|Entered date||Event description|
|ENS 49565||21 November 2013 04:23:00||On November 20, 2013 at 2240 (EST), secondary containment drawndown testing surveillance failed to meet acceptance criteria SR 188.8.131.52.5 due to maximum flow rate exceeding the allowable value. SSES (Susquehanna Steam Electric Station) previously entered SR 3.0.3 at 0900 on 11/15/2013 due to not meeting SR 184.108.40.206.4 and SR 220.127.116.11.5 because of an untested alignment of the 101 bay with ventilation aligned as a no zone during past performances of the drawdown testing surveillance. The surveillance being performed on 11/20/2013 was testing this previously unsurveilled alignment. Upon failure of the surveillance, secondary containment ventilation was realigned to the previously tested 818 hatch alignment. Upon restoration of secondary containment ventilation to a known operable alignment, secondary containment LCO 18.104.22.168 was cleared and operability restored. This event is being reported under 10 CFR 50.72(b)(3)(v)(c) and per the guidance of NUREG 1022, Rev. 3 section 3.2.7 as a loss of a Safety Function. There is no redundant Susquehanna Secondary Containment System. The licensee has placed administrative controls on the 101 bay doors to prevent loss of secondary containment during the investigation to determine the reason for the surveillance test failure. The licensee has notified the NRC Resident Inspector.|
|ENS 52509||24 January 2017 19:40:00||During the semi-annual inspection, the licensee discovered that the shutter on a fixed density gauge would not close. The gauge is used to monitor slurry flow at a gold mining operation and the shutter is normally open. The gauge remains in operation and does not present a safety concern to personnel. There are no exposures involved with this event. The licensee has notified the manufacturer of the gauge to request that an authorized technician repair the shutter. The gauge is Ronan SA-1 Density Gauge containing a Cs-137 1000 mCi source, Serial # M4884.|
|ENS 52508||24 January 2017 14:00:00||On January 24, 2017, at 1000 hours (CST), Operations was notified that two Secondary Containment interlock doors (between the Unit 2 Reactor Building and Unit 2 Turbine Building) were open simultaneously. The doors were immediately closed and Secondary Containment pressure remained negative. Unit 1 and Unit 2 share secondary containment. This condition represents a failure to meet Surveillance Requirement 22.214.171.124.2 given two doors in a single access opening were open. As a result, entry into Technical Specification 126.96.36.199, Condition A. was made momentarily due to Secondary Containment being inoperable. This event is reportable under 50.72(b)(3)(v)(C) as an event or condition that could have prevented the fulfillment of a safety function. The NRC Senior Resident Inspector has been notified. The cause of this event was due to an equipment interlock (solenoid) failure and the doors are currently blocked closed.|
|ENS 52506||23 January 2017 12:27:00||The following report was received from the State of Texas via email: On Sunday, January 22, 2017, at 1041 (CST), (the Texas Commission on Environmental Quality) CID (Critical Infrastructure Division) received a call from URI's Radiation Safety Officer (RSO), for the Kingsville Dome facility, that a fire occurred on the west side of well field 17 in Production Area 3, on Sunday morning, January 22, 2017. (The RSO) stated the cause of the fire appeared to be a cigarette butt. A URI operator working at the site noticed the fire. When the operator arrived at the fire location, he noticed that the fire department (FD) was already putting out the fire. At the time of the call, according to (the RSO), the fire was under control and the FD was making sure to put out the hot spots. At the time of the fire, the well field was on a shutdown status and there were no other activities at the well field. The preliminary findings were damage to some of the well field fences and wells in the well field. The extent of damage is still being evaluated and there are no reports of contamination or exposures, outside of the facility, at this time. The URI, Inc., Kingsville Dome is an In-Situ Uranium Recovery facility in Texas. Texas Report Number: None assigned at this time.|
|ENS 52490||14 January 2017 13:09:00||At 0613 EST on 1/14/2017, with the unit in Mode 2 at 0 percent power at the start of Refueling Outage 22, Drywell inspection identified a through-wall leak on the 3/4-inch vent line off the bonnet of valve 02MOV-43A, Reactor Water Recirc Pump A Suction Isolation Valve, in the Reactor Coolant System (RCS) loop inside the Primary Containment. This condition constitutes a defect in the primary coolant system. This event notification is being made in accordance with 10 CFR 50.72(b)(3)(ii)(A) as a condition that results in the condition of the nuclear power plant, including its principal safety barriers, being seriously degraded. The licensee has notified the NRC Resident Inspector.|
|ENS 52485||12 January 2017 16:35:00||The following report was received from the State of Colorado via email: Detail: Wibby Environmental sold the business to Phenova in 2010. Through a long process of trying to reach the company contact at Wibby, contact with Phenova was made in February of 2016. After several attempts to discover what occurred with the assets of Wibby Environmental, one device is being reported as lost/abandoned. According to Phenova they do not see the gas chromatograph as being on the asset sheet at the time of purchase. It is unknown as to what Wibby Environmental did with the device or if the device was returned for disposal. Our office has received no records from manufactures including CJ Bruyn & Company, showing the Gas Chromatograph was returned. Device information: Model # N610-0133 Source SN# 0767, NI-63, 15 mCi, Date shipped from CJ Bruyn 02-27-2004. Event Description: Due to the length of time for discovery this case is closed. Colorado Event Report ID No.: CO 17-0002 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 52488||13 January 2017 15:32:00|
The following report was received from the State of Utah via email: DWMRC (Division of Waste Management and Radiation Control) was notified of the event at about 1500 MST, on January 12, 2017. The licensee indicated that the shipper, Honeywell (Converdyne) had notified the NRC and the US DOT. This incident report is the initial notification of the NRC Operations Center from the DWMRC. The following description is the initial information received by DWMRC. The information will be updated as the Division obtains more information. A van carrying (three) barrels of (solid and wet) radioactive materials was received at the White Mesa Mill. When the employees were unloading the van they realized that a barrel, maybe two, had rusted bottoms and was leaking. Although the bottom of the barrels were not rusted through, they were characterized as 'being soft'. The barrels were on plastic sheeting that should have restricted the leaking materials to the plastic; however, when the barrels were unloaded, the employees noted that there was a hole in the sheeting. When the RSO was informed of the incident and arrived at the van, he noted that there was visible leakage on the girders and the siding of the van. Unfortunately, he was unable to take measurements of the area where the leak occurred to determine the radiation levels. The RSO (Radiation Safety Officer) stated that the van had been pulled in and cleaned prior to any measurements being taken, so the only measurements were taken after the van was cleaned.
On January 12, 2017, at approximately 1142 MST, a TAM International van carrying barrels of KOH alternate feed materials from Honeywell International arrived at the White Mesa Mill Scale house. According to Honeywell the materials are basically uranium ore concentrates. The White Mesa employees began unloading the van, but as they began offloading some of the last barrels, they noticed that some of the barrels were leaking. From what they could tell, three barrels were potentially leaking . The barrels were not rusted through, but were 'soft' and allowed liquid contained in the solid materials to leak from the barrel. The material leaked from the barrel onto a plastic sheet; however, the plastic sheet had an opening (rip, tear) through which the material was able to pass. The RSO (Radiation Safety Officer) was notified of the event at about 1300 MST. When the RSO reached the van, the employees had completed emptying the van and had 'cleaned' the van. The RSO indicated he was able to see visible evidence that the material had been able to leak from the van, but the area had been cleaned and no valid measurements or samples of the materials leaked from the van were able to be taken. Pictures were taken and from the pictures provided to the Division (Utah Division of Waste Management and Radiation Control), it appears that only a small quantity of liquid material was leaking from the barrels. Because the contents of the barrels are primarily solid and only contain small amounts of liquid and the liquid contents in the barrels were leaking through the bottom of the barrel, through a plastic sheet, across the van floor and out of the van along a girder and the siding, it is likely that only small quantities of materials were leaking from the van. The weather across the nation has been fairly stormy this past week and has included both rain and snow. Because of the rain and snow that has been occurring while the vehicle was traveling to the Mill site, it is likely that any small amounts of materials that leaked from the van were subsequently washed away and diluted. The Mill analyzes each alternate feed stream annually to verify the isotopic content of the alternate feed. At the last verification, the Honeywell KOH material contained, approximately 633,000 mg/kg uranium (about 61% uranium), 39 pCi/g Pb-210, 44.6 pCi/g Ra-226, 358 pCi/g Th-230, and 27 pCi/g Th-232. This event was also reported by National Response Center to the Nuclear Regulatory Commission , Report # 1168447, on 01/12/17 at 1720 EST. Utah Event Number: UT170001
|ENS 52486||12 January 2017 18:25:00||A non-licensed supervisor has been found in violation of the Duke Energy Fitness for Duty Policy during a random fitness for duty test. The individual's access to the plant has been suspended. The licensee has notified the NRC Senior Resident Inspector.|
|ENS 52507||23 January 2017 15:11:00||The following report was received from the State of Nebraska via email: (The licensee) contacted (the Nebraska Office of Radiological Health) on January 12, 2017. The licensee indicated that they purchased the theater from Perkins Delaware, LLC but were not informed of the Tritium exit signs or the related license. The licensee became aware of the signs and the related license when they received the annual renewal and related inventory. The licensee opted to dispose of the signs with SRB technologies. During the packing process, the licensee discovered two signs missing from the inventory and is unable to ascertain when they were lost. Lost Source Number: 1, Source/Radioactive Material: SEALED SOURCE LUMINOUS, Manufacturer: Isolite, Model Number: 2040, Serial Number: A4N1925, Radionuclide: H-3, 7.5 Ci 277.5 GBq. Lost Source Number: 2, Source/Radioactive Material: SEALED SOURCE LUMINOUS, Manufacturer: NRD, INC., Model Number: T4001, Serial Number: 61821, H-3, 25 Ci 925 GBq. Cause: Human Error. Corrective Action: None. The State of Nebraska considers this event to be closed. Nebraska Item Number: NE170001. 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 52487||13 January 2017 13:56:00|
The following summary was excerpted from report #EMD-GM-17-02, provided by the Curtis-Wright Electro-Mechanical Corporation: Curtiss-Wright Electro-Mechanical Corporation (CW-EMD) has identified a condition which is currently being evaluated to determine if a significant safety hazard exists in the casing for the AP1000 Reactor Coolant Pump (RCP). The investigation is underway, but will not be completed within the required 60 days from discovery as required by paragraph 21.21(a)(1) of the regulation. This submission constitutes CW-EMD's interim report on the condition in accordance with paragraph 21.21(a)(2) of the regulation. Casings for the AP1000 Reactor Coolant Pump. The casings have an excess material condition in the transition region between the cast bowl and the suction nozzle. The condition is a discontinuous (non-tangential) axisymmetric feature between the cast bowl region and machined suction nozzle OD. The as-built transition feature on the casings has not been analyzed and is not included in CW-EMD Engineering Memorandum 7242, Revision 1, Volume 1, which is the applicable generic design report for the AP1000 RCP casing and main closure. At present, the following domestic deliveries have been made: - 8 casings have been provided to V.C. Summer Units 2 and 3. - 8 casings have been provided to Vogtle Units 3 and 4. Name and address of the individual or individuals informing the Commission: Brian W. Eckels, General Manager Curtiss-Wright Electro-Mechanical Corporation 1000 Wright Way Cheswick, Pa 15024
The following update was received via email: Curtiss-Wright Electro-Mechanical Corporation (CW-EMD) submitted an interim report concerning a deviation in the AP1000 Reactor Coolant Pump casing via letter EMD-GM-17-02, dated January 12, 2017. This letter updates the status of the investigation. 1. Extent of condition evaluation has been completed. CW-EMD has confirmed that the discontinuous (non-tangential) axisymmetric feature between the cast bowl region and machined suction nozzle OD is limited (domestically) to the 8 AP1000 RCP casings delivered to V.C. Summer units 2 & 3. 2. AP1000 RCP casing stress and fatigue supplemental analyses have been performed to assess the excess material condition discovered on the suction nozzle OD. The deviation has been determined to still satisfactorily meet the pressure boundary stress requirements of the ASME Code and can therefore be accepted as-is. 3. The analysis included generation of a 2D axisymmetric FEA (Finite Element Analysis) of the casing with a bounding representation of the as-built geometry, based on extraction of cross-sectional profiles from the V.C. Summer laser scan data of 2 casings. The 2D FEA was used to determine the stress concentration factor(s) associated with the as-built deviated feature, enabling a fatigue strength reduction factor to be estimated and applied to the existing 3D FEA casing fatigue (nominal design) analysis. 4. The maximum fatigue usage at the location of the deviation on the casing suction nozzle was conservatively estimated to increase from 0.287 to 0.633. This condition still satisfies the 1.0 usage limit specified in the ASME Code. 5. Therefore, this deviation does not constitute a substantial safety hazard, and the condition has been determined to be not reportable under the provisions of Title 10 Code of Federal Regulations, Part 21. Notified the R2DO(Bartley) and Part 21 Group (via email).
|ENS 52489||13 January 2017 16:46:00||The following report was received from the State of Louisiana via email: Event type: Level density gauge on a process had a shutter malfunction. Eagle was performing their annual operational checks when the malfunction was discovered. The malfunction is the gauge shutter is stuck in the open position on an active process. The gauge Service Company, BBP Sales will make the repair or replacement after the assessment of the device. They discovered the shutters would not close. The Gauge is a RONAN SA1-C-10 device/source holder, S/N 9527GG, loaded with a 200 mCi Cs-137 source. Notifications: LDEQ (Louisiana Department of Environmental Control) was notified by Eagle US 2 in a message on the voice mail system on 1/10/2017. The notice was sent to the Compliance Radiation Assessment Section of LDEQ. The notification was readdressed when Eagle was contacted on 1/12/2017, to get the preliminary information report. Eagle US 2 will send LDEQ a corrective action and a final report no later than 30 days of the corrective action. The Eagle US 2 preliminary written report was received 1/12/2017 at 1457 CST. Event description: On 1/10/2017, Eagle US 2 was performing their annual inventory and operational checks of their licensed devices. During their routine annual maintenance checks, the shutter malfunction was discovered. The gauge shutter would not close. Eagle US 2 called a service contractor, BBP Sales, to evaluate the situation and determine the best course of action to correct the problem. (The contractor) was unable to close the shutter and will determine the course of corrective action. The cause appears to be the corrosive environment where it (the gauge) is installed and used. The source and device with shutter stuck open will remain installed and utilized on the process until the repairs are made. This is not a radiation exposure hazard and does not pose a health and safety situation for the Eagle US 2 employees or the general public. This event is considered closed by LDEQ. This event is being reported to the NRC as required by Regulatory Requirement 10 CFR 30.50(b)(2). Event Report ID No.: LA-17001.|
|ENS 52440||16 December 2016 09:27:00|
On December 15, 2016, a grid area employee was moving containers in the plating room, where his finger was pinched between two containers. He was taken to the emergency room where he was treated for an injury to the tip of his left ring finger. The event did not involve special nuclear material or contamination and is classified as an industrial safety incident.
This concurrent report is being made under Paragraph (c) of 10 CFR 70, Appendix A because a 24 hour report was made to the South Carolina Department of Labor (at 0700 EST on 12/16/16) per 29CFR1904.39. The employee was not admitted to a hospital and was sent home after treatment. The licensee has notified NRC Region II.
|ENS 52426||13 December 2016 14:40:00||On 12/13/16 at 1410 (EST), the following voluntary communication was made to the State of Tennessee in accordance with Tennessee Valley Authority's (TVA) guidance for communicating inadvertent radiological spills/leaks that are below regulatory reporting requirements to outside agencies and in alignment with NEI 07-07, 'Industry Ground Water Protection Initiative'. On 12/12/16, Sequoyah Nuclear Plant determined that a spill of greater than 100 gallons (approximately 3000 gallons) of condensate storage tank water with tritium levels of 1560 pCi/L (picocuries per liter) was spilled to a yard drain. The spill occurred on 12/5/16, during the filling of the Unit 1 #4 steam generator when a hose connection on a temporary fill skid failed. No elevated tritium levels have been detected at the Sequoyah Yard Drainage Pond before or after the event. This is reported in accordance with 10CFR50.72(b)(2)(xi), the required reportable threshold for tritium is 20,000 pCi/L. The licensee will notify the NRC Resident Inspector.|
|ENS 52431||14 December 2016 14:39:00||The following report was received from the State of California via email: On December 13, 2016, the RSO (Radiation Safety Officer) of Cal Land Engineering, contacted the RHB (California Radiologic Health Branch) at the Brea, CA office about a gauge that had been run over. The gauge was a CPN MC-3, S/N M310706239 (10 mCi Cs-137, 50 mCi Am:Be-241). The gauge operator had removed the gauge from it's transport case, but had not set up or used the gauge at the site (the Cs-137 source was still in the shielded position). The operator then stepped away from the gauge to answer his cell phone. While the operator was on the phone, a pickup truck backed into the gauge. The gauge body remained intact, but the guide tube and source rod had been sheared off the body of the gauge. After the incident, the area was isolated and the operator contacted the RSO, who then contacted RHB. Pictures were provided by the gauge operator and forwarded to RHB for review. After the pictures confirmed that the gauge body was intact, the RHB inspector authorized the operator to place the gauge in the storage case, secure it for transport, and then return the gauge to their (licensee's) office. An RHB inspector met the operator at their office to inspect the gauge. A Canberra Inspector 1000 with the LaBr (IPROL-1) and neutron (IPRON-N) probes was used to verify the sources remained in the gauge. The Cs-137 source was identified by the Inspector 1000 and a neutron dose of 20 counts per second (with a background of 0 counts per second) confirmed the Am:Be-241 source was present. A Victoreen 450 CHP was used to survey the gauge. The contact dose rate was 20 mR/hr (background 0.01 mR/hr), which is consistent for an MC-3 gauge. The gauge will remain in storage until disposal can be arranged. Maurer Technical Services will be used to assist in the disposal. California 5010 number: 121316|
|ENS 52409||5 December 2016 16:24:00||A non-licensed supervisory employee had a confirmed positive for a prohibited substance during a random fitness-for-duty test. The individual's unescorted access to the plant has been denied. The licensee has notified the NRC Resident Inspector.|
|ENS 52405||2 December 2016 13:23:00||A Supervisor employee failed to report for completion of random testing following notification. The employee's access to the plant has been suspended. The licensee has notified the NRC Resident Inspector.|
|ENS 52408||5 December 2016 15:11:00||The following report was received from the State of Tennessee via email: On December 3, 2016, M & EC (East Tennessee Materials & Energy Corporation) staff notified DRH (Tennessee Division or Radiological Health) regarding an event in which a (Fire Protection) water line outside an M & EC building began leaking (on 12/1/16). The water entered the building flooding a non-contaminated area (radiologically clean) and a bermed contaminated area. After the leak was stopped, water removal was performed Friday and over the weekend. M & EC staff were able to segregate removal of water from the clean area from removal of water from the bermed contaminated area. A total of (approximately 136,000 to 242,000) gallons was removed, a large percentage of which was in the contaminated area. M & EC is in process of completing the clean up, and is also in the process of performing radio-analysis of both the clean and the contaminated water. The contaminated bermed area had an average contamination of (approximately) 15,000 dpm/100 cm2. No release to the public has occurred and a follow-up investigation by the Tennessee DRH is being conducted and additional information will be provided by DRH when available. Report #: TN-16-173|
|ENS 52404||1 December 2016 15:05:00||On December 1, 2016, at 1014 hours (EST), the speakers for a portion of the bottom floor of Building 305, office and lunch room areas, were inadvertently disabled during construction activities to remove obsolete equipment. Special Nuclear Material (SNM) is not processed, handled, or stored within the areas where the speaker system was disabled; however, these areas require evacuation in the unlikely event of a nuclear criticality accident as described in the NFS (Nuclear Fuel Services) Emergency Plan. There were no actual radiological or other nuclear safety consequences. The potential consequence was that in the event of a nuclear criticality accident, evacuation could have been delayed for those personnel within these areas with a resultant increase of postulated doses. Additionally, awareness of a fire event could have been delayed for these same personnel. A series of compensatory actions were taken including restricting access to the affected areas to essential personnel, and establishing radio communications between the personnel in these areas and the personnel continuously monitoring the alarm station panel. On December 1, 2016, at approximately 1159 hours (EST), full compliance was restored by repairing a disabled cable in the speaker circuit and by successfully performing a speaker system test in the affected areas. The licensee notified the NRC Resident Inspector on. December 1, 2016.|
|ENS 52412||7 December 2016 13:12:00||In accordance with 10 CFR 52.99(c)(2), Vogtle Units 3 and 4 Construction is making this notification to NRC for determining that Inspection, Test, Analysis, and Acceptance Criteria (ITAAC) 2.6.01.02.ii (ECS System Seismic Category I Equipment Design Basis Loads) for both units requires additional actions to restore its completed status. The Closure Notification for this ITAAC (NRC Index No. 580) was originally submitted on May 17, 2016 (reference ML 16138A080 and ML 16166A030). On November 1, 2016, it was determined by the Vogtle 3&4 Contractor that modifications to the Reactor Coolant Pump (RCP) breaker (i.e., switchgear) cabinet design were required to ensure compliance with the applicable portions of IEEE 384, 'Standard Criteria for Independence of Class 1E Equipment and Circuits.' The modification involves an engineering change which adds different equipment to the RCP Breaker cabinet that function to trip the RCP. The new components were not previously qualified for use in the RCP breaker cabinet assembly. Additional seismic qualification type testing and analysis of components are being performed for the added components in the RCP breaker. Update of the Equipment Qualification Data Package and Equipment Qualification Summary Report for the RCP breaker to confirm the breaker withstands seismic design basis loads and Licensee's acceptance is in progress. The revised ITAAC Completion Notice will be submitted to the NRC once all related ITAAC activities have been completed. The licensee has notified the NRC Resident Inspector.|
|ENS 52342||2 November 2016 15:05:00||On November 2, 2016, while performing a re-evaluation of the radiological consequences of the Fermi 2 control rod drop accident (CRDA), DTE Electric Company (DTE) identified a non-conservatism in the current Fermi 2 design and licensing basis of the CRDA. As described in the Updated Final Safety Analysis Report (UFSAR) Section 15.4.9, the current design and licensing basis assumes that the post-CRDA release pathway consists of carryover with steam to the turbine condenser. The re-evaluation has identified that a forced release from the gland seal exhausters (GSEs) could also occur which could result in post-CRDA radiological consequences that exceed the current 10 CFR 100.11 offsite dose limits and Standard Review Plan 6.4 (General Design Criterion 19) main control room dose limits when operating at low power conditions. The unanalyzed condition described above only applies to low power operating conditions (i.e. less than 10% power) since the fuel damage postulated as a result of a CRDA is only credible under low power operating conditions when an individual control rod worth is high. Fermi 2 is currently at 97% power and, therefore, the plant condition is currently bounded by the design and licensing basis such that the condition currently does not exist and no immediate actions are required. However, Fermi 2 has operated at low power levels several times in the past three years. Those periods of operation at low power represent unanalyzed conditions that significantly degraded plant safety since the occurrence of a CRDA during those periods could have resulted in offsite and main control room doses exceeding regulatory limits. Therefore, this 8-hour notification is being reported in accordance with 10 CFR 50.72(b)(3)(ii)(B). There was no adverse impact to public health and safety or to plant employees. The licensee has notified the NRC Resident Inspector.|
|ENS 52343||2 November 2016 15:35:00|
During panel walkdown, it was discovered that a tag out for the 'C' Residual Heat Removal pump suction valve was active and the valve was open with its breaker open. This rendered the valve inoperable and Technical Specification 188.8.131.52 Action C for penetration with one inoperable PCIVs was entered. The action was to isolate the penetration by closing the valve within (4) hours or restore power. The event was discovered at 0845 (EDT) and the breaker was closed at 0925 (EDT). Technical Specification 184.108.40.206 Action C (Isolate penetration within 4 hrs.) was entered at 0130 (EDT) (time breaker was opened per tagout) and exited at 0925 (EDT). This condition of non-compliance existed from 0530 (EDT) on 11/02/16 until 0925 (EDT) on 11/02/16. This event is being reported under 10CFR50.72 (b)(3)(v)(C). NRC Resident has been notified.
In accordance with Technical Specification (TS) 220.127.116.11, Primary Containment Isolation Valves, the TS Basis states that one or more barriers are provided for each penetration so that no single credible failure or malfunction of an active component can result in a loss of isolation or leakage that exceeds limits assumed in the safety analyses. When two or more barriers are provided, one of these barriers may be a closed system. During this event, one of the barriers in the penetration became inoperable: 'C' Residual Heat Removal (RHR) pump suction valve 10MOV-13C. After the initial NRC notification, it was confirmed that the RHR system piping is classified as a closed system outside containment. The integrity of the closed-loop RHR system is verified by monitoring the keep-full system. Since the piping is maintained full of water during normal and post-accident modes of operation, a barrier against post-accident, gaseous, containment leakage is provided. Therefore, the affected penetration could have performed its intended safety function since there was redundant equipment in the same system which was operable. This event is not reportable under 10 CFR 50.72(b)(3)(v)(C) and the original notification may be retracted. Finally, the primary containment penetration with 10MOV-13C is with a closed system and the completion time per TS 18.104.22.168 Required Action C is 72 hours. The valve was restored to operable prior to exceeding this time. The licensee notified the NRC Resident Inspector. Notified the R1DO (Dentel).
|ENS 52340||1 November 2016 19:01:00|
At 1743 CDT on 11/1/16, readings were obtained that indicated ammonia above IDLH (Immediately Dangerous to Life and Health) values in the Unit 1 Auxiliary Building and an Alert was declared by the Emergency Director. The plant is stable with Unit 1 at 2% power. Unit 2 is at 100% power (and stable). The site is currently investigating to determine the source and ventilate the area." The licensee has informed the NRC Resident Inspector. Notified DHS SWO, FEMA Ops Center, USDA Ops Center, HHS Ops Center, DOE Ops Center, DHS NICC Watch Officer, EPA EOC, FEMA National Watch Center (email), FDA EOC (email), Nuclear SSA (email).
The following is the Emergency Preparedness 8-hour report for the Plant Farley ALERT declaration. An ALERT was declared at 1743 CDT on November 1, 2016, based on Emergency Action Level (EAL) HA3 (Release of toxic, asphyxiant, or flammable gases within vital areas which jeopardizes operation of systems required to maintain safe operations or establish or maintain safe shutdown) due to an ammonia discharge into the Auxiliary Building. Farley Unit 1 was in Mode 2 at 2 percent power and Unit 2 was in Mode 1 at 100 percent power throughout the event. Actions completed during and leading up to termination from the event included walkdown of the area to locate the ammonia source, and isolating the ammonia source while atmospheric monitoring was performed in all affected and adjacent areas. Installation of additional ventilation fans was completed. This event has been entered into the Farley Corrective Action Program under Condition Report 10293519 (Ammonia levels in 100 Feet Radside). A review of the Alert classification has determined that the classification was timely and accurate. All required notifications were completed accurately and in a timely manner. The ERO (Emergency Response Organization) notification system functioned as expected and all emergency response facilities were activated in a timely manner. There were no missing or injured personnel during the event. There is no evidence of tampering or sabotage of plant equipment leading to this event. At 2202 CDT on 11/1/2016, the ammonia source was identified as Valve N1P20V913. At 2228 CDT, the ammonia source was isolated. Final assessment of the Auxiliary Building showed that the ammonia levels were less than 5 ppm. It was determined that a deficiency in the valve component was the cause of the ammonia leak. The Alert was terminated on 11/1/2016 at 2340 CDT. The Licensee has notified the NRC Resident Inspector. Notified R2DO(Ernstes), IRD (Stapleton), NRR EO (Miller). Notified DHS SWO, FEMA Ops Center, USDA Ops Center, HHS Ops Center, DOE Ops Center, DHS NICC Watch Officer, EPA EOC, FEMA National Watch Center (email), FDA EOC (email), Nuclear SSA (email).
|ENS 52339||1 November 2016 18:15:00||The following report was received from the State of California via email: On November 1, 2016, the RSO (Radiation Safety Officer) of RMA Group, contacted RHB ICE (State of California, Radiation Health Branch, Inspection, Compliance, and Enforcement Section) to report the theft of a moisture density gauge from the gauge user's pickup truck while he was home for lunch (in Grand Terrance, CA). The nuclear gauge was inside a Type A transportation box was reportedly chained and secured properly but the chains were found cut. Troxler Labs model 3430, # 24568, contains 0.3 GBq of Cs-137 and 1.48 GBq of Am-241/Be. A power generator and other assorted equipment was also stolen. (The RSO) will notify the local newspaper to place a Reward Notice for information leading to the safe return of the gauge. (The RSO) was asked to obtain a detailed report from the authorized gauge user. The gauge's theft was reported to San Bernardino Police Dept. as well as the local calibration facilities were alerted of anyone bringing the gauge to them for service. (The State of California) RHB requested a copy of the police report and newspaper notice. California State Report 5010 Number: 110116 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 52336||31 October 2016 18:07:00||On October 31, 2016, an informal, voluntary communication was made to the Louisiana Department of Environmental Quality (LDEQ) and the Saint Charles Parish Emergency Operating Center concerning a spill of greater than 100 gallons (estimated 150 gallons total) of Tritium contaminated water to the storm drains located within the Protected Area. Tritium activity was 8.201E-06 micro Ci/ml and no detectable gamma activity. The spill lasted approximately 5 minutes before the pump moving the tritium contaminated water was secured. The contaminated water came from the Condensate Polisher Backwash Storage Tank and discharged from a (loose) hose connection going to a baker tank located within the protected area. The licensee has notified the NRC Resident Inspector.|
|ENS 52337||1 November 2016 12:00:00|
This 24 hour report is being made as per 10CFR72.75. On October 31, 2016, at 1100 (CDT), the Supplemental Cooling System was secured to the Dry Cask Storage Hi-Track Transfer cask, as allowed by Dry Fuel Storage T.S. 3.1.4, in preparation for moving the Hi-Track transfer cask. At approximately 1500, after moving the Hi-Track into position for downloading including securing the cleats to the mating device, the next step in the procedure was to lower the Hi-Track. When lowering the Hi-Track, a crane overload condition occurred. With the crane, attached the Supplemental Cooling System can not be operated and has remained secured. The Supplemental Cooling System is classified as 'Important to Safety Category B' and is required to be utilized, as necessary, to maintain the peak fuel cladding temperature below the allowed limits. T.S. 3.1.4 allows the Supplemental cooling system to be secured for up to 7 hours during the Hi-Track transfer process and then followed by a 7 day allowed outage time. This is being reported based on 10CFR72.75 which states in part that a 24 hour Non-emergency notification is required if; 'An event in which important to safety equipment is disabled or fails to function as designed'. The Hi-Track is presently in a safe condition with supplemental cooling still secured. The licensee has notified the NRC Resident Inspector.
On November 1, 2016 at 1100 CDT Waterford 3 notified the NRC of a 24 hour reportable event per 10CFR72.75. The notification is documented under EN# 52337 based on the information known at that time. Follow up investigation determined that the conditions required per 10CFR72.75 were not met and should not have been reported. 10CFR72.75 states in part that a 24 hour non-emergency notification is required if, 'An event in which important to safety equipment is disabled or fails to function as designed.' It also requires that, 'The equipment is required by regulation, license condition, or certificate of compliance to be available and operable to prevent releases that could exceed regulatory limits, to prevent exposures to radiation or radioactive materials that could exceed regulatory limits, or to mitigate the consequences of an accident.' Based on a review of the bases for Certificate of Compliance, T.S. 3.1.4 thermal analysis shows that the fuel cladding temperature would not exceed the short term temperature limits applicable to an off normal condition. The bases further states that because the thermal analysis is a steady-state analysis, there is an indefinite period of time available to make repairs to the Supplemental Cooling System. The completion time of 7 days to restore the system per T.S. 3.1.4 is considered an appropriate and reasonable amount of time to plan the work and complete repairs. Based on this information there was no possibility of exceeding regulatory limits or the need to mitigate the consequences of an accident. Waterford 3 is retracting event notification EN 52337. The licensee has notified the NRC Resident Inspector. Notified R4DO (Campbell).
|ENS 52338||1 November 2016 14:07:00||This report was received by the State of Washington via email: During preparation of a radiography exposure, the radiographer and another radiation worker from Bechtel, attempted to untangle their dosimetry from the camera apparatus. In doing so, they left their dosimetry next to the camera during a shot. The radiographer exposed his TLD (Thermo Luminescent Dosimetry) and pocket dosimeter, as well as a client's electronic dosimeter during the exposure. Their dosimetry minus the radiographer's rate alarm was left next to the camera during the exposure. The radiographer wears dosimetry issued by Northwest Inspection and their client, Bechtel, sub contractor of the US Department of Energy. The radiographer's pocket dosimeter was off scale. Bechtel's electronic dosimeter showed an exposure dose of 300 mrem. The radiographer reported the incident to the RSO (Radiation Safety Officer). The radiographer's TLD was sent to the dosimetry processing facility. Additional training for the radiographer has already taken place. In view of the fact that no 'persons' were overexposed, a spare TLD will be issued to the radiographer and will be allowed to continue to work. The radiography camera is a QSA Model Number A424-9, Serial Number 32886G, containing an Ir-192 34.5 Ci source. Washington State Incident Number: WA-16-045|
|ENS 52315||23 October 2016 23:41:00||At 1908 CDT on 10/23/16, elevated vibration readings were identified on Cooper Nuclear Station (CNS), Control Room Emergency Filter System (CREFS) supply fan A. Vibration readings were evaluated by Engineering and were determined to be indicative of bearing failure on supply fan A. The Control Room declared CREFS inoperable and entered LCO 3.7.4, Condition A, which requires restoration of CREFS to operable status in 7 days. Repair activities have been initiated for this condition. The plant is currently in Mode 5, with refueling activities and OPDRVs (Operation with Potential to Drain Reactor Vessel) in progress. CNS is not currently in the mode of applicability for a USAR defined accident. This condition is being conservatively reported under 10 CFR 50.72(b)(3)(v)(D) as a single train safety system that is required to be OPERABLE during situations under which significant radioactive releases can be postulated. The NRC Resident Inspector has been notified.|
|ENS 52341||2 November 2016 11:38:00||The following report was received from the State of Georgia via email: Northside Hospital's Radiation Safety Officer called the Department (Georgia Radioactive Materials Program Environmental Protection Division) on 10/21/2016, informing us of a misadministration with the HDR (High Dose Rate) that occurred approximately two weeks ago. The patient was to receive 5 vaginal treatments consisting of 1 cylinder, 1 capri and 3 capris. The misadministration occurred during the second treatment. The capri was inserted into the rectum instead of the vagina. The Authorized User (AU) was not certain if a misadministration occurred until 2 weeks after the treatment. The AU requested the assistance of the radiologist who confirmed that the rectum was treated instead of the vaginal area. Based on the calculations, the rectum received approximately 350 cGy, what is to be considered a low dose. Additional information from the licensee will be forthcoming. Treatment material used: Varian Medical Systems, model: Gamma Medplus iX, with an Ir-192 source of less than 22 Ci. 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 52360||10 November 2016 10:33:00||The following information was excerpted from a report received from the State of Georgia via email: On Wednesday, November 9th, it was reported to the Department (Georgia Radioactive Materials Program) by Theragenics Corporation in Buford, GA that an employee's dosimetry report indicated that she had exceeded the annual whole body dose limit of 5000 mRem. An investigation has been conducted by the employer and no possible explanation has been found as of yet. Our office has an ongoing investigation and further details will be provided as we receive them. (The licensee reports receiving) notification from their dosimetry processor, on October 12, an individual received a whole body dose of 5,215 mRem. (The licensee) spent the last month trying to recreate a situation that could have exposed this worker to a dose of this magnitude and can't make it happen. According to discussions, with the worker, she never left her dosimetry in a lab where it could have been exposed to an unshielded source of radiation. According to her supervisor, team leads in the lab in which she worked, and HP (Health Physics), she is a very good radiation worker and, there is never anything out of the normal at her work station. (The worker) did take her whole body dosimeter and both finger rings home with her one evening. Even more baffling is her ring dose is less than 200 mRem. (The licensee has) tried to recreate that scenario of taking the dosimeters home as well and have not had a dosimeter come back from processing that is above its minimal level of detection. It is (licensee's) interpretation of (the CFR,) Reportable Events, that this report of an overexposure is due to (the State of Georgia) 30 days after notification, or Friday, 11/11/16. (The licensee is) waiting on additional information (from the dosimetry processor) prior to report submittal.|
|ENS 52289||8 October 2016 05:47:00||On October 8, 2016, while reducing power for a planned refueling outage, the unit was taken offline by opening the main generator output breakers. With the reactor at approximately 7 percent power in MODE 1, an unplanned actuation of the reactor protection system occurred. At 0150 (EDT), an unexpected steam valve transient occurred while main turbine valve control was being transferred from throttle valve to governor valves during main turbine overspeed testing. This resulted in an automatic low steamline pressure Safety Injection and Reactor Trip. All safety systems functioned as expected. The operations staff responded to the event in accordance with applicable plant procedures. The plant stabilized at normal operating no-load reactor coolant system (RCS) temperature and pressure following the reactor trip, with decay heat being removed using steam generator power operated relief valves. Steam generator water levels are being maintained using auxiliary feedwater. All emergency core cooling system (ECCS) equipment is available. The cause of the steam valve transient is under investigation. This condition is being reported as an ECCS discharge to RCS, an unplanned reactor protection system actuation, and a specified system actuation in accordance with 10 CFR 50.72(b)(2)(iv)(A), 10 CFR 50.72(b)(2)(iv)(B). and 10 CFR 50.72(b)(3)(iv)(A). This condition does not affect the health and safety of the public or station employees. The NRC Resident Inspector has been notified. The Safety Injection occurred for approximately 6 minutes and Pressurizer level increased to approximately 71%. The Main Steam Isolation Valves closed as a result of the Safety Injection and Decay Heat is being removed using the Steam Generator Atmospheric Relief Valves. There is no known primary to secondary leakage.|
|ENS 52244||16 September 2016 00:14:00|
During a review of commercial grade dedication records for a Unit 1 (Emergency Core Cooling System ECCS) Centrifugal Charging Pump discharge pressure gauge, it was identified that the process side of the diaphragm seal utilizes a Teflon (PTFE) gasket. Further review found Teflon (PTFE) to be installed in the pressure gauge seal assembly for all four of the Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2. Teflon (PTFE) is a restricted material normally prohibited from use in contact with reactor coolant or in radiation environments. Teflon (PTFE) is not radiation tolerant and significantly degrades in a radiation environment. The Teflon (PTFE) used in these pressure gauges could fail during a LOCA (Loss of Coolant Accident) which could cause the (ECCS) Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2 to be inoperable, and exceed system leakage limits. Excessive leakage from systems which would contain post-LOCA recirculation fluid would challenge onsite and offsite dose estimates and in-plant post-accident accessibility. This represents an unanalyzed condition. Currently, the pressure gauges for all four of the (ECCS) Centrifugal Charging Pumps and both of the Positive Displacement Charging Pumps on Units 1 and 2 have been isolated until this issue can be further evaluated. Luminant Power believes that the Teflon (PTFE) has existed in the pressure gauges since initial plant licensing. Luminant Power is currently investigating the extent of the condition and repair techniques. The NRC Resident Inspector has been notified.
On 09/16/2016, Comanche Peak reported an ENS Report (no. 52244) related to the identification of teflon-containing pressure-seal assemblies installed on the suction and discharge sides of the centrifugal charging pumps and on the suction side of the positive displacement pump. The technical concern was the potential for the teflon-containing assemblies to leak if subjected to post-LOCA recirculation fluid and associated radiation levels. Subsequent investigations by Engineering have determined: (1) the centrifugal charging pumps were operable for all postulated non-LOCA design bases events which required their operation and (2) for postulated LOCA scenarios which would involve radiation levels sufficient as to call into question the ability of the teflon-containing assemblies to maintain system pressure boundary, the ECCS function would be fulfilled in the event one or all of the charging pumps had to be removed from service (due to system leakage) and limiting (control room) doses would have remained below applicable regulatory limits. Based on the above, the condition described in ENS report no. 52244 is not considered to be an un-analyzed condition as described in10 CFR 50.72(b)(3)(ii)(B), nor is it considered to be a condition which could have led to a potential uncontrolled radiation release per 10CFR 50.72(b)(3)(v)(C), nor is it considered to be a condition which could have prevented fulfillment of a safety function under 10 CFR 50.72.(b)(3)(v)(D). The NRC Resident Inspector has been notified. Notified the R4DO (Azua).
|ENS 52310||20 October 2016 11:45:00||The following Part 21 Report was received from the licensee via facsimile: 10 CFR Part 21 Notification - Westinghouse Life Line D Type LAC Induction Motor Model HSDP 4000V, 700hp. This is a non-emergency facsimile notification required by 10 CFR 21.21(d)(3)(i). A written notification in accordance with 10 CFR 21.21(d)(3)(ii) will be provided within 30 days. NextEra Energy Seabrook, LLC has determined there is evidence that the Westinghouse Life Line D Type LAC Induction Motor Model HSDP 4000V, 700 hp motors, original to plant construction, have a deviation from expected quality of construction. Of the four motors purchased for Unit 1, Primary Component Cooling Water (PCCW) pumps CC-P-11-D and CC-P-11-C failed after approximately 87,000 hours of operation on July 23, 2008, and November 21, 2008, due to a short caused by localized heating. On June 13, 2015, CC-P-11-B failed due to shorted windings following approximately 32,000 hours of operation. Failure analysis determined the heating was most likely caused by a turn-to-turn short circuit which led directly to the eventual failure of the entire coil-to-ground. Forensic examination identified that the coil insulation was not tightly wrapped, resulting in less than 100% resin penetration throughout the stator insulation system (i.e., voids). The voids led to poor thermal conductivity and localized hot spots that accelerated the degrading of insulation properties over time. Based on the failure analysis, it can be concluded that the undesirable coil quality is most likely attributed to workmanship, not motor design. The failure of motor insulation could cause phase-to-phase and phase-to-ground faults which ultimately would prevent motor and PCCW pump from performing their intended safety function. The identified condition appears to be a deviation from expected quality of construction and the three failures indicate that the condition is likely applicable to all the motors manufactured at the same time. The NRC Senior Resident has been notified.|
|ENS 52183||16 August 2016 09:14:00||On 8/12/16 at 1356 CDT, an ISFSI security officer noticed that two male children were walking west on the railroad tracks on the south end of the OCA, approximately 50 yards onto the property. The children appeared to be between 7 and 10 years of age. The children, presumably upon noticing Security's presence, then ran northeast and back towards the beach fence line. It appeared that the children somehow left the property through the east fence line back onto the beach. Industrial Security and shift RP tech began vehicle and foot patrols to search for the children and ensure that they did, in fact, leave the property. The children were not seen again. In accordance with 49 CFR 1580.105 - Reporting Significant Security Concerns, at 1520 on 8/15/16, the Radioactive Waste Group supervisor contacted the designated Transportation Safety Administration inspector to discuss the trespassing incident from 8/12/16. After the discussion with the inspector, the Radwaste supervisor was informed to contact the Department of Homeland Security Freedom Center and provide details of the incident. This contact was made at 1540 on 8/15/16. The notification is being made in accordance with 10 CFR 50.72(b)(2)(xi) notification to offsite government agency. The licensee has notified the NRC Resident Inspector.|
|ENS 52158||6 August 2016 18:29:00|
The following report was received from the State of Texas via email: On August 6, 2016, the Agency (Texas Department of State Health Services) was contacted by the licensee's radiation safety officer (RSO) who reported one of their crews had lost a Troxler model 3440 moisture density gauge. The gauge contains an 8 millicurie cesium-137 and a 40 millicurie americium-241 source. The crew had returned to their office in Dallas and when they went to get the gauge to place it in storage they realized the gauge had been left on the tailgate of the truck and was no longer there. The RSO was notified and both the crew and the RSO drove the 20 mile route used by the technicians returning to the office in an attempt to find the gauge, but the gauge was not located. The RSO stated the cesium source was located in the fully shielded position, but the operating rod was not locked. Local Law Enforcement in the area where the gauge was being used was notified of the lost gauge. The gauge does have a label on it identifying the licensee and listing the licensee's contact number. This number is to an answering service. Additional information will be provided as it is received in accordance with SA-300. Local law enforcement has been notified. Texas Report #: I-9423
The following was received from State of Texas via email: On August 10, 2016, the Agency (Texas Department of State Health Services) was notified by the licensee that it had recovered the missing gauge. The licensee's radiation safety officer (RSO) stated they found the gauge listed on Craig's List web site and contacted the seller. The RSO stated they paid the seller a reward and gained custody of the gauge at 1445 hours on August 10, 2016. The RSO stated the gauge was not damaged and the cesium source was still fully shielded. The RSO stated he was sending the gauge to a service company for inspection. Additional information will be provided in accordance with SA-300. Notified R4DO (O'Keefe), ILTAB (English), NMSS_Events Resource, CNSNS (MEXICO) 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 52157||5 August 2016 20:55:00|
At 1209 CDT on 8/5/16, during testing of Unit 2 Auxiliary Building Pre-action Sprinkler Systems, all zones on the Unit 2 Pyrotronics Fire Detection Panel went into an alarm state and were unable to be reset. This condition is reportable per 10 CFR 50.72(b)(3)(xiii) as a Major Loss of Assessment capability. The NRC Resident has been notified. The licensee has initiated all necessary compensatory and corrective actions.
At 1600 CDT on 8/6/16, the Unit 2 Pyrotronics Fire Detection Panel was declared functional following repair of master override reset test switch and supply fuse. The Pyrotronics Fire Detection Panel was successfully tested following maintenance. The emergency assessment capability for the site's Emergency Plan has been fully restored. The NRC Resident has been notified Notified the R2DO (Suggs).
|ENS 52153||4 August 2016 22:04:00|
At 1415 CDT on August 4, 2016, while performing a scheduled fire protection surveillance, it was discovered that a component within fire panel FZCP-7, BATTERY ROOM FIRE DETECTION had failed resulting in the inability of the installed fire detectors to detect a fire within the Division 1 and Division 2, 125 VDC battery rooms as well as the Division 2, 250 VDC battery room. This is being reported under 10 CFR 50.72(b)(3)(xiii) for a Loss Of Emergency Assessment Capability as the Control Room would not receive automatic notification of a fire in these areas for evaluation of HU2.1 and HA2.1 for fire within impacted battery rooms which are located within the Protected Area. There is no impact to the health and safety of the public. A 15 minute fire watch has been established for the affected fire zones. The NRC Resident Inspector has been notified.
Event Notification 52153 completed at 2204 EDT on 8/4/2016 shown above contains an error. The failure of FZCP-7, BATTERY ROOM FIRE DETECTION, resulted in the inability to detect a fire within the Division 1 and Division 2 125 VDC battery rooms as well as the Division 1 250 VDC battery room. The Division 2 250 VDC battery room was not affected by this issue. Additionally, the State of Minnesota was notified of this issue. The NRC Resident Inspector has been notified of this update. Notified R3DO (Skokowski)
|ENS 52089||14 July 2016 17:24:00||The following report was received from the Commonwealth of Virginia via facsimile: On July 12, 2016, Sims Metal Recycling, Richmond Virginia, reported to the Virginia Radioactive Material Program (VRMP) that a radioactive source had been found in a shipment of metal. They said the source was a gauge like the one found the previous day in their Petersburg Virginia Facility (see EN #52088). Refer to VA-16-008. VRMP Radiation Safety Specialists performed an onsite review. Labels indicated that the source was a general license device from EG&G Berthold, model not identified, serial number 001095, with a 100 millicurie (September 1990) cesium-137 source, serial number 3023. Maximum radiation levels at contact at the mounting near the shutter were 2 mR/hr, at three feet, maximum levels were less than 0.1 mR/hr. The gauge was placed in a drum in a secured area. Dose estimates by VRMP health physicists indicated no individual was likely to have received more than 1 mrem whole body or 5 mrem extremity dose. On-site tests for leakage indicated no removable contamination. The recycling facility will contact a waste broker to arrange for disposal. The gauge owner was not identified from review of the VRMP general license database. Berthold Technologies USA is reviewing records to identify the owner of the gauge. Virginia Event Report ID No: VA-16-009 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 52046||27 June 2016 12:21:00|
At 1050 CDT, (on 6/27/16), an Alert was declared at Dresden Unit 3. The Alert is due to Unit 3 experiencing a fire in the HPCI (High Pressure Coolant Injection) system, auxiliary oil pump motor. The fire is out. This notification is being made per 10 CFR 50.72(a)(1)(i). Dresden Unit 3 is stable and continues to operate at 100% power and HPCI has been declared inoperable. There is no impact on Dresden Unit 2. The licensee has notified the NRC Resident Inspector Notified DHS SWO, FEMA, USDA, HHS, DOE, DHS NICC, EPA, and FEMA National Watch Center, FDA EOC, NuclearSSA via email only.
Termination of MA-5 (Alert). Fire in HPCI room verified extinguished. HPCI system is inoperable. (Technical Specification) TS 3.5.1 condition G in effect, per 10 CFR 50.72(c)(i) - notification of termination of Alert. Dresden Unit 3 terminated the Alert at 1319 CDT, on 6/27/16. Dresden Unit 3 continues to operate at 100% power. The licensee will notify the NRC Resident Inspector. Notified R3DO (Kunowski), NRR (Miller), IRD (Grant), DHS SWO, FEMA, USDA, HHS, DOE, DHS NICC, EPA, and FEMA National Watch Center, FDA EOC, NuclearSSA via email only.
At 1042 (CDT) on 6/27/16, the U3 High Pressure Coolant Injection (HPCI) system was declared inoperable after the Auxiliary Oil Pump failed. This event is reportable per 10CFR50.72(b)(3)(v)(D); any event or condition that at the time of discovery could have prevented fulfillment of the safety function of structures or systems that are needed to mitigate the consequences of an accident. This is an eight hour reporting requirement. The Dresden NRC Resident (Inspector) has been notified. Notified R3DO (Kunowski).
|ENS 52125||26 July 2016 10:20:00||The following report was received from the State of New York via facsimile: Pace Analytical Services reported that a leak test of a Varian ECD (electron capture device), model 02- 001972-00, serial no. SP472 indicated removable contamination in excess of 0.005 microcuries. Test results measured 0.0132 millicuries of contamination associated with a 15 millicurie Nickle 63 source manufactured by Isotope Products Laboratory (IPL), model NBCB14746, serial# SP472. The licensee was advised to isolate the source and not to use it. (The licensee was) requested to arrange for retesting and provide results to the NYS DOH (New York State - Department of Health). Retesting was performed by (a Certified Health Physicist), Massachusetts License #13-3021, on July 14, 2016. The result of the second measurement was 0.0002 microcuries. Pace Analytical Services, Inc. stated that the ECO will be returned back to the manufacturer/ vendor and receipt will be sent to the NYS Department of Health. EVENT REPORT ID NO. NYDOH - NY-16-03|
|ENS 51974||3 June 2016 11:44:00||Description of the Event: On June 2, 2016 at 1245 (EDT), it was reported to the Environment, Health and Safety (EH&S) department that the two administrative verifications of ventilation clean-out containers were not performed. The Conversion Area nitric acid scrubber was scheduled for annual inspection and cleaning May 27 thru June 1. On May 18th, 25 disposable clean-out containers were obtained from the storeroom in preparation for the annual scrubber cleaning. The clean-out containers are used to collect special nuclear material (SNM) from scrubber and filter house cleanings. After obtaining new clean-out containers from the storeroom and prior to using them with SNM, one container from each lot is required to be measured by an operator to assure proper dimensions (IROFS STORAGE-GEN-126). An independent measurement by a process engineer to assure proper dimensions is also required prior to use with SNM (IROFS STORAGE-GEN-127). An additional 20 clean-out containers were obtained from the storeroom on May 31st. On June 2nd, a process engineer discovered that the measurement verifications were not performed. EH&S was notified of the event by phone and the 'Redbook' reporting system. (Redbook Issue #71225). At no time was there any actual or potential health and safety consequence to the workers, the public, or the environment. The safety function of these IROFS (Items Relied on for Safety) is to preclude using an incorrect container size. The ventilation clean-out containers are a standard stocked storeroom item. They have an approximate volume capacity of 1.5 gallons, providing a substantial safety margin to the minimum requirement of 5.7 gallons in the Criticality Safety Evaluation (CSE) which assumes an optimum uranium/water mixture and full 12-inch water reflection. The CSE also requires more than 64 close packed containers with an optimum uranium/water mixture; while the containers are limited to maximum array of 25 (IROFS STORAGE-GEN-112). Additionally, the clean-out containers remained spaced 18 inches apart at all times. Based on available IROFS, this accident sequence was unlikely, a failure probability of (10E-3), and not highly unlikely, a failure probability of (10E-4) or less. Therefore, this geometry accident sequence does not meet the performance requirements of 10CFR70.61. As stated above, the actual configuration remained safe at all times. Also, no external conditions affected the event. Immediate Corrective Actions: The clean-out container dimensions were verified as correct. This event has been entered into the facility Corrective Action Prevention And Learning system (CAPAL) #1003888517. The Licensee will notify NRC Region II.|
|ENS 51924||12 May 2016 20:01:00|
On May 12, 2016, at 16:47 EDT with the reactor at 100% power and the mode switch in RUN, an assessment of the Spent Fuel Pool racks containing neutron absorbing material concluded that some degradation had occurred. The result is that we cannot assure we are maintaining Keff < 0.95 as required per design. Conservative measures have been implemented to ensure public health and safety.
Recent planned testing conducted in the spent fuel pool determined that one rack panel had degradation of the neutron absorbing material in excess of what had been analyzed. An extent of condition review indicated additional potential at-risk locations may exist. Analyses are being performed to determine the potential impact and mitigating actions. Fuel pool conditions are safe and stable. Conservative measures have been implemented and spent fuel safety is maintained. Based on the above, the condition is reportable to the NRC as an Unanalyzed Condition under 10 CFR 50.72(b)(3)(ii), and requires an 8 hour notification. The condition is also reportable in accordance with 10 CFR 50.73 (a)(2)(ii). The licensee has notified the NRC Senior Resident Inspector. The licensee will notify the Commonwealth of Massachusetts.
|ENS 51886||27 April 2016 01:17:00||At 1736 (CDT) on 26 April 2016, a licensed operator performing a control room panel walkdown noted the green standby light for the HPCI (High Pressure Coolant Injection) Auxiliary Oil Pump (AOP) was not illuminated. The bulb was replaced and the replacement bulb did not illuminate. A non-licensed operator (NLO) was dispatched to the local 250VDC starter rack. The NLO discovered the green standby light on the 250VDC starter rack had failed. An attempt was made to start the AOP with the control switch. The pump did not start. The AOP is required to start in order to open the steam admission valves for the HPCI turbine. HPCI was declared inoperable at time 1754 (CDT) on 26 April 2016. Tech Spec LCO Conditions were entered and required actions completed. HPCI is a single train system. This report is submitted as a condition that at time of discovery could prevent the fulfillment of the safety function of an SSC (Structure, System, and Component) needed to mitigate the consequences of an accident. A similar condition was discovered on 25 April 2016 (see NRC Event #51882). Corrective maintenance was performed and HPCI was declared operable following satisfactory completion of post work testing of the AOP. Initial investigation indicates that the fault which occurred on 26 April is not the same as that which occurred 25 April. Investigation is on going. The licensee has notified the NRC Resident Inspector.|
|ENS 51878||22 April 2016 18:22:00||At 1359 CDT on April 22, 2016, Browns Ferry Units 1 & 2 experienced a partial loss of power during the transfer of Shutdown Bus 2 from the alternate power source back to the normal power source. During the transfer, the normal feeder breaker failed to close and provide power to the Shutdown Bus, resulting in an auto actuation of two Emergency Diesel Generators (EDGs). Power to Shutdown Bus 2 was immediately restored using the alternate feeder breaker. The EDGs did not tie to the boards. All systems responded as expected for the loss of power, and both Units 1 & 2 maintained 100% Rx Power. All systems have been restored to a normal lineup, and both Units 1 & 2 remain at 100% Rx Power. This event requires an 8 hour report per 10 CFR 50.72(b)(3)(iv)(A), 'Any event or condition that results in valid actuation of any of the systems listed in paragraph (b)(3)(iv)(B), (8) Emergency AC electrical power systems, including: Emergency diesel generators (EDGs).' The NRC resident inspector has been notified. The cause of the normal feeder breaker failure is being investigated. There was no impact on Unit 3.|
|ENS 51863||12 April 2016 14:36:00|
On April 12, at 1235 CDT, Duane Arnold Energy Center contacted officials with the State of Iowa (Bureau of Radiological Health and Department of Natural Resources) and Linn County Public Health Department in accordance with the nuclear industry voluntary reporting criteria contained in NEI 07-07 'Industry Ground Water Protection Initiative'. The site contacted the agencies as courtesy to notify them about the identification of low levels of tritium found within the site's protected area from a potential new source. Samples were taken, and no regulatory limits were exceeded. The site team is working with industry experts on pinpointing the cause and installing an extraction well to remediate the situation. This report is being made in accordance with 10 CFR 50.72(b)(2)(xi), as a result of notification to offsite agencies.
The Licensee has notified the NRC Resident Inspectors. There is no risk to plant employees, the public or drinking water.
|ENS 51860||9 April 2016 23:14:00||During pre-planned surveillance testing of the Supplementary Leak Collection and Release System (SLCRS), an issue was found affecting the Secondary Containment boundary. Millstone Unit 3 is being moved to Mode 5 for a refuel outage where investigation and repairs will be made. This is reportable under 10CFR50.72(b)(3)(v)(C), a condition that could have prevented the fulfillment of a safety function for systems or structures to control the release of radioactive material, and 10CFR50.72(b)(3)(v)(D) to mitigate the consequences of an accident. The surveillance testing was being performed while Millstone 3 was being removed from service for an upcoming Refueling Outage. The licensee has notified the NRC Resident Inspector and State and Local authorities.|
|ENS 51859||8 April 2016 20:50:00||At 1657 EDT on April 8, 2016, an oil leak developed from the station's switchyard transformer no. 4. Approximately 25,000 gallons of oil has leaked within the transformer's containment berm. At 1820 EDT on April 8, 2016, DC Cook environmental personnel determined that approximately 2000 gallons of oil had leaked outside of the containment berm onto the ground. None of the oil has made it to any nearby drains. Leak has stopped and cleanup is ongoing. D.C. Cook has notified the State of Michigan and local authorities. The NRC Resident Inspector was notified. This notification is being made in accordance with 10 CFR 50.72(b)(2)(xi) due to notification of offsite agencies.|
|ENS 51857||8 April 2016 13:10:00||The following report was received from the State of California via email: On April 7, 2016, (the licensee gauge operator) for Geo-Advantec, Inc., contacted the California Office of Emergency Services to report a moisture density gauge, a CPN Model MC3 Elite (10 mCi Cs-137, 50 mCi Am:Be-241, S/N30625), was run over by an unknown type of vehicle. The incident occurred at 2003 Speyer Lane, Redondo Beach, CA. After taking the report, the report was forwarded to the Los Angeles County Radiation Management which was then forwarded to Brea office of Radiologic Health Branch (RHB). (The licensee gauge operator) was contacted by a RHB inspector and stated that the gauge had been run over and that the source rod had been broken. He was also asked if the Cs-137 source had been exposed, and he stated that he was taking a standard count and that the source was in the shielded position at the time of the incident. (The licensee gauge operator) was then asked to take pictures of the gauge so that it could be evaluated to determine if the sources were still intact and to determine the extent of the damage to the gauge. The pictures indicated that the source rod was broken approximately 1-2 inches below the bottom of the guide tube, which had broken off the source shield, and that the rest of the source rod appeared to be inside the opening on the top part of the source shield. The Am:Be-241 source appeared to be intact and still attached to the body of gauge. (The licensee gauge operator) was instructed to place the body of the gauge into the transport case in its normal position and then place the rest of the pieces of the gauge into the case. He was also instructed to lock the case and secure it into the transport vehicle as normal. He was then instructed to check ground below the gauge to verify that the source was not exposed during the incident. After (the licensee gauge operator) verified that there were no signs of the source having been extended during the incident, he was instructed to mark the area where the gauge was damaged to ensure that area could be found if a survey was needed and then he was allowed to transport the gauge to Maurer Technical Services (MTS) to be inspected and surveyed to ensure the sources were still in the gauge and to verify they were intact. When the gauge arrived at Maurer Technical Services, a survey was performed by the MTS RSO (Radiation Safety Officer) (meter type not reported). The gauge had contact reading of 8.5 mR/hr, which was consistent with a CPN MC3 moisture/density gauge with both sources in their shielded positon. The RSO also took swipes of the device, the hands of the operator and the trunk of the operator's car. An RHB inspector arrived at MTS to perform a count of the swipes with a Ludlum general purpose meter and a 44-9 Geiger-Mueller probe and found all of the swipes were at background, indicating that there was not leakage and the sources were intact. The gauge will be stored at MTS until disposal of the gauge can be arranged. The investigation is on-going and any citations will be determined at a later date. CA 5010 # 040716.|
|ENS 51802||18 March 2016 09:12:00||The following report was received from the State of Ohio via email: At approximately 1315 (EDT) on 3/16/16, ODH (Ohio Department of Health) was notified that a licensee had a nuclear density gauge run over on a construction site. The incident occurred at about 0730 that morning. The licensee's technician instructed site personnel to stay 20 feet from the damaged gauge and contacted the RSO (Radiation Safety Officer) via telephone. The RSO arrived on site at approximately 0810 and determined that both sources were intact within the gauge housing. Readings attained around the damaged gauge with a radiation monitor and indicated normal levels. The gauge was placed back in the case and additional readings were (obtained) from the soil at the accident location with negative results. The gauge was then transported to a licensed service provider for inspection and storage. The service provider inspected the gauge and performed a leak test which indicated that the sources were not leaking. Investigation is ongoing. Sealed Source Gauge. Manufacturer: QSA GLOBAL, Model Number: X.2084, Serial Number: A-7879, 0.050 Ci Am/Be-241 source and Model Number: X.8, Serial Number: C-7879, 0.010 Ci Cs-137 source. Ohio: Item Number: OH160001|
|ENS 51782||9 March 2016 15:10:00||The licensee reported that a Siemens C-200 Gauge, model L 640 containing a Ra-226 0.005 Ci source, was run over by a pick up truck and damaged. Damage to the gauge was limited to the electronics package and battery. The source remains properly contained and un-damaged. A survey of the area indicates no leakage and no exposures to personnel. The gauge has been placed in a secure location and will be sent back to the manufacturer for repairs.|
|ENS 51781||9 March 2016 14:11:00||The following report was received from the Commonwealth of Pennsylvania via email and facsimile: Event Type: Loss of licensed material in a quantity greater than or equal to 1000 times the Appendix C quantities in part 20. Notifications: NRG Energy discovered the event on March 7, 2016, (at their Seward, PA location,) and submitted a report to the Department (Pennsylvania Department of Environmental Protection) on March 9, 2016. This event is reportable as per 10 CFR 20.2201(a)(1)(i). Event Description: On Monday March 7, 2016, while conducting the six month inventory check of the radioactive sources at the Seward Power Plant, one tritium exit sign was found missing. The exit sign was installed above a door located in Seward's Fuel Barn. The last inventory check was conducted September 11, 2015, and the exit sign was present at that time. The exit sign was manufactured by EMERG-LITE and was an Everlite series sign. The sign contained between 9.5 - 11.5 Ci of tritium gas at the time of manufacture and was to be replaced before February 2023. The sign was last known to be in good condition and not damaged. No cause for the missing sign has been identified and no exposures have been recorded at this time. Cause of the Event: Unknown at this time. The plant is currently searching the site and conducting interviews with personnel. Actions: The Department will be following up with the facility for any additional information. The plant is also conducting refresher radiation training to plant personnel. More information will be provided upon receipt. Pennsylvania Event Report ID No: PA160009. 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|