|Entered date||Event description|
|ENS 50115||14 May 2014 18:07:00||SONGS (is) making a 4-Hour notification per 10CFR72.75 (b) to the NRC Operations Center regarding the following: There has been a fire in the vicinity of the station. The fire is not on plant property and has not challenged station operations. Entry into SONGS' Emergency Plan and activation of the Emergency Response Organization is not required at this time. Because of the fire near the plant, Southern California Edison will be making a press release today to update the public as to the situation at the plant. The licensee notified an onsite NRC Inspector.|
|ENS 50114||14 May 2014 17:58:00||At 1705 (EDT) on May13th, 2014, approximately 5 gallons of diesel fuel oil was spilled onto the ground on the south side of the Service Water Intake structure at the Salem Generating Station. The spill of diesel fuel was caused by a leak from the fuel supply line to the service water hot air furnace. The leak was isolated at the time of discovery and the spill terminated. The diesel fuel oil cleanup is in progress by Clean Harbors personnel and will continue until the spill has been remediated. Nuclear Environmental Affairs Department determined a 4 hr report to the NRC under RAL 11.8.2.a. was warranted due to the 15 minute notification to the New Jersey Department of Environmental Protection at 1651 (EDT) on May 14, 2014. The licensee will notify the NRC Resident Inspector.|
|ENS 50109||12 May 2014 16:57:00||On May 12, 2014, at 1458 CDT, River Bend Station management determined that tritium confirmed to be present in water samples taken from in-leakage into the below-ground elevation of the turbine building basement, through an underground pipe penetration, was reportable in accordance with NEI 07-07, Ground Water Protection Initiative. The leakage was tested for gamma and tritium activity. No gamma contamination was detected. Tritium was measured at 28,270 picocuries per liter. The source and volume of the radioactive in-leakage is unknown. The leakage into the plant is currently being contained, and actions are in progress to identify the source. The Louisiana Department of Environmental Quality was notified at 1500 CDT today (5/12/14). This event is being reported in accordance with 10 CFR 50.72(b)(2)(xi) as an event requiring notification of the state government. The NRC Resident Inspector has been notified.|
|ENS 50108||12 May 2014 14:48:00||During a review from industry operating experience it was identified that there are three additional unprotected DC control circuits for non safety related DC motors which are routed from the turbine building to other separate fire areas (this is in addition to the one circuit that was previously identified and submitted under event #50059). Fuses used to protect the motor power conductors appear to be inadequate to protect the control conductors. The concern is that under fire safe shutdown conditions, it is postulated that a fire in one area can cause short circuits potentially resulting in secondary fires or cable fires in other areas where the cables are routed. The secondary fires or cable failures are outside the assumptions of the 10 CFR 50 Appendix R Safe Shutdown Analysis. This condition is reportable as an 8 hour ENS report 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 50110||12 May 2014 17:37:00||At 0845 (CDT) during planned maintenance, the annunciators which would indicate flooding in the Emergency Core Cooling System (ECCS) Pump Rooms to the max safe water level in the rooms were disabled. These max safe water level alarms are used to assess the emergency action level (EAL) entry condition (HA4) for flooding (ALERT classification). Compensatory measures are in place to perform periodic walk downs of RHR A, RHR B, RHR C, HPCS, LPCS, RCIC rooms and the fuel building basement once per shift to verify no leakage. Additionally, the annunciator for max normal water level is operational and pump run times are still available in the Main Control Room to indicate increased leakage into the ECCS rooms. This max normal water level alarm is used to assess the EAL entry condition (HU4) for flooding (Unusual Event classification) and EOP-8 entry. Since the operator compensatory actions that were established are not proceduralized, then this is reportable as a loss of emergency assessment capability under 10 CFR 50.72(b)(3)(xiii). The expected return to service time of all alarms is 2300 CDT. The licensee will notify the NRC Resident Inspector.|
|ENS 50113||14 May 2014 15:57:00|
On May 14, 2014, the licensee notified the Agency (Texas Department of Health) that when it was locking out sources on May 12, 2014, for a shutdown it was unable to close the shutter on one of its Ohmart-Vega model SHLG-2 fixed nuclear gauges which contained a 3 curie cesium-137 source. The licensee contacted the manufacturer. A service company came to the licensee's facility on May 14, 2014, and closed the shutter, removed the gauge, and placed it in storage. The licensee will send the gauge to the manufacturer for repair. The gauge normally operates with the shutter in the open position. No one received any exposure as a result of this event. Additional information will be provided as it is obtained in accordance with SA-300. TX # I-9193
A review of the licensee's written report found that a second gauge had failed during this event. The second gauge was an Ohmart-Vega model OHM-S-A2 gauge containing a 3000 milliCurie cesium - 137 source. The licensee stated that while trying to close the shutter on this gauge, the pin for the actuator handle broke and the handle came loose from the gauge. The handle was placed on the gauge and held in place with wire. The shutter was left in the open position. On May 14, 2014, the manufacturer was able to free the shutter and place it in the closed position. The gauge was removed from its mounting and package for shipment by the manufacturer. No significant exposure was received by any individual as a result of this event. Additional information will be provided as it is received in accordance with SA-300. Notified R4DO (Hay) and FSME Resources (email).
|ENS 50093||7 May 2014 15:19:00||A leak containing a low level tritium concentration assumed to be greater than 100 gallons of water, with the potential to reach groundwater, occurred at Oconee Nuclear Station. A water sample indicated that the tritium level was 3150 picocuries per liter; which is less than a quarter of the US Environmental Protection Agency drinking water standard of 20,000 picocuries per liter for tritium. While conducting a transfer from one chemical treatment pond to another, water was observed seeping from the ground at a location near the transfer piping between the ponds. The transfer was terminated and the ground seepage subsided. Actions have been taken to prevent further use of the chemical treatment pond discharge path at this time. Based upon the on-site location and low tritium levels, there is no health or safety impact to the public or employees. Voluntary notification of state and local agencies is being made via the industry groundwater protection initiative; measured tritium levels were below any required notification threshold. Agencies notified: South Carolina Department of Health and Environmental Control, South Carolina Emergency Management, City of Seneca, City of Anderson, Greenville Water System, Oconee County Administration, Pickens County Administration, Anderson County Administration, Oconee County Emergency Management, Pickens County Emergency Management, Greenville County Emergency Management, and Clemson University Utility Services. The licensee notified the NRC Resident Inspector.|
|ENS 50090||6 May 2014 13:27:00|
At 0830 (CDT) on 05/06/2014, the Unit 3 reactor automatically scrammed due to low reactor water level as a result of a trip of both recirculation pumps. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System Groups 2, 3, 6, and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. The Reactor Feedwater System controlled and maintained water level above the level 2 initiation setpoint. Prior to the Scram, the reactor was operating at 100% power. A Core and Containment Cooling Systems Analog Trip Unit Functional Test was in progress. The cause of the recirculation pump trip is under investigation. This event is reportable within 4 hours per 10CFR 50.72(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). The NRC Resident Inspector has been notified. U1 and U2 remained at 100% power and were unaffected.
Investigation revealed that a failed power supply caused an Anticipated Transient Without Scram/Alternate Rod Insertion (ATWS/ARI) signal to be generated when a level 2 Reactor Water Level was simulated on one instrument. All systems responded to the ATWS/ARI signal as designed. This signal opened the Recirc Pump Trip breakers for both Recirculation Pumps and opened the ARI valves to bleed air from the Reactor Protection System (RPS) scram air header. The resulting transient caused reactor water level to dip below the RPS trip setpoint (level 3 Reactor Water Level), a normal plant response, and the automatic scram signal occurred. At the time of the RPS scram signal, all rods were inserting and reactor power was approximately 2-3% and lowering. The NRC Resident Inspector has been notified. Notified R2DO (Bonser).
|ENS 50088||6 May 2014 11:47:00||The licensee Radiation Safety Officer (RSO) determined that the shutter to a Berthold Model LB7440-D-CR process gauge with a 500 mCi Cs-137 source (source s/n: 031-08) did not open despite repeated attempts to cycle and lubricate the shutter opening mechanism. The RSO made this determination using instrumentation after rotating the shutter opening handle 180 degrees to the normally open position. The shutter had apparently become disconnected from the opening mechanism. The instrument is installed on a hopper dredge and is pointed down in an area not normally accessible by personnel. A licensed Berthold technician has been contacted and will come to the site to repair the gauge.|
|ENS 50067||28 April 2014 15:28:00||On April 28, 2014, at approximately 0845 AM (PDT), (the licensee pending ARSO), an authorized user of radioactive gauges of Twining Laboratories of Southern California, Inc., RML (California Radioactive Materials License) #7780-37 located in San Bernardino, CA, contacted RHB Brea (Radiation Health Branch located in Brea, CA) concerning the moisture/density gauge, CPN model, MC-1, serial number MD10606201 (Cs-137, 0.370 GBq, Am-241, 1.85 GBq) that had been stolen from a temporary storage/construction site in San Bernardino County on the I-15 North freeway at the top of the Cajon Pass at the Oak Hills Road exit. (The licensee pending ARSO) arrived at the temporary storage area to begin work on Monday morning and noticed that the construction facility had been broken into and that other tools and equipment had been taken as well as the radioactive gauge, which was charging while locked in a gang box which was removed in it's entirety once the construction fence and locked cargo container were breached. (The licensee pending ARSO) has contacted the San Bernardino Sheriff's Department which has filled out a theft report which will be copied and sent to RHB Brea as part of this report. The RSO of Twining Laboratories of Southern California, Inc., is currently on vacation and (the licensee pending ARSO) is receiving assistance and recommendations from another RSO of Twining Laboratories of Southern California, Inc., RML #6872-19 located in Long Beach, CA. (The licensee pending ARSO) is contacting pawn shops and landfills in the area to determine if he can locate the stolen radioactive gauge in advance of utilizing local papers to attempt to retrieve the stolen gauge. The investigation will continue to determine if the radioactive gauge can be recovered in a reasonable time frame. CA 5010 Number: 042814 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 50060||26 April 2014 01:24:00||System Affected: Millstone Station Stack Radiation Monitor, RM-8169 Actuations and Their Signals: None Causes: Preplanned Maintenance Effect of Event on Plant: None Actions Taken or Planned: Completed testing and restored operability" (on 4/26/14 at 0037 EDT) Additional Information: None The licensee notified the NRC Resident Inspector.|
|ENS 50070||29 April 2014 15:48:00||A DOT special permit was issued by our Agency to allow the transport of unidentified radioactive material (RAM) from one waste transfer station to the originating transfer station. Subsequently, a contractor (Atlantic Nuclear) was hired by the waste services operator to isolate the RAM, identify it, label it accordingly, and store it in a secure onsite location. The contractor was not able to definitively identify the RAM. The contractor employed portable multi-channel analyzer's in the effort (both an ICS-4000, and LMJ-702), but no positive identification of the isotope(s) could be made. The material in question is encapsulated in what appears to be a metal enclosure, and the contractor postulated that the material might be Ra-226 due to the number of gamma energy lines measured. It was also postulated that metal encapsulation may be providing interferences, and further testing is required to establish the identity of the isotope. The dose rate measurements conducted by the contractor revealed: At contact 35-40 mR/hr; at 12 inches approx. 3 mR/hr. The contractor, Atlantic Nuclear (MA Lic. No. 56-0477), is planning to pick up the materials for storage pending disposal in accordance with their specific license. The Agency (Massachusetts Radiation Control Program) investigation remains open.|
|ENS 50007||8 April 2014 09:45:00||U1 entered Technical Specification 3.0.3, and initiated actions to reduce power via control rod insertion in preparations to enter startup within the following 6 hours. Entry into Tech Spec 3.0.3 was a result of both HPCI and RCIC injection systems being inoperable at the same time. During startup from the U1 refueling outage, RCIC and HPCI full flow testing was unable to be completed at rated reactor pressure prior to the expiration of the 12 hour allowance per Tech Spec Surveillances 4.7.3.b* and 4.5.1.b.3** respectively. This occurred due to testing issues encountered while attempting to perform the rated pressure pump valve and flow tests. U1 HPCI testing was completed satisfactorily at 0830, and HPCI was restored to operable. Plant shutdown was terminated at 0830 (EDT) as conditions for 3.0.3 no longer existed. The licensee will notify the NRC Resident Inspector.|
|ENS 49928||19 March 2014 03:07:00||At 2252 on 03/18/2014, the Unit 3 reactor automatically scrammed due to a turbine trip from a high Main Turbine moisture separator level. Initial indications show the level controller for 3B2 Moisture Separator failed to adequately maintain level. Additionally local manual control attempts failed to restore moisture separator level. Main Steam Isolation Valves remained open with main turbine bypass valves controlling reactor pressure. Reactor feedwater pumps are in service to control reactor water level. Primary Containment Isolation System Groups 2, 3, 6 and 8 containment isolation and initiation signals were received. Upon receipt of these signals all required components actuated as required. Neither High Pressure Coolant Injection nor Reactor Core Isolation Cooling initiation signals were received. The reactor had been operating near 35% power during scheduled power ascension. This event is reportable within 4 hours per 10CFR 50.72(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). NRC Resident Inspector has been notified.|
|ENS 49920||17 March 2014 09:25:00||On 3/17/2014 at 0514 (CDT) the reactor was manually scrammed from approximately 41% core thermal power due to a steam leak in the turbine building. All control rods fully inserted and all systems actuated and operated as designed. All Main Steam Isolation Valves were manually shut. The Reactor Core Isolation Cooling System was manually initiated to assist in level control and pressure control. No safety relief valves actuated automatically. Manual cycling of safety relief valves and Reactor Core Isolation Cooling are being used to maintain reactor water level and pressure within normal bands. Group 2 and 3 RHR isolation signals were received; however no valve movement occurred since the affected valves are normally closed. This event is reportable under 10CFR50.72(b)(2)(iv)(B) for the reactor trip and 50.72(b)(3)(iv)(A) for the manual start of the reactor core isolation cooling system. The licensee informed the NRC Resident Inspector.|
|ENS 49918||15 March 2014 11:32:00||On 3/14/2014 at 1330 hours, the Plant Shift Superintendent was notified by Fire Services that adequate pressure could not be verified during post maintenance testing of an inoperable High Pressure Fire Water System for the C-335 building. During investigation into the issue, it was discovered that eleven operable High Pressure Fire Water Systems were also impacted. The facility is in TSR mode three as all cells have been sampled UF6 negative and the cell motors are not energized. In this mode, the High Pressure Fire Water Systems are still required to be operable per TSR 22.214.171.124; however, due to inadequate pressure, the eleven systems may not have been able to perform the intended safety function. The eleven operable High Pressure Fire Water Systems were declared inoperable and hourly fire patrols were initiated according to TSR LCO 126.96.36.199.9.1 at 1520 hours on 3/14/2014. On 2/26/2014 at 1212 hours, three High Pressure Fire Water System Sectional Valves were isolated and declared inoperable in order to isolate a section of header which contained a small leak. Isolating these valves resulted in a single supply to twelve High Pressure Fire Water Systems in C-335. Upon investigation of the lack of pressure, it was determined that the sectional supply valve in the remaining supply loop was not operating as designed and is suspected to be the cause of the restricted water flow. Sectional supply valves that had been used for isolation of the small leak have been opened to provide unrestricted flow to the affected sprinkler systems. Following the confirmation of adequate flow, the sprinkler systems were declared operable and the hourly fire patrols were discontinued at 1739 hours on 3/14/2014. The small leak will be monitored until repairs can be initiated or the leak becomes unmanageable. At that point TSR LCO actions will be entered as necessary. This event is reportable as a 24 hour event in accordance with 10CFR 76.120(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 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 Region II (Marvin Sykes) has been notified of this event via voice mail.|
|ENS 49879||6 March 2014 19:33:00|
The condition described below is being reported as an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B) and per the guidance of NUREG-1022, Rev. 3. On 03/06/2014 at 0906 PST, Diablo Canyon Power Plant (DCPP) identified a nonconforming condition involving the Emergency Diesel Generator (EDG) ventilation exhaust plenums installed in Unit 1 and Unit 2. Specifically, the radiator exhaust plenums and exhaust piping need to be re-evaluated to ensure adequate protection against flying debris that could be generated by a tornado. The occurrence of such an event is highly unlikely and there is no imminent concern regarding severe weather involving tornados. The EDGs are located inside the power plant structure and are capable of performing their safety function. Compensatory measures are being developed to address the associated nonconformance. This event does not adversely affect the health and safety of the public. The licensee informed the NRC Resident Inspector."
This condition does not adversely affect the health and safety of the public. Based on an extent of condition review being performed for this event, the issue identified in the original event notification 49879 has also been determined to similarly affect the ventilation systems associated with the Unit 1 and 2 Vital 480 volt AC switchgear and battery/inverter equipment. The condition described in this update is being reported as an unanalyzed condition per 10 CFR 50.72(b)(3)(ii)(B) and as an event or condition that could have prevented the fulfillment of a safety function per 10 CFR 50.72(b)(3)(v)(A). Compensatory measures are being developed to address the associated condition. The licensee informed the NRC Resident Inspector. Notified R4DO (Azua)
|ENS 49878||6 March 2014 11:22:00|
Feed tube level sensor was found to be in a state such that it has failed as an IROFS (Item Relied On For Safety) for a fire accident sequence leading to the loss of a criticality control. Another IROFS is in place and the accident sequence continues to meet performance requirements. Less than a safe mass was always maintained. This event is being reported because only one item relied on for safety, as documented in the Integrated Safety Analysis summary, remains available and reliable to prevent a nuclear criticality accident, and has been in this state for greater than eight hours. Portion of the plant affected: FMO (Fuel Manufacturing Operation) Press. The licensee notified the North Carolina State Radiation Protection Branch, New Hanover County Emergency Management, and NRC Region II.
On 3/6/14, GNF-A (Global Nuclear Fuels - America) conservatively made a 1 hour event notification (EN 49878) due to a discovery that a feed tube level sensor had failed. After further review, it has been determined that a second control remained available, reliable, and the remaining IROFS was sufficient to meet performance requirements. As a result, the event notification is retracted. Notified R2DO (Vias) and NMSS EO (Lombard).
|ENS 49883||7 March 2014 16:52:00||The following was received from the State of Texas via email: On March 7, 2014, the Agency (Texas Department of Health) was notified by the licensee that one of its vehicles had been stolen on February 18, 2014. The vehicle contained a moisture density gauge, Troxler model 3340, serial number 17909, containing 1.48 GBq of Americium-241/Beryllium serial number 4713350 and 0.30 GBq of Cesium-137 serial number 507394. A local police report was filed. No recovery of vehicle or gauge. One violation cited for not reporting incident within regulatory timeframe. Any further information will be reported within SA 300 guidelines. Texas Incident #: I-9162 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 49833||16 February 2014 21:37:00||At time 1725 (EST) on 02/16/2014 a contract worker fell ten to fifteen (10-15) feet off of a ladder while descending the ladder above the 565 foot level in the Containment Building. The individual received lacerations to the forehead and potential other injuries. For his work activities the individual was dressed in anti-contamination clothing. The individual was transported via off-site ambulance in his anti-contamination clothing to H.B. Magruder hospital in Port Clinton, Ohio. A First Energy (FENOC) Radiation Protection (RP) Technician accompanied the injured person in the ambulance during the transfer to the hospital. A FENOC RP Supervisor along with the RP Technician assisted at the hospital while the injured individual received medical treatment. The protective anti-contamination clothing was removed from the individual. A full body frisk was performed and showed no contamination on the individual. Smear readings on the anti-contamination booties and gloves showed (approximately) 1500 dpm / 100 square cm. The clothing was bagged and subsequent readings and surveys on the outside of the bag showed no readings above background. The bagged anti-contamination clothing will be transported back to Davis-Besse NPS by FENOC RP Supervision. The H.B. Magruder Hospital Treatment Room and the off-site ambulance also were frisked and showed no residual contamination. The licensee notified the State of Ohio and Lucas and Ottawa Counties. The licensee notified the NRC Resident Inspector.|
|ENS 49831||14 February 2014 23:08:00|
An earthquake was felt in the control room at Virgil C. Summer Nuclear Station Unit 1 (VCSNS) at approximately 2223. The earthquake was confirmed with USGS at 2240. An unusual event in accordance with the emergency plan was declared at 2245 and NRC was notified at (2308 EST) via ENS per 10 CFR 50.72(a)(1)(i). The plant is stable and continues to operate at 100% power. The licensee has completed preliminary building walkdown inspections with no damage noted. There were no injuries. The licensee notified the NRC Resident Inspector.
(V.C. Summer Nuclear Station) VCSNS has terminated the (Unusual) Event after walkdowns of the plant were satisfactorily completed and no aftershocks were felt. An update to ENS has also been made. The event was terminated at 1045 (EST) on 2/15/2014. The plant continues to operate at 100% power and the licensee has notified the NRC Resident Inspector. Notified R2RA (McCree), R2DO (Guthrie), NRR EO (Monninger), NRR (Uhle), IRD (Grant), DHS SWO, FEMA Ops, DHS NICC, and NuclearSSA via email.
|ENS 49824||12 February 2014 16:17:00||At 1113 CST on February 12, 2014, R-21 Circulating Water Discharge Monitor failed during routine surveillance testing and was declared nonfunctional. 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. This results in a Loss of Emergency Assessment Capability while R-21 is out of service. This is a reportable condition in accordance with 10 CFR 50.72(b)(3)(xiii). There are no radioactive leaks that will impact the Circulating Water System as evidenced by normal readings on R-21 prior to its failure. In addition all other liquid effluent radiation monitors that monitor releases to the Circulating Water Discharge are in-service and have normal readings. Corrective maintenance is in progress and will continue until the monitor is returned to service. 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 49804||7 February 2014 08:19:00|
This event is being reported in accordance with 10CFR 50.72(b)(2)(i), 'Initiation of a Shutdown Required by Technical Specifications.' At 2043 hours (EST) on February 06, 2014, the Perry Nuclear Power Plant entered Technical Specification 188.8.131.52 Primary Containment Isolation Valves (PCIVs), action C.1, due to leakage identified during local leak rate testing of the containment penetration for the Containment and Drywell Purge system. Leakage was identified on the outboard containment isolation valve resulting in the plant exceeding the limit for secondary containment bypass leakage. The Containment and Drywell Purge system penetration is normally isolated and remains isolated in accordance with Technical Specifications. Action C.1 requires restoration of the leakage rate within four hours. At 0043 hours on February 7, 2014, the plant entered Technical Specification 184.108.40.206, 'Primary Containment Isolation Valves (PCIVs)', action E as the leakage rate was not restored. Action E requires the plant be in Mode 3 in 12 hours and Mode 4 in 36 hours. At 0600 hours on February 07, 2014, the Perry Nuclear Power Plant initiated a shutdown in accordance with Technical Specification 220.127.116.11, action E. Repairs to restore the penetration leakage to within allowable limits are in progress. The NRC Resident Inspector has been notified.
At 0943 hours (EST) the reactor shutdown to comply with Technical Specification 18.104.22.168 action E was terminated (with the reactor at 42% power). A blind flange was installed downstream of the outboard containment isolation valve. Local leak rate testing of the containment penetration for the Containment and Drywell Purge system verified that leakage was within the limits for secondary containment bypass leakage. The NRC Resident Inspector has been notified. The licensee has commenced increasing reactor power. Notified R3DO (Orlikowski)
|ENS 49803||7 February 2014 03:04:00|
At 03:00 EST on Friday, February 7, the Cook Nuclear Plant (CNP) Technical Support Center (TSC) air conditioning and charcoal filtration systems will be removed from service for scheduled maintenance. Under certain accident conditions the TSC may become unavailable due to the inability of the air conditioning and charcoal filtration systems to maintain a habitable atmosphere. Compensatory measures exist to relocate TSC personnel to the unaffected unit's control room if necessary. TSC ventilation system maintenance and post maintenance testing is scheduled to be completed by 16:00 EST on Friday, February 7. The licensee has notified the NRC Resident Inspector. This notification is being made in accordance with 10 CFR 50.72 (b)(3)(xiii) due to the loss of an emergency response facility.
TSC ventilation system was returned to service following successful maintenance and post maintenance testing at 14:30 EST on Friday, February 7, 2014. The licensee has notified the NRC Resident Inspector." Notified R3DO (Orlikowski)
|ENS 49805||7 February 2014 08:44:00||The following information was obtained from the State of Texas via email: On February 7, 2014, the Agency (Texas Department of State Health Services) was notified by the licensee's Radiation Safety Officer (RSO) a source disconnect had occurred on February 6, 2014. The disconnect occurred while radiographers were working at a field location using a QSA model 880D exposure device with a 39 curie iridium-192 source. The radiographers were working in a shooting bay and had been working for about five hours when the problem occurred. The RSO stated the radiographer had extended the source to the collimator and was attempting to retract the source when the source moved a few inches and then could not be moved in any direction. The radiographer contacted the RSO who directed them to secure the area and wait for his arrival. The RSO and an assistant arrived at the facility and attempted to retract the source, but could not. The RSO stated the guide tube had a sharp bend near the area of the collimator and believed that could have damaged the guide tube causing the hang-up. The RSO disconnected the guide tube from the exposure device and using a remote handling tool, slid the guide tube down the drive cable. When the end of the drive cable was exposed, the RSO noted the source was not attached. He then shook the source from the guide tube and onto the ground. The source was covered with bags of lead shot. The RSO connected the source to the drive cable and was able to retract the source into the exposure device. The exposure device and drive cable connectors were tested using a go-no-go device and both passed. The highest exposure received from the source retrieval was 38 millirem. No overexposures occurred and no member of the general public was exposed due to this event. The RSO stated the guide tube would be returned to the manufacturer for inspection. Additional information will be provided as it is received in accordance with SA-300. TX Incident: I-9155|
|ENS 49790||4 February 2014 10:26:00||The Millstone Station Stack Radiation Monitor, RM-8169, was removed from service at 0854 EST for pre-planned maintenance. The Stack Radiation Monitor will be restored to service after maintenance. The licensee informed the Connecticut Department of Environmental Protection, Waterford Dispatch and the NRC Resident Inspector.|
|ENS 49632||13 December 2013 20:07:00||During a feed pump shift from the motor driven reactor feed pump to the 'A' turbine driven reactor feed pump during power ascension, reactor pressure vessel (RPV) water level was approaching the Reactor Protective System (RPS) automatic scram set point when a manual reactor scram was inserted. All control rods fully inserted. RPV water level is being maintained by the normal condensate booster / feedwater systems. RPV pressure is being maintained via normal main steam system. There were no actuations of any Emergency Core Cooling System and no Safety Relief Valves actuations. All systems responded as expected. The licensee notified the NRC Resident Inspector.|
|ENS 49629||13 December 2013 16:43:00||In 2010, changes within the Dominion FFD (Fitness for Duty) program resulted in currently 11 individuals not being subjected to random FFD testing although required. Two individuals are actively badged at Surry Power Station and North Anna Power Station (NAPS). Both individuals accessed the protected area at NAPS recently. The other nine individuals were not badged but perform duties that require them to be subject to the FFD program. The affected individuals are now within the random testing program. This event is reportable per 10CFR26.719(b)(4), 'Any 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.'" A work review will be conducted for the two actively badged individuals. The program error has been corrected. The licensee has notified the NRC Resident Inspector and the Louisa County Administrator. See similar Surry report EN #49630.|
|ENS 49593||2 December 2013 11:03:00||At 0904 (EST) on Monday, December 2, 2013, Nine Mile Point Unit 2 was manually scrammed from approximately 40% thermal power due to the loss of both reactor recirculation pumps during a planned downpower evolution. Manual scram of the unit is procedurally required upon loss of both recirculation pumps to avoid potential power/flow oscillations. The reactor recirculation pumps failed to transfer to the low frequency motor generators when downshifted from fast speed. The cause of the loss of both reactor recirculation pumps is not known at this time. (Nine Mile Point Unit 2) NMP2 has commenced cooldown in preparation for the forced outage to investigate and commence repairs. 10 CFR 50.72(b)(2)(iv)(B) requires reporting within 4 hours of any event or condition that results in actuation of the reactor protection system (RPS) when the reactor is critical. All control rods fully inserted. No safety systems actuated. Decay heat is being removed via the main condenser. The "A" recirculation pump was restarted in low speed at 1045 EST. Unit 2 is in a normal shutdown electrical lineup. The licensee informed the NRC Resident Inspector and will inform the New York State Public Services Commission.|
|ENS 49631||13 December 2013 19:44:00|
On December 13, 2013 it was determined that a reportable condition has existed at Columbia Generating Station since 1500 hours (PST) on November 25, 2013. At 1500 hours on November 25, 2013, the Control Room Envelope (CRE) was declared inoperable based on the inability to ensure that the Control Room Emergency Filtration (CREF) System would be able to maintain a positive differential pressure with all areas surrounding the CRE boundary. Columbia does not have the installed instrumentation to directly monitor the differential pressure between the Main Control Room (MCR) and certain areas adjacent to the MCR. The pressure in the adjacent areas is controlled by placing conservative limits on allowed breach size for these adjacent areas. On November 25, 2013, it was identified that the combined breach size associated with several doors in these adjacent spaces resulted in exceeding the allowed limit. Based on exceeding the allowed breach size limit to the adjacent areas, the Control Room Envelope was declared inoperable, and Technical Specification Action Statement 3.7.3.B.1 was entered. An additional breach was discovered on 12/05/13 from a hole in ductwork passing through the cable spreading room, which is one of the adjacent areas to the CRE boundary, and that condition was added to the existing action statement 3.7.3.B.1. These are conditions that could have prevented fulfillment of a safety function of structures that are needed to mitigate the consequences of an accident and are reportable under 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector.
In the case of this event, the Control Room Envelope (CRE) was declared inoperable and Technical Specification Action Statement 3.7.3.B.1 was entered conservatively based upon limited knowledge at the time of discovery. There is no installed differential pressure indication between the CRE and this area adjacent to the Main Control Room (MCR), therefore conservatively this adjacent area is included in the CRE, and leakage is administratively controlled. The leakage between the MCR and the adjacent area included in the CRE exceeded this administrative limit. Testing performed on November 20, 2013, prior to the December 13, 2013 reported events, as well as testing after the event, January 10, 2014, has demonstrated that the leakage identified does not prevent the Control Room Envelope (CRE) from establishing and maintaining the required differential pressure to ensure fulfillment of its required safety function for Control Room Habitability. The ability of the Control Room Emergency Filtration (CREF) system to perform its function of pressurizing and maintaining the Main Control Room positively pressurized with respect to its surroundings was not lost due to the leaking doors and duct specified in the event. Performance of the surveillance without these breaches sealed validated this conclusion. The event described above should not have been reported, as the Control Room Envelope was always operable and capable of fulfilling its safety function with the existing breaches and did not constitute a reportable event as conditions that could have prevented fulfillment of a safety function of structures that are needed to mitigate the consequences of an accident under 10 CFR 50.72(b)(3)(v)(D). The licensee notified the NRC Resident Inspector. Notified the R4DO (Spitzberg).
|ENS 49568||21 November 2013 14:47:00||The State of Florida received a call from their licensee regarding a pick-up truck fire with a soil moisture density gauge in the back. On 11/20/13 at approximately 11:25 am EST on I-95 southbound near mile marker 169, a 2002 Ford F150 owned by the licensee caught fire. The exact cause of the fire is unknown. The fire spread so quickly, the licensee was unable to remove the Troxler Model 3440 portable density gauge (S/N: 77-4969) from the truck bed. The licensee notified emergency responders who arrived within 15 minutes and extinguished the fire. The gauge transportation container was damaged in the fire but there was no breach. The licensee RSO responded to the scene to conduct surveys. All surveys of the area were at or near background levels. Surveys of the gauge container and gauge were normal. The gauge itself showed no signs of damage from the fire. The licensee plans on installing the gauge into a spare transportation container to return it to the manufacturer for inspection and repair if necessary. The licensee will instruct employees to avoid driving through high grass and underbrush to avoid a similar incident in the future.|
|ENS 49511||7 November 2013 11:41:00||Technical Support Center (TSC) out of service due to planned maintenance. 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) because the work activity affects the functionality of an emergency response facility. Planned maintenance activities during the Unit 2 outage on 11/7/2013 will render TSC out of service for approximately 48 hours. The unit 2 work is on electrical equipment that will impact the power supply to the TSC. If an emergency is declared requiring TSC activation during this period, the alternate TSC will be used per existing emergency planning procedures. The NRC Resident Inspector has been notified. This event poses no threat to the public or station employees.|
|ENS 49518||8 November 2013 09:05:00||A non-licensed, contract, supervisory employee had a confirmed positive during a for-cause fitness-for-duty test. The employee's access to the plant has been terminated. The licensee notified the NRC Resident Inspector.|
|ENS 49508||6 November 2013 11:08:00||The following Agreement State Report was received via facsimile: The licensee contacted (the NC Radiation Protection) agency at approximately 0930 EST with the following info: A dump truck at a construction site ran over an lnstroTek Model 3500 Xplorer moisture density gauge. The area of the accident was cordoned off and the licensee RSO (was sent to) report back to agency with survey readings and leak test results. The gauge sources contain 11 mCi of Cs-137 and 44 mCi of Am-241/Be. A survey by the licensee (RSO) showed a maximum reading of 0.5 mrem/hr. The instrument make/model was of the survey meter was not specified. Sources appear to be intact. The gauge and debris were returned to the gauge case (doubles as shipping container- DOT 7A Yellow II). The gauge will be shipped to lnstroTek corporation for repair or disposition. NC State Report Number: ICD 13-20|
|ENS 49507||6 November 2013 03:26:00|
When transferring power supplies to a non-safety related cooling tower bus for planned outage maintenance, R-21, the Circulating Water Discharge Radiation Monitor was removed from service at 0058 (CST) and returned to service at 0111 (CST). There is no installed backup for R-21 which has an emergency response function to provide indication of gaseous liquid effluent release to the environment. This monitor has no compensatory measure that will allow timely classification of two NUE (Notification of Unusual Event) and Alert classifications when out of service. This resulted in a loss of emergency assessment capability while R-21 was out of service. There are no radioactive leaks that impact the Circulating Water System. The licensee notified the NRC Resident Inspector.
* * * RETRACTION AT 1155 EST ON 12/2/13 FROM WAYNE EPPEN TO DANIEL MILLS * * *
Based on further reviews of plant drawings and discussion with the Radiation Monitor System Engineer, R-21 was not inoperable when power supplies were transferred to the non-safety related Cooling Tower Bus. While R-21 was logged out of service during the transfer on November 6, 2013, it did not lose power and was not out of service. R-21 did not lose the ability to provide continuous monitoring of discharge canal effluent or monitoring in the event of an unplanned radiological release. Our defense in depth strategies are relied upon to take actions to protect the health and safety of the public. The licensee has notified the NRC Resident Inspector. Notified the R3DO (Duncan).
|ENS 49492||1 November 2013 09:51:00||On November 1, 2013 at 0309 EDT, 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 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 22.214.171.124 was entered for both units at 0309 EDT and exited at 0315 EDT. 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 notified the NRC Resident Inspector. See similar event number #49489.|
|ENS 49487||30 October 2013 16:44:00||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 Millstone Power Station Unit 2 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 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 short to the ground plane. Simultaneously, it is postulated that the fire causes another DC wire from the opposite polarity on the same battery to also short to the ground plane. 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. Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. This condition is being reported pursuant to 10 CFR 50.72(b)(3)(ii)(B). The licensee notified the NRC Resident Inspector.|
|ENS 49479||28 October 2013 16:03:00||The State of California submitted the following via email: At 1130 (PDT) on October 28, 2013, the RSO for AMEC called to report a stolen moisture density (M/D) gauge: Troxler, model 3440, S/N 36036 that contains 8 millicuries of Cs-137 and 40 millicuries of Am-241/Be. The stolen gauge was last seen on Friday, October 25, 2013 by the authorized user, who secured the M/D gauge in safe mode inside the locked yellow transportation box which was placed inside a locked metal box. The contractor's tools at the site in Mission Valley, CA was also burglarized and a police report was filed. The (metal) box's lock had been drilled out and the chain attaching it to the building was cut. RHB (Radiation Health Branch) advised the RSO to place a Lost/Found notice in the local paper offering a reward for return of the gauge to AMEC." CA 5010 Number: 102813 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 49485||30 October 2013 14:21:00|
The following report was received from the North Dakota Department of Health via e-mail: A portable moisture/density gauge possessed by ALLWEST Testing & Engineering, LLC of Hayden, Idaho containing a 10 mCi Cesium-137 sealed source and a 50 mCi Americium-241:Beryllium sealed source was crushed by a piece of heavy equipment (excavator bucket) at a temporary job site located southwest of Dickinson, North Dakota. The portable gauge user had failed to maintain constant surveillance of the gauge. Upon observing the heavy equipment running into the gauge, the user flagged down the heavy equipment operator to halt his activity. The operator lifted the bucket off the gauge and set it to the side. The gauge user instructed the operator to relocate to an area approximately 150 feet from the damaged gauge. Subsequently, he phoned his immediate supervisor and the Radiation Safety Officer (RSO) for guidance. An area at an approximate distance of 150 feet in three directions from the damaged gauge was roped off with 'Caution Radiation' tape. The fourth direction (east) consisted of a large hill of soil not readily accessible. 1:02 PM (MST): The gauge user placed a call to the ND State Radio emergency response number to report the event. 1:08 PM (MST): ND State Radio personnel notified the Stark County Emergency Manager (SCEM) who in turn notified the Southwestern District Health Unit Executive Officer (HUEO). 1:10 pm (MST): The HUEO notified the ND Department of Health Radiation Control Program Manager (RCPM) of the event. The RCPM informed the HUEO the gauge should remain in place until radiation surveys had been performed and the site evaluated by his department. 2:17 pm (MST): The HUEO and the SCEM were present at the event site. They met with the gauge user and the Westcon, Inc. HSE Coordinator for a briefing of the event. The HUEO performed an initial radiation survey using a calibrated SE International, Inc. Model Radiation Alert Inspector survey instrument (SN 35756). The survey was performed beginning at the outer boundary moving inwards toward the damaged gauge. The reading at a distance of 4 feet from the source was 0.063 mR/hr. The background reading was 0.013 mR/hr. The HUEO and SCEM instructed the gauge user to leave the gauge in place and wait for North Dakota Department of Health personnel to be on site the next morning to evaluate the site. Visual assessment of the gauge showed evidence of the two source housings to be physically intact. October 28, 2013: 6:00 am (MST): As instructed by ALLWEST Testing & Engineering's RSO, the gauge user relocated the damaged gauge and associated fragments to the gauge transport case. The case was secured in the box of his pickup truck. 9:00 am (MST): North Dakota Department of Health (NDDoH) personnel were on site to perform interviews, radiation surveys and evaluation of the site. Radiation surveys were performed by the NDDoH using a calibrated Ludlum Model 19 microR meter (SN 270378) and a Canberra Dineutron neutron meter (SN 18327). The background readings were 12 microR/hr and 0.027 mR/hr respectively. The highest gauge shipping container surface readings were 3.1 mR/hr (gamma) and 0.09 mR/hr (neutron). Surveys of the pathway from the initial impact of the bucket to the gauge's final resting spot revealed background readings. A leak test of the gauge was performed and shipped via overnight express for analysis. The leak test results demonstrated no leakage. The gauge was transported by licensee personnel back to their Idaho office to make arrangements for final disposal. Throughout the event, the gauge user had not worn personnel dosimetry. Exposure calculations will be performed by the RSO.
The North Dakota Department of Health has completed their investigation. Several non-compliances were identified during the reactive inspection. A letter of apparent non-compliance was sent to the licensee. Corrective actions will be required to be submitted in response to this letter. The North Dakota Department of Health is recommending the LER (License Event Report) for closure. Notified the R4DO (Vasquez), R1DO (Cook), and FSME Events Resource via email.
|ENS 49463||22 October 2013 16:59:00||A non-licensed contract employee supervisor had a confirmed positive for alcohol during a random fitness for duty test. The contractor's access to the plant has been terminated. The licensee notified the NRC Resident Inspector.|
|ENS 49459||22 October 2013 12:41:00||An employee experiencing chest pains reported to the on-site dispensary this morning. The plant nurse administered first aid and then sent the employee to an off-site medical facility. A whole body survey of the employee's plant clothing was performed. The maximum amount of contamination found was present on the employee's right boot, 197,000 dpm/100cm2. Prior to leaving the Restricted Area, the employee's plant clothing was removed, and he was whole body frisked out of the plant. The employee was free of contamination upon release.|
|ENS 49446||17 October 2013 16:25:00||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 the Perry Nuclear Power Plant resulting in a potentially unanalyzed condition with respect to 10 CFR 50 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, it is postulated that 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 (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). Interim compensatory measures (i.e., fire watches) have been implemented for affected areas of the plant. The licensee has notified the NRC Resident Inspector. See the following related Event Numbers: 49411, 49419, 49422, and 49444.|
|ENS 49413||5 October 2013 01:19:00||On October 4, 2013, during the Beaver Valley Unit 1 refueling outage, a planned visual examination of the interior containment liner and coatings was being performed. The containment design consists of an interior steel liner that is surrounded by reinforced concrete. An area approximately 0.4 inches by 0.28 inches was discovered that penetrated through the containment steel liner plate. With the plant currently shut down and in Mode 6 the containment as specified in Technical Specification 3.6.1 is not required to be operable. The cause of this discrepancy is currently being evaluated. This is reportable pursuant to 10 CFR 50.72(b)(3)(ii)(A). The Site Resident Inspector has been notified.|
|ENS 49407||4 October 2013 02:12:00|
On October 3, 2013, at 2045 (CDT) hours, a defect (pinhole through-wall leak) was identified on the drain line for the LS 2-2365, HPCI TURBINE INLET DRAIN POT LEVEL SWITCH. The defect was identified during investigation of leakage near LS 2-2365. The LS 2-2365, HPCI TURBINE INLET DRAIN POT LEVEL SWITCH, is provided to detect a failure of the HPCI steam trap during standby line-up. The location of the defect, is in Class 2 Safety related piping. HPCI is a single train safety system and this notification is being made in accordance with 10CFR50.72(b)(3)(v)(D). The instrument isolations for LS 2-2365 have been close and the leak has been isolated. There is no increase to plant risk and RCIC (Reactor Core Isolation Cooling) is available. The licensee will inform the NRC Resident Inspector.
The purpose of this notification is to retract the ENS report made on October 4, 2013, at 0212 EDT (ENS Report # 49407). Upon further investigation the pinhole through-wall leak discovered in the Unit 2 HPCI room was in a weld at a 'Tee' downstream of the Unit 2 HPCI turbine inlet drain pot level switch (LS 2-2365) on drain line 2-2386B-1-B. The defect was characterized as a 1/16-inch rounded hole due to gas porosity (with no evidence of cracking). A subsequent evaluation performed by Quad Cities Station considering the defect size, location, and characterization confirmed the Unit 2 High Pressure Coolant Injection (HPCI) system would have performed its safety function when required. Based on this subsequent evaluation, ENS Report # 49407 is being retracted. Note: On October 3, 2013, at 1155 CDT the Unit 2 HPCI drain line leak was isolated and HPCI was declared operable. The licensee has notified the NRC Resident Inspector. Notified R3DO (Lipa).
|ENS 49354||18 September 2013 16:28:00||The Commonwealth of Massachusetts submitted the following information via email: A dose that likely differs from the prescribed dose by more than 50 rem to an organ, the liver, and the total dose delivered differs from the prescribed dose by 20% or more. The licensee's radiation safety officer reported to the Agency (Massachusetts Radiation Control Program) on 9/18/2013 that on same date licensee administered yttrium-90 SIR-Spheres to patient's left lobe of liver for treatment of metastatic disease; that 16.8 millicuries of yttrium-90 was prescribed; and that 12.06 millicuries was administered resulting in an under dose of 28.1 percent. The licensee discovered the event upon measurement of waste material including dose vial as part of normal quality assurance. The licensee's radiation safety officer reported that the authorized user and the referring physician have been notified and it is unknown at time of the report whether the referring physician has elected to notify the patient. The licensee's radiation safety officer reported that medical staff does not believe that this error will adversely affect the patient treatment outcome. The licensee will submit a written report within 15 days in accordance with the (Massachusetts) requirements of 105 CMR 120.594(A)(4) to include why event occurred and actions to prevent recurrence. Root cause and corrective action are not known at this time and the Agency continues to investigate. System Name: Brachytherapy Product; Manufacturer: Sirtex Wilmington LLC 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 49349||16 September 2013 16:57:00||System(s) Affected: RM-8169 Vent Stack Radiation Monitor Actuations & Their Initiation Signals: None Causes (If known): Pre-Planned Maintenance Effect of Event on Plant: Loss of assessment capability Actions Taken or Planned: Compensatory samples - restore RM-8169 to service ASAP. Additional Information: None The licensee notified the NRC Resident Inspector, the State of Connecticut and local Waterford Dispatch.|
|ENS 49348||16 September 2013 13:58:00||On September 16, 2013, the licensee notified the Agency (Texas Department of Health) that on September 13, 2013, one of its radiography crews had been unable to retract a Iridium-192 source back into a SPEC 150 exposure device at a temporary work site in Tilden, Texas. Following the first exposure of the morning, the source pigtail would not retract fully into the device. An authorized person performed the source retrieval. This individual received 20 millirem and the individual that assisted him received 10 millirem. No member of the public received any exposure as a result of this event. The licensee's radiation safety officer inspected the inside of the source guide tube using a tube scope and found that the coil in the source guide tube was pushed out into the inside of the tube where the connector fitting was crimped onto the guide tube. The connector had just been replaced by a local service company. An investigation into this event is ongoing. Information will be provided as it is obtained in accordance with SA-300. Texas State Report # I-9114|
|ENS 49347||16 September 2013 13:58:00||On September 16, 2013, the Agency (Texas Department of Health) received notification from the licensee that on September 12, 2013, one of its radiography crews working at a temporary work site in Tilden, Texas, had been unable to retract an Iridium-192 source back into its SPEC 150 exposure device. After several attempts, the connector on the end of the cable came off which allowed the cable to come all the way through the device and left the source inside the guide tube. Source retrieval was performed by authorized person who received 120 millirem, an assistant who received 55 millirem, the radiographer trainer who also assisted received 100 millirem (including the day's work) and the radiographer trainee who received 32 millirem (including the day's work). The licensee will have the crank assembly, cable, and connector evaluated. No member of the public received any radiation exposure as a result of this event. An investigation into this event is ongoing. More information will be provided as it is obtained in accordance with SA-300. Texas State Report # I-9113|
|ENS 49335||9 September 2013 22:39:00|
At 1400 (CDT) on September 9, 2013, plant personnel found HELB barrier HATCH-1/TB blocked. The hatch is diamond plate steel located on the turbine floor. Pallets, a fan and a gantry were positioned on top of the hatch possibly preventing pressure relief during a HELB event. This issue is being reported as an unanalyzed condition per 10CFR50.72(b)(3)(ii)(B). All items were removed from the hatch by 1800 on September 9, 2013. The licensee notified the NRC Resident Inspector.
The licensee reviewed the design basis calculations and analyses for HELB events in the area where Hatch-1/TB is located. The review determined that it is acceptable to block or hold Hatch-1/TB down and that the painted markings on the hatch are overly restrictive. In looking at HELB calculation of record, feed water break at the feed water pumps, there is no flow path modeled between the HELB volumes. Therefore, if Hatch-1/TB is blocked, the physical flow path is in accordance with the Gothic model of the HELB volume. Further, the analyses also indicated that no credit is taken for the hatch to relieve as no HELBs are postulated in the room under the hatch. Finally, Hatch 1/TB structural integrity was verified assuming a HELB occurred with the hatch blocked as described in notification 49335 (pallets, fan and gantry) the barrier would function as designed. Licensee initiated a Work Request to remove any markings on Hatch-l/TB that indicate do not block or hold down. The licensee will notify the NRC Resident Inspector. The Region 3 Duty Officer (Valos) was notified.
|ENS 49328||6 September 2013 14:47:00|
At 0607 hours PDT on September 6, 2013, the Radwaste Building process radiation monitoring sample rack was declared non-functional due to a loss of power to the sample rack. The cause of the loss of power is under investigation. At 0945 hours PDT, auxiliary sampling equipment was installed to collect samples from the associated effluent release pathway. To compensate for the loss of assessment capability while the Radwaste Building process radiation monitoring sample rack is non-functional, field team survey results will be used if required. This event is being reported as a loss of emergency assessment capability in accordance with 10CFR50.72(b)(3)(xiii). The licensee has notified the NRC Resident Inspector.
Repairs have been completed and the Radwaste Building process radiation monitoring sample rack has been returned to service and declared functional at 1631 hours PDT. The licensee has notified the NRC Resident Inspector. Notified the R4DO (Gaddy).