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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 |
| 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. |
| 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-2018002-01 | 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 |
| 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. |
| 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). |
| 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). |
| 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). |
| 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). |
| 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. |
| 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. |
| 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. |
| 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. |
| 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. |
| 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. |
| 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-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. |
| 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. |
| 05000482/FIN-2015002-01 | 2015Q2 | Wolf Creek | Class 1E 4kV Feeder Breakers from Station Blackout Diesel Generators Current Transformer Wiring not Installed per Design Drawings | The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, for not installing the current transformer wiring in the Class 1E 4kV alternate feeder breaker cubicles from the station blackout diesel generators per the design drawings. As a result, testing performed seven months after the system was declared operational identified that the connections were unable to power the safety-related buses due to incorrect wiring of the current transformers. The licensee entered this issue into the corrective action program as Condition Report 83379. This finding was more than minor because it was associated with the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, due to the incorrect wiring of the current transformers, the SBO diesel generators were unable to power safety related buses as they were designed. The inspectors performed the initial significance determination for the finding using NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The finding required a detailed evaluation because it had the potential to degrade at least one train of a system that supports a risk significant system or function. Therefore, a senior reactor analyst performed a bounding detailed risk evaluation. The finding was of very low safety significance (Green) because the risk assessment programs quantified the change in core damage frequency less than 1.0x10-6. The inspectors determined that the finding had a teamwork cross-cutting aspect in the area of human performance. The licensee individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries. Specifically a drawing revision was not properly attached to the work order which resulted in the incorrect wiring of both trains, and because different groups were completing different components, parts of the wiring were incorrectly installed per a superseded revision. |
| 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. |
| 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). |
| 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). |
| 05000482/FIN-2014005-01 | 2014Q4 | Wolf Creek | Failure to Conduct and Evaluate Simulator Testing In Accordance with ANSI/ANS 3.5-2009 and ANSI/ANS 3.5-1998 | The inspectors identified a Green finding for the inadequate conduct and evaluation of simulator performance testing in accordance with the standards of ANSI/ANS 3.5-2009 and ANSI/ANS 3.5-1998, Nuclear Power Plant Simulators for Use in Operator Training and Examination. Specifically, Wolf Creek Nuclear Operating Corporation (WCNOC) did not adequately identify that the simulator responses during 2008 through 2014 tests of Transient 3, Simultaneous Closure of All Main Steam Isolation Valves, did not meet the acceptance criteria described in Section 4.1.4 of ANSI/ANS 3.5-2009 (or the 1998 edition), which if left uncorrected, could have resulted in negative training of licensed operators and call into question Wolf Creeks ability to conduct valid licensing examinations with the simulator. WCNOC initiated condition reports 90179 and 90417 and simulator discrepancy report A14-154. WCNOC also plans to conduct benchmarking at other sites to compare simulator responses during applicable testing, and is evaluating the need for additional procedure revisions or other corrective actions. The performance deficiency is more than minor because it adversely impacted 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. Additionally, if left uncorrected, the performance deficiency could have become more significant in that not correcting noticeable differences between the simulator and the reference plant could cause negative training of licensed operators and call into question WCNOCs ability to conduct valid licensing examinations with the simulator. Utilizing Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Tables 1 and 2 worksheets, issued June 19, 2012, and flowchart block 14 of Appendix I, Licensed Operator Requalification Significance Determination Process (SDP), issued December 6, 2011, the finding was determined to have very low safety significance (Green), because the deficiencies were associated with simulator testing, modifications, and maintenance, and there was no evidence that the plant-referenced simulator does not demonstrate the expected plant response or have uncorrected modeling and hardware deficiencies. This finding has a cross-cutting aspect in the area of problem identification and resolution, Identification, because WCNOC personnel did not implement a corrective action program with a low threshold for identifying issues. Specifically, this issue was first identified when the RETRAN-3D code analysis was first used in 2008 transient testing, and additional tests performed in 2008, 2009, 2010, and 2012 were opportunities to identify the performance deficiency; however, the issue was not entered into the corrective action program, a noticeable difference was not evaluated, a training needs assessment was not performed, and the process used to conduct simulator transient testing, as described in Procedure Al 30C-006, was not updated to include all of the minimum acceptance criteria described in the ANSI/ANS 3.5 standard. Hence, simulator issues expected to be identified during the testing process could potentially be missed by implementing the AI 30C-006 procedure, which did not include all of the minimum acceptance criteria described in the ANSI standard (P.1). |
| 05000482/FIN-2014004-01 | 2014Q3 | Wolf Creek | Failure to Maintain Control and Cognizance of Activities With the Potential to Impact Plant Conditions | A self-revealing finding was identified for failure to recognize the potential effects on supported plant equipment while manipulating electrical power distribution components. The finding resulted in an unplanned reactor pressure transient during solid plant operations because the charging flow control valve failed open. Plant pressure increased from 84 to 345 psig before operators were able to control charging flow and lower pressure. The inspectors also concluded that the Licensed Operator Watchstation Expectations in station procedure AP 21-001, Conduct of Operations, was not met. Specifically step 6.3.2 states that Control Room personnel are responsible for in-plant activities and maintain control and cognizance of any activities which have the potential to impact plant conditions. This issue was entered into the corrective action program as Condition Report 80870. Failure to maintain control and cognizance of activities which have the potential to impact plant conditions was a performance deficiency. Specifically, operators failed to recognize the potential effects on primary plant pressure while manipulating electrical power distribution system. The performance deficiency was more than minor because it affected the configuration control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. A regional senior reactor analyst performed a simplified risk evaluation and additionally considered guidance from Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 5, 2014, and determined that since this deficiency did not involve: 1) exceeding the pressure rating of low pressure piping; or 2) maintaining the low temperature over-pressure protection itself, this finding was of very low safety significance (Green). This was used to inform the assessment using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 15, 2005. The analyst determined that the risk deficit was much less than 1E-6/year. The inspectors determined that the finding had a cross-cutting aspect of teamwork in the area of human performance in that individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the licensee failed to coordinate the planned bus realignment with a replacement of a redundant power supply such that the momentary loss of power would not have occurred (H.4). |
| 05000483/FIN-2014004-01 | 2014Q3 | Callaway | Failure to Perform Nondestructive Testing on Essential Service Water Piping in Accordance with Procedures | The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform nondestructive testing on portions of the essential service water system known to be susceptible to wall thinning in accordance with procedures. As a result, the licensee failed to identify wall thinning prior to developing a through-wall leak that rendered train A inoperable. Specifically, despite procedural guidance to the contrary, technicians only used the low frequency electromagnetic technique testing, which cannot monitor bends and portions of welds. They also failed to properly calibrate this equipment, and failed to perform ultrasonic testing on the portions of essential service water system that could not be properly monitored by use of low frequency electromagnetic technique. The resultant through-wall leaks were repaired according to ASME code standards. The licensee entered this issue into their corrective action program as Callaway Action Request 201405200 and planned to re-perform testing during the fall of 2014. Failure to follow procedures while performing nondestructive testing on portions of the essential service water system was a performance deficiency. This performance deficiency is more than minor because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to perform nondestructive testing on portions of the essential service water system that were known to be susceptible to wall thinning resulted in the failure to prevent a through-wall leak affecting the availability of a safety related system. Using NRC Inspection Manual 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance because it only affected a single train, and resulted in a loss of function for less than its technical specifications allowed outage time. This finding has a procedure adherence cross-cutting aspect within the human performance area because the licensee failed to ensure that individuals followed processes, procedures, and work instructions. Specifically, licensee oversight failed to ensure that contractors followed specific guidance in their procedures for both ensuring that the low frequency electromagnetic technique tool was appropriately calibrated and areas unable to be scanned were tested utilizing ultrasonic testing. |
| 05000483/FIN-2014004-03 | 2014Q3 | Callaway | Licensee-Identified Violation | The licensee dose assessment methods are inaccurate under some circumstances. Title 10 of the Code of Federal Regulations, Section 50.54(q)(2) requires, in part, that power reactor licensees follow and maintain the effectiveness of an emergency plan that meets the requirements of Appendix E to Part 50 and the standards of 10 CFR 50.47(b). Section 50.47(b)(9) requires that adequate systems, methods, and equipment for assessing the actual and potential offsite consequences of a radiological emergency condition are in use. Contrary to the above, between December 5, 2009, and April 23, 2014, Callaway Plant did not maintain the effectiveness of an emergency plan that fully met the standards of 10 CFR 50.47(b). Specifically, the licensee failed to maintain adequate systems, methods, and equipment for assessing the actual and potential offsite consequences of a radiological release to the environment. The licensee identified circumstances which could cause their dose assessment program to overestimate offsite doses by a factor of 24. This finding was assessed using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and was determined to be of very low safety significance (Green). The finding was a failure to comply with NRC requirements, was not a loss of risk significant planning standard function, and was not a degraded planning standard function. The finding was determined not to degrade the planning standard function because the calculation was not inaccurate in its normal operating configuration, the circumstances in which the inaccurate calculation would be used were rare, a method existed for the operator to correct the error, and because the error could be readily detected from examination of the dose assessment report. The issue was entered into the licensees corrective action program as Corrective Action Request 201402814. |
| 05000483/FIN-2014004-02 | 2014Q3 | Callaway | Failure to Verify Material Properties Prior to Installation | The inspectors reviewed a self-revealing finding involving failure to verify the proper material was installed in the plant during a modification to the circulating water pumps. Specifically, Request for Resolution 201300416 specified the use of ASTM A276 410 stainless steel cap screws with a tensile strength around 186 ksi. Contrary to this, 410 stainless steel cap screws with a tensile strength between 201 ksi and 221 ksi were installed. Because the tensile strength was much higher, and thus more brittle and susceptible to stress corrosion, these cap screws were not appropriate for the application. This led to failure of the cap screws and the separation of the shaft coupling for circulating water pump B after less than one operating cycle in service, degrading condenser vacuum. The licensee removed the modification and installed the original type cap screws. This issue was entered into the licensees corrective action program as Callaway Action Request 201404722. The inspectors determined that failure to verify the correct materials were installed in the plant during a modification was a performance deficiency. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as during power operations. Specifically, failure to install the correct material resulted in failure of circulating water pump B and degrading condenser vacuum. The inspectors evaluated the finding using NRC Inspection Manual 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was of very low safety significance (Green) because the transient initiator did not cause a reactor trip and the loss of mitigating equipment. This finding has an avoid complacency cross-cutting aspect within the human performance area because the licensee relied on the vendor to provide the correct material and did not verify the cap screws met the material specification. |
| 05000482/FIN-2014004-02 | 2014Q3 | Wolf Creek | Failure to Ensure That Outage Work Could Be Safely Performed During the Existing Plant Conditions | A Green self-revealing non-cited violation of Technical Specification 5.4.1.a and Regulatory Guide 1.33, Section 9.e was identified for the failure to implement procedures for the control of maintenance involving motor operated valve testing to ensure that it did not affect safety-related equipment while the plant was aligned to support alternate decay heat removal. The activity resulted in unplanned reactor pressure transients during solid plant operations. The inspectors reviewed the clearance order paperwork and found that the precautions for dealing with potential fluid and energy sources, specifically out of service equipment were not clearly defined. The result was that the procedure assumed a normal refueling alignment of the residual heat removal system, when in fact the licensee had altered the system alignment to support an alternative reactor decay heat removal flow path using the spent fuel pool. This issue was entered into the corrective action program as Condition Report 81981. Failure to ensure that outage work could be safely performed during the existing plant conditions was a performance deficiency. Specifically, when the licensee revised the outage plan shortly before the start of the Mid-cycle Outage 20, they did not re-perform the risk evaluation for the potential fluid and energy sources to account for the unusual configuration established to allow for alternate decay heat removal. The performance deficiency is more than minor because it affected the configuration control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. A region based senior reactor analyst performed a simplified risk evaluation and additionally considered guidance from Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process. This was used to inform the assessment using Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 15, 2005. The analyst determined that the finding had very low safety significance (Green) because the risk deficit was less than 1E-6. The inspectors determined that the finding had a cross-cutting aspect of teamwork in the area of human performance in that individuals and work groups failed communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the licensee developed the alternate decay heat removal alignment shortly before the outage, however the effects of the implementation were not communicated to the schedulers and operators who had already made risk assumptions based on different anticipated plant conditions (H.4). |
| 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-2014002-04 | 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. |
| 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-2014002-01 | 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-2013005-04 | 2013Q4 | Wolf Creek | Failure to Preclude Repetition of a Significant Condition Adverse to Quality Affecting Class 1E Air Conditioning Unit | A self-revealing non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified for failure to preclude repetition of a significant condition adverse to quality. Specifically, the licensee failed to determine the cause of the train A Class 1E air conditioner slug flow in the refrigerant prior to experiencing multiple failures and plant shutdowns. The licensee eventually determined that the probable cause of the failures was the failure to adequately chemically clean the system following the failure of a charcoal filter, in that some charcoal and cleaning agent was not removed following the work activity. The remaining debris fouled the lubricating oil and instrumentation, causing four failures. This issue was entered into the corrective action program as Condition Report 78709. Failure to preclude repetitive failures of the train A Class 1E air conditioner, a significant condition adverse to quality, 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, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically the leftover debris in the system resulted in a subsequent maintenance outage to replace rapidly degrading components. Using the Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined 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 nontechnical specification trains of equipment designated as high safety significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined that the cause of the finding had a cross-cutting aspect in the area of human performance. The licensee did not ensure complete, accurate and up-to-date design documentation, procedures, and work packages, and correct labeling of components. Specifically applicable housekeeping requirements detailed in station procedure AP-12-002 Internal/External System Cleanliness were not met during the chemical cleaning of system piping on May 8, 2013, (H.2(c)). |
| 05000482/FIN-2013005-01 | 2013Q4 | Wolf Creek | Failure Rates Exceed Twenty Percent for Annual Requalification Operating Tests | The inspector reviewed a self-revealing finding associated with licensed operator performance on the annual requalification operating tests. Specifically, 2 of 8 crews (25 percent) failed the simulator scenario portion of the operating test; and 11 of 46 licensed operators (23 percent) either failed the scenario or failed the job performance measure portions of the operating tests. The licensee remediated and retested the staff prior to returning them to licensed duties. Wolf Creek entered this finding into their corrective action program as Condition Report 75336. In accordance with Inspection Procedure 71111.11, each of the following was a performance deficiency against expected licensed operator knowledge and abilities: 1) Greater than 20 percent of the crews failing their scenarios; and 2) greater than 20 percent of the licensed operator staff failing their operating tests. Using the Inspection Manual Chapter 0612, Appendix B, Issue Screening, the inspector determined that the finding was more than minor because the performance deficiency was associated with the Mitigating Systems Cornerstone attribute of human performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspector determined that this finding could be evaluated using Inspection Manual Chapter 0609, Appendix I, Licensed Operator Requalification Significance Determination Process. This finding was of very low safety significance (Green) because the finding was related to the requalification exam results, did not result in a failure rate of greater than 40 percent, and the licensed operators were remediated prior to returning to shift. This finding has a crosscutting aspect in the area of human performance associated with resources, because the licensee failed to ensure that personnel were adequately trained to assure nuclear safety (H.2(b)). |
| 05000482/FIN-2013005-02 | 2013Q4 | Wolf Creek | Failure to Assess Risk Prior to Performing Online Maintenance to an Offsite Power Circuit Component | The inspectors identified a Green non-cited violation of 10 CFR 50.65(a)(4) for the failure to assess risk associated with an emergent maintenance activity performed on one of the offsite power circuit components inside the Wolf Creek switchyard. Specifically, Wolf Creek arranged for the transmission system maintenance companies to recharge SF6 gas in a 13.8kV breaker actively feeding the train A Class 1E distribution system, without performing risk management actions to verify the readiness of onsite power sources. Loss of SF6 pressure would have caused this breaker to automatically open. This issue was entered into the corrective action program as Condition Report 00077139. The inspectors determined that failure to assess risk associated with an emergent maintenance activity in accordance with station procedure AP 22C-003, Online Nuclear Safety and Generation Risk Assessment, Step 6.1.9 was a performance deficiency. The performance deficiency was more than minor because it affects the switchyard activities area of the protection against external factors attribute of the Initiating Events Cornerstone. Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, was used to assess the significance of the finding because the finding resulted from the licensees underestimate of plant risk. The regional senior reactor analyst performed a bounding analysis of the incremental core damage probability and determined that the finding was of very low safety significance (Green) because the bounding analysis indicated that the effect of the finding on plant operations was less than the Green/White threshold. The inspectors determined that the finding did not have a cross-cutting aspect because the performance deficiency was caused by an inadequate procedure change in 1998, which was not representative of current performance. |
| 05000482/FIN-2013004-01 | 2013Q3 | Wolf Creek | Failure to Analyze Erected Scaffolding for Fire Impairment and Transient Combustible Loading | The inspectors identified a non-cited violation of Technical Specification 5.4.1.d, Fire Protection Program Procedures, for the failure to analyze scaffolding for fire protection impairments and transient combustible loading. The cause of the finding was a procedure change that allowed for a grace period of one working day to complete a fire protection review of newly erected scaffolding. As a result, there was no longer a direct interface with the scaffold builders and fire protection engineers, which complicated scoping and tracking the required inspections. This violation was entered into the corrective action program as Condition Report 71910. Failure to analyze scaffolding for fire impairment and transient combustible loading is a performance deficiency. The performance deficiency is more than minor because it affects the protection against external factors attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination for Findings at Power, which required screening using MC 0609, Appendix F, Fire Protection Significance Determination Process. Using screening question 1.3.1 of this Appendix, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not affect the ability to achieve and maintain safe shutdown. The inspectors determined that this finding had a crosscutting aspect in the human performance area of work control because the licensee failed to appropriately coordinate work activities by incorporating the need for planned compensatory actions. Specifically, Wolf Creek did not ensure that a fire protection assessment of scaffold 13-S100 and 13-S134 was performed in a timely manner which resulted in compensatory measures for the impaired sprinkler heads and transient combustible material not being established. |
| 05000482/FIN-2013004-02 | 2013Q3 | Wolf Creek | Licensee-Identified Violation | Technical Specification (TS) 3.8.4, DC Sources Operating, requires two operable trains of DC sources while in Modes 1 through 4. TS 3.8.7, Inverters Operating, requires two operable trains of inverters while in Modes 1 through 4. Technical Specification 3.8.9, Distribution Systems Operating, requires the Train A and Train B subsystems of the DC and AC vital bus systems be operable in Modes 1 through 4. Technical Specifications 3.7.11, Control Room Air Conditioning System, requires two trains of Control Room Air Conditioning be operable in Modes 1 through 6. Technical Specification Limiting Condition for Operation (LCO) 3.0.3 requires, in part, that when an LCO is not met and the associated actions are not met or an action is not provided, the unit shall be placed in a Mode in which the LCO is not applicable, or take action within one hour to place the plant in Mode 3 within the next 7 hours, Mode 4 within the next 13 hours, and Mode 5 within 37 hours. Contrary to the above, two trains of Class 1E electrical equipment air conditioning were not functional in Modes 1 through 4, but the licensee did not take action within one hour to place the plant in Mode 3 within the next 7 hours, Mode 4 within the next 13 hours, and Mode 5 within 37 hours. Specifically, the licensee over-torqued mounting screws on air conditioning units SGK04B and SGK05B during their respective maintenance activities performed on August 12, 2011, and May 30, 2012, rendering these air conditioning units and their supported systems inoperable. The licensee failed to restore air conditioning units SGK04B and SGK05B to operable status until March 22 and March 23, 2013; and thus exceeded the allowed action times of the respective Technical Specifications. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, the inspectors performed a significance determination screening and determined that a more detailed risk evaluation was required because the finding potentially represented an actual loss of safety function for a single train for greater than the technical specification allowed outage time. The Senior Reactor Analyst determined that since the mounting screws did not actually fail, and their function is to prevent failure during a seismic event, and because of the low probability of a seismic event leading to core damage or early release, this violation is of very low safety significance (Green). This violation was entered into the licensees corrective action program as Condition Report 65421. |
| 05000482/FIN-2013004-03 | 2013Q3 | Wolf Creek | Licensee-Identified Violation | Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed by procedures appropriate to the circumstances. Contrary to the above, a procedure for an activity affecting quality was not appropriate to the circumstances. Specifically, as of May 6, 2013, Procedure MPE GK-004 GK Unit Preparation for Work, Revision 4, was inadequate to prevent over-tightening of the wing screws on the filter dryer assembly, in that the procedure did not specify how tight to leave the wing screws. This led to failure of the filter drier, pieces of which blocked the thermostatic expansion valves and overheated the Class 1E electrical equipment air conditioning chiller. Using Inspection Manual Chapter 609, Appendix A, Exhibit 2, Mitigating Systems Cornerstone screening questions, Section A, the finding was determined to be of very low safety significance (Green) because only one train of Class 1E electrical equipment was affected and room temperatures remained below maximum limits. This violation was entered into the licensees corrective action program as Condition Report 68818. |
| 05000482/FIN-2013003-01 | 2013Q2 | Wolf Creek | Failure to Follow Station Procedures | The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstances and accomplished in accordance with these procedures. Contrary to the above, the licensee failed to ensure procedures related to the boric acid corrosion control program were adequate and properly implemented. Specifically, prior to February 19, 2013, the licensee failed to: (1) resolve discrepancies within the boric acid corrosion control program procedure; (2) resolve discrepancies between the boric acid corrosion control program procedure and the boric acid leak management procedure; and (3) failed to track and resolve leakage for locations where health physics had installed drip catch containments, to review the Health Physics Drip Bag Log as part of the quarterly outside containment walkdown, and to add component locations to the program. Further, the licensee failed to periodically assess the effectiveness of the program on a refueling frequency. The violation was entered into the licensees corrective action program as Condition Report 65212. The inspectors determined that the failure to recognize discrepancies between boric acid control procedures and the failure to follow boric acid program procedures was a performance deficiency. The performance deficiency was more than minor because it affected the Initiating Events Cornerstone attribute of procedure quality and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations, and if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, failure to resolve discrepancies within procedures or track and resolve leak locations where health physics had installed drip catch containments had the potential to mischaracterize leaks or allow leaks to corrode safety-related systems. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding was determined to be of very low safety significance (Green), because the finding was a procedure quality problem that did not represent a loss of system safety function, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to ensure supervisory and management oversight of work activities, including procedure appropriateness and compliance, such that nuclear safety is supported. |
| 05000482/FIN-2013003-07 | 2013Q2 | Wolf Creek | Notice of Enforcement Discretion (NOED) 13-4-002 for a Non-Functional Class 1E Air Conditioning Unit | On June 17, 2013, an oil sample taken from the train A Class 1E air conditioning unit was found to have unacceptable levels of aluminum particulate, indicating that internal parts were degrading and long term reliability was not assured. The unit was declared non-functional and Wolf Creek again entered Technical Specification 3.0.3 at 11:11 a.m. Wolf Creek requested a NOED that was granted by the NRC staff at 4:07 p.m. The inspectors reviewed the documentation, plant status information, the equipment history, as well as the Inspection Manual Chapter 0410 process. Consistent with NRC policy, the NRC agreed not to enforce compliance with the specific technical specifications in this instance, but will further review the cause(s) that created the apparent need for enforcement discretion to determine whether a violation of NRC requirements existed. This will be tracked under unresolved item (URI) 05000482/2013003-06, NOED 13-4-002 for a Non-functional Class 1E Air Conditioning Unit. |
| 05000482/FIN-2013003-06 | 2013Q2 | Wolf Creek | Failure to Maintain Complete and Accurate Housekeeping Records | The inspectors identified a Severity Level IV violation of 10 CFR 50.9, Completeness and accuracy of information, for the Wolf Creek Nuclear Generating Stations failure to maintain complete and accurate records required by a license condition. Title 10 CFR 50.9 requires, in part, that information required by statute, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. Contrary to the above, between October and December 2008, the licensee failed to maintain records required by License Condition 2.C.5 that were complete and accurate in all material respects. Specifically, the Housekeeping Inspection Card for the spent fuel pool area indicated that the inspection had been completed when security access logs indicate that the individual that completed the record had not entered the area. The NRC investigation determined that the assigned individual did not walk down the assigned area, and did not assign a designee to do so (EA-13-084). The failure to maintain records required by License Condition that are complete and accurate in all material respects in accordance with 10 CFR 50.9 was a violation. Because the violation is associated with willfulness and impacted the regulatory process it was evaluated under the traditional enforcement process as set forth in the NRC Enforcement Policy. Since this violation was the result of a willful action, the NRC considers the violation to be more than minor, and therefore, the NRC has classified the violation at Severity Level IV, in accordance with the NRC Enforcement Policy. |
| 05000482/FIN-2013003-05 | 2013Q2 | Wolf Creek | Failure to Properly Manage Reactivity Changes when Swapping Turbine Steam Admission Modes from Full to Partial Arc | Inspectors identified a Green non-cited violation of Technical Specification 5.4.1.a for the failure to follow Conduct of Operations and Reactivity Management procedures. The inspectors reviewed an unplanned 11 percent power increase during a shift in turbine control modes, and identified that pre-job briefings did not adequately discuss expected plant response, operators did not take action to limit the power increase when an unexpected response was observed, and management was not adequately involved in decision making prior to continuing power ascension before the details of an apparent turbine control malfunction were fully understood. This issue was entered into the licensees corrective action program under Condition Report 68711. Failure to provide contingency actions for a greater than anticipated reactor transient in the pre-job reactivity brief, and continuing with power ascension without understanding the cause of the unexpected turbine control system behavior is a performance deficiency. The performance deficiency is more than minor because it affected the human performance attributes of the Initiating Events cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609 Appendix A, Checklist 1, Initiating Events Screening Questions, and the inspectors determined that the finding was of very low safety significance (Green) because the finding did not result in a reactor trip coincident with the loss of mitigation equipment. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance area of work practices because the licensee failed to communicate human error prevention techniques, such as holding pre-job briefings, self and peer checking, and proper documentation of activities such that work activities were performed safely. In addition, personnel proceeded in the face of uncertainty or unexpected circumstances. Specifically, in the first example control room operators pre-job reactivity brief was not appropriate commensurate with the risk of the assigned task; in the second example station personnel proceeded in the face of uncertainty. |