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	Quarter	Site	Title	Description
05000528/FIN-2010008-01	2010Q4	Palo Verde	Failure to Correct and Prevent Recurrence of a Significant Condition Adverse to Quality Associated with the Emergency Diesel Generator Fuel Oil Transfer Pumps	10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. In the case of significant conditions adverse to quality, the measures shall assure that the cause of the condition is determined and corrective action taken to preclude repetition. Contrary to the above, from April 2009 through September 2010, the licensee failed to correct a significant condition adverse to quality and implement adequate corrective actions to preclude repetition. Specifically, the licensee failed to correct a water intrusion path to the Unit 2 motor termination boxes for the emergency diesel generator fuel oil transfer pumps, resulting in degraded electrical connections and a pump trip.
05000482/FIN-2010006-09	2010Q3	Wolf Creek	Failure to Translate Design Information into a Procedure	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to translate criteria from the atmospheric relief valve accumulator leakage calculation into proceduralized leakage criteria. Specifically, engineering personnel did not translate the calculated design basis leakage criteria and the required minimum pressure to start the test into the procedure. The licensee entered this in to the corrective action program as Condition Report 26771, and the licensee was developing plans to revise the leakage criteria in the procedure. This issue was more than minor because it affected the design control 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. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because licensee personnel failed to take appropriate corrective actions to previously identified problems (P.1(d))
05000482/FIN-2010006-08	2010Q3	Wolf Creek	Notice of Unusual Event Due to Loss of Both Emergency Diesel Generators	The inspectors reviewed a self-revealing noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to identify a degraded equipment condition in December 2006. As a result, the emergency diesel generator system experienced a failure on October 22, 2009, which caused the plant to make a notice of unusual event emergency declaration. Licensee personnel missed an opportunity to identify the condition because they did not thoroughly evaluate a surveillance failure and post-mortem testing data available in December 2006. The finding is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. A crosscutting aspect was identified in the problem identification and resolution in that the licensee did not thoroughly evaluate problems such that the resolution addressed causes (P.1(c))
05000482/FIN-2010006-07	2010Q3	Wolf Creek	Failure to Determine if a Deficiency Existed in the Ultimate Heat Sink	The team identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to follow the requirements of Procedure AP 26C-004, Technical Specification Operability, Revision 20. Specifically, Wolf Creek Generating Station failed to confirm if a deficiency existed with the ability of the ultimate heat sink to perform its safety function after delaying the 5-year scheduled dredging of the channel. The licensee initiated Condition Report 27080 and performed an operability determination to evaluate the deficiency. The issue was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of problem identification and resolution associated with the corrective action program component because licensee personnel failed to identify a potential deficiency in the ultimate heat sink in a timely manner (P.1(a))
05000482/FIN-2010006-05	2010Q3	Wolf Creek	Failure to Perform Adequate Evaluation for Significant Conditions	The inspectors identified a cited violation 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because the licensee failed to perform an adequate evaluation to determine the cause of loss of offsite power induced water hammers and internal corrosion in the essential service water system and did not take corrective actions to preclude repetition of additional water hammer events and system leaks. Specifically, the licensee performed an apparent cause evaluation instead of a root cause evaluation as required, and the licensees evaluation did not consider metallurgical evaluations that were performed outside the corrective action program. The inspectors found that the licensee had not corrected a previous NCV 05000482/2009007-03, Failure to Correctly Screen ESW Piping Leaks for Significance, which resulted in the licensee failing to perform a root cause evaluation. Because the licensee failed to restore compliance within a reasonable time after NCV 05000482/2009007-03 was identified, this violation is being cited in a Notice of Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy. The licensees corrective action to this cited violation was to initiate Condition Reports 27212, 26466, and 27075, to evaluate and correct the identified conditions, to start a root cause evaluation and, separately, to evaluate the licensees failure to properly respond to NCV 05000482/2009007-03. The issue was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and is therefore a finding. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the component of corrective action program because the licensee failed to thoroughly evaluate problems such that the resolutions address causes and extent of conditions (P.1(c))
05000482/FIN-2010006-04	2010Q3	Wolf Creek	Failure to Update an Operability Evaluation	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure to implement Procedure AP 26C-004, Technical Specification Operability, Revision 20, to adequately evaluate the operability of a degraded essential service water system. Specifically, operations and engineering personnel failed to adequately evaluate the operability of the essential service water system when relevant new information was identified that challenged a previously performed operability determination and which challenged the reasonable expectation for operability. Condition Report 27288 was initiated to evaluate the failure to perform adequate operability determinations. The issue was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding is associated with the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of human performance associated with resources because the licensee failed to provide complete, accurate, and up-to-date procedures for performing operability evaluations (H.2(c))
05000482/FIN-2010006-03	2010Q3	Wolf Creek	Failure to Adequately Monitor Control Room Deficiencies	The inspectors identified a finding for the failure to follow Procedure AI 22A-001, Operator Work Arounds/Burdens/Control Room Deficiencies, Revision 8, to adequately identify, document, and track control room deficiencies associated with instruments and controls to ensure proper prioritization and timely corrective actions. Specifically, inspectors observed that the licensee had approximately 52 WR (work request) buttons on the control boards indicating that work requests had been initiated to correct problems on instruments and controls. However, not all deficiencies were logged, and some of the deficiencies had existed for years without correction or justification. The licensee initiated Condition Report 27034 to document and evaluate this concern. The deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern, in that, the deficient condition could cause an operator to take an inappropriate action based on expected plant response or conversely cause an operator not to take action when action is required. The finding is associated with the Mitigating Systems Cornerstone. The senior reactor analyst determined that this finding was not appropriate to be evaluated using the significance determination process since this finding was associated with numerous equipment issues and associated human performance aspects that might impact equipment operation. Using Inspection Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, the finding is determined to have very low safety significance because there was no adverse impact to plant equipment. The inspectors determined that the cause of the finding has a crosscutting aspect in the area of problem identification and resolution associated with the component of corrective action program because the licensee did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance (P.1(a))
05000482/FIN-2010006-02	2010Q3	Wolf Creek	Unqualified Scaffolding Erected Near Safety-Related Equipment	The inspectors identified a noncited violation of Technical Specification 5.4.1.a for failure to properly implement Procedure AP 14A-003, Scaffold Construction and Use, Revision 17, when scaffolding was erected near operable safety-related equipment. On July 14, 15, and 28, the inspectors identified a total of four instances where the minimum separation distance between scaffolding and safety-related components was less than the minimum allowed by procedure and an approved engineering evaluation to justify the deviation was not performed. The licensee entered the issue into its corrective action program as Condition Reports 26752 and 27010, corrected each scaffolding deficiency, and performed comprehensive walkdowns of all scaffolding around safety-related structures, systems, and components. The deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding was associated with the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined the finding has a crosscutting aspect in the area of problem identification and resolution associated with corrective action program because the licensee did not take appropriate corrective actions to address previously identified scaffolding construction issues in a timely manner (P.1(d)).
05000482/FIN-2010006-01	2010Q3	Wolf Creek	Failure to Resolve Degraded Conditions in a Timely Manner	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly correct degraded or nonconforming conditions in that the conditions were not corrected at the first available opportunity or appropriately justify a longer completion schedule. Some examples of affected degraded or nonconforming conditions included degraded atmospheric relief valve discharge line silencer, essential service water system water hammer events and internal corrosion, and 23 items on the Operability Evaluation Database that had not been corrected prior to the start of the last refuel outage. As corrective actions for this issue, the licensee implemented interim procedural guidance and initiated Condition Report 27071 to evaluate the adequacy of tracking methods used for degraded, nonconforming, or unanalyzed conditions. In addition, the licensee initiated a review of work requests, condition reports, and other items for degraded, nonconforming, or unanalyzed conditions and is assessing the justification for delayed implementation of these corrective actions. This issue was more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, the issue is determined to have very low safety significance because the finding is not a design or qualification issue confirmed not to result in a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors determined that the finding has a crosscutting aspect in the area of human performance associated with the component of resources because the licensee failed to provide adequate procedures to assure timely resolution of degraded or nonconforming conditions (H.2(c)).
05000361/FIN-2009009-01	2010Q1	San Onofre	Failure to Follow a Level 1 Quality Assurance Program Affecting Human Performance Procedure	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, \"Instructions, Procedures, and Drawings,\" for the failure of training personnel to ensure activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, between September 27, 2009 and November 17, 2009, training personnel failed to follow Level 1 Quality Assurance Program Affecting Procedure SO123-XXI-1.11.23, \"Human Performance Training Program Description,\" Revision 0, to ensure workers received human performance training before hands-on work was performed in the plant, which resulted in over 80 employees not receiving human performance training and contributed to at least two human performance events. This finding was entered into the licensees corrective action program as Nuclear Notification NN 200670169. The finding is greater than minor because, if left uncorrected, the failure to follow procedures to provide human performance training, would have the potential to lead to more significant safety concerns as is evidenced by the two human performance events that occurred by untrained individuals. This finding is associated with the Initiating Events Cornerstone. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding is determined to have very low safety significance because the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity (P.1(d)) (Section 4OA2).
05000361/FIN-2009009-02	2010Q1	San Onofre	Failure to Maintaing Written Procedures Covered in Regulatory Guide 1.33	The inspectors identified a non-cited violation of Technical Specification 5.5.1, \"Procedures,\" for the failure of procedure writer personnel to maintain written procedures covered in Regulatory Guide 1.33. Specifically, from initial plant startup of Units 2 and 3 to November 2009, no process requirement or procedure existed to identify procedures that required technical changes so that those procedures could be suspended or put an administrative hold until the required changes were made. This resulted in a quality controlled procedure requiring technical changes available to use on a safety-related system without flagging the required changes. This finding was entered into the licensees corrective action program as Nuclear Notification NN 200671179. The finding is greater than minor because, if left uncorrected, the failure to maintain and control procedures would have the potential to lead to a more significant safety concern by having technically inaccurate procedures being used on safety-related systems. This finding is associated with the Mitigating Systems Cornerstone. Using Manual Chapter 0609.04, \"Phase 1 Initial Screening and Characterization of Findings,\" the finding was determined to have a very low safety significance because the finding did not result in a loss of a system safety function, an actual loss of safety function of a single train for greater than its technical specification allowed outage time, or screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program because problems were not thoroughly evaluated such that the resolutions addressed the causes and extent of conditions. This includes properly classifying and prioritizing conditions adverse to quality (P.1(c)) (Section 4OA2).
05000298/FIN-2009005-17	2009Q4	Cooper	Licensee-Identified Violation	Technical Specification Limiting Condition for Operation 3.10.4 requires, in part, that to allow withdrawal of a single control rod with the reactor in Mode 4, all other control rods in a five by five array centered on the control rod being withdrawn are disarmed. Condition B.2.2 requires when a limiting condition for operation is not met with the affected control rod not insertable to immediately initiate actions to satisfy the requirements of this limiting condition for operation. Contrary to the above, on November 1, 2009, the licensee discovered that the control rods in the five by five array around a withdrawn control rod were not disarmed for over two hours without immediately taking the actions required by technical specification action statement B.2.2. This was documented in the licensees corrective action program by CR-CNS-2009-9138. Because the plant was shutdown at the time this performance deficiency occurred, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Checklist 7 in Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, the inspectors determined that the finding had very low safety significance because it did not require quantitative assessment for a phase 2 or 3 analysis
05000298/FIN-2009005-14	2009Q4	Cooper	Licensee-Identified Violation	10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, that activities affecting quality shall be accomplished in accordance with procedures that are appropriate for the circumstances. Contrary to this requirement, on October 4, 2009, the licensed operators failed to follow the guidance of Section 7.3 of Administrative Procedure 0.40, Work Control Program, Revision 68. Specifically, the licensed operators performed Surveillance Procedure 6.1RPS.313, RPS Channel Test Switch Functional Test (Div 1), instead of the scheduled Surveillance Procedure 6.2RPS.313, RPS Channel Test Switch Functional Test (Div 2). This performance deficiency was discovered by licensed operators during closeout of the work order and was documented in CR-CNS-2009-07618. This event demonstrated failure to effectively use error prevention tools. Specifically, the licensees two minute drill card specifically challenges workers to ensure they are working on the right division. Despite continued emphasis on human error prevention, the entire watchteam agreed to perform a surveillance test on the wrong division. The inspectors determined that this issue was of very low safety significance because no loss of system safety function resulted from the performance deficiency
05000298/FIN-2009005-15	2009Q4	Cooper	Licensee-Identified Violation	10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, that activities affecting quality shall be accomplished in accordance with procedures that are appropriate for the circumstances. Contrary to this requirement, during preparation for refueling outage 25, the plant staff failed to follow the guidance of Section 5 of Administrative Procedure 0.50.5, Outage Shutdown Safety, Revision 12. Specifically, the licensed operators failed to manage the risks associated with an operation with the potential to drain the reactor vessel. This performance deficiency was discovered on September 27, 2009, while shut down in Mode 4, by a control room operator who noted reactor vessel water level lowering and was documented in CR-CNS-2009-07191. Reactor vessel water level control was lost for five minutes when an inadvertent drain path was established lowering vessel level four inches prior to restoring a positive rising level. The inspectors determined that this issue was of very low safety significance because no loss of system safety function resulted from the performance deficiency
05000298/FIN-2009005-16	2009Q4	Cooper	Licensee-Identified Violation	10 CFR Part 50.72(b)(3)(v)(B) requires that any condition resulting in a loss of the residual heat removal safety function be reported to the NRC as soon as practical and in all cases within eight hours of the occurrence. Contrary to this requirement, on November 7, 2009, a human performance error resulted in an automatic isolation of the shutdown cooling system and a loss of the residual heat removal safety function and this loss of safety function was not reported as required. The licensee discovered this missed report during management review of the event on November 9, 2009 and identified the performance deficiency in CR-CNS-2009-09537. The inspectors determined that this issue is consistent with the examples of a SLIV violation in Supplement I, paragraph D.4 of the Enforcement Policy
05000298/FIN-2009005-13	2009Q4	Cooper	Procedure Noncompliance Causes Fire in Heater Bay	A self-revealing noncited violation of Technical Specification 5.4.1.d, Fire Protection Program Implementation, was identified for the licensees failure to follow Administrative Procedure 0.39, Hot Work. Specifically, contractors under the licensees control failed to consider weld pre-heating as an activity requiring hot work controls, and as such did not take the appropriate precautions for a pre-heating activity. As a result, a degraded pre-heating blanket failed in service, started a fire in the heater bay and resulted in declaration of a Notice of Unusual Event. The licensee entered this issue in their corrective action program as CR-CNS-2009-08061. The performance deficiency associated with this finding involved the licensees failure to follow the requirements of Administrative Procedure 0.39, Hot Work. Specifically, contractors performing work in the turbine building heater bay failed to consider weld pre-heating as an activity requiring hot work controls and did not take the appropriate precautions for the pre-heating activity. The finding is more than minor because it affected the external events aspect of the initiating events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, could not be applied to shutdown plant conditions. Because the plant was shutdown at the time this performance deficiency occurred, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Checklist 7 in Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, the inspectors determined that the finding had very low safety significance because every item on the checklist was met. This finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee personnel failed to maintain adequate supervisory control over contractors performing welding in the turbine building heater bay (H.4(c)
05000298/FIN-2009005-12	2009Q4	Cooper	Failure to Follow Procedure For Control of Material	A self-revealing finding was identified for the licensees failure to follow Administrative Procedure 0.47, Control of In-Process Material, Specifically, a maintenance technician violated the procedure by obtaining a spare o-ring from an uncontrolled toolbox and that o-ring was then installed in the Main Turbine Control Valve 3 hydraulic fitting. The o-ring was the wrong size and caused a hydraulic leak that required taking the turbine off line and shutting down the reactor from 70 percent power. The licensee entered this issue in their corrective action program as CR-CNS-2009-09606. The finding is more than minor because it adversely affected the configuration control attribute of the initiating events cornerstone, and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, in that this finding resulted in a condition that prompted a plant shutdown from 70 percent power. In accordance with Manual Chapter 0609, Attachment 4, the inspectors used the Phase 1 Initial Screening and Characterization worksheet to determine that the finding has very low safety significance because it did not result in the loss of any system safety function. The cause of this finding is related to human performance cross cutting component of work practices because the involved maintenance personnel proceeded in the face of uncertainty when obtaining replacement o-rings (H.4(a)
05000298/FIN-2009005-11	2009Q4	Cooper	Failure to Preclude Repetition of Loss of Shutdown Cooling	A self-revealing noncited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to preclude repetition of a significant condition adverse to quality, namely the loss of shutdown cooling caused by drawing a vacuum in the reactor pressure vessel. Specifically, corrective actions taken after a March 17, 1994, loss of shutdown cooling event were inadequate to prevent a similar event from occurring on November 7, 2009. The licensee entered this issue in their corrective action program as CR-CNS-2009-09486. The finding is more than minor because it affected the procedure quality attribute of the mitigating systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors determined that Manual Chapter 0609, Appendix G was applicable due to the fact that at the time of the performance deficiency was discovered, the plant was in a forced outage with residual heat removal system in service. Using Checklist 8 in Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, the inspectors determined that although the residual heat removal mitigation capability on the checklist was not met, the criteria for requiring a phase 2 or phase 3 analysis were not satisfied. The inspectors determined that no cross cutting aspects were appropriate for this finding due to the fact that the performance deficiency occurred in 1994 and is not reflective of current performanc
05000298/FIN-2009005-10	2009Q4	Cooper	Failure to Correct Diesel Generator 2 Oil Leakage	A self-revealing noncited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, occurred for the licensees failure to assure that a condition adverse to quality was corrected. Specifically, the licensee identified oil leakage on Diesel Generator 2 mechanical overspeed governor drive flange as a condition adverse to quality on June 23, 2009, and failed to correct the condition of oil leakage as demonstrated by a September 9, 2009, failure of the Diesel Generator 2 due to loose fasteners at this location. The licensee entered this issue in their corrective action program as CR-CNS-2009-06716. The finding is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Using the screening worksheet in Manual Chapter 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined that the finding has very low safety significance because it was not a design or qualification deficiency and did not result in the loss of any system safety function. This finding has a crosscutting aspect in the corrective action program component of the Problem Identification and Resolution area because the licensees periodic trends and assessments did not identify programmatic and common cause problems, in that the licensees periodic trends and assessments did not recognize the significance of precursor events related to fasteners loosening and prompt action to prevent further problems on the emergency diesel generators (P.1(b)
05000298/FIN-2009005-09	2009Q4	Cooper	Failure to Follow Radiation Work Permit Requirements in Two Instances	The inspectors reviewed a self-revealing, noncited violation of Technical Specifications 5.4.1 involving two examples of a failure to follow Radiation Work Permit requirements. In the first example, workers were not monitored with telemetry and constant coverage by a radiation protection technician was not provided as required by the radiation work permit. In the second example, a worker was not monitored with telemetry as required by the special work permit. As a result, the licensee conducted a stand-down to reinforce expectations for compliance with radiation work permits, instituted management challenges at the access control point, and began conducting an apparent cause evaluation. This was entered into the licensees corrective action program as Condition Report CR-CNS-2009-08197 and CR-CNS-2009-08623. The inspectors determined that the failure to meet radiation and special work permit requirements was a performance deficiency. The finding is more than minor because it involved multiple failures of radiation protection measures which, if left uncorrected, could become a more significant safety concern. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined this finding had very low safety significance because the finding involved an ALARA planning and work controls and the licensees average collective dose is less than 240 person-rem per unit. The finding had a crosscutting aspect in the area of human performance associated with work practices because of the lack of self and peer checking to ensure work activities were performed safely (H.4(a)
05000298/FIN-2009005-08	2009Q4	Cooper	Failure to Establish an Adequate Procedure to Ensure Constant Communications in a Locked High Radiation Area	The inspectors identified a noncited violation of Technical Specifications 5.4.1 for a failure to establish a procedure with adequate provisions to control work inside a locked high radiation area. Specifically, although the licensees procedure required constant communications with workers in a locked high radiation area, the procedure had no provisions for providing a reasonable assurance that constant communications was being maintained during the duration the workers were inside the area. As a result, on October 6, 2009, the licensee lost constant communications with workers inside a locked high radiation area when the workers unknowingly bumped the cell phone and de-energized it. The licensees immediate corrective action was to lock the keyboard on the cell phones to prevent them from inadvertently being turned off. The licensee entered the finding into the corrective action program as Condition Report CR-CNS-2009-07718. The inspectors determined that the failure of licensee procedures to contain adequate provisions that work inside a locked high radiation area would be controlled through constant communications is a performance deficiency. The finding was more than minor because, if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Using the Occupational Radiation Safety Significance Determination Process the inspectors determined this finding had very low safety significance because the finding did not involve ALARA planning and work controls, did not result in an overexposure, did not involve a substantial potential for overexposure, and did not compromise the licensees ability to assess dose. Additionally, the finding had a crosscutting aspect in the area of human performance, resources component, because the licensee failed to ensure that equipment used to control work inside a posted locked high radiation area was adequate for environment and working conditions (H.2(d)
05000298/FIN-2009005-07	2009Q4	Cooper	Procedure Violation Results in Loss of Fuel Pool Cooling	A self-revealing noncited violation of Technical Specification 5.4.1.a was identified regarding the licensees failure to follow the requirements of System Operating Procedure 2.2.18, 4160V Auxiliary Power Distribution System. Specifically, operators preparing the 4160 F bus for a maintenance outage secured the wrong fuel pool cooling pump. When the bus was subsequently deenergized, a loss of fuel pool cooling occurred. The licensee entered this issue in their corrective action program as CR-CNS-2009-07770. The finding is more than minor because it is associated with barrier integrity cornerstone attribute of configuration control, and adversely affected the cornerstone objective of maintaining functionality of the spent fuel pool cooling system to provide reasonable assurance that the fuel cladding physical design barrier protects the public from radionuclide releases caused by accidents or events. Because the plant was shutdown at the time this performance deficiency occurred, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Checklist 7 in Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, the inspectors determined that the finding had very low safety significance because every item on the checklist was met. The finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to effectively use required self-checking error prevention tools (H.4(a)
05000298/FIN-2009005-06	2009Q4	Cooper	Maintenance Error Results in Recirculation Pump Trip	The inspectors identified a finding for the licensees failure to follow the requirements of Administrative Procedure 0.40, Work Control Program, Revision 68. Specifically, a maintenance technician violated the procedure by attempting corrective maintenance on the Reactor Recirculation Motor Generator A lubricating oil system without notifying the control room, resulting in a trip of the motor generator and the supported reactor recirculating pump. The licensee entered this issue in their corrective action program as CR-CNS-2009-09023. The performance deficiency associated with this finding was the licensees failure to follow the requirements of Administrative Procedure 0.40, Work Control Program, on October 29, 2009. The finding is more than minor because it adversely affected the configuration control attribute of the initiating events cornerstone, and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Because the plant was shutdown at the time this performance deficiency occurred, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Checklist 7 in Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, the inspectors determined that the finding had very low safety significance because every item on the checklist was met. The finding has a crosscutting aspect in the area of human performance associated with work practices because the licensees maintenance technician did not use the procedurally-required Stop-Think-Act-Review step (error prevention tool) which would have required him to ensure that all energy had been removed from the recirculation motor generator oil system prior to performing maintenance on the system (H.4(a)
05000298/FIN-2009005-05	2009Q4	Cooper	Failure to Identify Foreign Material in the Reactor Core	A self-revealing noncited violation of Technical Specification 5.4.1.a was identified regarding the licensees failure to follow the requirements of System Operating Procedure 2.2.18, 4160V Auxiliary Power Distribution System. Specifically, operators preparing the 4160 F bus for a maintenance outage secured the wrong fuel pool cooling pump. When the bus was subsequently deenergized, a loss of fuel pool cooling occurred. The licensee entered this issue in their corrective action program as CR-CNS-2009-07770. The finding is more than minor because it is associated with barrier integrity cornerstone attribute of configuration control, and adversely affected the cornerstone objective of maintaining functionality of the spent fuel pool cooling system to provide reasonable assurance that the fuel cladding physical design barrier protects the public from radionuclide releases caused by accidents or events. Because the plant was shutdown at the time this performance deficiency occurred, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Checklist 7 in Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, the inspectors determined that the finding had very low safety significance because every item on the checklist was met. The finding has a crosscutting aspect in the area of human performance associated with work practices because the licensee failed to effectively use required self-checking error prevention tools (H.4(a)
05000298/FIN-2009005-04	2009Q4	Cooper	Failure to Implement a Prescribed Risk Mitigating Action	The inspectors identified a noncited violation of 10 CFR 50.65.a (4) for the licensees failure to manage the increase in risk that may result from proposed maintenance activities. Specifically, inspectors discovered that after the licensee had designated Core Spray Pump B as protected in accordance with Administrative Procedure 0-PROTECT-EQP, Protected Equipment Program, the licensee removed the protected core spray pump from service for a maintenance activity. The licensee entered this issue in their corrective action program as CR-CNS-2009-09243. The performance deficiency associated with this finding involved the licensees failure implement prescribed risk mitigating actions. Specifically, inspectors discovered that a protected train core spray pump had been made unavailable for a maintenance activity. The finding is more than minor because the licensee failed to implement a prescribed significant compensatory measure. A senior reactor analyst assisted with the significance determination process. For this finding, the analyst used the guidance in NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The analyst determined that the finding associated with an inoperable core spray pump, while that pump was specified as protected equipment, screened as having very low safety significance in both the Appendix K and Appendix G significance determination processes. This finding has a crosscutting aspect in the area of human performance associated work practices because operations personnel failed to follow the procedural requirements of Administrative Procedure 0-PROTECT-EQP (H.4(b)
05000298/FIN-2009005-03	2009Q4	Cooper	Failure to Set Goals and Monitoring for the Diesel Generator Lubricating Oil System	The inspectors identified a noncited violation of 10 CFR 50.65(a)(1) for the failure to monitor the performance of the diesel generator lubricating oil system against licensee-established goals in a manner sufficient to provide reasonable assurance that the diesel generator lubricating oil system was capable of fulfilling its intended safety functions. Specifically, a revision to the root cause investigation report for a diesel generator 2 lubricating oil pipe crack failure resulted in an undetected repeat maintenance preventable functional failure that required an automatic (a) (1) status of the associated maintenance rule function. Although the diesel generator system was already in (a) (1) status for other reasons, the appropriateness of the existing goals required evaluation under 10 CFR 50.65(a) (1). The licensee entered this issue in their corrective action program as Condition Report CR-CNS-2009-06392 and determined it was appropriate to establish and monitor an additional goal for the emergency diesel generator lubricating oil system. This finding is more than minor because it affected the reliability objective of the equipment performance attribute under the mitigating systems cornerstone. The inspectors determined that this performance deficiency was an additional, but separate consequence of the degraded performance of the diesel generators lubricating oil systems. Following the guidance of Appendix B to Manual Chapter 0612 and Appendix D to Inspection Procedure 71111.12, the inspectors determined that this finding occurred as a consequence of actual problems with the diesel generator lubricating oil system, and that those actual problems were not attributable to this finding. The inspectors used Manual Chapter 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, to conclude that the finding was of very low safety significance. The finding has a crosscutting aspect in the area of human performance associated with resources because the licensee did not ensure that procedures were available and adequate to assure nuclear safety, in that the licensee did not ensure that Administrative Procedure 0.5.NAIT required reevaluation of maintenance rule failures following revisions of equipment cause analyses (H.2(c)
05000298/FIN-2009005-02	2009Q4	Cooper	Multiple Examples of a Failure to Follow Procedure For Extension Cord Configuration Control	The inspectors identified multiple examples of a finding for the licensees failure to initiate condition reports as required by Administrative Procedure 0.36.7, Electrical Cord Control/GFCI Program, to resolve extension cords which had been in place longer than 90 days. Had the condition reports been initiated, design engineering would have evaluated whether permanent power receptacles were needed to power plant equipment, such as security cameras. The licensee entered this issue in their corrective action program as CR-CNS-2009-08610. The performance deficiency associated with this finding was the licensees failure to initiate condition reports for multiple examples of extension cords being used as a substitute for permanent wiring for greater than 90 days. The finding is more than minor because, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern, such as electrical shock, equipment damage or fire. Because the plant was shutdown at the time this performance deficiency occurred, the inspectors used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. Using Checklist 7 in Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists For Both PWRs and BWRs, the inspectors determined that the finding had very low safety significance because every item on the checklist was met. The finding has a crosscutting aspect in the area of human performance associated with resources because the licensees procedure for control of extension cords does not require tracking of extension cord use to ensure that condition reports are initiated for cords in use greater than 90 days (H.2(c)
05000298/FIN-2009005-01	2009Q4	Cooper	Failure to Follow Surveillance Procedure Causes Near Toxic Gas Release	A self-revealing noncited violation of Technical Specification 5.4.1.d, Fire Protection Program Implementation was identified for the licensees failure to follow the requirements of Surveillance Procedure 6.FP.306, Fire Detection Systems Semi-Annual Examination. Specifically, licensee technicians actuated the wrong thermal detector during surveillance testing, causing the CO2 fixed flooding system timer to actuate. Technicians recognized the error when the local and remote alarms actuated, and removed the heat source from the detector prior to release of the CO2 gas. The licensee entered this issue in their corrective action program as CR-CNS-2009-07008. The performance deficiency associated with this finding involved the licensees failure to follow the requirements of Surveillance Procedure 6.FP.306, Fire Detection Systems Semi-Annual Examination. Specifically, licensee technicians actuated the wrong thermal detector during surveillance testing, causing the CO2 fixed flooding system timer to actuate. The finding affects the initiating events cornerstone and is more than minor because it could be reasonably viewed as a precursor to a significant event, namely a toxic CO2 release in the Diesel Generator 1 room. Using the Manual Chapter 0609, Appendix F, Phase 1 screening worksheet, the inspectors determined that the finding has very low safety significance because it was associated with a low degradation rating. The finding has a crosscutting aspect in the area of human performance associated with work practices because maintenance technicians failed to use appropriate self or peer checking techniques, and proceeded in the face of uncertainty when unlabeled components were encountered (H.4(a)
05000483/FIN-2009004-04	2009Q3	Callaway	Licensee-Identified Violation	Technical Specification 5.4.1, Procedures, required that written procedures be established and implemented covering activities specified in Appendix A, Typical Procedures for Pressurized Water Reactors, of Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation), February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a, required procedures for performance of maintenance. Contrary to the above, from October 20, 2008, to September 9, 2009, Procedure MTM-AL-QP002, Turbine-Driven Auxiliary Feedwater Pump, Revision 3, was inadequate for assembly of the turbine-driven auxiliary feedwater pump outboard thrust bearing. Specifically, step 6.15.5 of the procedure incorrectly directed the bearing retaining ring to be installed on the pump shaft instead of within the thrust bearing housing outboard end cover. This finding was entered in the licensees corrective action program as Callaway Action Request 200907931. This finding is greater than minor - 29 - Enclosure because it was associated with the Mitigating Systems Cornerstone attribute of procedural quality and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events
05000483/FIN-2009004-03	2009Q3	Callaway	Licensee-Identified Violation	Technical Specification 3.8.1, AC Sources Operating, Action A.1, required that when one offsite source is inoperable, the licensee shall verify the remaining offsite source is operable within 1 hour and every eight hours thereafter. Contrary to the above, on July 4, 2009, the offsite power source to bus NB01 was inoperable and action was not taken within 1 hour to verify the other offsite source was operable. This finding was entered in the licensees corrective action program as Callaway Action Request 200905373. This finding is greater than minor because it was associated with the Mitigating Systems Cornerstone attribute of human performance and it affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, the issue screened as very low safety significance because it was not a design or qualification deficiency that resulted in a loss of operability or functionality, did not create a loss of system safety function of a single train for greater than the technical specification allowed outage time and did not affect seismic, flooding, or severe weather initiating events
05000397/FIN-2009009-01	2009Q3	Columbia	Failure to Provide Adequate Respiratory Protection Equipment for Emergency Response	The inspectors identified a non-cited violation of 10 CFR 50.47(b)(10) for the failure to provide adequate respiratory protection equipment for emergency response, compromising the protective actions developed for the plume exposure pathway for emergency workers. Adequate quantities of small sized self-contained breathing apparatus respirator masks were not available in the control room for licensed plant operators that were fit-tested for small sizes. This issue was entered into the licensees corrective action program as Action Request 00201679. This finding is greater than minor because it is associated with the Emergency Preparedness Cornerstone attribute of response organization performance and adversely affects the cornerstone objective of ensuring the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was evaluated using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Sheet 1, Failure to Comply. The issue described was a planning standard problem, was not a risk-significant planning standard problem, and did not involve a planning standard function failure. Therefore, the finding is of very low safety significance. This finding has a crosscutting aspect in the area of human performance, associated with resources because the licensee did not have enough small sized selfcontained breathing apparatus respirator masks available in the control room for licensed plant operators that were fit-tested for small sizes (H.2(d)
05000483/FIN-2009004-02	2009Q3	Callaway	Failure to Correctly Identify Safety System Functional Failures in a Licensee Event Report	The inspectors identified a Severity Level IV noncited violation of 10 CFR 50.73(a)(2)(v), Licensee Event Report System, for a failure to report two examples of safety system functional failures in licensee event reports within 60 days after discovery of events requiring a report. The two examples were: March 26, 2008, discovery that operation of containment air coolers in fast speed, during a period of higher than normal containment pressure, could open the air coolers fast speed thermal overload device rendering all the coolers incapable of automatically restarting in slow speed May 21, 2008, discovery of a 6.6 cubic foot void of air in the common suction piping capable of affecting the function of both of the safety injection system pumps For each example, the inspectors reviewed the licensees reportability evaluation and associated past operability reviews and determined each event was reportable per 10 CFR 50.73(a)(2)(v) since each example resulted in a condition which affected both trains of a system described in the Final Safety Analysis Report that was needed to mitigate the consequences of an accident. Alternate safety systems accident mitigation is not permitted as a reason to not report the discovery of the conditions. The licensee also failed to report these failures to the NRC performance indicator database because of the failure to include the safety system functional failure in each respective licensee event report. This finding affects the Mitigating Systems Cornerstone and is greater than minor because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Consistent with the guidance in Section IV.A.3 and Supplement VII, Paragraph D.1 of the NRC Enforcement Policy, this finding was determined to be a Severity Level IV noncited violation. The licensee planned to update the associated license event reports as described in Callaway Action Request 200904980. This finding has a crosscutting aspect in the area of human performance associated with the resources component because the licensee failed to ensure, through adequate training, that its staff understood the guidance documents pertaining to the 10 CFR 50.73 rule (H.2.(b)
05000483/FIN-2009004-01	2009Q3	Callaway	Inadequate Corrective Actions for Essential Service Water Pump Cable Underground Electrical Vault Seals	The inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to take prompt corrective actions prevent continuous submergence of essential service water pump kerite insulated power cables. The continuously submerged environment for these cables existed because the two vaults containing these cables (MH-01N and MH-01S) had inadequate seals needed to protect the vaults from incoming surface water. Callaway Action Request 200201916 stated that all medium voltage cables of concern were located more than 4 feet above the basemat of the vault and thus were not in a submerged condition. The Callaway action request noted that the seals at the top lid were the source of water intrusion and that the seal design was inadequate. On July 9 and 22, 2009, the resident inspectors, along with Callaway plant engineers, inspected the two essential service water underground vaults. The north vault (train A) was found to have water covering the two safety related upper cable trays. Contrary to the Callaway Action Request 200201916 evaluation, the cable trays were about 2.5 feet and 3.5 feet from the basemat of the vault floor. During these 2009 inspections it was noted that the same lid design deficiency identified in Callaway Action Request 200201916 still existed. This led to the discovery that the Callaway Action Request actions from 2002 had not been completely performed. The only significant corrective action had been to increase the inspection frequency to once every three years. The licensee has subsequently taken measures to improve the seals and written Callaway Action Request/Request for Resolution 200905838 to further evaluate this issue. This finding is more than minor because it affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of safety systems. Using Manual Chapter 0609.04, Phase 1 Initial screening and Characterization of Findings, this finding was determined to be of very low safety significance because the degraded seals were a design or qualification deficiency confirmed not to result in loss of operability. The inspectors determined that the finding has no crosscutting aspect as the performance deficiencies were not - 3 - Enclosure reflective of current performance. The licensee entered this item into their corrective action program as Callaway Action Request 20090885
05000285/FIN-2009007-04	2009Q2	Fort Calhoun	Failure to report a potential defect of breaker trip bars per 10 CFR Part 21	The team identified an unresolved item concerning the extent of a deviation originally discovered in a failed safety-related breaker. An inadequate evaluation of the deviation was performed that could result in an event or condition not being properly reported under 10 CFR Part 21, 10 CFR Part 50.72, 10 CFR Part 50.73or 10 CFR Part 73.71.Description. On August 24, 2007, safety-related Breaker MCC-4B1-B01, Pressurizer Backup Heaters Bank 3 Group 8 failed its instantaneous trip setting on one phase. The failure analysis determined the failure to be curvature of the trip bar, likely due to a material defect. This failure was a deviation as defined by 10 CFR Part 21 (a departure from the technical requirements included in a procurement document) and the licensees governing procedure SO-R-1, Reportability Determinations. In order for this deviation to be reportable under 10 CFR Part 21, 10 CFR 50.72, 10 CFR 50.73 or 10 CFR 73.71, the deviation must be determined to be a defect. As defined by 10 CFR Part 21, a defect includes deviations in a basic component delivered to a purchaser for use in a facility or an activity subject to the regulations in this part if, on the basis of an evaluation, the deviation could create a substantial safety hazard. In evaluating the deviation, the licensee arbitrarily determined that the deviation only applied to breakers with the same date code as the failed breaker. This conclusion was reached with no engineering basis and without consultation with the vendor of the breaker. In evaluating deviations, only the vendor can fully determine the extent of the deviation and its potential effect on other plant components. Since Procedure SO-R-1does not direct vendor notification unless the initial deviation is potentially associated with a substantial safety hazard, it was not possible to determine whether the deviation existed in other components. The licensee determined there were no other breakers with the same date code located anywhere on site, thus the only breaker assumed to have the deviation was the initial breaker that failed. Due to safety-related function of the particular breaker, it was determined that there was no substantial safety hazard, and the event was not reportable under 10 CFR Part 50.72 or 10 CFR Part 50.73. Thus the licensee determined that any reporting requirements required under Part 21 were satisfied, as described in 10 CFR Part 21.2(c). However, since the extent of the deviation was measured against breakers only with the same date code, and without consultation with the vendor, the evaluation was inadequate to determine if the event was reportable under 10 CFR Part 50.72 or 10 CFR Part 50.73. In addition, a proper evaluation of components stored in the warehouse could not be made resulting in an inadequate evaluation to determine if the condition was reportable under 10 CFR Part 21.On November 14, 2007, safety-related breaker MCC-3C1-B01, Pressurizer Backup Heaters Bank 2 Group 4 failed its 300 percent thermal test, instantaneous trip setting, on all three phases. This breaker was the same make and model as the breaker that failed on August 24, 2007, but was a different date code. The failure analysis of this breaker was documented in the same report as the initial breaker failure. While the failure mechanism of this breaker was different than the previous breaker failure, the failure analysis noted that the trip bar was curved, though it did not contribute to the failure. The first breaker failure was determined to be curvature of the trip bar, and the second breaker was exhibiting the same characteristics. Since the two failures occurred so close together in time and the failure analyses were documented in the same report, the licensee could have reasonably questioned the extent to which the deviation present in the first breaker occurred. After a review of the two breaker failure events, the team asked the licensee to determine if other breakers were installed in the plant or stored in the warehouse that contained the same deviation. This issue is unresolved pending review of potentially affected breakers after the licensee consults with the vendor to determine if a substantial safety hazard exits (Unresolved Item 05000285/2009007-04)
05000285/FIN-2009007-01	2009Q2	Fort Calhoun	Failure to Perform an Operability Evaluation of a Degraded Condition	The team identified a Green non-cited violation for the licensees failure to meet 10 CFR Part 50, Appendix B, Criterion V in that the licensee failed to perform an operability determination for a degraded condition. The licensee determined that certain relays classified as Functional Importance Determination 1, should be replaced every 9or 15 years depending on the duty cycle and environmental conditions. Most of the relays in the emergency diesel generator had been in service since initial installation, over 35 years ago. Subsequent to the inspection, the licensee performed an operability determination that showed all the effected relays were operable. This condition has been entered into the licensees corrective action program as Condition Reports 2009-2319 and 2342.The finding was determined to be greater than minor because the performance deficiency is associated with the procedure quality attribute (maintenance procedures)of the mitigating system cornerstone, and the performance deficiency adversely affected the associated cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated this finding using Manual Chapter 0609, Attachment 4, Phase 1 Significance Determination, and determined that it was of very low safety significance (Green) because the failure to perform the operability determination did not result in loss of operability or functionality and because the finding did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding has a crosscutting aspect in the area of human performance, decision-making, in that the licensee did not make safety-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions to ensure safety is maintained H.1(a
05000285/FIN-2009007-03	2009Q2	Fort Calhoun	Managing Gas Accumulation in Emergency Core Cooling System, Decay Heat Removal, and Containment Spray System	The team identified an unresolved item concerning the licensees program to identify and manage gas accumulation in emergency core cooling, decay heat removal, and containment spray systems. Specifically, on April 30, 2009, the licensee identified that a section of piping was inappropriately excluded from the scope of its Gas Management Program. Based on this, the licensee was reviewing the program to determine if additional piping was excluded that could cause voided piping, thereby resulting in the inoperability of a safety-related system. In response to NRC Generic Letter 2008-08, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, dated January 11, 2008 (ML072910759), the licensee developed a program to manage gas accumulation in the identified systems. By letter, dated October 14, 2008, Omaha Public Power District described the results of its analyses and concluded that gas accumulation in safety systems was unlikely to create conditions adverse to safety at the Fort Calhoun Station. However, on April 30, 2009, while performing ultrasonic examination of system piping under Work Order QC-ST-HPSI-0001, the licensee identified a gas void on the suction line to high pressure safety injection Pump SI-2B, downstream of Valve HCV-349. In its review, Omaha Public Power District found that it had inappropriately omitted this section of piping from the scope of the Gas Management Program. The team noted that the Updated Safety Analysis Report, section 6.2, page 11 of 35, revision 34, stated, in part, that this section of piping was not necessary to meet the core cooling requirements. However, opening Valve HCV-349 is in the Emergency Operating Procedures, and could introduce the void into the suction piping of high pressure safety injection Pump SI-2B.When discovered, the licensee conservatively declared that section of high pressure safety injection suction piping inoperable and entered Technical Specifications 2.3(2)(e), a 24-hour Limiting Condition for Operation. The licensee took actions to immediately vent and fill that section of piping and declared the system operable. The licensee initiated Condition Report 2009-2069 to determine the cause of the event and to evaluate whether other sections of piping were inappropriately excluded from the scope of its analyses that could render safety-related systems inoperable. At the conclusion of this inspection, the licensee had not completed its reviews. This issue is unresolved pending further NRC review of the licensees Gas Management Program Basis to determine if similar sections of piping were inappropriately excluded such that gas voids could render safety-related systems inoperable (Unresolved Item05000285/2009007-03)
05000285/FIN-2009007-02	2009Q2	Fort Calhoun	Failure to Perform Vendor and Industry Recommended Testing on Safety-Related and Risk Significant 4160 and 480 V Circuit Breakers	The team identified an unresolved item associated with inadequate maintenance procedures for 4160 and 480 V safety-related breakers. The team determined that maintenance procedures used to ensure that 4160 and 480 V safetyrelated breakers were being maintained and overhauled in a timely manner were inadequate. The licensee had no engineering analysis or technical basis to justify the deviation from vendor/Electric Power Research Institute guidance. At the end of the inspection, the licensee identified approximately 20 breakers that had failed over the last15 years and the team was waiting for additional information to determine if the failures were related to the inadequate maintenance. The team identified that the licensee was not performing the maintenance on the breakers as recommended by the vendor or Electric Power Research Institute guidelines. The licensee had completed a review of its breaker maintenance programs in November 2007 and modified it based on Electric Power Research Institute Documents TR-106857-V2 and TR-106857-V3, which are preventive maintenance program bases for low and medium voltage switchgear. The licensee only implemented portions of the recommended maintenance program, and had no engineering analysis or technical basis to justify the changes. Additionally, the guidance states in part that, this program assumes breakers are in nominally good condition to begin with. Breakers that have not been serviced for a very long time may need an overhaul or have a detailed inspection performed before this program is applied. The licensee had not been performing the entire vendor or Electric Power Research Institute recommended tests, inspections, and refurbishments on the breakers since installation. The team reviewed the licensee\'s circuit breaker maintenance procedures and records. The team determined that the licensee had not refurbished Asea Brown Boveri 4160 or General Electric 480 V safety-related and risk significant non-safety-related circuit breakers within the vendor specified 10-year maximum overhaul periodicity or the Electric Power Research Institute guidance of 12 years and had no engineering basis or evaluation to justify the deviation. The team compared the Electric Power Research Institute guidance and vendor-recommended maintenance requirements against the licensee\'s maintenance procedures and found that the licensee was not performing some of the recommended activities or had extended the periodicity of some inspections beyond even the Electric Power Research Institute recommended guidelines. The Fort Calhoun Station program for medium and low voltage switchgear and circuit breakers did not include most of the recommended testing and trending. Specifically, no testing of the operation of the 125-V DC control circuitry was performed at the voltages postulated to exist at the device terminals during design basis events. Contemporary industry standards and Electric Power Research Institute guidance recommend reduced control voltage testing as part of breaker maintenance. Vendor overhaul procedures include reduced control voltage testing on the as-found and as-left control circuit. While there is not an explicit requirement to perform reduced voltage testing on breaker control circuitry, the Electric Power Research Institute guidance recommends reduced voltage testing on breaker control circuitry in order to have reasonable assurance of reliable operation of control circuitry at the postulated minimum control voltage. Additional recommended testing per the preventative maintenance program basis DocumentsTR-106857-V2 and TR-106857-V3 that were not being performed included: Thermography inspections of the breakers and switchgear at recommended periodicity and trending, and: Measurement of the electrical resistance of coils and relays, trended over time to detect progressive failure of winding insulation and give an indication of the condition of these electrical devices. As a result, the team requested the basis for not performing all of the recommended maintenance activities. The licensee was unable to produce an engineering evaluation that allowed the use of the Electric Power Research Institute guidance versus the vendor guidance. Additionally, the team found that the licensee failed to update their in-use guidance when operating experience or new vendor information were issued. Because the licensee was unable to produce documentation demonstrating recommended maintenance had been performed at the appropriate intervals or which qualified the practice of extending the maintenance and refurbishment intervals, the team was concerned about the reliability of the safety-related and safety significant breakers that had not been overhauled within 10 years.n The licensee stated that the 10-year vendor requirement was based on breakers manufactured and lubricated with petroleum-based grease and that their Asea Brown Boveri circuit breakers were lubricated with synthetic-based grease, Anderol 757, which does not dry out as fast and extends the useful life of the lubrication. The licensee cited a May 11, 1995, letter from Asea Brown Boveri/Combustion Engineering that implied grease hardening was not an issue with Anderol 757 lubricant. The team identified operating experience which showed that other licensees had experienced grease hardening in Asea Brown Boveri breakers that contained the Anderol 757.Following the10 CFR Part 21 report issued by D. C. Cook on March 3, 1989, Asea Brown Boveri established the 10 year overhaul frequency. This report was issued after two Asea Brown Boveri 4160 V breakers failed to close because of hardened grease in their operating mechanism. Additional operating experience from Perry supported that grease hardening can occur in less than ten years, pertaining to the 4160 V C residual heat removal (RHR) pump breaker. It stated in part, Various anomalies were identified during the process of disassembling the breaker, and the lubricant within the operating mechanism appears to be hardened. Based on the breaker serial number it was determined that this breaker would have used the synthetic lubricate. This provided further evidence that synthetic grease can degrade in less than 10 years. Asea Brown Boveri breaker historical industry data showed that the lubrication in the operating mechanism tended to harden within 10 years and that this condition can cause sluggish breaker operation. The issue was entered into the licensees corrective action program- 14 V Enclosure and was being evaluated under Condition Report 2009-2306. This issue is unresolved pending review of the causes of the breaker failures as related to the improperly performed maintenance (Unresolved Item 05000285/2009007-02)
05000275/FIN-2009002-02	2009Q1	Diablo Canyon	Inadequate Procedure	The inspectors reviewed a self-revealing noncited violation of Technical Specification 5.4.1 for failure to develop a procedure for removing the reactor head from the reactor pressure vessel and the subsequent filling of the reactor coolant system in a manner that would minimize the potential for airborne contamination. Specifically, on March 5, 2009, while lifting the reactor vessel head in preparation for reloading the reactor core, the licensee experienced airborne radioactivity as high as 4.8 derived air concentrations due to the delay in flooding the reactor refuel cavity. The delay allowed the radioactive contamination on the reactor upper internal structure to dry and subsequent air flow around the upper internal structure caused the contamination to become airborne. The licensee evacuated unnecessary personnel from the containment, initiated containment purge to reduce airborne contamination, and obtained air samples until airborne contamination levels were reduced to normal levels (less than 0.2 derived air concentrations). The licensee entered this item into the corrective actions program as Notification 50209442 and is conducting an apparent cause evaluation of the event. The failure to develop and implement procedures for removing the reactor head and filling the reactor coolant system in a manner that minimized the potential for airborne radioactivity is a performance deficiency. The finding is greater than minor because it is associated with the Occupational Radiation Safety Cornerstone attribute of the program and process and affected the cornerstone objective of exposure/contamination control in that failure to develop and implement adequate procedures for removing the reactor vessel head and fill the reactor coolant system resulted in workers unplanned, unintended dose. Using the Occupational Radiation Safety Significance Determination Process, the inspectors determined this finding had very low safety significance because the finding involved as low as is reasonably achievable planning and work controls, and the licensees 3-year rolling average collective dose is less than 135 person-rem per unit. Because the AMS-4 on the refuel floor in containment alarmed at an airborne concentration of greater than 0.5 derived air concentrations, the finding is self-revealing. Additionally, the finding had a crosscutting aspect in the area of human performance, work control component, because the licensee failed to plan and coordinate work activities by incorporating job site conditions which may impact radiological safety (H.3(a)).
05000275/FIN-2009002-01	2009Q1	Diablo Canyon	Failure to Follow Power Ascension Procedures	The inspectors identified a noncited violation of Technical Specification 5.4.1, Procedures, after plant operators failed to stabilize reactor power and perform a comparison between the calorimetric heat balance calculation and the power range output prior to exceeding 30 percent power. The inspectors concluded several human performance factors contributed to the procedure violation, including less than adequate pre-job brief and poor operational command and control of the reactor power ascension. This finding is greater than minor because the failure to follow procedure is associated with the human performance attribute of the Mitigating Systems Cornerstone and affected the cornerstones objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors used Inspection Manual Chapter 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, to analyze the significance of this finding. The inspectors concluded the finding is of low safety significance because the violation is not a design or qualification deficiency, did not represent a loss of a system safety function or risk significant equipment, and did not screen as potentially risk significant due to a seismic, flooding, or a severe weather initiating event. This finding has a crosscutting aspect in the area of human performance and the work practices component because the licensee failed to ensure adequate supervisory oversight of power ascension activities (H.4(c))
05000275/FIN-2009002-03	2009Q1	Diablo Canyon	Licensee-Identified Violation	Title 10 CFR 50, Appendix B, Criteria X, Inspection, required PG&E to perform examinations or measurements where necessary to assure quality. Contrary to this, Work Package 1-3055C, Reinstall Lower Supports 1-3, completed on March 16, 2009, did not include an examination of the gap between the seismic mounting plates and the load bearing surfaces for the Unit 1 replacement steam generators. As a result, Steam Generator 1-3 was placed in service on March 20, 2009 in an unanalyzed condition due to excessive gaps between two seismic mounting plates and corresponding support columns. On March 22, 2009, after establishing the reactor coolant system at normal operating temperature and pressure, PG&E identified the excessive gaps after a temporary worker raised the concern during an exit interview. The licensee declared the reactor coolant system inoperable and applied the provisions of Technical Specification 3.0.3. PG&E took corrective actions to repair seismic mounting plates. The licensee entered this condition into the correction action program as Notification 50214618, SG 1-3 Column Bearing Surface Issue, This finding is of very low safety significance because the condition did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event
05000275/FIN-2009002-04	2009Q1	Diablo Canyon	Licensee-Identified Violation	Technical Specification 5.4.1.a, Procedures, required that PG&E implement written procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33 includes procedures for draining the reactor coolant system. Contrary to this, on March 13, 2009, PG&E failed to properly align the mid-loop level monitoring system in accordance with Procedure MP I-2.28, Activation of the Reactor Vessel Refueling Level Indication System, prior to start of reactor drain down for mid-loop operations. During the reactor coolant system drain down, a maintenance technician identified that the narrow range level transmitter isolation valve was closed when required to be opened. PG&E stopped the drain down operations and performed a system walk down. During the walk down, PG&E also identified that the flex hose used for the wide range level instrument vent was not connected as required. A maintenance technician and independent verifier had signed off both procedure steps as completed. The inspectors concluded that less than adequate pre-job brief, failure to maintain the procedures in-hand, inadequate use at place keeping and peer checking during the system alignment contributed to this violation. PG&E entered this issue into the corrective action program as Notification 50212379. This finding is of very low safety significance because PG&E did not lose all reactor vessel level indications during midloop operations.
05000275/FIN-2009002-05	2009Q1	Diablo Canyon	Licensee-Identified Violation	Technical Specification 5.7.2.b states, in part, that each entryway to an area with dose rates greater than 1 rem/hour at 30 centimeters from the source shall be conspicuously posted as a high radiation area. Access to and activities in such area shall be controlled by means of a radiation work permit or equivalent that includes specification of radiation dose rates in the immediate work area(s). Contrary to the above, at approximately 7:00 a.m. on March 12, 2009, an individual inadvertently crossed the boundary for the locked high radiation area at the entrance to the stairway to the cavity. The individual was not signed in on a radiation work permit that allowed access to locked high radiation areas; and therefore, did not get a briefing on the dose rates in the area. The violation was identified by the radiation protection technician who immediately informed the individual to exit the area. This issue has been documented as Notification 50211054. The finding was determined to be of very low safety significance because it did not involve as low as is reasonably achievable planning and controls, did not involve an overexposure, did not have a substantial potential for overexposure, and did not result in an impaired ability to assess dose
05000298/FIN-2007007-01	2007Q2	Cooper	Inadequate Procedures Result in Failure of Emergency Diesel Generator Voltage Regulator	The team identified an apparent violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions Procedures, and Drawings, for the failure to establish procedural controls for evaluating the use of parts of indeterminate quality prior to their installation in safety-related applications. This procedural deficiency resulted in the installation of a voltage regulator circuit board of indeterminate quality that adversely affected the function of Emergency Diesel Generator 2. Specifically, following installation of the part on November 11, 2006, failure of the part occurred following 35 hours of operation resulting in an over-voltage trip of Emergency Diesel Generator 2 on January 18, 2007. The finding is greater than minor because it is associated with the equipment performance cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using NRC Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet, a Phase 2 evaluation was required because the finding resulted in the loss of the safety function of Emergency Diesel Generator 2 for greater than the Technical Specification completion time. The Phase 2 evaluation concluded that the finding was of low to moderate safety significance. A Phase 3 preliminary significance determination analysis also determined the finding was of low to moderate safety significance.

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