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05000255/FIN-2018003-012018Q3PalisadesWire Not Landed on Safety Injection Initiation Relay CircuitThe inspectors identified a Green finding and an associated non-cited violation (NCV)of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish an activity affecting quality in accordance with the implementing procedure. Specifically, only one of two required wires was landed on terminal 13 of relay SIS2 in the right channel of the safety injection system (SIS) actuation logic following surveillance testing that was performed on May 8, 2017. As a result, the right channel of the safety injection system actuation logic was inoperable until the problem was discovered during troubleshooting and the wire was subsequently re-landed onMay 3, 2018
05000374/FIN-2017010-012017Q4LaSalleFailure To Ensure Fire Door Was Engaged And PinnedThe inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation of License Condition 2.C.15 for Unit 2, for the licensees failure to ensure all fire rated assemblies (i.e., fire doors) were operable. Specifically, during a plant walk down, the inspectors found Fire Door 282 inoperable. The lower pin of the stationary part of the double door was not engaged, because the pin was broken.The licensee entered the issue into their Corrective Action Program and as an immediate action, declared the door inoperable, established hourly fire watch, and subsequently installed a new pin.The inspectors determined that the performance deficiency was more-than-minor because the finding was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the issue screened as having very-low safety significance (Green) by answering Yes to Question 1.4.3.A of IMC 0609, Appendix F, Attachment 1 based on no combustible within 10 feet of Door 282 on the 5A4 side and one pin should still provide sufficient defense-in-depth for several hours before buckling or moving out of the frame. The finding had a cross-cutting aspect in the Procedure Adherence component of the Human Performance cross-cutting area. Specifically, the licensee failed to follow procedural guidance to thoroughly verify that fire doors were pinned when challenging the doors. (H.8)
05000237/FIN-2017004-012017Q4DresdenFailure to Follow Procedure,Results in Non-Functional Fire DoorThe inspectors identified a finding of very-low safety significance and associated NCV of Technical Specification 5.4.1.c for the licensees failure to implement the established Fire Protection Program procedures which ensure Fire Barrier Integrity. Specifically, the licensee ran an electrical cable through the doorway of an automatically closing fire door. This was contrary to Procedure DFPP 417501, which requires in part that fire doors must not be blocked open by props or any other material in its closing path. The licensee took immediate actions to restore the fire door, by removing the obstruction and entered the issue into their Corrective Action Program (CAP). The inspectors determined that the performance deficiency was more-than-minor because it affected the Mitigating Systems cornerstone objective since the electrical cable could have prevented the fire door from performing its function. The finding was of very-low safety significance per Task 1.4.3A of IMC 0609, Appendix F. Specifically, the total combustible loading on both sides of the affected fire door was representative of a fire duration less than 1.5 hours. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, associated with the Training component, because the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee believed the performance deficiency was caused by the one of the new temporary contractors brought onto the site to work in support of the D2R25 refueling outage. (H.9)
05000373/FIN-2017008-012017Q3LaSalleFailure to Correctly Evaluate/Justify Post-Accident Operability Qualification for the Reliance Motor 1(2)VY03C RHR Pumps RoomCooling FanThe inspectors identified a finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50.49, Paragraph (f)(4), for the licensees failure to provide adequate analysis in combination with partial type test data to qualify an Environmental Qualification (EQ) component. Specifically, EQ-LS068 failed to provide adequate analysis to justify the Post-Accident Operability Qualification for the Reliance Electric motor utilized for 1(2)VY03C. The EQ Binder incorrectly relied on test values that was strictly performed for thermal aging (for normal plant conditions) to justify a Post-Accident Qualification. The licensee captured the inspectors concern into their Corrective Action Program (CAP) as Action Request (AR) 04030532.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, as an immediate corrective action, the licensee performed a preliminary assessments and concluded that the motors could be EQ qualified for the environmental conditions for which they could be exposed. The finding was associated with a cross-cutting aspect in the area of Human Performance, Design Margin. (H.6)
05000373/FIN-2017008-022017Q3LaSalleFailure to have Adequate Justification for Extending the life of Lubricant used in EQ Motor BearingsThe inspectors identified a finding of very-low safety significance and an associated NCV of 10 CFR Part 50.49, Paragraph (j), Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, for the licensees failure to have adequate justification for extending the service-life for grease used in the bearing for EQ motors installed in harsh environment. Specifically, the licensee extended the bearing grease qualified service life for several EQ motors installed in Zone H4A, H6 and H5A from 31.5, 20.5 and 19.5 years respectively to 60 years based on incorrect assumptions. The justification for 60 years extension incorrectly assumed that the calculated service-life was based on continuous operation of the motor. The licensee captured the inspectors concern into their CAP as AR 04030538. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, as an immediate corrective action, the licensee performed a preliminary evaluation that concluded that the grease remained qualified based on test data which showed that the grease consistency remained within acceptable range during the thermal age test. The finding did not have a cross-cutting aspect associated with it because it was not representative of current performance.
05000266/FIN-2015008-012015Q1Point BeachFailure to Promptly Correct Conditions Adverse to Quality Regarding Electrical Power Cable Sizing and ProtectionThe inspectors identified a finding of very-low safety significance, and an associated Non-Cited Violation of Title 10, Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to implement timely corrective actions to address the longstanding issue of electrical power cables that have not been verified to be sized or protected in accordance with their design bases, as described in PBNPs Final Safety Analysis Report Section 8.0.1. Specifically, the licensee failed to correct known deficiencies regarding: (1) power cables with operating currents in excess of their current-carrying capacities; (2) power cables that are not protected against overload in accordance with the National Electrical Code; and (3) power cables for which their current-carrying capacities are undetermined. Although various corrective action documents have been initiated since these issues first came to light in the 1990 to 1991 time period, the licensee has not taken appropriate actions to correct the conditions adverse to quality to this date. The licensee entered this finding into their Corrective Action Program as Condition Report (CR) 02035020 and CR 02035680, with recommended actions to perform ampacity analysis for applicable cables, verify cables are protected against overload in accordance with the National Electrical Code, verify cable ampacities are higher than their respective load currents, and perform an evaluation to determine why this issue has not been resolved and address the safety culture aspect. The inspectors determined the licensees failure to promptly correct the conditions adverse to quality regarding electrical power cables was a performance deficiency warranting a significance determination. The performance deficiency was determined to be more than minor, and a finding in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Design Control attribute of the Reactor Safety, Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding in accordance with IMC 0609.04, Phase 1, Initial Screening and Characterization of Findings. The finding screened as having very-low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function on the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The inspectors identified a crosscutting aspect associated with this finding in the area of Human Performance, associated with the Design Margin component, because the licensee failed to ensure equipment is operated within design margins, and margins are carefully guarded and changed only through a systematic and rigorous process.
05000461/FIN-2015001-012015Q1ClintonFailure to perform adequate channel calibration on seismic instrumentationThe inspectors identified a Green Finding associated with the licensees failure to perform a channel calibration with appropriate acceptance criteria to determine the functionality of the stations seismic monitoring equipment used for evaluating earthquakes. Specifically, station procedure CPS 9437.21, Triax Time-History Accelerometer Channel Calibration, Revision 39c, did not include steps with clear acceptance criteria to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source as part of the operability requirements. The licensee documented the issue in the corrective action program (CAP) as action request (AR) 02454630. The licensee performed an operability evaluation and determined that the original voltage read by the technicians was outside the band described in the procedure but was within the band described by the vendor and therefore was operable. The licensee also planned to change the surveillance test procedure to clarify the acceptance criteria. The licensees failure to perform a channel calibration with appropriate acceptance criteria to determine the functionality of the stations seismic monitoring equipment used for evaluating earthquakes was a performance deficiency. Specifically, station procedures did not include steps with clear acceptance criteria to ensure that battery backup power was provided to operate the equipment on a loss of the normal power source. The performance deficiency was more than minor because it adversely impacted the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using IMC 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, issued June 19, 2012, the inspectors answered Yes to the Mitigating Systems cornerstone question, Does the finding involve the ...degradation of equipment...specifically designed to mitigate a seismic...initiating event.... Therefore, the inspectors addressed the questions in Exhibit 4, External Event Screening Questions. The inspectors answered No to the two questions in Exhibit 4. Specifically, 1) if completely failed the seismic monitor would not cause an initiating event or degrade multi-trains or risk-significant systems; and 2) the finding does not involve the total loss of any safety function. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias where individuals use decision making-practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee documented the issue of the voltage being high out of specification and instead of performing additional corrective actions to determine if leaving the voltage out of specification was appropriate marked the step as not applicable and proceeded with the rest of the procedure. (H.14)
05000461/FIN-2015001-022015Q1ClintonUnqualified safety-related cables used in a submerged environmentThe inspectors identified a finding and an associated non-cited violation (NCV) of 10 CFR Part 50 Appendix B, Criterion III, Design Control, for the failure to maintain safety-related cables for the shutdown service water (SX) system in an environment for which they were designed. Specifically, the licensee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. The design of the system used monitor and pump down the cable vaults did not indicate that the system had failed, leading the licensee to believe that the environmental conditions in which the cables were stored were acceptable even when they were not. The licensee documented this issue in their CAP as AR 02474543. Corrective actions included draining the cable vaults so that the cables were no longer submerged and planned to repair all affected solar powered pumps and associated alarm systems. The licensees failure to maintain safety-related cables for the SX system in an environment for which they were designed was a performance deficiency. Specifically, the licensee failed to maintain SX safety-related cables in an environment for which they were designed when the cables were allowed to be submerged in water inside cable vaults. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered Yes to the question does the SSC maintain its operability or functionality. Specifically, the SX system submerged cables did not cause the SX system to be inoperable or nonfunctional. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution in the aspect of resolution, where the organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee failed to implement effective corrective actions to address an adverse trend of water in cable vaults which led to SX safety-related cables being submerged in water. (P.3)
05000461/FIN-2015001-032015Q1ClintonFailure of the Division 3 Shutdown Service Water Pump due to an inadequate Bushing DesignA self-revealed finding, preliminarily determined to be of low to moderate safety significance (White) and an associated apparent violation (AV) of 10 CFR Part 50 Appendix B, Criterion III, Design Control was identified for the failure to verify the suitability of the replacement pump design for the Division 3 Shutdown Service Water system. Specifically, the licensee failed to verify the design of the suction bell bushing for the replacement pump would pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water, resulting in the failure of the pump. This finding was self-revealed on September 16, 2014, during a surveillance test to ensure operability of the Division 3 shutdown cooling water pump, after the pump failed to start due to a damaged bushing rendering the pump inoperable. This finding does not represent an immediate safety concern because the licensee replaced the pump in September of 2014 with a pump of similar design in combination with additional monitoring of pump performance and provided adequate documentation that assures the pump will remain operable until a different design can be installed by June of 2016. The inspectors determined that the licensees failure to verify the suitability of the design for the Division 3 SX replacement pump for conditions under which it was to be used, as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency. Specifically, the licensee failed to verify the design of the suction bell bushing for the replacement pump would pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water, resulting in the failure of the pump. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A Significance and Enforcement Review Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross-cutting aspect was identified with this finding. (Section 4OA3)
05000346/FIN-2014007-012015Q1Davis BesseFailure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis AnalysisThe inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control for the licensees failure to install and control the Reactor Coolant Pump (RCP) seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress Analysis: Reactor Coolant Pump 1-1-1 Seal Cavity Vent, into specifications, drawings, procedures, an instructions. The licensee entered this finding into their Corrective Action Program (CAP) to review the lack of controls over the installation of the flexible hoses, but determined that the flexible hoses remained operable. The performance deficiency was determined to be more than minor because, if lef uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits. The finding screened as of very-low safety significance (Green) because the finding could not result in exceeding the Reactor Coolant System (RCS) leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation. The inspectors determined this finding had an associated cross-cutting aspect, Design Margins, in the Human Performance cross-cutting area. Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. (H.6)
05000331/FIN-2014005-012014Q4Duane ArnoldConstruction Code Used during a Replacement Activity Not Reconciled with the Owner's RequirementsA finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Section 50.55a, Codes and Standards, was identified by the inspectors for the failure to reconcile the construction code and owners requirements when replacing rod hangers associated with the high pressure coolant injection (HPCI) system. The licensee subsequently performed a code reconciliation and concluded the applicable construction code requirements were met. The licensee captured this issue in its Corrective Action Program as condition report 01999594. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of HPCI to respond to initiating events to prevent undesirable consequences. Specifically, the failure to reconcile the construction code and owners requirements when replacing HPCI support rod hangers reduced confidence in the systems capability to meet its mitigating function consistent with its design basis. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. This finding had a cross-cutting aspect of procedure adherence in the area of Human Performance because the licensee failed to follow American Society for Mechanical Engineers Section XI, Administrative Manual for Repair, Replacement, and Modification. (H.8)
05000331/FIN-2014005-022014Q4Duane ArnoldLiquid Penetrant Testing Procedures Not Qualified for their Full Applicability RangeA finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors for the failure to properly qualify nondestructive testing procedures in accordance with applicable codes. Specifically, liquid penetrant testing procedures were not qualified for their full applicability temperature ranges in accordance with American Society for Mechanical Engineers (ASME) Code, Section V, Nondestructive Examination. The licensee entered this issue into the Corrective Action Program as condition report 01950601 and 01999596. As an immediate corrective action, the licensee reviewed completed liquid penetrant examination records and did not find an example where the procedures were implemented at the unqualified temperature range. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, since the liquid penetrant testing procedures were not qualified for their full applicability temperature ranges, liquid penetrant examinations were not assured to detect flaws in the unqualified temperature ranges. As a consequence, the potential would exist for a rejectable flaw to go undetected affecting the operability of the affected system. This finding affected the Initiating Event, Mitigating System, and Barrier Integrity cornerstones. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the inadequate qualifications were performed more than three years ago and was not confirmed to reflect current performance.
05000331/FIN-2014005-072014Q4Duane ArnoldIneffective Radiological Engineering Controls Resulted in Unplanned and Unintended Radiological Intakes to WorkersA finding of very-low-safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Section 20.1701 was self-revealed during work activities associated with the failure to implement effective radiological engineering controls during reactor pressure vessel (RPV) disassembly that resulted in personal contaminations and unplanned and unintended radiological intakes to workers. Specifically, on October 5, 2014, several individuals working on the refuel floor were contaminated and several received small intakes of radioactive material while venting the RPV head. The licensee entered the issue into the Corrective Action Program as condition report 01996216. Corrective actions included revising applicable procedures for RPV flood-up with the RPV vented to atmosphere on the refuel floor. The finding was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected th cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to implement effective radiological engineering controls during RPV disassembly resulted in personal contaminations and low dose intakes to several workers. The inspectors also concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The finding was determined to be of very-lowsafety significance because it was not an ALARA planning issue; there was neither overexposure nor a substantial potential for an overexposure; and the licensees ability to assess dose was not compromised. This finding was associated with the crosscutting aspect of operating experience in the area of Problem Identification and Resolution because the licensee did not systematically implement relevant external operating experience in a timely manner. (P.5)
05000331/FIN-2014005-062014Q4Duane ArnoldFailure to Determine Dose Rates Prior to Entry into a High Radiation AreaA finding of very low safety significance and an associated non-cited violation of Technical Specification 5.7.1.e was identified by the inspectors following entry into the fuel pool heat exchanger room which was a high radiation area (HRA). The inspectors determined that the licensee failed to determine the radiological conditions in the HRA in accordance with the Technical Specifications and plant procedures to ensure the workers were accurately briefed on the current conditions prior to entry. As a result, an individual was permitted entry into areas with greater than expected dose rates. This issue was entered into the licensees Corrective Action Program as condition report 02000258. The licensee subsequently performed a follow-up survey of the HRA and coached the individual that performed the brief. The performance deficiency was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, worker entry into HRAs without knowledge of the radiological conditions placed them at increased risk for unnecessary radiation exposure. The finding was determined to be of very low safety significance (Green) because the performance deficiency was not an as-low-as-reasonably-achievable (ALARA) planning issue; there was neither an overexposure nor a substantial potential for an overexposure; and the licensees ability to assess dose was not compromised. The finding was associated with the cross-cutting aspect of challenge the unknown in the area of Human Performance because the licensee failed to challenge the adequacy of the January 19, 2014, radiological survey as the fuel pool cooling heat exchanger room contained equipment that continuously transported radioactive liquid and was subject to changing radiological conditions. (H.11)
05000331/FIN-2014005-052014Q4Duane ArnoldInadequate Containment Isolation Valve Leak Tightness Test ProcedureThe inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish an adequate procedure for an activity affecting quality for a system penetrating the primary containment pressure boundary. Specifically, Surveillance Test Procedure STP 3.6.1.1-09, Containment Isolation Valve Leak Tightness Test Type C Penetrations TIP (traversing in-core probe) Valves, Revision 4, failed to include leak rate testing instructions for all of the fittings inboard of the outboard TIP valves tested, which constituted part of the primary containment pressure boundary. The licensee entered the issue in their Corrective Action Program as condition report 02003580. As part of their corrective actions, the licensee re-performed a local leakage rate test to verify the fittings were leak tight. The inspectors determined that the finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of procedure quality and adversely affected the cornerstone 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. As it related to the finding, procedure STP 3.6.1.1-09 lacked adequate instructions to ensure no leakage of a system penetrating the primary containment pressure boundary. The finding was of very low safety significance (Green) because it did not represent an actual open pathway of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding was associated with the cross-cutting aspect of resources in the area of Human Performance because STP 3.6.1.1-09 did not include testing of the fittings inboard of the outboard TIP valve as required. (H.1)
05000331/FIN-2014005-032014Q4Duane ArnoldFailure to Evaluate Maintenance Activities for PreconditioningA finding of very low safety significance and an associated non-cited violation of Technical Specification 5.4.1.a, Procedures, was identified by the inspectors for the licensees failure to maintain maintenance planning procedures appropriate for the circumstances that could affect performance of safety related equipment. Specifically, procedure MA-AA-203-1001, Work Order Planning, did not ensure that maintenance activities performed on secondary containment components between surveillance testing intervals (2012 and 2014) was properly evaluated for the potential for preconditioning. The licensee entered the inspectors concerns into the Corrective Action Program as condition report 02008529. Corrective actions included the performance of a condition evaluation to evaluate the work that had been done over the previous cycle for preconditioning and an apparent cause evaluation for the work planning procedural gap with respect to preconditioning and its possible impact on work activities. The performance deficiency was determined to be more than minor because the finding impacted the Barrier Integrity cornerstone attribute of procedural quality, and adversely affected the cornerstone 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 and events. Furthermore, the finding wa determined to be more than minor because if left uncorrected, failing to properly and consistently evaluate the potential for unacceptable preconditioning would have the potential to lead to a more significant safety concern. The inspectors applied IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012, to this finding. The inspectors answered No to all questions within Table 3, Significance Determination Process Appendix Router, and transitioned to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, June 19, 2012. The inspectors answered No to questions C.1 and C.2 in Exhibit 3, Barrier Integrity Screening Questions. Therefore, the finding was screened as very low safety significance (Green). This finding was associated with the cross-cutting aspect of work management in the area of Human Performance because the licensees work order planning process was not appropriate for the circumstances to evaluate the impact of maintenance activities on Technical Specification equipment and surveillance tes results. (H.5)
05000341/FIN-2014403-042014Q4FermiSecurity
05000341/FIN-2014403-032014Q4FermiSecurity
05000331/FIN-2014005-042014Q4Duane ArnoldFailure to Accomplish Procedure for Leaking Pipe SnubberA finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, on May 8, 2014, the licensee failed to properly evaluate functionality of a leaking pipe snubber associated with the A core spray subsystem, the resultant operability impact on the Technical Specification affected systems, and the extent of condition. The licensee entered the inspectors concerns into the Corrective Action Program as condition report 02003867 and 02010686. Corrective actions included coaching/training of licensed operators during requalification training and management review committee members, and changes to applicable snubber program procedures. The performance deficiency was determined to be more than minor because it impacted the Mitigating System cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding did not involve the total loss of any safety function, the finding screened as very low safety significance (Green). This finding was associated with the cross-cutting aspect of consistent process in the area of Human Performance, because the licensees inconsistent application of the systematic operability/functionality determination process to evaluate the leaking snubber led to prolonged exposure of the extent of cause that affected several safety-related systems. (H.13)
05000341/FIN-2014403-022014Q4FermiSecurity
05000461/FIN-2014005-022014Q4ClintonPotential Failure to Maintain Safety Related Cables in a Qualified EnvironmentThe inspectors identified an unresolved item (URI) associated with maintaining safety related cables in a qualified environment. Specifically, cable vaults that contain safety related SX cables were found to be full of water, when opened to perform cable vault inspections. DOn October 14, 2014, the licensee commenced their periodic inspection of safety related cable vaults. These cable vaults house cables related to all 3 divisions of SX and the division 3 emergency diesel generator. When the licensee opened the cable vaults, they found that 6 out of the 10 vaults contained water at a level that would cause the cables to be in a submerged condition. In response to a violation from the 2007 NRC component design basis inspection, th licensee installed solar powered sump pumps in all of the cable vaults on site. The system included a float mechanism that would provide local indication when water was accumulating in the vaults; a yellow light would indicate water had reached a level requiring pump down of the vault. A red light would indicate water had reached the cables, and therefore immediate actions should be taken to pump down the cable vaults. 12 The licensee discovered the lights were not lit when the vaults were found to be full of water. After further investigation, the licensee concluded that in some cases the float switches that trigger the lights were not functioning properly and in other cases the light bulbs were not functional. The only maintenance performed on the sump pump stations was to replace light bulbs periodically The amount of time the cables were submerged is unknown. The licensees cable monitoring program procedure ERAA300150, Cable Condition Monitoring Program, states that cables should not be submerged for greater than 2 months. The inspectors questioned whether there was an evaluation for establishing a 2-month limit. Due to the probability the cables were submerged for greater than 2 months, the inspectors questioned whether the cables were still operable and what the technical justification was for operability. The licensee indicated the cables were currently operable because the SX pumps are tested periodically and have run without any issues. The inspectors concluded the licensees failure to maintain the safety related cable vaults in a dry condition is a performance deficiency. In order to determine whether the performance deficiency is more than minor the inspectors need to review additiona information, such as the procurement documents, cable testing results, submergence qualification records and the evaluations performed in response to this issue. The URI is opened pending the inspectors review of licensee document (URI 05000461/201400502, Potential Failure to Maintain Safety Related Cables in a Qualified Environment)
05000331/FIN-2014005-082014Q4Duane ArnoldFailure to Evaluate Several Safety Related Relays Installed Beyond Their Design LifeA finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, when new degraded or non-conforming conditions adverse to quality were identified. Specifically, the licensee failed to evaluate operability and the acceptability for continued operation when an extent of condition review identified several safety-related time delay relays installed beyond the vendor recommended design life. The licensee documented the inspectors concerns in condition report 02015742. The affected relays were immediately declared operable but non-conforming, and a prompt operability determination and apparent cause evaluation to determine corrective actions were in progress at the end of the inspection period. The performance deficiency was determined to be more than minor because, if left uncorrected, failing to properly assess the operability of degraded or non-conformin conditions and evaluating the necessity for compensatory measures would have the potential to lead to a more significant safety concern. Because the finding was a qualification deficiency confirmed not to result in loss of operability, the finding screened as very low safety significance (Green). This finding was associated with the crosscutting aspect of identification in the area of Problem Identification & Resolution because the licensee did not identify or capture the extent of the relay aging condition within the corrective action program to ensure that new conditions adverse to quality were properly screened for significance and potential operability impacts. (P.1)
05000331/FIN-2014005-092014Q4Duane ArnoldFailure to Test of Evaluate the Seismic Critical Characteristic for a Commercial Grade Circuit BreakerThe inspectors identified a finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, "Design Control," for the licensee's failure to verify commercial grade circuit breakers were suitable for use in safety-related applications. Specifically, the licensee failed to verify, either through seismic testing or justification, that the circui breakers being dedicated on purchase order 02309726 would be able to perform their intended safety function during a seismic event. The licensee entered this finding into their Corrective Action Program as condition report 01986727 and 01987616. An extent of condition review was performed and concluded that these circuit breakers were not yet installed at Duane Arnold and a seismic test would be performed on these types of breakers prior to installation. The performance deficiency was determined to be more than minor because, if lef uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Because the finding did not represent an actual loss of function (circuit breakers were not currently installed), the finding screened as very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of licensee's current performance.
05000341/FIN-2014403-012014Q4FermiSecurity
05000331/FIN-2014005-102014Q4Duane ArnoldLicensee-Identified Violation10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established for the selection and review for suitability of application of materials and parts that are essential to the safety-related functions of structures, systems, and components. Contrary to the above, in October 2012, the licensee failed to properly select an review the suitability of application of several electrical cable splices and terminal strips during the replacement of safety-related electrical cables associated with the A and B SBDGs. Specifically, two modification packages associated the cable replacements during the 2012 RFO 23 did not appropriately evaluate the impacts of the stations environmental qualification (EQ) program and the effects of an internal turbine building flood. Because the SSCs maintained operability based on the deficiency affecting the qualification of the SSCs, the finding screened as very low safety significance (Green). This issue was documented in the licensees CAP as CR 01979556. Immediate corrective actions included a determination of operability (the SBDGs had no specified safety function for the EQ and turbine building flood events in question per the UFSAR), installation of a temporary flood barrier to compensat for the non-conforming condition for the A SBDG, cable splice and terminal strip replacement for the B SBDG, and the performance of a root cause evaluation.
05000263/FIN-2014403-032014Q3MonticelloLicensee-Identified Violation
05000341/FIN-2014007-012014Q3FermiIncorrect Valve Location in ProcedureThe inspectors identified a finding of very low safety significance (Green) and associated NCV of Technical Specifications (TS) Section 5.4.1.a for the licensees failure to maintain Procedure 20.000.23, High RPV (Reactor Pressure Vessel) Water Level to address an RPV overfill event. Specifically, the licensee provided an incorrect location of a manual valve in the Standby Feedwater (SBFW) system. The procedure described the valve as being located in the turbine building basement, while the valve was actually located in a locked high radiation area in the north heater room. The licensee revised the procedure to include the correct location of the valve. The inspectors determined that the issue was more than minor because a reactor overfill event could impair the RCIC and HPCI systems during a fire in fire zone RW. The finding affected the Mitigating Systems cornerstone. The finding was determined to be of very low safety significance based on a detailed risk-evaluation. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not take effective corrective actions to address a potential reactor pressure vessel overfill event.
05000263/FIN-2014403-022014Q3MonticelloSecurity
05000263/FIN-2014403-012014Q3MonticelloSecurity
05000282/FIN-2014008-022014Q2Prairie IslandLicensee-Identified ViolationThe licensee identified a Severity Level IV violation of 10 CFR 50.59, (Changes, Tests, and Experiments, for the failure to demonstrate in a written evaluation that prior NRC-approval was not required for changes made to an accident analysis. Specifically, the licensee incorrectly concluded in written Evaluation 1102, Waste Gas Tank Rupture Dose Analysis, Revision 0 that higher activity levels and dose rates at the Exclusion Area Boundary and Low Population Zone associated with extended plant life due to license extension did not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR. The performance deficiency was determined to be more than minor because it was associated with the Radiation Safety cornerstone attribute of program and process and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors determined the violation was of Severity Level IV because the associated finding was of very low safety significance (Green) as there was no actual radioactive material release. The licensee entered this issue into their Corrective Action Program as AR 1417573 and AR 1427150 and intended to submit a license amendment request for review by the NRC.
05000282/FIN-2014008-012014Q2Prairie IslandNo Compensatory Measure were established for Lack of Fuses Coordination associated with Safe Shutdown Power SuppliesThe inspectors identified a finding of very low safety significance and associated NCV of the Prairie Island Nuclear Generating Plant Facility Operating License Condition 2.C.(4) for the licensees failure to implement the requirements as specified in the Fire Protection Program (FPP) for impaired safe shutdown equipment. Specifically, the licensee failed to establish appropriate compensatory measures when they identified lack of coordination between DC panel fuses and upstream panels supply fuse under fault conditions for several safe shutdown power supplies. The licensee replaced all miss-coordinated fuses and entered the issue into their Corrective Action Program. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to fire events prevent undesirable consequences (i.e., core damage). Specifically, the failure to establish compensatory measures for lack of fuse coordination degraded the defense and depth element of the Fire Protection Program. The finding represented a low degradation and therefore the inspectors determined that the finding screened as having very low safety significance (Green) in Task 1.3.1 of IMC 0609, Appendix F. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence for the licensees failure to follow instructions as specified in Procedure FP-E-CAL-01 Calculations.
05000315/FIN-2014002-012014Q1CookDegraded Latch Prevents Closure of Fire DoorThe inspectors identified a finding of very low safety significance (Green) and associated non-citied violation of License Condition 2.C.4 for Unit 1, for the licensees failure to ensure that a fire door would be closed at the time of a fire. Specifically, fire door 1-DR-AUX387 was found with a degraded latch that prevented the door from closing. Donald C. Cook is required to comply with the National Fire Protection Association (NFPA) 80, 1970 which requires a closing device to ensure fire doors close and latch at the time of a fire. Contrary to this requirement, fire door 1-DR-AUX-387 would not close and latch because the latching mechanism for the inactive leaf had failed in a manner preventing the door from closing. As immediate corrective action, the licensee started hourly fire watches on the door and performed an interim repair to restore the door to a functional status. The licensee has entered the condition into the corrective action program as AR 2014-0802. The inspectors determined the finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events (Fire) and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to maintain door 387 such that it could perform its required function as a 3 hour fire barrier. Using IMC 0609, Appendix F, the inspectors concluded that the finding was of very low safety significance (Green) because the fire loading was below the screening criteria of 120,000 btu/ft2. The inspectors concluded the finding included a cross-cutting aspect of H.5, Work Planning, in the area of human performance because the licensee did incorporate risk insights.
05000315/FIN-2014002-022014Q1CookFailure to Establish Procedures for Vacuum FillA finding and associated non-cited violation of technical specification (TS) 5.4.1, Procedures, self-revealed pertaining to establishing and maintaining procedures to ensure reliable indication of reactor vessel level during reduced RCS inventory and vacuum fill operations. Specifically, the licensee failed to include in procedures for vacuum fill methods to ensure the level detection system sensing lines were vacuum tight and to include provisions to normalize level indications. During the vacuum fill evolution for Unit 1, the licensee made 5 attempts to draw vacuum because of diverging level indications. The additional time spent in reduced inventory as well as the additional drain downs resulted in increased plant risk. As immediate corrective actions, the licensee corrected the leaking fitting, normalized level readings, and completed the vacuum fill evolution. The licensee has entered this issue into the corrective action program (CAP) as action request (AR) 2013-6907. The inspectors concluded the finding was more than minor because it adversely affected the Initiating Event cornerstone objective of limiting the likelihood of events that upset plant stability while shutdown. Specifically, the issue impacted the Procedure Quality attribute. Based on the screening criteria of IMC 0609, the inspectors and regional SRA concluded a phase 2 or 3 evaluation was needed. The Office of Nuclear Reactor Regulatory (NRR) performed a phase 3 assessment and estimated the conditional core damage probability at 5.9E-7. Therefore, the finding is of very low safety significance (Green). The finding included a cross-cutting aspect of H.9, Training, in the human performance area because the licensee lacked understanding of the precision level instruments.
05000316/FIN-2014002-032014Q1CookFailure to Establish Procedures for Vacuum FillA finding and associated non-citied violation of TS 5.4.1, Procedures, self-revealed pertaining to establishing and maintaining procedures to ensure reliable indication of reactor vessel level during reduced RCS inventory and vacuum fill operations. Specifically, the licensee failed to include in procedures for vacuum fill methods to ensure the level detection system sensing lines were vacuum tight. Although the licensee implemented some corrective actions prior to the scheduled vacuum fill evolution, the actions taken failed to prevent recurrence. During the vacuum fill evolution for Unit 2, the licensee made 2 attempts to draw vacuum because of diverging level indications. The additional time spent in reduced inventory as well as the additional drain down resulted in increased plant risk. As immediate corrective actions, the licensee corrected the leaking fitting, normalized level readings, and completed the vacuum fill evolution. The licensee has entered this issue into the CAP as AR 2013-18146. The inspectors concluded the finding was more than minor because it adversely affected the Initiating Event cornerstone objective of limiting the likelihood of events that upset plant stability while shutdown. Specifically, the issue impacted the Procedure Quality attribute. Based on the screening criteria of IMC 0609, the inspectors and regional SRA concluded a phase 2 or 3 evaluation was needed. Since the issue in Unit 2 was bounded by the phase 3 assessment performed for Unit 1, the inspectors and SRA concluded the finding was of very low safety significance, (Green). The finding included a cross-cutting aspect of P.3, Resolution, in the corrective action area because the licensee failed to implement corrective actions that prevented recurrence.
05000440/FIN-2013007-022013Q4PerryFailure To Promptly Correct a Non-conservative Technical SpecificationThe inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly correct a non-conservative Technical Specification. Specifically, the inspectors identified on November 14, 2013, that the licensee failed to promptly correct the non-conservative Technical Specification 3.4.11 by not submitting a license amendment request in accordance with NRC Administrative Letter 98-10, which required submittal within 1 year or 1 operating cycle. The licensee had determined Technical Specification 3.4.11, RCS Pressure and Temperature (P/T) Limits, to be non-conservative on October 16, 2009, and implemented administrative controls as allowed by the Administrative Letter. As of November 14, 2013, the licensee had not submitted the license amendment request, over 4 years and 2 operating cycles after determining the Technical Specification was non-conservative. The licensee entered the finding into the corrective action program as Condition Report CR 2013-18983. The performance deficiency was determined to be more than minor because the finding was associated with the area of Routine Operations Procedures within the Procedure Quality attribute of the Barrier Integrity Cornerstone and had the potential to adversely affect the associated cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding was screened as very low safety significance because it was determined that operators followed the appropriate reactor coolant system P/T curves even though the Technical Specification was non-conservative. The finding has a cross-cutting aspect in the area of human performance, decision-making, where licensee decisions demonstrate that nuclear safety is an overriding priority. Specifically, from the time of discovery of the non-conservative technical specification until now, various decisions had been made by the licensee that have delayed the timely submittal of the license amendment request.
05000440/FIN-2013007-012013Q4PerryFailure To Comply With Technical Specification 3.4.11The inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of Technical Specification 3.4.11, RCS Pressure and Temperature (P/T) Limits, for failure to comply with reactor pressure vessel pressure/temperature limits. Specifically, in 2011 the inspectors identified the pressure/temperature limits in Technical Specification 3.4.11 only contained values for reactor pressure vessel pressures greater than 0 pounds per square inch gauge. However, between June 2011 and July 2013, the licensee operated the plant with a vacuum in the reactor pressure vessel during 5 cold startups and 1 cooldown. The licensee entered the finding into its corrective action program as Condition Report CR 2013-18689. The performance deficiency was determined to be more than minor because the finding was associated with the area of Routine Operations Performance within the Human Performance attribute of the Barrier Integrity Cornerstone and had the potential to adversely affect the associated cornerstone objective of providing reasonable assurance that a physical design barrier (reactor coolant system) protects the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance because it was determined that there was no change in risk due to the performance deficiency. This finding has a cross-cutting aspect in the area of human performance, resources. Specifically, complete, accurate, and up-to-date procedures were not available to operators to ensure operations within the requirements of Technical Specification 3.4.11.
05000443/FIN-2013005-012013Q4SeabrookLicensee-Identified Violation10 CFR Part 50.65, paragraph a(4), Requirement for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, states, in part, that the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. NextEra procedure WM 10.1. On-Line Maintenance, Section 3.3.1, requires that an evaluation of the risk impact of planned maintenance tasks be performed. Contrary to the above, on September 24, 2012, NextEra failed to adequately assess and manage the impact to plant risk during a planned maintenance activity. Specifically, NextEra identified during internal reviews that they had failed to recognize an elevated online maintenance risk level (Yellow) during the performance of the 1-EDE-B-1-B Battery Service Test due to incorrect coding in NextEras PRAX risk model program. The inspectors determined NextEras failure to assess and manage risk during the period when the Battery Service Test was reasonably within NextEras ability to foresee and correct, and was identified as a performance deficiency. This performance deficiency is more than minor, and considered a finding, because it is associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because this finding represents a violation of 10 CFR Part 50.65 Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Section a(4), the inspectors used IMC 0609, Appendix K, Flowchart 1 Assessment of Risk Deficit, to analyze the finding. The regional Senior Reactor Analyst determined the incremental core damage probability (ICDP) for the surveillance period (~5-10 minutes) to be several orders of magnitude below the 1E-6 threshold due to the short duration of the systems unavailability. As this finding is not related to Risk Management Actions only, and the ICDP Risk Deficit is not >1E-6, the inspectors determined that the finding is of very low safety significance (Green). The issue was entered into NextEras CAP as AR 1906782.
05000315/FIN-2013005-012013Q4CookLicensee-Identified ViolationSub-paragraph 50.36a (a)(2) of 10 CFR Part 50, requires the licensee to submit a report to the Commission annually that specifies the quantity of each of the principal radionuclides released to unrestricted areas in liquid and in gaseous effluent during the previous 12 months, including any other information as may be required by the Commission to estimate maximum potential annual radiation doses to the public resulting from effluent releases. The NRC Safety Guide 23 Onsite Meteorological Programs states that knowledge of meteorological conditions in the vicinity of the reactor is important in providing a basis for estimating maximum potential annual radiation doses resulting from radioactive materials released in gaseous effluents. The Safety Guide also described a suitable onsite meteorological program to provide meteorological data needed to estimate potential radiation doses to the public as a result of the routine or accidental release of radioactive material to the atmosphere and to assess other environmental effect. The Safety Guide states that meteorological instruments should be inspected and serviced at a frequency which will assure at least a 90 percent data recovery and which will minimize extended periods of instrument outage. Contrary to sub-paragraph 50.36a (a)(2) of 10 CFR Part 50, the licensee failed to submit information required by the Commission to estimate maximum potential annual radiation doses to the public resulting from effluent releases. Specifically, the licensee was not able to maintain Meteorological Tower instrumentation so that data recovery remained above 90 percent for the calendar year 2012, information that the Commission required to estimate maximum potential doses to the public. After identifying the error, the licensee took corrective action to prevent further loss of meteorological data due to equipment failure by troubleshooting and repairing Meteorological Tower instrumentation and instituting additional data recovery efforts in 2013. The licenses corrective actions were entered into the CAP as Condition Reports AR 2013-12764 and AR 2013-15116. Because the licensee identified the failure to properly recover data, the inspectors determined that the violation met the requirements of a licensee identified NCV.
05000440/FIN-2013008-042013Q3PerryInsufficient Controls to Prevent Common Mode Flooding of ECCS RoomsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to control drainage of the emergency core cooling system room sumps in a manner that prevents common mode flooding of these rooms. Specifically, procedures did not ensure appropriate controls to prevent backflow from the floor drain system. The licensee entered the issue into their Corrective Action Program and revised procedures to prevent opening more than one emergency core cooling system room sump isolation valve at the same time. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of the emergency core cooling system to respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because it did not result in either the loss of operability or an actual loss or degradation of a function designed to mitigate flooding. Specifically, a review of recent plant history did not find an instance where the configuration of the floor drain system allowed common mode flooding of the emergency core cooling system rooms when operability of this system was required. The inspectors determined that this finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee did not conduct a self-assessment of sufficient depth. Specifically, the licensee evaluated a flooding incident during a self-assessment conducted in 2013 and failed to thoroughly evaluate the cause that resulted in common mode flooding of the rooms.
05000440/FIN-2013008-022013Q3Perry10 CFR 50.59 Evaluation Did Not Consider the Freeze Seal Effect to the RCPBThe inspectors identified a finding of very low safety significance and associated Severity Level IV Non-Cited Violation of Title 10 Code of Federal Regulations (CFR) 50.59, Changes, Test, and Experiments, for the failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis for not applying for a license amendment associated with the use of a freeze seal in the reactor coolant pressure boundary when its integrity was required to protect irradiated fuel. The finding was entered into the licensees Corrective Action Program with recommended actions to, in part, revise the associated 10 CFR 50.59 documents. The inspectors determined that the violation was more than minor because they could not reasonably determine the changes would not have ultimately required NRC prior approval. The finding affected the Initiating Events cornerstone attribute of equipment performance and 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. The inspectors determined that the underlying technical issue was of very low safety significance (Green) using a Phase II evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency.
05000440/FIN-2013008-032013Q3PerryLack of Alternate Methods of Decay Heat RemovalOn May 21, 2004, the A ESW pump became inoperable due to a failure of the uppermost shaft coupling. Technical Specification Limiting Condition for Operation (LCO) 3.7.1, ESW System Divisions 1 and 2, required the licensee to restore operability within 72 hours. Because this action could not be met, TS required the licensee to be in Mode 3 within 12 hours and Mode 4 within 36 hours. While performing plant shutdown, LCO 3.4.10, Residual Heat Removal (RHR) Shutdown Cooling System Cold Shutdown, became applicable. It required, in part, two shutdown cooling subsystems operable in MODE 4 when heat losses to the ambient were not sufficient to maintain average reactor coolant temperature below 200oF. Because ESW is the heat sink of shutdown cooling, the A train of shutdown cooling was also inoperable. With one or two shutdown cooling subsystems inoperable, TS 3.4.10, Required Action A.1, required the licensee to verify an alternate method of decay heat removal was available for each inoperable shutdown cooling subsystem within one hour and once per 24 hours thereafter. The associated TS Basis described the alternate method as one that re-establishes backup decay heat removal capabilities similar to the requirements of the LCO. However, the licensee was unable to identify an alternate method of decay heat removal to satisfy TS 3.4.10, Required Action A.1. Moreover, during repairs on the ESW A pump, the licensee concluded that sufficient doubt existed regarding the ESW B pump; thus, they declared the pump inoperable. Consequently, the B train of shutdown cooling also became inoperable requiring two alternate methods of decay heat removal available. This incident resulted in an NCV which was documented in IR 05000440/2004011 and Licensee Event Report (LER) 05000440/2004-001. On October 19, 2009, the B ESW pump tripped off due to failure of the motor power supply cable. Again, the licensee was required to perform a plant shutdown by TS 3.7.1, declared the B shutdown cooling train inoperable when TS 3.4.10 became applicable, and was unable to verify an alternate method of decay heat removal within one hour to satisfy TS 3.4.10, Required Action A.1. This incident was captured in the CAP as CR 2009-66216 and resulted in LER 05000440/2009003. Following these two incidents, the licensee installed the Alternate Decay Heat Removal (ADHR) system. During this inspection period, the inspectors reviewed the associated 10 CFR 50.59 evaluation (i.e., Evaluation 05-04712, Installation of ADHR System ) which stated The intent of the ADHR system is to assure TS compliance in MODE 4 by providing an additional alternate decay heat removal option that does not rely upon RHR or ESW. However, the inspectors noted its design was limited to a heat removal rate which bounds the approximate decay heat production rate of the core 24 hours after a scram from sustained 100 percent power. During normal shutdown conditions, the licensee transitions from 100 percent power to MODE 4 in a few hours. For instance, this transition occurred in about five hours during refueling outage 1R13. In addition, the licensee revised procedure ONI-E12-2, Loss of Decay Heat Removal, by adding Attachment 11, Cold Shutdown Decay Heat Removal by Steaming. This attachment contained instructions to establish an alternate method of decay heat removal independent of ESW. However, the attachment included a note stating, It will be necessary to validate the effectiveness of this attachment to maintain or reduce RPV temperature (by Engineering calculation or demonstration) if qualifying this as an alternate decay heat removal method per TS 3.4.9 and 3.4.10. As a result, the inspectors questioned the effectiveness of this approach given it had not been verified. The licensee consequently, performed a calculation that determined Attachment 11 was limited to a heat removal rate which bounds the approximate decay heat production rate of the core three days after a shutdown from sustained 100 percent power. The procedure contained other alternatives but these either relied on ESW or lacked enough capacity to serve as backup methods during periods of high decay heat loads. Based on this information, the inspectors were concerned the plant lacked two alternate methods of decay heat removal that have been verified to be effective should a loss of shutdown cooling result from ESW inoperability while in MODE 4 with high decay heat load. The inspectors were particularly concerned because this condition had occurred in the past at least twice. The licensee captured the inspectors concerns in their CAP as CR 2013-11480. This issue is unresolved pending further review and determination of NRC actions to resolve the issue.
05000440/FIN-2013008-012013Q3Perry10 CFR 50.59 Evaluation Did Not Consider the Freeze Seal Effect to the RCPBThe inspectors identified a finding of very low safety significance and associated Severity Level IV Non-Cited Violation of Title 10 Code of Federal Regulations (CFR) 50.59, Changes, Test, and Experiments, for the failure to perform a written evaluation, which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis for not applying for a license amendment associated with the use of a freeze seal in the reactor coolant pressure boundary when its integrity was required to protect irradiated fuel. The finding was entered into the licensees Corrective Action Program with recommended actions to, in part, revise the associated 10 CFR 50.59 documents. The inspectors determined that the violation was more than minor because they could not reasonably determine the changes would not have ultimately required NRC prior approval. The finding affected the Initiating Events cornerstone attribute of equipment performance and 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. The inspectors determined that the underlying technical issue was of very low safety significance (Green) using a Phase II evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency.
05000315/FIN-2013004-042013Q3CookReactor Trip Due to Improper Control Valve SetpointA self-revealed finding of very low safety significance occurred because the licensee failed to program the automatic controller for the condensate heater condensate bypass control valve, 2-CRV-224, with the correct setpoint. Specifically, the automatic controller (2-RU-2) setpoint was not set at the required 240 psig because licensee personnel incorrectly interpreted information in SD-ENG-05400, System Description Condensate System. Consequently, an incorrect set point of 188 psig was incorporated in procedure 2-OHP-4021-001-006, Power Escalation, which was used to program the automatic controller. As a result, 2-CRV-224 did not open as designed to mitigate the lowering main feedwater pump suction pressure, which resulted in the west main feedwater pump tripping on low suction pressure and a subsequent manual reactor trip. For corrective actions, the licensee programmed the correct setpoint into the automatic controller; revised the associated procedures to ensure setpoint changes are accurately incorporated and reviewed prior to implementation; developed plans to communicate lessons learned to the site; and entered the condition into the CAP. Using IMC 0612 the inspectors determined that this issue was more than minor because it was associated with the design control attribute of the Initiating Events Cornerstone and it 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, failure to set the 2-CRV-224 automatic controller to the design setpoint of 240 psig resulted in the subsequent loss of the west main feedwater pump during a feedwater heater level transient, which caused steam generator water levels to lower and required the operators to manually trip the reactor. The inspectors determined the finding was of very low safety significance (Green) using Exhibit 1 of IMC 0609, Appendix A, because the finding did not cause both a reactor trip and a loss of mitigating equipment. The inspectors concluded that this finding was associated with an aspect in the Resources component of the Human Performance cross-cutting area. Specifically, the procedure used to program the automatic controller for 2-CRV-224 was not accurate in that it did not contain the correct design setpoint.
05000443/FIN-2013004-012013Q3SeabrookInadequate Operability Determination Regarding Service Water Leakage and Associated TS ViolationThe inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and an associated violation of technical specification (TS) 3.7.4, because NextEra did not follow the requirements of station procedure EN-AA-203-1001, Operability Determinations/ Functionality Assessments. Specifically, NextEra did not properly evaluate and document an adequate basis for operability, when relevant information was available that would have challenged the reasonable expectation of operability threshold for a service water (SW) through-wall leak that degraded incrementally from weepage on August 7, 2013, to a significantly larger leak on August 28, 2013. NextEra completed a temporary non-code repair of the flaw with the installation of a weldolet on September 1, 2013, following NRC review and approval of a relief request. Additionally, under the corrective action process, NextEra completed apparent cause evaluations for the piping flaw, as well as engineering decision-making during the non-destructive examinations and evaluations, and are currently evaluating the fundamental issue of decision-making regarding TS operability and TS compliance. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the prompt operability determination incorrectly concluded the B cooling tower (CT) SW header and the B SW (ocean) pumps were operable, but degraded, versus inoperable. IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and Exhibit 4, External Events Screening Questions, were used to assess this issue and a detailed risk evaluation was completed. The inspectors assumed that functionality of the SW system, based upon the as-found wall thinning, would only be challenged when aligned to the cooling tower basin when the SW piping is subjected to a higher overall sytem pressure. This system configuration is used to mitigate a seismic event following the loss of the normal SW intake structure. Based on low probability of SW piping system failure due to a seismic event and the overall low likelihood of a seismic event of a magnitude sufficient to cause structure, system, and component (SSC) damage, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the decision making component because NextEra failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate it is unsafe in order to disapprove the action. Specifically, NextEra personnel had not considered relevant information in the form of UT data and actual leak propagation to conclude that they no longer had reasonable assurance of operability and did not declare the B header of ocean and CT SW systems inoperable.
05000315/FIN-2013004-032013Q3CookImproper Setting in Digital Control SystemA self-revealed finding of very low safety significance (Green) occurred because the licensee failed to adjust a key parameter, (KWINIT), in the turbine digital control system after replacing and calibrating the turbine control system linear variable differential transformers. Vendor documents for the generator recommended an initial load of 2 to 5 percent of full load when the turbine generator is synchronized to the grid. For Cook Unit 1, this equates to 22 to 54 megawatts. However, the licensee did not adjust the KWINIT parameter, which is used to determine control valve position, after the turbine control system linear variable differential transformers were replaced and subsequently calibrated using a tighter tolerance than previously used. Consequently, when the turbine generator was synchronized to the grid the turbine control valves opened more than on previous synchronizations, which resulted in picking up excessive load. As a result, reactor cooling system (RCS) temperature momentarily lowered below the minimum temperature for criticality. As an immediate corrective action, the licensee stabilized the plant by taking manual control of the turbine generator. The licensee has entered the condition into the corrective action program (CAP) as AR 2013-7472. Using IMC 0612 the inspectors concluded that this issue was more than minor because it is associated with the equipment performance attribute in the Initiating Events Cornerstone and it adversely impacted the cornerstone objective of limiting the likelihood of events that upset plant stability. Using IMC 0609, Appendix A, Exhibit 1, the inspectors concluded the finding was of very low safety significance (Green) because it did not cause both a reactor trip and a loss of mitigating equipment. The inspectors concluded the finding had an aspect in the Work Control component of the Human Performance cross-cutting area because the licensee did not coordinate work activities to address the impact of changes to work activities on plant performance.
05000315/FIN-2013004-022013Q3CookFaulted 4KV Qualified Off-site CircuitA finding of very low safety significance was self-revealed on April 24, 2013, because the licensee failed to comply with requirements contained in procedure PMI-7030, Corrective Action Program, prior to restoring power to the Unit 1 reserve auxiliary transformer CD-101. Specifically, following multiple trips of supply breaker 12 CD, the licensee failed to correct an issue, defined as a condition adverse to quality in their corrective action program, prior to restoring power to the transformer on April 21. This ultimately led to the supply breaker to the Unit 1 and 2 reserve auxiliary transformers opening due to a faulted cable. No violations of NRC requirements were identified for this issue since the degraded cable was on a non-safety related portion of the electrical system. The licensee entered the issue into the corrective action program as AR 2013-6194. The corrective actions for this issue included replacing the faulted cables and testing the unaffected cables. Using IMC 0612, the inspectors concluded that the issue was more than minor because it was associated with the equipment performance attribute of the Mitigating System Cornerstone and it adversely impacted the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded insulation failed causing a loss of the qualified circuit; a condition which lessened the likelihood of its availability for some events. Using IMC 0609, Appendix A, Section 6, a detailed risk evaluation, assuming inoperability of four days, determined the delta Core Damage Frequency was less than 1E-6; therefore the finding screens as very low safety significance (Green). The inspectors concluded this finding was associated with an aspect in Operating Experience component of the Problem Identification and Resolution cross-cutting area because the licensee did not implement and institutionalize operating experience information from the Electric Power Research Institute (EPRI) and Institute of Electrical and Electronics Engineers (IEEE) into processes and procedures.
05000315/FIN-2013004-052013Q3CookUnreliable Level IndicationsThe inspectors identified an Unresolved Item (URI) regarding the requirements pertaining to establishing and maintaining procedures to ensure reliable indication of reactor vessel level during reduced RCS inventory and vacuum fill operations. During the recent refueling outages, the licensee experienced numerous issues monitoring the RCS water level during the vacuum fill activities. Each of the issues described below resulted in additional entries into RCS mid-loop operations with an associated increase in plant risk during each entry. Specific issues include: During Unit 2 refueling outage 20, the licensee made three vacuum fill attempts. The licensee terminated the first attempt due to a failed camera. The licensee terminated the second attempt due to level variations on NLI-122. The third attempt succeeded, During the Unit 1 refueling outage 25, the licensee made five attempts to vacuum fill the RCS before succeeding: Attempt 1: A divergence between RCS level indications caused the licensee to terminate the evolution; Attempt 2: Licensee terminated the vacuum fill attempt due to similar level indication issues observed in Attempt 1; Attempt 3: Licensee checked level instrumentation fittings for tightness and then made a drain down attempt; instrument deltas again resulted in terminating the attempt; The licensee found a leak in a level instrument transducer box; the licensee replaced the transducer and tightened the fittings; the licensee terminated the attempt due to deviations between level indications; The licensee performed a zero adjust on the level indications instruments; the attempt succeeded. The licensee performed a root cause analysis on the vacuum fill attempts described above and determined the root cause to be that the site did not have ownership of the level indicators. Based on review of the root cause analysis and observation of the licensees performance during the vacuum fill evolution, the inspectors concluded that the licensee identified the root cause. The inspectors also noted that the root cause analysis included several examples where procedures were not appropriate to the circumstances. For example, procedures did not: provide direction for a vacuum leak test; provide detailed direction calibration/protection of transducers from shock, or; include requirements to normalize readings between level indications. The inspectors noted that including the above items into procedures would have reduced the time the licensee spent in a mid-loop condition as well as the number of drain downs needed to complete the RCS vacuum fills. At the conclusion of this inspection, the inspectors had not completed their review to determine if a performance deficiency or violations of regulatory requirements occurred. The URI will remain open pending this review. (URI 05000315/316/2013004-05, Unreliable Level Indications)
05000255/FIN-2013003-022013Q2PalisadesFailure to Follow Corrective Action Process for Service Water LeaksA finding of very low safety significance with an associated non-citied violation of 10 CFR 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the failure to adhere to the requirements of the sites corrective action process. Specifically, the station failed to complete corrective actions to address cavitation-induced erosion of service water system components, which resulted in additional through-wall leaks and other adverse conditions in that safety-related system. Since 1993, this phenomenon caused several through wall leaks and the failure of a valve, which isolated normal service water flow to a component cooling water heat exchanger. Corrective actions to replace valves susceptible to this type of erosion were not implemented, and actions to utilize more effective non-destructive examination (NDE) techniques to assess piping or development of pre-emptive repair/replacement strategies were not performed, resulting in further leaks from the service water system. The current corrective action process procedure, EN-LI-102, states that corrective actions are determined, implemented, and adequate to resolve conditions. The licensee entered the issue in their corrective action program (CAP) as CR- PLP-2013-05813. The issue was determined to be greater than minor in accordance with IMC 0609 Appendix B, Issue Screening, issue date September 7, 2012, because it adversely affected the equipment performance attribute of the mitigating systems cornerstone whose objective is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a through wall leak can challenge the integrity of the piping and system function. The inspectors concluded the finding was of very low safety significance (Green) utilizing IMC 0609, Significance Determination Process, issue date June 2, 2011. Specifically, in Attachment 4, issue date June 19, 2012, utilizing Exhibit 2 of Appendix A, all questions in Section A were answered no since the leaks did not result in a loss of safety function. The finding had an associated cross-cutting aspect in the area of problem identification and resolution for the operating experience component. Specifically, the licensee did not implement and institutionalize operating experience through changes to station processes and procedures.
05000255/FIN-2013003-012013Q2PalisadesInadequate Control of Welding at the F-East Nozzle Reinforcement PlateThe inspectors identified a finding of very low safety significance and an associated non-citied violation of 10 CFR 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to perform adequate pre-weld cleaning and control the welding process in a manner that ensured proper weld fusion of the F-East nozzle reinforcement plate weld joint within the safety injection refueling water storage tank (SIRWT). Consequently, this weld failed in service causing leakage from the SIRWT. The licensee subsequently replaced the floor of the SIRWT and included instructions in the floor replacement work order that required pre-weld cleaning with acetone or other approved solvents. The licensee entered the issue in their corrective action program (CAP) as CR- PLP-2013-03185. The finding was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the inspectors answered yes to the More-than-Minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern ? Absent NRC identification, the failure to adequately clean aluminum prior to welding and adequately control the repair welding techniques may have been repeated during future repairs to the SIRWT and resulted in lack of fusion type weld defects/cracks returned to service. Unstable cracks could propagate and create failure of the SIRWT pressure boundary resulting in loss of inventory and increase the risk for insufficient core cooling for post Loss-of-Coolant Accident (LOCA) conditions. Therefore, this finding adversely affected the mitigating systems cornerstone attribute of equipment performance (reliability). The inspectors determined this finding was of very low safety significance (Green) based on answering no to the questions in Part A of Exhibit 2, Mitigating Systems Screening Questions, in IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Specifically, the small amount of leakage from the SIRWT weld leak did not result in loss of a mitigating system function. Therefore, this finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance for the resources component because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety was supported.
05000266/FIN-2013003-022013Q2Point BeachFailure to Follow Operability Evaluation Process Following Water Leakage into the Control RoomThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V for the licensees failure to follow procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, following water leakage into the control room, the licensees immediate operability determination failed to evaluate the effect the leakage had on the control room envelope operability. Additionally, the licensee did not address the functionality of the degraded flood barrier and its impact on operability. This issue was entered into the CAP as AR01877185. Corrective actions for this issue included performing a test of the control room envelope to demonstrate that appropriate positive pressure could be maintained with the known degraded barrier, and repair of the degraded flood barrier following performance of a functionality assessment. The inspectors determined the finding to be more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Protection Against External Factors attribute of the Initiating Event Cornerstone, and adversely affected the Cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The inspectors determined the finding to be of very low safety significance in accordance with IMC 0609, Appendix A, Exhibit 1, because they answered No to the questions under Transient Initiators and External Event Initiators. The inspectors concluded that this finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to thoroughly evaluate this problem such that the resolution addressed the cause and evaluated the condition for operability (P.1(c)).
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