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05000298/FIN-2018003-032018Q3CooperFailure to Provide Adequate Lubrication for Drywell Fan Coil UnitsThe inspectors reviewed a self-revealed finding for the licensees failure to implement Work Order 5060136 during maintenance on the drywell fan coil units. Specifically, on October 26, 2016, during bearing replacement work on drywell fan coil, unit D, maintenance personnel failed to properly reinstall auto-lubricator injection connectors after removing the interferences per the work order instructions. This error resulted in the failure of drywell fan coil, unit D, due to inadequate bearing lubrication, and ultimately led to a downpower and reactor shutdown.
05000298/FIN-2018003-022018Q3CooperFailure to Perform Process Applicability DeterminationThe inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow Administrative Procedure 0.9, Tagout, Revision 88, for performing a monthly audit and Process Applicability Determination. Specifically, the inspectors noted that a clearance order on the safety-related residual heat removal service water booster pump room fan coil unit was hanging for greater than 90 days with no Process Applicability Determination performed, which resulted in the power switch for the fan coil unit being unintentionally tagged out of its normal configuration for almost 2 years
05000298/FIN-2018003-012018Q3CooperFailure to Provide Complete and Accurate Information in a License Amendment RequestThe inspectors identified that the licensee provided inaccurate information to the NRC in a license amendment request for an emergency action level scheme change. Specifically, the licensee provided information about the measurement ranges of a liquid effluent radiation monitor used in emergency action levels that was not accurate.
05000482/FIN-2018007-022018Q2Wolf CreekMinor ViolationPerformance Deficiency: Failure to promptly identify and correct known-defective switches in inservice safety-related breakers, or to control nonconforming breakers accepted into warehouse stores, as required by 10 CFR 50 Appendix B Criteria XV and XVI. In February 2008, the licensee received a notification from GE Hitachi of reduced reliability of some safety-related circuit breakers due to defective cutoff switches internal to the breakers. The licensee incorrectly screened this information as not applicable to the Wolf Creek Generating Station. In August 2011, after licensee engineers received the information again from industry peers, the licensee screened the information as applicable. The licensee then added steps to its overhaul and pre-install test procedures to check for the defective subcomponent. These steps were performed during subsequent regularly scheduled overhaul or pre-install tests, with the last affected switches being replaced in June 2014 and the last potentially susceptible safety-related breaker being inspected in March 2015. The team determined that because the station had information on the defect in February 2008, but did not correct the condition until 2014 and did not confirm that it was corrected until 2015, the licensee had failed to promptly identify and correct a condition adverse to quality. Further, the licensee failed to inspect or place administrative controls on potentially affected spare breakers that had been accepted into warehouse stores, though the added steps in the pre-install procedure likely would have prevented a defective component from being installed. However, by failing to segregate the potentially affected components until they were inspected, the licensee failed to comply with quality assurance requirements for control of nonconforming components. On June 26, 2018, the licensee put a hold on four potentially affected breakers that were in warehouse stores. The licensee documented this performance deficiency in CR 124693. Screening: The performance deficiency was minor because the licensee did not experience an inservice failure as a result of the defect during the 6 years they remained in service and had a procedure in place that would likely have prevented a defective spare from being issued for installation. Therefore, there was no adverse effect on the mitigating systems cornerstone objective and there was no potential to create a more significant safety concern. Enforcement: This failure to comply with 10 CFR 50 Appendix B Criteria XV and XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000482/FIN-2018007-012018Q2Wolf CreekFailure to Provide Adequate Work Instructions for Preventive Maintenance on Safety-Related EquipmentThe team reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a to establish, implement, and maintain written procedures recommended by Regulatory Guide 1.33, Appendix A, Revision 2. Specifically, work instructions for the preventive maintenance for the train B Class 1E electrical equipment A/C unit SGK05B, lacked adequate guidance for preventive maintenance and calibration of the associated thermostat. This resulted in the loss of cooling failure of the A/C unit SGK05B,on February 12, 2018.
05000416/FIN-2018002-082018Q2Grand GulfPerformance of Surveillance Testing Following Maintenance on Containment AirlockThe inspectors identified a Green non-cited violation of 10CFRPart50,AppendixB, Criterion XI, Test Control, for the licensees failure to perform surveillance testing of containment airlock seals under appropriate conditions. The licensee failed to appropriately control the sequence of maintenance and testing activities to ensure that surveillance testing was not performed subsequent to maintenance which could affect the validity of surveillance test results.
05000416/FIN-2018002-072018Q2Grand GulfLoss of Shutdown CoolingA self-revealed,Green non-cited violation of Technical Specification 5.4, Procedures,for the licensees failure to follow written procedures was identified when the residual heat removal (RHR) system automatically isolated due to an inadvertent emergency core cooling system (ECCS) actuation. While the plant was shut down with the RHR system in decay heat removal mode, maintenance personnel inadvertently opened an incorrect valve during a transmitter calibration activity, which caused a false low reactor pressure vessel (RPV) water level signal, an ECCS actuation, and a loss of decay heat removal for approximately 31 minutes
05000416/FIN-2018002-062018Q2Grand GulfImproper Evaluation and Resolution of Intermediate Range MonitorNoise Leads to Manual Reactor ShutdownA self-revealed, Green non-cited violation of 10CFRPart50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality. Specifically, the licensee failed to implement appropriate corrective actions related to intermediate range monitor (IRM) nuclear instrument (NI) electronic noise spiking. The failure to implement adequate corrective actions over the course of at least 5 years resulted in a plant shutdown due to declaration of multiple IRM channels inoperable while in Mode 2.
05000416/FIN-2018002-052018Q2Grand GulfFailure to Follow Procedure Requirements Resulting in Unplanned DoseA self-revealed, Green non-cited violation of Technical Specification 5.4.1 was identified when an individual alarmed a personnel contamination monitor upon exit from the radiologically controlled area. Specifically, the licensee failed to follow procedures to establish a decontamination plan or procedure, conduct a specific pre-job brief addressing appropriate contamination risk, and receive approval by radiation protection supervision prior to conducting decontamination activities on thereactor pressure vessel(RPV) O-rings
05000416/FIN-2018002-042018Q2Grand GulfHigh Radiation Area Boundary ViolationA self-revealed, Green non-cited violation of Technical Specification 5.7.1 was identified when an individual received a dose rate alarm when the individual failed to comply with established radiological barriers and protective measures and entered a high radiation area. Specifically, an individual leaned over a high radiation area barricade rope, thereby entering the high radiation area. The individuals radiation work permit (RWP) did not permit entry into a high radiation area.
05000416/FIN-2018002-032018Q2Grand GulfFailure to Adequately Test NUS Temperature SwitchA self-revealed,Green non-cited violationof 10CFRPart50, AppendixB, CriterionIII, Design Control, was identified when the reactor core isolation cooling (RCIC) system automatically isolated due to an inadvertent high temperature input from the leakage detection system. Specifically, the licensee failed to fully test a modification that installed a new type of temperature switches, and the system inappropriately isolated the RCIC system when a loss and subsequent restoration of power occurred.
05000416/FIN-2018002-022018Q2Grand GulfFailure to Follow ASME Requirements for Maintaining Inservice Inspection (ISI) Cycles and Perform ASME Required Inservice Inspections within the Scheduled ISI CycleThe inspector identified 15 examples of a Green non-cited violation (NCV)of 10 CFR 50.55(a)(g)(4)(ii), which requires that inservice examination of components classified as American Society of Mechanical Engineers (ASME), Section XI, Code Class 1, Class 2, and Class 3 be conducted during successive 120-month inspection intervals, and requires compliance with the requirements of the latest edition and addenda of the ASME Code (and all its paragraphs) applicable to the specific interval, including maintaining the 120-month inspection interval in accordance with the ASME Code, Section XI, Paragraph IWA-2430. Specifically, the licensee inappropriately adjusted its second inservice inspection 120-month cycle, and failed to perform VT-3 and MT examinations of 15 class 1, class 2, and class 3 components, including the high pressure core spray pump attachment weld and reinforcing band before the third inservice inspection cycle expired on November 30, 2017, as required by 10CFR50.55(a)(g)(4)(ii).
05000416/FIN-2018002-012018Q2Grand GulfFailure to Institute Effective Corrective Action to Preclude RepetitionAn NRC-identified,Green non-cited violation of 10CFRPart50, AppendixB, CriterionXVI, Corrective Action, was identified when the licensee failed to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee left a secondary containment personnel hatch in an open configuration for approximately 30 minutes while performing a roof inspection, which rendered secondary containment inoperable. This issue had also previously occurred in 2016, but corrective actions to prevent it from occurring again were ineffective.
05000416/FIN-2017011-062018Q1Grand GulfFailure to Perform Functionality Assessments as Required by ProceduresThe inspectors identified a finding for the licensees failure to follow Procedure EN-OP-104, Operability Determination Process, Revisions 10 through 12. Specifically, the licensee did not perform functionality assessments for adverse conditions on the offgas system as required by the procedure. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11265. The licensee initiated corrective actions to perform functionality assessments for the conditions identified and to evaluate any potential programmatic issues. The failure to perform functionality assessments required by Procedure EN-OP-104 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to perform functionality assessments could affect the availability and reliability of the offgas system to maintain the doses associated with releases to the environment as low as reasonably achievable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it involved the Effluent Release Program, it did not impair the ability to assess dose, and did not exceed the 10 CFR Part 50, Appendix I, or 10 CFR 20.1301(d) limits. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently disposition adverse conditions associated with the offgas system resulted in the station not performing required functionality assessments (H.13)
05000416/FIN-2017011-052018Q1Grand GulfFailure to Correct Control Room Boundary Door Resulted in Loss of Safety FunctionThe inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Criterion XVI, Corrective Action, for the licensees failure to appropriately correct a condition adverse to quality. Specifically, the control room envelope door had been documented in several condition reports for not consistently working properly. Subsequent to these condition reports, on July 9, 2017, the door was opened and did not close automatically, and therefore the door was left in an unsecured position. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-06705. The licensee restored compliance by securing the door and replacing the hinge bushings to ensure the door would close properly. The failure to correct a condition adverse to quality for a control room envelope boundary door was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the structures, systems, and components and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (functionality of the control room) protect the public from radionuclide releases caused by accidents or events. Specifically, on July 9, 2017, since the licensee had not corrected the adverse conditions identified on the control room envelope door, the door was left in an unsecured position and resulted in the station declaring both trains of standby fresh air inoperable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system, and did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The period of the barrier in the open position was of short duration, approximately 1 minute, and therefore did not result in significant risk input. This finding had a cross-cutting aspect in the area of problem identification and resolution, resolution, because the licensee did not take corrective actions in a timely manner commensurate with their safety significance. Specifically, the licensee did not ensure proper priority of corrective actions on the degraded control room envelope boundary door (P.3).
05000416/FIN-2017011-042018Q1Grand GulfFailure to Install the Residual Heat Removal Pump A Mechanical Seal in Accordance with ProceduresThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures. Specifically, on September 22, 2016, maintenance did not install seal assembly drive pins in accordance with Step 7.8.2 of Procedure 07-S-14-279, Revision 007. The licensee entered this issue into their corrective action program as Condition Reports CR-GGN-2017-08269 and CR-GGN-2017-11009. The licensee implemented immediate corrective actions by declaring the pump inoperable and replacing the mechanical seal. The failure to perform maintenance on the residual heat removal pump A mechanical seal in accordance with written procedures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, on September 22, 2016, mechanical maintenance installed the residual heat removal pump A seal drive pins backwards. As a result, the drive pins damaged the seal and on August 23, 2017, caused an unisolable leak from the seal. This resulted in unplanned inoperability and unavailability of the residual heat removal pump A from August 23, 2017, through August 25, 2017, when the mechanical seal was replaced. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, avoid complacency, because the licensee failed to recognize and plan for the possibility of mistakes, and individuals failed to implement appropriate error reduction tools. Specifically, the licensee failed to use appropriate error reductions tools such as self-check or peer checking which resulted in incorrect performance of procedural steps (H.12)
05000416/FIN-2017011-032018Q1Grand GulfFailure to Conduct Common Cause Failure Evaluation in Response to Inoperable Emergency Diesel GeneratorThe inspectors identified three instances of a non-cited violation of Technical Specification 3.8.1, AC Sources Operating, for the licensees failure to take required actions for an inoperable emergency diesel generator. Specifically, after classifying the Division I or Division II emergency diesel generator as inoperable on the basis of nonconforming conditions, and after failing to either verify that the opposite train emergency diesel generator was not inoperable due to common cause failure within 24 hours or conduct a surveillance run on the opposite train emergency diesel generator within 24 hours, the licensee failed to enter Mode 3 within 12 hours as required by Technical Specification 3.8.1, Actions B.3.1, B.3.2, and G.1, respectively. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-11393. The licensee initiated corrective actions to conduct an adverse condition analysis. The failure to take required actions for an inoperable emergency diesel generator was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment reliability 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. Actions B.3.1 and B.3.2 of Technical Specification 3.8.1 exist to ensure the availability, reliability, and capability of at least one emergency diesel generator in scenarios where there is a potential for a common cause failure of both emergency diesel generators, and the licensee took neither action. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent an actual loss of function of either the Division I or Division II emergency diesel generator for greater than its technical specifications allowed outage time. The finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee failed to use a consistent, systematic approach to make decisions. Specifically, the licensee failed to review the required actions of the applicable technical specification to ensure that all of those actions would be properly carried out (H.13).
05000416/FIN-2017011-022018Q1Grand GulfFailure to Disposition Adverse Conditions as Required by ProceduresThe inspectors identified a finding for the licensees failure to disposition conditions as required by Procedure EN-LI-102, Corrective Action Program, Revisions 24 through 30. Specifically, the licensee did not identify 72 conditions as either Adverse Category B, C, or D as required by the procedure. As a result, the licensee failed to perform the required cause evaluations and develop corrective actions to address the conditions. The licensee entered the conditions into their corrective action program as Condition Report CR-GGN-2017-10896. The licensee initiated corrective actions to re-categorize the conditions and perform the required evaluations. The failure to disposition conditions as adverse (B, C, or D) as required by Procedure EN-LI-102 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, condition reports associated with deficiencies or potential deficiencies involving safety-related equipment are required to be categorized as adverse and appropriate corrective actions are assigned including causal analyses appropriate to the circumstances per licensee Procedure EN-LI-102. The inspectors performed an initial screening of the finding in accordance with Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently disposition identified conditions as adverse led to the failure to fully evaluate the conditions (H.13).
05000416/FIN-2017011-012018Q1Grand GulfFailure to Categorize Condition Reports for Significant Conditions Adverse to Quality as Required by ProceduresThe inspectors identified five examples of a finding for the licensees failure to categorize and evaluate conditions in accordance with procedural requirements. Specifically, the licensee did not categorize adverse conditions that represented the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102, Corrective Action Program, Revisions 24 through 28. The licensee entered the conditions into their corrective action program as Condition Report CR-GGN-2017-10896. The licensee initiated corrective actions to re-categorize the conditions and perform the required evaluations. The failure to categorize conditions that represent the loss of a safety function as significant conditions adverse to quality as required by Procedure EN-LI-102 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, root cause evaluations, corrective actions to prevent recurrence, and effectiveness reviews are used per licensee Procedure EN-LI-102 to ensure availability and reliability of structures, systems, and components are maintained. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it was related to, but was not itself: (1) a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) a loss of system and/or function; (3) an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. This finding had a cross-cutting aspect in the area of human performance, consistent process, because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensees failure to consistently evaluate the conditions during initial screening led to the incorrect categorization of the condition reports (H.13)
05000313/FIN-2017003-012017Q3Arkansas NuclearFailure to Maintain Service Water Train SeparationThe inspectors identified a non- cited violation of Technical Specification 5.4.1.a for the licensees failure to maintain train separation between safety -related service water trains when swapping the swing high pressure injection (HPI) pump between trains. Specifically, by following procedure OP 1104.002, Makeup and Purification System Operation, Revision 89, operators cross -tied service water trains, placing the system in an unanalyzed condition. This condition resulted in the train A electrical equipment room emergency chiller and train B reactor building emergency cooling coils being inoperable for a maximum of 25 minutes per occurrence. Additionally, it was determined that service water temperatures over the past 3 years did not result in an actual loss of function associated with these components if a design basis accident would have occurred. The immediate corrective actions were to assess past operability for not maintaining service water train separation and to revise Operating Procedure 1104.002 with adequate work instructions to maintain service water train separation. The licensee entered this deficiency into the corrective action program as Condition Report CR -ANO -1-2017- 02518. The licensees failure to maintain safety -related service water train separation when swapping the swing HPI pump between trains was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensees failure to maintain service water train separation placed the system in an unanalyzed condition and was subsequently determined to cause the train A electrical equipment room emergency chiller and train B reactor building emergency cooling coils to be inoperable for a maximum of 25 minutes per occurrence . Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding s At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety -significant , non -technical specification train. Specifically, inspectors confirmed that service water temperatures were never high enough to result in an actual loss of function for either limiting component. The finding had 3 a cross -cutting aspect in the area of human performance associated with conservative bias because the licensee failed to determine whether the proposed action was safe to proceed, rather than unsafe in order to stop. Specifically, in December 2015 when this approach was revise d to declare only the non- protected service water train inoperable, the licensee did not ensure that the transition lineup was analyzed to be within safety analyses before adopting the revised steps. (H.14)
05000458/FIN-2017009-012017Q2River BendFailure to Obtain Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection PointGreen. The NRC identified a Severity Level IV violation for the licensees failure to restore compliance for a non-cited violation (NCV) associated with failure to obtain NRC approval prior to making a change to the reactor core isolation cooling injection point. Specifically, as of April 28, 2017, the licensee had not restored compliance with a violation the NRC identified on October 8, 2015. This violation described a previously made change to the facility without prior NRC approval in violation of 10 CFR 50.59, Changes, Tests, and Experiments. The team determined that the licensees failure to restore compliance within a reasonable amount of time was a performance deficiency. Title 10 CFR 50, Appendix B, Criterion XVI, requires in part that, measures shall be established to assure that conditions 3 adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2017-03505. The finding was more than minor because it is associated with the initiating events aspect of the reactor safety cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The finding is of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a human performance cross-cutting aspect associated with procedural adherence because individuals failed to follow the procedures delineated by the corrective action program (H.8). Originally, the licensee met the criteria for dispositioning the issue (50.59) as a NCV. However, based upon the fact that the condition report, which documented the NCV, was closed without restoring compliance, the licensee no longer met the criteria for a NCV and therefore, this violation is being cited in a notice of violation
05000458/FIN-2015007-022015Q4River BendFailure to Obtain Prior NRC Approval for a Change in Reactor Core Isolation Cooling Injection PointThe team identified a Severity Level IV, Green, non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section (c)(2) which states, in part, A licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the final safety analysis report (as updated). Specifically, prior to October 8, 2015, the licensee failed to correctly evaluate that a spurious reactor core isolation cooling actuation injecting into the feedwater line resulted in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the updated final safety analysis report. In response to this issue, the licensee initiated a condition report to document completion of a new evaluation under current regulatory guidelines. This finding was entered into the licensees corrective action progam as Condition Report CR-RBS-2015-7259. The team determined that the failure to perform an adequate evaluation of a design change was a performance deficiency. This finding was also evaluated using traditional enforcement because it had the potential to impact the NRCs ability to perform its regulatory function. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, the licensee failed to correctly evaluate that a spurious reactor core isolation cooling actuation injecting into the feedwater line resulted in a more than minimal increase in the frequency of occurrence of the loss of feedwater heating accident previously evaluated in the updated final safety analysis report. In accordance with Inspection Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency where the mitigating structure, system, or component maintained its operability or functionality. Since the violation is associated with a Green reactor oversight process violation, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. There is no cross-cutting aspect assigned to this performance deficiency because the performance deficiency is not indicative of current performance and also because cross-cutting aspects are not assigned to traditional enforcement violations.
05000285/FIN-2015009-012015Q4Fort CalhounFailure to Take Adequate Corrective Action to Preclude Repetition of a Significant Condition Adverse to Quality Associated with Emergency Diesel Generator Room Water IntrusionsThe team identified an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take corrective actions to prevent repetition of a significant condition adverse to quality. Specifically, since February 2009, the licensee failed to take corrective actions to prevent repetitive water intrusions from the Auxiliary Building HVAC room (Room 82) into the number one Emergency Diesel Generator room (Room 63). The inspectors determined that the licensees failure to implement corrective actions to preclude repetitive water intrusions into Room 63 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone. Specifically, water intrusion events from Room 82 into Room 63 could challenge the reliability of the emergency diesel generator when relied upon during a loss of offsite power. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Question, inspectors determined that the finding was of very low safety significance (Green). The finding has a problem identification and resolution cross-cutting aspect within the area of Resolution, because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3).
05000285/FIN-2015009-022015Q4Fort CalhounFailure to Revise Procedures and Perform Additional TrainingThe team evaluated a self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies... are promptly identified and corrected. Specifically, prior to September 30, 2015, the licensee failed to revise procedures, and perform additional operator training, to prevent the inadvertent opening of steam bypass and steam dump valves during plant startup, and any subsequent plant impacts. In response to this issue, the licensee initiated a condition report to document these corrective actions. This finding was entered into the licensees corrective action program as Condition Report CR-FCS-2015-13718. The team determined that the failure to take timely corrective actions to revise procedures and complete additional training to correct a condition adverse to quality, was a performance deficiency. This finding was more than minor because it was associated with the initiating events cornerstone objective of configuration control to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to take recommended corrective actions to revise procedures and perform additional operator training to ensure proper alignment of the steam dump and bypass valves controller during startup. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the team determined that the finding was determined to have very low safety significance (Green) since the transient did not result in a reactor trip or loss of mitigation equipment. The finding has a problem identification and resolution cross-cutting aspect in the area of Operating Experience, because the licensee failed to systematically and effectively collect, evaluate, and implement relevant internal operating experience in a timely manner (P.5).
05000416/FIN-2015003-012015Q3Grand GulfFailure to Have Appropriate Instructions for Preventative Maintenance on the Division II Diesel Generator Fuel Rack Control LeverThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, for the failure to establish appropriate maintenance instructions to perform maintenance activities on the fuel rack control lever of the division II diesel generator. Specifically, the preventative maintenance instruction did not inspect for foreign material between the fuel rack control lever and the adjacent clamp, which caused the fuel rack control lever to be stuck in the open position. As a result, the division II diesel generator was rendered inoperable and unavailable. On June 28, 2015, the licensee cleaned and lubricated the fuel rack control lever and performed the preventative maintenance instruction to return the division II diesel generator to operable status. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2015-3741. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely 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, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program. The mechanical standard was last updated in 2006, and the preventative maintenance instruction was last updated in 2012 for editorial purposes only. The inspectors determined that the cause of the deficiency occurred in 2006, and therefore, determined the finding did. not have a cross-cutting aspect since it is not indicative of current licensee performance.
05000298/FIN-2015002-012015Q2CooperFailure to Prevent Reactor Thermal Power from Exceeding 2419 MWt for Preplanned ActivityThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the licensees failure to appropriately implement General Operating Procedure 2.1.10, Station Power Changes, Revision 107. Specifically, the procedure required in Step 10.3 that the licensee, Ensure any pre-planned evolution (e.g., pressure change, flow change, etc.) will not result in operation greater than 2419 MWt. On May 8, 2015, the licensee failed to implement Step 10.3 of General Operating Procedure 2.1.10, when they failed to reduce power to ensure that reactor power did not exceed 2419 MWt as the reactor recirculation motor generator B scoop tube was unlocked. As a result of this failure to reduce power for this planned evolution, reactor power increased to 2422 MWt. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-04259. The performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown as well as power operations. Specifically, the licensee did not know the condition of the reactor recirculation motor generator set B potentiometer prior to unlocking it and failed to reduce power such that when the scoop tube was unlocked, the resulting power increase would not exceed 2419 MWt. The inspectors screened the finding using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Section C, Reactivity Control Systems, which resulted in a Yes answer to Question 2 since the finding involved control manipulations that unintentionally added positive reactivity. This referred the inspectors to Inspection Manual Chapter 0609, Appendix M, Significance Determination Using Qualitative Criteria. A Senior Reactor Analyst performed a bounding qualitative evaluation and determined that the finding was of very low safety significance (Green) because of the relatively small magnitude of the overpower event, the prompt operator actions to return power to below the licensed limit upon discovery, and the fact that the overpower event did not result in any failure of the fuel cladding. This finding has a cross-cutting aspect in the area of human performance associated with conservative bias. Specifically, the affected evolution was known in advance to have the possibility of a positive reactivity impact; however, operators did not take appropriate actions to reduce power sufficiently prior to unlocking the reactor recirculation motor generator set B scoop tube in order to prevent the reactor from exceeding 2419 MWt (H.14).
05000298/FIN-2015001-012015Q1CooperInadequate Operations ProcedureThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the inadequate Operations Procedure 2.2.7, Condensate Storage and Transfer System, Revision 56. Specifically, the procedure did not require that the affected system, either the high pressure coolant injection system or the reactor core isolation cooling system, be declared inoperable when one or more of the high pressure coolant injection or reactor core isolation cooling test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, or RCIC-MOV-33, were moved off of their closed (passive safety function position) seats. The license entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-00274. The failure to establish and maintain a correct filling procedure required by Technical Specification 5.4.1.a. was a performance deficiency and resulted in the licensees failure to declare the high pressure coolant injection and reactor core isolation cooling systems inoperable when required to do so. The performance deficiency is more than minor, and therefore a finding, because it is associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the high pressure coolant injection and reactor core isolation cooling systems were not declared inoperable when their test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, and RCIC-MOV-33, were taken off their normally closed (passive safety function position) seats. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-ofservice for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction techniques. Specifically, licensee personnel fell into a pattern of acceptance regarding Procedure 2.2.7. This resulted in a failure to question the lack of an operability caution statement, even though there was other guidance in the inservice inspection program to that effect (H.12).
05000416/FIN-2015001-012015Q1Grand GulfFailure to Take Timely Corrective Actions Associated with Division 1 and 2 Standby Service Water Pump House Ventilation System Due to Degraded RelaysThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee's failure to take timely corrective actions to correct a condition adverse to quality associated with the division 1 and 2 standby service water pump house ventilation systems. Specifically, in June 2011, the licensee identified that relays associated with the standby service water system pump house ventilation system failed due to age/environmental degradation, which resulted in an unplanned inoperability of the standby service water system. However, the licensee did not implement timely corrective actions for replacing these relays, which resulted in the inoperability of the division 1 standby service water system in December 2014, and again in January 2015. The licensee documented this issue in their corrective action program as Condition Report CR-GGN-2015-00739. The short-term corrective actions included replacing all of the division 1 and 2 standby service water ventilation pump house relays in February and early March 2015. The inspectors determined that the failure to take timely corrective actions to replace degraded relays in the standby service water pump house ventilation system was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating System Cornerstone and adversely affected the 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, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, the inspectors determined the issue to be of very low safety significance (Green) because all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. The inspectors determined that this performance deficiency was not indicative of current plant performance, and therefore no cross-cutting aspect was considered.
05000416/FIN-2015001-022015Q1Grand GulfFailure to Follow a Procedure Resulting in the Unplanned Inoperability of the Reactor Core Isolation Cooling SystemThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a, for failure to follow a procedure which resulted in the unplanned inoperability of the reactor core isolation cooling system. This occurred when licensee technicians tested for continuity between incorrect points, while performing surveillance activities related to the residual heat removal system. This resulted in an invalid group 4 isolation signal and an isolation of the reactor core isolation cooling steam supply. The licensee entered this issue into the corrective action program as Condition Report CR-GGN- 2015-01532, and took immediate corrective actions to stop the residual heat removal system surveillance activity and restore the reactor core isolation cooling system to service. The failure to properly follow the surveillance procedure, which resulted in the unplanned inoperability of the reactor core isolation cooling system, was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Mitigating Systems Cornerstone. Specifically, the licensees failure to properly follow the surveillance procedure resulted in the unplanned inoperability of the reactor core isolation cooling system, which adversely affected the 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, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) in that the issue did not affect the design or qualification of the reactor core isolation cooling system; did not represent a loss of the reactor core isolation cooling system function (in that the isolation could have been promptly reset by procedures, had the system operation been required); and did not represent loss of function for greater than the Technical Specification allowed outage time. The inspectors determined this finding had cross-cutting aspect in the area of human performance associated with avoiding complacency, in that the I&C technicians did not implement appropriate error reduction tools to ensure the meter was connected to the correct points, which resulted in the invalid group 4 isolation signal, and inoperability of the reactor core isolation cooling system (H.12).
05000416/FIN-2015001-062015Q1Grand GulfFailure to Adequately Establish Commercial-Grade Items as Basic ComponentsThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the suitability of replacement parts that were procured from commercial suppliers. Specifically, the inspectors noted that none of the tests specified by the licensee were sufficient to ensure that the seismic qualification of an auxiliary relay had been maintained. The finding was entered into the licensees corrective action system as Condition Report CR-GGN-2014-05049. The performance deficiency is more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because the licensee performed an operability determination, which evaluated the safety impacts of postulated relay chatter during a seismic event, for the applications in which these relays were installed. The licensees subsequent operability evaluation determined that potential relay chatter would not impact the safety-related functions of the relays in the applications in which they were installed. Thus, all applicable screening questions in Manual Chapter 0609, Appendix A, Exhibit 2, were answered no. A cross-cutting aspect is not being assigned to this finding.
05000416/FIN-2015001-032015Q1Grand GulfEmergency Action Level Scheme for Nonfunctional Seismic MonitorThe inspectors identified a non-cited violation of 10 CFR 50.54(q)(2) for the licensees failure to follow and maintain the effectiveness of an emergency plan that meets the requirements of the planning standard 50.47(b)(4), which requires that a standard emergency classification and action level scheme, is in use by the licensee. Specifically, the licensee had identified, on October 15, 2013, that the seismic monitoring instrumentation was non-functional, but had not further evaluated the plant configuration, and the effect on emergency action level declaration capabilities for seismic events. The licensee documented this issue in Condition Report CR-GGN-2015-00713. The corrective actions, based on CR-GGN-2013-06514, were implemented, and a new seismic monitor was installed, tested, and brought into service on January 30, 2015. The licensees inability to promptly declare Emergency Action Level (EAL) HA6, as required in the approved emergency classification and action level scheme per 10 CFR Part 50.47(b)(4), was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Emergency Preparedness Cornerstone and adversely affects the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, it negatively impacts the cornerstone attribute of procedure quality in that the plant configuration prohibited the timely declaration of the facility EALs, as written. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that the issue affected the Emergency Preparedness Cornerstone. In accordance with NRC Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 23, 2014, the inspectors determined that the issue is of very low safety significance (Green) because an Emergency Action Level was rendered ineffective such that HA6 would not be declared, consistent with Table 5.4-1 and Figure 5.4-1. The inspectors determined the finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation, in that the organization did not thoroughly evaluate issues to ensure that resolutions address causes, and extent of conditions, commensurate with their safety significance; in that while following Technical Requirements Manual requirements for a non-functional piece of equipment (seismic monitor), the complete effect was not evaluated to ensure the EALs were still capable of being implemented (P.2).
05000416/FIN-2015001-042015Q1Grand GulfFailure to Properly Calibrate Main Steam Line Radiation Monitors and Containment/Drywall High Range Radiation MonitorsThe inspectors identified a non-cited violation of 10 CFR 20.1501(c) for the licensees failure to properly calibrate the main steam line radiation monitors and the containment/drywell high range radiation monitors. The violation was of very low safety significance and was entered into the licensees corrective action program as Condition Report CR-GGNS-2015-01832. The failure to properly calibrate radiation monitors was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it adversely affects the cornerstone objective to ensure adequate protection of employee health and safety and is associated with the cornerstone attribute of plant instrumentation. Specifically, the failure to properly calibrate radiation monitors impacts their ability to be used to assess dose rates. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding to be of very low safety significance because it was not an as low as reasonably achievable (ALARA) issue, there was no overexposure or substantial potential for overexposure, and the licensees ability to assess dose was not compromised. This finding has a cross-cutting aspect in the resources component of the human performance area because the licensee did not ensure that calibration procedures were adequate, nor was proper calibration equipment designed, characterized, and made available (H.1).
05000416/FIN-2015001-052015Q1Grand GulfFailure to Establish, Implement, and Maintain Appropriate Changes to the Offsite Dose Calculation Manual For REMP Airborne SamplingThe inspectors identified a non-cited violation of Technical Specification 5.5.1, Offsite Dose Calculation Manual (ODCM). Specifically, when changes were made to the Offsite Dose Calculation Manual in 1997, the licensee failed to establish an airborne sampling location for a community with the highest deposition factor (D/Q) for the site. As immediate corrective actions, the licensee evaluated their Offsite Dose Calculation Manual, evaluated the dose differential for the monitoring locations, and developed a plan to meet the environmental sampling requirements. The issue was documented in Condition Report CR-GGNS-2015-01835. The failure to establish an air sampling location in the vicinity of a community having the highest D/Q was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it adversely affects the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the environment and public domain. Specifically, the failure to maintain the Offsite Dose Calculation Manual with appropriate airborne radionuclide sampling requirements adversely impacts the licensee's ability to validate offsite radiation dose assessments for members of the public under certain effluent release conditions. Using Inspection Manual Chapter 0609, Appendix D, dated February 12, 2008, Public Radiation Safety Significance Determination Process, the inspectors determined that the violation had very low safety significance because it involved the environmental monitoring program. This finding has a cross-cutting aspect in the procedure adherence component of the human performance area because licensee personnel failed to follow procedures when they determined the airborne sampling locations for the updated Radiological Environmental Monitoring Program (H.8).
05000285/FIN-2014005-012014Q4Fort CalhounFailure to establish Appropriate Preventive Maintenance and Failure to Identify Raw Water SSC Maintenance Rule Performance Criteria Exceeded and thereby establish Monitoring Requirements for the SSCThe inspectors identified an NCV of very low safety significance of 10 CFR 50.65 paragraph (a)(2) Requirements for Monitoring the Effectiveness of Maintenance of Nuclear Power Plants, because the licensee did not demonstrate that performance of a component was being effectively controlled through appropriate preventive maintenance, and did not monitor the performance of the component against licensee-established goals to provide reasonable assurance that the component was capable of fulfilling its intended function. Specifically, the licensee failed to demonstrate that the performance of raw water system valve HCV-2875A was being effectively controlled through appropriate preventive maintenance and failed to monitor the valves performance against licensee established goals when performance criteria were exceeded. Corrective actions taken for this violation included revising the Maintenance Rule performance criteria assessment for this component, classifying the component as 10 CFR 50.65 (a)(1), and specifying goals, corrective actions, and additional monitoring for the component. The licensees failure to demonstrate component performance through appropriate preventive maintenance, and the failure to identify that system performance criteria had been exceeded, and as a result, the failure to perform an evaluation of the system for 50.65 (a)(1) goals, corrective actions, and monitoring, was a performance deficiency within the licensees ability to foresee and correct. The finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the Cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify that valve HCV-2875A performance criteria had been exceeded resulted in a delayed assessment of this component and additional failures occurred in the intervening timeframe which adversely affected the overall reliability of the raw water system. The inspectors screened the finding in accordance with NRC IMC 0609, Appendix A, the Significance Determination Process (SDP) for Findings at Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its TS allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution and the Evaluation aspect because the licensee failed to appropriately evaluate the preventive maintenance for valve HCV-2875A to demonstrate component performance and failed to correctly evaluate a functional failure against system performance criteria to ensure system goals, corrective actions, and monitoring were identified.
05000298/FIN-2014005-022014Q4CooperImplementation of Enforcement Guidance Memorandum 11-003, Revision 2, Causes Conditions Prohibited by Technical SpecificationsDuring Refueling Outage 28, Cooper Nuclear Station performed Operations with a Potential for Draining the Reactor Vessel (OPDRV) activities while in Mode 5 without an operable secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measure to terminate the uncovering of fuel. Secondary containment is required by TS 3.6.4.1 to be operable during OPDRV. The required action for this specification is to suspend OPDRV operations. Therefore, entering the OPDRV without establishing secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). The NRC issued Enforcement Guidance Memorandum (EGM) 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliances with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to: (1) adhere to the NRC plain language meaning of OPDRV activities, (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times, (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5, and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities. The inspectors reviewed this Licensee Event Report for potential performance deficiencies and/or violations of regulatory requirements. The inspectors reviewed the stations implementation of the Enforcement Guidance Memorandum 11-003, Revision 2, during operations with a potential for draining the reactor vessel. Specific observations included: 1. The inspectors observed that the operations with a potential for draining the reactor vessel activities were logged in the control room narrative logs, and that the log entry appropriately recorded the standby source of makeup designated for the evolutions. 2. The inspectors noted that the reactor vessel water level was maintained at least greater than 21 feet above the top of the reactor pressure vessel flange as required by Technical Specification 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designed in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours. 3. The inspectors verified that the operations with a potential for draining the reactor vessels were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the operations with a potential for draining the reactor vessels. The inspectors verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events. Technical Specification 3.6.4.1 requires, in part, that secondary containment shall be operable during operations with a potential for draining the reactor vessel. Technical Specification 3.6.4.1, Condition C, requires the licensee to initiate actions to suspend operations with a potential for draining the reactor vessel immediately when secondary containment is inoperable. Contrary to the above, from October 3, 2014 to October 22, 2014, Cooper Nuclear Station performed operations with a potential for draining the reactor vessel activities while in Mode 5 without an operable secondary containment. Specifically, the station conducted the following seven operations with a potential for draining the reactor vessel activities without an operable secondary containment: Draining reactor recirculation pump without the jet pump plugs fully installed Control rod drive maintenance Removal of jet pump plugs associated with reactor recirculation pump B maintenance Venting the control rod drives Defeating the scram function for two control rod drives and support IVVI inspections Reactor recirculation pump A seal maintenance Control rod drive freeze seal These conditions were reported as conditions prohibited by Technical Specifications. The licensee entered this issue into its corrective action program as Condition Reports CR-CNS-2014-06293. Since this violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy, and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request within 4 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The Licensee Event Report is closed.
05000285/FIN-2014005-022014Q4Fort CalhounFailure to determine the availability of local population data for use in estimating changes in the EPZ populationThe NRC identified a Green non-cited violation for the licensees failure to determine the availability of year 2013 state and local population data in estimating annual changes in the plume exposure emergency planning zone population. The failure to determine whether State and/or local population data was available for 2013 was a performance deficiency within the licensees ability to forsee and correct. Appendix E to 10 CFR Part 50, Section IV.5, states, in part, that during the years between decennial censuses, nuclear power reactor licensees shall estimate emergency planning zone permanent resident population changes once a year using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. Contrary to the above, Fort Calhoun Station failed in 2013 to estimate emergency planning zone permanent resident population changes using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. Specifically, Fort Calhoun Station failed to determine whether State and local government population data was available prior to performing the analysis. The issue was entered into the licensees corrective action system as Condition Report 2014-12474. This finding is more than minor because the issue is associated with procedure quality and offsite Emergency Preparedness cornerstone attributes and adversely affected the Emergency Preparedness cornerstone objective. The finding was evaluated using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated February 24, 2014, and was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements, was not a loss of planning standard function, and was not a degraded planning standard function. The planning standard function was not degraded because including state and local 2013 data would not have required the current emergency planning zone time estimate to be updated. There are no immediate safety or security concerns associated with this finding. This finding was assigned a cross-cutting aspect in the area of human performance associated with work management because the licensee failed to understand the scope of work performed by a contractor on their behalf, and failed to ensure the contractor fully complied with regulatory requirements.
05000298/FIN-2014005-012014Q4CooperFailure to Follow Procedure for Post Maintenance TestingThe inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to follow Special Procedure GEH-TP-116, Procedure for the Operation and Maintenance of the REM*TAKE-2/D-100 Modified REM*TAKE 2, Revision 3, for postmaintenance testing following corrective maintenance. Specifically, the licensee did not follow post-maintenance testing requirements associated with the calibration of the bleeder valve for the REM*TAKE-2/D-100 tool following corrective maintenance to address water intrusion. This resulted in the bleeder valve being misadjusted and nullifying the fail-safe feature of the REM*TAKE-2/D-100 tool. With the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when the supplemental employee inadvertently pressed the disengage button. No reactor fuel was damaged as indicated by normal radiation levels and air samples on the refuel floor and reactor water coolant samples. The licensees immediate corrective actions for the event was to suspended all in-vessel maintenance activities and remove REM*Take-2/D-100 grapple from service and determined functionality of the tool. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2014-06809. The licensees failure to follow the post-maintenance testing requirements in Special Procedure GEH-TP-116 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the associated objective of maintaining functionality of fuel cladding. Specifically, with the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when a supplemental employee inadvertently pressed the disengage button. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 09, 2014, inspectors determined that the finding was of very low safety significance (Green) because the finding did not impact the fuel barrier because it: (1) does not increase the potential for failure of the freeze seal or if unmitigated have the potential to cause a disruption of residual heat removal/decay heat removal or a loss of inventory event; (2) does not involve two or more adjacent control rods with the potential to, or actually, add postive reactivity; and (3) does not degrade the ability to isolate a drain down or leakage path. The finding has a cross-cutting aspect in the area of human performance associated with the field presence component because the licensee failed to ensure supervisory and management oversight of work activities including contractors and supplemental personnel (H.2).
05000458/FIN-2014003-012014Q2River BendFailure to Follow Tagging Clearance InstructionsThe inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a., "Procedures," for the failure to adhere to procedural requirements to ensure that other fire suppression ring header valves are/are not correctly positioned. Specifically, on May 19, 2014, the licensee failed to follow the specified instructions in tagging clearance 1C16 / 251-001-O-FPW-P1A, to verify that there were no other ring header valves isolated before implementing the clearance, resulting in the inadvertent isolation of the fire protection ring header. The licensee entered this issue into their corrective action program as Condition Report CR-RBS-2014-02489. The failure to follow procedures is a performance deficiency. The performance deficiency is more than minor and, therefore, a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone, in that the licensee isolated the fire suppression header to the majority of the plant for approximately 36 hours. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," dated June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was in operation. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013. The inspectors determined that Appendix F did not address the loss of the fire protection ring header to most of the facility and Appendix F, "Assumptions and Limitations," states "the SDP approach is intended to support the assessment of known issues only in the context of an individual fire area. A systematic plant-wide search and assessment effort is beyond the intended scope of the fire protection SDP." Therefore, a senior reactor analyst (SRA) performed a detailed risk evaluation. The total exposure period was 36 hours. The bounding change to the core damage frequency was 2E-7/year. The bounding change to the large early release frequency was 4E-8 per year. The finding was of very low safety significance (Green). The dominant core damage sequences included a fire-induced loss of offsite power, failure of operators to suppress the fire, and damage to Division I, II, and III components. The reactor core isolation cooling system and the short exposure period helped to minimize the risk. The finding has a cross-cutting aspect in the area of human performance associated with avoiding complacency because the licensee failed to recognize and plan for the possibility for mistakes and did not implement appropriate error reduction tools (H.12).
05000458/FIN-2014003-022014Q2River BendLicensee-Identified ViolationLicense Condition 2.C(10), "Fire Protection," requires the licensee to "...comply with the requirements of the fire protection program as specified in Attachment 4 (of the license)." Provision 1 of Attachment 4 states in part: "EOI shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility through Amendment 22 and as approved in the SER dated May 1984 and Supplement 3 dated August 1985 subject to provisions 2 and 3." Section 9.5.1 of the Updated Final Safety Analysis Report (UFSAR), "Fire Protection System," Subsection 9.5.1.4., "Inspection and Testing Requirements," states that "Periodic operational checks, inspections, and servicing required to maintain fire protection systems that protect equipment that is important to safety, including the alarm and detection systems, conform with the RBS Technical Requirements Manual." Technical Requirements Manual, Section TRM 3.7.9.2, Action A.2, states that if "One or more of the...required spray or sprinkler systems (are) inoperable," the licensee would be required to "Establish a continuous fire watch with backup fire suppression equipment for those areas in which redundant systems or components could be damaged," and "Establish an hourly fire watch patrol for other areas," within a completion time requirement of one hour. Contrary to the above, on May 20, 2014, the licensee failed to establish fire watches within the one hour requirement, specified in TRM 3.7.9.2., after it was determined that the fire protection ring header was inoperable. The failure to adhere to the requirements of TRM 3.7.9.2 action A.2, to ensure that all required hourly fire watches are posted within one hour from the time of entry into the TRM, is a performance deficiency. The performance deficiency was more than minor and, therefore, a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone, in that the failure to post fire watches, in a timely manner, could result in preventing prompt detection and extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," dated June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was in operation. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix F, "Fire Protection Significance Determination Process," dated September 20, 2013. Since the finding affected many fire areas, the inspector consulted with a senior reactor analyst. The analyst determined that Appendix F was not a suitable tool to process this finding, and that a detailed risk evaluation needed to be performed. Although the exposure period for this finding was just a few hours, the risk analyst determined that the detailed risk evaluation performed for the finding described above (Section 4OA2.3.b.) fully bounded this finding as well. As a result of the referenced evaluation, the finding was found to be of very low safety significance (Green). The dominant core damage sequences included a fire-induced loss of offsite power, failure of operators to suppress the fire, and damage to Division I, II, and III components. The reactor core isolation cooling system and the short exposure period helped to minimize the risk. This violation is being treated as a non-cited violation (NCV), consistent with Section 2.3.2.a of the Enforcement Policy because it was of very low safety significance (Green) and it was entered into the licensee's corrective action program as Condition Report CR-RBS-2014-02489 to address recurrence.
05000416/FIN-2014002-022014Q1Grand GulfFailure to Control a Locked High Radiation Area Due to Unsecured Highly Radioactive Materials Stored in the PoolThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.7.3, resulting from the licensees failure to control a high radiation area with radiation levels greater than 1000 millirem per hour. As immediate corrective actions, the licensee stopped the work activity, placed a senior radiation protection technician in control of the area, surveyed all affected areas, and properly posted and controlled the area. The licensee also checked qualifications of the involved individuals and conducted a root cause evaluation for the event. This event was documented in the licensees corrective action program as Condition Reports CR-GGN-2014-02219, CR-GGN-2014-02221, and CR-GGN-2014-02224. The failure to control a high radiation area with radiation levels greater than 1000 millirem per hour was a performance deficiency and a violation of Technical Specification 5.7.3. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because it removed a barrier intended to prevent the worker from receiving unexpected dose. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not an as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with procedure adherence, because the licensee failed to follow process, procedures, and work instructions when they did not inventory and ensure control of the dry tube plunger end as it was stored in the horizontal fuel transfer system pool within containment.
05000416/FIN-2014002-012014Q1Grand GulfFailure to Ensure Scaffold Activity Would not Interfere with Fire Brigade ResponseThe inspectors identified a non-cited violation of License Condition 2.C(41), Fire Protection Program, for the failure to adhere to procedural requirements to ensure that scaffold installed in the plant would not prevent or restrict the fire brigade from accessing a certain route used for response to a fire in the area. On February 4, 2014, the licensee installed a scaffold in the containment building for an inspection. The licensees procedure required a walkdown of proposed scaffold to determine if the scaffold would prevent or restrict fire brigade access. The initial reviewer identified that the ladder to access the scaffold would restrict fire brigade access, thus the ladder was not installed until it was required. On March 1, 2014, the ladder was installed for the four hour inspection. Once completed, the licensee failed to remove the scaffold ladder to restore normal access to the area. On March 4, 2014, the inspectors identified that the scaffold ladder was still installed. The inspectors brought their concern to the licensee, who determined that the scaffold would adversely affect the response of fire brigade members to that area of containment. As an immediate corrective action, the licensee removed the scaffold ladder to allow adequate access for the fire brigade members. The licensee documented this issue in Condition Report CR-GGN-2014-02363. The failure to ensure fire brigade members had adequate access passed a scaffold installed in the containment building was a performance deficiency. The performance deficiency was more than minor and therefore a finding because it adversely impacted the protection against external factors attribute of the Mitigating System Cornerstone in that the fire brigades inability to gain access to certain areas in containment could result in preventing prompt extinguishing of fires. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, the inspectors determined that the issue affected the Mitigating Systems Cornerstone and that the finding pertained to a degraded condition while the plant was shutdown for refueling outage RF19. As a result, the inspectors were directed to Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, dated February 28, 2005. The inspectors determined that Appendix G did not address fire brigade issues and solicited input from the senior reactor analyst. The senior reactor analyst performed a detailed risk evaluation and determined that Inspection Manual 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, June 19, 2012, Exhibit 2, Mitigating System Screening Questions, adequately bounded the performance deficiency. The inspectors determined that the finding involved the response time of the fire brigade to a fire, and the finding was of very low safety consequence (Green) because the fire brigades response time was mitigated by other defense-in-depth elements such as area combustible limits were not exceeded, installed fire detection systems were functional, and alternate means of safe shutdown were not impacted. Specifically, there were no combustibles in the area beyond limits, all fire detectors for the area were functional, and the plant was in a shutdown condition with the cavity flooded at the time. The apparent cause of this finding was the work groups involved did not communicate the significance of the impact the scaffold ladder had on fire brigade access to the area and the importance of having the ladder removed upon completion of the work. Therefore, the finding has a cross-cutting aspect in the human performance area associated with team work, in that the individuals and workgroups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained.
05000416/FIN-2014002-032014Q1Grand GulfLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that design control measures be established and implemented to assure that applicable regulatory requirements and the design basis for structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, the licensee failed to implement applicable design bases for the Standby Service Water System Pump 4160 VAC cables being submerged. Specifically, on January 31, 2014, the licensee did not prevent water from submerging the cables in Manhole MH-01 due to a failed sump pump. The inspectors verified that the latest megger tests for the standby service water pump cables were acceptable for demonstrating operability. This finding has been entered into the licensees corrective action program as Condition Reports CR-GGN-2014-00616 and CR-GGN-2014-00768. Using Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that this finding had very low safety significance (Green) because it did not result in the standby service water system becoming inoperable.
05000416/FIN-2014002-042014Q1Grand GulfLicensee-Identified ViolationTechnical Specification (TS) 3.3.6.1, Primary Containment and Drywell Instrumentation, requires the primary containment and drywell isolation instrumentation be operable while in Modes 1, 2, and 3. Contrary to the above, on August 3, 2013, the licensee failed to ensure the primary containment and drywell isolation instrumentation was operable prior to changing from Mode 4 (Cold Shutdown) to Mode 2 (Startup). On August 6, 2013, during a supervisory review of procedures in progress, the licensee determined that they were not incompliance with TS 3.3.6.1 due to jumpers that were installed to disable the function of the instrumentation. The licensee immediately entered the TS 3.3.6.1 Limiting Condition for Operation and associated actions. The licensee restored compliance with the TS by removing the jumpers and restoring the primary containment and drywell instrumentation to operable status and documented this issue in the corrective action program under Condition Report CR-GGN-2013-5101. Using Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, dated June 19, 2012, the inspectors determined that this finding had very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment or drywell and did not involve the hydrogen igniters in the reactor containment.
05000382/FIN-2013003-012013Q2WaterfordFailure to provide design control measures to withstand the effects of flooding on the reactor auxiliary building roofThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, because the licensee did not provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a probable maximum precipitation (PMP) flooding event on the reactor auxiliary building (RAB) roof areas. Specifically, the licensee did not provide an analysis to demonstrate that adequate flood protection existed from the effects of a PMP flooding event on safety-related components and electrical equipment located on the roof of the RAB in the main steam isolation valve (MSIV) wing areas. As a result, the licensee did not perform an analysis to determine if expected ponding levels from a PMP flooding event would challenge safety-related components and electrical equipment such as the emergency feedwater flow control and isolation valves and cables, main steam isolation valves and cables, atmospheric dump valves, and back-up nitrogen accumulator components. The licensee entered this issue into their corrective action program as CR-WF3-2012- 7520. The immediate corrective actions taken to restore compliance included the performance of a preliminary analysis to show that the installed scuppers and roof drains have margin to protect against a local PMP flooding event and that the ponding depth would have little or no affect on the safety-related equipment and cables located in the MSIV wing areas. The failure to provide design control measures for verifying or checking the adequacy of the features designed to withstand the effects of a local PMP on the RAB roof areas was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating System cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the safety-related equipment located on the RAB roof in the MSIV wing areas are required to safely shutdown and maintain the reactor in a cold shutdown condition following accidents and anticipated operational occurrences. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the corrective action program component of the problem identification and resolution area in that the licensee did not identify potential flooding issues completely, accurately, and in a timely manner commensurate with their safety significance.
05000313/FIN-2013003-012013Q2Arkansas NuclearFailure to Evaluate and Correct Excessive Containment Isolation Valve LeakageThe inspectors identified a non-cited violation of Unit 2 Technical Specification 6.5.16, Containment Leakage Rate Testing Program, for the failure to evaluate and take appropriate corrective actions to achieve acceptable performance for containment isolation valves that exceed the local leak rate administrative limit. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-2-2013-01370. The failure to perform a cause determination and take appropriate corrective actions for containment isolation valves that exceed the local leak rate administrative limit was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events, and is therefore a finding. Specifically, the failure to perform a cause determination and take appropriate corrective actions adversely affected the licensee\'s ability to ensure containment isolation valves function properly. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, the finding is determined to have very low safety significance because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components, and the finding did not involve an actual reduction in function of hydrogen igniters in the reactor containment. Since the cause of the performance deficiency occurred more than three years ago, the inspectors concluded that the finding was not representative of current licensee performance and no cross-cutting aspect was assigned.
05000382/FIN-2013003-022013Q2WaterfordFailure to Update Fuel Handling Accident Analysis in the Updated Final Safety Analysis ReportThe inspectors identified a Severity Level IV non-cited violation for the licensees failure to update the final (updated) safety analysis report in accordance with 10 CFR 50.71(e). Specifically, from July 1981 to April 18, 2013, the licensee failed to update the methodology, the data input, and the resulting limits for the fuel bundle drop accident analysis in the Waterford Steam Electric Station, Unit 3, Updated Final Safety Analysis Report (UFSAR), Section 15.7.3.4, Design Basis Fuel Handling Accidents. This violation was entered into the licensees corrective action program as Condition Report CR-WF3-2013-0193. The failure to update the methodology, the data input to the calculation, and the resulting limits for the fuel bundle drop accident analysis in Section 15.7.3.4 of the UFSAR in accordance with 10 CFR 50.71(e) is a performance deficiency. This performance deficiency was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. Consistent with the NRC Enforcement Policy, the inspectors determined that the performance deficiency is a Severity Level IV non-cited violation. This noncited violation had no cross-cutting aspect because there was no finding associated with this traditional enforcement violation.
05000313/FIN-2013003-022013Q2Arkansas NuclearFailure to Correctly Install Control Room Emergency Chiller Supply BreakerInspectors documented a Green self-revealing non-cited violation of Technical Specification 6.4.1.a for the licensees failure to implement procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, the licensee failed to follow procedures for the replacement of the supply breaker for control room emergency chiller 2VE-1A. As a result, the breaker was installed incorrectly and the chiller was inoperable for over two months. Immediate corrective actions included proper installation of the breaker and procedural requirements for visual verification of breaker configuration. The licensee documented the issue in their corrective action program as CR-ANO-2-2013-00233. Inspectors concluded that the failure to follow Procedure 1403.179 for replacement of the train A control room emergency chiller breaker is a performance deficiency. The performance deficiency is more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the 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. Specifically, the loose breaker connection adversely affected the availability and reliability of the control room emergency chiller A. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding required a detailed risk evaluation because it represented an actual loss of function of a single train for longer than its technical specification allowed outage time. The senior reactor analyst performed a detailed risk evaluation using the Arkansas Nuclear One Standardized Plant Analysis Risk models. The dominant risk sequences include a seismically-induced loss of offsite power with the failure of control room emergency chiller A. The analyst assumed that the operators and control room instrumentation could survive a peak control room temperature of 120 F, and that chiller A was susceptible to failure during a seismic event for the 83 days. None of the core damage sequences affected by this performance deficiency were important to the large, early release frequency. Therefore, based on the combined internal and seismic ICCDP of 2.9 x 10-7, this finding was of very low safety significance (Green). The finding was determined to have a cross-cutting aspect in the area of human performance, associated with work practices, in that the licensee failed to use work practices that support human performance. Specifically, licensee personnel were aware of the possibility of misaligning the wire grip style lug, but failed to use adequate self and peer checking to ensure the lug was correctly installed.
05000382/FIN-2013003-032013Q2WaterfordFailure to comply with Action 4 of TS 3.3.1 during shutdown in Modes 4 and 5The inspectors identified a non-cited violation of Waterford Steam Electric Station, Unit 3, Technical Specification (TS) Limiting Condition of Operation (LCO) 3.3.1 because the licensee did not take action to suspend operations that involved reactivity changes to accomplish startup activities with only one excore nuclear instrumentation (ENI) logarithmic (log) channel operable. Specifically, the licensee did not take action to suspend operations involving diluted water additions to the volume control tank and temperature increases with a positive moderator temperature coefficient (MTC) without the required number of operable log channels. As a result, the licensee did not comply with Action 4 of TS LCO 3.3.1 because they did not suspend all operations involving positive reactivity changes with the exception of minimum reactivity additions due to temperature fluctuations or operations, which are necessary to maintain fluid inventory. The licensee entered this issue into their corrective action program as CR-WF3-2013-2166 and CR-WF3- 2013-3182. The immediate corrective actions taken to restore compliance included the discontinued use of water additions to the volume control tank and the increase of RCS temperatures with a positive MTC until the licensees personnel returned an additional log channel to service. The failure to comply with TS LCO 3.3.1, Action 4, was a performance deficiency. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the LCO for the log power channels ensures that adequate information is available to verify core reactivity conditions while shutdown to minimize the probability of the occurrence of postulated events. The inspectors used Checklist 4 contained in Attachment 1 of the NRC Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists, to evaluate this finding. The inspectors determined that the finding did not meet the reactivity guidelines because the licensee did not comply with TS LCO 3.3.1, Action 4. The inspectors determined that the finding was of very low safety significance (Green) because it did not require a quantitative assessment and was not similar to any of the examples requiring a phase two or phase three analyses. The inspectors also determined that the licensee maintain the required shutdown margin to preclude inadvertent criticality in the shutdown condition. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the decision-making component of the human performance area in that the licensee did not make a safety-significant decision using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety was maintained. This included obtaining interdisciplinary input and reviews on safety-significant decisions.
05000382/FIN-2013003-042013Q2WaterfordFailure to submit an LER after discovery that manual handwheels on AOVs were not functionalThe inspectors identified a non-cited violation of 10 CFR 50.73(a)(1) because the licensee did not submit a Licensee Event Report (LER) in a timely manner after the discovery of a reportable event. Specifically, the licensee failed to submit a required LER within 60 days after the discovery of a condition that affected the manual hand-wheel operation of safety related air operated valves following a loss of their corresponding back-up nitrogen accumulators. The licensee determined that the manual hand-wheel function on the essential chiller and emergency feedwater isolation and backup flow control valves did not work. The licensee was aware of the condition that existed but did not adequately evaluate the condition as a part of their reportability review. The licensee entered this issue into their corrective action program as CR-WF3-2013-2564. The immediate corrective actions taken to restore compliance included a new reportability review of the condition and the development of an LER. The failure to submit a required LER within 60 days after discovery of a condition that required a report was a violation of NRC requirements. The inspectors determined that this violation was also a performance deficiency. However, the inspectors determined that the performance deficiency was minor. The inspectors considered this issue to be within the traditional enforcement process because it had the potential to impact the NRC\'s ability to perform its regulatory oversight function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors determined that the violation was a Severity Level IV because it was similar to an example provided in Section 6.9 of the NRC Enforcement Policy. The inspectors did not assign a cross-cutting aspect to this non-cited violation because there was no finding associated with this traditional enforcement violation.
05000382/FIN-2013003-052013Q2WaterfordFailure to provide adequate post modification testing instructions for vibration monitoring on the feedwater piping systemThe inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, because the licensee did not provide post modification testing instructions for activities affecting quality that were appropriate to the circumstances and that included appropriate acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee did not provide adequate post modification testing instructions for vibration monitoring of the feedwater piping system that included appropriate acceptance criteria following the installation of the new replacement steam generators. As a result, the plant experienced an automatic reactor trip and a subsequent down power due to an increase in vibrations on the feedwater piping system without appropriate acceptance criteria and monitoring during power ascension. The licensee entered this issue into their corrective action program as CR-WF3-2013-0445. The immediate corrective actions taken to restore compliance included the implementation of a revised vibration-monitoring plan to include appropriate acceptance criteria and the development of engineering changes to mitigate vibration effects on the feedwater piping system. The failure to provide adequate post modification testing instructions for vibration monitoring of feedwater piping system following steam generator replacement was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors used the NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, to evaluate this issue. The inspectors determined that the finding was of very low safety significance (Green) because the transient initiator did not contribute to the likelihood that mitigation equipment or functions would not be available. The inspectors concluded that the finding reflected current licensee performance and involved a cross-cutting aspect in the operating experience component of the problem identification and resolution area in that the licensee did not implement operating experience through changes to station equipment to support plant safety.
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