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05000010/FIN-2012012-01 +During an NRC investigation completed on November 22, 2011, and a supplemental investigation completed on October 10, 2012, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR 50.75(a) establishes requirements for indicating to the NRC how a licensee will provide reasonable assurance that funds will be available for the decommissioning process and states that for power reactor licensees, reasonable assurance consists of a series of steps as provided in paragraphs (b), (c), (e), and (f) of 10 CFR 50.75. 10 CFR 50.75(f)(2) states, in part, that power reactor licensees shall report at least every 2 years on the status of its decommissioning funding for each reactor or part of a reactor that it owns; and, that the information in this report must include, at a minimum, the amount of decommissioning funds estimated to be required pursuant to 10 CFR 50.75(b) and (c). 10 CFR 50.75(b)(1) states, in part, that for a holder of an operating license under 10 CFR Part 50, financial assurance for decommissioning shall be provided in an amount which may be more, but not less, than the amount stated in the table in paragraph (c)(1) adjusted using a rate at least equal to that stated in paragraph (c)(2). 10 CFR 50.75(c)(1) states the minimum amount required to demonstrate reasonable assurance of funds for decommissioning by reactor type and power level. 10 CFR 50.75(c)(2) requires, in part, that an adjustment factor be applied, which is based on escalation factors for labor and energy, and waste burial. 10 CFR 50.9(a) states, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on March 31, 2005, March 31, 2006, March 31, 2007, and March 31,2009, Exelon Generation Company, LLC (Exelon) provided information on the status of its decommissioning funding that was not complete and accurate in all material respects, when it submitted the decommissioning funding status (DFS) reports pursuant to 10 CFR 50.75. Specifically, the March 31, 2005, March 31, 2007, March 31, 2006, and March 31, 2009, DFS reports stated that the decommissioning funds estimated to be required for each of the reactors, as listed in the report, were determined in accordance with 10 CFR 50.75(b) and the applicable formulas of 10 CFR 50.75(c). However, in multiple instances, the amount reported was a discounted value that was less than the minimum required amount specified by 10 CFR 50.75(b) and (c). This is a Severity Level IV violation.  +
05000029/LER-2001-001 +Yankee Nuclear Power Station ceased power operation in February 1992 and is being decommissioned. On 06/26/01 during the conduct of a Nuclear Safety (Quality Assurance) Audit a discrepancy regarding the alarm setpoints for the Spent Fuel Pit (SFP) Area Radiation Monitor (ARM) was identified. The SFP ARM is an instrument required by Technical Specification 3.3 to ensure early detection of inadvertent criticality during fuel handling activities. The Technical Specification requires the alarm setpoints for the ARM be set at less than 5 mr/hr or two times the background radiation level, whichever is greater, while moving irradiated fuel, control rods or sources. The discrepancy identified was that the background radiation level annotated on procedure OP-4816, "Functional Test and Alarm Setting of the Area Radiation Monitoring System" was 2 mr/hr while the alarm setpoint for both the alert and high alarms was set at 7 mr/hr, thus greater than the Technical Specification requirement. As such, this LER is submitted in accordance with 10CFR50.73(a)(2)(i)(B) as a condition of non-compliance with a Technical Specification. No fuel handling evolutions were in progress at the time of discovery of this issue.  +
05000133/LER-2010-001 +On June 24, 2010, while conducting the quarterly inventory of radioactive sources in accordance with Humboldt Bay Power Plant (HBPP) Unit 3 Radiation Control Procedure (RCP) -6D, "Inventory and Controls for Radioactive Sources," it was discovered that source number HBS-498 was missing from the count room. HBS-498 is a mixed gamma source with an activity of 0.35 micro-curies as of August 9, 2010, and is used for the calibration of gamma detectors. The calibration source radionuclide composition yields an aggregate quantity of missing licensed material of 53 times the quantity specified in 10 CFR 20 Appendix C, which exceeds the reporting criterion of 10 times the quantity specified in Appendix C. At the time Revision 0 to this LER was submitted, it was reported that an extensive search had not been successful in locating the source to date (i.e., as of August 20, 2010); however, the missing source was located in Unit 3 on February 1, 2011. The root cause for this event has been determined to be inadequate Radiation Protection procedures to ensure control of radioactive sources. Procedure RCP-6Lt was revised to strengthen the source control process.  +
05000219/FIN-2003005-02 +A finding was identified by the NRC involving the failure to comply with 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, specifically the failure to take appropriate corrective actions following medium voltage cable failures in 1996 and 2001. As a result, on May 20, 2003, a ground fault on the cable running from the No. 1 EDG output breaker to the 4160 volt emergency bus 1C caused an unrecoverable loss of the 1C electrical bus. The loss of bus 1C, as well as the mitigating system and barrier integrity equipment powered from this bus, forced a Technical Specification required shutdown. This finding was more than minor because it was associated with the equipment performance attribute of the initiating events and mitigating systems cornerstones. Specifically, the failure of the cable associated with the 1C 4160 volt emergency bus resulted in the loss of the bus, which increased the likelihood of a reactor transient and caused the mitigating equipment powered from the bus to be unavailable to perform its safety function. A significance determination process (SDP) Phase 1 screening of this finding was performed and determined that the finding degraded both the initiating event and mitigating systems cornerstones. Therefore, an SDP Phase 2 evaluation was performed, which determined that the finding was of low to moderate safety significance (White). (Section 40A5)  +
05000219/FIN-2004009-02 +An apparent violation associated with emergency planning standard 10 CFR 50.47(b)(4) and Technical Specification 6.8, along with its associated implementing procedures related to configuration control, was identified which has low to moderate safety significance because the Emergency Action Level, Fission Product Barrier Matrix contained an incorrect threshold value used for making a General Emergency and/or Site Area Emergency declaration. The finding is more than minor because it is associated with the EP cornerstone attribute of standard emergency classification and action level scheme and offsite EP. It affects the cornerstone objective of ensuring the capability to implement measures to protect the health and safety of the public during an emergency. The finding is potentially greater than very low safety significance because an untimely General Emergency could delay actions directed by State and local response plans.  +
05000219/FIN-2005006-03 +The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59 Changes, Tests, and Experiments, requirements for the failure to perform an adequate safety evaluation of a change to the facility. Specifically, the safety evaluation did not evaluate the potential for a new type of malfunction of an installed liner associated with the 30-inch overboard discharge line on the emergency service water (ESW) system. This finding was addressed using traditional enforcement since it potentially impacts or impedes the regulatory process in that a required 10 CFR 50.59 evaluation was not adequate. This is contrary to the regulatory process that allows licensees to make changes without a license amendment provided that licensees comply with 10 CFR 50.59 process. The finding is more than minor because there was a reasonable likelihood that the change could have required Commission review and approval prior to implementation. However, the finding has been evaluated as very low safety significance (Green) because the liner was subsequently determined to have not have introduced a new malfunction that would impact on the ESW system.  +
05000219/FIN-2005011-02 +An NRC-identified apparent violation (AV) of 10 CFR 50.47(b)(4) was identified. This AV, which has low to moderate safety significance, occurred because the Oyster Creek E-Plan EAL matrix was not properly utilized to determine if a plant parameter met the EAL threshold for declaring an emergency classification. This resulted in not recognizing during an actual event, that plant parameters met the EAL thresholds for declaring a UE and a subsequent Alert. Immediate corrective actions were taken in which shift crews were retrained on the implementation of E-Plan requirements. The finding is greater than minor because it is associated with the EP cornerstone attribute of response organization (RO) performance (actual event response). It affects the cornerstone objective of ensuring the capability to implement measures to protect the health and safety of the public during an emergency. The licensee did not use the Oyster Creek E-Plan EAL matrix when plant parameters met the EAL thresholds for declaring a UE and a subsequent Alert. As a consequence, both the onsite and offsite EROs were not activated during actual Alert conditions. Had the event degraded further, the onsite ERO would not have been readily available to assist in the mitigation of the event and the offsite agencies could have been prevented from taking initial offsite response measures. This finding is of low to moderate safety significance because it constituted a failure to implement a Risk Significant Planning Standard during an actual event in which plant conditions met an Alert. The cause of the finding is related to the cross-cutting element of human performance (personnel). (Section 3.1)  +
05000219/FIN-2007004-01 +The inspectors performed two maintenance effectiveness inspection activities. The inspectors reviewed the following degraded equipment issues to assess the effectiveness of maintenance: C C RFP motor failure on July 17, 2007 (IR 650654); and C Bank 6 startup transformer B phase voltage regulator failed upscale on July 17, 2007 (IR 650702). The inspectors verified that the systems or components were monitored in accordance with AmerGens maintenance rule program requirements. The inspectors compared documented maintenance preventable functional failure (MPFF) determinations and unavailable hours to those being tracked by AmerGen to evaluate the effectiveness of AmerGens monitoring activities and determine whether performance goals were being met. The inspectors reviewed completed maintenance work orders and procedures to determine if inadequate maintenance contributed to equipment performance issues. The inspectors reviewed applicable work orders, corrective action program condition reports, preventive maintenance tasks, vendor manuals, and system health reports. Documents reviewed for this inspection activity are listed in the Supplemental Information attachment to this report. No findings of significance were identified. An unresolved item (URI) was identified to review AmerGens corrective action program evaluation (IR 650654) regarding the C RFP motor failure on July 17, 2007. The inspectors plan to review this evaluation after it is completed, which had not occurred by the end of this inspection period. (URI 05000219/2007004-01, C Reactor Feedpump Motor Failure)  +
05000219/FIN-2007005-04 +On December 19, 2007, a reactor power reduction to approximately 50% was commenced to perform planned maintenance on the reactor recirculation pump MG sets and to find and repair condenser tube leaks in the A north water box. Shortly after reducing power to 55%, the plant experienced a loss of vacuum in the A condenser and a trip of the A reactor feedpump due to low suction pressure. Operations personnel responded in accordance with abnormal operating procedures ABN-14,Loss of Condenser Vacuum, and ABN-17, Feedwater System Abnormal Conditions; and performed a manual reactor scram (shutdown) due to the plant conditions. Specifically, the B reactor feedwater pump was removed from service during the power reduction per operating procedures; and with only the C reactor feedwater pump in service, operators performed a manual scram per abnormal operating procedure ABN-17 guidance. Operators mitigated the reactor scram and stabilized the plant in accordance with abnormal operating procedure ABN-1, Reactor Scram and emergency operating procedure (EOP) EMG-3200.01A, RPV Control - No ATWS. Operations personnel and equipment responded as expected during the event. The plant was maintained in hot shutdown while investigation into the cause of the event was determined. At the time of the event, Oyster Creek was operating with two of its four circulating water pumps in-service. In accordance with AmerGens environmental plan and work management schedule for the downpower, the circulating water system was reduced to two pump operation to maximize discharge water temperatures and to minimize the thermal shock impact to aquatic life in the discharge canal during winter conditions. AmerGens preliminary investigation (IR 713652) into the cause of the event determined that two circulating water pump operation, combined with draining of the A north water box, resulted in degraded condenser vacuum, reduced performance of the A condensate pump, and the subsequent trip of the A reactor feedwater pump on low suction pressure. AmerGen reported this event to the NRC in Event Notification 43854, Manual Reactor Scram Due to Lowering Reactor Level. The inspectors responded to the control room following site announcement of a loss of condenser vacuum and observed the response of AmerGen personnel to the event, including operator actions in the control room. At the time of the event, the inspectors verified that conditions did not meet the entry criteria for an emergency action level (EAL) as described in the Oyster Creek EAL matrix. In addition, the inspectors reviewed 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, to verify that AmerGen properly notified the NRC during the event. The inspectors also reviewed technical specification requirements to ensure that Oyster Creek operated in accordance with its operating license. This also included a review of Oyster Creeks environmental technical specifications and AmerGens environmental discharge permit NJ0005550 (issued by New Jersey Department of Environmental Protection) due to the impact on the aquatic life (fish) due to the unplanned shutdown. The inspectors reviewed PPC data, control room logs, and discussed the event with AmerGen personnel to gain an understanding of how operations personnel and plant equipment responded during the event. The inspectors evaluated AmerGens program and process associated with event response to ensure they adequately implemented station procedures OP-AA-108-114, Post Transient Review and OP-AA-106-101-1001, Event Response Guidelines. The inspectors also observed the PORC meeting prior to plant startup to evaluate whether AmerGen understood the cause of the event and appropriately resolved issues identified during the event. The inspectors reviewed AmerGens post-trip review report (IR 713652) to gain additional information pertaining to the event, and ensure that human performance and equipment issues were properly evaluated and understood prior to plant startup. No findings of significance were identified. An unresolved item (URI) was identified to review AmerGens corrective action program root cause evaluation (IR 714203) regarding the manual reactor scram on December 19, 2007. The inspectors plan to review this evaluation after it is completed, which had not occurred by the end of this inspection period. (URI 05000219/2007005-04, Loss of A Condenser Vacuum and Trip of A Feedwater Pump Results in a Reactor Scram)  +
05000219/FIN-2008002-01 +The inspectors identified that AmerGen performed an inadequate risk assessment for a planned, but not yet commenced, maintenance activity on the Bank 6 startup transformer in February 2008; which resulted in an under-estimation of the risk associated with performing the activity. This finding was determined not to be a violation of NRC requirements. AmerGens corrective actions for this issue included reassessing the risk for the activity and discussing this issue with work management personnel. The finding was more than minor because the risk assessment did not account for the unavailability of a single train of a system that provides a shutdown key safety function. This finding was also similar to more than minor example 7.e in NRC Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, because when the activity was correctly assessed the plant would have been in a higher, licensee-established risk category. In accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, the inspectors evaluated the significance of this issue and determined that the incremental core damage probability deficit (ICDPD) associated with this activity was less than 1.0 E-6 and noted that the incorrectly assessed maintenance activity did not occur. Therefore, in accordance with Appendix K this finding screened as very low safety significance. The performance deficiency had a cross-cutting aspect in the area of human performance because AmerGen did not plan a maintenance activity, consistent with nuclear safety because risk insights were not properly incorporated into the work planning (H.3(a)). (Section 1R13)  +
05000219/FIN-2008003-01 +A self-revealing finding was identified when AmerGen improperly reassembled the inlet valve actuator on the C & D instrument air dryers which damaged its o-ring and subsequently resulted in an instrument air transient on March 24, 2008. This finding was determined not to be a violation of NRC requirements. AmerGens corrective actions included repairing the air dryer inlet valve by replacing the failed o-ring and providing training on o-ring installation to maintenance personnel. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. In accordance with inspection manual chapter (IMC) 0609.04, Phase 1 Initial Screen and Characterization of Findings, the inspectors conducted a Phase 1 SDP screening and determined that a detailed Phase 2 evaluation was required to assess the safety significance because the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment would not be available. The finding was determined to be of very low safety significance based upon the Phase 2 evaluation. The performance deficiency had a cross-cutting aspect in the area of human performance because training was not adequate to ensure proper reassembly of the valve actuator by maintenance personnel (H.2(b)). (Section 1R12  +
05000219/FIN-2008003-02 +A self revealing finding occurred when the suction air filters to the 1-1 and 1-2 service air compressors became clogged with debris which affected the availability and reliability of the compressors on April 25, 2008. In 2001, AmerGen implemented a modification which involved replacing the service air compressors. During the modification process, AmerGen removed preventive maintenance tasks for the suction air filters without adequate technical justification. AmerGens corrective actions included replacing the inlet air filters, taking action to create Preventive Maintenance (PM) to inspect/replace the air filters and reviewing the extent of condition with respect to similar plant modifications. This finding was of very low safety significance and determined not to be a violation of NRC requirements. The finding was more than minor because it was associated with the equipment performance attribute of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was assessed in accordance with IMC 0609.04, Phase 1 Initial Screen and Characterization of Findings. The inspectors performed a Phase 1 screening and determined that a Phase 2 evaluation was required to assess safety significance because the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment would not be available. A Region 1 senior reactor analyst (SRA) determined that a Phase 2 evaluation was not suited to assess this event. A Phase 3 analysis was performed by the SRA and the finding was determined to be of very low safety significance. The inspectors did not identify a cross cutting aspect for this finding because the performance deficiency had occurred several years ago and is not indicative of current performance  +
05000219/FIN-2008003-03 +The inspectors identified that AmerGen had scheduled surveillance tests in a sequence that would have resulted in unacceptable preconditioning of valves within the core spray system on May 19, 2008. This finding was determined not to be a violation of NRC requirements. AmerGens corrective actions involved reordering the scheduling sequence of the tests and reviewing upcoming (next 60 days) work control schedules to identify potential preconditioning. The finding was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, preconditioning of valves could mask their actual as-found condition and result in an inability to verify their operability, as well as make it difficult to determine whether the valves would perform their intended safety function during an event. In accordance with IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance because it was not a design or qualification deficiency which resulted in a loss of operability or functionality, did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk-significant for greater than 24 hours, and was not potentially risk significant due to a seismic, flooding or severe weather initiating event. The performance deficiency had a cross-cutting aspect in the area of human performance because AmerGen did not appropriately coordinate work activities to support long term equipment reliability (H.3(b)). (Section 1R22  +
05000219/FIN-2008003-04 +A self revealing finding occurred when AmerGen did not properly implement a functional test procedure for the 1-1 diesel driven fire pump on November 7, 2007. Specifically, operations personnel did not accurately measure the speed of the pump while performing the functional test, which resulted in the pump being declared inoperable and unavailable for greater than three weeks during troubleshooting by AmerGen personnel. This finding was of very low safety significance and determined to be a non-cited violation (NCV) of technical specification 6.8, Procedures and Programs. AmerGens corrective actions included providing additional training to operators to accurately monitor speed of the diesel with a stroboscope and revising the procedure to include vendor guidance for measuring diesel speed. The finding was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and affected the objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings, the inspectors conducted a Phase I SDP screening and determined that the finding was of very low safety significance (Green). The finding was of low safety significance because there was no loss of safety function due to the availability of the redundant diesel driven fire pump. The inspectors also reviewed this issue in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, to confirm the above results. The finding was determined to be of very low safety significance (green) because it was assigned a low degradation rating due to availability of other fire protection pumps. The performance deficiency had a cross-cutting aspect in the area of human performance because training was not adequate to ensure the proper use of the stroboscope by operations personnel during testing (H.2 (b)). (Section 4OA2  +
05000219/FIN-2008004-01 +The inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because AmerGen did not properly implement scaffolding control procedural requirements on August 11, 2008. Specifically, AmerGen did not perform engineering evaluations for scaffolding constructed within the minimum allowed distance of safety-related equipment to determine its acceptability. AmerGens corrective actions included modifying or removing scaffold, conducting a briefing on this issue to all scaffold builders and supervisors, and scheduling a second brief for scaffold builders who arrive at Oyster Creek prior to the upcoming refueling outage. This finding was more than minor because it was associated with the external factors attribute of the mitigating systems cornerstone and affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding was also similar to example 4.a in NRC Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, because AmerGen routinely did not perform evaluations for scaffolds constructed within the minimum allowed distance of safety related equipment. In accordance with IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings, the finding was determined to be of very low safety significance because it was not a design or qualification deficiency which resulted in a loss of operability or functionality, did not represent a loss of system safety function, did not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time, did not represent an actual loss of safety function of one or more non-technical specification trains of equipment designated as risk-significant for greater than 24 hours, and was not potentially risk significant due to a seismic, flooding or severe weather initiating event. The performance deficiency had a cross-cutting aspect in the area of human performance because AmerGen did not follow procedures and obtain engineering evaluations for scaffold that did not meet the requirements contained in procedures for scaffold installation in the plant (H.4(b)). (Section 1R15)  +
05000219/FIN-2008004-02 +No findings of significance were identified. An unresolved item (URI) was identified to review AmerGens corrective action program evaluation (IR 815415) regarding the identification of tritium in a September 8, 2008 surface water sample from a puddle of water in an excavation area within the protected area. Based on preliminary reviews and observations, the inspectors did not identify an immediate public health and safety concern. In addition, no indication of contamination of drinking water aquifers was identified by the inspectors. AmerGen performed an initial evaluation to determine the potential source of the tritium in the samples collected. The evaluation included a review of potential active or historical leakage in the area; potential occurrence of a water spill associated with cutting of the torus water storage tank piping located in the excavation area; problems with water sampling (including sample collection, preparation, and counting); and potential washout of tritium from station effluents. To support their evaluation, AmerGen collected groundwater samples from various wells surrounding the area and collected multiple soil samples along the length of the TWST piping, including the location of the puddle. AmerGens evaluation, which continued at the conclusion of this inspection period, indicated the likely cause of the indication of tritium in the water sample was due to an inadvertent cross-contamination of the sample from the use of glassware in the chemistry laboratory. The evaluation also initially concluded that the indication of tritium was not due to a past spill or an active pipe leak, that water had not spilled during cutting of piping in the excavation area, or that rain water washout caused tritium to appear in the sample of the water puddle. AmerGen initiated corrective actions to improve sample preparation and counting, including detection of potential sample cross-contamination. The inspectors plan to review this evaluation after it is completed. (URI 05000219/2008004-02, Water with Tritium Identified in Excavation Area within the Protected Area)  +
05000219/FIN-2008004-03 +10 CFR 50.55a, Codes and Standards, requires that systems and components of boiling and pressurized water-cooled nuclear power reactors shall meet the requirements of the ASME Boiler and Pressure Vessel Code. Contrary to this, AmerGen did not perform an accelerated IST of the 1-1 service water pump in August 2008, as required by the ASME code when the pumps performance was evaluated to be in the alert range during its quarterly IST on June 19, 2008. The following quarterly IST on September 16, 2008 showed the pumps performance to be in the action range. The increased frequency testing was not performed because the work order to perform the testing was not properly coded in the work management system. AmerGen performed a technical evaluation and determined that the cause of the degradation in the pumps performance was due to a system degradation, specifically system leakage, and verified that the pump was operating satisfactorily and was not degraded. The violation was of very low safety significance because the deficiency was a qualification issue that did not result in the loss of operability of the system or component. The issue is described in corrective action program condition reports IR 820700 and CR 787909. Corrective actions involved AmerGen performing an extent of condition review on work orders involving surveillance testing and ISTs to ensure they were properly coded and scheduled correctly.  +
05000219/FIN-2008005-01 +A self-revealing non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, occurred when Exelon did not perform an adequate self-check and did not properly use test equipment during 480 VAC breaker maintenance on November 7. Specifically, during the maintenance, a human performance error occurred causing a phase to phase fault and an arc flash, and resulted in the loss of safety related equipment and an automatic halon system actuation in the 480 VAC room. In response, Exelon entered this issue into the corrective action program and implemented actions to address work practice deficiencies. The finding is more than minor because it is associated with the human performance attribute of the initiating events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Appendix G, Shutdown Operations Significance Determination Process, of Manual Chapter 0609, Significance Determination Process, the finding was determined to have very low safety significance (Green) because it did not increase the likelihood of a loss of reactor coolant system (RCS) inventory, did not affect the licensees ability to terminate a leak path or add inventory to the RCS, or degrade the licensees ability to recover decay heat removal in the event it was lost. The performance deficiency had a cross-cutting aspect in the area of human performance because Exelon did not properly implement human error prevention techniques, such as self and peer checking (H.2(c)). (Section 1R12  +
05000219/FIN-2008005-02 +The inspectors identified an NCV of Technical Specification 3.9.D Refueling, when Exelon performed core alterations without the required configuration of operable source range monitors (SRM). Specifically, Exelon installed two fuel assemblies in a reactor quadrant when the required configuration of SRMs was not operable. In response, Exelon entered this issue into the corrective action program and implemented actions to revise the reactor refueling procedure. The finding is more than minor because it is associated with the configuration control attribute of the barrier integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, during a time of decreased availability of physical barriers (refueling outage), Exelon performed core alterations without the required configuration of operable SRMs. Using Appendix G, Shutdown Operations Significance Determination Process, of Manual Chapter 0609, Significance Determination Process, the finding was determined to have very low safety significance (Green) because it did not increase the likelihood of a loss of reactor coolant system (RCS) inventory, did not affect the licensees ability to terminate a leak path or add inventory to the RCS, or degrade the licensees ability to recover decay heat removal in the event it was lost. The performance deficiency had a cross-cutting aspect in the area of human performance, because Exelon did not ensure that the reactor refueling procedures accurately implemented the neutron monitoring requirements contained in the Technical Specifications (H.2(c)). (Section 1R20  +
05000219/FIN-2008005-03 +An unresolved item (URI) was identified because additional information and a review of system sampling capabilities is required to determine if there are any impacts to Oyster Creeks emergency plan and to determine if any performance deficiency exists. The inspectors plan to review the additional information after it is complied by Exelon, which had not occurred by the end of this inspection period. As a result of the difficulties experienced with the in-line auto filter change apparatus, Exelon has initiated action to replace the system with an updated model. In the interim, the installed conventional direct in-line particulate and iodine sampling system was placed in service to implement the sampling provisions of Oyster Creeks Offsite Dose Calculation Manual (ODCM). The sampling system continues to autoadjust flow rates to ensure proper collection of isokinetic samples. A separate sampling apparatus provides for sampling of tritium. A review of the availability of the system indicates the system has met requirements for sampling in accordance with the ODCM and has maintained high reliability. At the time of this inspection, the inspectors had not completed a review of potential emergency response implications associated with collection and analysis of particulate and iodine samples during accident conditions, relative to the changes in sample collection methodology (i.e., inability to use the autofilter change-out capability). Exelon is providing additional information regarding collecting samples using the installed in-line particulate and iodine sampling system. Inspectors will review the additional information against the requirements of the emergency plan to determine if a performance deficiency exists. (URI 05000219/2008005-03: Stack Radiation Monitoring System Sampling Capabilities  +
05000219/FIN-2008005-04 +An unresolved item was identified to review Exelons root cause assessment and licensee event report (LER) regarding the failure of the M1A main transformer and subsequent load reject scram to determine whether a performance deficiency existed which contributed to the transformer failure. The inspectors plan to review Exelons evaluation after it is completed, which had not occurred by the end of this inspection period. At 2101, November 28, Oyster Creek experienced a generator trip due to an A-phase and B-phase differential relay actuation, which resulted in a reactor shutdown due to a load reject scram. All safety systems operated as expected during the scram. The grid disturbance report provided by Jersey Central Power & Light, combined with information from the Oyster Creek Digital Protective Relay System, differential voltage and current indication data, and dissolved gas in oil analysis indicated that the fault occurred on the B phase of the M1A Main Power Transformer. Exelon entered this issue into their corrective action program in condition report IR 850348. (URI 05000219/2008004-04: Failure of M1A Transformer Causes an Automatic Load Reject Scram  +
05000219/FIN-2008005-05 +TS 6.8.1, Procedures and Programs, requires that written procedures be established and implemented covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33 as referenced in the Quality Assurance Topical Report (QATR). Regulatory Guide 1.33, Rev.2, as referenced in the QATR, recommends procedures for chemical and radiochemical control including validity of calibration techniques and adequacy of analyses. Contrary to this, the licensee modified its tritium analysis method to achieve an improved environmental lower limit of detection and did not follow its method development process outlined in station procedure CY-AA-130-200, Rev. Section 4.1, for analysis of tritium using the method described in procedure CY-OC-130-530, Rev.4, including completion of Attachment 6 of the procedure. This was identified in the licensees corrective action program as IR 832750. This finding is of very low safety significance because it involves the area of environmental monitoring and the radiological environmental monitoring program did not identify unexpected conditions in the environment  +
05000219/FIN-2008007-01 +As noted in the detailed observations of this report, a number of issues were observed which Exelon placed into its corrective action program. The specific issues for further review include (1) Exelon applied a strippable coating to the refuel cavity liner to prevent water intrusion into the gap between the drywell steel shell and the concrete shield wall. The strippable coating unexpectedly de-laminated, resulting in increased refuel cavity seal leakage. As a result, water entered the gap and subsequently flowed down the outside of the shell and into four sand bed bays. In addition, Exelon had established an administrative limit for cavity seal leakage that was higher than the actual leakage rate at which water intrusion into the gap occurred. (Sections 3.1 & 3.5) (2) While the reactor cavity was being filled, Exelon frequently monitored the cavity seal leakage by observing flow in the cavity trough drain line. Subsequently, Exelon determined that the trough drain line had been left isolated during a previous maintenance activity. As a result, cavity seal leakage had not been monitored as intended. (Section 3.2) (3) During the refueling outage, Exelon monitored for water leakage from the sand bed bay drains by checking poly bottles connected via tygon tubing and funnels to the sand bed drain lines. Exelon subsequently discovered that the poly bottle tubing was not connected to the drain lines for two sand bed bays. (Section 3.4) (4) Exelon identified four blisters on the epoxy coating in one sand bed bay. Exelon\'s evaluation to determine the cause of the blisters was still in-progress at the time this inspection was completed. In addition, a video recording from 2006 appeared to indicate that one of the blisters existed at that time, but was not identified during Exelon\'s 2006 visual inspection. (Section 3.9) The inspectors will review these issues in a future inspection to determine whether the individual issues are acceptable or constitute a CLB performance deficiency. The inspectors\' assessment will, in part, determine whether these items are consistent with design specifications and requirements, the conduct of operations, and whether appropriate administrative controls were utilized. (URI 05000219/2008007-01: Drywell Sand Bed Water Intrusion, Drain Monitoring, and Coating Deficiency)  +
05000219/FIN-2008008-01 +The team identified that in July 2002, AmerGen failed to review a change to personnel resources that would increase the time necessary to complete an NRC approved hot shutdown repair after a fire in the A 480V switchgear room. Specifically, AmerGen eliminated the need for onsite electrical or instrument and controls technician staffing at all times. This finding was determined to be of very low safety significance (Green) and a NCV of Oyster Creek Nuclear Generating Station Facility Operating License condition 2.C.(3) Fire Protection. AmerGens immediate corrective actions for this issue included assessing current call-in processes to verify the hot shutdown repair would be completed by qualified personnel within the safe shutdown analysis time requirement. The team determined that this finding was more than minor because it was associated with the external factors attribute (fire) of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, AmerGen did not analyze the reduction in personnel readiness for an adverse impact on implementing a hot shutdown repair to Bus USS 1B2 within the safe shutdown analysis time requirement. This finding was also similar to more than minor example 3.i in NRC Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues. The team assessed this finding in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process. This finding screened to very low safety significance (Green) in phase 1 of the SDP because it was assigned a low degradation rating. A low degradation rating was assigned because actual emergency response organization call-in and drive-in data demonstrated that the hot shutdown repair would most likely be completed within the safe shutdown analysis time requirement. (Section 1R05.01  +
05000219/FIN-2009002-01 +The inspectors identified a non-cited violation (NCV) of 10CFR50.54(q), Conditions of Licenses, because Exelon did not properly maintain the conditions of the Oyster Creek Emergency Plan. Specifically, Exelon did not implement timely corrective or compensatory actions when the radioactive gas effluent monitoring system (RAGEMS) automatic sampling system was taken out of service from November 2006 through March 2009. Exelons corrective actions included replacing solenoid valves in the automatic sampling system and placing the automatic system back in service. The finding was more than minor because it affected the Emergency Response Organization Performance attribute of the Emergency Preparedness (EP) Cornerstone to ensure that the licensee is capable of implementing adequate measures to protect the public health and safety of the public in the event of a radiological emergency. In accordance with Inspection Manual Chapter (IMC) 0609, Appendix B, Emergency Preparedness Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green). Specifically, the inspectors utilized IMC 0609, Appendix B, Section 4.9 and Sheet 1, Failure to Comply, to determine that the failure to satisfy 10 CFR 50.47(b)(9) was a risk-significant planning standard (RSPS) problem; but it was not a RSPS functional failure of the Oyster Creek dose assessment process. Because a time-motion study concluded that a manual iodine and particulate sample could have been obtained under accident conditions without exceeding regulatory dose limits, the inspectors determined that the RSPS function had not been degraded and the failure of the automatic sampling system ultimately would not have affected the outcome of protecting the health and safety of the public. The performance deficiency had a cross-cutting aspect in the area of problem identification and resolution, because Exelon did not take appropriate corrective actions in a timely manner commensurate with its safety significance and complexity. Specifically, the RAGEMS sampling system was not able to satisfy the functions required by the Oyster Creek Emergency Plan for over two years before Exelon took adequate steps to initiate corrective actions P.1(d)  +