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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000530/FIN-2018003-01 | 2018Q3 | Palo Verde | Failure to Maintain Command and Control During a Feedwater Control Valve Malfunction | While reviewing the licensee response to a Unit 3 feedwater pump trip, reactor cutback, reactor trip, and main steam isolation system actuation on June 27, 2018, the inspectors identified that the licensee did not meet the command and control standards outlined in station Procedure 40DP-9OP02 Conduct of Operations, Revision 72. Specifically, senior reactor operators in the control room did not effectively coordinate manual main feedwater output adjustments in the control room or operator actions in the field in response to an apparent valve failure with the activities of non-licensed operators locally evaluating the equipment condition in the field. These uncoordinated actions resulted in a significant plant transient |
| 05000483/FIN-2018002-01 | 2018Q2 | Callaway | Failure to Adequately Assess and Manage Risk Associated with Switchyard Work During a Planned Risk Significant Turbine-Driven Auxiliary Feedwater Pump Equipment Outage | The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at Nuclear Power Plants, for the licensees failure to adequately assess and manage risk associated with switchyard work during a planned risk significant turbine-driven auxiliary feedwater pump equipment outage. Specifically, the licensee failed to properly classify switchyard work and manage the risk as required by Procedures APA-ZZ-00322, Appendix F, Online Work Integrated Risk Management, Revision 16, and ODP-ZZ-00002, Appendix 2, Risk Management Actions for Planned Risk Significant Activities, Revision 13. |
| 05000483/FIN-2018002-02 | 2018Q2 | Callaway | Failure to Establish Maintenance Procedures for Doors that Provide Safety-Related Functions | The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to establish, implement, and maintain procedures associated with door maintenance. Specifically, the licensee failed to establish, implement, and maintain maintenance procedures for doors that provide safety-related functions such as ventilation pressure boundaries. As a result, 15 safety-related doors were identified that either had degraded conditions or that did not have a periodic maintenance task to inspect the doors. |
| 05000483/FIN-2018002-03 | 2018Q2 | Callaway | Failure to Critique an Inaccurate Emergency Classification During a Simulator Training Scenario | The inspectors identified a non-cited violation of 10 CFR 50.47(b)(14) for the licensees failure to critique an inaccurate emergency classification made during licensed operator training. |
| 05000483/FIN-2018002-04 | 2018Q2 | Callaway | Failure of an Analysis of the Impact of Changes to Emergency Action Levels to Demonstrate the Changes Did Not Reduce the Effectiveness of the Emergency Plan | The inspectors identified a non-cited violation of 10 CFR 50.54(q)(3) for the failure of an analysis of the impact of changes to licensee emergency action levels to demonstrate that the changes did not reduce the effectiveness of the emergency plan. |
| 05000528/FIN-2018002-01 | 2018Q2 | Palo Verde | Failure to Re-baseline Valve Stroke Times as Required by ASME OM Code | The inspectors identified a Green, non-cited violation of Palo Verde Technical Specification 5.5.8, Inservice Testing Program, which requires inservice testing of ASME Code Class 1, 2, and 3 components in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). On October 22, 2017, the licensee failed to establish new stroke time reference values for Unit 1 safety injection (SI) valve 660 following maintenance which could affect the valves performance |
| 05000530/FIN-2018002-03 | 2018Q2 | Palo Verde | Failure to Assess the Operability of a Degraded or Nonconforming Structure, System, or Component | The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to evaluate conditions adverse to quality for impacts on the operability of the essential spray ponds. |
| 05000528/FIN-2018002-02 | 2018Q2 | Palo Verde | Failure to Implement and Maintain Procedures Regarding Breathing Air Quality | The inspectors identified a Green, non-cited violation of 10 CFR 20.1703 for failing to implement and maintain written procedures to ensure that respiratory protection equipment (air compressors and bubble hood suites) supplied respirable air of grade D quality or better to radiation workers. |
| 05000483/FIN-2018002-05 | 2018Q2 | Callaway | Minor Violation | Contrary to Technical Specification 3.6.3, Containment Isolation Valves, the licensee failed to maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4. Specifically, the licensee failed to shut the reactor building service air header supply outer containment isolation valve KAV0118 after the fall 2017 refueling outage. As a result, isolation valve KAV0118 was left open from November 25, 2017, through January 11, 2018, which rendered the valves containment isolation function inoperable. The as-found testing demonstrated that the overall containment isolation function, for that penetration, was met with inner containment isolation valve KAV0039 in the normally shut position. Additional information can be found in Licensee Event Report 05000483/2018-001-00, Violation of 20 Technical Specification 3.6.3, Containment Isolation Manual Valve Found in Open Position (ADAMS Accession Number ML18071A208). The licensees failure to comply with Technical Specification 3.6.3, Containment Isolation Valves, and maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4 was a performance deficiency. Screening: The inspectors determined the performance deficiency was minor because it was not a precursor to a significant event, did not have the potential to lead to a more significant safety concern, did not relate to a performance indicator that would have exceeded a threshold and did not adversely impact any of the cornerstone objectives. Specifically, the as-found local leak rate testing demonstrated that containment isolation function was met with inner containment isolation valve KAV0039 in the normally shut position. Enforcement: The failure to comply with Technical Specification 3.6.3, Containment Isolation Valves, and maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000528/FIN-2018001-01 | 2018Q1 | Palo Verde | Inadequate Post Maintenance Test Instructions for Diesel Fuel Oil Transfer Pump | The inspectors reviewed a self-revealed, Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to prescribe appropriate work instructions for maintenance on the Unit 1 diesel fuel oil transfer pump A. Specifically, following power cable maintenance on November 9, 2017, the instructions for conducting a post-maintenance test for the transfer pump were inadequate to detect a high resistance connection in the associated motor control center. |
| 05000528/FIN-2017003-02 | 2017Q3 | Palo Verde | Loss of Refrigerant Failure of Essential Chiller Unit due to Installation of Incorrect Parts | The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 3.7.10 Condition A for exceeding the allowed outage time of 72 hours to restore one inoperable train of essential chilled water system to an operable status. Specifically, the Unit 1 essential chiller B was inoperable from April 11, 2017, to April 18, 2017, due to a refrigerant leak. The licensee entered this issue into their corrective action program as Condition Report 17-05605. The licensees corrective actions included: isolating the automatic purge unit, thereby stopping the leak; refilling the essential chiller with refrigerant; and retesting the essential chiller unit to return it to an operable status on April 18, 2017. Additionally, the licensee checked the other five essential chillers across the station and found no additional material deficiencies.The inspectors determined that the failure to ensure the correct Swagelok fitting was being installed in accordance with station procedure is a performance deficiency. The performance deficiency is more than minor and a finding because it is 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. Specifically, on April 18, 2013, the licensee installed the incorrect Swagelok fitting during maintenance on the essential chiller. When the licensee placed the auto purge system in service, this resulted in the refrigerant leaking out of the Swagelok fitting rendering the essential chiller inoperable.The inspectors performed the initial significance determination 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 Questions, Step A.3 which required a senior reactor analyst to perform a detailed risk evaluation because essential chiller B was incapable of performing its safety function for greater than its technical specification allowed outage time. A regional senior reactor analyst performed a detailed risk evaluation and determined that the finding was of very low safety significance (Green). Essential Chiller 1B was assumed to be unavailable for 8 days and the potential for common cause failure on the remaining essential chiller was assumed. This resulted in a change in core damage frequency of 3.6E-7 per year. Losses of offsite power comprised the most dominant core damage sequences. The emergency diesel generators and the emergency feed water systems remained available for mitigation of the dominant sequences.The inspectors determined this finding had a cross-cutting aspect in the area of human performance, avoid complacency, in that the licensee failed to recognize and plan for the possibility of latent issues or mistakes. Specifically, the licensee failed to provide an appropriate post-maintenance testing procedure as required by station procedure. The work order executed on April 11, 2017, gave no direction to test for leaks on the filter assembly (H.12). |
| 05000529/FIN-2017003-03 | 2017Q3 | Palo Verde | Failure to Follow Conduct of Operations Procedure | The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 5.4.1.a. Procedures, for the licensees failure to implement their Conduct of Operations procedure. Specifically, licensee personnel improperly performed a reactor coolant pump seal injection filter flushing evolution as a skill of the craft task without written instructions. Consequently, Unit 2 experienced a loss of letdown and exceeded the pressurizer level technical specification limit of 56 percent. Licensed operators took immediate corrective actions to restore letdown and lower pressurizer level to within acceptable limits. The licensee entered this issue into their corrective action program as Condition Report 17-09326.The inspectors determined that the failure to follow the Conduct of Operations procedure for performance of skill of the craft tasks is a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it was associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the decision to perform the reactor coolant pump seal filter flushing evolution without a controlled procedure allowed operators to place the system in a configuration causing an automatic isolation of the letdown system that challenged the availability of the pressurizer to respond to reactor coolant system pressure transients. The inspectors evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the finding was of very low safety significance (Green) because it only contributed to the likelihood of a reactor trip and not the likelihood that mitigation equipment or functions would not be available. The inspectors determined 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, latent issues, and inherent risk. Specifically, licensee personnel did not recognize the inherent risks associated with the reactor coolant pump seal filter flushing evolution before proceeding to perform the task without formal written instructions (H.12). |
| 05000530/FIN-2017003-01 | 2017Q3 | Palo Verde | Failure to Initiate Corrective Actions for Thermography Tests | The inspectors reviewed a self-revealed, Green finding for the licensees failure to initiate corrective actions to address elevated temperature measurements identified during thermography inspections of the Unit 3 Phase C main transformer control cabinet. As a result, an extended loss of cooling to the Phase C main transformer resulted in a manual trip of the main turbine and a reactor power cutback. This issue was entered into the licensees corrective action program under Condition Report 17-09022, and the licensee took immediate actions to reinsert and tighten a loose wire associated with the transformer cooling control circuitry. The inspectors determined that the failure to follow procedure 37TI-9ZZ01, Thermography Inspection of Plant Components, Revision 8, Step 4.5.10.1 to initiate a condition notification report following the identification of elevated temperatures during thermography inspections is a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it was associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions duringshutdown as well as power operations. Specifically, the failure to initiate corrective actions following the identification of the hot spot on the Unit 3 Phase C main transformer 4-8 contactor resulted in a reactor power cutback that upset plant stability. Using NRC Manual Chapter 609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the finding screened as having very low safety significance (Green) because the deficiency resulted in a reactor trip, but mitigation equipment remained unaffected. The inspectors determined this finding had a cross-cutting aspect in the area of problem identification and resolution, identification, in that the licensee failed to identify issues completely, accurately, and in a timely manner in accordance with the corrective action program. Specifically, on three occasions in 2016 and 2017, the licensee collected data indicating potential loose connections at the 4-8 contactor, but failed to recognize and communicate the data in accordance with the corrective action program (P.1). |
| 05000530/FIN-2017003-04 | 2017Q3 | Palo Verde | Reactor Trip due to Pressurizer Spray Valve Failing Open due to Volume Booster Internals Not Environmentally Qualified for Anticipated Ambient Operating Temperatures | The inspectors reviewed a self-revealed, Green, non-cited violation of Technical Specification 5.4.1.a Procedures, for the licensees failure to follow station procedure 73DP-0EE05, Engineering Preventive Maintenance Program. The licensee did not consult design basis resources and operating experience when changing the preventive maintenance frequency of the pressurizer spray valve air-operated volume boosters. The valve internals were not rated for ambient operating temperature conditions, as a result a pressurizer spray valve failed open, requiring operators to trip the reactor. The licensee entered this condition into their corrective action program as Condition Report 16-14219. The licensees corrective actions included replacing the affected pneumatic volume boosters with high temperature qualified soft parts and by revising procedure 73DP-0EE05 to ensure a more thorough engineering management oversight of the equipment reliability engineering template process. The inspectors determined that the failure to follow station procedure 73DP-0EE05, Engineering Preventive Maintenance Program, Revision 6, Step 3.4.7, to consult design basis information including internal operating experience resources when determining a required preventive maintenance frequency is a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the pressurizer spray valve failed open requiring the operators to trip the reactor. The inspectors evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors determined that the finding was of very low safety significance (Green) because the finding 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. Specifically after the reactor trip, control room operators were able to regain pressure control by securing the reactor coolant pumps driving pressurizer spray, and initiating auxiliary spray through the charging system. The inspectors determined this finding had a cross-cutting aspect in the area of human performance, consistent process, in that the licensee failed to use a systematic approach to make decisions including incorporating risk insights. Specifically, the pressurizer spray valves are designated as critical components and single point vulnerabilities in 73DP-0EE05, which requires a technical basis to allow for a preventive maintenance frequency change. The licensee did not document the technical basis to increase the service life from one to four cycles (H.13). |
| 05000528/FIN-2017002-01 | 2017Q2 | Palo Verde | Inoperable Containment Isolation Valve Due toNot Operating Valve in Accordance with Station Procedures | The inspectors reviewed a Green self-revealing non-cited violation of Technical Specification 3.6.3 Condition C for exceeding the allowed outage time of 4 hours to isolate the flow path of an inoperable containment isolation valve. Specifically, Unit 1 containment isolation valve SG-1134 was inoperable from June 28, 2016, to September 21, 2016, due to improper restoration from planned maintenance. The licensee entered this condition in their corrective action program and performed a Level 2 cause analysis under Condition Report 16-14896. The licensee also undertook immediate actions to restore the valve from the neutral position and remotely stroke the valve per procedure.The inspectors concluded the failure to restore Unit 1 containment isolation valve SG-1134 from maintenance in accordance with station procedures was a performance deficiency. The performance deficiency was more-than-minor and a finding because it is associated with the configuration control attribute of maintaining functionality of containment under the Barrier Integrity cornerstone which affects the cornerstone objective to provide reasonable assurance that physical design barriers will protect the public from radionuclide releases caused by accidents or events. Specifically, the inoperability of this containment isolation valve allowed the potential of a radioactive release during a design basis accident. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix H, Containment Integrity Significance Determination Process, Issue Date: 05/06/04. Section 4.1 determined this to be a Type B finding since the degraded condition did not affect the likelihood of core damage. Table 4.1 shows that containment isolation valves in lines connecting reactor coolant systems to environments with small lines would not contribute to large early release frequency. Since valve SG-1134 is a small (one-inch) valve, this finding screened to Green using the flow chart in Figure 4.1 LERF-based Significance Determination Process. This finding has a cross-cutting aspect in the area of human performance associated with the documentation component. Specifically, the licensee failed to provide a work package that was complete, thorough, accurate, and current in accordance with station procedure 40OP-09OP01, Operation of Air Operated Valves, when returning SG-1134 to its normal operating condition following maintenance. As a result, the valve handwheel was left out of neutral, thereby preventing remote operation (H.7). |
| 05000528/FIN-2017002-02 | 2017Q2 | Palo Verde | Licensee-Identified Violation | Title 10 CFR 50.55a(g)(4), Inservice Inspection Standards Requirement for Operating Plants, states, in part, Throughout the service life of a pressurized water-cooled nuclear power facility, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the ASME Code. The ASME Code, Section XI, Article IWA-2610, requires that a reference system be established for all welds and areas subject to a surface or volumetric examination. This includes identifying each weld that is subject to ASME Section XI requirements.Contrary to the above, prior to April 12, 2017, the licensee failed to establish a reference system for all welds and areas subject to a surface or volumetric examination. Specifically, five welds located in an ASME Code, Section XI, Class 2, train A and train B refuel water suction lines were not identified as applicable ASME Section XI welds. The licensee restored compliance by correctly reclassifying the subject welds and entering them in the ASME Section XI program. The finding was of very low safety significance(Green) because the finding did not represent an actual loss of safety function of a system or train and did not result in the loss of a single train for greater than technical specification allowed outage time. This issue was entered into the licensees corrective action program as Condition Report 17-05607. |
| 05000529/FIN-2017001-01 | 2017Q1 | Palo Verde | Failure to establish station procedure instructions for denial work authorizations | The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the failure to establish procedure instructions for work authorization denials or deferrals. Specifically, this led to a 60 day extended unavailability of the diverse auxiliary feedwater actuation system when corrective maintenance was inappropriately deferred by the operations department. Failure to provide adequate procedural guidance in the event of a denied work authorization, a circumstance anticipated to occur, is a performance deficiency. The performance deficiency is more than minor, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability and reliability of equipment that responds to an initiating event. Specifically, because the corrective maintenance was not performed in a timely manner, both trains of the diverse auxiliary feedwater actuation system remained in bypass for an additional 60 days whereby the system was not capable of performing its required safety function. The inspectors evaluated the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, Section A, Question 2, which required a detailed risk evaluation because the finding involved a loss of system safety function. A Region IV senior reactor analyst performed a detailed risk assessment of the finding and determined that the finding was of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect in the human performance area of Work Management. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the Unit Operations Managers decision to deny the work authorization was based on conservative but faulty assumptions, and if other work groups with greater specific technical knowledge had been involved, the corrective maintenance likely would have proceeded (H.5) |
| 05000529/FIN-2016004-01 | 2016Q4 | Palo Verde | Inadequate monitoring of MSIV nitrogen pre-charge pressure | The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.7.2 for exceeding the Condition A completion time for an inoperable main steam isolation valve (MSIV) single actuator train and not immediately declaring the affected main steam isolation valve inoperable in accordance with Condition E. Specifically, the Unit 2 main steam isolation valve 171 actuator A was inoperable from July 30, 2016, to August 9, 2016, when a known nitrogen leak was not adequately monitored. The licensees inadequate monitoring allowed the nitrogen pre-charge pressure in the actuator to decrease to below the minimum acceptable limit for operability. The licensee restored the pre-charge pressure and entered this issue into their corrective action program as Condition Report 16-12740. The failure to perform adequate monitoring for a degraded condition as required by procedure 40DP-9OP26, Operations Condition Reporting Process and Operability Determination/Functional Assessment, was a performance deficiency. The performance deficiency was more-than-minor and therefore a finding because it affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically the failure to adequately monitor a known nitrogen leak resulted in depressurizing one of two hydraulic accumulators thereby reducing the reliability of the system to initiate a fast closure of MSIV 171 upon receipt of a main steam isolation signal. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, Issue Date: June 9, 2012. The finding required a detailed risk evaluation since it represented a loss of function for a single train for greater than the Technical Specification allowed outage time. A Region IV senior reactor analyst determined the finding was of very low safety significance (Green) since the MSIV remained capable of performing its safety function with the alternate actuator. The finding has a cross-cutting aspect in the area of human performance associated with the teamwork component. Specifically, the licensee failed to coordinate activities across organizational boundaries in that the operations personnel did not obtain engineering input to ensure that additional monitoring requirements for the nitrogen pre-charge leak were adequate to verify continued MSIV 171 operability (H.4). |
| 05000528/FIN-2016002-02 | 2016Q2 | Palo Verde | Failure to Implement High Radiation Area Controls in an Area with a Dose Rates Greater Than 1 rem per Hour | The inspectors reviewed a Green, self-revealing, non-cited violation of Technical Specification 5.7.2, which was caused by the licensees failure to control a high radiation area with radiation levels greater than 1 rem per hour in the Unit 1 containment. A radiation protection technician received an unexpected dose rate alarm while conducting surveys on piping in the 87-foot elevation of the 2B reactor coolant pump bay area near a high efficiency particulate air unit in containment. Licensee personnel corrected the error by guarding the area, posting the area, and changing the pre-filters in the adjacent portable a high efficiency particulate air units to reduce the dose rates. This issue was entered into the licensees corrective action program as Condition Reports 16-06515 and 16-07479. The inspectors determined that the failure to identify a locked high radiation area through timely surveys and adequate a high efficiency particulate air maintenance procedures that could have revealed changing radiological conditions was a performance deficiency. 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 licensee personnel did not implement barriers intended to prevent workers 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 had very low safety significance (Green) because: (1) it was not an as low as is reasonably achievable 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 finding has a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee leaders failed to ensure that personnel, equipment, and procedures were available and adequate to support nuclear safety. Specifically, the licensee failed to ensure that procedures were adequate to ensure radiation levels around portable high efficiency particulate air units were monitored to evaluate changing radiological conditions in a timely manner such that hazards were appropriately controlled (H.1). |
| 05000528/FIN-2016002-03 | 2016Q2 | Palo Verde | Inadequate Engineering and Radiological Controls Resulting in a Unit 1 Containment Building Airborne Radioactivity Event with Unplanned Intakes | The inspectors identified a non-cited violation of 10 CFR 20.1701 due to the licensees failure to implement adequate processes and engineering controls necessary to reduce airborne radioactivity and prevent internal dose to workers in Unit 1. On April 20, 2016, inspectors identified that procedures and instructions for monitoring high efficiency particulate air ventilation filter unit to prevent worker exposures to radiation and airborne radioactivity were being inadequately implemented. On April 21, 2016, the licensees inadequate engineering and radiological controls during a high efficiency particulate air operations caused an airborne radioactivity event in containment, resulting in the evacuation of 41 potentially contaminated workers of whom 8 had measurable intakes of radioactive material. The licensees immediate corrective actions included stopping work in the Unit 1 containment, evacuating workers in containment, assessing workers for external and internal contamination, and investigating the cause and source of the contamination event. This matter was placed in the licensees corrective action program as Condition Reports16-06499 and 16-06578 and the licensee initiated a root cause investigation. The inspectors determined that the failures to implement adequate engineering and radiological controls to reduce airborne radioactivity during a high efficiency particulate air unit operations in accordance with 10 CFR 20.1701 and radiation protection procedures were performance deficiencies. The performance deficiencies were more than minor because they were associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. This was evident by the Unit 1 containment airborne radioactivity event on April 21, 2016, that resulted in at least eight workers with unplanned intakes. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding had very low safety significance (Green) because: (1) it was not an as low as is reasonably achievable planning and controls 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. The inspectors concluded that the finding has a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee leaders failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, procedures and radiation exposure permits failed to have adequate instructions for ensuring a high efficiency particulate air filter loading and dose rates were monitored to prevent overloading, and safe handling of loaded a high efficiency particulate air filters (H.1). |
| 05000528/FIN-2016002-01 | 2016Q2 | Palo Verde | Leakage From Reactor Coolant Pump 2B Discharge Pipe Instrument Nozzle | The inspectors identified an unresolved item for pressure boundary leakage from reactor coolant pump 2B discharge pipe instrument nozzle. On April 10, 2016, during the Unit 1 Refueling Outage 19, the licensee discovered reactor coolant system pressure boundary leakage at instrument nozzle 1JRCETW0121Y on the 2B reactor coolant pump discharge piping. The leakage was discovered during a planned visual inspection of Unit 1 hot and cold leg nozzles. The leak was not detectable by either the reactor coolant system leak rate procedure or the containment radiation monitor trend reviews while the unit was operating. Additionally, the leak had not been visually detected during the previous refueling outage. The leakage was consistent with a small leak characterized by moderate boric acid accumulation at the leakage site. The licensee determined that the cause of the leakage was primary water stress corrosion cracking of the Alloy 600 instrument nozzle. The licensee corrected the leakage using a mechanical nozzle seal assembly repair method utilizing ASME Code Case N-733, Mitigation of Flaws in NPS 2 (DN 50) and Smaller Nozzles and Nozzle Partial Penetration Welds in Vessels and Piping by Use of a Mechanical Connection Modification, Section XI, Division 1. The evaluation of the 2B cold leg RTD nozzle leakage is being evaluated by the licensee as part of Palo Verde Action Request 15-01640-012. The inspectors reviewed the circumstances surrounding the discovery of the leak and observed portions of the repair activity during the refueling outage. Once the licensee completes their evaluation, the inspectors will review and complete an inspection to determine if a performance deficiency exists as a result of the nozzle failure. |
| 05000530/FIN-2016001-01 | 2016Q1 | Palo Verde | Failure to use adequate engineering and radiological controls resulting in two unplanned intakes | A self-revealing non-cited violation of 10 CFR 20.1701 was identified for the licensees failure to implement adequate processes or engineering controls to control the concentration of radioactive material in air and prevent internal dose to workers. Specifically, on April 14, 2015, the licensee implemented inadequate engineering and radiological controls to remove a pre-filter and Y-connector from a high efficiency particulate air (HEPA) ventilation unit resulting in an airborne radioactivity condition and two intakes. The licensee was alerted to this issue when two radiation protection technicians alarmed PM12 portal monitors upon their exit from the radiologically controlled area. The licensee took immediate corrective actions and instructed these technicians to report to dosimetry for whole body counting and evaluation. The licensee entered this issue into their corrective action program as Condition Report (CR) CR 16-01093. The failure to implement adequate engineering and radiological controls during HEPA unit maintenance in accordance with procedures and the radiological exposure permit requirements was a performance deficiency. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety attribute of Program and Process and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. This was evident by two workers receiving unplanned intakes. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issue date 8/19/2008, the finding was determined to be of very low safety significance (Green) because it did not involve: (1) as low as reasonably achievable (ALARA) planning and controls, (2) an overexposure, (3) a substantial potential for an overexposure, or (4) an impaired ability to assess dose. The inspectors concluded that the finding has a Conservative Bias cross-cutting aspect in the Human Performance area because the licensee failed to use decision-making practices that emphasized prudent choices over those that are simply allowable when they changed out the HEPA pre-filter and Y-connector components (H.14). |
| 05000530/FIN-2016001-02 | 2016Q1 | Palo Verde | Fatigue failure of pneumatic fitting due to excessive vibrations | The inspectors documented a self-revealing non-cited violation of Technical Specification 3.7.2 Condition A for exceeding the allowed outage time of seven days. Specifically Unit 3s MSIV-181 actuator B was found to be inoperable from May 1, 2015 until August 15, 2015 when a design change installed a new swivel type fitting on an air-line without taking into account vibrational forces, as required by the stations procedure. This eventually resulted in the fatigue failure of the fitting, depressurizing the actuator B to less than 5000 psig. The licensee entered this condition in their corrective action program and performed a Level 2 cause evaluation under Condition Report 15-02686. The inspectors concluded that the failure to take into account excessive vibrational stresses as required by procedure 81DP-0EE10, Design Change Process Step J.2.9.1, when implementing the design change was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Mitigating Cornerstone to ensure the availability, reliability, and the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically the failure to account for the vibrational stresses resulted in the fatigue failure of the air-line fitting which depressurized one of two hydraulic accumulators thereby reducing the reliability of the system to initiate a fast closure of MSIV-181 upon receipt of a Main Steam Isolation Signal. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, Issue Date: 06/19/12. The finding screened as Green since the MSIV remained capable of performing its safety function with the alternate accumulator. The finding has a cross-cutting aspect in the area of human performance associated with the avoid complacency component. Specifically the licensee assumed there were no factors affecting the mechanical design requirements beyond the performance requirements. As a result the licensee failed to perform a thorough review of the mechanical conditions (such as vibrations) the air-line was subjected. |
| 05000528/FIN-2016008-01 | 2016Q1 | Palo Verde | Operations Department Failure to Document Conditions Adverse to Quality in Condition Reports | The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to document conditions adverse to quality in the corrective action program. Previous similar failures to initiate condition reports led to, or contributed to, two significant conditions adverse to quality over the last 15 months. The failure of the operations department to document identified conditions adverse to quality in condition reports, as required by Procedure 01DP-0AP12, Condition Reporting Process, Revision 23, was a performance deficiency. This 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, on two other occasions since January 2015, failures by operations personnel to write condition reports for equipment-related problems resulted in or contributed to significant conditions adverse to quality. This performance deficiency demonstrated a continued gap within Palo Verde Nuclear Generation Stations operations department in understanding condition report initiation criteria. This performance deficiency is associated with the mitigating systems cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, the team determined that this finding was of very low safety significance (Green) because it did not affect the operability or functionality of a mitigating structure, system, or component. This finding has a resolution cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3). |
| 05000529/FIN-2016001-03 | 2016Q1 | Palo Verde | Licensee-Identified Violation | Technical Specification 5.4.1, Procedures, requires that procedures be established, implemented, and maintained covering the applicable procedures in Regulatory Guide 1.33. Regulatory Guide 1.33, Appendix A, Section 9 requires, in part, that maintenance that can affect the performance of safety-related equipment be properly preplanned and performed in accordance with written procedures. Contrary to the above, prior to October 1, 2015, licensee work management personnel failed to perform an activity affecting quality in accordance with written procedures. Specifically, the licensee did not conduct an adequate review of technical specification LCO implications of a planned Unit 2 essential spray pond outage in accordance with procedure 51DP-9OM08, Look Ahead Process. Work planners did not recognize that the removal of two spray pond piping spool pieces was an activity required to restore spray pond system operability and therefore did not establish a tracking mechanism to ensure that the spool pieces were removed before the Unit 2 essential spray pond A was declared operable. Consequently, the Unit 2 essential spray pond A would not have been able to provide cooling to the essential cooling water heat exchanger following a seismic event. The inspectors evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012. Inspectors concluded the finding was of very low safety significance (Green) because all questions in Exhibit 2 could be answered no. The licensee entered the issue into the corrective action program as CR 15-08352. The licensee now plans and controls the removal and re-installation of spray pond spool pieces using the stations temporary modification process. |
| 05000397/FIN-2016001-01 | 2016Q1 | Columbia | Licensee-Identified Violation | Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, prior to November 17, 2015, the licensee failed to establish measures to assure that conditions adverse to quality are promptly identified and corrected. Specifically, in October 2012, the licensee identified in AR 271801 that the maintenance procedure for Square D QMB electrical disconnects, specified in procedure PPM 10.25.53, Inspection of Lighting Panels and Power Panels, Revision 10, did not include steps to clean and burnish contacts that are susceptible to corrosion that may yield a high-resistance connection. However, the licensee failed to identify that several installed QMB disconnects may be vulnerable to failure since the previous maintenance performed did not include the steps to clean and burnish the contacts. Consequently, on November 17, 2015, the 125 VDC circuit (E-DP-S1/2D circuit 6) associated with under voltage trips of the division 2 vital bus failed a monthly surveillance test due to degraded voltage from high-resistance connections on corroded contacts. The licensee implemented corrective action by declaring affected components inoperable per technical specifications, identified high-resistance contacts as the cause, burnished the contacts to restore the circuit, and re-performed the surveillance to establish operability. The licensee also performed relay testing to demonstrate 125 VDC circuit availability at the observed, degraded voltages. The inspectors assessed the finding in accordance with Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012. Using Exhibit 2 of IMC 0609, the inspectors determined the finding was of very low safety significance (Green) because the finding did not represent a loss of safety function, did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time, and did not represent an actual loss of function of one or more non-technical specification equipment for greater than 24 hours. This violation was entered into the licensees corrective action program as AR 340134. |
| 05000528/FIN-2016008-02 | 2016Q1 | Palo Verde | Licensee-Identified Violation | Title of 10 CFR 50.55a(f)(4), requires, in part, that pumps and valves classified as ASME Code Class 1, 2, or 3 must meet the inservice test requirements set forth in the ASME Operation and Maintenance (OM) Code and addenda to the extent practical within the limitations of design, geometry, and materials of construction of the components. The inservice testing program is incorporated into the Palo Verde Nuclear Generation Station licensing basis under Technical Specification 5.5.8 and governed by the procedures controlled under that specification. ASME OM Code Case OMN-1 was adopted by Palo Verde Nuclear Generating Station per Valve Relief Request number 1, and approved by the NRC as an alternative for performing Code-required valve and pump testing for the second and third 10-year testing intervals (January 1998-2018). ASME OMN-1, Section 3.3.1(b) requires that, if insufficient data exist to determine the inservice test frequency...then (motor operated valve) MOV inservice testing shall be conducted every two refueling cycles or three years until sufficient data exist to determine a more appropriate test frequency. Palo Verde Nuclear Generating Station Procedure 73DP-9ZZ12, Motor Operated Valve Program, Section 4.5.4.5, and Appendix H, define when sufficient test data exists to justify increasing test frequencies beyond 3 years. This criteria includes completing at least two complete diagnostic testing cycles constituting a baseline pre-service test and two subsequent as-found tests. These testing requirements are invoked after complete replacement of the valve, installation of a new valve, or major maintenance, which could substantially change the valve/actuator performance. Contrary to the requirements listed above, the licensee failed to perform Code-required testing for a total of 17 valves between 2008 and 2016. The licensee identified an issue in May 2015 with the testing frequency of five valves after a modification installed new motor operated valves in the charging system. An extent of condition was performed and 11 additional valves were identified as being noncompliant. An engineering evaluation was performed to assess and manage the risk of not completing the required ASME testing per Technical Specification Surveillance Requirement 3.0.3. A prompt operability determination was also performed to provide reasonable assurance of operability until the valves could be tested again. In January 2016, an additional valve was identified as being non-compliant and a separate operability evaluation was completed to provide reasonable assurance that the valve would still perform its function. There are currently seven valves that are still in non-compliance with the Coderequired testing frequency; all other valves have been tested satisfactorily and are now in compliance. Those still requiring testing are scheduled during their respective next available system windows. This violation is of very low safety significance (Green) because the non-conforming valves were determined to have reasonable assurance of operability. The licensee entered the condition into its corrective action program and initiated corrective actions to restore compliance under Condition Report 15-02470. |
| 05000530/FIN-2015004-01 | 2015Q4 | Palo Verde | Licensee-Identified Violation | Technical Specification 3.0.4 requires, in part, that when an LCO is not met, entry into a mode or other specified condition in the applicability shall only be made when the associated actions in the mode permit continued operation; a risk assessment is performed and accepted for the inoperable components; or when an allowance is stated. Technical Specification 3.7.4, Atmospheric Dump Valves, requires that four ADV lines shall be operable in Modes 1, 2, 3, and 4 when the steam generator is relied upon for heat removal. Contrary to the above, on May 1, 2015, Unit 3 operators entered a mode with an LCO not met. Specifically, one atmospheric dump valve line was not operable as required by Technical Specification 3.7.4 prior to entering Mode 3. The licensees investigation concluded that the valve failure was a result of inadequate reassembly following maintenance. The licensee reported this condition in Licensee Event Report 05000530/2015-002-00 as a condition prohibited by Technical Specifications due to entering a mode in the applicability of LCO 3.7.4 while the LCO was not met. The inspectors concluded that the finding is of very low safety-significance (Green) because it was not a design or qualification deficiency, did not result in a loss of safety function, did not result in a loss of function of a train of safety equipment out greater than its allowed outage time, or a loss of function of high importance maintenance rule equipment greater than 24 hours. The licensee has entered the issue in the corrective action program as CRDR 4654422. |
| 05000528/FIN-2015004-02 | 2015Q4 | Palo Verde | Licensee-Identified Violation | Title 10 CFR 55.49, Integrity of examinations and tests, requires, in part, that facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Contrary to the above, during the week of November 9, 2015, the licensee caused a compromise of examination integrity when two licensed operators, who had previously validated portions of the 2015 annual operating test and had signed the examination security agreement, administered emergency preparedness (EP) job performance measures (JPMs) to a total of three licensed operators who had not yet taken their annual operating test. Specifically, the two licensed operators validated and/or approved simulator scenarios and EP JPMs for the annual operating test and then subsequently administered JPMs to three other licensed operators for the purpose of supporting EP program indicators. If not for detection, this activity could have affected the equitable and consistent administration of the annual operating examination. The failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV violation because it involved a nonwillful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because it had no actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Requalification. The licensee entered this issue into their corrective action program as Condition Report 15-10910. |
| 05000285/FIN-2015003-01 | 2015Q3 | Fort Calhoun | Failure to Maintain Safety Injection Tank Boron Concentration within Technical Specification Limits | A Green, self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI Corrective Action was identified because the licensee failed to identify and evaluate an adverse trend related to boron concentration in Safety Injection Tank (SIT) SI-6A and to take corrective actions to prevent boron concentration from going below the minimum concentration required by Technical Specifications. The licensees immediate corrective actions included documenting this condition in their corrective action program in Condition Report (CR) 2015-10181, declared SI-6A inoperable, and raised SI-6A boron concentration. The finding is more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone, in that this finding resulted in the SIT becoming inoperable when boron concentration fell below TS limits for approximately 8.5 days prior to August 20, 2015. Analysis conducted by a Senior Reactor Analyst determined the finding to be of very low safety significance (Green), primarily because the SIT function is needed only for mitigation of a postulated large-break loss of coolant accident, and the initiating-event frequency for such accidents is 2.5 x 10-6/year. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution and the Evaluation aspect, because the licensee did not thoroughly evaluate the issue and ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance. |
| 05000285/FIN-2015003-02 | 2015Q3 | Fort Calhoun | Failure to Maintain Fire Watch and Fire Watch Logs | Inspectors identified a Green, Severity Level IV, non-cited violation of 10 CFR 50.9(a), Completeness and Accuracy of Information, for the licensees failure to maintain the required fire watch logs complete and accurate in all material respects. The licensee entered this into their corrective action program as Condition Reports (CR) 2014-06416 and 2014-06680. This finding is more than minor because it adversely affected the human performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding has very low safety significance (Green) because it did not impact the ability to achieve safe shutdown. This findings severity level is based on an example in the Enforcement Policy, Section 6.1.d.2, which states, in part, that Severity Level IV violations involve violations of 10 CFR 50.59 (which) result in conditions evaluated as having very low safety significance. |
| 05000529/FIN-2015002-02 | 2015Q2 | Palo Verde | Failure to Take Timely Corrective Actions to Prevent Charging Pump Discharge Bladder Failure | The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50 Appendix B, Criterion XVI for the failure to take timely corrective actions associated with failure of the discharge pulsation dampener poppet valves in the positive displacement charging pump. The charging system. is designated as quality related for its function to provide a boration flowpath to the reactor coolant system. Specifically, following the investigation of a degrading discharge dampener bladder on the Unit 2 charging pump E and the discovery that the poppet valve stem was galled and stuck in the poppet valve seat, the licensee incorrectly concluded that routine monthly monitoring and the 5-year bladder replacement maintenance would identify further failures in the other charging system trains. The licensee entered this issue into the corrective action program as Condition Report 15-4230. Failure to take timely corrective actions to replace the charging pump discharge dampener poppet valve assemblies was a performance deficiency. The performance deficiency was more-than-minor and is a finding because it is associated with the equipment performance attribute and directly affected the Initiating Event Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to correct this condition adverse to quality resulted in a reactor coolant system transient and challenged normal plant operations. Using Manual Chapter 0609, Appendix A, "Significance Determination Process (SDP) for Findings At Power," the inspectors determined the finding was of very low safety significance (Green) because the finding did not result in a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has an evaluation cross-cutting aspect in the area of problem identification and resolution because the organization did not thoroughly evaluate issues to ensure that resolutions address causes and extent of condition commensurate with their safety significance. Specifically, the corrective actions taken in response to the January 2014 poppet galling event included a number of engineering judgements and assumptions regarding both the degradation mechanism, and the internal workings of the sytem components were used to justify not performing additional poppet assembly inspections. These assumptions were known to be incorrect by uninvolved technical experts inside the licensee and vendor organization. Had those assumptions been properly vetted and verified by vendor or other industry experts at the time, the extent-of-condition inspections likely would have been accelerated (P.2). |
| 05000529/FIN-2015002-05 | 2015Q2 | Palo Verde | Failure to Establish Adequate Procedures to Respond to a Total Loss of Charging Event | The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, through Regulatory Guide 1.33, Revision 2, Appendix A, Section 6.t, February 1978 for the licensees failure to establish adequate procedures for combating emergencies and other significant events regarding a total loss of charging pumps due to gas binding that affected reactor coolant system pressure and level control. On March 20, 2015, after Unit 2 experienced a total loss of charging, operators relied on a normal operating procedure which did not address how to combat a total loss of charging flow due of gas binding from a failed discharge pulsation dampener. The licensee entered this issue into the corrective action program as Condition Report 15-4230. The failure to provide adequate procedures for combating emergencies and other significant events regarding a total loss of charging pumps due to gas binding that affected reactor coolant system pressure control was a performance deficiency. The performance deficiency was more-than-minor and is a finding because it is associated with the procedure quality attribute and directly affected the Initiating Event Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the lack of adequate procedural guidance challenged reactor operators during the loss of charging event. In accordance with Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process (SDP) for Findings AtPower," the performance deficiency was determined to be of very low safety significance (Green) because the finding did not result in a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance because the decision to eliminate the abnormal operating procedure and not to train reactor operators was made in 1997. |
| 05000530/FIN-2015002-04 | 2015Q2 | Palo Verde | Notice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System Operating Conditions B and C | (Open) Unresolved Item 05000530/2015002-04, TAC Number MF6276 - NOED Number 15-4-01. Notice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System - Operating Conditions B and C On May 27, 2015, the licensee removed Unit 3 high pressure safety injection train A for planned maintenance. The following morning, during the maintenance, the licensee noted lube oil contamination, and determined that an outboard motor bearing had apparently failed during the last run following maintenance during the last refueling outage which involved disassembling and reassembling the bearing. The licensee identified procedural guidance inadequacies in the reassembly procedure that were the likely cause of the failure. The licensee could not perform required repairs in a controlled manner within the remaining action statement completion time, so on May 29, 2015, the licensee requested a Notice of Enforcement Discretion for a one-time action statement extension of 24 hours to allow time to reassemble and test the replacement bearings prior to restoring operability. The NRC granted that request as NOED 15-4-01. The licensee completed maintenance, testing, and restoration approximately 11 hours into the 24-hour extension window. In accordance with Inspection Manual Chapter 0410, Unresolved Item (URI) 05000530/2015002-04 is opened for NOED 15-4-01, and remains open pending further inspection and disposition in a future inspection report. |
| 05000528/FIN-2015002-01 | 2015Q2 | Palo Verde | Failure to Verify the Design of the Essential Spray Pond System Crosstie Valves | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to maintain adequate design control measures associated with the ultimate heat sink. Specifically, the essential spray pond crosstie valves did not meet design requirements established in Regulatory Guide 1.117, "Tornado Design Classification," as described in the Updated Final Safety Analysis Report. Consequently, if the crosstie valves were damaged by a tornado, the licensee would not have enough available water inventory to meet the mission time of the essential spray pond system. The licensee has added steps to their emergency operating procedure to instruct operators to open the crosstie valves to achieve and maintain long-term cooling subsequent to a design-basis tornado event, and is evaluating potential plant modifications. The licensee has entered this issue into the corrective action program as Palo Verde Action Request 4633058. The failure to verify the design of the essential spray pond system in accordance with Regulatory Guide 1.117 was a performance deficiency. This performance deficiency was more-than-minor and is a finding because it affected the protection against external factors attribute of the Mitigating Systems Cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, if the crosstie valves were damaged by a tornado, the licensee would not have enough available water inventory to meet the mission time for one train of the essential spray pond system during accident conditions. The inspectors performed the initial significance determination for the performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating System Screening Questions," dated July 1, 2012. The finding required a detailed risk evaluation because it involved the potential loss of a safety system, in that after at least 13 days of spray pond operation, operators were required to open the spray pond cross-connect valve to enable one train of the ultimate heat sink to use both trains of spray pond inventory. A Region IV senior reactor analyst performed a detailed risk evaluation. The design basis accident mission time was 30 days. However, the probabilistic risk assessment mission time was only 24 hours. Since the spray ponds could still perform the probabilistic risk assessment function for the probabilistic risk assessment mission time, this finding was of very low safety significance (Green). The change to the core damage frequency was much less than 1E-7/year. The finding did not contribute to the large early release frequency. Because the most likely cause of the finding does not reflect current licensee performance, no cross-cutting aspect is assigned to this finding. |
| 05000529/FIN-2015002-03 | 2015Q2 | Palo Verde | Failure to Identify and Correct Engineered Safety Features Actuation System Steam Generator Differential Pressure Setpoint Drift | The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.3.5 condition A.1 for failure to place a failed steam generator differential pressure in bypass or trip. Specifically, on January 11, 2015, after Unit 2 received a steam generator pressure difference setpoint alarm on channel B, operators failed to determine the cause of the alarm. As a result, the auxiliary feedwater actuation signal channel was inoperable for a period of 13 days, which was longer than the technical-specification allowed outage time of one hour, during which time the failed channel would provide a false negative under valid actuation setpoint conditions. The licensee entered this condition in their corrective action program and performed a root cause evaluation under Condition Report Disposition Request 4618033. The failure to provide adequate alarm procedures was a performance deficiency. The performance deficiency was more-than-minor and is a finding because it affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the control room operators did not have an alarm response procedure for plant monitoring system (RJ) alarm on point SASB22, which resulted in the channel B auxiliary feedwater actuation signal steam generator 2 drifting out of tolerance for a period of 13 days. This exceeded the allowed outage time specified in the technical specifications. The inspectors performed the initial significance determination using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating Systems Screening Questions." The finding screened to a detailed risk evaluation because it involved the actual loss of function of at least a single train for greater than its technical specification allowed outage time. A Region IV senior reactor analyst performed a detailed risk evaluation and determined that the change in core damage frequency CDF < 5E -9 corresponds to very low (Green) safety significance. This finding has a cross-cutting aspect in the area of human performance associated with the change management component in that the licensee did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the licensee did not use a systematic process to identify and correct the lack of alarm procedures associated with this parameter along with 76 other alarms that have technical specification implications during the design modification process for the plant computer alarm system (H.3). |
| 05000528/FIN-2015001-01 | 2015Q1 | Palo Verde | Failure to conduct required in-service testing in accordance with ASME OM Code | The inspectors identified a Green, non-cited violation of Palo Verde Technical Specification 5.5.8 Inservice Testing Program which requires the in-service testing of ASME Code Class 1, 2, and 3 components in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2001 Edition with Addenda through 2003. On April 26, 2013, the licensee did not test Unit 1 train A shutdown cooling isolation valve SIA-UV-651, an ASME Code Class 1 valve, in accordance with ASME OM Code Section ISTC-3310.The licensee entered this issue into the corrective action program as Palo Verde Action Request 4398843. The failure to complete ASME OM Code required in-service testing on a Class 1 motor operated valve is a performance deficiency. This performance deficiency is more than minor, and therefore is a finding, because it affected the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Specifically, by not performing the required testing, the licensee did not maintain the requisite level of assurance of the equipments capability of performing its intended function. Using 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 the condition was not a design or qualification deficiency, did not involve an actual loss of safety function for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Because the most-significant contributor to the finding was that Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained, the finding has a cross-cutting aspect in the Human Performance area and the aspect of Teamwork (H.4). |
| 05000528/FIN-2014005-01 | 2014Q4 | Palo Verde | Failure to Verify the Adequacy of the Design of the Diesel Fuel Oil Cooler | The inspectors reviewed a self-revealing Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to adequately review the suitability of materials of the diesel fuel oil cooler. Specifically, the Unit 2 A diesel generator fuel oil cooler design allowed for the interface of two dissimilar metals which promoted galvanic corrosion. This corrosion ultimately affected the structural integrity of the cooler and rendered the A Essential Spray Pond inoperable. In response to this, the licensee has replaced all six of the fuel oil cooler covers and initiated a design change to remove the fuel oil cooler from service. The licensee has entered the issue into the corrective action program as Condition Report Disposition Request 4543394. The failure to verify the adequacy of the design of the diesel fuel oil cooler was a performance deficiency. The performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone to ensure the availability, reliability, capability of systems that respond to initiating events to prevent undesirable consequences. Specifically the Unit 2 A diesel fuel oil cooler design allowed for the interface of two dissimilar metals which promoted galvanic corrosion. The corrosion ultimately affected the structural integrity of the cooler and rendered the Unit 2 A spray pond inoperable. In accordance with NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The finding screened to a detailed risk evaluation because it involved a potential loss of one train of safety related equipment for longer than the outage time allowed by technical specifications. By performing a detailed risk evaluation, a Region IV senior reactor analyst determined that the associated change to the core damage frequency was 1.5E- 7/year (Green). The dominant core damage sequences included loss of offsite power events that lead to station blackout conditions. The gas turbine generators and the auxiliary feedwater system helped to minimize the risk. This finding has no cross-cutting aspect because it is not indicative of current performance. |
| 05000528/FIN-2014005-02 | 2014Q4 | Palo Verde | Licensee-Identified Violation | Title 10 CFR 55.49, Integrity of Examinations and Tests, requires, in part, that facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Contrary to the above, during the week of November 12, 2013, the licensee caused a compromise to examination integrity by exceeding, 50 percent overlap on exam items during the same examination cycle. Specifically, the licensee repeated three of the required five job performance measures from one week to the next. The failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV violation because it involved a nonwillful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because there was not an actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Requalification. The licensee entered this issue into their corrective action program as Condition Report 4578169. |
| 05000528/FIN-2014005-03 | 2014Q4 | Palo Verde | Licensee-Identified Violation | Title 10 CFR 50.59(d)(1) requires, in part, that the licensee shall maintain records of changes in the facility, of changes in procedures, and of tests and experiments made pursuant to paragraph (c) of this section. These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does not require a license amendment pursuant to paragraph (c)(2) of this section. Contrary to the above, prior to August 28, 2014, the licensee failed to perform an evaluation against the criteria in 10 CFR 50.59(c)(2) for a change to the facility. Specifically, the licensee identified that Licensing Document Change Request 04-F020, performed on March 4, 2005, had changed the FSAR description of the auxiliary feedwater system. The new revision stated that portions of the auxiliary feedwater system, which are not contained within a Seismic Category I structure or installed underground, have been analyzed to show that the probability of being struck by a tornado missile is sufficiently low and do not require tornado missile protection. Previously, the FSAR described that all components of the auxiliary feedwater system were either enclosed by a Seismic Category I structure or are installed underground. This change had been inappropriately screened out of the 50.59 process in 2005. The licensees 50.59 screening did not recognize that this change to the FSAR description constituted a de facto change to the design of the facility. Consequently, the licensee failed to perform an evaluation against the criteria in 10 CFR 50.59(c)(2). On August 28, 2014, the licensee recognized the auxiliary feedwater recirculation lines do not meet the original FSAR criteria of being protected from tornado missiles. The licensee initiated PVAR 4568732 to document the lack of tornado missile protection for the auxiliary feedwater minimum flow recirculation lines. The licensee performed an immediate operability determination on August 29, 2014 and determined that there was a reasonable expectation that the auxiliary feedwater system would provide adequate decay heat removal following a tornado. The inspectors reviewed the licensees operability determination and verified that the licensee intends to submit a license amendment request for acceptance of the as-built configuration of the auxiliary feedwater system. Because the failure to implement the requirements of 10 CFR 50.59 had the potential to impact the NRCs ability to perform its regulatory function, the team evaluated the performance deficiency using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the inspectors evaluated this finding using the significance determination process to assess its significance. The finding required a detailed risk evaluation because it involved the failure of two or more trains in a multi-train system. A Region IV senior reactor analyst performed a bounding detailed risk evaluation and determined that the bounding delta-CDF was less than 3.5E-8/year. In accordance with Section 6.1.d of the NRC Enforcement Policy, this violation is categorized as Severity Level IV violation because the resulting change was evaluated by the SDP as having very low safety significance (i.e., Green finding). This issue has been entered into the licensees corrective action program as CRDR 4570021. |
| 05000528/FIN-2014004-01 | 2014Q3 | Palo Verde | Failure to Translate Design Basis Requirements for Establishing Operability of Spray Pond System | The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the mission time of the essential spray pond system into a procedure used to determine operability. In response to the inspectors concerns, the licensee re-evaluated essential spray pond operability determinations that had used the erroneous 26-day mission time and concluded that acceptable margin was available to ensure the system would remain operable for the 30-day mission time. The licensee entered this issue into the corrective action program as Palo Verde Action Request 4550539. The failure to ensure that design basis information associated with the mission time of the essential spray pond system was correctly translated into a procedure used to determine operability was a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to use the correct mission time when determining operability could establish nonconservative results that could lead to the essential spray pond system not being able to meet its design safety function. The inspectors performed an initial screening of the finding in accordance with NRC Manual Chapter 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 a single train for greater than its technical specification 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 human performance because the licensee failed to create and maintain complete, accurate, and up-to-date documentation. Specifically, after initially recognizing the adverse condition, the licensee did not document a standing order or temporary procedure change to prevent operability evaluations from using the incorrect essential spray pond mission time. |
| 05000528/FIN-2014004-02 | 2014Q3 | Palo Verde | Failure to Provide Adequate Technical Justification for Operability | The inspectors identified a Green non-cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to provide an adequate technical justification for continued operation of a degraded structure, system, or component. Specifically, after discovering that the turbine driven auxiliary feedwater pump exhaust line did not have any tornado missile protection, operators performed an immediate operability determination and declared the system operable. The inspectors determined that the licensee did not provide adequate technical justification for continued operation with this condition because: (1) the evaluation relied on a probabilistic risk assessment that assumed the turbine driven auxiliary feedwater pump fails due to impact from a tornado missile, and (2) the evaluation assumed that a future analysis would provide satisfactory results. In response to the inspectors concerns, plant personnel subsequently completed an analysis that provided a reasonable expectation that the turbine driven auxiliary feedwater pump would be able to perform its safety function if impacted by a tornado missile. The licensee entered this issue into the corrective action program as Palo Verde Action Request 4255816. The inspectors concluded that the failure of plant personnel to adequately evaluate the operability of a safety-related structure, system, or component was a performance deficiency. The inspectors concluded the performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors performed the initial significance determination for the performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 4, External Events Screening Questions, dated July 1, 2012. The finding required a detailed risk evaluation because the turbine driven auxiliary feedwater pump is one train of a system that supports a risk significant function. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The change to the core damage frequency was determined to be 7E-10/year (Green). The dominant core damage sequences included a tornado induced loss of offsite power initiating event, failure of the turbine driven auxiliary feedwater pump, and random failures of the motor driven auxiliary feedwater pumps. The low frequency for the tornadoinduced loss of offsite power initiating event helped to minimize the risk significance. The inspectors determined this finding has a cross-cutting aspect in the area of human performance because the licensee failed to utilize a conservative bias in its evaluation of the missing tornado missile protection, considering the risk significance of the turbine driven auxiliary feedwater pump and lack of any technical evaluation. |
| 05000528/FIN-2014004-03 | 2014Q3 | Palo Verde | Inadequate Calculations to Support the Degraded Voltage Relay Setpoint | The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure the adequacy of degraded voltage relay (DVR) setpoints. Specifically, the team identified that the licensee failed to perform calculations to demonstrate the voltage setpoints for the installed degraded voltage relays would afford adequate voltage to safety-related loads during worst case accident loading. The failure to assure the adequacy of DVR setpoints for voltage and the time delay by performing adequate voltage drop calculations was a performance deficiency. This finding is more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and it adversely impacted to the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Specifically, the failure to properly ensure that safety-related electrical devices had adequate voltage could impact their safety function. The basis for this conclusion was that despite the non-conservative voltage inputs to voltage calculations and, therefore, loss of design margin for available voltage, there was still adequate voltage for the circuits to perform their safety function based on worst case voltage as demonstrated in the updated calculations. The licensee developed design basis calculations for its DVR voltage setpoints and committed to develop a plant design change and an associated license amendment to shorten the existing time delay in Technical Specication 3.3.7.3(a). There is no cross-cutting aspect associated with this finding because it is a historical condition and not indicative of current performance. |
| 05000382/FIN-2014003-01 | 2014Q2 | Waterford | Failure to Control Entry into a High Radiation Area | The inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638. The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. 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 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 an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12). |
| 05000382/FIN-2014003-02 | 2014Q2 | Waterford | Failure to Maintain Adequate Public Address System to Implement Onsite Protective Actions | The inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10). The licensee implemented compensatory measures while the system was being restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860. The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the objective of ensuring 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. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings; and the corresponding Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes (P.2). |
| 05000482/FIN-2014002-02 | 2014Q1 | Wolf Creek | Failure to Maintain Licensed Power Limits During Planned Evolutions Affecting Reactivity | A self-revealing non-cited violation, with two examples, of Technical Specification 5.4.1.a, Procedures, was identified for the failure to follow the reactivity management procedures. On two occasions, operators failed to take prudent actions to ensure that reactor power did not exceed the licensed limit of 3565 megawatts thermal while performing activities known to cause power increases. On February 17, 2014, while performing chemical and volume control system inservice check valve testing on the discharge check valve of the train A centrifugal charging pump, operators performed a dilution of the reactor coolant system for normal power maintenance while reactivity was also being affected by the testing of the charging pump check valve, resulting in exceeding 100 percent power. On March 6, 2014, while returning the reactor to full power following data collection on the main turbine control valves, operators used an automatic power ramp to a setpoint of only 3 megawatts below 100 percent, without accounting for the overshoot that would result from the selected ramp rate, resulting in exceeding 100 percent power. In both cases, operators were alerted by an alarm indicating that the 1-minute average power level exceeded 100 percent. The inspectors reviewed station procedure GEN 00-004 Power Operation, and noted a requirement in Attachment A: For pre-planned evolutions that are expected to cause a transient rise in reactor power that could exceed the licensed power level, prudent actions should be taken to reduce power prior to the evolution. Failure to take prudent action to maintain the reactor within licensed power limits prior to performing activities known to cause an increase in reactor power levels is a performance deficiency. The performance deficiency was more than minor because it affected both the configuration control attribute of reactivity control as well as the human performance attribute of procedure adherence of the Barrier Integrity Cornerstone, and impacted the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors screened the finding using the reactivity control screening questions found in Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Section C; question number 3 referred the inspectors to Inspection Manual Chapter 0609, Appendix M, Significance Determination Using Qualitative Criteria. NRC Management performed the qualitative assessment and determined that the finding was of very low safety significance (Green) because the relatively small magnitude of the overpower events, the prompt operator actions to return power to below the licensed limits upon discovery, and the fact that overpower events did not result in any failure of the fuel cladding. The inspectors determined that the finding had a conservative bias cross-cutting aspect in the area of human performance. Specifically, the affected evolutions were known in advance to have positive reactivity impacts; however, operators did not consider reducing power in the case of the check valve testing, nor was a slow approach to the maximum reactor power level used to avoid overshoot during dynamic turbine loading for the turbine valve data collection in order to prevent licensed power levels from being exceeded. |
| 05000482/FIN-2014002-01 | 2014Q1 | Wolf Creek | Inadequate Work Instructions for Reinstallation of ESW Expansion Joints | The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, Procedures, for maintenance instructions inappropriate to the circumstances. Specifically, Work Orders 11-341986-005 and 11-342065-002 did not contain adequate instructions for reassembling essential service water Garlock expansion joints to ensure proper joint alignment. As a result, on February 11, 2014, the inspectors identified that the inlet expansion joint for the essential service water intercooler heat exchanger, which provides cooling to emergency diesel generator B jacket water system, was misaligned by 0.5 inches, which exceeded the vendor specification of less than 0.125 inch. This item was entered into the corrective action program as Condition Reports 79352 and 79623, and the fitting was replaced during the mid-cycle 2014 outage. The licensee also conducted an extent of condition inspection and identified three additional Garlock expansion joints that were not made with the approved liner material. The failure to properly reinstall essential service water expansion joints consistent with the vendor approved and analyzed configuration was a performance deficiency. The performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the misaligned Garlock expansion joint in the essential service water system degraded its long-term operability and its ability to withstand a seismic event. Using the 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 the finding did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time and the finding 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 for greater than 24 hours. Specifically, although the expansion joint was in a degraded condition, it was determined to be operable based on an engineering evaluation and seismic test data. The inspectors determined that the finding had a cross-cutting aspect in the human performance area of resources because the licensee did not ensure that personnel equipment, procedures, and other resources were available and adequate to support nuclear safety. |
| 05000482/FIN-2014002-03 | 2014Q1 | Wolf Creek | Failure to Maintain Seismic and Missile Protection Design Basis Requirements During Essential Service Water Construction | A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for the failure to conduct excavation work such that it would ensure that design basis requirements for tornado missile protection and seismic qualification of safety-related cables were maintained during construction near the essential service water pump house. Specifically, when excavation near underground essential service water cables caused a loss of safety-related backfill over the cables, the licensee did not plan and execute the work in a manner that ensured that the qualified soil coverage around the train B essential service water duct bank was maintained by protecting against trench cave-ins. Failure to maintain adequate soil coverage of the essential service water duct banks during construction is a performance deficiency. The deficiency is more than minor because it affected the protection against external factors and design control attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. The inspectors determined that the finding had a cross-cutting aspect of work management in the area of human performance in that the process for planning, controlling, and executing work did not adequately include the identification and management of risk. Specifically, work planning did not account for adequate shoring material to prevent design basis ground cover from caving in during planned excavations in the vicinity of operable safety related equipment. |
| 05000482/FIN-2014002-04 | 2014Q1 | Wolf Creek | Licensee-Identified Violation | A licensee-identified violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified for failure to ensure that design basis requirements were maintained in response to a cave-in of required essential service water ground cover. Specifically, the licensee did not ensure that the qualified soil coverage around the train B essential service water duct bank was maintained when they re-covered the duct banks with an unapproved material. Contrary to these requirements, on January 20, 2014, upon a loss of essential service water duct bank soil coverage due to a cave-in, the licensee refilled the voided area with an unapproved material that was not qualified to withstand seismic and missile design basis accidents. The performance deficiency was failure to ensure the appropriateness of seismic and missile-qualified material. This violation was more than minor because it affected the protection against external factors and design control attributes of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 4, External Events Screening Questions, the inspectors determined that the finding was of very low safety significance because it did not involve the total loss of any safety function that contributes to external event initiated core damage accident sequences. Since the finding was licensee-identified, no cross-cutting aspect is assessed. The finding was entered into the licensees corrective action program as Condition Report 79089. |
| 05000498/FIN-2013005-01 | 2013Q4 | South Texas | Failure to Include Appropriate Acceptance Criteria in a Quality Procedure | The inspectors identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, the licensee did not include sufficient criteria to identify and evaluate new critical tasks created for operator performance on the simulator scenario portion of the biennial requalification examination to enable the evaluators to correctly assess licensed operator performance. The licensee has entered this issue into their corrective action program as Condition Report 2013-13857. The failure to include appropriate qualitative acceptance criteria in Procedure LOR-GL-002, to ensure evaluators can correctly identify and evaluate critical tasks based on operator performance was a performance deficiency. The performance deficiency was more than minor, therefore, a finding, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the failure to include the appropriate criteria to identify and evaluate critical tasks during biennial requalification examinations could result in operators returning to licensed operator duties without being properly remediated and retested on performance deficiencies. Using Manual Chapter 0609, Attachment 0609.04, Appendix I, Operator Requalification Human Performance Significance Determination Process, starting at block 9, the finding was determined to be of very low safety significance (Green) because the finding is associated with licensee administration of an annual requalification operating test. The finding had a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee failed to make safety-significant or risk-significant decisions using a systematic process (H.1(a)) (Section 1R11.3.b.1). |