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05000424/FIN-2018003-01Licensee-Identified Violation2018Q3This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 CFR Part 50.54(q)(2), required, in part, the licensee shall follow and maintain the effectiveness of its emergency plan that meet the standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(4), required, in part, a standard emergency classification and action level scheme, the bases of which include facility and system effluent parameters, is in use by the nuclear facility licensee. Contrary to the above, from January 30, 2018 to July 20, 2018, the licensee failed to maintain the effectiveness of its emergency plan. Specifically, Units 1 and 2 procedure 19200, F-O Critical Safety Function Status Tree, version 1.0, specified over-conservative reactor coolant system (RCS) temperature values for determining a critical safety function RED Path on RCS Integrity used to evaluate emergency classification FA1 (Alert), potential loss of RCS barrier, in response to a rapid RCS cooldown event.
05000424/FIN-2018002-02High Vibrations on Unit 2 NSCW Pump No. 3 Result in Pump Inoperability2018Q2An NRC-identified Green NCV of 10 CFR 50 Appendix B, Criterion III, Design Control, was identified for the licensees failure to ensure that design control measures for the Unit 2 train A (2A) nuclear service cooling water (NSCW) pump no. 3 motor replacement, conducted in May 2015, adequately evaluated and addressed structural resonance of the pump, commensurate with the original pumps. As a result, the pump operated at higher than desired vibrations, since installation, causing accelerated bearing wear and premature failure of the motor in February 2018. The licensees failure to ensure that design control measures for the 2A NSCW pump no. 3 motor replacement adequately evaluated and addressed structural resonance of the pump, commensurate with the original pumps was a performance deficiency.
05000424/FIN-2018002-01Failure to Adequately Load Emergency Deisel Generator (EDG) During 24-Hour Endurance Test2018Q2An 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-2018410-03Security2018Q1
05000424/FIN-2018410-02Security2018Q1
05000424/FIN-2018410-01Security2018Q1
05000425/FIN-2018001-03Inadequate Refurbishment of Emergency Diesel Generator Pneumatic Control System Logic Boards2018Q1A Green self-revealing NCV of TS Section 5.4.1.a, Procedures, was identified for the licensees failure to properly preplan and perform maintenance work on the Unit 2 B train (2B)emergency diesel generator (EDG) pneumatic control shutdown logic board. The inadequate shutdown logic board refurbishment resulted in a pneumatic control system air leak that generated an EDG shutdown signal during testing and de-energized the safety-related emergency power bus.
05000424/FIN-2018001-02Inadequate Acceptance Criteria for Testing of NSCW Pump Discharge Valves2018Q1An NRC identified Green NCV of 10 CFR 50.55a(f), "Inservice testing(IST)requirements," subsection (4), American Society of Mechanical Engineers (ASME) Operation and Maintenance of Nuclear Power Plants (OM) code Subsection ISTC-5122, Stroke Acceptance Criteria, was identified for the licensees failure to incorporate adequate acceptance criteria for exercise testing of NSCW pump discharge valves into procedures. Specifically, the licensee failed to incorporate acceptance criteria for stroke close exercise testing into in-service test procedures and used inadequate reference values when determining the HIGH/LOW code allowable limits for the stroke open exercise testing.
05000424/FIN-2018001-01Failure to Provide Work Instructions for Sealing Around NSCW System Pump Shaft Well Access Openings2018Q1An NRC identified Green NCV of Vogtle Electric Generating Plant Technical Specification(TS), Section 5.4.1.a, Procedures, was identified for the licensees failure to provide work instructions for the sealing of gaps around cover plates for the nuclear service cooling water (NSCW) system pumps shaft well access openings and for the failure to follow work instructions for NSCW tower clean/inspect. Specifically, the licensee failed to provide instructions for sealing around the well plate covers following well plate cover removal/reinstallation in work orders SNC737852 (Unit 1 NSCW pump #3) and SNC737853 (Unit 1 NSCW pump #5). Also, during the performance of a NSCW tower clean/inspect work order, the licensee failed to generate condition reports, as required by the work instructions, upon the discovery of cracks or gaps in the Foreign Material Exclusion (FME) barrier. As a result, gaps were left around the NSCW pumps which could allow foreign material to enter the NSCW system and adversely affect cooling water flow to essential component coolers.
05000424/FIN-2017004-03Licensee-Identified Violation2017Q4The following violations of very low safety significance (Green) or Severity Level IV were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation.Title 10 CFR 50.55a(f), Inservice testing requirements, subsection (4) required, in part, that pumps and valves which are classified as ASME Class 1, Class 2, and Class 3 must meet the inservice test requirements set forth in the ASME OM Code. The ASME Code of record for Vogtle for Operation and Maintenance of Nuclear Power Plants (OM) is the 2004 edition through 2006 addendum. Subsection ISTC-1300, Valve Categories, required in part, that valves within this subsection shall be placed in one or more of the following categories. Category A is for valves for which seat leakage is limited to a specific maximum amount in the closed position for fulfillment of their required function(s), as specified in ISTA-1100. Contrary to the above, since 1991, the licensee did not categorize valves in ECCS recirculation flow paths to the RWST as Category A valves to ensure the ASME OM test requirements were met by leak testing the valves to demonstrate that their seat leakage would limit the consequences of an accident to control room operators and to the public at the site boundary per Title 10 CFR Part 100 limits. The inspectors determined this finding was of very low safety significance (Green) because the issue would only have the potential to represent a degradation of the radiological barrier function provided for the control room. This issue was documented in the licensees CAP as CR 829367 and TE 886122.
05000424/FIN-2017004-02Failure to Maintain NEMA Type 4 Qualification for the Nuclear Service Cooling Water Pumps2017Q4A 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-2017004-01Failure to Implement and Establish Appropriate Work Instructions for PMT of Namco Limit Switch on 2HV-89202017Q4A 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.
05000425/FIN-2017009-01Failure to Install Drain Hole2017Q4The NRC identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for licensees failure to verify drain holes were installed as assigned, following the licensees evaluation of Information Notice (IN) 89-63, Possible Submergence of Electrical Circuits Located Above the Flood Level Because of Water Intrusion and Lack of Drainage. Specifically, the licensee did not verify that that junction box 2BTJB0486 was equipped with a weep hole consistent with the assigned corrective action and the corrective action was closed without corrective action being taken. In response to the issue, the licensee initiated Condition Report (CR) 10439858, performed an immediate determination of operability, and determined that equipment associated with the cables in the junction box were Operable but Degraded/Non-conforming (OBDN), and plans to return the affected equipment to fully conforming status. This 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 of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, not correcting the condition could cause submergence of the unqualified cables during events, which affects the reliability of the equipment. The inspectors 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. No cross cutting aspect was assigned because the inspectors determined that the finding was not indicative of current licensee performance, because the error occurred on January 25, 1990.
05000424/FIN-2017003-04Licensee-Identified Violation2017Q310 CFR 50, Appendix B, Criterion XI, Test Control stated, in part, that test programs shall be established to assure that all testing required to demonstrate that structures, systems and components will perform satisfactorily in service. UFSAR Section 8.1.4.3.C.2 stated that the onsite electrical system was designed in accordance with IEEE 308 -1974, Criteria for Class 1E Power System at Nuclear Generating Stations. IEEE 308 -1974 Section 6.3 recommended periodic tests be performed at scheduled intervals to detect deterioration of equipment to demonstrate operability of the components that are not exercised during normal operation. Contrary to the above, the licensee did not establish adequate test control measures to assure that the protective function of all 1E lockout relays were periodically verified. Specifically, there was no preventative maintenance to test the 1E lockout relays for non- MSPI loads. This condition has existed since plant initial operation and was identified during a licensee Nuclear Oversight audit on July 13, 2017. The inspectors determined this finding was of very low safety significance (Green) because the inspectors found no documented history of in- service failures of 1E lockout relays rendering safety -related equipment inoperative. This issue was documented in the licensees corrective action program as CR 10381797.
05000425/FIN-2017003-02Failure to Maintain ECCS Flow Balance and Check Valve Inservice Test Procedure2017Q3An 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-2017003-03Failure to Maintain Cleanliness of Motor Operated Valve Limit Switch Compartment2017Q3A 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-2017406-01Security2017Q3
05000424/FIN-2017003-01Failure to Implement and Establish Appropriate Work Instructions Affecting Safety-Related Chiller2017Q3A 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-2017003-05Licensee-Identified Violation2017Q310 CFR 20.1501 requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in Part 20 and that are reasonable under the circumstances to evaluate the extent of radiation levels, concentrations or quantities of radioactive materials, and the potential radiological hazards that could be present. Contrary to the above, on June 28, 2017, the licensee failed to evaluate radiological conditions in room 1- AB -C-94, Back flushable Filter Crud Tank Pump Room, following the tank being placed in recirculation by Operations. On July 2, 2017, during a routine survey of room 1- AB- C-94, general area dose rates in the area were found to be as high as 600 mrem/hr. On the previous survey, conducted on June 19, 2017, maximum dose rates were found to be as high as 60 mrem/hr. This finding was evaluated using IMC 0609, Appendix C, Occupational Radiation Safety SDP, and was determined to be of very low safety significance (Green) because the finding is not related to ALARA dose planning, did not result in an overexposure or the substantial potential for overexposure, and the ability to assess dose was not compromised due to the use of appropriate personnel dosimetry. Therefore, the inspectors determined the finding to be of very low safety significance (Green). This issue was entered into the licensees corrective action program as CR 10383067.
05000424/FIN-2017002-02Failure to Follow Work Instructions for Implementation of Open Phase Protection System2017Q2(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-2017002-01Failure to Correct a Condition Adverse to Quality involving an MSIV Manufacturing Deficiency2017Q2(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-2017008-01Fire Protection Program Change did not meet VEGP License Condition Requirement 2.G for Units 1 and 22017Q2The 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-2017007-01Failure to identify a Degraded Atmospheric Relief Valve2017Q1The 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-2016007-02Failure To Ensure Adequate Unit 1 Emergency Diesel Generator Surveillance Acceptance Criteria2016Q4The 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-2016007-08Failure to Promptly Identify Nonconformances with Tornado Missile Protection2016Q4Enforcement Guidance Memorandum (EGM) 15-002 dated 6/10/2015, (ADAMS Accession No. ML15111A269) provided guidance to exercise enforcement discretion when an operating power reactor licensee does not comply with the plants current site-specific licensing basis for tornado-generated missile protection. Specifically, discretion would apply to the TS limiting conditions for operation (LCO) which would require a reactor shutdown or mode change, if a licensee could not meet TS LCO required action(s) within the TS completion time. The EGM background discussed Regulatory Issue Summary (RIS) 2015-06, Tornado Missile Protection, dated 6/10/2015, (ADAMS Accession No. ML15020A419) to remind licensees of the need to conform their facility to the current, site-specific licensing basis for tornado-generated missile protection. In addition the EGM stated, that upon reviewing the above-noted RIS, some licensees may discover that a TS-controlled SSC at their facility does not comply with the plants current licensing basis (CLB) and that an operability determination (or functional assessment) will be necessary. The EGM actions section specified that the NRC would exercise this enforcement discretion only when a licensee implements initial compensatory measures prior to the expiration of the time allowed by the LCO that provide additional protection such that the likelihood of tornado missile effects are lessened. The licensee initiated CR10087558 on 06/23/2015, to evaluate the RIS and conducted at least two walk-downs to identify tornado missile nonconformances. The licensee discovered potential nonconformances during these walk-downs and itemized them in a list. However, the licensee failed to identify all of these items as conditions adverse to quality (CAQs), in accordance with Appendix B, Criterion XVI. The team determined that the CAP required the evaluation of these items, CRs to document the nonconformances, and operability determinations for items affecting TS. Procedure NMP-GM-002, Corrective Action Program, Section 2, defined a condition adverse to quality in part, as an all-inclusive term used in reference to any of the following: ..., deficiencies, ..., and nonconformances potentially impacting Nuclear Safety. Nonconformances are deficiencies in characteristic, documentation, or procedure that renders the quality of an item or activity unacceptable or indeterminate. The team determined that, at the time of discovery, the itemized tornado missile vulnerabilities rendered the quality of SSCs indeterminate and thus a nonconformance in accordance with the definition in the procedure. Procedure NMP-GM-002-001, Corrective Action Program Instructions Section 4 specified that personnel should initiate a CR to identify an event, condition, problem, or process that needs correcting. (This included) nonconforming items. In addition, Section 4 specified to immediately contact the Shift Support Supervisor or Work Week Coordinator (Dispatcher) when a condition is discovered that has the potential to impact plant operation or reportability. (This included) equipment or process issues related to Technical Specifications (tech specs). The team noted that the licensee did not create any additional CRs for the itemized potential vulnerabilities as required by their corrective action instructions procedure. On October 4, 2016, the inspectors conducted plant walk downs of the SSCs selected in the CDBI inspection plan and identified potential tornado missile issues. These issues were previously highlighted as potential nonconformances by the licensee, but not identified as CAQs. As a result of these observations, the licensee initiated CRs: CR10291142, Unit 1 TDAFW Exhaust nonconformance CR10291143, Unit 2 TDAFW Exhaust nonconformance CR10291144, Unit 1 Condensate Storage Tanks nonconformance CR10291145, Unit 2 Condensate Storage Tanks nonconformance CR10291146, Unit 1 Main Steam Safety Valve Exhaust nonconformance CR10291148, Unit 2 Main Steam Safety Valves Exhaust nonconformance The licensee determined that the TDAFW Exhaust and Condensate Storage Tanks were not operable because of nonconformances with these components tornado missile protection design bases. Additionally, the licensee submitted a 10 CFR 50.72 notification report (52319) to the NRC in accordance with plant procedures and NRC requirements.
05000424/FIN-2016007-07Failure to Update the UFSAR with the Complete and Accurate Information2016Q4The 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-2016007-06Turbine Driven Auxiliary Feedwater (TDAFW) Pumps 1/2-1302- P4-001 and Motor Driven Auxiliary Feedwater (MDAFW) Pumps 1/2-1302-P4-002/0032016Q4The 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.
05000424/FIN-2017503-01Transposition Results in Significantly Different EAL Threshold Values2016Q4TBD: 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-2016007-01Failure to Verify Capability of EDGs under Maximum Voltage and Frequency2016Q4The 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-2016007-03Failure to Meet Isolation Requirements When Incorporating Non- Class 1E Components into Class 1E electrical Circuits2016Q4The 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-2016007-05Failure to Perform Periodic Testing Of Safety-Related Valve Interlocks2016Q4The 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-2016007-04Failure to Perform Required In-Service Testing of Unit 2 CST Swap over Valves2016Q4The 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.
05000425/FIN-2016004-01Failure to Implement Maintenance Procedure for Electrical Grayboot Connectors2016Q4(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-2016003-02Licensee-Identified Violation2016Q3Title 10 CFR Part 50.54(q)(2) required, in part, that a licensee shall follow and maintain the effectiveness of its emergency plan that meets the planning standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(4) required, in part, that a standard emergency classification and action level scheme is in use by the nuclear facility licensee. Contrary to these requirements, since 2008, emergency action level EAL HA1 #5 was not translated from the emergency plan to implementing procedure NMP-EP-110 GL03 (formally 91001-C) during the 2008 revision of the emergency plan. The licensee entered the issue into their corrective action program as CR 10251396. The inspectors determined that the finding was of very low safety significance (Green) because the finding constituted an ineffective EAL rather than a failed risk-significant planning standard.
05000424/FIN-2016403-01Security2016Q3
05000424/FIN-2016003-01Failure to Properly Implement Fire Door Inspections2016Q3An 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-2016403-02Licensee-Identified Violation2016Q3
05000425/FIN-2016002-01Failure to properly implement a maintenance procedure caused a Reactor Trip2016Q2A 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-2016502-01Failure to Adequately Maintain Emergency Response Facilities2016Q2The 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-2016407-01Security2016Q2
05000424/FIN-2015004-02Licensee-Identified Violation2015Q4Title 10 CFR Part 50.54(q)(2), required, in part, a licensee shall follow and maintain the effectiveness of its emergency plan that meet the standards of 10 CFR 50.47(b). 10 CFR 50.47(b)(4), required, in part, a standard emergency classification and action level scheme, the bases of which include facility and system effluent parameters, is in use by the nuclear facility licensee. Contrary to the above, from April 2002 to September 2015, the licensee failed to maintain the effectiveness of its emergency plan. Procedure 43014-C, Special Radiological Controls, version 53, specified non-conservative dose rates used to verify if RCS activity exceeded the threshold value for an emergency classification FA1 (Alert), loss of fuel clad barrier, in response to a chemical volume control system (CVCS) letdown process line high radiation alarm. The licensee entered this violation into the corrective action program as CR 10124780. The inspectors determined this violation was of very low safety significance (Green) because the finding did not constitute a failed risk significant planning standard (RSPS).
05000424/FIN-2015004-01Failure to Implement Preventive Maintenance Procedure for 7300 Process Protection and Control System Printed Circuit Board2015Q4A 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-2015003-03Unauthorized Entry into a High Radiation Area2015Q3A 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-2015003-04Licensee-Identified Violation2015Q3Technical Specifications 5.4.1.b, Procedures, required, in part, that written procedures shall be established, implemented, and maintained for the emergency operating procedures (EOPs) required to implement NUREG-0737, Clarification of TMI Action Plan Requirements, and Supplement 1 to NUREG-0737. Contrary to this requirement, as of August 15, 2007, the licensee failed to maintain EOPs required to implement the requirements of NUREG-0737 and NUREG-0737, Supplement 1. The licensee failed to maintain EOP 19100-C, ECA-0.0 Loss of All AC Power, version 39, consistent with revised Pressurized Water Reactor Owners Group/Westinghouse Owners Group emergency response guidelines that restricted the RCS cooldown rate to less than 100 degrees Fahrenheit per hour to prevent thermal shock to the reactor coolant pump (RCP) seals following a loss of all alternating current power (SB) event. The licensee entered this violation into the CAP as CR 10066747 and revised EOP 19100-C consistent with the updated guidance. A bounding detailed risk evaluation was performed by an NRC regional senior risk analyst (SRA) who determined the finding to be of very low risk significance (Green). The dominant result was a grid-related Loss of Offsite Power that then proceeds to an SBO event and RCP seal failure due to thermal shock.
05000424/FIN-2015003-02NRC Biennial Written Examinations did not Meet Qualitative Standards2015Q3An 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-2015003-05Licensee-Identified Violation2015Q310 CFR 55.21, Medical examination, states, in part, that a licensee shall have a medical examination by a physician every two years. Contrary to the above, on March 24, 2015, the licensee identified that a licensed operator did not complete the required biennial NRC medical examination by February 2015, which was the two year due date. The licensed operators requirement to have a medical examination was incorrectly removed from the licensees learning management system (LMS) database when the operator entered the initial license training program to upgrade to a senior operator. The inspectors determined that the violation was not greater than very low safety significance (Green) because the licensed operator was not actively performing licensed duties in the control room. This issue was entered in the licensees corrective action program as CR 10045159.
05000424/FIN-2015007-01Failure to Fully Close and Latch Plant Fire Doors2015Q3An 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-2015003-01Failure to Maintain Requalification Examination Integrity2015Q3An 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-2015007-02Failure to Identify and Repair a Degraded Fire Penetration Seal2015Q3An 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-2015404-01Security2015Q2
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