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05000313/FIN-2018003-062018Q3Arkansas NuclearReactor Power Transient Caused by the Turbine Bypass Valve Failing OpenThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly pre-plan maintenance for the replacement of air supply tubing for turbine bypass valve CV-6687, which resulted in the failure of the air tubing, causing valve CV-6687 to fail open, which led to a manual reactor trip and a subsequent loss of the main condenser.
05000313/FIN-2018003-052018Q3Arkansas NuclearFailure to Maintain Main Feedwater Pump B Discharge Pressure in Band Caused a Reactor TripThe inspectors reviewed a self-revealed, Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specifications 5.4.1.a, for the licensees failure to implement Procedure OP-1102.002, Plant Startup, Revision 106. Specifically, control room operators failed to maintain main feedwater pump discharge pressure in the required band to control flow to the steam generators during a plant startup. As a result, the only operating main feedwater pump tripped on high discharge pressure, causing an automatic reactor trip.
05000313/FIN-2018003-042018Q3Arkansas NuclearFailure to Verify Safety-Related 4160 V Breaker Operability Following Maintenance ActivitiesThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to perform post-maintenance testing to demonstrate component operability for the train A safety-related 4160 V switchgear A-303 breaker that provides power to the swing service water pump B (P-4B) after the breaker was racked in. The breaker subsequently failed to close when attempting to start the pump.
05000313/FIN-2018003-032018Q3Arkansas NuclearFailure to Provide Complete and Accurate Information in a License Amendment Request to Change Emergency Action Level RequirementsThe inspectors identified a Severity Level IV non-cited violation because the licensee provided inaccurate information to the NRC in a license amendment request for an emergency action level scheme change. Specifically, the licensee provided information about the availability of the postaccident sampling system building radiation monitor and the Unit 1 level instrumentation that was material to the licensing decision, but not accurate. The NRC approved an emergency action level scheme change on November 9, 2012 (ADAMS Accession No. ML12269A455) to allow Arkansas Nuclear One to adopt the Nuclear Energy Institute (NEI) 99-01, Revision 5, scheme. Subsequently, the licensee identified that two of their current emergency action level thresholds could not be implemented in accordance with their emergency classification procedure: On May 26, 2017, Condition Report CR-ANO-2-2017-03161 documented that postaccident sampling system building radiation monitor 2RX-9840 should be removed from all regulatory commitments because the postaccident sampling system had been removed from service, and its building would not be monitored for radiological releases. Radiation monitor 2RX-9840 was being used as a means to evaluate emergency action levels AU1, AA1, AS1, and AG1. In addition, it was used in the loss/potential loss of containment (CNB6) for fission product emergency action levels. The condition report noted that requirements for the postaccident sampling system had been removed from Arkansas Nuclear One licenses in August 2000 and the licensee had abandoned the systems valves (March 2003, EC-ANO-1779), removed power from the postaccident sampling system ventilation system (January 2004), and made radiation monitor 2RX-9840 nonfunctional (May 2008, Condition Report CR-ANO-2-2008-01439 and Work Order 150817). On March 15, 2018, Condition Report CR-ANO-C-2018-01121 documented that the Unit 1 level instrumentation set point used in emergency action level CA1 was below the indicating range of the instrument. The emergency action level indicated that a loss of Unit 1s reactor vessel inventory was shown by an indicated level less than 368 feet, 0 inches. Therefore, the lowest level indicated on the instrument would be higher than the level used in making the emergency classification decision. The inspectors reviewed the licensees license amendment request, dated December 1, 2011 (ADAMS Accession No. ML113350317), Proposed Emergency Action Levels Using NEI 99-01, Revision 5, Scheme, and the licensees response to a request for additional information dated July 9, 2012, (ADAMS Accession No. ML12192A090) to determine whether the conditions identified in the corrective action program existed at the time the licensee requested the license amendment and whether the request correctly described the instruments. The inspectors identified: The December 1, 2011, submittal incorrectly indicated that radiation monitor 2RX-9840 was a viable means of classifying emergency action levels AU1, AA1, AS1, and AG1, as well as providing input for the evaluation of fission product barrier emergency action levels. In the response to NRCs request for additional information (RAI) dated July 9, 2012, the licensee provided additional details about the super particulate iodine noble gas (SPING) radiation monitors used in this application. Response to Question 3 associated with emergency action levels AA1, AS1, and AG1 stated: Each SPING is associated with a particular ventilation pathway and provides continuous monitoring of air discharged via the respective release pathway. The license reviewer concluded that all of the SPING monitors included in the license amendment request were operable and continuously monitoring the specified release pathways, thereby being capable of measuring the radiation levels described in the proposed emergency action levels. 17 The December 1, 2011, submittal indicated that loss of Unit 1 reactor vessel inventory for emergency action level CA1 was a vessel level less than 368 feet, 0 inches. This issue was NRC-identified because when the licensee identified the emergency action level errors, they took action to correct the errors, but failed to address the failure to ensure that technical information provided to the NRC in support of the license amendment request was complete and accurate in all material respects. Corrective Actions: To correct the Unit 1 reactor vessel level emergency action level threshold error, the licensee issued communications regarding correct application of the emergency action level on March 15, 2018, followed by implementation of a change to Procedure OP-1903.010, Emergency Action Level Classification, Revision 56, dated June 26, 2018, with the corrected level. The use of radiation monitor 2RX-9840 is being removed from the emergency action levels as part of an emergency action level scheme change submitted to the NRC on March 29, 2018 (ADAMS Accession No. ML18088B412 and ML18094A155). In the interim, the licensee issued communications to emergency director-qualified staff members to ensure they are aware of the error, how to address it if implementing emergency action levels, and to inform them of the corrective actions in progress. Additionally, the licensee issued Condition Report CR-ANO-C-2018-03597, dated September 13, 2018, for the incomplete and inaccurate emergency action level submission examples to address the completeness and accuracy issues identified by the inspectors.
05000313/FIN-2018003-022018Q3Arkansas NuclearFailure to Implement Welding Standard Guidance and Examination ProceduresThe inspectors reviewed a self-revealed Green finding and associated non-cited violation of Arkansas Nuclear One, Unit 1, Technical Specification 5.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to implement welding standard guidance and examination procedure guidance during the installation of the high pressure injection system drain line containing drain valves MU-1066A and MU-1066B. The drain line weld developed a crack that caused a leak shortly after plant startup that was determined to have been caused by grinding during the welding process, which was not permitted by the welding standard.
05000313/FIN-2018003-012018Q3Arkansas NuclearFailure to Translate the Design Requirements into Instructions for Refueling Emergency Diesel GeneratorsThe inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to translate current design into instructions for Unit 1 and Unit 2 diesel fuel oil transfer system. Specifically, the licensee failed to translate the current diesel fuel oil transfer system design into instructions to refuel Unit 1 and Unit 2 safety-related fuel bunkers, T-57 and 2T-57, if the non-safety bulk diesel fuel oil tank T-25 was unavailable following a design basis event (e.g., tornado, external flooding, or earthquake) for which it was not designed to withstand.
05000250/FIN-2017007-072017Q3Turkey PointFailure to Identify ICW Pipe CorrosionThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to inspect Intake Cooling Water (ICW) piping in accordance with license renewal commitments. For immediate corrective actions, the licensee entered the issue into their corrective action program as AR 02218430 and AR 02218437, planned to perform localized corrosion wall thickness measurements to ensure the ICW system remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, unmonitored corrosion affects the reliability of the ICW systems. The team determined the finding to be of very low safety significance because it did not represent an actual loss of function of one or more non-Tech Spec trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for >24 hrs. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because the licensee failed to implement a corrective action program with a low enough threshold for identifying issues (P.1). Specifically, individuals routinely failed to identify corrosion issues on CCW system area walk downs that exceeded proceduralized acceptance criteria of light surface rust specified in 0-ADM- 564, during the July 5, 2017, August 11, 2016, and April 11, 2016 CCW area walk downs.
05000250/FIN-2017007-092017Q3Turkey PointFailure to Update the UFSAR with the Latest Information DevelopedThe NRC identified a Severity Level-IV non-cited violation of Title 10 Code of Federal Regulations 71(e), Maintenance of Records, Making of Reports, for the failure to assure that the Updated Final Safety Analysis Report (UFSAR) contained the latest information developed, including all changes made in the facility or procedures as described in the UFSAR. The team determined that the licensee failed to update the UFSAR to include the latest information regarding several design features associated with turbine runback. For immediate corrective actions, the licensee entered this issue into their corrective action program as AR 2218695 to update the UFSAR. The NRC determined this violation was associated with a minor performance deficiency in accordance with the screening criteria in IMC 0612, Appendix E. Because the failure to update the UFSAR impacted the NRCs ability to perform its regulatory process, the team evaluated the violation using the traditional enforcement process. The team determined that this met the criteria for a SLIV violation because not accurately describing turbine runback design features in the UFSAR could have a material impact on licensed activities, and met the SLIV violation criteria in 6.1.d.3 of the NRC Enforcement Policy. The violation represented a failure to update the UFSAR as required by Title 10 Code of Federal Regulations Part 50.71(e), but the lack of up-to- date information has not resulted in any unacceptable change to the facility or procedures. Cross-cutting aspects are not assigned to traditional enforcement violations
05000250/FIN-2017007-082017Q3Turkey PointFailure to Correct a Non-Conforming Condition Impacting ContainmentThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to take timely corrective action to maintain the unit 3 and 4 containment cathodic protection systems. These systems have been non-functional on both units since 2009. The cathodic protection systems purpose is to protect the containments interconnected liner, reinforcing bars, and tendon trumplates. For immediate corrective actions, the licensee entered the issue into their corrective action program as AR 2216534 and performed a prompt operability determination. The licensee concluded that the containment structure was operable but non-conforming and established plans to monitor the potentially impacted inaccessible areas through continued performance of the American Society of Mechanical Engineers (ASME) IWL and IWE programs until actions are taken to restore the Cathodic Protection System. The performance deficiency was determined to be more than minor, because it is associated with the Design Control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of maintaining the containment structural integrity and operational capability to provide reasonable assurance that the containment protects the public from radionuclide releases caused by accident or events. Specifically, the failure 6 to implement timely corrective actions to maintain the protection of the containments interconnected liner, reinforcing bars, and tendon trumplates affected the structural integrity and operational capability of the containment structure. The team determined the finding to be of very low safety significance because the finding was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance
05000250/FIN-2017007-062017Q3Turkey PointFailure to Verify the Adequacy of Design for the ECC and CCW SystemsThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the Emergency Containment Cooler (ECC) unit 4A auto start circuitry would not result in exceeding the thermal limits of the CCW system during a design basis accident. Specifically, the licensee failed to verify that a single active failure of the motor starter auxiliary contacts would not result in exceeding the design basis limits for CCW as described in updated final safety analysis report (UFSAR) Section 9.3. For immediate corrective actions, the licensee entered the issue into their corrective action program as AR 2219505, performed a prompt determination of operability, and determined that the CCW system remained operable. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, three ECC fans running during a during a design basis accident would result in exceeding the design basis temperature of 158.6 F for the 5 CCW supply and a significant reduction in margin for the SI pump lube oil cooler. The team determined the finding to be of very low safety significance because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC) and the SSC maintained its operability. This finding was not assigned a cross- cutting aspect because the issue did not reflect current licensee performance.
05000250/FIN-2017007-052017Q3Turkey PointFailure to Adequately Perform Discharge Testing on Battery 3BThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to perform surveillance testing on station battery 3B in accordance with the requirements of Institute of Electrical and Electronic Engineers (IEEE) 450-1987. For immediate corrective actions, the licensee entered this issue into their corrective action program as AR 2219948 and performed an extent of condition review, which determined that none of the station batteries were currently in a degraded condition, and placed surveillance procedure 0-SME-003.15 on administrative hold until the corrective actions are completed. The performance deficiency was determined to be more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the performance deficiency could result in masking degradation of the battery on future performance discharge tests and adversely affect the ability to trend when the testing periodicity should be increased to once a year as required by Technical Specifications (TS). The team determined the finding to be of very low safety significance because the finding did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hrs. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance.
05000250/FIN-2017007-042017Q3Turkey PointFailure to Verify the Adequacy of Design for Component Protective CoversThe NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design for the non-safety related component protective covers attached to safety related equipment. For immediate corrective actions, the licensee entered this into their corrective action program as AR 02220993 and removed visibly degraded protective covers. 3 The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute and of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as pow er operations. Specifically, the failure to ensure the quality and qualification of commercial components and assemblies to maintain adequate mounting to Class 1E equipment increased the likelihood of inadvertent component failures, and thus increased the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The team determined the finding to be of very low safety significance because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition (e.g. loss of condenser, loss of feedwater). This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance.
05000250/FIN-2017007-032017Q3Turkey PointFailure to Verify the Adequacy of CCW isolation from Supplemental Cooling System (SCS)The NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design of temperature set points used for isolation of the Component Cooling Water (CCW) from the CCW supplemental cooling system (SCS) during an accident. For immediate corrective actions, the licensee entered this into their corrective action program as AR 2218834, performed an operability determination, which determined the system is operable but non-conforming, and issued engineering change (EC) 289598 to account for uncertainties in the CCW SCS temperature isolation setpoint. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and 4 capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by not ensuring prompt isolation or adjusting the isolation setpoint to account for instrument uncertainties and temperature lag, the licensee failed to ensure that the SCS loop would be isolated at onset of an accident, which affected the reliability and capability of the CCW system when called upon. The determined the finding to be of very low safety significance because the findings were a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained their operability or functionality. The finding had a cross-cutting aspect in the area of Human Performance because the licensee failed to ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values (H.9).
05000250/FIN-2017007-022017Q3Turkey PointFailure to Perform Design Verification for Under Frequency Trip of the Main Generator BreakersThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for failure to verify or check the adequacy of design of the under frequency trip feature of the main generator circuit breakers with regard to the effect of its operation on plant stability and the maintenance of critical safety functions. The licensee entered this issue into their corrective action program as AR 2220874 and AR 2224998. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions. Specifically, opening of the main generator breakers due to an under frequency condition on the offsite power system would cause the generator load to suddenly drop from full power to the level of the plant loads, and there was no verification that plant stability and critical safety functions would be maintained. The team evaluated the finding with Inspection Manual Chapter 0609, Appendix A, and determined the finding met the Support System Initiators screening criteria for requiring a detailed risk evaluation. The team determined that this issue increased the likelihood of the support system initiator loss of offsite power (LOOP). The regional senior risk analyst conducted a detailed risk evaluation with a one year exposure and determined the change in core damage frequency was less than 1E-6, which was of very low safety significance (Green). The team did not assign a cross-cutting aspect because the issue did not reflect current licensee performance.
05000250/FIN-2017007-012017Q3Turkey PointInadequate Verification of Electrical Protective Device Selective CoordinationThe NRC identified a non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for failure to verify that coordination exists between the protective devices on safety related switchgear in order to minimize the probability of losing a safety related power bus. For immediate corrective actions, the licensee entered this issue into their corrective action program as Action Request (AR) 2220956 and performed an operability determination, which determined the system was operable, and was performing a reevaluation of the calculation to determine adequate coordination. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failing to verify short circuits in non-safety related SSCs downstream of the safety related switchgear would not cause a lockout of the safety related bus affected its availability and reliability. The team determined the finding to be of very low safety significance because the finding was a deficiency affecting the design of a mitigating structure, system, or component (SSC), and the SSC maintained their operability or functionality. This finding was not assigned a cross-cutting aspect because the issue did not reflect current licensee performance.
05000250/FIN-2017007-102017Q3Turkey PointPotential failure of 125 Vdc Bus 3B Class 1E componentsUFSAR Section 8.2.2.3.1 stated that the emergency power for vital instrumentation and controls is supplied by a station DC power system which contains five safety related 125Vdc batteries and four DC distribution panels. 125 Vdc distribution panel 3B supplies safety related power to several safety-related equipment including sequencers, reactor trip switchgear, inverter 3Y06, and control power to 480Vac load centers 3B and 3D and 4160 Vac switchgears 3AB01 and 4AB20. UFSAR Section 7.2 stated that the reactor protection system was designed in accordance with IEEE 279- 1968. Section 4.5 of IEEE 279-1968, Channel Integrity, requires all protection system channels be designed to maintain necessary functional capability under extremes of conditions relating to malfunctions. During the review of calculation 5177-265-EG-22, Circuit Breaker/Fuse Coordination Study, Rev. 8, the team questioned if there were instances where class 1E cables associated with DC Bus 3B (3D23) would not be adequately protected given a short circuit on the load side of the breakers. The failure to ensure the Class 1E protective devices would not allow the maximum available short circuit to permanently damage cabling to safety-related equipment associated with DC Bus 3B could result in additional loss of Class 1E equipment. Unresolved Item (URI) 05000250/2017007-01 and 05000251/2017007-01, Potential failure of 125 Vdc Bus 3B Class 1E components,) is opened for additional review to determine if the Class 1E cables on DC Bus 3B can withstand the maximum possible short circuit and to determine if a performance deficiency exists.
05000269/FIN-2017007-012017Q1OconeeFailure to Identify and Correct Broken Cable Trench Cover.Green. The NRC identified a non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to assure that a condition adverse to quality associated with a damaged trench cover on the yellow trench was identified and corrected. Specifically, the seismic design function of the trench cover was not identified or recognized at the time of the licensees original identification of the issue and subsequent NCR generation, and, due to this error, appropriate corrective actions were not assigned or completed. In response to the issue, the licensee replaced the broken trench cover on the yellow trench with a temporary cover on March 22, 2017 , and planned work order 20147282 to replace it with a permanent cover to restore the design configuration . This performance deficiency was more than minor because it was associated with the Design Control Attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, inadequate identification and correction of this condition adverse to quality adversely impacted the trench covers reliability and capability to perform its function during and following a seismic event. The team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), and the SSC maintained its operability or functionality. The team determined that the finding was indicative of current licensee performance, because the issue was first identified in January 2017. A cross -cutting aspect of Consistent Process (H.13.) in the Human Performance Area was assigned because individuals did not use a consistent, systematic approach to make decisions.
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