Semantic search
| Entered date | Event description | |
|---|---|---|
| ENS 49541 | 14 November 2013 20:35:00 | During analysis of Watts Bar Nuclear (WBN) Unit 2 fire protection features, it was revealed that a potential fire induced failure of centrifugal charging pumps could occur in Unit 1. Specifically, a potential fire induced failure of both Unit 1 Chemical and Volume Control System centrifugal charging pumps (CCPs) (1-PMP-62-108-A and 1-PMP-62-104-B) could occur due a fire in either auxiliary building room 737.0-A1 (general area for elevation 737.0) or 757.0-A2 (6.9 kV and Shutdown Board Room A). It is postulated that a fire in these rooms could cause a spurious closure of the CCP suction valve (1-LCV-62-133-B) from the volume control tank (VCT) (1-TANK-62-129) and could disable the control circuit which opens the flow from the refueling water storage tank (RWST) suction valve (1-LCV-62-135-A). The fire safe shutdown analysis (Fire Protection Report, Part VI) currently addresses this occurrence via the performance of a prompt main control room operator action to open the RWST suction path. However, this procedurally directed action may require several minutes to complete and due to the potentially short duration (possibly as short as a few seconds) for CCP survivability without suction flow, the action has now been determined to be unacceptable. As a result, the loss of charging flow could result in a loss of injection to the reactor coolant pump (RCP) seals which could subsequently lead to a RCP seal failure and a small break loss of coolant event. WBN engineering is continuing to validate whether the CCP minimum flow recirculation would protect the pumps with both suction paths (VCT and RWST) isolated and with the reactor at normal operating pressure. WBN has established compensatory measures to ensure that a fire in affected rooms will not cause a spurious closure of the CCP suctions valves. The licensee has notified the NRC Resident Inspector. |
| ENS 49542 | 15 November 2013 13:27:00 | On the afternoon of 11/14/2013, a patient was scheduled for a Y-90 microsphere radioembolization treatment at Indiana University Medical Center under NRC License 13-02752-03. The treatment consisted of two separate doses of Y-90 for which two separate written directives were prepared. Segment 4 of the patient's liver was prescribed a dosage of 27.0 mCi, and the right lobe of the liver was prescribed a dosage of 88.0 mCi. At 13:50 on 11/14/2013, following measurement of the remaining activity after injection of the Y-90 microspheres, it was determined that a dose of 19.5 mCi was delivered to segment 4 (72.2% of the intended dose). Shortly thereafter, at 14:06, a dose of 87.5 mCi of Y-90 was delivered to the right lobe (99.4% of the intended dose). Both procedures appeared to proceed in accordance with standard operating procedures, and no abnormalities were identified during the procedure by the Interventional Radiology Physician or the Health Physicist supporting the procedure. Following the procedure, personnel and area surveys were performed using an SE International GM meter and no contamination of personnel, the room or equipment was identified. The container holding residual activity from the segment 4 treatment has been set aside for decay and further analysis. As the activity delivered to segment 4 of the liver meets the criteria in 10 CFR 35.3045(a)(1) and 10 CFR 35.3045(a)(1)(i), a report to the NRC Operations Center shall be made in accordance with 10 CFR 35.3045(c). The attending physician and patient will be contacted in accordance with 10 CFR 35.3045(e). 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 49539 | 14 November 2013 16:08:00 | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits has determined the described condition is applicable to the Salem Nuclear Power Plant resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R. The original plant wiring design and associated analysis for ammeters associated with the station batteries are not provided with overcurrent protection features to limit the fault current. A postulated fire that results in a short to ground concurrent with an opposite polarity short from the same battery could result in excessive current flow (i.e., heating) in the ammeter wiring. This excessive current could result in a secondary fire in another fire area. The secondary fire could adversely affect safe shutdown equipment and cause loss of the ability to conduct a safe shutdown as required by 10 CFR 50 Appendix R. The areas affected are the Control Room, Relay Rooms and 460 Volt Switchgear Rooms. This condition is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). Interim compensatory measures (i.e., fire watches) have been implemented for the affected areas of the plant. The licensee has notified the NRC Resident Inspector. |
| ENS 49537 | 14 November 2013 14:40:00 | The following report was received from the Washington Department of Health (WA DOH) via e-mail: The licensee was performing an administration of Nordion TheraSpheres, a procedure performed without incident for over sixty administrations so far, when measurements indicated that an inordinate amount of the material remained in the waste/tubing. The licensee has confirmed the catheter used had an internal diameter (ID) of 0.68mm, which equals and exceeds the manufacturer's specifications of equal or greater than 0.5 mm ID. 1. On Friday, November 8, 2013 at approximately 1040 (PST), the RSO received notification of a possible medical event involving the administration of Y-90 microspheres (TheraSpheres). The Radiation Safety Specialist contacted WA DOH at approximately 1100 on Friday, November 8, 2013 to report a medical event in accordance with requirements in WAC 246-240-651. 2. The routine procedure for Y-90 microsphere administration requires the measurement of the materials used for the administration at a fixed geometry both before and after administration. The ratio of the exposure rate measured (minus background) indicates the percentage of the microspheres remaining in the tubing. This value subtracted from the originally prescribed activity determines the percentage of activity actually administered to the patient. The prescribed dose was 129 Gy to the left lobe, equivalent to an administered activity of 5.0 GBq. The measured activity (via Capintec CRC-15R) was 5.04 GBq which resulted in a pre-administration exposure rate of 5.3 mR/hr. Post-administration, the residual waste exposure rate measurement was 3.8 mR/hr (using the same geometry). The post-administration measurement was taken about 2 hours after the pre-administration measurement. Based on the post administration measurement, it is estimated that at least 73% of the prescribed dose was still present in the waste materials implying that only 27% of the prescribed dose was administered or 1.36 Gbq which would result in a target dose of 35 Gy. 3. Investigation into the root cause indicated that the use of a Surefire Catheter may have been the underlying factor of this medical event. The interventional radiologist (RA) reported that this was the first time he had used the Surefire Catheter in conjunction with a TheraSphere case. This catheter was chosen due to medical need. The interventional radiologist wanted to minimize the amount of auxiliary embolization required for this case and this catheter satisfied that requirement. Prior to administration, contrast was administered to verify the integrity of the infusion system. No issues were noted during the contrast administration. During the administration of the microspheres, the interventional radiologist noted that the feel of the syringe was different from past administrations and that it was more difficult to push the plunger, however it did appear that the infusion was occurring. After completion of the infusion, the interventional radiologist noted that the syringe plunger pushed back. Final measurement of both the patient and the waste materials by medical physics and nuclear medicine indicated that the dose was not properly infused and consequently an underdose had occurred. 4. During the investigation it was noted that this case was the first time the Surefire Catheter was utilized. There were no other changes in the set up. The vendor representative mentioned that he had previously observed issues with the use of this catheter, however there are no documents or other notices issued with regards to catheter usage combinations. Review of the infusion materials did not reveal any physical issues with the setup, though based on the physician's report regarding the feel of the plunger during the infusion, it is possible that there was a kink or other similar issue in either the catheter or infusion system that resulted in incomplete administration of the TheraSphere dose. 5. The interventional radiologist communicated directly with the patient regarding this medical event. Both the interventional radiologist and the radiation oncologist (who is also the physician authorized user for this material) do not anticipate any additional medical issues that would be a result of this incomplete administration. This is based on the history of the patient who had previously received a TheraSphere administration to his right primary lobe vs. the current administration to the left secondary lobe. 6. In order to minimize the possibility of future medical events, the Surefire Catheter will not be used for future procedures. Since there have been no cases during our experience with this procedure prior to this event that have resulted in an underdosing due to equipment malfunction and since the only significant change in equipment set up was the use of the Surefire catheter, it is expected that a return to our previous catheter will insure that underdosing will not occur. 7. In accordance with WAC 246-240-651 a full report regarding this event has been transmitted to WA DOH. Washington Incident : WA-13-055 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 49502 | 5 November 2013 05:00:00 | At 0041 hours (EST) on 11/05/2013, while placing the Condensate Polishers into service, a secondary side perturbation occurred, resulting in a loss of the one running Main Feedwater Pump ('A' Pump) on low suction pressure. At the time, the plant was in Mode 2 with Startup Low Power Physics Testing in progress. By design, this condition resulted in an automatic start of Auxiliary Feedwater. Both 'A' and 'B' Motor-Driven Auxiliary Feedwater (AFW) Pumps started as designed. Steam generator water levels were maintained by the Auxiliary Feedwater flow. The 'B' Motor Driven AFW pump was secured following its automatic start to stabilize steam generator water levels and reactor coolant system temperature. Plant conditions including steam generator water levels have been stabilized. At 0250 hours (EST) 'A' Main Feedwater Pump was restarted and at 0252 hours (EST), the 'A' Motor-Driven AFW pump was secured. At this time, the cause of the secondary side perturbation is being investigated. The plant remains in Mode 2 and Startup Low Power Physics Testing has resumed. Due to the valid actuation of AFW, this event is being reported as an 8-hour non-emergency per 10CFR50.72(b)(3)(iv)(A). At no time during this occurrence was the public or plant staff at risk as a result of this event. The NRC Resident Inspector has been notified. |
| ENS 49497 | 2 November 2013 01:44:00 | At 2147 CDT on 11/1/2013, Unit 2 Reactor tripped during Solid State Protection System Slave Relay Testing. This test utilizes a blocking circuit to verify the operability of the slave relay which trips the Main Turbine and both Main Feedwater pump turbines, on a Hi-Hi Steam Generator level or Safety Injection. No valve actuation is expected to occur. While positioning the Slave Relay switch in a testing lineup, the relay actutated. The Unit 2 Turbine tripped as well as both Main Feedwater Pumps. The Turbine Trip actuated the Reactor Trip since power was above 50%. The trip of both Main Feedwater Pumps started both Motor Driven Auxiliary Feedwater Pumps. The Steam Generator Lo Lo Levels started the Turbine Driven Auxiliary Feed Water Pump. All systems responded as expected. Currently, Unit 2 is being maintained in Hot Standby (Mode 3) in accordance with Integrated Plant Operating Procedure IPO-007B and the Emergency Response Guideline Procedure Network has been exited. Decay Heat is being rejected to the Main Condenser via Steam Dump Valves (Turbine Bypass Valves). The licensee has notified the NRC Resident Inspector. |
| ENS 49489 | 31 October 2013 05:00:00 | On October 31, 2013 at 0251, Secondary Containment Zone I (Unit 1 Reactor Building) differential pressure was lost following a routine transfer of Reactor Protection System Power supplies. Upon restoration from the power supply transfer, one of the Reactor Building Equipment Compartment Exhaust Fans tripped. There were no obvious malfunctions associated with the equipment and fan was able to be restarted. Zone II (Unit 2 Reactor Building) and III (Common Refuel Floor Area) ventilation remained in service and stable. Zone I differential pressure recovered within a few minutes and was verified to be stable. LCO 3.6.4.1 was entered for both units at 0251 and exited at 0255. Tech Spec Secondary Containment Operability requires a negative pressure of at least 0.25 inches water gauge. There have been no further perturbations in differential pressure and secondary containment remains operable. This event is being reported under 10 CFR 50.72(b)(3)(v) 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 has notified the NRC Resident Inspector. |
| ENS 49493 | 1 November 2013 10:57:00 | The following report was received from the Kansas Department of Health and Environment via facsimile: Initial notification of an overexposure to the extremities of a radiopharmacy nuclear medicine technician at St. Francis Health Center, Topeka, KS, was made by the radiation safety officer. A Landauer report received 10/31/2013 indicated a right hand dosimeter (dose) at 55.85 rem and a left hand dosimeter (dose) at 54.29 rem. Whole body badge (dose) indicated 36 mrem. The tech has been removed from any job duties involving occupational radiation exposure. A more detailed report is being prepared (by the licensee). Kansas Report Number KS130009 |
| ENS 49471 | 24 October 2013 09:45:00 | At 0553 EDT on 10/24/2013, Oconee Unit 3 was manually tripped due to oscillations in the feedwater system in anticipation of an automatic reactor trip. At 0549 EDT, Unit 3 began experiencing small feedwater oscillations. The feedwater control system was placed in manual in an attempt to stabilize feedwater flows. Feedwater oscillations continued to grow in magnitude and at 0553 EDT, a manual trip was directed to prevent an automatic reactor trip. Due to an RPS actuation, this event is being reported as a 4 and 8 hour Non-Emergency per 10 CFR 50.72 (b)(2)(iv)(B) and 10 CFR 50.72 (b)(3) Following the reactor trip, four main steam relief valves failed to reseat. Procedure guidance was utilized to reduce main steam system pressure by approximately 30 psig to reseat the main steam relief valves. All main stream relief valves are now reseated. All other post trip conditions were normal and all other systems performed as expected. Unit 3 is currently in Mode 3 and stable. Operations have been stabilized on Unit 3. A post-trip investigation is in progress, per site procedures and directives. The licensee has notified the NRC Resident Inspector.
Duke Energy reviewed NRC Event Number 49471 against NUREG 1022, Rev 3, section 3.2.6, "System Actuation" and determined this event should have been reported only per 10 CFR 50.72 (b)(2)(iv)(B), RPS Actuation (while critical). Thus, Duke Energy is revising NRC Event Number 49471 to remove the 8-hour report criteria 10 CFR 50.72 (b)(2)(iv)(A). The NRC Resident Inspector was notified of this revised report. This update has no effect on safety significance. Notified R2DO (Shaeffer). |
| ENS 49465 | 23 October 2013 10:20:00 | On October 23, 2013 at 0620, Susquehanna Steam Electric Station operators observed secondary containment differential pressure was at negative 0.17 inches water gauge for Zone II (Unit 2 Reactor Building). Tech Spec Secondary Containment Operability requires a negative pressure of at least 0.25 inches water gauge. Zone I (Unit 1 Reactor Building) and III (Common Refuel Floor Area) ventilation remained in service and stable. Zone II differential pressure was restored to within the required band by manual damper adjustment in about 15 minutes and was verified to be stable. LCO 3.6.4.1 was entered for both units at 0620 and exited at 0635. This event is being reported under 10 CFR 50.72(b)(3)(v) 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 has notified the NRC Resident Inspector. |
| ENS 49438 | 14 October 2013 14:50:00 | A general license holder reported a leaking Nickel-63 source that is used in an Agilent Electron Capture Detector. The source was wipe tested and discovered to have a surface contamination of 0.011 micro-curies. The Ni-63 source for the detector has a rated source strength of 15 micro-curies. The company sent the source to a contractor for further determination if the source is leaking or can be cleaned and returned to service. |
| ENS 49443 | 15 October 2013 13:53:00 | The following report was received from the Colorado Department of Public Health and Environment Radioactive Materials Unit via e-mail: On Tuesday, October 15, 2013, at approximately 8:30 am MDT (Mountain Daylight Time), the Radiation Safety Officer (RSO) for a portable gauge licensee (Martinez Associates, LLC; Colorado License No. 1076-01, Amendment 09) based in Denver, Colorado notified the Colorado Department of Public Health and Environment (CDPHE) Radioactive Materials Unit that a Troxler moisture density gauge was stolen on Sunday, October 13, 2013 from a private vehicle parked at the private residence of an employee. The private residence was located in Thornton, Colorado. The stolen gauge was last used on Friday, October 11, 2013 and the theft was reported (by the employee) to the RSO and local law enforcement (Thornton, Colorado, Police) on Sunday, October 13, 2013 at approximately 8:30 pm MDT. The stolen gauge was a Troxler model 3430 (gauge serial number 23358), containing approximately 8 mCi (03/17/1994 assay date) of Cs-137 (source serial #75-5374) and 40 mCi (03/08/1994 assay date) of Am-241:Be (source serial #47-19240). CDPHE will conduct an investigation and evaluation of the circumstances surrounding the reported theft and intends to issue a press release regarding the incident. Further details and information from the licensee is pending.
On October 15, 2013, the Radiation Program of the Colorado Department of Public Health and Environment (CDPHE) reported to the NRC Operations Center that a gauge had been stolen from an apartment complex in Thornton, Colorado. The City of Thornton, Colorado Police reported to CDPHE at approximately 10 pm on October 15, 2013 that the gauge had been recovered and that suspects were in custody. The gauge was transported to the City of Thornton Police for further criminal evaluation. The CDPHE will assist the Thornton Police as they evaluate the device. Notified R4DO (Hay), FSME (Henderson), ILTAB (Wray) via email, and FSME Event Resource via email.
In our email dated 10/16/13 (7:43 am), it was incorrectly reported that the Thornton, Colorado Police Department (PD) had suspects in custody. The Thornton PD had questioned suspects at the time of their notification to our agency, but did not have the individuals in custody nor were any arrests made. All other information is correct to the best of our knowledge and the gauge remains in the custody of the police at this time. Notified R4DO (Azua), ILTAB (Wray) 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 49439 | 14 October 2013 15:16:00 | The following report was received from the Oklahoma Department of Environmental Quality via e-mail: On Oct. 10 we were informed by the Oklahoma Dept. of Transportation that one of their Troxler Model 3440 gauges (S/N 32291) had been run over at a road construction site on Hwy 412 east of Woodward, OK. The Radiation Safety Officer responded to the site that day, recovered the gauge and transported it back to their facility in Oklahoma City. Department of Environmental Quality staff inspected the gauge the next day. Although the source rod had been completely snapped off, it did not appear that either of the sealed sources had been damaged. |
| ENS 49419 | 8 October 2013 15:42:00 | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined the described condition to be applicable to Comanche Peak Nuclear Power Plant resulting in a potentially unanalyzed condition with respect to 10CFR50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1 E batteries control room ampere indications do not include overcurrent protection features to limit the fault current. In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane. Simultaneously, the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This could cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (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 10CFR50 Appendix R. This condition is reportable in accordance with 10CFR50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector. See related Event #49411. |
| ENS 49414 | 6 October 2013 14:26:00 | Today at approximately 1040 (EDT), during a planned reactor power ascension with reactor power at approximately 20% of rated thermal power, main condenser vacuum began to lower. In accordance with the abnormal operating procedure for degrading vacuum, Operators inserted a manual scram of the reactor at 1130 (EDT). The cause of the degraded vacuum is currently under investigation. All rods inserted into the core and all systems functioned as expected during the scram. No electromatic (EMRVs) or safety relief valves lifted during the transient. The plant is currently shutdown and parameters are stable. The plant is in its normal shutdown electrical lineup and decay heat is being removed via steam bypass valves to the main condenser. 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. The licensee has notified the NRC Resident Inspector and will be notifying state authorities. |
| ENS 49411 | 4 October 2013 18:20:00 | A review of industry operating experience regarding the impact of unfused Direct Current (DC) ammeter circuits in the control room has determined the described condition to be applicable to Palo Verde Nuclear Generating Station resulting in an unanalyzed condition with respect to 10 CFR 50 Appendix R analysis requirements. The original plant wiring design and associated analysis for the Class 1E Train B and D batteries and chargers (including the BD Swing charger) control room ampere indications do not include overcurrent protection features to limit the fault current. In the postulated event, a fire in the control room could cause one of the ammeter wires to hot short to the ground plane; simultaneously, the fire causes another DC wire from the opposite polarity on the same battery to also hot short to the ground plane. This would cause a ground loop through the unprotected ammeter wiring. This event could result in excessive current flow (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 10CFR50 Appendix R. The train B and D DC circuits were reviewed first since they are part of alternate safe shutdown capability for the control room fire event. An extent of condition review is ongoing for the A and C train DC circuits and other similar circuit designs that could potentially cause a secondary fire. This condition is reportable in accordance with 10 CFR 50.72(b)(3)(ii)(B) as an unanalyzed condition. Compensatory measures (fire watches) have been implemented for affected areas of the plant. The NRC Resident Inspector has been notified. |
| ENS 49410 | 4 October 2013 17:13:00 | The following report was received from the South Carolina Department of Health and Environmental Control via e-mail: The SC Department of Health and Environmental Control was notified on Friday, October 4, 2013, at 1117 hrs (EDT), that a CPN Model MC-1-DR had been damaged at 1110 hrs (EDT) on October 4, 2013. (The) RSO, stated that the gauge had been run over by a bulldozer during a density test. (The South Carolina) Duty Officer responded to the scene and arrived at 1223 hrs (EDT). The source rod was extended and the gauge had sustained considerable damage. The inspector placed a small lead pig over the exposed source rod and placed the source rod and the damaged components in the transport container. The inspector surveyed all parts and was unable to verify that the Am-241:Be source was still contained with the gauge. A thorough survey of the area was made and the area under the damaged gauge was dug up but the inspector was unable to verify that the Am-241:Be source was with the damaged gauge. The inspector traveled to the permanent storage location of the licensee to ensure that the gauge was properly secured. (The licensee) was advised to contact the gauge manufacturer for further instruction regarding disposal and to keep all personnel away from the gauge. (The licensee's RSO) was advised by (the South Carolina inspector) to submit a written report detailing this event to the Department within 30 days. The event is open and pending the licensee's investigation and report to the Department. Updates will be made through the national NMED system. |
| ENS 49406 | 3 October 2013 21:10:00 | The Browns Ferry Nuclear Plant (BFN) minimum Operations shift staffing was evaluated for response to a fire in the Control Bay that ultimately leads to entry into Appendix R Safe Shutdown Instructions (SSIs). The result of this evaluation revealed that the minimum Operations shift staffing does not provide sufficient staffing to support SSI required staffing levels. In the event of an Appendix R fire, one BFN unit would be without a Senior Reactor Operator to direct the implementation of the time critical manual actions specified in the SSI procedures. On October 3, 2013, this condition was determined to be an 8 hour non-emergency report in accordance with 10 CFR 50.72(b)(3)(ii)(B) since sufficient shift staffing levels to implement the SSI procedures in the event of an Appendix R fire were not provided. Compensatory measures have been implemented to ensure sufficient shift staffing is provided to implement Appendix R SSIs. The event was entered into the licensee corrective action program as Service Request number 788812. The NRC Resident Inspector has been notified. |
| ENS 49405 | 3 October 2013 15:46:00 | The following report was received from the Wisconsin Radiation Protection Section via e-mail: On October 1, 2013 the licensee identified a tritium exit sign as missing and reported it to the state of Wisconsin on October 2, 2013. Last inventoried during 2009, the sign was not recorded on an inventory conducted on September 17, 2013. The missing sign is a double-sided Isolite exit sign model 2040-07R-20BA SN: 227163. According to the SSD sheet, this device contains up to 25 Curies of tritium. The licensee stated it might have gone missing after a lighting retrofit project in 2010. Wisconsin Report ID: WI130020 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 49412 | 4 October 2013 17:19:00 | The following information was received from the Oklahoma Department of Environmental Quality via e-mail: On Sept. 27, (2013) the (Oklahoma Department of Environmental Quality) was notified by the University of Oklahoma Health Science Center that they had found an Electron Capture Detector with approximately 1 microCi of removable contamination. The unit in question is a Varian 3000 Series (P/N 02-001972-00) S/N A8075 (SSDR #CA-8253-D-80 1-B). Activity was 8 mCi (Ni-63) in Aug. 1990. The unit has been removed from service and is being stored prior to disposal. |
| ENS 49377 | 23 September 2013 10:25:00 | Ventilation process radiation monitors for the Turbine Building Stack Discharge (monitors HVR*RE10A and HVR*RE10B) will be out of service for preplanned maintenance. This is reportable in accordance with 10 CFR 50.72(b)(3)(xiii). The licensee stated that this condition is being reported based on the inability to sufficiently identify and classify an Emergency Action Level for radiation releases that utilize input from the affected monitors while they are out of service. The licensee will notify state and local authorities. The licensee will also notify the NRC Resident Inspector. |
| ENS 49376 | 22 September 2013 23:48:00 | On 09/22/2013 at 1940 CDT, Prairie Island Nuclear Generating Plant (PINGP) requested an offsite ambulance via the 911 system for medical assistance for an individual in the radiologically controlled area. The person was treated as potentially contaminated because a complete survey to confirm the absence of contamination was not completed prior to transport of the individual. An ambulance arrived on site at 1956 CDT and departed the site at 2036 CDT to the Red Wing Minnesota Hospital. PINGP radiation protection personnel accompanied the individual to the hospital. A survey at the hospital determined that the individual was not contaminated. This is considered a transport of a potentially contaminated individual requiring an 8 hour ENS Notification per 10CFR50.72(b)(3)(xii). The Prairie Island Indian Community was notified of the transport of an individual by ambulance. This is considered a notification of another government agency and an event that may have potential interest to the media requiring a 4 hour ENS Notification per 10CFR50.72(b)(2)(xi). The licensee notified the NRC Resident Inspector. |
| ENS 49364 | 20 September 2013 05:42:00 | On 9/19/13 at approximately 2230 EDT, Emergency Siren 602 inadvertently actuated. Berrien County Dispatch was notified by local residents at 2237 EDT. The cause of the actuation is under investigation but believed to be due to rain water intrusion. The siren was subsequently disconnected by a station technician to prevent further erroneous actuation. The siren remains out of service and is the only siren out of service within the 10 mile Emergency Planning Zone (EPZ). There are a total of 70 sirens. This notification is being made under 10 CFR 50.72(b)(2)(xi), Offsite Notification, as a four (4) hour report. The Operations Shift Manager was notified of the actuation on 9/20/13 at 0340 EDT. The licensee has notified the NRC Resident Inspector. |
| ENS 49356 | 19 September 2013 03:10:00 | While performing the secondary containment airlock door interlock surveillance, the interlock to the main plenum room did not prevent the opening of both doors to the plenum room airlock (DOOR-85 and DOOR-86). The plenum room airlock doors were immediately closed. The time both doors were opened is estimated to be approximately one (1) second. When both doors open, Technical Specification surveillance requirement SR 3.6.4.1.3 was not met and secondary containment was declared inoperable. Secondary containment was declared operable after independently verifying at least one secondary containment access door was closed. There were no radiological releases associated with this event. The NRC Resident Inspector has been notified.
This update provides additional information on the initial notification of the event. On 9/18/13, while testing secondary containment airlock doors, the interlocks did not prevent opening of both doors simultaneously. With the outer door to the main plenum room open, the inner door was able to be opened. At this point, Technical Specification SR 3.6.4.1.3 was not met and secondary containment was inoperable. The inner door was closed immediately. While in this condition, the inner door was then opened, and the interlock did not prevent the opening of the outer door. The outer door was closed immediately. Secondary containment was declared operable after verifying at least one of the airlock doors was closed. There were no radiological releases associated with this event. The NRC Resident Inspector has been notified. Notified R3DO (Reimer). |
| ENS 49341 | 13 September 2013 10:55:00 | At about 1045 EDT on 9/13/13, it was discovered that the feed tube level sensor on a press operation is not fail safe upon loss of signal. The sensor is a sole IROFS (Item Relied On For Safety) for a particular sequence. Criticality controls remained in place. Affected equipment has been shut down. No unsafe condition exists. Feed tube level sensor is in place for the sequence to limit mass. At no time was the mass limit exceeded. We are reporting under Part 70, Appendix A, (a)(4) which states that credited IROFS must remain available and reliable. We cannot evaluate reliability in the time required for a 1 hour report. In addition, the affected equipment has been secured. An investigation is underway to determine corrective actions and extent of condition. The license will notify NRC Region 2.
On 9/13/13, GNF-A conservatively made a 1 hour event notification (EN 49341) due to a discovery that a feed tube level sensor is not fail safe upon loss of signal. After further review, it has been determined that the control remained available, reliable and continued to meet performance requirements. As a result, the event notification is retracted. The licensee has notified NRC Region 2. Notified R2DO (Sykes), NMSS EO (Rubenstone), and IRD (Grant). |
| ENS 49318 | 3 September 2013 14:10:00 | During planned maintenance activities, station personnel simultaneously opened the inner and outer airlock doors from unit 2 reactor enclosure to the U2 reactor enclosure HVAC room, resulting in a lowering of reactor enclosure delta pressure to below the tech spec minimum required value. The airlock doors were closed within approximately 5 seconds and reactor enclosure delta pressure recovered to greater than the tech spec minimum required value within approximately 20 seconds. Unit 2 secondary containment was declared inoperable for the time that reactor enclosure delta pressure was below the tech spec minimum required value, and was declared operable when reactor enclosure delta pressure recovered to greater than the tech spec minimum required value. Total Limiting Condition of Operation time was approximately 20 seconds. The licensee has notified the NRC Resident Inspector. |
| ENS 49310 | 27 August 2013 18:00:00 | The following report was received via e-mail from the Louisiana Department of Environmental Quality: On 08/27/2013, the RSO for Tulane University Hospital called to notify the Department that their facility had a Medical Event involving (exposure to unintended tissue greater than) 50 Rem. The event was discovered on 08/27/2013 when an application was not able to be applied to the intended tissue. The HDR source had 'dog legged' into the bowel area when it was intended to apply the radiation dose to the cervical area. The films were pulled for the application on 08/22/2013 and revealed that the application had 'dog legged' also. The cervical tissue did not receive the initial intended dose. The HDR (High Dose Rate Brachytherapy Afterloader) unit was a Nucletron Micro-Selectron, loaded with (an) Ir-192 (source). The therapy dose was 8.4 Gray (840 rads) given in fractions. The patient is to receive the entire corrected therapy dose prescribed. This is believed to be (under investigation) a positioning problem and not an equipment malfunction. The patient's physician has been notified. However, the patient was heavily sedated and has not been notified. Updates will be made when additional information is available. Louisiana Report ID: LA-130001 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 49253 | 9 August 2013 08:21:00 | A contractor supervisor tested positive for alcohol on a follow-up fitness for duty screening. The individual's construction site access was revoked. The Unit 2/3 Resident Inspector has been notified. |
| ENS 49249 | 8 August 2013 03:15:00 | The licensee reported that there was an explosion in a non-safety related electrical bus (Bus 1E2). The bus is designated as a safe shut down bus. The licensee has classified this event as an Alert based on an EAL entry condition of "an explosion of sufficient force to damage permanent structures or equipment within the Protected Area." The licensee is still investigating what loads come off this bus but no safety-related loads have been identified at this time. The plant is currently stable at 92% power. Power was reduced due to the loss of moisture separator reheaters on the secondary side as a result of this event. There is no ongoing fire as a result of the bus explosion. No personnel were injured. No damage to other equipment has been identified at this time. The licensee has notified state and local authorities and the NRC Resident Inspector. Notified DHS SWO, DOE, FEMA, HHS, DHS NICC, USDA, EPA, FDA and NuclearSSA via email. NRC PAO (Brenner) notified.
Auxiliary bus 1E2 (safe shutdown equipment bus) is de-energized due to the explosion. The Alert was declared due to the explosion affecting safe shutdown equipment." The licensee also noted that the event was initially classified as an Unusual Event at 0234 EDT and then upgraded to an Alert classification at 0305 EDT when the bus was determined to supply safe shutdown equipment. The licensee has notified the NRC Resident Inspector.
No offsite assistance was requested during this event. The licensee verified that there was no fire. It was noted that the licensee is in Technical Specification Action Statement 3.8.1.1 for the "B" Emergency Diesel Generator being out of service. The EDG is out of service due to loss of power to the "C" and "D" air compressors. The licensee was uncertain if the loss of these air compressors was related to the event. The licensee has notified the NRC Resident Inspector. R2DO (Shaeffer) and NRR EO (Hiland) notified.
The licensee terminated the Alert declaration at 0533 EDT based upon no fire at Bus 1E2 and verification that Bus 1E2 is de-energized and disconnected from all electrical systems. The licensee remains at 92% power. The only noted safety related equipment that was impacted is the "1B" Emergency Diesel Generator which has been declared inoperable due to loss of power to associated air compressors "1C-SB" and "1D-SB. The licensee has notified the NRC Resident Inspector. R2DO (Shaeffer), NRR EO (Hiland) and IRD MOC (Morris) notified. Notified DHS SWO, DOE, FEMA, HHS, DHS NICC, USDA, EPA, FDA and NuclearSSA via email. NRC PAO (Brenner) notified.
It has been determined that operability of the 1B Emergency Diesel Generator (EDG) was not affected by the damage to 480 volt auxiliary bus 1E2. The 1C and 1D starting air compressors for 1B EDG did not lose power and thus the 1B EDG was operable. The NRC Resident Inspector has been notified. Notified R2DO (Shaeffer). |
| ENS 49246 | 7 August 2013 09:01:00 | At approximately 0100 (CDT) on 7 August, 2013, severe thunderstorms and high winds in the area resulted in a loss of power to the Point Beach Nuclear Plant (PBNP) Emergency Operations Facility (EOF) and to Alert and Notification System (ANS) sirens. Loss of ANS sirens resulted in a loss of greater than 50% of the Emergency Planning Zone (EPZ) population coverage (14 out of 22 lost). Note: PBNP has installed a new siren system that is currently in operational acceptance testing and remained fully functional on backup battery power. During the loss of power to the EOF, station personnel verified that power remained available to the Alternate Emergency Operations Facility (AEOF). At approximately 0545 CDT, station personnel verified that power to the EOF was restored. Efforts are ongoing to restore power to sirens. The licensee has notified State and local authorities and the NRC Resident Inspector. See related EN #49247.
At 1525 CDT on 8/7/13, the licensee returned 9 of 14 emergency sirens to service which brings the percentage of population covered by the sirens below the reporting threshold. The licensee has notified the NRC Resident Inspector. Notified R3DO (Lara). |
| ENS 49250 | 8 August 2013 11:09:00 | The following report was received from the State of Tennessee Division of Radiological Health via e-mail: Tennessee's Division of Radiological Health was notified on Wednesday August 7, 2013 by the RSO from Aerojet Ordnance Tennessee, regarding the failure of a primary ventilation system. On August 5, 2013 at 10:00 AM, an employee at the Deflash Station observed smoke outside the operation booth while grinding burrs off radiography camera castings. A supervisor was notified and investigation showed the ventilation system was working, but the belt connecting the pump and fan was broken. Operations were suspended and personnel evacuated. The belt was replaced on the pumps and the ventilation system was up and operational by 10:30 AM. Event resulted in elevated airborne uranium concentrations. All personnel in the building submitted urinalysis. Workers at the Deflash station were wearing respiratory protection during time of event as part of standard procedures. Area air samplers along with environmental air samplers were pulled and analyzed; initial results identified no concerns of elevated concentrations. The State will follow-up and keep NRC informed of the status of our investigation. Tennessee Report Number:TN-13-134 |
| ENS 49320 | 4 September 2013 15:07:00 | The following report was received from the Kansas Bureau of Environmental Health via facsimile: During radiography operations on the night of 5 Aug 2013, two employees, a radiographer and an assistant radiographer of Coder Welding & X-ray Service were finishing an 8 hour shift (which is 2nd shift for Alstom), when the source and crank assembly would not operate correctly. At that time, they could only determine that the source had not fully retracted into the camera correctly. After several additional attempts, they concluded a malfunction of unknown causes prevented the source from retracting to a full and locked condition. Following Coder Operating and Emergency protocol, the first action was to secure and maintain a 2mr/hr boundary. A complete survey of the area determined that the original roped and placarded boundaries were still correct and valid. Next the RSO was contacted as well as the assistant RSO. The RSO advised the radiographer to secure and maintain the 2mr/hr boundary and asked if any personnel or workers had been exposed to radiation levels in excess of those in Kansas Radiation Protection Regulations, Part 4 and following, and they stated no. They were advised that the RSO would be on site in 90 minutes to oversee the incident and resolve the situation. While waiting for the RSO to arrive, the two Coder employees decided to extend the roped boundaries an additional 75 feet in addition to the existing roped area and inform Alstom management of the situation. Since this was the end of 2nd shift for Alstom workers, and no Alstom workers were in the area at the time, there was no disruption of production or evacuation needed. Any Alstom employees on site were advised not to enter plant area as a precaution. At no time were Coder or Alstom employees at risk or in danger of overexposure. The RSO arrived on site at approximately 12:15 p.m. and made a radiation survey of the area and boundaries. The RSO found correct actions had been taken and 2mr/hr boundaries were maintained. In fact, 1mr/hr was the highest reading. The RSO then walked up to the crank-out reel and found the radiation level to be 5mr/hr. He concluded that the source was in fact in the tungsten collimator (4.3 hvl (half value integers)) secured on the pipe weld where it was during earlier radiographic operations. Several attempts to return the source to its shielded and locked condition failed, so plans were made to allow for closer inspection of the cables and source tube. At this point, the assistant RSO was contacted and advised to bring additional drive cables and source tube in the event they could be needed. While the radiographer and assistant radiographer maintained security over the boundaries and source, the RSO and several Alstom management, who had arrived on site, went outside to look for suitable shielding that could be brought in. Two one inch plates were chosen for use. The plates were tack welded together and moved to an area where overhead cranes could be used to move into position. This was accomplished by using the remote controls of the crane system so no person would have to be in a high radiation environment. With help from Alstom personnel, the steel plates were directed into place by the RSO next to the source camera, providing additional shielding. It was then possible for the RSO to walk up to the source camera with a survey level of 32mr/hr. It was then possible to inspect the drive cables and look at the source tube for possible causes of the return failure. A small depression was noted some 6 to 7 feet from the camera and cable attachment. Using a hacksaw, pliers, and other tools, the outer cable shielding was removed and the RSO returned to the cable crank, and was able to retract the source into the camera in the full and locked position. During the entire operation the RSO received a whole body dose of 35 mr. Due to the time and distance for the RSO, late arrival of the assistant RSO arrival on site, and the caution taken to resolve the situation, some 3-1/2 hours elapsed from the start to the end. The help and materials provided by Alstom management aided in the safe and satisfactory conclusion of this incident. At this time, it is unknown what might have caused the depression in drive cables or why it suddenly caused a failure to retract situation. Kansas Report Number: KS130006 |
| ENS 49176 | 7 July 2013 06:49:00 | At 0400 EDT on July 7, 2013, Nine Mile Point Unit 2 isolated the Reactor Building to comply with Tech Spec 3.3.6.2.A.1 Secondary Containment Isolation Instrumentation. Isolating the Reactor Building was the result of Refuel Floor Radiation Monitor 2HVR*CAB14A becoming inoperable on 7/6/13 at 0430 hours. As a result of the Reactor Building being isolated, the Radwaste/Reactor Building Vent Gaseous Effluent Radiation Monitor has been declared non-functional based on sample flow rates. The Radwaste/Reactor Building Vent Gaseous Effluent Radiation Monitor is necessary for accident assessment and is credited for Emergency Action Level (EAL) classification. The inability to classify an EAL due to the out-of-service Radwaste/Reactor Building Vent Gaseous Effluent Monitor is considered a loss of emergency assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii). Repairs and testing are currently in progress to correct the cause and restore functionality of Radiation Monitor 2HVR*CAB14A. The expected return to service time is within the next 12 hours. Compensatory actions for the non-functional Radwaste/Reactor Building Vent Gaseous Effluent Radiation Monitor are in place to take and analyze periodic grab samples and flow estimates in accordance with the Offsite Dose Calculation Manual. The NRC Resident Inspector has been notified." The licensee will also be notifying state authorities. * * * UPDATE ON 7/8/13 AT 1145 EDT FROM DAN CIFONELLI TO DANIEL MILLS * * * "Radiation Monitor 2HVR*CAB14A has been repaired and was declared operable at 1920 on July 7, 2013. Normal Reactor Building ventilation was subsequently re-established, and the Radwaste/Reactor Building Vent Gaseous Effluent Radiation Monitor was retumed to service at 2100 on July 7, 2013, thereby restoring emergency assessment capability. The NRC Resident Inspector has been notified. |
| ENS 49177 | 7 July 2013 11:15:00 | At 0326 EDT on 07/07/2013, it was determined that 14 of 96 Ginna Nuclear Power Plant Emergency Offsite Sirens were nonfunctional. The apparent cause is loss of power to the sirens due to storm related power outages. The weather has since returned to normal conditions, and residential power restoration is in progress. (At the time of this report, only 3 sirens remained without power). This event is being reported as a Loss of Emergency Preparedness Capabilities pursuant to 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified. |
| ENS 49175 | 6 July 2013 23:47:00 | The Millstone Unit 1 spent fuel pool island ventilation system tripped offline for reasons unknown at the time of this report. The spent fuel pool island radiation monitor requires the ventilation system to be in service to function. Therefore, the radiation monitor was declared inoperable. The loss of the radiation monitor results in a loss of assessment capability. The licensee's chemistry department is taking compensatory samples in accordance with the Radiation Effluent Monitoring Offsite Dose Calculation Manual action statement until the spent fuel pool island ventilation can be restored. The licensee has notified the NRC Resident Inspector, state and local authorities. |
| ENS 49158 | 29 June 2013 18:30:00 | At 1214 EDT on June 29, 2013, Beaver Valley Power Station (BVPS) Unit 2 determined that the Digital Radiation Monitoring System (DRMS) was operating intermittently resulting in an intermittent loss of control room radiation monitor indication and alarm capability. BVPS Unit 2 DRMS was declared non-functional. Repair efforts were initiated and at 1415 EDT DRMS was restored to functional status. Compensatory measures were established during the period of time that the DRMS was operating intermittently. The period that BVPS Unit 2 DRMS was non-functional resulted in a loss of emergency assessment capability and is reportable per 10 CFR 50.72(b)(3)(xiii). The NRC Resident Inspector has been notified. |
| ENS 49156 | 28 June 2013 20:13:00 | Two employees reported to the on-site dispensary this afternoon with chemical burns to their shoulders at approximately 1645 CDT. The plant nurse and doctor administered treatment to the employees and then sent them home. A whole body survey of the employees in their plant clothing was performed while in the dispensary. The maximum amount of contamination found was present on one of the employee's boots, 20,745 dpm/100cm2. Prior to leaving the Restricted Area, the employees removed all plant clothing, changed into their personal clothing, and exit monitored from the facility. The employees were free of contamination upon release. The licensee plans to notify NRC R2. |
| ENS 49153 | 28 June 2013 14:18:00 | At 0749 CDT on June 28, 2013, indication was received in the Control Room that two Secondary Containment doors, in one access opening, were opened simultaneously. The interlock mechanism preventing both doors from operating simultaneously did not operate as expected. This condition represents a failure to meet Surveillance Requirement 3.6.4.1.2. As a result, entry into Technical Specifications 3.6.4.1 Condition A was made due to Secondary Containment being declared inoperable. The doors were secured at 0749 CDT (total time with both doors open was 7 seconds) and Secondary Containment was declared operable. This event is being reported in accordance with 10 CFR 50.72(b)(3)(v)(C) as a condition that could have prevented the fulfillment of a safety function. The NRC Resident Inspector has been notified. |
| ENS 49440 | 14 October 2013 16:10:00 | The following information was received from the Kentucky Radiation Health Branch via facsimile: Kentucky Radiation Health Branch (RHB) was notified by telephone on 10/2/13 by a representative from a generally licensed facility, Momentive Specialty Chemicals, of the fact that the shutter ON-OFF control mechanisms had been stuck in the open position on two level gauges undergoing routine servicing almost four (4) months previously. The two gauges, Thermo TN Technologies model 7063P each with 100 mCi Cs-137 (sources) (S/N S92BO601 and S94K1101) were being routinely serviced by a licensed manufacturer, VEGA Americas, on June 17, 2013 when the shutters were found to be stuck in the open position. The VEGA service representatives returned to the licensee's two days later (6/19/13) and repaired the sticking shutters so they would operate freely. The VEGA representatives informed the general licensee's representative that a report to RHB was required but the person responsible for the gauges and for making the report was not notified of the problem until almost four months later. When he did learn of the event and the reporting requirement, he reported the event to RHB by telephone on 10/2/13 and followed up with a written report on 10/11/13. The reporting criteria is defined in 902 KAR 100:050 Section 3(3)(c)5 and 10 CFR 31.5(c)(5). |
| ENS 49121 | 16 June 2013 02:42:00 | This event is being reported in accordance with 10 CFR 50.72(b)(2)(i) and 50.72(b)(3)(ii)(A). On June 16, 2013 at 0200 EDT, the Perry Nuclear Power Plant commenced a controlled plant shutdown. The shutdown was due to a small leak through the base of a vent line on the 'B' Reactor Recirculation Flow Control Valve. On June 15, 2013 at 2250 EDT, the leak was identified and was subsequently determined to require a plant shutdown in accordance with Technical Specification 3.4.5, Action (C) which requires the plant to be in Mode 3 within 12 hours. The NRC Resident Inspector has been notified." The licensee will also be notifying state and local authorities. The licensee had come down in power to make a drywell entry and investigate drywell leakage indications. Steam was observed to be coming from a vent line that comes off the top of the recirc flow control valve. The licensee was unable to characterize the leak rate other than a small leak. The licensee stated that the steam appeared be coming from a weld location where the vent line comes out of the flow control valve which would classify it as pressure boundary leakage. |
| ENS 49073 | 28 May 2013 10:32:00 | The Station Stack Radiation Monitor, RM-8169, and the Unit 1 Spent Fuel Pool Island Rad Monitor have been removed from service for preplanned maintenance for filter change out. During this time period, the licensee will not have normal assessment capability for radiation releases via these pathways. Compensatory monitoring methods are in place for the duration of the maintenance activity. The radiation monitors will be restored upon completion of the filter change-out. The expected completion time is before the end of day shift on 5/28/13. The licensee has notified the NRC Resident Inspector along with state and local authorities. |
| ENS 49052 | 19 May 2013 18:51:00 | During Refueling Outage 21, at approximately 0700 PDT on May 20, 2013 the following plant's gaseous effluent radiation monitoring systems and seismic monitoring systems will be removed from service due to a planned power outage: - Reactor Building Stack Radiation Monitor: Low Range; Intermediate Range; and High Range Detectors. (NOTE: The Reactor Building Stack Radiation Monitors were removed from service as of 1421 PDT on 19 May 2013 (24 hours before the power outage) to allow for a gradual warming up of the sensors) - Rad Waste Building Vent Exhaust Low Range Radiation (Rate Meter) and Exhaust Air Monitor Radiation Indicating Switch - Turbine Building Radiation Indicating Switch and Exhaust Air Radiation Indicating Switch - Seismic Instrument Accelerometers - Seismic Instrument Accelerographs The listed equipment is expected to be re-energized at approximately 1400 PDT on May 22, 2013. The Reactor Building Stack Radiation Monitors is expected be operational approximately 48 hours after they are re-energized to allow for sensor cooling requirements to be established. To compensate for the loss of the radiation monitoring equipment, an additional Health Physics (HP) Technician trained to acquire offsite dose assessment information on offsite releases will be on shift. The additional personnel will be pre-staged in support of the radiation monitoring system outage and will be deployed in accordance with guidance in site procedures and the compensatory measure instructions. To compensate for the loss of the seismic monitoring capability, an entry into the abnormal operating procedure 'EARTHQUAKE' will be made when an earthquake is felt in the control room or when information is received from plant personnel that an earthquake has been felt. Earthquake severity will be estimated in accordance with abnormal operating procedure 'EARTHQUAKE' in lieu of instrumentation being available. Information from the US Geological Survey (USGS), if available, will supplement the estimation of earthquake severity. This event is being reported as a loss of emergency assessment capability in accordance with 10 CFR 50.72(b)(3)(xiii). A follow up notification will be made when the equipment has been returned to service. The licensee has notified the NRC Resident Inspector.
The Columbia Seismic Instruments (Accelerometers and Accelerographs) were returned to service on May 30, 2013, at 0440 PDT. The Columbia radiation monitoring instruments listed below were returned to service on June 2, 2013, at 1814 PDT: - Reactor Building Stack Radiation Monitor: Low Range; Intermediate Range; and High Range Detectors - Rad Waste Building Vent Exhaust Low Range Radiation (Rate Meter) and Exhaust Air Monitor Radiation Indicating Switch - Turbine Building Radiation Indicating Switch and Exhaust Air Radiation Indicating Switch The licensee will notify the NRC Resident Inspector. Notified R4DO (Spitzberg). |
| ENS 49002 | 5 May 2013 01:45:00 | At 0112 EDT on May 5, 2013, the plant commenced a shutdown due to water leakage from the SIRW (Safety Injection Refueling Water) Tank exceeding the operational decision-making issue process trigger point of 34 gallons per day, causing it (the SIRW) to be declared inoperable and requiring entry into Technical Specification LCO 3.5.4 Condition B. LCO 3.5.4 Condition B requires the SIRW Tank to be returned to Operable status within one hour or entry into Condition C which requires the plant to be in Mode 3 within 6 hours and Mode 5 within 36 hours. This event had no impact on the health and / or safety of the public. The NRC Resident has been notified. The exact location of the leakage has not been determined at this time. The Plant will be taken to Mode 5. The licensee has been operating with SIRW leakage at a rate of less than 34 gallons per day. The leakage has increased for unknown reasons to a calculated value of approximately 90 gallons per day. See EN #48018 dated 06/12/12 for similar report on a technical specification shutdown for the SIRW tank leakage. |
| ENS 48976 | 26 April 2013 17:25:00 | The following report was received from ITT Engineered Valves, LLC via facsimile: It is my duty as the Responsible Officer of ITT Engineered Valves, LLC (ITT) to inform the Nuclear Regulatory Commission of a defect with certain items of our nuclear diaphragm valve product line which may be considered Basic Components. The components are ITT's Nuclear M1 diaphragms, sizes 3 inch and 4 inch that may have been sold to certain customers for specific design conditions. The defect does not affect all 3 inch and 4 inch M1 diaphragms that have been sold. It only applies to those that were sold for a particular service condition of Code Case N31 (250?F and 220 psi with 40 year radiation exposure of 1E8 Rad). The nature of the defect is best described by 10 CFR Section 21.3 Defect Definition #5, as 'an error, omission or other circumstance in a design certification or standard design approval that... could create a substantial safety hazard.' In this case, ITT inadvertently qualified the 3 inch and 4 inch M1 diaphragms for a design condition that includes the effect of radiation when in fact our recommendation was erroneously based on diaphragm testing that did not include irradiated diaphragm test results for those sizes. The potential safety hazard stems from the fact that if one of these diaphragms sees radiation in this particular service, there is no data to indicate that the diaphragm will perform its function in that service condition. Until such time that we can conduct additional irradiated diaphragm testing to additional sample diaphragms and test for this condition, we need to consider the parts that are in this service as potentially unsafe. ITT is in the process of identifying all facilities for which the diaphragms were sent, either as spare parts or diaphragms incorporated into valve assemblies. We are also preparing to do further verification tests of the 3 inch and 4 inch M1 diaphragms in an attempt to ascertain the true performance rating at the noted condition. Per 10 CFR 21 policy guidelines, this initial notification will be followed by a written notification by May 27, 2013. |
| ENS 48969 | 25 April 2013 22:23:00 | This report is being made pursuant to 10CFR50.72(b)(2)(iv)(B), RPS Actuation (scram). At 2019 CDT on April 25, 2013, LaSalle Unit 2 was manually scrammed due to a loss of Condenser Circulating Water. The Unit was manually scrammed after the condenser circulating water pumps tripped due to high level in the turbine building condenser pit. The high level in the condenser pit was caused by a leak on the upper manway of the condenser water box during a maintenance activity. MSIV's were isolated due to loss of heat sink. The safety relief valves were used in pressure control mode. Current plant status: reactor level is stable and reactor pressure is stable. The condenser water box manway leak has been isolated. The plant will remain in hot shutdown pending investigation and repairs. Reactor Core Isolation Cooling (RCIC) is being used in the pressure control mode. The licensee has notified the NRC Resident Inspector. |
| ENS 48968 | 25 April 2013 20:51:00 | This is a report of a loss of emergency assessment capability as required by 10 CFR 50.72(b)(3)(xiii). On April 25, 2013 at 1759 CDT, with Unit 2 in Mode 6 during a refueling outage, power was interrupted to all Unit 2 vent stack radiation monitors as part of a pre-planned activity to connect the radiation monitors to an alternate temporary power supply to support de-energizing the normal power source for preventative maintenance. The connection to the alternate supply was completed and power was restored to the vent stack radiation monitors at 1845 CDT. While the radiation monitors were without power, pre-planned compensatory measures were implemented to monitor vent stack discharge and to minimize activities that posed a potential for release. At the completion of the preventive maintenance on the normal power supply, power to the vent stack radiation monitors will again be briefly interrupted to reconnect the normal power source to the monitors. The pre-planned compensatory measures will again be utilized during this power interruption. An update to this report will be provided following the restoration of normal power to the radiation monitors. The NRC Resident inspector has been informed.
On 28 April, 2013, power was again interrupted to the Unit 2 vent stack radiation monitors to restore the connection to their normal power supply. The radiation monitors were out of service from 2315 until 2340 CDT. Pre-planned compensatory measures were again implemented to monitor vent stack discharge and minimize potential for vent stack release. The reported time for the initial loss of vent stack radiation monitoring on April 25, 2013 was incorrect. The correct time was 1759 CDT. The NRC Resident has been notified. Notified R2DO (Sykes). |
| ENS 48963 | 24 April 2013 16:50:00 | A non-licensed, supervisory employee had a confirmed positive for alcohol during a random fitness-for-duty test. The employee's access to the plant has been restricted. The licensee has notified the NRC Resident Inspector. |
| ENS 48955 | 22 April 2013 23:23:00 | On 4-22-13, at 1623 (CDT), the ANO Unit 2 control room was notified of a loss of ventilation capability to the Emergency Operations Facility (EOF). The main control boards associated with the variable speed drives on both air handling units at the EOF have failed. Therefore, there are no means to filter air for the EOF. If the EOF is staffed, the EOF will be required to relocate to the Alternate EOF in the event of a release that causes the EOF evacuation criteria to be exceeded, as directed by approved emergency response procedures. The on-site Operations Support Center, on-site Technical Support Center and off-site Alternate EOF remain fully functional to perform emergency assessment activities. Efforts are underway to expedite repairs. This notification is required by 10CFR50.72(b)(3)(xiii). The licensee has notified the NRC Resident Inspector.
At 1637 EDT on 4/23/13, the EOF ventilation has been restored and the EOF has full functionality. The licensee has notified the NRC Resident Inspector. Notified R4DO (Whitten). |
| ENS 48953 | 22 April 2013 14:58:00 | The Comanche Peak Nuclear Power Plant Emergency Operations Facility (EOF) is not available due to the loss of HVAC and filtering capabilities resulting from a failed Emergency Operations Facility (EOF) ventilation fan. The condition was discovered at 0900 CDT on 4/22/13. Repair parts are expected by the morning of 4/23/13 and the EOF is projected to be available by the end of the day on 4/23/13. Compensatory measures are in place to staff and activate the Alternate EOF in the event of a declared emergency. The NRC Resident Inspector has been informed.
The Emergency Operations Facility vent fan was returned to service at 1400 EDT on 4/23/13. The licensee has notified the NRC Resident Inspector. Notified R4DO (Whitten). |
| ENS 48971 | 26 April 2013 12:00:00 | Prefix Corporation uses static control ionizers for applying coatings in some of its product processes. These static control ionizers are leased from NRD LLC and contain less than 0.1 millicuries of Polonium-210. In early April, 2013, Prefix was preparing to return one of its leased Nuclecel Ionizers to NRD and discovered it was missing. The source was last used in November of 2012. A search was conducted for the source without success and Prefix considers the source to be lost. The source is small and cylindrical and could have rolled off its storage location onto the floor and been swept-up as trash. Prefix sent a written notification to NRC Region 3 on April 18, 2013, but had not previously reported the event to the NRC Operations Center. 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 |