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ENS 41074 + (Crystal River Unit 3 site has declared an … Crystal River Unit 3 site has declared an Unusual Event due to the declaration of a hurricane warning for Citrus County including the Crystal River Unit 3 site.Current wind speed on-site is 20 mph with 40 mph gusts. The licensee's current plan is to remain in power operation.The licensee notified both state and local agencies and the NRC Resident Inspector.* * * UPDATE FROM AL STERN TO MIKE RIPLEY 0349 ET 09/27/04 * * *Hurricane warning has been discontinued. Unusual Event exited at 0345 ET. The licensee notified the NRC Resident Inspector.Notified R2 Response Manager (S. Cahill), R2 DO (R. Haag), NRR EO (S. Richards), IRD Manager (S. Frant), FEMA (J. Canupp), DHS (Sr Watch Officer) FEMA (J. Canupp), DHS (Sr Watch Officer))
ENS 40299 + (Crystal River Unit 3 was at 37 percent pow … Crystal River Unit 3 was at 37 percent power starting up from a refueling outage. During troubleshooting of control problems with the B Main Feedwater Pump, a feedwater transient occurred which underfed the steam generators. This resulted in a reactor trip on high Reactor Coolant System pressure and Emergency Feedwater Actuation on low steam generator levels. This event is reportable as a 4-Hour Non-Emergency Notification per 10CFR50.72 (b)(2)(iv)(B) for Reactor Protection System Actuation and as an 8-Hour Non Emergency Notification per 10CFR50.72 (b)(3)(iv)(A) for Emergency Feedwater Actuation.The licensee reported that all control rods fully inserted on the trip and that steam generator safety valves lifted and reseated as expected. The primary system is currently at 2155 psi, 549 degrees F with steam generator feedwater being supplied by auxiliary feedwater. The main condenser is available and is being used for primary system heat removal via steam dump. The station electrical grid is stable and in normal configuration; the emergency diesel generators are operable and in standby.The licensee has notified the NRC Resident Inspector.e has notified the NRC Resident Inspector.)
ENS 41022 + (Crystal River declared a Notification of U … Crystal River declared a Notification of Unusual Event due to hurricane warning in effect for the plant and surrounding areas. All safety systems are operable.The licensee notified the NRC Resident Inspector and the State emergency response organization.* * * UPDATE AT 1027 ON 9/7/04 R TYRIE TO W GOTT * * *The licensee terminated the Notification of Unusual Event at 1017 today, 9/7/04. The basis of the termination is that the hurricane warning was lifted yesterday 9/6/04 and offsite power has been restored. Notified R2 (Pribish), NRR (Reis), IRD (Wessman), DHS (Akers), FEMA (Kuzia). IRD (Wessman), DHS (Akers), FEMA (Kuzia).)
ENS 49324 + (Current design basis calculations indicate … Current design basis calculations indicate the Low Pressure Safety Injection (LPSI) pumps could potentially operate in a run-out condition under certain worst case design basis conditions. The LPSI pumps could operate in a run-out condition beyond the analyzed time by 20 minutes. Current design basis calculation assumes LPSI Pump would be shutdown by (the) RAS (Recirculation Actuation Signal) in less than one hour, however due to past changes to Containment Spray Pump Start Logic, the time was lengthened to 80 minutes which is beyond the one hour analyzed. This represents a reportable unanalyzed condition.The licensee notified the NRC Resident Inspector.* * * RETRACTION FROM LUKE JENSEN TO HOWIE CROUCH AT 1722 EDT ON 10/31/13 * * * Fort Calhoun completed additional analysis which verified that the LPSI pumps will not go into run-out as previously reported. Therefore Fort Calhoun is withdrawing the event notification.The licensee will notify the NRC Resident Inspector. Notified R4DO (Drake).esident Inspector. Notified R4DO (Drake).)
ENS 48730 + (Current design basis calculations indicate … Current design basis calculations indicate the high pressure safety injection (HPSI) pumps could potentially operate in a run-out condition under certain worst case design basis conditions. The calculated flow is beyond the pump curves. The HPSI pumps could operate in a run-out condition for an extended period following a design basis accident. The pump vendor indicates that long term operation in this condition could not be supported due to accelerated wear of pump internal components. This represents a reportable unanalyzed condition.The NRC Resident Inspector has been notified. NRC Resident Inspector has been notified.)
ENS 44507 + (D.C. Cook Unit 1 declared an Unusual Event … D.C. Cook Unit 1 declared an Unusual Event (EAL H4 & H5) due to a fire in the Main Turbine. The reactor was manually tripped from 100 percent. The fire occurred at the upper level of the turbine building (Level 633), and was extinguished by the fire suppression system and local fire brigade. Three fire pumps are running at this time. No injuries were reported.All rods fully inserted, auxiliary feed water initiated and decay heat is being removed via atmospheric relief valves. Unit 1 is currently shutdown and stable in Mode 3, Hot Standby. Main steam stop valves are closed. Main condenser vacuum was broke. Unit 2 was not affected. All Unit 1 safety-related equipment is in-service and available. The licensee is currently assessing the extent of damage.The licensee will inform the NRC Resident Inspector.Notified DOE (Morrone), USDA (Shaf) and HHS (Mammarelli) that the NRC entered Monitoring Mode at 2045 EDT.* * * UPDATE AT 0005 EDT ON 09/21/08 FROM BRADDOCK LEWIS TO V. KLCO * * * On September 20, 2008 at 20:05 the DC Cook Unit 1 Reactor was manually tripped after a malfunction occurred on the main turbine generator causing high vibration. A fire in the Unit 1 Main Generator resulted from this malfunction. A Notification of Unusual Event was declared September 20, 2008 at 20:18 due to Event classifications H-4, Fire within the protected area not extinguished within 15 minutes and H-5, Toxic or Flammable gas release affecting plant operation. The Unit 1 Main Generator Fire was reported extinguished at September 20, 2008 at 20:28.The Unit 1 plant trip was uncomplicated and all Automatic Control systems functioned as expected. All control rods inserted on the Reactor trip. The Turbine and both Motor Driven Auxiliary Feedwater pumps automatically started and fed all four Steam Generators as designed. The Steam Generator Stop Valves were manually closed to arrest plant cooldown. The cause of the Main Generator fire has not yet been determined, but the investigation is ongoing. No radiological release resulted from this event.This event is being reported as a four hour report required by 10CFR50.72(b)(2)(iv)(B) due to the Reactor Protection System automatic actuation and as an eight hour report required by 10CFR50.72(b)(3)(iv)(A) for the automatic actuation of the Auxiliary Feedwater system. The Notification of Unusual Event was reported separately.Unit 1 is stable in Mode 3. Shutdown Margin was satisfactorily verified. The main condenser was isolated as the primary heat sink. Steam Generator Power Operated Relief Valves are removing core decay heat in automatic control due to breaking main condenser vacuum. Main condenser vacuum was broken to stop the Unit 1 main turbine generator due to high vibration. Preparations are in progress to cooldown Unit 1 to Mode 5, Cold Shutdown.The Fire Suppression Water System was actuated and one of two 565,000 gallon tanks was drained. The second 565,000 gallon Fire Suppression Water tank was placed in service to restore the Fire Suppression Water system function.The licensee notified the NRC Resident Inspector, local and state authorities. The licensee will likely make a press release.Notified R3RA (Caldwell), R3DO (Stone), NRR (Leeds and Galloway) IRD (McDermott), DHS (Gomez) and FEMA (Kuzia).* * * UPDATE PROVIDED BY PAUL LEONARD TO JASON KOZAL AT 0414 ON 09/21/08 * * *At 0409 the licensee terminated from the Notice of Unusual Event. The licensee has established the forced outage recovery team. No fires exist and no conditions conducive to fires exist due to the event. The licensee has established the integrity of the fire protection system.Notified R3RA (Caldwell), R3DO (Stone), NRR (Galloway), IRD (Grant), DHS (Jason), DOE (Maroni), FEMA (Sweetser), USDA (Phillip), and HHS (Nathan). (Sweetser), USDA (Phillip), and HHS (Nathan).)
ENS 42536 + (D.C. Cook Unit 2 was in Mode 4 and heating … D.C. Cook Unit 2 was in Mode 4 and heating up after Unit 2 Cycle 16 refueling outage when 2-IMO-340 (East RHR pump to Charging Pump Suction Header) was throttled open. RHR suction was aligned to Loop 2 Hot Leg with wide range RCS pressure at 337 psig and RCS average temperature 280 degrees Fahrenheit. Charging header safety valve 2-SV-56 lifted with a setting of 220 psig and relieved to the Pressurizer Relief Tank (PRT) between 10:56:00 and 10:58:30. 2-IMO-340 was then closed. Approximately, 120 gallons of reactor coolant were directed to the PRT during the approximate 2.5 minutes, resulting in a flow rate of approximately 48 gpm, greater than the 25 gallon per minute identified leak rate limit for an Unusual Event (10 CFR 50.72(a)(1)(i)). Time of discovery for leakage quantity and reportability was 13:21 on 4/29/06.Therefore, for approximately 2.5 minutes, D.C. Cook Unit 2 met the conditions for an Unusual Event. However, this fact was discovered after leakage was stopped. Unit 2 is stable in Mode 4 at 340 degrees Fahrenheit and 405 psig RCS pressure.An Unusual Event was NOT declared, but the conditions for an Unusual Event were met under Emergency Condition Criteria S-8 between 10:56:00 and 10:58:30 on 4/29/06. This ENS notification is being made within one hour of the discovery of the undeclared event.The licensee will notify the NRC Resident Inspector.* * * RETRACTION ON 4/30/06 AT 1419 EDT FROM D. TURINETTI TO J. MACKINNON * * *EN# 42536 reported that D.C. Cook Unit 2 satisfied the conditions for entry into an Unusual Event under Emergency Condition Criteria S-B, identified RCS leakage exceeding 25 GPM. This was identified after system alignment had been restored, ending the event, and was reported as an after the fact declaration that the conditions for an Unusual Event had been satisfied, but not declared. Subsequent review of this event has determined that the conditions for Emergency Condition Criteria S-B were not satisfied in that no identified RCS leakage occurred during this event. The diversion of 120 gallons of CVCS (Chemical Volume and Control System) inventory was the result of plant alignment by Operations personnel which was immediately recognized and terminated. Plant systems and components functioned as designed to terminate this event.The licensee will notify the NRC Resident Inspector. Notified R3DO (H. Peterson).t Inspector. Notified R3DO (H. Peterson).)
ENS 40716 + (D.C. Cook Unit 2 was performing post-maint … D.C. Cook Unit 2 was performing post-maintenance testing on a solenoid valve associated with a letdown orifice valve and realigning letdown flowpath when safety valve 2-SV-51 (Regenerative Heat Exchanger Letdown Outlet Safety Valve) lifted on 4/29/04 at 15:57 EDT, relieving 323 gallons of water to the Pressurizer Relief Tank over a five (5) minute period before letdown was isolated. Average leak rate was approximately 65 gallons per minute, greater than the 25 gallon per minute identified leak rate limit for an Unusual Event (10CFR50.72(a)(1)(i)). This was a rapidly concluded event of a five (5) minute duration. Time of discovery for leakage quantity and reportability was 17:10 on 4/29/04.Therefore, for a five (5) minute period, D.C. Cook Unit 2 met the conditions for an Unusual Event. However, this fact was discovered after leakage was stopped. Unit 2 is stable at approximately 100% reactor thermal power with the excess letdown system in service.An Unusual Event was NOT declared, but the conditions for an Unusual Event were met between 15:57 and 16:02 on 4/29/04. This ENS notification is being made within one hour of the discovery of the undeclared event.The licensee notified the NRC Resident Inspector.ensee notified the NRC Resident Inspector.)
ENS 48063 + (D1 and D2 Diesel Generators (DG) were decl … D1 and D2 Diesel Generators (DG) were declared inoperable at 1551 CDT due to exceeding the maximum outside air temperature limit of 97 degree F. This limit was based on heatup analysis for a HELB (High Energy Line Break) in the turbine building. The temperature is measured using the 15 min average of air temperature at the 10 meter meteorological tower, which approximates the intake height for room ventilation.LCO 3.8.1 Condition B was entered for both DGs and LCO 3.8.1 Condition E was entered for two DGs inoperable on the same unit. The DGs could start and run if required at the time of entry.This condition is reportable per 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function.During review of the analysis, two pressure switches on D2 and one on D1 were identified as being limiting and a modification was approved to replace them. Upon replacement, the limiting outside temperature would increase to 100.5 degree F. Replacement of components had been completed for D2 diesel generator prior to LCO entry; however, the temperature limit of 100.5 degree F had not yet been approved.Shortly after D1 was declared inoperable, it was isolated to complete the component replacement which is expected to take less than four hours. D1 will return to operable once the component replacement and PMT is completed.At 1630 (CDT) D2 was declared operable based on approval of a revised operability recommendation that raised the outside temperature limit to 100.5 degree F.Outside air temperature peaked at 97.1 degree F and is slowly decreasing.The licensee has notified the NRC Resident Inspector.* * * RETRACTION FROM GENE DAMMANN TO DONALD NORWOOD AT 1555 EDT ON 8/28/12 * * *Event Notification 48063 reported that D1 and D2 Diesel Generators (DG) were inoperable due to exceeding the maximum outside air temperature limit of 97 degree F.An additional approved evaluation determined that the D2 lube oil pressure switches were replaced prior to exceeding the outside ambient air temperature limit. The new lube oil pressure switches have a higher temperature rating that supports a higher outside ambient air temperature, therefore, D2 was able to fulfill its safety function and Event Notification 48063 is retracted.The NRC Resident Inspector has been informed.Notified R3DO (Cameron).as been informed. Notified R3DO (Cameron).)
ENS 50170 + (D1 and D2 Emergency Diesel Generators (DG) … D1 and D2 Emergency Diesel Generators (DG) were declared inoperable at 1902 CDT due to the station identifying that the Turbine Building HELB (High Energy Line Break) heatup analysis temperature of 270 F exceeded the maximum temperature limit of 200 F for both diesel generators supply and exhaust fan blade positioners.LCO 3.8.1 Condition B was entered for both DGs and LCO 3.8.1 Condition E was entered for two DGs inoperable on the same unit. The DGs could start and run if required at the time of entry.This condition is reportable per 10 CFR 50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function.Unit 1 continued to operate in Mode 1 at 100% power while bypasses were being put in-service to fail the diesel generators supply and exhaust fan blade positioners in the full cooling mode to allow the LCO conditions to be exited.At 2002 CDT the bypasses were installed on the D1 Diesel Generator Supply and Exhaust Fan Blade Positioners therefore LCO 3.8.1 Condition E was exited for both D1 and D2 Emergency Diesel Generators being inoperable. In addition LCO 3.8.1 Condition B was exited for D1 Diesel Generator due to it being declared Operable.At 2056 CDT the bypasses were installed on the D2 Diesel Generator Supply and Exhaust Fan Blade Positioners therefore LCO 3.8.1 Condition B was exited for D2 Diesel Generator due to it being declared Operable.At no time was the health and safety of the public threatened.The Senior NRC Resident has been notified.The Senior NRC Resident has been notified.)
ENS 48122 + (D5 and D6 Diesel Generators (DG) were decl … D5 and D6 Diesel Generators (DG) were declared inoperable at 1400 CDT due to a missing internal flood control barrier seal. The missing flood control barrier seal is required to be installed with Unit 2 in Modes 1, 2, 3, and 4 when the MSIV's are open.LCO 3.8.1 Condition B was entered for both DGs and LCO 3.8.1 Condition E was entered for two DGs inoperable on the same unit. The DGs could start and run if required at the time of entry.This condition is reportable per 10CFR50.72(b)(3)(v) as an event or condition that could have prevented the fulfillment of a safety function.At 1505 CDT the missing internal flood barrier seal was verified in place restoring the safety function. LCO 3.8.1 Condition B and E were exited.The NRC Resident Inspector has been informed.At this time, it is not known how long the seals had been missing. An investigation is on-going.en missing. An investigation is on-going.)
ENS 42479 + (DAEC (Duane Arnold Energy Center) abnorma … DAEC (Duane Arnold Energy Center) abnormal operating procedure AOP 301.1 for Station Blackout specifies that 30 minutes are allowed to establish alternate ventilation for RCIC (Reactor Core Isolation Cooling) /HPCI (High Pressure Coolant Injection) rooms, switchgear rooms, battery rooms, and the Main control room. During validation demonstration conducted on April 5, 2006 for the NRC Components team (from NRC Region 3 Office) the 30 minutes requirement was not met with the control room alternate ventilation taking about 60 minutes and with the other areas also exceeding their time requirements. This event is reportable as an unanalyzed condition that significantly degraded plant safety pursuant to 10 CFR 50.72(b)(3)(ii) reportability notification.The NRC Resident Inspector was notified of this event by the licensee.* * * RETRACTION FROM MURRELL TO HUFFMAN AT 1442 EDT ON 6/01/06 * * *Duane Arnold Energy Center is retracting event number 42479 which was reported to the NRC Operations Center on 4/5/06 at 1855. This event is now determined to be not reportable because further evaluation to assess the significance of the delays in establishing alternate control room ventilation determined that the delay did not result in an adverse temperature increase in the affected areas. Specifically, the control room and back panel area temperature rise was evaluated using a detailed mass and heat transfer model of the affected areas. The evaluation confirmed that delays in establishing alternate control room ventilation did not adversely impact station commitments to 10 CFR 50.63 or accident responses outlined in UFSAR Section 15.3.2. Other areas required to have alternate ventilation established within 30 minutes had previously been successfully validated. Therefore, this event is not reportable as an unanalyzed condition that significantly degraded plant safety. The analyses and the bases for this retraction can be found under the plant corrective action program.The licensee notified the NRC Resident Inspector. R3DO (Kozak) notified.esident Inspector. R3DO (Kozak) notified.)
05000331/FIN-2008004-03 + (DAEC TS 5.4.1 provides, in part, that writ … DAEC TS 5.4.1 provides, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33 Revision 2, Appendix A, February 1978. Section 2 of Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), Revision 2, February 1978, provides, in part, that the licensee establish written procedures for preparation for refueling, refueling equipment operation, and core alterations. Section 3.6.3 of DAECs ACP 1407-2, Material Control in the Spent Fuel Pool and Cask Pool, Revision 15, a procedure that implements TS 5.4.1 and Regulatory Guide 1.33, provides in part, that health physics must be made aware of what and where any material is being relocated in the spent fuel pool, cask pool or cavity water prior to movement. Contrary to the above, on February 7, 2007, a contract employee relocated reactor cavity lights from the spent fuel pool to the reactor cavity without notifying health physics. The NRC reviewed the incident, interviewed applicable staff, and concluded that the violation was not willful. This inspection activity closed URI 2007002-06. Because the violation is of very low safety significance, it meets the criteria in Section VI.A.1 of the NRC Enforcement Policy, and it has been entered into the Corrective Action Program (CAP 047066), it is being treated as an NCV.AP 047066), it is being treated as an NCV.)
05000390/FIN-2018010-02 + (DCN 66459 added relays and wiring to chang … DCN 66459 added relays and wiring to change the actuation system that initiated the Auxiliary Building Isolation (ABI) and Containment Ventilation Isolation (CVI) protective functions. The new control elements (relay contacts and wiring) bypassed the Unit Solid State Protection System (SSPS) circuitry. The intent was to actuate the CVI function on an ABI actuation from the opposite unit while it was at full power operation and the other unit was in refueling mode. By bypassing the refueling units protection system and controlling the components in the refueling unit, the modification in effect made the protection systems (CVI) a shared system. The CVI was classified as part of the Engineered Safeguards Protection Systems (ESFAS). The ESFAS was not listed in the UFSAR Chapter 3.1.2 WBNP Conformance with GDCs (General Design Criteria), as shared systems under GDC 5. The UFSAR compliance with GDC 5 Sharing of Structures, Systems, and Components, specified all shared systems are sized for all credible initial combinations of normal and accident states for the two units, with appropriate isolation to prevent an accident condition in one unit from carrying into the other. The new control elements integrated in the ESFAS logic, apparently on both units. The licensee did not perform a failure modes and effects analysis to determine the negative effects that could degrade the ESFAS isolation functions when they are required to operate. The inspectors are concerned that the integration of the two ESFAS circuitry could have a detrimental effect. Additional failure modes appear to have been introduced into these systems. The inspectors need to determine the extent to which each units protection system and CVI were exposed to additional failures including common cause failures to determine whether there could be more than a minorissue and a potential failure to perform an adequate 50.59 evaluation in accordance with NPG-SPP-09.3 Plant Mods and Engineering Change Control, Section III, was a performance deficiency. This URI, is being opened to determine whether the PD is more than minor. This modification was complete on June 16, 2017. This issue was captured in CR 1398935, Potential violation of 10 CFR 50.59(d)(1) via DCN 66459.ation of 10 CFR 50.59(d)(1) via DCN 66459.)
ENS 39714 + (DEGRADED CONDITION DUE TO LEAKAGE AT SMALL … DEGRADED CONDITION DUE TO LEAKAGE AT SMALL BORE INCONEL 600 PENETRATION The following information was received from the licensee via facsimile "The following event description is based on information currently available. If through subsequent reviews of this event, additional information is identified that is pertinent to this event or alters the information being provided at this time, a follow-up notification will be made via the ENS or under the reporting requirements of 10CFR50.73 "On March 29, 2003, at approximately 04:00 Mountain Standard Time (MST) engineering personnel performing preplanned visual examinations of reactor coolant System (RCS) piping in accordance with procedure requirements discovered boric acid residue on three RCS small bore Inconel 600 penetrations. One location was the RCS hot leg in-service thermowell 3JRCBTW0112HB. The visual observation was characterized as a small white trail of boron residue running down the hot leg approximately 2 inches. There were no signs of dripping, spraying, puddles of liquid, or liquid running down the nozzle or hot leg. The residue appeared dry. The other two were on pressurizer heater sleeves (nozzles) A-4 and A-18. The visual observation at these two locations was characterized as a small white buildup of boron residue around the heater sleeve as the sleeve enters the pressurizer bottom head. There does not appear to be residue running down the outside of the sleeves. There were no signs of dripping, spraying, puddles of liquid, or liquid running down the nozzle or (pressurizer). The residue appeared dry "Technical Specifications (TS) Limiting Condition for Operation (LCO) 3.4.14 permits no reactor coolant system (RCS) pressure boundary leakage and therefore, the discovery of leakage (presumed boric acid residue) from the hot leg thermowell and the pressurizer heater sleeves was a degradation of a principal safety barrier. Therefore, the ENS notification of this event is in accordance with 10CFR50.72(b)(3)(ii). The control room personnel entered LCO 3.4.14 Condition B and are continuing to place the plant in Mode 5. The RCS was being cooled down in preparation for Unit 3's tenth refueling outage. At the time of discovery, the RCS was at approximately 520 degrees F and 2218 psia "An investigation of this event will be conducted in accordance with the PVNGS (Palo Verde Nuclear Generating Station) corrective action program. The cracking of Alloy 600 components both at Palo Verde and industry-wide has been attributed to axially oriented, primary water stress corrosion cracking (PWSCC). PWSCC is not considered a significant threat to the structural integrity of the RCS boundary, thermowell, or heater sleeves as this type of cracking typically results only in small leaks "The bases for this conclusion is that if PWSCC occurred at Palo Verde, the cracks would be predominately axial in orientation. As in this case, the cracks would result in visibly detectable leakage that would be apparent during visual examinations, performed as part of walkdown inspections, before significant damage to the reactor coolant boundary occurred "Palo Verde has a program for replacing the Alloy 600 hot leg thermowells. This thermowell was scheduled for replacement during this refueling outage. The plans to replace the thermowell remain unchanged "A mechanical nozzle sleeve assembly (MNSA) will be installed on each of the pressurizer heater sleeves "No ESF actuations occurred and none were required. There were no structures, systems, or components that were inoperable at the time of discovery that contributed to this condition. There were no failures that rendered a train of a safety system inoperable and no failures of components with multiple functions were involved. The event did not result in the release of radioactivity to the environment and did not adversely affect the safe operation of the plant or health and safety of the public. The licensee has informed the NRC Resident Inspector.nt or health and safety of the public. The licensee has informed the NRC Resident Inspector.)
ENS 45066 + (DESCRIPTION OF DEFICIENCY Configuration co … DESCRIPTION OF DEFICIENCYConfiguration control was not maintained and physical equipment issues were not documented under a Quality Assurance Plan for the period of time from in which Construction Permits CPPR-122 and CPPR-123 were withdrawn until they were reinstated.SAFETY IMPLICATIONSThere are no safety implications because controls are in place under TVA's corrective action program that will prohibit current plant structures, systems or components, including those affected in the course of resource recovery activities, from being placed into service without being further evaluated and having been fully restored or replaced.This deficiency has been entered into TVA's Corrective Action Program as Problem Evaluation Report 170988.ogram as Problem Evaluation Report 170988.)
ENS 41171 + (DESCRIPTION: 11/2/04 at 1735 the Plant Shi … DESCRIPTION:11/2/04 at 1735 the Plant Shift Superintendent (PSS) office was notified of a tanker truck located outside the east corner of the X-700 has uranium results at 322.0 +/- 64.4 PPM and 4.87 +/- 0.49 U-235 assay. No NCSA applies to this operation. This is reportable per NRCB 91-01 as a 24 hour event.SAFETY SIGNIFICANCE:Very Low Safety Significance. The amount and concentration of uranium involved cannot possibly achieve a critical configuration.POTENTIAL CRITICALITY:There are no criticality pathways involved due to the limited mass and concentration of material.CONTROLLED PARAMETERS:No NCSA was established for this operation, so there are no control parameters. However, the limited mass and concentration of the solution does not warrant double contingency control because a criticality is not credible.NUCLEAR CRITICALITY SAFETY CONTROLS:No NCSA controls were established for this operation.CORRECTIVE ACTIONS:Upon discovery the PSS office directed entering an anomalous condition. Additional samples are being taken to confirm results. Contents of tanker are to be removed.lts. Contents of tanker are to be removed.)
ENS 40133 + (DHP-VR350, 1200A, circuit breakers were ma … DHP-VR350, 1200A, circuit breakers were manufactured by Cutler-Hammer Inc., commercially dedicated for Class 1E applications by Westinghouse and shipped to the Columbia Generating Station. During installation acceptance testing, one breaker charging motor failed to charge. Columbia Generating personal performed, a 100% inspection on all 16 breakers with safety-related functions. There was a loose screw associated with this position switch found in seven of these breakers.Westinghouse will be issuing a Nuclear Safety Advisory Letter (NSAL) to all PWR and BWR plants informing them of this issue so they can determine applicability to their plant.an determine applicability to their plant.)
ENS 52692 + (Dale E. Porter GE-Hitachi Nuclear Energy S … Dale E. PorterGE-Hitachi Nuclear EnergySafety Evaluation Program Manager3901 Castle Hayne Rd., Wilmington, NC 28401(910) 819-4491Dale.Porter@GE.ComThe inappropriate addition of chlorinated water from container box desiccants into the CRDMs (Control Rod Drive Mechanisms) during leak testing after rebuild could potentially initiate Intergranular Stress Corrosion Cracking (IGSCC) or Transgranular Stress Corrosion Cracking (TGSCC). These two types of SCC could cause a separation of the stop piston or separation of the index tube contained within the CRDM. The stop piston separation could cause a slower scram speed and damage the drive so it could not be withdrawn. An index tube separation could result in a similar type of rod uncoupling event that would have the potential to result in a rod drop accident (RDA). The piston tube located within the CRDM is a reactor coolant pressure boundary (RCPB) and is an ASME component. There is a possibility of cracking causing a RCPB leak. SCC initiation on the Cylinder Tube and Flange (CTF) area of the CRDM could result in a separation that could prevent a scram or normal insertion of a CRDM.Reports have been issued to River Bend, LaSalle Unit 2, and Hatch Unit 2 providing the results of an evaluation that concludes that the condition will not create a substantial safety hazard or potentially cause a Technical Specification Safety Limit violation for a minimum of one operating cycle. The Browns Ferry Unit 2 drives were shipped but were not installed prior to recall, thus a short-term evaluation for Browns Ferry has not been completed.River Bend, Entergy, Shipped Date: 2017, Quantity Shipped: 15, Customer PO Number: 10478763LaSalle Unit 2, Exelon, Shipped Date: 2017, Quantity Shipped: 24, Customer PO Number: 00414787-66Hatch Unit 2, Southern Nuclear, Shipped Date: 2017, Quantity Shipped: 15, Customer PO Numbers: SNG50295-0001 & SNG50295-0002Browns Ferry Unit 2, TVA, Shipped Date: 2017, Quantity Shipped: 32, Customer PO Number: 2424171* * * UPDATE ON 7/12/17 AT 1035 EDT FROM LISA SCHICHLEIN TO BETHANY CECERE * * *Pursuant to 10 CFR 21.21(d)(4), GEH is providing the final report with the conclusion that, in limited cases, the chloride contamination could create a substantial safety hazard. Attachment 1 identifies the potentially impacted plants and Attachment 2 contains the final report information. The enclosure provides additional details of the evaluation.Updated Attachment 1 notes:A portion of the CRDMs at the Hatch Plant are Reportable while the remainder are not.Summary of updates to Attachment 2:The inappropriate addition of chlorinated water from container box desiccants into the CRDMs during leak testing after rebuild could potentially initiate Intergranular Stress Corrosion Cracking (IGSCC) or Transgranular Stress Corrosion Cracking (TGSCC). These two types of SCC were initially considered for the potential to cause a separation of the stop piston or separation of the index tube contained within the CRDMs that are constructed of 304 Stainless steel. The completed evaluation indicates that a stop piston separation could cause a slower scram speed and damage the drive so it could not be withdrawn. The potential exists for the control rod to drift out. The piston tube located within the CRDM is a reactor coolant pressure boundary (RCPB) and is an ASME component. The possibility of cracking causing RCPB leakage was eliminated by the evaluation. An index tube separation was eliminated as a potential failure mode. Likewise, the potential for SCC initiation on the Cylinder Tube and Flange (CTF) area of the CRDM resulting in a separation that could prevent a scram or normal insertion/withdrawal of a CRD was eliminated.The long-term evaluation concluded there is no concern for TGSCC. This evaluation also determined the stop piston to piston tube separation is the only failure mechanism that could occur, and only if those components were manufactured from 304 SS.GEH initiated a Root Cause Evaluation (RCA) to determine why this event occurred and has implemented process changes to ensure that the condition does not reoccur. Actions to prevent recurrence, such as eliminating the desiccant material and flushing the closed loop water system, have been completed.For Hatch Unit 2, the 12 CRDMs that have the 304 SS piston tubes should be replaced prior to those CRDMs exceeding 10 years in service. See table below:CRDM S/N, Piston TubeSE0474, 304 SSA8737, 304 SSA9423, 304 SSA6791, 304 SS3095, 304 SSA8729, 304 SS7253, 304 SSA6786, 304 SSSE0368, 304 SSA5409, 304 SS7080, 304 SSA9484, 304 SSInterim Reports were issued to River Bend, LaSalle Unit 2, and Hatch Unit 2 providing the results of an evaluation that concluded the condition would not create a substantial safety hazard or potentially cause a Technical Specification Safety Limit violation for a minimum of one operating cycle. The Browns Ferry Unit 2 drives were shipped but were not installed prior to recall, thus a short-term evaluation for Browns Ferry was not provided.GEH has completed the long-term CRDM evaluation with the following results:A Safety Information Communication is being issued to River Bend, LaSalle Unit 2, and Browns Ferry Unit 2 stating that the CRDMs exposed to the chloride intrusion will not cause a substantial safety hazard, or cause a Technical Specification Safety Limit violation and is therefore not reportable (see enclosure 1 for details).For Hatch Unit 2, the introduction of chlorides could cause a substantial safety hazard for the 12 CRDMs that were manufactured from 304 SS material, and is therefore a reportable condition per 10CFR 21.21(d); however, the 3 CRDMs manufactured from XM-19 material would be considered not reportable.The results of the short-term evaluations for all three plants where CRDMs were installed remains valid.Notified the R2DO (Bonser), R3DO (Peterson), R4DO (Proulx), and Part 21/50.55 Reactors Group by email.x), and Part 21/50.55 Reactors Group by email.)
ENS 50699 + (Dassault Falcon Jet Wilmington Corp. has l … Dassault Falcon Jet Wilmington Corp. has lost a leased piece of equipment that contains radioactive materials. The tool in question is an air blow gun leased from NRD, LLC. The blow gun model number is P-2021-1000, Serial number A2JT781 and was on lease number 055641 shipped on 12/03/2013. The blow gun (contains) 1.7 mCi (61 MBq) (of Po-210). The tool was used to remove static from aircraft interior windows and also to remove static from interior wood paneling between decorative finish coats. It was used primarily in the cabinet department of the company for the purposes described above. The tool was believed to last be used in June of 2014.Since discovering the tool was lost, every effort has been made to locate it by posting lost tool messages throughout the company and searching all tool storage locations and employee tool boxes. To prevent this from happening again, the company has placed tools like this into our inventory tool control program where employees will be required to sign out tools and return them after jobs are complete.The tool was in use at the company's New Castle, DE facility.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.org/MTCD/publications/PDF/Pub1227_web.pdf)
ENS 42067 + (Declaration of an Unusual Event due to Tur … Declaration of an Unusual Event due to Turkey Point Nuclear Site in a confirmed hurricane warning due to the projected path of Hurricane Wilma.The licensee indicated all 4 EDG's were available if required, and all safety systems were also available if required.The licensee will notify the NRC Resident Inspector.UPDATE FROM NRC (WILSON) TO NRC (KNOKE) AT 17:00 ON 10/23/05The NRC entered into Monitoring Mode due to hurricane force winds expected to be at the site within 12 hours.xpected to be at the site within 12 hours.)
ENS 41743 + (Declared Unusual Event due to RCS leakage … Declared Unusual Event due to RCS leakage greater than 10 gallons per minute. VCT level is lowering with constant plant temperature, pressure and pressurizer level. VCT level lowering at approximately 19 gallons per minute with associated rise in RWST level approximately 20 gallons per minute.The licensee will notify the NRC Resident Inspector.* * * UPDATE FROM MELIN TO W. GOTT AT 0120 ON 06/03/05 * * * The licensee briefed NRC Management (McGinty, P. Hiland, R. Haag, J Munday) that the leakage was verified to be CVCS leakage and not RCS leakage. The leakage was stopped at 0055 and the licensee is evaluating termination/retraction criteria. NRC will not enter Monitoring Mode.* * * UPDATE FROM G MELIN TO W. GOTT AT 0140 ON 06/03/05 * * *At 0055, leakage stopped. Leakage verified to be CVCS leakage, not RCS. The leakage is stopped. Validating no other anomalies exist then will evaluate termination and retraction of UE notification.* * * UPDATE FROM G MELIN TO P. SNYDER AT 0221 ON 06/03/05 * * * Notification of termination of Unusual Event at 0221. Plant validation complete. RCS gross leakage is 0.07 gallon per minute and plant parameters are stable. Leakage confirmed to be from CVCS system which has been isolated.The licensee will notify the NRC Resident Inspector.Notified R2DO (R. Haag), NRR EO (P. Hiland), IRD (T. McGinty), R2 (J. Munday), DHS (I. Lee), FEMA (C. Ligget), and DOE (M. Smith)* * * UPDATE ON 6/15/05 @ 1243 FROM JIM SPEICHER TO CHAUNCEY GOULD * * * RETRACTIONEvent Notice 41743 reported declaration of an Unusual Event at 00:33 on 6/3/05 for RCS leakage in excess of Technical Specification Limits. The source of leakage was subsequently determined to be from the Chemical and Volume Control System (CVCS) and not from the RCS. The Unusual Event was terminated at 02:00 6/3/05 after the leak was isolated.This event is retracted since the source of the leakage was not from the RCS and upon further review the event did not meet reporting requirements.The NRC Resident Inspector was notified. RDO (Tom Decker), EO (M. Mayfield) and IRD (Peter Wilson) were notifiedield) and IRD (Peter Wilson) were notified)
ENS 53644 + (Degraded or unanalyzed condition due to th … Degraded or unanalyzed condition due to the possibility for a postulated fire induced hot short to cause a secondary fire in a different fire area, which would be outside the boundaries analyzed for safe shutdown in calculation SSC-001 due to an unfused circuit associated with the 1M43C0001A, Diesel Generator Building Ventilation Fan. This condition is not bounded by existing design and licensing documents. Without overcurrent protection for this circuit, the potential exists that an initial fire event affecting this circuit could cause a short circuit without protection that would cause excessive current through the circuit beyond the capacity rating of the conductors. This could lead to a secondary fire in another plant area where this circuit is routed challenging the ability to achieve and maintain safe shutdown.The postulated event would affect the following fire zones: 1CC-3c (Unit 1, Division 1 4160V and 480V Switchgear Room, 620 feet 6 inch elevation), 1CC-3e (Unit 1 West Corridor North of Elevator, 620 feet 6 inch elevation), DG-1d (Hallway Diesel Generator Building 620 feet 6 inch elevation), and 1DG-1c (Unit 1, Division 1 Diesel Generator Building 620 feet 6 inch elevation).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 Senior Resident Inspector.otified the NRC Senior Resident Inspector.)
ENS 40551 + (Del Monte Foods, a general licensee has be … Del Monte Foods, a general licensee has been unable to locate a fixed gauge containing 100 milliCuries of Americium-241. The unit was one of 4 generally licensed gauges used as fill/level detectors at the Del Monte facility at Terminal Island. Del Monte had the 4 lines removed by contractors and after the completion of the project Del Monte realized one gauge remains unaccounted for. They have searched their facility and checked with all contractors and no information was obtained regarding the whereabouts of the gauge. The licensee has been requested to provide (the California Radiation Control Program Office) additional information regarding the dates/times involved, the names of the contractors who removed the lines, and the names of individuals who may have purchased the equipment.uals who may have purchased the equipment.)
ENS 41203 + (Department of Veterans Affairs, National H … Department of Veterans Affairs, National Health Physics Program report the following information:I am calling to report packages that exceeded the limits for removable radioactive surface contamination. The packages were received between 0700 to 1200, November 17, 2004, by the permittee (VA Medical Center, Salem, Virginia) authorized under the master license issued to the Department of Veterans Affairs, NRC License 03-23853-01VA.The basis for reporting the loss is under 10 CFR 20.1906(d)(1) in that the packages, ammo boxes being used to deliver radioactive materials, had approximately 28, 70, and 110 DPM per centimeter squared removable radioactive contamination on the outside.The contamination was identified as Technetium-99M.The Cardinal Health Nuclear Pharmacy Services delivered the packages. The status of whether the vendor delivery vehicle was exclusive use was not determined.The permittee Radiation Safety Officer notified the vendor about the package and has taken appropriate health and safety steps to prevent any possible spread of contamination.vent any possible spread of contamination.)

DocketNumber/FIN-Item Number + (Description)

ENS 49117 + (Description of Event: During operations at … Description of Event:During operations at full power (500 kWth), the Senior Reactor Operator (SRO) on duty noticed that the fuel temperature thermocouple reader indicated a temperature of 202 (degrees) C, approximately 60 - 70 (degrees) C below the expected value. The SRO recognized that the problem was likely caused by a fuel thermocouple wire grounding to its conduit. A trainee was instructed to move the wires to avoid grounding. Following this action, the thermocouple reader indicated the proper value.Upon review of the log book, the SRO noticed that the faulty fuel temperature reading had been logged for several days without corrective action. The facility Technical Specifications (TS) require at least one fuel temperature indication to be operable during operation, and define a system as 'operable' when it is capable of performing its intended function in a normal manner. Therefore the fuel temperature indication was not operable as defined in the TS. Since the reactor was not immediately secured nor was the indication immediately fixed, the event constitutes a Reportable Occurrence per facility TS 6.9.2.Background:For the four days of reactor operations during which the problem with the thermocouple existed, the reactor was operated for short amounts of time (approximately 10 -15 minutes) at various power levels in order to characterize a new beam port configuration and test an experimental apparatus. The total time above 100 kWth was 2 hours and 51 minutes. The reactor was typically staffed by a trainee at the panel, supervised by a licensed reactor operator (RO).The indicated temperature is the average reading of three thermocouples. One thermocouple is at the fuel midplane, one is 2 (inches) above the midplane, and one is 2 (inches) below the midplane. The fuel temperature indication, when partially grounded, is approximately correct below the point of adding heat (approximately 10 kWth). It differs from the expected value by approximately 15 (degrees) C at 100 kWth, and by approximately 65 (degrees) C at 500 kWth. The fuel temperature readout at the control panel is used to provide an automatic scram at 400 (degrees) C. This setpoint is set well below the Safety Limit of 750 (degrees) C fuel temperature during steady state reactor operations. The fuel temperature scram is NOT required by TS. Normal operations at the reactor do not approach this scram setpoint or the Safety Limit.The logbooks are reviewed daily as part of the pre-operation reactor checkout procedure. The staff is trained to review logs back to the most recent time they were on duty, to check for changes to the reactor, problems with instruments, etc. The staff is not trained to audit the previous days' logs for anomalous readings.Timeline:The following timeline of operations is taken from the reactor logbook. Only operations at or above 100 kWth are listed, because the difference between measured and expected temperature is small at lower power levels. All times are local (Central Daylight Time).Date Time Power T (Measured, (degrees) C) T (Expected, (degrees) C)6/7/2013 0946-1001 500kW 200 2656/7/2013 1037 - 1050 500kW 202 2656/7/2013 1441 - 1446 500kW 202 2656/10/2013 0937 - 0950 100kW 83 1006/10/2013 1005 - 1015 100kW 87 1006/10/2013 1029-1041 100kW 84 1006/12/2013 0906 - 0919 100kW 84 1006/12/2013 0934 - 0959 530kW 217 2706/12/2013 1611 - 1633 100kW 74* 1006/13/2013 1103 - 1143 500kW 201 2656/13/2013 1352 - 1406 530kW 209 270*Temperature was logged while it was still rising toward an equilibrium value.6/13/2013 - 1559 - Problem observed and corrected by repositioning thermocouple wires.6/14/2013 - 1020 - Reportable occurrence reported to NRC Headquarters Operations Center.Causes:The facility has identified the following as contributing causes to the event. 1. Licensed operators were not sufficiently attentive when supervising trainees at the control panel. The licensed operators were focused on reactor power indications and did not pay sufficient attention to other TS related indications.2. The log book review required prior to daily operations was not conducted with sufficient rigor to detect the improper thermocouple readings logged the previous day. The review was instead focused on noting changes to the reactor facility and problems with instrumentation since the operators' previous duty at the panel. 3. Only one functioning instrumented fuel element was used to provide the required fuel temperature indication channel. Therefore no redundant readout was available to check against the indicated fuel temperature.4. The sharp edge on the instrumented fuel element conduit can cut through the insulation on the thermocouple wires, causing grounding.Corrective Actions:The facility will perform the following corrective actions. All changes to the reactor systems, such as thermocouple wire insulation, are subject to review per the requirements of 10CFR50.59.Time: Prior to operation;Action: Attempt to improve insulation on thermocouple wires, using electrical tape, shrink tubing, or spray-on insulation.Time: Prior to operation;Action: Attempt to repair thermocouple wires for a currently installed but non-functional instrumented fuel element to provide an independent fuel temperature indication channel.Time: Prior to operation;Action: Install a new instrumented fuel element. This will bring the total number of independent channels of fuel temperature indication to 2 - 3.Time: Prior to operation and as part of requalification training program;Action: Train reactor staff on the importance of vigilance when supervising trainees and the importance of attentiveness to all channels of information at the control console, as opposed to focusing on a few specific indicators, such as reactor power channels.Time: Prior to operation and as part of requalification training program;Action: Train reactor staff to check for anomalous values in the prior days' log entries during the daily reactor checkout.Time: Upon approval by Reactor Safeguards Committee, but not necessarily prior to operation;Action: Append Procedure 15 - Reactor Startup with a list of observed instrument values for different reactor power levels to be used as a reference by trainees and licensed staff.A copy of this report will be provided to the Kansas State University Reactor Safeguards Committee for review.y Reactor Safeguards Committee for review.)
ENS 40763 + (Description of Incident: Approximately 11: … Description of Incident: Approximately 11:00 central time on 5-20-04, The Division of Radiological Health, MS State Dept. of Health, was notified that a leaking Cesium 137 source was found when a consultant was preparing sealed sources for disposal. The leaking source was found in the bottom of a storage container (original pig from Du Pont) that was used to store a Du Pont Merck Model NER-401H 592 microcurie Cs-137 dose calibrator source. The leaking source was found after the dose calibrator source had been removed from the storage pig and determined to be free of contamination. The leaking source was discovered in a crack at the bottom of the storage pig and was in excess of .005 microcuries removable contamination. A consultant stated that the leaking source appeared to be about the size of a pencil lead. Instead of trying to remove the contaminated source from the pig, he put epoxy in the pig to seal it in place to ready it for disposal. He estimated that the source was probably used as a check source. The dose calibrator source was originally purchased from Du Pont Merck in 1982. All tests for leakage/contamination were negative for over 20 years. It is our belief that the leaking source was in the pig when it was purchased from Du Pont Merck. The licensee had never been licensed for any other Cs-137 sources.Describe clean-up actions: Consultant had removed all sources and completed close-out surveys and wipes for termination of license.List Radiation measurements taken: Consultant stated that survey readings at top of storage pig were approximately 0.2 mRem/hr.List any other actions required of DRH: follow up as required and notify NMED.Case Closed: Yesequired and notify NMED. Case Closed: Yes)
ENS 49649 + (Description of incident: HDR brachytherapy … Description of incident: HDR brachytherapy treatment of the bronchial, tracheal area. The source was placed incorrectly, resulting in underdose to intended treatment area. An additional treatment fraction will be prescribed to compensate for underdose to intended treatment area. Underdose was discovered on 1st of 3 fractions, prior to delivery of 2nd fractions. Sequelea: No adverse effects expected as a result of medical event, since the patient is terminally ill.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.necessarily result in harm to the patient.)
ENS 49155 + (Description of the Deviation or Failure to … Description of the Deviation or Failure to comply that is being evaluated:Linear indications/cracks were discovered in the flange forging on four N-E13DM pressure transmitters recently supplied by Ultra Electronics, NSPI to Omaha Public Power District (OPPD), Fort Calhoun Station FC-1-5 Plant. These transmitters with the identified indications had been received by Fort Calhoun but were not installed. Three of the four transmitters have been returned to Ultra Electronics, NSPI for evaluation and analysis. At this time, the fourth transmitter is in process of being returned for additional evaluation.Evaluation Status:Ultra Electronics, NSPI has performed an initial assessment and determined that the indication/crack penetrates the forging beyond the specified minimum wall thickness. At this time Ultra Electronics, NSPI is in the process of contracting a third party metallurgical research laboratory to assist with evaluation and root cause analysis. Ultra Electronics, NSPI will supply forging samples to the metallurgical research laboratory for analysis including material/composition characteristics as well as the structure and physical characteristics of the indications. Ultra Electronics, NSPI will continue to work with the forging vendor to determine root cause and the extent of the affected material.Completion of the Evaluation:The evaluation is expected to be completed on or before August 31, 2013 be completed on or before August 31, 2013)
ENS 51974 + (Description of the Event: On June 2, 2016 … 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.003888517. The Licensee will notify NRC Region II.)
ENS 46444 + (Description of the Event: This report is p … Description of the Event: This report is performed pursuant to 10 CFR 40.60(b)(2) reporting requirements.On Tuesday, November 30, 2010, at approximately 0500, the failure of F-Substation caused a loss of instrumentation in operating units in the Feed Materials Building. At 0505, various weight, temperature and pressure indications were not available. Due to the loss of power to a load cell component, the UF6 Vaporizer weight exceeded the administrative limit. During the power loss the vessel's weight measurement was not available to the operator. At 0538, power was restored and instrumentation readings returned to normal. The loss of the Vaporizer weight control did not have any safety consequences.Isotope, Quantities and Chemical Form: No material release. Personnel Radiation Exposure Date (if applicable): No additional exposure to radiation or radioactive materials.The licensee will inform the NRC Region II Office.* * * RETRACTION FROM MICHAEL GREENO TO JOE O'HARA AT 1659 ON 12/20/10 * * *On December 1, 2010, Honeywell Metropolis Works made notification of Event #46444 to the NRC Operations Center. This November 30, 2010 incident occurred due to the loss of power to a load cell component. During the power loss the weight measurement was not available to the operator, and the UF6 Vaporizer weight exceeded the administrative limit. This twenty-four hour report was made following 10 CFR 40.60(b)(2) reporting requirements:Each licensee shall notify the NRC within 24 hours after the discovery of any of the following events involving licensed material: ...(2) An event in which equipment is disabled or fails to function as designed when:(i) The equipment is required by regulation or license condition to prevent releases exceeding regulatory limits, to prevent exposures to radiation and radioactive materials exceeding regulatory limits, or to mitigate the consequences of an accident;(ii) The equipment is required to be available and operable when it is disabled or fails to function; and(iii) No redundant equipment is available and operable to perform the required safety function.As determined by Honeywell's safety analysis, potential UF6 release may occur due to failure to control process resulting in process vessel failure. UF6 Vaporizer load cells, which are identified in the safety analysis as a safety feature to control the total weight in this vessel, were not operable during the incident. However, load cells are not the only component designed to perform the equivalent safety function - to prevent UF6 Vaporizer failure. Thus, UF6 Vaporizer failure due to over-pressurization is prevented by the existing relief system which is designed to relieve at a pressure below the maximum allowable working pressure. This Vaporizer relief system was available and operable during the event. As a result of these additional considerations, Honeywell determined that the redundant equipment (Vaporizer relief system) was available and operable to perform the required safety function (Vaporizer failure prevention), and therefore the initially reported incident (# 46444) does not meet the reporting requirement (iii). Since this incident does not constitute a reportable event, Honeywell requests to withdraw its event notification #46444.Notified R2DO(Henson) and NMSS EO(Davis). Notified R2DO(Henson) and NMSS EO(Davis))
ENS 44882 + (Description of the material involved, incl … Description of the material involved, including kind, quantity, and chemical and physical form: GE Vapor Tracer 2 detection system (SN 07024933861). Approx. 8 mCi Nickel-63 (#09-3896). Description of the circumstances under which the loss occurred: On 29 Jan 09, during a routine semi-annual audit, the Andrews Air Force Base (AAFB) Radiation Safety Officer (RSO) identified the missing device used by the Aerial Port Squadron (APS). The device was used to screen luggage for explosives/narcotics. A review of records indicated the device was sent to the Defense Reutilization and Marketing Service (DRMS) through base supply by the user on 1 Jul 08. On 4 Feb 09, the RSO contacted the USAF Radioisotope Committee (RIC) to report the event. NOTE: The RSO is a member of the Bioenvironmental Engineering (BE) office. An investigation shows the APS (user) contacted the BE office on 24 Jan 08 and received instructions for disposing of the device via the DRMS. A technician from the BE office certified the device did not contain any hazardous material and would be suitable for turn-in to the DRMS. On 25 Jan 08, an officer from the BE office (not the RSO) sent a letter to the user affirming the technician's assessment. The technician and officer had no knowledge the device contained radioactive material (RAM). No RAM warning labels were visible on the exterior surface of the device. During the investigation of the loss of material, the RSO found the RAM warning label was located under the battery pack and not visible. Being unfamiliar with the device, the technician evaluated it for chemical hazards (and found none) but didn't consider radiological aspects. After review of all transfer/receipt and inventory documents over a span of three years, the RSO found no evidence that another event of this type had occurred at AAFB. This is an isolated event. Air Force Instruction 40-201, Managing Radioactive Material in the U.S. Air Force, expressly prohibits GLDs from being sent to the DRMS. A statement of disposition, or probable disposition, of the material involved: The investigation by the RSO revealed the device left AAFB and traveled to DRMS at FT Meade, MD and then to the DRMS at Mechanicsburg, PA. It was sold to a de-manufacturing contractor, Global Investment Recovery, where it was shredded for scrap. The scrap was sold through the DRMS scrap sales partner, Government Liquidation, LLC. It could not be tracked further. Exposures of individuals to radiation, circumstances under which the exposures occurred, and the possible total effective dose equivalent to persons in unrestricted areas: Nickel-63 is a pure beta emitter with a 100 year half life. The ingestion annual limit on intake (ALI) is 9 millicuries (mCi). Ingestion of the source would not exceed the ALI and the committed effective dose equivalent (CEDE) would be less than 5 rem. The (Class D) inhalation ALI is 2 mCi. Using the EPA's inhalation dose conversion factor of 3.1 millirem/microCuries, the CEDE for 8 mCi of Ni-63 (Class D) is 25 rem. The size of the source is approximately that of a pencil eraser. It is unlikely an industrial shredder could have pulverized it to such an extent so as to present an inhalation hazard. The possible total effective dose equivalent to persons in unrestricted areas is presumed to be less than 1% of the ALI. Actions that have been taken, or will be taken, to recover the material: The RSO tracked the device from AAFB (1 Jul 08) to the DRMS at Fort Meade, MD and to the DRMS at Mechanicsburg, PA (7 Aug 08). It was shredded by Global Investment Recovery, the DRMS de-manufacturing contractor and sold as scrap through the DRMS scrap sales partner, Government Liquidation, LLC. It could not be tracked further. Procedures or measures that have been, or will be, adopted to ensure against a recurrence of the loss of material: The RSO has determined the root cause of the event to be inadequate training. The users of the device and the BE office did not, collectively, possess information about GL material. The RSO had not shared his knowledge about the material with the BE office.The following corrective measures have been taken: Photographs of GLDs will be taken and provided to the BE staff, the users, base supply and DRMS. If warning labels are considered to be inadequate, new labels will be affixed to GLDs to alert individuals that RAM is present. The RSO, working with BE staff, has initiated awareness training for users of GLDs. Such will be conducted annually. Users will be required to possess binders in which to maintain information about their devices (e.g., safety data sheets, owner's manuals, policy, pictures, inventory, leak test results, contact numbers, etc.).The RSO has informed management of the event. Lessons learned will be passed down to targeted audiences. Andrews AFB instructions will be evaluated for gaps in policy and strengthened as necessary. The means for purchasing and transferring GLDs back to the manufacturer will be addressed in detail. The RIC Secretariat has contacted the GE Radiation Safety Officer regarding the placement of the radiation warning label. The RIC Secretariat informed Ms. Rachel Browder, NRC Region-IV, about the loss of a GLD on 10 Feb 09.THIS MATERIAL EVENT CONTAINS A "LESS THAN CAT 3" LEVEL OF RADIOACTIVE MATERIALSources 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.This source is not amongst those sources or devices identified by the IAEA Code of Conduct for the Safety & Security of Radioactive Sources to be of concern from a radiological standpoint. Therefore is it being categorized as a less than Category 3 sourceg categorized as a less than Category 3 source)
05000321/FIN-2017004-03 + (Description: During the February 2017 Uni … Description: During the February 2017 Unit 2 refueling outage, all eleven 3-stage safety relief valves (SRVs) were removed and replaced. The SRVs were Target Rock model 0867F, a 3-stage valve design which was in its first use on Unit 2. This design was adopted as a corrective action to address corrosion bonding experienced by 2-stage SRV model 7687F valves which were previously in use at Hatch. "As-found" testing results indicated two of the eleven SRVs had experienced a setpoint drift during the previous operating cycle which resulted in their failure to meet the Technical Specification (TS) opening setpoint pressure as required by TS Surveillance Requirement (SR) 3.4.3.1. The SRV pilot valves were disassembled and inspected to determine the reason for the drift. The licensee determined that the abutment gap closed pre-maturely most likely due to loose manufacturing tolerances. For the 3-stage design, the pilot disc seating stresses should increase proportionally as reactor pressure increases to where a mechanical gap within the valve stem mechanism, referred to as the abutment gap, is closed. Additional pressure increases will cause the valve stem mechanism to reduce the disc seat pressure until the valve eventually opens. This same cause was previously identified in 2016 (CAR 264544) after two of eleven SRVs removed from Unit 1 also experienced setpoint drift. Because the Unit 2 valves were already installed when the cause was initially identified, there was no opportunity for the licensee to take corrective actions for the valves that are the subject of this LER. Additionally, there were no symptoms available to operators or maintenance personnel to indicate the potential for the set point drift prior to post-service testing. As a corrective action, when the eleven valves were removed for post-service testing, the licensee installed eleven refurbished pilot valves that underwent the corrective actions identified by CAR 264544 which included the vendors usage of revised tolerances.Enforcement: Hatch Unit 2 TS limiting condition for operation 3.4.3, Safety/Relief Valves, required 10 of 11 SRVs be operable in MODES 1, 2 and 3. With two or more SRVs inoperable, the required TS action must be taken by the applicable completion time. Contrary to the above, Unit 2 operated from the initiation of the degraded condition until February 6, 2017, with two SRVs inoperable. The inspectors concluded that the violation would normally be characterized as a Severity Level IV violation because it was of very low safety significance (Green). However, the NRC is exercising enforcement discretion (EA-18-006) in accordance with Section 3.10 of the Enforcement Policy because the violation was not associated with a licensee performance deficiency. This issue was documented in the licensees corrective action program as CR 10382586. corrective action program as CR 10382586.)
ENS 44065 + (Description: Three patients were schedule … Description: Three patients were scheduled for transperineal permanent prostate seed brachytherapy implantations on March 14, 2008. Three separate packages of seeds in preloaded needles were received; surveys showed no surface contamination or contamination outside the inner sterile containers. On March 14, 2006, after 12 of 106 seeds were implanted in the first patient, a survey meter showed a small amount of radioactive contamination on the inside of the sterile packaging. This implantation was stopped. The survey meter showed contamination on the tips of three of the four needles that had been used, the greatest being 5000 cpm (420 Bq if an efficiency of 20% is assumed). This patient was administered stable iodine to block his thyroid in case of a leaking seed. The seed vendor was notified by telephone.To determine if the remaining patients should be implanted, the remaining two packages of seeds were opened and the interiors of the sterile packaging were surveyed. No contamination was found. An implant procedure was performed on the second patient. At the end of the procedure, the used needles were surveyed. A survey meter showed contamination on the tips of two of the needles; it was about 1000 cpm (83 Bq if an efficiency of 20% is assumed) on each. The seed vendor was again notified by telephone. A urine bioassay of this patient showed no radioactivity.Implantation of the third patient was cancelled.The needles of preloaded seeds were supplied by Best Medical International. The seeds contained I-125 and were Best Model 2301. The three batches of seeds were Lot Numbers 23017, 23019, and 23018 (not implanted). At this time, it is uncertain whether any seeds were leaking. A possibility is that the contents of the sterile packages were contaminated by the vendor, but no seeds were leaking.Effect on Patients: The VA is still evaluating this event. At this time, no adverse effects to the patients are expected.Patient notification: The permittee is in the process of ensuring that the referring physicians and patients were notified.We will notify the NRC Project Manager, Cassandra Frasier, of NRC Region III.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.necessarily result in harm to the patient.)
ENS 47310 + (Description: At 0854 CDT on 09-30-11, the … Description: At 0854 CDT on 09-30-11, the Plant Shift Superintendent (PSS) was notified that water was observed in the #5 withdrawal position scale pit during the completion of the monthly test of the C-310 scale pit water detection system alarm module. The alarm module was being tested per procedure, when the module was found with the visual alarm (a red light) on at the local panel in the #5 withdrawal position room. In response to the alarm, the scale pit hatch was opened and the water detection sensor cable was observed to be at least partially submerged. Immediate investigation found the sump pump breaker to be tripped; when the breaker was reset the pump actuated and water was immediately removed. At the time of the occurrence, product withdrawal was in progress in the #3 and #4 withdrawal position room, and no cylinder was present in the #5 withdrawal position room. The source of the water was found to be a leaking steam condensate valve above the #5 withdrawal room ceiling. The water had drained to the concrete pad outside the building and then along the scale cart rails, eventually finding its way into the #5 scale cart pit where it gradually accumulated. Because the C-310 Scale Pit Water Detection Alarm did not function as credited, it is in violation of NCSE 032 (NCSA 310-004). Since one leg of double contingency was lost, this is being reported to the NRC as a 24-hour Event Report in accordance with NRC BL 91-01 Supplement 1.The NRC Senior Resident Inspector has been notified of this event. PGDP Problem Report No. ATRC-11-2610; PGDP EventReport No. PAD-2011-17SAFETY SIGNIFICANCE OF EVENTS: The safety significance of this event is low, even though the event made it possible for the level of pre-existing water to exceed the safe geometry limit. Although it is normal case for overall PGDP operations to have assay up to 5.5 wt.% 235U, the actual assay of product withdrawal operations during the period in question remained no higher than 2.0 wt. % 235U. At that actual assay, the depth of water necessary to support a criticality would have been more than 7.21 inches, which might have been credible but in itself would have remained a very unlikely possibility due to the slow ingress rate and high probability of detection and mitigation by personnel performing routine activities in that area.POTENTIAL CRITICALITY PATHWAYS INVOLVED (BRIEF SCENARIO(S) OF HOW CRITICALITY COULD OCCUR): In order for a criticality to happen, a significant breach in the process system integrity would have to occur. After the breach, fissile UF6 and its reaction products would have to react with pre-existing water to form fissile solution. There would have to be a sufficient depth of water in the pit to support a criticality (e.g. more than 3.68 inches of water at 5.5 wt. % 235U).CONTROLLED PARAMETERS (MASS, MODERATION, GEOMETRY, CONCENTRATION, ETC.): The two process conditions relied upon for double contingency are mass and geometry.ESTIMATED AMOUNT, ENRICHMENT, FORM OF LICENSED MATERIAL (INCLUDE PROCESS LIMIT AND % WORST CASE CRITICAL MASS): Product withdrawal assay at the time of the event was no higher than 2.0 wt% U235. However, no UF6 release occurred.NUCLEAR CRITICALITY SAFETY CONTROL(S) OR CONTROL SYSTEM(S) AND DESCRIPTION OF THE FAILURES OR DEFICIENCIES: Geometry is controlled in the second leg of double contingency by limiting the level of pre-existing water that might be present in the scale pit. Water accumulation is considered normal case in the NCSE by acute or by chronic sources. The NCSE credits the Scale Pit Water Detection Alarm to provide detection of chronic water accumulation in the scale pit. The alarm is set to actuate before the water level exceeds 2.5 inches in the pit. The geometry parameter limit is 3.68 inches assuming the worst-case possible enrichment of 5.5 wt.% 235U. The alarm is credited to provide early indication, and result in prompt mitigation, of water ingress to the pit before the NCS parameter limit is exceeded. Since the alarm was not functional, and the ingress rate was sufficiently slow that ingress was not easily detectable, there was no reliable means in place to detect and mitigate the ingress of water into the pit. The sensor and local panel light performed their intended function; however, it is the ACR audible and visible alarms that are controlled as AQ-NCS equipment and not the local panel light and buzzer. With the alarm out of service, continued ingress of water to the pit could have resulted in exceeding the geometry parameter limit for water depth before detection and mitigation. NCS entered the scale pit for inspection shortly after notification of the discovery and after the water had been drained. NCS observed that the water level at the lowest point in the pit may have reached 2.5 inches. Based on those inspections, it is likely that the water level remained below the 3.68-inch level, but there was no definitive way to prove the maximum height that might have occurred throughout the period of time when the alarm was not functional. Therefore, for conservatism it is assumed credible that the geometry parameter limit was violated during the lime the alarm was not functional.CORRECTIVE ACTIONS TO RESTORE SAFETY SYSTEMS AND WHEN EACH WAS IMPLEMENTED: The sump pump was immediately activated by resetting its breaker, which restored the water level in the pit below the 2.5 inch administrative limit, thus removing the hazard of this incident. The #5 withdrawal position scale pit will be checked twice per shift beginning on 9-30-11 in accordance with procedure due to the ACR alarm being out-of-service.due to the ACR alarm being out-of-service.)
ENS 43927 + (Detection of toxic gases (CO) inside the t … Detection of toxic gases (CO) inside the turbine building. Source not identified.Initial Readings were greater than 50 parts per million in turbine building lower elevations.EALHU5.1Detected by safety department random sampling.Turbine building has been evacuated, all personnel have been accounted for with no injures.Follow-up notification will be made with determination of cause of event.The unit is currently defueled. The carbon monoxide source is being investigated.The NRC Resident Inspector as well as state and local authorities were notified.* * * UPDATE FROM ERIC PICKRELL TO JOE O'HARA AT 1932 EST ON 1/23/08 * * * Notified J. O'Hara of termination of the declared NOUE. Termination is based on subsequent air monitoring sampling revealing normal Oxygen levels and no detectable Carbon Monoxide levels.(unknown) Source of abnormal sample readings is not yet determined. Believed to be result of Hydrogen Water Chemistry system purge which was in progress at the time of sampling.Licensee terminated the declaration of notification of unusual event at 1827 CST following two subsequently satisfactory air samples. Notified R4DO(Johnson), NRR(Case), and IRD(Blount)* * UPDATE FROM TIM SCHENK TO JOHN KNOKE AT 1515 EST ON 01/24/08 * * It has been determined that hazardous levels of toxic gases (Carbon Monoxide) inside the turbine building did not exist at the time of emergency declaration. Actual carbon monoxide levels were less than detectable. Gas sensor cross contamination led to false readings of high carbon monoxide level (60 ppm), when actual gas measurements were reading hydrogen levels of 20 ppm. This level of hydrogen is well below the 40,000 ppm Lower Explosive Limit for hydrogen. The source of the hydrogen readings was attributed to a leaking ventilation rig for the Hydrogen Water Chemistry system during system purge with Nitrogen.The licensee notified the NRC Resident Inspector. Notified R4DO (C. Johnson).nt Inspector. Notified R4DO (C. Johnson).)
ENS 40344 + (Deviation related to upper stud assemblies … Deviation related to upper stud assemblies for General Electric Nuclear AKR-30 low voltage circuit breakers. The deviation is specific to upper stud assemblies supplied under part number Q139C4632G1 and consist of an incorrect angle between the stud and pivot. Of the fifteen assemblies supplied with possible deviations to AmerenUE, five were returned and ten had been installed in Callaway Plant. The deviation will not prevent the circuit breakers from performing their design basis function at the Callaway Plant, however, the capability of the assemblies is indeterminate for severe faulted conditions. A circuit breaker could fail if an upper stud assembly with identified deviation was installed and the circuit breaker was called upon to interrupt a severe fault.Callaway has concluded that this deviation does not constitute a "defect" as defined in 10CFR Part 21 because the breakers would still perform their design basis requirements and would not create a substantial safety hazard. However, Callaway can not determine if the potential for a significant safety hazard or exceeding of a technical specification safety limit could exist at another nuclear power plant.Received written documentation that the licensee has notified the NRC resident inspector.e has notified the NRC resident inspector.)
05000275/FIN-2011002-05 + (Diablo Canyon Facility Operating License D … Diablo Canyon Facility Operating License DPR 80/DPR 82, License Condition 2.C(5), Fire Protection, required Pacific Gas and Electric to implement and maintain all provisions of the approved fire protection plan as described by the FSARU. FSARU, Appendix 9.5a, Fire Hazards Analysis, and Equipment Control Guideline 18.7, Fire Rated Assemblies, required that all fire penetration seals necessary to protect safe shutdown equipment be operable. Contrary to the above, on January 13, 2011, the licensee identified that the penetration seal for Conduit K2026, providing a 4-kV power cable between Fire Area 20, Unit 2 12-kV Cable Spreading Room, and Fire Area 19A, Unit 2 Turbine Building, was missing the internal seal. The licensee subsequently identified that several additional penetrations were not installed per the design specification. In both Unit 1 and Unit 2 12-kV cable spreading rooms, several conduits were installed with nonconforming materials that did not meet the 3-hour fire barrier rating. The licensee implemented appropriate compensatory measures following discovery of the condition and entered the problem into the corrective action program as Notifications 50370048, 50371018, 50371019, 50370858 and 50370859. The inspectors determined that the finding was of very low safety significance (Green) because the degraded barriers would provide a minimum of 20 minutes fire endurance protection and in situ fire ignition sources and combustible or flammable materials were positioned such that the degraded barrier would not be subject to direct flame impingement.ot be subject to direct flame impingement.)
05000275/FIN-2011004-07 + (Diablo Canyon Power Plant Technical Specif … Diablo Canyon Power Plant Technical Specification 3.7.10, Control Room Ventilation System (CRVS) , required two CRVS trains to be operable. Contrary to this, the licensee failed to maintain the control room ventilation system in the design configuration. On August 29, 2011 plant operators identified that the integrity of the control room habitability envelope had been lost after maintenance personnel erroneously removed a blank flange supporting maintenance on the Unit 2 ventilation system inlet isolation dampers. This condition could have prevented fulfillment of the control room to meet the safety function to provide adequate radiation protection to prevent the occupants from exceeding 5 rem whole body (equivalent) radiation exposure for the duration of an accident. The operators identified that the boundary was inoperable and entered Technical Specification 3.7.10-B.1 and took corrective action to reinstall the blank flange to restore integrity of the control room envelope. The performance deficiency was more than minor because it was similar to example 4.a, in Manual Chapter 0612, Appendix E, Examples of Minor Issues, because the inspectors concluded that the removal of the blank flange adversely affected safety-related equipment. The inspectors used Inspection Manual Chapter 0609.04, Phase 1 - Initial Screening and Characterization of Findings, to analyze the finding. The inspectors concluded that the failure to maintain the operability of the control room ventilation system in accordance with Technical Specification 3.7.10, was a finding of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room. Pacific Gas and Electric entered the issue into the corrective action program as Notification 50427567 and 50428115.ram as Notification 50427567 and 50428115.)
ENS 41774 + (Diablo Canyon Units 1 and 2 declared a Not … Diablo Canyon Units 1 and 2 declared a Notification of Unusual Event at 2050 PDT 06/14/05 due to the receipt of a Tsunami Warning from the California State Warning Center.The licensee notified State and local agencies and will notify the NRC Resident Inspector. The NRC remained in the Normal Mode.* * * UPDATE FROM D. BAHNER TO M. RIPLEY 0045 EDT 06/15/05 * * *The Tsunami Warning was cancelled and Diablo Canyon Units 1 and 2 terminated the Notification of Unusual Event at 2138 PDT 06/14/05. The licensee notified State and local agencies and the NRC Resident Inspector.l agencies and the NRC Resident Inspector.)
ENS 41612 + (Dianon Systems is a diagnostic laboratory … Dianon Systems is a diagnostic laboratory licensed by the NRC to handle P-32 and I-125. On April 13, 2005, Dianon received a shipment of two radioactive sources without any information as to what the sources were, their strength (i.e., radioactivity), or what diagnostics where to be performed. Dianon immediately contacted the originator of the shipment, a doctor out of Brooklyn, NY. The doctor was not willing to provide any information on the sources other than to say that the sources were "6 months old, safe, and no threat." Dianon surveyed the sources and found that they were still radioactive, however, the caller was unable to provide a meaningful millirem/distance reading. Dianon has safely secured the sources in lead shielding and is uncertain on how to proceed in resolving disposition of the sources.d in resolving disposition of the sources.)
ENS 40306 + (Discovered vulnerability in a safeguard sy … Discovered vulnerability in a safeguard system that could allow access to a controlled access area for which compensatory measures were not employed. The licensee notified the NRC Resident Inspector. TAS notified Region IAT member who will followup for NRC.Contact the Headquarters Operations Officer for additional details.* * * UPDATE ON 11/13/03 @ 1340 BY T. MANZELLA/ R. MURILLO TO GOULD * * * RETRACTIONAfter further review it was determined that there was no vulnerability that would allow access to a controlled access area, therefore the licensee is retracting this event.The NRC Resident Inspector was notified.. The NRC Resident Inspector was notified.)
ENS 40273 + (Discovered vulnerability in a safeguard sy … Discovered vulnerability in a safeguard system that could allow unauthorized or undetected access to safeguards material. Compensatory measures have been employed. The licensee notified the NRC Resident Inspector. Contact Headquarters Operations Officer for additional details.Operations Officer for additional details.)
ENS 40276 + (Discovered vulnerability in a safeguard sy … Discovered vulnerability in a safeguard system that could allow access to a controlled access area for which compensatory measures were secured prematurely. The licensee notified the NRC Resident Inspector, the CT State Police and local law enforcement. Contact the Headquarters Operations Officer for additional details.The NRC Resident Inspector was notified of this event by the licensee.as notified of this event by the licensee.)
ENS 40505 + (Discovered vulnerability in a safeguard sy … Discovered vulnerability in a safeguard system that could allow unauthorized or undetected access to safeguards material for which compensatory measures have not been employed. The licensee will be notifying the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.* * * * RETRACTION FROM D. RITTER TO M. RIPLEY 1748 ET 02/04/04 * * * * Seabrook Station is retracting the one-hour notification made at 1924, 3 Feb 04 under EN # 40505. Upon further review of reporting criteria, the incident will be documented in the station's Security Event Log.The licensee will notify the NRC Resident Inspector and NRC Region 1 office. Notified R1 DO (J. Kinneman) and NSIR TAS (A. Danis). DO (J. Kinneman) and NSIR TAS (A. Danis).)
ENS 40382 + (Discovered vulnerability in a safeguard sy … Discovered vulnerability in a safeguard system that could allow access to a controlled access area for which compensatory measures have not been employed. The licensee will notify the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.* * * UPDATE from R. Lowery to R. Jolliffe at 1100 EST on 12/11/03 * * * The licensee provided additional information about this event. Contact the Headquarters Operations Officer for additional details. Notified R4DO (D. Graves) and TAS (J. Whitney)fied R4DO (D. Graves) and TAS (J. Whitney))
ENS 42115 + (Discovered vulnerability in a safeguards s … Discovered vulnerability in a safeguards system that could allow access to a controlled access area. Compensatory measure have been initiated. The licensee will notify the NRC Resident Inspector. Contact the Headquarters Operations Officer for additional details.*** UPDATE AT 11:47 EST ON 11/08/05 FROM EWERS TO KNOKE ***This is a retraction to a notification made in accordance with 10CFR73.71 on November 3, 2005, regarding a security safeguards breach of a vital area at Millstone 2, (NRC Event No. 42115). Upon further investigation and review, it has been determined that there was no breach of the security barrier in that the required barriers were visually inspected and found to be intact. Therefore, this event has been determined to be not reportable in accordance with 10CFR73.71." The NRC Resident Inspector has been notified. Notified R1DO (Miller)has been notified. Notified R1DO (Miller))
ENS 51514 + (Discovery of a PCB (Polychlorinated Biphen … Discovery of a PCB (Polychlorinated Biphenyls) leak on a 440 gallon deenergized transformer. Leak was approximately 0.5 gallons (3-5 lbs of PCB). (The leak was) contained onsite within unit 2, 617 turbine building. Contractor clean-up (is) enroute. No personnel injury or exposure. No offsite release. (The licensee) contacted IEMA (Illinois Emergency Management Agency), National Response Center, and the local emergency planning center.The licensee has notified the NRC Inspector. Notified FEMA, USDA, HHS, DOE, and EPA EOC.otified FEMA, USDA, HHS, DOE, and EPA EOC.)
ENS 40919 + (Division 2 Emergency Uninterruptible Power … Division 2 Emergency Uninterruptible Power Supply (UPS) Inverter is inoperable. The 24-hour completion time, for Technical Specification 3.8.7.A, has expired and Technical Specification 3.8.7.B was entered. Technical Specification 3.8.7.B requires the plant to be in Hot Shutdown by 0503 on August 4, 2004 and Cold Shutdown by 0503 on August 5, 2004.A plant shutdown was initiated at 1930.The licensee has notified the NRC Resident Inspector.e has notified the NRC Resident Inspector.)
ENS 43872 + (Doctor ordered iodine (didn't specify isot … Doctor ordered iodine (didn't specify isotope) thyroid uptake & scan for a patient. This licensee uses I-123 for this purpose. Patient was incorrectly scheduled for a I-131 (2.2mCi) whole body scan and that was done on 12/17/07. Error was discovered 12/25. Patient & doctor have been notified and no adverse health effects are expected. Licensee will submit a written report. Scheduling personnel will be re-educated to verify an order before scheduling patients for a procedure. The tech will be re-educated to read the script before dosing a patient. Any further action on this incident is referred to Radioactive Materials.Incident Number FL07-205A "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.oes not necessarily result in harm to the patient.)

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