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05000528/FIN-2018008-042018Q3Palo VerdeMinor ViolationFailure 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-2018008-032018Q3Palo VerdeLicensee-Identified ViolationThis 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-2018008-022018Q3Palo VerdeLicensee-Identified ViolationThis 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-2018008-012018Q3Palo VerdeInadequate Corrective Actions For Missing Control Room Hand-Switch Operator KnobThe 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.
05000528/FIN-2018008-052018Q3Palo VerdeMinor ViolationFailure 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.
05000382/FIN-2018002-032018Q2WaterfordLicensee-Identified ViolationThis violation of very low safety significance 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: Technical Specification 3.6.3, Containment Isolation Valves, requires, in part, that when an isolation valve for containment penetrations associated with an open system are inoperable, the licensee must restore the inoperable valve(s) to operable status within 4 hours, isolate the affected penetration within 4 hours, or be in hot standby within the next 6 hours. Contrary to the above, between December 8, 2017, and December 11, 2017, with containment isolation valves inoperable, the licensee did not restore the inoperable valves to operable status within 4 hours, isolate the affected penetrations within 4 hours, or place the unit in hot standby within the next 6 hours. The licensee restored the valves to operable status on December 20, 2017, exceeding the Technical Specification 3.6.3 allowed outage time by approximately 70 hours. Significance/Severity Level: The finding was of very low safety significance (Green) because the containment isolation valves were maintained closed during the period and did not represent an actual open pathway in the physical integrity of the reactor containment. Corrective Action Reference: CR-WF3-2018-00983
05000382/FIN-2018002-022018Q2Waterford10 CFR 50.59 Evaluation Associated with Emergency Feedwater Logic ModificationThe licensee changed the emergency feedwater logic, as described in the Updated Final Safety Analysis Report (UFSAR), Section 7.3.1.1.6, from flow control mode to level control mode during a safety injection actuation signal. To accomplish this change, the licensee had to modify the following logic system signals and setpoints: steam generator critical level, steam generator lo level, steam generator lo-lo level, safety injection actuation, control board manual control, and the steam generator lo-lo level annunciator. The NRC team questioned whether the emergency feedwater modification required additional information to determine if the 10 CFR 50.59 evaluation was adequate, or if NRC approval was needed for the change. Specifically, the NRC team questioned if the emergency feedwater logic change: used a method of evaluation other than what was described in the UFSAR (e.g. the use of the TRANFLOW program) or would result in a more than minimal increase in the likelihood of occurrence of a malfunction of a system important to safety. Specifically, because the emergency feedwater logic change introduced the potential to overcool the reactor, and substituted a previous automatic action for manual operator action, the NRC team questioned if the change and associated 50.59 evaluation addressed these concerns. Planned Closure Actions: The NRC and the licensee are working to gather more information related to the Final Safety Analysis Report-described methods for steam generator analyses and if the change resulted in a more-than-minimal increase in risk. Specifically, the licensee plans to provide an analysis that demonstrates the emergency feedwater logic change would not result in a more than minimal increase in the likelihood of an overcooling accident. Licensee Actions: The licensee has implemented a compensatory measure to take manual control of the emergency feedwater system during a safety injection signal such that an overcooling event will be prevented. Corrective Action References: CR-WF3-2017-06067, CR-WF3-2017-05882, CR-WF3-2017-05173
05000382/FIN-2018002-012018Q2WaterfordFailure to Ensure Appropriate Chemistry Controls on the Component Cooling Water Heat ExchangersThe inspectors reviewed a self-revealed, Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which occurred because the licensee did not prescribe procedures for preventing fouling of the component cooling water heat exchangers that were appropriate to the circumstances. Specifically, the licensee did not require in its instructions for adding biocide to the auxiliary component cooling water system that additions be coupled with running the associated auxiliary component cooling water pump or other means of ensuring that the biocide would be sufficiently circulated through the system. As a result, on February 8, 2018, component cooling water heat exchanger B failed a performance test and therefore would not maintain required design basis temperatures under all accident conditions due to biological fouling.
05000528/FIN-2018001-012018Q1Palo VerdeInadequate Post Maintenance Test Instructions for Diesel Fuel Oil Transfer PumpThe 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.
05000382/FIN-2018001-012018Q1WaterfordFailure to Obtain NRC Staff Authorization Prior to Changing a Procedure that Impacts Implementation of Technical SpecificationsThe inspectors identified a Severity Level IV, non-cited violation of 10CFR50.59, Changes, Tests, and Experiments, Section (c)(1), for the licensees failure to submit and obtain authorization prior to implementation procedures described in the Final Safety Analysis Report
05000530/FIN-2017003-042017Q3Palo VerdeReactor Trip due to Pressurizer Spray Valve Failing Open due to Volume Booster Internals Not Environmentally Qualified for Anticipated Ambient Operating TemperaturesThe 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-2017003-012017Q3Palo VerdeFailure to Initiate Corrective Actions for Thermography TestsThe 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-2017003-032017Q3Palo VerdeFailure to Follow Conduct of Operations ProcedureThe 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-2017003-022017Q3Palo VerdeLoss of Refrigerant Failure of Essential Chiller Unit due to Installation of Incorrect PartsThe 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-2017002-012017Q2Palo VerdeInoperable Containment Isolation Valve Due toNot Operating Valve in Accordance with Station ProceduresThe 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-2017002-022017Q2Palo VerdeLicensee-Identified ViolationTitle 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.
05000382/FIN-2017002-012017Q2WaterfordFailure to Prepare the Site for Impending Adverse WeatherThe inspectors identified multiple examples of a non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, for the licensees failure to follow Licensee Procedure OP-901-521, Severe Weather and Flooding, Revision 323. Specifically, on three occasions, the licensee did not close exterior doors when required by the procedure due to potential severe weather conditions. As a result, plant equipment was at an increased failure risk due to severe weather at the site. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2017-03961 and CR-WF3-2017-04944. The licensee is planning corrective actions to ensure doors do not remain blocked open during conditions that require their closure.The performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to maintain all of the doors required by Licensee Procedure OP-901-521 with all fuel offloaded to the spent fuel pool threatened the licensees ability to maintain the functionality of the spent fuel pool cooling system. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process, and determined that a qualitative analysis by a senior reactor analyst was required. The senior reactor analyst determined that the finding was of very low safety significance (Green). Using Inspection Manual Chapter 0609, Appendix M, Signifiance Determination Process Using Qualitative Criteria, the senior reactor analyst performed a bounding analysis indicated that the total increase in core damage frequency from the failure to close the doors during severe weather was less than 1E-6. The finding had a work management cross-cutting aspect in the area of human performance because the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority and the work process did not include the identification and management of risk commensurate to the work and the need for coordination with different groups of job activities. Specifically, during the planning and executing of work activities associated with Refueling Outage 21, the licensee did not consider the nuclear safety implications of blocking open exterior watertight and tornado doors and the work process did not include the identification and management of the risk associated with the blocked-open doors (H.5).
05000382/FIN-2017002-022017Q2WaterfordFailure to Ensure Containment Equipment Hatch Closure Prior to RCS Time to BoilThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, which occurred because the licensee did not implement instructions for maintaining containment integrity. Specifically, on April 18, 2017, the licensee did not ensure that the containment equipment hatch could be closed within the calculated reactor coolant system time to boil as required by Licensee Procedure OP-010-006, Outage Operations, Revision 330. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-02541. The licensees corrective actions included exiting the applicable condition, re-performing the equipment hatch closure drill to show the equipment hatch could be closed prior to the reactor coolant system time to boil, and performing repairs to the containment equipment hatch. The performance deficiency was more than minor because it was associated with thehuman performance attribute of the Barrier Integrity Cornerstone and adversely affected its objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee must close containment penetrations prior to the reactor coolant system time to boil in order to minimize radionuclide releases under accident conditions. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, instructed the inspectors to use Appendix H, Containment Integrity Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because licensee maintained in-depth shutdown capability and because the duration of the performance deficiency was less than 8 hours. The inspectors concluded that the finding had a teamwork cross-cutting aspect in the area of human performance because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, personnel performed work resulting in a short calculated reactor coolant system time to boil without first communicating their actions to operations or the outage control center, resulting in an unexpected plant condition (H.4).
05000382/FIN-2017002-032017Q2WaterfordFailure to Ensure Appropriate Testing of TSP Baskets Inside ContainmentThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure that testing required to demonstrate that structures, systems, and components will perform satisfactorily while in service was identified and performed in accordance with written test procedures incorporating the requirements and acceptance limits contained in the applicable design documents. Specifically, prior to performing Licensee Procedure OP-903-027, Inspection of Containment, Attachment 10.3, Trisodium Phosphate Storage Basket Inspection, the licensee routinely performed a preliminary check to fill the trisodium phosphate storagebaskets, thereby ensuring the successful completion of the technical specification-required surveillance. As a result, following unsatisfactory preliminary checks, the trisodium phosphate storage baskets were not evaluated for past operability. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05108. The licensees corrective actions will include performing the surveillance procedure as an as-found check and evaluating failed surveillances for past operability.The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, conducting preliminary checks of the trisodium phosphate storage baskets and refilling them prior to performing the technical specification surveillance can mask the as-found condition of the test and preclude an evaluation of past operability if the levels are below the technical specification-required values. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, Attachment 1, Exhibit 3, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; (2) did not represent a loss of system safety function; (3) did not represent an actual loss of safety function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; (4) with the cavity flooded, it did not represent an actual loss of safety function of one or more nontechnicalspecification trains of equipment during shutdown designated as risk-significant, for greater than 24 hours; (5) did not degrade the reactor coolant system level indication and/or core exit thermal couples when the cavity was not flooded; (6) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event; (7) did not involve fire brigade training and qualification requirements, or brigade staffing; (8) did not involve the response time of the fire brigade to a fire, and; (9) did not involve fire extinguishers, fire hoses, or fire hose stations. The finding had a change management cross-cutting aspect in the area of human performance because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, when the licensee implemented the preliminary check practice in 2012, they did not evaluate the unintended consequences of how that practice would impact the results of the technicalspecification surveillance. Additionally, the licensee performed the preliminary check during each successive refueling outage between 2012 and 2017 giving the licensee an opportunity to identify the improper practice. As a result, the inspectors concluded this performance deficiency was indicative of current performance (H.3).
05000382/FIN-2017002-042017Q2WaterfordFailure to Perform a Post Maintenance Test on a Main Steam Isolation Valve Solenoid ValveThe inspectors identified a non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, for the licensees failure to perform operability testing on a safety-related component. Specifically, following the coil replacement of main steam isolation valve 2 solenoid valve, a safety-related component, the licensee did not perform a retest of the solenoid valve. As a result, main steam isolation valve 2 was returned to service without the assurance that no new deficiencies had been introduced, calling into question its operability. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05507. The licensees corrective action was to perform a voltage check of the solenoid valve to ensure it would energize in the event that a main steam isolation valve 2 closure was needed.The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee restored main steam isolation valve 2 to an operable status without ensuring that its solenoid valve, which is a main steam isolation valve support system, was properly retested following maintenance.The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix A, Significance Determination Process for Findings At-Power. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) 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 licensees maintenance rule program for greater than 24 hours.The finding had a conservative bias cross-cutting aspect in the area of human performance because individuals did not use decision making-practices that emphasized prudent choices over those that were simply allowable. Specifically, the licensee did not make a conservative decision when determining whether the main steam isolation valve or its solenoid valve should be tested prior to proceeding with plant startup (H.14).
05000382/FIN-2017002-052017Q2WaterfordFailure to Perform Maintenance on the Correct Safety-Related ComponentThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, which occurred due to the licensees failure to perform field work on reactor coolant loop 2 shutdown cooling warm-up valve, SI-135A. Specifically, mechanical maintenance technicians, who were assigned work on safety injection train A, erroneously performed work on safety injection train B on reactor coolant loop 1 shutdown cooling warm-up valve, SI-135B. As a result, both trains of emergency core cooling systems were simultaneously inoperable, which placed the plant in a 1-hour technical specification shutdown action statement. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-01433. The licensees corrective actions included a revision of the model work order to require concurrent verification for component identification, and adding the valves to the protected equipment list for when the opposite train is inoperable.The performance deficiency was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the mechanics worked on valve SI-135B instead of valve SI-135A, they simultaneously made both trains of emergency core cooling systems inoperable. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix A, Significance Determination Process for Findings At-Power. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, and component; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) 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 licensees maintenance rule program for greater than 24 hours.The finding had an avoid complacency cross-cutting aspect in the area of human performance because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, and did notimplement appropriate error reduction tools. Specifically, maintenance technicians repeatedly visited the incorrect work location and didnt properly verify the valve number to ensure they would work on the correct component (H.12).
05000382/FIN-2017002-062017Q2WaterfordLicensee-Identified ViolationLicensee Audit LO-WLO-2016-00037, Bioassay Program, dated November 21, 2016, identified that during Refueling Outage 20, staff reviewing air sample and lapel air sampler results had not been identifying positive results. The audit revealed that two positive lapel air samples from Refueling Outage 20 had not been identified nor had estimated personnel exposures been calculated. In addition, the audit identified seven positive air sample results which had no documented estimated exposures. As a result, dose was not assigned to individuals exposed to airborne radioactivity. As a result of the audit findings, the licensee retroactively assigned dose to three individuals working the October 25, 2015, cavity drain job in the amount of 36 mrem committed effective dose equivalent (CEDE) and 700 mrem committed dose equivalent (CDE) to bone surfaces and to one individual working on a November 8, 2015, decontamination job in theamount of 33 mrem CEDE and 661 mrem CDE to bone surfaces.Title 10 CFR 20.1703 states, in part, the licensee shall implement and maintain a respiratory protection program that includes: (1) air sampling sufficient to identify the potential hazard and estimate doses, and (2) surveys and bioassays, as necessary, to evaluate actual intakes.Contrary to the above, on November 21, 2016, the licensee failed to implement and maintain their respiratory protection program to include air sampling sufficient to identify the potential hazard and estimate doses, and surveys and bioassays, as necessary to evaluate actual intakes. Specifically, for two jobs and four individuals, the licensee failed to identify positive air sample results and assign internal dose to the subject individuals.In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, the inspectors determined that the performance deficiency was more than minor. The finding adversely affected the Occupational Radiation Safety Cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the failure to adequately assess internal exposure affects the licensees ability to control and limit radiation exposure to the worker. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it did not involve: (1) as low as reasonably achievable (ALARA) planning and controls; (2) a radiological overexposure; (3) a substantial potential for an exposure; or (4) a compromised ability to assess the dose.The licensees immediate corrective action was to coach all technicians on surveying airborne areas, ensure all air sample and lapel results were discussed with management, and count all air and lapel samples for alpha and beta to evaluate any potential internal radiation exposure. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2016-07300.
05000382/FIN-2017001-012017Q1WaterfordFailure to Perform Field Changes in Accordance with Design Control MeasuresGreen . The inspectors reviewed a self -revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to perform field changes in accordance with design control measures. Specifically, following maintenance on reactor coolant pump 1B , the licensee performed unauthorized field changes by not reinstalling two design supports for the differential pressure instrument line. As a result, the instrument line developed a vibration- induced fl aw, which caused an increase in reactor coolant system unidentified leakage, and consequently , an unplanned reactor shutdown. The licensee entered this condition into their corrective action program as Condition Report CR- WF3 -2016 -06698. The licensees corrective actions included replacing the damaged instrument line and installing the missing supports. The performance deficiency was more than minor , and therefore a finding, because it affected the equipment performance attribute of the Initiating Events Cornerstone and its 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 licensees failure to reinstall the required supports on the reactor coolant pump 1B instrumentation line resulted in plant operation with increased reactor coolant system unidentified leakage, requiring an unplanned reactor shutdown to perform repairs. The inspectors screened the finding in accordance wit h NRC Inspection Manual Chapter 0609, Significance Determination Process . Using 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 instrument line flaw, after a reasonable assessment of degradation, could not result in exceeding the reactor coolant system leak rate for a small loss -of-coolant accident , and could not likely affect other systems used to mitigate a loss-of-coolant accident , resulting in a total loss of their function. Because the licensees review indicated that no work had been performed in this instrument line within the last three years, and a specific date for the performance deficiency was not identified, the inspectors concluded that the finding does not reflect current licensee performance, and therefore, did not assign a cross -cutting aspect .
05000529/FIN-2017001-012017Q1Palo VerdeFailure to establish station procedure instructions for denial work authorizationsThe 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)
05000285/FIN-2016004-022016Q4Fort CalhounLicensee-Identified ViolationTechnical Specification 2.0.1 requires the unit to be shut down within 6 hours in the event a limiting condition for operation and/or associated action requirement cannot be satisfied because of circumstances in excess of those addressed in the specification. Contrary to the above, the licensee failed to enter Technical Specification 2.0.1 and take the prescribed actions on several occasions when shutdown cooling heat exchanger valves were opened which impacted component cooling water (CCW) flow to the containment air cooling units under certain accident conditions. On May 10, 2016, an unanalyzed condition was discovered during scheduled maintenance on the shutdown cooling heat exchanger valves. As part of the maintenance, HCV-484, Shutdown Heat Exchanger AC-4A Component Cooling Water Outlet Valve, and HCV-481, Shutdown Cooling Heat Exchanger AC-4B Component Cooling Water Inlet Valve, were failed open which rendered both valves inoperable. Under these conditions, with the assumed single failure loss of DC control power during a loss of coolant accident (LOCA), CCW would be allowed to flow through both shutdown cooling heat exchangers, effectively reducing CCW system flow to the containment air cooling units. These conditions are not assumed under plant design basis calculations and placed the plant in an unanalyzed condition. It has not been demonstrated that the CCW system would adequately perform its design function of providing a cooling medium for the containment atmosphere under LOCA conditions with CCW flow diverted through the shutdown cooling heat exchangers. With two containment air cooling units inoperable, Technical Specification 2.4, does not provide an associated action; therefore, Technical Specification 2.0.1 applies. Upon completion of the maintenance activity, both valves were returned to service which eliminated the condition. The licensee conducted an extent of condition review and identified that they had created this unanalyzed condition six times within the last 3 years and had exceeded the Technical Specification 2.0.1 6-hour shutdown action statement on March 8, 2016; April 21, 2016; and May 10, 2016. In addition, the licensee determined this condition was first identified on February 3, 2015, in Condition Report 2015-01388. Procedure TDB-VIII, Equipment Applicability Guidance, Revision 64, incorrectly stated the valves had a required safety function in the open direction. The licensee initiated procedure change EC-68088 on September 26, 2015, to correct the procedure; however, the proposed change did not accurately reflect the safety function of the valves to remain closed for all LOCA conditions. This procedure change was still under review on May 10, 2016. The failure to promptly correct Procedure TDB-VIII was a contributing cause of the violation. The violation is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone. On March 8, 2016; April 21, 2016; and May 10, 2016, the plant was placed in a condition prohibited by technical specifications and exceeded the Technical Specification 2.0.1, 6-hour shutdown action statement. This adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A senior reactor analyst qualitatively determined that this finding was of very low safety significance (Green) for increases in core damage frequency and large early release frequency because of the short exposure time of less than 3 days and because of the low frequency of events where a LOCA with an independent and coincidental loss of DC control power would occur. Therefore, this finding screens to Green. The licensee entered the issue into their corrective action program as Condition Reports 2016-05340 and 2016-04468.
05000285/FIN-2016004-032016Q4Fort CalhounLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, though the licensee identified a potential vulnerability to raw water pumps from a missile hazard striking diesel driven fire pump FP-1B or associated piping during review of missile hazards during the 2013 tornado missile project, the licensee failed to evaluate this condition or specify a modification to the plant to protect the raw water pumps at that time. This was discovered on August 25, 2016, by the licensee during a design review. This finding is of very low safety significance (Green) considering compensatory measures that were put in place to disable pump FP-1B and isolate associated piping when severe weather is forecast and the very low probability of the postulated event. This issue was entered into the licensees corrective action program as CR 2016-06972.
05000285/FIN-2016004-042016Q4Fort CalhounLicensee-Identified ViolationTitle 10 of the Code of Federal Regulations, Part 50.9(a), requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on December 26, 2014, Fort Calhoun Station provided information to the Commission which was not complete and accurate in all material respects. Specifically, a license amendment request (ML14365A123) to adopt a scheme of emergency action levels based on Nuclear Energy Institute Document 99-02, Revision 6, contained inaccurate information about the characteristics of the cask used in the licensees Independent Spent Fuel Storage Installation and, as a result, incorrect external radiation levels were incorporated into emergency action level E-HU1. Subsequently, while preparing another emergency action level submittal, the emergency preparedness staff discovered the incorrect information that had previously been submitted. The issue was determined to be a Severity Level IV violation of NRC requirements, in accordance with Section 6.9 of the Enforcement Policy, dated November 1, 2016, because the inaccurate information would not have caused the NRC to reconsider a regulatory position or undertake substantial further inquiry. The issue was documented in the licensees corrective action program as Condition Report CR-2016-08400. Because the Severity Level IV violation has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000529/FIN-2016004-012016Q4Palo VerdeInadequate monitoring of MSIV nitrogen pre-charge pressureThe 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).
05000382/FIN-2016004-012016Q4WaterfordFailure to Ensure Appropriate Post-Maintenance Testing on Essential Chiller BGreen. A self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, occurred because the licensee did not assure that the procedures for post-maintenance testing of activities affecting quality included appropriate quantitative or qualitative acceptance criteria for determining that maintenance activities were satisfactorily accomplished. Specifically, the licensee did not assure that post-maintenance testing of essential chiller B would identify inappropriately assembled guide vanes, following maintenance on April 11, 2016, resulting in the unexpected inoperability of essential chiller B on August 12, 2016. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2016-05155. The corrective action taken to restore compliance was to issue work instructions for post-maintenance testing of the essential chillers that ensures the guide vanes respond to load changes. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance 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 perform maintenance with procedures appropriate to the circumstances resulted in the inoperability of essential chiller B. The inspectors determined the significance of the finding using NRC Inspection Manual Chapter 0609, Significance Determination Process. Using Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green) because all the screening questions in Exhibit 2, Mitigating Systems Screening Questions, were answered No. The finding had a cross-cutting aspect in the area of human performance, teamwork, because the licensee did not ensure that individual and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, electrical and mechanical maintenance personnel did not communicate and coordinate work to ensure that the guide vane arm and actuator linkage were assembled appropriately (H.4).
05000285/FIN-2016004-012016Q4Fort CalhounFailure to Provide Training on Changes to Protective Action Recommendation ProceduresThe inspector reviewed a self-revealed non-cited violation associated with Fort Calhoun Stations failure to provide radiological emergency response training to those who may be called upon to assist in an emergency, as required by 10 CFR 50.47(b)(15). Specifically, in December 2014, 10 shift managers and 6 Technical Support Center and Emergency Operations Facility staff, responsible for making and reviewing protective action recommendations, were not trained on Procedure EPIP-EOF-7, Protective Action Recommendations, Revision 26, and flowchart EP-FC-111-AD-F-02, before they were implemented on December 23, 2014. As immediate corrective actions, the licensee issued a reading package covering the new protective action recommendation process to the 16 individuals who had not been trained. The issue was entered into the licensees corrective action program as Condition Report CR-2015-08951. The failure to provide radiological emergency response training to those who may be called upon to assist in an emergency is a performance deficiency within the licensees ability to foresee and correct. The performance deficiency was more than minor because it was associated with the procedure quality attribute of Emergency Prepardness Cornerstone and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was evaluated using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 22, 2015, and was determined to be of very low safety significance (Green) because it was a failure to comply, was not a risk significant planning standard function, was not a loss of the planning standard function, and was a degraded planning standard function. This finding had a cross-cutting aspect in the area of human performance associated with change management because the emergency preparedness department failed to identify all of the emergency response organization staff who required training on revisions to the process for making protective action recommendations (H.3).
05000382/FIN-2016003-022016Q3WaterfordLicensee-Identified ViolationTitle 10 CFR 50.72, Immediate notification requirements for operating nuclear power reactors, section (b)(3)(v), requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (A) shut down the reactor and maintain it in a safe condition; (B) remove residual heat; (C) control the release of radioactive material; or (D) mitigate the consequences of an accident. Contrary to the above, on August 12, 2016, the licensee experienced a loss of the essential chilled services water safety function, which is needed to mitigate the consequences of an accident, and did not notify the NRC within 8 hours. The licensee identified this issue and entered it into their corrective action program as CR-WF3-2016-05188 and made the required notification on August 15, 2016. This violation was assessed using Section 2.2.4 of the NRCs Enforcement Policy, revised February 4, 2015. Using the example listed in Section 6.9.d.9, A licensee fails to make a report required by 10 CFR 50.72, the issue was determined to be a Severity Level IV violation.
05000285/FIN-2016002-022016Q2Fort CalhounFailure to Develop Adequate Procedures for Post Modification TestingThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.8.1., for failure to establish, implement, and maintain a procedure recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Specifically, the licensee failed to develop adequate procedures for testing equipment important to safety. The licensee failed to identify and mitigate all possible turbine logic trip signals when testing Distributed Control System logic. Following the logic modification of the Turbine Control System, post modification testing inserted two Emergency Trip System test signals which caused an automatic turbine trip resulting in an automatic reactor protective system scram actuation. Failure to establish, implement, and maintain procedures as required by technical specifications is a performance deficiency. The performance deficiency is more than minor because it adversely affected the procedure quality attribute of the initiating event cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to mitigate all possible turbine trip signals while testing the Distributed Control System which caused the turbine to trip and thereby caused a loss of load reactor trip. Using NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, the finding screened as having very low safety significance (Green) because although the deficiency resulted in a reactor trip, the trip was uncomplicated and mitigating equipment remained unaffected. This finding has a cross-cutting aspect in the teamwork component of the human performance cross-cutting area because the licensee did not ensure that individuals and work groups communicate across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Distributed Control System expert did not review the post modification testing procedure prior to implementation (H.4).
05000528/FIN-2016002-022016Q2Palo VerdeFailure to Implement High Radiation Area Controls in an Area with a Dose Rates Greater Than 1 rem per HourThe 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-2016002-032016Q2Palo VerdeInadequate Engineering and Radiological Controls Resulting in a Unit 1 Containment Building Airborne Radioactivity Event with Unplanned IntakesThe 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).
05000382/FIN-2016002-022016Q2WaterfordFailure to Properly Assess and Manage Risk When Performing Dry Cooling Tower MaintenanceThe inspectors identified a non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, section (a)(4) because the licensee did not properly assess and manage risk associated with maintenance on the dry cooling tower fans train B. Specifically, the licensee failed to adequately assess risk and take appropriate risk management actions when replacing a logic card associated with the dry cooling tower train B fans. As a result, an electrical transient occurred that caused unexpected valve movements in component cooling water and auxiliary component cooling water train B systems, an unexpected start of the auxiliary component cooling water pump train B, and the unexpected shutdown of essential chiller train AB. The licensee entered this issue into their corrective action program as condition report CR-WF3-2016-04084. Corrective actions included reassessing the risk associated with the maintenance and identifying appropriate risk management actions to use when performing similar maintenance activities in the future. The inspectors determined that the performance deficiency was more than minor because it was associated with the Configuration 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 take appropriate risk management actions resulted in unexpected valve movements, an unexpected start of auxiliary component cooling water pump B, and an unplanned entry into Technical Specification 3.7.4, Ultimate Heat Sink. The inspectors used Inspection Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, Flowchart 2, Assessment RMAs, and determined the need to calculate the incremental core damage probability to determine the significance of this issue. The Waterford probabilistic risk assessment model yielded an incremental core damage probability, or actual increase in risk during this work window, of 1.5x10-8. In accordance with Flowchart 2 in Appendix K, because the incremental core damage probability was less than 1x10-6, the finding screened as having very low safety significance (Green). This finding had a Procedure Adherence cross-cutting aspect in the area of Human Performance because individuals did not follow processes, procedures and work instructions. Specifically, the licensee did not assess and manage the risk associated with the maintenance in accordance with EN-WM-104, On Line Risk Assessment (H.8).
05000382/FIN-2016002-032016Q2WaterfordFailure to Perform Drills Required by the Site Emergency PlanThe inspectors identified a non-cited violation of 10 CFR 50.54(q)(2), which requires a power reactor licensee to follow and maintain the effectiveness of the site emergency plan. Specifically, Waterford Steam Electric Station, Unit 3, failed to conduct two proficiency drills in calendar year 2015 as required by the Site Emergency Plan, Revision 46, Section 8.1.2.4. The licensee has initiated work tracker surveillances to ensure all drills required in 2016 are performed. The issue is more than minor because the finding was associated with the Emergency Response Organization Performance attribute and adversely affected the Emergency Preparedness cornerstone objective to ensure the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was evaluated using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 23, 2014, and was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements, was not a risk-significant planning standard function, and was not a lost or degraded planning standard function. The inspectors determined that the finding had a Work Management cross-cutting aspect in the area of Human Performance, because the emergency preparedness department did not properly schedule, oversee, and manage required activities (H.5).
05000382/FIN-2016002-012016Q2WaterfordFailure to Properly Pre-Plan and Perform Maintenance on the Cable Vault and Switchgear Ventilation SystemThe inspectors identified a non-cited violation of Technical Specification 6.8, Procedures and Programs, associated with the licensees failure to properly pre-plan and perform maintenance on safety-related components in accordance with EN-DC-335, Preventative Maintenance Basis Template. Specifically, the licensee did not follow the required preventive maintenance basis template for the safety-related cable vault and switchgear ventilation system, and was performing vibration monitoring of these components on an 18-month frequency instead of the required 3-month frequency. As a result, the licensee was deviating from the industry standard preventive maintenance recommendations without documented technical bases, and the required preventive maintenance tasks on these safety-related components were not performed. The licensee entered this condition into their corrective action program as condition report CR-WF3-2016-02353. The licensee restored compliance by assigning the proper preventive maintenance activities for the components in this system and instituting the appropriate frequency. In addition, a maintenance scope review is being performed. The performance deficiency was more than minor because it affected the Equipment Performance attribute of the Mitigating Systems cornerstone and its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, actions to detect, preclude and address degradation of the safety-related components were delayed. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low significance (Green) because all the screening questions in Exhibit 2 Mitigating Systems Screening Questions were answered No. The finding had an Identification cross-cutting aspect in the area of Problem Identification and Resolution because individuals did not identify issues completely, accurately, and in a timely manner in accordance with the corrective action program. Specifically, during previous vibration tests, the licensee had opportunities to identify the incorrect classification of the preventive maintenance task but did not do so (P.1).
05000382/FIN-2016002-042016Q2WaterfordFailure to Account for Starting Air Design Features in Emergency Diesel Operating ProceduresA self-revealing, Green, non-cited violation of Technical Specification 6.8, Procedures and Programs, occurred because the licensee did not establish adequate procedures for the operation of the emergency diesel generators. Specifically, prior to July 7, 2015, the licensees procedure for operating the emergency diesel generators allowed lube oil pressure to be maintained low enough to activate a design feature of the starting air system that injects starting air into the diesel cylinders, which could damage the emergency diesel generator turbocharger. The licensee entered this issue into their corrective action program as condition report CR-WF3-2015-04459. The corrective action taken to restore compliance was to increase the procedure requirement for operating lube oil pressure from 35 psig to 45 psig. The inspectors concluded that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the procedural allowance to run the emergency diesel generator lube oil pressure at the starting air injection setpoint could have resulted in the failure of the emergency diesel generators when they were called upon to perform their safety function. The inspectors used NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, to determine the significance of the finding. The inspectors determined that the finding required a detailed risk evaluation because it represented the loss of a system or function. The detailed risk evaluation determined that the finding is of very low safety significance (Green). The senior reactor analyst estimated the increase in core damage frequency to be 4.6E-7/year and the increase in large early release frequency to be 3.9E-8/year. Dominant core damage sequences were medium break losses of coolant accidents and steam generator tube ruptures with associated losses of off-site power. Core damage was mitigated by the remaining emergency diesel generator. This finding had an Evaluation cross-cutting aspect in the area Problem Identification and Resolution, because the licensee did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensees previous evaluation performed for operating the emergency diesel generators with low lube oil pressures did not thoroughly evaluate the risk associated with the starting air system (P.2).
05000285/FIN-2016002-012016Q2Fort CalhounFailure to Perform an Adequate Evaluation of Service Life for Component Cooling Water Pump MotorsThe inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion III, design control, associated with the licensees failure to perform an adequate evaluation of the service life of component cooling water pump motors. Specifically, the licensee operated component cooling water pump motors beyond the vendor recommended horsepower, temperature and voltage limits for the pumps which resulted in the potential for early winding failure of the motors. The licensees existing calculation determined a component cooling water pump motor life of 16.9 years. During the inspection, the licensee re-evaluated component cooling water pump motor life and determined the expected motor life was actually between 6.8 (if degraded voltage is considered) and 7.2 years. Actual in-service life of the longest operating component cooling water pump was approximately 6.13 years. The licensee entered this issue into the corrective action program as Condition Report 2016-04319. The inspectors determined that the failure to adequately evaluate the service life of the component cooling water pump motors is a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The inspectors determined the finding was of very low safety significance in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, because although the finding was a deficiency affecting the design of a mitigating system, it did not result in a loss of operability or functionality. Specifically, although there was a significant reduction in the calculated service life of the component cooling water pump motors, the actual in-service life of the longest operating component cooling water pump was approximately 6.13 years, which is still encompassed by the revised service life calculation. The finding does not have a cross-cutting aspect because the failure to perform an adequate service life evaluation for component cooling water pump motors is not indicative of current licensee performance. The licensees current design process requires reviews of in-service temperature effects on equipment service life including pump motors.
05000528/FIN-2016002-012016Q2Palo VerdeLeakage From Reactor Coolant Pump 2B Discharge Pipe Instrument NozzleThe 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.
05000285/FIN-2015004-032015Q4Fort CalhounLicensee-Identified ViolationTechnical Specification (TS) 2.5(1) requires two trains of auxiliary feedwater (AFW) to be operable when cold leg temperature is above 300F. In the event that both trains become inoperable, immediate action is required to restore one AFW train to operable status. Technical Specification 2.0.1 and all TS actions requiring mode changes are suspended until one AFW train is restored to operable status. Operation with the main and auxiliary feedwater cross-tied was a violation of the technical specification requirements to maintain operability of AFW systems. The violation is more than minor because it is associated with the configuration control attribute of the mitigating systems cornerstone because the failure to prevent cross-tying these systems resulted in unrecognized inoperability of both trains of AFW. This adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The violation was of very low safety significance because although MFW and AFW were momentarily cross-tied, this condition existed for only a brief period of time as operators restored system line-ups following system testing. In addition, a Senior Reactor Analyst evaluated the postulated main feedwater line break frequency and exposure time of the condition and determined the likelihood of this event during the exposure time is less than the Green/White threshold and of very low safety significance. The licensee entered the issue into their corrective action program as Condition Report 2015 03698.
05000285/FIN-2015004-022015Q4Fort CalhounLicensee-Identified ViolationTechnical Specification (TS) 2.4(1)a.iv requires that all valves, piping, and interlocks associated with the components of the containment cooling system required to function during accident conditions be operable. In the event that any of these components, required to function during accident conditions become inoperable, the reactor shall be placed in a hot shutdown condition within 12 hours. The containment spray pumps and the associated piping are part of the containment cooling system. Prior to making modifications to containment spray piping in 2015, the operability of this piping would have been challenged by a main steam line break or a loss of coolant accident due to thermal stresses induced in the piping after a rise in containment temperature after the postulated event. Operation prior to the implementation of the modifications was a violation of the technical specification requirements to maintain operability of containment cooling systems. The violation is more than minor because it is associated with the design control attribute of the mitigating systems cornerstone because the failure to anticipate the rise in containment spray piping temperature dates back to the original design of the plant. This adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The violation was of very low safety significance because although the subject piping was inoperable due to exceeding code specified stress limits, analysis showed that the piping would have been able to perform its safety function to deliver adequate containment spray flow in the event of an accident. The licensee entered the issue into their corrective action program as Condition Report 2015-04578.
05000285/FIN-2015004-012015Q4Fort CalhounLicensee-Identified ViolationTitle 10 CFR 72.174 requires that each licensee maintain sufficient records to furnish evidence of activities affecting quality. Records pertaining to the design, fabrication, erection, testing, maintenance, and use of structures, systems, and components important to safety must be maintained by or under the control of the licensee until the NRC terminates the license. Contrary to the above, as of June 21, 2013, Fort Calhoun failed to maintain sufficient records to furnish evidence of activities affecting quality. Specifically, the licensee did not maintain records for loading activities associated with DFS-HSM-06 that was placed on the ISFSI pad in July of 2009. This violation was identified by FCS and placed in their corrective action program (CR 2013-12884). The fuel assembly data was reconstituted based on records from the Reactor Engineering group, and the canister helium leak-test data was reconstituted based on the helium leak-test technician's field notes. The remaining canister records associated with the canister processing, sealing, and transportation to the ISFSI, including several TS requirements, were not found. Fort Calhoun reconstituted the fuel and helium leak test data and conducted interviews with cask loading personnel to conclude that there was no evidence to suggest that loading activities did not comply with the licensee's procedures and the licensed Technical Specifications. This violation did not have any safety impact because all fuel assemblies met the requirements for burn-up, decay heat, and cooling time and the licensee demonstrated that the canister integrity was intact based on the reconstituted helium leak test records. All the fuel inside the canister and the cask remain in a safe condition. This finding was reviewed by NRC Headquarters Division of Spent Fuel Managements Spent Fuel Licensing Branch. Based on the reconstituted records and interviews with the dry fuel loading staff, the NRC found no evidence to demonstrate that the canister did not meet the required license conditions and as such, found the canister acceptable for continued storage under FCSs general Part 72 license. However, though the canister is acceptable for storage, the licensee must track this issue to identify that further analyses may be required for this canister to meet all applicable Part 71 requirements to be acceptable for transportation. In accordance with the NRC Enforcement Policy Section 2.2 and IMC 0612 Section 03.23, Part 72 ISFSI inspection findings follow the traditional enforcement process and are not dispositioned through the Reactor Oversight Process or the Significance Determination Process. The violation screened as having very low safety significance, Severity Level IV, and is being treated as a non-cited violation, consistent with Section 2.3.2.a. of the Enforcement Policy. The violation was determined to be more than minor since the licensee failed to establish, maintain, or implement adequate controls over procurement, construction, examination, or testing processes that are important to safety. The violation was entered into the licensees corrective action program as CR 2013-12884. Following identification of the issue the licensee performed an assessment that showed the cask would continue to perform its design function. Corrective actions for this issue included performing an extent of condition review, performing an apparent cause analysis report, reconstitution of the missing documents, conducting interviews with the dry cask loading personnel, providing training to the staff involved, and changing processes and responsibilities within FCS Records Management Group.
05000285/FIN-2015003-022015Q3Fort CalhounFailure to Maintain Fire Watch and Fire Watch LogsInspectors 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.
05000382/FIN-2015003-012015Q3WaterfordFailure to Establish Design Control Measures for Safety-Related Emergency Feedwater System ValvesThe inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to verify the adequacy of the design of the emergency feedwater system. As a result, on June 3, 2015, following a manual plant trip that occurred due to a loss of the main feedwater system, the emergency feedwater back-up flow control valves oscillated so severely that control room personnel removed the system from automatic operations and manually controlled flow to the steam generators. The licensee entered this condition into their corrective action program as condition report CR-WF3-2015-03565. Long term corrective actions are to develop a modification to the system for better flow control, and complete testing that would demonstrate the automatic function of these valves. The performance deficiency is more than minor because it is 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 ensure that the safety-related emergency feedwater back-up flow control valves would perform as designed, impacted the systems ability to perform its safety function during the feedwater loss event on June 3, 2015. A bounding detailed risk evaluation determined that the finding was of very low safety significance (Green) and was not significant to the large early release frequency. The dominant sequences included losses of off-site power, failure of the backup essential feedwater valves in the closed direction, and random failures of the primary essential feedwater flow control valves in the closed direction. The primary essential feedwater flow control valves and the diversity of the emergency feedwater system helped to minimize the risk. The finding does not have a cross-cutting aspect because the most significant contributor to the performance deficiency of not identifying the design flaws or the need for a test occurred more than two years ago and did not reflect current licensee performance.
05000285/FIN-2015003-012015Q3Fort CalhounFailure to Maintain Safety Injection Tank Boron Concentration within Technical Specification LimitsA Green, self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI Corrective Action was identified because the licensee failed to identify and evaluate an adverse trend related to boron concentration in Safety Injection Tank (SIT) SI-6A and to take corrective actions to prevent boron concentration from going below the minimum concentration required by Technical Specifications. The licensees immediate corrective actions included documenting this condition in their corrective action program in Condition Report (CR) 2015-10181, declared SI-6A inoperable, and raised SI-6A boron concentration. The finding is more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone, in that this finding resulted in the SIT becoming inoperable when boron concentration fell below TS limits for approximately 8.5 days prior to August 20, 2015. Analysis conducted by a Senior Reactor Analyst determined the finding to be of very low safety significance (Green), primarily because the SIT function is needed only for mitigation of a postulated large-break loss of coolant accident, and the initiating-event frequency for such accidents is 2.5 x 10-6/year. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution and the Evaluation aspect, because the licensee did not thoroughly evaluate the issue and ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance.
05000285/FIN-2015011-032015Q3Fort CalhounFailure to Correct a Non-Conforming Condition Associated with Auxiliary Feedwater ValvesThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality. Specifically, the licensee failed to take corrective actions after identifying that the steam generator auxiliary feed containment isolation valves were not rated for the maximum temperature they would experience in service. The inspectors determined that on February 2, 2015, an NRC inspector questioned the licensee whether valves HCV-1107A and HCV-1108A were adequately designed for containment temperatures. The licensee determined that the design specification for the valves was 180F, and the containment temperature following a main steam line break was evaluated to be 374F. The fact that the valve was not designed for the most limiting conditions was a non-conforming condition of a safety related component, and was a condition adverse to quality. However, the licensee did not initiate a condition report to resolve and correct the condition. Additionally, the inspectors determined that in 2002, the licensee initiated Condition Report CR-2002-02124 after identifying elevated temperatures in the auxiliary feedwater piping. This condition report documented that the design specification for the two valves was 180F and had been exceeded in service. Although the condition report description recommended modifying the design of the valves, the licensee did not take actions to correct the condition. In both of these instances, the licensee recognized that the valve design temperature was not adequate for its application, but did not take action to resolve the discrepancy. The inspectors determined that although the inadequate design was a non-conforming condition, the valves were not inoperable until the licensee installed inappropriate elastomer material during the 2015 refueling outage as a result of inadequate design control. The licensee entered the failure to identify and correct the non-conforming design in their corrective action program as Condition Report CR-2015-08523. The licensees failure to take corrective action for a non-conforming condition was a performance deficiency. The performance deficiency is more than minor, and therefore 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 (i.e. core damage). Specifically, the licensee failed to take corrective actions to ensure an adequate design for the steam generator auxiliary feed containment isolation valves HCV-1107A and HCV-1108A. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because although the finding was a deficiency affecting the design or qualification of a mitigating system, structure, or component, the system, structure, or component maintained its operability. The finding has a basis for decisions cross-cutting aspect in the human performance cross-cutting area since leaders and individuals did not verify their understanding or question the basis of decisions. Specifically, the licensee failed to understand the potential significance of the non-conforming design of the valves and the basis for not taking corrective actions (H.10). (Section 4OA5.6)
05000285/FIN-2015011-022015Q3Fort CalhounFailure to Establish a Technical Basis for Operability of the Auxiliary Feedwater SystemThe 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 follow the operability determination procedure. Specifically, the licensee failed to establish a valid technical basis for operability of auxiliary feed containment isolation valves HCV-1107A and HCV-1108A. Following the valves failure on June 5, the licensee replaced the failed valve elastomers with new PTFE seals and nitrile O-rings. The licensee then performed an operability evaluation that considered the effect of high temperatures from a main steam line break on the valve elastomers. The inspectors found that the evaluation was not sufficient because it did not determine that the new O-rings would function under all potential temperature conditions and did not consider the function of the other valve components. The licensee entered these issues in their corrective action program as Condition Report CR-2015-08362 and revised their operability evaluation. The licensees failure to follow the operability determination procedure was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the licensee failed to sufficiently address the capability of the steam generator auxiliary feed containment isolation valves HCV-1107A and HCV-1108A to perform their safety function, requiring significant further analysis to demonstrate operability. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because although the finding was a deficiency affecting design or qualification, but the mitigating structure, system or component maintained its operability. The finding has a consistent process cross-cutting aspect in the human performance cross-cutting area since the organization did not use a consistent, systematic approach to make decisions and incorporate risk insights appropriately. Specifically, the licensee failed to re-evaluate the operability decision when new information on the conditions and susceptibility affecting valves HCV-1107A and HCV-1108A during normal operations was available (H.13).
05000285/FIN-2015011-012015Q3Fort CalhounFailure to Ensure the Suitability of Replacement Materials during the Design Review ProcessThe inspectors reviewed a Green, self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the suitability of materials utilized during the design review process. Specifically, the licensee failed to identify during the design review process that replacement valve internal seal materials for the steam generator auxiliary feed containment isolation valves would not be suitable for high temperature conditions that the valves would experience in service, and as a result, caused both trains of the safety-related auxiliary feedwater system to become inoperable during hot standby conditions. The licensee entered this issue into their corrective action program as Condition Report CR-2015-07564 and replaced the valve internals with material that had been previously installed in valves HCV-1107A and HCV-1108A before the modification. The inspectors determined that the licensees failure to evaluate the suitability of the materials used during the design review process for the steam generator auxiliary feed containment isolation valves was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the licensees failure to properly evaluate the suitability of CTFE for use in the steam generator auxiliary feed containment isolation valves led to the failure of HCV-1107A and HCV-1108A and rendered both safety-related trains of auxiliary feedwater inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation since the finding represented a loss of system and/or function. A Region IV senior reactor analyst performed the detailed risk evaluation in accordance with Appendix A, Section 6.0, Detailed Risk Evaluation. The detailed risk evaluation result is a finding of very low safety significance (Green). The calculated change in core damage frequency of 2.3 x 10-7 was dominated by a loss of offsite power; common cause failure of the auxiliary feedwater discharge air-operated valves; failure of diesel-driven auxiliary feedwater pump FW-54; failure of the feed and bleed operation; and failure of operators to manually override a steam generator isolation signal and establish a flowpath for the main feedwater system. The analyst determined that the finding did not involve a significant impact to external initiators because of the short exposure time, or a significant increase in the risk of a large, early release of radiation. The finding has an operating experience cross-cutting aspect in the problem identification and resolution cross-cutting area since the organization did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, readily available internal operating experience on the high temperature conditions that valves HCV-1107A and HCV-1108A experienced during normal operations was not utilized during the design change process (P.5).
05000382/FIN-2015003-022015Q3WaterfordFailure to Follow Procedures when Changing Materials Used for Feedwater Heater Level Control ValvesThe inspectors reviewed a self-revealing finding of very low safety significance that occurred because the licensee did not follow procedural guidance when changing materials used for feedwater heater level control valves. As a result, a feedwater heater normal level control valve failed unexpectedly, causing a trip of feedwater pump A and ultimately resulted in a plant trip. The licensee entered this issue into their corrective action program for resolution as condition report CR-WF3-2015-03563. The immediate action taken to restore compliance was to replace the valve internals with those of appropriate materials. The performance deficiency was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. A detailed risk evaluation determined that the finding was of very low safety significance (Green). The bounding change to the core damage frequency was less than 4E-7/year, and the finding was not significant with respect to the large early release frequency. The dominant core damage sequences included transients with the common-cause failure of the essential chilled water system and the failure of the turbine driven emergency feedwater pump. This finding has a cross-cutting aspect in the Evaluation aspect of the Problem Identification and Resolution area because the licensee did not thoroughly evaluate the causes of several previous feedwater level control valve failures.
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