Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000313/FIN-2018003-052018Q3Arkansas NuclearFailure to Maintain Main Feedwater Pump B Discharge Pressure in Band Caused a Reactor TripThe inspectors reviewed a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to implement Procedure OP-1102.002, Plant Startup, Revision 106. Specifically, control room operators failed to maintain main feedwater pump discharge pressure in the required band to control flow to the steam generators during a plant startup. As a result, the only operating main feedwater pump tripped on high discharge pressure, causing an automatic reactor trip.
05000313/FIN-2018003-042018Q3Arkansas NuclearFailure to Verify Safety-Related 4160 V Breaker Operability Following Maintenance ActivitiesThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to perform post-maintenance testing to demonstrate component operability for the train A safety-related 4160 V switchgear A-303 breaker that provides power to the swing service water pump B (P-4B) after the breaker was racked in. The breaker subsequently failed to close when attempting to start the pump.
05000313/FIN-2018003-032018Q3Arkansas NuclearFailure to Provide Complete and Accurate Information in a License Amendment Request to Change Emergency Action Level RequirementsThe inspectors identified a Severity Level IV non-cited violation because the licensee provided inaccurate information to the NRC in a license amendment request for an emergency action level scheme change. Specifically, the licensee provided information about the availability of the postaccident sampling system building radiation monitor and the Unit 1 level instrumentation that was material to the licensing decision, but not accurate. The NRC approved an emergency action level scheme change on November 9, 2012 (ADAMS Accession No. ML12269A455) to allow Arkansas Nuclear One to adopt the Nuclear Energy Institute (NEI) 99-01, Revision 5, scheme. Subsequently, the licensee identified that two of their current emergency action level thresholds could not be implemented in accordance with their emergency classification procedure: On May 26, 2017, Condition Report CR-ANO-2-2017-03161 documented that postaccident sampling system building radiation monitor 2RX-9840 should be removed from all regulatory commitments because the postaccident sampling system had been removed from service, and its building would not be monitored for radiological releases. Radiation monitor 2RX-9840 was being used as a means to evaluate emergency action levels AU1, AA1, AS1, and AG1. In addition, it was used in the loss/potential loss of containment (CNB6) for fission product emergency action levels. The condition report noted that requirements for the postaccident sampling system had been removed from Arkansas Nuclear One licenses in August 2000 and the licensee had abandoned the systems valves (March 2003, EC-ANO-1779), removed power from the postaccident sampling system ventilation system (January 2004), and made radiation monitor 2RX-9840 nonfunctional (May 2008, Condition Report CR-ANO-2-2008-01439 and Work Order 150817). On March 15, 2018, Condition Report CR-ANO-C-2018-01121 documented that the Unit 1 level instrumentation set point used in emergency action level CA1 was below the indicating range of the instrument. The emergency action level indicated that a loss of Unit 1s reactor vessel inventory was shown by an indicated level less than 368 feet, 0 inches. Therefore, the lowest level indicated on the instrument would be higher than the level used in making the emergency classification decision. The inspectors reviewed the licensees license amendment request, dated December 1, 2011 (ADAMS Accession No. ML113350317), Proposed Emergency Action Levels Using NEI 99-01, Revision 5, Scheme, and the licensees response to a request for additional information dated July 9, 2012, (ADAMS Accession No. ML12192A090) to determine whether the conditions identified in the corrective action program existed at the time the licensee requested the license amendment and whether the request correctly described the instruments. The inspectors identified: The December 1, 2011, submittal incorrectly indicated that radiation monitor 2RX-9840 was a viable means of classifying emergency action levels AU1, AA1, AS1, and AG1, as well as providing input for the evaluation of fission product barrier emergency action levels. In the response to NRCs request for additional information (RAI) dated July 9, 2012, the licensee provided additional details about the super particulate iodine noble gas (SPING) radiation monitors used in this application. Response to Question 3 associated with emergency action levels AA1, AS1, and AG1 stated: Each SPING is associated with a particular ventilation pathway and provides continuous monitoring of air discharged via the respective release pathway. The license reviewer concluded that all of the SPING monitors included in the license amendment request were operable and continuously monitoring the specified release pathways, thereby being capable of measuring the radiation levels described in the proposed emergency action levels. 17 The December 1, 2011, submittal indicated that loss of Unit 1 reactor vessel inventory for emergency action level CA1 was a vessel level less than 368 feet, 0 inches. This issue was NRC-identified because when the licensee identified the emergency action level errors, they took action to correct the errors, but failed to address the failure to ensure that technical information provided to the NRC in support of the license amendment request was complete and accurate in all material respects. Corrective Actions: To correct the Unit 1 reactor vessel level emergency action level threshold error, the licensee issued communications regarding correct application of the emergency action level on March 15, 2018, followed by implementation of a change to Procedure OP-1903.010, Emergency Action Level Classification, Revision 56, dated June 26, 2018, with the corrected level. The use of radiation monitor 2RX-9840 is being removed from the emergency action levels as part of an emergency action level scheme change submitted to the NRC on March 29, 2018 (ADAMS Accession No. ML18088B412 and ML18094A155). In the interim, the licensee issued communications to emergency director-qualified staff members to ensure they are aware of the error, how to address it if implementing emergency action levels, and to inform them of the corrective actions in progress. Additionally, the licensee issued Condition Report CR-ANO-C-2018-03597, dated September 13, 2018, for the incomplete and inaccurate emergency action level submission examples to address the completeness and accuracy issues identified by the inspectors.
05000313/FIN-2018003-022018Q3Arkansas NuclearFailure to Implement Welding Standard Guidance and Examination ProceduresThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to implement welding standard guidance and examination procedure guidance during the installation of the high pressure injection system drain line containing drain valves MU-1066A and MU-1066B. The drain line weld developed a crack that caused a leak shortly after plant startup that was determined to have been caused by grinding during the welding process, which was not permitted by the welding standard.
05000313/FIN-2018003-012018Q3Arkansas NuclearFailure to Translate the Design Requirements into Instructions for Refueling Emergency Diesel GeneratorsThe inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate current design into instructions for Unit 1 and Unit 2 diesel fuel oil transfer system. Specifically, the licensee failed to translate the current diesel fuel oil transfer system design into instructions to refuel Unit 1 and Unit 2 safety-related fuel bunkers, T-57 and 2T-57, if the non-safety bulk diesel fuel oil tank T-25 was unavailable following a design basis event (e.g., tornado, external flooding, or earthquake) for which it was not designed to withstand.
05000313/FIN-2018003-062018Q3Arkansas NuclearReactor Power Transient Caused by the Turbine Bypass Valve Failing OpenThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly pre-plan maintenance for the replacement of air supply tubing for turbine bypass valve CV-6687, which resulted in the failure of the air tubing, causing valve CV-6687 to fail open, which led to a manual reactor trip and a subsequent loss of the main condenser.
05000382/FIN-2018002-032018Q2WaterfordLicensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification 3.6.3, Containment Isolation Valves, requires, in part, that when an isolation valve for containment penetrations associated with an open system are inoperable, the licensee must restore the inoperable valve(s) to operable status within 4 hours, isolate the affected penetration within 4 hours, or be in hot standby within the next 6 hours. Contrary to the above, between December 8, 2017, and December 11, 2017, with containment isolation valves inoperable, the licensee did not restore the inoperable valves to operable status within 4 hours, isolate the affected penetrations within 4 hours, or place the unit in hot standby within the next 6 hours. The licensee restored the valves to operable status on December 20, 2017, exceeding the Technical Specification 3.6.3 allowed outage time by approximately 70 hours. Significance/Severity Level: The finding was of very low safety significance (Green) because the containment isolation valves were maintained closed during the period and did not represent an actual open pathway in the physical integrity of the reactor containment. Corrective Action Reference: CR-WF3-2018-00983
05000313/FIN-2018002-012018Q2Arkansas NuclearFailure to Implement Procedural Guidance to Close Spent Fuel Pool Cooler Outlet Crosstie ValveThe inspectors reviewed a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One (ANO) Unit 1 Technical Specification (TS) 5.4.1.a for the licensees failure to implement Procedure OP-1102.015, Filling and Draining the Fuel Transfer Canal, Revision 44. Specifically, operators failed to close spent fuel pool cooler outlet valve SF-9 while lining up to fill the fuel transfer canal (FTC) from the borated water storage tank (BWST). As a result, the licensee drained approximately 2600 gallons from the SFP to the FTC.
05000382/FIN-2018002-022018Q2Waterford10 CFR 50.59 Evaluation Associated with Emergency Feedwater Logic ModificationThe licensee changed the emergency feedwater logic, as described in the Updated Final Safety Analysis Report (UFSAR), Section 7.3.1.1.6, from flow control mode to level control mode during a safety injection actuation signal. To accomplish this change, the licensee had to modify the following logic system signals and setpoints: steam generator critical level, steam generator lo level, steam generator lo-lo level, safety injection actuation, control board manual control, and the steam generator lo-lo level annunciator. The NRC team questioned whether the emergency feedwater modification required additional information to determine if the 10 CFR 50.59 evaluation was adequate, or if NRC approval was needed for the change. Specifically, the NRC team questioned if the emergency feedwater logic change: used a method of evaluation other than what was described in the UFSAR (e.g. the use of the TRANFLOW program) or would result in a more than minimal increase in the likelihood of occurrence of a malfunction of a system important to safety. Specifically, because the emergency feedwater logic change introduced the potential to overcool the reactor, and substituted a previous automatic action for manual operator action, the NRC team questioned if the change and associated 50.59 evaluation addressed these concerns. Planned Closure Actions: The NRC and the licensee are working to gather more information related to the Final Safety Analysis Report-described methods for steam generator analyses and if the change resulted in a more-than-minimal increase in risk. Specifically, the licensee plans to provide an analysis that demonstrates the emergency feedwater logic change would not result in a more than minimal increase in the likelihood of an overcooling accident. Licensee Actions: The licensee has implemented a compensatory measure to take manual control of the emergency feedwater system during a safety injection signal such that an overcooling event will be prevented. Corrective Action References: CR-WF3-2017-06067, CR-WF3-2017-05882, CR-WF3-2017-05173
05000382/FIN-2018002-012018Q2WaterfordFailure to Ensure Appropriate Chemistry Controls on the Component Cooling Water Heat ExchangersThe inspectors reviewed a self-revealed, Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which occurred because the licensee did not prescribe procedures for preventing fouling of the component cooling water heat exchangers that were appropriate to the circumstances. Specifically, the licensee did not require in its instructions for adding biocide to the auxiliary component cooling water system that additions be coupled with running the associated auxiliary component cooling water pump or other means of ensuring that the biocide would be sufficiently circulated through the system. As a result, on February 8, 2018, component cooling water heat exchanger B failed a performance test and therefore would not maintain required design basis temperatures under all accident conditions due to biological fouling.
05000313/FIN-2018001-032018Q1Arkansas NuclearLicensee-Identified ViolationTitle10CFR20.1501(a) requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20, and that are reasonable under the circumstances to evaluate the magnitude and extent of radiation levels, concentrations, or quantities of radioactive materials, and the potential radiological hazards that could be present.Contrary to the above, on August 7, 2017, the licensee failed to make necessary surveys of the Unit 2, 2T-15 tank room, that were reasonable to evaluate the magnitude and extent of radiation levels that could be present. Consequently, workers were allowed access to an area with dose rates up to 1000 millirem per hour at 30 cm without a proper briefing or oversight 17 Significance: Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because: (1) it was not associated with as low as is reasonably achievable (ALARA) planning or work controls; (2) there was no overexposure; (3) there was no substantial potential for an overexposure; and (4) the ability to assess dose was not compromised.Corrective Action Reference(s): CR-ANO-2-2017-04634 and CR-ANO-2-2017-0533
05000368/FIN-2018001-022018Q1Arkansas NuclearFailure to Preplan and Perform Service Water Pre-Screen MaintenanceThe inspectors reviewed a self-revealed,non-cited violation and associated finding of Arkansas Nuclear One, Unit 2, Technical Specification 6.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly preplan pre-screen cleaning maintenance, causing the trainB service water system to become inoperable
05000313/FIN-2018001-012018Q1Arkansas NuclearFailure to Establish Adequate Criteria for Flood Seal TestingThe inspectors identified a Green finding and associated non-cited violation of Unit1 Technical Specification 5.4.1.a and Unit 2 Technical Specification 6.4.1.a for the licensees failure to establish the criteria for ensuring the necessary conditions existed for a successful test of hatch flood seals. Specifically, Procedure OP 1402.240, Inspection of Watertight Hatches, Revision 1, did not contain adequate guidance to ensure that the auxiliary building was at a lower pressure than the turbine building such that puffing smoke on the turbine building side would allow a seal leak to be detectable.
05000313/FIN-2017015-012017Q4Arkansas NuclearSecurity
05000313/FIN-2017015-022017Q4Arkansas NuclearSecurity
05000313/FIN-2017003-012017Q3Arkansas NuclearFailure to Maintain Service Water Train SeparationThe inspectors identified a non- cited violation of Technical Specification 5.4.1.a for the licensees failure to maintain train separation between safety -related service water trains when swapping the swing high pressure injection (HPI) pump between trains. Specifically, by following procedure OP 1104.002, Makeup and Purification System Operation, Revision 89, operators cross -tied service water trains, placing the system in an unanalyzed condition. This condition resulted in the train A electrical equipment room emergency chiller and train B reactor building emergency cooling coils being inoperable for a maximum of 25 minutes per occurrence. Additionally, it was determined that service water temperatures over the past 3 years did not result in an actual loss of function associated with these components if a design basis accident would have occurred. The immediate corrective actions were to assess past operability for not maintaining service water train separation and to revise Operating Procedure 1104.002 with adequate work instructions to maintain service water train separation. The licensee entered this deficiency into the corrective action program as Condition Report CR -ANO -1-2017- 02518. The licensees failure to maintain safety -related service water train separation when swapping the swing HPI pump between trains was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensees failure to maintain service water train separation placed the system in an unanalyzed condition and was subsequently determined to cause the train A electrical equipment room emergency chiller and train B reactor building emergency cooling coils to be inoperable for a maximum of 25 minutes per occurrence . Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding s At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety -significant , non -technical specification train. Specifically, inspectors confirmed that service water temperatures were never high enough to result in an actual loss of function for either limiting component. The finding had 3 a cross -cutting aspect in the area of human performance associated with conservative bias because the licensee failed to determine whether the proposed action was safe to proceed, rather than unsafe in order to stop. Specifically, in December 2015 when this approach was revise d to declare only the non- protected service water train inoperable, the licensee did not ensure that the transition lineup was analyzed to be within safety analyses before adopting the revised steps. (H.14)
05000298/FIN-2017003-032017Q3CooperLoss of Control Room Ventilation Due to Inadequate Post-Maintenance Testing ActivitiesThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure that all testing required to demonstrate that the control room emergency filter system would perform satisfactorily in service was identified and performed in accordance with written test procedures. Specifically, on May 25, 2017, following corrective maintenance to replace bent positioning rods for the A and B discharge dampers for the control room supply fans, the licensee failed to ensure that all testing described in Maintenance Procedure 7.0.5, CNS Post-Maintenance Testing, Revision 53, was identified and performed, in order to assure that the control room filter system would be able to perform its safety function. As a result, on May 26, 2017, after the licensee restored the system back to service, the in-service B discharge damper was found partially closed, resulting in the supply fan failing to meet minimum flow requirements and the control room emergency filter system being declared inoperable. Corrective actions to restore compliance included replacement of the damper positioning arm, interim actions requiring post-maintenance testing after each repositioning of the dampers, and long term actions to modify the damper control arms to prevent bending and improve position verification methods. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2017-05794.The licensees failure to assure that adequate post-maintenance testing was identified and performed for work on the control room supply fan discharge dampers was a performance deficiency. Using Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined the performance deficiency was more than minor, and therefore a finding, because it was associated with the structure, system, and component, and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (control room envelope) protect the public from radionuclide releases caused by accidents or events. Specifically, the finding resulted in control room supply fan B failing to meet minimum flow requirements and the control room emergency filter system being declared inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation. Specifically, the licensee failed to ensure that the organization thoroughly evaluated indications of degraded supply fan flow that occurred during testing, and failed to properly assess bent discharge damper positioning rod deficiencies discovered during the maintenance activities, to ensure that resolutions addressed causes and extent of conditions were commensurate with their safety significance ( P. 2).
05000298/FIN-2017003-022017Q3CooperFailure to Account for Instrument Uncertainty in Safety-Related Ventilation Surveillance ProceduresThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for multiple examples of the licensees failure to assure that required testing was performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, on July 12, 2017, the inspectors identified that Surveillance Procedure 6.1SGT.501, Standby Gas Treatment A Carbon Sample, Carbon Adsorber and HEPA Filter In-place Leak Test, and Components Leak Test, Revision 16, failed to account for test instrument uncertainty in the surveillance acceptance criteria. In response to the inspectors question, the licensee discovered that instrument uncertainty was not accounted for in several standby gas treatment system surveillance procedures, as well as surveillance procedures for the control room emergency filter system; diesel generator ventilation system; control building essential ventilation system; emergency core cooling essential ventilation systems; and several emergency preparedness ventilation systems. Corrective actions to restore compliance included incorporation of instrument uncertainty into procedure changes for the affected surveillance procedures and verification that the new acceptance criteria did not challenge past operability for the affected systems. The licensee entered this issue into the corrective action program as Condition Report CR-CNS-2017-04229.The inspectors determined that the licensees failure to assure surveillance test procedures for safety-related ventilation systems incorporated test instrument uncertainty into acceptance criteria was a performance deficiency. Because the systems involved in this performance deficiency were systems that mitigate the consequences of accidents, the inspectors evaluated the finding under the Mitigating Systems Cornerstone. In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was more than minor, and therefore a finding, because it was a programmatic deficiency which adversely impacted the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the acceptance criteria for the licensees safety-related ventilation systems did not assure the availability of these systems to respond to accident conditions, as required by the technical specifications. The inspectors assessed the significance of this finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, and determined this finding was of very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant, nontechnical specification train. The finding had a cross-cutting aspect in the area of human performance associated with documentation because the licensee failed to ensure that the organization created and maintained complete, accurate, and up-to-date documentation (H.7).
05000298/FIN-2017003-012017Q3CooperFailure to Ensure Suitability of Materials for the Reactor Building Northeast Fan Coil UnitThe inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that appropriate measures were established for the selection and review for suitability of application of materials, parts, equipment, and processes that were essential to the safety-related functions of a reactor building fan coil unit. Specifically, on March 9, 2016, the licensee installed a new coil for the reactor building northeast quad fan coil unit, but failed to assure the suitability of application of the materials, parts, and equipment associated with the new coil design in that the new component had measurably higher air resistance across the coil than the previous design. As a result, on August 1, 2017, the fan coil unit failed air flow surveillance testing during the next performance of the test, resulting in the fan coil unit being declared inoperable. Corrective actions to restore compliance included cooling coil cleaning activities, implementation of compensatory measures to restore operability, and generation of a work order to replace the degraded cooling coil. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2017-04701.The licensees failure to assure that the newly designed coil installed in the northeast quad fan coil unit was appropriately reviewed for suitability and adequacy was a performance deficiency. The performance deficiency was evaluated using Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, and was associated with the Mitigating Systems Cornerstone. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the northeast quad fan coil unit being declared inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency where the component maintained operability; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant, nontechnical specification train. The finding had a cross-cutting aspect in the area of human performance associated with design margins, because the licensee failed to ensure that the organization operated and maintained equipment within design margins, and failed to ensure that these margins were carefully guarded and changed only through a systematic and rigorous process with special attention placed on maintaining safety-related equipment. Specifically, although the new fan coil units air flow immediately degraded from 7950 scfm to 7360 scfm after coil installation in 2016, which significantly degraded the margin to the minimum flow requirements, the licensee did not take action to address the degraded performance until it failed subsequent air flow testing (H.6).
05000298/FIN-2017011-012017Q3CooperEmergency Transformer Bus Failure due to Inadequate Inspection and Testing ActivitiesAV. The inspectors identified a preliminary low -to-moderate safety significance (White) finding with two NRC- identified apparent violations of Technical Specification 5.4.1.a, for the licensees failure to implement and maintain Maintenance Procedure 7.3.41, Examination and High Pot Testing of Non- Segregated Buses and Associated Equipment, Revision 10, during testing and inspection of the emergency station service transformer 4160 V bus bars. Specifically, the inspectors identified: 1. A violation of Technical Specification 5.4.1.a, for the failure to implement inspection instructions to examine the emergency transformer bus insulation for discoloration and repair the associated components on March 23, 2015; and 2. A violation of Technical Specification 5.4.1.a, for the failure to maintain adequate instructions for performing high potential testing of the emergency transformer bus bars between March 23, 2015 , and April 18, 2017. As a result, the licensee did not properly assess corona- related degradation on the emergency transformer bus, which resulted in a n emergency transformer bus fault and a loss of the emergency transformer and the supplemental diesel generator on January 17, 2017. Corrective actions to restore compliance included replacement of the faulted portions of the emergency transformer bus, extent of condition inspection and cleaning of the remainder of the emergency transformer bus bars, long term corrective actions to replace the emergency transformer bus insulation, and revision of high potential testing procedure instructions. The licensee entered these issues into the corrective action program as Condition Reports CR- CNS -2017- 00223 and CR -CNS -2017 -02164. The licensees failure to implement and maintain Maintenance Procedure 7.3.41 to properly assess degradation of the emergency station service transformer bus, in violation of Technical Specification 5.4.1.a, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and 3 challenge critical safety functions during shutdown, as well as power operations. Specifically, the finding resulted in an emergency transformer bus fault and a loss of the emergency transformer and the supplemental diesel generator. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, the inspectors determined that the finding required a detailed risk evaluation because it involve d the partial loss of a support system that contributes to the likelihood of, or causes, an initiating event (loss -of-offsite power), and the finding affected mitigation equipment (supplemental diesel generator). A senior reactor analyst performed a detailed risk evaluation in accordance with Inspection Manual Chapter 0609, Appendix A, Section 6.0, Detailed Risk Evaluation. The calculated increase in core damage frequency was dominated by station blackout initiators. The NRC preliminarily determined that the increase in core damage frequency for internal and external initiators was 6.3E -6/year, a finding of low -to-moderate risk significance (White). The performance deficiency had a cross -cutting aspect in the area of problem identification and resolution associated with evaluation, because the licensee failed to thoroughly evaluate emergency transformer electrical bus discoloration and high potential test failures to ensure that resolutions addressed the causes and extent of conditions commensurate with their safety significance (P.2 ).
05000298/FIN-2017002-042017Q2CooperLicensee-Identified ViolationTechnical Specification 5.7.1 states, in part, that high radiation areas w ith dose rates greater than 0.1 rem/hr at 30 centimeters shall be barricaded and conspicuously posted as a high radiation area. Contrary to the above, on November 2, 2016, a high radiation area with does rates greater than 0.1 rem/hr at 30 centimeters was not barricaded and conspicuously posted as a high radiation area. Specifically, a radiation protection technician (RPT) identified an unposted high radiation area at the control rod drive (CRD) A pump filter area on r eactor building 881 feet southea st quadrant. D ose rates of 120 mrem/hr at 30 centimeters from the CRD filter were identified. This issue was identified as a result of a RPTs deliberate and focused observations during the course of performing their normal duties of performing radiological surveys. The licensee documented this issue in the corrective action program as Condition Report CR- CNS -2016 -00788. The finding was determined to be of very low safety significance (Green) because it was not an ALARA planning issue, there was no overexposure or potential for overexposure, and the licensees ability to assess dose was not compromised.
05000298/FIN-2017002-032017Q2CooperLoss of Control Room Ventilation Due to Improper Switch ManipulationThe inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a , for the licensees f ailure to implement System Operating Procedure 2.2.38, HVAC Control Building, Revision 43, during control building ventilation testing. Specifically, on December 7, 2016, when directed to turn off control building ventilation recirculation fan, RF- C-1A, operations personnel instead inadvertently turned off the operating control room emergency filtration system supply fan, 1 -SF -C-1A, resulting in the loss of the control room emergency filtration system function. Corrective actions to restore compliance included restoration of the control room emergency filtration supply fan and procedure changes to require peer checks for this surveillance test and similar 4 activities. The licensee entered this deficiency into the corrective action program as Condition Report CR -CNS -2016- 08744. The licensees failure to implement System Operating Procedure 2.2.38 , in violation of Technical Specification 5.4.1.a , was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers ( control room envelope) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The finding had a cross -cutting aspect in the area of human performance associated with challenge the unknown, because the licensee did not stop when faced with uncertain conditions, and did not ensure that risks we re evaluated and managed before proceeding. Specifically, despite noting several a bnormalities with the switch being manipulated, operations personnel did not stop to evaluate the uncertain conditions nor did they evaluate the risks associated with proceeding (H.11).
05000298/FIN-2017002-022017Q2CooperLoss of Control Room Ventilation Due to Ineffective Preventive Maintenance StrategyGreen . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a , for the licensees failure to maintain work order instructions for control room supply fan maintenance resulting in the loss of the control room emergency filtration system. Specifically, prior to October 23, 2016, work order instructions for periodic preventive maintenance on the SF- C-1A supply fan failed to include industry recommended checks to ensure that the bearings were adequately engaged with the fan shaft, and failed to include proper work sequencing to ensure vibration data trending was meaningful. The ineffective preventive maintenance strategy resulted in the failure of the control room supply fan i nboard bearing during operation and a loss of the control room emergency filtration system function. Corrective actions to restore compliance included repair of the s upply fan and changes to improve the effectiveness of the fans preventive maintenance strategy. The licensee entered this deficiency into the corrective action program as Condition Report CR- CNS -2016- 07426. The licensees failure to maintain work order instructions for control room supply fan maintenance , in violation of Technical Specification 5.4.1.a , was a performance deficiency. The performance deficiency was more than minor , and therefore a finding, because it was associated with the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers ( control room envelope) protect the public fro m radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the inspectors determined that the finding had very low safety significance (Green) because the inspectors answered no to all of the Barrier Integrity screening questions. The finding had a cross -cutting aspect in the area of human performance associated with resourc es, because the licensee failed to ensure that personnel, equipment, procedures, and other resources we re available and adequate to support nuclear safety (H.1).
05000298/FIN-2017002-012017Q2CooperFailure to Assess Operability of Technical Specification System Functions during Surveillance TestingGreen . The inspectors identified a non- cited violation of Technical Specification 5.4.1.a, for the licensees fail ure to follow Station Procedure 0.26, Surveillance Program, Revision 70, and to assess the operability of alternate shutdown reactor pressure instrumentation during surveillance testing. Specifically, the licensee failed to assess the operability of the hig h pressure coolant injection turbine steam inlet pressure instrument that provides indications of reactor pressure for the alternate shutdown panel when the instrument was isolated during surveillance testing. As a result, operations personnel failed to r ecognize that the instrument was inoperable and failed to enter the appropriate technical specification action statements . As immediate corrective actions, the licensee validated that the alternate shutdown reactor pressure function was inoperable and that Technical Specification 3.3.3.2, Altern ate Shutdown System, Condition A, should have been entered, and generated a procedure change request to ensure T echnical Specification 3.3.3.2 would be entered during future surveillances . The licensee entered this deficiency into the corrective action program as Condition Report CR -CNS -2017- 02280. The licensees failure to assess the operability of alternate shutdown reactor pressure instrument ation when the high pressure coolant injection turbine inlet steam pr essure instrument was isolated for surveillance testing, in violation of Station Procedure 0.26, was a performance deficiency. The performance deficiency was determined to be more than minor , and therefore a finding, because it was associated with the hum an performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Specifically, the alternate shutdown reactor pressure instrument was inoperable when the high pressure coolant injection turbine inlet pressure instrument was isolated for surveillance testing, and the appropriate technical specification action statement was not entered. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not repr esent a loss of system and/or function; did not represent an 3 actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety -significant nontechnical specification train. The finding had a cross -cutting aspect in the area of human performance associated with work management. Specifically, the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the identification and management of risk commensurate with the isolation of the high pressure coolant injection turbine inlet pressure instrument during surveillance testing (H.5).
05000298/FIN-2017001-012017Q1CooperFailure to Maintain Alternate Shutdown Emergency ProcedureThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a for the licensees failure to maintain Emergency Procedure 5.1ASD, Alternate Shutdown, Revision 17, for establishing reactor equipment cooling system flow to the high pressure coolant injection system fan coil unit. Specifically, the licensee failed to maintain Emergency Procedure 5.1ASD with adequate instructions to place the reactor equipment cooling system north or south critical loop in service and verify reactor equipment system flow to the high pressure cooling injection system fan coil unit during some control room evacuation scenarios. The immediate corrective actions were to assess operability of the high pressure coolant injection system during control room evacuations that are not related to fire scenarios, and to revise Emergency Procedure 5.1ASD with instructions to open the criticalloop supply valves (REC-MOV-711 or REC-MOV-714) in the control room or locally, and verify reactor equipment system flow to the high pressure coolant injection fan coil unit. The licensee entered this deficiency into the corrective action program as Condition ReportCR-CNS-2017-01403. The licensees failure to maintain Emergency Procedure 5.1ASD to establish reactor equipment cooling system flow to the high pressure coolant injection fan coil unit during some control room evacuation scenarios, in violation of Technical Specification 5.4.1.a, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee did not provide instructions to establish reactor equipment cooling system flow to the high pressure coolant injection system fan coil unit, which would have complicated operator response during a control room evacuation. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant nontechnical specification train. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with identification. Specifically, the licensee failed to implement a corrective action program with a low threshold for identifying issues during the required annual review of emergency procedures (P.1).
05000298/FIN-2017001-022017Q1CooperFailure to Identify a Condition Adverse to Quality Associated with the 250 Vdc Electrical SystemThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to identify a condition adverse to quality associated with Station Procedure 2.2.24.1, 250 Vdc Electrical System (Div 1), Revision 14, in accordance with Station Procedure 0-CNS-LI-102, Corrective Action Process, Revision 6. Specifically, the licensee failed to identify that Station Procedure 2.2.24.1 contained inadequate instructions to ensure the oncoming charger 1C output voltage was matched with the bus 1A voltage when transferring bus 1A from charger 1A to charger 1C, so that technical specification bus voltage requirements would remain met. This resulted in an unexpected and initially unrecognized decline in voltage on the bus to below the required minimum of 260.4 Vdc. This condition required the licensee to declare the Division 1 250 Vdc electrical system and Division 1 residual heat removal low pressure coolant injection system inoperable. The immediate corrective action was to adjust the charger 1C float voltage greater than 260.4 Vdc to restore operability of the Division 1 250 Vdc electrical and residual heat removal low pressure coolant injection systems. The licensee entered this deficiency into the corrective action program as Condition Reports CR-CNS-2016-08658 and CR-CNS-2017-00750. The licensees failure to identify a condition adverse to quality associated with Station Procedure 2.2.24.1, to ensure technical specification bus voltage requirements were met, in violation of Station Procedure 0-CNS-LI-102, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the procedural quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, charger 1C, when in service, did not maintain battery 1A terminal voltage within the requirements of Surveillance Requirement 3.8.4.1, which required the licensee to declare the Division 1 250 Vdc electrical system and the Division 1 residual heat removal low pressure coolant injection system inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant, nontechnical specification train. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation. Specifically, the licensee failed to thoroughly evaluate the charger 1C float voltage issue to ensure that the resolution addressed the cause and extent of condition commensurate with the safety significance (P.2).
05000298/FIN-2017001-032017Q1CooperFailure to Identify a Condition Adverse to QualityThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to identify a condition adverse to quality for Division 1 residual heat removal service water booster pump A, in accordance with Station Procedure 0-CNS-LI-102, Corrective Action Process, Revision 6. Specifically, on January 5, 2017, the inspectors identified an oil level lower than normally expected, oil on the pump skid, and an oil droplet formed on the Division 1 residual heat removal service water booster pump A inboard bearing sight glass. The inspectors informed the control room of this condition, and the licensee determined the oil leakage from the pumps sight glass would have prevented the pump from operating for the required 30 days during a design basis accident. The immediate corrective action was to repair the Division 1 residual heat removal service water booster pump A inboard bearing sight glass, restoring operability of the pump. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2017-00054. The licensees failure to identify a condition adverse to quality for Division 1 residual heat removal service water booster pump A, in violation of Station Procedure 0-CNS-LI-102, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, the oil leakage from the service water booster pump A inboard bearing sight glass would have prevented the pump from operating for its required 30-day mission time during a design basis accident and resulted in the pump being declared inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety significant nontechnical specification train. The finding had a cross-cutting aspect in the area of human performance associated with challenge the unknown because the licensee failed to stop when faced with uncertain conditions and failed to ensure that risks are evaluated and managed before proceeding. Specifically, the licensee did not maintain a questioning attitude during job-site reviews to identify and resolve unexpected conditions, including lower than the expected oil level in the service water booster pump A inboard bearing sight glass, oil on the pump skid, and an oil droplet formed on the bottom of the sight glass (H.11).
05000298/FIN-2017001-042017Q1CooperFailure to Address Nonconforming Pipe Thinning in Accordance with the ASME CodeThe inspectors identified a non-cited violation of 10 CFR 50.55a(g)(4) for the licensees failure to use an approved method to disposition an American Society of Mechanical Engineers Code nonconforming condition in the residual heat removal service water system. Specifically, the licensee identified multiple locations with localized pipe thinning below the American Society of Mechanical Engineers Code B31.1 design minimum pipe-wall thickness during an ultrasonic examination but failed to use an approved method to calculate a new acceptable pipe-wall thickness. As a corrective action to restore compliance, the licensee replaced this section of piping on November 1, 2016, during Refueling Outage 29. The licensee entered this issue into the corrective action program as Condition Reports CR-CNS-2016-05558 and CR-CNS-2016-05963. The licensees failure to use an approved method to calculate a new minimum allowable pipe-wall thickness, in violation of 10 CFR 50.55a(g)(4), was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating System Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, calculating an allowable minimum pipe-wall thickness value that is below the American Society of Mechanical Engineers code design minimum value reduces the pipings structural integrity, potentially leading to the failure of the piping. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, inspectors determined the finding screened as having very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant nontechnical specification train. This finding had a cross-cutting aspect in the area of human performance associated with design margins because the licensee failed to operate and maintain the residual heat removal service water system within the American Society of Mechanical Engineers code minimum pipe-wall thickness. Specifically, having identified that the affected pipe location was below the allowable pipe-wall thickness, the licensee opted to calculate and accept a new minimum pipe-wall thickness value that was not consistent with code requirements instead of repairing the affected piping at the time of discovery (H.6).
05000298/FIN-2017001-052017Q1CooperLoss of Shutdown Cooling due to Relay MaintenanceThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to implement Maintenance Procedure 7.3.16, Low Voltage Relay Removal and Installation, Revision 22, for relay replacement work. Specifically, on October 28, 2016, the licensee failed to evaluate the potential impact of primary containment isolation system relay PCIS-REL-K27 work on shutdown cooling relay PCIS-REL-K30, which was mounted next to K27 and shared a common mounting rail. As a result, the licensee did not identify the potential of losing residual heat removal shutdown cooling, and while installing the K27 relay and snapping it into the mounting rail, workers caused a momentary actuation of relay K30 and a loss of residual heat removal shutdown cooling. Corrective actions to restore compliance included restoration of shutdown cooling, completion of the K27 relay maintenance with shutdown cooling out of service, and an outage risk management procedure change that prohibited work on or near shutdown cooling relays while the system was required to be in service. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-07645. The licensees failure to implement Maintenance Procedure 7.3.16, in violation of Technical Specification 5.4.1.a, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Initiating Events Cornerstone and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown operations. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the inspectors determined that the finding did not require a quantitative assessment because the event occurred when the refuel canal/cavity was flooded. Therefore, the finding screened as very low safety significance (Green). The finding had a cross-cutting aspect in the area of human performance associated with work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the need for coordination with different work groups or job activities. Specifically, the licensee failed to control, execute, and coordinate safety-related primary containment isolation system relay work activities to ensure residual heat removal shutdown cooling was not adversely impacted (H.5).
05000298/FIN-2017001-062017Q1CooperFailure to Install Correct Mechanical Stop and Verify Proper OperationThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 3.0.4 for the licensees failure to install the correct reactor core isolation cooling pressure control valve, RCIC-AOV-PCV23, mechanical stop and verify proper operation of the system prior to entering a mode of applicability for Technical Specification 3.5.3. This condition resulted in RCIC-AOV-PCV23 going fully open during surveillance testing following Refueling Outage 29, causing a pressure transient. This transient caused a failure of the reactor core isolation cooling turbine lube oil cooler gasket, lifting of a pressure relief valve, and a water leak. The licensee immediately shut down the reactor core isolation cooling system and declared it inoperable. The immediate corrective actions were to restore RCIC-AOV-PCV23 from the closed mechanical stop to the required open mechanical stop and to replace the turbine lube oil cooler gasket to restore operability of the system. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-08122 and initiated a root cause evaluation to investigate this condition. The licensees failure to install the correct reactor core isolation cooling pressure control valve, RCIC-AOV-PCV23, mechanical stop and verify proper operation of the system prior to entering a mode of applicability for Technical Specification 3.5.3, in violation of Technical Specification 3.0.4, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee installed RCIC-AOV-PCV23 with the incorrect mechanical stop, and proper valve operation was not verified after installation during Refueling Outage 29, which caused the reactor core isolation cooling system to lose function during surveillance testing. This transient caused a failure of the reactor core isolation cooling turbine lube oil cooler gasket and an associated water leak. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding required a detailed risk evaluation because it represented a loss of system and/or function. In the detailed risk evaluation, the analyst assumed the reactor core isolation cooling system was unavailable for 50 hours. The analyst used the Test/Limited Use Version COOPER-DEESE-HCI03 of the Cooper SPAR model run on SAPHIRE, Version 8.1.5. The analyst updated the initiating event frequencies for transients, losses of condenser heat sink, losses of main feed water, grid related losses of offsite power, and switchyard centered losses of offsite power to the more recent values from the 2014 update to the industry data found in INL/EXT-14-31428, Initiating Event Rates at U.S. Nuclear Power Plants, 1998-2013, Revision 1. From this, the finding was determined to have an increase in core damage frequency of 8.4E-8/year and to be of very low safety significance (Green). Transients, losses of condenser heat sink, and losses of main feed water were the dominant core damage sequences. The automatic depressurization system and the reactor protection system remained to mitigate these sequences. The finding had a cross-cutting aspect in the area of human performance associated with documentation because the licensee failed to create and maintain complete, accurate, and up-to-date documentation associated with RCIC-AOV-PCV23 design drawings and the maintenance procedure for setting and testing the mechanical stop (H.7).
05000298/FIN-2016004-012016Q4CooperFailure to Maintain Reactor Vessel Assembly Procedure to Ensure Adequate Moisture Separator ShieldingThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a for the licensee's failure to ensure sufficient radiological work controls were in place when the reactor pressure vessel moisture separator was installed during vessel reassembly. Specifically, the licensee failed to maintain sufficient detail in Station Procedure 7.4Reassembly, Reactor Vessel Reassembly, Revision 13, to ensure that the moisture separator had adequate water shielding during lifts, such that radiation fields were appropriately controlled. The licensee took immediate corrective action to ensure resubmergence of the radiologically significant sections of the moisture separator and restore the requisite water shielding, thereby restoring ambient refuel floor radiological conditions. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-07552. The licensee's failure to ensure sufficient radiological work controls were in place when the reactor pressure vessel moisture separator was lifted during vessel reassembly, in violation of Technical Specification 5.4.1.a, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the failure to have sufficient procedural guidance to maintain adequate water shielding on the moisture separator resulted in unanticipated elevated dose rates on the refuel floor and unplanned radiological exposures to workers in the immediate work area. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined that the violation had very low safety significance (Green) because: (1) it was not an as low as reasonably achievable (ALARA) finding; (2) there was no overexposure; (3) there was no substantial potential for an overexposure; and (4) the ability to assess dose was not compromised. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance associated with avoiding complacency. Specifically, the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes and failed to implement appropriate error reduction tools (H.12).
05000298/FIN-2016004-032016Q4CooperFailure to Maintain Service Water Pump Maintenance ProcedureThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 3.6.4.2, Secondary Containment Isolation Valves, for the licensees failure to maintain secondary containment isolation valve HV-AOV-265 operable as a result of erecting scaffolding that interfered with valve operation. Specifically, between June 29, 2016, and September 14, 2016, the licensee erected scaffolding in close proximity of valve HV-AOV-265, such that, during valve stroking, the scaffolding would pinch the actuator air line and prevent the valve from closing, rendering the valve inoperable for approximately 10 weeks. This resulted in the licensees need to reduce power to approximately 50 percent in order to comply with technical specifications upon discovery. Immediate corrective actions included removal of the scaffolding, replacement of the pinched air line, and restoration of the valve to operable status. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-05608 and initiated a root cause evaluation to investigate this condition. The licensees failure to implement Procedure 7.0.7, Scaffolding Construction and Control, Revision 34, to ensure scaffolding did not adversely affect plant equipment, in violation of Technical Specification 3.6.4.2, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the improperly erected scaffolding prevented the operation of a secondary containment isolation valve, rendering it inoperable for approximately 10 weeks. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room, reactor building, spent fuel pool building, or standby gas treatment system. The finding had a cross-cutting aspect in the area of human performance associated with resources. Specifically, the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety (H.1).
05000298/FIN-2016004-042016Q4CooperFailure to Maintain Main Steam System Operating ProcedureThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to maintain Station Procedure 2.2.56, Main Steam System, Revision 49, to prevent a main steam line high flow Group 1 primary containment isolation signal when opening an inboard main steam isolation valve. Specifically, the licensee failed to maintain Station Procedure 2.2.56 with adequate differential pressure limits for reopening closed main steam isolation valves during plant shutdown, which caused the unexpected closure of all the open main steam isolation and drain valves during the plant cooldown process. This resulted in a loss of the main steam line decay heat removal path, which caused reactor coolant system pressure and temperature to increase by approximately 13 psig and 3 degrees Fahrenheit, respectively, during the event. The immediate corrective actions were to reset the Group 1 isolation signal and open the main steam line drain valves to recommence plant cooldown. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-05835, and the licensee initiated an apparent cause evaluation to investigate this condition. The licensees failure to maintain Station Procedure 2.2.56 to prevent a main steam line high flow Group 1 isolation signal when opening an inboard main steam isolation valve, in violation of Technical Specification 5.4.1.a, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the procedural quality attribute of the Initiating Events Cornerstone and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. Specifically, the Group 1 isolation signal closed the main steam line drain valves, which resulted in a loss of the main steam line decay heat removal path and caused reactor coolant system pressure and temperature to increase. The inspectors determined Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, was not applicable because plant temperature and pressure were not within the normal residual heat removal/decay heat removal system operating parameters. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding screened as having very low safety significance (Green) because it did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. A cross-cutting aspect was not assigned to this finding because the performance deficiency occurred in 1988 when the licensee changed the procedural limits for differential pressure across the main steam isolation valves when reopening them, and therefore, was not indicative of current licensee performance.
05000298/FIN-2016004-022016Q4CooperFailure of an Analysis to Demonstrate that Changes Did Not Reduce the Effectiveness of the Emergency PlanThe inspectors identified a non-cited violation of 10 CFR 50.54(q)(3) for the licensees failure to perform an analysis demonstrating that proposed emergency plan implementing procedure changes did not reduce the effectiveness of the emergency plan. Specifically, the licensees 50.54(q) evaluation failed to demonstrate that Emergency Plan Implementing Procedure 5.7.1, Emergency Classification, Revision 54, changes, associated with Emergency Action Level SG2.1 and the fission product barrier matrix, did not result in a reduction in effectiveness. The corrective action was to revise 10 CFR 50.54(q) Evaluation 2016-011 to provide additional information about the ability of emergency coordinators in the Technical Support Center and Emergency Operations Facility to classify using the revised emergency action levels. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-05697. The licensees failure to perform an analysis demonstrating that proposed changes to Emergency Plan Implementing Procedure 5.7.1 did not reduce the effectiveness of the emergency plan, in violation of 10 CFR 50.54(q)(3), was a performance deficiency. The finding was more than minor, and therefore a finding, because it was associated with the procedure quality attribute (emergency action level changes) of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the licensees ability to ensure that adequate measures are taken to protect the health and safety of the public is degraded if the licensee performs inadequate analyses of the effects of changes to the emergency plan. Using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, Attachment 2, dated September 22, 2015, the inspectors determined that the finding was of very low safety significance (Green) because it was not associated with a risk-significant planning standard function or a planning standard function. This finding had a cross-cutting aspect in the area of human performance, associated with change management, because the licensee failed to use a systematic process for evaluating and implementing changes so that nuclear safety remains the overriding priority. Specifically, the licensee did not have an adequate understanding of the licensing basis for making changes to emergency action levels (H.3).
05000298/FIN-2016002-012016Q2CooperFailure to Meet Technical Specification Requirements for Traversing In-Core Probe B Ball Valve (The inspectors identified a non-cited violation of Technical Specification 3.6.1.3, Primary Containment Isolation Valves, for the licensees failure to maintain traversing incore probe B ball valve, a primary containment isolation valve, operable for its containment isolation function. Specifically, on May 5, 2016, from 5:20 a.m. until 1:08 p.m., the licensee failed to maintain the traversing in-core probe B ball valve operable or isolate its flow path within 4 hours of indications that the mechanical in-shield limit switch had failed. This failure prevented the ball valve from performing its containment isolation function. The licensee took immediate corrective actions upon discovery to restore compliance with Technical Specification 3.6.1.3 by de-energizing the ball valves solenoid operating valve, causing it to close. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2016-03665. The licensees failure to maintain the traversing in-core probe B ball valve, a primary containment isolation valve, operable for its containment isolation function, in violation of Technical Specification 3.6.1.3, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases and that the radiological barrier functionality of containment is maintained. Specifically, the traversing in-core probe B ball valve was unable to perform its primary containment isolation function with a failed mechanical inshield limit switch. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), and heat removal components; and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect in the area of human performance associated with conservative bias because the licensee failed to use decision making practices that emphasized prudent choices over those that were simply allowable and failed to ensure proposed actions were determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee failed to validate the assumption that the traversing in-core probe B ball valve would fulfill its containment isolation function with a failed mechanical in-shield limit switch, and failed to validate the degraded condition prior to exceeding the 4-hour completion time of Technical Specification 3.6.1.3 (Section 1R12). (H.14)
05000298/FIN-2016002-022016Q2CooperFailure to Follow Work Instructions for Post-Maintenance Testing of Safety-Related Ventilation SystemsThe inspectors identified two examples of a non-cited violation of Technical Specification 5.4.1.a, associated with the licensees failure to perform required postmaintenance testing for safety-related ventilation systems in accordance with documented instructions, prior to system restoration. Specifically, the licensee failed to follow work order instructions contained in Work Orders 5062878 and 5065112 for (1) performing surveillance testing to measure the airflow of emergency diesel generator supply fan coil unit HV-DG-1C following maintenance, and (2) performing leak testing of a newly created control room ventilation boundary penetration. Corrective actions included performing the required surveillance test for the diesel generator ventilation unit, retesting the control room penetration in accordance with the procedure, and initiating site-wide communications discussing the errors and reemphasizing procedural adherence. The licensee entered these deficiencies into their corrective action program for resolution as Condition Reports CR-CNS-2016-02207 and CR-CNS-2016-02232. The licensees failure to perform required post-maintenance testing for safety-related ventilation systems, in accordance with documented instructions, was a performance deficiency. This performance deficiency was associated with multiple cornerstones. The first example of the performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to measure supply fan coil unit HV-DG-1C airflow resulted in delayed identification that the maintenance had resulted in degraded flow through the ventilation unit. The second example of the performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases and that the radiological barrier functionality of the control room is maintained. Specifically, the licensees failure to follow post-maintenance testing instructions resulted in a challenge to the operability of the newly created control room boundary penetration seal. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because it did not represent a design or qualification deficiency; did not represent a loss of safety function; did not represent a loss of a single train for greater than its technical specification allowed outage time; did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events; did not represent an actual open containment pathway; and did not involve a reduction in function of hydrogen igniters. The finding had a crosscutting aspect in the area of human performance associated with work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the need for coordination with different work groups or job activities. Specifically, the licensee failed to control, execute, and coordinate safety-related ventilation work activities to ensure all required post-maintenance testing was completed satisfactorily prior to declaring the associated equipment operable (Section 1R19). (H.5)
05000298/FIN-2016002-032016Q2CooperFailure to Maintain Design Control for High Pressure Coolant Injection System Electrical CircuitThe inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design of the high pressure coolant injection auxiliary lube oil pump 125 Vdc starter circuit. Specifically, in 1984, the licensee modified the design of the starter circuit and eliminated a resistor that served to protect the circuit from shorting due to indication light bulb failures. As a result, on April 26, 2016, a shorted light bulb resulted in the loss of power to the auxiliary lube oil pump, rendering the high pressure coolant injection system inoperable and unavailable. Immediate corrective actions included replacing the light socket and blown fuse and changing out the nonessential light bulb with an essential bulb. This event was entered into the licensees corrective action program as Condition Report CR-CNS-2016-02318, and the licensee initiated a root cause evaluation to investigate the failure. The licensees failure to verify the adequacy of design of the high pressure coolant injection auxiliary lube oil pump starter circuit in accordance with 10 CFR Part 50, Appendix B, Criterion III, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, at the time the modification was installed, the licensee had not taken sufficient actions to ensure that the electrical circuit was protected from light bulb shorting failures, resulting in the high pressure coolant injection system ultimately being rendered inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, inspectors determined that the finding required a detailed risk evaluation because it represented a loss of the system and function of high pressure coolant injection. The inspectors determined that the finding was of very low safety significance (Green) through performing a detailed risk evaluation. A cross-cutting aspect was not assigned to this finding because the performance deficiency occurred in 1984, and therefore, is not indicative of current licensee performance (Section 4OA3).
05000298/FIN-2016001-022016Q1CooperFailure to Assess Operability of Technical Specification System Functions during Surveillance TestingThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to follow Station Procedure 0.26, Surveillance Program, and assess the operability of high pressure coolant injection steam line isolation instrumentation during surveillance testing. Specifically, the licensee failed to assess the operability of required isolation instrumentation when maintentance personnel opened terminal box 392 during surveillance testing and temporarily invalidated its environmental qualification. Licensee procedures required operations personnel to either establish compensatory measures to restore the terminal box during an event, or declare the instrumentation inoperable and enter the applicable technical specification actions when the terminal box was opened. As an immediate corrective action, the licensee implemented Standing Order 2016-03, which directed operators to establish compensatory measures, if applicable, or declare the affected equipment inoperable when environmentally qualified terminal boxes would be opened during testing. The licensee entered this issue into their corrective action program for resolution as Condition Reports CR-CNS-2016-00320 and CR-CNS-2016-00476. The licensees failure to assess the operability of high pressure coolant injection instrumentation when the associated terminal box was opened during surveillance testing, in violation of Station Procedure 0.26, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the structure, system, component, and barrier performance attribute of the Barrier Integrity Cornerstone, and adversely affected the cornerstone objective to ensure the radiological barrier functionality of containment isolation. Specifically, with terminal box 392 open, its environmental qualification was temporarily invalidated, making the high pressure coolant injection low steam pressure and high steam flow containment isolation instrumentation inoperable during surveillance testing. In addition, two other terminal boxes and their associated surveillances were impacted by the performance deficiency. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the finding had very low safety significance (Green) because it: (1) did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components; and (2) did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect in the area of human performance associated with work management. Specifically, the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the identification and management of risk commensurate with opening terminal box 392 during surveillance testing (H.5).
05000298/FIN-2016001-012016Q1CooperFailure to Follow ASME Code Requirements when taking Corrective Actions for a Pump in the Required Action RangeThe inspectors identified a non-cited violation of 10 CFR 50.55a, Codes and Standards, for the licensees failure to follow the ASME Code for Operation and Maintenance of Nuclear Power Plants when addressing the performance of reactor equipment cooling pump A within the high required action range of the inservice testing program. Specifically, on February 11, 2016, the licensee failed to follow ASME Subsection ISTB 6200(b) when engineering personnel, taking corrective action to address pump performance, failed to either correct the cause of the deviation or establish new reference values for the pump. Instead of establishing new reference values, the licensee performed an analysis to administratively raise the upper required action range limit, creating a wider range of acceptable pump operation than allowed by Table ISTB-5100-1, Centrifugal Pump Test Acceptance Criteria. The licensee entered this issue into the corrective action program as Condition Report CR-CNS-2016-00920, took action to reevaluate and rebaseline the pump with new reference values, and performed an extent of condition review to determine if other equipment was impacted by similar interpretations of the code. The licensees failure to establish new reference values for reactor equipment cooling pump A in accordance with the ASME Code was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the actions initially taken by the licensee would have required a relief request; could have delayed identification of a degrading pump trend due to the creation of a wider range of acceptable operation; and the licensees generic interpretation, that the Table ISTB-5100-1 acceptable range could be administratively expanded, represented a programmatic vulnerability. The inspectors used Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined that the finding had very low safety significance (Green) because it did not represent a design or qualification deficiency, did not represent a loss of safety function for a single train for greater than its technical specification allowed outage time, and did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation. Specifically, the licensee failed to thoroughly evaluate performance of reactor equipment cooling pump A in the required action range to ensure that the resolution correctly addressed the causes of the degraded performance (P.2).
05000298/FIN-2015004-022015Q4CooperFailure to Update the Updated Safety Analysis ReportThe inspectors identified two examples of a non-cited violation of 10 CFR 50.71(e), Maintenance of Records, Making Reports, for the licensees failure to update the Updated Safety Analysis Report for the reactor equipment cooling system and fire protection program to ensure that the report contained the latest information. Specifically, licensing personnel failed to update the Updated Safety Analysis Report when implementing License Amendment 232, in May 2009, for changes associated with the reactor equipment cooling system and again in April 2015, when the licensee implemented License Amendment 248 for the fire protection program transition to meet the requirements of NFPA-805. The licensee initiated corrective actions to update the affected sections, and initiated an extent of condition evaluation to identify other similar portions of the Updated Safety Analysis Report that may not have been updated. The licensee entered these deficiencies into the corrective action program as Condition Reports CR-CNS-2015-05948, CR-CNS-2015-06240, and CR-CNS-2015-06483. The licensees failure to update the Updated Safety Analysis Report for the reactor equipment cooling system and fire protection program to ensure that the information included within the report contained the latest information developed in accordance with 10 CFR 50.71(e) was a performance deficiency. This performance deficiency was screened using Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, and was determined to be minor in the Reactor Oversight Process, and therefore, it was not evaluated as a finding using the significance determination process. In accordance with the NRC Enforcement Policy, the performance deficiency was evaluated using the traditional enforcement process because it had the potential for impacting the NRCs ability to perform its regulatory function. Under the traditional enforcement process, this performance deficiency was determined to be more than minor and a Severity Level IV violation because it was consistent with the example in Paragraph 6.1.d.3 of the NRC Enforcement Policy, dated February 4, 2015. Specifically, the licensee failed to update the Updated Safety Analysis Report as required by 10 CFR 50.71(e), but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. No cross-cutting aspect was assigned to this violation because there was no Reactor Oversight Process finding associated with the performance deficiency.
05000298/FIN-2015004-032015Q4CooperFailure to Perform a Complete Evaluation of the Licensee Interface With Offsite OrganizationsThe inspectors identified a non-cited violation of 10 CFR 50.54(t)(2), for the licensees failure to include an evaluation of the adequacy of the interfaces with state and local governments in a review of emergency preparedness program elements in Audit 2014-06, dated November 7, 2014. Specifically, the licensee failed to include an evaluation of this interface when audit personnel did not provide offsite officials with an opportunity to provide their view of the adequacy of the interface to the audit team. Corrective actions included development of lessons learned for future audits and reengagement with state and local governments to assure adequate interface existed during the most recent emergency preparedness audits. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2015-06403. The failure to perform an evaluation for adequacy of the interface with state and local governments was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the offsite emergency preparedness attribute of the Emergency Preparedness Cornerstone, and affected the cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the ability to implement adequate measures to protect the health and safety of the public could be affected if communication and coordination problems between the licensee and offsite agencies are not detected and corrected. The finding was evaluated using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 22, 2015, and was determined to have very low safety significance (Green) because it was a failure to comply with NRC requirements, was not a loss of planning standard function, and was not a degraded planning standard function. The finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the audit team failed to fully evaluate the potential for problems to exist with the adequacy of the interface with state and local governments (P.2).
05000298/FIN-2015004-012015Q4CooperDiesel Fuel Oil Cloud Point Acceptance Criteria not in accordance with ASTM D975, Revision 1989aThe inspectors identified a non-cited violation of Technical Specification 5.5.9, Diesel Fuel Oil Testing Program, for the licensees failure to establish an emergency diesel generator fuel oil cloud point acceptance criterion in accordance with ASTM D975, Standard Specification for Diesel Fuel Oils. Specifically, the diesel fuel oil cloud point acceptance criterion of = 32F specified in the licensees diesel fuel oil testing program procedures was not in accordance with the ASTM limit of = 3.2F and was not technically justified as described by the standard. Corrective actions included development of an evaluation which concluded that the appropriate acceptance criterion was = 15F based on the most limiting day tank room temperatures during accident conditions; verification that the cloud point of the fuel onsite at the time was 8.6F, which met this criterion; and establishment of compensatory measures to monitor and administratively control the cloud point until fuel oil program procedures could be revised. The licensee entered this deficiency into the corrective action program as Condition Reports CR-CNS-2015-06745, CR-CNS-2015-06717, CR-CNS-2015-06718, and CR-CNS-2015-7150. The licensees failure to establish a diesel fuel cloud point acceptance criterion in accordance with ASTM D975, in violation of Technical Specification 5.5.9, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to establish a diesel fuel cloud point acceptance criterion in accordance with ASTM D975 could result in formation of wax crystals affecting the capability to transfer the fuel oil from the storage tanks to the emergency diesel generator engine cylinders. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding had a cross-cutting aspect in the area of human performance associated with documentation because the licensee failed to create and maintain complete, accurate, and up-to-date documentation for the worst case temperature at which the emergency diesel generator fuel oil would be stored (H.7).
05000298/FIN-2015003-062015Q3CooperFailure to Preclude Repetition for a Significant Condition Adverse to QualityThe inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, associated with the inadequate extent of condition and extent of cause evaluations to preclude repetition for a significant condition adverse to quality identified in a 2012 root cause evaluation documented CR-CNS-2012- 07174 for the isolation of shutdown cooling system isolation in valves RHR-MOV-17 and RHR-MOV-18 due to localized pressure perturbations at the pressure sensors. Specifically, in 2012, the licensee failed to conduct an adequate extent of cause and condition evaluation to preclude repetition of this event from occurring on May 30, 2015 with the reactor plant in Mode 4. On May 30, 2015, isolation of shutdown cooling system isolation valves RHR-MOV-17 and RHR-MOV-18 due to localized pressure perturbations at the pressure sensors, led to the isolation of the shutdown cooling system for approximately 22 minutes. The station entered Station Procedure 2.4SDC, Shutdown Cooling Abnormal, Revision 14, and restored shutdown cooling. The reactor coolant system temperature increased approximately 20 degrees Fahrenheit but did not exceed 212 degrees Fahrenheit, maintaining the reactor plant in Mode 4. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2015-03188. The licensees failure to conduct an adequate extent of cause and condition evaluation to preclude repetition of a significant condition adverse to quality identified in a 2012 root cause evaluation documented in CR-CNS-2012-07174 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Initiating Events Cornerstone, and affected the associated cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. Specifically, the failure to preclude repetition of the isolation of shutdown cooling system isolation valves RHR-MOV-17 and RHR-MOV-18 due to localized pressure perturbations at the pressure sensors led to the isolation of the shutdown cooling system for approximately 22 minutes when the reactor plant was in Mode 4 on May 30, 2015. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, inspectors determined that the finding did not require a quantitative assessment because adequate mitigating equipment remained available, and the finding did not constitute a loss of control, as defined in Appendix G. Therefore, the finding screened as a very low safety significance (Green). The inspectors determined that the finding did not have a cross-cutting aspect because the most significant contributor of this finding occurred in 2012, and does not reflect current licensee performance.
05000298/FIN-2015003-042015Q3CooperFailure to Make a 10 CFR 50.72(b)(2)(xi) NotificationThe inspectors identified a non-cited violation of 10 CFR 50.72(b)(2)(xi) because the NRC Operations Center was not notified within four hours of a reportable event related to the health and safety of the public for which notification to other government agencies had been made. Specifically, in May 2013, the licensee did not notify the NRC of its notification to the State of Nebraska about an inadvertent release of 14 bags of radioactively contaminated dirt and debris to a public landfill. To correct this condition, the licensee notified the NRC Operations Center of this event on August 26, 2015. This violation was evaluated using traditional enforcement because the failure to make a required report could adversely impact the NRCs regulatory process. Using the criteria contained in Section 6.9(d)(9) of the NRCs Enforcement Policy, this violation was determined to be Severity Level IV. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2015-0544. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000298/FIN-2015003-082015Q3CooperLicensee-Identified ViolationTechnical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, Quality Assurance Program Requirements, of February 1978. Section 8.b of Regulatory Guide 1.33 recommends that specific procedures for surveillance tests, inspections, and calibrations should be written (implementing procedures are required for each surveillance test, inspection, or calibration listed in the technical specification) for containment leak-rate and penetration leak-rate tests. The licensee maintains Station Procedure 6.PC.524, Primary Containment Airlock Local Leak Rate Tests, Revision 21 for containment and penetration local leak-rate testing for the primary containment personnel airlock. Contrary to the above, until June 3, 2015, the licensee failed to maintain procedure 6.PC.524 to provide surveillance testing guidance to test the inner personnel airlock equalization valve in the accident direction. This condition resulted in surveillance tests not being performed within their specified frequency and questioned operability of the inner personnel airlock equalization valve. The station implemented the requirements of Surveillance Requirement 3.0.3 and conducted a risk evaluation to determine that integrated leak rate test conducted in Refueling Outage 27 tested the inner personnel airlock equalization valve in the accident condition providing reasonable expectation of operability. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the containment functionality was maintained. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that the finding screened as having very low safety significance (Green) because it did not represent an actual: (1) open pathway in the physical integrity of reactor containment (valves, airlocks, etc.) containment isolation system (logic and instrumentation), and heat removal components; and (2) reduction in function of hydrogen igniters in the reactor containment. The license entered this deficiency into the corrective action program Condition Report CR-CNS-2015-00986.
05000298/FIN-2015003-072015Q3CooperLicensee-Identified ViolationTechnical Specification 3.3.6.1, Primary Containment Isolation Instrumentation, Action A.1, requires that inoperable high main steam line flow isolation channel(s) be placed in trip in 12 hours. Contrary to the above, from November 21, 2014, to February 19, 2015, the licensee failed to place three inoperable Division II high main steam line flow isolation channels in trip within 12 hours, because the licensee failed to properly implement the requirements of TSTF-493, Clarify Application of Setpoint Methodology for Limiting Safety System Settings, for main steam differential pressure indicating switches. The licensee recalibrated the main steam differential pressure indicating switches in question and re-stored the setpoints to the pre TSTF-493 values. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the containment functionality was maintained. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that the finding screened as having very low safety significance (Green) because it did not represent an actual: (1) open pathway in the physical integrity of reactor containment (valves, airlocks, etc.) containment isolation system (logic and instrumentation), and heat removal components; and (2) reduction in function of hydrogen igniters in the reactor containment. The license entered this deficiency into the corrective action program as Condition Report CR-CNS-2015-03315.
05000298/FIN-2015003-052015Q3CooperFailure to Follow Primary Containment Atmosphere Sampling ProcedureThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to appropriately implement a procedure required by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Specifically, on June 2, 2015, a chemistry technician failed to implement Station Procedure 8.8.12, Primary Containment Oxygen or Noble Gas Activity Grab Sample Analysis, Revision 14. This resulted in the incorrect primary containment isolation sample valve being operated, which resulted in both divisions of primary containment H2O2 analyzers tripping on low pressure/flow. Operations personnel declared both divisions of primary containment H2O2 analyzers inoperable and entered Limiting Condition for Operation 3.3.3.1, Post Accident Monitoring Instrumentation, Conditions A and C, and restored them to an operable status in accordance with station procedures. The licensee entered this deficiency into the corrective action program as Condition Reports CR-CNS-2015-03292. The licensees failure to operate the correct primary containment isolation sample valve, in support of primary containment atmosphere sampling, in violation of Station Procedure 8.8.12, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Question, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The finding has a human performance cross-cutting aspect within the avoid complacency area because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, which resulted in individuals not implementing appropriate error reduction tools (H.12).
05000298/FIN-2015003-032015Q3CooperFailure to Control Licensed MaterialThe inspectors reviewed a self-revealing non-cited violation of 10 CFR 20.1802 for the failure to control licensed material not in storage when the licensee sent 14 bags of radioactively contaminated dirt and debris to an off-site landfill for disposal. Immediate corrective actions included the licensee retrieving the contaminated material and returning it to site. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2013-03392. The failure to control licensed material that was not in storage was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Public Radiation Safety Cornerstone and adversely affected the cornerstone objective of assuring 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 used IMC 0609, Significance Determination Process, Attachment D, Public Radiation Safety Significance Determination Process, February 12, 2008. The inspectors determined the finding to be of very low safety significance (Green) because the finding involved radioactive material control but it did not result in an exposure to the public in excess of five millirem. The finding has a cross-cutting aspect in the area of human performance, associated with work management, because the licensee did not implement a process of planning, controlling, and executing work activities such that safety was the priority. Specifically, the licensee did not control work activities involving multiple organizations such that radioactive material remained controlled on site (H.5).
05000298/FIN-2015003-022015Q3CooperFailure to Ensure Measurement Conditions Were Consistent With Instrument CalibrationThe inspectors identified a non-cited violation of 10 CFR 20.1501(c) for the failure to ensure measurement conditions were consistent with instrument calibration parameters for the elevated release point monitor, compromising the ability to accurately determine the concentration of radioactive effluents released. Specifically, water intrusion/condensation in the elevated release point Kaman normal range effluent monitor noble gas sample chamber introduced discrepancies relative to the calibration geometry and water in the particulate filter and iodine cartridge adversely affected the sample media collection efficiencies. Immediate corrective actions included the licensee performing a functionality assessment of the monitor. The licensee entered this deficiency into the corrective action program as Condition Reports CR-CNS-2015-05051 and CR-CNS-2015-05067. The failure to ensure measurement conditions were consistent with instrument calibration parameters for the elevated release point monitor was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Public Radiation Safety Cornerstone attribute of plant equipment/process radiation monitoring and adversely 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 used IMC 0609, Significance Determination Process, Attachment D, Public Radiation Safety Significance Determination Process, February 12, 2008, and determined the finding to be of very low safety significance (Green) because it was associated with the effluent program; however, it was not a substantial failure to implement the effluents program and it did not result in a public dose greater than an Appendix I criterion or 10 CFR 20.1301(e). The finding has a cross-cutting aspect in the area of problem identification and resolution associated with identification, because the organization failed to implement the corrective action program with a low threshold for identifying issues. Specifically, plant personnel failed to initiate condition reports, as required by procedure, on 89 occasions since the discovery on March 24, 2015 (P.1).
05000298/FIN-2015003-012015Q3CooperFailure to Ensure Turbine Building Design Calculation was Correct and JustifiedThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to September 29, 2015, the licensee prepared Calculation NEDC 13-028, Ultimate Internal Pressure of Turbine Building Blowout Panels and Metal Wall System, Revision 1, in accordance with Engineering Procedure 3.4.7, to ensure pressure relief in the turbine building due to a main steam line break would occur at less than or equal to 0.5 pounds per square inch differential pressure as stated in Amendment 25 to the Cooper Nuclear Station Final Safety Analysis Report. However, the inspectors determined that the methodology and assumptions employed in Calculation NEDC 13-028 were not adequate and could not conclude that it ensured siding failure as required. In response to this issue, the licensee performed an operability determination to ensure that safety-related structures, systems, and components and the control room were not adversely affected by a main steam line break. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2015-05705. The licensees failure to ensure that a turbine building design calculation was correct and justified was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, 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. Specifically, Calculation NEDC 13-028 did not ensure that safety-related structures, systems, and components and the control room, which are necessary for responding to initiating events, would not be adversely affected by a main steam line break in the turbine building. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Question, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding had a crosscutting aspect in the area of human performance associated with conservative bias because individuals failed to use decision making practices that emphasize prudent choices over those that are simply allowed. (H.14).
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Christopher-20Henderson-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"