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05000528/FIN-2018008-04Minor Violation2018Q3Failure to promptly identify and correct conditions adverse to quality as required by 10 CFR 50, Appendix B, Criterion XVI. The team identified a backlog of conditions adverse to quality that the licensee had failed to timely correct. The oldest of these conditions was approximately 10 years old, with several hundred having been identified at least two operating cycles prior to the inspection. The team determined that the licensee was appropriately addressing degraded components that had an impact on safety or security, but was not always tracking or timely correcting nonconformances with its design bases in cases where these nonconformances had been assessed as not impacting safety-related functions. Further, the licensee was unable to initially determine the scope of its nonconformance backlog. The licensee documented this deficiency as Condition Reports 18-13549 and 18-14426. Screening: The performance deficiency was minor because if left uncorrected it would not have led to a more significant safety concern and it did not adversely affect any cornerstone objectives. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000528/FIN-2018008-03Licensee-Identified Violation2018Q3This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: 10 CFR 50.73(a)(2)(i)(B) requires, in part, that the holder of an operating license shall submit an licensee event report within 60 days of discovery of the event, which includes any operation or condition which was prohibited by technical specifications. Contrary to the above, the licensee failed to submit a licensee event report within 60 days of April 23, 2016, after discovering that the Unit 1 channel C excore was in a condition which was prohibited by technical specifications. The detector was found in a configuration without o-rings at two electrical connection interfaces. Condition Report 16-06735 documented the non-conforming condition, but was closed without performing a reportability review. Significance/Severity Level: This violation was considered as traditional enforcement because the failure to notify the NRC had the potential for impacting the NRCs ability to perform its regulatory function. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the failure to report the condition prohibited by technical specifications was determined to be a Severity Level IV violation. Corrective Action Reference(s): Condition Report 18-02569
05000528/FIN-2018008-02Licensee-Identified Violation2018Q3This violation of very low safety-significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR Part 50, Appendix B, Criterion V requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, on May 24, 2007, the licensee failed to perform the installation of the Unit 1, channel C excore nuclear instrument preamplifier connection, an activity affecting quality, in accordance with these instructions, procedures, or drawings. The licensee determined that a human performance error occurred during the performance of the 2007 work order which explicitly stated that the o-rings were required for environmental qualification. As a result, the excore detector would not have performed its safety function during a design basis main steam line break. Significance/Severity Level: The team determined this finding was of very low safety significance (Green) because a minimum of two excore detector channels always remained available to trip the reactor during a main steam line break. Redundant channels were not affected and were available to perform the required safety function to trip the reactor. Corrective Action Reference(s): Condition Report 18-12217
05000528/FIN-2018008-01Inadequate Corrective Actions For Missing Control Room Hand-Switch Operator Knob2018Q3The team reviewed a Green, NRC identified, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct the failures of multiple control room hand-switch operator knobs.
05000530/FIN-2018003-01Failure to Maintain Command and Control During a Feedwater Control Valve Malfunction2018Q3While 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-2018008-05Minor Violation2018Q3Failure to control the issuance of documents, such as instructions, procedures, and drawings, including changes thereto, which prescribe all activities affecting quality as required by 10 CFR 50, Appendix B, Criterion VI. The team identified that the CAP procedure directed the use of the Cause Analysis Manual in performing some cause evaluations. This cause evaluation process is an activity affecting quality required by 10 CFR 50, Appendix B and the licensees Quality Assurance Program. The licensee failed to control the Cause Analysis Manual in accordance with the Palo Verde Nuclear Generating Station Operations Quality Assurance Program Description, Revision 0, Section 2.6, Document Control. The licensee documented this violation in Condition Report 18-13996. Screening: The performance deficiency is minor because if left uncorrected it would not have led to a more significant safety concern and it did not adversely affect any cornerstone objectives. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion VI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000528/FIN-2018002-02Failure to Implement and Maintain Procedures Regarding Breathing Air Quality2018Q2The 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-2018002-03Failure to Assess the Operability of a Degraded or Nonconforming Structure, System, or Component2018Q2The 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-2018002-01Failure to Re-baseline Valve Stroke Times as Required by ASME OM Code2018Q2The 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-2018001-01Inadequate Post Maintenance Test Instructions for Diesel Fuel Oil Transfer Pump2018Q1The inspectors reviewed a self-revealed, Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to prescribe appropriate work instructions for maintenance on the Unit 1 diesel fuel oil transfer pump A. Specifically, following power cable maintenance on November 9, 2017, the instructions for conducting a post-maintenance test for the transfer pump were inadequate to detect a high resistance connection in the associated motor control center.
05000528/FIN-2018012-01Failure to provide adequate guidance to personnel to assure degraded or deficient emergency lighting required for post-fire safe shutdown was corrected in a timely manner2018Q1The licensee provides emergency lighting for access and egress paths, and to illuminate required safe shutdown components to safely shut down the reactor in case of fire requiring control room evacuation, in accordance with 10 CFR 50, Appendix R, Section III.J., Procedure 40DP-9ZZ16, Administrative Controls for Appendix R Equipment, Revision 13, Steps 4.5.1 and 4.6.1, limit the out-of-service time for Appendix R equipment and require restoring the equipment to service within 30 days. The licensee provides guidance to operations department watch standers to check for deficient lighting throughout the plant. Procedure 40DP-9OP20, Watch Standing Practices, Revision 48, Step 4.3.1.1 instructs watch standers to check for adequate or sufficient lighting, among other items during routine tours of the plant. On January 30, 2018, during the walkdown of the control room evacuation due to fire Procedure 40AO-9ZZ19, Control Room Fire, Revision 35, the inspectors identified multiple examples of Appendix R emergency light fixtures that were not functional. The licensee confirmed no condition report had documented these deficient emergency lights at that time. During operator rounds through the area, with offsite power available, the area had adequate and sufficient lighting from the normal AC lighting system. The guidance provided in Procedure 40DP-9OP20 was not adequate to instruct watch standers to identify when an Appendix R emergency light is not functional and to promptly identify the failure to ensure timely restoration of the light.Corrective Actions: The licensee documented the emergency light deficiencies in the corrective action program to initiate repairs, and also issued an Operations Communication Newsflash titled, Actions needed on Emergency (App R) Lighting, dated February 1, 2018. The Newsflash required all operations crews to review procedures for identifying emergency lighting deficiencies and to focus on the emergency light fixtures. The licensee has 30 days to repair and restore the emergency lights to operation. The licensee has a standing compensatory measure for emergency lighting requiring operators to obtain flashlights from the emergency storage locker during a control room evacuation due to fire. The licensee initiated a revision request to provide additional guidance in Procedure 40DP-9OP20, concerning the Appendix R emergency lighting.
05000528/FIN-2017404-01Security2017Q4
05000530/FIN-2017003-04Reactor Trip due to Pressurizer Spray Valve Failing Open due to Volume Booster Internals Not Environmentally Qualified for Anticipated Ambient Operating Temperatures2017Q3The 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-2017003-01Failure to Initiate Corrective Actions for Thermography Tests2017Q3The 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-2017003-03Failure to Follow Conduct of Operations Procedure2017Q3The 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-2017003-02Loss of Refrigerant Failure of Essential Chiller Unit due to Installation of Incorrect Parts2017Q3The 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-2017002-01Inoperable Containment Isolation Valve Due toNot Operating Valve in Accordance with Station Procedures2017Q2The 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-2017002-02Licensee-Identified Violation2017Q2Title 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-2017001-01Failure to establish station procedure instructions for denial work authorizations2017Q1The 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)
05000528/FIN-2017007-01Failure to Analyze Shutdown Cooling and Feedwater Lines for High-Energy Line Break Pipe Whip Effects2017Q1Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, from August 11, 1982, to March 3, 2017, the licensee did not analyze dynamic pipe whip effects of a main feedwater line for a high-energy line break of a shutdown cooling line. In response to this issue, the licensee performed immediate and prompt operability evaluations and determined that the piping systems remained operable and could withstand the effects of a high-energy line break. This finding was entered into the licensees corrective action program as Condition Report CR-17-02815. The team determined that the failure to perform an adequate analysis for shutdown cooling and feedwater lines for high-energy line break pipe whip effects was a performance deficiency. This finding was more-than-minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to analyze the main feedwater piping for high-energy line break effects called the operability of the piping system into question. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as hav ing very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross- cutting aspect because the most significant contributor to the performance deficiency did 3 not reflect current licensee performance. Specifically, the licensee performed the calculation in 1982 and revised it in 1991; therefore, the performance deficiency occurred outside of the nominal three-year period for present performance.
05000529/FIN-2016004-01Inadequate monitoring of MSIV nitrogen pre-charge pressure2016Q4The 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-2016404-01Licensee-Identified Violation2016Q4
05000528/FIN-2016002-01Leakage From Reactor Coolant Pump 2B Discharge Pipe Instrument Nozzle2016Q2The 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.
05000528/FIN-2016002-02Failure to Implement High Radiation Area Controls in an Area with a Dose Rates Greater Than 1 rem per Hour2016Q2The 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-2016002-03Inadequate Engineering and Radiological Controls Resulting in a Unit 1 Containment Building Airborne Radioactivity Event with Unplanned Intakes2016Q2The 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-2016008-02Licensee-Identified Violation2016Q1Title of 10 CFR 50.55a(f)(4), requires, in part, that pumps and valves classified as ASME Code Class 1, 2, or 3 must meet the inservice test requirements set forth in the ASME Operation and Maintenance (OM) Code and addenda to the extent practical within the limitations of design, geometry, and materials of construction of the components. The inservice testing program is incorporated into the Palo Verde Nuclear Generation Station licensing basis under Technical Specification 5.5.8 and governed by the procedures controlled under that specification. ASME OM Code Case OMN-1 was adopted by Palo Verde Nuclear Generating Station per Valve Relief Request number 1, and approved by the NRC as an alternative for performing Code-required valve and pump testing for the second and third 10-year testing intervals (January 1998-2018). ASME OMN-1, Section 3.3.1(b) requires that, if insufficient data exist to determine the inservice test frequency...then (motor operated valve) MOV inservice testing shall be conducted every two refueling cycles or three years until sufficient data exist to determine a more appropriate test frequency. Palo Verde Nuclear Generating Station Procedure 73DP-9ZZ12, Motor Operated Valve Program, Section 4.5.4.5, and Appendix H, define when sufficient test data exists to justify increasing test frequencies beyond 3 years. This criteria includes completing at least two complete diagnostic testing cycles constituting a baseline pre-service test and two subsequent as-found tests. These testing requirements are invoked after complete replacement of the valve, installation of a new valve, or major maintenance, which could substantially change the valve/actuator performance. Contrary to the requirements listed above, the licensee failed to perform Code-required testing for a total of 17 valves between 2008 and 2016. The licensee identified an issue in May 2015 with the testing frequency of five valves after a modification installed new motor operated valves in the charging system. An extent of condition was performed and 11 additional valves were identified as being noncompliant. An engineering evaluation was performed to assess and manage the risk of not completing the required ASME testing per Technical Specification Surveillance Requirement 3.0.3. A prompt operability determination was also performed to provide reasonable assurance of operability until the valves could be tested again. In January 2016, an additional valve was identified as being non-compliant and a separate operability evaluation was completed to provide reasonable assurance that the valve would still perform its function. There are currently seven valves that are still in non-compliance with the Coderequired testing frequency; all other valves have been tested satisfactorily and are now in compliance. Those still requiring testing are scheduled during their respective next available system windows. This violation is of very low safety significance (Green) because the non-conforming valves were determined to have reasonable assurance of operability. The licensee entered the condition into its corrective action program and initiated corrective actions to restore compliance under Condition Report 15-02470.
05000530/FIN-2016001-02Fatigue failure of pneumatic fitting due to excessive vibrations2016Q1The 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.
05000530/FIN-2016001-01Failure to use adequate engineering and radiological controls resulting in two unplanned intakes2016Q1A 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).
05000528/FIN-2016008-01Operations Department Failure to Document Conditions Adverse to Quality in Condition Reports2016Q1The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to document conditions adverse to quality in the corrective action program. Previous similar failures to initiate condition reports led to, or contributed to, two significant conditions adverse to quality over the last 15 months. The failure of the operations department to document identified conditions adverse to quality in condition reports, as required by Procedure 01DP-0AP12, Condition Reporting Process, Revision 23, was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, on two other occasions since January 2015, failures by operations personnel to write condition reports for equipment-related problems resulted in or contributed to significant conditions adverse to quality. This performance deficiency demonstrated a continued gap within Palo Verde Nuclear Generation Stations operations department in understanding condition report initiation criteria. This performance deficiency is associated with the mitigating systems cornerstone. Using NRC Inspection Manual Chapter 0609, Appendix A, the team determined that this finding was of very low safety significance (Green) because it did not affect the operability or functionality of a mitigating structure, system, or component. This finding has a resolution cross-cutting aspect in the area of problem identification and resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3).
05000529/FIN-2016001-03Licensee-Identified Violation2016Q1Technical 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-2015008-01Inadequate Flow Test Procedure2016Q1The team identified a Green non-cited violation of License Conditions 2.C.7, 2.C.6, and 2.F for Units 1, 2, and 3, respectively, because the licensee had not established criteria for determining when a fire main loop had degraded and had not properly tested all portions of the fire main loop. Specifically, the licensee had not established a differential pressure that would initiate actions to evaluate the cause for a degradation and the licensee had not determined the flow through individual flow paths in their auxiliary and control buildings. The licensee documented these issues in Condition Reports 15-00513 and 16-00686 and initiated actions to correct the procedure and perform the flow test of the individual loops. The team identified a performance deficiency related to the procedure used to test their fire main loop. Specifically, the licensee had not established criteria for determining a degraded fire main loop and had not properly tested all portions of the fire main loop. This performance deficiency was more than minor because it was associated with the protection against external factors attribute (fire) and adversely affected 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 failure to test the fire main loops inside the control/auxiliary building separately and failure to establish appropriate acceptance criteria affected the ability to demonstrate the continued capability to deliver adequate flow and pressure to the fire suppression systems. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012. The inspectors determined that an IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, review was required as the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013, the finding was screened as a Green finding of very low safety significance in accordance with Task 1.4.7, Fire Water Supply, Question A. The inspectors determined that although the licensee failed to test portions of the fire main system in accordance with code requirements, the inspectors determined that at least 50 percent of required fire water capacity would be available based on the testing is done with only one fire pump in service and there are three available fire pumps. Since these fire main loops inside the control/auxiliary building had not been monitored for pressure changes when flow tested since initial testing and nothing caused the licensee to reevaluate the test, the team determined that this failure did not reflect current performance.
05000528/FIN-2015004-02Licensee-Identified Violation2015Q4Title 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-2015004-01Licensee-Identified Violation2015Q4Technical 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.
05000529/FIN-2015002-05Failure to Establish Adequate Procedures to Respond to a Total Loss of Charging Event2015Q2The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, through Regulatory Guide 1.33, Revision 2, Appendix A, Section 6.t, February 1978 for the licensees failure to establish adequate procedures for combating emergencies and other significant events regarding a total loss of charging pumps due to gas binding that affected reactor coolant system pressure and level control. On March 20, 2015, after Unit 2 experienced a total loss of charging, operators relied on a normal operating procedure which did not address how to combat a total loss of charging flow due of gas binding from a failed discharge pulsation dampener. The licensee entered this issue into the corrective action program as Condition Report 15-4230. The failure to provide adequate procedures for combating emergencies and other significant events regarding a total loss of charging pumps due to gas binding that affected reactor coolant system pressure control was a performance deficiency. The performance deficiency was more-than-minor and is a finding because it is associated with the procedure quality attribute and directly affected the Initiating Event Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the lack of adequate procedural guidance challenged reactor operators during the loss of charging event. In accordance with Inspection Manual Chapter 0609, Appendix A, "Significance Determination Process (SDP) for Findings AtPower," the performance deficiency was determined to be of very low safety significance (Green) because the finding did not result in a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance because the decision to eliminate the abnormal operating procedure and not to train reactor operators was made in 1997.
05000530/FIN-2015002-04Notice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System Operating Conditions B and C2015Q2(Open) Unresolved Item 05000530/2015002-04, TAC Number MF6276 - NOED Number 15-4-01. Notice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System - Operating Conditions B and C On May 27, 2015, the licensee removed Unit 3 high pressure safety injection train A for planned maintenance. The following morning, during the maintenance, the licensee noted lube oil contamination, and determined that an outboard motor bearing had apparently failed during the last run following maintenance during the last refueling outage which involved disassembling and reassembling the bearing. The licensee identified procedural guidance inadequacies in the reassembly procedure that were the likely cause of the failure. The licensee could not perform required repairs in a controlled manner within the remaining action statement completion time, so on May 29, 2015, the licensee requested a Notice of Enforcement Discretion for a one-time action statement extension of 24 hours to allow time to reassemble and test the replacement bearings prior to restoring operability. The NRC granted that request as NOED 15-4-01. The licensee completed maintenance, testing, and restoration approximately 11 hours into the 24-hour extension window. In accordance with Inspection Manual Chapter 0410, Unresolved Item (URI) 05000530/2015002-04 is opened for NOED 15-4-01, and remains open pending further inspection and disposition in a future inspection report.
05000528/FIN-2015002-01Failure to Verify the Design of the Essential Spray Pond System Crosstie Valves2015Q2The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to maintain adequate design control measures associated with the ultimate heat sink. Specifically, the essential spray pond crosstie valves did not meet design requirements established in Regulatory Guide 1.117, "Tornado Design Classification," as described in the Updated Final Safety Analysis Report. Consequently, if the crosstie valves were damaged by a tornado, the licensee would not have enough available water inventory to meet the mission time of the essential spray pond system. The licensee has added steps to their emergency operating procedure to instruct operators to open the crosstie valves to achieve and maintain long-term cooling subsequent to a design-basis tornado event, and is evaluating potential plant modifications. The licensee has entered this issue into the corrective action program as Palo Verde Action Request 4633058. The failure to verify the design of the essential spray pond system in accordance with Regulatory Guide 1.117 was a performance deficiency. This performance deficiency was more-than-minor and is a finding because it affected the protection against external factors attribute of the Mitigating Systems Cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, if the crosstie valves were damaged by a tornado, the licensee would not have enough available water inventory to meet the mission time for one train of the essential spray pond system during accident conditions. The inspectors performed the initial significance determination for the performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 2, "Mitigating System Screening Questions," dated July 1, 2012. The finding required a detailed risk evaluation because it involved the potential loss of a safety system, in that after at least 13 days of spray pond operation, operators were required to open the spray pond cross-connect valve to enable one train of the ultimate heat sink to use both trains of spray pond inventory. A Region IV senior reactor analyst performed a detailed risk evaluation. The design basis accident mission time was 30 days. However, the probabilistic risk assessment mission time was only 24 hours. Since the spray ponds could still perform the probabilistic risk assessment function for the probabilistic risk assessment mission time, this finding was of very low safety significance (Green). The change to the core damage frequency was much less than 1E-7/year. The finding did not contribute to the large early release frequency. Because the most likely cause of the finding does not reflect current licensee performance, no cross-cutting aspect is assigned to this finding.
05000529/FIN-2015002-03Failure to Identify and Correct Engineered Safety Features Actuation System Steam Generator Differential Pressure Setpoint Drift2015Q2The 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-2015301-01Licensee-Identified Violation2015Q2The following violation of very low safety significance (Green) and Severity Level IV was identified by the licensee and is a violation of NRC requirements, which meets the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited 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, on April 14, 2015, the licensee engaged in an activity that compromised the integrity of the examination. Specifically after administrative JPMs had been administered to the applicants by the examination team, the licensee, upon performing their examination security walk-down, neglected to secure a three-ring binder that contained two reactor operator and two senior reactor operator administrative JPMs that were to be performed the next day. All four JPMs were left unattended and unsecured until 5:00 a.m. on April 15, 2015, when they were discovered as part of the licensee examination security preparation procedure. The four compromised JPMs were replaced by new administrative JPMs as required by NUREG-1021. 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 non-willful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because there was not an actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Integrity of Examinations and Tests. The licensee entered this issue into their corrective action program as PVAR 4645293.
05000529/FIN-2015002-02Failure to Take Timely Corrective Actions to Prevent Charging Pump Discharge Bladder Failure2015Q2The 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-2014404-08Licensee-Identified Violation2015Q1
05000528/FIN-2014404-01Security2015Q1
05000528/FIN-2014404-03Security2015Q1
05000528/FIN-2014404-04Security2015Q1
05000528/FIN-2014404-02Security2015Q1
05000528/FIN-2014404-05Security2015Q1
05000528/FIN-2015001-01Failure to conduct required in-service testing in accordance with ASME OM Code2015Q1The inspectors identified a Green, non-cited violation of Palo Verde Technical Specification 5.5.8 Inservice Testing Program which requires the in-service testing of ASME Code Class 1, 2, and 3 components in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) 2001 Edition with Addenda through 2003. On April 26, 2013, the licensee did not test Unit 1 train A shutdown cooling isolation valve SIA-UV-651, an ASME Code Class 1 valve, in accordance with ASME OM Code Section ISTC-3310.The licensee entered this issue into the corrective action program as Palo Verde Action Request 4398843. The failure to complete ASME OM Code required in-service testing on a Class 1 motor operated valve is a performance deficiency. This performance deficiency is more than minor, and therefore is a finding, because it affected the equipment performance attribute of the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Specifically, by not performing the required testing, the licensee did not maintain the requisite level of assurance of the equipments capability of performing its intended function. Using Inspection Manual Chapter 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the condition was not a design or qualification deficiency, did not involve an actual loss of safety function for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. Because the most-significant contributor to the finding was that Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained, the finding has a cross-cutting aspect in the Human Performance area and the aspect of Teamwork (H.4).
05000528/FIN-2014404-10Licensee-Identified Violation2015Q1
05000528/FIN-2014404-07Security2015Q1
05000528/FIN-2014404-06Security2015Q1
05000528/FIN-2014404-11Licensee-Identified Violation2015Q1
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