was administered the initially determined dose of 20 milliCuries of I-131. Medical personnel determined that there was no impact on the patient. Hospital supervision notified the on-site Authorized User, the Radiation Safety Officer and the Medical Physicist. 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.
The actual event date was May 1, 2017, and the discovery date was September 25, 2017, at approximately 1500 MDT. Notified R4DO (Proulx) and NMSS Events Notification (via email). |
ENS 52720 | 1 May 2017 11:30:00 | The following information was received from the Commonwealth of Kentucky via facsimile: KY RHB (Kentucky Radiation Health Branch) Inspector, Christopher Keffer, was performing a routine health and safety inspection of the licensee when the RSO (Radiation Safety Officer) discovered that a stored device was missing. According to the RSO, the laboratory where the device was stored was cleaned out the week before; it is currently believed that the device has been thrown away and is now in a landfill. The sealed source identification number is NR-536-D-808-B associated with a Perkins Elmer Clarus Model 500. The source is a Ni-63, 15 microCurie source. 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 |
ENS 52746 | 11 May 2017 12:16:00 | The following information was received from the State of New Jersey via facsimile: Notifications: Phone call was made to the State of New Jersey Department of Environmental Protection (NJDEP) Bureau of Environmental Radiation on 5/10/17. The event occurred on 4/26/17. Event Description: PADEP (Pennsylvania Department of Environmental Protection) staff notified NJDEP staff of a package that was transported by a (New Jersey) pharmacy from a (Pennsylvania) nuclear medicine office. It appears that the package was accompanied by an inaccurate bill of lading and package label/(Transport Index). and brought to the licensee's (Somerset, New Jersey) facility. Investigation is ongoing. |
ENS 52686 | 18 April 2017 10:17:00 | The following information was received from the State of New Jersey via facsimile: The RSO (Radiation Safety Officer) for this cardiology office called to report a lost/missing Cs-137 dose calibrator vial source. Control of this facility was recently transferred to a medical center. When the new RSO visited the cardiology office to become familiar with it, it was discovered that the Cs-137 source was missing. The source, as listed on the cardiology office's inventory, contained 199.04 uCi (microCuries) of activity as of its calibration date of 9/1/2005. The decayed source would contain approximately 152.2 uCi (microCuries) of activity as of the date of this notification. The manufacturer and model # of the source were not immediately available. The RSO will follow-up with a written report within 30 days. 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 |
ENS 52673 | 11 April 2017 14:28:00 | The following information was received from the State of Utah via email: (University) Radiological Health personnel responded to an incident involving a damaged tritium exit sign at the University Guest House. It was determined the damaged exit sign was leaking tritium and the licensee notified the (Utah) Division of Waste Management and Radiation Control. This incident report is the initial notification of the NRC Operations Center. Utah Event Report: UT170003 |
ENS 52778 | 31 May 2017 07:50:00 | This 60-day optional telephone notification is being made in lieu of an LER submittal, as allowed by 10 CFR 50.73(a)(1). This notification is made pursuant to the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of one of the systems listed in 10 CFR 50.73(a)(2)(iv)(B). On April 6, 2017, at 1212 Eastern Daylight Time (EDT), an invalid actuation of emergency diesel generators (EDGs) 1, 2. 3. and 4 occurred. In support of maintenance associated with the onsite electrical distribution system, activities were in progress to power the 2C balance-of-plant (BOP) bus from the startup auxiliary transformer (SAT) followed by de-energization of the 2D BOP bus. However, flexible links between the SAT and the 2D BOP bus had not been installed. As a result, under voltage sensing relay (27SX) was not energized and an invalid SAT secondary side under voltage EDG auto start signal was generated. There was no actual under voltage on the SAT, no loss of power, and all emergency buses continued to be powered by the unit auxiliary transformer (UAT). The EDGs responded properly to the auto-start signal. The actuation was complete, in that the EDGs successfully started and ran unloaded. The EDGs were returned to standby status by 1415 EDT. Since no actual under voltage condition existed which required the EDGs to start, and the start was not in response to actual plant conditions satisfying the requirements for initiation, this event has been determined to be an invalid actuation. This event did not result in any adverse impact to the health and safety of the public. The licensee notified the NRC Resident Inspector. |
ENS 52655 | 31 March 2017 19:14:00 | On March 31, 2017 at 1155 hours (EDT), with the reactor at 97% core thermal power and steady state conditions, operators inadvertently caused water level to rise in the Pressure Suppression Pool (TORUS). Pilgrim Nuclear Power Station (PNPS) was restoring normal system valve line-ups after performing flushing of the suction piping of the Core Spray system in accordance with station procedures. During the process of restoring the appropriate valve line-ups, water was inadvertently transferred to the TORUS from the Condensate Storage Tank. The cause of the event is understood. The Technical Specification (TS) Limiting Condition for Operation (LCO) Action Statement (AS) 3.7.A.5 was entered. The LCO AS was exited at 1540 when TORUS water level was restored to the limits specified in LCO's 3.7.A.1.b and 3.7.A.1.m. Because the TORUS was declared inoperable, PNPS is providing an 8-hour non-emergency notification in accordance with 10 CFR 50.72(b)(3)(v)(D), an event or condition that could have prevented fulfillment of a safety function needed to mitigate the consequences of an accident. This was a case of the water level in the TORUS being above the TS limit. The TORUS was potentially available to provide cooling to the reactor if required. The NRC Resident Inspector has been notified. The licensee notified the Commonwealth of Massachusetts and Plymouth County. |
ENS 52652 | 30 March 2017 21:58:00 | At 1630 PDT on March 30, 2017, a non-work related fatality occurred on the Diablo Canyon Power Plant property. The individual's work location was outside of the Protected Area. The fatality was not related to the health and safety of the public. Specifically, a contractor for Pacific Gas and Electric (PG&E) was found in the Security Training Building unresponsive. The individual was promptly attended to by Diablo Canyon personnel. The individual was subsequently pronounced dead by the San Luis Obispo County Paramedics. PG&E has not observed any heightened public, media, or government concerns as a result of the fatality. Because the fatality is unrelated to Diablo Canyon Power Plant industrial or radiological health and safety, no news release is planned. Because the fatality was not work related, nor the result of an accident, no notification to other government agencies was made at the time. However, PG&E will make a notification to the California Occupational Safety and Health Administration. Thus this ENS notification is in response to a notification to another government agency in accordance with 10 CFR 50.72(b)(2)(xi). The NRC Senior Resident Inspector and Resident Inspector have been notified. |
ENS 52654 | 31 March 2017 16:15:00 | The following information was received from the Commonwealth of Kentucky via facsimile: On 3/28/2017 the licensee left a Cs-137 brachytherapy sealed source at (address provided). On 3/30/2017 the licensee discovered the source was not in its shielded container and immediately determined the location of the source and took steps to retrieve and secure it. This event is actively being investigated by the licensee. Kentucky Event: KY170003 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 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) |
ENS 52555 | 16 February 2017 13:03:00 | On February 15, 2017 at 1515, it was discovered by corporate Fitness for Duty (FFD) personnel that an unescorted access reactivation feature in the security database (Illuminate) does not reset the flag to include an individual in the random FFD pool due to a database coding error. The Illuminate database was implemented fleet-wide 1/3/17. Review by corporate FFD personnel found one individual currently badged at Clinton Power Station was affected by the coding error. The individual was not in the FFD random pool from 1/3/17 until 2/15/17. Corporate security personnel found no other individuals to be affected by this issue. Affected individual was added to the FFD random pool. Corporate security personnel notified all Exelon sites of the issue. Sites were notified that the ability to use the re-activation feature in Illuminate would be removed from use by site personnel. Pending removal, a daily query would be run in the database to assure the re-activation feature had not been used by site personnel. The licensee informed the NRC Resident Inspector. |
ENS 52437 | 15 December 2016 11:47:00 | On December 15, 2016, at 1010 EST, the startup of the Reactor Building HVAC (Heating Ventilation and Air Conditioning) system resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met for approximately 1 second. The maximum secondary containment pressure observed during that time was approximately 0.044 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit of 0.125 inches of vacuum water gauge (TS SR 3.6.4. 1.1) by Reactor Building HVAC and Standby Gas Treatment System already in operation. There were no radiological releases associated with this event. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector.
In this event notification, DTE Electric Company (DTE) reported conditions whereby the Fermi 2 secondary containment was believed to have exceeded Technical Specification Surveillance Requirements due to high winds. DTE hereby retracts this event notification as the Fermi 2 secondary containment has been determined to have been operable during this event as described below. The Fermi 2 secondary containment pressure is maintained at a pressure less than the external pressure to contain, dilute, hold up, and reduce the activity level of fission products prior to release to the environment, and to isolate and contain fission products that are released during a Design Basis Accident or certain operations. Secondary containment pressure is monitored by a number of differential pressure (dP) sensors. High wind gusts have resulted in momentary negative pressure on the leeward side of the building, causing a more positive pressure indication from one or more dP sensors. The secondary containment building pressure remains relatively constant during these 'wind events.' In December 2016, DTE implemented a software design change to display a 120-second rolling average for secondary containment dP indication. A 120-second rolling average recorded every second provides the operator a more accurate report of actual secondary containment conditions, while mitigating the signal noise and wind gust effects. The conditions associated with the subject event notification were re-reviewed in light of the improved secondary containment dP indication and it was determined that the Fermi 2 secondary containment was operable during this event. Specifically, the secondary containment pressure did not exceed Technical Specification Surveillance Requirements during this event. In summary, the above event notification is retracted because the Fermi 2 secondary containment was determined to have been fully operable during the conditions identified in the subject report. The licensee notified the NRC Resident Inspector. Notified R3DO (Stoedter). |
ENS 52432 | 14 December 2016 15:10:00 | On December 14, 2016, at 1314 EST, the startup of the Reactor Building HVAC (Heating, Ventilation and Air Conditioning) system resulted in the Technical Specification (TS) for secondary containment pressure boundary not being met for approximately 1 second. The maximum secondary containment pressure observed during that time was approximately 0.07 inches of vacuum water gauge. Secondary containment pressure was returned to within the TS operability limit of 0.125 inches of vacuum water gauge (TS SR 3.6.4.1.1) by Reactor Building HVAC and Standby Gas Treatment System already in operation. There were no radiological releases associated with this event. Declaring secondary containment inoperable is reportable under 10CFR50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function needed to control the release of radioactive material. The licensee has notified the NRC Resident Inspector.
In this event notification, DTE Electric Company (DTE) reported conditions whereby the Fermi 2 secondary containment was believed to have exceeded Technical Specification Surveillance Requirements due to high winds. DTE hereby retracts this event notification as the Fermi 2 secondary containment has been determined to have been operable during this event as described below. The Fermi 2 secondary containment pressure is maintained at a pressure less than the external pressure to contain, dilute, hold up, and reduce the activity level of fission products prior to release to the environment, and to isolate and contain fission products that are released during a Design Basis Accident or certain operations. Secondary containment pressure is monitored by a number of differential pressure (dP) sensors. High wind gusts have resulted in momentary negative pressure on the leeward side of the building, causing a more positive pressure indication from one or more dP sensors. The secondary containment building pressure remains relatively constant during these 'wind events.' In December 2016, DTE implemented a software design change to display a 120-second rolling average for secondary containment dP indication. A 120-second rolling average recorded every second provides the operator a more accurate report of actual secondary containment conditions, while mitigating the signal noise and wind gust effects. The conditions associated with the subject event notification were re-reviewed in light of the improved secondary containment dP indication and it was determined that the Fermi 2 secondary containment was operable during this event. Specifically, the secondary containment pressure did not exceed Technical Specification Surveillance Requirements during this event. In summary, the above event notification is retracted because the Fermi 2 secondary containment was determined to have been fully operable during the conditions identified in the subject report. The licensee notified the NRC Resident Inspector. Notified R3DO (Stoedter) |
ENS 52433 | 14 December 2016 16:12:00 | On 12/13/2016 at approximately 1500 (PST), the AREVA Nuclear Criticality Safety Staff was notified that an administrative IROFS (Item Relied On For Safety) control had not been performed in the ELO raffinate treatment process. The ELO raffinate treatment process requires the sampling of a favorable geometry process tank (IROFS 306) that is discharged to one of two sets of favorable geometry quarantine tanks. When one of the two sets of quarantine tanks is full, the input is diverted to the other set of tanks and the set that is full is recirculated and sampled for U (Uranium) concentration (IROFS 307). When both sample results have been confirmed to be acceptable the discharge valve on the transfer line may be unlocked and the raffinate solution transferred to a filter press. Sampling of the process tank was completed as required, however; the quarantine tank transfer line was unlocked and contents were pumped to the filter press without completing the required independent sampling of the quarantine tank. AREVA is conservatively reporting this plant condition under 10CFR70 Appendix A, because an accident sequence that could result in accidental nuclear criticality may not have remained highly unlikely in the absence of IROFS 307. The licensee will notify NRC Region 2. |
ENS 52356 | 8 November 2016 17:36:00 | At 1331 (CST) on November 8, 2016, Farley Nuclear Plant Unit 1 manually tripped from 32% reactor power. The plant was ramping down to remove the main generator from service due to an unrelated issue. 1A SGFP did not respond to control Steam Generator (SG) level as expected when the miniflow was opened per procedure. SG levels lowered due to lower feed flow and the reactor was manually tripped in accordance with plant procedures. All control rods fully inserted and Auxiliary feedwater (AFW) auto started as expected. The Main Steam Isolation Valves were closed to minimize the cool-down. Decay heat is being removed through the Atmospheric Relief Valves. All other systems responded as expected. The plant is currently stable in Mode 3 (Hot Standby). The failure of the 1A SGFP control is under investigation. Unit 2 was not affected. The NRC Resident Inspector has been notified. There is no primary to secondary leakage. |
ENS 52351 | 7 November 2016 16:50:00 | The following was excerpted from information received from the State of New Jersey by email: Event Narrative: On November 7, 2016, during the six-month shutter checks, the pneumatically operated shutter on the Vega source holder Phillips 66-Bayway Tag # PBL002 (source capsule S/N 0321CG) failed to close when tested. Several attempts were made wherein the shutter position indicator seemed to move slightly. It was concluded that the issue was not a failure of the air system controlling the pneumatic shutter actuator. The manufacturer was contacted to assess the problem. Root cause(s) and contributing factors: The source remained in the holder attached to the vessel in its normal operating position. The integrity of the source holder remains intact so there should be no exposures. Source/Radioactive Material/Devices: radioactive level gauge Isotope and activity; manufacturer, model and serial number, leak test results as applicable: The source is a 50 mCi Cs-137 solid sealed source, S/N 0321CG. The last leak test was 10/19/15. The equipment is an Ohmart/Vega model SH-F1A source holder mounted to a vessel. New Jersey Case Number: 161107162023
The following update was received from the State of New Jersey via email: Vega serviced the unit on November 11, 2016, with lubrication. It was considered to work properly after service. Phillips 66 notified NJDEP (New Jersey Department of Environmental Protection) on November 30, 2016, that a new unit will be purchased to replace the old unit. Event status is 'closed'. Notified R1DO (Bower) and NMSS_Events_Notification via email. |
ENS 52410 | 7 December 2016 09:15:00 | Agilent Technologies is a manufacturer of a part containing an electron capture detector (ECD) that fits into a gas chromatograph. The ECD contains an embedded sealed source (Ni-63; 15 milliCuries) and is manufactured in Shanghai, China and is transferred into the United States through the JFK Airport Worldwide Flight Services Warehouse. When the licensee's Philadelphia truck shipping service attempted to retrieve the two ECD sources at the JFK warehouse, the sources were discovered missing. The warehouse was searched without success. The ECD source serial numbers are U30355 and U30356. The model number is 62397AECD. Agilent Technologies holds an NRC license and is located at 2850 Centerville Road, Wilmington, DE 19808.
As a corrective action, the licensee has changed warehouse operations to a different warehouse used to receive devices manufactured in China. Notified R1DO (Welling) and NMSS Events Notification via email. 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 |
ENS 52359 | 9 November 2016 17:50:00 | The following information was excerpted from a facsimile received by SOR: Pursuant to the requirements of 10CFR Part 21, this letter notifies the NRC of a Part 21 condition. Irradiation testing performed since 1984 did not take into account all of the uncertainties associated with reported doses of gamma radiation to nuclear test specimens for qualification testing. SOR contracted services with lsomedix in 1992 for the radiation aging that was performed per SOR nuclear qualification report 9058-102 Revision 1. Although SOR imposed Part 21 reporting requirements, lsomedix did not include SOR as part of their Part 21 notification. The Part 21 was brought to SOR's attention through an inquiry by a nuclear power station. SOR requested a conference with Steris lsomedix which occurred on November 3, 2016. The teleconference confirmed that the subject radiation aging test results report would be affected by the Steris lsomedix Part 21. As a result, corrections are underway per qualification test report 9058-102 regarding the uncertainty calculations. SOR does not have the capability to perform further evaluations to determine if a safety hazard exists as the specific customer application is unknown. The end user must confirm for each application that the qualified life dose + accident dose + 10% of accident dose is less than or equal to the corrected values. SOR is currently identifying all customers potentially affected by this deviation. At the conclusion of this activity, SOR will notify the customers and the U.S. Nuclear Regulatory Commission in accordance with the requirements of 10 CFR Part 21.
The following is an excerpt of an updated Part 21 received via email: Corrections are now complete to test report 9058-102 regarding the uncertainty calculations. The calculations changed from 8% uncertainty to 9.6% uncertainty for the minimum irradiation aging. SOR does not have the capability to perform further evaluations to determine if a safety hazard exists as the specific customer application is unknown. The end user must confirm for each application that the qualified life dose + accident dose + 10% of accident dose is less than or equal to the above noted values. Should you have any additional questions regarding this matter, please contact: Linda Coutts Inside Sales Representative Email: lcoutts@sorinc.com Tel 91.3�-956�-3071 Charles Lautner Customer Service Manager Email: clautner@sorinc.com Tel 913-956-3070 Notified R1DO (Bower), R2DO (McCoy), R3DO (Stoedter), R4DO (Haire), NMSS_Events_Notification, and Part 21/50.55 Reactors via email. |
ENS 52354 | 8 November 2016 11:44:00 | In accordance with 10CFR52.99(c)(2), V.C. Summer Units 2 and 3 Construction is making this notification to NRC for determining that Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) 2.6.01.02.ii (Seismic Qualification of Reactor Coolant Pump Switchgear) for both units requires additional actions to restore its completed status. The Closure Notification for this ITAAC (NRC Index No. 580) was originally submitted on February 29, 2016 (reference ML16060A344 and ML16060A345). On November 2, 2016, it was determined that modifications to the RCP switchgear cabinet design were required to ensure compliance with the applicable portions of IEEE 384, Standard Criteria for Independence of Class 1E Equipment and Circuits. The modification involved an engineering change which adds different equipment to the RCP Switchgear cabinet which function to trip the RCP. The new components were not previously seismically qualified for use in the RCP switchgear cabinet assembly. The additional components have now undergone seismic qualification testing for use in the RCP switchgear. The Equipment Qualification Data Package and Equipment Qualification Summary Report for the RCP switchgear will be revised based on the results of the testing to confirm the switchgear withstands seismic design basis loads. The revised testing report has been completed on November 8 2016. The licensee notified the NRC Resident Inspector. |
ENS 52348 | 7 November 2016 12:28:00 | The following information was received by email: On November 2, (2016), the registrant's representative reported (to the State of Illinois) that a generally licensed gauge under their control had failed to operate as designed. A Gamma Instruments model GR100 device exhibited signs of the shutter not closing properly. A contractor was notified and repairs affected. The gauge was subsequently returned to service with no additional engineering issues noted. Procedures at the plant were modified such that the area around the gauge is now cleaned following each shift. The fixed gauge serial number was 930706, and the sealed source was 37 GBq of Am-241. Illinois Event: IL 16010 |
ENS 52269 | 28 September 2016 14:57:00 | The following was excerpted from an email received from WECTEC LLC: Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply. The two flanges identified with deviations on Passive Core Cooling System pipe spools for the Vogtle Unit 3 AP1000r project had incorrect raised-face dimensions. This appears to have been caused by the two flanges being transposed due to an inadvertent fabrication error that occurred at the pipe spool supplier's facilities (CB&I Laurens). The error was subsequently discovered after delivery to the fabrication facility (Aecon Industrial). This error resulted in conditions where the two flanged connections would not have met the design configuration. If the flanged connections had been assembled in the delivered configuration, it is not known if system integrity and operability would have been maintained during operation. The incorrect configuration could have also led to subsequent failure after installation and operation. Hydrostatic testing of these connections is required, but had not yet been performed because the condition was discovered prior to the assembly and testing of these portions of the system. The condition is being corrected prior to the performance of that hydrostatic testing, therefore it is not known if the flanges in the incorrect configuration would have been able to pass hydrostatic testing. Due to the possibility that system integrity and operability could have been impacted by the use of the incorrect flanges, it has been conservatively concluded that this condition should be reported under 10 CFR Part 21. This conservative conclusion is based on the possibility that the Passive Core Cooling System could have been adversely impacted by the identified deviations, if the deviations had been left uncorrected. The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action. . . . The flange configuration was corrected and the Q223 Mechanical Module was delivered to the Vogtle Unit 3 site on September 23, 2016. A corrective action report has been entered into the Westinghouse/WECTEC system to further evaluate the circumstances that led to the identified deviations.
WECTEC LLC determined that additional pipe spools with incorrect flange configurations were fabricated for V.C. Summer Unit 3 and Vogtle Unit 4. None of the pipe spools were installed in either of the facilities. Corrective actions have been taken to prevent re-occurrence. Notified R2DO (Ehrhardt) and Part 21 group via email. |
ENS 52266 | 27 September 2016 22:27:00 | On September 27, 2016 at 1644 (EDT), damaged ductwork was identified in the secondary containment boundary associated with reactor building zone 3 (Units 1 and 2) recirculation plenum. The size of the hole in the secondary containment boundary was determined to be 22.5 square inches. Due to exceeding allowable total leakage in the current secondary containment isolation configuration, a violation of SR 3.6.4.1.5 (occurred). Action to establish a tested configuration with sufficient inleakage margin to restore compliance with SR 3.6.4.1.5 was completed September 27, 2016 at 2115 hrs. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and per the guidance of NUREG 1022, Rev 3, Section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System. The licensee notified the NRC Resident Inspector.
Following the 8 hour 10 CFR 50.72 notification made on September 27, 2016 (EN 52266), further engineering analysis determined that the as-found tear in the Zone 3 ductwork did not impact the ability of Secondary Containment to perform its safety function and that Secondary Containment was not inoperable as a result of the condition. To support the determination, a drawdown test was conducted in a limiting configuration (i.e. least inleakage margin). No substantial change in drawdown testing results were observed over the last three tests. These tests spanned over seven years. Additionally, repairs were promptly made to the affected area. As a result, this event notification is being retracted as it is not reportable pursuant to 10 CFR 50.72(b)(3)(v)(C). The licensee notified the NRC Resident Inspector. Notified the R1DO (Dwyer). |
ENS 52268 | 28 September 2016 13:51:00 | Maintenance was being performed on a Berthold moisture analyzer in a coke bin vessel because a moisture tube inside of the vessel was damaged. Two maintenance technicians took the source holder off the vessel and then proceeded to work on the vessel. The source holder was placed nearby the work area and the source was not able to be fully retracted back into the gauge during the maintenance evolution. Maintenance then re-mounted the source holder back onto the vessel. The licensee contacted the manufacturer in order to investigate the maintenance evolution and assist estimating dosage to the two technicians. The licensee contacted Region 4 (Janine Katanic). The licensee investigation continues.
On 09/29/16 Berthold Technologies was onsite to assist in the investigation. A leak test was performed and the contractors involved were interviewed. The information will be provided to the Berthold RSO in order to estimate the dose received. NRC R4 (Dykert) was also present to observe the investigation. The licensee does not anticipate receiving the results until next week. Notified R4DO (Groom) and NMSS Events Notification by email. |
ENS 52263 | 26 September 2016 16:14:00 | The following information was provided by the State of Arkansas via email: On September 23, 2016 at 1300 CDT, the licensee contacted the Department (Arkansas Department of Health) reporting that during operations on September 22, 2016, the licensee discovered that the shutter on a Vega Americas Model BAL density gauge would not open. The gauge contains 300 milliCuries of Krypton-85. The gauge was replaced with a spare gauge and has been placed in a secure storage area and radiation exposure is maintained at less than 2 mR/hr. The licensee is contacting the manufacturer to request repair or disposal of the gauge. In accordance with RH-1502.f.2 (10 CFR 30.50(b)(2)) the malfunctioning shutter is reportable within 24 hours. The State of Arkansas is awaiting a written report from the licensee. The State's event number is AR-2016-011.
The following update was received via e-mail: The licensee stated that the shutter solenoid was 24 years old and had been in continuous service over that time. The shutter failed to open on September 22, 2016 and the gauge was removed from service and placed in storage. On November 28, 2017, the manufacturer took possession of the gauge for disposal. The Department considers this event to be CLOSED. Notified the R4DO (Deese) and NMSS Events Notification (E-mail). |
ENS 52253 | 19 September 2016 21:40:00 | At 1550 (CDT) on September 19, 2016, Dresden received the Methyl Iodide Penetration test results for the Control Room Emergency Ventilation (CREVS) charcoal. The test results did not meet technical specification acceptance criteria. This results in the inoperability of CREVS. CREVS is a single train system and therefore is reportable per 10CFR50.72(b)(3)(v)(D). The Air Filtration Unit (AFU) is required to operate during a design basis accident to maintain Main Control Room habitability. This places unit 2 and unit 3 in a 7 day LCORA (Limiting Condition of Operation Required Action) per Tech Spec 3.7.4 Required Action A.1. The licensee notified the NRC Resident Inspector.
The licensee is retracting this report based on the following: The purpose of this notification is to retract ENS notification 52253 made on September 19th, 2016, for Dresden Nuclear Power Station. After further evaluation and testing, it has been determined that the Control Room Emergency Ventilation System (CREVS) charcoal would have fulfilled its safety function given the Methyl Iodide Penetration test results. The initial tests were performed with a 2 inch bed depth due to a difference in batches used in each charcoal filter, but testing at a 4 inch bed depth is the correct testing methodology for Dresden's configuration. At a 4 inch bed depth, the test results met the Technical Specification acceptance criteria with significant margin. Therefore, this event does not meet the criteria of 10 CFR 50.72(b)(3)(v)(D) and the ENS report is being retracted. The NRC Resident Inspector has been notified. Notified R3DO (Orlikowski). |
ENS 52251 | 19 September 2016 14:30:00 | EVENT DESCRIPTION: It was determined at approximately 6 AM today (Eastern) that an Item Relied on for Safety (IROFS) associated with a Fuel Manufacturing Operation (FMO) exhaust system was not operating as required. An FMO scrubber exhaust system blower was determined to be not operating and resulted in a failure to meet performance requirements. The safety function of the scrubber exhaust system is to limit the release of uranium hexafluoride (UF6) and its byproducts to the environment in the unlikely event of an accidental airborne release in a process area. Other upstream controls remained available and reliable and prevented significant quantities of UF6 and its byproducts from being released into the scrubber exhaust system. There was no release of material and at no time was an unsafe condition present. The Dry Conversion Process has been shutdown. An investigation is continuing which will provide additional corrective actions and extent of condition. While this did not result in an unsafe condition, the event is being reported pursuant with the reporting requirements of 10CFR70 Appendix A (b)(2) within 24 hours of discovery. SAFETY SIGNIFICANCE OF EVENT: At no time was an unsafe condition present SAFETY EQUIPMENT STATUS: The Dry Conversion Process (DCP) was shutdown. STATUS OF CORRECTIVE ACTIONS: Additional corrective actions, extent of condition, and extent of cause are being investigated. There was no offsite release of UF6 as a result of the IROF failure. The licensee will inform the State of North Carolina, New Hanover County and the NRC Resident Inspector. |
ENS 52249 | 19 September 2016 13:19:00 | The following information was received from the State of Illinois by email: The licensee's radiation safety officer contacted the Agency (Illinois Emergency Management Agency) on September 16, 2016 to advise that while a licensed contractor was conducting calibration of sensors, the shutter failed on an older KayRay fixed gauge. The gauge is mounted on a process line in a remote location that is normally not accessible. Workers were putting the gauge back into service when it appeared that a shear pin broke on the lever that controls the shutter position. The area was cleared of workers and radiation measurements performed. Those measurements showed the shutter was not completely closed. No exposures to workers were noted. Arrangements are being made with a specifically licensed contractor to make repairs to the gauge as soon as can be managed. Fixed gauge is manufactured by KayRay/Sensall; Model 7063P; Serial Number S98L2001; Sealed Source Gauge (.331 Ci of Cs-137). Illinois Event: IL 16006 |
ENS 52246 | 16 September 2016 16:26:00 | The following information was received by the Commonwealth of Massachusetts: The licensee called the Agency (Massachusetts Radiation Control Program) to report a fixed fill level gauge stuck in the open position. The gauge is a Berthold LB7400 Series gauge, Model LB7440, and contains a 150 mCi Cs-137 source (S/N:1476). The gauge is mounted to a liquid holding tank which was emptied and was scheduled for routine cleaning. Before cleaning of the tank could begin, the gauge shutter was attempted to be put into the closed position. When attempting to close the shutter, the handle broke off and left the gauge stuck in the open position. Cleaning was not able to be performed and no one was able to enter the tank. The gauge remains in the installed position and is in an inaccessible area with safety controls in place to prevent access. The manufacturer was contacted and will be servicing the gauge 09/27/16 thru 09/28/16 by either repairing or replacing the shielding as necessary. The Agency continues to investigate and considers this event to be open. Manufacturer of gauge notified and will respond to assess the device; repair or replacement as necessary. |
ENS 52243 | 15 September 2016 16:10:00 | The following information was received from the State of Oklahoma via email: On 9/12/2016 at approximately 1500 CDT, a portable gauge was struck by a private vehicle at a jobsite. This is preliminary information, and (The Oklahoma Environmental Agency) has not received a full report yet from the licensee. There does not appear to be any leakage or exposures from this incident. Licensee: Oklahoma Department of Transportation (OK-15794-01) Reported by: Larry Hawkins, ODOT RSO Device: Troxler Model 4640 Isotope: 8 mCi Cesium-137 Location: Intersection of Highways 183 & 152, Cordell, OK Description: Licensee reports that worker was using the gauge to take surface density measurements on an asphalt highway. The gauge shield was open, but the rod was not extended. A vehicle travelling approximately 25 mph ignored warnings, 'straddled' the highway centerline, and struck the gauge. The driver who struck the gauge left the scene after discovering what had happened. The licensee reports that there was no contamination found on the highway or the gauge. The gauge sustained damage to the case and circuit board. The licensee stated that debris was cleared from the shield area and then the shield was closed. The gauge was shipped to Troxler. The licensee is waiting on leak test results, and will submit an incident report to the DEQ (Oklahoma Department of Environmental Quality). NMED Item Number: OK160005 |
ENS 52250 | 19 September 2016 13:33:00 | The following information was received from the State of Illinois by email: The radiation safety officer, (RSO) at GE Healthcare, called to advise that a package of I-125 seeds which had been given to (a common carrier) to deliver to Sydney, Australia had not reached their destination by the desired date. The package is 6x6x6 inches box of 'excepted package - limited quantity' which contains 10 sealed sources with a total activity of 6.8 milliCi (actual) in a box calibrated for Saturday, 8/27/16, at noon. The RSO has been in touch with (common carrier) since September 3, 2016 when GE's customer service became concerned while routinely performing package tracking. (The common carrier) last had positive tracking for this package in San Francisco, CA on August 28th. (The common carrier) is still conducting visual surveys in San Francisco and Sydney in an attempt to locate the box and has no evidence to indicate the materials have been delivered to an outside entity. Surveys conducted at O'Hare Airport in Chicago where the package was initially offered for transport were negative. (The common carrier) reports that it is still conducting searches and has put a 'world wide tracer' into effect to locate the material as it still believes it is within their delivery system. Illinois Event: IL 16007 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 |
ENS 52203 | 26 August 2016 10:34:00 | The following information was received from the Commonwealth of Pennsylvania via facsimile: Notifications: On August 25, 2016, the Department (Pennsylvania Bureau of Radiation Protection) was notified that a licensee had a cobalt-60 (Co-60) source rod that became bent and unusable during installation into a gauge at a work location. This is reportable per 10 CFR 30.50(b)(2). Event Description: During the process of installing a Co-60 source rod and moving it from a transfer shield to a gauge and industrial process mold, the licensee observed difficulty in getting the source rod to insert into the mold. Using a remote handling device, the Co-60 source was partially removed from the transfer shield and briefly (about 30 seconds) inspected. It was noted that the rod was partially bent. The rod was immediately retracted into the fully shielded position in the transfer shield. The transfer shield containing the Co-60 rod was then taken to a designated storage area where it is secured from unauthorized access. The calculated dose to the workers was estimated to be 0.44 millirem. No exposures over regulatory limits occurred. Radionuclide: Co-60; Manufacturer: Berthold Technologies; Source Model: P 2608-100; Gauge Model: LB 300 ML; Source Serial Number: 1766-10-13; Activity: 3.8 milliCuries ACTIONS: The rod has been securely stored and placed out of service until repairs or return can be accomplished by a licensed service provider. The Department has scheduled a reactive inspection. More information will be provided when available. Pennsylvania Event: PA160024 |
ENS 52188 | 17 August 2016 21:30:00 | At 1826 (CDT), Operators identified that off-site 161 kV power source predicted post-trip voltage was below the operability limit of 161.3 kV. (Operators) entered AOP-31, 161 kV Grid Malfunctions, Section 1, 161 kV Grid Instability, and declared House Service Transformers T1A-3 and T1A-4 inoperable. (Operators) entered Technical Specification 2.7(2)c. Per Technical Specification 2.7(2)c, 'Both house service transformers T1A-3 and T1A-4 (4.16kV) may be inoperable for up to 72 hours. The loss of the 161kV incoming line renders both transformers inoperable. The NRC Operations Center shall be notified by telephone within 4 hours after inoperability of both transformers.' Per OPPD (Omaha Public Power District) Transmission, grid conditions are currently stable. OPPD Transmission has successfully raised predicted post-trip 161 kV voltage with all predicted voltages meeting or exceeding operability requirement of 161.3 kV as of 1834. Current post-trip predicted voltage is 162.2 kV as of 2029. Lowest observed actual voltage was 163.7 kV. The 161 kV line and transformers T1A-3 and T1A-4 remained available at all times. The licensee notified the NRC Resident Inspector. |
ENS 52187 | 17 August 2016 17:46:00 | At 1722 (EDT) on 8/17/16, a Past Operability Evaluation (POE) determined the configuration of the Emergency Gas Treatment System (EGTS) flow controllers that existed prior to 0420 on 8/6/16 constituted an Unanalyzed Condition due to not meeting single failure criteria. This POE examined the condition where EGTS may auto-swap from the flow control path in A-Auto to the Standby flow control path upon the start of a Design Basis Event (DBE). The intended design of the EGTS swap over flow control path in Auto to Standby was to detect and respond to an actual failure of the A-Auto flow control path. The unnecessary auto-swap to Standby could prevent the EGTS train configured in Auto from performing its required safety function during a DBE. The POE performed a detailed calculation to determine the release effects due to the failure of the redundant trains of EGTS controllers. These calculations concluded that failure of both trains of EGTS controllers would not result in exceeding the 10CFR100 limits, however this condition was unanalyzed and failed to meet single failure criteria. This condition is reportable under 10CFR50.72(b)(3)(ii)(B), Unanalyzed Condition due to a system required to meet the single failure criterion does not do so. This condition had no impact to the health and safety of the public. The NRC Resident Inspector has been notified. |
ENS 52186 | 16 August 2016 15:49:00 | The following information was received from the State of New Jersey by email: Event Narrative: During a walkthrough of the facility, the night shift RSO (Radiation Safety Officer) discovered that an ABB Industrial Systems, Ltd model LS100 fixed measuring gauge had fallen off of a coal feeder to which it was attached. The area, which is not normally accessed by staff, was cordoned. Readings were taken directly on top of the device, and measured 0.5 to 1.0 mR/h. The device does not have a shutter mechanism. The direct beam was surrounded with lead that was available for shielding. Root cause(s) and contributing factors: Vibration from coal feeder caused the metal mounting bracket to shear. Semiannual preventative maintenance did not identify the issue. Scaffolding is required to reach the gauge for maintenance, which makes it difficult to do more regularly. Isotope and activity; manufacturer, model and serial number: Ra-226, 0.5 mCi, ABB Industrial Systems, Ltd model LS100, serial number R868. New Jersey Incident Number: C612607
The following was received from the State of New Jersey via email: On September 14, 2016, a 30-day written report from BL England Generating Station was received by NJDEP (New Jersey Department of Environmental Protection). This report outlined the corrective actions and the current status of the gauge. As a corrective step, BL England updated their maintenance schedule to include more frequent checks of their gauges. The device has been secured on a lead plate, and staged in a labeled 55-gallon steel drum to be disposed of at a later date when disposal of other site sources is scheduled. It will remain in storage under lock and key until that time. The licensee's next inspection will include a review of the implanted changes. This incident can be closed. NMED incident Number: 160359. Notified R1DO(Noggle) and NMSS Events Resource via email. |
ENS 52182 | 15 August 2016 21:13:00 | While at a construction jobsite at Kenai Airport, a technician using a Troxler moisture density gauge observed a large equipment grader approaching in reverse mode. The technician retreated from the area and the gauge was run over by the grader. Personnel roped off the damaged gauge area and proceeded to monitor for any contamination. The gauge is a Troxler, Model 3440; S/N 37310; Sources: Cs-137 (8 mCi) and Am-241/Be (44 mCi).
The licensee placed the damaged gauge into an over pack container loaded with sand and transported the damaged gauge to a local office permanent storage facility. The storage area is barricaded and is being monitored. The licensee is consulting with the manufacturer for final damaged gauge disposition. Notified the R4DO (Proulx) and NMSS Events via email. |
ENS 52185 | 16 August 2016 15:50:00 | The NIST (National Institute of Standards and Technology) irradiator has several Cs-137 sources used to calibrate instruments. During a calibration process, a 3.6 Ci source did not return to its shielded position. The event occurred in a portion of the building where such events are expected, therefore there were no health or safety consequences to employees, public or the environment. Licensee corrective actions include manually installing a lead plug into the beam port (opening) of the irradiator. The licensee is contacting the manufacturer in order to assist with troubleshooting and repairs. The irradiator (Model 81-12; JL Shepherd; S/N 7132) is currently in a safe and stable configuration. It is noted that the irradiator is NOT a Part 36 irradiator. |
ENS 52181 | 15 August 2016 17:48:00 | On Monday, August 15, 2016 at 1552 (EDT), with the reactor at (about) 70 percent core thermal power (CTP), Pilgrim Nuclear Power Station (PNPS) entered a 24-hour shutdown Limiting Condition for Operation Action Statement (LCO-AS) for Salt Service Water (SSW) inlet temperature exceeding the Technical Specification (TS) limit in TS 3.5.B.4. The LCO-AS was subsequently exited at 1651 hours when the temperature of SSW trended to below the TS limit. Under certain design conditions, the SSW system is required to provide cooling water to various heat exchangers such as the Reactor Building Closed Cooling Water (RBCCW) and Turbine Building Closed Cooling Water (TBCCW) systems. When the inlet temperature to these supplied loads exceeds the 75 degrees F limit established in the TS, the SSW system is conservatively declared inoperable until the temperature trends below this value. This condition existed for approximately 60 minutes. The SSW temperature is being closely monitored and trended on a continuous basis. This event has no impact on the health and safety of the public. The licensee has notified the NRC Senior Resident Inspector. This notification is being made in accordance with 10 CFR 50.72(b)(3)(v)(B) and 10 CFR 50.72(b)(3)(v)(D) due to an event or condition that could have prevented fulfillment of a safety function. The licensee will be notifying the Commonwealth of Massachusetts Emergency Management Agency. |
ENS 52180 | 15 August 2016 15:32:00 | This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. The Technical Support Center (TSC) was removed from service on 08/15/2016 at 1030 (EDT) for a scheduled facility upgrade project, which will improve the overall functionality of the facility. The duration of the upgrade is expected to be 26 days. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii), as the TSC will be unavailable for greater than 72 hours. In the event of an emergency, McGuire's alternate TSC will be used while the TSC is upgraded. During this period, the alternate TSC will be staffed and activated using existing emergency planning procedures. The Emergency Response Organization team has been notified that the TSC will be unavailable during the upgrade and to report to the alternate TSC in the event of an emergency. This upgrade does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. |
ENS 52179 | 15 August 2016 14:10:00 | This is a non-emergency notification from Waterford 3. On August 12, 2016, at 1704 CDT, the shift operating crew entered Technical Specification (TS) 3.0.3 due to both trains of Essential Services Chilled Water being inoperable. Essential Services Chilled Water Loop A had previously been declared inoperable for maintenance on August 11, 2016. On August 12, 2016, at 1704, the shift operating crew noted that Loop B Essential Services Chilled Water outlet temperature exceeded the allowed TS limit of 42 degrees Fahrenheit. Essential Chiller AB was subsequently aligned to Loop B and TS 3.0.3 was exited on August 12, 2016 at 1802 when outlet temperature was verified less than or equal to 42 degrees Fahrenheit. On August 15, 2016, subsequent review of this event determined that this event was reportable under 10 CFR 50.72(b)(3)(v)(D), 'event or condition that could have prevented fulfillment of a safety function of structures or systems that are needed to (D) mitigate the consequences of an accident' due to both Essential Services Chilled Water Loops being inoperable. The NRC Resident Inspector has been notified. |
ENS 52156 | 5 August 2016 13:58:00 | On 8/5/2016 at 1014 (CDT), the Monticello Nuclear Generating Plant (MNGP) was notified by the Minnesota Department of Health (MDH) of a notice of violation for exceeding the drinking water limit for carbon tetrachloride in the drinking water well that supplies the Security Access Facility. Additionally the MDH will be notifying the Minnesota Pollution Control Agency regarding the violation. As a result, this issue is being reported under 10CFR50.72(b)(2)(xi) for notifications to other offsite government agencies. There was no impact to the health and safety of the general public as a result of this issue. The drinking fountains in the Security Access Facility have been isolated. The NRC Resident Inspector has been notified. |
ENS 52152 | 4 August 2016 17:54:00 | This is a non-emergency eight-hour notification for a loss of Emergency Assessment Capability. This event is reportable in accordance with 10 CFR 50.72(b)(3)(xiii) as the discovered condition affects the functionality of an emergency response facility. A condition impacting functionality due to a loss of cooling of the Technical Support Center (TSC) Ventilation system was discovered on 8/4/16 at 1100 EDT. Repairs are complete. If an emergency would have been declared requiring TSC activation during this period, the TSC would have been staffed and activated using existing emergency planning procedures unless the TSC had become uninhabitable. If relocation of the TSC had been necessary, the Site Emergency Coordinator would have relocated the TSC staff to the Alternate TSC in accordance with applicable emergency plan implementing procedures. The Emergency Response Manager and Site Emergency Coordinator were notified of the condition and the possible need to relocate during an emergency. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. |
ENS 52150 | 3 August 2016 21:38:00 | On August 3, 2016 at approximately 1300 CDT during review of NFPA 805 requirements, it was determined that the NFPA 805 analysis and Fire Safe Shutdown (FSS) procedures do not consider the potential for fire-induced failure of the 4kV Shutdown Board under-voltage trip functions for Emergency Diesel Generator (EDG) power supply alignments. As such, a condition could possibly exist during a postulated fire where a required EDG's 4kV loads would not trip on an undervoltage condition. Current procedures and timeline analysis do not consider operator actions that could be necessary to manually strip the 4kV Safe Shutdown (S/D) board prior to subsequent EDG restart. As such, a subsequent restart, manual or automatic, of the EDG under these conditions, with its associated loads still connected to the 4kV S/D board, could potentially over load the EDG on restart. This notification is to report a condition involving a deficiency in FSS procedures affecting restoration of power to safe shutdown busses under certain postulated fire scenarios. The condition could result in an adverse impact on the ability of operators to implement FSS procedures in response to a postulated fire in 6 fire areas. Therefore, this notification is being made pursuant to 10 CFR 50.72(b)(3)(ii)(B), any event or condition that results in the nuclear power plant being in an unanalyzed condition that significantly degrades plant safety. Compensatory fire watches have been established in the affected areas and this deficiency has been added to the current fire protection impairment plan. The (NRC) Resident Inspector has been notified. |
ENS 52124 | 26 July 2016 06:37:00 | A moisture-density gauge (Troxler - Model 3440, Serial Number 34924) was stolen from a technician's vehicle. The vehicle trunk was broken into, the lockdown chains were cut, and the entire gauge was removed. This model Troxler typically contains 8 mCi Cesium-137 and 40 mCi Americium-241:Beryllium. The licensee has notified local law enforcement and an investigation continues.
The following is a synopsis of information received from the Connecticut Department of Energy and Environmental Protection, Radiation Division (CDEEPRD) via telephone: At 1637 EDT, CDEEPRD received a call from the Connecticut State Police, informing them that the Troxler moisture-density gauge reported stolen this morning, had been located and recovered. It was reported that the gauge was recovered in an intact condition, outside its storage container. Details of how the gauge was located are unknown at this time. Details concerning suspect(s) are unknown at this time. CDEEPRD has an inspector en route to survey the gauge. If the gauge checks out as undamaged, it will be returned to the licensee at that time. Notified R1DO (Welling). Notified via E-mail only - NMSS Events Notification and ILTAB.(Tucker).
The following is a synopsis of information received from CDEEPRD via telephone: An individual attempted to pawn the moisture-density gauge at a pawn shop in Bridgeport, CT. Pawn shop personnel immediately notified the Bridgeport Police Department. Officers arrived on the seen and arrested the individual who was attempting to pawn the gauge. CDEEPRD performed a leak test on the moisture-density gauge. The gauge passed the leak test. The gauge was then turned over to licensee personnel. The moisture-density gauge was not in its storage box when brought into the pawn shop. The licensee placed the gauge in another storage box which they possessed. Notified R1DO (Welling). Notified via E-mail only - NMSS Events Notification and ILTAB.(Tucker). 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 |
ENS 52091 | 15 July 2016 10:27:00 | The following information was received from the State of Colorado via email: Detail: The project manager for a newly constructed apartment complex ordered exit signs to be installed. The distributor (LEI Companies) ordered them from a local supplier (Gexpro). The distributor is unable to locate the exit signs. It is unknown as to what occurred with the signs not being installed at the apartment complex nor being returned to their company. Exit signs are reported as lost. Manufacturer: SRB Technologies, Winston-Salem, NC; Model # BX-10-WH; Serial # C121961, C121962, C121963, C121964, and C121965; Date Shipped: 4-16-15. Designated for: Denizen Apartments, Denver, CO. Ordered through Gexpro, Denver, CO by LEI Companies. Event Description: Gexpro reported the exit signs were picked up by an employee of LEI Companies and were not returned. LEI Companies has documentation to return them, however, no record of exit signs being returned has been located. Denizen Apartments never received nor had exit signs installed. Colorado Event: CO16-I16-12 NMED: C160006 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 |
ENS 52117 | 22 July 2016 11:10:00 | The following information was received by email: Event Description: On July 14, 2016, 3M Corporate Health Physics was notified by laboratory personnel that a static eliminator could not be located during a semiannual physical inventory. The static eliminator is an NRD model P-2001 with an initial activity of 9.56 mCi on September 1, 2015 (activity on July 21, 2016 was 1.89 mCi). The room was recently renovated and the contents of the room were moved to another building with some contents moved to a 3M Distribution Center. Both buildings and the distribution center were searched, but the source was not located. 3M Corporate Health Physics notified the Minnesota Department of Health on July 21, 2016 of the missing static eliminator. Minnesota Event: MN160002 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 |
ENS 52069 | 6 July 2016 21:33:00 | In accordance with 10 CFR 50.72(b)(2)(xi), Duke Energy is notifying the NRC of a report made to the Department of Transportation concerning the identification of removable contamination in excess of 49 CFR 173.443(a) limits. This report was made at 1807 Eastern Daylight Time (EDT). On July 6, 2016, an EnergySolutions 3-60B Transportation Package was received onsite. As a result of receipt surveys, Brunswick Health Physics personnel confirmed removable surface contamination on the transportation package in excess of 49 CFR 173.443(a) limits. The package was shipped as UN2910, Radioactive material, excepted package-limited quantity of material, 7, and was consigned as a non-exclusive use shipment. Surveys identified mixed beta/gamma contamination ranging from approximately 2500 to 4500 dpm/100 sq cm on the surface of the transportation package. All other smears taken on the cask raincover, trailer bed and tires were less than minimum detectable activity for removable contamination. The transportation package is located in a radiological controlled area and access is controlled by Radiation Protection. Surveys have confirmed that the contamination is limited to the surface of the cask. In addition, no personnel contamination events have been attributed to the contamination found on the transportation package. This event did not result in any adverse impact to the health and safety of the public. The NRC Resident (Inspector) has been notified. The safety significance of this condition is minimal. There is no indication of onsite or personnel contamination as a result of this event. The transportation package is controlled in a radiological controlled area and access is controlled by Radiation Protection. The originator of the empty cask arriving at the site (Westinghouse-Pittsburgh) was notified of the contamination. The cask is used for control rod blades and local power range neutron monitoring string shipping. |
ENS 52068 | 6 July 2016 16:48:00 | The following information was received from the State of Tennessee via email: A licensee reported a generally licensed device as missing. Further details will be reported as the information becomes available. The specifics on the device are as follows: Manufacturer- Industrial Dynamics, Model-FT-50B, Serial-113439, Isotope-Am-241, Activity-100 mCi. State Event Report ID No.: TN-16-098 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 |
ENS 52064 | 5 July 2016 20:01:00 | On July 5, 2016, at 1640 Eastern Daylight Savings Time (EDT) the Unit 2 HPCI system was declared inoperable due to apparent failure of the HPCI Auxiliary Oil Pump after the 'HPCI Aux Oil Pump Motor Overload' control room annunciator was received. Failure of the HPCI Auxiliary Oil Pump prevents the HPCI system from performing its design safety function. As such, this event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety function of a system that is needed to mitigate the consequences of an accident. This event did not result in any adverse impact to the health and safety of the public. The NRC Senior Resident Inspector has been notified. The safety significance of this condition is minimal. All other Emergency Core Cooling Systems and the Reactor Core Isolation Cooling (RCIC) system remain operable. Troubleshooting activities are in progress. The HPCI system will remain inoperable until the cause of the failure has been corrected. |
ENS 52047 | 27 June 2016 22:14:00 | A non-licensed contract employee had a confirmed positive for alcohol following a for-cause fitness-for-duty test. The employee's access to the plant has been terminated. This information will be sent to the NRC Region 4 Office as San Onofre no longer has a resident.
In the event report, the individual was described as a contract supervisor, thus meeting the reporting threshold. (the violation of the FFD policy was that he failed a for-cause test for alcohol). Recently, an NRC review of SONGS' 2016 Annual Fitness for Duty report found that SONGS described that test failure as 'not reportable' and that the labor category was 'Maintenance (Craft).' After reviewing the data, Access Authorization confirmed that the individual was a Union Maintenance Foreman (not a supervisor). The event was not reportable." SONGS will be notifying NRC Region 4. Notified R4DO (O'KEEFE) and FFD E-mail Group. |
ENS 52045 | 26 June 2016 23:08:00 | The United States Coast Guard reported an oil sheen in the vicinity of the station's circulating water system effluent. Investigation by station personnel has not determined the source. The circulating water pumps were secured to mitigate the potential source. The United States Coast Guard response Center, and New York State Department of Environmental Conservation have been notified. The licensee notified the NRC Resident Inspector. Notified DOE, EPA, USDA, HHS, FEMA.
The source of the oil sheen has been identified. The source, main turbine lubricating oil, has been stopped and cleanup efforts are underway. Notified R1DO (Gray), DOE, EPA, USDA, HHS, and FEMA. |
ENS 52044 | 25 June 2016 18:30:00 | At 1407 (CDT), during power ascension to 100 percent, turbine control valves closed unexpectedly causing reactor protection trip signals that in turn caused a reactor scram. Reactor scram, turbine trip ONEPs (Off Normal Event Procedure), and EP2 (Emergency Procedure for Level Control) were entered. Reactor water level was stabilized at 36 inches narrow range on startup level and reactor pressure stabilized at 935 psig using bypass valves. No other safety system actuations occurred and all systems performed as designed. All control rods inserted. Reactor level is maintained by feedwater. Normal electrical shutdown configuration is through offsite electrical power sources. The Safety Relief Valves lifted, then closed. The plant is stable at normal level and pressure and remains in Mode 3. The event is under licensee investigation. The licensee notified the NRC Resident Inspector. |
ENS 52043 | 24 June 2016 19:45:00 | On 06/24/2016 at 1511(CDT), an unexpected trip of a Fuel Building ventilation supply fan occurred followed by an exhaust fan trip and secondary containment differential pressure became positive. At 1512 (CDT), the standby fuel building ventilation fans auto started and secondary containment differential pressure was restored to Technical Specification required conditions. Secondary containment was declared INOPERABLE when Technical Specification-required differential pressure was not being maintained and LCO 3.6.4.1 Action A.1 was entered and exited for the given time period. Emergency Operating Procedure (EOP) - 8 was entered due to Secondary containment differential pressure reading positive (greater than 0 inches of water). This loss of secondary containment is reportable under 10CFR 50.72(b)(3)(v)(C) as an event or condition that could have prevented fulfillment of a safety function needed to control the release of radioactive material. The cause of the fuel building supply fan trip is under investigation. The NRC Resident Inspector has been informed. |
ENS 52042 | 24 June 2016 16:06:00 | At 1215 (EDT) on 6/24/2016, James A. FitzPatrick (JAF) was at 100% power when Breaker 710340 tripped and power was lost to L-gears L13, L23, L33, and L43. These provide non-vital power to Reactor Building Ventilation (RBV), portions of Reactor Building Closed Loop Cooling (RBCLC), and 'A' Recirculation pump lube oil systems. Off-site AC power remains available to vital systems and Emergency Diesel Generators (EDG) are available. Due to the loss of RBV, Secondary Containment differential pressure increased. At 1215 (EDT), Secondary Containment differential pressure exceeded the Technical Specifications (TS) Surveillance Requirement SR-3.6.4.1.1 of greater than or equal to 0.25 inches of vacuum water gauge. The Standby Gas Treatment (SBGT) system was manually initiated and Secondary Containment differential pressure was restored by 1219 (EDT). The 'A' Recirculation pump tripped at 1215 (EDT) and reactor power decreased to approximately 50%. 'B' Recirculation pump temperature began to rise due to the degraded RBCLC system. At 1236 (EDT), a manual scram was initiated. Reactor Pressure Vessel (RPV) water level shrink during the scram resulted in a successful Group 2 isolation. All control rods have been inserted. The RPV water level is being maintained with the Feedwater System and pressure is being maintained by main steam line bypass valves. A cooldown is in progress and JAF will proceed to cold shutdown (Mode 4). Due to complete loss of RBCLC system, the Spent Fuel Pool (SFP) cooling capability is degraded but the Decay Heat Removal system remains available. SFP temperature is slowly rising and it is being monitored. The time (duration) to 200 degrees is approximately 117 hours. The initiation of reactor protection systems (RPS) due to the manual scram at critical power is reportable per 10 CFR 50.72(b)(2)(iv)(B) and 10 CFR 50.72(b)(3)(iv)(A). The general containment Group 2 isolations are reportable per 10 CFR 50.72(b)(3)(iv)(A). In addition, the temporary differential pressure change in Secondary Containment is reportable per 10 CFR 50.72(b)(3)(v)(C), as an event that could have prevented fulfillment of a safety function. The licensee notified the NRC Resident Inspector and the State of New York. |
ENS 52038 | 23 June 2016 19:46:00 | The following information was received by the State of Washington via email: The operator of a portable moisture-density gauge temporarily left the gauge unattended at a construction site in Redmond, Washington, and the gauge was run over by the wheel of a roller. The top of the gauge handle was broken off, but both radioactive sources are in safe condition inside the body of the gauge. The gauge operator is maintaining security around the damaged gauge, and has called a nuclear gauge calibration and servicing company to come to the construction site to assess the scene, package the gauge for transport to a safe location, and conduct radiation surveys to verify that there is no radioactive contamination of the construction site and the roller equipment, and to verify that the radioactive sources are undamaged and inside the gauge case. The portable gauge is a Campbell Pacific Nuclear; Model MC-1-DR; Serial Number MD51008063; Sources-Cs-137 (.010 Ci), Am/Be (.050 Ci). Washington Item Number: WA160002 |
ENS 52037 | 23 June 2016 17:13:00 | The following information was received from the State of Texas by email: On June 23, 2016, the licensee notified the Agency (Texas Department of State Health Services) that a radiography camera had failed to lock in position after retracting the source. The ball stop moved about 3/16 of an inch causing the camera to not lock in position after the source was retracted into position. The licensee's radiation safety officer (RSO) obtained the following information about this component failure. The camera was a delta 880 source serial number S7340 at an activity of 52.6 curies. No overexposures were reported to the RSO. An investigation into this event is being conducted by the RSO. The camera has been secured and is located at one of the licensee's sites. Updates will be provided as obtained in accordance with SA300. Texas Incident: I-9415 |
ENS 52036 | 23 June 2016 11:45:00 | The following was received by email: Per your request, I'm sending this email as a follow-up to my telephonic notification to your office that occurred earlier today (re: 10 CFR 20.2201). We're reporting the apparent loss (presently unknown whereabouts) of an aggregate quantity of approximately 95 mCi of Ni-63 housed in the following seven generally licensed devices (GLDs): - 3 ea. x Smiths Detection SABRE 2000s (up to 15 mCi Ni-63 ea.) - 2 ea. x Smiths Detection SABRE 4000s (up to 15 mCi Ni-63 ea.) - 2 ea. x GE VaporTracers (up to 10 mCi Ni-63 ea.) - 19 May 2016: Tinker AFB installation RSO attempted to schedule semiannual swipe sample collection for May 2016. The unit possessing/using GLDs indicated that the GLDs had been transferred. Installation RSO advised unit to present documentation. - 23 May 2016: Unit reported to installation RSO that they were still looking for paperwork. - 13 Jun 2016: After an exhaustive search of records and information management systems, neither the installation RSO or the unit found any documentation for GLD transfer. - 16 Jun 2016: Installation RSO notified the unit commander; continues to investigate. - 20 Jun 2016: Installation RSO notified the USAF Radioisotope Committee Secretariat by telephone and then by email to report situation. The above report of a possible lost or missing source involves sources not specifically licensed under the MML (Master Material License). 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 |
ENS 51960 | 25 May 2016 19:26:00 | The following information was received from the State of Oklahoma by email: Approximately (1300 CDT) today, (Oklahoma was) notified by Advanced Inspection Technologies (AIT) (OK-27588-02) that a radiography truck belonging to them had been involved in a collision with a tractor/trailer truck at (1118 CDT) today at mile marker 178 on I-44 near Stroud, OK. The driver of the radiography truck was killed and the truck partially burned. At the time it was carrying a 30 Ci Ir-192 source and a 25 Ci Ir-192 source. The truck was taken to the Oklahoma Highway Patrol facility in Stroud where (Oklahoma personnel) arrived at about (1400 CDT). The darkroom, where the cameras were stored (QSA Model 880s), was partially burned but had not been opened. Initial surveys of the exterior indicated the cameras, which were normally secured near the darkroom door, had been thrown forward and come to rest just behind the cab. Shortly after (Oklahoma personnel) arrived, the AIT RSO arrived and the darkroom door was forced open. The cameras were recovered and surveys indicated the shielding was intact. Wipe tests of each were also collected. One camera was damaged but the sources were secure inside each (camera). The cameras have been returned to the AIT facility in Sand Springs, OK and will be returned to QSA for repair or disposal. |
ENS 51949 | 23 May 2016 15:36:00 | A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. |
ENS 51931 | 15 May 2016 12:34:00 | Unit 2 experienced RCS Leakage, potentially pressure boundary leakage, or Unidentified Leakage greater than 10 gpm for (greater than) 15 minutes. Reactor Vessel Flange Temperature High Leakoff alarm was received. This met threshold for declaration of an Unusual Event at 1118 CDT per MU6. Unit 2 is currently in MODE 3. Investigation is in progress to identify specific leakage location. The plant is stable, leakage indicates about 32 gpm and the startup has been stopped. There is no impact on Unit 1. The licensee notified the NRC Resident Inspector and State authorities. Notified DHS SWO, FEMA Ops Center, DHS NICC,. Notified FEMA National Watch and Nuclear SSA via email.
At 1459 CDT, the conditions under MU6 are no longer met. The site has terminated the Unusual Event. In addition, a press release will be made of this event. Unit 2 is stable and in Mode 3 pending further evaluation. The leak stopped when a loop drain isolation valve was closed. The licensee notified the NRC Resident Inspector and State and local authorities. Notified the R3DO (Duncan), NRR EO (Morris), IRD MOC (Grant), DHS SWO, FEMA Ops Center, and DHS NICC. Notified the FEMA National Watch and Nuclear SSA via email. |
ENS 51928 | 13 May 2016 20:02:00 | At 1200 (CDT) May 13, 2016, while the plant was operating at 100% power, it was brought to the attention of the River Bend Station Main Control Room staff that an existing design inadequacy could prevent both trains of the Standby Gas Treatment System (GTS) from performing its design function. Under certain specific conditions, the installed Masterpact breakers may not close to allow energization of the filter train exhaust fans. A start signal (reactor level 2, drywell pressure 1.68 psid, annulus high radiation, annulus low flow) combined with a trip signal within a certain time differential, could result in a failure of the breakers to close. As a result of this condition, both Standby Gas Trains were declared inoperable, which required entry into LCO 3.6.4.3 Condition C (requires entering Mode 3 in 12 hours). Declaring both trains of Standby Gas Treatment System inoperable resulted in loss of the safety function since a system that has been declared inoperable is one in which the capability has degraded to the point where it cannot perform with reasonable expectation or reliability. The Standby Gas Treatment System (GTS) limits release to the environment of radioisotopes, which may leak from the primary containment, ECCS systems, and other potential radioactive sources to the secondary containment under accident conditions. At 1240 (CDT) May 13, 2016, one division of GTS, GTS 'A', was manually started from the Main Control Room. This action prevents the breaker failure mode, restored the operability of one train and restored the safety function of the GTS system. LCO 3.6.4.3 Condition A (restore Operability in 7 days) is currently entered for Standby Gas Train 'B'. During the 40 minutes of inoperability, both trains of Standby Gas remained available. At no time was the health or safety of the public impacted. This condition is being reported in accordance with 10CFR50.72(b)(3)(v)(C) as an event that could have caused a loss of safety function to control the release of radioactive material. The Senior NRC Resident was notified.
Further review has determined that the design inadequacy discussed in EN #51928 could adversely effect the ability of the main control building heating, ventilation, and air conditioning (HVAC) system to perform its design safety function, based upon a particular sequence of events occurring within a short window of time (approximately 75 milliseconds). River Bend has implemented compensatory actions to ensure operability of the main control building HVAC system. The Resident Inspector has been notified by the licensee. Notified the R4DO (Miller). |
ENS 51923 | 12 May 2016 19:23:00 | The following information was a licensee received facsimile; Pursuant to 10CFR 21.21(d)(3)(ii), AZZ/NLI is providing written notification of the identification of a potential defect or failure to comply. On the basis of our evaluation, it has been determined that there is sufficient information to determine if the subject condition is left uncorrected could potentially create a Substantial Safety Hazard or could create a Technical Specification Safety Limit violation as it relates to the subject plant applications. The plants will need to evaluate their application to determine if the identified condition could have an impact to the plant operation. The following information is required per 10CFR 21.21(d)(4): (i) Name and address of the individual or individuals informing the Commission. Tracy Bolt, Director of Quality Assurance Nuclear Logistics, Inc. 7410 Pebble Drive Ft. Worth, TX 76118 (ii) Identification of the facility, activity, or the basic component supplied for such facility or such activity within the United States which fails to comply or contains a defect. Masterpact NT and NW style circuit breakers. -The failure of the breaker being ready to electrically close after being subjected to an 'Anti-Pump condition'. Note: The specific application where the failures have occurred is when the breaker is being utilized as a starter for closing into an inductive load like a fan motor. (iii) Identification of the firm constructing or supplying the basic component which fails to comply or contains a defect. AZZ/ Nuclear Logistics Fort Worth, Texas 76118 (iv) Nature of defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply. Possible 'failure to close' condition of Masterpact breakers NT and NW style, that are being used with specific logic schemes that are subjected to 'anti-pump' conditions during normal operation. These breakers have a higher susceptibility to not return to the ready to close position after the close signal has been removed. PSEG reported approximately 14 instances with different breakers in different cubicles where they initiated an electric close order, and the breakers failed to close. All of the 14 instances were in applications of being used to start an inductive load. NLI inspected three of the breakers (all NWs) that were returned by PSEG and could not fully replicate the problem as described by the plant. NLI was only able to repeat the failure to close when performing an 'anti-pump' test. The failure to close was intermittent, but could be duplicated. When the anti-pump condition was not present, NLI could not duplicate a failure to close. Visual inspections of the tested breakers did not reveal any visible damage to the breaker linkages, latches, shunt close or shunt trip assemblies. Schneider Electric (SE) performed testing of three Masterpact NW08 breakers (operated to beyond design life) and duplicated the fail to close condition as described by the plant. It was determined that a standing close signal with a trip/open signal applied is determined to be the root cause of the fail to close issue. The SE testing confirms that the presence of this condition can cause the breaker anti-pump latch to receive excessive forward pressure. When the nose of the latch impacts the close coil plunger, it will 'rock' up in the rear, catching on the top of the mechanism plate. Once the close voltage is removed, and the plunger retracts, the latch may or may not let go. If the latch does not release, then application of the close coil voltage will simply activate the close coil plunger and without the latch underneath the plunger, the breaker will not close. PSEG performed extensive troubleshooting at the Hope Creek plant and discovered that all of the affected breakers were in an anti-pump condition when the breakers failed to close. (v) The date on which the information of such defect or failure to comply was obtained. This revised notification is being submitted based on the information gathered on 5/10/2016 after additional testing, at the request of River Bend, was performed. This additional testing was requested following the notification that was provided to the plants listed below, in the original issue of this letter in February 2016. The evaluation of the condition was originally completed in September of 2012. The issue was originally determined at that time to not be a reportable condition based on the breaker not containing a defect and the condition was believed to be attributed to the specific logic scheme at the plant. To date, this issue has only been reported to NLI from the following plants, PSEG Hope Creek and River Bend Station. No other plants have reported this specific fail to close condition. NLI was in direct communication with the plants when this issue was first being evaluated and the failure analysis were being conducted. The two affected plants were knowledgeable of the condition. (vi) In the case of a basic component which contains a defect or fails to comply, the number and location of these components in use at, supplied for being supplied for, or may be supplied for, manufactured or being manufactured for one or more facilities or activities subject to the regulations In this part. Plants which have been supplied the Masterpact circuit breakers. PSEG Hope Creek - Issue Identified for NW style River Bend - Issue identified for NT style Callaway - This issue has not been identified however, the potential should be evaluated. St. Lucie - This issue has not been identified however. the potential should be evaluated. Turkey Point - This issue has not been identified however, the potential should be evaluated. Beaver Valley - This issue has not been identified however, the potential should be evaluated. Davis Besse - This issue has not been identified however, the potential should be evaluated. Three Mile Island - This issue has not been identified however, the potential should be evaluated. Calvert Cliffs - This issue has not been identified however, the potential should be evaluated. Hatch -This issue has not been identified however, the potential should be evaluated. STP - This issue has not been identified however, the potential should be evaluated. SONGS - This issue has not been identified however, the potential should be evaluated. KHNP Ulchin - This issue has not been identified however, the potential should be evaluated. KHNP Kori - This issue has not been identified however, the potential should be evaluated. Duke Oconee - This issue has not been identified however, the potential should be evaluated. Duke McGuire - Non-safety (not supplied by NU), This issue has not been identified. (vii) The corrective action which bas been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action. NLI originally created a technical bulletin to address the issue and recommendations. However, since new information has been recently identified, NLI TB-12-007 will be revised, as the proposed solution will not reliably solve the problem for all postulated events. Upon completion of the revised technical bulletin, it will be re-submitted to the plants which have been supplied the Masterpact breakers from NLI. (viii) Any advice related to the defect or failure to comply about the facility, activity, or basic component that has been, is being, or will be given to purchasers or licensees. NLI is currently working with the OEM of the circuit breaker to determine the permanent solution to correct the possible failure to close event after the breaker is subjected to an Anti-Pump condition. Advice for plants with breakers currently installed: Evaluate the applications where the breakers may be potentially subjected to an Anti-Pump condition; where the close coil will be energized for an extended period of time. The circuit breaker will continue to operate if this condition is present however there may need to be human interaction with the circuit breaker by manually pressing the trip/open button on the front of the circuit breaker to free the mechanism. Please contact NLI with any questions or comments. Sincerely, Tracy Bolt Director of Quality Assurance
Added 4 additional plants that were inadvertently left off the list. Browns Ferry - This issue has not been identified however, the potential should be evaluated. Fort Calhoun - This issue has not been identified however, the potential should be evaluated. Wolf Creek - This issue has not been identified however, the potential should be evaluated. Seabrook - This issue has not been identified however, the potential should be evaluated. Notified R1DO (Burritt), R2DO (Heisserer), R3DO (Duncan), R4DO (Campbell), and Part 21 Group via email.
The following information was received via facsimile: Additional information in attachment has been updated since the original report provided on 5/13/2016. Additional facility identified as impacted: St. Lucie - Issue identified For additional information contact: Tracy Bolt Director of Quality Assurance AZZ/NLI Nuclear Logistics 7410 Pebble Drive Fort Worth, Texas 76118 Notified the R1DO (Ferdas), R2DO (Rich), R3DO (Kunowski), R4DO (Gaddy) and Part 21 Reactor group (via email).
On the basis of our evaluation, it has been determined that there is sufficient information to determine if the subject condition is left uncorrected could potentially create a Substantial Safety Hazard or could create a Technical Specification Safety Limit violation as it relates to the subject plant applications. The plants will need to evaluate their application to determine if the identified condition could have an impact to the plant operation. Plants which have been supplied the Masterpact circuit breakers: PSEG Hope Creek - Issue Identified for NW style River Bend - Issue identified for NT style Callaway - This issue has not been identified however, the potential should be evaluated. St. Lucie - This issue has been identified. Turkey Point - This issue has not been identified however, the potential should be evaluated. Beaver Valley - This issue has not been identified however, the potential should be evaluated. Davis Besse - This issue has not been identified however, the potential should be evaluated. Three Mile Island - This issue has not been identified however, the potential should be evaluated. Calvert Cliffs - This issue has not been identified however, the potential should be evaluated. Hatch - This issue has not been identified however, the potential should be evaluated. STP - This issue has not been identified however, the potential should be evaluated. SONGS - This issue has not been identified however, the potential should be evaluated. KHNP Ulchin - This issue has not been identified however, the potential should be evaluated. KHNP Kori - This issue has not been identified however, the potential should be evaluated. Duke Oconee - This issue has not been identified however, the potential should be evaluated. Duke McGuire - Non-safety (not supplied by NLI). This issue has not been identified. Browns Ferry - This issue has not been identified however, the potential should be evaluated. Fort Calhoun - This issue has not been identified however, the potential should be evaluated. Wolf Creek This issue has not been identified however, the potential should be evaluated. Seabrook This issue has not been identified however, the potential should be evaluated. NLI originally created a technical bulletin to address the issue and recommendations. However, since new information has been recently identified, NLI TB-12-007 has been revised. The solution for this potential problem is to replace the XF (shunt close assembly) with the XFCOM shunt close assembly. The part numbers are: 847323 (100-130VAC/DC) 847324 (200-240VAC/DC) Additional details regarding the replacement device are contained in the NLI technical bulletin TB-12-007. Notified the R1DO (Krohn), R2DO (Blamey), R3DO (Jeffers), R4DO (Deese) and Part 21 Reactor group (via email). |
ENS 51899 | 3 May 2016 01:50:00 | At 2229 (CDT) on 05-02-2016, River Bend Station declared the High Pressure Core Spray system INOPERABLE in accordance with Technical Specification 3.8.9, Condition E (Declare High Pressure Core Spray System and Standby Service Water System Pump 2C inoperable immediately) due to Division 1 Control Room Air Conditioning System HVK-CHL1C being INOPERABLE due to a trip of the chiller on high inboard bearing temperature. Actions taken to exit the LCO: Alternated divisions of Control Room Air Conditioning System to Division 2 HVK-CHL1D in service and Division 1 HVK-CHL1A in standby. The licensee notified the NRC Resident Inspector.
Supplement: An operability evaluation has been performed based on system operating procedures in place at the time of this event, and on calculations regarding heat-up rates of the spaces served by the main control room air conditioning system. Operating procedures already in place on May 2 specified the operator actions required to restore the air conditioning system to service following the unanticipated trip of a chiller. The normal shift complement was on duty at the time of the event, and could have provided an adequate number of operators to accomplish this task. The operability evaluation made no new assumptions regarding availability of operators. The manual actions to be performed for the start of an alternate chiller following a trip of an in-service chiller system have been determined to require 2.15 hours, based on ANSI 58.8 guidance. (ANSI/ANS 58.8, Time Response Design Criteria for Nuclear Safety Related Operator Actions, provides the industry guidance In this regard.) Calculations of building heat-up rates have demonstrated that the loss of the air conditioning system can be sustained for 19 hours before temperatures in the rooms containing the Division 3 electrical equipment that support operability of the HPCS system exceed their maximum allowable ambient value. Based on the conclusions of the operability evaluation, the trip of the 'C' HVK chiller on May 2 had no actual adverse effect on the ability of the electrical distribution systems in the main control building to support the safety function of the HPCS system. Event Notification No. 51899 is hereby withdrawn. The licensee has notified the NRC Resident Inspector. Notified R4DO (Rollins). |
ENS 52041 | 24 June 2016 13:20:00 | The following information was excerpted from a received licensee fax: On April 26, 2016, ABB confirmed a customer complaint regarding a 60Q relay received with an internal short circuit. The cause was determined to be a manufacturing deviation from specification. Records show a total of 53 suspect relays were provided to 2 customers (Prairie Island and Tennessee Valley Authority). Based on the nature of the deviation, any installed relay would have failed upon application of power in the relay. Therefore, the primary concern is that any relay, either in storage or installed with no power applied, may have this defect and will create a system malfunction upon power application. If you have any questions regarding this notice, please contact the ABB Technical Support at (954) 752-6700. |
ENS 51877 | 22 April 2016 00:03:00 | Missing fire barrier between Fire Area (FA) 59 and 85. During a walk down of fire barriers for the NFPA 805 project, it was determined that the fire barrier between Fire Area 59 (Unit 1) and 85 (common) is not a rated barrier due to unsealed penetrations in the barrier. Evaluation FPEE 12-006 evaluated the acceptability of the barrier being unrated based on separation of safe shutdown equipment however a review of equipment credited for Appendix R safe shutdown identified that the redundant credited Appendix R equipment is on either side of the fire barrier which is not 3 hour rated. The conclusion of the FPEE is therefore no longer valid. Fire Hazard Analysis Drawings Do Not Match Boundary Description. The plant layout in F5 Appendix F, Rev. 28, Fire Hazard Analysis (FHA), does not agree with the boundary description in the FHA for the Unit 1 and 2 Containment Annulus fire areas, Fire Area (FA) 68 and 72. The layout should but does not show the fire area boundary between the annulus and adjacent fire areas, FA 60 and 75 on 735 (foot) and 61A on 755 (foot), as an Appendix R boundary. The annulus airlock doors are 3-hour fire rated and the airlock is constructed of concrete thick enough to qualify as a 3 hour fire barrier however, there are penetrations in the barrier that are not sealed with fire rated materials or inspected as required by the Fire Protection Program. Therefore, this is an unanalyzed condition reportable under 10 CFR 50.72(b)(3)(ii)(B). This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. |
ENS 51872 | 20 April 2016 01:52:00 | On April 19th, 2016 at 2159 (EDT), Secondary Containment became inoperable due to failure to meet Surveillance Requirement (SR 3.6.4.1.1) on Unit 1 and Unit 2. The inoperability was caused when Reactor Building differential pressure was discovered to be less than Technical Specification requirements (-0.25 inches of water gauge). Secondary Containment was restored April 19, 2016 at 2222 by adjusting intake louvers in accordance with off normal operating procedure ON-RBHVAC-201. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment system. The licensee notified the NRC Resident Inspector. |
ENS 51920 | 11 May 2016 17:37:00 | The following information was received by email: On 4/13/16 CPN Model MC Series gauge was left in marked licensee vehicle parked outside technician's home in Columbus at end of work day, reportedly properly secured in vehicle with two independent locking devices. Gauge contains 10 mCi Cs-137 and 50 mCi Am-241:Be sources. Technician found gauge missing when came out to go to work on 4/14/16. Technician claims RSO was notified, RSO does not recall. No police report was filed and no report was made to ODH (Ohio Department of Health) at that time. On 5/10/16, gauge was found in vacant lot during separate police investigation. Police called fire department HAZMAT unit. Transport case was not locked, but gauge rod was locked. Licensee was identified by paperwork in the transport case. Licensee was contacted by fire department to retrieve gauge. Licensee RSO took possession of gauge and returned it to office in Columbus. On 5/11/16 licensee reported theft and recovery of gauge to ODH. ODH investigators visited site to determine cause of incident and reasons for lack of notifications. Ohio Item Number: OH160003 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 |
ENS 51861 | 11 April 2016 15:37:00 | 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. At approximately 2335 (MST) on April 10, 2016, while performing planned routine visual examinations of Unit 1 components in the Containment Building, engineering personnel identified white residue on the piping instrument nozzle for the reactor coolant system (RCS) 2B cold leg resistance temperature detector, 1JRCET121Y. The white residue was dry and no active leakage was noted on the instrument nozzle. Preliminary chemical analysis for radionuclide and boron content of the white residue determined the existence of boron and the isotopic content was consistent with RCS fluid. At 0535 (MST) on April 11, 2016, it was determined the residue resulted from RCS pressure boundary leakage, based on results of the chemical analysis and additional examination of the piping and instrument nozzle by qualified engineering personnel. Technical Specifications Limiting Condition of Operation (LCO) 3.4.14 permits no RCS pressure boundary leakage and therefore, the discovery of leakage from the instrument nozzle represents a degradation of a principal safety barrier. This notification is being made for a degraded condition pursuant to the requirements of 10CFR 50.72(b)(3)(ii)(A). The unit has been shut down for its 19th refueling outage since 4/9/16 at 0000. The NRC resident inspectors have been informed of this condition. |
ENS 51838 | 31 March 2016 12:45:00 | On March 31, 2016 at 0603 (EDT), with Susquehanna Unit 1 in its 19th Refueling and Inspection Outage, Unit 1 received a valid isolation signal. Preliminary investigation indicates the isolation signal was the result of a human performance error. The systems affected by the isolation signal responded as designed for the current shutdown plant conditions. This isolation of multiple primary containment isolation systems is being reported under 10CFR50.72(b)(3)(iv)(A) and per the guidance of NUREG 1022, Rev. 3, section 3.2.6 as a system actuation. The licensee notified the NRC Resident Inspector. |
ENS 51840 | 31 March 2016 16:12:00 | On 3/31/2016 at approximately 0342 CDT, a worker within the Protected Area self-reported a can of beer had been packed in the worker's lunchbox. The worker reported after opening the can and taking a sip it was discovered to be a beer. This event is reportable under 10 CFR 26.719(b)(1). The worker notified Security who immediately escorted the worker from the Protected Area and disposed of the beer. The worker is not an Operator or a Supervisor. The investigation of this event is in progress. The public health and safety are not impacted. The NRC Resident Inspector was notified. |
ENS 51842 | 31 March 2016 20:03:00 | The following excerpted information was received from the State of Washington by email: Event Narrative: The Washington Department of Health received a call today (3/31/2016) at noon from the RSO of the licensee, to report a stolen gauge from their conex unit at their SeaTac location. The RSO first noticed the missing gauge (the only one stored at this facility) when he went to check it out at 1030 PDT this morning. There was no indication of it being checked out by other users, and the RSO called the other users to make sure. The RSO also noticed the broken lock on the conex door. The licensee's contractor called police for this incident as well as other containers that had been broken into. More information to come. Make/Model/Serial Number- Troxler 3440, SN 31182. Isotopes/Activity - Cesium 137/0.37 GBq and Americium 241 Beryllium/1.85 GBq 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 |
ENS 51839 | 31 March 2016 14:07:00 | The following information was received from the State of Ohio by email: On March 30, 2016 at approximately (1250 EDT) the Bureau of Environmental Health and Radiation Protection (BEHRP) received a phone call from licensee's RSO that an industrial radiography crew working on a water tank construction job in Groveport, Ohio had a source stuck in the guide tube of a QSA Model 880D camera and were unable to retrieve the back into the camera. The source in use was 75 Curies of Iridium-192. The incident occurred at approximately (1225 EDT). The stuck source was discovered after a shot time had ended and the radiography crew attempted to crank the source back in to the camera. The radiography crew conducted surveys of the area and moved boundaries out 2 mR/hr or less. The cause of the stuck source was due to a magnetic stand becoming dislodged during radiography operations, which fell onto the guide tube, crimping it, and preventing retraction of the source.
A BEHRP inspector was immediately dispatched to the job-site and arrived there at approximately (1320 EDT). The inspector met with the licensee's customer and reviewed the actions taken by the radiography crew to establish new barriers and prevent access to the site. The licensee's trained retrieval personnel dispatched to the site arrived a short time later. After a thorough review of the incident and work area, the licensee's response team was able to retrieve the source, which was completed at approximately (1600 EDT) that afternoon. The maximum dose received by any individual involved in the recovery effort was 50 mR. The camera and guide tube will be returned to manufacturer for repair. QSA Global Camera,Serial Number-D8042; Source Serial Number-29222G. Ohio Event- OH160002 Notified the R3DO (Pelke) and the NMSS Events Notification via email. |
ENS 51841 | 31 March 2016 17:58:00 | The following information was received from the State of Colorado via email: The Colorado Department of Public Health and Environment - Radioactive Materials Unit was notified on March 25, at (1041 MDT) of a misadministration that occurred on the evening of March 24, 2016. The licensee is Presbyterian St. Luke's Medical Center. A patient was given approximately 70% of the prescribed SIR-Spheres dose due to a clogged catheter. The investigation is ongoing and a corrective action has not yet been determined. Colorado Report: CO16-M16-02 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. |
ENS 51809 | 20 March 2016 23:51:00 | During a planned Unit 1 shutdown for a refueling outage, a 0.5 gpm 'pressure boundary leak' was identified on a 1 inch pipe connected to the '1A' RHR-Shutdown Cooling return line by the drywell leak inspection team during a drywell inspection at approximately 15% power. The leak exceeded the TS 3.4.3.2 'Operational Leakage' LCO of no pressure boundary leakage. TS action 'a' was entered which requires to be in at least Hot Shutdown within 12 hours and Cold Shutdown within the next 24 hours. Therefore, the event is reportable within 4 hours per 10CFR50.72(b)(2)(i) due to the initiation of a plant shutdown required by the plant's TS. The event is also reportable within 8 hours per 10CFR50.72(b)(3)(ii) due to an event that resulted in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. TS 1.28 defines Pressure Boundary Leakage as leakage through a nonisolable fault in a reactor coolant system component body, pipe wall or vessel wall; therefore, the leak is a 'pressure boundary leak' as defined in TS. The licensee notified the NRC resident Inspector. |
ENS 51795 | 17 March 2016 08:02:00 | On March 17, 2016, at 0115 (EDT), Watts Bar Unit 1 (WBN1) entered Technical Specification Limiting Condition of Operation (TS LCO) 3.0.3 due to the inoperability of both trains of the Emergency Gas Treatment System (EGTS). TS LCO 3.7.12 Condition B was also entered at this time due to the inoperability of both trains of the Auxiliary Gas Treatment System (ABGTS). The train B EGTS and train B ABGTS had been removed from service for scheduled maintenance, when at 0115, the train A Auxiliary Air Compressor became inoperable. On March 17, 2016, at 0133, the train A Auxiliary Air Compressor was declared OPERABLE, and TS LCO 3.0.3 and 3.7.12 Condition B were exited. The auxiliary air system supports the EGTS by providing a safety grade air supply. When train A auxiliary air became inoperable, the supported train A EGTS and ABGTS became inoperable, creating a condition where both trains of EGTS and ABGTS were unavailable. In the event of an accident, the EGTS establishes a negative pressure in the annulus between the shield building and the steel containment vessel and the ABGTS establishes a negative pressure in the Auxiliary Building Secondary Containment Enclosure (ABSCE). Filters in these system mitigate the release of radioactive contaminants to the environment. WBN1 remained in Mode 1 at 100% power and no safety functions were required during the event. This event is reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified.
The purpose of this notification is to retract event report no. 51795 made on 3/17/16 at 0802 (EDT). Previously, Tennessee Valley Authority (TVA) reported a loss of the Emergency and Auxiliary Building Gas Treatment Systems (EGTS/ABGTS) at Watts Bar Nuclear Plant Unit 1 (WBN1). Both trains of EGTS and ABGTS were declared INOPERABLE when the train A auxiliary air system cooling water supply bypass valve was isolated, prior to completing the requisite post maintenance testing following repairs to the normal cooling water supply solenoid valve. Upon recognition, WBN1 operations personnel declared the train A auxiliary air system INOPERABLE, resulting in inoperability of Train A EGTS and ABGTS and forcing entry into TS LCO (Limiting Condition for Operation) 3.0.3 (from TS LCO 3.6.9 EGTS) and 3.7.12 Condition B for ABGTS. At the time the condition was recognized, train B EGTS and train B ABGTS were INOPERABLE for scheduled maintenance. Subsequently, TVA completed the post maintenance testing of the train A auxiliary air system ERCW (Emergency Raw Cooling Water) normal supply solenoid valve and determined that the valve, while not fully qualified at the time, was in fact operable and capable of performing its safety function. Therefore, entry into TS LCO 3.0.3 and 3.7.12 Condition B was not necessary and the event is no longer reportable under 10 CFR 50.72(b)(3)(v)(C) and (D) as a condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified. Notified the R2DO (McCoy). |
ENS 51780 | 9 March 2016 04:01:00 | Watts Bar Unit 2 declared an Unusual Event at 0342 EST based on a fire greater than 15 minutes in the turbine building - 2B Hotwell pump motor. The fire was extinguished by 0401 EST, at the time of notification. Unit 2 is currently shutdown in Mode 5 making preparations for startup. No offsite assistance was requested. All personnel are accounted for and there are no personnel injuries reported. The licensee notified the NRC Resident Inspector. Notified DHS SWO, DOE, FEMA OPS, FEMA National Watch (email), DHS NICC, and Nuclear SSA (email).
The licensee terminated the Unusual Event at 0508 EST based on verification that the fire was out and that the fire response team had been secured. The licensee notified the State and local agencies and the NRC Resident Inspector. Notified R2DO (Suggs), NRR EO (Morris) and IRD (Grant). Notified DHS SWO, DOE, FEMA OPS, FEMA National Watch (email), DHS NICC, and Nuclear SSA (email). |
ENS 51678 | 23 January 2016 19:48:00 | At 1703 (EST) on 1/23/16, with Unit 1 and Unit 2 operating at 100% power, the North Anna 34.5 kv Bus 3, off-site power feed to the 'C' Reserve Station Service Transformer, was lost which resulted in the loss of power to the Unit 1 'H' Emergency Bus and the Unit 2 'J' Emergency Bus. Loss of 34.5kV Bus 3 resulted from feeder breaker L102 opening. As a result of the power loss, the 1H Emergency Diesel Generator and the 2J Emergency Diesel Generator automatically started as designed and restored power to the associated emergency bus. During this event, the Unit 1 'B' Charging Pump, 1-CH-P-18 automatically started as designed due to the loss of power event. The valid actuation of these ESF components due to the loss of electrical power is reportable per 10 CFR 50.72 (b)(3)(iv)(A). The Unit 1 'H' Emergency Bus off-site power source was restored to service and the 1H Emergency Diesel Generator was secured and returned to Automatic. The Unit 2 'J' Emergency Bus power feed continues to be from the 2J Emergency Diesel Generator. Restoration of offsite power to operable status is currently being pursued. The Unit 1 'B' Charging Pump has been secured and returned to automatic. Both units are in a stable condition. An investigation is underway to determine the cause of the L102 feeder breaker opening resulting in the 34.5 kv Bus 3 loss of power. The licensee notified the NRC Resident Inspector |
ENS 51813 | 22 March 2016 11:07:00 | The following report was excerpted from an Oklahoma Department of Environmental Quality via email: On January 22, 2016 the RSO (Radiation Safety Officer) reported that a package received by the University of Oklahoma Health Science Center pharmacy (OK-03176-04MD) was found to have 2,118,858 cpm of removable contamination. The outside of the package was surveyed at 30 mR/hr, the reading at 1 meter was 0.12 mR/hr. The contaminant was identified as Tc-99m and was confined to the outside of the package. The package contained a number of empty unit dose syringes in lead pigs and had been returned from HCA Health Services of Oklahoma which administers the University of Oklahoma Medical Center. The package was transported by the pharmacy. Refer to NRC Event #51661. |
ENS 51646 | 12 January 2016 11:14:00 | Event Report per 10 CFR 26.719(b)(2)(ii) On January 11, 2016, Callaway determined a violation of two provisions of the site Fitness For Duty policy were committed offsite by a non-licensed supervisory employee. Unescorted access for the employee has been denied. The licensee notified the NRC Resident Inspector. |
ENS 51558 | 20 November 2015 21:49:00 | On (11/20/2015 at 1808) CST it was noted that the MET tower (both primary and backup) was offline and not communicating with the Plant Management Information System(PMIS). This results in a major loss of emergency assessment capabilities with respect to meteorological conditions and is reportable under 10CFR50.72(b)(3)(xiii). Communications technicians responded to the plant and the MET Tower communications were restored to PMIS on (11/20/2015 at 1937). The licensee notified the NRC Resident Inspector. |
ENS 51556 | 20 November 2015 14:56:00 | The following information was received from the State of Arizona by facsimile: At approximately (0800 MST), November 20, 2015, the Agency (Arizona Radiation Regulatory Agency) was informed that the Licensee had a Campbell-Pacific Model MC-3, Serial Number MC310700331, portable moisture- density gauge stolen from a truck at an apartment parking lot. The theft occurred before (0500 MST), November 20, 2015. The gauge was the only thing stolen from the truck which had expensive tools in it. The sources were Cs-137, 10 millicuries and Am-241:Be, 50 millicuries. The Agency continues to investigate the event. The AZ Governor's Office, U.S. NRC, the States of CA, CO, NV, and NM are being notified of this event (by the Agency)." Arizona First Notice: 15-025 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 |
ENS 51554 | 19 November 2015 19:10:00 | A review of operations log data revealed the Unit 2 Spent Fuel Pool temperature was allowed to drift approximately two degrees below the accepted analyzed temperature of 68 degrees F. The Extent of Condition is under review and the number of instances has not been fully determined. It is possible Unit 3 may have also been affected. As described in the San Onofre UFSAR, Section 9.1.2.3 'Design Basis for Fuel Storage and Handling,' one of the conditions assumed in meeting the design basis, is as follows: the most adverse delta-k (reactivity change) was analyzed for a pool temperature range from 68 degree F to 160 degree F. Although the temperature decrease is considered unlikely to have a significant impact on the results of the analysis, it is being reported under 10 CFR 50.72(b)(2) as an unanalyzed condition because there is at this time no analysis of record for the reduced temperature condition. There were and continue to be no observed, abnormal effects from the temperature drop below 68 degrees F, and no impact on the health or safety of plant personnel of the public. Re-analysis is in progress and is expected to demonstrate that the criticality acceptance criteria of 10CFR50.68 were met at all times. The licensee notified the Regional NRC Contact. |
ENS 51523 | 9 November 2015 09:10:00 | On Monday, November 9, 2015 at 0800 (EST), planned routine maintenance was initiated on the Technical Support Center/Operations Support Center (TSC/OSC) ventilation system. The planned maintenance is to replace the charcoal filters and test the HVAC trains. All other TSC/OSC functions remain available. Under certain accident conditions the TSC/OSC may become unavailable as a result of the ventilation system not being available. Existing Emergency Procedures direct the responsible Emergency Plant Manager to relocate the TSC/OSC staff to the designated alternate location. The affected Emergency Response Organization facility leads have been informed. The licensee has notified the NRC Senior Resident Inspector. The Commonwealth of Massachusetts will be notified. This notification to the USNRC Operations Center is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the potential loss of an Emergency Response Facility (ERF).
At 1730 EST on Monday, November 9, 2015, the TSC/OSC ventilation system was restored to service. The licensee will notify the NRC Resident Inspector. Notified R1DO (Arner). |
ENS 51628 | 30 December 2015 13:16:00 | This 60-day telephone notification is being made per the reporting requirements specified by 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to describe an invalid actuation of the Emergency AC electrical power systems, due to invalid start of an emergency diesel generator (EDG). On November 6, 2015, at 0815 (CST), EDG A received a fast start signal inadvertently. Maintenance Fix-It-Now (FIN) personnel, working in the vicinity of the relay panel, made incidental contact with one of the relays in the Diesel Generator Logic Relay Panel which initiated the engine start. The engine started and obtained rated speed and voltage in single unit mode. This was not a valid initiation of EDG A. Operations personnel responded to the EDG actuation by ensuring that the engine was shut down and placed in standby condition in accordance to plant operating instructions. The plant conditions at the time of the EDG A initiation would not have actuated the EDG; therefore, the actuation of EDG A was invalid. There were no safety consequences or impact to the health and safety of the public as a result of this event. This event was entered into the Corrective Action Program (CAP) as Condition Report (CR) 1101730. The NRC Resident Inspector was notified of this event. |
ENS 51464 | 9 October 2015 21:59:00 | At 1800 (CDT) on October 9, 2015 a polling test was initiated in Saint Johns Parish to test the circuitry of the installed sirens. During the polling test no sirens are expected to sound as it is only a circuitry test. Siren number SJ39 inadvertently sounded for 15 to 20 minutes, and no others. Saint Johns Parish notified the parish residents that the sounding of the siren was inadvertent via a Parish wide cable television channel and a press release. A contract vendor has disabled the siren and will troubleshoot and repair starting on October 12, 2015. All remaining sirens within Saint Johns Parish remain operational and capable of being activated when required. 0% of the population is affected by the loss of this siren due to siren overlap. Time to repair and restore siren SJ39 to service is still being investigated. This event is reportable pursuant to 10CFR 50.72 (b)(2)(xi), News Release or Notification of Other Government Agency. The NRC Resident Inspector has been notified. |
ENS 51450 | 5 October 2015 08:41:00 | Braidwood Unit 2 was performing a planned plant shutdown for refueling outage A2R18. In accordance with plant shutdown procedures while in Mode 1 (Power Operations) at approximately 15% power, operators attempted to start the Start Up Feedwater (SFWP) pump and the pump immediately tripped on Phase A Overcurrent. The 2A Motor Driven Feedwater pump (MDFWP) was manually started to maintain Steam Generator Water Level during the shutdown and subsequent plant cooldown. While in Mode 3 (Hot Standby) at (550 Degree-F), the 2A MDFWP was manually secured due to pump inboard journal bearing temperature exceeding its (200 Degree-F) operating limit. At 0105 (CDT) an anticipated automatic Auxiliary Feedwater actuation signal was generated on low Steam Generator level (36.3%) and both the 2A and 2B Auxiliary Feedwater pumps (AFP) auto-started. Also at 0105 (CDT) a Reactor Protection System (RPS) Reactor trip signal was received due to low Steam Generator level (36.3%) with the reactor not critical. Both Auxiliary Feedwater trains operated as designed with the Main Steam Dumps in service and the Main Condenser providing the heat sink. All systems operated as designed with the exception of the SFWP and the MDFWP described above. The plant is currently stable in Mode 5 with both AFPs secured. This report is being made per 10 CFR 50.72(b)(3)(iv)(A) for automatic actuation of the (1) RPS Reactor Trip with the reactor not critical and (6) Auxiliary Feedwater System, 8 hour notification. The licensee notified the NRC Resident Inspector. |
ENS 51406 | 17 September 2015 21:32:00 | A miniature alcohol bottle, containing trace amounts of liquid, was discovered inside the protected area. Site security took possession of the bottle and removed it from the protected area. The licensee notified the NRC Resident Inspector. |
ENS 51407 | 18 September 2015 11:59:00 | The following information was received from the State of Texas via email: On September 18, 2015, the Agency (Texas Department of State Health Services) was informed by the licensee's radiation safety officer (RSO) that a radiography crew had experienced a source disconnect at a temporary field site (Galveston, Texas). The RSO stated the crew was working inside a vessel using a QSA 880D exposure device containing a 52.9 curie Iridium-192 source. The device fell from a distance of 30 feet and hit the floor of the vessel. The source was in the fully shielded position when the device fell. The radiographers noted the guide tube had a small kink in it and replaced the guide tube. The radiographers tested the source by cranking the source out, but when they attempted to retract the source, the drive cable did not stop at the rear outlet of the camera. The radiographers contacted their supervisor and performed a dose rate survey at their barrier. The dose rate was 1 millirem per hour. An individual qualified in source recovery was able to remove the source from the guide tube and place it in a source changer for storage. The RSO stated their inspection of the source drive cable found the connecter on the drive cable had separated from the drive cable. The RSO stated all equipment involved in the event will be returned to the manufacturer for inspection. No individual received an over exposure as a result of this event. No member of the general public received an exposure due to this event. The licensee is conducting an investigation into the event. Additional information will be provided as it is received in accordance with SA-300. Texas Incident: I-9339 |
ENS 51405 | 17 September 2015 16:24:00 | At 1220 (EDT) on 9/17/2015, both doors of a Secondary Containment airlock were reported to be simultaneously open for approximately five seconds during the normal passage of personnel. The brief time that the doors were simultaneously open constitutes an inoperable condition of Secondary Containment. Secondary Containment differential pressure was maintained throughout the time period that the doors were open. This event is being reported under 10 CFR 50.72(b)(3)(v)(C) and per the guidance of NUREG-1022, Rev. 3, '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 control the release of radioactive material.' The NRC Resident Inspector has been notified. |
ENS 51408 | 18 September 2015 15:34:00 | When requested by the NRC, the licensee was unable to locate a 15 mCi Ni-63 source. The source was used in a Perkin-Elmer gas chromatograph electron capture detector. The licensee noted that the device may have been decommissioned, but a fire at their facility effectively destroyed all records that could be used to located the source. 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 |
ENS 51452 | 5 October 2015 09:58:00 | The following information was received by the State of Ohio via facsimile: A patient had a prostate volume study done on 7/24/15 at Mount Carmel St. Ann's (MCSA). The mass that was observed on the ultrasound is in the location of a typical prostate and takes the shape of a typical prostate. Implant (I-125) procedure was performed on 8/21/15. During the implant, ultrasound guidance was used when placing the seeds and the images that were seen matched what was taken at the time of the volume study on 7/24/15. A CT (Computerized Axial Tomography) scan was taken on 9/23/15 for post implant study to verify the seed placement and target coverage. The post implant study was performed on 10/1/15 and revealed that the prostate that is visible by CT is not adequately covered by the seeds, and that the seeds may be in the rectum. The images were reviewed further and there is a mass located between the rectum and the prostate. This is what is believed to have been visualized on the ultrasound, and what was treated in the OR (Operating Room). The licensee is still investigating to determine whether this tissue is part of the prostate, possibly rectum, or something else altogether. The patient has been scheduled for an MRI (Magnetic Resonance Imaging) to better visualize the tissues in this area. What is known at this time is that a large part of the prostate was not treated, the coverage to the intended target organ is below 80%, and this is being treated as a medical event. The licensee will submit a complete report when the analyses are finalized. ODH (Ohio Department of Health) intends to send an investigator to the licensee for follow-up. This NMED record will be updated as more information becomes available. Ohio Incident: OH150010 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. |
ENS 51451 | 5 October 2015 10:00:00 | This 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to report an invalid actuation of the Unit 2, Train B Containment Ventilation Isolation (CVI) at Sequoyah Nuclear Plant. At 1919 EDT on August 7, 2015, during planned performance of a Unit 2 containment vent, the Train B CVI actuated due to an invalid Hi Rad signal from 2-RM-90-131, Containment Vent Radiation Monitor. In addition to the Train B CVI alarm, unexpected alarms were received for 2-RM-90-106, Lower Containment Radiation Monitor and 2-RM-90-112, Upper Containment Radiation Monitor instrument malfunctions as they isolated for the CVI and 2-RM-90-131 Hi Rad alarm. Prior to the invalid Hi Rad alarm, all radiation monitors were stable at their normal values. All required automatic actuations occurred as designed. Upon investigation, the cause of the invalid Hi Rad alarm was due to an exposed shield wire at the 2-RM-90-131 detector. Preventative maintenance had been performed the week prior to the CVI and it is believed the damage occurred at that time. Control Room Operators performed Annunciator Response actions and verified by diverse indications that the subject condition was an invalid Hi Rad signal. There were no indications of degraded reactor coolant system parameters or fuel failure. Applicable Technical Specification (TS) Limiting Condition for Operations (LCOs) were entered and the radiation monitors declared inoperable. No Emergency Response criteria were applicable with the subject radiation monitors inoperable. Radiological surveys performed in the vicinity of 2-RM-90-131 verified no abnormal radiological conditions. Radiation Monitor 2-RM-90-131 was removed from service, the shield wire was repaired and returned to service with no issues. Radiation Monitors 2-RM-90-106 and 2-RM-90-112 were tested and returned to service. The applicable TS LCOs were exited. At the time of the event, plant conditions for a Hi Rad alarm did not exist; therefore, the CVI was invalid. The NRC Resident Inspector was notified. |
ENS 51236 | 16 July 2015 04:41:00 | This notification is being made due to a loss of emergency assessment capability in accordance with 10CFR 50.72(b)(3)(xiii). At 2332 (CDT), on 07/15/2015, the meteorological tower computer system software failed which resulted in a loss of meteorological data to the plant. Proceduralized compensatory measures for dose assessment include use of National Weather Service followed by historically determined default values. Information Technology personnel reported to the plant and successfully reset the software. Meteorological data to the plant was restored at 0216 (CDT) on 07/16/2015. The NRC Resident Inspector has been notified. |
ENS 51232 | 15 July 2015 01:04:00 | At 0004 (CDT) on Wednesday, July 15, 2015, the Dresden Nuclear Power Station (DNPS) Technical Support Center (TSC) emergency ventilation system will be removed from service for planned maintenance activities. During the maintenance, the TSC Ventilation will be shut down. The TSC air filtration fan and dampers will be non-functional, rendering the TSC HVAC accident mode non-functional. This maintenance is scheduled to minimize out of service time. The planned TSC ventilation outage is scheduled to be completed in approximately 24 hours. Contingency plans are in place so that if an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing Emergency Planning (EP) procedures and checklists. If radiological or environmental conditions require TSC facility evacuation during ventilation system restoration; the Station Emergency Director will relocate the TSC staff in accordance with station procedures. The NRC Resident Inspector has been notified.
At 1347 CDT on July 17, 2015, Dresden TSC Ventilation was restored. The Dresden TSC Ventilation is Functional at this time. The NRC Resident Inspector has been notified. Notified R3DO (Orth). |
ENS 51233 | 15 July 2015 10:50:00 | The following information was received by the State of Ohio via email: On July 14, 2015, the licensee reported that the intended delivery of Y-90 SirSpheres went to the small bowel instead of the right lobe of the liver during a procedure that morning. The intervention physician felt that the dose delivery was not going where it should be going and discontinued the treatment. Scanning the patient identified that the Y-90 microspheres were delivered to the small bowel. The original prescribed dose to the right lobe of the liver was 78 Gy with 20.5 mCi. The delivered dose of 36 Gy with 7.79 mCi went to the small bowel instead of the liver right lobe. The patient was notified at the time of the event. The interventional physician was the referring physician and AU (Authorized User). Ohio Report: OH150007 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. |
ENS 51213 | 8 July 2015 21:53:00 | This telephone notification is provided in accordance with Exelon Reportability manual SAF 1.10, 'Major Loss of Emergency Preparedness Capabilities', and 10CFR50.72(b)(3)(xiii). On July 8th 2015 at 1837 (CDT), it was determined that the onsite Technical Support Center (TSC) Ventilation System Supply Fan belts had failed, resulting in loss of ventilation for the facility. Repairs were not completed within the time required had the TSC needed to be staffed. There is currently no emergency event in progress requiring TSC staffing. If an emergency is declared and the TSC ERO (Emergency Response Organization) activation is required, the TSC will be staffed and activated unless the TSC becomes uninhabitable due to ambient temperatures, radiological, or other conditions. If relocation of the TSC staff becomes necessary, the Station Emergency Director will relocate the staff to an alternate TSC location in accordance with applicable site procedures. The licensee has notified the (NRC) Senior Resident Inspector of the issue.
After repairs were completed, the TSC Ventilation was restarted on 7/9/15 at 0625 EDT for a maintenance run, the TSC Ventilation was restored to operable status at 1500 EDT on 07/11/2015. The licensee has notified the NRC Resident Inspector. Notified R3DO (Stone). |
ENS 51211 | 8 July 2015 15:37:00 | Indian Point Unit 3 was manually tripped at 1427 EDT due to lowering steam generator water levels. At 1425 EDT, #31 condensate pump tripped, causing the lowering water levels. There were no immediate complications on the trip and the unit is stable in Mode 3. Auxiliary feedwater actuated as expected and is in service. All rods inserted and decay heat is being rejected to the condensers. Offsite electrical power is in service. Unit 2 is stable at 100% power. The licensee plans on issuing a press release. The licensee notified the NRC Resident Inspector and New York Public Service Commission. |
ENS 51202 | 7 July 2015 16:07:00 | On 7/7/2015 at approximately 1435 EDT, the Technical Specification for Secondary Containment Pressure Boundary was not met when vacuum could not be maintained greater than or equal to -0.125 inches of water gauge for approximately 41 seconds. As part of post-maintenance testing for the non-safety related Reactor Building HVAC Center Exhaust Fan, the fan was started while the safety-related Standby Gas Treatment system was also in operation. Shortly after the fan was started, operators observed degrading vacuum in secondary containment and subsequently secured the center exhaust and supply fans. Vacuum continued to degrade momentarily after the fans were secured, and then returned to a Technical Specification allowable value. Subsequent inspections discovered that the affected fan was operating in the reverse direction. This is believed to have caused Secondary Containment pressure to increase. Since vacuum could not be maintained with the safety-related Standby Gas Treatment system operating, the plant operated in an unanalyzed condition. The cause of the reverse rotation is under investigation. There were no radiological releases associated with this event. The NRC Senior Resident Inspector has been notified.
Based on plant configuration at the time of the event and further review of the Fermi 2 UFSAR, the plant did not operate in an unanalyzed condition. The Reactor Building HVAC fans would have tripped, as designed, upon receipt of a safety-related Standby Gas Treatment actuation signal during the time of the event. Therefore, the fans' pressurizing effect on secondary containment would have ceased within the time limits assumed in the existing accident analysis. The reporting criteria of 10CFR50.72(b)(3)(v)(C) remains valid. The licensee notified the NRC Resident Inspector. Notified the R3DO (Stone). |
ENS 51203 | 7 July 2015 16:20:00 | The following information was received from the State of Texas by email: On July 7, 2015, the Agency (Texas Department of State Health Services-Radiation Branch) was notified by the licensee's radiation safety officer (RSO) that the shutter on a Ronan GS-400 level gauge containing a 50 millicurie cesium - 137 source was stuck in the open position. The stuck shutter was discovered during the start up of a system component. Open is the normal position for the shutter. The gauge does not possess an exposure risk to any individuals. The manufacturer has been contacted and will replace the gauge. Additional information on this event will be provided as it is received in accordance with SA-300. Texas Incident: I-9324 |
ENS 51204 | 7 July 2015 17:01:00 | The following information was received from the State of Mississippi by email: Location of Incident: Brandon, Mississippi, I-20 exit ramp 56 construction. Description of Incident: July 3, 2015 at approximately (1630 CDT), a Humboldt Model 5001 portable nuclear moisture density gauge, Serial Number 3339, was run over by a heavy equipment truck while on a jobsite. The gauge was severely damaged, but the sources remained intact and in shielded position. DRH (Mississippi Division of Radiological Health) was notified and provided with a survey measurement of 8mR/hr at the surface of the damaged gauge. The RSO was able to safely load the fragmented gauge back into its approved transport case and return it to the MDOT (Mississippi Department of Transportation) storage facility. DRH personnel visited the MDOT storage facility and observed an 8mR/hr surface survey reading with an NDS ND-2000 survey meter, serial number 24562, calibration date 7-1-2015. Leak test wipes were taken from both sources and sent for analysis by the licensee. The measured activity was well below the regulatory upper limit of .005 microcuries. Wipe analysis was received 7-7-2015. Gauge is being returned to the manufacturer for disposal. Radioisotopes: Cs-137(11 mCi); Am-241/Be(44 mCi). Radiation measurements taken by the Mississippi Division of Radiological Health: 8 mR/hr at surface; Less than 1 mR/hr at 1 meter. Mississippi Event: MS-15001 |
ENS 51180 | 25 June 2015 14:18:00 | The following information was received from the State of Louisiana via email: Event date and time: On 06/25/2015 (at 0930 CDT, the licensee Radiation Safety Officer) called the Radiation Section of LDEQ (Louisiana Department of Environmental Quality) to report a lost/missing radiography camera. The camera was to be loaded on a rig truck and (to) be transported to a temporary site. The crew and site RSO had been looking for the camera since it was discovered missing at 0830 (CDT). A radiography exposure device was left on the bumper of a rig truck and not secured in the vault/overpack on the truck. The crew left the yard on Highway LA #30 and headed to I-10 and then East on I-10. About 5 miles down I-10, the crew remembered that they had not secured the camera in the rig truck. They stopped and found the camera missing. They backed tracked I-10 to LA # 30 and back to the office. They did not locate the missing camera. The LA State Police was notified in addition to LDEQ. The Radiation Section (of LDEQ) was notifying the staff and dispatching a Radiation (Environmental Scientist) individual to respond. The media put out an alert of the missing camera. Homeland Security, QSA Global, and the Ascension Parish Sheriff responded and were aiding in locating the lost camera. They were combing the area on ATVs and utilizing sensitive radiation detection instruments. At about 1130 (CDT) the LDEQ was given an update by (the licensee Radiation Safety Officer) which only includes responding agencies and the Site RSO for contact. About 1150 (CDT) the NRC Region IV was notified of the incident (Latisha Hanson) and given a preliminary notification and told the (NRC Operations Center) was being notified. Event Location: A rig truck was dispatched from TIS at 37568 Hwy # 30, Gonzales, LA down LA #30 to Interstate-10. The rig went east on I-10 for about 5 miles when the crew remembered that the camera had not been secured in the rig. The crew check the back of the rig and backed tracked the I-10 to LA #30 route. The camera was not located. Notifications were made. Event type: Loss of control over an exposure device. A QSA 880 Delta exposure device S/N 4586. The exposure device was loaded with about 30 Ci of Ir-192, QSA source Model # 84-9. The Category II, Quantity of Concern was released or lost into the general public. The radioactive exposure device was released into the general public by the two individuals not following the TIS's (Team Industrial Services) Radiation Safety Procedures or the IC (Increased Controls) Security Procedures. The exposure device was released to the general public unsupervised and not in direct control of an authorized company representative. Event description: The equipment was a QSA 880 Delta exposure device S/N D4586. The source was about 30 Ci of Ir-192, model # 84-9. At this time LDEQ consider this incident still open and updates will be given when available. Transport Vehicle: This was a TIS company crew truck being dispatched to a temporary jobsite. Media attention: News Media was alerted and reporting agencies were notified. Louisiana Event: LA150010 Notified DHS SWO, FEMA, USDA, HHS, DOE, DHS NICC, EPA, FDA(email), Nuclear SSA (email) FEMA National Watch Center (email), DNDO-JAC (email). THIS MATERIAL EVENT CONTAINS A "CATEGORY 2" LEVEL OF RADIOACTIVE MATERIAL Category 2 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 a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioactive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachytherapy. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf
The following information was received from the State of Louisiana via email: The exposure device was recovered at about 7:00pm (CDT) on 06/25/2015. Device QSA/AEA Technologies, Model Delta 880, S/N 4586 Source QSA/AEA Technologies, Model A424-9, S/N 166306; 48.2 Ci Ir-192 Recovered (approximately) 1.5 (Miles) on LA 61 east of US I-10 east. It was on the side of the road in a wet-muddy ditch area. (The device was found) by backtracking the trucks GPS device. A health and safety survey was conducted and the shielding appeared to be intact. The exposure device was loaded on to another Team Industrial vehicle, blocked and braced, and returned to the vault at the highway LA-30 address. The device and source were leak tested and analysis was performed on the test. A QSA Global representative stated it appears the device (DU) and the Ir-192 source were not leaking or compromised. With the exception of corrective actions and enforcement issues the department, LDEQ (Louisiana Department of Environmental Quality), considers this incident closed. Notified R4DO(O'Keefe), IRD MOC(Grant), ILTAB (Wray), and NMSS Events Resource (via email). Notified DHS SWO, FEMA, USDA, HHS, DOE, DHS NICC, EPA, FDA(via email), Nuclear SSA (via email) FEMA National Watch Center (via email), DNDO-JAC (via email). |
ENS 51127 | 5 June 2015 09:27:00 | The licensee nuclear medicine technologist ordered a package containing radioisotopesTc-99m and Xenon for a patient. The package was delivered by an offsite nuclear pharmacy to the Indiana University Health- Ball Memorial Hospital. When the technician surveyed the package, the surface contamination exceeded specified limits and nominally measured about 14,000 dpm. The package was quarantined and the vendor/shipper was notified concerning the surface contamination. The package was placed in a gamma camera and the indicated camera spectrum indicates Xenon package contamination. The Radiation Safety Officer has ordered the package to remain quarantined. No personnel contamination resulted from this event. |
ENS 51073 | 15 May 2015 21:03:00 | A South Texas Project Offsite Emergency Notification siren was (inadvertently) going off. The Matagorda County Sheriff's office notified the Emergency Response organization at the station that a siren had actuated during a severe thunderstorm moving through the area. Station personnel are addressing the issue with the siren. The Matagorda County Sheriffs office was the only offsite agency that was contacted during this event. The licensee notified the NRC Resident Inspector. |
ENS 51041 | 5 May 2015 00:48:00 | 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. This event is being reported pursuant to 10 CFR 50.72(b)(3)(xiii) for a loss of emergency assessment capability at the Palo Verde Nuclear Generating Station (PVNGS). On May 4, 2015 at 0320, seismic monitoring (SM) system force balance accelerometer R0006 was determined to be non-functional due to an emergent equipment failure. On May 4,2015, at approximately 1600, further review of this equipment failure and the related impact to the capability of the SM system determined that this was a reportable loss of emergency assessment capability. This specific accelerometer functions to provide indication that the Operational Basis Earthquake threshold has been exceeded following a seismic event and is used in the PVNGS Emergency Plan to perform classification for emergency action level HA1.1, Natural or Destructive Phenomena affecting Vital Areas. As a compensatory measure, PVNGS procedures for seismic event evaluation provide alternative methods for HA1.1 event classification with accelerometer R0006 out of service. Maintenance to correct the condition is in-progress. The NRC Resident Inspector has been informed of this condition. |
ENS 51035 | 3 May 2015 00:05:00 | Loss of assessment capability due to unplanned removal from service of a radiation monitor due to process flow monitor indication failing hi. The normal and hi range ventilation vent process radiation monitors (3HVR*RE10A/B are out of service. This condition was discovered during control room rounds. The condition is reportable per 10CFR50.72(b)(3)(xiii). Compensatory measures are in place. The licensee notified the NRC Resident Inspector and applicable State and Local authorities. |
ENS 51008 | 24 April 2015 00:15:00 | This notification is being reported to NRC in accordance with 10 CFR 50.72(b)(2)(xi) for notification of an on-site fatality of a contract employee. In addition, the contracting company plans to notify the Occupational Safety and Health Administration (OSHA) of a fatality per 29 CFR 1904.39. At approximately 1717 CDT on 4/23/15, a 911 call was received in the Control Room regarding a contract employee who was found unresponsive and unattended in a temporary break room set up on the Turbine Deck during the Unit 1 refueling outage. Resuscitation by first responders and paramedics from a nearby town was unsuccessful. Resuscitation efforts were suspended at 1750. The Houston County Sheriff's Office was notified at approximately 1800 and they responded to the site at 1822. The county coroner was notified and arrived on site at 1850. (Farley Nuclear Plant) received notification at approximately 2035 that the contractor company intended to notify OSHA. A press release is not planned at this time. The NRC Resident Inspector has been notified. Unit 1 remains in Mode 6 and Unit 2 remains in Mode 1 at 100% power. |
ENS 51004 | 23 April 2015 05:25:00 | On April 23, 2015 DC Cook Unit 2 Reactor was manually tripped due to an uncontrolled cooldown due to two (2) failed open steam dump valves. The cause of the failure is still under investigation. This event is reportable under 10 CFR 50.72(b)(2)(i) Tech Spec Required Shutdown, as a four (4) hour report; 10 CFR 50.72(b)(2)(iv)(B), Reactor Protection System (RPS) actuation, as a four (4) hour report; and under 10 CFR 50.72(b)(3)(iv)(A), specified system actuation of the Reactor Protection System (RPS), as an eight (8) hour report. The electrical grid is stable and Unit 2 continues to be supplied by offsite power. All control rods fully inserted. Decay heat is being removed via steam generator Power Operated Relief Valves due to steam dump valves being manually isolated. Preliminary evaluation indicates all plant systems functioned normally following the Reactor Trip. DC Cook Unit 2 remains stable in Mode 3 while conducting the post Trip Review. No radioactive release is in progress as a result of this event. The DC Cook Resident NRC Inspector has been notified. There is no indication of primary to secondary leakage and there is no impact on Unit 1. |
ENS 51009 | 24 April 2015 09:22:00 | The following information was received from the Commonwealth of Massachusetts via email: The licensee's Radiation Safety Officer (RSO) reported on April 23, 2015 that, on the morning of April 22, 2015, the licensee mistakenly administered to a patient the wrong radioactive drug, a 118 mCi Tc-99m bulk dose instead of the prescribed 12.9 mCi Tc-99m Sestamibi dose, at the licensee's Baystate Franklin Medical Center facility. The wrong radioactive drug administered was reported by the licensee's RSO to have resulted in 5.6 rem effective dose equivalent to the patient, a reportable medical event in accordance with 105 CMR 120.594(A)(1)(b)1. The licensee's RSO reported that the patient and the referring physician have been notified and that the RSO did not expect any harm to the patient. The RSO reported the cause included that proper procedures were not followed. The Agency (Massachusetts Radiation Control Program) plans to perform a special inspection and considers this event to be open. 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. |
ENS 51001 | 22 April 2015 01:58:00 | On April 21, 2015 at 2258 (EDT), Secondary Containment became inoperable requiring a Technical Specification 3.6.4.1 entry for failure to meet SR 3.6.4.1.1 on Unit 1 and Unit 2. The inoperability was caused by Zone 3 differential pressure lowering to less than 0.25 (inches Water Column) when Zone III fans tripped during 30mph wind gusts. Fans were restarted and differential pressure restored to greater than 0.25 (inches Water Column) at 2314 hrs. April 21, 2015. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022 Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment system. The licensee notified the NRC Resident Inspector. |
ENS 51000 | 22 April 2015 00:18:00 | The following information was received by the State of Texas via email: On April 21, 2015 at 2209 (CDT) hours, the Agency (Texas Department of State Health Services) was contacted by Andrews County Emergency Management (ACEM). They informed the Agency that an accident had occurred 12 miles south of Andrews, Texas, on highway 385, which involved a radiography truck. The Agency contacted ACEM chief who stated he was at the scene of a three vehicle accident which included a truck from Desert NDT. The driver was killed in the accident and the truck cab had separated from the frame. He stated the dark room had separated from the truck bed. He stated a person from the Andrews County WCS (Waste Control Specialist) was there and had performed a radiation survey and measured a dose rate of 15 millirem 10 feet from the truck. He stated the licensee had been contacted. He stated they had taken care of the survivors and had backed out of the area until the licensee's radiation safety officer arrived on the scene. He stated they had not seen the shipping papers, only the radiation symbol on the truck. I asked him to have the licensee contact the Agency as soon as they arrived on site. The licensee's (Desert NDT) RSO arrived at the scene at 2223 hours and contacted the Agency. He stated his priority was to locate the source. He agreed to call the Agency as soon as he had control of the source. At 2240, the RSO contacted the Agency and stated he had control of the source. The iridium source was inside a INC 100 radiography camera and the RSO believed the activity was between 20 and 26 curies. He stated the camera did not appear to be damaged. He stated the dark room had separated from the truck and split into two pieces. The camera was located still in its transport box in a section of the darkroom. He stated the dose rate on contact with the camera was 16 millirem an hour and 0.4 millirem at 1 foot. The dose rate at 1 meter was not distinguishable from background. He stated no individual at the scene would have received an exposure to radiation that would have exceeded any limits. The RSO stated he was taking the source back to the licensee's office for storage. The RSO stated they would send the exposure device to the manufacturer for inspection. Additional information will be provided as it is received in accordance with SA-300. Texas Incident: I-9305
The following information was received by the State of Texas via email: The licensee's corporate radiation safety officer contacted the Agency (Texas Department of State Health Services) and informed them that two radiographers were killed in this event. He stated the source activity was only 13 curies. He stated that local law enforcement in Andrews, Texas will not release any details of the accident until their investigation is completed. Additional information will be provided as it is received in accordance with SA-300. Notified the R4DO (Drake), IRD MOC (Gott) and the NMSS Events Notification via email. |
ENS 50986 | 15 April 2015 16:02:00 | The following information was excerpted from a Commonwealth of Massachusetts facsimile: A radiation source was detected in a trash load at the Roxbury Transfer Station (RTS) by radiation detectors at the transfer station entrance. The RTS consultant performed a survey to separate the radiation source from the remainder of the trash. The consultant transported radioactive trash to Atlantic Nuclear (MA license #56-0477) to perform isotope identification. Atlantic Nuclear' s analysis indicates the radiation source contains about 90 microCuries of Ra-226. A dose rate of 15 millirem/hour was measured at about 1 inch from the object. The consultant separated the single object from the trash bag. The source is stored at Atlantic Nuclear and is waiting for proper disposal. The Agency (Massachusetts Radiation Control Program) continues to investigate and considers this event to be open. 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 |
ENS 50983 | 14 April 2015 11:44:00 | A medical event involving Y-90 microspheres (TheraSpheres) occurred at approximately 1000 EDT on 4/14/15. The prescribed dosage was 34.6 mCi and the delivered dosage was 25.5 mCi. This equates to a 26.3 percent underdose. The patient was notified by the authorized user following treatment and before discharge on 4/13/15. The referring physician was notified by the authorized user via electronic mail at 1149 EDT on 4/13/15. The initial hypothesis on cause may have been related to difficult access to an anatomical region in the liver, resulting in the need to use lower than normal pressure on the syringe used for microsphere delivery. All established administration procedures were followed. A written report to the appropriate NRC offices will follow within 15 days.
The following information was excerpted from the licensee email: The reason for this retraction is based upon discussions with the Authorized User (AU) who performed the Y-90 treatment and additional questions raised and clarifications made by the NRC Region III Office. During that discussion, the AU indicated that he utilized a lower syringe pressure than normal to prevent reflux of the Y-90 microspheres which would have resulted in a less than optimal treatment. The AU acknowledged that the amount administered was acceptable, given the need to use the lower syringe pressure and that he will modify the written directive to appropriately reflect a change in the written directive based upon those circumstances. Notified the R3DO (McCraw) and NMSS Events Notification via email. 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. |
ENS 51117 | 4 June 2015 11:18:00 | This 60-day telephone notification is being submitted in accordance with 10 CFR 50.73(a)(1) and 10 CFR 50.73(a)(2)(iv)(A) to report an invalid actuation of the Train B Phase A Containment Isolation at Sequoyah Nuclear Plant. At 1320 EDT on April 12, 2015, during planned performance of the Containment Isolation Train-A, Phase A Isolation Testing and Emergency Gas Treatment System (EGTS) Cleanup System Test, the main control room received several Train-B annunciators. Upon investigation, it was determined that an invalid signal to the Train-B Solid State Protective System (SSPS) actuated the Train B, Phase A Containment Isolation. The invalid isolation signal was the result of a human performance error during the performance of the Phase A Isolation Test surveillance procedure. Operations personnel responded to the SSPS initiation, testing was aborted, ensured that all equipment operated as designed and restored affected systems in accordance with plant procedures. Approval to restart testing was obtained. All prerequisites were met and testing of the SSPS Train-A, Phase A Isolation was completed satisfactorily. As part of the prerequisite test alignment of the Train-A, Phase A, Unit 2 had entered a planned 7 day action for EGTS being inoperable. During the test when the Train-B of Phase A actuated, the suction dampers for Unit 1 supply to EGTS were closed per plant procedures. This prevented Train-B EGTS from aligning to Unit 1 and allowed Train-B of EGTS to remain operable for Unit 2. An SSPS Phase A signal can be generated automatically by a Safety Injection Signal (SIS) or manually. At the time of the event, plant conditions for an SIS did not exist; therefore, the Phase A actuation was invalid. The licensee notified the NRC Resident Inspector. |
ENS 50961 | 7 April 2015 15:45:00 | A loss of Main Generator Load which caused a Reactor Trip on Units 1 & 2. A switchyard voltage transient from a highline occurred, which caused an undervoltage condition on both units' safety related 4KV buses. Unit 1 is on normal heat removal to the condenser. Unit 2 is on auxiliary feedwater and normal condenser bypass valves for temperature control. An Auxiliary Feedwater Actuation System (AFAS) actuation occurred on Unit 2. The (Unit 2) 2B emergency diesel generator did not start and load on its respective 24-4 KV bus. The 24-4KV Bus was repowered from the alternate feeder breaker. Cause of the emergency diesel failure to start is under investigation. All safety functions are met for both units. All control rods fully inserted. The site is in a normal shutdown electrical configuration powered from offsite. The site plans to stay in Mode 3 pending restart. The licensee notified the NRC Resident Inspector, State and local authorities. A press release is planned.
During post trip review, it was determined that the 21 saltwater pump had to be manually started. With the failure of 2B emergency diesel generator, there were no saltwater pumps running for approximately 12 minutes. Additional troubleshooting determined the 2A emergency diesel generator sequencer did not automatically start 21 saltwater pump. The 2B emergency diesel generator was returned to service on 4/8/2015 at 1730 (EDT). The loss of saltwater (pump) and emergency diesel generator is reportable as an event that could have prevented fulfillment of a safety function and is also an unanalyzed condition. The licensee has notified the NRC Resident Inspector. Notified R1DO (Ferdas), IRD MOC (Grant), NRR EO (Morris). |
ENS 50948 | 2 April 2015 17:29:00 | Northern States Power Company - Minnesota (NSPM) has completed a review of seismic monitor performance at the Prairie Island Nuclear Generating Plant (PINGP) over the past 3 years. The emergency preparedness plan requires seismic monitoring instruments to diagnose an earthquake for emergency action levels (EAL) HA1.1 (Seismic Event Greater Than Operating Basis Earthquake (OBE) as indicated by 'OBE Exceedance' alarm on Seismic Monitoring Panel) or HU1.1 (Earthquake felt in plant as indicated by Valid 'Event' alarm on Seismic Monitoring Panel). Contrary to that requirement, this review identified 6 unplanned instances where the seismic monitor was non-functional that were not previously reported, and 3 planned instances where the seismic monitor was non-functional for greater than 24 hours that were not previously reported. Since there was no compensatory measure that could be credited when the seismic monitor was non-functional, an emergency classification at the ALERT or UNUSUAL EVENT level could not be obtained with site instrumentation for a seismic event. The seismic monitor is currently functional, however it was determined to be non-functional on the following dates: Unplanned out of service: 1. August 14, 2012 2. November 16, 2012 3. November 18 2012 4. November 21, 2012 5. December 5, 2012 6. January 16, 2013 Planned greater than 24 hour out of service: 1. December 14, 2012 2. September 3, 2014 3. September 30, 2014 The unplanned non-functional conditions of the seismic monitor have been corrected and were entered into the NSPM Corrective Action Program. The loss of assessment capability is reportable to the NRC within 8 hours of discovery in accordance with 10 CFR 50.72(b)(3)(xiii). This report is required per 10 CFR 50.72(a)(1)(ii) as an event that occurred within 3 years of the date of discovery. Corrective actions are in progress to address the missed reporting of seismic monitor unavailability. The licensee notified the NRC Resident Inspector. |
ENS 50945 | 1 April 2015 18:53:00 | At 1210 CDT the transformer supplying power to the Emergency Operating Facility (EOF) stopped working due to the failure of a capacitor bank. The EOF is located adjacent to OPPD's (Omaha Public Power District) North Omaha facility, approximately 17 miles south of Fort Calhoun Station. The event caused a small grass fire which was quickly extinguished. The local fire department was called. The backup emergency diesel generator for the EOF started and supplied power to the facility, as designed. With the EOF diesel operating, the facility is able to function as required during emergency conditions. At 1440 CDT the EOF emergency diesel generator stopped running. At 1545 CDT the Conference Operations (COP) network phone system failed. The COP network is the primary emergency notification system between OPPD, state and county agencies. It is used to provide initial and updated notifications and for general information flow between these agencies. Alternate means of communication have been established (commercial lines) and a dedicated communicator is stationed in the control room to ensure that we can facilitate communication should the need arise. Power to the EOF was restored at 1713 CDT. At time 1720 CDT the COP tested as normal. The licensee notified the NRC Resident Inspector. |
ENS 50917 | 23 March 2015 11:16:00 | At approximately 0605 EDT on March 23, 2015, the Oconee Nuclear Station main control room and Security received an emergency call for an employee experiencing a non-work related medical issue. Site first responders were dispatched in conjunction with a request for off-site medical assistance. The individual was transported by ambulance to the Oconee Medical Center and was pronounced dead at 0717 EDT. The individual was outside of the protected area (within the owner controlled area) and no radioactive material or contamination was involved. The cause of death has not been determined. This notification is being made in accordance with 10 CFR 50.72(b )(2)(xi) for situations related to the health of on-site personnel for which a notification to other government agencies has been made. The South Carolina Occupational Safety and Health Administration (SCOSHA) was notified at 0920 EDT. The NRC Resident Inspector has been notified. The licensee notified Pickens County, South Carolina and Oconee County Emergency Managements. |
ENS 50914 | 21 March 2015 04:15:00 | At 0030 CDT, on March 21, 2015, both doors of a Secondary Containment Airlock were opened concurrently by two separate individuals. The doors being open at the same time caused a failure to meet SR (Surveillance Requirement) 3.6.4.1.2 to verify that either the outer door(s) or the inner door(s) in each Secondary Containment access opening are closed. The identified condition caused Secondary Containment to be considered inoperable per TS LCO (Technical Specification Limiting Condition for Operation) 3.6.4.1. Upon discovery, immediate action was taken to close the doors. The doors were open concurrently for a momentary amount of time. The action to close the door allowed SR 3.6.4.1.2 to be met, and restored Secondary Containment to an operable status. This notification is being made pursuant to 10 CFR 50.72(b)(3)(v)(C). The NRC Resident Inspector has been notified. |
ENS 50903 | 19 March 2015 10:51:00 | At 0702 EDT on March 19, 2015, Fermi 2 received an automatic scram due to actuation of the Reactor Protection System (RPS) function of Oscillation Power Range Monitor (OPRM) Upscale. The plant had recently transitioned to Single Loop Operation after securing the 'A' Reactor Recirculation Pump due to loss of normal and emergency cooling water supply. The lowest reactor water level was 134 inches above top of active fuel. Reactor water level is being maintained in the normal band by the Feedwater and Control Rod Drive Systems. No Safety Relief Valves (SRV) actuated. Reactor pressure is being maintained via the Main Turbine Bypass Valves and Main Condenser. Reactor Pressure Vessel Level 3 isolation occurred. No additional safety system actuations occurred. All off-site power sources were available throughout the event. The plant is currently in Mode 3 and in a stable condition. Investigation into the cause of the event is ongoing. This event is being reported under the four hour Non-Emergency reporting criteria of 10CFR50.72(b)(2)(iv)(B). The NRC Resident Inspector has been notified. |
ENS 51075 | 17 May 2015 11:28:00 | MEASURING AND TEST EQUIPMENT DEFICIENCIES The following information was received by facsimile: This report is being provided as an interim report in accordance with 10CFR 21.21. (i) Name and address of the individual or individuals informing the Commission. Adam Mohr; President, Fabrication and Manufacturing; CB&I; One CB&I Plaza; 2103 Research Forest Drive; The Woodlands, TX; 77380. (ii) Identification of the facility, the activity, or the basic component supplied for such facility or such activity within the United States which fails to comply or contains a defect. This is an interim report. The deviations being evaluated pertain to deficiencies identified within the Measuring and Test Equipment program at Chicago Bridge and Iron (CB&I) Laurens, 366 Old Airport Road, Laurens, SC. (iii) Identification of the firm constructing the facility or supplying the basic component which fails to comply or contains a defect. This is an interim report. The construction activities for the V.C. Summer (Units 2 and 3) and Vogtle AP1000r (Units 3 and 4) nuclear projects, which include procurement of the piping assemblies, are being performed by CB&I Power, 128 S. Tryon St., Charlotte, NC 28202. (iv) Nature of the defect or failure to comply and the safety hazard which is created or could be created by such defect or failure to comply. This is an interim report. The evaluations of vendors and previously performed calibrations is under evaluation and equipment in question is being re-calibrated by an approved lab and evaluated for extent of condition. Additionally, causal analysis is being performed that is expected to provide relevant information pertaining to the cause of the deviations and if any quality assurance breakdowns may have occurred that could have produced a defect. Evaluation of the condition is expected to be completed by June 26, 2015. (v) The date on which the information of such defect or failure to comply was obtained. The discovery date of the deviations that require evaluation is March 18, 2015, based on the nonconformance reports and C/PAR that identify the deviations. Evaluation of reportability in accordance with 10 CFR Part 21 was not able to be completed within the 60 day evaluation period. Additional time is needed to collect additional data pertaining to the identified nonconformances, perform causal analysis, and complete the evaluation. (vi) In the case of a basic component which contains a defect or fails to comply, the number and location of these components in use at, supplied for, being supplied for, or may be supplied for, manufactured. or being manufactured for one or more facilities or activities subject to the regulations in this part. No basic components have been determined to fail to comply or contain a defect. This is an interim report. (vii) The corrective action which has been, is being, or will be taken; the name of the individual or organization responsible for the action; and the length of time that has been or will be taken to complete the action. No basic components have been determined to fail to comply or contain a defect. This is an interim report. (viii) Any advice related to the defect or failure to comply about the facility, activity. Or basic component that has been, is being, or will be given to purchasers or licensees. None at this time. (ix) In the case of an early site permit, the entities to whom an early site permit was transferred. Not applicable." |
ENS 50882 | 13 March 2015 08:43:00 | On March 13, 2015 at 0100 (CDT), it was identified that fuel assembly QAD224 was mis-oriented 180 degrees at core location 51-40. The intended orientation was (southeast). However the assembly was identified as being (northwest). This issue was identified during the core verification process. In the current core configuration, there is not a bounding analysis that assures adequate Shutdown Margin. This event is reportable under 50.72(b)(3)(ii) as an unanalyzed condition that significantly degraded plant safety. The NRC Resident Inspector has been notified.
The purpose of this notification is to retract the ENS notification made on March 13, 2015 (ENS 50882). An evaluation has determined that Shutdown Margin was met with the mis-oriented fuel bundle. Therefore, the threshold for reporting the issue as a degraded or unanalyzed condition was not met (NUREG 1022 Revision 3 - Event Report Guidelines Section 3.2.7). The NRC Resident Inspector has been notified. Notified the R3DO (Skowkowski). |
ENS 50883 | 13 March 2015 11:17:00 | A licensed employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. |
ENS 50878 | 11 March 2015 09:30:00 | At (0621 EDT) on 3/11/2015, Sequoyah Unit 1 reactor/turbine automatically tripped. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically initiated as expected from the Feedwater Isolation Signal. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 547 F and 2235 psig, with auxiliary feedwater supplying decay heat removal via the steam generators and condenser steam dumps. The immediate cause of the trip was a Power Range Nuclear Instrumentation negative rate signal, caused by a malfunction in the rod control system. There was no associated work in progress related to this and all systems were normally aligned. Current Temperature and Pressure - temperature is 547 degrees F and stable, pressure is 2235 psig and stable. There is no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal, supplied from Off-Site power. There is no operational impact to Unit 2. Unit 2 continues to operate in Mode 1 at 100%. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart timeline has not yet been determined. The licensee notified the NRC Resident Inspector. |
ENS 50860 | 3 March 2015 15:18:00 | Nine Mile Point Unit 1 (NMP1) had a momentary loss of Secondary Containment due to both Reactor Building Airlock doors being opened at the same time. At 0837 (EST) on 03/03/2015, both Reactor Building Airlock doors at NMP1 were opened simultaneously for approximately 2 seconds. This results in a momentary loss of Secondary Containment operability (TS 3.4.3). The doors were closed and operability was restored. Secondary Containment being inoperable is an 8 hour notification per 10CFR50.72(b)(3)(v)(C), '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 control the release of radioactive material.' The condition has been entered into the station's corrective action program and the Senior Resident NRC Inspector was notified. The licensee notified the State of New York. |
ENS 50841 | 21 February 2015 21:35:00 | The Millstone site stack radiation monitor, RM-8169, failed and was declared inoperable at 1950 EST on February 21, 2015. Repairs are in progress. This event is reportable pursuant to 10 CFR 50.72(b)(3)(xiii) as any event that results in a major loss of emergency assessment capability, off-site response capability, or off-site communications capability. The Instrument and Controls Department is conducting troubleshooting and repair. The cause of the radiation monitor failure was sample pump failure. The licensee has notified the NRC Resident Inspector and applicable State and Local authorities. |
ENS 50846 | 23 February 2015 17:05:00 | The following information was received by facsimile: (The Curtiss-Wright) letter is issued to provide notification of a potential defect in Socket (PIN: GB-1 A-5) sold separately or supplied as part of a GRAYBOOT 'A (GB-1A) Connector Kit. There is one affected lot of Sockets (Lot#: 092413). The potential defect is that the socket was not zone annealed which will possibly affect the ability to crimp the socket onto the wire conductor. Zone annealing is specified for the crimp barrel to return the crimp barrel to a soft condition to counteract the effects of previous heat treating required to develop spring tension in the socket tines that mate with the pin contact. The nine affected Customers and their associated Purchase Orders are listed below. All Customers will be notified by February 24, 2015 and supplied with the information we have at this time. 1. Vattenfall- Forsmark Kraftgrupp AB, P.O. 4500301916 2. Vattenfall- Forsmark Kraftgrupp AB, P.O. 4500301916 3. Decon International (HK) Ltd., P.O. DIL00590 4. Ringhals AB, P.O. 641511-090 5. Dremel Inc., P.O. 13-040 6. Bruce Power L.P., P.O. 00185336 7. Kanata Electronic Services Limited, P.O. 34513 8. Meggitt Safety Systems Inc., P.O. 104057 9. Meggitt Safety Systems Inc., P.O. 104345 Based on our recent in-house testing of socket Lot #: 092413, the defective sockets appear to be more difficult to crimp than sockets that have been zone annealed. Some of the Lot #: 092413 sockets showed cracking or bending under magnification. We did not have any sockets actually break apart during the crimping process. All the test specimens crimp barrels remained intact. If a severe fracture had occurred at the time of crimping we believe it would have been very noticeable to the technician installing the contact. Once crimped, the socket appears to function well by conducting electricity and exhibiting acceptable electrical resistance. Many of these sockets from Lot #: 092413 may already be installed. Customers will need to determine if replacing the sockets is required. All customers listed may choose to replace already installed sockets. It is requested that any uninstalled sockets be returned for replacement. All customers requesting replacement sockets should contact Bambi Rhoades at brhoades@curtisswright.com or by phone at 256-924-7424 for arrangements or Jim Tumlinson, contact information listed below. Send all defective sockets to QualTech NP; 120 West Park Loop; Huntsville, AL 35806. Customer can contact Cindy Tidwell at ctidwell@curtisswright.com or by phone at (256) 924-7436 for help with international or freight collect shipping instructions. Ship all defective sockets collect to UPS Account No. 35254E. Additional details, corrective actions and root causes will be provided once our report of the recent testing complete. If you require additional information or would like to discuss this further please do not hesitate in contacting (Jim Tomlinson; Office Phone: (256) 924-7429; email:jtomlinson@curtisswright.com).
The following report was received via fax: This letter provides for the formal close out of notification QTHuntsville 10CFR21-2015-01. The initial notification was made on February 20, 2015 (NRC Event# 50846). All customers were notified and received the test data. All customers were given the opportunity to receive replacement sockets, as they deemed necessary. These replacement parts have been delivered. All actions have been completed. To prevent reoccurrence in the future, the following internal documents have been revised. The procurement and dedication guidelines for both the original GRAYBOOT and GRAYBOOT 'A' connectors were revised. GRAYBOOT 'A' Report No.: SAIC-TR-l038.2-03 currently at Revision P, dated July 16, 2015. GRAYBOOT Report No.: EGS-TR-880708-01 currently at Revision Y, dated July 16, 2015. Both the reports above now contain additional destructive testing as well as visual inspection under magnification to provide more reasonable assurance that the annealing is acceptable. Also, additional training was conducted at our sub-supplier and additional requirements have been added to our purchase orders to prevent any reoccurrence . Based on the above information and corrective actions, this part 21 file is considered closed. If you would like to discuss this further, please contact Tony Gill (QualTech NP Quality Assurance Manager) at 256-924-7438 (office). 256-426-4558 (mobile) or tgill@curtisswright.com. Notified the R2DO (Ernstes) and Part-21 Group (via e-mail). |
ENS 50838 | 20 February 2015 13:14:00 | The following information was received from the State of Texas via email: On February 20, 2015, the Agency (Texas Department of State Health Services) was notified by the licensee that on February 19, 2015, one of its radiography crews working at a remote field site (near Kennedy, Texas) was unable to retract a 31.9 Curie Iridium 192 source into a QSA 880D exposure device. The radiographers were examining a pipe on a pipe pad with the collimator being held in place with a magnetic stand. As the radiographer began to retract the source after a shot, the stand fell and struck the source guide tube crimping the tube to a point where the source could not be moved. The radiographers stopped work in the area and moved their boundaries to prevent exposures to members of the general public. The radiographers contacted their radiation safety officer (RSO), but he was located 8 hours from the work site. The RSO contacted the licensee's office in Corpus Christi, Texas and the RSO from that location responded to the event. The Corpus Christi RSO is (at the location) to perform source retrieval. The event occurred at 1630 (CST) and the source was retracted at 2400 (CST). No over exposures occurred and no member of the general public received any additional exposure from this event. The guide tube has been removed from service for inspection. The dosimetry badges for the individuals involved in the event have been sent to the licensee's processor for reading. The licensee is investigating the event. Additional information will be provided as it is received in accordance with SA-300. On February 20, 2015, the licensee agreed to send the source involved in this event to the manufacturer for inspection. Texas Incident: I-9281 |
ENS 50802 | 10 February 2015 11:35:00 | The following was received from the Ohio Bureau of Radiation Protection via email: A crew working near Cambridge, Ohio this morning experienced a source disconnect on a QSA Model 880D camera containing 60.5 Curies of Iridium-192, which occurred at 9:26 AM EST. The disconnect was discovered after a shot, when the crew's survey instrument indicated that the source was still exposed after the guide cable had been fully retracted. The cause for the source disconnect has not yet been determined. The area has been secured, roped off, and is under constant surveillance by the radiography crew. Two Acuren supervisors trained in source recovery are enroute from their Akron office. The customer has been advised and is cooperating in keeping all personnel away from the area. There has been no exposure to workers or members of the public from the disconnect. An ODH (Ohio Department of Health) Investigator is enroute to the site to observe recovery options. The QSA Global Camera (Model: 880D; Serial number: 4192) contained an Ir-192 source of 60.5 Ci (Serial number:13665G) |
ENS 50800 | 10 February 2015 01:25:00 | On October 14, 2013 a calculation for the containment internal structural analysis was revised and accepted by the station. This calculation limited the Safety injection tank level to 74%. On October 16, 2013 Safety injection tank level was raised to 100% for approximately 13 hours in preparations for plant start-up. While the plant was safely in a cold shutdown condition, this represents a reportable unanalyzed condition. This issue is of a historical nature and does not question the current operability of any plant systems or structures. This was self identified during a Fort Calhoun calculation review. The licensee notified the NRC Resident Inspector.
Following review of the reported event, attendant calculations and associated documentation, engineering personnel determined that the condition described in event notification EN50800 did not place the plant in an unanalyzed condition. Revision 1 of a calculation for the containment internal structural analysis demonstrated that when the safety injection tanks 'B' and 'D' are 100% filled in an outage condition, approximately a 10% safety margin is maintained. This revision was the calculation of record at the time the safety injection tank levels were raised above 74%, in October, 2013. Revision 2 of the calculation was completed to remove excess conservatism and to provide a closer representation of available margin. In addition, margin was also improved by limiting tank level to 74%. However, improving margin by limiting tank level to 74% does not result in an unanalyzed condition when tank level is 100%, as adequate margin remains. Therefore this event is being retracted. The licensee will notify the NRC Resident Inspector. Notified the R4DO (Okeefe). |
ENS 50797 | 9 February 2015 11:05:00 | The Unit 2 Stack High Range Radiation Monitor (RM-8168) was removed for service for planned maintenance. There is no significant effect of this planned maintenance on the plant. The licensee notified the NRC Resident Inspector, the State of Connecticut and Waterford township. |
ENS 50794 | 7 February 2015 07:37:00 | On 2/6/15 at 2300 (CST) the Division 1 Reactor Water Cleanup (RT) system differential flow instrument was declared inoperable due to erratic indication. The Division 1 RT differential flow instrument was declared inoperable in accordance with Technical Specification 3.3.6.1 Action D.1. At time 2355 Division 2 RT differential flow instrument failed downscale and was declared inoperable in accordance with Technical Specification 3.3.6.1 Action D.1 and also Technical Specification 3.3.6.1 Action E.1 (entered due to Division 1 RT differential flow already inoperable). Since this condition renders the Leakage Detection System incapable of performing its safety function, it is reportable under 10CFR50.72(b)(3)(v)(C). Division 1 RT differential flow was declared Operable at time 0036 on 2/7/15. Division 2 RT differential flow was restored to Operable at time 0225 on 2/07/2015. The NRC Resident (Inspector) has been notified. |
ENS 50837 | 20 February 2015 11:30:00 | The following information was received from the State of Oklahoma via email: (The Oklahoma Department of Environmental Quality has) been notified by Globe X-Ray Services (OK-15194-02) that one of their assistant radiographers received a reported dose of 5.083 R for the month of January, 2015. The assistant has been suspended and stated that he dropped his badge at some point during the monitoring period but did not report it until now. Landauer reported that the reading was 'inconclusive'. Investigation is ongoing. |
ENS 50773 | 28 January 2015 11:38:00 | The Comanche Peak Primary Emergency Operations Facility (EOF) will be unavailable during planned maintenance on the EOF ventilation system. On January 28, 2015, CPNPP (Comanche Peak Nuclear Power Plant) began planned work to improve the reliability of the EOF ventilation system. The EOF will be unavailable for approximately three weeks. During the time the primary EOF is unavailable, the affected ERO members will respond to the Backup EOF in Granbury, Texas for any declared emergency event. Therefore, it is expected that appropriate assessment of plant conditions, notifications, and communications could still be made, if required, during the time the primary EOF is unavailable. The extended unavailability of the primary EOF is being reported in accordance with 10CFR50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. The NRC Resident (Inspector) has been notified. A follow-up ENS communication will be made when the primary EOF availability is restored.
The availability of the Comanche Peak Primary Emergency Operations Facility (EOF) has been restored following planned maintenance on the EOF ventilation system. On January 28, 2015, CPNPP (Comanche Peak Nuclear Power Plant) began planned work to improve the reliability of the EOF ventilation system. The unavailability of the primary EOF was reported in accordance with 10 CFR 50.72(b)(3)(xiii). This report serves as a follow-up to Event Number 50773. The licensee notified the NRC Resident Inspector. The R4DO (Okeefe) was notified. |
ENS 50772 | 28 January 2015 08:56:00 | At 0535 CST on 1/28/15, control room staff identified that valve EJHV8716A, RHR A To SIS (Safety Injection System) Hot Leg Recirc Loops 2&3 (isolation valve), had been closed per clearance order C20-D-EJ-A-005 to support maintenance on the A RHR system. Closing valve EJHV8716A placed Wolf Creek in TS 3.0.3. At 0550 CST on 1/28/15, power was restored to valve EJHV8716A and the valve was opened, allowing the unit to exit from TS 3.0.3. Subsequent reviews of clearance order C20-D-EJ-A-005 identified that valve EJHV8809A had been energized and closed per direction of the clearance order. TS 3.0.3 was reentered at 0635 CST due to discovery of this condition. At 0650 CST, valve EJHV8809A was opened and deenergized allowing exit from TS 3.0.3. The licensee notified the NRC Resident Inspector. |
ENS 50747 | 21 January 2015 13:05:00 | The following information was received by email: On January 21, 2015, the licensee notified the Agency (Texas Department of State Health Services) that on January 20, 2015, during the process of closing fixed nuclear gauge shutters at its facility in order to perform detector calibrations, it discovered that the shutter on one of its Ohmart-Vega SH-F2 gauges, containing a 500 millicurie cesium-137 source, would not close. The gauge normally operates with the shutter in the open position and the failure does not pose a risk of exposure to any person. The licensee is coordinating to have repairs made. Further information will be provided as it is obtained in accordance with SA-300. Texas Incident: I 9269 |
ENS 50744 | 19 January 2015 14:24:00 | The missile door (door 33012) protecting Class 1E Engineered Safety Features (ESF) buses NB01/NB02 switchgear rooms was discovered misaligned on its hinge and stuck partially open and not capable of being closed. The missile door has since been repaired and closed. Technical Specification (TS) 3.8.9, 'Distribution Systems- Operating,' was declared not met and Condition F entered when the immediate operability determination identified that buses NB01 and NB02 were inoperable. Condition F of TS 3.8.9 requires immediate entry into Limiting Condition for Operation (LCO) 3.0.3. LCO 3.0.3 was entered at 1100 CST and subsequently exited when the missile door was repaired at 1118 CST. The unit was in and still is in MODE 1 at 100% power. No actions were initiated to commence a unit shut down. The NRC resident inspector was contacted regarding this event. All systems functioned as expected.
The licensee is retracting this event based on the following: An engineering evaluation concluded that the weather conditions during the period of the event did not result in the threat of a tornado. Given that the weather during the event would not have presented a valid threat of a tornado, the stuck open missile door would not have prevented the ESF busses and the DGs (Diesel Generators) from performing their specified safety function. The ESF busses and the DGs were considered OPERABLE but degraded. This is analogous to Example 4 in RIS 2001-09, 'Control of Hazard Barriers', with the exception that this event did not occur as a result of planned maintenance or a plant modification. As such, this event has been determined to not be reportable per 10 CFR 50.72(b)(3)(v)(D). The licensee has notified the NRC Resident Inspector. Notified R4DO (Gepford). |
ENS 50745 | 20 January 2015 15:19:00 | The following report was received from the State of Louisiana via email: Event Date and Time: 01/16/2015, around (1515 CST) a radiography crew was working at the ExxonMobil Refinery on Scenic Highway, Baton Rouge, LA. The event was reported (about 1645 CST) on January 17, 2015, by a phone call from (an individual) who represented himself as the Corporate RSO. He stated he drove down on January 17, to evaluate and investigate this incident. He reported this incident appears to be a Human Error Potential Excessive Exposure. Event Location: ExxonMobil Refinery 4999 Scenic Highway. Baton Rouge, LA 70805. A temporary jobsite for Acuren Inspection. Event type: This is a potential excessive exposure involving a radiographer attempting to breakdown a radiography exposure setup. He attempted to disconnect the guide tube from the exposure device and the source was not locked in the shielded position. It was noticed that the locking device was red after the guide tube was handled to disconnect it from the exposure device. Notifications: LA DEQ (Department of Environmental Quality), Assessment, Radiation by direct phone call to our after hours answering system. The notification came in around (1645 CST) on January 17, 2015. Event Description: The radiography crew was making exposures on lower level equipment at the ExxonMobil Refinery. The crew was utilizing (about) 38 Ci of Ir-192. The crew attempted to breakdown/disconnect the equipment after the exposures. The guide tube would not disconnect. The 2nd hand of the crew manipulated the drive cable that returned the source into the shielded position. A quarter turn on the crank shielded the source. The radiographer and his equipment were checked. His pocket dosimeter was off scale, but he claims his Alarm Rate Alarm meter did not alarm. A second check of the Alarm Rate Meter revealed the unit did alarm, but it was a weak alarm. Estimated dose calculations were done for his whole body and extremities. His whole-body estimated dose was 3.3 Rads and his extremity dose was estimated at 206 Rads to his hands. These were calculated on a one minute exposure where a .5 minute is more realistic. The exposed radiographer was taken to Core Occupational Medicine for examination, x-rays and blood work. He is being monitored and examined every other day. At this time he has been asymptomatic for an excessive radiation exposure. The Licensee is conducting reenactments. This incident is not considered closed by the Department (LA DEQ). The investigation findings will be updated when they become available. The equipment was all QSA equipment loaded with 38 Ci Ir-192. This appears to be an operator error exposure. The source is secure from removal and unnecessary exposure. This event is not closed and additional investigation and evaluation will continue. The source is in a safe shielded position and no threat to workers or the general public. Transport vehicle description : N/A This was at a temporary job site inside the ExxonMobil Refinery located in Baton Rouge, LA. License Numbers: LA-7072-L01, AI 126755 Louisiana Event Number: LA1500002 |
ENS 50746 | 20 January 2015 15:44:00 | The following information was received by email: Event Description: A specific licensee reported the loss (potential theft) of two generally licensed tritium (H-3) exit signs. The signs had been removed from installation and are missing from the storage location. The licensee became aware of the missing material on 1/14/2015, however, it is possible the material was missing for approximately one month. The devices were last seen by licensee staff in mid-December 2014. Specific details about the device model and activity have not yet been provided, potentially up to 15 Ci H-3 per device. An NJDEP (New Jersey Department of Environmental Protection) inspector will visit the site to investigate the incident. New Jersey Event: #C545467
The following update was received from the New Jersey Department of Environmental Protection via email: The College of New Jersey (TCNJ) reported the loss or possible theft of two radioluminescent exit signs. The signs had been removed from installation and were placed into storage. The signs were discovered to be missing from the storage location on 1/14/2015. However, TCNJ stated that it was possible the signs had been missing for approximately one month. The signs were last seen by TCNJ in mid-December 2014. Each sign contained approximately 137 GBq (3.7 Ci) of H-3. A New Jersey Department of Environmental Protection inspector visited the site on 1/21/2015 and determined the most likely scenario was accidental disposal. The licensee submitted a full report in accordance with N.J.A.C. 7:28-6.1 (10 CFR 20.2201). The missing signs were not located. Although there was no evidence of contamination, the licensee estimated a dose of 8.5 mrem if the signs had been broken. NJDEP tracked the event as Incident # C545467 and Investigation #: 507375-INV15001. Corrective Actions: Commitment to increased security should they be in possession of other sources. (The College of New Jersey) reports no inventory of generally licensed devices. Device Number: 1 Device/Equipment: RADIOLUMINESCENT EXIT SIGN Model Number: LE Manufacturer: SAFETY LIGHT CORP Serial Number: 201864 Device Number: 2 Device/Equipment: RADIOLUMINESCENT EXIT SIGN Model Number: LE Manufacturer: SAFETY LIGHT CORP Serial Number: 201865 Notified R1DO(Powell), NMSS Events Notification and ILTAB (email). 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 |
ENS 50732 | 12 January 2015 23:17:00 | At 1939 (EST) on January 12, 2015, Nine Mile Point Unit 2 entered Tech Spec 3.6.4.1 when secondary containment was declared inoperable due to secondary containment differential pressure being above the Tech Spec Surveillance Requirement of -0.25 inches vacuum water gauge. This condition is related to sustained high winds. At 1956 on January 12, 2015 the differential pressure was restored, the secondary containment was declared operable and the Tech Spec 3.6.4.1 exited. Secondary containment being inoperable is a 8-hour report for 10 CFR 50.72(b)(3)(v)(c), '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'. The NRC Resident Inspector has been notified. The licensee notified the New York Public Service Commission. |
ENS 50731 | 12 January 2015 13:37:00 | The following information was received from the State of Texas via email: On January 12, 2015, the licensee notified the Agency (Texas Department of State Health Services) that during routine fixed nuclear gauge inspections and shutter checks, it discovered that the shutter on one of its Ronan Model SA1-C5 gauges, containing a 300 millicurie cesium-137 source (SN: 6409GK), was stuck in the open position. (The event occurred at the licensee site located in Evadale, Texas.) This gauge normally operates with the shutter in the open position. The gauge is mounted on a vessel that the licensee does not enter. There is no risk of exposure to any individual. The licensee is contacting the manufacturer to schedule repair. Further information will be provided as it is obtained in accordance with SA-300. Texas Incident: I-9266 |
ENS 50718 | 6 January 2015 17:18:00 | In accordance with 29 CFR 1904.39(a)(2), notification was made to the Occupational Safety and Health Administration regarding the in-patient hospitalization of an individual while in the owner controlled area. The licensee has notified the NRC Resident Inspector. The licensee notified the State of Ohio and local authorities The individual employee is currently under medical treatment and is not contaminated. |
ENS 50715 | 3 January 2015 14:14:00 | At 0538 (CST) on 01/03/2015, a partial loss of the 25KV Power Distribution System caused a loss of both the Primary and Backup Meteorological Towers at the Comanche Peak Nuclear Power Plant. This resulted in a major loss of emergency assessment capabilities in regard to meteorological conditions. The 25KV Plant Support Power Loop feeds certain non-safety-related support equipment and did not result in an impact to plant/unit operation. Meter and Relay technicians are performing ongoing investigation of the cause of the breaker trip. A similar incident occurred 01/02/2015 at 0314 (CST). Reference (NRC) EN# 50713. The licensee will notify the NRC Resident Inspector. |
ENS 50714 | 2 January 2015 16:45:00 | At 0925 (EST) on 1/2/2015, Engineering personnel identified a gas void in each of two Unit 4 cold leg High Head Safety Injection (HHSI) discharge lines which exceeded procedural gas accumulation acceptance criteria. This condition rendered the cold leg HHSI flow path inoperable and required entry into Technical Specification 3.0.3. The voids were vented and Technical Specification (TS) 3.0.3 was exited at 1032. The testing for gas voids conducted on 1/2/2015 was a follow-up to a gas void found in one Unit 4 HHSI line on 12/26/2014. On 12/26/14, TS 3.0.3 was entered at 1020 and exited at 1048 after the gas void was vented. After further review, the 12/26/2014 event was also reportable in accordance with 10 CFR 50.72(b)(3)(v)(D). Engineering evaluation will be performed for both events to determine the specific impact of the gas voids on HHSI system function. Cause evaluation is being conducted to determine the source of the gas and any needed corrective actions. Unit 3 was verified to not have this voiding issue. The licensee notified the NRC Resident Inspector.
On 1/2/2015 at 1645 (EST), Event Notification 50714 reported to the NRCOC (Nuclear Regulatory Commission Operations Center) gas voids detected on 12/26/2014 and 1/2/2015 in Unit 4 cold leg High Head Safety Injection (HHSI) piping which exceeded procedural gas accumulation acceptance criteria. Subsequent analysis has determined that if a HHSI pump started with the measured gas voids present, the resulting system conditions would not have impacted the integrity of the Unit 4 HHSI discharge flow path to the reactor coolant system (RCS) and therefore its safety related function would not be impaired. The Unit 4 HHSI system discharge flow path to the RCS had been operable with the voids present, Technical Specification (TS) requirements were met, and entry into TS 3.0.3 was not required. The HHSI system remained capable of fulfilling the safety function to mitigate the consequences of an accident on Unit 4. Therefore, the immediate notification to the NRCOC on 1/2/2015 at 1645 in accordance with 10 CFR 50.72(b)(3)(v)(D) is hereby retracted. The NRC Resident Inspector has been informed. Notified R2DO (Shaeffer). |
ENS 50713 | 2 January 2015 11:43:00 | At 0314 (CST) on 01/02/2015, a partial loss of the 25KV Power Distribution Systems caused a loss of both the Primary and Backup Meteorological Towers at the Comanche Peak Nuclear Power Plant. This resulted in a major loss of emergency assessment capabilities in regard to meteorological conditions. The 25 KV Plant Support Power Loop feeds certain non-safety-related equipment and does not affect plant operation. An investigation by Meter and Relay Technicians revealed no abnormal conditions/damaged equipment. The supply breaker was re-closed at 0915 on 01/02/2015. Power was restored to both Meteorological Towers and proper operation was verified. The licensee notified the NRC Resident Inspector. |
ENS 50712 | 1 January 2015 22:02:00 | During (surveillance) checks of Control Room doors, a boundary door did not latch after being accessed until the door was opened and closed. This is being reported as it could have prevented the fulfillment of a safety function to mitigate the consequences of an accident per 10CFR50.72(b)(3)(v)(D). The door is currently closed and latched. The door was in this condition for between 5 and 10 seconds. The licensee notified the Connecticut Department of Environmental Protection, Town of Waterford and the NRC Resident Inspector.
Upon further review, Millstone Power Station Unit 2 has concluded that there was no loss of safety function, because even with the control room door latch degraded, the control room door and its closing mechanism would still be able to maintain the control room envelope's boundary intact. Therefore, this condition is not reportable and NRC Event Number 50712 is being retracted. The basis for this conclusion will be provided to the NRC Resident Inspector. The licensee notified the NRC Resident Inspector. Notified R1DO (Jackson). |
ENS 50609 | 13 November 2014 11:58:00 | The following information was received from the State of Texas via email: On November 13, 2014, the licensee notified the Agency (Texas Department of State Health Services, Radiation Branch) that on November 12, 2014, at approximately (1500 CST) it was notified by local law enforcement that one of its industrial radiography cameras had been found on the side of a road approximately 3 miles from the licensee's facility. The licensee determined that the SPEC 150 camera containing a 38 Curie Iridium-192 source had fallen from one of its trucks at approximately (1245 CST) while in route to a temporary job site. The camera had not been secured inside the truck before it left the licensee's facility but had been left on the tailgate. The licensee retrieved the camera and there was no apparent damage to the device. There are no known exposures resulting from this event. Further information will be provided as it is obtained in accordance with SA- 300. Notified DHS SWO, DOE, FEMA, HHS, NICC, USDA and EPA. Notified FDA, NuclearSSA and DNDO-JAC via email. Texas Incident: I-9251 THIS MATERIAL EVENT CONTAINS A "CATEGORY 2" LEVEL OF RADIOACTIVE MATERIAL Category 2 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 a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioactive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachytherapy. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
ENS 50568 | 25 October 2014 13:42:00 | The normal and high range ventilation process radiation monitors (3HVR*RE 10 A/B) were removed from service for preplanned maintenance for greater than 72 hours. The radiation monitors will be returned to service following maintenance and completion of operational testing. The licensee has notified the NRC Resident Inspector, the State of Connecticut and Waterford township.. |
ENS 50567 | 24 October 2014 20:59:00 | The normal range Supplemental Leakage Collection and Release System (SLCRS) was removed from service due to a sample pump failure. Compensatory sampling and monitoring is being established per the Radiological Effluent Monitoring and Offsite Dose Calculation Manual. Following the sample pump repairs and restoration, the Radiation Monitor will be returned to service. The licensee notified the NRC Resident Inspector, the State of Connecticut and Waterford township. |
ENS 50533 | 14 October 2014 07:37:00 | This notification is being made as required by 10 CFR 50.72(b)(2)(iv)(B) due to a Farley Nuclear Plant Unit 2 manual reactor trip. The trip was initiated when the in service train of CCW cooling to the Reactor Coolant Pumps was lost due to a loss of the 2B Start Up Transformer (SUT). The control room team manually tripped the reactor then tripped all three reactor coolant pumps as required by station procedure. There was a line of severe thunderstorms with lightning passing through the plant site at the time of loss of the 2B Start Up Transformer. The 2B emergency diesel generator was out of service for maintenance therefore there was a loss of the 'B' train emergency power 4160V electrical bus ('B' train LOSP (Loss of Offsite Power)). 'A' train emergency power remained energized from offsite sources. The plant is stable at normal operating pressure and temperature. At 0433 (CDT), 2B Reactor Coolant Pump was re-started when support conditions were re-established. Heat sink is adequate using the 2A Motor Driven Auxiliary Feedwater Pump. Unit 2 'B' train power was restored by starting the 2C emergency diesel generator at 0523 (CDT). This restored power to the Digital Rod Position Indication system, and control rod K-8 in control bank 'C' indicated full out, and all other control rods fully inserted. An emergency boration is in progress to compensate for the stuck rod. Additionally, the reactor trip resulted in a valid actuation of the Aux Feedwater system which is an eight hour non-emergency report per 10 CFR 50.72(b)(3)(iv)(A). During the transient, one primary PORV momentarily opened, then reseated. Decay heat is being directed to the atmospheric relief valves with no indicated primary to secondary leakage. There was no impact on Unit 1. The licensee notified the NRC Resident Inspector.
Digital rod position indication troubleshooting was conducted on 10/14/2014 and confirmed all control rods, including control rod K-8, fully inserted following the reactor trip. The licensee notified the NRC Resident Inspector. Notified the R2DO (Ayres), NRR EO (Davis) and IRD (Gott). |
ENS 50532 | 14 October 2014 01:30:00 | During the plant response to the trip of the B Recirculating water pump, reactor water level rose to the HPCI (High Pressure Core Injection) high water level trip setpoint as indicated on the associated instrumentation. With this high water level trip actuated, the HPCI high drywell pressure initiation signal would not have allowed the HPCI system to perform its intended safety function if required. If the HPCI system received the low water level initiation signal, the system would have been able to perform Its intended safety function. This high water level signal was actuated from 1935 (EDT) until reset at 1940 (EDT). This is reportable under 50.72(b)(3)(v). The licensee notified NRC Resident Inspector.
Further review has determined that the condition was not a result of procedural errors/inadequacies, equipment failures, or design / analysis inadequacies. Plant systems responded as per design when the HPCI system high water level trip actuated when reactor vessel water level rose to the HPCI high water level trip setpoint. HPCI initiation has two logics: one for low-low vessel water level and the other for a high drywell pressure. A vessel low-low water level is an indication that reactor coolant is being lost with a need for HPCI injection for core cooling. High drywell pressure could indicate a line break in the Reactor Coolant Pressure Boundary inside the drywell. The HPCI level instrumentation is designed to shut down the HPCI system upon high water level to prevent HPCI turbine damage due to gross moisture carryover and will re-initiate HPCI if vessel water level drops to the initiation water level setpoint. A HPCI high drywell pressure initiation signal, above setpoint, would have made up the logic for HPCI initiation and as per design, HPCI would have injected at the vessel low low level setpoint without operator action to reset the trip. In this instance, the trip was reset as prescribed by station procedures. HPCI was capable of performing its safety function after the high water level trip reset either by operator action or instrumentation (low low level initiation). The licensee will be notifying the NRC Resident Inspector. Notified R1DO (Rogge). |
ENS 50443 | 10 September 2014 11:55:00 | This is a non-emergency eight hour notification for a loss of Emergency Preparedness Capabilities. This event is reportable in accordance with 10 CFR50.72(b)(3)(xiii) as the condition affects the functionality of an emergency response facility. At approximately 0453 EDT on September 10, 2014, a fire alarm in an area near the primary Technical Support Center (TSC) triggered an automatic shutdown of the TSC ventilation system. The system was restored at 0755 EDT. The alternate TSC was available at all times. This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. |
ENS 50324 | 30 July 2014 13:27:00 | On July 30, 2014, at (0940 CDT), with the plant operating at 100% power, a review of an engineering analysis of the ultimate heat sink (UHS) determined that the UHS had been in an unanalyzed condition that degraded plant safety. This condition was the result of a design basis deficiency for the UHS that did not account for the adverse effects of system leakage on compliance with the 30-day inventory required by Regulatory Guide 1.27. The system design basis requires that 30-day inventory be maintained, with the assumption that no replenishment of the UHS occurs for the entire duration of the postulated event. In support of the development of the engineering analysis, compensatory measures have been implemented which provide adequate assurance that the UHS will perform its design safety function. Corrective actions to restore full compliance with design basis requirements are in development. This event is being reported in accordance with 10 CFR 50.72 (b)(3)(ii) as an unanalyzed condition that degraded the safety function of the UHS. The licensee notified the NRC Resident Inspector. |
ENS 50323 | 30 July 2014 12:06:00 | The following information was received from the State of Texas by facsimile: On July 29, 2014, the Agency (Texas Department of State Health Services) was notified by the licensee that a Humboldt Scientific Inc. model 5001 moisture density gauge containing a 10 millicurie Cesium-137 and a 40 millicurie Americium-241 source was damaged at a field site. The technician had placed the device at a sample location and extended the cesium source into the inspection hole. The area had been compacted and the technician did not believe there was any heavy equipment in the area he was working. While waiting for the results of the sample, the technician received an email. The technician needed his glasses to read the email so he walked 70 feet to his truck to get his glasses. While at the truck, the device was run over by a soil compactor (steam roller). The technician went to the gauge and restricted access to the area. He then contacted his Radiation Safety Officer (RSO). The licensee's RSO contacted a service provider (SP) who responded with the RSO to the scene. The device case was severely damaged, but the licensee was able to return the Cesium source to the shielded position and secure it in position. The SP's technician verified the Americium source was still in the device. The SP's technician surveyed the device and did not find any abnormal dose rates. The SP took the damaged device to their facility for disposal. No individual received any significant additional exposure due to this event. Additional information will be provided as it is received in accordance with SA-300. Texas Incident: I-9216 |
ENS 50452 | 12 September 2014 12:05:00 | The following information is provided as 60 day telephone notification to the NRC in accordance with 10 CFR 50.73(a)(1) reported under 10 CFR 50.73(a)(2)(iv)(A) for an invalid actuation of Division 1 Shutdown Service Water (SX) and an invalid actuation signal for Division 1 Containment Isolation Group 12. This event occurred on July 26, 2014, at 1648 CDT. As allowed by 10 CFR 50.73(a)(1) this notification is made via telephone. (a) The specific train(s) and system(s) that actuated were: On July 26th, during lightning strikes on the switchyard grid system, the Division 1 SX (Shutdown Service Water) auto-started as a result of momentary loss of power to the Low Pressure Auto Start Relay. A lightning strike causing voltage transients also caused a Division 1 Group 12 Containment Isolation signal affecting DIV 1 Hydrogen monitor. (b) Whether each train actuation was complete or partial. Upon receiving the invalid signal from momentary loss of power, for Division 1 SX and Group 12 Containment Isolation signal, the systems responded as expected for existing plant conditions. For group 12 isolation, (Containment Monitoring) 1CM011/12/47 and 48 valves closed from their normally open position. For DIV 1 Shutdown Service Water (SX), the start of the Shutdown Service Water (SX) pump and alignment of valves operated as expected. The actuation was considered a complete Division 1 SX and Division 1 Group 12 Containment Isolation actuation. Containment Isolation Signals: The following Group 12 valves closed and associated shunt trips occurred on a loss of power to (Radiation Monitors) 1RIX-PR001A/1C: 1CM011, 1CM012, 1CM047, and 1CM048. (c) Whether or not the system started and functioned successfully. Upon receiving the invalid signal from momentary loss of power, Division 1 SX and Containment Isolation signals started and functioned successfully. The NRC Senior Resident Inspector has been notified. |
ENS 50312 | 26 July 2014 10:44:00 | (A Honeywell) employee with dust in his right eye reported to the on-site dispensary this morning. The plant nurse administered first aid. A whole body survey of the employee in his plant clothing was performed; the maximum amount of contamination present was on the employee's work boots, 331,597 dpm/100 sq.cm. The plant nurse allowed the employee to return to work following treatment. The employee remained inside the Restricted Area over the entire course of the event. The licensee will notify the NRC Resident Inspector and Region 2 (David Hartland via email). |
ENS 50303 | 24 July 2014 07:44:00 | On July 24, 2014 at 0800 EDT, the Unit 2 Radiation Monitoring Computer System will be inoperable and unavailable due to pre-planned maintenance to implement an upgrade to the system. Alternate methods for monitoring are being utilized. The Radiation Monitor Computer System is expected to be restored to available status at approximately 1700 EDT on July 24, 2014. This notification is being made in accordance with 10CFR 50.72 (b)(3)(xiii), as an event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. An update will be provided once the Radiation Monitoring Computer System has been restored to normal operation. The NRC Resident Inspector has been notified.
The Unit 2 Radiation Monitoring Computer System has been restored to an operable status as of July 24, 2014 at 1650 EDT. The NRC Resident Inspector has been notified. Notified R2DO(Musser) |
ENS 50284 | 18 July 2014 08:28:00 | The following information was received from Velan Inc by facsimile: SUBJECT NOTIFICATION: 2 INCH BONNETS, VELAN PART NUMBER 8943-014 On May 16, 2014 (Velan) received notification from Westinghouse Electric Co. (WES) that 2 bonnets supplied by Velan to WES in early 2013 for installation at Comanche Peak exhibited the following issues: - The bonnets were intended to be exact replacements for the bonnets built to drawing E73-020 Rev E (OEM is Velan) except for material change to SA-182 FXM-19. Bonnets were visually inspected when received at site. No issues were noted; both bonnets appeared to be identical. - In April 2013, Unit 1 bonnet was installed in valve 1-8109. No issues were noted with the installation. The new bonnet was put into service. -In April 2014, Unit 2 installation was scheduled to begin. After the disassembly of the valve, the old and new bonnets were compared. It was noted that the backseat dimensions are different between the 2 bonnets. The increase in backseat diameter on the new bonnet would cause the stem to not backseat. The decision was made to re-install the old bonnet and send the new bonnet back to the OEM, Velan. On June 10, 2014 the bonnet, identified in the last bullet above, arrived at Velan Plant 2. The review by the (Velan) Evaluation Committee was finalized on July 17 and concluded that: -Four similar bonnets were delivered to WES on three different occasions in 1988 and early 90's -The stem head diameter is 01.312 (inches) so, when opening, the stem may pass through the stem bore of the bonnet and not seat on the backseat. -On opening, if the limit switches on the actuator do not function, the stem may enter the packing chamber. The packing may be deformed and a leak may develop. Stem travel is limited by the disc contacting the bonnet and/or the end of the stem thread stopping on the actuator drive nut. -If the actuator and packing flange nuts are removed, there is the potential for the stem to blow out of the valve. -The packing chamber depth will result in more packing being installed in the valve. This may result in a higher packing friction load on the actuator when operating and reduce the actuator margin. -The smaller packing chamber will not affect safety. A different diameter packing may be required. The gland bushing diameter (01.744 inches) is less than the packing chamber diameter and will work correctly. These bonnets were fabricated against ASME Sec. Ill for installation in Class 2 systems. Not knowing exactly the nature of the application we cannot determine if the (above identified) potential issues may pose a significant safety hazard and therefore we have informed WES by way of a similar letter.
Reporting Organization/Supplier who made the original event report on 07/18/2014 reported to the NRC Operations Center that the Event Notification posted has a typographical error regarding the Velan, Inc part number described in the report. The original documentation provided was concerning Velan Part Number 8943-014 which was mistakenly transcribed as 6943-014 in the original report. This error has been corrected in this updated report. Notified R4DO (Okeefe) and Part 21 Group via email. |
ENS 50261 | 7 July 2014 22:12:00 | On July 7, 2014 at approximately (2040 EDT), an issue was discovered with currently removed Electromatic Relief Valves (EMRVs) that calls the operability of the currently installed EMRVs into question. Based on this new information, all 5 of the currently installed EMRVs were conservatively declared inoperable. With the potential of 5 EMRVs inoperable a Technical Specification shutdown is required under Technical Specification 3.4.b, whereby reactor pressure shall be reduced to 110 psig or less within 24 hours. This event is immediately reportable under: 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' 50.72(b)(3)(v)(D), '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: (D) Mitigate the consequences of an accident.'" The licensee notified the NRC Resident Inspector. The licensee will notify the State of New Jersey and issue a press release.
The original concern and July 7, 2014 EN (report) was caused by abnormal wear found on removed Electromatic Relief Valve actuators. However, on July 8, 2014, upon completion of in situ testing as well as visual examination of the installed EMRV actuators, it was determined that the 5 currently installed EMRVs were fully operable and capable of performing their safety function. Therefore, Oyster Creek is retracting the notifications made under 50.72(b)(2)(i) and 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector and the State will be notified. The licensee will be issuing a press release. Notified the R1DO (DeFrancisco). |
ENS 50236 | 28 June 2014 14:53:00 | At 1215 PM CDT on June 28, 2014, Xcel Energy notified the Minnesota State Duty Officer that approximately 1000 dead fish of assorted species were discovered in the site's discharge canal. The fish perished due to rising temperatures in the discharge canal caused by an orderly shutdown of Cooling Towers as directed by procedure for response to flooding of the Mississippi River. The discharge canal is a body of water with a barrier from the Mississippi River. The Mississippi River temperature across the site has only risen 0.9F which is well within the permit limit of the maximum of 5F. There is no impact to the availability of safety systems or safety system performance. This condition has no impact to public health and safety. This event is reportable per 10 CFR 50.72(b)(2)(xi), an event related to protection of the environment for which a notification to other government agencies has been made. The licensee has notified the NRC Resident Inspector (and local authorities). |
ENS 50235 | 27 June 2014 20:33:00 | At approximately 1535 Pacific Daylight Time (PDT) the Diablo Canyon Power Plant (DCPP) Emergency Planning Supervisor received a notification from an offsite DCPP employee that one of the emergency plan sirens had inadvertently actuated. The DCPP Shift Manager was notified of the situation by approximately 1545 PDT. The County of San Luis Obispo was notified of the inadvertent actuation of the single siren. At approximately 1550 PDT the County of San Luis Obispo sent out a county wide alert stating, 'Civil Emergency in this area until (1610) PDT prepare for action'. At approximately 1600 PDT the County of San Luis Obispo sent out another county wide alert stating, 'An early warning system siren was sounded in error. There is no emergency.' Field technicians have been sent out to the siren location and will shut the siren off. The cause of the inadvertent actuation of the siren is not known at this time. The licensee notified the NRC Resident Inspector and the County of San Luis Obispo. |
ENS 50234 | 27 June 2014 16:21:00 | On June 27, 2014, at approximately 1131 hours EDT, it was determined that an individual who is licensed under Part 10 CFR Part 55 to operate a power reactor was in violation of the FENOC (FirstEnergy Nuclear Operating Company) Fitness for Duty policy. This condition is being reported pursuant to 10 CFR 26.719(b)(2)(ii) . The NRC Resident Inspector has been notified. |
ENS 50230 | 26 June 2014 15:29:00 | A non-licensed employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The employee's access to both the McGuire and Oconee plants has been restricted. The NRC Resident Inspectors at both the McGuire and Oconee sites have been notified. |
ENS 50232 | 27 June 2014 12:56:00 | Notified by the licensee that three tritium exit signs were improperly disposed of as trash. The signs were manufactured by SRBT with an estimated source strength of 20 Ci for each sign. The licensee noticed the missing signs following an equipment inventory taken after a renovation. The licensee determined that the three signs were mixed with general construction debris and sent to a landfill. The licensee investigation is continuing. 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 |
ENS 50228 | 26 June 2014 12:28:00 | The following information was received by facsimile from the State of Oregon: Event Description: OR Radiation Protection Services (RPS) was notified by facsimile on June 24, 2014 at 0813 (PDT), by Schnitzer Steel Industries, 12005 N. Burgard Way (Portland, Oregon) of one cesium-137 fixed gauge received in a truckload of scrap from Idaho that triggered their site entry detectors. RPS personnel investigated, found the gauge housing to be severely crushed, shutter damaged and partially open. One side of housing is split but compressed together. Identification plate still attached and mostly legible and given as follows: Manufacturer: Texas Nuclear; Model: 5197; Gauge housing serial number: B7951; Source: Cesium-137; Activity: Currently 65 mCi according to Thermo Fisher Scientific. Gauge is being stored in secured metal bin in restricted and remote area on company site. Highest exposure reading at bin surface measured at 1.48 mR/hour. Schnitzer personnel reported 60,000 microR/hour (60 mR/hour) at approximately 4 inches from split side of gauge housing surface when first discovered. Schnitzer personnel (one person) used shovel and 4 foot steel rod with hook to move gauge from truck to storage bin on June 23rd. Estimated dose to company person from reconstruction of gauge move is 1.50 mrem based on a 6/25/14 exposure measurement. No other company employees received a dose from this device and RPS personnel did not receive any appreciable dose. RPS personnel took contamination wipes of gauge housing and found no evidence of exterior contamination after analysis. The manufacturer, now Thermo Fisher Scientific, was contacted and found the gauge was originally sold to a Martell, California company, Wheelabrator Martell, Inc., on August 23, 1996. Thermo added the last contact for leak test services was December, 1999. RPS is currently investigating further to trace gauge history since 1999 as well as location in Idaho where truck originated. The gauge will not be returned to the manufacturer according to Thermo and classified as waste. A waste broker will be contacted by Schnitzer for packaging and disposal. RPS will be monitoring this process. Oregon State Incident: OR-14-0028.
State of Oregon provided clarification of source activity and exposure estimate. Notified the R4DO (Allen) and FSME Resources via email. 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 |
ENS 50171 | 4 June 2014 09:38:00 | The following information was received by the State of Texas by email: On June 3, 2014, the Agency (Texas Department of State Health Services) received notice that the licensee had a damaged Berthold LB300ML gauge (S/N 10004). The gauge had been hit with molten steel and damage had been done to attachment points ('ears') for a locking mechanism and carrying handle. No damage to the 2.5 milliCurie cobalt-60 source (S/N 1374-07-11) or the gauge shutter occurred. The gauge is scheduled for repair on June 5, 2014. Additional information will be supplied as it is received in accordance with SA-300. Texas Incident: I-9198
On June 23, 2014, the Agency (Texas Department of State Health Services) received a written report from the licensee stating the gauge shutter was not damaged during this event. The licensee stated they were able to close the shutter at the time of the event. The gauge was placed in a locked storage box. The manufacturer has completed repairs to the gauge and the gauge has been returned to service. The licensee stated no additional exposure was received by their employees or members of the general public. "Additional information will be provided in accordance with SA-300. Notified R4DO (Allen) and FSME Events Resource email. |
ENS 50146 | 28 May 2014 16:39:00 | The following information was received from the State of Louisiana by facsimile: Event Location: The radiography crew was at the MARATHON PETROLEUM COMPANY REFINERY, 4663 West Airline Highway, Garyville, LA, St. John the Baptist Parish. The crew was in the refinery to do radiography weld testing. The weather for the day was stormy and raining. The crew was in the vehicle waiting out the weather when a tornado ripped the darkroom off of the company vehicle and took the camera and source with the darkroom. Event type: The lost/found gamma camera is an AMERSHAM, MODEL 880 DELTA, S/N4045 exposure device housing 39 Ci of Ir-192 AEA Technology sources, Model #A424-9, S/N # 1040C. The camera was located within the refinery boundary and appeared to be undamaged and the source remained in the shielded area. The camera was surveyed and only background radiation levels were detected. The camera was then transported to QSA Global on Langley Dr. in Baton Rouge, LA for leak tests, radioactive source and DU (Depleted Uranium) mechanical evaluation. (QSA Global) stated the device would be disassembled for the evaluation. If the results of the evaluation are negative, the equipment will be removed from service. The leak test results for radiation and DU were negative for removable contamination. Notifications: LDEQ (Louisiana Department of Environmental Quality) was notified of the lost/found radiography source in a QSA Global, Amersham Delta 880 gamma camera/source holder housing a 39 Ci Ir-192 source. The source exposure device was located within the refinery boundary. After a field evaluation, the exposure device was moved to QSA Global in Baton Rouge, LA for a professional mechanical evaluation. The results of the evaluation will determine if this equipment will be returned to service or remain out of service. Event Description: On May 28, 2014, (a senior trainer and business ops manager) called LDEQ and reported that a radiography crew dispatched to the Marathon Refinery in Garyville, LA had been hit by a tornado on the jobsite. The incident was at the Marathon Refinery, 4663 West Airline Highway in Garyville, LA. The tornado ripped the darkroom off of the truck with the gamma camera in the darkroom. After the storm had passed, the crew and the Marathon RSO surveyed the area in an attempt to locate the radiography source. The source and exposure device were located, field test/inspected for damage and radioactive contamination. When the equipment was deemed safe, it was transferred to QSA Global for a professional mechanical evaluation. ALL EQUIPMENT AND PERSONNEL APPEAR TO BE SAFE AND UNHARMED. LDEQ CONSIDERS THIS EVENT CLOSED. THIS MATERIAL EVENT CONTAINS A "CATEGORY 2" LEVEL OF RADIOACTIVE MATERIAL Category 2 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 a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioactive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachytherapy. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
ENS 50240 | 30 June 2014 15:33:00 | Originally, an Ionscan 400, S/N 10586, was sent to Smith Detection for repair in October 2011. From that time until May 2014, Security Forces personnel believed that the device was still at Smith Detection awaiting repair. In May 2014, (The Air Force) was able to determine that according to Smith Detection records, following repair, the device had been returned to Lackland AFB in October 2012. Security Forces has no record of receiving the device and no notification that the device had been shipped. Thorough searches of Security Forces storage and use areas have been conducted without finding the device. One theory is that the device was sent to an old address for Security Forces. The building at that address has been searched without finding the device. Ionscan 400 contains a NI-63 source with an activity of 15 milliCi. The Ionscan 400 serial number is 10586. The sealed source and device registry number (SS&DR) associated with the Ionscan 400 is NR-0163-D-801-G. The device is used to screen passengers for hazardous material. The licensee will be notifying the NRC Region IV (Cook). 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 |
ENS 50087 | 6 May 2014 09:26:00 | The following information was received from the Commonwealth of Virginia: The licensee discovered a shutter stuck in the open position during a routine test of a fixed gauge on May 5, 2014. The gauge is a Ronan Engineering Model SA1, serial number M-7299. It is used as a low-level indicator in a pre-dryer vessel and contains a 26.9 milliCurie (decay corrected) cesium-137 source. The licensee indicated that using unusual force to try to close the shutter would likely damage the actuator rod mechanism. The shutter is kept in the open position during operations and does not pose an additional radiation exposure to personnel. The licensee performs radiation surveys at one foot from the gauge surface during routine tests. The maximum reported result for this gauge was 300 microR per hour. The licensee has contacted the manufacturer to repair the gauge. The Agency (Virginia Radioactive Materials Program) will continue to monitor the situation until the shutter is repaired. Virginia Event: VA-2014-004 |
ENS 50118 | 16 May 2014 16:15:00 | The following information was received from the State of Nevada by email: (A Common Carrier) misdelivered an Ir-192 source for 645 North Arlington Avenue, Suite 120, Reno, NV 89503. The address was correct, but the source was inadvertently delivered to the Main Hospital receiving at 235 West Sixth Street, Reno, NV 89503. The source bucket was then delivered to the Radiation Safety Officer (RSO) - at the correct address. Once received by the RSO, the delivering employee was contacted and questioned about length of contact and it was estimated that 2 mrem of dose was received for 10 minutes of contact. The employee as well as the Director for the receiving department were contacted and told to refuse shipment of anything with a radioactive label and to call the RSO immediately. The shipper (Varian) was contacted and they filed a formal complaint with (the Common Carrier). Nevada Event: NV140011 THIS MATERIAL EVENT CONTAINS A "CATEGORY 2" LEVEL OF RADIOACTIVE MATERIAL Category 2 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 a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioactive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachytherapy. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
ENS 50144 | 27 May 2014 16:22:00 | The following information was received from the State of Texas by email: On May 19, 2014, the Agency (Texas Department of State Health Services) received a complaint alleging multiple rule violations had been committed by an industrial radiographer working for the licensee. On May 27, 2014, during the complaint investigation, the licensee's site radiation safety officer (SRSO) stated that on April 25, 2014, the radiographer had taken one of the licensee's radiography trucks which contained a SPEC 150 exposure device with a 23 curie iridium-192 source following completion of a job in San Antonio, Texas. The radiographer had possession of the truck and exposure device with its source until April 30, 2014. The licensee's SRSO stated they attempted during that time to contact the radiographer and to locate the truck using its GPS locator system, but they were unsuccessful. The licensee located the truck at a recreational vehicle park in Portland, Texas. The licensee's SRSO stated the exposure device was found laying in the floor of the dark room, the dark room door was not locked, and the dark room's alarm was not activated. The SRSO stated to the Agency investigator during the May 27th telephone interview that it did appear to him that an exposure could have resulted to persons in an unrestricted area during the event. An investigation into this event is ongoing and more information will be provided as it is obtained in accordance with SA-300. Texas Incident: I-9196
The following was received from the State of Texas via email.: Investigation has determined that no theft occurred and this was not a reportable event. Investigation will continue in order to identify and address other potential regulatory violations. Notified R4DO (Taylor), Grant (IRD), Wray (ILTAB), and FSME Event Resource via email. THIS MATERIAL EVENT CONTAINS A "CATEGORY 2" LEVEL OF RADIOACTIVE MATERIAL Category 2 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 a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioactive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachytherapy. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
ENS 50055 | 24 April 2014 07:50:00 | On April 24, 2014 at 0230 (EDT), Secondary Containment Zone 3 (Unit 1&2 Reactor Building) differential pressure lowered to 0.10 (negative inches) WG (Water Gauge) when restoring Unit 1 Zone 3 HVAC during a routine swap of RPS power supplies, due to a trip of the Unit 1 Zone 3 Supply fan. Zone 3 differential pressure was restored to > 0.25 (negative inches) WG at 0243 hours. Zone I (Unit 1 Reactor Building) ventilation is isolated with secondary containment relaxed for refuel outage on Unit 1. Zone II (Unit 2 Reactor Building) ventilation remained in service and stable. Zone 3 differential pressure recovered to SR 3.6.4.1.1 requirements of 0.25 WG (negative inches) at 0243 hours and was verified to be stable. LCO 3.6.4.1 was entered at 0230 hours and exited at 0313 hours. Tech Spec Secondary Containment Operability requires a negative pressure of at least 0.25 (negative inches) WG for all three Reactor Building Ventilation Zones when secondary containment is required. This event is being reported under 10CFR50.72(b)(3)(v)(C) and per the guidance of NUREG-1022,Rev 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System. The licensee notified the NRC Resident Inspector. |
ENS 50036 | 14 April 2014 18:21:00 | The following information was received by email: On April 14, 2014 at approximately 0600 MST, the licensee discovered that a portable gauge was stolen out of the bed of a pick-up truck at a technician's home. The gauge was being housed in a transport box, which was chained to the bed of the truck. The lock was cut and the entire box was stolen. The gauge is a CPN MC Series, serial #MC380404231, and contained 50 millicuries of Cs-137 and 10 millicuries of Am-241. The Agency (Arizona Radiation Regulatory Agency) continues to investigate the event. The USNRC, California, Nevada, Utah, Colorado, New Mexico, FBI, and Governor's Office are being notified of this event (by the Agency). Arizona Event: 14-007 * * * UPDATE AT 1722 EDT ON 5/5/14 FROM GORETZKI TO SNYDER * * * The following information was received by email: On May 3, 2014, at approximately 0300 (MST), the Agency received a telephone call from the Phoenix Fire Department that they had responded to an incident at an apartment complex where a radioactive gauge was found. The gauge was found next to a dumpster by the apartment security guard while he was doing his rounds. The Agency responded and took possession of the gauge. The lock on the handle was still intact and the shutter was closed. It was noticed that the electronics looked like they had been tampered with but the gauge seemed to be in good condition otherwise. A survey of the gauge was performed, with nothing out of the ordinary noted. The investigation into this event is closed. Arizona Supplemental Notice: 14-007 Notified R4DO (Whitten), FSME Events Resource (email) and Mexico (email). 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 |
ENS 50035 | 14 April 2014 15:58:00 | The following information was received from the State of Texas by email: On March 14, 2014, the Agency (Texas Department of State Health Services) received a call from the radiation safety officer (RSO) of a licensee. The RSO stated that during an instrument check on Friday, April 11, 2014, technicians found two line level indicators with stuck shutters. The open position is the normal operating position for these types of instruments. The technicians stated there was no apparent damage to the rod or the shutter mechanisms. The level indicators are located on the sides of a very tall product receiver bin containing polyethylene resin. There were no exposures and employees are not routinely around these sources. The instruments contained Cesium-137 at 5000 milliCuries each. The level indicators are tagged as inoperable and a service company has been contacted to inspect and repair the instruments. Further information will be entered in accordance with SA300.
The first line of the initial report contained the wrong report date. The first line should have read: On April 14, 2014, the Agency received a call from the radiation safety officer (RSO) of a licensee. Additional information: The gauges are Ohmart model SHLM-BR-4 gauges. The sources are in the out position, but the licensee's RSO stated there has not been any additional exposure to radiation to any individual due to this event. Additional information will be provided as it is received in accordance with SA300. Notified the R4DO (Gaddy) and FSME Resources via email.
On April 15, 2014, the Agency was contacted by the licensee and informed they were able to retract the source and fully close the shutter placing the source in the fully shielded position. Additional information will be provided as it is received in accordance with SA-300. Notified the R4DO (Gaddy) and FSME Resources via email. Texas Incident: I-9182 |
ENS 49997 | 4 April 2014 12:13:00 | The Millstone Station Stack Radiation Monitor, RM-8169 was removed from service for pre-planned maintenance. This is reportable as a loss of assessment capability in accordance with 10 CFR 50.72 (b)(3)(xiii). RM-8169 was restored following pre-planned maintenance at 1150 EDT on 4/4/2014. The licensee notified the NRC Resident Inspector and applicable State and local agencies. |
ENS 49999 | 4 April 2014 16:37:00 | The following information was received by facsimile from the State of Texas: On April 14, 2014, the Agency (Texas Department of State Health Services) was notified by the licensee's Corporate Radiation Safety Officer (CRSO) that one of their radiography crews had experienced source disconnect while using an INC IR 100 exposure device. The device contains a 70 curie iridium - 192 source. The CRSO stated the radiographers had completed a shot and noted the dose rates on the camera had not returned to normal after retracting the source. The radiographers increased the barriers around the exposure device and contacted their supervisor for assistance. The CRSO stated the source had been recovered by an individual listed on their license to retrieve sources. The CRSO stated he did not believe any individual received an exposure which exceeded any limits. The CRSO stated no member of the general public received any exposure. The CRSO stated he had just learned of the event and was beginning his investigation and would provide the Agency with additional information as it was obtained. This Agency will provide additional information as it is received in accordance with SA-300. Texas Incident: I-9178 |
ENS 49948 | 24 March 2014 05:14:00 | At 0031 (EDT) on March 24, 2014, Nine Mile Point Unit 2 was lowering power for the planned refueling outage. The loss of reactor building heating resulted in the isolation of the reactor building to maintain building temperature. Isolation of the reactor building resulted in the isolation of the reactor building vent radiation monitor (Vent WRGMS) which is a loss of emergency assessment capability. The isolation of the reactor building also resulted in declaring secondary containment inoperable due to secondary containment differential pressure being positive. Secondary containment was declared operable at 0034 (EDT) when differential pressure was restored to greater than negative 0.25 inches water gauge. Secondary containment being inoperable is a 8-hour report for 10 CFR 50.72(b)(3)(v)(c), 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. The unplanned isolation of Vent WRGMS is a 8-hour report for 10 CFR 50.72(b)(3)(xiii), any event that results in a major loss of emergency assessment capability. The NRC Resident Inspector has been notified. The licensee will notify the New York Public Service Commission. |
ENS 49934 | 20 March 2014 08:50:00 | A non-licensed contract employee supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector. |
ENS 49901 | 12 March 2014 10:34:00 | At 0841 (CDT) on March 12, 2014, the Unit 1 Pyro Panel (fire/smoke detection panel) was removed from service for required maintenance. The pyro panel was declared non-functional when it was removed from service. Compensatory measures have been established for all affected areas except the unit 1 Containment Building. Since a fire in the Containment Building is an entry condition for the site's Emergency Plan, this is considered a loss of emergency assessment capability and is being reported per 10CFR50.72(b)(3)(xiii). Containment temperature is being monitored while the pyro panel is out of service however this is not considered a satisfactory compensatory measure for maintaining effective assessment capability. A courtesy follow up notification will be sent when the pyro panel is returned to service and functional. The NRC Resident Inspector has been informed.
The repairs to the Unit 1 Pyro Panel have been completed and the panel was returned to service on 3/12/14 at 1010 CDT. The NRC Resident Inspector has been informed. Notified R2DO (O'Donohue). |
ENS 49876 | 5 March 2014 16:08:00 | This report is being made to provide information to the NRC regarding Monticello Dry Shielded Canister (DSC)-16. On February 17, 2014, dye penetrant examinations were performed on the outer top cover plate (OTCP) to shell weld on dry shielded canister (DSC)-16. This was a re-examination of a linear indication identified on January 24, 2014. The results of the re-examination identified a 1.6 inch linear indication that remained after surface conditioning. This indication had not been previously detected by nonconforming nondestructive examination previously reported by TriVis Inc. Xcel Energy is evaluating the condition and will remedy prior to moving the cask to the ISFSI (Independent Spent Fuel Storage Installation) pad. The associated DSCs loaded during the current campaign successfully passed their helium leak tests. Helium leak checks are performed to demonstrate confinement and boundary integrity. Thus, public health and safety is not affected. Since the licensee communication plan also notified other government agencies, this report is being made pursuant to 10CFR50.72(b)(2)(xi) and 72.75(b)(2). The above referenced dry shielded canister is currently located on the refuel floor in the reactor building. The licensee notified the NRC Resident Inspector. |
ENS 49875 | 5 March 2014 14:10:00 | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined that a similar condition is applicable to the Waterford 3 Nuclear Station resulting in a potentially unanalyzed condition with respect to 10 CFR 50 Appendix R requirements. The original plant wiring design and associated analysis for an ammeter measuring current from the train AB Class 1E battery to its associated power distribution panel does not include overcurrent protection features to limit the fault current and is routed through multiple fire areas. The ammeter is located on the train AB power distribution panel in the train AB switchgear room. In the postulated event, a fire could cause one of the ammeter wires to short to ground. Simultaneously, it is postulated that the fire could cause another DC wire from the opposite polarity on the same battery to also short to ground. This could cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (i.e., heating) in the ammeter wiring to the point of causing a secondary fire in the raceway system. The secondary fire could adversely affect safe shutdown equipment and potentially cause the loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). There is no effect on plant operation. Fire watches have been implemented for affected areas of the plant as an interim compensatory measure. The licensee notified the NRC Resident Inspector.
Subsequent engineering evaluation has determined that the circuit for an ammeter measuring current from its Class 1E battery to its associated power distribution panel is not routed through multiple fire areas. Therefore, the IER 13-54 related condition is not, and was not, an unanalyzed condition at Waterford 3 that significantly degraded plant safety, and thus not required to be reported under 10 CFR 50.72(b)(3)(ii)(B). The licensee notified the NRC Resident Inspector. Notified R4DO (Gaddy). |
ENS 49850 | 22 February 2014 19:32:00 | (The New Jersey Department of Environmental Protection) and Ocean County Health (Department were) notified of a spill of approximately 1 cup of hydraulic fluid (oil) that leaked from a vehicle and entered a storm drain at approximately 1615 (EST). The licensee notified the NRC Resident Inspector.
Upon subsequent review of the Material Safety Data Sheet associated with the fluid that leaked from the vehicle, it was determined that the fluid was not a hazardous substance as defined by the State of New Jersey (Appendix A of N.J.A.C. 7:1E). Additionally, it was determined that the fluid did not reach surface water. Based on this information, the New Jersey Department of Environmental Protection (NJDEP) was notified that the fluid that leaked from the vehicle did not reach surface water and was not a hazardous substance. The NJDEP notification was updated for the record. For the above reasons, the 4 hour non-emergency ENS notification made under EN # 49850 is hereby retracted The NRC Resident Inspector has been informed. Notified R1DO (DeFrancisco). |
ENS 49848 | 21 February 2014 15:45:00 | A report was made to the facility Corrective Action Program (PIRCS) at 0839 (EST) on 2/21/2014 regarding difficulty hearing plant announcements and alarms in the recently renovated Building 110B restroom. Testing of the Public Address (PA) system confirmed that the PA system was difficult to hear. The speakers associated with the PA system are also used for annunciating the site criticality accident alarm evacuation warning. Subsequent testing of the criticality evacuation alarm indicated the alarm was difficult to hear as well. Safety management personnel were notified of the problem and the restroom was locked and posted with signs indicating the area was not to be occupied, pending resolution of the audibility issue. The licensee notified the NRC Resident Inspector. |
ENS 49877 | 5 March 2014 17:05:00 | The following information was received by email: On February 21, 2014 a physician in the department of Interventional Radiology (IR) at Renown Regional Medical Center in Reno, NV prescribed a patient with 26.73 mCi of Y-90 to the liver. This isotope is listed on the RAM license 16-12-0016-01 as item K, sealed sources (Sirtex Medical Limited SIR-Spheres microspheres). This case was approached in the same manner as the previous 20+ cases. The physician felt that the entire dose was appropriately delivered, therefore he went to air and flushed the catheter. The case was ended without incident. After the dose calculations were performed, it was found out that only 54.2% of the dose was delivered. The technicians investigated the delivery system and found that the majority of the undelivered isotope was in/around the 3-way stop system. The company representative (SIRTex) was (at the licensee's site) and after lengthy discussion with the physician, it is felt that the stop might have been defective. The patient and the referring physician were notified. Nevada Event: NV140006 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. |
ENS 49810 | 9 February 2014 08:10:00 | On 02/09/2014, at 0100 (EST), a major portion of Peach Bottom Atomic Power Station's (PBAPS) offsite communication capability was discovered to be non-functional due to an offsite communications equipment failure outside the control of PBAPS. Peach Bottom continues to investigate and resolve the cause of this loss of offsite communications. Peach Bottom Main Control Room ENS and offsite communication lines have been restored and are currently available. The Technical Support Center is limited to satellite phones for offsite communications. On-site communications were not affected. This report is being submitted pursuant to 10CFR 50.72(b)(3)(xiii) as a result of a major loss of offsite communications capability. The NRC (Resident Inspector) has been informed of this notification. |
ENS 49799 | 6 February 2014 08:21:00 | At approximately 0730 CST on February 6, 2014, 1R-22 Shield Building Vent Gas Radiation Monitor will be removed from service for planned maintenance. This monitor has no compensatory measure that will allow timely classification of two Emergency Action Levels (EALs), NUE (Notification of Unusual Event) and Alert classifications when out of service. It is also used for offsite dose projection calculations. This results in a Loss of Emergency Assessment Capability while 1R-22 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii). Unit 1 Shield Building Ventilation Stack is also monitored by high range monitor, 1R-50, which is used for the same purpose in Site Area or General Emergency classifications. 1R-50 is being monitored and is indicating normal values. There are no radioactive leaks that will impact the Shield Building as evidenced by normal readings on 1R-22 prior to removing it from service. This planned maintenance will not result in the unplanned release of radioactivity to the environment and will not adversely affect the safe operation of the plant or health and safety of the public. The licensee has notified the NRC Resident Inspector. |
ENS 49784 | 2 February 2014 11:01:00 | On 2-2-2014 at 0859 (EST), with Unit 2 operating at 100% power, a manual reactor trip was initiated by the control room staff following a trip of the 'A' main feedwater pump and automatic start of the 'C' feedwater pump due to crew concerns that both motors of the 'C' feedwater pump had not actuated. When the 'C' feedwater pump auto started, the running indicator light for one of the 'C' feedwater pump motors failed to illuminate. Both motors of the 'C' feedwater pump had started as designed. Following the reactor trip, all control rods fully inserted into the core and Unit 2 was stabilized in Mode 3 at normal reactor coolant system temperature and pressure. Decay heat is being removed using the normal condenser steam dump system. Unit 2 is in a normal shutdown electrical alignment with power being supplied from the Reserve Station Service Transformers. This event is reportable per 10CFR50.72(b)(2)(iv)(B) for actuation of the reactor protection system. Following the reactor trip, the auxiliary feedwater pumps automatically started as designed and provided makeup flow to the steam generators. The steam generator levels were returned to normal operating level and the auxiliary feedwater pumps were returned to the normal standby automatic alignment. This event is reportable per 10CFR50.72(b)(3)(iv)(A) for actuation of an ESF system. Unit 1 is operating at 100% power and was not affected by the event. The licensee informed the NRC Resident Inspector and will inform the Louisa County Administrator. |
ENS 49780 | 31 January 2014 11:44:00 | This 4-hour notification is being made to report that Salem Unit 2 has performed an unplanned manual reactor trip. The trip was initiated due to reactor coolant temperature approaching the minimum temperature for criticality, 543 degree F, due to boration to achieve shutdown margin requirements following identification of a misaligned control rod. All control rods inserted on the reactor trip. All ECCS (Emergency Core Cooling System) and ESF (Engineered Safety Features) systems functioned as expected with no equipment actuated. The 21 safety injection pump was out of service for scheduled maintenance during the event as was the 2R41 plant vent radiation monitor. Salem Unit 2 is currently in Mode 3. Reactor coolant system pressure is at 2235 psig and temperature is 547 degrees F with decay heat removal via the main steam dump and auxiliary feedwater systems. Unit 2 has no active shutdown technical specification action statements in effect. There was no major secondary equipment tagged for maintenance prior to the event. Prior to the event, the licensee was conducting their monthly control rod surveillance. No primary or secondary relief valves lifted during the transient. The electrical grid is stable and the plant is in its normal shutdown electrical lineup. There was no effect on Unit 1. The licensee notified the NRC Resident Inspector, the State of New Jersey, the State of Delaware and the Lower Alloways Creek Township. |
ENS 49779 | 31 January 2014 04:09:00 | Operators commenced a Technical Specification required shutdown due a 480 volt supply breaker trip resulting in degradation of the Pressurizer Proportional Heaters Group 1. The licensee plans on making a containment entry to investigate the cause of the breaker trip. The B-Emergency Diesel-Generator is currently inoperable for planned maintenance. Unit 2 is currently ramping down and proceeding to a shutdown. The plant is in a normal electrical configuration. There is no indication of any primary to secondary leakage. Unit 3 is unaffected. The licensee notified the NRC Resident Inspector. |
ENS 49773 | 29 January 2014 11:06:00 | On January 29, 2014, during planned inspections of control rod drive mechanism (CRDM) upper housings, it was determined that the inspection results for some housings did not meet the applicable acceptance criteria. That is, evaluations of the housing indications are being performed under ASME Code, Section XI, IWB-3600, 'Analytical Evaluation of Flaws,' and indications were identified that exceed acceptance criteria specified in the Code. None of the indications were through-wall and there was no evidence of leakage. The housing indications, varying in depth and length characteristics, were identified in 17 of the 45 CRDM housings inspected. All 45 CRDM housings were inspected, which constituted 100% extent of condition inspection. The plant was in cold shutdown at 0% power for a planned refueling outage at the time of discovery. Replacement or repair actions are in progress and will be completed prior to plant startup from the outage. This condition has no impact to the health and safety of the public. This report is being made in accordance with 10CFR 50.72(b)(3)(ii)(A), since indications were found that did not meet acceptance criteria referenced in ASME Code, Section XI. The licensee notified the NRC Senior Resident Inspector. |
ENS 49753 | 21 January 2014 21:14:00 | At 1340 (EST) on 1/21/2014, Susquehanna Control Room Operators received an alarm due to the lockup of a monitoring panel utilized for Vent Stack Radiation Monitors (known as SPINGs - System Particulate, Iodine, and Noble Gas). There are two monitor panels, one in the control room and one in the Technical Support Center (TSC). Both panels are fully redundant. At the time, the control room panel was out of service for maintenance. This event rendered all five SPINGs inoperable. Additionally, control room alarm capability was not available. The associated Technical Requirement (TROs) were entered for this inoperability. In order to clear the lockup, the TSC terminal needed to be powered down and rebooted. The TSC monitoring terminal was restored to operable status at 1352 and the associated TROs were cleared. The SPING function and control room alarm functions were inoperable for a total of 12 minutes. In accordance with the guidance of NUREG 1022, Revision 3, Item 3.2.13, this event is being reported as an 8 hour ENS notification due to a loss of emergency assessment capability. The loss of the Stack Monitoring System was not part of a pre-planned evolution. This equipment is relied upon by control room operators and emergency response personnel to implement procedures addressing classification, assessment, and protective actions associated with the emergency plan. The licensee notified the NRC Resident Inspector. |
ENS 49755 | 22 January 2014 15:58:00 | The following information was received from the Commonwealth of Virginia by email: Event description: On January 20th the Radioactive Materials Program (RMP) received a call from Simms Metal Recycling in Petersburg, that a load of scrap metal being received set off the radiation monitoring detector. A DOT exemption form was completed and the scrap load returned C&C Cullet, Inc. in Ashland, where it originated from. They dumped the scrap load and found the item using a survey meter, which indicated 87 microrem/hr. Pictures were sent to the RMP and upon review concluded that the device was a liquid scintillation analyzer. The RMP contacted the manufacturer, Perkin Elmer, and began a conversation regarding the analyzer. The analyzer is secured at C&C Cullet, Inc. as an investigation is ongoing to determine the serial number and owner of the analyzer. It is believed to contain an 18 mCi source of either Ba-133 or Ni-63. There are no health or safety impacts as the source is secured in the analyzer. Virginia Event: VA-14-01
The following was received from the Commonwealth of Virginia via email: On Tuesday January 28th, the source was removed by the manufacturer for disposal. The company will try to ascertain a serial number from the source to determine the General Licensee whom the device was provided to. Notified R1DO (Burritt) and FSME Event Resource via email.
The following was received from the Commonwealth of Virginia via email: On Tuesday January 28th, the source was removed by the manufacturer for disposal. The source serial number (432228) was ascertained by the manufacturer which allowed for tracking to Virginia State University (VSU) as the recipient of the device. The source activity was incorrectly stated at 18 mCi in the initial report, the actual activity is 18 microcuries (uCi) of Ba-133. An investigation was performed by VSU in regards to when and how the device was disposed of. VSU stated the device was given to the Department of General Services (DGS) as surplus equipment in 2012 and was then subsequently sold as scrap metal. The RMP will contact DGS and discuss the proper disposal methods of radioactive material. Notified R1DO (Welling) and FSME Event Resource via email. 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 |
ENS 49751 | 21 January 2014 12:37:00 | The following information was received via facsimile: Subject: Non-Conformance on (CCI Thermal Technologies) Purchase Order 244232-1-MRA-MEEW-00001 Element We notify you that we have initiated a notification process today to: Bechtel Project Quality Manager EQB1A/Watts Bar Nuclear Plant Hwy 68 Spring City, TN 37381 Regarding finding on January 16, 2014 related to non-conforming items shipped under purchase order 244232-1-MRA-MEEW-00001 Element. Manufactured by: CCI Thermal Technologies Inc. 2721 Plymouth Dr. Oakville, (Ontario) (Canada), L6H5R5 Non-conforming items are under investigations with further pending review of action with the issuance party on the purchase order indicated. The nonconformance is related to a borated water storage tank electrical heater that was shipped in 2011 for use on Watts Bar Unit 2. |
ENS 49681 | 31 December 2013 10:35:00 | The licensee reported a shutter handle was bent during fracking field operations that are located in Marion County, West Virginia. The shutter is stuck closed in a safe operations position in a fixed location and presents no danger to personnel. The malfunctioning device is a Berthold Model LB8010; Shield serial number (#10102); Source serial number (0561/10); Source of 10 milliCuries of Cesium-137. |
ENS 49683 | 31 December 2013 10:37:00 | |
ENS 49666 | 20 December 2013 11:23:00 | While in Mode 3 in preparing the Unit 2 secondary plant for startup, conditions occurred where it became necessary to break vacuum on the main condenser. Procedures directed closing of the main steam isolation valves. Instead of shutting each main steam isolation valve individually, a manual main steam isolation actuation was initiated through the solid state protection system (SSPS) to close the valves. This actuation of SSPS was a valid signal. The actuation was not part of a preplanned sequence. This notification is supported by the guidance of NUREG-1022, Revision 3, 'Event Reporting Guidelines 10CFR50.72 and 50.73.' In part, the guidance states: 'The Commission is interested in both events in which a system was needed to mitigate the consequences of an event (whether or not the equipment performed properly) and events in which a system actuated unnecessarily.' The manual actuation was not initiated to mitigate the consequences of an actual event. However, the method of closing the main steam valves for this condition did not specifically require that the valves should be closed by initiating a main steam isolation signal and therefore, the safety system was unnecessarily actuated. Therefore, this notification is being made pursuant to 10 CFR 50.72(b)(3)(iv)(A) as an event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B) that was not part of preplanned sequence during testing or reactor operation. The system listed in paragraph (b)(3)(iv)(B) is (2) main steam isolation valves. The licensee notified the NRC Resident Inspector. |
ENS 49648 | 18 December 2013 03:41:00 | Pursuant to 10CFR50.72(b)(3)(iv)(A), notifications are being performed for a valid actuation of the reactor protection system resulting in a full scram. The actuation was a result of the reactor mode switch being placed from the refuel position to the shutdown position without the scram bypass jumpers installed. The reactor was subcritical with all rods inserted at the time of the actuation. All systems functioned as designed. The licensee notified the NRC Resident Inspector. |
ENS 49624 | 11 December 2013 15:51:00 | Two Trijicon gun scopes, source ID# 6-53953 and 110-59164, both non-exempt sources, were identified as being missing from a Security rifle scopes inventory. Each scope had a (tritium) activity quantity of 57.766 milliCuries (mCi) and 60.528 milliCuries (mCi), respectively. The radioactive material is an integral part of the scope from manufacture. Scope source ID# 110-59164 was recovered, and scope source ID# 6-53953 remains unrecoverable. The activity quantity for the these gun scope exceeds the 10 times the 10 CFR 20, Appendix C quantity, but is (less than) 1,000 times the regulatory criteria of Appendix C and therefore is being reported in accordance with 10 CFR 20.2201(a)(ii). The licensee informed the State of Connecticut, and the NRC Resident Inspector. 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 |
ENS 49621 | 10 December 2013 16:43:00 | The following forty eight (48) lost licensee tritium signs were submitted by the State of Nebraska via email: 1. ShopKo Store #47 located in Lincoln, NE; (1) SRB Technologies; Model Safety-10; S/N ST18770; H-3; 21 Ci. (10) Safety Light; Model #2040; S/N A44267, A44269, A44272 thru A44274, A44276, A42278 thru A44281; H-3; 11.5 Ci each. Nebraska Event - NE130006. 2. ShopKo Store #48 located in Norfolk, NE; (6) Safety Light Model #2040; S/N Unknown; H-3; 11.5 Ci. Nebraska Event - NE130007. 3. ShopKo Store #56 located in Omaha, NE; (29) SRB Technologies; Model Safety-10; S/N 24303 thru 24318, 24327 thru 24333, 24398, 24299, 24400, 24401, 24399H-3; 21 Ci each. (2) Safety Light Model #2040; S/N AON845, AON847; H-3; 12 Ci each. Nebraska Event - NE130008.
The following thirteen (13) lost licensee tritium signs were submitted by the State of Nebraska via email: 1. ShopKo Store #53 located in North Platte, NE; (13) Safety Light Model #2040; S/N Unknown; H-3; 11.5 Ci. Nebraska Event - NE130009. Notified R4DO (Walker), FSME Events Resource and ILTAB via email. 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 |
ENS 49605 | 3 December 2013 20:04:00 | At approximately 1330 (EST) on Tuesday, December 03, 2013, while performing a table top drill, Pilgrim Nuclear Power Station (PNPS) discovered that EP (Emergency Preparedness) bridge conferencing lines were unavailable. The conference lines affected included the mitigation line, plant data phone, radiation data phone, emergency conferencing line, and the back up conference bridge line. Reviews to determine the cause of the event and efforts to restore the system are ongoing. The licensee has determined the Emergency Plan to be functional based on other communication methods that are available between onsite and offsite facilities. These include direct telephone lines, portable handheld radios, satellite phones and cell phones. Immediate actions to establish compensatory conferencing lines have been completed. On-going actions are in-progress to ensure procedure instruction is provided at each facility to enable use of the compensatory conference lines. At the time of this report, the plant is currently operating at 82% power due to a planned power maneuver unrelated to the reported communication event. The licensee has notified the NRC Senior Resident Inspector (and will notify the Commonwealth of Massachusetts). This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the loss of emergency response communication capability.
The primary communications method has been restored. The backup communications method should be restored tomorrow. The licensee will notify the NRC Resident Inspector. Notified the R1DO (Cook). |
ENS 49589 | 27 November 2013 18:14:00 | At approximately 1446 hours (CST) on November 27, 2013 a badged employee was discovered non-responsive inside a vehicle within the Owner Controlled Area. (The vehicle was located outside the protected area and in the parking lot). On-site and off-site emergency personnel responded to the medical issue. The individual was pronounced dead at 1604 hours. No radioactive material or contamination was involved. The death was not work-related nor the result of an occupational event. This event is being reported under 10CFR50.72(b)(2)(xi). The NRC Resident lnspector has been notified. |
ENS 49620 | 10 December 2013 15:10:00 | The following information was received by facsimile: This serves as notification to the U.S. Nuclear Regulatory Commission (NRC) of a possible deviation or defect in nuclear qualified High Efficiency Particulate Air (HEPA) Filters manufactured by Flanders Filters. Flanders Nuclear HEPA Model T-007-W-43-05-NU-51-23-GG-FU5 failed requalification testing at Edgewood Chemical Biological Center (ECBC). This failure was related to the 'Resistance to Pressure' test required per ASME AG-1, Section FC5140. On November 19, 2013, Flanders received a report from ECBC indicating failure of this filter model to re-qualify per the requirements of ASME AG-1, Section FC. Flanders is investigating the failure, but at this time do not believe Flanders has the ability to determine if this could create a substantial safety hazard because we do not have specific knowledge regarding the actual application the filters are being used in at each site. Customers supplied this model Filter, or filters with the same media type and pack style, since the last qualification of the media type and pack style on July 18, 2012 are being notified. In accordance with 10CFR Part 21.21(d)(4), Flanders will provide written notification to the NRC within 30 Days with additional information including corrective actions where applicable. The customers include the following: SCANA; Nuclear Fuel Service; Premier Technology Incorporated of Blackfoot, Idaho; and Energy Northwest. |
ENS 49723 | 14 January 2014 18:33:00 | On 10/16/2013 at 14:20 (CDT), the Technical Support Center (TSC) Emergency Makeup Filter Flow Control Damper (OVV145Y) was discovered degraded and non functional during planned maintenance activities. The degraded damper adversely impacted the TSC ventilation system function. On 10/16/2013 at approximately 15:00 (CDT), administrative actions were developed and briefed to isolate instrument air to the degraded nonfunctional damper OVV 145Y providing the capability to restore TSC Emergency Makeup Filter functionality. Isolating instrument air to the degraded damper places the damper in the failed open position, thereby restoring TSC Emergency Makeup Filter functionality. On 10/16/2013 at 20:25 (CDT), instrument air was isolated to the degraded damper OVV 145Y under administrative controls restoring TSC Emergency Makeup Filter functionality. This action was taken after determining no adverse affects on system operation. On 10/30/2013, degraded damper OVV145Y was repaired, instrument air restored and post maintenance testing completed, thereby restoring the degraded TSC emergency Makeup Filter Unit flow control damper functionality. This event is reportable per 10CFR50.72(b)(3)(xiii) for a major loss of emergency assessment capability because the emergent degraded TSC flow control damper OVV145Y adversely impacted the function of TSC Emergency Makeup Filter and was not restored within the TSC activation time (60 minutes). This event was originally determined not reportable because the capability to restore within the TSC activation time (60 minutes) existed from the time of discovery. Upon further review it was determined to be reportable because the degraded damper 0VV145Y was not restored to service within the TSC activation time (60 minutes). The licensee has notified the NRC Resident Inspector. |
ENS 49390 | 27 September 2013 17:42:00 | On 9/27/13, Monticello Nuclear Generating Plant personnel identified that when the Service Water Radiation Monitor or the Discharge Canal Radiation Monitors are removed from service for planned maintenance activities, a loss of emergency assessment capability may occur. The adequacy of compensatory measures for when these radiation monitors are removed from service is under evaluation. The Service Water Radiation Monitor is taken out of service weekly for flushing. This last occurred at 0221 (CDT) on 9/26/13. The Discharge Canal Radiation Monitors are taken out of service approximately once a year for cleaning. The last time these monitors were removed from service was at 1250 (CDT) on 9/18/13. In accordance with 10CFR50.72(b)(3)(3)(xiii) these past occurrences are being reported as a loss of emergency assessment capability. The NRC Resident Inspector has been notified. |
ENS 49595 | 2 December 2013 13:52:00 | This is a 10 CFR 50.55(e) initial notification for a significant breakdown in the Quality Assurance (QA) Program of Chicago Bridge & Iron (CB&I) Lake Charles facility, a sub-supplier of CB&I. CB&I Lake Charles supplies safety-related structural sub-modules for the Vogtle 3 & 4 construction project. In September 2013, CB&I Lake Charles issued a root cause analysis report for deviations associated with sub-modules being supplied to domestic AP1000 construction projects. An evaluation of the root cause analysis results concluded that a significant QA program breakdown had occurred that could have produced a defect. No defect has been identified. This initial notification is being made in accordance with 10 CFR 50.55(e)(4)(iii) and 10 CFR 50.55(e)(5)(i). The licensee will notify the NRC Resident Inspector. Reference similar Summer Event (EN#49582). |
ENS 49314 | 29 August 2013 18:35:00 | On August 28, 2013, Monticello Nuclear Generating Plant (MNGP) was notified of the NRC's final significance determination for a finding involving the failure to maintain a procedure addressing all of the effects of an external flooding scenario on the plant. Specifically, MNGP failed to maintain flood Procedure A.6, 'Acts of Nature,' such that it could support the timely implementation of flood protection activities within the 12 day timeframe credited in the design basis as stated in the updated safety analysis report. The finding is not a current safety concern. On February 15, 2013, actions were completed to reduce the flood mitigation plan timeline to less than 12 days by developing an alternate plan for flood protection features, pre-staging equipment and materials, improving the quality of the A.6 procedure, and preplanning work orders necessary to carry out Procedure A.6 actions. The NRC Resident Inspector has been notified. |
ENS 49307 | 26 August 2013 01:05:00 | Reactor Building (Secondary Containment) pressure increased to above the Technical Specification Surveillance requirement of 0,25 inches vacuum water gauge. This event is reportable as an event that could have prevented fulfillment of a safety function needed to control the release of radiation and mitigate the consequences of an accident. The Reactor Building differential pressure controller was placed in manual operation and Secondary Containment pressure was restored to normal (greater than 0.25 inches vacuum water gauge) returning Secondary Containment to operable status. Secondary Containment pressure was outside the allowable Technical Specification requirement for 4 minutes. There were no radiological releases associated with the event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector. |
ENS 49280 | 14 August 2013 12:41:00 | This is a non-emergency notification. During a routine plant inspection, plant personnel discovered a leak from a pipe on plant property on Tuesday, August 13, (2013). An isolation valve, 3WN-38, leaked causing approximately 9,900 gallons of water containing low levels of tritium (5,780 picocuries per liter) to leak onto the surrounding soil. The valve was secured and the leak stopped. Based upon the leak location and low tritium levels, there is no health or safety risk to the public or to employees on the site. There is no impact to public health and safety due to this condition. Due to NPDES requirements at 0840 EDT on August 14, a report of this condition was made to the state of (North Carolina). The State has evaluated the event and classified the event as a spill-non-sewage (no environmental impact). The NRC Resident Inspector has been notified.
Additional voluntary notifications have been made to state agencies in accordance with the Groundwater Protection Initiative following the guidance of NEI 07-07. The licensee has notified the NRC Resident Inspector. Notified R2DO (Musser). |
ENS 49247 | 7 August 2013 09:05:00 | On August 7, 2013, at 0345 (CDT), Kewaunee Power Station was notified that a loss of alert sirens impacting greater than 50% of the Emergency Planning Zone population are non-functional due to a loss of power. The sirens are used to provide a public prompt notification. The sirens lost are those that Point Beach Nuclear Power Plant takes primary responsibility. Power was lost to them during a severe thunderstorm and the expected out of service time is unknown at this time. This condition is being reported in accordance with 10CFR50.72(b)(3)(xiii) as an event that results in a major loss of emergency assessment capability. The NRC Resident Inspector has been notified. See related EN #49246 * * * UPDATE FROM SCOTT CIESLEWICZ TO PETE SNYDER AT 1652 EDT ON 8/7/13 * * * Most emergency sirens have been returned to service such that the lost coverage is now only 24%. The licensee notified the NRC Resident Inspector. Notified R3DO (Lara). |
ENS 49232 | 3 August 2013 06:50:00 | (On August 3, 2013, at 0520 CDT), Farley Unit 1 declared an ALERT emergency based on (EAL) HA3, 'Release of Toxic, Asphyxiant, or Flammable Gas'. This was due to a tagout of the CO2 Fire Protection System, which resulted in an unexpected discharge of CO2 on the Unit 1 Radiation Side of the Auxiliary Building. The 100 foot elevation level on Unit 1 had an oxygen concentration below the required amount. CO2 has been isolated at the source tank in order to terminated the leak. No personnel have been affected. There is no radiation release in progress. ERDS has been activated. The licensee is assembling a team to enter the affected area and verify oxygen levels are within allowable limits in order to terminate the ALERT emergency. The licensee has notified the NRC Resident Inspector and the State and local organizations. Notified DHS SWO, DOE, FEMA, HHS, DHS NICC, USDA, EPA, FDA and NuclearSSA via email.
Farley Unit 1 terminated from the ALERT at 1110 CDT on 8/3/13. Oxygen levels are acceptable in all areas of the plant. The CO2 tank has been isolated and is stable at 68%. Plant conditions are stable and improving and all necessary compensatory measures are in place. The licensee has notified the NRC Resident Inspector and applicable state and local authorities. Notified R2DO (King), NRR EO (Hiland), IRD (Kozal), and PAO (Couret), DHS SWO, DOE, FEMA, HHS, DHS NICC, USDA, EPA, FDA and NuclearSSA via email. |
ENS 49173 | 3 July 2013 12:51:00 | At 0156 PDT on July 3, 2013, Diablo Canyon Power Plant experienced a loss of power to ancillary facilities on site. This affected power supplies to data systems used by the Emergency Response Facilities (EOF). All plant status and performance data in the TSC (Technical Support Center) and EOF is available via telephone bridge line and control room phone talker. Additionally, the Emergency Response Data System (ERDS) has been determined to be unavailable. The data system loss occurred following depletion of backup batteries in the affected data center." The licensee notified the NRC Resident Inspector and will notify state and local agencies.
At 2120 PDT on 6/23/13, Diablo Canyon Power Plant (DCPP) experienced a loss of the offsite 230 kV startup power source due to an offsite transmission system relay actuation (see EN# 49143). At 1000 PDT on 7/3/13, PG&E identified that the conditions present on 7/3/13 that resulted in loss of plant status and performance data in the Technical Support Center (TSC) and Emergency Operations Facility (EOF) (see EN# 49173) also existed on 6/23/13. This update reports the 6/23/13 loss of plant status and performance data in the TSC and EOF due to loss of power in the data center that supports those facilities. All plant status and performance data required for emergency assessment and communication remained available via a telephone bridge line and control room phone talker. The data became unavailable in the TSC and EOF following depletion of backup batteries in the affected data center. On 7/3/13 PG&E reconfigured the TSC workstations to ensure they would function and provide data for emergency assessment capability independent of the status of the data center power supplies. The Emergency Response Data System was also unavailable following the loss of 230 kV startup power on 6/23/13 due to loss of power to the data center. This unavailability was as designed and as expected. This delayed event notification has been entered into the DCPP corrective action program. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Gaddy) and ERDS Group via email. |
ENS 49143 | 24 June 2013 04:35:00 | At 2120 PDT, Diablo Canyon Power Plant experienced a loss of the offsite 230 kV startup power source due to an offsite transmission system relay actuation, resulting in valid anticipatory starts of Units 1 and 2 three emergency diesel generators on each unit. All diesels successfully started but were not loaded. All systems operated as designed with no problems observed. The 230 kV startup power source is the only offsite power system designed to be immediately available following an accident. However, the safety related onsite emergency diesel system would have provided power to mitigate the consequences of an accident while the 230 kV was unavailable. Restoration of the 230 kV offsite power system is in progress." The licensee will notify the NRC Resident Inspector.
230 kV was restored to Operable on 06/24/13 at 0200 PDT. Normal 500 kV Offsite Power remained operable and was unaffected by this event. Following further review, Diablo Canyon Power Plant (DCPP) has determined that the loss of the 230 kV system was not a condition that could have prevented fulfillment of the safety function of a system credited to mitigate the consequences of an accident. This power source is not considered to be a safety-related system that is credited to mitigate any accident as described in the DCPP UFSAR, Chapters 6 and 15 accident analyses. PG&E concludes that the emergency diesel generators are the only power source needed to fulfill the accident mitigation function, and they did not become inoperable as a result of this event. The licensee will be notifying the NRC Resident Inspector. Notified R4DO (Werner). |
ENS 49101 | 7 June 2013 20:21:00 | This is a report of an on-site fatality and planned press release per 10 CFR 52(b)(2)(xi). An off-duty Farley employee was supporting a non-work related activity on the plant site (within the Owner Controlled Area) and sustained a suspected weather related fatal injury. The individual was located outside of the protected area but on the plant property. The individual was transported to the plant nurses station and CPR initiated. CPR was terminated at 1550 (CDT) at the direction of ambulance personnel with the individual being unresponsive. The Houston County Sheriff was notified at approximately 1600 (CDT) and responded to the site with the county coroner. The NRC resident was notified on 6/7/2013 at 1650 (CDT). A press release is planned. |
ENS 49100 | 7 June 2013 15:11:00 | On June 7, 2013 at 1351 EDT, Millstone Unit 2 rendered the Main Steam Line Radiation Monitors (RM-4299 A/B/C) out-of-service for planned maintenance. The maintenance activity was completed and the radiation monitors were returned to service at 1355 EDT. The licensee has notified the NRC Resident Inspector as well as the State and local agencies. |
ENS 49099 | 7 June 2013 13:50:00 | At approximately 1203 (EDT) on June 7, 2013, Susquehanna Steam Electric Station Unit One reactor was manually scrammed during reactor startup. Pressure setpoint was being adjusted to the normal operating setpoint, from 750 psig to 934 psig, when all turbine bypass valves unexpectedly opened. Reactor Feed Pumps, Main Turbine, HPCI and RCIC tripped on the high level setpoint Level 8 (+54 inches) due to the resultant reactor level swell. The reactor operator then inserted a manual scram. All control rods inserted. Reactor water level lowered to approximately -10 inches causing Level 3 (+13 inches) isolations. There were no automatic emergency core cooling system initiations. No steam relief valves opened during the event. All safety systems operated as expected. The cause of the Turbine Bypass valve opening is under investigation. Unit 2 was unaffected. The licensee notified the NRC Resident Inspector and will notify the Pennsylvania Emergency Management Agency. |
ENS 49097 | 6 June 2013 16:39:00 | The following information was received by facsimile: On June 4, 2013, the Radiation Safety Officer for the University of South Alabama, Mobile, Alabama, notified the Alabama Office of Radiation Control of a fetal/embryo dose that was in excess of 500 millirem. On April 2, 2013, a 30 (year old) female was referred to the University of South Alabama Medical Center for treatment of symptomatic hyperthyroidism via Sodium Iodide-131. The patient was interviewed regarding pregnancy by the authorized user and a blood test was collected for qualitative serum hCG testing. After a negative pregnancy testing and the patient's statements, the patient was given 15 millicuries of Sodium Iodide-131. The patient was also counseled to avoid pregnancy for six months. On May 30, 2013, eight weeks and two days later, the patient reported to her physician a positive pregnancy diagnosis by her OB/GYN physician. The patient reported that her OB/GYN physician determined that she was in the tenth week of pregnancy. This would place the patient approximately 10 days pregnant at the time of administration. Alabama Incident: 13-27 |
ENS 49075 | 28 May 2013 18:09:00 | On May 28, 2013, at 1507 (EDT), Unit 2 was manually tripped from approximately 98 percent power due to decreasing steam generator levels as a result of a main feedwater system transient. The main feedwater system transient was initiated when the 'C' Main Feedwater Pump Discharge Motor-Operated Valve, 2-FW-MOV-250C, spuriously closed. The cause of the spurious closure of 2-FW-MOV-250C is unknown at this time. The Operations crew entered the reactor trip procedure and stabilized Unit 2 in Mode 3 at normal operating temperature and pressure. All control rods fully inserted into the core following the reactor trip. The reactor protection system actuation is reportable per 10CFR50.72(b)(2)(iv)(B). The Auxiliary Feedwater (AFW) pumps actuated as designed as a result of the reactor trip and provided makeup flow to the steam generators. The automatic start of the AFW system is reportable per 10CFR50.72 (b)(3)(iv)(A) for a valid actuation of an ESF system. The AFW pumps were subsequently secured and returned to automatic. Decay heat is being removed by the condenser steam dump system. Unit 2 is in the normal shutdown electrical line-up. Unit 1 was not affected by this event. The licensee notified the NRC Resident Inspector. |
ENS 49072 | 27 May 2013 18:49:00 | At 1545 hours on 05/27/2013, the North Anna Control Room was notified by local authorities that a potential drowning had taken place at the number 3 Dike in Lake Anna. This incident has been reported to the FERC (Federal Energy Regulatory Commission) Regional Engineer under FERC requirements. Therefore, this is reportable to the NRC under 10CFR50.72(b)(2)(xi). In addition, this incident has received significant media interest. The identity of the victim is not known at this time. The licensee notified the NRC Resident Inspector and the Louisa County Administrator. |
ENS 49036 | 15 May 2013 15:00:00 | This is a non-emergency 30 day notification for missing licensed material. This event is reportable in accordance with 10CFR20.2201(a)(1)(ii). On April 16, 2013, while performing the required semi-annual source leak check and inventory, Radiation Protection personnel could not locate the source label tag or cable for source RMC-0169 on radiation monitor 1EMF44H (This monitor has not been in use since 1995). The monitor's pre-amplifier box had been removed as well. 1EMH44H was inspected and the source was not found within the housing. A search was conducted for this missing source, however it could not be located. During the previous source leak check and inventory on October 4, 2012, the source was in its expected location. Source RMC-0169 is a 200 milligram depleted uranium source. The total original activity of the source was 9.998E-02 microCuries (3.03E-02 micro curies; U-234; 1.98E-03 microCuries; U-235, 6.77E-02 microCuries; U-238). The reportable limit for U-234 (the shortest-lived isotope in the source is U-234 with a half-live of 2.46E+05 years) is 0.01 microCuries per 10CFR20, Appendix C. Based on the activity of U-234 present in the source of 0.03 microCuries, this 30 day phone notification to the NRC is being provided pursuant to 10CFR22.2201(a)(1)(ii). The external dose to an individual from this source is negligible due to the small quantity and the type of material involved. Therefore, this event has no adverse effect upon the health and safety of employees or the public. The required written report pursuant to 10CFR20.2201(b)(1) will be provided to the NRC within 30 days of the telephone notification. The NRC Resident Inspector has been notified. The licensee with notify State and local authorities. 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 |
ENS 48962 | 24 April 2013 10:58:00 | Based on a review of qualifications of the current RSO (Radiation Safety Officer) at the Rhode Island Nuclear Science Center, it was determined that the individual did not meet the licensee's Technical Specification 6.2.2 for education or experience requirements. This review is a follow-up to an NRC inspection report dated March 25, 2013. This non-compliance is reportable in accordance with licensee Technical Specifications 1.25, item 8, which delineates administrative and procedural requirements. Immediate actions was to shut down operations until such time that inadequacies can be remedied. |
ENS 48940 | 17 April 2013 23:19:00 | On April 17, 2013 at 1600 (EDT), while performing a valve inspection/repair of the Unit 2 'A' Reactor Coolant Loop Fill Valve (2- RC-HCV-2556A), the as-found inspection results identified evidence of a suspected flaw causing leakage from the valve body to the threads of a stud housing of the valve. This valve is a 2 (inch) 316 SS (Stainless Steel) cast ASME XI (Class 1) 1500 psi valve body of a globe style design. Due to this design and the installed orientation, the RCS pressure medium fills the upper portion of the valve bonnet where the leak is located during normal plant operations. Therefore, this leakage would be considered pressure boundary leakage. 2-RC-HCV-2556A is currently isolated from the Reactor Vessel and is at atmospheric pressure. This inspection was performed in response to dry discolored boric acid identified during the normal operating pressure boric acid accumulation inspection procedure during the Spring 2013 Unit 2 refueling outage shutdown. An engineering evaluation of the suspected defect will be performed and corrective actions implemented. This event is reportable in accordance to 10CFR50.72(b)(3)(ii)(A) for, 'Any event or condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded'. The licensee notified the NRC Resident Inspector and local County Commissioners.
Event Number 48940 was made on April 17, 2013 in accordance with 10CFR50.72(b)(3)(ii)(A) to document a suspected flaw resulting in RCS pressure boundary leakage on Unit 2 'A' Reactor Coolant Loop Fill Valve (2-RC-HCV-2556A). North Anna Power Station is retracting this notification following completion of a cause analysis and metallurgical examination. The analysis determined that the valve leakage was due to the body-to-bonnet gasket joint. The original valve body was especially susceptible to gasket creep, which lead to a loss of sufficient sealing stress. This resulted in body-to-bonnet leakage, not a through-wall leak. Based on this analysis, the reporting requirements of 10CFR50.72(b)(3)(ii)(A) are not met and this event report is being retracted. The licensee has notified the NRC Resident Inspector. Notified R2DO (Ehrhardt). |
ENS 48938 | 17 April 2013 15:53:00 | The following information was excerpted from an email from the licensee: On April 16, 2013 at 1701 EDT, the research test reactor automatically shutdown from 100% power (1 MW) due to a valid high power condition. The duty Senior Reactor Operator removed a timed irradiation sample from the core that added positive reactivity. Both the digital (non-safety system) and the analog safety system acted on the high power condition and initiated the shutdown. All systems functioned as designed. The short duration power transient reached a peak power of about 1.3 MW. There was no increase in radiation levels, personnel radiation exposure, or release of radiation from the facility. No emergency event entry criteria were met. The plant was placed in a secured condition and an event review investigation was conducted. The event is (potentially) reportable in that the Maximum Power Level observed during the short duration (< 1 second) transient exceeded the steady state power limit for non-pulse mode operation as described in Technical Specification(TS) 3.1.1 Non-pulse mode operation sub-section b. The maximum power level shall be no greater than 1.1 MW (thermal). The reactor was returned to routine service at approximately 1300 EDT on April 17, 2013. |
ENS 48925 | 15 April 2013 12:27:00 | At approximately 1018 CDT on April 15, 2013, the licensee was notified that the Pierce County Sheriff Dispatch reported an inadvertent activation of an emergency siren (P-43), in Pierce County, WI. The cause of the siren activation is unknown. The siren was deactivated at 1022 CDT after sounding for approximately 28 minutes. The siren vendor (NELCOM) has been contacted to repair the siren. The siren remains out of service and is the only siren out of service within the 10 mile Emergency Planning Zone (EPZ). NRC Resident has been informed (by the licensee). |
ENS 48896 | 8 April 2013 15:58:00 | The Salem Generating Unit 1 and 2 control room ENS phone system was down for greater than one hour from 1140EDT to 1240EDT. The licensee notified the NRC Resident Inspector. |
ENS 48885 | 3 April 2013 16:40:00 | During remodeling of a bathroom on the third floor of the Administrative Building which is located inside the Protective Area, workers discovered two very old containers of blackberry brandy after removing the ceiling tiles. This item will be entered into the licensee corrective actions program for follow up. The licensee informed the NRC Resident Inspector. HOO Note: A similar report (EN #48877) was received on 4/2/2013. |
ENS 48889 | 4 April 2013 17:43:00 | The following information was from the State of California via email: On 04/04/13, (the licensee) RSO, called RHB (California Radiologic Health Branch) to report a lost gauge. The gauge is a Model 503 DR CPN moisture gauge (S/N H33064926) containing 50 mCi (max) of Americium 241:Be. (The RSO) stated that he placed the gauge in the back of his truck next to the tailgate, then got distracted by a telephone call and started driving from 4181 Brew Master Dr, Suite 4, Ceres, CA, on Crows Landing, Hwy N 99 and then onto Hwy 88 without securing the gauge. He discovered the gauge missing (at 1600 PDT) on 04/03/13, upon arrival at a jobsite. The Ceres Police Department was notified of the lost gauge (case # 213-001653). Licensee was advised to place an advertisement in a local paper/craigslist offering a reward of $4000.00 for the safe return of the gauge. RHB will be investigating the incident and licensee will be cited for the items of non compliance associated with this incident. California Report: #040413 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 |
ENS 48853 | 27 March 2013 04:21:00 | On March 26, 2013, at 1635 (CDT), with Unit 1 shutdown for refueling, leakage was identified from the 2-inch reactor head vent line during a Reactor Pressure Vessel (RPV) pressure test. The leakage was approximately 20 drops per minute. The RPV pressure test was stopped and the reactor vessel depressurized to facilitate examination of the piping and associated flange connections. At 0245 hours on March 27, 2013, the leak was confirmed to be through-wall originating from a socket weld (i.e., pipe elbow). 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 |
ENS 49055 | 21 May 2013 16:46:00 | On March 24, 2013, at 1009 EDT, while personnel were entering the torus compartment to perform planned maintenance activities via permanently installed plant ladder, the sensing line to transmitter 2E41-N062D was inadvertently bumped. (This) was confirmed to be the cause for an invalid torus high water level alarm and a HPCI (High Pressure Core Injection) pump suction swap. This resulted in the HPCI suction swapping from its normal lineup, condensate storage tank (CST), to the torus as designed. Once actuated the suction swap occurred as designed. The cause was attributed to the close proximity between an individual descending/ascending the fixed ladder and the affected sensing line in conjunction with a loose tubing restraint which made the line more sensitive to being bumped. After confirming that the actuation on high torus level was invalid, HPCI suction was realigned to the CST. The HPCI pump suction was subsequently realigned to the CST and the loose tubing restraints were tightened. The licensee notified the NRC Resident Inspector. |
ENS 48846 | 24 March 2013 09:54:00 | At 0458 (EDT), a lightning strike caused a voltage disturbance on the 115 KV offsite power source and resulted in the automatic start of the 'A' emergency diesel generator on an undervoltage relay actuation. The 'A' emergency diesel generator came up to rated frequency and voltage. The voltage disturbance (duration) was not long enough to cause the normal incoming breaker to the 'A' train 7.2 KV safety related bus (XSW1DA) to open or the 'A' emergency diesel generator output breaker to close. The voltage disturbance caused the 'A' safety related chiller and some ventilation fans to trip. These loads were subsequently restarted with no issue. The 'A' emergency diesel generator was secured and made ready for autostart at 0520 (EDT). The station is in Mode 5 for a mid cycle outage to repair a reactor coolant pump seal. This notification is being made per 10CFR50.72(b)(3)(iv)(A) as a valid actuation of an emergency diesel generator. The unit is in a normal electrical shutdown plant configuration. The licensee notified the NRC Resident Inspector and will notify the State and Local authorities. |
ENS 48842 | 22 March 2013 06:10:00 | During performance of surveillance OSP-CONT-M102, secondary containment had been declared inoperable and Technical Specification 3.6.4.1.A had been entered. Operations personnel performing the surveillance were holding the inner door open and testing the outer door interlock. When the individual presented his badge to the badge reader and attempted to open the outer door, the door opened. Per the surveillance contingency actions, the individual immediately shut the door. Momentarily, both the inner and outer reactor building access doors were open. The surveillance was completed with no further issues. Technical Specification 3.6.4.1.A was exited and secondary containment was declared operable. There were no radiological releases associated with this event. No safety system actuations or isolations occurred. The licensee notified the NRC Resident Inspector.
The licensee is retracting this event based on the following: During subsequent evaluation for reportability, guidance in NUREG-1022 was reviewed. NUREG-1022 provides that removal of a system or part of a system from service as part of a planned evolution for maintenance or surveillance testing when done in accordance with an approved procedure and the plant's TS (unless a condition is discovered that could have prevented the system from performing its function) is generally not reportable. In the case of this event, the secondary containment system was removed from service as part of a planned surveillance test. TS Action 3.6.4.1.A was entered as directed by the governing surveillance procedure in anticipation of the airlock interlock testing which, by the very nature of the test, could result in both doors in a given airlock being open simultaneously. The event described above should not have been reported in consideration of the guidance provided in NUREG-1022 and did not constitute a reportable event as a condition that could have prevented the fulfillment of a safety function under 10CFR50.72(b)(3)(v). The licensee will notify the NRC Resident Inspector. Notified R4DO (Pick). |
ENS 48888 | 4 April 2013 16:08:00 | The following information was received from the Washington State Division of Radiation Protection by email: PermaFix Northwest received a shipment from Perkin Elmer, Inc. that consisted of 32 packages, 4 plastic drums and 28 metal drums, and was shipped as an exclusive use shipment. Upon receipt, the drums were surveyed and 2 plastic drums were found to exceed the 49 CFR 173.443 non-removable contamination limit of 2,200 dpm/cm2 for an exclusive use shipment. The drum survey results were reported as 44,391 dpm/100 cm2 H-3 and 18,080 dpm/100 cm2 C-14; 20,127 dpm/100 cm2 H-3 and 18,508 dpm/100 cm2 C-14, and 13,323 dpm/100 cm2 H-3 and 10,019 dpm/100 cm2 C-14. This most contaminated drum was manifested with only H-3 and C-14, the other 2 drums were manifested with only C-14. Washington Incident Number: WA-13-021 |
ENS 48823 | 15 March 2013 01:48:00 | On 03/14/2013 at 1800 CDT, Browns Ferry Nuclear Plant experienced a partial loss of offsite communication to include the Emergency Notification System to the NRC and normal phone communication to offsite. This loss of communications appears to be the result of an unplanned degraded voltage condition on the 48VDC Telephone Battery supplying power to the communications network. Main Control Room indications and assessment capability were not affected. The Nextel cellular phone system, the satellite phones in the Control Rooms, and the Health Physics Radio Network were available onsite to communicate offsite. The plant radio system was also unaffected. Required corrective maintenance is being evaluated. Communication capability is being periodically restored during maintenance troubleshooting, but has not yet been permanently corrected. Operation of Unit 1, Unit 2, and Unit 3 was not affected by the event. At 2145 CDT, the NRC Operations Center was notified of alternate contact numbers for (Browns Ferry) until the Telephone Battery is fully restored. The NRC Senior Resident Inspector has been notified. This event is reportable under 10CFR50.72(b)(3)(xiii).
A spare battery charger was aligned to the 48 VDC Telephone Battery and offsite communications was restored. The licensee notified the NRC Resident Inspector. Notified the R2DO (O'Donohue). |
ENS 48820 | 13 March 2013 03:11:00 | At 0149 CDT on 3/13/2013, Wolf Creek declared an Unusual Event due to inoperability of both on-site emergency diesel generators. While the A-EDG (Emergency Diesel Generator) was out of service due to planned maintenance, the B-EDG experienced a loss of control power. The cause is currently under investigation. Offsite power is available and providing electrical power to the Unit. The licensee notified state and local agencies and the NRC Resident Inspector. Notified DHS, FEMA, DHS NICC and NuclearSSA (email).
Repairs are complete on the 'B' EDG (Emergency Diesel Generator). The 'B' EDG has been tested and restored to operable status at 0221 (CDT) on 03/14/2013. The 'A' EDG remains out of service for scheduled maintenance. NUE (Notification of Unusual Event) terminated at 0239 CDT on 03/14/2013. The licensee notified state and local agencies and will notify the NRC Resident Inspector. Notified the R4DO (Powers), IRD (Grant) and NRR (Lubinski). Notified DHS, FEMA, DHS NICC and Nuclear SSA (email). |
ENS 48809 | 6 March 2013 12:52:00 | At 0401 CST on 3/6/2013, while in RHR-High (Residual Heat Removal-High) water level the plant experienced a momentary Loss of Shutdown Cooling which resulted in a loss of safety function for Residual Heat Capability. Division 2 RHR shutdown cooling was restored within approximately 90 seconds without issue. No changes were experienced in refuel volume temperature or level during the loss of RHR shutdown cooling. This occurred shortly after a flow adjustment on the system was made utilizing the outboard valve. The inboard valve was reopened and an investigation is in progress. At the time of the valve closure, decay heat removal continued from Reactor Water Cleanup in heat reject mode and fuel pool cooling (with the fuel pool gates removed) is in service. Division 1 RHR (Shutdown Cooling) was available (not Operable) at the time of the loss. It is not currently understood why the injection valve closed. All systems functioned as required except for the spurious closing of MO-2015 (the Div 2 RHR inboard injection valve). The following make-up sources are available: Divisions 1 and 2 RHR, Divisions 1 and 2 Core Spray, CRD (Control Rod Drive), CST (Condensate Storage Tank) via a Core Spray with pressurizing station bypassed. The licensee notified the NRC Resident Inspector.
On March 6, 2013 (Notification No. 48809) NSPM (Northern States Power Monticello) reported in accordance with 10 CFR 50.72 (b)(3)(v)(B), a momentary closure of valve MO-2015 in the operating Residual Heat Removal (RHR) subsystem as an event or condition that at the time of discovery could have prevented the fulfillment of a safety function. Following the event, the RHR SDC (shut down cooling) subsystem was removed from operation for equipment forensics and troubleshooting. Results validated that valve MO-2015 was operable and no issues were identified with the associated electrical circuitry, or the RHR SDC subsystem. The decay heat removal requirements of LCO 3.9.7, RHR - High Water Level, were met and there was not a loss of safety function. Therefore, NSPM retracts the March 6, 2013 notification for this event. The licensee notified the NRC Resident Inspector, state and local authorities, and may make a press release. Notified the R3DO (Passehl). |
ENS 48806 | 4 March 2013 18:32:00 | It has been determined that the mechanical seals used in two Low Pressure Safety Injection Pumps and three Containment Spray Pumps are made of a material that may not maintain the designed integrity of the systems under certain accident conditions. These seals have been installed since original plant construction. This issue was discovered by plant personnel while researching requirements for the replacement parts during scheduled outage activities. The licensee notified the NRC Resident Inspector. |
ENS 48805 | 4 March 2013 11:58:00 | At 0927 CST on March 4, 2012, XCEL Energy Environmental Services made a report to the State of Minnesota regarding 475 fish killed from the discharge canal temperature transient following reactor shutdown on March 2, 2013. Monticello was in the process of performing a planned shutdown in preparation for a refueling outage. The NRC Resident Inspector, Wright County Sheriffs Department and Sherburne County Sheriffs Department have been notified by the licensee. |
ENS 48786 | 26 February 2013 19:20:00 | At approximately 1630 PST on February 26, 2013, Pacific Gas & Electric (PG&E) will be performing repairs on a vital inverter. The clearance will remove power from various inputs to the Safety Parameter Display System (SPDS), Emergency Response Data System (ERDS), and Emergency Response Data and Recall Recorder Subsystem (ERFDS). PG&E expects to have the equipment repaired and returned to service within 12 hours. During this time, a dedicated licensed operator will be available to provide plant data to the NRC's Emergency Operations Center. DCPP is making this 8-hour, non-emergency notification under 10 CFR 50.72(b)(3)(xiii) as any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. The licensee notified the NRC Resident Inspector. |
ENS 48782 | 25 February 2013 17:49:00 | At 1313 (CST) on 02/25/2013, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System from a turbine trip. Preliminary indications show the turbine tripped on low condenser vacuum. Cause of loss of condenser vacuum has been identified as Reactor Feedwater recirculation piping separation. Main Steam Isolation Valves (MSIVs) were manually closed to isolate the leak. None of the Safety Relief Valves (SRVs) automatically cycled during the transient, and one Safety Relief Valve (SRV) was manually operated to maintain Reactor Pressure due to the Main Turbine Bypass Valves unavailability because of loss of condenser vacuum. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC), reactor water level initiation set points were reached. Reactor water level is being controlled by the RCIC system and Reactor Pressure is being controlled with the High Pressure Coolant Injection (HPCI) system. All expected containment isolation and initiation signals (Groups 2, 3, 6, and 8) were received. Upon receipt of these signals all required components actuated, with the exception of one valve in Group 6. Drywell Continuous Air Monitor (CAM) Inboard Return Isolation Valve 3-FSV-90-257 did not have indication following isolation signal and was not able to be verified locally. Indication was subsequently restored following restoration of containment isolation signals, and the Drywell CAM was manually isolated at 1422 (CST) with positive indication of isolation, and isolation valves deactivated at time 1514 (CST) to satisfy TS LCO 3.6.1.3 required actions. This event is reportable within 4 hours per 10CFR50.72(b)(2)( iv)(B), 'Any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation'. It is also reportable within 8 hours per 10CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR 50.73(a)(2)(iv)(A). At 1415 (CST), Suppression Pool Water level exceeded -1 inch due to operation with HPCI in pressure control mode, and required entry into TS LCO 3.6.2.2 condition A to restore level within 2 hours. Efforts are being made to lower suppression pool water level within limits. At 1615 (CST), water level remains above -1 inch requiring entry into TS LCO 3.6.2.2 condition B requiring action to be in MODE 3 in 12 hours and MODE 4 within 36 hours. This event is reportable within 4 hours per 10CFR 50.72(b)(2)(i), 'The initiation of any nuclear plant shutdown required by the plant's Technical Specifications.' The NRC Resident Inspector has been notified. All control rods fully inserted and electrical offsite power is in a normal shutdown configuration. Residual Heat Removal is aligned for suppression pool cooling. There was no impact on either Unit 1 or 2. |
ENS 48778 | 24 February 2013 15:20:00 | On February 24, 2013 at 1205 (EST) with reactor power at 25% and the turbine offline, a manual reactor trip for Sequoyah Unit 2 was initiated due to loss of condenser vacuum indication causing closure of condenser steam dumps, opening of the Steam Generator Atmospheric Relief Valves, and lowering hotwell level resulting in imminent loss of hotwell pumps. The cause of the event was determined to be a faulty test connection on B Condenser vacuum pressure switch. During the event, steam pressure rose to the setpoint for the first Steam Generator code safety valve (1064 psig). (The safety valve opened, then reseated). Following the reactor trip, all safety related equipment operated as designed. Auxiliary feedwater actuated as expected on loss of the operating main feedwater pumps. The reactor trip was uncomplicated. Unit 2 is currently being maintained in Mode 3 at NOP/NOT (Normal Pressure and Temperature), with auxiliary feedwater supplying the steam generators and maintaining level at approximately 33% narrow range. Method of decay heat removal is via atmospheric reliefs to the atmosphere. Current RCS conditions: temperature (is) 547 degrees F and stable. Pressure (is) 2235 psig and stable. (There is) no indication of any primary/secondary leakage. All rods are inserted. Electrical alignment is normal and supplied from offsite power. (There is) no impact to Unit 1. Unit 1 is operating at 100% power / Mode 1. There was no impact on public health and safety. Post-trip investigation is in progress and planned restart is 02/25/2013. (The licensee plans a press release.) The licensee notified the NRC Resident Inspector. |
ENS 48777 | 22 February 2013 20:13:00 | On February 22, 2013 at approximately 1430 (EST), Susquehanna identified a computer program error that affected the Susquehanna Fitness for Duty (FFD) program. Specifically, two Behavioral Observation Program (BOP) inquiries were accepted without proper documentation of the required need to continue unescorted access authorization (UAA) and without verification of an actual observation within the required thirty day timeframe. The computer error resulted in answers for two of the three questions on the Behavior Observation Inquiry form not being recorded when the form was submitted by the supervisor. This resulted in the two security accounts being re-zeroed and allowing UAA for an additional 15 days. The BOP supervisor was contacted and verified that these individuals were intended to continue with UAA. At no time were these individuals removed from the FFD or Behavior Observation Program. In accordance with 10 CFR 26.719(b)(4), this report is being made based on being a potential programmatic failure, degradation, or discovered vulnerability of the FFD program that may permit undetected drug or alcohol use or abuse by individuals within a protected area, or by individuals who are assigned to perform duties that require them to be subject to the FFD program. The (NRC Resident Inspector) and the Branch Chief for the Region I Division of Reactor Safety were notified. |
ENS 48904 | 9 April 2013 17:07:00 | On February 11, 2013, at 0613 hours (CDT), the Reactor Core Isolation Cooling (RCIC) system was manually started during a planned Unit 3 reactor shutdown. A Reactor Feedwater recirculation piping separation resulted in the loss of condenser vacuum and subsequent unavailability of the Main Turbine Bypass Valves. The RCIC system was manually started at 9.2" of condenser vacuum in order to control reactor water level in anticipation of loss of Reactor Feedwater Pumps (RFPs) which occurs at 7" of condenser vacuum. Safety Relief Valves (SRVs) were manually operated to maintain reactor pressure. The reactor water level was controlled in the normal band by RCIC, and Reactor Pressure was controlled with a combination of Reactor Core Isolation Cooling (RCIC) system and SRV manual operation. All systems operated as designed and Reactor water level was maintained in the prescribed band. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. RCIC operation was secured at 1449 (CDT) on 2/11/2013. This event is reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A). During a review of operating logs it was identified that this event met reporting requirements and had not been reported. Therefore, this report does not comply with the 8 hour requirement. This condition has been entered into the corrective action program. Additionally, an LER is required within 60 days per 10CFR50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. |
ENS 48696 | 23 January 2013 22:03:00 | On 1/23/2013 at 1516 (EST), Nine Mile Point 2 (NMP2) had a failure of a Reactor Building General Area temperature trip unit occur resulting in the closure of an isolation valve on the Reactor Core Isolation Cooling (RCIC) system steam supply line. Concurrent with this failure, the High Pressure Core Spray (HPCS) system was inoperable for planned surveillance testing. With both the RCIC and HPCS systems inoperable, NMP2 entered a Technical Specification Required Action to be in Mode 3 within 12 hours. At 1550, the HPCS system was restored to OPERABLE. Based on the concurrent loss of the high pressure reactor makeup capability of these two systems, it was determined that the condition is reportable under section 50.72(b)(3)(v) as the following safety functions were impacted: (A) Shutdown the reactor and maintain it in a safe shutdown condition; and (D) Mitigate the consequences of an accident. NMP2 remains in a stable condition at rated power. The offsite grid is stable with no restrictions or warnings in effect. The licensee notified the NRC Resident Inspector. |
ENS 48691 | 22 January 2013 13:10:00 | The following information was received by email: This First Notice constitutes EARLY notice of events of POSSIBLE safety or public interest significance. The information is as initially received WITHOUT verification or evaluation, and is basically all that is known by the (Arizona Radiation Regulatory) Agency Staff at this time. At approximately 0830 (MST on 1/22/13), the Agency was notified by PCI (Patient Care Infusion), the Licensee, that yesterday (1/21/13) they distributed a pharmaceutical that failed the quality control testing. Several licensees reported to them they were receiving improper data from patients injected with the material. A total of 9 licensees were shipped the radiopharmaceutical and the licensee estimates as many as 13 patients may have received the deficient material. Additionally, a like number were scheduled to receive the material but did not receive the material. The Licensee is preparing the 30 day written report of this event. The Agency continues to investigate this event. Arizona First Notice: 13-003 |
ENS 48784 | 26 February 2013 12:42:00 | A female patient with a previously removed thyroid, was treated for remnant thyroid cancer on January 18, 2013. She received an I-131 therapy dose of 58 mCi. On February 21, 2013, the radiation oncologist was informed by the primary physician that the patient had tested positive on a pregnancy test. The patient received an ultrasound on February 25, 2013 and the conception date was determined to be January 14, 2013. Based on the patient therapy activity received, an embryo dose of 15 Rem was calculated. Physician review of the ultrasound was determined to be normal for the embryo. The radiation oncologist is in contact with the primary physician and the patient.
Based on a telephone notification made by Douglas Heidorn, Radiation Safety Officer (RSO) at 1950 EDT on 3/25/2013, it was agreed that a written justification for a retraction would be submitted. The following is a synopsis of what Dr. Heidorn submitted. The Radiation Oncologist (Prescribing Physician) and RSO originally thought that the fetal dose was 14.6 rem, but upon further investigation, it was determined that the fetal dose was only 52 mrem. The discrepancy in the dose estimate was due to confusion about the exact conception date (fetal age). The Prescribing Physician has been corresponding with a Professor of Pediatrics and Radiology (Expert Physician) regarding this issue. The Expert Physician believes that conception occurred between 1/27/2013 and 1/29/2013. Another obstetrician at SJRMC and the Referring Physician concur with the Expert Physician's date range for conception. The 52 mrem fetal dose rate is based on this conception date. On 3/11/2013, a second ultrasound was performed that confirms the above information. Both ultrasound reports indicate the fetus is growing and has normal characteristics. As a corrective action, the Radiation Department will immediately adopt a policy that requires any female aged 12-60 shall receive a pregnancy test 48 hours prior to an I-131 administration. Notified R4DO (Pick) and FSME Event Resources (via E-mail). |
ENS 48673 | 14 January 2013 22:04:00 | Actuation of RPS (Reactor Protection System) with reactor critical. The Reactor Scram occurred at 1805 (CST) 01/14/13 from 100% CTP (Core Thermal Power). The cause of scram appears to be a Turbine Generator Trip. 05-S-01-EP-2 RPV Control, Reactor Scram ONEP (Off Normal Event Procedure) 05-1-02-I-1, and Turbine and Generator Trips ONEP 05-1-02-1-2 were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF power occurred. No ECCS initiation signals were reached and no ECCS or Diesel Generator initiation occurred. All control rods are fully inserted. MSIVs remained open and SRVs lifted and reseated as designed. Currently, reactor water level is being maintained by the Condensate and Feedwater system in the normal band and reactor pressure is being controlled via Main Turbine Bypass valves to the main condenser. There are no challenges to Primary or Secondary Containment at this time. The licensee notified the NRC Resident Inspector. |
ENS 48674 | 15 January 2013 12:45:00 | The following information was received by email: On Monday, January 14, 2013, the (State of Tennessee) Division of Radiological Health received a report from Methodist University Hospital regarding a stuck HDR source. A patient was to be treated with the high dose rate remote afterloader (Nucletron model 105.999) on January 14th. The radiation source became stuck in the applicator/transfer tube at the beginning of treatment before reaching the patient. The physicists and physician followed the policy and procedure for removal of the source and tubing. The source was placed in a shielded container. A Nucletron engineer was notified by phone and arrived at 1600 CST, but was unable to dislodge the source from the transfer tube. The source and transfer tube will be sent back to Nucletron and replacements have been ordered. A written report is being prepared and will be sent to the Division of Radiological Health. Inspectors from the Memphis field office will follow-up on this incident and will continue to keep NRC informed of the status of our investigation. Tennessee Event: TN-13-013 |
ENS 48675 | 15 January 2013 17:54:00 | An item was processed for shipment (from Norfolk, VA) to the manufacturer for repair and was delivered to a contracted carrier on 9 January 2013. The carrier notified the government representative on Monday, 14 January 2013, reporting that the subject material was involved in a police investigation for a potential theft in Riverside, CA. The government contractor reported the loss of freight to the depot who reported the incident to the DLA (Defense Logistics Agency) Distribution Radiation Safety Officer at 1503 hrs on 15 January 2013. DLA Distribution (Norfolk, VA) will work with all parties to obtain further information related to this incident and update the NRC as information becomes available. The shipped item is a pressure indicator (NSN#6620-01-125-8904) containing a Sr-90 source (500 microCuries). DLA Incident Number: 2013-DLA-001 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 |
ENS 48666 | 11 January 2013 12:27:00 | The following report was received via fax: Fisher Information Notice: FIN 2013-01; 9 January 2013 Subject: Type 546NS Transducers Equipment Affected By This Information Notice: Type 546NS Transducers shipped prior to 19 December 2012. Purpose: The purpose of this Fisher Information Notice (FIN) is to alert users of the Type 546 and 546NS Transducers, shipped prior to 19 December 2012, that Fisher Controls International LLC (Fisher) was made aware of a situation which may affect the performance of the aforementioned equipment. We are informing you of this circumstance in accordance with Sections 21.21(b) and 50.55 (e) of 10CFR21. Applicability: This notice applies only to Fisher Type 546 and 546NS Transducers shipped prior to 19 December 2012 that are not in operation, installed or in service. Discussion: Recently, while a Fisher maintenance engineer was installing a Type 546NS Transducer in a non US customer's plant, the instrument did not perform as expected. Upon investigation of the unit, it was determined that the vent hole to the relay was plugged which resulted in the build up of pressure inside the housing. This increased pressure will cause the output pressure to ramp up to supply pressure. In the event the relay is plugged, users will easily detect ramping up of the output pressure immediately after the sealed unit is put in service, which is why this FIN applies to units that are not yet in service. Action Required: All Type 546 and 546NS Transducer units shipped to customers prior to 19 December 2012 and not already-in-service should be checked for this restriction of the case vent. Units in-service with the cover installed and properly working will not have a plugged case vent. 10CFR21 Implications: Fisher requests that the recipient of this notice review it and take appropriate action in accordance with 10CFR21. If there are any technical questions or concerns, please contact: George Baitinger Manager; Quality Fisher Controls International LLC 205 South Center Street Marshalltown. IA 50158 Fax; (641) 754-2854 Phone: (641) 754-2026 George.Baitinger@Emerson.com |
ENS 48637 | 29 December 2012 05:05:00 | Actuation of RPS (Reactor Protection System) with reactor critical. Reactor Scram occurred at 0018 (CST), 12/29/12, from 100% CTP (Core Thermal Power). The cause of scram appears to be a Generator/Turbine trip. Appropriate off normal event procedures were entered to mitigate the transient with all systems responding as designed. No loss of offsite or ESF (engineered safety feature) power occurred. No ECCS initiation signals were reached and no ECCS or Diesel Generator initiation occurred. MSIVs (Main Steam Isolation Valves) remained open and SRVs (Safety Relief Valves) lifted. Currently, reactor water level is being maintained by the condensate and feedwater system in normal band and reactor pressure is being controlled to limit cooldown. All control rods inserted. The plant is in hot shutdown with decay heat removal to the condenser and the electrical line-up is in a normal configuration. The cause of the turbine/generator trip is under investigation. The licensee notified the NRC Resident Inspector. |
ENS 48617 | 21 December 2012 07:06:00 | Salem Unit 1 has experienced an automatic reactor trip at 0528 (EST) on 12/21/12. The unit tripped due to turbine trip above P-9 (Greater than 49% power). All shutdown and control rods fully inserted on the reactor trip. Prior to the trip, the unit was operating at 100% when the crew received the OHA (Overhead Alarm) for main power transformer over excitation which actuated generator protection, which initiated the turbine trip. The auxiliary feed water (AFW) system auto started on low steam generator levels as expected on a reactor trip. Numbers 11, 12 & 13 AFW pumps all automatically started to provide feed to the steam generators. The (Operating Crew) entered EOP-Trip-1, then transitioned to EOP-Trip-2 and stabilized the plant. The unit is currently in mode 3. The OCC (Outage Control Center) is manned and the cause of the main power transformer over excitation is under investigation at this time. RCS temperature is 547 degrees F, RCS pressure is 2235 psig. The 11-14 RCP's are in service. There is one shut down technical specification action statement in effect. Unit 1 containment APD (Containment Radiation Monitor) is inoperable for DCP (Design Change Package) work. This is a 30 day shutdown LCO that expires on 1/16/2013 at 0830. All ECCS and ESF systems are available. Decay heat removal is being provided by 11 and 12 AFW pumps and the main steam dump system. The 13 AFW pump operation is not required and has been removed from service. The plant is aligned with a normal electrical line-up from offsite power sources. There were no personnel injuries associated with this event This event is reportable per 10CFR50.72(b)(2)(iv)(b) due to the automatic reactor trip. This event is reportable per 10CFR50.72(b)(3)(iv)(a) due the AFW actuation on low steam generator levels. There was no lifting of PORVs or primary to secondary leakage. There was no impact on Unit 2. The licensee notified the NRC Resident Inspector. |
ENS 48616 | 21 December 2012 00:06:00 | During a walkdown on December 20, 2012 at 1600 CST, two degraded Appendix R fire barriers (walls) were identified. These barriers separate the Torus Room (Fire Area IV)/ 'A' RHR Room (Fire Area I) and the Torus Room (Fire Area IV)/ 'B' RHR Room (Fire Area II). The walls separate Appendix R fire safe shutdown divisional equipment. A fire watch was established as a compensatory measure immediately following identification of the issue on December 20, 2012. The barrier affecting the 'B' RHR Room has been repaired on both sides. The barrier affecting the 'A' RHR Room has been repaired on the Torus Room side. The discovery of this non-compliance is being reported as an unanalyzed condition as defined by 10CFR50.72(b)(3)(ii)(B). The fire watch remains in place until verification of the completed repair is performed. The licensee notified the NRC Resident Inspector.
An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was conservatively reported on December 21, 2012 for degraded fire barriers between the Torus Room (Fire Area IV) and 'A' RHR Room (Fire Area I), and the Torus Room (Fire Area IV) and 'B' RHR Room (Fire Area II). Subsequent engineering analysis determined that the degraded fire barriers maintained the required degree of separation for redundant safe shutdown trains and plant safety was not significantly degraded. The 10 CFR 50.72(b)(3)(ii)(B) report is retracted. The licensee will notify the NRC Resident Inspector. Notified R3DO (Kunowski). |
ENS 48611 | 20 December 2012 09:22:00 | This is a non-emergency notification. At 0735 (EST) on December 20, 2012, radiation monitors RM-1CR-3561A, RM-1CR-3561B, RM-1CR-3561C and RM-1CR-3561D, Containment Ventilation Isolation Radiation Monitors, were declared inoperable for preplanned maintenance. These monitors are the only monitors credited in the EALs for monitoring elevated radiation levels inside containment for irradiated fuel. These radiation monitors are necessary for accident assessment and are credited for Emergency Action Level (EAL) classification in the Harris Nuclear Plant Emergency Plan. Inability to classify an EAL due to these monitors being out of service is considered a loss of accident assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2. This condition does not affect the health or safety of the public or the operation of the facility. The NRC Resident Inspector has been notified. |
ENS 48605 | 19 December 2012 08:54:00 | Voluntary ENS Notification Regarding Maintenance Affecting the Ability to Assess an EAL Classification. Kewaunee Power Station (KPS) is preparing to perform routine planned calibrations on Wednesday, December 19, and Thursday, December 20, 2012, on the forebay level indication. This maintenance will impact the forebay level computer and Control Room indication normally used to make the Unusual Event or Alert emergency classifications in the event the lake or forebay threshold is exceeded for greater than 15 minutes. The maintenance is planned to occur from 0800 (CST) on 12/19/12 until 1230 (CST) on 12/20/12. There are four level transmitters each with local indication. Each transmitter will be calibrated, but only one at a time, so that the capability to locally determine forebay level using the other three transmitters will remain. This activity has been evaluated against the requirements of 10 CFR 50.72(b)(3)(xiii) and it has been determined that the condition does not rise to the significance necessitating an event notification since Kewaunee retains the capability to make the necessary emergency classifications. However, this information is being conservatively provided so that NRC may be aware of the condition should an emergency develop. The licensee notified the NRC Resident Inspector. |
ENS 48551 | 2 December 2012 17:28:00 | The raw water pumps (AC-10A/B/C/D) base plate support anchors were discovered by Fort Calhoun Station personnel to have inadequate embedment to support existing analysis. Plant drawing specify a j-bolt type of anchor with a required 16 inch embedment. Actual plant configuration was found to be a j-bolt type anchor with a 9 inch embedment. Plant design analysis requirements are not being met for the existing configuration. Existing analysis requires a minimum embedment of 60 inch for a j-bolt type anchor. There are a total of 4 anchors for each raw water pump, totaling 16 anchors. The as found condition renders all four raw water pumps inoperable. In the current plant Mode 5 (De-fueled), Shutdown Condition, the raw water pumps are considered available per the station's Shutdown Operations Protection Plan. Raw water pumps AC-10B and AC - 10D are in service providing cooling to the Component Cooling Water System. The core is offloaded and the Component Cooling Water System is maintaining Spent Fuel Pool temperature. The licensee notified the NRC Resident Inspector. |
ENS 48550 | 2 December 2012 14:06:00 | The Unit-2 AMSAC (Anticipated Transient Without Scram Mitigation Actuation Circuitry) actuation generated a Unit-2 turbine trip signal. The Unit-2 turbine was already tripped, but the 2A and 2B Auxiliary Feedwater and 2B Nuclear Service Water pump started. AMSAC actuation occurred during calibration of AMSAC actuation pressure switches. The licensee notified the NRC Resident Inspector.
This notification retracts an eight (8) hour notification per 10CFR50.72(b)(3)(iv)(A) which was made December 2, 2012 and documented as NRC Event Notification (EN) # 48550. The December 2, 2012 actuation of the Auxiliary Feedwater System was initially reported under 10CFR 50.72(b)(3)(iv)(A), valid actuation of the Auxiliary Feedwater system as listed in paragraph 10CFR50.72(b)(3)(iv)(B)(6). McGuire Nuclear Station (MNS) has subsequently determined that the actuation signal, which was caused by a test signal during calibration of the AMSAC (Anticipated Transient Without Scram Mitigation Actuation Circuitry) actuation pressure switches, constitutes an invalid actuation as described in Revision 2 of NUREG-1022 (Event Reporting Guidelines 10CFR50.72 and 50.73). Specifically, valid actuations result from signals initiated in response to actual plant conditions or parameters satisfying the requirements for initiation of the safety function of the system. A test signal is not representative of an actual plant condition and was therefore an invalid actuation signal and does not meet the NRC 8-hour reporting criteria under paragraph 10 CFR50.72(b)(3)(iv)(B)(6). In addition to retracting the 10CFR50.72(b)(3)(iv)(B)(6) notification, MNS is notifying the NRC Operations Center that the December 2, 2012 event met the reporting criteria specified in 10CFR50.73(a)(2)(iv)(A). Specifically, an invalid actuation of the auxiliary feed water system occurred while the systems was in service. The event did not involve an actuation of the reactor protection system (RPS) when the reactor was critical; therefore, MNS is opting to notify the NRC Operations Center within 60 days in lieu of a 60 day written licensee Event Report (LER), as allowed by 10CFR 50.73(a)(1) and 10CFR50.73(a)(2)(iv)(A). This notification satisfies 10CFR50.73(a)(2)(iv)(A) reporting requirements in lieu of a written LER. The 2A and 2B trains of the auxiliary feed water were actuated by an invalid signal. Each train's actuation was complete and systems operated as designed. The licensee notified the NRC Resident Inspector. Notified the R2DO (Sykes). |
ENS 48548 | 1 December 2012 16:28:00 | The Technical Support Center (TSC) Ventilation failed to enter 'Incident Mode' as required during scheduled testing. That is, the air cannot be re-directed through the charcoal filter in an event requiring the TSC atmosphere to be filtered due to an on-site release. The TSC ventilation system remains in service providing proper temperature control to maintain the facility habitable during normal operation; however with the 'Incident Mode' being unavailable, it may not remain habitable during all postulated scenarios and is, therefore, considered non-functional. The licensee will notify the NRC Resident Inspector and the New York Public Service Commission. |
ENS 48522 | 20 November 2012 11:02:00 | At 1625 CST on 11-19-12, Electrical Maintenance and Power Operations were in the process of changing a circuit breaker in the C-409 facility which supplies power to the C-409 Criticality Accident Alarm System (CAAS) Uninterruptable Power Supply (UPS). This UPS supplies power to the C-409 CAAS horns. Procedure CP2-CO-ON3031 was being utilized and the CAAS Horn Power UPS was being monitored continuously to maintain the CAAS operable. As soon as the breaker was opened it was identified that the UPS failed to throw over to battery power. At that point the breaker was re-closed and AC power restored to the CAAS horns. The time that AC power was off the CAAS horns was approximately 3-5 seconds. There were no fissile material operations in progress in the C-409 facility. This event is reportable as a 24 hour event in accordance with 10CFR76.120(c)(2)(i). This is an event in which equipment is disabled or fails to function as designed when: a.) the equipment is required by a TSR (Technical Safety Requirements) to prevent releases, prevent exposures to radiation and radioactive materials exceeding specified limits, mitigate the consequences of an accident, or restore this facility to a pre-established safe condition after an accident; b.) the equipment is required by a TSR to be available and operable and either should have been operating or should have operated on demand, and c.) no redundant equipment is available and operable to perform the required safety function. The NRC Senior Resident Inspector has been notified of this event. |
ENS 48518 | 18 November 2012 11:10:00 | On 11/18/2012 at approximately 1130 EST, the Summer Safety Parameter Display System (SPDS) and the Emergency Response Data System (ERDS) will be taken out of service for approximately 12 hours to support a planned maintenance on an inverter that feeds the station's Integrated Plant Computer System (lPCS). During this time frame. ERDS and SPDS will be unavailable. Should the need arise, plant status information will be communicated to the NRC using other available communication systems. SPDS and ERDS are expected to be restored on 11/18/2012 at 1800 EST. This event is reportable under 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified.
On November 18, 2012, at approximately 1730, IPCS was returned to service. Completing this planned maintenance activity restored the VCSNS Safety Parameter Display System and the Emergency Response Data System (ERDS). ERDS was tested this morning, November 19, 2012, by VCSNS and the NRC ERDS Help Desk to verify system operability. The NRC Resident Inspector has been notified. Notified R2DO (Widmann). |
ENS 48496 | 9 November 2012 03:03:00 | At approximately 0118 hours (EST) on November 9, 2012, Susquehanna Steam Electric Station Unit Two reactor was scrammed by plant operators due to a loss of ICS (Integrated Control System; which controls the reactor feed and reactor recirculation systems). The reactor operator placed the mode switch in shutdown when reactor water level reached +25 inches and lowering. All control rods inserted and both reactor recirculation pumps tripped at -38 inches. Reactor water level lowered to -52 inches causing Level 3 (+13 inches) and level 2 (-38 inches) isolations. HPCI and RCIC both automatically initiated. HPCI was overridden prior to injection and RCIC was utilized to restore reactor water level to the normal band. All isolations and initiations at this level occurred as expected. No steam relief valves opened. Pressure was controlled via turbine bypass valve operation. All safety systems operated as expected. The (Unit 2) reactor is currently stable in Mode 3. An investigation into the cause of the loss of ICS is underway. Unit One continued power operation (at 78% power). The NRC Resident Inspectors were notified. A press release will occur. The licensee will inform the State of Pennsylvania. Decay heat removal is being maintained through the main condenser. On-site electrical power is in the normal configuration. |
ENS 48485 | 7 November 2012 09:39:00 | A planned outage of all FCS (Fort Calhoun Station) sirens is to occur at 0900 CST today to replace required router power supplies. Based on the planned maintenance, all sirens for the Alert Notification System within the Emergency Planning Zone (EPZ) will be nonfunctional. Prior notifications and coordination with Local Law Enforcement have been completed with compensatory measures established to support notification of the public in case of an actual emergency during the scheduled maintenance. The planned maintenance is expected to take three hours with a projected completion time of 1200 CST. Also, contingencies have been established with the maintenance to back out if required in support of Law Enforcement activities. 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'. The licensee notified the NRC Resident Inspector and local counties.
At 1129 CST, maintenance was completed with all sirens restored to functional status. Local Law Enforcement has been notified and relaxed required compensatory actions. NRC Resident informed. Notified R4DO (Farnholtz). |
ENS 48492 | 8 November 2012 10:11:00 | The following information was received by facsimile: While reviewing an order for a Type 9200 Butterfly Valve, it was discovered that some parts (shipped from Marshalltown, Iowa from 2009-present) had not been identified for Commercial Grade Dedication, These parts were considered Essential-to-Function and were needed for the valve assembly to perform its safety-related function(s). (The vendor) then extended its review to all safety-related orders going back to 2009. This review confirmed that the failure to dedicate was not confined to Type 9200 butterfly valves. The beginning of 2009 was selected because (an audit follow-up was) performed in January, 2009. A key point of emphasis during the audit was commercial grade dedication, with a recommended outcome being the addition of guidance on commercial grade dedication with all new quotations. Therefore, the 2009 audit brought clarity and consistency in approach, that was not always applied correctly between January, 2009 thru October, 2012. It is (vendor's) opinion that while the affected equipment identified as safety-related was not properly dedicated, the failure to dedicate does not appear to pose an inherent safety risk based on currently available Information. The reason being that much of the equipment affected by (the vendor's) error were items such as mounting kits, wherein (the vendor) failed to dedicate the hardware used to mount the instrument (i.e., the bracket and screws), but did properly dedicate the instrument. Additionally, in forming this opinion, (the vendor) has taken into account that there are no known field issues with the affected equipment and all such possibly non-dedicated equipment passed the required standard testing. U.S. nuclear plants affected include the following: D.C. Cook; Millstone; Surry; McGuire; San Onofre; Indian Point; Palisades; Clinton; Peach Bottom; Beaver Valley; Cooper; Palo Verde; Brunswick; Hope Creek/Salem and Watts Bar. It is noted that numerous foreign facilities are also affected. If there are any technical question, contact the Fisher Quality Manager (George Baitinger) : Fax: (641) 754-2854 or Phone: (641) 754-2026. |
ENS 48523 | 20 November 2012 11:41:00 | An ICAM (Improved Chemical Agent Monitor) is missing at Fort Riley, KS. The ICAM contains a 10 mCi Nickel-63 source. The licensee has initiated a formal investigation and notified NRC Region 3 Inspector (McGraw). 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 |
ENS 48435 | 23 October 2012 20:59:00 | On October 23, 2012, V. C. Summer Station Unit 1 (VCSNS) identified two reactor vessel head penetrations (19 and 52) that did not meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1. The station is in a refueling outage (RF20) and the plant is currently shutdown and in Mode 6. The reactor vessel head (RVH) contains a total of 66 penetrations and inspection efforts are approximately 50% percent complete. There have been no previous repairs to the reactor vessel head penetrations and/or j-groove welds. The indications are not through wall as indicated by volumetric and bare metal visual inspections. The penetrations will be repaired to meet the requirements of 10CFR50.55a prior to returning the vessel head to service. The inspection results are reportable pursuant to 10CFR50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. The licensee will notify the State of South Carolina and local counties. * * * UPDATE FROM BETH QUATTLEBAUM TO PETE SNYDER AT 1800 ON 10/25/12 * * * On October 25, 2012, V. C. Summer Station Unit 1 (VCSNS) finalized our inspections of the reactor vessel head, which has identified a total of four reactor vessel head penetrations (19, 52, 31, and 37) that did not meet the requirements of 10CFR50.55a(g)(6)(ii)(D) and ASME Code Case N-729-1. The station is in a refueling outage (RF20) and the plant is currently shutdown and defueled. The indications are not through wall as indicated from volumetric and bare metal visual inspections. The penetrations will be repaired to meet the requirements of 10CFR50.55a prior to returning the vessel head to service. The inspection results are reportable pursuant to 10CFR50.72(b)(3)(ii)(A). The NRC Resident Inspector has been notified. Notified R2DO (Musser). |
ENS 48433 | 22 October 2012 19:29:00 | At 1846 EDT on 10/22/12, the Main Control Room received a report of an arc flash and smoke in the screen house. The fire brigade leader was dispatched and confirmed that smoke was coming from heater board H-111. The board was de-energized at 1854 EDT and smoke was observed to be reduced. The licensee declared an Unusual Event at 1901 EDT per EAL 8.2.1. The fire was out at 1908 EDT. The Unusual Event was terminated at 1932 EDT. Unit 1 continues running at full power operation and all parameters are stable. There was no impact on Unit 2. The licensee notified the NRC Resident Inspector, State of New York and Oswego County. Notified DHS SWO, FEMA, DHS NICC and NuclearSSA via email. |
ENS 48431 | 22 October 2012 14:46:00 | At 0745 CDT on 10/22/2012, Dominion announced its plan to permanently shutdown and decommission the Kewaunee Power Station. A local press release was made at 1100 (CDT) on 10/22/2012. At 1139 (CDT), security reported that members of the local media were gathering on company property at the plants training facility which is located outside the Owner Controlled Area. At this time security reports no security issues and no other press releases are planned at this time. This event is being conservatively reported under 10 CFR 50.72(b)(2)(xi) for offsite notifications based on the press release and the notification to Midwest Independent System Operator (MISO) of the planned shutdown of the Kewaunee Power Station. The licensee notified the NRC Resident Inspector. |
ENS 48432 | 22 October 2012 16:48:00 | Incident: At approximately 1130 CDT, at the new Joplin High School Construction Site in Joplin, Jasper County, Missouri, a nuclear density (gauge) was run over by a tire of a towed scraper. The damage was primarily to the housing of the meter. The technician immediately secured the area and contacted his supervisor. (The supervisor) went to the site and took readings using a TROX-A-LERT Model 3105B Radiation Survey Meter of the gauge and the soil where the gauge was run over and of the scraper which ran over the (gauge). Readings at 1 meter from the (gauge) ranged from 0.12 to 0.20 mrem/hr. A reading of 2.5 mrem/hr (was indicated at the) housing of the (gauge) . The source did not appear to be damaged and was still encapsulated in the (gauge). The gauge was placed in a truck box mounted in the bed of a pickup truck and taken back to (the licensee's lab). Once at the lab the gauge was wrapped in lead sheeting and placed in a nuclear (gauge) storage box. A reading of 0.1 mrem/hr was obtained at 1 meter with the lead sheeting wrapped around the meter. The damaged gauge is a CPN, serial number MD1205992. |
ENS 48428 | 21 October 2012 13:03:00 | The control room was notified that an unknown quantity of H2 gas (classified as a minor coupling leak (identified with Snoop liquid leak detector)) is currently being released to the air from the Unit 2 Full Flow Hydrogen skid. The Environmental Protection Group reported the leak to the California Emergency Management Agency (Cal EMA) at 0809 PDT and the San Diego Department of Environmental Health at 0812 PDT lAW (plant) procedure S0123-XV-17.3, 'Spill Contingency Plan'. The Hydrogen Gas leak is currently still in progress. There is no gas collection areas. Maintenance is in the process of taking action to terminate the leak. The licensee notified the NRC Resident Inspector. |
ENS 48422 | 19 October 2012 15:51:00 | The following information was received by facsimile: DRH (Tennessee Division of Radiological Health) was notified on 10/19/12 by the licensee that preliminary analytical results of a 30-gallon waste container of tritium contaminated waste oil and absorbents indicate approximately 8,600 curies of tritium. The container was manifested by the generator (SRS) (Savannah River Site) as containing 14.9 curies. Post sampling of this container, DSSI (Diversified Scientific Services, Inc.) had a controlled but elevated level of tritium contamination in the process room; this room is under negative ventilation. The room has been decontaminated back to normal processing levels. Contents of the container have been overpacked and placed in a safe state, (and) not a source of tritium contamination to the air or surface. Tennessee Event: TN-12-267 |
ENS 48421 | 18 October 2012 18:57:00 | During BFNP (Browns Ferry Nuclear Plant) NFPA (National Fire Protection Association) 805 transition review, it was determined in the event of an Appendix-R fire, fire induced circuit damage could potentially result in the Residual Heat Removal System Division II inboard isolation valve being prevented from opening or cause the valve to spuriously close on units 2 and 3. The current Appendix R safe shutdown analysis credits opening of these valves. Failure to open these valves results in loss of ability to provide long term cooling to the core. Compensatory actions in the form of fire watches to mitigate this condition are in place in accordance with the BFNP Fire Protection Report. This condition is being reported pursuant to 10CFR50.72(b)(3)(ii)(B) and 10CFR50.72(b)(3)(v). The NRC Resident Inspector has been notified. |
ENS 48458 | 30 October 2012 13:24:00 | The following information was received by email: On October 18, 2012, the Agency (State of Texas Radiation Branch) received a request from an exposure device manufacturer for reciprocity to retrieve a . . . cobalt 60 source into a Spec 300 radiography device. The Agency contacted the Texas licensee and was told that the source had been retracted and that no reportable event had occurred. The licensee stated it needed assistance in disconnecting the drive cable from the source pig tail. On October 23, 2012, the Agency was informed by the manufacturer that the dose rates measured at the front of the exposure device indicated that the source may not be in the fully shielded position. On October 30, 2012, the Agency was informed by the manufacturer that the source was stuck inside the device approximately three inches from the locked and fully shielded position. The licensee has not reported this event to the Agency. The Agency will conduct an on-site investigation at the licensee's facility on November 1, 2012. There does not appear to have been any exposures to members of the general public. There were no overexposures to employees of the manufacturer. Exposures to the licensee's employees have not been determined. The investigation into this event is ongoing. Additional information will be provided as it is received in accordance with SA 300. Texas Incident: I-9000 |
ENS 48425 | 19 October 2012 17:34:00 | A patient was being treated for a liver tumor and the dosage was delivered to the wrong lobe. The physician prescribed 17 mCi of Y-90 to the left lobe of the liver. The prescribing physician was not present during the dose administration procedure. The intervention radiologist examined a fluoroscope of the patient's liver and noted a larger tumor on the right lobe. The prescribed dosage was then delivered to the right lobe of the liver, not in accordance with the prescribed dosage plan. Patient examination detected no observable impact, and the physician is developing another dosage plan for the patient. The Iowa Department of Public Health will perform an onsite investigation. 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.
The following was received from the State of Iowa via email: The University of Iowa nuclear medicine staff notified the Environmental Health & Safety (EHS) office on October 19, 2012 that they had discovered a reportable medical event that occurred during an October 18, 2012, therapy administration of Yttrium-90 microspheres (SIRSpheres) to a patient with liver metastases from a carcinoid tumor. The Nuclear Medicine authorized user had prescribed for the delivery of 17 millicuries of the Y-90 microspheres to the left lobe of the patient's liver with plans to also administer 34 millicuries of the Y-90 microspheres to the right lobe of the patient's liver at a later time. Y-90 activity to be administered was calculated based on body surface area and tumor involvement. However, the interventional radiologist performing the Y-90 microsphere administration under fluoroscopic guidance noted that the angiogram of the patient's liver showed more tumor blood flow in the right lobe of the patient's liver and decided it would be medically more advantageous for the patient to treat the right lobe of the liver first. He proceeded to treat the right lobe without consulting the authorized user. The interventional radiologist was only able to deliver 96% of the 17 millicurie Y-90 dose to the right lobe of the patient's liver before the right artery occluded. The authorized user discovered the administration discrepancy the next day when the patient was scheduled for post therapy imaging on October 19th. Both the patient and referring physician were informed of the medical event on October 19th. The physicians involved concluded that the patient's health and outcome were not affected by treating the right lobe of the liver first with only 17 millicuries of Y-90 microspheres since only 96% of activity could be injected before full embolization of the right hepatic artery occurred. The physicians plan to treat the left lobe of the patient's liver in approximately 4 weeks. The Iowa Department of Public Health (IDPH) performed an onsite investigation of this medical event on October 23, 2012. The IDPH inspector interviewed the University Radiation Safety Officer, the Nuclear Medicine Authorized User, the Interventional Radiologist performing the Y-90 administration, the Chief Nuclear Pharmacist, the Chairman of the Medical Radiation Protection Committee, and Hospital Radiation Safety Review Group. The cause of the event was a lack of understanding of the requirements for administering radioactive material under the supervision of an authorized user. The interventional radiologist made a medical decision to alter the administration site without consulting the prescribing authorized user. To avoid recurrence the following actions are being taken by the University of Iowa. All nuclear medicine authorized users prescribing Y-90 microsphere therapy will review the supervision requirements specified in Iowa Administrative Code 641-41.2(11) with all interventional radiologists on staff to ensure that they understand that they are required to follow the instructions of the prescribing authorized user. A roster of the individuals receiving this training will be forwarded to the IDPH upon completion. Additionally, nuclear medicine and interventional radiology will develop a written policy regarding the proper steps to be taken in the event that any deviation from the authorized user's written directive for the medical administration of radioactive materials is required. A copy of the policy will be forwarded to IDPH upon completion. Reporting Requirement: 35.3045(a)(1(i) - Total dose delivered that differs from the prescribed dose by 20% or more; and differs from the prescribed dose by more than 0.05 Sv (50 rem) SDE. IA Report: IA120006 Notified R3DO (Kozak) and the FSME Event Resource via email. |
ENS 48420 | 18 October 2012 16:56:00 | The following information was received by email: On October 18, 2012, the Agency (Texas Radiation Branch) was notified by the licensee that on October 17, 2012, a radiographer was unable to retract a (source) into the QSA 880F exposure device. The guide tube for the device was damaged during radiography operation in the fixed facility when a part fell on it, crimping the guide tube to a point where the source could not pass through it. The licensee stated that they were able to repair the guide tube enough to retract the source and lock it in the exposure device. The licensee stated that no over exposure occurred from the event and no member of the general public received any exposure from the event. Additional information will be provided as it is received in accordance with SA 300. Texas Incident: I-8998 |
ENS 48575 | 10 December 2012 21:40:00 | On Saturday, October 13, 2012, Unit 1 was operating at 100% power. At 1841 (EDT), the 1A RPS/UPS inverter tripped and the automatic transfer of the RPS and UPS 120 VAC distribution panel (1A-Y160) loads to the primary alternate AC power source was delayed. The delay in automatic load transfer caused the RPS series breakers to trip on undervoltage. The failure caused a loss of power to Division IA and IIA RPS relays and Division IA and IIA NS4 relays. This caused primary containment isolation valves (PClVs) to automatically close on more than one system. The IB and IIB channels were unaffected. The most probable cause for the delayed load transfer was a failed logic power supply with a momentary loss of synchronization. Troubleshooting continues (in order) to confirm the specific cause of the component failure. The distribution panel loads are currently supplied by an installed alternate AC power source. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The isolation was a partial actuation. This 60-day ENS report is being made per 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report invalid automatic actuations of systems listed in paragraph (a)(2)(iv)(B). The listed system that actuated was general containment isolation signals affecting containment isolation valves in more than one system. Primary containment isolation valves (PCIVs) closed on drywell chilled water (DWCW), reactor enclosure cooling water (RECW), primary containment instrument gas (PCIG), Unit 1 containment leak detector, and Unit 2 containment leak detector. The licensee notified the NRC Resident Inspector. |
ENS 48581 | 12 December 2012 12:30:00 | Event Summary: On October 13, 2012, at approximately 0218 (EST), a full reactor scram signal was received in the James A. FitzPatrick (JAF) control room. At the time of this event, the plant was in cold shutdown (Mode 4) and refueling outage 20 (R20) was in progress. The scram signal occurred because Reactor Vessel Scram & Primary Containment Isolation Level Transmitter (02-3LT-101C) and Reactor Vessel Scram & Primary Containment Isolation Level Transmitter EQ (02-3LT-101D) momentarily failed downscale, and then immediately recovered. 02-3LT-101C is an 'A' division component and 02-3LT-101D is a 'B' division component. Therefore, both the 'A' and 'B' divisions of reactor protection actuated providing a full reactor scram and outboard primary containment isolation signal. AOP-15, 'Isolation Verification and Recovery' verified that the proper containment isolation response was received. This event is reportable under 10 CFR 50.73(a)(2)(iv)(A) because it resulted in the invalid actuation of the reactor protection system while the reactor was already shut down. Apparent Cause: A Failure Mode Analysis and Apparent Cause Evaluation were performed to determine the most likely cause of both level transmitters to momentarily spike downscale. It was concluded that the most probable cause was due to a worker inadvertently coming into contact with the level transmitters' exposed sensing lines. Contributing to this event was the misjudgment by the Operations individual reviewing the work package, on the risk significance of the instrumentation in the vicinity of the work area. As a result, the work area was not constructed in a manner to preclude interference with the level transmitters or associated sensing lines. Corrective Actions: Immediate corrective actions were to walk down the affected instrument lines to ensure no damage had been caused. Additional corrective actions were to install signs near the level transmitters and exposed sensing lines. The signs denote that sensitive instrument lines are present. Future corrective actions include a walk down by engineering and operations to identify other areas in the plant where sensitive instrument lines are present and place additional signs or barriers as appropriate. The licensee notified the NRC Resident Inspector. |
ENS 48388 | 8 October 2012 02:37:00 | On October 7, 2012, with Unit 2 in a defueled condition, a differential current lockout occurred on the 2B3 4.16kV essential bus, causing a deenergization of the 2B3 4.16kV essential bus. At the time of the event, the 2B Emergency Diesel Generator (EDG) was loaded to the essential bus. Due to the differential current lockout, all bus loads were lost and the 2B EDG output breaker feeding the essential bus opened and the 2B EDG transferred to emergency mode. The 2A EDG is operable and in standby. All equipment responded as expected. The plant is currently being maintained in a defueled condition. Decay heat removal is being supplied by the 2A Fuel Pool Cooling train. The cause of the differential current lockout of the 2B3 4.16kv bus is under investigation. This event is reportable pursuant to 10CFR 50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector. |
ENS 48381 | 5 October 2012 07:37:00 | At approximately 0800 EDT, on October 5, 2012, the Emergency Response Facility Information System (ERFIS) will be removed from service to perform a planned modification for the improvement of site wide data communications between various plant process computing platforms. The expected duration of ERFIS non-functionality should not exceed 24 hours and during this time would not be able to be restored within one hour. The ERFIS computer system provides monitoring and communications capability for plant data systems including the Emergency Response Data System (ERDS) and the Safety Parameter Display System (SPDS). The loss of ERFIS requires alternate methods, as described in plant procedures, to be used for the above described functions. Therefore, assessment of plant conditions, notifications, and communications could still be made, if required, during the time that the ERFIS computer system is non-functional. The on call Emergency Response Organization has been notified of the ERFIS outage. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Rev. 2. The NRC Resident Inspector has been notified. |
ENS 48354 | 28 September 2012 09:00:00 | This is a non-emergency eight hour notification for a loss of Emergency Assessment Capability. On September 28, 2012, the EOF/TSC air handler chiller unit was removed from service to perform planned maintenance. This maintenance activity will not affect the air filtration portion of the system and these facilities remain available for use during an emergency. This maintenance activity will be performed in a manner to minimize the time that the air handler chiller is out of service. This maintenance activity impacts the ability to maintain ambient air temperature in the facilities. The (estimated) duration of this activity is planned to be 4 hours. If an emergency condition occurs that requires activation of the emergency response facilities, the EOF and TSC will be utilized. The Emergency Response Organization team members have the ability to relocate to alternate locations in accordance with emergency implementing procedures based on conditions. Alternate emergency response facilities will remain available in the event that relocation is necessary. This report is being made in accordance with 10 CFR 50.72(b)(3)(xiii), which is any event that results in a major loss of emergency assessment capability, offsite response capability, or offsite communications capability. An update message will be provided when the emergency response facilities are restored. The licensee notified the NRC Resident Inspector, the State of South Carolina and the local counties of Lee, Chesterfield and Darlington.
The EOF/TSC Chiller is back in service as of 1102 (EDT) on 9/28/12. The ability to maintain ambient air temperature in the EOF/TSC facilities has been restored. The licensee notified the NRC Resident Inspector. Notified the R2DO (Nease). |
ENS 48350 | 27 September 2012 11:01:00 | The following information was received by facsimile: GEH (General Electric Hitachi) has recently discovered that calculations of choked flow rate in the Main Steam Line (MSL) of GEH BWRs may not be conservative, with the potential impacts to be evaluated for existing MSL high-flow setpoints and Analytical Limits (ALs). GEH has not completed the evaluation of this condition to determine reportability under 10CFR Part 21 and is therefore issuing this 60-day Interim Notification. GEH will close or issue an update on this matter on or before December 12, 2012. Given the early status of the evaluation, GEH has no recommended actions at this time. This 60-day Interim Notification is issued in accordance with 10CFR Part 21.21(a)(2), and will be sent to all GE BWR/2-6 plants and ABWR plants. Affected plants include the following: Nine Mile 1-2, Fermi 2, Columbia, Grand Gulf, River Bend, FitzPatrick, Pilgrim, Vermont Yankee, Clinton, Dresden 2-3, LaSalle 1-2, Limerick 1-2, Oyster Creek, Peach Bottom 2-3, Quad Cities 1-2, Perry 1, Duane Arnold, Cooper, Susquehanna 1-2, Brunswick 1-2, Hope Creek, Hatch 1-2, Browns Ferry 1-3, and Monticello. |
ENS 48345 | 26 September 2012 09:47:00 | On 9/26/12, the Monticello Nuclear Generating Plant's TSC power supply will be isolated to perform a planned maintenance activity. The maintenance activity requires implementation of compensatory measures to maintain TSC functions during the planned activity. Compensatory measures include having the Emergency Director report to Control Room and relocating the remaining TSC staff at the EOF should an event be declared requiring Emergency Response Organization (ERO) activation. Maintenance activity is scheduled to be complete with the TSC fully functional by end of dayshift on 9/26/12. Site ERO has been notified of maintenance activity and instructed on planned compensatory measures to be implemented during activity if required. This event is considered reportable per 10CFR50.72(b)(3)(xiii). The licensee notified the NRC Resident Inspector. The licensee will notify the Minnesota State Duty Officer, and the Sherriff Departments for both Wright and Sherburne Counties.
At approximately 1520 Central time, a disturbance occurred in the 12.5kV system during restoration activities for the TSC power supply. This would have caused a Major Loss of Emergency Assessment Capability and thus is reportable under 10CFR50.72(b)(3)(xiii). The licensee is investigating the disturbance. The EOF and all other emergency assessment capabilities were verified functional. The licensee notified NRC Resident Inspector. The disturbance to the 12.5kV system did not affect plant operations. Notified R3DO (Lipa). |
ENS 48349 | 27 September 2012 07:17:00 | New York City was notified by the licensee Radiation Safety Officer of a patient receiving an underdose of Yttrium-90 TheraSpheres for liver treatment. The prescribed dose was 120 gray, but only 11.4 gray was delivered to the patient. A review of the licensee's patient delivery system indicates that about 90% of the prescribed dosage remained in the catheter. 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. |
ENS 48317 | 18 September 2012 05:59:00 | This report is being made pursuant to 10CFR50.72(b)(3)(v)(C), Event or Condition that could have prevented fulfillment of a Safety Function needed to Control the Release of Radioactive Material. An employee entered a secondary containment interlock and identified that both doors of the interlock opened simultaneously when the door on the reactor building side was opened. The employee immediately secured both doors in the interlock and notified the Main Control Room Supervisor. Both doors in the interlock were open for approximately 10 seconds. With both doors open, TS SR 3.6.4.1.2 was not met. This rendered secondary containment inoperable per TS 3.6.4.1. Reactor Building differential pressure, as observed in the Main Control Room, has remained less than -0.25" H20 at all times. Initial investigation determined that a mechanical interlock for the doors was malfunctioning. Administrative controls have been put in place to ensure the doors remain closed pending repairs to the mechanical interlock. The licensee notified the NRC Resident Inspector. |
ENS 48195 | 16 August 2012 10:58:00 | A HDR (High Dose Rate) treatment was planned for a patient's mammary area, but the delivery of the source position was offset from the intended delivery. The patient and referring physician were informed of the difference between the planned and the delivered dosage. The patient will receive follow-up to discuss potential options. 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. |
ENS 48191 | 15 August 2012 10:37:00 | The following information was received by email: On August 15, 2012, the licensee notified the Agency (Texas Department of State Health Services) that sometime during the evening, a QSA Global 880D radiography camera was stolen out of the dark room of one of their trucks. The camera contained Iridium-192. The truck was parked at the licensee's office, however the Radiographer left the camera in the truck instead of transferring the camera to the vault. The thieves broke into five radiography trucks taking various items including generators and the one camera. A police report was filed with the Chamber County Sheriff's Department. The thieves did not take the crank outs or source tube. The State of Texas event number for this event is I- 8979. Additional information will be provided in accordance with SA 300. Notified DHS, FEMA, USDA, HHS, DOE, DHS NICC and EPA EOC.
On August 15, 2012, around 1130 (CDT), the licensee and Local Law Enforcement recovered the stolen QSA Global 880D Radiography Camera. The camera is stored in the vault at the licensee facility and is in normal working condition. (A person of interest has been identified). The truck used by the thief was identified on video surveillance tape. The camera was in the back of truck at the residence of the thief. Investigation will continue for the company and radiographers (concerning) Increased Controls (IC) security violations. Notified R4DO (Hagar), FSME (McIntosh), ILTAB (Whitney), IRD (Gott), Mexico via fax, DHS, FEMA, USDA, HHS, DOE, DHS NICC and EPA EOC. THIS MATERIAL EVENT CONTAINS A "CATEGORY 2" LEVEL OF RADIOACTIVE MATERIAL Category 2 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 a short time (minutes to hours). It could possibly be fatal to be close to this amount of unshielded radioactive material for a period of hours to days. These sources are typically used in practices such as industrial gamma radiography, high dose rate brachytherapy and medium dose rate brachytherapy. For additional information go to http://www-pub.iaea.org/MTCD/publications/PDF/Pub1227_web.pdf |
ENS 48190 | 15 August 2012 04:37:00 | At 2045 (CDT) on 8/14/12, MNGP (Monticello Nuclear Generating Plant) Operations determined that valves RHR-82 and RHR-84 had been inappropriately closed as part of an isolation clearance order for work on shutdown cooling suction piping. These valves are required to be open to provide overpressure protection for RHR piping passing through primary containment penetration X-12. Upon discovery of the condition, Primary Containment was declared Inoperable and the Required Actions of Tech Spec 3.6.1.1 were entered. Following discovery, the isolation was restored and the valves opened. At 0001 (CDT) on 8/15/12, Primary Containment was declared Operable. This issue is being reported in accordance with 10CFR50.72(b)(3)(v)(C) and 10CFR50.72(b)(3)(v)(D) as a condition that at the time of discovery could have prevented the fulfillment of the safety functions of a system needed to control the release of radioactive material or to mitigate the consequences of an accident. The MNGP Senior NRC Resident Inspector has been notified of this issue. The licensee will contact the Minnesota State Duty Officer.
This notification is a retraction of ENS 48190 based on further engineering evaluation. Monticello had previously evaluated penetration X-12 for thermally induced over pressurization. The evaluation qualified the piping components in the penetration for a maximum pressure of 3,306 psig using ASME Section III Appendix F operability criteria. The peak pressure calculated for the penetration was 2,743 psig based on Reactor pressure of 1000 psig with Reactor in Mode 1, and at worse case LOCA conditions for the Drywell. These assumptions and parameters envelop those that were present when valves RHR-82 and RHR-84 were closed on August 14, 2012. Therefore, this event would not have prevented the fulfillment of the safety function reported. The NRC Resident Inspector has been notified. Notified R3DO (Duncan). |
ENS 48423 | 19 October 2012 16:25:00 | The licensee misplaced a static eliminator device containing 10 mCi of Po-210. The device model number is P-2021-8301; Serial number is A2HS590. The device was used to neutralize static charge on plastic parts prior to entering the paint process. The licensee considers the device as lost. 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 |
ENS 48134 | 25 July 2012 22:14:00 | At 1602 (CDT), Engineering personnel notified the control room that during review of a pipe stress calculation it was identified that non-conservative or incorrect methodologies were used in the calculation. This calculation was for a modification to install four; 3 (inch) drain lines between the Essential Service Water (ESW) (safety) and the Service Water (SW) (non-safety) in 1991. A preliminary ME101 stress analysis performed, which corrects the above-identified discrepancies, indicates that the pipe stresses at the drain line weldolet connection exceed the ASME code of record allowable stresses by approximately 50%, when the revised Stress Intensification Factor (SIF) is applied. This modification affected both trains (A & B) ESW trains. The normal system alignment uses the SW water to supply the ESW, then during accident conditions the SW and ESW systems isolate from each other so that two redundant separate train isolation valves isolate the ESW system. These 3 (inch) drain lines are located in the section of piping that is isolated from the ESW and SW systems. At the time of notification 'A' ESW was isolated from SW and 'B' ESW was in normal system alignment. 'B' ESW was declared inoperable and action was taken to separate the SW and ESW and isolate the 3 (inch) drain valves. With this action complete the non-conforming components have been removed from service and OPERABILITY of the ESW has been restored. This condition is been reported per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified.
Further engineering evaluation determined that the four drain lines (3-inch) between the Essential Service Water (ESW) (safety) and the Service Water (SW) (non-safety) were found to be within the allowable limits for operability and are acceptable. As a result, the condition has been determined to not be reportable per 10 CFR 50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. Notified R4DO (Miller). |
ENS 48133 | 25 July 2012 17:32:00 | During annual physical inventory of Special Nuclear Material (SNM) conducted at Grand Gulf Nuclear Station (GGNS), a replacement Source Range Monitor (SRM) detector could not be accounted for. The SRM detector was being stored in the refuel floor storage locker after maintenance activities during RF18. This detector did not function and was removed prior to the start-up therefore, the SRM detector was not subjected to a critical reactor. Actions continue to locate the missing SRM detector. There is no evidence of sabotage or tampering associated with this activity. The licensee notified the NRC Resident Inspector. 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 |
ENS 48132 | 25 July 2012 14:45:00 | TMI is issuing a press release and performing courtesy stakeholder communications as a result of an elevated level of tritium detected in one of 55 on-site ground water monitoring wells near the plant structure. The tritium concentration in the well is contained to the site and well (within) EPA drinking water standards. TMI is performing voluntary communications in accordance with NEI 07-07 Industry Ground Water Protection Initiative and contacting local and state stakeholders as a courtesy. As a result of the press release, a four-hour notification is being made per 10 CFR 50.72 (b)(2)(xi). The licensee notified the NRC Resident Inspector, the Commonwealth of Pennsylvania and local counties (Dauphin and York). |
ENS 48127 | 24 July 2012 13:43:00 | At approximately 0555 (EDT) on 07/24/2012, the AH-17 TSC Cooling Fan was found with the cooling system not fully working. The fan is running, but the condensing compressor is not. Repairs are being planned and will be worked immediately. This event is reportable per 10 CFR 50.72(b)(3)(xiii) as described in NUREG-1022, Revision 2. The on call Site Emergency Coordinator and Emergency Response Manager have been notified. The Alternate TSC is available per plant procedure, if required. The licensee notified the NRC Resident Inspector.
TSC cooling system repairs have been completed and the system has been returned to normal. The licensee notified the NRC Resident Inspector. Notified the R2DO (Franke). |
ENS 48128 | 24 July 2012 14:23:00 | The following information was received by email: Wisconsin Department of Health Services (DHS) received notification via voicemail from the licensee on July 23, 2012 regarding the discovery of a brachytherapy medical event that occurred on July 17, 2007. The medical event involved a procedure during which the administered dose differed from the prescribed dose by more than 20% (underdose) for permanent prostate implants using Cs-131 seeds. The prescribed dose was 115 Gy. The delivered dose was 60 Gy. The medical event was identified by DHS inspectors during a recent routine inspection, at which time the inspectors also determined that the licensee had not conducted adequate review of their prostate brachytherapy cases against their medical event criteria. Per the licensee's procedures, a post-implant CT is performed the same day as the procedure and analyzed within two working days after. If the post-implant analysis reveals an underdose, which occurred in this situation, additional seeds are implanted and another post-implant CT/analysis process follows. The licensee is currently reviewing additional cases dating back to 2003 for medical events against the revised medical event criteria submitted to DHS on June 22, 2012. DHS is investigating the event and is currently in the process of communicating with the licensee regarding decisions on patient notifications. A special inspection team will be sent following the licensee's review. Wisconsin Event: WI 120009 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. |
ENS 48130 | 24 July 2012 15:50:00 | The following information was received by email: On the evening of July 18, 2012 the licensee's radiation safety officer contacted the (Illinois Emergency Management) Agency's Radiation Duty Officer to advise that a fixed gauge had failed to perform as expected when an attempt was made to close the shutter in advance of work scheduled to be performed on the vessel it was mounted on. The RSO reports that although the spindle of the handle rotated, the shutter did not close. The shutter remained stuck in the 'open' position and would not close. As a result, the planned work was deferred and access to the confined space has been prohibited. The gauge is mounted such that there are no adjacent work stations and unrestricted access is not possible. Radiation monitoring performed by the radiation safety officer showed no abnormal readings. The gauge shutter had been successfully tested in March with no abnormalities at that time. The gauge is mounted in such a way that although it is exposed to ambient temperatures, it is not directly exposed to the elements by virtue of an overhead cover. The gauge has not been subjected to any corrosives or solvents or other conditions which exceed the gauge prototype's test conditions. Arrangements have been made to have the manufacturer's representative on-site to conduct an evaluation and effect repairs as necessary at the end of the week. The fixed gauge is a model SA-1 manufactured by Ronan Engineering (serial number 69094). The gauge contains a Cs-137 sealed source (.082 Ci). Illinois Incident: IL 12011 |
ENS 48292 | 10 September 2012 15:47:00 | The following information was received on 9/7/2012 via email: The Corporate RSO for Bed, Bath and Beyond (B3) called to advise that an exit sign containing H-3 appears to be missing from their Illinois store. The sign had been taken down by an electrician in July in anticipation of recovery and disposal by Shaw Environmental. A representative from Shaw had arrived this past week to collect and package the sign only to find that the sign was missing from the storage location where it had been placed at the time of removal. The electrical company hired by the licensee to perform the un-installation had noted that there were no signs of damage or loss of contents at the time of removal on July 13, 2012. The manager of the licensee's store can only definitely recall having seen the sign on July 14 or 15th in secure storage. Subsequent visual surveys of the store were conducted in an attempt to locate the sign with no success. The sign involved is made by Safety Light Corporation (m/n SLX-60). It was purchased in November of 2004. Based on the serial number involved (289354) it was determined that the sign was sold with a nominal H-3 content of 11.5 Ci. As of this date, that activity is approximately 7.4 Ci. Currently, it is believed that the device has been inadvertently disposed with the facilities normal trash stream. Illinois Number: IL-12014 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 |
ENS 48544 | 28 November 2012 19:47:00 | At 1930 (EST) on November 28, 2012, it was determined that a Notification of Unusual Event (NOUE) was not declared for an event that occurred on June 3, 2012, when an equipment failure resulted in a deposit of ion exchange resin onto the floor of the Radioactive Waste building. Subsequent radiological surveys indicated that conditions met the requirements for a NOUE in accordance with the Perry Nuclear Power Plant Emergency Plan. This report is being provided within one hour of the recognition of the undeclared event. As discussed in NUREG 1022, Revision 2, an actual declaration of an Unusual Event is not necessary. The Initiating Conditions for the emergency classification no longer existed at the time of recognition. The NRC Resident Inspector has been notified. |
ENS 47972 | 29 May 2012 07:22:00 | On 5/29/2012 at 0331 (CDT) the Unit 3 reactor scrammed due to turbine control valve fast closure initiated by a load reject signal on the Main Generator. The cause of the load reject signal is Main Transformer differential relay 387T. Reactor power at the time of the SCRAM was approximately 75%. All systems responded as expected to the load reject signal. Main Steam Isolation Valves remained open and reactor pressure is being controlled on the Main Turbine Bypass Valves. No Main Steam Relief Valves lifted during the transient. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation setpoints were reached. Primary Containment Isolation Signals (PCIS) Groups 2, 3, 6 and 8 were received. The lowest reactor water level observed was -41 inches. Reactor water level was restored to and is being controlled by the Feedwater system in the normal band. This event is reportable within 4 hours per 10CFR50.72(b)(2)(iv)(B), 'any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10CFR50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10CFR50.73(a)(2)(iv)(A). This event is documented in the station corrective action program on SR# 557947. The NRC Resident Inspector has been notified. All control rods inserted into the reactor core. Electrical power is being back fed from offsite power through the 161 KV feeder line. The reactor is being cooled down within the Technical Specification rates. |
ENS 47968 | 26 May 2012 02:28:00 | On May 26, 2012, at 0200 EDT with the reactor at 42% core thermal power, the following press release is being issued by Entergy Nuclear regarding the status of the ongoing labor negotiations. Pilgrim Station Negotiation Team Agrees to Present Proposed Package to Union Membership for Ratification Vote. Plymouth, Mass. - Entergy Nuclear and the Utility Workers Union of America (UWUA), which have been negotiating a four-year Pilgrim Station labor contract, agreed early Saturday to present a proposed package of pay and benefits to the membership for a vote on Saturday, June 7. Additionally, the parties have agreed to a 10-day (contract) extension that will expire at midnight on Tuesday, June 5. The previous contract for the approximately 240 workers will remain in effect during the extension. Pilgrim's Site Vice President Robert Smith said, 'We appreciate the hard work of both parties to get to this point and believe this proposal represents fair and equitable terms both for our employees and the company.' Entergy owns and operates power plants with more than 30,000 megawatts of electrical generating capacity and has about 14,000 employees. Entergy's nuclear businesses comprise six reactors at five sites in Massachusetts, New York, Vermont and Michigan and five reactors at four locations in Arkansas, Mississippi and Louisiana. Entergy Nuclear also manages operations at a nuclear generating plant in Nebraska. The (NRC) Resident Inspector staff has been informed of this press release and notification. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi). The licensee will notify the Massachusetts Emergency Management Agency. |
ENS 47930 | 17 May 2012 12:11:00 | At 0438 (CDT) on Thursday, May 17, the Callaway Plant Emergency Operations Facility (EOF) was declared non-functional when the building's return fan was found not running. Loss of the EOF return fan, results in an inability to maintain a positive pressure on the facility. Efforts are underway to return this fan to service. If an emergency is declared requiring EOF activation while the EOF is non-functional, EOF emergency response personnel will report to their backup locations in accordance with Callaway Plant emergency planning procedures. This notification is being made in accordance with 10 CFR 50.72(b)(3)(xiii) due to the unavailability of an emergency response facility. The NRC Resident Inspector has been notified.
At 1330 (CDT) on 05/17/12, the Callaway Plant EOF was been returned to service. The licensee will notify the NRC Resident Inspector. Notified R4DO (Powers). |
ENS 47932 | 17 May 2012 20:23:00 | During a review of records on May 10, 2012, the licensee was unable to account for 4 NRD Nuclecel In Line Ionizer model number P-2021 1000. A thorough sweep of the facilities was completed on May 16, 2012 at 1600 HST. The sweep failed to locate these devices and they were declared lost. The materiel involved were four devices, each containing 10mCi of Polonium-210 and encased in a steel pressure welded cylindrical case, each approximately 2.3 inches long and .5 inches in diameter. The circumstances under which the loss occurred were that one shipment of 2 devices (Serial Numbers: A2HE600A/A2HE601A) were last known to have been delivered to the licensee logistics center in August 2010 and there were no record of those two items having been returned to NRD, LLC. These two items were received from NRD on 7/8/2010. A second set of two devices (Serial Numbers: A2HT541/A2HT540) were last known to have been in a sealed package in the possession of a Boeing employee in March, 2012. These items were received from NRD on 7/14/2011. All devices are presumed lost and potentially disposed of as ordinary waste. The casing of all devices were known to be intact at the time of loss and no exposure has occurred. A thorough review of shipping records and site wide physical search have been conducted and failed to account for the devices. In the future, all such devices will be subject to the accountability rules, requirements, and procedures established by the USAF Bio-Environment office, Hickam Air Force Base, Honolulu, Hawaii. Internal company procedures will be updated to conform to the requirements of the USAF Bio-Environmental Office, Hickam Air Force Base. 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 |
ENS 48008 | 8 June 2012 18:23:00 | The following information was received by e-mail: Licensee notified DRH (Mississippi Department of Radiological Health) about a Ronan X92 Continuous Level Gauge, Serial No. 9479GG with source Holder Model SA1-F37. Licensee suspects the source shutter may not be closing 100% due to elevated readings of 7 mR/hr with the shutter in the closed position. * * * RETRACTION FROM JASON MOAK TO PETE SNYDER ON 6/26/12 AT 1558 EDT * * * This report is retracted based on the fact that the gauge "did not fail to function as designed. No maintenance was performed on the gauge. The service representative did not find a problem with the gauge shutter or radiation fields. Notified R4DO (Clark), FSME (e-mail). Mississippi Incident Number: MS-12002 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 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) |
ENS 48003 | 7 June 2012 14:18:00 | The following information was received by facsimile: A biannual tritium sign inspection during the week of 5/7/2012 disclosed four tritium signs had been removed during a renovation of Mendocine Hall during the third week of January 2012 by the contractor. All attempts to locate the signs have been exhausted. The missing signs are as follows: Serial # 263374 Model # BX1711OWHSGN 7.5 curies Serial # 263373 Model # BX1711OWHSGN 7.5 curies Serial # 263387 Model # BX1711OWHSGN 7.5 curies Serial # 263388 Model # BX1711OWHSGN 7.5 curies Date of manufacture for all signs is 5/2/2004. California 5010 Number: 060712 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 |
ENS 47869 | 27 April 2012 13:19:00 | 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. |
ENS 47866 | 26 April 2012 13:50:00 | 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. |
ENS 47863 | 25 April 2012 22:42:00 | 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.
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).
On April 25, 2012, at 2106 CDT, Point Beach Nuclear Plant declared an Alert under EAL HA3.1 (EN #47863) 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. This retraction is based on the fact that station personnel incorrectly used an Alert trigger value for toxic gas concentration that was well below the actual value which would pose an immediate danger to life and health. In a subsequent evaluation, NextEra concluded that the station used the carbon monoxide (CO) OSHA limits (OSHA Permissible Exposure Limit 8-hour time weighted average of 50 ppm to 200 ppm short term exposure limit) that are normally monitored for confined spaces. The recommended CO limit value of 1200 ppm was determined to be the correct IDLH for the plant air compressor room, the actual reading in the room was 124 ppm. The plant air compressor room is not a confined space, and is not required to be monitored to the same limits as a confined space. Therefore, the Emergency Notification made on April 25, 2012, documenting an Alert due to toxic gas in a vital area is retracted. The NRC Resident Inspector has been notified. Notified R3DO (Kunowski). |
ENS 51957 | 25 May 2016 15:55:00 | The following information was received from the Commonwealth of Virginia by email: Event description: On May 23, 2016 a Virginia Radioactive Material Program (VRMP) inspector performed a routine unannounced inspection of Superior Paving Corporation. The inspector discovered that on April 20, 2012, two Troxler Model 4640-B Portable Gauges (serial numbers 1384 and 845) containing 8 mCi of Cs-137 each, were damaged by fire. The fire burnt the storage box which contained the two gauges. The transport containers and plastics on the gauges were melted. However, according to the Radiation Safety Officer statement, the integrities of the sources were intact. On May 10, 2012 the two gauges were transported by the licensee to North East Technical Services (NETS) for disposal. A transfer record was available for review. The licensee will provide the agency (VRMP) a detailed report for review. There was no public health exposure or environmental release from this event. Virginia Event Report: VA-16-005 |
ENS 47838 | 16 April 2012 07:38:00 | 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. |
ENS 47814 | 7 April 2012 15:27:00 | 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.
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). |
ENS 47812 | 6 April 2012 20:33:00 | 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. |
ENS 47813 | 6 April 2012 22:18:00 | 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. |
ENS 47809 | 5 April 2012 18:21:00 | 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. |
ENS 47806 | 4 April 2012 21:32:00 | 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. |
ENS 47769 | 25 March 2012 23:58:00 | On March 25, 2012 at 2336 EDT, an Unusual Event was declared due to an earthquake felt on site. The site entered EAL HU1.1. No plant systems were affected. The National Earthquake Information Center reported a magnitude 3.1 seismic event 6 miles south-south west of Mineral, Virginia. A plant inspection is on-going to determine any plant issues related to the seismic event. Unit 1 is in a refueling outage and containment integrity was maintained. Unit 2 continues in full power operation. The licensee notified the NRC Resident Inspector, State and local agencies. Notified DHS SWO, FEMA, NICC and Nuclear SSA via email.
On March 26, 2012 at 0410 EDT, the Unusual Event was terminated. The basis for the termination was that all equipment walkdowns are complete with no damage discovered. The licensee will notify the NRC Resident Inspector. Notified the R2DO (Haag), NRR EO (Brown), IRD (Grant), DHS SWO, FEMA, NICC and the Nuclear SSA via email. |
ENS 47767 | 25 March 2012 01:44:00 | On March 24, 2012, at 1855 (EDT) during the performance of work activities to support Alloy 600 dissimilar metal weld overlay work on the 'B' Reactor Coolant loop hot leg to the 'B' Steam Generator nozzle weld, two through-wall defects were identified. The workers noted a small amount of water seeping from the indications in the nozzle weld area. The indications are in the area of excavation that was being performed for the weld overlay project. Approximately 1 (inch) of weld material had been removed prior to the seepage being identified. Entered Technical Requirement 3.4 .6, 'ASME Code Class 1, 2 and 3 Components' and immediately initiated actions to isolate the 'B' Reactor Coolant loop. The 'B' Reactor Coolant loop stop valves were closed at 2312 hours on March 24, 2012, which isolated the defects from the reactor coolant system . An engineering evaluation of the defects will be performed and corrective actions implemented. This event is reportable in accordance to 10CFR50.72(b)(3)(ii)(A) for 'any event or condition that results in the condition of the nuclear power plant, including its principle safety barriers, being seriously degraded'. The licensee notified the NRC Resident Inspector and will notify Louisa County. |
ENS 47748 | 16 March 2012 10:56:00 | The Refueling Water Storage Tank (RWST) was placed on purification in accordance with (site procedure) OP-913, Refueling Water Purification Pump Operation, as directed from OP-301-1, Chemical and Volume Control System (Infrequent Operation), at 0400 (EDT) on 3/16/2012 to support make up of level to the RWST. This condition, connection of the purification loop, is not currently allowed based on unresolved seismic concerns with purification piping to the RWST. This was later discovered during a log review at 0545, and operators were immediately directed to remove the RWST from purification. ITS 3.5.4 was applied from 0400 based on when it was determined that this condition had been entered. ITS 3.5.4 was exited at 0622 when the RWST was removed from purification. This is being reported pursuant to 50.72(b)(3)(v), Event or Condition That Could Have Prevented Fulfillment of a Safety Function. The licensee notified the NRC Resident Inspector. |
ENS 47746 | 15 March 2012 08:32:00 | A planned maintenance evolution at the Oyster Creek Nuclear Generating Station will remove the Technical Support Center (TSC) ventilation system from service. Therefore, the TSC ventilation system will be rendered non-functional during the course of the work activities. The TSC ventilation is expected to be out of service for approximately six hours from 0800 (EDT) to 1400 today. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures unless the TSC becomes uninhabitable due to ambient temperature, radiological, or other conditions. If relocation of the TSC becomes necessary, the Emergency Director will relocate the TSC staff to an alternate location in accordance with applicable site procedures. This notification is being made in accordance with 10CFR 50.72(b)(3)(xiii) due to potential loss of the TSC. An update will be provided once the TSC ventilation has been restored to normal operation. The NRC resident has been notified.
The TSC was restored to service on March 15, 2012 at 1107 EDT. The licensee will notify the NRC Resident Inspector. Notified the R1DO (Krohn) |
ENS 47721 | 6 March 2012 19:33:00 | At 1613 (CST), FCS (Fort Calhoun Station) Control Room was notified by corporate communications that 2 non-adjacent sirens out of 101 total sirens lost communication. These were Siren #1 at 1602 (CST) and Siren #50 at 1609 (CST) due to potential router issues. Required compensatory actions were established at 1630 (CST) for loss of the 2 sirens. Initial troubleshooting revealed radio communication failed at 1543 (CST) and manual testing of the remaining sirens was instituted. At 1643 (CST), corporate communications notified the Control Room initial testing on Siren #2 was not responding and indicative that all sirens were lost. Based on that report, all sirens for the Alert Notification System within the Emergency Planning Zone (EPZ) were declared nonfunctional and notifications were completed. Local Law Enforcement has been notified in the required surrounding counties to perform required actions in case of an emergency with the sirens unavailable. Compensatory measures are in place to ensure notification of the public by local law enforcement in case of an actual emergency. Troubleshooting of the siren's communication system revealed that a peripheral router dual power supply had failed at 1539 (CST) and has been replaced. All repairs completed and retested satisfactorily with proper communications confirmed with each siren. As of 1750 (CST), all sirens (were) restored to functional status. The power supply failure resulted in 2.2 hours with the sirens being unavailable. Notifications have been completed with compensatory actions by local law enforcement secured. 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'. The licensee notified the NRC Resident Inspector. |
ENS 47723 | 7 March 2012 16:00:00 | On March 01, 2012, it was discovered that a (Contractor) program manager intentionally failed to implement a procedure change as instructed by management. An initial investigation has determined that interim compensatory actions were in place prior to this pending procedural change. The involved manager's access authorization has been revoked. Southern Nuclear Operating Company, Inc. (SNC) was notified by (the Contractor) of this discovery on March 06, 2012 at 19:00 EST. SNC is providing this notification under the provisions of 10 CFR 26.719(b)(3) as an intentional act that casts doubt on the integrity of the FFD program. At the time of this report, no regulatory barriers for individuals assigned to perform duties that require them to be subject to the FFD program have been breached regarding this event. The licensee notified the NRC Resident Inspector. |
ENS 47707 | 28 February 2012 17:59:00 | At 1544 CST on February 28. 2012, it was identified that under certain fire conditions, the ability to meet the performance criteria of the approved Fire Protection Program may be challenged. Specifically the cabling within the Dedicated Fire Zone may be affected under certain fire conditions resulting in fire induced spurious operation of valves RC-46 (pressurizer/reactor head vent), PR-33A (pressurizer head vent) and RC-45A (reactor head vent). The ability to maintain pressurizer level within the indicated range may be challenged due to these spurious valve openings during these postulated conditions. Contingency actions per the Fire Protection Program to address the fire in the area of concern have been established (hourly fire watch). This condition is reported pursuant to 10CFR50.72 (b)(3)(ii)(B). Similar conditions have been previously reported by the licensee in EN #44482 and EN #47686. The compensatory measures that had been put in place for those earlier conditions remain applicable to this current condition. The licensee notified the NRC Resident Inspector |
ENS 47699 | 26 February 2012 20:15:00 | Calvert Cliffs will be performing planned maintenance to the U-1 Plant Process Computer to install isolation transformers. This maintenance window was expected to start at 2100 (EST) today, 2/26/2012, but the U-1 Plant Process Computer failed at 1546 this afternoon and it was decided to commence the planned maintenance window at that time versus spending resources to attempt recovery of the computer for just a short time period. The planned maintenance window is expected to be 54 hours long and end February 28, around 2300. This will impact the Unit 1 data dissemination to the Safety Parameter Display System (SPDS), TSC Computer. PI (Plant Trending Software) and ERDS will also be out of service for both Unit-1 and Unit-2. Should an emergency be declared during this period, the Control Room will continue to have the capability to retrieve plant data inputs to assess plant conditions and perform core damage assessment. Control Room Emergency Response Organization personnel will use backup methods already captured in emergency response procedures to disseminate plant parameter data to the effected Emergency Response Facilities and NRC during the plant data network outage. MIDAS (Meteorological Data) will continue to be operational at the site. Applicable Reporting Requirement: 10 CFR 50.72 (b) (3) (xiii); 8 Hour report. The licensee notified the NRC Resident Inspector.
At 1000 (EST) this morning planned maintenance was completed to the Unit 1 plant process computer. All plant assessment capabilities have been restored. The licensee informed the NRC Resident Inspector. Notified R1DO (Ferdas). |
ENS 47695 | 23 February 2012 14:28:00 | The Technical Support Center will be unavailable for approximately 10 hours due to planned maintenance on motor control center SMXE. Maintenance is expected to begin at 0500 on February 24, 2012. The licensee will notify the NRC Resident Inspector. |
ENS 47694 | 23 February 2012 12:52:00 | On February 23, 2012, at 0900 CST, an oil leak was identified from a gear box in the intake structure that had leaked into the circulating water cell. Prompt action was taken to stop the leak. Less than or equal to 1 pint of oil was spilled into the circulating water cell and, subsequently, to the Missouri River. The spill has been stopped and clean up of the oil sheen in the intake cell is proceeding. A visible oil sheen was noted on the Missouri River downstream of the station's intake structure near the station's outfall. The gear box has been drained to prevent any further leakage. Notifications will be made per Fort Calhoun Nuclear Station's National Pollutant Discharge Elimination System (NPDES) permit. Samples will be taken for off site analysis to determine the quantity of oil discharged. The State of Nebraska, Department of Environmental Quality and National Response Center will be notified. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi), 4 hour non-emergency notification due to a notification being made to a Government Agency (State of Nebraska, Department of Environmental Quality and National Response Center). The licensee has notified the NRC Resident Inspector. |
ENS 47687 | 22 February 2012 23:36:00 | At 1859 hours EST, the Brunswick site experienced a loss of balance of plant (BOP) bus Common C. As a result, makeup pumps to the ECCS discharge line keepfill systems lost power. At 1905 on Unit 1, 'A' loop of the Core Spray (CS) system received a low discharge pressure alarm and was declared inoperable. At 1916 hours, 'B' loop of the Residual Heat Removal (RHR) system received a low discharge pressure alarm and was declared inoperable. With the loss of the second low pressure ECCS system, Condition J of Technical Specification 3.5.1, 'ECCS Operating,' was entered, which requires the Unit 1 to enter LCO 3.0.3 immediately. At 1931 hours, 'A' loop of RHR was declared inoperable due to low discharge pressure. Power reduction of Unit 1 was initiated at 2014 hours. At 2055 hours on Unit 2, 'A' loop of the Residual Heat Removal (RHR) system received a low discharge pressure alarm and was declared inoperable. At 2128 hours, 'B' loop of the Core Spray (CS) system received a low discharge pressure alarm and was declared inoperable. With the loss of the second low pressure ECCS system, Condition J of Technical Specification 3.5.1, 'ECCS Operating,' was entered, which requires the Unit 2 to enter LCO 3.0.3 immediately. Power reduction of Unit 2 was initiated at 2219 hours. This event reportability is in accordance with 10CRF50.72(b)(2)(i), Technical Specification Required Shutdown, due to inoperability of ECCS systems. The initial safety significance of this event is minimal. Offsite power and the Emergency Diesel Generators are operable. The High Pressure Coolant Injection (HPCI) system remains operable on both Unit 1 and Unit 2. The Reactor Core Isolation Cooling (RCIC) system remains operable on Unit 1 and is being restored following maintenance on Unit 2. Troubleshooting activities to determine the loss of the BOP Common C bus are in progress. Efforts are in progress to install temporary power to the keepfill makeup pumps. The licensee will notify the NRC Resident Inspector.
Unit 1 - At 2315 hours, temporary power was provided to the ECCS keepfill makeup pump and the ECCS systems were restored. LCO 3.0.3 was exited on Unit 1 at 0041 hours with restoration of the 'A' and 'B' loops of the RHR systems. The 'A' loop of the Core Spray system was restored at 0058 hours on 2/23/2012. During the shutdown, Unit 1 was manually scrammed due to high delta-pressure across the Circulating Water Pump traveling screens. See EN #47690 for details. Unit 2 - At 2315 hours, temporary power was provided to the ECCS keepfill makeup pump and the ECCS systems were restored. LCO 3.0.3 was exited on Unit 2 at 2354 hours with restoration of 'B' loop of the RHR system. The 'A' loop of the Core Spray system was restored at 0039 hours. Unit 2 was at 96% of Rated Thermal Power when the shutdown was terminated. The licensee notified the NRC Resident Inspector. Notified R2DO (Ernstes). |
ENS 47686 | 22 February 2012 18:10:00 | At 1307 CST on February 22, 2012, it was identified that under certain fire conditions, the ability to meet the performance criteria of the approved fire protection program may be challenged. Specifically the cabling for Train A Pressurizer Power Operated Relief valve and Train B Pressurizer Power Operated Relief valve alternate circuits may be affected under certain fire conditions affecting the relay room. The ability to maintain pressurizer level within the indicated range may be challenged due to a spurious opening of a pressurizer power operated relief valve during these postulated conditions. Compensatory actions to address the fire in the area of concern are in place. This condition is reported pursuant to 10CFR50.72(b)(3)(ii)(B). A similar condition was previously reported (by the licensee) in EN #44482. The compensatory measures that had been put in place for that earlier condition remain applicable to this current condition. The licensee notified the NRC Resident Inspector. |
ENS 47698 | 24 February 2012 16:54:00 | The following information was received from the State of Texas via email: On 2/22/12 the licensee notified the Agency (Texas Department of State Health Services) of a medical event that occurred on 2/16/12. The licensee reported that the wrong radiopharmaceutical had been administered to a patient. Tc-99m MAA was administered rather than the prescribed Tc-99m Pertechnitate. During the administration of the wrong drug, an estimated 1/3-1/2 of the dose infiltrated the skin causing an estimated 400 rad exposure subcutaneously. The referring physician and patient have been notified. No signs or symptoms have been observed in the patient as a result of the incident. The estimated exposure is being reevaluated by the licensee. Texas Incident: I-8937 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. |
ENS 47693 | 23 February 2012 11:08:00 | 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. |
ENS 47657 | 10 February 2012 07:31:00 | On February 10, 2012, with Unit 1 in Mode 5, while performing scheduled maintenance, a technician inadvertently made contact with a component that caused an undervoltage condition on an essential bus, resulting in the automatic start and loading of the 1B Emergency Diesel Generator (EDG). Prior to the event the 1B EDG was inoperable and not required by Technical Specifications; however, the 1B EDG was available. All equipment responded as expected. Currently maintaining the plant in Mode 5. Decay heat removal is being supplied by the 1A Shutdown Cooling train and was never interrupted. There was no impact on the Shutdown Safety Assessment. Unit 2 was unaffected and remains in Mode 1 at 100% power. This event is reportable pursuant to 10CFR 50.72(b)(3)(iv(A). The licensee notified the NRC Resident Inspector. |
ENS 47659 | 11 February 2012 10:32:00 | The following information was received by email: On February 10, 2012, the Agency (Texas Department of State Health Services) was notified by the licensee that the shutter on an Ohmart model SH-LG 2 nuclear gauge containing 8.5 curies of Cesium-137 was found to be stuck in the open position. Open is the normal position for the shutter and the failure does not pose any additional exposure risk. The licensee is trying to lubricate the operating arm in an effort to free the shutter. The licensee has contacted the manufacturer for assistance. Additional information will be provided in accordance with SA-300. Texas Incident: I-8932 |
ENS 47652 | 9 February 2012 09:37:00 | A non-licensed employee supervisor had a confirmed positive for alcohol during random testing. The employee's access to the plant has been terminated and his badge deactivated. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. |
ENS 47632 | 3 February 2012 08:22:00 | LOSS OF PLANT COMMUNICATIONS DUE TO SCHEDULED MAINTENANCE
"Calvert Cliffs will be implementing scheduled maintenance to the plant data network to install a data diode to meet the new cyber security requirements listed in 10 CFR 5.71. This work will require the TSC, OSC and subsequently the EOF to lose normal data flow from the plant data network for a period of approximately 6 to 8 hours. ERDS will also be unavailable during this maintenance. Should an emergency be declared during this period, the Control Room will continue to have the capability to retrieve plant data inputs to assess plant conditions and perform core damage assessment. Control Room Emergency Response Organization personnel will use backup methods already captured in emergency response procedures to disseminate plant parameter data to the effected Emergency response Facilities and NRC during the plant data network outage. MIDAS (Meteorological Data) will continue to be operational at the site." The licensee notified the NRC Resident Inspector. |
ENS 47622 | 25 January 2012 21:39:00 | Planned preventive maintenance activities are being performed on the Braidwood Nuclear Station Technical Support Center (TSC) Ventilation System. These work activities are planned to be performed and completed expeditiously within 8 hours. This maintenance activity includes the performance of preventive maintenance on the TSC outside air supply fan unit which affects the TSC emergency filter train and air handling unit. During a portion of the time these activities are being performed, this equipment will not be available for operation. As such, the TSC Ventilation will be rendered non-functional during the performance of portions of the work activity. If an emergency condition occurs that requires activation of the Technical Support Center, during the time this work activity is being performed, it will take no more than 4 hours to return the equipment back to functional status, dependent on the stage of the work activity at the time an emergency occurs. Plans are to utilize the TSC for any declared emergency during the time this work activity is being performed as long as radiological conditions allow. 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 licensee notified the NRC Resident Inspector.
Braidwood Nuclear Station TSC ventilation was restored to available status at 0635 CST on January 26, 2012. The previously reported system preventative maintenance has been completed. The licensee notified the NRC Resident Inspector. Notified R3DO (L. Kozak). |
ENS 47599 | 17 January 2012 13:04:00 | 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.
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)
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. |
ENS 47595 | 14 January 2012 21:37:00 | Salem Generating Station made a 15 minute notification of a chemical discharge to the State of New Jersey Department of Environmental Protection at 1806 (EST). The Salem Non-Rad Waste Chemical Treatment Building sump overflowed out of the building to a catch basin that discharges to the Delaware River. Approximately 100 gallons of chlorinated water was reported to the State of New Jersey as being discharged, which was terminated at 1806 (EST). There were no personnel injuries associated with this event. There was no impact to any Salem Station Safety-Related systems and all Safety-Related systems are available. Investigation into the cause of the event is in progress. The licensee notified the NRC Resident Inspector and Lower Alloways Township.
On 1/14/2012 Salem Generating Station made a 15 minute notification of a chemical discharge to the State of New Jersey Department of Environmental Protection at 1806. The Salem Non-Rad Waste Chemical Treatment Building sump overflowed out of the building to a catch basin that discharges to the Delaware River through DSN 488. Subsequent investigation has indicated that the spill was entirely contained within the onsite storm drainage system and that there was no discharge to the Delaware River. The storm drain system was plugged, flushed and pumped out for appropriate disposal of the waste water. Additionally, the spill to the ground was cleaned up within 24 hours. The licensee will notify the NRC Resident Inspector and has notified the New Jersey Department of Environmental Protection. |
ENS 47593 | 13 January 2012 18:27:00 | A patient of Memorial Hermann in Houston received the wrong radioisotope. A dose of Gallium-67 was ordered, but a dose of Thallium-201 was delivered. Because the dose was improperly labeled as Gallium-67, the dose calibration process indicated an acceptable radioisotope and dose. The patient was injected with the wrong radioisotope on January 11, 2012. During patient imaging on January 13, 2012, it was realized that the patient received the wrong radioisotope. The pharmacy was notified of the error and admitted to delivery of the wrong isotope. The physicist at the hospital estimates that the patient received a dose of about 6 REM whole body. Texas Incident Number: I-8921
The following information was received as an update: The licensee called to report that the wrong isotope was administered to a patient. Thallium 201 had been injected in a patient instead of Gallium 67 that was ordered. Apparently, the pharmacy sent the wrong isotope. 8 mCi of Gallium was ordered and a estimated 4.7 mCi of Thallium was delivered. Dose activities were similar and the dose calibrator didn't pick up the difference in isotope. Patient was injected on 1/11/12 and imaged 1/13/12. The mistake was discovered in imaging. The licensee will file a written report within 15 days. Notified R4DO (Pick) 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. |
ENS 47592 | 13 January 2012 15:35:00 | The following information was received via facsimile: (The) Environmental Safety Manager of Louisiana Scrap Metal Recycling notified the (Louisiana) Department of Environmental Quality of a lost Niton XLI 818 Q/serial number 6066 analyzer with 30 mCi of Am-241 on January 6, 2012. FINDINGS: The investigation was conducted at Louisiana Scrap Metal Recycling in Lafayette. Contact was made with the Environmental Safety Manager and the (Non-Ferrous Manager), who provided the following information of their internal investigation. The facility's investigation revealed that on January 5, 2012, (Employee #1) was scanning a customer's trailer to identify the type metal in the load, with a Niton XLI 818Q/serial number 6066 analyzer with 30 mCi of Am-241. The analyzer was discovered missing the next morning, January 6, 2012 by (Employee #2) at approximately (0900 CST and notified the department at 1048 CST on January 6, 2012 of the missing source. (Employee #2) called (Employee #1) to try to reconstruct his actions of the previous day since he had signed out the analyzer on the utilization log. (Employee #1) stated that he believed that he must have left the analyzer on the trailer of a customer. (Employee #2) called the customer to ask if he had found the analyzer and also the route he took to go home. The customer stated that he had not found the analyzer. (Employee #2) then retraced the route to search for the missing analyzer for approximately eight hours but was not successful. The analyzer is still missing to date, however does not pose a health hazard to the general public. In conclusion the licensee did not secure licensed radioactive material from unauthorized removal or access. The above area is contrary to LAC 33:XV.445.A. The licensee failed to maintain constant surveillance to prevent unauthorized use of licensed radioactive material that is in a controlled or unrestrictive area. The above area is contrary to LAC 33:XV.445.B. The licensee also failed to have the minimum of two independent physical controls that form a tangible barrier to secure portable gauges from unauthorized removal, whenever portable gauges are not under the control and constant surveillance of the licensee. The above area is contrary to LAC 33:XV.326.B. Louisiana Incident Number: LA120002 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 |
ENS 47569 | 5 January 2012 05:18:00 | A section of piping between sodium hypochlorite tanks(tanks 121 and 122) broke and resulted in a unisolable leak in one of the tanks. The spill was contained within the berm of the chlorine house. The spill resulted in airborne concentrations exceeding Immediately Dangerous to Life and Health (IDLH) conditions in the chlorine house, which is contiguous to the plant screen house. Alert declared based upon HA3.1 - Report or detection of toxic gases within or contiguous to Table H-1 areas in concentrations that may result in an atmosphere Immediately Dangerous to Life and Health (IDLH). The licensee will remain in the Alert until IDLH conditions have been terminated. An offsite HAZMAT team has been notified and is expected to be on-site today at 0630 CST to clean up the spill. The licensee will notify the Resident Inspectors.
The licensee reports that the HAZMAT contractor has arrived on-site. Licensee personnel are preparing a clearance for the sodium hypochlorite system in order to support HAZMAT contractor work. The licensee will notify the NRC Resident Inspector. Notified the R3DO (Valos)
The Alert declaration due to HA3.1, Report or detection of toxic gases within or contiguous to Table H-1 areas in concentrations that may result in an atmosphere Immediately Dangerous to Life and Health (IDHL): Prairie Island terminated from the Alert declaration at 1408 CST on Jan 5, 2012. The following additional conditions were established prior to termination: 1. The permanent berm containing the sodium hypochlorite released by 121 Sodium Hypochlorite Tank piping break was pumped down. 2. Clean-up and remediation of the affected area is in progress. 3. Inspection of the unaffected sodium hypochlorite tank piping has been conducted to ensure satisfactory integrity. NRC Resident has been informed. Notified NRR EO (Ruland), R3DO (Valos), IRD (Morris), DHS (Beach), FEMA (Blankenship), DOE (Jackson), USDA (Timmons), and HHS (Fajardo). |
ENS 47591 | 13 January 2012 15:34:00 | The following information was received by facsimile: This is a report of a transportation incident where an Industrial Radiography Camera was not properly secured. The RSO stated that two radiographers, one an instructor, conducted a radiography job in Bay St. Louis, Mississippi. On January 4, 2012, while returning to the Baton Rouge, LA office, they decided to meet with another radiographer who is an instructor on the radioactive materials license for Mistras and one of the other radiographer's father. The father offered to take the radiography camera to the office in Baton Rouge, LA. The instructor from the radiography job in Bay St. Louis agreed, but did not realize that the father did not have his radiography truck. The father put the camera in the trunk of his personal vehicle unsecured and unbraced. After noticing that the father and son did not have some of the required paperwork, the instructor pursued them. Approximately 2 miles down the road, the father and son in the same vehicle were pulled over for speeding. After they were pulled over, the son, who was driving, was suspected of intoxication and tested. The RSO received a call from the father regarding the impending arrest of the son for DUI at (2130 CST). The son was arrested for DUI and the instructor from the radiography job secured the radiography camera. The father was arrested for outstanding warrants. Both the father and son were suspected of being under the influence, but the father refused to be tested by law enforcement. The RSO arrived to the site at (2200 CST). The camera was placed in the Mistras storage vault around (0030 CST on January 5, 2012). Mistras is conducting an internal investigation. Louisiana Department of Environmental Quality is investigating. So far, the son's employment has been terminated. The father's Trustworthy and Reliability status has been suspended. All radiographers will be drug tested. Additional information will be forthcoming. Louisiana Incident Number: LA120001 |
ENS 47549 | 23 December 2011 10:10:00 | At 0610 CST on 12/23/2011 with the reactor at 100% power, River Bend Station experienced an (automatic) reactor scram resulting from a RPS actuation. Following the scram, reactor water level briefly lowered below level 3, resulting in the automatic closure of containment isolation valves in the suppression pool cooling system. This isolation was confirmed to have occurred as designed. The reactor is stable with pressure and temperature being controlled by the feed water system and main steam bypass valves, respectively. The cause of the scram was due to a turbine trip. Initial indications are that the turbine tripped due to a loss speed sensor. All control rods inserted and Reactor Core Isolation Cooling was manually operated for approximately 1 minute and secured, The plant is conducting causal investigations to fully understand the cause of the turbine trip. As information becomes available River Bend Station will provide additional information. This event is being reported in accordance with 10CFR50.72(b)(iv)(B) as an automatic RPS actuation with the reactor critical. The safety relief valves momentarily lifted immediately following the scram. The plant electrical distribution system is in a normal shutdown configuration. The scram was uncomplicated. The licensee notified the NRC Resident Inspector. |
ENS 47548 | 23 December 2011 09:41:00 | The following information was received by e-mail: On December 22, 2011, at 2102 hours, the Agency was notified by Harris County Hazmat that a radiography truck had been involved in an accident. The truck was carrying a Spec 150 camera serial # 153 containing a 32 curie Iridium 192 source. The radiographer informed the Hazmat officer that the source in the camera was not in the locked position. The radiographer stated that the source had separated from the drive cable and he had used a pair of pliers to insert the source into the camera backwards to shield it. The camera opening for the guide tube connection was covered with duct tape to prevent the source from coming out of the camera. The radiographer could not get the crank out device to separate from the camera so the camera and the crank out devices were placed between the transport container and the darkroom wall to transport them back to the storage location. The accident caused the source to move further to the back of the 'S' tube in the camera and the dose rates from the camera increased slightly. The crank out device was used to push the source back to the shielded position. Lead sheets were placed above the camera to help decrease the dose rates from the camera. The dose rates at this point were between two and four millirem at one meter. The radiographer took the camera and source to the licensee's storage location. The source was removed from the radiography camera and placed in a source changer shield. The radiographer stated that he had received 180 millirem for the event. No member of the general public received exposure due to this event. The licensee will return the source and camera to the manufacturer for inspection. Additional information will be provided as it is received in accordance with SA-300. Texas Incident: I-8914 |
ENS 47546 | 22 December 2011 09:56:00 | On December 22, 2011, at 1000 EST, due to pre-planned refueling outage maintenance, the Control Room Emergency Ventilation system on St. Lucie Unit 1 is inoperable. The Unit 1 Control Room Emergency Ventilation system provides filtered air to the Technical Support Center (TSC) ventilation system, therefore, the TSC ventilation system has been rendered unavailable during the course of the work activities. The Control Room Emergency Ventilation system is expected to be returned to service in 38 hours, and TSC filtered ventilation restored. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures. Should the TSC become uninhabitable, the TSC staff will relocate to an alternate TSC location in accordance with applicable site procedures. This notification is being made in accordance with 10CFR 50.72 (b)(3)(xiii) due to the potential loss of an emergency response facility (ERF). An update will be provided once the Control Room Emergency Ventilation system has been restored to normal operation. The NRC Resident Inspector has been notified. Unit 1 is currently de-fueled.
Update 12/23/2011, 0418 EST: The U1 Control Room Emergency Ventilation has been restored to operable (status) and the TSC filtered ventilation has been restored (to operable status). The licensee notified the NRC Resident Inspector. Notified the R2DO (McCoy). |
ENS 47523 | 14 December 2011 16:40:00 | The reactor was manually tripped at 1510 EST on 12/14/11 due to loss of both main feedpumps. Both feedpumps tripped on low suction pressure due to an apparent unplanned opening of the 'A' main feedpump recirculation valve. The cause of the main feedpump recirculation valve opening has not been determined. All full length control rods fully inserted. Auxiliary feed pump P-8A automatically started at 1511 EST on steam generator level as designed (10CFR50.72(b)(3)(iv)(A)). The turbine bypass valve is in service maintaining reactor coolant system temperature (by directing steam flow to the main condenser). The plant is stable in mode 3 (and the reactor trip was considered uncomplicated). The Van Buren County Sherriff was notified (per other plant requirements) concerning use of the atmospheric steam dump causing excessive noise in the vicinity of the plant (immediately following the plant trip). The plant electric power is in the normal shutdown configuration. There was no primary to secondary leakage. A press release is planned for the local media. The licensee notified the NRC Resident Inspector. |
ENS 47522 | 13 December 2011 19:43:00 | A degraded fire barrier between Fire Area (FA) 118 (Bus 26 Room) and FA 128 (Bus 27 Room) has existed during the last three years. The top of the wall between the Bus 26 and Bus 27 room had a missing/degraded fire barrier. At the time of discovery on December 9, 2011 a fire watch was in place. However, it was determined that at various times during the last three years a fire watch was not established. Bus 27 has been conservatively aligned to Bus 26 to provide the required degree of separation for redundant safe shutdown trains (between Bus 26 and Bus 25). The fire watch will remain in place as a compensatory measure until the fire barrier is repaired. (The) NRC Resident (Inspector) has been informed.
An eight hour report per 10 CFR 50.72(b)(3)(ii)(B) was reported on December 13, 2011 for a degraded fire barrier between Fire Area (FA) 118 (Bus 26 Room) and FA 128 (Bus 27 Room.) 'Subsequent engineering analysis determined that the degraded fire barrier maintained the required degree of separation for redundant safe shutdown trains and plant safety was not significantly degraded. The 10 CFR 50.72(b)(3)(ii)(B) report is retracted. The NRC Resident Inspector has been informed. Notified R3DO (L. Kozak) |
ENS 47518 | 12 December 2011 12:31:00 | The following information was provided via email: On 12/8/11 at 1430 PST, the Nevada Radiation Control Program (NRCP) received a call from the Alternate Radiation Safety Officer (ARSO) for Renown South Meadows Medical Center, that a truck carrying linen had been detained by the California Highway Patrol (CHP) at the Truckee inspection station after setting off a radiation alarm. The ARSO explained that the truck was not carrying any materials from the Nuclear Medicine Department or decay room and she knew of no reason that it would be radioactive. She provided the name and number for the CHP at Truckee. The NRCP contacted the CHP and were told that they had surveyed the truck with a Ludlum meter and measured .400 millirem per hour (or 400 microR/hour) outside the trailer. The CHP said that their protocols dictate that anything above three times background is treated as a hazmat incident and must have proper packaging and manifest. They had not run an identification spectrum on the truck. They had no capability to unload a trailer on site for further investigation. They were holding the vehicle pending instructions from their departmental radiation specialist and agreed to call (the NRCP) when they had a decision. The Aramark (linen service) representative confirmed that the vehicle in question was a tractor trailer and was carrying only linen from Renown. At approximately 1545 PST, the CHP called and indicated that they had released the trailer with orders to return to Renown Medical Center. They were unable to identify the radioisotope present and the dose rate reading was now .100 millirem per hour (100 micro R/hr). The NRCP Incident Response Supervisor arrived at Renown Medical center at approximately 1645 PST, shortly after the truck, and met with Renown staff. After conducting a radiation survey on the outside of the truck, he determined the general location of the high radiation and identified Technetium-99m (Tc-99m) as a suspected isotope. Linens are transported in large plastic bins, which were removed with a pallet jack, by the staff at Renown. When the radioactive bin was identified, it was segregated, a thorough survey was done and the radioisotope was positively identified as Tc-99m. The bin was placed in Renown's decay room where it will remain for 2 to10 days until the Tc-99m decays. The most likely cause of this contamination was that a recently treated patient soiled the sheets and the possibility of radioactive contamination was not recognized by floor staff. Linens are not routinely screened for radiation. Aramark recently began transporting linens to Sacramento, rather than processing locally, so they are now subject to inspection upon entering California. The RCP will assist Renown with some procedure changes that will minimize the chances of this happening in the future. Dose rate readings were one order of magnitude lower than what CHP reported. It is unknown why CHP was unable to obtain an accurate identification of the isotope. Radiation readings: RadEye - Background 6 uR/hr; at bin 48 uR/hr; Ortec MicroDetective - Background 65 counts per second; at bin 1800 counts per second; Renown 451P Ion Chamber - Background 6 uR/hr; at bin 52 uR/hr; Ludlum 14c with pancake G-M detector - at bin 1800 counts per minute. Nevada Report Number: NV110024 |
ENS 47496 | 3 December 2011 16:53:00 | On December 3, 2011 during operator rounds, it was discovered that an Enclosure Building (secondary containment) door sweep had failed and was not providing an adequate seal. Technical Specification (TS) 3.6.5.2, 'Secondary Containment Enclosure Building', is applicable in Modes 1,2,3, and 4 was entered at 1235 (EST). Since secondary containment was rendered inoperable, Dominion is reporting this as a condition that could have prevented the fulfillment of the safety function to control the release of radioactive material. The door has subsequently been repaired, the Enclosure Building declared OPERABLE and the TS Action Statement exited at 1524 (EST). Further engineering review will be conducted to more fully evaluate the impact on radiological controls. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(v)(C). The licensee has notified the Connecticut Department of Environmental Protection, surrounding towns and the NRC Resident Inspector. |
ENS 47494 | 2 December 2011 13:21:00 | A non-licensed employee supervisor had a confirmed positive drug test during random testing. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee informed the NRC Resident Inspector. |
ENS 47493 | 1 December 2011 16:05:00 | At 1344 hours (EST) on December 1, 2011, the Control Building Instrument Air Dryer failed resulting in loss of control air. As a result, the three Control Room Air Conditioning subsystems required by Technical Specification (TS) 3.7.4, 'Control Room Air Conditioning (AC) System,' and the two Control Room Emergency Ventilation (CREV) subsystems required by TS 3.7.3, 'Control Room Emergency Ventilation (CREV) System,' became inoperable. As a result, this condition could have prevented the fulfillment of the safety function for these systems. Because Brunswick has a shared control room, Unit 1 and Unit 2 entered TS 3.7.3 Required Action C.1, for two CREV subsystems inoperable (i.e., be in Mode 3 within 12 hours) and TS 3.7.4, Required Action E.1, for three Control Room AC subsystems inoperable (i.e., enter LCO 3.0.3 immediately). Operability of two Control Room AC subsystems and one CREV subsystem was restored and LCO 3.0.3 was exited, at 1410 hours, when the Instrument Air Dryer was bypassed. No power reduction took place as a result of the LCO 3.0.3 entry. This report applies to both Units 1 and 2 and is being made in accordance with 10 CFR 50.72(b)(3)(v)(D), as a condition that at the time of discovery could have prevented fulfillment of the safety function of systems that are needed to mitigate the consequences of an accident. The safety significance of this event is considered minimal. The condition existed for approximately 26 minutes. Plant staff took immediate actions to return the equipment to service. For the brief time the Control Room AC and CREV systems were inoperable, performance of plant personnel and equipment in the Control Room was not adversely affected. The maximum Control Room back panel temperature during this event was approximately 68 degrees F. Troubleshooting activities are under way to determine the cause of the Instrument Air Dryer failure. The licensee notified the NRC Resident Inspector. |
ENS 47460 | 20 November 2011 02:10:00 | While performing a regularly scheduled Turbine Bypass Valve surveillance, prior to Turbine Bypass Valve movement, a 'B' half scram (signal) was received. Operators immediately suspended testing. Approximately 10 seconds later, a full Reactor Protection System actuation occurred. Following the reactor scram, reactor water level lowered below the Group II isolation initiation setpoint of +9 inches, (resulting in containment valve isolations). There were no radioactive releases associated with this event. No other alarms were received prior to the RPS actuation. The cause of the reactor scram is under investigation at this time. Also, due to the reactor scram, discharge canal temperature rate of change exceeded plant requirements. As a result, the State of Minnesota, and appropriate local agencies will be notified. All control rods inserted and the scram is considered uncomplicated. The plant is in a normal shutdown electrical configuration. The licensee notified the NRC Resident Inspector. |
ENS 47430 | 11 November 2011 10:01:00 | This event is being reported because a non-licensed contractor employee's direct supervisor authorized the individual to return home instead of initiating a for cause Fitness For Duty test. Both the direct supervisor and the contractor employee had their unescorted access to the protected area terminated. Contact the Headquarters Operations Officer for additional details. |
ENS 47427 | 10 November 2011 00:59:00 | 'This telephone notification is provided in accordance with 10CFR50.72(b)(3)(xiii), loss of emergency assessment capability. On 11-10-11 at 0100 Central Standard Time, the Technical Support Center Emergency Diesel Generator and Technical Support Center regular lighting will be taken Out of Service to support a scheduled temporary power supply return to normal configuration. The temporary power had been in place since the last refueling outage in support of planned bus maintenance. The Technical Support Center Emergency Diesel Generator and regular lighting are scheduled to be restored at 1100 (CST) on 11-10-11. The licensee has notified the Senior Resident Inspector of this scheduled work. |
ENS 47547 | 23 December 2011 03:03:00 | This 60-day telephone notification is being made per the reporting requirements specified in 10 CFR 50.73(a)(2)(iv)(A) and 10 CFR 50.73(a)(1) to describe an invalid actuation signal affecting containment isolation valves in more than one system. On November 7, 2011, Nine Mile Point Unit 2 (NMP2) received a Division 2 reactor building pipe chase high ambient temperature isolation signal, which resulted in closure of isolation valves in the reactor water cleanup (RWCU) system, the reactor core isolation cooling (RCIC) system, and the residual heat removal (RHR) system (isolation valve groups 5, 6, 7, and 10). All affected isolation valves responded as designed. The Division reactor building pipe chase high ambient temperature isolation signal was generated by a new NUS/Scientech ambient temperature indicating switch that had recently been installed as a replacement for the original Riley temperature switch, which is no longer being manufactured. Two temperature switches (one for Division 1 and one for Division 2) monitor the reactor building pipe chase area to detect a rise in area temperature, which is indicative of a leak in the RWCU, RCIC, or RHR system piping that exists in the area. At the time of the event, operations personnel confirmed that conditions requiring isolation of the RWCU, RCIC, and RHR systems did not exist, based on a check of the Division 1 reactor building pipe chase high ambient temperature channel and area radiation monitors. Therefore, the isolation signal was determined to be invalid. The NUS/Scientech temperature switch was subsequently removed and the original Riley temperature switch was re-installed. The apparent cause for generation of the trip signal from the new temperature indicating switch was determined to be the presence of signal noise that was not adequately filtered. This event was entered into the (Nine Mile Point) corrective action system as Condition Report (CR) 2011-010062. There were no safety consequences or impact on the health and safety of the public as a result of this event. The licensee notified the NRC Resident Inspector. |
ENS 49282 | 15 August 2013 14:05:00 | The following information was received by email: Based on protocol, a dose of 120 Gy (1.79 GBq) was prescribed. Upon completion of the treatment, survey of the Nalgene waste container measured higher than expected. Ensuing calculations resulted in dose delivered to be 85 Gy (1.24 GBq); greater than 20% variation from prescribed dose. All drapes, towels, etc were surveyed with no evidence of radioactivity present, therefore assuring no contamination present. Contents of the waste container were measured separately to locate the source of residual activity. The readings indicated minimal activity in the Y -90 vial; readings of the patient delivery microcatheter were indicative of residual microspheres. The treatment protocol was followed with no variations of procedure. As is typical, 3 saline flushes were made of the catheter including several vigorous flushes to dislodge any microspheres as recommended by Nordion, the product manufacturer. No high pressure was detected at any point during infusion which would trigger the pressure valve and deliver saline in the overflow vial. There was no build up of particles in the hub of the delivery catheters as inspected throughout the procedure. Measurements over the length of the catheter revealed greatest activity in the proximal portion of the catheter with little-to-no activity in the tip. Nordion has been contacted. In the future, survey of the catheter prior to disconnecting it for disposal may help detect the build-up of particles. Oregon Incident: 11-0037 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. |
ENS 47403 | 2 November 2011 20:41:00 | Ongoing field inspections as a result of Industry Operating Experience have identified certain fire barriers that are not in conformance with required tested configurations. Specifically, some of the flexible conduit that penetrates these barriers has a coating that does not exhibit flame retardant characteristics and therefore does not meet the requirements for fire barriers at Beaver Valley Power Station, Unit 1. In the event of a postulated fire, this non-conformance has the potential to affect fire barriers separating the two independent trains required for post fire Safe Shutdown equipment. This issue is being reported per 10CFR 50.72(b)(3)(ii)(B). Compensatory actions have been established in accordance with the approved Fire Protection Program. The NRC Resident Inspector has been notified. |
ENS 47401 | 1 November 2011 18:08:00 | On November 1, 2011 at 1450 PDT, San Onofre Unit 3 declared an ALERT and entered EAL HA3.1 due to an ammonia leak that prevented access to local areas. The plant is in a stable condition while the leakage is being secured. The turbine building on Unit 3 has been evacuated. Plant personnel are in the process of verifying no presence of ammonia in the turbine building. There was no impact on Unit 2.
At 1807 PDT licensee exited the ALERT and EAL HA3.1. The leak was at the Ammonia day tank and was flowing through an overflow vent into the berm. The high level in the Ammonia day tank was due to a leaking closed valve between the Demineralizer system and the ammonia bulk storage. The berm area was drained of all fluids. The wind direction caused the ammonia fumes to travel to the Unit 3 turbine deck. No off-site HAZMAT personnel came on-site. At 1756 PDT the precautionary evacuation of on-site personnel was terminated. Unit 2 was not affected from this event. The NRC Resident Inspector was notified. Notified NRR EO (Fredrick Brown), R4DO (Vincent Gaddy, IRDMOC (Jeff Grant), DHS (Hill), DOE (Doyle), USDA (Krauf), FEMA (Fuller) and HHS (Fajardo).
On November 1, 2011, Southern California Edison notified the California Emergency Management Agency at 1755 PDT and the San Diego Department of Environmental Health at 1810 PDT that approximately 25 gallons of Ammonium Hydroxide was spilled under the Ammonium Hydroxide day tank located outside the Unit 3 turbine building. The spill was contained in a berm under the tank and subsequently cleaned up. Both Units 2 and 3 were at approximately 100% power at the time of the event. Notified the R4DO (Gaddy). |
ENS 47398 | 1 November 2011 13:56:00 | A person working in an office area began feeling excessive heartburn and nausea. The individual was taken to the site dispensary for evaluation. An ambulance was requested to take the employee to the hospital for further evaluation. The employee had contamination on his coveralls and boots. Before leaving the site for the hospital, the employee's plant clothing was removed and the individual was surveyed. There was no detectable contamination on the employee when he was transferred to the hospital. Isotope, Quantities and Chemical Form: Uranium Ore Concentrates, U308 NRC Region II informed: Richard Gibson- Senior Fuel Cycle Inspector |
ENS 47397 | 1 November 2011 12:34:00 | An outgoing shipment of scrap metal from the Yaffe Iron and Metal Company detected a radioactive source when going through the monitoring process. It is thought an orphaned radioactive source entered the scrap metal yard with an unmonitored shipment of aluminum. Upon further investigation, a radioactive metal rod of about 15 inches long was discovered in the outgoing shipment of scrap metal. Initial readings indicate a dose about a 200 mRem at 2 inches from a metal box that contains the source. Based on the use of a G-M detector, the activity is estimated to be 3.75 Ci. The metal box is constructed of one quarter inch steel. Initial portable gamma spectrometry indicates the source is Radium-226. The source is currently locked in the metal box. The State of Oklahoma is currently on the scene investigating and will determine a list of potential individuals who may have been exposed to the source. A preliminary assessment has determined that one individual received about 600 mRem to the hand. 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 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 |
ENS 47399 | 1 November 2011 14:52:00 | The following information was received by facsimile: On Friday, October 28th, two NVI (Nondestructive and Visual Inspection, LLC) employees were performing radiography on a pipeline project in Wyalusing, PA. While performing radiography on a main line the crew approached the pipe after cranking in the source to set-up and mount for their next shot. While strapping the next weld with film, one of the crew noticed the indicator which shows full retraction of the source on their Amersham Model D880 had not popped out. At this time both crew members confirmed their survey meters read zero. However, also at this point they realized one member's rate alarm was chirping, but not very loudly and the other's rate alarm was not chirping at all. It was noted both rate alarms were inspected and working properly at the beginning of the shift. The crew then approached the crank controls where one was able to make approximately one turn with the crank, fully retracting the source back into the camera. They inspected their dosimeters which were both off-scale. They informed the RSO and were immediately removed from work. The badges were sent for emergency processing and whole body dosimetry results were 5133mR and 1447mR. CAUSE OF THE EVENT: Undetermined at this time, expected faulty equipment. ACTIONS: The licensee will be submitting a written report within 30 days. The Department (PA DEP Bureau of Radiation Protection) plans to do a reactive inspection. Pennsylvania Event Report: PA110032 |
ENS 47400 | 1 November 2011 16:06:00 | The following information was received by e-mail: On 10/31/11, the ARSO (Alternate Radiation Safety Officer) at TC Inspection informed RHB (California Radiation Health Branch) via email of an incident occurring on 10/26/11 at Valero Refinery in Benicia, CA during one of their radiography operations. The email written by the ARSO is as follows: On October 26, 2011, there was an incident involving RAM material; one of (the licensee's crew was) performing radiography at the Valero refinery when, while cranking out the source, the trainer noticed the crank handle started free-spinning. When he tried to crank the source back in it was still free spinning so the source was stuck out of the shielded position. When the trainer called, (the licensee) advised him to loosen one of the nuts on the crank assembly, pull back the tube and then grab the cable and pull the source back into the exposure device and into the shielded position and that (the licensee was on his way). When (the trainer) did this he noticed that the end of the cable was inside the tube, he was able to grab it with a pair of needle nose pliers and retrieve the source back into the shielded position. Two things happened here, the first; the trainer or assistant (still not sure which one) did not fully connect the guide tube to the camera. This allowed the source and cable to go out of the camera into air, thus allowing the cable to reach the end where the stop at the end of the cable did not stop the cable from coming out of the crank assembly. After further investigation (the licensee) found that the aluminum body of the crank assembly was worn right at the exit hole thus allowing the stop to go through. (The licensee) just did a maintenance inspection on those cranks on 10/1/11 and saw some wear on it but not as much as was there this time. (The licensee has) been in the process of replacing the aluminum body on all of (the licensee's) INC crank assemblies with stainless steel bodies when the techs tell (the ARSO) their cranks are getting hard to crank (That is usually the first sign that the aluminum body is wearing). (The apparent cause of the event is a technician forgetting to connect all of the equipment pieces due to production pressures or) equipment failure. CA 5010 Number: 103111
The following was received via email: On the day of the event, the operators pocket dosimeters indicated 10 mR. Camera information: INC IR-100, S/N 4301, with a source activity of 40.8 Ci. The crank assembly has been sent to INC and we'll be visiting INC today. We requested written statements from the trainer assistant. The ARSO already received them and he'll be sending a copy to RHB today. TC was requested to process the dosimetry badges worn by trainer and the assistant. Notified R4DO (Gaddy) and FSME EO (Camper). |
ENS 47372 | 24 October 2011 20:14:00 | At 1715 hours on 10/24/2011, replacement of a router and circuit at the Emergency Operation Facility (EOF) commenced. This action will cause a temporary outage of network devices such as personal computers, telephones, printers, routers, switches, alarm panels, and wireless communications. The duration of work is expected to be approximately 8 hours. If use of the EOF is subsequently required, this condition may delay activation of the EOF until equipment is restored. Restoration of the equipment will take less than one hour. This is considered a Loss of Emergency Assessment Capability and reportable under 10CFR50.72(b)(3)(xiii). The licensee notified the State of Pennsylvania, Luzerne and Columbia Counties and the NRC Resident Inspector.
Systems have been restored and the EOF is operable. The licensee has notified the NRC Resident Inspector. Notified the R1DO (DeFrancisco). |
ENS 47370 | 24 October 2011 17:24:00 | On 10/24/11 at 1035 CDT, a maintenance technician performing a field walk-down of a clearance order inadvertently opened the bus potential fuse drawer for Unit 2 Safety Related Bus 23. This caused all loads of Bus 23 to receive an undervoltage (UV) load shed signal. Division I Containment Cooling Service Water (CCSW) pumps are powered from Bus 23 and are currently aligned to the Main Control Room Emergency Ventilation (CREV) Air Conditioning system. On a loss of CCSW, the CREV Air Conditioning system is inoperable due to a loss of its emergency cooling water supply. This condition affects both Dresden Units 2 and 3. This condition lasted for approximately 7 seconds before the bus potential fuse drawer was reinserted and the Bus 23 UV condition cleared. This event is reportable under 10CFR50.72(b)(3)(v)(D) as an 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. The licensee notified the NRC Resident Inspector. |
ENS 47366 | 21 October 2011 20:25:00 | This event is being reported in accordance with 10CFR50.72(b)(2)(xi). On October 21, 2011, at approximately 1723 hours EDT, notification of a fuel oil spill was made to the US Environmental Protection Agency (EPA), National Response Center. At the time of the event, the plant was in Mode 1 at 100 percent rated thermal power. The fuel oil spill was caused by a leak in an underground pipe outside the protected area, but inside the owner controlled area. The spill is estimated between 1000 to 1500 gallons and is contained onsite. However, a certified wetlands specialist was contacted and determined the area met the criteria for wetlands designation, which in turn made the event reportable. Clean Harbors Incorporated is assisting with the onsite clean-up and remediation. Additionally, the Ohio EPA: State Emergency Response Commission, Perry Township Fire Department, Lake County Emergency Planning Committee, and the U.S. Coast Guard were notified in accordance with plant procedures. This event is also being reported in accordance with the plant's Operating License, Appendix B, Environmental Protection Plan, which states, in part, that any occurrence of an unusual or important event that indicates or could result in significant environmental impact causally related to plant operation shall be recorded and reported to the NRC within 24 hours followed by a written report. The licensee notified the NRC Resident Inspector. |
ENS 47364 | 21 October 2011 16:42:00 | The station experienced a lockout of the 2R Auxiliary Power Transformer. The resulting transient caused an automatic actuation of the RPS system. All control rods fully inserted. A Group 2 Primary Containment isolation occurred. Both 11 and 12 Emergency Diesel Generators started on a loss of voltage signal. Equipment response was that the 11 ESW (Emergency Service Water) pump (cooling for the #11 Emergency Diesel) failed to develop required pressure. The #13-4160V non-safety related bus failed to restore after the transient (and feed the Division 1 Essential Bus). Additionally, the #15 bus transferred to the 1AR transformer (and is feeding the Essential Bus). The #11 Emergency Diesel Generator is currently tagged out of service. Electrical supply is being provided by offsite power. Reactor heat is being removed through the main steam line to the main condenser and reactor water inventory is being provided by the feedwater system. The SRVs lifted and reseated. The HPCI system was manually place into a pressure control mode. The Minnesota Pollution Control Agency is being notified due to the licensee violating the site discharge canal temperature rate of change limit. The licensee notified the NRC Resident Inspector.
Prior to this event the 'B' Control Room Emergency Filtration (CREF) and 'B' Control Room Ventilation (CRV) Systems were inoperable for planned maintenance. On 10-21-11 at 1325 CDT, the #11 EDG ESW Pump was declared inoperable due to low cooling water pump flow, resulting in the #11 EDG being inoperable, which in turn resulted in the 'A' CREF and 'A' CRV being inoperable. Contrary to reporting requirements this condition was not identified and reported pursuant to 10 CFR 50.72(b)(3)(v)(D) as required within 8-hours in the previous event notification. This condition resulted in a loss of safety function for both divisions of CREF and CRV. This update amends the 10-21-11 event notification to include this as an 8-hour non-emergency event pursuant to 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified the R3DO (Nick Valos) |
ENS 47362 | 20 October 2011 20:38:00 | On October 20, 2011 at approximately 1335 hours, a South Texas Project employee suffered an apparent heart attack offsite while attending a fire brigade training exercise. The person was transported to Matagorda County Regional hospital. The individual was evaluated by emergency room personnel and later pronounced deceased. OSHA is being notified pursuant to the requirements of 29CFR1904.39. This ENS report is being made in accordance with 50.72(b)(2)(xi). There was no radioactive contamination involved in this event. The licensee does not plan any media or press release. The NRC Resident Inspector has been notified. |
ENS 47360 | 20 October 2011 13:03:00 | At approximately 0910 (PST) on 10/20/2011, during discussions with an NRC inspector, the AREVA Richland NCS Manager concluded that a previous determination made on Friday, 7/22/2011 that an IROFS (a criticality drain on a HEPA filter housing) declared to be in a degraded state was, in fact, failed in that a pre-filter in the HEPA housing was sitting on the drain and could have prevented it from performing the required safety function. This original discovery was made during a routine PM of the system (a management measure prescribed to this IROFS). On 7/22/2011 (the initial day of discovery) the criticality drain was modified so that it would remain unobstructed and free flowing. At 1303 EDT, AREVA's EHS&L Manager notified the NRC Operations Center of this condition per the requirements of 10CFR70 Appendix A criterion (a)(5) (1 hour report) which requires reporting if '... only one IROFS remains available and reliable to prevent a nuclear criticality and has been in that state for greater than 8 hours,' and indicated that potentially criterion (b)(1) (24 hour report) was also met. This 24 hour criterion requires reporting '... the facility being improperly analyzed, or different than analyzed in the Integrated Safety Analysis, and which results in failure to meet the performance requirements of 10CFR 70.61.' Subsequent to the initial report, AREVA has concluded that the performance requirements of 10CFR 70.61 were met and the 24 hour reporting criteria (b)(1) do not apply. Safety Significance of Event: The safety significance of this event is low. This process is used to recover incinerator ash and the ISA team was unable to identify any mechanism whereby a critical concentration of approximately 285 g U/liter of solution could be created in the process connected to the duct work. Potential Nuclear Criticality Pathways Involved: The only potential pathway is for uranium bearing liquids that exceed 285 g U/liter to gradual build up in the HEPA filter housing. Controlled Parameters (Mass, Moderation, Geometry, Concentration, Etc.): Uranium mass/concentration is controlled. Nuclear Criticality Safety Control(s) or Control System(s) and description of the failures or deficiencies: No control system failures occurred. Corrective Actions to Restore Safety Systems and When Each Was Implemented: A grate was installed over the criticality drain inlet on July 22, 2011 and has been present since that time.
After careful consideration of the guidance contained in FCSS ISG-12 and NUREG 1520, Rev. 1, AREVA has concluded that this condition did not meet the reporting requirements set forth in 10CFR70 Appendix A and formally retracts the previously made report. Notified R2DO (Freeman) and NMSS EO (Hiltz). |
ENS 47359 | 20 October 2011 13:46:00 | During follow-up inspections of flood barrier penetrations into two rooms in the plant it was determined that some of the water tight conduit fittings were not filled with the material required to make them water tight. Inspection caps were removed from the fittings to perform the inspections. Three fittings into room 19 (auxiliary feedwater and plant air compressors) and fittings into room 56E (electrical switchgear) were found to contain no filling material. One additional fitting into room 56E that was thought to be capped was found to be open with a sheet metal box covering the inside access thereby obscuring inspections. All of the affected penetrations have modifications in progress to assure that they are modified and qualified for design basis flood levels. Of the 16 penetrations 6 have been verified to be made water tight by other means, specifically fire foam barrier installed in the conduit from the room 56E side. The remaining 10 penetrations will leak with a 1014 flood, although the plate will restrict flow to some degree. The stations auxiliary feedwater and safety related electrical switch gear could be affected. This eight-hour notification is being made pursuant to 10CFR50.72 (b)(3)(v). The licensee notified the NRC Resident Inspector. |
ENS 47357 | 19 October 2011 23:35:00 | With Unit 1 in Mode 4 and Unit 2 in Mode 1 an issue was identified with the body material of existing installed pressure instruments for both the Personnel and Emergency Airlocks of both units. The pressure instruments were determined to have an aluminum body which is not suited for safety related use in containment. Aluminum is a restricted/limited material in containment because it is not compatible with accident conditions and has failures with multiple adverse effects. Due to this condition, the pressure instruments would potentially lose pressure integrity during a LOCA with containment spray actuation. These pressure instruments are located inside containment and are connected to tubing that penetrates the airlock barrel. In event of a failure of any pressure instrument the integrity of the airlock would be compromised. The containment air locks form part of the containment pressure boundary and, as such, a loss of pressure boundary integrity would no longer meet general design criteria. Compensatory measures have been taken to prevent a failure of the airlock integrity due to containment spray actuation and at this time the airlock is operable. Luminant power determined this to be reportable at 2002 on 10/19/11 per 50.72(b)(3)(ii)(B) Comanche Peak Units 1 and 2 being in an unanalyzed condition that significantly degrades plant safety. The licensee notified the NRC Resident Inspector. |
ENS 47521 | 13 December 2011 16:55:00 | The following information was received by facsimile: Pursuant to 10 CFR 21.21(a)(2), Fisher Controls International, LLC (Fisher) is providing the required written interim notification of a possible deviation or failure to comply. On October 13, 2011, Fisher became aware of a possible issue with the disc pin engagement of a 20" type A11 butterfly Valve, serial number 19102243. The affected valve was returned to Fisher for evaluation and correction and, upon correction, was returned to the customer (Clinton Power Station) on November 21, 2011. Fisher expects to complete its evaluation by January 31, 2012. At that time, if the evaluation reveals that a potential issue exists with the disc pin engagement, Fisher will issue a notification per the requirements of 10 CFR 21.21(b).
This report is retracted based on the following: Pursuant to 10 CFR 21.21(a)(2), Fisher Controls International LLC (Fisher) is providing the required written final notification of a possible deviation or failure to comply. On December 13, 2011, Fisher provided an interim notification of a possible deviation or failure to comply concerning a possible issue with the disc pin engagement of a 20 inch Type A 11 butterfly valve, serial number 19102243. The affected valve was returned to Fisher for evaluation and correction and, upon correction, was returned to the customer on November 21, 2011. Fisher has completed the review of the valve design and has determined that this potential issue would not have negatively affected the subject valve or its performance. Therefore, Fisher will not be issuing a notification per the requirements of 10 CFR 21 21.21(b). Notified R3DO (Kozak) and Part 21 Group via email. |
ENS 47396 | 1 November 2011 11:11:00 | The following information was received by facsimile: NYS Incident 935 - On 10/31/2011 a NY radioactive materials licensee reported a diagnostic misadministration which occurred on 10/13/2011 and discovered on 10/28/2011. A patient undergoing diagnostic imaging of the thyroid using Iodine-123 was administered 4.21 mCi instead of the intended 400 uCi. The estimated dose to the patient's thyroid is 58 rem. This is a preliminary 24 hour notification report. The facility is performing an investigation and root cause analysis. Telephone communications with the facility (and the State of New York) are ongoing. The facility is required to submit a written report within 15 days. New York Event: NY-11-25 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. |
ENS 47334 | 9 October 2011 10:27:00 | The following is a compilation of information received from a verbal report from the State of New Jersey; an e-mail from the State of California; and discussions with the RSO involved in the incident: On October 8, 2011 at approximately 1000 EDT, three contractor personnel were loading Co-60 sources on a GammaCell 220 irradiator at a Johnson & Johnson facility located in Raritan, New Jersey. During the activity, an 8 inch 1210 Curie Co-60 source was somehow dislodged from its shielded position and approximately 4 inches of the source were exposed in an unshielded configuration. In addition, the insertion tool prevented the workers from re-inserting the source quickly. The workers were able to take a titanium shield and force the source into place by breaking off the insertion tool. The source was then confirmed to be in a shielded configuration and safely secured. The total time the source was exposed was estimated to be 25 to 30 seconds. The two individuals loading the source are employed by Foss Therapy, a California Agreement State licensed service provider. The third individual involved in the activity was the RSO employed by Gamma Irradiator Service (GIS) under a Pennsylvania Agreement State license. Based on readings from the electronic dosimetry being used by the workers, the two Foss Therapy employees are estimated to have received whole body doses of 8.1 R and 7.7 R respectively. The RSO working for GIS is estimated to have received a dose of 8.5 R. Extremity exposure has been estimated at between 50 R to 60 R for the two Foss Therapy workers. The workers were also wearing fresh TLDs and finger badging. More detailed dose information will be forthcoming after analysis of this dosimetry. See event report from the State of California: EN # 47335 |
ENS 47307 | 30 September 2011 10:50:00 | |
ENS 47306 | 30 September 2011 00:18:00 | Monticello has discovered that it has not met Technical Specification Surveillance Requirement (SR) 3.8.1.7 relating to the largest single post-accident load reject for the Emergency Diesel Generators (EDG). Although the current test designated post-accident load is successfully load rejected during the surveillance, the test load rejection must be higher to bound all post-accident load scenarios. The capability of an EDG subsystem to recover from a reject of the largest single post-accident load testing has not met the requirements of SR 3.8.1.7. Therefore, both EDGs have been declared inoperable. Both EDGs are considered Functional and Available for use at this time. There were no automatic EDG initiation signals associated with this event. The licensee notified the NRC Resident Inspector and will notify the State. |
ENS 47308 | 30 September 2011 11:00:00 | During a review of Integrated Safety Analysis (ISA) and criticality evaluation assumptions for a floor cleaning scrubber used in the Dry Conversion Process facility, it was determined that an equipment configuration was different than that analyzed in the ISA. Field verification of the floor cleaning scrubber recovery tank determined that its capacity was greater than the safe volume limit referenced in the analysis. The tank volume is credited as an IROFS (Items Relied On For Safety) in the ISA summary. Based on a review of this as-found condition, discovered at approximately 1120 EDT on September 29, 2011, it was determined that the system was different than analyzed in the ISA and resulted in a failure to meet performance requirements. The tanks for the floor cleaning scrubbers were inspected and no unsafe condition existed. Operation of the equipment has been suspended pending additional review and implementation of corrective actions. Additional corrective actions and extent of condition are being evaluated. This event is being reported pursuant to the reporting requirements of 10CFR70 Appendix A (b)(1) within 24 hours of discovery. The licensee will notify the NRC Region 2, North Carolina Radiation Protection Agency and New Hanover County Emergency Management. |
ENS 47302 | 29 September 2011 06:56:00 | During plant startup, a single train fire alarm was received in containment at 0540 EDT on 9/29/2011. An Unusual Event was declared at 0602 EDT based on the containment not being accessible within 15 minutes. An inspection in containment revealed no fire or smoke or the cause for receipt of the alarm. The licensee has notified the State, Counties and the NRC Resident Inspector.
The Unusual Event was terminated at 0714 EDT on 9/29/2011. The licensee has notified the State, Counties and the NRC Resident Inspector. Notified the R2DO (Nease), NRR (Thomas), IRD (Gott), FEMA (Casto) and DHS (Rickerson). |
ENS 47299 | 28 September 2011 08:26:00 | At 0414 (CDT) on 9/28/2011, the Unit 3 reactor automatically scrammed due to actuation of the Reactor Protection System (RPS) from a turbine trip. Preliminary indications show the turbine tripped on a generator trip with generator neutral overvoltage (359GN) relay actuation. Cause of relay actuation is under investigation. Seven Safely Relief Valves (SRVs) cycled due to the reactor pressure transient with reactor pressure automatically controlled by the Main Turbine Bypass Valves. All systems responded as expected to the turbine trip. No Emergency Core Cooling System (ECCS) or Reactor Core Isolation Cooling (RCIC) reactor water level initiation set points were reached. Primary containment isolation and initiation signals for groups 2, 3, 6 & 8 were received as expected. Reactor water level is being automatically controlled by the feedwater system. This event is reportable within 4 hours per 10 CFR 50.72(b)(2)(iv)(B) 'any event or condition that results in actuation of the RPS when the reactor is critical except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation.' It is also reportable within 8 hours per 10 CFR 50.72(b)(3)(iv)(A) and requires an LER within 60 days per 10 CFR 50.73(a)(2)(iv)(A). The NRC Resident Inspector has been notified. All control rods fully inserted. The plant is being supplied from offsite power and is in a normal shutdown configuration. The MSIVs are open with decay heat being removed via steam to the main condenser using the bypass valves. There was no impact on Units 1 or 2. |
ENS 47288 | 22 September 2011 22:20:00 | On Thursday, September 22, 2011 at 1452 (EDT), with the reactor at 100% core thermal power and steady state conditions, Pilgrim Nuclear Power Station (PNPS) conservatively declared both Salt Service Water (SSW) subsystems inoperable when engineering analysis determined that inrush current on pump restart may exceed the thermal overload trip relay settings during certain degraded voltage conditions. A 24-hour Limiting Condition for Operation action statement was entered. Entergy/Pilgrim is in the process of implementing temporary modifications to correct this issue. This potential concern has no impact on public health and safety. This 8-hour notification is being conservatively reported in accordance with 10 CFR50.72(b)(3)(ii)(B). The NRC Resident Inspector has been notified. The licensee will notify the Massachusetts Emergency Management Center.
Event Notification 47288 was conservatively made to ensure that the Eight-Hour Non-Emergency reporting requirements of 10 CFR 50.72 were satisfied pending further evaluation of Salt Service Water (SSW) subsystem operability during certain degraded voltage conditions. Pilgrim Station has subsequently evaluated the impact of the Salt Service Water (SSW) pump motor operating load; potential degraded voltage conditions on thermal overload settings; and pump motor restart requirements; and determined that reasonable assurance of SSW subsystem operability existed. Both SSW subsystems were evaluated to be capable of satisfying the system safety function of providing cooling water to the Reactor Building Closed Cooling Water System heat exchangers during accident and transient conditions. Pilgrim 's evaluation considered manufacturer's data for the thermal overload relays, site specific shop testing, and reliable grid conditions which minimize the potential for extended operation with 4 kV buses operating at just above degraded voltage relay trip settings. This past operability evaluation concluded that the SSW pumps would support continuous operation of the SSW subsystems. Therefore, the initial 50.72 report is being retracted. The licensee will notify the NRC Resident Inspector. Notified the R1DO (Caruso). |
ENS 47261 | 13 September 2011 16:18:00 | A non-licensed employee supervisor had a confirmed positive drug test during random testing. The employee's access to the plant has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee informed the NRC Resident Inspector. |
ENS 47253 | 9 September 2011 14:01:00 | A temporary diesel generator has emitted droplets of oil to the ground which was washed to a storm drain. The exhaust droplets were contained in secondary containment. An oil-only absorbent boom was placed around the storm drain to capture residual oil and prevent oil to the storm drain. It is estimated that 0.05 gallons of diesel fuel entered a storm drain. The licensee notified the Maryland Department of Environment, National Response Center, U.S. Coast Guard and the NRC Resident Inspector. |
ENS 47252 | 8 September 2011 19:08:00 | On September 8, 2011 at 1538 PDT, the San Onofre Units 2 and 3 reactors tripped due to a grid disturbance. All control rods fully inserted. The EFAS (Emergency Feed Actuation System) initiated as expected for a reactor trip. Steam generators are being fed by the main feedwater system and decay heat is being removed through the steam bypass system to the main condensers. There is no primary to secondary leakage and no safety relief valves lifted. Site electrical power sources are being fed from off-site power and both units are in a normal shutdown configuration. The emergency diesel-generators are in standby/operable status and were not required during the event. Both units are stable (NOT/NOP) and in Mode 3. The reactor trip response is considered uncomplicated. The licensee has notified the NRC Resident Inspector. |
ENS 47255 | 9 September 2011 17:03:00 | On 9/8/2011, during site cleanup activities, workers at the Greenfields coal processing facility near Pageton, WV unexpectedly discovered two generally licensed, fixed-type gauges in a remote, back lot area of the plant property. One of the gauges had a stuck open shutter. The gauges are currently clamped to pipes, roped off in an area restricted from personnel access, at about a 300 foot distance from personnel. No threat to health and safety exists at this time. The gauges are designed to be affixed to piping for industrial process control, but were found among various discarded metal parts. An inspector from the West Virginia Radiological Health Program took survey measurements and verified there was no spread of contamination. Dose rates outside of the gauge casing are about 0.4 mRem/hr. Dose rates inside of the pipes are about 40 mRem/hr, which indicates that the shutters are likely open. Facility personnel will take photos of the gauges and send to NRC Region I. Additionally, they intend to place the gauges with the piping in a locked gang box that will be chained and locked to a fixed metal structure. The facility owner, Deep Green, is taking steps to have the material removed from the site. Gauge information: (2) Berthold Technologies gauges, Model No. LB 7440 D; 30 mCi, Cs-137 source (Sep. 21, 2001); Source serial # HE681 manufactured by AEA Technologies, model CDC.p4; SSD Safety Evaluation No. TN-1031-D-101-B. |
ENS 47251 | 8 September 2011 16:54:00 | The following information was received by e-mail: On September 8, 2011, Arrow Engineering contacted California Office of Emergency Services to report a stolen moisture/density gauge, Troxler model 3440 Plus, 8 mCi Cs-137 source and 40 mCi Am:Be-214 source, S/N 39529. The gauge was locked in the transport case, secured in the back of a vehicle, a 2001 Chevy Silverado. The gauge operator discovered the vehicle stolen at approximately (0500 PDT), when he was leaving his home, to proceed to the worksite. The operator then notified police and then the Radiation Safety Officer (RSO). The Los Angeles County Sheriffs Department arrived approximately (0545 PDT) and the operator made his report of the stolen vehicle and gauge. After the operator had made the report, he then proceeded to return to work. After the RSO was notified of the incident, he proceeded to contact Los Angeles County Radiation Management and left a detailed message and then contacted the Office of Emergency Services to report the incident. Approximately (0940 PDT) a LA Co. inspector notified the ICE RAM South of the incident. An ICE inspector then contacted the RSO to obtain information on the incident. The RSO confirmed the incident took place at the residence of the operator sometime in the early morning of September 8, 2011, prior to (0500 PDT), when the operator discovered that the vehicle was missing. He stated that the operator had left the gauge in the locked transport case, locked in the bed of the pickup with two chains and three locks. He also stated that the operator had an alarm on the vehicle, but did not hear it since he was likely asleep at the time. He also stated that the vehicle was not able to be stored in the garage since it was a 4 wheel drive vehicle. The gauge was left in the vehicle since the operator had returned from a jobsite late the prior evening and had an early appointment the next morning. The RSO was informed to provide a written report within 30 days, to place an ad and a reward in the local newspaper and to notify ICE if the gauge is found. At approximately 1100 (PDT), the RSO contacted RHB reporting that the vehicle was found with the gauge and transport case still locked in the bed of the truck and with a broken drivers side window. The RSO stated that the vehicle was left in the driveway of a neighbors house. It was found by the owner while returning home from the office at approximately 1030 (PDT). The owner then contacted the LA Co. Sheriff and was told not to touch the vehicle before the Sheriff arrived. When they arrived, he was allowed to check the gauge case to confirm that the gauge was still inside and intact. The operator was able to confirm this and the Sheriff were able to inspect the vehicle and gather evidence. The licensee was informed to take a leak test immediately and keep the gauge out of service until the leak test results were returned and confirmed to be below the contamination limit. While the investigation is still ongoing, pending licensee's findings in their 30 day report, the licensee is likely to be cited for storing the gauge in an unauthorized storage location, for improperly securing the gauge, or leaving the gauge unattended. 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 |
ENS 47254 | 9 September 2011 16:26:00 | Wisconsin Department of Health Services (DHS) received notification by phone call from the licensee on September 8, 2011 about a medical event involving I-125 permanent prostate seed implants. During a standard review conducted September 7, 2011 of a post seed implant report the Authorized Medical Physicist determined that the dose delivered differed from the prescribed dose by 20% or more. Specifically it was found that only 76% (110 Gy) of the prescribed dose was delivered to 90% of the CTV for an implant completed on July 22, 2011. The licensee had established the dose based criteria that by post-operation CT, prostate D90 values are < 80% or >130% for classifying medical events. The licensee has notified the Authorized User, referring physician and will notify the patient during a scheduled examine the week of September 11, 2011. There is no expected immediate harm to the patient and the Authorized User and referring physician will discuss with the patient to determine if supplemental radiation (implant or external beam) will be done. DHS conducted an investigating of this medical event on September 9, 2011 by sending a special inspection team. The preliminarily conclusion after reviewing the licensee's procedures and discussion with the Authorized User and Authorized Medical Physicist is that the under dose was directly caused by edema of the prostate, i.e. post implant procedure swelling. Wisconsin Event Number: WI110014 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. |
ENS 47213 | 28 August 2011 11:13:00 | Oyster Creek has lost 10 sirens in the Emergency Plan Zone. Communications with the Ocean County Office of Emergency Management has been conducted to inform them of the outage. The licensee has notified the NRC Resident Inspector. |
ENS 47211 | 28 August 2011 06:35:00 | PBAPS (Peach Bottom Atomic Power Station) Control Room was notified of a loss of greater than 25% of sirens after severe storms in the area associated with Hurricane Irene. 51 of 97 Emergency Plan Zone (EPZ) sirens are unavailable in York County and Hartford County. Actions are currently being taken to restore unavailable sirens. The licensee notified Pennsylvania and Maryland Emergency Management and York and Harford Counties. The licensee notified the NRC Resident Inspector. |
ENS 47210 | 28 August 2011 03:38:00 | At 2130 (EDT) on August 27, 2011, an automatic actuation of the Unit 1 Train A emergency diesel generator occurred due to an actuation signal from the load sequencer. The Train A 4160 kV emergency bus transferred to the emergency diesel generator and all Train A emergency loads required for Mode 2 started and sequenced onto the Train A 4160 KV emergency bus except the 480 volt center breaker to the bus E1A2 that did not close. The load sequencer is designed upon the receipt of a safety injection actuation and/or loss of offsite power to provide a signal to strip loads from the 4160 kV emergency bus and then, in sequence, to re-energize the associated 480 volt buses and to load engineering safety feature components onto the 4160 kV and associated 480 volt emergency buses in a predetermined sequence. Per 10 CFR 50.72(b)(3)(iv)(B), additional emergency safety features loads that actuated were the Train A reactor containment fan coolers and auxiliary feedwater system. Unit 1 remains critical at 100 percent power. No emergency core cooling system injection occurred into the reactor coolant system. The event occurred during surveillance testing when the Train A sequencer was taken from the AUTO Test position to the local position. It is not understood why the actuation occurred. In addition, it is not understood why the 480 volt load center breaker to the bus E1A2 did not close. The 480 volt bus E1A2 was re-energized at 2308 (EDT) on August 27, 2011. The Train A 4160 kV bus was restored to the offsite power source at 0150 (EDT) on August 28, 2011 and the Train A emergency diesel generator and engineering safety features loads were restored to their normal condition at 0201 (EDT) on August 28, 2011. With the Train A sequencer non-functional, the following Train A components are inoperable: 1) High Head Safety Injection Pump 1A; 2) Low Head Safety Injection Pump 1A; 3) Containment Spray Pump 1A; 4) RCFC (Reactor Containment Fan Cooler) Fan 11A; 5) RCFC Fan 12A; 6) Component Cooling Water Pump 1A; 7) Essential Cooling Water Pump 1A; 8) Aux Feedwater Pump 11; 9) Control Room/Elect. Aux Bldg HVAC; 10) Ess (Essential) Chiller 12A; and 11) ESF Diesel Generator 11. Although these components will not automatically start on a safety injection signal or loss of offsite power, these loads can be manually actuated. Engineered Safety Features Trains B and C remain operable. The licensee notified the NRC Resident Inspector.
The licensee is updating this event report to retract the originally reported valid specified system actuation and report it instead as a 60-day invalid specified system actuation report made by telephone: This update is a 60-day telephone notification in lieu of a written licensee event report being made under 10CFR50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1). This event was originally reported per 50.72(b)(3)(iv)(A) as a valid actuation of the Unit 1 Train A emergency diesel generator and sequencing of Mode II (Loss of Offsite Power) emergency loads. The actuation occurred during surveillance testing when the Train A load sequencer was taken from the AUTO test position to the local position. Subsequent investigation has determined that the actuation occurred due to a faulty integrated chip within the sequencer's load sequence auto test module, and was not due to sensed or simulated plant conditions that would require a Mode II actuation. Unit 1 was at 100% power and no loss of offsite power occurred. The Train A equipment response to this invalid actuation is described in the original notification information provided on 08/28/2011. Additionally, the 10CFR50.72 Notification originally reported under Event Number 47210 is being retracted, since the actuation has been determined to be not valid. The licensee will notify the NRC Resident Inspector. R4DO (Whitten) notified. |
ENS 47169 | 19 August 2011 01:10:00 | At 2250 EDT on 8/18/2011, Unit 1 Reactor/Turbine automatically tripped on RCP (Reactor Coolant Pump) Busses UV (Under-Voltage) trip. Following the reactor trip, all safety-related equipment operated as designed. Auxiliary Feedwater automatically actuated as expected from the Feedwater Isolation Signal. No primary PORVs and/or Safety Valves opened during or after this trip. Unit 1 is currently being maintained in Mode 3 at NOT/NOP, approximately 548 F (degrees) and 2233 psig, with Auxiliary Feedwater supplying the Steam Generators. At the time of the trip, a 50G (instantaneous overcurrent ground) relay flag was found dropped on the '1A' 6.9 KV unit board. Subsequently, the '1A' 6.9 KV start bus was found to have transferred to its alternate supply, 'B' CSST (Common Station Service Transformer). 1A condenser circulating water pump motor trip out was also received in the MCR (Main Control Room). The method of decay heat removal is via steam dumps to the condenser with MSIVs open. The current temperature and pressure is stable. There is no indications of any primary/secondary leakage. All control rods inserted. The electrical alignment is normal with the exception of the above mentioned items, supplied from off-site power. There is no impact to Unit 2. Unit 2 is operating at 100% power/ Mode 1. The licensee notified the NRC Resident Inspector. |
ENS 47139 | 10 August 2011 14:42:00 | At approximately 0730 (EDT) on 8/9/11 a siren system communication problem was discovered by Emergency Preparedness personnel at the Harris Nuclear Plant (HNP). Telecommunications was contacted to troubleshoot the problem and at approximately 0820 (EDT) the communications system was restored. A previously scheduled quarterly growl test was conducted with 83 sirens performing satisfactorily. Upon further investigation, it was determined that there were communication failures recorded during the 0600 (EDT) report. 20 of the 83 sirens were impacted by the communications errors. The previous 0000 (EDT) report indicated no communications errors. Investigation is continuing as to the ability of the sirens to be activated during the period the errors were received. It is likely that the ability to activate the sirens had been lost, therefore, this event is reportable per 10 CFR 50.72(b)(3)(xiii) due to the loss of a significant portion of the offsite notification system. The cause of the communication errors are believed to be due to radiofrequency disturbances however the investigation is ongoing. In the event that the sirens had been needed, the State of North Carolina and all four counties within the 10-mile emergency planning zone would have implemented mobile route alerting, as detailed in the Emergency Plan. The licensee notified the NRC Resident Inspector. |
ENS 47137 | 9 August 2011 16:25:00 | (There was an) inadvertent actuation of siren 46 due to water intrusion into the control box. The door was knocked open by a falling tree branch. Power to the siren has been isolated to prevent re-occurrence. Repair is scheduled for August 10, 2011 The licensee notified both Wayne and Monroe counties and the NRC Resident Inspector. |
ENS 47119 | 2 August 2011 19:52:00 | At 0400 (CDT) on Wednesday, August 3, 2011, the Dresden Nuclear Power Station (DNPS) Technical Support Center (TSC) emergency ventilation system will be removed from service for planned preventative maintenance activities on the TSC ventilation radiation monitoring system (PING). The TSC air handing and filtration units will be nonfunctional, rendering the TSC HVAC accident mode non-functional. This maintenance is scheduled to minimize out of service time. The planned TSC ventilation outage is scheduled to be completed within 48 hours. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing Emergency Planning procedures and checklists. If radiological conditions require TSC facility evacuation during ventilation system restoration; the Station Emergency Director will evacuate and relocate the TSC staff in accordance with, EP-AA-112-200-F-01. The licensee notified the NRC Resident Inspector.
At 1905 CDT on August 4, 2011, Dresden TSC ventilation was restored. The emergency response facility is fully functional at this time. The licensee has notified the NRC Resident Inspector. Notified R3DO (Skokowski). |
ENS 47120 | 2 August 2011 21:27:00 | The following condition is being reported by Arkansas Nuclear One, Unit 2 (ANO-2) in accordance with 10CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition' and in accordance with 10CFR 50.72(b)(3)(v)(D), 'A Condition That Could Have Prevented Fulfillment of a Safety Function.' On 08/02/2011 at 1346 CDT, the ANO Unit 2 Control Room was notified by Engineering that a postulated High Energy Line Break (HELB) could potentially cause both the Red and Green Train Emergency Safeguard Features (ESF) Rooms to exceed their environmentally qualified temperature limits. This postulated condition would be possible due to normally open room purge dampers exposing ESF equipment in these rooms to a common area impacted by HELB conditions. The ESF Rooms contain the Red and Green Trains of High Pressure Safely Injection Pumps, Low Pressure Safety Injection Pumps, Containment Spray Pumps, and Shutdown Cooling Heat Exchangers. Until further Engineering evaluation can be performed to validate this postulated scenario, ANO-2 has closed ESF room purge dampers to provide Red and Green ESF train separation during a potential HELB event. Refer to (ANO-2) Condition Report CR-ANO-2-2011-02772 for further information. The NRC Resident has been notified.
The purpose of this notification is to retract a previous report made by Arkansas Nuclear One, Unit 2 (ANO-2) on 08/02/2011 at 2127 (EDT) (EN# 47120). The initial report documented that a postulated High Energy Line Break (HELB) could potentially cause rooms containing both trains of Emergency Safeguard Features (ESF) equipment to exceed their environmentally qualified temperature limits. The ESF rooms contain the High Pressure Safely Injection Pumps, Low Pressure Safety Injection Pumps, Containment Spray Pumps, and Shutdown Cooling Heat Exchangers. Specifically, normally open ESF room purge dampers exposing both trains of ESF equipment to a common area impacted by postulated HELB conditions were not modeled in the ANO-2 HELB analysis. This condition was reported in accordance with 10CFR 50.72(b)(3)(ii)(B), 'Unanalyzed Condition' and 10CFR 50.72(b)(3)(v)(D), 'A Condition That Could Have Prevented Fulfillment of a Safety Function'. Since the initial report, Engineering has revised the ANO-2 HELB model to include the effects of the open ESF room purge dampers. The resulting analysis shows that a HELB event will not cause the required ESF equipment to exceed analyzed temperature limits with the room purge dampers in the open configuration. Therefore, the condition did not result in 'a condition that could have prevented the fulfillment of a safety function' and did not result in an 'unanalyzed condition that significantly degrades plant safety'. Based on the revised HELB analysis, the previous report (EN#47120) describes a condition that does not meet the reporting requirements of 10CFR 50.72(b)(3)(v)(D) or 10CFR 50.72(b)(3)(ii)(B) and is therefore retracted. The NRC Resident Inspector has been informed of the retraction. Notified R4DO (Hay). |
ENS 47116 | 2 August 2011 15:30:00 | A utility non-licensed supervisor had a confirmed positive for alcohol during a random fitness-for-duty test. The individual's access to the site has been terminated. Contact the Headquarters Operations Officer for additional details. The licensee informed the NRC Resident Inspector. |
ENS 47114 | 1 August 2011 19:43:00 | On August 02, 2011, at 0700 EDT, the Technical Support Center will be unavailable due to pre-planned maintenance to maintain the Technical Support Center and Emergency Response Data Acquisition and Display System ventilation system. The TSC is expected to be restored to available status in approximately 12 hours. If an emergency is declared requiring TSC activation during this period, the TSC will be staffed and activated using existing emergency planning procedures, and the TSC staff will relocate to an alternate TSC location in accordance with applicable site procedures. This notification is being made in accordance with 10CFR 50.72 (b)(3)(xiii) due to the potential loss of an emergency response facility (ERF). An update will be provided once the TSC has been restored to normal operation. The NRC Resident Inspector has been notified.
On August 2, 2011 at 1645 EDT the TSC was restored and is now fully functional. The licensee notified the NRC Resident Inspector. Notified the R2DO (Desai). |
ENS 47140 | 10 August 2011 16:23:00 | The following information was received by e-mail: On August 8, 2011 the licensee's radiation safety officer called to advise the Agency (Illinois Emergency Management Agency) of a gauge shutter failure. During the performance of routine operability checks and area surveys on July 22, 2011 it was found that the shutter lever on a Kay- Ray/Sensall 7062P could not be completely closed. The gauge is permanently mounted to a 12 inch pipeline in a remote area of the refinery. The line runs overhead at a height of approximately 10 to 12 feet. Arrangements were made to have personnel with duties in the area to be notified of the shutter failure and to prohibit any work in the immediate area. The nearest work station was determined to be over 200 feet away. Radiation levels at one foot from the gauge were measured at less than 1 milliR/h. After contacting the manufacturer on August 8 to determine the best course of action, the licensee applied a lubricant to the shutter area. Subsequently, after waiting a day, attempts to close the shutter were successful. The shutter is now operating as expected. Arrangements have been made to have a service licensee visit the site to ensure operation is as expected. As the failure was noted during performance of routine tests designed to detect such failures, no changes to the licensee's procedures or protocols are anticipated. The licensee was reminded of the importance of promptly reporting such events in the future. The sealed source gauge is a Kay-Ray/Sensall model 7700-Y containing 0.14 Ci of CS-137. The fixed gauge is a Kay-Ray/Sensall, Model number 7063P; Serial number S96G1816. Illinois Event Number IL11105 |
ENS 47069 | 18 July 2011 16:38:00 | The following information was received by e-mail: A routine inspection was performed of the licensee on July 11 and 12, 2011. Based on records reviewed at the time of inspection and additional documentation provided to the Kentucky Radiation Health Branch on July 15, 2011. Three unreported Medical Event(ME)s appeared to have occurred during the past three years. 1. On June 16, 2008 a Written Directive(WD) signed by an Authorized User (AU) for the administration for a total activity of 192.00 U (148.492 mCi) and a Prescription Dose of 125.0 Gy of Pd-103 seeds to the prostate. Manufacturer: Theragenics Model Number: Not Available Seed Lot No: NOT AVAILABLE Source Activity: 2.000 U (1.547 mCi )per seed Number of Seeds (Sources): 96 Based on the licensees use of the nationally recognized Report of AAPM (American Association of Physicists in Medicine) Task Group Report No. 64 for determining the dose received by the prostate a ME was identified. The post operative CT indicated the prostate received: D90 of 81.17 Gy (64.94%) of the prescribed dose. D100 of 43.71 Gy (34.97%) of the prescribed dose. 2. On March 23, 2009 a Written Directive(WD) signed by an Authorized User for the administration for a Total Activity of 133.560 U (103.295 mCi) and a Prescription dose of 90. Gy of PD-103 seeds to the prostate. Manufacturer: Theragenics Model Number: 200 Seed Lot No: 0910E Source Activity: 1.590 U (1.230 mCi) per seed Number of Seeds (Sources): 84 Based on the licensees use of the nationally recognized Report of AAPM Task Group Report No. 64 for determining the dose received by the prostate a ME was identified. The post operative CT indicated the prostate received: D90 of 55.41 Gy (61.57% ) of the prescribed dose. D100 of 28.03 Gy (31.15% of the prescribed dose). 3. On April 12, 2010 a Written Directive (WD) signed by an Authorized User for the administration for a Total Activity of 140.736 U (108.848 mCi) and a Prescription Dose of 125.0 Gy of PD-103 seeds to the prostate. Manufacturer: Theragenics Model Number: 200 Seed Lot No: 0910E Source Activity: 2.199 U (1.701 mCi ) per seed Number of Seeds (Sources): 64 Based on the licensees use of the nationally recognized Report of AAPM Task Group Report No. 64 for determining the dose received by the prostate a ME was identified. The post operative CT indicated the prostate received: D90 of 70.71 Gy (56.57% )of the prescribed dose. D100 of 34.21 Gy (27.36% of the prescribed dose). The licensee and the RSO (Radiation Safety Officer) were unaware the three procedures were reportable ME's. The RSO and the AU reviewed the Pre and Post implant plan and the AU was satisfied to follow the patients progress with repeated PSA testing. The licensee is currently reviewing the findings and implementing preventive measures agreed to by the licensee and the (Agreement State). 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. |
ENS 47053 | 15 July 2011 10:35:00 | The following special report is being made in accordance with the requirements of Technical Specification (TS) 6.7.2(1)(c) which states in part: There shall be a report not later than the following day by telephone or similar conveyance to the NRC Headquarters Operation Center of any reportable occurrence as defined in TS 1.3. TS 1.3 defines an observed inadequacy in the implementation of administrative or procedural controls, such that the inadequacy caused the existence of a condition which results in a violation of technical specifications as a reportable occurrence. Specifically, on Thursday July 14, 2011, while performing routine surveillance checks, an operator left the control room while the console key remained in the console key switch and the switch was in the TEST position. This is in direct violation of the staffing requirements defined in TS 6.1.3(1)(a) which states a licensed reactor operator must be in the control room when the reactor is not secured. The reactor was not secured by the fact that the console key switch was not in the OFF position and the key remained in the console key switch. However, all control elements were fully inserted, the reactor was shut down, no work was in progress involving core fuel, core structure, control elements or drives and no experiments were being moved. This condition remained for a period of approximately 7 minutes until the cognizant senior reactor operator entered the control room and secured the reactor. * * * UPDATE FROM ROBERT AGASIE TO PETE SNYDER ON 7/18/11 AT 0846 EDT * * * After a records review the licensee determined that: the actual event time was 1055 AM CDT and the actual duration of time that the operator was not in the control room was 85 seconds. Notified R3DO (Duncan) and NRR (Tran). |
ENS 47037 | 8 July 2011 13:34:00 | A gauge handle on a portable Troxler Gauge Model 3440 (S/N #16851) was run over by a truck in the Hastings, Michigan area. The measurement and source handle rod was bent. The licensee performed an area survey and no leakage was detected. The density gauge source was not exposed. The licensee plans to ship the damaged gauge to the Troxler Company for repair or replacement. |
ENS 47034 | 8 July 2011 12:03:00 | The following information was received from the State by facsimile: Notifications: Licensee called the Pennsylvania Emergency Management Agency (PEMA), who then contacted Department of Environmental Protection (DEP) via phone on July 7, 2011 @ 1540. This event is 24 hour reportable under 10CFR 30.50(b) (2). Event Description: A CPN Model MC3 soil moisture Nuclear Density Gauge with cesium-137 and americium-241 sources was damaged while in use. Specifically, the rod with the Cs-137 source was damaged during a field measurement. However, the licensee was able to pull the rod back enough to get the nuclear gauge in the DOT ship container. Per CPN specs, the source is an industrial type 10 millicurie Cs-137 source with double encapsulation. Cause of the Event: A truck owned and operated by GA Covey Engineering drifted back over a CPN moisture density gauge bending one of the probe rods, leaving the Cs-137 source extended. Actions: The gauge was inspected by the company's regional RSO. It was determined that the source was still contained within the probe rod and managed to fit it back into the shipping container and secure it in the company truck. The licensee secured the area around the truck until a confirmatory survey was performed by PA DEP Emergency Response staff. The gauge was then brought to a secure site in PA for storage. The licensee will work with CPN to have the nuclear gauge repaired. PA Event: PA110015 |
ENS 47033 | 7 July 2011 14:50:00 | Power has been removed from 10 (out of 101) sirens due to flooding conditions. (Three) out of 5 sirens in Pottawattamie county, 4 of 18 in Harrison county and 3 of 78 in Washington county. There are compensatory measures in place to ensure notification to any members of the public that may still be in these areas. The station is also suspending testing and reporting (performance indicator) data for these sirens in accordance with NEI 99-02. All of these sirens serve areas for which there are no residents requiring evacuation. This is being reported per 10CFR50.72(b)(3)(xiii)) for 'Any event that results in a major loss of emergency assessment capability, offsite response capability, or communications capability'. The licensee notified the States of Nebraska and Iowa, the Counties of Harrison, Washington and Pottawattamie and the NRC Resident Inspector of this report. |
ENS 47039 | 8 July 2011 16:00:00 | This notification is being made under 10 CFR 50.72(b)(2)(xi) of a situation related to the protection of the environment for which a notification to a state agency has been made. On July 8, 2011, South Carolina Electric & Gas Company (SCE&G) made a voluntary notification to the South Carolina Department of Health and Environmental Control (SCDHEC) in accordance with the Nuclear Energy Institute (NEI) Groundwater Protection Initiative for an on-site leak that may exceed 100 gallons. The leak was entirely contained within the V. C. Summer Nuclear Station site boundary. On July 7, 2011 during a routine inspection, a leak was discovered in a liquid radwaste line upstream from its final discharge point and downstream of the release radiation monitors. This line contains processed liquid radwaste that has been sampled and permitted for release. The leak occurred inside a containment enclosure which overflowed. The enclosure was drained and the liquid was returned to the plant for disposal. The leak has been isolated and all radwaste release via this pathway has been suspended. The last inspection of this equipment was conducted on February 14, 2011, and revealed no external leakage. The licensee notified the NRC Resident Inspector. |
ENS 47112 | 1 August 2011 12:03:00 | The following information was received by facsimile: COMPONENT: Air start motors with friction-clutch inertia drives (not pre-engagement type) Engine Systems Inc. (ESI) began a 10CFR21 evaluation on 06/16/11 upon receipt of a failed air start motor from Exelon - Clinton. According to information provided by Clinton Nuclear Plant, they experienced a slow start of an EDG during one of their surveillance runs. Subsequent troubleshooting revealed that one of the air start motors had failed. Specifically, it was found that the starter drive had failed which prevented the air start motor from applying full starting torque to the engine. The inspections performed by Clinton indicated the failed motor had what appeared to be excessive preload applied to the springs of the drive clutch which led to spring roll and fragmentation. Clinton had also supplied six remaining air start motors from their inventory to ESI for evaluation of the drive clutch torque setting. The air starters are Pow-R-Quik type LS-60-RH1 and LS-60-LH1 (where RH and LH designate right hand and left hand rotation respectively). The evaluation was concluded on 08/01/11 and determined the air motor failure to be a reportable defect as defined by 10CFR21. The starter drive failure was attributed to excessive torque adjustment of the clutch assembly. This overstressed the clutch disc plates which lead to eventual failure of the friction-clutch inertia drive assembly. Failure of the clutch drive renders the air start motor inoperative. Depending upon the configuration of the engine's starting system, an inoperative air start motor could cause excessive starting times or a start failure. Either scenario has the potential to prevent the emergency diesel generator from performing its safety related function. Clutch torque testing was performed on the six (6) starters returned from Clinton's inventory. (Results indicate inconsistent clutch settings). The specified torque setting for the clutch drive is 275-300 ft-Ibs. Of the six starters tested, only one was within specification. Based on the test data obtained from the six (6) Clinton starters, all previous shipments of air motors with the friction-clutch inertia drive are suspect for having inconsistent clutch drive torque settings. Affected users of the air start motors that contain friction-clutch inertia drives include the following: Exelon-Clinton, Entergy-Indian Point 3 and PSE&G-Salem. Engine Systems, Inc Report Number: 10CFR21-0102, Rev. 0. |
ENS 47118 | 2 August 2011 19:52:00 | While preparing to restore AOV 3-1601-22, Drywell /Torus Vent Valve, to operable status following solenoid replacement, operations personnel attempted to open AOV 3-1601-22 in accordance with plant procedures. When the control switch was taken to the open position, the U3 control room received a partial U3 Containment System Isolation. Light indication was lost for several Containment Isolation valves. Troubleshooting revealed that a control power fuse had blown during the attempt to open AOV 3-1601-22. Operations personnel entered the appropriate Technical Specification and Technical Requirement Manual Required Actions. Operations personnel replaced the blown fuse and indication was restored to the affected valves. Following the fuse replacement, the following containment air sample valves were observed to have repositioned during the event: 3-8501-3B, 3-8501-1B, 3-8501-5B, 3-9205B, 3-9206B, 3-9207B, and 3-9208B. Subsequent troubleshooting revealed that during the AOV solenoid replacement, the conduit cover screw penetrated the tape of an electrical splice and caused a short to ground. The solenoid was repaired and systems were restored as required. During this event, all valves repositioned to their intended safety position. There was no loss of function as a result of this condition. Therefore, the health and safety of the public were not compromised. Since this ESF actuation was caused by an invalid signal, it is being reported pursuant to 10 CFR 50.73(a) (2) (iv) (A) as specified by 10 CFR 50.73(a) (1), which allows a telephone notification in lieu of a written licensee event report within 60 days. The licensee notified the NRC Resident Inspector. |
ENS 46920 | 3 June 2011 15:19:00 | The following information was sent by the State of Florida Bureau of Radiation Control via email: (The licensee's two) fill detectors were bought in the 1990's and were not in use. The fill detectors were stored in a spare room full of scrap metal. The room was cleaned out in early May and the scrap sent to Alpha Metal Recycling, 2392 NW 147 Street, Opa-Locka, Florida 33054. The recycling plant manager said that the load has already been sent overseas to either China, Pakistan or India. Loss of the material was found when the application for license renewal was being filled out. The licensee will send a written report to Radioactive Materials. Any further action is referred to Radioactive Materials. This office will take no further action on this incident. The two fill detector sealed sources each contained 100 mCi of Am-241 and were manufactured by Industrial Dynamics (Fil Tech; Model Number FT50; Serial Numbers 1296 and 126). Florida Incident: FL11-045 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) |
ENS 46921 | 3 June 2011 15:40:00 | The following information was sent by the State of Florida Bureau of Radiation Control via email: (On June 2, 2011), a patient with over 100 tumors was being treated. One lesion was supposed to get 16 gray, but received only .85 gray. The physicist forgot to adjust for the weight factor. The physicians have been notified. The patient will receive the correct dose during the next visit. The licensee will submit a written report to the Radioactive Materials. Any further action is referred to Radioactive Materials. No further action will be taken on this incident by this office. The patient was undergoing Gamma Knife Stereotactic Radiosurgery using 3073 Ci of Co-60. Florida Incident: FL11-046 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. |
ENS 46908 | 1 June 2011 13:05:00 | The following information was received by fax: Radiographers set up a shot on a 2 inch diameter pipe which was laying in jackstands. They were working with a SPEC 150 exposure device containing a 42 Ci Iridium-192 sealed source. The jobsite was a large laydown yard in an industrial setting. The pipe fell from the stands and hit the guide tube. The guide tube was hit about 2 inches from the exposure end, and attempts at retracting the source were unsuccessful. The radiographers re-established boundaries and contacted the local (Assistant) RSO (ARSO) from the new boundary at approximately 1200 (EDT). The ARSO is authorized to perform source retrieval for JANX. (The licensee) maintained surveillance while awaiting his arrival. The ARSO arrived onsite at 1240 (EDT), assessed the situation and interviewed the crew. He noted that the source tube appeared to have 2 crimps. The ARSO was in contact with the Corporate RSO in Michigan during the operation. The ARSO used lead blankets to shield the source and surveyed. Survey revealed need for more shielding that was delivered to the site by (1345 EDT) and acceptable dose rate was achieved. He proceeded to remove the crimped section from the guide tube, observed that the drive cable was unaffected, connected the tube together with tape, and was able to retract the source into the exposure device. The event concluded by (1520 EDT). The licensee will be making a full report with corrective actions within 30 days of the occurrence. New Jersey Event: NJ 11003 |
ENS 46875 | 23 May 2011 23:41:00 | On May 23, 2011, at 2200 CDT, the station needed to notify the State of Nebraska, Department of Environmental Quality and the National Response Center due to an oil leak from the stations lube oil system. A quantity of oil spilled into the turbine building basement and an undetermined amount drained to the turbine building sump which discharges to the Missouri River. Some amount of oil was discharged to the river, which resulted in discoloration and a visible sheen noted on river sampling. Per Fort Calhoun Nuclear Station's National Pollutant Discharge Elimination System (NPDES) permit, the notifications will be made and samples will be taken for off site analysis to determine the quantity discharged involved. Also, the oil quantity in the turbine building sump (out fall L02) is expected to have exceeded 20 ppm oil and grease which does violate the State NPDES permit NE00000418. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi), 4 hour non-emergency notification due to a notifications being made to Government (Agencies) (State of Nebraska, Department of Environmental Quality and the National Response Center). The licensee notified the NRC Resident Inspector. |
ENS 46851 | 13 May 2011 12:39:00 | The following information was received by e-mail: Empty (Model) LR-230 shipping containers received from off-site where one (1) LR-230 had removable surface contamination above limits for alpha/beta activity on the outer surface of the container (~7,800 dpm/100 sq cm alpha and 13,300 dpm/100 sq cm beta). The LR-230 shipping containers are used to transport uranyl nitrate (<5.0 wt % U-235). Contamination was not related to container integrity. Contamination appears to be related to minor drips during unloading. Areas have been successfully decontaminated. There were no actual or potential safety consequences to workers, the public, or the environment. The licensee notified the NRC Resident Inspector. |
ENS 47036 | 8 July 2011 14:24:00 | The following information was received by facsimile: After a valve failed a leak test, it was determined that a fixture to machine the valve disc was missing one of the locating pins that maintained the integrity of the shaft bore centerline when machining. It was determined that the pin was missing from the fixture from January 8, 2011, to April 7, 2011 . Fisher has conducted a review and determined that the only ASME Section III code and commercial grade dedicated parts machined during this time that would use this fixture were the parts used for the orders referenced (below). Fisher Controls has implemented a corrective action program to address this issue. Although this notice is dated July 5, 2011 , Palo Verde was officially notified May 13, 2011, when Fisher first became aware of this issue. In the mean time, all affected discs have been returned to Fisher Controls for evaluation. Subject: NPS 16 Type A11 Disc/Shaft Assemblies for Palo Verde Orders 500541604 and 500548617 (Fisher Orders 019-F514977 and 019-F515346). |
ENS 46841 | 11 May 2011 15:50:00 | During an Emergency Preparedness drill on May 11, 2011 the Hatch Nuclear Plant's Technical Support Center (TSC) HVAC (Heating, Ventilation and Air-Conditioning) system's Roll Filter Indicating Light on the 1X75B001 TSC Air Handling Unit was illuminated. The ARP (Alarm Response Procedure) response actions in the 73EP-EIP-016-0 indicated that the roll filter needed to be replaced. Reference (Hatch) Condition Report 2011106603. The TSC HVAC is being considered non-functional during the performance of this corrective work activity. If an emergency condition occurs during the time these work activities are being performed, which requires activation of the TSC, the contingency plan calls for utilization of the TSC, as long as radiological conditions allow for habitability of the facility. Procedure 73EP-EIP-063-0, Technical Support Center Activation, provides instructions to direct TSC management to the Control Room and TSC support personnel to the Simulator Building to continue TSC activities if it is necessary to relocate from the primary TSC so that TSC functions can be continued. This event is reportable per 10CFR50.72 (b)(3)(xiii) as described In NUREG-1022, Rev. 1 since this activity affects the functionality of the TSC emergency response facility for the duration of the evolution. The licensee notified the NRC Resident Inspector.
The TSC has been returned to service following completion of filter maintenance. The licensee notified the NRC Resident Inspector. Notified the R2DO (Guthrie). |
ENS 46805 | 2 May 2011 11:28:00 | At 0626 (CDT) on 05/2/2011, with BFN U1 in Mode 4, Browns Ferry Nuclear Plant, received an 'A' Emergency Diesel Generator output breaker trip for unknown reasons that de-energized the 'A' 4kV Shutdown Board, resulting in a loss of power to RPS 'A' and a subsequent half scram, 'B' CREV and 'B/C' SGT automatic initiation, and Primary Containment isolation for Groups 2, 3, 6 and 8. The PCIS isolation (Group 2) also caused a loss of Shutdown Cooling on U1 which was restored at 0723 following restoration of power to the 'A' 4kV Shutdown Board from offsite 161kV power. The general containment isolation signals affecting containment isolation valves in more than one system is reportable as an 8 hour notification to the NRC IAW 10CFR50.72(b)(3)(iv), as any event or condition that results in a valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), except when the actuation results from and is part of a pre-planned sequence during testing or reactor operation. This is also reportable as 60 day written report IAW 10CFR50.73(a)(2)(iv). The NRC Resident (Inspector) has been notified of this event. This event was entered into the licensee's Corrective Action Program as PER# 362340. |
ENS 46910 | 1 June 2011 17:03:00 | The following information was received by fax: On May 31, 2011 the Department (Wisconsin Radiation Protection Section) received notification via email from the licensee's RSO that the licensee had reported a portable nuclear gauge stolen from their storage location in Lodi, WI (Columbia County). The Department called the licensee for additional information on June 1, 2011. The gauge was a Seaman model R-50 roofing gauge containing a maximum of 40 mCi of Americium-241:Be. In the call the licensee stated that on May 1, 2011 they had discovered their storage location had been broken into, their security barriers to the gauge had been defeated, and their portable nuclear gauge had been removed from the premises. Law enforcement had been notified on May 1, 2011 and investigated the theft. The gauge was recovered on May 28, 2011 in Madison, WI and had already been returned to the licensee at the time this conversation took place. The Department performed a reactive inspection on June 1, 2011, and confirmed that the gauge was recovered in good condition and placed in secure storage. The Department is waiting for additional information regarding this theft from both the Madison Police Department and the Columbia County Sheriff's Office. |
ENS 47038 | 8 July 2011 15:45:00 | The following information was received from the Commonwealth by facsimile: The RSO reported that a patient was 60% underexposed by an Ir-192 remote after-loader. The prescribed dose is not known. The error occurred because the physicist did not calculate the effect of the tube used to deliver the after-loader source. The target organ was the skin. 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. |
ENS 46777 | 22 April 2011 14:21:00 | A non-licensed contractor supervisor tested positive for alcohol during a follow-up fitness-for-duty test. The employee's access to the plant has been suspended. Contact the Headquarters Operations Officer for additional details. The licensee notified the NRC Resident Inspector. |
ENS 46778 | 22 April 2011 15:29:00 | On April 21, 2011, at approximately 1530 CDT, a malfunction occurred when a repaired gauge was being installed on a process vessel at the licensee's facility. The device has two 2-curie Cesium-137 sources, each attached to a separate chain, that are lowered into place in the insertion tube via hand crank. When the first source was being lowered into the tube, the mechanism seized and the source could not be raised or lowered. The device was removed from the vessel, the chain was cut, and the source was pulled up and placed into shielded container. The device and two sources were transported back to the service company for repair. Gauge information: Thermo Scientific (TFS) model 5220. Device SN: B-36. Source SNs: MB-3946 and MB-3956. Texas Incident: I-8837 |
ENS 46774 | 21 April 2011 19:16:00 | A manual trip of Salem Unit 1 was initiated due to a loss of circulating water pumps from heavy grassing at the circulating water intake. All rods fully inserted on the trip and all systems responded as designed with decay heat being removed via the steam dump system. All Auxiliary Feed Water (AFW) pumps auto started as expected on low steam generator levels from the trip. During a period of heavy grassing with one circulating water pump out of service for maintenance and another out of service for water box cleaning, a third circulating water pump's screen stopped and this circulating water pump was emergency tripped. When this circulating water pump was emergency tripped, the adjacent circulating water pump received a large intake of grass and automatically tripped. The abnormal operating procedure provides guidance to trip the plant if less than three circulating water pumps are in service and power is above P-10. At this point the manual reactor trip was initiated. The crew entered EOP-TRIP-1, appropriately transitioned to EOP-TRlP-2 and stabilized the plant at no load conditions. Salem Unit 1 is currently in mode 3. Reactor Coolant System temperature is 547F with pressure at 2235 psig. All ECCS and ESF Systems are available. Salem Unit 2 is currently defueled in a planned refueling outage and there is no impact to Salem Unit 2. No personnel injuries have occurred as a result of the trip. This event is reportable per 10CFR50.72(b)(3)(iv)(A) due to the auto start of the AFW pumps. This event is reportable per 10CFR50.72(b)(2)(iv)(B) due to the manual reactor trip. The plant is aligned with a normal electrical line-up from offsite power sources. The Licensee has notified the NRC Resident Inspector. The Licensee will notify Lower Alloways Creek Township and the States of New Jersey and Delaware.
During the post-trip data review for the manual reactor trip that occurred on April 21, 2011 at 1601 (EDT), 2 additional automatic Auxiliary Feedwater (AFW) pump start signals were identified. Operators were controlling AFW level to maintain steam generator levels within the band required by the emergency operating procedures. At approximately 1615 (EDT) the 13 steam generator level had cleared the low level setpoint but subsequently dipped below the low level setpoint at 1623 (EDT). At approximately 1640 (EDT), the 14 steam generator had recovered above the low level setpoint and subsequently dipped below the low level setpoint. The Unit 1 reactor had been previously tripped at 1601 (EDT) and all AFW pumps had automatically started. The 13 turbine driven AFW pump was stopped at approximately 1625 (EDT). The two motor driven AFW pumps continued to run throughout this event. Low level in one steam generator generates an automatic start of the motor driven AFW pumps and is reportable per 10CFR50.72(b)(3)(iv)(A). The licensee notified the NRC Resident Inspector and will notify Lower Alloways Creek Township and the States of New Jersey and Delaware. Notified the R1DO (Cook). |
ENS 46773 | 21 April 2011 16:57:00 | On (Thursday), 4/21/2011, at 1405 (EDT), Oconee Nuclear Station notified the South Carolina Department of Health and Environmental Control (SCDHEC) that sanitary sewage had been collected and inadvertently discharged without processing through the sanitary sewage system. The event was identified at 0730 (EDT) on 4/21/2011. The suspected cause is plumbing blockage. It is estimated that intermittent discharges over approximately three months resulted in a total discharge of approximately 4500 gallons. The Oconee NRC Senior Resident Inspector was notified. This event notification is being made pursuant to 10 CFR 50.72 (b)(2)(xi) due to notification of SCDHEC. Immediate corrective action was taken to isolate the affected portion of the sanitary sewage system. This event has no significance with respect to the health and safety of the public. |
ENS 46744 | 10 April 2011 06:40:00 | On April 09, 2011, at 2349 (EDT), Beaver Valley Power Station (BVPS) Unit No. 2 was operating at 15% power while preparing to synchronize the main unit generator to the grid. At that time, the 'A' Auxiliary Feedwater Injection Header was declared inoperable due to a water leak identified from a vent valve fillet weld between the inside and outside Containment Isolation Valves (outside of containment) for containment penetration X-79. In accordance with Technical Specification 3.7.5, Auxiliary Feedwater (AFW) System, Condition D, at 0345 (EDT), April 10, 2011, BVPS Unit 2 commenced a Reactor Shutdown to Mode 3. Required action is to be in Mode 3 within 6 hours. This event is being reported as a Technical Specification required shutdown pursuant to 10CFR50.72(b)(2)(i), 4 hour notification. Repairs are in progress. The following additional shutdown actions may be required from the time the Injection Header/Containment Penetration was declared inoperable: Technical Specification 3.7.5, Condition D, Mode 4 in 18 hours and Technical Specification 3.6.1, Condition A, Mode 5 within 37 hours. This event is also being reported as a degraded condition for Containment pursuant to 10CFR50.72(b)(3)(ii)(A), 8 hour notification. Additionally, at 0357 (EDT), during the Reactor Shutdown, at 4.6% Reactor Power, the BVPS Unit 2 Reactor was manually tripped due to reaching a pre-established manual trip criteria of 25% Steam Generator Level for the 21A Steam Generator. This was conservative criteria set above automatic actuation setpoint of 20.5% level. This event is being reported as a RPS Actuation pursuant to 10CFR50.72(b)(2)(iv)(B), 4 hour notification. Control room personnel entered Emergency Operating Procedure E-0, 'Response to Reactor Trip and Safety Injection.' Safety systems and equipment functioned as designed following the manual reactor trip. Due to the cooldown and subsequent shrink of level in the 21A Steam Generator, an automatic start of the Steam Driven Auxiliary Feedwater Pump (2FWE-P22) occurred at 20.5%. This event is being reported as an Auxiliary Feedwater System Actuation pursuant to 10CFR50.72(b)(3)(iv)(A), 8 hour notification. All control rods fully inserted into the core. The plant electrical system is aligned to normal offsite power sources. Decay heat from the reactor coolant pumps is being directed to atmospheric dump valve. There is no primary to secondary leakage. There was no impact on Unit 1. The licensee notified the NRC Resident Inspector. |
ENS 46883 | 25 May 2011 11:36:00 | On Saturday, April 2, 2011, Unit 2 refueling outage activities were in progress. The 2A RPS/UPS Static Inverter was out of service and bypassed with loads transferred to the primary alternate power supply. At 1218 hours, a post maintenance test was performed on the secondary alternate power supply. The inverter alternate power manual transfer switch was transferred from the 'primary alternate' to 'secondary alternate' position to support the post maintenance test. Since the transfer switch is 'break before make' the alternate power supply was interrupted momentarily. This deenergized the 2A RPS/UPS power distribution panel loads including the Division IA and IIA RPS relays and Division IA and IIA NSSSS (Nuclear Steam Supply Shutoff System) relays. Primary containment isolation valves (PCIVs) automatically closed on more than one system. The IB and IIB channels were unaffected. The portion of the primary containment isolation system that received an actuation signal functioned successfully. All of the affected open isolation valves automatically closed. The isolation was a partial actuation of the isolation actuation instrumentation. This 60-day ENS report is being made per 10CFR 50.73(a)(2)(iv)(A) and 10CFR50.73(a)(1) to report invalid automatic actuations of systems listed in paragraph (a)(2)(iv)(B). The listed system that actuated was general containment isolation signals affecting containment isolation valves in more than one system. Primary containment isolation valves (PCIVs) closed on drywell chilled water (DWCW), reactor enclosure cooling water (RECW), primary containment instrument gas (PCIG), and suppression pool cleanup. The licensee has notified the NRC Resident Inspector. |
ENS 46712 | 30 March 2011 17:00:00 | An extent of condition review of Braidwood Unit 2 unplanned loss of safety system annunciators Emergency Plan Unusual Event on March 24, 2011 (ENS number 46694) was performed for both Units of Braidwood Station. During this review it was identified that a previous unknown loss of annunciators had also occurred on August 10, 2010 from 1024 to 1136 CT on Unit 2. This condition occurred during planned maintenance on annunciator cabinet 2PA19J power supply capacitors. The maintenance performed on August 10, 2010 would normally not cause a loss of all Unit 2 annunciators. During the work, it was expected to lose approximately one third of the annunciators. Latent annunciator system problems identified from the March 24, 2011 event caused a loss of all Unit 2 annunciators and contributed to this condition being unknown to Main Control Room operators. All Unit 2 indications and computer points to the sequence of events recorder remained available and Unit 2 was stable during this timeframe. At 1538 CT on 3/30/11, it was determined that the August 10, 2010 condition met the threshold for Emergency Action Level MU6, UNPLANNED loss of most or all safety system annunciation or indication in the control room for greater than 15 minutes. This notification is being made as an undeclared Unusual Event Emergency Plan Classification per 10 CFR 50.72(a)(1)(ii). Per NUREG 1022, a 1- hour notification is required when a condition existed which met the emergency plan criteria but no emergency was declared and the basis for the emergency class no longer exists at the time of the discovery. The licensee notified the NRC Resident Inspector. |
ENS 46711 | 30 March 2011 16:24:00 | City of San Clemente inadvertently secured power to 19 Community Alert Sirens. Trouble alarms were received at 0700 (PDT) but it is not known at what time specifically power was removed. All sirens were functional the day before when an inspection was performed on 3/29/2011 at 1213 (PDT). Power was restored to the sirens at 1145 (PDT) 3/30/2011. This event resulted in 20 community sirens being non-functional for greater than 1 hour. Siren State Parks 4 (SP-4) was also non-functional for unrelated reasons and remains out of service. Power was restored and all but the state park siren capabilities were declared operable. The licensee notified the NRC Resident Inspector. |
ENS 46779 | 22 April 2011 18:07:00 | The following information was received by e-mail from the State of Nevada: A vial with serial # S358015-011, containing 198 microcuries of Cs-137, belonging to the licensee, is unaccounted for. Cardinal Health removed all equipment from the licensee's office on the last day of patient care and stored it at Advanced Isotopes. The source was originally from Biotech (currently Cardinal Health). The licensee allegedly tried working with Cardinal Health to clear this up, but to no avail. The licensee plans to contact the RSO of Advanced Isotopes, who is currently out of town, and let the (Nevada) Radiation Control Program know if they have any additional information. The matter is being taken very seriously and is being investigated. This incident is documented as Nevada Incident- NV110012 and was initially reported to the State of Nevada on 4/12/2011.
The following update was received from the State of Nevada via email: NV 110012 has been closed. The source has been deemed missing. The license had been revoked. The Responsible Person for the license had been prohibited from being listed on a RML in Nevada, for a period of 3 years, from the date of revocation, though his eligibility to perform work under the supervision of an authorized user is not affected by this prohibition. Since the radiological material license is being revoked, not terminated, the licensee remains responsible for the proper disposal of the lost Cs-137 source. Should the source be found or discovered, the licensee is responsible and liable for the costs associated with disposal of the source, any damages or losses caused by source, and, if necessary, any cleanup, remediation, and/or decommissioning efforts that may be required as a result of the source. Notified R4DO (Proulx) and FSME (Hsueh). 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 |
ENS 46687 | 20 March 2011 07:08:00 | During testing of the High Pressure Core Spray (HPCS) system minimum flow valve breaker, it was discovered that the overcurrent trip setpoint was out of tolerance. This testing was being performed as a result of a breaker trip that occurred during a surveillance. The discovery occurred while the system was inoperable for maintenance and no TS (Technical Specification) limits or action times were exceeded. The overcurrent trip setpoint was placed within tolerance and HPCS is in operable status. The licensee notified the NRC Resident Inspector. |
ENS 46685 | 19 March 2011 06:54:00 | Following a scheduled plant shutdown for refueling the operators were forced to close the Main Steam Isolation Valves (MSIV's) to limit plant cooldown. While opening MSIV's to restore steam to the secondary, a Reactor Trip and Safety Injection (SI) occurred. The MSIV bypass valves were opened to equalize pressure across the MSIV's. Steam header pressure dipped when the MSIV for 'C' Steam Generator (S/G) was opened. The low steamline pressure bistables are rate sensitive and actuated to cause the SI when steam pressure dipped. Lowest steamline pressure was 1040 psig, the low steam line pressure SI actuates at 615 psig. During the SI the PZR (Pressurizer) PORV's cycled approximately 10 times to limit RCS pressure. When the PORV's opened the 'B' PZR Code Safety Main Control Board (MCB) and plant computer alarm actuated but the actual MCB indication did not change nor does plant response indicate that a PZR Code Safety opened. This appears to be an indication problem related to the PORV's cycling. All equipment functioned as required. The station electric buses are aligned to normal offsite power. Decay heat removal is being discharged to the atmospheric relief valves with no indication of primary to secondary leakage. The licensee notified the NRC Resident Inspector.
1. The expected system actuations that occurred when the plant experienced a Safety Injection (SI) 03/19/11 at 04:04 CDT, previously reported on EN 46685 for 10CFR50.72(b)(2)(iv)(A), is also reportable under 10CFR50.72(b)(3)(iv)(A) for Specified System Actuation. 2. During the recovery of the Safety Injection (SI) actuation that occurred 03/19/11 at 04:04 CDT and previously reported on EN 46685, the Safety Injection Signal was reset which blocked any further automatic actuation. This was directed per the appropriate procedure step. There is no Technical Specifications allowed condition for both trains of ECCS to be inoperable, therefore the unit entered Tech. Spec. 3.0.3 due to the Auto SI feature being blocked. LCO 3.5.2 action C.1. directs immediate entry into LCO 3.0.3. The entry into TS 3.0.3 was made at 0411 CDT and exited at 0639 CDT when the Reactor Trip Breakers were reclosed which re-enabled the automatic SI signal. This is reportable under 10CFR50.72(b)(3)(v)(D) for Accident Mitigation. NRC Resident was notified of the update. Notified R4DO(Cain). |
ENS 46647 | 28 February 2011 10:00:00 | During the dedication process of nuclear E7000 relays, Tyco Electronics experienced an issue with a component. This component, a coil internal to the relay, failed to operate during quality assurance testing. Tyco purged this lot of coils from their inventory and also from all "work in process." Tyco was able to account for all the coils in this lot of E7000 relays with serial numbers beginning with 1046. Tyco records indicate that some relays shipped to the Exelon/Quad Cities facility were made using this specific coil lot. Tyco sold this safety-related item for specified and unspecified applications. |
ENS 46646 | 28 February 2011 03:45:00 | At 2159 CST on 2/27/2011, during the testing of the 'A' Train Safety Injection System, an Auxiliary Operator in the field identified that the oiler for 2P-15B, 'B' Train Safety Injection Pump, had rotated and the oil had drained out of the oiler. The Auxiliary Operator immediately reported this condition to control room personnel. The 'B' train safety injection pump was declared inoperable and LCO 3.0.3 was entered based upon the condition of both trains of safety injection being out of service. The Unit 2 'A' Train Safety Injection System was being tested in accordance with inservice testing procedure IT-535C, Leakage Reduction and Preventive Maintenance Program Train 'A' HHSI and RHR Piggyback Test Mode 1,5,6 (Refueling) Unit 2, which placed the Unit 2 ECCS in TSAC (Technical Specification Action Condition) 3.5.2.A, One ECCS Train Inoperable. Unit 2 exited LCO 3.0.3 at 2211 CST, upon completion of the 'A' train inservice test. The 2P-151B safety injection pump remains inoperable in accordance with TSAC 3.5.2.A, One ECCS Train Inoperable for troubleshooting and repair. This condition is reported in accordance with 10 CFR 50.72(b)(3)(v)(D) Accident Mitigation. There was no impact on Unit 1 and the licensee notified the NRC Resident Inspector. |
ENS 46627 | 19 February 2011 22:24:00 | At 1920 (EST) on 2-19-11, a chemistry technician notified the Shift Manager that more than the expected amount of Sodium Bisulfite was inadvertently added to the Circulating Water System discharge. The Spill Prevention Control & Countermeasures (SPCC) Plan was referenced for required actions. 17 additional gallons of Sodium Bisulfite was discharged to surface waters of Lake Michigan. This amount of Sodium Bisulfite is less than the reportable quantity of 417 gallons per the SPCC Plan. However, since it was released to surface water (Lake Michigan), a notification to the District Water Quality Division (DWQD) was made at 2020 (EST) by the site's Senior Environmental Specialist (SES). The DWQD requested that additional notifications be made to state (Pollution Emergency Alert System - PEAS) and local (911 operator) agencies. PEAS was notified at 2020 (EST) by the SES and Van Buren County dispatch (911) was notified by the Control Room staff at 2047 (EST). The licensee notified the NRC Resident Inspector. |