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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000313/FIN-2018003-06 | 2018Q3 | Arkansas Nuclear | Reactor Power Transient Caused by the Turbine Bypass Valve Failing Open | The 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-2018003-05 | 2018Q3 | Arkansas Nuclear | Failure to Maintain Main Feedwater Pump B Discharge Pressure in Band Caused a Reactor Trip | The inspectors reviewed a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to implement Procedure OP-1102.002, Plant Startup, Revision 106. Specifically, control room operators failed to maintain main feedwater pump discharge pressure in the required band to control flow to the steam generators during a plant startup. As a result, the only operating main feedwater pump tripped on high discharge pressure, causing an automatic reactor trip. |
| 05000313/FIN-2018003-04 | 2018Q3 | Arkansas Nuclear | Failure to Verify Safety-Related 4160 V Breaker Operability Following Maintenance Activities | The inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to perform post-maintenance testing to demonstrate component operability for the train A safety-related 4160 V switchgear A-303 breaker that provides power to the swing service water pump B (P-4B) after the breaker was racked in. The breaker subsequently failed to close when attempting to start the pump. |
| 05000313/FIN-2018003-03 | 2018Q3 | Arkansas Nuclear | Failure to Provide Complete and Accurate Information in a License Amendment Request to Change Emergency Action Level Requirements | The 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. |
| 05000298/FIN-2018003-03 | 2018Q3 | Cooper | Failure to Provide Adequate Lubrication for Drywell Fan Coil Units | The inspectors reviewed a self-revealed finding for the licensees failure to implement Work Order 5060136 during maintenance on the drywell fan coil units. Specifically, on October 26, 2016, during bearing replacement work on drywell fan coil, unit D, maintenance personnel failed to properly reinstall auto-lubricator injection connectors after removing the interferences per the work order instructions. This error resulted in the failure of drywell fan coil, unit D, due to inadequate bearing lubrication, and ultimately led to a downpower and reactor shutdown. |
| 05000298/FIN-2018003-01 | 2018Q3 | Cooper | Failure to Provide Complete and Accurate Information in a License Amendment Request | The inspectors identified that 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 measurement ranges of a liquid effluent radiation monitor used in emergency action levels that was not accurate. |
| 05000313/FIN-2018003-01 | 2018Q3 | Arkansas Nuclear | Failure to Translate the Design Requirements into Instructions for Refueling Emergency Diesel Generators | The 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-2018003-02 | 2018Q3 | Arkansas Nuclear | Failure to Implement Welding Standard Guidance and Examination Procedures | The 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. |
| 05000298/FIN-2018003-02 | 2018Q3 | Cooper | Failure to Perform Process Applicability Determination | The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow Administrative Procedure 0.9, Tagout, Revision 88, for performing a monthly audit and Process Applicability Determination. Specifically, the inspectors noted that a clearance order on the safety-related residual heat removal service water booster pump room fan coil unit was hanging for greater than 90 days with no Process Applicability Determination performed, which resulted in the power switch for the fan coil unit being unintentionally tagged out of its normal configuration for almost 2 years |
| 05000298/FIN-2018002-01 | 2018Q2 | Cooper | Failure to Maintain Alarm Procedure for Service Water Booster Pump Ventilation Manual Actions | The inspectors identified a Green non-cited violation of Technical Specification 5.4, Procedures, when the licensee failed to maintain Procedure 2.3_R-1 with the bounding time restrictions for required manual ventilation actions identified in Engineering Evaluation NEDC 92-064, Transient Temperature Rise in SWBP Room After Loss of Cooling, Revision 3C2. As a result, the licensee relied on procedure guidance that contained an incorrect, less restrictive allowance of 13 hours for completion of manual actions rather than the bounding 5.8-hour allowance described in NEDC92-064. |
| 05000298/FIN-2018002-02 | 2018Q2 | Cooper | Failure to Maintain Adequate Work Instructions for Traversing In-Core Probe System Limit Switches | A self-revealed, Green non-cited violation of Technical Specification 5.4, Procedures, was identified when the licensee failed to maintain Procedure 14.2.14, TIP Chamber Shield Maintenance, with adequate instructions for reinstalling the traversing in-core probe system in-shield limit switches. As a result, the licensee experienced multiple failures of the shield limit switches resulting in inoperable primary containment isolation valves. |
| 05000313/FIN-2018002-01 | 2018Q2 | Arkansas Nuclear | Failure to Implement Procedural Guidance to Close Spent Fuel Pool Cooler Outlet Crosstie Valve | The 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-2018001-01 | 2018Q1 | Arkansas Nuclear | Failure to Establish Adequate Criteria for Flood Seal Testing | The 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-2018001-02 | 2018Q1 | Arkansas Nuclear | Failure to Preplan and Perform Service Water Pre-Screen Maintenance | The 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-2018001-03 | 2018Q1 | Arkansas Nuclear | Licensee-Identified Violation | Title10CFR20.1501(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-2017003-01 | 2017Q3 | Arkansas Nuclear | Failure to Maintain Service Water Train Separation | The 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) |
| 05000368/FIN-2017002-02 | 2017Q2 | Arkansas Nuclear | Failure to Install Set Screw Leads to Breaker Failure | Green . 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 |
| 05000368/FIN-2017002-01 | 2017Q2 | Arkansas Nuclear | Failure to Follow Fire Protection Program Procedures | Green . 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-2017002-03 | 2017Q2 | Arkansas Nuclear | Failure to Comply with ECCS Technical Speci fications | Green . 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-2017002-04 | 2017Q2 | Arkansas Nuclear | Licensee-Identified Violation | Title 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. |
| 05000313/FIN-2017008-01 | 2017Q2 | Arkansas Nuclear | Inadequate FLEX Power Supply Connections | Green. 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) |
| 05000313/FIN-2017008-02 | 2017Q2 | Arkansas Nuclear | Inadequate FLEX Procedures | Green. 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). |
| 05000382/FIN-2017001-01 | 2017Q1 | Waterford | Failure to Perform Field Changes in Accordance with Design Control Measures | Green . The inspectors reviewed a self -revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to perform field changes in accordance with design control measures. Specifically, following maintenance on reactor coolant pump 1B , the licensee performed unauthorized field changes by not reinstalling two design supports for the differential pressure instrument line. As a result, the instrument line developed a vibration- induced fl aw, which caused an increase in reactor coolant system unidentified leakage, and consequently , an unplanned reactor shutdown. The licensee entered this condition into their corrective action program as Condition Report CR- WF3 -2016 -06698. The licensees corrective actions included replacing the damaged instrument line and installing the missing supports. The performance deficiency was more than minor , and therefore a finding, because it affected the equipment performance attribute of the Initiating Events Cornerstone and its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to reinstall the required supports on the reactor coolant pump 1B instrumentation line resulted in plant operation with increased reactor coolant system unidentified leakage, requiring an unplanned reactor shutdown to perform repairs. The inspectors screened the finding in accordance wit h NRC Inspection Manual Chapter 0609, Significance Determination Process . Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, the inspectors determined that the finding was of very low safety significance (Green) because the instrument line flaw, after a reasonable assessment of degradation, could not result in exceeding the reactor coolant system leak rate for a small loss -of-coolant accident , and could not likely affect other systems used to mitigate a loss-of-coolant accident , resulting in a total loss of their function. Because the licensees review indicated that no work had been performed in this instrument line within the last three years, and a specific date for the performance deficiency was not identified, the inspectors concluded that the finding does not reflect current licensee performance, and therefore, did not assign a cross -cutting aspect . |
| 05000313/FIN-2017001-01 | 2017Q1 | Arkansas Nuclear | Failure to Identify Damaged Lugs | Green. 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-2017001-02 | 2017Q1 | Arkansas Nuclear | Failure to Evaluate All Required Functions for Operability | Green. 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-2017001-04 | 2017Q1 | Arkansas Nuclear | Licensee-Identified Violation | The 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-2017001-05 | 2017Q1 | Arkansas Nuclear | Licensee-Identified Violation | On 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-2017001-03 | 2017Q1 | Arkansas Nuclear | Inadvertent Reactivity Addition | Green. 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-2016004-03 | 2016Q4 | Arkansas Nuclear | Licensee-Identified Violation | The 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-2016004-01 | 2016Q4 | Arkansas Nuclear | Failure to Pre-plan Walkdown to Avoid Impacting Safety Bus | Green. 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-2016004-02 | 2016Q4 | Arkansas Nuclear | Failure to Design Pipe Support for Vibration | Green. 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 |
| 05000368/FIN-2016011-01 | 2016Q4 | Arkansas Nuclear | Failure to Ensure Adequate Lubication for Emergency Diesel Generator Bearing | The 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-2016004-04 | 2016Q4 | Arkansas Nuclear | Licensee-Identified Violation | During 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-2016003-04 | 2016Q3 | Arkansas Nuclear | EA-16-143, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliances | Appendix 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-2016002-01 | 2016Q2 | Arkansas Nuclear | Failure to Incorporate Vendor Guidance in Work Order | The 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) |
| 05000368/FIN-2016002-02 | 2016Q2 | Arkansas Nuclear | Failure to Clean Main Feedwater Lube Oil Reservoir Leads to Reactor Power Reduction | The inspectors documented a self-revealing finding for failure to clean the main feedwater turbine lube oil reservoir. Specifically, the main feedwater turbine lube oil reservoir had not been cleaned since 2006, causing clogged filters and low main feedwater turbine bearing oil pressure on February 5, 2016. The licensee entered this finding into their corrective action program as Condition Report CR-ANO-2-2016-00470 and implemented the necessary preventive maintenance. The failure to perform preventive maintenance to ensure cleanliness on the main feedwater pump turbine bearing oil reservoir as required by the preventive maintenance program is a performance deficiency. The performance deficiency is more than minor because it impacted the equipment performance attribute and adversely affected the initiating events cornerstone objective to limit the likelihood of events that upset plant stability and challenged critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency resulted in operators lowering reactor power and rendered a main feedwater pump unavailable. Using NRC Inspection Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors screened the finding as having very low safety significance because the finding affected a transient initiator but did not result in a reactor trip. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. Specifically, the maintenance strategy changed in 2009 |
| 05000368/FIN-2016002-03 | 2016Q2 | Arkansas Nuclear | Licensee-Identified Violation | Unit 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-2016001-04 | 2016Q1 | Arkansas Nuclear | Failure to Inject Service Water with Corrosion Inhibitors during Refueling Outages | The inspectors reviewed a self-revealing Green finding and an associated non-cited violation for the failure to follow Procedure OP-1052.007, Secondary Chemistry Monitoring, Revision 040. Specifically, the licensee failed to inject corrosion inhibiting chemicals into Unit 2 service water during refueling outages, which resulted in increased corrosion of the service water system. This issue was entered into the corrective action program as Condition Report CR-ANO-2-2016-02879. The failure to inject corrosion inhibitors into Unit 2 service water during refueling outages resulted in increased corrosion of the service water system is a performance deficiency. The performance deficiency is more than minor because it adversely affected the human performance attribute of the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the performance deficiency adversely affected the structural strength of service water system boundaries. Using NRC Inspection Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Dated June 19, 2012, the inspectors screened the finding as having very low safety significance because it is a deficiency affecting the design or qualification of a mitigating SSC, but the SSC maintained its operability. The inspectors determined that this finding had a cross-cutting aspect in the human performance area of Avoid Complacency, because the licensee failed to recognize the potential consequences of isolating chemical injection to the service water during outages, which contributed to degradation. |
| 05000313/FIN-2016007-17 | 2016Q1 | Arkansas Nuclear | Determine Impact of Modifying Fire Seals for Flood Protection | The team identified an unresolved item related to ability to meet the requirements of License Condition 2.C.(8) and 2.C.(3)(b), Fire Protection Program, in Units 1 and 2, respectively. Specifically, the team identified ANO had modified numerous fire rated seals to also provide a flood protection barrier without ensuring existing fire protection requirements continued to be met. ANO Units 1 and 2 used a 3- hour fire rated silicon foam material to seal floor and walls penetrations in order to provide adequate separation to prevent the spread of fire between fire areas. ANO determined that numerous exiting fire seals were also required to provide flood protection. To provide an 3-hour fire barrier and also be capable of withstanding a design basis flood, ANO issued design changes to use several materials, such as Polywater FST Foam Sealant, Promatec Product 12 (P12), Sylgard, and Promatec High Density Silicone Elastomer (HDSE and HDSE-IR), to create dual purpose seals. The team determined that HDSE, HDSE-IR and Sylgard have been tested as a 3-hour fire barrier and tested satisfactorily to provide adequate flood protection. However, ANO could not produce documentation to show that fire rating testing or qualification testing had been performed for the new dual function seals using P12 and Polywater. This was documented in CR-ANO-C-2016-00490. ANO has determined that the population of the non-qualified seals was 139 (96 containing Polywater and 43 containing P12). ANO stated that all of the new dual function seals using P12 consist of the flood protective layer of P12 being placed on top of the existing originally qualified 10 inch fire silicone seal, and that no credit was given to the P12 layer to provide any additional fire protection capabilities. The P12 has been tested by Promatec with silicone seals for flood and was flood tested by the station for use with silicone foam seals. Therefore, ANO believes that no negative chemical reactions can be expected. ANO installed Polywater material either on top of the currently installed fire barrier seal, or in electric conduits that are not required to have a fire seal present. Polywater is designed to create an air and watertight barrier suitable for use in conduits. ANO did not remove any portion of the originally qualified silicon foam fire seals, therefore the flood protection layer of Polywater was applied on top of the existing qualified fire seal. As part of the approved Fire Protection Program, a periodic visual inspection of fire penetration seals is required by TRM 3.7.12.3 and TRM 3.7.5, for Units 1 and 2 respectively, such that 10 percent of the total fire seal population is inspected each year. These inspections are conducted per Unit 1 procedure OP 1405.016, U-1, Penetration Fire Barrier Visual Inspections, and Unit 2 procedure OP 2405.016, U-2, Penetration Fire Barrier Visual Inspections. The team reviewed the inspection procedures and interviewed the fire protection engineers. The team was concerned that for many of the new dual function seals, the original fire rated and qualified seal was no longer accessible for performance of required visual inspections. The team was concerned that because the silicone fire seals are no longer accessible for inspection, the intent of the required fire seal inspection to detect surface flaws or damage to indicate potential underlying damage has occurred to the qualified fire penetration system per the fire protection program could not be met. The team concluded that not having fire rating qualification testing for the existing configuration of some fire seals, and the inability to perform required periodic visual inspections for newly modified fire seals, was a performance deficiency that was reasonably within ANOs ability to foresee and prevent. Since ANO has not yet completed the evaluation or fire testing qualification of the modified seals, the team was unable to evaluate the overall impact of this condition or classify the performance deficiency. ANO intended to complete the evaluation of these issues and document the results in CR-ANO-C-2016-00490. Some of the actions being considered include performing required 3-hour fire testing in representative dual function configurations containing Polywater or P12; and doing a feasibility study for removal and replacement of these seals with fire and flood qualified materials. The team concluded that further review is necessary in order to properly evaluate and disposition the significance of this condition. Specifically, the NRC will need to review the following: ANOs evaluation, extent of condition, and disposition and/or testing results of the non-qualified dual function fire/flood seals; and the significance of the non-qualified population (139 seals containing Polywater or P12). This item is being treated as an unresolved item (URI) 05000313/2016007-17 and 05000368/2016007-17, Fire Seals Modified for Flood with Material not Qualified for Fire and Inability to Perform Required Periodic Visual Inspection. |
| 05000313/FIN-2016001-06 | 2016Q1 | Arkansas Nuclear | Licensee-Identified Violation | Title 10 CFR 50.65(a)(4), states in part, that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, before performing maintenance activities on February 24, 2016, the licensee failed to assess and manage the increase in risk that resulted from maintenance activities in the switchyard. Specifically, the licensee performed maintenance on the supervisory control circuits associated with the startup transformer breakers during the Unit 2 forced outage. This work had already begun when Entergy executives on a fleet call questioned the impact of maintenance on the breakers that supply power to safety-related buses while Unit 2 is shutdown. Further review indicated that the impact was more extensive than previously thought. For immediate corrective actions, control room operators contacted the switchyard coordinator and rescheduled the supervisory control circuit work. Because the finding affects the licensees assessment of risk associated with performing maintenance activities, NRC Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012 directs significance determination using NRC Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005. The finding was determined to be Green because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release frequency probability deficit was less than 1E-7. A senior reactor analyst estimated incremental core damage probability deficit to be 1.9E-8 for Unit 1 and 1.2E-8 for Unit 2 using the Standardized Plant Analysis Risk models for Unit 1 (Revision 8.19) and Unit 2 (Revision 8.26) run on SAPHIRE, Version 8.1.2. The licensee entered the issue into the corrective action program as Condition Report CR-ANO-C-2016-00908. Licensee-identified violations are not assigned cross-cutting aspects. |
| 05000368/FIN-2016001-03 | 2016Q1 | Arkansas Nuclear | Blocked Drain Results in Emergency Feedwater Pump Inoperability | The inspectors documented a self-revealing Green finding with an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to verify that the floor drains in the Unit 2 turbine-driven emergency feedwater pump room would pass the amount of water added to the drain during operation of the pump in order to prevent the pump from becoming submerged. As a result, the licensee was unaware that the turbine-driven emergency feedwater pump room drain had become blocked until water began pooling in the room during a pump test. Upon discovery, the licensee stopped the pump, declared the train inoperable, and cleared the drain. This finding was entered into the licensees corrective action program as Condition Report CR-ANO-2-2016-0384. The failure to verify that the Unit 2 turbine-driven emergency feedwater pump room drain would pass the water added to the drains during operation of the turbine-driven emergency feedwater pump is a performance deficiency. The finding is more than minor because it adversely affected the protection against external factors (i.e., flood hazard) attribute of the mitigating systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to detect a clogged drain affected the availability of the emergency feedwater system by potentially flooding the room and failing the pump. The inspectors evaluated the finding using Manual Chapter 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power. 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. A senior reactor analyst performed a detailed risk evaluation and estimated the total increase in core damage frequency to be 7.7E-7/year, and therefore the finding had very low safety significance (Green). The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor, inadequate documentation of the pump design requirements during initial plant construction, does not reflect current licensee performance. |
| 05000368/FIN-2016001-02 | 2016Q1 | Arkansas Nuclear | Failure to Follow Design Control Requirements for Pump Seal Cooler Replacements | The inspectors identified a Green finding and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to ensure the suitability of materials used in safety-related equipment. Specifically, the licensee made a change to the material used in ten safety-related pump bearing coolers without considering the potential effects of corrosion. As a result, a drain plug corroded and caused service water to spray, rendering two safety-related pumps inoperable. This issue was entered into the corrective action program as Condition Report CR-ANO-2-2016-00550. The failure to consider the potential for corrosion in the design of safety-related equipment is a performance deficiency. The performance deficiency is more than minor because it adversely affected the design control attribute of the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, in each of the three examples, the licensee made changes to the plant where the potential effects of corrosion on safety-related equipment was not considered. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors screened this finding as Green, because the finding did not represent an actual loss of safety function. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. |
| 05000313/FIN-2016001-05 | 2016Q1 | Arkansas Nuclear | Failure to Identify and Repair Intermittent Card Failure Leads to a Reactor Trip | The inspectors reviewed a self-revealing Green finding for the failure to fully understand a malfunction which resulted in putting susceptible cards back into the Unit 1 integrated control system. In 2014, a failure caused a feedwater transient, which operators successfully mitigated. Troubleshooting identified and repaired some of cards susceptible to the intermittent problem. The licensee reinstalled cards that had not been repaired in the integrated control system, which later caused a feedwater transient and subsequent manual reactor trip on December 15, 2015. The licensee documented the issue in Condition Report CR-ANO-1-2015-04178 and replaced the cards. The failure to fully understand a malfunction, which resulted in putting susceptible cards back into the Unit 1 integrated control system, is a performance deficiency. The finding is more than minor because it adversely affected the equipment performance attribute of the initiating event cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee placed the suspect cards back into the integrated control system, which caused a feedwater transient and contributed to a subsequent manual reactor trip. Using NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the finding screened as having very low safety significance (Green) because the deficiency resulted in a reactor trip, but mitigation equipment remained unaffected. Specifically, main feedwater remained available. The inspectors determined this finding has a problem identification and resolution cross-cutting aspect in the area of Evaluation, because the primary cause of the performance deficiency involved the failure to thoroughly evaluate a 2014 integrated control system failure so that the resolution addressed the cause commensurate with safety significance. |
| 05000313/FIN-2016001-01 | 2016Q1 | Arkansas Nuclear | Failure to Assess and Manage Hot Work Risk | The inspectors identified a Green finding and an associated non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the failure to assess and manage the increase in risk due to performing hot work near risk-significant Unit 1 non-vital switchgear. Specifically, the licensee failed to identify the work as having low integrated risk, and implement required risk management actions to protect available fire pumps and brief the fire brigade. As immediate corrective actions, the licensee stopped the hot work until they completed a risk assessment and risk management actions. This finding was entered into the licensees corrective action program as Condition Report CR-ANO-1-2016-00348. The failure to assess and manage the increase in risk of performing hot work near risk-significant Unit 1 non-vital switchgear is a performance deficiency. The finding is more than minor because it adversely affected the protection against external factors (i.e., fires) attribute of the initiating event cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to assess the potential for hot work to cause a fire, and manage the risk to critical safety functions. Because the finding affects the assessment of risk associated with performing maintenance activities, NRC Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, directs significance determination using NRC Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. A regional senior reactor analyst used Manual Chapter 0609, Appendix K, Flowchart 2, Assessment of Risk Management Actions, dated May 19, 2005, to assess the significance of the finding. The licensee site probabilistic risk assessment engineer provided information which estimated the incremental core damage probability deficit of 3.3E-10. The analyst confirmed similar results using the NRC probabilistic risk assessment model. The incremental large early release probability deficit was conservatively estimated to be equal to the incremental core damage probability deficit. Since this issue dealt only with the failure to take risk management actions, Flowchart 2, Assessment of Risk Management Actions, of Appendix K was used. In accordance with Flowchart 2, because the incremental core damage probability deficit was less than 1E-10 and the incremental large early release probability deficit was less than 1E-7, the finding screened as having very low safety significance (Green). The inspectors determined this finding has a problem identification and resolution cross-cutting aspect in the area of Teamwork, because the most significant contributor involved the failure to communicate and coordinate activities across organizational boundaries to ensure nuclear safety is maintained. Specifically, work groups did not inform operations work control personnel that hot work was part of the intended work. |
| 05000313/FIN-2015004-03 | 2015Q4 | Arkansas Nuclear | Failure to Properly Translate the Design Requirements for the Unit 1 Decay Heat Vault Rooms Being Sealed ( | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to correctly translate the regulatory requirements and design basis into specifications, drawings, procedures, and instructions to ensure the Unit 1 decay heat vault boundary components could perform their safety-related function. Inspectors identified that the Unit 1 decay heat vaults had a safety-related function to limit accident dose consequences to the public and the control room operators, but some boundary components had not been classified as safety-related. In response to this issue, the licensee performed an immediate operability determination and reviewed previous leakage testing on the containment spray and low pressure injection systems. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-1-2015-04195. The inspectors determined that the failure to correctly translate the design requirement that the Unit 1 decay heat vaults be sealed to mitigate the dose consequences of an accident into specifications, drawings, procedures, and instructions was a performance deficiency. This performance deficiency was more than minor because it was associated with the design control and safety-related structures, systems, and components and barrier performance attributes 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 for the auxiliary building. Specifically, the licensee failed to ensure that Unit 1 decay heat vault boundary components were designated as safety-related components and met the applicable requirements needed to assure the reliability and integrity of the barrier function. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the issue screened as having very low safety significance (Green) under the Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building questions because the finding only represented a degradation of the radiological barrier function provided for the control room and the auxiliary building and it did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance since this condition had existed since construction. |
| 05000313/FIN-2015004-01 | 2015Q4 | Arkansas Nuclear | Failure to Assess Risk for Switchyard Work | The inspectors identified a non-cited violation of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for failure to assess the risk impact of switchyard maintenance. Specifically, the station failed to properly classify some switchyard work and assess risk as specified in Procedure COPD-024, Risk Assessment Guidelines, Revision 055 during multiple periods of switchyard work between October 2 and 15, 2015. The work involved the repair of damaged conduit on the voltage regulators, transformer refurbishment, relay calibrations, and motor operated disconnect replacement. For immediate corrective actions, each operations shift manager provided training to their crews to ensure they were familiar with required station risk updates. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-C-2015-04147. The failure to assess the increase in risk due to switchyard maintenance is a performance deficiency. The finding is more than minor because it adversely affected the protection against external factors attribute of the Initiating Event cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to evaluate the potential impact of maintenance in the switchyard which could result in plant upsets or transients. Because the finding affects the licensees assessment of risk associated with performing maintenance activities, NRC Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, directs significance determination via the use of NRC Manual Chapter 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005. A regional senior reactor analyst screened the change in core damage frequency to be <1E-6 for Unit 1 and calculated the change in core damage frequency to be 1.5E-7 for Unit 2. In accordance with Flowchart 1 of Appendix K, the significance of this finding was determined to be of very low safety significance (Green), because the calculated Incremental Core Damage Probability Deficits for both units were not greater than 1.0E-6. The inspectors determined this finding has a cross-cutting aspect in the area of Consistent Process, because the primary cause of the performance deficiency involved the failure to use a consistent, systematic approach to manage work decisions in the switchyard (H.13). |
| 05000313/FIN-2015004-02 | 2015Q4 | Arkansas Nuclear | Failure to Identify and Correct Rain Water Accumulation in the Emergency Diesel Generator System Exhausts | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to identify a condition adverse to quality. Specifically, the licensee failed to identify rain water accumulation in the exhaust systems for the Units 1 and 2 emergency diesel generators due to clogged water drains. As a result, rainwater in the exhaust piping may have caused the emergency diesel generators to exceed the seismic rating of the exhaust systems during a seismic event. The inspector identified that when ANO removed the rain shields in 1998, they planned to implement periodic drain line cleaning to avoid clogging, but never created the preventive maintenance item to implement the cleaning. In response, the licensee cleaned the drain lines, drained the exhaust pipes, and implemented preventative maintenance activities to periodically clean the drain lines. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-C-2015-04570. The failure to identify that rainwater was accumulating in all four emergency diesel exhaust systems and could impact the availability of the system is a performance deficiency. The performance deficiency is more than minor because it affected the protection against external factors attribute of the Mitigating Systems Cornerstone objective 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, operators failed to recognize that drain lines were blocked during routine operations to drain the exhaust lines, which allowed rain water to accumulate that exceeded the allowed seismic loading of the piping. Using NRC Manual Chapter 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations, the inspectors determined that a detailed risk evaluation was required. 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 dominant risk was determined to involve seismically induced losses of offsite power. Emergency feedwater and a Unit 2 emergency diesel generator remained available to successfully avoid core damage. The inspectors determined this finding has a cross-cutting aspect in the area of Avoid Complacency because the primary cause of the performance deficiency involved the failure to plan for or recognizing latent conditions involving clogged drain lines (H.12). |
| 05000313/FIN-2015003-02 | 2015Q3 | Arkansas Nuclear | Failure to Promptly Correct a Condition Adverse to Quality Involving Motor Control Center Bus Stabs | The inspectors reviewed a self-revealing violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to correct conditions adverse to quality. Specifically, the licensee failed to promptly replace short bus stabs with longer bus stabs in six 480V safety-related motor control centers as planned following a 2007 motor control center fault. Subsequently, safety-related motor control centers remained susceptible to a fault because corrective actions had not been implemented. This issue was entered into the licensees corrective action program as Condition Report 2015-2661. The licensee has completed the modifications to all breakers except those requiring an outage. The failure to promptly correct conditions adverse to quality associated with 480V breaker connections to bus bars was a performance deficiency. The performance deficiency is more than minor because it is associated with the equipment performance attribute of Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Specifically, untimely corrective actions allowed an increased likelihood of a fault to continue to exist that would result in the loss of the associated safety-related motor control centers if the fault occurred. Using NRC Inspection Manual Chapter 0609 Appendix A, Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green) because the finding was not a deficiency affecting design or qualification, did not represent a loss of system and/or function, and did not represent an actual loss of function. This finding was not assigned a cross-cutting aspect because it was not indicative of current plant performance; the licensee decided to remove the corrective actions from the corrective action program more than 3 years ago. |
| 05000368/FIN-2015003-01 | 2015Q3 | Arkansas Nuclear | Failure to Follow Procedure Results in Increased Reactor Coolant Activity | The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to follow the instructions in the chemical volume control system charging pump pulsation dampener bladder charging procedure. Specifically, maintenance personnel used a gas cylinder containing argon, carbon dioxide, and oxygen rather than a pure nitrogen cylinder to charge the dampener as required by procedure 2411.066, Charging Pump Dampener Bladder 115A, B, C and 2M-116A, B, C Checking and Depressurization, Revision 05. Because the dampener had an existing bladder leak, the gas leaked into the reactor coolant system and the argon subsequently activated when it passed through the reactor. Reactor coolant system activity significantly increased, which elevated dose rates in the auxiliary building. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-2-2015-02576. The licensee revised the procedure to require an independent verification of the gas before charging the pulsation dampeners. The failure to follow the dampener charging procedure, which resulted in increased reactor coolant system activity and elevated dose rates in the auxiliary building, was a performance deficiency. The performance deficiency is more than minor because it is associated with the human performance attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, charging argon into a pulsation dampener with a known bladder leak caused elevated dose rates in several plant areas. Using NRC Inspection Manual Chapter 0609 Appendix, C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, the inspectors determined that the finding was of very low safety significance (Green) because it did not involve ALARA planning or work controls, did not involve an overexposure, did not have a substantial potential to be an overexposure, and the ability to assess dose was not compromised. The inspectors determined this finding had a cross-cutting aspect in the area of Avoid Complacency, because the plant maintenance mechanics failed to implement appropriate error reduction tools such as self-checking and peer-checking (H.12). |
| 05000368/FIN-2013012-05 | 2014Q1 | Arkansas Nuclear | Failure to Follow the Materials Handling Program during the Unit 1 Generator Stator Move | The inspectors reviewed a self-revealing apparent violation of 10 CFR 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 and shall be accomplished in accordance with these instructions, procedures or drawings. The licensee did not follow the requirements specified in Procedure EN-MA-119, Material Handling Program, in that, the licensee did not perform an adequate review of the subcontractors lifting rig design calculation and the licensee failed to conduct a load test of the lifting rig prior to use. The licensee initiated Condition Report CR-ANO-C-2013-00888 to capture this issue in the corrective action program. The licensees corrective actions included repairing damage to the Unit 1 turbine deck, fire main system, and electrical system. In addition, changes were made to various procedures including Procedure EN-DC-114, Project Management, to provide guidance on review of calculations, quality requirements, and standards associated with third party reviews. The inspectors determined that this finding was more than minor because it was associated with the procedural control attribute of the initiating event cornerstone, and adversely affected the cornerstones objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations. The stator drop caused a reactor trip on Unit 2 and damage to the fire main system which resulted in water intrusion into the electrical equipment causing a loss of startup transformer 3. This resulted in the loss of power to various loads, including reactor coolant pumps, instrument air compressors, and the safety-related Train B vital electrical bus. The inspectors used Inspection Manual Chapter 0609, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, to evaluate the significance of the finding. Since this was an initiating event, the inspectors used Exhibit 1 of Appendix A and determined that Section C, Support System Initiators, was impacted because the finding involved the loss of an electrical bus and a loss of instrument air. The inspectors determined that Section E, External Event Initiators, of Exhibit 1 should also be applied because the finding impacted the frequency of internal flooding. Since Sections C and E were impacted, a detailed risk evaluation was required. The NRC risk analyst used the Arkansas Nuclear One, Unit 2 Standardized Plant Analysis Risk Model, Revision 8.21, and hand calculation methods to quantify the risk. The model was modified to include additional breakers and switching options, and to provide credit for recovery of emergency diesel generators during transient sequences. Additionally, the analyst performed additional runs of the risk model to account for consequential loss of offsite power risks that were not modeled directly under the special initiator. The largest risk contributor (approximately 96 percent) was a loss of all feedwater to the steam generators, with a failure of once-through cooling. The result of the analysis was a conditional core damage probability of 2.8E-5; therefore, this finding was preliminarily determined to have substantial safety significance (Yellow). This finding had a cross-cutting aspect in the area of human performance associated with field presence, because the licensee did not ensure adequate supervisory and management oversight of work activities, including contractors and supplemental personnel. Specifically, the licensee did not provide a sufficient level of oversight in that, the requirements in Procedure EN-MA-119, for design approval and load testing of the temporary hoisting assembly, were not followed (H.2) (Section 4OA3.9). |