Semantic search

Jump to navigation Jump to search
 Start dateSiteIdentified byTitleDescription
05000313/FIN-2018003-0130 September 2018 23:59:59Arkansas NuclearNRC identifiedFailure 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-2018003-0530 September 2018 23:59:59Arkansas NuclearSelf-revealingFailure 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-2018405-0130 September 2018 23:59:59Arkansas NuclearNRC identifiedSecurity
05000313/FIN-2018003-0230 September 2018 23:59:59Arkansas NuclearSelf-revealingFailure 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-2018003-0630 September 2018 23:59:59Arkansas NuclearSelf-revealingReactor 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.
05000313/FIN-2018405-0230 September 2018 23:59:59Arkansas NuclearLicensee-identifiedLicensee-Identified Violation
05000313/FIN-2018011-0130 September 2018 23:59:59Arkansas NuclearNRC identifiedFailure to Properly Size the Unit 1 Emergency Diesel Generator Room Ventilation SvstemsAn NRC identified Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," was identified for failure to properly size the Unit 1 emergency diesel generator room ventilation systems to be capable of removing the design heat load during the most limiting design conditions while maintaining redundancy of the exhaust fans.
05000313/FIN-2018003-0330 September 2018 23:59:59Arkansas NuclearNRC identifiedFailure 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-2018011-0230 September 2018 23:59:59Arkansas NuclearNRC identifiedFailure of Both Arkansas Nuclear One Units to Establish Adequate Corrective Actions Resulting in Excessive Instances of Damaged and Broken Internals of the Emergency Feedwater Pum o Turbine Steam Admission Check Valves.An NRC identified Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, "Corrective Actions," was identified for failure to establish an adequate corrective action program and the resulting inability to correct a deficient system design which resulted in damaged and broken internals of the check valves admitting steam to the emergency feedwater turbine.
05000313/FIN-2018003-0430 September 2018 23:59:59Arkansas NuclearSelf-revealingFailure 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-2018011-0330 September 2018 23:59:59Arkansas NuclearNRC identifiedFailure to Evaluate the Effects and the Suitability of Components in Containment from a Main Steam Line Break.The team identified an unresolved item (URI) related to the containment environment that would result from a main steam line break. Specifically, for ANO Unit 1 the licensee did not analyze the containment temperature, or evaluate the suitability of components in containment for the effects of a main steam line break (MSLB) accident. The Final Safety Analysis Report states, in part, that "At the end of Cycle 19, the original once through steam generators (OTSGs) were replaced. In support of Cycle 20 operation, an evaluation of the containment pressure/temperature response with the replacement OTSGs for loss of coolant accidents (LOCA) and MSLB was performed. For the MLSB, the containment pressure response with the replacement OTSGs was bounded by the current analysis. The post-MSLB temperature response w ith the replacement OTSGs would be worse. Entergy Operations, Inc. has adopted NUREG-0458 into the AN0-1 licensing basis which recognizes that the post-MSLB atmosphere may become superheated, but the temperature spike is of such short duration that the thermal lag of any SSC inside containment will not increase significantly. Consequently, the initial temperature peak does not define operating limits on any system, structure, or component (SSC) and the long-term containment temperature (which is essentially the saturation temperature) dominates the temperature response of SSCs. Therefore, as long as the peak MSLB pressure is less than the peak pressure following a LOCA, the temperature response of SSCs will still be defined by the LOCA." The NRC issued several bulletins subsequent to the issuance of NUREG-0458. Specifically IEB-79-01, as supplemented, and NRC Order CLI 80-21 state, in part, that "The Guidelines leave open the question of what standard will be applied to replacement parts in operating plants. Unless there are sound reasons to the contrary, the 1974 standard in NUREG-0588 will apply. The Guidelines and NUREG-0588 apply progressively less strict standards to the older plants. The justification for this position was not articulated at the time the older plants were grandfathered from the provisions of Reg. Guide 1.89." The NRC issued a Safety Evaluation Report to ANO, which states, in part, "A final rule on environmental qualification of electric equipment important to safety for nuclear power plants became effective on February 22, 1983. This rule, Section 50.49 of 10 CFR 50, specifies the requirements of electrical equipment important to safety located in a harsh environment. In accordance with this rule, equipment for Arkansas Unit 1 may be qualified to the criteria specified in either the DOR Guidelines or NUREG-0588, except for replacement equipment. Replacement equipment installed subsequent to February 22, 1983 must be qualified in accordance with the provisions of 10 CFR 50.49, using the guidance of Regulatory Guide 1.89, unless there are sound reasons to the contrary." The NRC issued Information Notice 85-39 states, in part, that the "Qualification of some replacement equipment was based on previously allowed DOR guidelines that stated "equipment is considered qualified for main steam line break environmental conditions if it was qualified for a loss-of-coolant accident environment in plants with automatic spray systems not subject to disabling single component failures." This basis of qualification is not acceptable without additional justification for replacement equipment that was procured and installed after February 22, 1983." The replacement steam generators have several design differences compared to the original steam generators. Specifically, the replacement steam generators were designed with larger secondary volumes, more tubes, flow-restricting venturis, and different materials (Alloy 690 vs. Alloy 600). Because the replacement steam generators were installed in 2005 (after 10 CFR 50.49 became effective on February 22, 1983) all replacement equipment must be qualified using the guidance of NUREG-0588 or Regulatory Guide 1.89. In addition, as stated above the licensee did not analyze or quantify the containment temperature that would result from a MSLB, and instead compared the containment pressures and the mass/energy releases that would result from a MSLB using the superseded guidance of NUREG-0458. The NRC team identified that there are several parameters that could have changed with the replacement steam generators which could impact the containment response. Specifically, input parameters such as: sub-compartment analysis, net positive suction head analysis, containment volume, heat sinks, properties of materials, heat transfer coefficients, initial conditions, and possibly cooling water temperature may affect the containment temperature response.
05000313/FIN-2018002-0130 June 2018 23:59:59Arkansas NuclearSelf-revealingFailure 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.
05000313/FIN-2018001-0131 March 2018 23:59:59Arkansas NuclearNRC identifiedFailure 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.
05000368/FIN-2018001-0231 March 2018 23:59:59Arkansas NuclearNRC identifiedFailure 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-2018001-0331 March 2018 23:59:59Arkansas NuclearLicensee-identifiedLicensee-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
05000313/FIN-2017015-0231 December 2017 23:59:59Arkansas NuclearNRC identifiedSecurity
05000313/FIN-2017015-0131 December 2017 23:59:59Arkansas NuclearNRC identifiedSecurity
05000313/FIN-2017007-0130 September 2017 23:59:59Arkansas NuclearNRC identifiedFailure to Promptly Identify and Correct an Inadequate Design Bases CalculationThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, from 1996 until August 10, 2017, the licensee failed to properly resolve the environmental conditions in room 38 following a high-energy line break, even when challenged during a self-assessment by members of the quality assurance group in June 29, 2015. In response to this issue, the licensee determined that in the event of a break in the letdown line, an engineered safety feature signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 40 seconds, preventing room 38 from going harsh. This finding was entered into the licensees corrective action program as Condition Report CR-ANO-1-2017-02441. The inspectors determined that the licensees failure to adequately evaluate and take prompt corrective actions to resolve an identified condition adverse to quality related to the high energy line break analysis for room 38 was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the licensee identified that the environmental conditions in room 38 of the auxiliary building were harsh, as determined by Design Bases Calculation CALC-01-EQ-1002-02, they failed to properly resolve the condition adverse to quality. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and 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 allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human resources, training, because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, there was a lack of understanding of the current licensing bases for the plant displayed by engineering, operations, and management (H.9).
05000313/FIN-2017003-0130 September 2017 23:59:59Arkansas NuclearNRC identifiedFailure 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)
05000313/FIN-2017008-0230 June 2017 23:59:59Arkansas NuclearNRC identifiedInadequate FLEX ProceduresGreen. The team identified a finding with three examples for the licensee failing to assure that FLEX procedures were adequate for implementation of the strategies credited in the licensees Final Implementation Plan. This issue was entered into the licensees corrective action program as Condition Reports CR -ANO -C-2017- 00341, CR- ANO -C 2017- 00344, CR- ANO -1-2017 -00250, and CR -ANO -C-2017 00295. The failure to provide adequate procedures for responding to an extended loss of all AC power due to a flooding or high wind event was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the finding was evaluated using NRC Inspection Manual Chapter 0609, Appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA -12-049 and EA -12-051), dated October 7, 2016, and Appendix M , Significance Determination Process Using Qualitative Criteria, dated April 12, 2012. A bounding evaluation was performed using the exposure time, frequency of random failure of both emergency diesel generators , and tornado frequency or flood frequency . The licensees compliance date wit h the order was January 12, 2016, so an exposure time of one year was used. The random failure frequency of both units emergency diesel generators (3.15E -3/year) was selected since the emergency diesel generators are protected from damage during high wind and flood events. For the two examples impacted by flood events, t he flood frequency selected was for a flood exceeding the site grade elevation (8.47E -5/year). The change in core damage frequency for the se examples was determined to be 2.67E -7/year. For t he example which would only impact the licensee s response to a high wind event , the tornado frequency selected was for an F2 or greater tornado striking the site (5.31E -5/year). The change in core damage frequency for th is example was determined to be 1.67E -7/year. Therefore, the three examples of the finding were determined to of very low risk significance. The findi ng had a cross-cutti ng aspect in the Procedure Adherence co mponent of Human Performance becau se the lice nsee failed to adequately perform reviews required by the licensees procedure control program to confirm that : (1) instructions for implementing the strategies in the licensees Final Implementation Plan were complete and appropriate; and (2) reviews for affected procedures relat ed to other procedure revisions identified impacts on the implementing strategies and revised them appropriately (H.8).
05000313/FIN-2017002-0330 June 2017 23:59:59Arkansas NuclearSelf-revealingFailure to Comply with ECCS Technical Speci ficationsGreen . The inspectors reviewed a Green self -revealing finding and associated non -cited violation of Unit 1 Technical Specification 3.5.2, Emergency Core Cooling System (ECCS) Operating, for the licensees failure to ensure the operability of the P36A high pressure injection pump after reinstalling its feeder breaker during a unit outage. A violation of Unit 1 Technical Specification 3.0.4 was also identified for making a mode change without meeting the requirements to do so. Following unit restart, the pump failed to start during routine equipment rotation, resulting in one train of emergency core cooling system being inoperable for long er than allowed by Unit 1 Technical Specifications. The licensee subsequently identified that the feeder breaker had not been fully racked into position. Inspectors also noted that the breaker had been racked in manually rather than using the normal electric racking tool, and no special precautions had been taken to ensure this infrequently -used method was successful. When the breaker was correctly racked in, the pump was satisfactorily tested. The licensee subsequently verified that all similar breakers were correctly racked into position. The licensee entered this issue into their corrective action program as Condition Report CR- ANO -1-2017- 01764. The inspectors determined that the failure to verify that the P36A high pressure injection pump was operable after racking its feeder breaker into the switchgear cubicle was a performance deficiency. The performance deficiency was more than minor 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. 4 The inspectors performed the initial significance determination for the performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012 , and concluded that it required a detailed risk evaluation because it involved the loss of a single train of mitigating equipment for longer than the technical specification allowed outage time. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The estimate in the increase in core damage frequency is 4.4 E-8 per year, or of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because the licensee failed to ensure that individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to verify that the pump was operable after its breaker was rein stalled, even though an infrequently-used method was employed (H.12).
05000313/FIN-2017002-0430 June 2017 23:59:59Arkansas NuclearLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50.55a(g)4, Inservice Inspection Standards Requirement for Operating Plants, states in part, Throughout the service life of a pressurized water -cooled nuclear power facility, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the ASME Code. The ASME Section XI, Article IWA - 2610, requires that all welds and components subject to a surface or volumetric examination be included in the licensees inservice inspection program. This includes identifying system supports in the inservice inspection plan, per ASME Section XI, Article IWA -1310. Contrary to the above, prior to March 9, 2017, the licensee did not ensure that all welds and components subject to a surface or volumetric examination were included in the licensees inservice inspection. Specifically, the licensee did not apply the applicable inservice inspection requirements for surface or volumetric examination to all portions of the Unit 2 emergency feedwater system within the system ASME Code Class 3 boundary. The licensee identified that they failed to include the emergency feed pump supports in their inservice inspection program. The licensee entered this issue into their corrective action program as Condition Report CR- ANO -2-2016 -01023 and reasonably determined the emergency feedwater system remained operable. The licensee restored compliance by inspecting the supports, with no degradation identified, and entering the emergency feedwater pump supports into the ASME Section XI program. The finding was of very low safety significance (Green) because the finding did not 34 represent an actual loss of safety function of a system or train and did not result in the loss of a single train for greater than technical specification allowed outage time. This issue was entered into the licensees corrective action program as Condition Report CR- ANO -2-2016- 01023.
05000368/FIN-2017002-0130 June 2017 23:59:59Arkansas NuclearNRC identifiedFailure to Follow Fire Protection Program ProceduresGreen . The inspectors identified a finding and associated non -cited violation of License Conditions 2.C.( 3)(b), Fire Protection, for Arkansas Nuclear One Unit 2, associated with the failure to adequately implement the fire protection program. Specifically, the licensee failed to follow the requirements for control of flammable liquid lockers and compressed hydrogen gas cylinders. The licensee immediately removed the hydrogen cylinders and stored them in an approved location and began processing the flammable liquid lockers through the design change process. The licensee entered these issues into their corrective action program as Condition Reports CR -ANO -2-2017- 01525 and CR -ANO -C-2017 -01508 . The failure to properly control transient combustible material in accordance with the approved fire protection program was a performance deficiency. The finding was considered more than minor because storing unanalyzed flammable material could result in the potential to exceed combustible material limits , and is associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to follow procedures resulted in conditions that increased the risk of fire which could upset plant stability and challenge critical safety functions. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and assigned the finding to the Fire Prevention and Administrative Controls category; because it affected the licensees combustible materials control. The finding was determined to be Green, or very low safety significance, in accordance with Inspection Manual Chapter 0609, Appendix F, Question 1.3.1, because the reactor would have been able to reach and maintain safe shutdown since the postulated fires would not have affected both trains of safe shutdown equipment . This finding had a cross -cutting aspect associated with teamwork within the human performance area since multiple groups in the licensee staff were involved in the decisions that resulted in the improper introduction of the flammable liquids lockers and the improper storage of the hydrogen cylinders (H.4).
05000313/FIN-2017008-0130 June 2017 23:59:59Arkansas NuclearNRC identifiedInadequate FLEX Power Supply ConnectionsGreen. The team identified a finding for the fail ure to assure that FLEX power supply connections would be reliable following all required postulated beyond design basis external events . Specifically, the team identified that one installed cable configuration could potentially be damaged during high wind events preventing operation of the portable diesel generator required to operate plant equipment. This issue was entered into the licensees corrective action program as Condition Report CR- ANO -C-2017- 00316. The failure to adequately ins tall the electrical modification for connecting the portable diesel generator was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating S ystems Cornerstone and adversely affected the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of the find ing was evaluated using NRC Inspection Manual Chapter 0609, Appendix O, Significance Determination Process for Mitigating Strategies and Spent Fuel Pool Instrumentation (Orders EA -12- 049 and EA -12-051), dated October 7, 2016, and Appendix M , Significance Determination Process Using Qualitative Criteria, dated April 12, 2012. A bounding evaluation was performed using the exposure time, tornado frequency, and frequency of a random failure of both emergency diesel generators. The licensees compliance date with the order was January 12, 2016, so an exposure time of one year was used. The tornado frequency selected was for an F2 or greater tornado striking the site (5.31E -5/year). The random failure frequency of both units emergency diesel generators (3.15E -3/year) was selected since the emergency diesel generators are protected from damage during high wind events. This is a conservative bounding analysis because it assumes that any tornado would result in damage causing a loss of offsite p ower and damage the cables in terminal panel 2TB1011 on the roof. The change in core damage frequency for the finding was determined to be 1.67E -7/year. Therefore, the finding was determined to a very low risk significance . The findi ng had a cross-cutti ng aspect in the challenge to the unknown co mponent of Human Performance becau se the lice nsee failed to adequately address all potential damage scenarios when developing the modification design requirements for beyond design basis external events (H.11)
05000368/FIN-2017002-0230 June 2017 23:59:59Arkansas NuclearSelf-revealingFailure to Install Set Screw Leads to Breaker FailureGreen . The inspectors documented a Green self -revealing finding and associated non- cited violation of Unit 2 Technical Specification 6.4.1.a, for failure to properly pre-plan and perform maintenance on the Unit 2 containment spray pump B breaker in accordance with written procedures. Specifically, the licensee failed to install a cam shaft set screw during the breakers last overhaul. The cam eventually became displaced on the shaft, and the breaker failed to close. To correct the issue, the licensee replaced the breaker and installed a cam shaft set screw in the failed breaker. The licensee also inspected all other similar breakers to verify the cams were properly secured. The licensee entered the issue in to their corrective action program as Condition Report CR -ANO -2-2017- 03168. The failure to install a cam shaft set screw during the overhaul of the Unit 2 containment spray pump B breaker is a performance deficiency. The performance deficiency is more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the failure of a Unit 2 containment spray pump breaker. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was not a design or qualification deficiency; did not represent a loss of system; did not result in the actual loss of function of a train of technical specification equipment for greater than its allowed outage time; and did not screen as potentially risk significant due to seismic, flooding, or severe weather events. The inspectors determined this finding did not have a cross -cutting aspect because the most significant contributor did not reflect current licensee performance. Specifically, the error occurred during the breakers last overhaul, which occurred in 2011
05000313/FIN-2017001-0531 March 2017 23:59:59Arkansas NuclearLicensee-identifiedLicensee-Identified ViolationOn January 16, 2017, Unit 1 operators noticed reduced pressure and flow from service water pump C while placing it in service. The licensee declared the pump inoperable, found and removed approximately 10 feet of 12-inch polymer tube that was obstructing the suction path of the pump, and completed a successful test and inspection of the pump before returning it to service. The licensee determined that the hose was inadvertently introduced while the service water bay was open for maintenance during the fall 2016 Unit 1 refueling outage. The inspectors reviewed the licensees evaluation of pump functionality and concluded that the pump could produce enough flow and pressure to fulfill its safety function, and that the pump could withstand fully ingesting the hose without significant damage to the pump or system. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings of a type appropriate to the circumstances. Licensee Procedure EN-MA-118, Foreign Material Exclusion, Revision 10, an Appendix B quality-related procedure, provides instructions for controlling foreign material, an activity affecting quality. Procedure EN-MA-118, Step 5.4, requires, in part, that only necessary material be allowed in the foreign material exclusion zone. Contrary to the above, between September 14, and November 25, 2016, the licensee failed to only allow necessary material in the foreign material exclusion zone. Specifically, when the Unit 1 service water pump C bay was open for maintenance, a hose was unnecessarily introduced and then left in the bay after the maintenance. The licensee documented the issue in the licensees corrective action program as Condition Report CR-ANO-1-2017-00164. To correct the issue, the licensee removed the hose, inspected and tested the pump, and inspected all other potentially affected service water bays to verify no foreign material was present. Using NRC Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012, the inspectors determined the finding to be of very low safety significance (Green) because the degraded pump would still be able to perform its safety function, despite the flow capability reduction.
05000313/FIN-2017001-0331 March 2017 23:59:59Arkansas NuclearSelf-revealingInadvertent Reactivity AdditionGreen. Inspectors documented a Green self-revealing finding and associated non-cited violation of Unit 1 Technical Specification 5.4.1.a. Specifically, the licensee failed to properly pre-plan and perform maintenance of the integrated control system equipment that can affect the performance of safety-related equipment. The licensee failed to plan and perform post-maintenance testing on newly installed integrated control system cards before returning the system to service. As a result, the licensee failed to detect a failed card. When the associated controller was placed into automatic mode, the system responded to a false demand signal that resulted in an inadvertent rod withdrawal that required prompt operator action to terminate the power increase and restore power to the original level. To correct the failed card, the licensee installed a new card that had been tested and validated prior to installation. The licensee documented this issue in Condition Report CR-ANO-1-2016-05551. Inspectors concluded that the failure to perform a post-maintenance test prior to placing a component in service is a performance deficiency. Specifically, the work order for replacing the steam generator reactor demand circuit card did not include a verification that the system was functioning properly after the replacement card was installed in the plant. The performance deficiency is more than minor because if left uncorrected, the performance deficiency has the potential to become a more significant safety concern. Specifically, if the operator had not taken prompt action to mitigate the event, it could have resulted in a more significant plant transient and could have challenged plant equipment. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, and Exhibit 1 of IMC 00609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Issued June 19, 2012, the inspectors determined the finding to be of very low safety significance (Green) because the finding is associated with the initiating events cornerstone and did not cause a reactor trip. The finding was determined to have a cross-cutting aspect in the area of human performance associated with Work Management, because the licensee did not ensure that they followed a process of planning, controlling, and executing the work activities in a formalized manner, allowing the work order to not have complete instructions for a post-maintenance test. (H.5)
05000313/FIN-2017001-0131 March 2017 23:59:59Arkansas NuclearSelf-revealingFailure to Identify Damaged LugsGreen. The inspectors documented a self-revealing finding and associated non-cited violation of Unit 1 Technical Specification 5.4.1.a, for the failure to properly perform maintenance on the Unit 1 suction valve to the emergency core cooling system B and containment spray B. Specifically, the licensee failed to identify a damaged electrical lug on the valve actuator during maintenance. The lug subsequently failed and the valve failed to stroke fully open after being returned to service. The licensee repaired the lug and restored the valve to service. The licensee documented this issue in Condition Report CR-ANO-1-2017-00270. The licensee failed to identify a damaged electrical lug on a motor-operated valve during maintenance, which is a performance deficiency. The performance deficiency is more than minor because it is associated with the human performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the failure of a suction valve for one train of emergency core cooling systems and containment spray systems after the valve was returned to service from the maintenance. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation because the finding represented an actual loss of function of a single train for greater than its technical specification allowed outage time. The analyst determined in a detailed risk evaluation that by combining internal and external event inputs yielded an estimate of the total increase in core damage frequency of 8.5E-7/year, or of very low safety significance (Green). The finding was determined to have a cross-cutting aspect in the area of human performance associated with Avoid Complacency because the primary cause of the performance deficiency involved the failure to plan for the possibility of mistakes and use appropriate error reduction tools. (H.12)
05000313/FIN-2017001-0231 March 2017 23:59:59Arkansas NuclearNRC identifiedFailure to Evaluate All Required Functions for OperabilityGreen. The inspectors identified a finding and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to evaluate the impact of all the required safety functions for operability when the valve failed to fully open during a valid demand. Specifically, the licensee failed to evaluate the operability impact on the safety function to close for the Unit 1 motor-operated borated water storage tank outlet valve CV-1408 before de-energizing and locking open the valve and declaring it operable. After the inspectors questioned this decision, the licensee declared the valve inoperable and repaired the valve operator. The licensee documented this issue in Condition Report CR-ANO-1-2017-00324. The failure to evaluate the operability impact of all required safety functions for Unit 1 motor-operated valve, CV-1408, before de-energizing and locking open the valve is a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by locking the valve open, the licensee prevented Train B of the emergency core cooling system from being able to be remotely isolated from the borated water storage tank during the containment recirculation phase of a potential loss of coolant accident, which could have allowed air binding of the pumps. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was not a design or qualification deficiency; did not represent a loss of system; did not result in the actual loss of function of a train of technical specification equipment for greater than its allowed outage time; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The inspectors determined that this finding has a cross cutting aspect in the human performance area of Consistent Process, because the performance deficiency was caused by not following a consistent, systematic approach to making a decision concerning operability of the affected train. (H.13)
05000313/FIN-2017001-0431 March 2017 23:59:59Arkansas NuclearLicensee-identifiedLicensee-Identified ViolationThe licensee identified that four seal injection check valves to the Unit 1 reactor coolant pumps (RCPs), which functioned as containment isolation valves, were missing internal springs required per original design. Due to the vertical orientation of the valves, the valves needed these springs to ensure that the valve disc would seat properly during reverse flow. The licensee also identified they had failed to test these ASME Code Class C check valves close safety function in accordance with ASME Code for Operation and Maintenance of Nuclear Power Plants (OM) Code. The licensee had been testing the close function by manually closing the check valves with their handwheels. Title 10 CFR Part 50.55a.(f)(4)(ii), requires in part, that ASME Code Class 3 pumps and valves must meet the inservice test requirements of ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). The 2003 Addenda to the 2001 ASME OM Code, Subsection ISTC, Inservice Testing of Valves in Light-Water Reactor Nuclear Power Plants, Section ISTC-5220, Check Valves, Subsection ISTC-5221, Valve Obturator Movement, Paragraph (a)(1), states in part, that check valves shall be exercised by verifying that on cessation or reversal of flow, the obturator has traveled to the seat. Contrary to the above, prior to November 29, 2016, the inservice tests to verify operational readiness of RCP seal injection check valves did not comply with the applicable version of the ASME OM Code requirement to exercise check valves by verifying that on cessation or reversal of flow, the obturator has traveled to the seat. Specifically, the licensee was manually closing these stop check valves in accordance with their test procedure to satisfy inservice testing. The licensee immediately installed springs for these valves as required and wrote a test procedure to test these valves in accordance with ASME OM Code. The licensee documented the issue in their corrective action program as Condition Report CR-ANO-2016-05149. Using NRC Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012, the inspectors determined the finding to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system and heat removal components.
05000313/FIN-2016008-0631 December 2016 23:59:59Arkansas NuclearNRC identifiedReadiness to Cope with External FloodingGreen. The team identified three examples of a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part that, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances. Specifically, prior to December 2, 2016, Unit 1 Operating Procedure OP 1203.025, Natural Emergencies, Revision 60 and Unit 2 Operating Procedure OP 2203.008 Natural Emergencies, Revision 42 failed to ensure all actions required to establish external flood protection, as specified by flood protection design basis engineering report CALC-ANOC-CS-00003, Revision 00 were implemented. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-2-2016-4265. The licensees failure to prescribe procedures appropriate to the circumstances for combating emergencies or other significant acts of nature such as flooding was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences, and would have the potential to lead to a more significant safety concern. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it does not involve the loss or degradation of equipment or function specifically designed to mitigate 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 identification because the licensee failed to identify issues, completely, accurately, and in a timely manner in accordance with the corrective action program. Specifically, the licensee failed to identify these deficiencies during a review of these same procedures as part of actions to close significant performance deficiencies as documented in Arkansas Nuclear One Area Action Plan FP-6 (P.1).
05000313/FIN-2016004-0331 December 2016 23:59:59Arkansas NuclearLicensee-identifiedLicensee-Identified ViolationThe licensee identified that the Unit 1 emergency diesel generator governors were left in droop mode at all times, so that during a loss of offsite power the speed and frequency of the EDGs would decrease as loading increased and cause a reduction in speed and capability from safety-related motors. The licensee determined that some EDG-powered safety-related motors would not have been capable of providing the required flow rate for a short period of time, but this did not prevent them from performing their safety function. Title 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, & Drawings, states, in part, that activities affecting quality shall be prescribed by procedures of a type appropriate to the circumstance. Contrary to the above, as of November 2, 2016, the procedure for Unit 1 EDG operations, an activity affecting quality, was not appropriate to the circumstance. Specifically, Procedure OP-1104.036, Emergency Diesel Generator Operation, Revision 74, did not state to set the speed droop settings for both A and B EDGs to zero when not load sharing with another power source and did not specify this as a requirement for the EDGs when in an emergency standby condition. The licensee immediately set the speed droop settings for both EDGs to zero and changed the procedure. The licensee documented the issue in their corrective action program as Condition Report CR-ANO-1-2016-04333. Using NRC Inspection Manual Chapter 0609, Appendix A, "The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012, the inspectors determined the finding to be of very low safety significance (Green) because the deficiency did not result in a loss of a safety function.
05000313/FIN-2016008-0531 December 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Ensure Safety Systems Would Survive Sustained Degraded Voltage ConditionsGreen. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, from December 17, 1979, to December 2, 2016, the licensee did not verify that the design of the protective devices for the loads required at the beginning of a loss-of-coolant accident were adequate to prevent tripping these devices under degraded voltage conditions, which would render the affected loads non-functional. In response to this issue, the licensee performed a preliminary analysis to determine that the protective overload devices would not cause safety equipment to fail at degraded voltages allowed by technical specifications. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-5027 and CR-ANO-C-2016-5191. The team determined that the failure to ensure that safety-related electrical components would not fail during the allowable time duration of a degraded voltage condition (in accordance with NRC Multi-Plant Action B-23, Position 1.C) was a performance deficiency. The finding was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that the protective devices for the loads required at the beginning of a Loss of Control Accident would not fail under degraded voltage conditions did not ensure that these loads would be available to perform their mitigating functions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000368/FIN-2016011-0131 December 2016 23:59:59Arkansas NuclearSelf-revealingFailure to Ensure Adequate Lubication for Emergency Diesel Generator BearingThe inspectors reviewed a self-revealing finding that was preliminarily determined to have low to moderate safety significance (White) for the failure to perform maintenance activities in a manner that ensured adequate lubrication to Unit 2 emergency diesel generator A. This finding involved a violation of Technical Specification 6.4.1.a, because the licensee failed to provide adequate work instructions for maintenance on the inboard generator bearing oil sight glass to ensure that the scribe mark indicated the minimum acceptable oil level to ensure adequate lubrication to the bearing. As a result, the licensee reinstalled the sight glass with the oil level scribe mark below the bottom of the bearing rollers. Subsequently, on June 22, 2016, the oil was drained and replaced with oil level close to the sight glass scribe mark, and the bearing failed on September 16, 2016, during a 24-hour surveillance. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-2-2016-03307. The licensee resolved the safety concern by repairing the bearing, successfully testing the diesel, and verifying the condition did not exist in any other safety-related equipment. The failure to ensure adequate lubrication to the inboard generator bearing so that the Unit 2 emergency diesel generator A would be capable of performing its safety functions for the intended mission time is a performance deficiency. This performance deficiency is more than minor, and therefore is a finding, because it is associated with 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 licensee failed to properly pre-plan and perform work that could affect this safety-related system in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances such that the minimum bearing oil level was correctly marked and maintained. This performance deficiency subsequently affected the availability and reliability of the emergency diesel generator, a mitigating system. The inspectors evaluated the finding with NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding required a detailed risk evaluation because an actual loss of function of a single train of mitigating equipment occurred for greater than its technical specification allowed outage time. As determined by a Significance and Enforcement Review Panel (SERP), the total increase in core damage frequency for the performance deficiency was preliminarily estimated to be between 3.0E-6 per year and 9.6E-6 per year, or of low to moderate safety significance. The inspectors determined this finding has a cross-cutting aspect in the human performance area of Work Management, because the primary cause of the performance deficiency involved the failure to plan, control, and execute work activities such that nuclear safety is the overriding priority (H.5).
05000313/FIN-2016004-0431 December 2016 23:59:59Arkansas NuclearLicensee-identifiedLicensee-Identified ViolationDuring the fall 2016 Unit 1 refueling outage, the licensee foreign object search and retrieval (FOSAR) inspections in the steam generator bowls and reactor vessel identified a number of foreign objects, including an 8-inch metal rod. Discussions with the licensee indicated that some of the debris constituted foreign material that should have been prevented from being introduced into the RCS by the foreign material exclusion program. The inspectors concluded that the foreign material was most likely introduced during the previous refueling outage. During the prior operating cycle, the licensees chemistry sampling identified increased RCS activity, and subsequent fuel bundle examinations of fuel removed from the core identified wear marks through the cladding of two adjacent fuel pins. The fuel assembly with the damage was not placed back into the RCS. Since there was no evidence of broken components inside the RCS, the licensee concluded that the most likely cause was the introduction of foreign material. While it was not possible to determine whether any of the foreign material had actually caused the fuel damage, the inspectors concluded that the licensee had failed to control foreign material and prevent it from entering the RCS. Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings of a type appropriate to the circumstances. Licensee Procedure EN-MA-118, Foreign Material Exclusion, Revision 10, an Appendix B quality-related procedure, provides instructions for controlling foreign material. Procedure EN-MA-118, Step 5.5, requires, in part, that all material and tools that were introduced to the FME zone are accounted for. Contrary to the above, between January 25, and March 1, 2015, the licensee failed to ensure that all material and tools that were introduced to the FME zone were accounted for. Specifically, the licensee failed to maintain adequate FME control, leading to two damaged cladding pins and slightly elevated dose rates in the RCS piping, as well as another piece of metallic FME in the vessel, as documented in CR-ANO-1-2016-03340. This issue was documented in the licensees corrective action program under CR-ANO-1-2016-03521. Corrective actions taken include a search for the foreign material and permanent removal of the fuel assembly from the core. Prior to 2012, the NRCs Significance Determination Process in IMC 0609, Attachment 4, Phase 1 - Initial Screening and Characterization of Findings, contained guidance to screen all more than minor performance deficiencies affecting fuel barriers to very low safety significance. The inspection manual chapters were restructured in 2012, and the screening was inadvertently omitted, though the NRC was in the process of reinstating that same guidance. Therefore, after consultation with the Office of Nuclear Reactor Regulation, the inspectors determined that this finding is of very low safety significance (Green).
05000313/FIN-2016404-0131 December 2016 23:59:59Arkansas NuclearNRC identifiedSecurity
05000313/FIN-2016008-0231 December 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Incorporate NRC Safety Guide 9 Criteria into Surveillance ProceduresGreen. The team identified Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Additionally, Test results shall be documented and evaluated to assure that test requirements have been satisfied. Specifically, as of December 2, 2016, Units 1 and 2 emergency diesel generator surveillance procedures failed to incorporate the applicable voltage and frequency limits of NRC Safety Guide 9, and did not consistently document or evaluate results to assure test requirements have been satisfied. In response to this issue, the licensee provided the team test results which demonstrated that an immediate safety concern was not present. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-1-2016-4785 and CR-ANO-2-2016-4257. The team determined that the failure to incorporate the acceptance limits of NRC Safety Guide 9 into surveillance test procedures for emergency diesel generators and assure that test requirements have been satisfied in accordance with 10 CFR Part 50, Appendix B, Criterion XI, Test Control, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences, and would have the potential to lead to a more significant safety concern. Specifically, the failure to incorporate appropriate acceptance criteria in test procedures and assure that the criteria have been satisfied had the potential to lead to a worse condition, if left uncorrected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016404-0231 December 2016 23:59:59Arkansas NuclearNRC identifiedSecurity
05000313/FIN-2016004-0131 December 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Pre-plan Walkdown to Avoid Impacting Safety BusGreen. The inspectors documented a self-revealed finding and associated non-cited violation of Unit 1 Technical Specification 5.4.1.a, for the failure to properly pre-plan and perform a pre-modification walkdown in the Unit 1 train A safety-related switchgear room so that the walkdown would not adversely affect the performance of train. As a result, licensee personnel inadvertently de-energized the A3 switchgear and associated ac buses, which resulted in the loss of one train of spent fuel pool cooling. Operators restored spent fuel pool cooling, the licensee evaluated the human error and performed a training stand-down to ensure pre-job walkdowns did not impact plant equipment. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-1-2016-04356. The failure to perform a plant walkdown in a manner that did not impact safety-related switchgear is a performance deficiency. The performance deficiency is more than minor because it adversely affected 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 caused by accidents or events. Specifically, de-energizing the safety-related switchgear resulted in the loss of one train of spent fuel pool cooling and an increase in risk level from Green to Yellow. The inspectors evaluated the finding with NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 3, Barrier Integrity Screening Questions, because the appendix provides the most applicable guidance, regardless of whether the unit was at-power or shutdown. The inspectors determined that the finding screened as having very low safety significance (Green) because the finding did not cause the spent fuel pool to exceed the maximum analyzed temperature, did not damage fuel cladding, did not result in a loss pool water inventory below the minimum analyzed level, and did not affect the pool neutron absorber or soluble boron concentration. The inspectors determined this finding has a cross-cutting aspect in the human performance area of Avoid Complacency, because the primary cause of the performance deficiency involved the failure to plan for the possibility of mistakes and use appropriate error reduction tools. (H.12)
05000313/FIN-2016008-0331 December 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Monitor Startup Transformers 1, 2, and 3 Voltage Regulator/Tap Changer FunctionGreen. The team identified a Green finding for the failure to meet the surveillance standards of IEEE 308-1971, Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations, Section 5.2.3, Preferred Power Supply. Specifically, from 2001 to December 2, 2016, the licensee failed to monitor the operation of the voltage regulator/load tap changer functions on startup transformers 1, 2, and 3. In response to this issue, the licensee provided reasonable assurance that the voltage regulator/load tap changer was operating properly based on review of plant computer voltage plot data following an Arkansas Nuclear One, Unit 1 trip that occurred on December 14, 2015. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-4777, CR-ANO-C-2016-4879, and CR-ANO-C-2016-5015. The team determined that the failure to monitor startup transformers 1, 2, and 3 voltage regulator/load tap changers to the extent that they are shown to be ready to perform their intended function, in accordance with IEEE Standard 308-1971, was a performance deficiency. The finding was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to monitor the adequacy of the voltage supplied from startup transformers 1, 2, and 3 voltage regulator/load tap changer did not ensure that offsite power would be available to perform its necessary functions to provide power to the safety-related mitigation equipment. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016008-0131 December 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Verify the Adequacy of Motor Operated Valve Thermal Overload DevicesGreen. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to December 2, 2016, the licensee failed to use appropriate assumptions in thermal overload device calculations and failed to establish a suitable periodic test program for safety-related Unit 1 motor operated valve thermal overload device trip setpoints, as discussed in Regulatory Guide 1.106, Regulatory Position C.2. In response to this issue, the licensee demonstrated reasonable assurance of operability by using the results of the 18-month high pressure injection system valve testing which required multiple stroking of block valves to obtain various flows without tripping the thermal overload devices. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-5017 and CR-ANO-1-2016-5130. The team determined that the failure to meet the intent of Regulatory Guide 1.106, Regulatory Position C.2 was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify the adequacy of the design and perform suitable testing for thermal overload device setpoint drift did not ensure that the safety-related motor operated valves would be available to throttle the associated system flows during a design basis accident. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluations 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 licensee failed to thoroughly evaluate Condition Report CR-ANO-1-2016-0778 which documented NRC inspector concerns associated with design and testing of motor operated valve thermal overload devices (P.2).
05000313/FIN-2016004-0231 December 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Design Pipe Support for VibrationGreen. The inspectors documented a self-revealed finding and associated non-cited violation of 10 CFR 50 Appendix B Criterion III for the licensees failure to verify that the decay heat removal (DHR) system drain piping configuration and supports could withstand vibrations created during low pressure and high flow conditions. As a result, a cracked weld and unisolable leak in the DHR system occurred due to high cycle fatigue caused by those conditions. To correct this issue, the licensee repaired the leaking weld and designed and installed a new piping support and piping configuration to reduce vibrations during the expected operating conditions. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-1-2016-03225. The failure to design the decay heat removal system piping to withstand expected vibrations from the systems cavitating venturis is a performance deficiency. The performance deficiency is more than minor because it was associated with the design control attribute of the initiating events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, inadequate design of the DHR system piping support resulted in a leak that could have challenged the capability of both trains of the DHR system during shutdown on September 29, 2016. The inspectors performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," issued October 7, 2016, and were directed to IMC 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase 1 Screening and Characterization of Findings, since the finding pertained to a degraded condition while the plant was shutdown. Using IMC 0609, Appendix G, Attachment 1, dated May 9, 2014, the inspectors determined that the finding required a Phase 2 evaluation. A senior reactor analyst performed a Phase 2 evaluation in accordance with IMC 0609, Appendix G, Attachment 2, Phase 2 Significance Determination Process Template for PWR during Shutdown, dated February 28, 2005. The senior reactor analyst performed a Phase 2 evaluation which used realistic break characteristics and plant configuration changes to determine the significance to be of very low safety significance (Green). The inspectors determined this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. Specifically, the licensee last reviewed and modified the pipe support configuration in 1996
05000313/FIN-2016008-0431 December 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Perform an Adequate Emergency Feedwater Pump Suction Transfer Design Calculation or Testing (EA 2017-017)Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part that, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to December 22, 2016, the licensee failed to verify the adequacy of the emergency feedwater suction transfer procedure by determining if the qualified condensate storage tank will be completely empty of water, possibly causing an air ingestion failure of the Unit 1 emergency feedwater pumps, prior to transferring to the credited safety-related alternate suction source. In response to this issue, the licensee resolved the immediate safety concern by revising the emergency feedwater pump operating procedure, removing the steps that were the cause of the concern. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-1-2016-5166, CR-ANO-1-2016-5725, and CR-ANO-1-2017-0040. The team determined that the failure to verify the adequacy of the design of the Unit 1 emergency feedwater suction from the qualified condensate storage tank to alternate sources of water by performance of design review, by use of calculational methods, or by performance of a suitable testing program in accordance with 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to have adequate measures in place to ensure an acceptable design analysis or a suitable test program would verify that the process of transferring emergency feedwater suction from the qualified storage tank to the alternate sources ensures the capability of the Unit 1 emergency feedwater system to perform its safety function. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the team determined this finding affected the secondary short term heat removal function of the Mitigating Systems Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the finding represented a loss of the emergency feedwater system and function. Therefore, a detailed risk evaluation was necessary. The senior reactor analyst determined that the change in core damage frequency of this finding was 7 x 10-7 per year, therefore the significance was of very low safety significance (Green). This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016003-0430 September 2016 23:59:59Arkansas NuclearNRC identifiedEA-16-143, Enforcement Discretion for Tornado-Generated Missile Protection NoncompliancesAppendix A to 10 CFR 50, General Design Criteria for Nuclear Power Plants, Criterion 2, Design Bases for Protection Against Natural Phenomena, states, in part, that SSCs important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes. Criterion 4, Environmental and Dynamic Effects Design Basis, states, in part, that SSCs important to safety shall be appropriately protected against dynamic effects including missiles which may result from events and conditions outside the nuclear power unit. As part of their response to external flood boundary degradation, the licensee performed a review of external hazard protection at the site, which included protection against tornado-generated missiles required by the current licensing basis for each unit. During the review, on four separate occasions, the licensee identified plant areas containing safety-related SSCs that could be susceptible to tornado missiles: Unit 1 Upper South Electrical Penetration Room Unit 1 Cable Spreading Room Unit 1 Controlled Access Area Unit 1 Vital Switchgear In each case, the licensee identified low-probability scenarios where one or more tornado-generated missiles could penetrate doors, walls, and other building features that were not fully qualified, and subsequently damage equipment that was important to safety inside the affected rooms. Details about the date of discovery, affected SSCs, condition report numbers, compensatory actions taken by the licensee, notifications made to the NRC, and affected technical specification actions for each susceptible area are listed in Attachment 3 of this report. Relevant Enforcement Discretion Policy On June 10, 2015, the NRC issued Enforcement Guidance Memorandum (EGM) 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance. (ML15111A269) The EGM referenced a bounding generic risk analysis performed by the NRC staff that concluded that tornado missile vulnerabilities pose a low risk significance to operating nuclear plants. Because of this, the EGM described the conditions under which the NRC staff may exercise enforcement discretion for noncompliances with the current licensing basis for tornado-generated missile protection. Specifically, if the licensee could not meet the technical specification required actions within the required completion time, the EGM allows the staff to exercise enforcement discretion provided the licensee implements initial compensatory measures prior to the expiration of the time allowed by the limiting condition for operation. The compensatory actions should provide additional protection such that the likelihood of tornado missile effects are lessened. The EGM then requires the licensee to implement more comprehensive compensatory measures within approximately 60 days of issue discovery. The compensatory measures must remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. In addition, the issue must be entered into the licensees corrective action program. Because EGM 15-002 listed Arkansas Nuclear One as a Group A plant, enforcement discretion will expire on June 10, 2018. However, the EGM did not provide for enforcement discretion for any related underlying technical violations; the EGM specifically requires that any associated underlying technical violations be assessed through the enforcement process. Licensee Actions For each of the examples listed above, the licensee declared the affected systems inoperable and complied with the applicable technical specification action statement(s), initiated a condition report, invoked the enforcement discretion guidance, implemented prompt compensatory measures, and returned the SSCs to an operable status. The licensee instituted compensatory measures intended to reduce the likelihood of tornado missile effects that included developing actions to be taken if a tornado watch is predicted or issued for the area to ensure the operability or restore redundant equipment during severe weather, and actions to be taken if a tornado warning is issued, including pre-staging operators in safe, strategic locations to promptly implement mitigative actions, and verifying the readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX). Other specific compensatory actions for the individual areas are listed in Attachment 3. NRC Actions The inspectors review addressed the material issues in the plant, and whether the measures were implemented in accordance with the guidance in EGM 15-002. The inspectors also evaluated whether the measures would function as intended and were properly controlled. The inspectors verified through inspection that the EGM 15-002 criteria were met in each case. Therefore, the staff determined that it was appropriate to exercise enforcement discretion and not take enforcement action for the technical specification requirements listed in Attachment 3 of this report, provided the noncompliances are resolved by June 10, 2018 (EA-16-143). The inspectors did not fully review the underlying circumstances that resulted in the technical specification violations. As stated in EGM 15-002, violations of other requirements which may have contributed to the technical specification violations will be evaluated independently of EGM implementation. The inspectors will verify restoration of compliance and assess the underlying circumstances in a follow-up inspection tracked under Licensee Event Reports 05000313/2016-002-00 and 05000313/2016-003-00, and any updates or additional licensee event reports that the licensee issues.
05000368/FIN-2016002-0330 June 2016 23:59:59Arkansas NuclearLicensee-identifiedLicensee-Identified ViolationUnit 2 Technical Specification Limiting Condition for Operation 3.3.3.1, Radiation Monitoring Instrumentation, requires that the radiation monitoring instrumentation channels shown in Table 3.3-6, Radiation Monitoring Instrumentation, shall be operable with their alarm/trip set points within the specified limits. Table 3.3-6, Item 2.a requires that the containment purge and exhaust radiation monitoring instrumentation be capable of isolating containment when process radiation equals or exceeds two times the background radiation rate. Contrary to the above, on October 26, 2015, the licensee failed to ensure that the required containment purge and exhaust radiation monitor remained operable to isolate containment when process radiation equals or exceeds two times the background radiation rate. Specifically, the licensee failed to restart the containment purge and exhaust isolation radiation monitor sample pump, which supplies process sample flow to the radiation monitor, following an electrical bus transfer which removed power to the sample pump. As a result, the containment ventilation system would not have automatically isolated to prevent a release of radioactive material in the event of a fuel handling accident. However, operators could manually isolate the ventilation system if a fuel accident occurred. An operator restarted the process sample pump and documented the issue in Condition Report CR-ANO-2-2015-04197. Because the finding degraded the ability to close or isolate the containment, NRC Inspection Manual Chapter 0609 Appendix G, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, directed the inspectors to use NRC Inspection Manual Chapter 0609 Appendix H, Containment Integrity Significance Determination Process, dated May 6, 2004. The inspectors classified the finding as a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage (Type B). Using the Phase 1 screening for Type B findings, the inspectors determined the finding to be of very low safety significance or Green, because the deficiency did not occur within eight days of the start of the refueling outage.
05000368/FIN-2016009-0430 June 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Adequately Establish Ignition Frequencies for the Risk-Informed Fire Protection ProgramThe team identified two examples of a Severity Level IV non-cited violation of License Condition 2.C(3)(b), Fire Protection, for the licensees failure to properly implement their risk-informed fire protection program and accurately capture component ignition frequencies in their fire probabilistic risk assessment. Specifically, the component ignition frequencies for air compressors and ventilation equipment were found to be lower than expected because the licensee misapplied the guidance in NUREG/CR-6850. The licensee entered this issue into their corrective action program as Condition Reports CR ANO-C-2016-2600, CR-ANO-C-2016-2528, and CR-ANO-2-2016-02356, with the intent to perform an extent of condition relating to other potential components that are misclassified in the fire ignition frequency analysis, correct the fire Ignition frequency report and the associated Fire Probabilistic Risk Assessment model calculations to incorporate the correct ignition frequency and the appropriate scenarios. The licensees failure to adequately implement the prescribed guidance in NUREG/CR-6850 to estimate the ignition frequencies for their risk-informed fire protection program was a performance deficiency. The performance deficiency was minor because the answer to all the IMC 0612, Appendix B, more than minor questions was No. The team determined that this issue was a traditional enforcement violation because it impacted the regulatory process when the only NRC-approved framework for conversion to NFPA 805 was not fully followed. The NRC determined that this violation was associated with a minor performance deficiency. The team determined that this violation was a Severity Level IV in the traditional enforcement process when comparing it to the violation examples in Section 6.1, Reactor Operations, of the NRC Enforcement Policy, specifically noting it was similar to Example 6.1.d.4 for failing to adequately assess the baseline risk of plant operations associated with implementation of a risk-informed program (NFPA 805) such that the program was implemented inappropriately. The finding did not have a cross-cutting aspect because traditional enforcement violations are not assessed for cross-cutting aspects.
05000368/FIN-2016002-0130 June 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Incorporate Vendor Guidance in Work OrderThe inspectors identified a finding for the failure to incorporate vendor instructions in a work order. Specifically, the licensee exceeded the vendor specified torque values and performed the work with the component in service, contrary to vendor cautions, breaking the glass, wetting the auxiliary feedwater pump, and necessitating the unplanned shutdown of the main feedwater pump. The licensee replaced the ruptured sight glass and repaired and tested the wetted components. The licensee documented the issue in Condition Report CR-ANO-2-2015-04832. The failure to incorporate vendor instructions in a work order is a performance deficiency. The finding is more than minor because it adversely affected the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the performance deficiency resulted in the Unit 2 auxiliary feedwater pump and main feedwater pump B being rendered unavailable. The inspectors evaluated the finding with NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding required a detailed risk evaluation because the finding involved an actual loss of function of auxiliary feedwater and one train of main feedwater, designated as having high safety significance in accordance with the licensees maintenance rule program, for greater than 24 hours. A senior reactor analyst performed a detailed risk evaluation and determined that the increase in core damage frequency was 1.3E-7/year (Green). The analyst assumed that all feedwater pumps were available until the time of the leak and that any increase in core damage frequency resulted from the unavailability of the pumps after the leak. The emergency feedwater system remained available to mitigate the increase in core damage frequency of this finding. The inspectors determined this finding has a cross-cutting aspect in the human performance area of Work Management because the primary cause of the performance deficiency involved the failure to identify and manage risk commensurate to the work and the need for coordination with different groups or job activities (Section 1R12). (H.5)
05000313/FIN-2016009-0130 June 2016 23:59:59Arkansas NuclearNRC identifiedInadequate Loop Flow TestingThe team identified a non-cited violation of License Conditions 2.C.(8), Fire Protection, for Unit 1; License Condition 2.C.(3)(b), Fire Protection, for Unit 2; and the technical requirements manuals because the licensee did not properly test all portions of the underground fire piping. Specifically, the licensee did not determine the flow rates through two headers that provided water to the ring header supplying the Unit 2 auxiliary building as designed. The licensee entered this violation into their corrective action program as Condition Report CR-ANO-C-2016-02613 and initiated actions to conduct a flow test of the headers. The failure to implement an adequate procedure to test underground fire piping was a performance deficiency. Specifically, the licensee did not test two headers included and designed as part of their underground fire piping to demonstrate that no faults had occurred. This performance deficiency was more than minor because it was associated with the protection against external factors attribute (fire) and adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to test two underground fire piping headers failed to demonstrate the capability to deliver adequate flow and pressure to the fire suppression systems as designed. The finding was screened in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012. Because the finding affected fixed fire protection systems or the ability to confine a fire, the team reviewed the finding using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013. The finding was screened as a Green finding of very low safety significance in accordance with Task 1.4.7, Fire Water Supply, Question A. Although the licensee failed to test all portions of the underground fire piping in accordance with their license and technical requirements manual, the team determined that at least 50 percent of required fire water capacity would be available based on the testing that is done. As a result, the finding was determined to be of very low safety significance (Green). The team determined that this finding did not have a cross-cutting aspect since it did not reflect current performance. Specifically, the licensee had not flow tested all underground fire piping headers since initial installation.
05000368/FIN-2016009-0230 June 2016 23:59:59Arkansas NuclearNRC identifiedInadequate Procedure Used as a Compensatory MeasureThe team identified a non-cited violation of License Condition 2.C.(3)(b), Fire Protection, for use of an inadequate procedure as a compensatory measure. Specifically, a procedure for providing temporary cooling to the safety parameter display system room when the normal room cooler is unavailable did not adequately address the impact of the temporary configuration on the ability to maintain safe and stable plant conditions for fires that require shutdown from outside the control room. The temporary room cooler did not have a power supply assured to remain available during a shutdown from outside the control room. The licensee entered this violation into their corrective action program as Condition Reports CR-ANO-2-2016-02143 and CR-ANO-C-2016-02638. In response to this issue, the licensee developed a thermal analysis of the safety parameter display system room temperature during this scenario and confirmed that the maximum room temperature would not challenge the operation to the safety parameter display system. The failure to provide an adequate procedure for use as a compensatory measure was a performance deficiency. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of procedure quality and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events (fire) to prevent undesirable consequences. Specifically, loss of cooling to the safety parameter display system room could adversely affect the availability, reliability, and capability of the safety parameter display system which is required to respond to a fire resulting in the evacuation of the Unit 2 control room. A senior reactor analyst performed a detailed risk evaluation of this finding because IMC 0609, Appendix F, does not include explicit treatment of fires in the control room. An evaluation of the survivability of the safety parameter display system compared to the best estimate of the heatup of the room housing its equipment demonstrated that the safety parameter display system would survive with high probability until the plant reached a safe and stable condition for the postulated fires. As a result, the finding was determined to be of very low safety significance (Green). This finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago.
05000368/FIN-2016009-0330 June 2016 23:59:59Arkansas NuclearNRC identifiedFailure to Ensure that the Assumptions in the Engineering Analysis Remain ValidThe team identified a non-cited violation of License Condition 2.C(3)(b), Fire Protection, for the failure to establish an appropriate monitoring program in accordance with National Fire Protection Association Standard 805, Section 2.6. Specifically, the licensee failed to set the action level for the availability of some plant components to ensure that the assumptions in the engineering analysis remained valid and also failed to establish a monitoring plan for the concurrent unavailability of one set of two components. The licensee entered the issues into their corrective action program as Condition Reports CR-ANO-2-2016-02355 and was in the process of developing corrective actions to address the monitoring of the components and work with industry organizations and Office of Nuclear Reactor Regulation to determine long-term resolution. The failure to adequately monitor unavailability of the plant components to ensure that the assumptions in the engineering analysis remained valid was a performance deficiency. The performance deficiency was more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the performance deficiency could adversely affect the acceptable level of availability of the components which are used to respond to fire initiating events, in that the action levels for availability in the monitoring program were greater than the assumptions in the fire probabilistic risk assessment. The finding was screened in accordance with Inspection IMC 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, dated June 19, 2012. Because the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire, the team reviewed the finding using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, dated September 20, 2013. The finding was screened as a Green finding of very low safety significance in accordance with Step 1.3, Ability to Achieve Safe Shutdown, B Question. Based on the criteria in Appendix F, Attachment 2, Degradation Rating Guidance Specific to Various Fire Protection Program Elements, dated February 28, 2005, the finding was assigned a low degradation rating. Using Table A2.3, the inspectors assigned the low degradation rating because the issue involved monitoring of components that did not appreciably degrade below acceptable levels during the exposure period. This finding had a cross-cutting aspect associated with change management within the human performance area since leaders did not use a systematic process (e.g., assigning an overall owner) for evaluating and implementing change during the development of the monitoring program for the fire probabilistic risk assessment model for Unit 2 (H.3).