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05000424/FIN-2018002-0130 June 2018 23:59:59VogtleFailure to Adequately Load Emergency Deisel Generator (EDG) During 24-Hour Endurance TestAn NRC-identified Green NCV of Vogtle Nuclear Station TS, Section 5.4.1.a, Procedures, was identified for the licensees failure to implement the EDG 24-hour endurance surveillance procedure 14668A-1, Train A Diesel Generator Operability Test, revision 7.2, to operate the EDG as close as practicable to 3390 kVAR. Specifically, the licensee failed to carry out procedure steps and provisions that would assist in loading the EDG closer to the TS value of 3390 kVAR. The failure to follow procedure 14668A-1 and get as close as practicable to 3390 kVAR was a performance deficiency.
05000424/FIN-2018410-0331 March 2018 23:59:59VogtleSecurity
05000425/FIN-2017004-0131 December 2017 23:59:59VogtleFailure to Implement and Establish Appropriate Work Instructions for PMT of Namco Limit Switch on 2HV-8920A Green, self-revealing, non-cited violation (NCV) of TS 5.4.1.a, Procedures, was identified for the licensees failure to implement maintenance work instructions and establish appropriate procedures concerning the post-maintenance testing (PMT) of the Namco limit switch on Unit 2 for 2HV-8920 following removal and reinstallation of the limit switch. As a result, during ECCS interlock testing, 2HV-8804B (RHR Pump B to SI Pump B Isolation Valve) failed to open due to 2HV-8920 Namco limit switch being installed improperly. The licensees failure to perform a PMT on the Namco limit switch for 2HV-8920 following removal and reinstallation, as required by NMP-MA-014-001 (Post Maintenance Testing Guidance), was a performance deficiency (PD). The licensee reinstalled the limit switch correctly and performed the interlock testing satisfactory following the corrective maintenance. The issue was entered into the corrective action program (CAP) as condition report (CR) 10410863.The PD was 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the PD affected the reliability of the ECCS valve interlock system. The finding was of very low safety significance (e.g. Green) because while logic path II (2HV-8920 and 2HV-8814) for the opening of 2HV-8804B was inoperable, the system maintained its functionality due to the availability of logic path I (2HV-8813). The inspectors determined there was no cross-cutting aspect since the finding is not indicative of current performance.
05000424/FIN-2017004-0231 December 2017 23:59:59VogtleFailure to Maintain NEMA Type 4 Qualification for the Nuclear Service Cooling Water PumpsA Green, self-revealing, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to properly implement and establish procedures to maintain watertight requirements of the nuclear service water system (NSCW) pumps motor main power cables termination box. As a result, the Unit 2 B train NSCW pump no. 4 failed due to aphase-to-ground fault caused by water and moisture intrusion into the power cable splice connections. Failure to adequately implement and establish procedures to maintain watertight requirements of the NSCW pumps motor main power cables termination box during maintenance, as required by maintenance procedures and specifications, was a performance deficiency. The licensee replaced the motor and faulted cable; and sealed all potential water and moisture intrusion enclosure locations until watertight enclosure standards are fully restored. This issue was entered into the licensees CAP as CRs10399125, 10404327, and corrective action report 270905.The performance deficiency was 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (e.g. core damage). Specifically, the Unit 2 NSCW pump no. 4 was rendered inoperable, adversely affecting the NSCW system reliability. The finding was determined to be of very low safety significance (Green) because it did not result in an actual loss of safety system function, and it did not represent a loss of function of one or more than one train for more than its Technical Specification (TS) allowed outage time or greater than 24 hrs. The finding was assigned a cross-cutting aspect of Resources, because procedures and/or work instructions were not available to maintenance personnel for properly verifying motor termination boxes were installed in compliance with NEMA 4 specifications. (H.1)
05000425/FIN-2017003-0230 September 2017 23:59:59VogtleFailure to Maintain ECCS Flow Balance and Check Valve Inservice Test ProcedureAn NRC- Identified, Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to maintain a Unit 2 surveillance procedure that demonstrated satisfactory performance of the forward flow safety function of emergency core cooling system ( ECCS ) check valves. The licensee revised and performed the test to verify satisfactory valve performance. This issue was entered into the licensees CAP as CR10410794. The failure to maintain procedure 14721D -2 to ensure test conditions that adequately demonstrated satisfactory performance of ECCS check valves 2- 1205- U6 -001/00 2, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD ). The performance deficiency was more than minor because if left uncorrected, it could result in degradation of ECCS check valves to go undetected. The finding was associated with the mitigating system cornerstone. The finding was determined to be of very low safety significance (Green) because the performance deficiency did not result in a loss of operability or functionality of ECCS check valves. The finding was assigned a cross cutting aspect of Resources, because the licensee did not ensure that an ECCS surveillance procedure was adequate to support nuclear safety . (H.1)
05000425/FIN-2017003-0330 September 2017 23:59:59VogtleFailure to Maintain Cleanliness of Motor Operated Valve Limit Switch CompartmentA Self -Revealing , Green, NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to perform an adequate cleanliness inspection of the Unit 2 nuclear service cooling water (NSCW) system pump no. 6 discharge motor -operated -valve (MOV) limit switch compartment, as required by the maintenance procedure. As result , the valve failed to operate when demanded and rendered the NSCW pump inoperable. The failure to perform an adequate cleanliness inspection of NSCW pump no. 6 discharge MOV limit switch compartment following preventive maintenance, as required by maintenance procedure NMP -ES- 017- 008, was a performance deficiency (PD). The licensee cleaned affected MOV sub -components, verified proper operation, and restored operability of the pump. This issue was entered into the licensees CAP as CR10399054 . 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 (i.e. core damage). The finding was determined to be of very low safety significance (Green) because although the performance deficiency affected the qualification and operability of the NSCW pump, it did not represent a loss of function of an NSCW train for greater than its TS Allowed Outage Time . The finding was assigned a cross cutting aspect of Avoid Complacency, because maintenance technicians did not recognize the possibility of making mistakes when performing routine tasks of inspecting and manipulating grease containing components inside the limit switch compartment. (H.12)
05000424/FIN-2017406-0130 September 2017 23:59:59VogtleSecurity
05000424/FIN-2017003-0130 September 2017 23:59:59VogtleFailure to Implement and Establish Appropriate Work Instructions Affecting Safety-Related ChillerA Self -Revealing, Green, non- cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow measurement and test equipment (M&TE) in support of essential safety features (ESF) chilled water pumps in- service testing (IST). As a result, the Unit 1 A train safety -related chiller was inadvertently rendered inoperable when technicians isolated a flow transmitter associated with the chillers auto -start control logic when installing and removing M&TE in support of the IST. The licensee entered this issue into their corrective action program (CAP) under condition report (CR) 10390340 and corrective action report 270610 and planned to revise the procedure. Failure to implement maintenance work instructions and establish appropriate procedures concerning the use of flow M&TE in support of ESF chilled water pumps IST, which can affect ESF chiller performance, as required by Regulatory Guide (RG) 1.33, Quality Assurance Program Requirements, of February 1978, was a performance deficiency (PD). The performance deficiency was 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance (Green) because while the unit 1 A train ESF chiller was rendered inoperable, it did not represent a loss of function of the train for greater than its TS Allowed Outage Time. The finding was assigned a cross cutting aspect of Challenge the Unknown because questions and risks regarding the use of flow M&TE for the test were not properly evaluated and managed before proceeding. (H.11)
05000424/FIN-2017002-0130 June 2017 23:59:59VogtleFailure to Correct a Condition Adverse to Quality involving an MSIV Manufacturing Deficiency(Green). A self -revealing, Green, non -cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to identify and correct a condition adverse to quality (i.e., manufacturing deficiency), which led to a repetitive failure of main steam isolation valve ( MSIV ) 1HV -3006B. The fail ure to determine the cause of a significant condition adverse to quality and take corrective action to preclude repetition was a performance deficiency. Specifically, the licensee failed to identify the root cause of an MSIV actuator failure on April 12, 2014, that resulted in a reactor trip. As a result, appropriate corrective actions were not taken and a repeat failure of the valve actuator caused another reactor trip on February 3, 2017 . The licensee has entered this issue into the corrective action pr ogram as condition report 10326456. This performance deficiency was more than minor because it was associated with the Human Performance attribute of the Initiating Events Cornerstone, and adversely affected the cornerstone objective of limiting the likeli hood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding was of very low safety significance (Green) because the finding did not result in a loss of mitigation equipment use d to transition the reactor to a stable shutdown condition. The finding was not assigned a cross cutting aspect since it was not indicative of current licensee performance due to the root cause evaluation in question being performed greater than three years ago
05000424/FIN-2017002-0230 June 2017 23:59:59VogtleFailure to Follow Work Instructions for Implementation of Open Phase Protection System(Green). A self -revealing, Green, non -cited violation of Technical Specifications 5.4.1.a, Procedures, was identified for the licensees failure to redline new wiring installation associated with an open phase protection system modification, as required by work instructions . As result, control circuit wires were not installed per wiring diagrams and caused a loss of the offsite power feed to the B train 4160- volt emergency power bus. The licensee's failure to redline new wiring installation associated with an open phase protection system modification installation, as required by work instruction SNC804606 and 3 maintenance procedure NMP -MA -017 was a performance deficiency. The licensee entered this issue into their corrective action program under condition reports 10343972 and 10344136 and restored offsite power to the emergency bus by correcting the wiring configuration . 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. The finding was determined to be of very low safety significance (Green) because the in- service train of shutdown cooling (i.e. , 'A' train of the residual heat removal system ) was not affected. The finding was assigned a cross -cutting aspect of Procedure Adherence, in the Human Performance area becaus e individuals did not follow work instructions and redline procedures when installing new wiring for the open phase protection system (H.8)
05000424/FIN-2017008-0130 June 2017 23:59:59VogtleFire Protection Program Change did not meet VEGP License Condition Requirement 2.G for Units 1 and 2The inspectors identified a Severity Level IV (SL IV) non-cited violation (NCV) and associated Green finding of Vogtle Units 1 and 2 Operating License Conditions 2.G, for the licensees failure to perform an evaluation of the impact of a change to the approved fire protection program (FPP). The failure to adequately evaluate the impact of the change resulted in the implementation of a change to the FPP that could have adversely affected the ability to achieve and maintain safe shutdown. The licensee initiated condition report (CR) 10382461 to evaluate the issue and make necessary correction to the program. The inspectors determined that the licensees failure to adequately evaluate the impact of the change to the FPP was a performance deficiency (PD). The PD was determined to be more than minor because if left uncorrected, the PD could have the potential to lead to a more significant safety concern. Specifically, if degraded fire doors are not evaluated for functionality, the doors could potentially be left in a condition where it would not perform its design function in the case of a fire. The licensees failure to submit the FPP change to the NRC was determined to impede the regulatory process because the FPP change required NRC review and approval prior to implementation. The finding was screened as Green because, based upon inspection of the affected barriers, the inspectors determined that, either, the combustible loading on both sides of the barrier represented a fire duration of less than 1.5 hours, there was a fully functional automatic suppression system on either side of the barrier, or the barrier separated rooms that utilized the same SSD strategy. This violation was determined to be a Severity Level IV violation because the associated finding was evaluated by the SDP as having very low safety significance (i.e., Green finding). No cross cutting aspect was assigned because the finding was not indicative of current licensee performance.
05000424/FIN-2017007-0131 March 2017 23:59:59VogtleFailure to identify a Degraded Atmospheric Relief ValveThe NRC identified a Green finding for the licensees failure to identify the reduced reliability of Unit 1 loop 3 atmospheric relief valve (ARV) 1PV-3020 as a degraded/nonconforming condition, as required by NMP-AD-012, Operability Determinations and Functionality Assessments, Version 12.5. As a result, corrective maintenance was not prioritized nor conducted at the next available opportunity and led to an additional valve failure in March 12, 2016. The failure to identify aging of 1PV-3020 #285 pilot-to-check valve as a degraded/non conforming condition, as required by NMP-AD-012, was a performance deficiency. The performance deficiency was 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 availability, reliability, and capability o systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the performance deficiency prevented the license from prioritizing and conducting corrective maintenance of 1PV-3020 at the next available opportunity, and led to an additional valve failure in March 2016. Using Exhibit 2 of IMC 0609, Appendix A, the inspectors determined that this finding is of very low safety significance (Green) because, although the performance deficiency (PD) affected the design/qualification of the 1PV3020 operability, it did not result in an actual loss of safety system function, and it did not represent a loss of function of one or more than one train for more than its technical specification (TS) allowed outage time or greater than 24 hours. The finding was assigned a cross cutting aspect of Resolution in the Problem Identification and Resolution area, because the licensee failed to take effective corrective actions to address aging of the #285 pilot-to-check valve in a timely manner.
05000424/FIN-2016007-0231 December 2016 23:59:59VogtleFailure To Ensure Adequate Unit 1 Emergency Diesel Generator Surveillance Acceptance CriteriaThe NRC identified a Green 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 have adequate instructions and acceptance criteria to confirm the emergency diesel generators capability to reject the largest single load without exceeding predetermined frequency and voltage while maintaining a specified margin to the overspeed trip. The violation was entered into the licensees corrective action program as condition report 10294395. An immediate determination of operability was performed and concluded that the Emergency Diesel Generators were operable but degraded nonconforming. The licensee was evaluating corrective actions, which may include a final determination of the most severe single largest load and re-performing the surveillance tests. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, without adequate acceptance criteria in surveillance procedure SR 3.8.1.8, the procedure could not ensure availability, reliability, and capability of the EDG under the most severe power demand characteristics for electric power used by components. The team determined the finding to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of technical specification or non-technical specification equipment. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000424/FIN-2017503-0131 December 2016 23:59:59VogtleTransposition Results in Significantly Different EAL Threshold ValuesTBD: The inspectors identified an apparent violation (AV) of Title 10 CFR Part 50.54(q)(2) for failure to follow and maintain the effectiveness of emergency plans which met the requirements of 10 CFR Part 50.47(b)(4) and Part 50 Appendix E, to have a standardized emergency action levels (EAL) scheme in use based on facility system and effluent parameters. Specifically, the licensee's emergency classification scheme for Radiological Effluent EAL RG1 (General Emergency) and RS1 (Site Area Emergency), contained radiation monitor threshold values which were significantly different (forty-two times different) due to a transposition of the threshold values. The licensee took immediate corrective actions by entering the issue into the corrective action program as condition report (CR) 10283097 and providing corrected EAL declaration threshold values to appropriate management and decision-makers (shift managers/emergency directors) via Standing Order C-2016-008. The performance deficiency was determined to be more than minor because it was associated with the Emergency Preparedness cornerstone attribute of Procedure Quality and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the licensees ability to declare a Site Area Emergency (SAE) and General Emergency (GE) based on effluent radiation monitor values was degraded in that event classification could be delayed and unnecessary Protective Action Recommendations could be provided to the public. The finding was assessed for significance in accordance with NRC Inspection Manual Chapter (IMC) 0609, Appendix B, Emergency Preparedness Significance Determination Process. The inspectors determined that the finding constituted a degraded rather than lost risk significant planning standard function and accordingly is assigned White significance. Additionally, the overconservative threshold values could result in an over classification and unnecessary PARs to the public. In accordance with IMC 0609, Appendix B, an EAL over-classification that would result in unnecessary PARs for the public is assigned White Significance. Because these two findings resulted from the same performance deficiency, one White finding with two examples will be cited. The cause of the finding was determined to be associated with a cross-cutting aspect in the change management component of the human performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority (H.3).
05000424/FIN-2016007-0731 December 2016 23:59:59VogtleFailure to Update the UFSAR with the Complete and Accurate InformationThe NRC identified a severity level IV non-cited violation of Title 10 Code of Federal Regulations Part 50.71(e)(4) for the failure to reflect all changes made in the facility or procedures as described in the Updated Final Safety Analysis Report (UFSAR). The licensee failed to update UFSAR with the design basis of a new digital emergency diesel generator sequencers installed in 2007. This violation was entered into the licensees corrective action program as condition reports 10288350, 10293456, 10291633. The licensee planned to update the UFSAR with the applicable design basis. The failure to update the UFSAR was a performance deficiency that was determined to be a minor reactor oversight program violation because it did not meet the more than minor screening criteria. Because the issue impacted the NRCs ability to perform its regulatory process, the inspectors evaluated the violation using the traditional enforcement process. The inspectors determined the issue was a severity level IV violation because it met violation example 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.
05000424/FIN-2016007-0331 December 2016 23:59:59VogtleFailure to Meet Isolation Requirements When Incorporating Non- Class 1E Components into Class 1E electrical CircuitsThe NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III Design Control, for installing non-safety related Individual Cell Equalizer devices into the Class 1E battery charging circuits without isolation as specified by Institute of Electrical and Electronics Engineers standard 384 as amended by RG 1.75. The violation was entered into the licensees corrective action program as condition report 10294321. The licensee was evaluating corrective actions, which included the removal of the non-Class 1E components. The performance deficiency was determined to be more than minor because it affected the Design Control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to conform to Class 1E design requirements for independence affected the reliability of the Class 1E battery systems. The team determined the finding to be of very low safety significance (Green), because it was a deficiency affecting the design or qualification of a SSC, and the SSC maintained its operability or functionality. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000424/FIN-2016007-0531 December 2016 23:59:59VogtleFailure to Perform Periodic Testing Of Safety-Related Valve InterlocksThe NRC identified a Green, non-cited violation of Title 10 Code of Federal Regulations Part 50.55a(h)(2) Protection Systems, because the licensee failed to perform periodic testing of safety-related valve interlocks to ensure an adequate single failure analysis by identifying detectable failures in accordance with Institute of Electrical and Electronics Engineers standard (IEEE) 379-1972, IEEE Trial-Use Guide for the Application of the Single-Failure Criterion to Nuclear Power Generating Station Protection Systems. The violation was entered into the licensees corrective action program as condition report 10293749. The licensee performed an immediate determination of operability and determined that the affected systems were operable but degraded nonconforming. The licensee was in the process of determining and developing adequate corrective actions to conform with Institute of IEEE Standard 379-1972. 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to periodically test safety-related valve interlocks affected the adequacy of the licensees single failure analysis. The team determined the finding to be of very low safety significance (Green) because the finding was not a design deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of technical specification or nontechnical specification equipment. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000425/FIN-2016007-0431 December 2016 23:59:59VogtleFailure to Perform Required In-Service Testing of Unit 2 CST Swap over ValvesThe NRC identified a Green non-cited violation of Technical Specification 5.5.8, Inservice Testing Program, for Vogtle Unit 2 failure to perform the required testing in accordance with the American Society of Mechanical Engineers Operation and Maintenance Code for nine valves that had active safety functions. Specifically, these valves were required to operate when aligning the AFW pumps from Condensate Storage Tank (CST) 1 to CST 2. The violation was entered into the licensees corrective action program as condition report 10293900. The licensee performed an immediate determination of operability and determined that the CST valves were operable but degraded nonconforming. The licensee planned to register the CST valves into the IST program and exercise those valves that that have never been exercised at the first available opportunity. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, degraded valve performance could go undetected without periodic testing and trending. The team determined the finding to be of very low safety significance (Green) because the finding was not a design or qualification deficiency, did not represent a loss of system and/or function, and did not represent the loss of any trains of TS or Non-TS equipment. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000424/FIN-2016007-0631 December 2016 23:59:59VogtleTurbine Driven Auxiliary Feedwater (TDAFW) Pumps 1/2-1302- P4-001 and Motor Driven Auxiliary Feedwater (MDAFW) Pumps 1/2-1302-P4-002/003The NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control for the licensee's failure to translate the Auxiliary Feedwater (AFW) pumps design bases into adequate acceptance criteria for technical specifications SR 3.5.7.2 and for the failure to verify the adequacy of the design of the same AFW pumps. The licensee entered the violation into the corrective action program as condition reports 10293456 and 10294168. As an immediate corrective action, the licensee evaluated the operability of the Unit 1 and 2 AFW pumps, modify the allowed diesel frequency acceptance criteria, and initiated corrective action to develop new acceptance criteria and monitor pump performance for degradation. The performance deficiencies were more-than-minor because they were associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, when the quality of the established surveillance criteria was considered, there was a reasonable doubt on the operability of the Unit 1 and 2 turbine driven AFW and 2A and 1B motor driven AFW pumps. The team determined the finding to be of very low safety significance (Green) because it did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time. The team determined that the finding had a crosscutting aspect in the Human Performance area of Design Margins (H.6), because engineers did not demonstrate the characteristic of ensuring that design margins were guarded and changed only through a systematic and rigorous process.
05000425/FIN-2016004-0131 December 2016 23:59:59VogtleFailure to Implement Maintenance Procedure for Electrical Grayboot Connectors(Green). A self-revealing non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to properly install shims when assembling electrical connectors on Unit 2 main steam isolation valve (MSIV) HV-3026B, in accordance with maintenance procedure 25709-C, Instructions for EGS Grayboot Connection Kit Installation, Ver. 21.1. The licensee replaced the affected connectors and entered the issue in their corrective action program under condition reports (CR) 10279411, and 10268507, and technical evaluations (TE) 970299, 968149, and 970300, to evaluate and develop additional training for maintenance technicians, enhance the maintenance procedure, and conduct extent of condition. The performance deficiency (PD) was more-than-minor, because it adversely effected the Initiating Events cornerstone objective when Unit 2 received an automatic reactor trip and safety injection on March 14, 2015. Also, if left uncorrected, the PD would result in moisture intrusion and degradation of MSIV connectors and potentially lead to a more significant safety concern. The finding was determined to be Green, because the PD did not result in a loss of mitigation equipment used to transition the reactor to a stable shutdown condition. The finding was assigned a cross cutting aspect of Procedure Adherence, because maintenance technicians failed to adhere to procedural guidance in Attachment 1 of 25709-C for installing the connector shims. (H.8)
05000424/FIN-2016007-0131 December 2016 23:59:59VogtleFailure to Verify Capability of EDGs under Maximum Voltage and FrequencyThe NRC identified a Green non-cited violation of Title 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for failure to correctly translate the appropriate permissible limits for frequency and voltage from technical specifications into the emergency diesel generators design loading calculations as required by the licensing and design bases. The violation and related issues were entered into the licensees corrective action program as condition reports 10288732 and 10293810. The licensee was evaluating corrective actions, which included determining acceptable loads at the more limiting power demands and developing procedural guidance. 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 the emergency diesel generators to respond to initiating events to prevent undesirable consequences. Specifically, failing to evaluate the impact from the frequency and voltage limits allowed by technical specification could result in overloading the diesel generator if operators manually loaded additional plant protection systems during an event. The team determined the finding was of very low safety significance (Green) because it was a design deficiency that did not result in a loss of emergency diesel generators operability. The team did not assign a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000424/FIN-2016003-0130 September 2016 23:59:59VogtleFailure to Properly Implement Fire Door InspectionsAn NRC-identified Green non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.d, Procedures, was identified for the licensees failure to correctly verify fire door gaps at the strike plate area and between meeting edges of double swinging metal doors were within acceptable limits. The licensee initiated hourly roving fire watches for these fire doors and took corrective maintenance action to restore affected fire doors within limits. The licensee documented this condition in condition reports 10254221 and 10252774. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Hazards (i.e. fire) and adversely affected the cornerstone objective in that door gaps outside the required limits compromised the doors fire rating qualification. The finding was determined to be of very low safety significance (i.e. Green) because either the combustible loading on both sides of each door was representative of a fire duration of less than 1.5 hours or each door maintained at least a 1-hour fire endurance rating. The finding had a cross-cutting aspect of Training in the Human Performance area because the licensee did not ensure there was adequate training to properly inspect station fire doors (H.9).
05000424/FIN-2016403-0130 September 2016 23:59:59VogtleSecurity
05000425/FIN-2016002-0130 June 2016 23:59:59VogtleFailure to properly implement a maintenance procedure caused a Reactor TripA self-revealing non-cited violation (NCV) of Technical Specifications (TS) 5.4.1.a, Procedures, was identified for the licensees failure to properly implement procedure 24750- 2, Steam Generator Level (Narrow Range) Protection Channel II 2L-519 Channel Operational Test and Channel Calibration. During testing of Unit 2 loop 1 steam generator (S/G) narrow range channel 2L-519 the channel was not removed from scan resulting in a reactor trip. The licensees immediate corrective actions were to remove the technicians performing the calibration from maintenance duties for formal remediation. The licensee documented this condition in CR 10230073. The performance deficiency (PD) was more than minor because it adversely affected the Initiating Events cornerstone objective in that the failure to properly remove channel 2L-519 from scan resulted in a reactor trip. The finding was determined to be Green because the PD did not result in a loss of mitigation equipment used to transition the reactor to a stable shutdown condition. The finding was assigned a cross cutting aspect of Avoid Complacency because maintenance technicians failed to implement appropriate error reduction tools to verify that the correct channel was removed from scan for testing.
05000424/FIN-2016502-0130 June 2016 23:59:59VogtleFailure to Adequately Maintain Emergency Response FacilitiesThe inspectors identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50.54(q)(2), for the licensees failure to maintain the effectiveness of its emergency plan by ensuring that adequate emergency facilities and equipment to support emergency response are provided and maintained as required by 10 CFR 50.47(b)(8). Specifically, the effectiveness of the emergency plan was reduced by a change to the Technical Support Center (TSC) functionality requirements in Technical Requirements Manual (TRM) TR 13.13.1, Emergency Response Facilities, Revision 1. The requirement to maintain climate control was removed without an adequate basis to support removal. The procedure change had been in place since September 2013, and until a corrected revision is issued, a Standing Order has been put in place. The licensee entered this finding into the corrective action program (CAP) as condition report (CR) 10221041. The inspectors determined that the performance deficiency was more than minor because it was associated with the procedure quality attribute of the Emergency Preparedness (EP) cornerstone, adversely affected the associated cornerstone objective, and would have affected the emergency response organizations ability to effectively perform their duties had an emergency been declared and TSC climate control non-functional. The finding was evaluated using the EP significance determination process and was identified as having very low safety significance (Green) because it was a failure to comply with NRC requirements and was not a loss of the planning standard function or the overall function of the TSC. The finding was associated with a cross-cutting aspect in the Change Management component of the Human Performance area because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000424/FIN-2015004-0131 December 2015 23:59:59VogtleFailure to Implement Preventive Maintenance Procedure for 7300 Process Protection and Control System Printed Circuit BoardA Green self-revealing NCV of TS 5.4.1, Procedures, was identified for the licensees failure to implement replacement schedules for 7300 process protection and control (PP&C) system cards in accordance licensee fleet maintenance procedures. As a result, failure of a 7300 PP&C card rendered the Unit 2 B train of nuclear service water system (NSCW) inoperable. The violation was entered into the licensees corrective action program as condition report (CR) 10124315 and corrective action report (CAR) 261373. The failure to implement replacement schedules for 7300 PP&C system cards in accordance with maintenance procedure NMP-MA-015 was a performance deficiency. 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 in that the failure of the 7300 PP&C card affected the availability of the Unit 2B train of NSCW. The finding screened as having very low safety significance (i.e. Green) because it did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time. No cross-cutting aspect was assigned to this finding because the inspectors determined that the cause of the finding was not indicative of current licensee performance because the licensee has established a change management process that would prevent the Performance Deficiency from occurring.
05000425/FIN-2015003-0330 September 2015 23:59:59VogtleUnauthorized Entry into a High Radiation AreaA self-revealing NCV of Technical Specification (TS) 5.7.1, High Radiation Area, for an unauthorized entry into a high radiation area (HRA). The radiological aspects were not discussed in the pre-job brief, the health physics (HP) technician in containment did not challenge the crew as to whether or not they received their HRA briefing, and the crew did not follow adequate radiological safety practices, such as reading instructions on the HRA posting prior to entry and not leaning against piping. The licensee entered this issue into the CAP as CR 870060 The entry into a HRA without meeting the entry requirements specified in T.S. 5.7.1 was a performance deficiency. This performance deficiency was more than minor because it was associated with the Occupational Radiation Safety cornerstone attribute of Human Performance and adversely affected the cornerstone objective in that workers who enter HRAs with inadequate knowledge of current radiological conditions could receive unintended occupational exposures. The finding was evaluated using the Occupational Radiation Safety Significance Determination Process and determined to be of very low safety significance (Green). This finding does not involve a cross-cutting aspect because it is not current license performance.
05000424/FIN-2015007-0130 September 2015 23:59:59VogtleFailure to Fully Close and Latch Plant Fire DoorsAn NRC-identified Green non-cited violation of Vogtle Units 1 and 2 Operating License Conditions 2.G, was identified for the licensees failure to ensure that fire doors V22108L1A67, V12111L1238, and V12111L1A41 in 3-hour rated fire barriers were fully closed and latched, as required by the approved fire protection program (FPP) and National Fire Protection Association (NFPA) Code 80-1983, Fire Doors and Windows (Vogtle NFPA Code of Record). The licensee took corrective actions and declared fire door V22108L1A67 inoperable and established a roving fire watch. The inoperable door was entered into the licensees corrective action program as condition report (CR) 10067247 and was repaired the next day. For doors V12111L1238 and V12111L1A41, the licensee immediately removed materials that were interfering with the latching of the doors and entered these into their corrective action program as CR 10096004 and CR10096008 respectively. Because these two conditions were corrected as soon as they were brought to the licensees attention by the inspectors, no fire watch was required to be established. The licensees failure to ensure the three fire doors were fully closed and latched as required by the approved FPP and NFPA Code 80-1983 was determined to be a performance deficiency. This performance deficiency was more than minor because it affected the reactor safety mitigating systems cornerstone attribute of protection against external events (i.e., fire) and adversely affected the fire protection defense-in-depth element involving fire confinement and control of fires that do occur to protect systems important to safety. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, which determined that an IMC 0609, Appendix F, Fire Protection Significance Determination Process, review was required as the finding involved the ability to confine a fire. The finding category of Fire Confinement was assigned, based upon that element of the FPP being impacted. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) at Task 1.4.3, Question C, based upon observation that a fully functioning, automatically actuated, fire suppression system was installed on both sides of fire doors V12111L1238 and V12111L1A41 and on one side of fire door V22108L1A67. The inspectors determined that the finding had a cross-cutting aspect of Procedure Adherence in the Human Performance area because individuals did not follow processes and procedures for ensuring that fire doors were properly closed and latched after passing through the doors.
05000424/FIN-2015007-0230 September 2015 23:59:59VogtleFailure to Identify and Repair a Degraded Fire Penetration SealAn NRC-identified Green non-cited violation of Vogtle Unit 1 Operating License Condition 2.G was identified for the licensees failure to identify and repair degraded fire penetration seal 1-11-759A, as required by the approved fire protection program (FPP). The licensee took corrective actions to declare the penetration seal inoperable, entered the issue in their corrective action program as condition report 10102010 and established a roving fire watch. The licensees failure to identify and repair the degraded fire penetration seal 1-11-759A was a performance deficiency. This performance deficiency was more than minor because it affected the reactor safety mitigating systems cornerstone attribute of protection against external events (i.e., fire) and adversely affected the fire protection defense-in-depth element involving fire confinement and control of fires that do occur to protect systems important to safety. The finding was screened in accordance with NRC Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 4, Initial Characterization of Findings, which determined that an IMC 0609, Appendix F, Fire Protection Significance Determination Process, review was required as the finding involved the ability to confine a fire. The finding category of Fire Confinement was assigned, based upon that element of the FPP being impacted. Using the criteria contained in IMC 0609 Appendix F, Attachment 2, Table A2.2, the inspectors concluded that the seal degradation level was low because the silicone foam seal depth and a fully intact damming board on one side of the barrier would have been sufficient to provide at least two hours of fire resistance. In addition, it was noted that the fire zones on each side of the degraded fire penetration seal were protected with an automatic fire suppression system. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding was of very low safety significance (Green) at Task 1.4.3, Question C. The inspectors determined that the finding had a cross-cutting aspect of Avoid Complacency in the Human Performance area because individuals inspecting the seals failed to recognize and plan for the possibility of the penetration seal being damaged.
05000424/FIN-2015003-0130 September 2015 23:59:59VogtleFailure to Maintain Requalification Examination IntegrityAn NRC-identified Non-cited Violation (NCV) of 10 CFR 55.49, Integrity of examinations and tests, was identified for the licensees failure to adhere to requirements of NMP-TR-424, License Operator Continuing Training Exam Development, Version 3.1. NMP-TR-424 was the procedure that the licensee used to implement industry standard ACAD 07-001, Guidelines for the Continuing Training of Licensed Personnel. ACAD 07-001 is a methodology which can be used to fulfill 10 CFR 55.59(c), Requalification program requirements and 10 CFR 55.4, Systems approach to training (SAT). This violation has been entered into the licensees corrective action program (CAP) as condition report (CR) 10115484. The inspectors determined that the licensees failure to adhere to overlap standards in NMP-TR-424 was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective in that the failure to adhere to examination overlap standards adversely affected the quality of the administration of the operating exams. The finding was determined to be of very low safety significance (Green) because there was no evidence that a licensed operator had actually gained an unfair advantage on an examination required by 10 CFR 55.59. The finding was directly related to the cross-cutting aspect of procedure adherence of the cross-cutting area of Human Performance because the training staff did not follow the guidance for all licensed operators 2014 annual operating exam.
05000424/FIN-2015003-0230 September 2015 23:59:59VogtleNRC Biennial Written Examinations did not Meet Qualitative StandardsAn NRC-identified finding was identified when between 20 and 40 percent of the written examination questions administered to licensed operators during the biennial requalification examination did not meet the requirements of NMP-TR-424, Licensed Operator Continuing Training Exam Development, and NUREG-1021, Operator Licensing Examination Standards For Power Reactors, Revision 10. The inspectors determined that the failure to ensure that biennial written examinations met the qualitative standards for written examinations was a performance deficiency (PD). The PD was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that the quality of biennial written examinations potentially impacted the licensees ability to appropriately evaluate licensed operators. The significance of the finding was determined to be Green because between 20 and 40 percent of the questions reviewed did not meet the standard. No cross-cutting aspect was identified that would be considered a contributor to the cause of the finding.
05000424/FIN-2015002-0130 June 2015 23:59:59VogtleFailure to Identify and Correct Degraded Foreign Material Cover Plates for the NSCW Pump WellsAn NRC-identified, non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for the licensees failure to identify and correct conditions adverse to quality associated with the cover plates for the nuclear service cooling water (NSCW) system pumps shaft well access openings. Specifically, the licensee failed to identify degraded conditions on the NSCW pump well cover plates (e.g. openings from uncovered holes and degraded periphery) that could result in foreign material (FM) entering the pumps well and impact cooling water flow to safety related heat exchangers. The licensee entered the issue into their corrective action program (CAP) under CR10033287, CR10085803 and CR10091171, installed temporary FM exclusion covers, and removed debris near the pump cover wells. The finding was more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the openings in the degraded pump well covers could allow FM to enter the NSCW system and adversely affect cooling water flow to essential component coolers. The finding was evaluated using the mitigating systems cornerstone column of Attachment 4 and Exhibit 2 of Appendix A to Inspection Manual Chapter 0609, Significance Determination Process, (SDP) dated April 29, 2015. The finding was of very low safety significance (i.e. Green) because the inspectors answered No to all of the screening questions in the exhibit. The inspectors determined the finding had a cross-cutting aspect of Evaluation in the Problem Identification and Resolution (PI&R) area because the organization did not thoroughly evaluate the NSCW debris-blocking event of the 1B safety injection (SI) lube oil (LO) cooler, in February 27, 2015, to ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance (P.2).
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05000425/FIN-2015001-0231 March 2015 23:59:59VogtleFailure to Implement Maintenance Procedure for Containment Spray PumpA self-revealing NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to verify that the total indicated run-out (TIR) for the Unit 2 B train containment spray pump was within the limits of procedure 27052-C, Gould 3415 Pump Maintenance Procedure, Ver. 6.0. This violation was entered into the licensees corrective action program as CR 855892. The failure to implement maintenance procedure 27052-C was a performance deficiency. The performance deficiency was more than minor because it was associated with the SSC and Barrier Performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective in that the failure to verify the 2B CS pump shaft TIR was within the procedural and vendor recommendation limits affected the CS system availability and reliability. 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, or heat removal components, and it did not involve a reduction in function of hydrogen igniters in the reactor containment. No cross-cutting aspect was assigned to this finding because the inspectors determined that the cause of the finding was not indicative of current licensee performance.
05000425/FIN-2014005-0131 December 2014 23:59:59VogtleFailure to Correctly Implement Control Rod Drive System Procedure During Reactor Startup ActivitiesA self-revealing non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement system operating procedure SOP 13502-2, Control Rod Drive and Position Indication System, version 42, when resetting the control rod drive system bank overlap unit (BOU). This caused an out-of-sequence control rod insertion that resulted in operators manually tripping the reactor. The licensee correctly reset the BOU prior to restarting the unit and enhanced the procedural guidance for resetting the BOU. The violation was entered into the licensees corrective action program (CAP) as condition report (CR) 879125. The performance deficiency (PD) was more than minor because it was associated with the Configuration Control and Equipment Performance attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective in that improper rod control system equipment lineup affected the licensees ability to control reactivity. The finding screened as Green because the finding did not affect reactor protection system trip capability or result in an unintentional positive reactivity addition. The inspectors determined the finding had a cross-cutting aspect of training in the human performance area because the organization had not provided sufficient practical or hands-on training on resetting the BOU.
05000425/FIN-2014005-0331 December 2014 23:59:59VogtleFailure to Follow Procedures Renders Safety Related Battery Charger InoperableA self-revealing NCV of TS 5.4.1.a, Procedures was identified for the licensees failure to properly implement maintenance procedures and work order instructions and inadvertently removed the 2AD1CB safety-related battery charger from service while attempting to perform routine battery surveillance on the 2CD1B battery. Upon discovery, the licensee immediately stopped the work and returned the battery charger to service. The licensee entered the condition into their corrective action program as CR 10002493. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective in that the opening of the power supply breaker to the incorrect battery charger (2AD1CB) resulted in the charger being inoperable for a total of 30 minutes. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. Because the inspectors answered No to all of the Exhibit 2, Mitigating Systems Screening Questions, the inspectors concluded that the finding was (Green).The inspectors determined the finding had a cross-cutting aspect of Challenge the Unknown in the Human Performance area because the station operator proceeded in the face of uncertainty.
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05000425/FIN-2014004-0530 September 2014 23:59:59VogtleNOED 14-2-03 to allow mechanical seal replacement and testing of the Unit 2 B Containment Spray PumpThe inspectors identified an unresolved item (URI) regarding NOED 14-2-03 granted on August 21, 2014. The inspectors reviewed NOED 14-2-03 and related documents to determine the accuracy and consistency with the licensees assertions and implementation compensatory measures and commitments, those of which included ensuring the availability of both trains of the emergency core cooling systems, both trains of the containment cooler units, and the remaining train of the CS system. Additional inspection is required to conduct a review of the licensee event report (LER) and licensee root cause analysis to determine if the 2B CS pump inboard mechanical seal failure was associated with a performance deficiency and violation of NRC requirements. This URI is identified as URI 05000425/2014004-05 NOED 14-2-03 to allow mechanical seal replacement and testing of the Unit 2 B Containment Spray Pump.
05000424/FIN-2014004-0130 September 2014 23:59:59VogtleFailure to Correctly Implement a Condensate and Feedwater Systems Procedure for StartupA self-revealing non-cited violation (NCV) of Technical Specification (TS) 5.4.1.a, Procedures, was identified for the licensees failure to implement system operating procedure (SOP) 13615-1, Condensate and Feedwater Systems, Version 84. Specifically, on July 30, 2014, the licensee conducted a power increase from Mode 2 (approximately 3 percent reactor power) to Mode 1 (approximately 8 percent reactor power) with main condenser hotwell level control in manual versus automatic as directed by procedure. This resulted in a main feedwater transient and a subsequent reactor shutdown. The licensee initiated an incident response team and entered this event into their corrective action program as condition report (CR) 847734. Additional corrective actions included revising the SOP to include specific instructions for the control of main condenser hotwell level with corresponding number of operating condensate pumps. The performance deficiency was more than minor because it was associated with the human performance 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 performance deficiency was associated with a human error during implementation of SOP 13615-1, resulting in a main feedwater transient event (i.e. loss of condensate pump net positive suction head (NPSH) in the condenser hotwell resulting in lowering steam generator water levels), that subsequently upset plant stability. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. The finding screened as Green because it did not cause a reactor trip. The inspectors determined the finding had a cross-cutting aspect of procedure adherence in the human performance area because the unit operator did not implement SOP 13615-1 procedure Step 4.1.1.5, which required the UO to verify condenser hotwell control, 1LIC- 4415, is in auto maintaining normal level.
05000424/FIN-2014004-0230 September 2014 23:59:59VogtleInoperability of Unit 1 Emergency Containment Coolers due to Incorrect TagoutA self-revealing NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to specify and verify the correct unit designation in clearance and tagout instructions for removing the Unit 2 nuclear service cooling water (NSCW) system B train from service, as required by Administrative Procedure NMP-AD-003, Equipment Clearance and Tagging, Ver. 17.4. As a result, on September 23, 2014, operators isolated the NSCW supply valve to the B train containment coolers on the wrong unit (i.e. Unit 1), rendering it inoperable. Following closure of the valve, operators in the Unit 1 control room received containment coolers low flow alarms and took actions to reposition the valve and restored NSCW flow. The licensee entered this issue into their corrective action program as CR 870005. The performance deficiency was more than minor because it was associated with the SSC and barrier performance attribute of the barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that the containment barrier to protects the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency affected the availability of the B train of the emergency containment coolers which support the capability of the containment barrier to protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. The finding screened as Green because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components, and it did not involve a reduction in function of hydrogen igniters in the reactor containment. The inspectors determined the finding had a cross-cutting aspect of challenge the unknown in the human performance area because neither of the individuals that reviewed the tagout documentation stopped, after questioning appropriateness of manipulating 1HV-11689, and evaluated the situation before proceeding.
05000425/FIN-2014004-0330 September 2014 23:59:59VogtleLoss of Plant Effluent Monitoring CapabilityA self-revealing NCV of TS 5.5.4, Radioactive Effluent Controls Program, occurred when the licensee failed to maintain continuous, representative monitoring of the Unit 2 plant vent gaseous effluents as required by the offsite dose calculation manual (ODCM) for approximately ten days, between March 16 and March 26, 2014. The license entered the event in the corrective action program as CR 8284999, and took immediate corrective actions to establish continuous monitoring of the Unit 2 plant vent gaseous effluents. Corrective actions planned, completed, or under evaluation include, changes to the vent sampling procedure, impact assessment on ODCM requirements, departmental stand downs to share lessons learned, work control process changes for equipment tagouts, and training. The performance deficiency was more than minor because it was associated with the public radiation safety cornerstone attribute of plant facilities, equipment and instrumentation availability and adversely impacted the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain. This finding was assessed for significance using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, issued February 12, 2008, and determined it to be of very low safety significance because the licensee was able to assess the dose to the public by correlating other plant radiation monitoring equipment and programs to demonstrate this dose was less than the values in Appendix I to 10 CFR Part 50 and/or 10 CFR 201301(e). This finding had a cross-cutting aspect of identification in the problem identification and resolution area because the licensee failed to recognize the impact a loss of vacuum indication had on the operability of 2RE12444 (the continuous monitoring equipment) completely, accurately, and in a timely manner (P1).
05000424/FIN-2014004-0430 September 2014 23:59:59VogtleFailure to Correctly Implement a Chemical and Volume Control System Procedure for Reactor Water MakeupA self-revealing NCV of TS 5.4.1.a, Procedures, was identified for the licensees failure to implement SOP 13009-1, CVCS Reactor Makeup Control System, Version 50.1. Specifically, on July 9, 2014, the licensee conducted a blended makeup to the volume control tank (VCT) at a boric acid concentration lower than what the procedure required, which resulted in an inadvertent boron dilution of the reactor coolant system (RCS), and caused a subsequent power excursion. Upon recognition, the unit operator took immediate actions to reduce power to an acceptable level. The licensee entered this issue into their corrective action program as 837899. The performance deficiency was more than minor because it was associated with the human performance 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 performance deficiency was associated with a human error during implementation of SOP 13009-1, resulting in a reactivity event (i.e. inadvertent boron dilution), that subsequently upset plant stability. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. The finding screened as Green because it did not cause a reactor trip. The inspectors determined the finding had a cross-cutting aspect of avoid complacency in the human performance area because the reactor operator did not implement error reduction tools, such as STAR (Stop, Think, Act, Review), as self-check to ensure that work activities were performed safely.
05000424/FIN-2014003-0230 June 2014 23:59:59VogtleUnauthorized Entry into a High Radiation AreaA self-revealing NCV of Technical Specification (TS) 5.7.1, High Radiation Area, was identified for an entry into a high radiation area (HRA) without meeting the entry requirements as specified therein. Specifically, on March 17, 2014, an operator was authorized to enter an HRA on Unit 1 under conditions where dose rates were known to be changing. This allowed the operator entry into an HRA without knowledge of actual radiological conditions. He was not provided with a radiation monitoring device that continuously indicated dose rates in the area, nor was he accompanied by an individual qualified in radiation protection procedures with a radiation monitoring device providing positive control over his activities. Upon discovery of the condition, the licensee secured access to the area, performed follow-up surveys and convened a human performance review board to examine causal factors and identify corrective actions. The licensee entered this issue into the corrective action program as CR 787908. This finding was more than minor because it was associated with the occupational radiation safety cornerstone attribute of human performance and adversely affects the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, workers permitted entry into HRAs with inadequate knowledge of current radiological conditions could receive unintended occupational exposures. The finding was evaluated using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process (SDP), dated August 19, 2008. The finding was not related to As Low As Reasonably Achievable (ALARA) planning, nor did it involve an overexposure or substantial potential for overexposure and the ability to assess dose was not compromised. Therefore, the finding was determined to be of very low safety significance (Green). This finding had a cross-cutting aspect of avoid complacency in the human performance area because health physics (HP) personnel failed to verify plant conditions through available means when an evolution was in progress that was known to increase area dose rates prior to authorizing entry into an HRA.
05000425/FIN-2014003-0430 June 2014 23:59:59VogtleInadequate Maintenance Procedures and Usage Results in a Failed MFRV and an Automatic Reactor TripA self-revealing NCV of 10 CFR 50 Appendix B Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide adequate work instructions as well as failure to follow the maintenance procedure used to install flexible and rigid conduit. Specifically, the work instructions did not provide adequate directions and/or precautions to properly slope conduit during installation to prevent water intrusion into a valve positioner. The work instructions referenced maintenance procedure 25008- C, Flexible and Rigid Conduit Installation. The maintenance procedure referenced Vogtle design specification X3AR01 Section E-8, Raceway Systems, which provided sloping and tightness criteria for conduit installations. The licensee conducted a root cause investigation and entered the event into their corrective action program (CR 797929). The licensee repaired the improperly sloped conduit, replaced the positioner, and revised procedure 25008-C to specify standards for proper sloping of conduits. The finding was more than minor because it was associated with the procedure quality and human performance attributes of the reactor safety - initiating events cornerstone and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to provide adequate work instructions as well as failure to follow procedure 25008-C, Flexible and Rigid Conduit Installation, resulted in the Unit 2 loop 3 main feedwater regulating valve (MFRV) positioner failing closed, causing a subsequent automatic reactor trip due to low-low steam generator (SG) water level. Because the inspectors answered No to all of the IMC 0609 Appendix A (dated June 19, 2012) Exhibit 1, Section B, Initiating Events Screening Questions, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect of procedure adherence in the human performance area because the maintenance electricians did not follow Vogtle design specification procedures or drawings resulting in the improper sloping of the MFRV flexible conduit.
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05000424/FIN-2014003-0130 June 2014 23:59:59VogtleInadequate Maintenance Procedure Results in a Failed MSIV and a Manual Reactor TripA self-revealing non-cited violation (NCV) of 10 CFR 50 Appendix B Criterion V, Instructions, Procedures, and Drawings, was identified for failure to provide adequate work instructions in the maintenance procedure used for main steam isolation valve (MSIV) maintenance. Specifically, maintenance procedure 26854-C, Main Steam Isolation Valve Actuator Maintenance, used to perform maintenance on Rockwell MSIV(s), did not provide adequate instructions for installing the lower manifold/cylinder O-ring during reassembly. This resulted in a pinched O-ring on 1HV3006B, a subsequent failure of the O-ring causing the MSIV to fail closed, and a manual reactor trip. The licensee conducted a root cause investigation and entered the event into their corrective action program (condition report (CR) 800018). The licensee replaced the Oring, performed an extent of condition evaluation for all other MSIVs, and revised the maintenance procedure to include specific instructions for the installation of the lower manifold/cylinder O-ring. The finding was more than minor because it was associated with the procedure quality attribute of the reactor safety - initiating events cornerstone and it adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to provide an adequate procedure with adequate instructions for ensuring proper O-ring installation resulted in the failure of the Unit 1 loop 1 outboard MSIV hydraulic actuator causing the loop 1 MSIV to fail closed and a subsequent manual reactor trip due to lowering steam generator water level. Because the inspectors answered No to all of the IMC 0609 Appendix A (dated June 19, 2012) Exhibit 1, Section B, Initiating Events Screening Questions, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors determined the finding had a cross-cutting aspect of resources in the human performance area, because the maintenance procedure used to install manifold/cylinder O-ring did not provide adequate instructions for the proper installation of the O-ring.
05000424/FIN-2013404-0231 December 2013 23:59:59VogtleSecurity
05000424/FIN-2013404-0131 December 2013 23:59:59VogtleSecurity
05000424/FIN-2013005-0131 December 2013 23:59:59VogtleFailure to Meet the Conditions of TS LCO 3.8.4A Green, self-revealing non-cited violation (NCV) of plant Technical Specification (TS) 3.8.4, DC Sources - Operating, was identified for failure to meet the conditions of TS limiting condition for operation (LCO) 3.8.4. Specifically, placing the 1AD1CA battery charger out of service during performance of the 18 month load test surveillance, concurrent with the failure of the 1AD1CB battery charger, caused the 1A train chargers to be unable to fulfill their specified safety function. As a result, the 1AD1 safety-related 1E 125 VDC source was inoperable. The 1AD1CB battery charger was repaired, functionally tested, and placed back in service. This violation was entered into the licensees corrective action program as condition report (CR) 735160. The finding was more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, removing the 1AD1CA battery charger from service to conduct a 18 month load test while the 1AD1CB battery charger was not capable of performing its specified safety function resulted in the loss of a single train for greater than its TS allowed outage time. The inspector evaluated the finding in accordance with IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. Since the inspectors answered Yes to the question A.3 of the Mitigating Systems Screening Questions, Does the finding represent an actual loss of function of at least a single Train for greater than its TS allowed outage time, a detailed risk evaluation was required. A detailed risk evaluation was performed by resident inspectors and reviewed by a regional senior reactor analyst in accordance IMC 0609 Appendix A guidance using the NRC Vogtle Standardized Plant Analysis Risk (SPAR) model and the NRC Saphire 8 risk analysis code. An SDP Module Condition Analysis was run with the Unit 1 A train battery chargers, 1AD1CA and 1AD1CB failed with no recovery allowed for a 14 hour exposure period. The dominant sequence was a transient consisting of a reactor trip coincident with the common cause failure of auxiliary feed pumps (AFW) to run and the inability of an operator to restore main feedwater (MFW). The detailed risk evaluation determined that the risk due to the performance deficiency was an increase in core damage frequency (?CDF) of <1E-7/year, a GREEN finding of very low safety significance. Because the increase in ?CDF was <1E-7/year no external events analysis was required. The risk was mitigated by the availability of alternate trains of components and the short exposure period. The detailed risk evaluation was reviewed by a regional senior reactor analyst. The inspectors determined that the cause of this finding was related to the corrective action program (CAP) component of the problem identification and resolution (PI&R) cross-cutting area due to less-than-adequate problem evaluation techniques. Specifically, licensee failed to adequately investigate why the wires were rolled during initial functional testing.