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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000275/FIN-2018003-02 | 2018Q3 | Diablo Canyon | 4 kV Vital Switchgear Room Ventilation Degraded or Non-Conforming Condition and Associated Compensatory Measure Not Corrected in a Timely Manner | The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because PG&E personnel failed to promptly correct a degraded or non-conforming condition associated with an open operability condition. Specifically, PG&E personnel did not promptly correct a degraded condition associated with an open operability determination and corresponding compensatory measure related to Unit 1 and Unit 2, 4 kV vital switchgear ventilation for a period of over 4 years. This time period included two refueling outages for Unit 1 and three refueling outages for Unit 2. |
| 05000482/FIN-2018003-01 | 2018Q3 | Wolf Creek | Failure to Correct Degraded Performance of a Safety-Related Tornado Damper | The inspectors identified a Green non-cited violation of 10 CFR Part 50, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality associated with a safety-related tornado damper. Specifically, damper GTD0002 failed tests in 2012 and 2015, and following maintenance on the damper in 2017, again failed its next as-found test on February 8, 2018. As a result, this safety-related tornado dampers ability to close during a design basis tornado event was adversely impacted. |
| 05000275/FIN-2018003-01 | 2018Q3 | Diablo Canyon | Multiple Examples of Scaffolding in Place Greater Than 90 Days Without Required Evaluation | The inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion V, Procedures, because PG&E personnel failed to follow the requirements of AD7.ID5, Scaffold Material Structure. Specifically, 20 instances of scaffold structures installed in the plant were identified that had been in place for greater than 90 days without required 10 CFR 50.59 reviews being completed. |
| 05000482/FIN-2018003-02 | 2018Q3 | Wolf Creek | Failure to Submit a Licensee Event Report for a Condition Prohibited by Technical Specifications | The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(2)(i)(B), because the licensee did not provide a written licensee event report (LER) to the NRC within 60 days. Specifically, the licensee did not provide a written LER to the NRC within 60 days of identifying a condition prohibited by the plants Technical Specifications associated with inoperability of control room emergency ventilation system train B for longer than its Technical Specification allowed outage time. As a result, the NRCs ability to regulate was impacted. |
| 05000483/FIN-2018002-05 | 2018Q2 | Callaway | Minor Violation | Contrary to Technical Specification 3.6.3, Containment Isolation Valves, the licensee failed to maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4. Specifically, the licensee failed to shut the reactor building service air header supply outer containment isolation valve KAV0118 after the fall 2017 refueling outage. As a result, isolation valve KAV0118 was left open from November 25, 2017, through January 11, 2018, which rendered the valves containment isolation function inoperable. The as-found testing demonstrated that the overall containment isolation function, for that penetration, was met with inner containment isolation valve KAV0039 in the normally shut position. Additional information can be found in Licensee Event Report 05000483/2018-001-00, Violation of 20 Technical Specification 3.6.3, Containment Isolation Manual Valve Found in Open Position (ADAMS Accession Number ML18071A208). The licensees failure to comply with Technical Specification 3.6.3, Containment Isolation Valves, and maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4 was a performance deficiency. Screening: The inspectors determined the performance deficiency was minor because it was not a precursor to a significant event, did not have the potential to lead to a more significant safety concern, did not relate to a performance indicator that would have exceeded a threshold and did not adversely impact any of the cornerstone objectives. Specifically, the as-found local leak rate testing demonstrated that containment isolation function was met with inner containment isolation valve KAV0039 in the normally shut position. Enforcement: The failure to comply with Technical Specification 3.6.3, Containment Isolation Valves, and maintain each containment isolation valve operable or enter applicable conditions and required actions for an inoperable containment isolation valve in Modes 1, 2, 3, and 4 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. |
| 05000483/FIN-2018002-04 | 2018Q2 | Callaway | Failure of an Analysis of the Impact of Changes to Emergency Action Levels to Demonstrate the Changes Did Not Reduce the Effectiveness of the Emergency Plan | The inspectors identified a non-cited violation of 10 CFR 50.54(q)(3) for the failure of an analysis of the impact of changes to licensee emergency action levels to demonstrate that the changes did not reduce the effectiveness of the emergency plan. |
| 05000483/FIN-2018002-03 | 2018Q2 | Callaway | Failure to Critique an Inaccurate Emergency Classification During a Simulator Training Scenario | The inspectors identified a non-cited violation of 10 CFR 50.47(b)(14) for the licensees failure to critique an inaccurate emergency classification made during licensed operator training. |
| 05000483/FIN-2018002-02 | 2018Q2 | Callaway | Failure to Establish Maintenance Procedures for Doors that Provide Safety-Related Functions | The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to establish, implement, and maintain procedures associated with door maintenance. Specifically, the licensee failed to establish, implement, and maintain maintenance procedures for doors that provide safety-related functions such as ventilation pressure boundaries. As a result, 15 safety-related doors were identified that either had degraded conditions or that did not have a periodic maintenance task to inspect the doors. |
| 05000483/FIN-2018002-01 | 2018Q2 | Callaway | Failure to Adequately Assess and Manage Risk Associated with Switchyard Work During a Planned Risk Significant Turbine-Driven Auxiliary Feedwater Pump Equipment Outage | The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at Nuclear Power Plants, for the licensees failure to adequately assess and manage risk associated with switchyard work during a planned risk significant turbine-driven auxiliary feedwater pump equipment outage. Specifically, the licensee failed to properly classify switchyard work and manage the risk as required by Procedures APA-ZZ-00322, Appendix F, Online Work Integrated Risk Management, Revision 16, and ODP-ZZ-00002, Appendix 2, Risk Management Actions for Planned Risk Significant Activities, Revision 13. |
| 05000445/FIN-2018002-03 | 2018Q2 | Comanche Peak | Failure to Incorporate Design Information Into System Test Procedures | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Controls, for the licensees failure to ensure that the stations in-service testing program for main steam isolation valves (MSIVs) incorporated the requirements and acceptance limits contained in applicable design documents. Specifically, the licensees in-service procedures did not direct testing of the valves be performed at the minimum required pressure and this resulted in the licensees failure to identify two degraded MSIVs during in-service testing. The licensee entered this issue into the corrective action program as Condition Report CR-2018-003229. |
| 05000445/FIN-2018002-02 | 2018Q2 | Comanche Peak | Unacceptable Preconditioning of Main Steam Isolation Valves | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Controls, for the licensees unacceptable preconditioning of the Unit 1 main steam isolations valves (MSIV) prior to performing as-found in-service stroke time testing. Specifically, the licensee raised accumulator pressure prior to stroke time testing and this potentially masked an issue with MSIV 1-01. The licensee entered this issue into the corrective action program as Condition Report CR-2018-002405. |
| 05000445/FIN-2018002-01 | 2018Q2 | Comanche Peak | Failure to Identify and Correct a Condition Adverse to Quality | The inspectors identified a Green,non-cited violation of 10CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to identify and correct a condition adverse to quality associated with unacceptable main steam isolation valve (MSIV) stroke times. Specifically, during stroke time testing of MSIV 2-02 the valves stroke time was outside of the acceptance limit and the licensee failed to determine why the stroke time was out of specification and correct the issue prior to declaring the valve operable and placing it in service. The licensee entered this issue into the corrective action program as Condition Report CR-2018-002189. |
| 05000416/FIN-2018002-08 | 2018Q2 | Grand Gulf | Performance of Surveillance Testing Following Maintenance on Containment Airlock | The inspectors identified a Green non-cited violation of 10CFRPart50,AppendixB, Criterion XI, Test Control, for the licensees failure to perform surveillance testing of containment airlock seals under appropriate conditions. The licensee failed to appropriately control the sequence of maintenance and testing activities to ensure that surveillance testing was not performed subsequent to maintenance which could affect the validity of surveillance test results. |
| 05000416/FIN-2018002-07 | 2018Q2 | Grand Gulf | Loss of Shutdown Cooling | A self-revealed,Green non-cited violation of Technical Specification 5.4, Procedures,for the licensees failure to follow written procedures was identified when the residual heat removal (RHR) system automatically isolated due to an inadvertent emergency core cooling system (ECCS) actuation. While the plant was shut down with the RHR system in decay heat removal mode, maintenance personnel inadvertently opened an incorrect valve during a transmitter calibration activity, which caused a false low reactor pressure vessel (RPV) water level signal, an ECCS actuation, and a loss of decay heat removal for approximately 31 minutes |
| 05000416/FIN-2018002-06 | 2018Q2 | Grand Gulf | Improper Evaluation and Resolution of Intermediate Range MonitorNoise Leads to Manual Reactor Shutdown | A self-revealed, Green non-cited violation of 10CFRPart50, Appendix B, Criterion XVI, Corrective Action, was identified for the failure of the licensee to identify and correct a condition adverse to quality. Specifically, the licensee failed to implement appropriate corrective actions related to intermediate range monitor (IRM) nuclear instrument (NI) electronic noise spiking. The failure to implement adequate corrective actions over the course of at least 5 years resulted in a plant shutdown due to declaration of multiple IRM channels inoperable while in Mode 2. |
| 05000416/FIN-2018002-05 | 2018Q2 | Grand Gulf | Failure to Follow Procedure Requirements Resulting in Unplanned Dose | A self-revealed, Green non-cited violation of Technical Specification 5.4.1 was identified when an individual alarmed a personnel contamination monitor upon exit from the radiologically controlled area. Specifically, the licensee failed to follow procedures to establish a decontamination plan or procedure, conduct a specific pre-job brief addressing appropriate contamination risk, and receive approval by radiation protection supervision prior to conducting decontamination activities on thereactor pressure vessel(RPV) O-rings |
| 05000416/FIN-2018002-04 | 2018Q2 | Grand Gulf | High Radiation Area Boundary Violation | A self-revealed, Green non-cited violation of Technical Specification 5.7.1 was identified when an individual received a dose rate alarm when the individual failed to comply with established radiological barriers and protective measures and entered a high radiation area. Specifically, an individual leaned over a high radiation area barricade rope, thereby entering the high radiation area. The individuals radiation work permit (RWP) did not permit entry into a high radiation area. |
| 05000416/FIN-2018002-03 | 2018Q2 | Grand Gulf | Failure to Adequately Test NUS Temperature Switch | A self-revealed,Green non-cited violationof 10CFRPart50, AppendixB, CriterionIII, Design Control, was identified when the reactor core isolation cooling (RCIC) system automatically isolated due to an inadvertent high temperature input from the leakage detection system. Specifically, the licensee failed to fully test a modification that installed a new type of temperature switches, and the system inappropriately isolated the RCIC system when a loss and subsequent restoration of power occurred. |
| 05000416/FIN-2018002-02 | 2018Q2 | Grand Gulf | Failure to Follow ASME Requirements for Maintaining Inservice Inspection (ISI) Cycles and Perform ASME Required Inservice Inspections within the Scheduled ISI Cycle | The inspector identified 15 examples of a Green non-cited violation (NCV)of 10 CFR 50.55(a)(g)(4)(ii), which requires that inservice examination of components classified as American Society of Mechanical Engineers (ASME), Section XI, Code Class 1, Class 2, and Class 3 be conducted during successive 120-month inspection intervals, and requires compliance with the requirements of the latest edition and addenda of the ASME Code (and all its paragraphs) applicable to the specific interval, including maintaining the 120-month inspection interval in accordance with the ASME Code, Section XI, Paragraph IWA-2430. Specifically, the licensee inappropriately adjusted its second inservice inspection 120-month cycle, and failed to perform VT-3 and MT examinations of 15 class 1, class 2, and class 3 components, including the high pressure core spray pump attachment weld and reinforcing band before the third inservice inspection cycle expired on November 30, 2017, as required by 10CFR50.55(a)(g)(4)(ii). |
| 05000416/FIN-2018002-01 | 2018Q2 | Grand Gulf | Failure to Institute Effective Corrective Action to Preclude Repetition | An NRC-identified,Green non-cited violation of 10CFRPart50, AppendixB, CriterionXVI, Corrective Action, was identified when the licensee failed to institute effective corrective actions to preclude repetition of a significant condition adverse to quality. Specifically, the licensee left a secondary containment personnel hatch in an open configuration for approximately 30 minutes while performing a roof inspection, which rendered secondary containment inoperable. This issue had also previously occurred in 2016, but corrective actions to prevent it from occurring again were ineffective. |
| 05000313/FIN-2018001-03 | 2018Q1 | Arkansas Nuclear | Licensee-Identified Violation | Title10CFR20.1501(a) requires that each licensee make or cause to be made surveys that may be necessary for the licensee to comply with the regulations in 10 CFR Part 20, and that are reasonable under the circumstances to evaluate the magnitude and extent of radiation levels, concentrations, or quantities of radioactive materials, and the potential radiological hazards that could be present.Contrary to the above, on August 7, 2017, the licensee failed to make necessary surveys of the Unit 2, 2T-15 tank room, that were reasonable to evaluate the magnitude and extent of radiation levels that could be present. Consequently, workers were allowed access to an area with dose rates up to 1000 millirem per hour at 30 cm without a proper briefing or oversight 17 Significance: Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because: (1) it was not associated with as low as is reasonably achievable (ALARA) planning or work controls; (2) there was no overexposure; (3) there was no substantial potential for an overexposure; and (4) the ability to assess dose was not compromised.Corrective Action Reference(s): CR-ANO-2-2017-04634 and CR-ANO-2-2017-0533 |
| 05000382/FIN-2018001-01 | 2018Q1 | Waterford | Failure to Obtain NRC Staff Authorization Prior to Changing a Procedure that Impacts Implementation of Technical Specifications | The inspectors identified a Severity Level IV, non-cited violation of 10CFR50.59, Changes, Tests, and Experiments, Section (c)(1), for the licensees failure to submit and obtain authorization prior to implementation procedures described in the Final Safety Analysis Report |
| 05000323/FIN-2018001-01 | 2018Q1 | Diablo Canyon | Improper Troubleshooting Results in Reactor Trip Signal and Loss of Source Range Nuclear Instrument Power | The inspectors reviewed a Green,self-revealed non-cited violation of Technical Specification 5.4.1.a Procedures, because PG&E personnel failed to follow the requirements of MA1.DC54, Conduct of Maintenance, Revision 15. Specifically, on March 20, 2018, with the reactor in Mode 3 during informal troubleshooting of high background count rate on source range nuclear instrument (NI) NI-32, PG&E personnel caused a short in NI cabinet B resulting in a blown fuse and the loss of power to the cabinet. This resulted in the loss of power to power range NI-42, intermediate range NI-36, source range NI-32, a reactor trip signal, a turbine trip signal, and all associated reactor protection interlocks. Power was automatically removed from the remaining source range NI due to reactor protection interlock P-10, resulting in no safety-related source range NI indication being available for control room operators. |
| 05000323/FIN-2018001-02 | 2018Q1 | Diablo Canyon | Failure to Follow Operating Experience Procedures Results in Inadequate Screen of Operating Experience Report | The inspectors identified a finding of very low safety significance (Green) because PG&E personnel failed to follow the requirements of OM4.ID3, Operating Experience Program, Revision 20. Specifically, PG&E personnel failed to screen relevant operating experience relating to a safety-related centrifugal charging pump (CCP) journal bearing failure due to non-metallic anti-rotation pin shear failure. This operating experience notice was received by PG&E September 2011 and was not screened per OM4.ID3, Operating Experience Program, preventing actions from being identified and implemented that could have eliminated vulnerabilities and prevented a similar event from occurring at DCPP. On November11,2017, CCP 2-1 was declared inoperable and determined to be non-functional due to a damaged journal bearing caused by non-metallic, anti-rotation pin shear failur |
| 05000368/FIN-2018001-02 | 2018Q1 | Arkansas Nuclear | Failure to Preplan and Perform Service Water Pre-Screen Maintenance | The inspectors reviewed a self-revealed,non-cited violation and associated finding of Arkansas Nuclear One, Unit 2, Technical Specification 6.4.1.a, for the licensees failure to properly preplan maintenance that can affect the performance of safety-related equipment. Specifically, the licensee failed to properly preplan pre-screen cleaning maintenance, causing the trainB service water system to become inoperable |
| 05000313/FIN-2018001-01 | 2018Q1 | Arkansas Nuclear | Failure to Establish Adequate Criteria for Flood Seal Testing | The inspectors identified a Green finding and associated non-cited violation of Unit1 Technical Specification 5.4.1.a and Unit 2 Technical Specification 6.4.1.a for the licensees failure to establish the criteria for ensuring the necessary conditions existed for a successful test of hatch flood seals. Specifically, Procedure OP 1402.240, Inspection of Watertight Hatches, Revision 1, did not contain adequate guidance to ensure that the auxiliary building was at a lower pressure than the turbine building such that puffing smoke on the turbine building side would allow a seal leak to be detectable. |
| 05000416/FIN-2017012-02 | 2017Q3 | Grand Gulf | Failure to Operate the Gaseous Radwaste System Within Design Specifications | The inspectors identified a finding associated with the licensees failure to operate the gaseous radwaste system within design specifications. These deficiencies in design specifications were associated with the off gas charcoal adsorber and vault refrigeration components of the gaseous radwaste system, which has impacted the systems reliability and efficiency since at least 2007. The design parameters for offgas flow rate into the charcoal adsorbers and vault refrigeration temperature were 30 scfm and 0 degrees Fahrenheit, respectively. In contrast, the gaseous radwaste system is being operated with an approximate flow rate is 80 scfm and vault refrigeration temperature is 15 degrees Fahrenheit. The licensee has developed a system improvement plan to address resolution of these issues during the next scheduled outages. This performance deficiency was entered into the licensees corrective action program as Condition Report CR-GGN-2017-06875. 5 The failure to operate the offgas gaseous radwaste system within design specifications, resulting in elevated radiological effluent releases, is a performance deficiency. The finding is more than minor because it is associated with the plant equipment attribute of the Public Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure adequate protection of public health and safety from exposure of radioactive materials released into the public domain as a result of routine civilian nuclear plant operation. Using Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because it involved the Effluent Release Program, it did not impair the ability to assess dose, and did not exceed the 10 CFR Part 50, Appendix I, or 10 CFR 20.1301(d) limits. The finding has a cross - cutting aspect in the area of problem identification and resolution, associated with the resolution component, because the licensee failed to take effective corrective actions in a timely manner to minimize the unreliability and inefficiency of the gaseous radwaste system (P.3). |
| 05000416/FIN-2017012-01 | 2017Q3 | Grand Gulf | Failure to Correct Instrument Calibration Process in a Timely Manner | The inspectors identified a violation of 10 CFR 20.1501(c) for the failure to properly calibrate installed radiation monitors using industry accepted calibration methods and tolerances. Specifically, since January 2012, the licensee failed to properly calibrate the following radiation monitors: main steam line, containment high range, and the drywell high range. This violation was originally entered into the licensees corrective action program in March 2015 as Condition Report CR-GGNS-2015-01832. However, in 2017, inspectors determined that subsequent to 2015, the licensee failed to implement corrective actions to properly calibrate the instruments. The licensee entered this repetitive issue into their corrective action process as Condition Report CR-GGN-2017-06826. The failure to properly calibrate radiation monitors is a performance deficiency. The performance deficiency is more than minor because it is associated with the cornerstone attribute of plant instrumentation and adversely affects the cornerstone objective to ensure adequate protection of employee health and safety during routine civilian nuclear reactor operation and is therefore a finding. Specifically, the failure to properly calibrate radiation monitors impacts the licensees ability to assess dose rates. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding to be of very low safety significance because it was not an as low as reasonably achievable (ALARA) issue, there was no overexposure or substantial potential for overexposure, and the licensees ability to assess dose was not compromised. This finding has a cross-cutting aspect in the resources component of the Problem Identification and Resolution area because the licensee did not ensure that effective corrective actions were implemented to address issues in a timely manner commensurate with the safety significance (P.3). |
| 05000397/FIN-2017010-01 | 2017Q3 | Columbia | Failure to Transfer Byproduct Material to a Disposal Facility in Accordance with the Terms of the Facilitys License | The inspectors reviewed a self-revealed non-cited violation of 10 CFR 30.41(b)(5) for the failure to transfer byproduct material to an authorized waste disposal facility in accordance with the terms of the facilitys license. Specifically, License Condition No. 21.C of the US Ecology license requires that all radwaste shall be packaged in such a manner that waste containers received at the facility do not show an increase in the external radiation levels as recorded on the manifest, within instrument tolerances. On July 20, 2017, Columbia Generating Station personnel transferred byproduct material to US Ecology for disposal (Shipment 17-51). The disposal facilitys surveys identified that the dose rate on contact with the waste liner was 1100 millirem per hour, whereas the manifest for this shipment recorded a dose rate of 12 millirem per hour. The licensees corrective actions included providing a corrected shipment manifest to US Ecology and proposed enhancements to the Columbia Generating Station procedures for shipping. This issue was documented in the licensees corrective acti on program as Action Request AR 00369215. The failure to transfer byproduct material to a low-level radwaste disposal facility in accordance with the facilitys license was a performance deficiency. The performance deficiency was more than minor because it was associated with the program and process attribute of the Public Radiation Safety Cornerstone and adversely affected the associated cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Using NRC Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined that the finding had very low safety significance (Green) because it was a low-level burial ground nonconformance; however, it was not Class C waste or greater and the waste did conform to the waste characteristics of 10 CFR 61.56. The finding has a cross-cutting aspect in the area of Human Performance, Resources, because licensee leaders failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety (H.1). |
| 05000458/FIN-2017003-01 | 2017Q3 | River Bend | Failure to Account for Delayed Closure of Isolation Valves in the Ultimate Heat Sink Inventory Analysis | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in Section 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to September 28, 2017, the licensees current calculation for assuring adequate ultimate heat sink inventory did not support the acceptability of the timing of a critical operator action in the abnormal operating procedure for the loss of standby service water. The potential safety consequence is that sufficient ultimate heat sink inventory might not be available to safely shut down the plant and maintain it in a cold shutdown condition for a 30-day period with no external makeup water source available. In response to this finding, the licensee performed an initial analysis and determined that the ultimate heat sink had sufficient inventory to account for the losses associated with the delayed closure of the normal service water return isolation valves and that the losses would likely be less than those previously calculated. This finding was entered into the licensee's corrective action program as Condition Report CR-RBS-2017-06998.The inspector determined that the failure to account for delayed closure of isolation valves in the ultimate heat sink inventory analysis was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in a condition where the current analysis to determine the acceptability of the ultimate heat sink with respect to the 30-day inventory requirement needed to be re-performed to assure that accident analysis requirements were met. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with design margins because the failure to account for delayed closure of isolation valves in the 30-day ultimate heat sink inventory analysis resulted in a significant reduction in the available margin (H.6). |
| 05000397/FIN-2017010-02 | 2017Q3 | Columbia | Failure to Control a High Radiation Area with Dose Rates Greater Than 1000 Millirem Per Hour at 30 Centimeters | The inspectors identified a non-cited violation of Technical Specification 5.7.2 for the failure to control a high radiation area with dose rates greater than 1000 millirem per hour at 30 centimeters. Specifically, equipment boxes placed adjacent to high radiation area barrier fencing in the reactor building 471 elevation, which created a natural ladder into the area, resulted in an uncontrolled entryway to a high radiation area with dose rates greater than 2500 millirem per hour. Once informed, the licensee immediately removed the equipment boxes from the barrier and added signage that restricted the placement of any items adjacent to the fencing forming the high radiation area barrier. This issue was documented in the licensees corrective action program as Action Request AR 00355646. The failure to properly control and barricade an entryway to a locked high radiation area in the reactor building, 471' elevation, was a performance deficiency. The performance deficiency was more than minor because it was associated with the program and process (exposure control) attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material. Using NRC Inspection Manual Chapter 0609, Appendix C, "Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding had very low safety significance (Green) because the finding was not an as low as reasonably achievable planning or work control issue, there was no overexposure or potential for an overexposure, and the licensee's ability to assess dose was not compromised. The finding had a cross- cutting aspect in the area of Human Performance, Field Presence, because leaders were not commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations, resulting in a lack of oversight of work activities, to include contractors and supplemental personnel (H.2). |
| 05000458/FIN-2017003-02 | 2017Q3 | River Bend | Manual Reactor Scram Initiated in Response to Increase in Steam Pressure during Steam Leak Troubleshooting | The inspectors reviewed a self-revealed finding for the licensees failure to properly complete steps of an approved procedure during the installation of a modification to the turbine electro-hydraulic control system. Specifically, the licensee failed to properly install a tee connection in a steam supply line to turbine pressure transmitters in the system, creating conditions for an eventual steam leak that led to a reactor scram. Corrective actions included properly installing the tee connection and writing specific procedural guidance on compression fitting inspection, installation, remake, and repair (CR-RBS-2017-02405).The failure to properly complete steps of an approved procedure during the installation of a modification to the turbine electro-hydraulic control system was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensees failure to properly install the tee connection caused a steam leak that led to a reactor scram. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The finding had a cross-cutting aspect in the area of human performance, work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority (H.5). |
| 05000482/FIN-2017002-02 | 2017Q2 | Wolf Creek | Failure to Declare Train A Component Cooling Water Inoperable | The inspectors identified a Green non-cited violation of Technical Specification Limiting Condition for Operation 3.7.7 for the licensees failure to place the unit in MODE 3 within 78 hours with the train A component cooling water system inoperable. Specifically, the essential service water emergency make-up to component cooling water train A valve was not declared inoperable when it was out of service, and as a result, train A component cooling water was out of service for longer than its Technical Specification allowed outage time. The licensees planned actions include revising Technical Specification Bases 3.7.7 and training operators on the proposed Technical Specification Bases revisions, and the licensee issued an Essential Reading document for operators to review. The licensee entered the issue into the corrective action program as Condition Report 111808. The failure to declare train A component cooling water inoperable is a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is 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 inspectors evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, and determined the finding was of very low safety significance (Green). The inspectors determined that the finding has a human performance cross-cutting aspect in the area of challenge the unknown because individuals did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding. This issue is indicative of current performance because the creation and implementation of the subject clearance order occurred in the last three years (H.11). |
| 05000313/FIN-2017002-03 | 2017Q2 | Arkansas Nuclear | Failure to Comply with ECCS Technical Speci fications | Green . The inspectors reviewed a Green self -revealing finding and associated non -cited violation of Unit 1 Technical Specification 3.5.2, Emergency Core Cooling System (ECCS) Operating, for the licensees failure to ensure the operability of the P36A high pressure injection pump after reinstalling its feeder breaker during a unit outage. A violation of Unit 1 Technical Specification 3.0.4 was also identified for making a mode change without meeting the requirements to do so. Following unit restart, the pump failed to start during routine equipment rotation, resulting in one train of emergency core cooling system being inoperable for long er than allowed by Unit 1 Technical Specifications. The licensee subsequently identified that the feeder breaker had not been fully racked into position. Inspectors also noted that the breaker had been racked in manually rather than using the normal electric racking tool, and no special precautions had been taken to ensure this infrequently -used method was successful. When the breaker was correctly racked in, the pump was satisfactorily tested. The licensee subsequently verified that all similar breakers were correctly racked into position. The licensee entered this issue into their corrective action program as Condition Report CR- ANO -1-2017- 01764. The inspectors determined that the failure to verify that the P36A high pressure injection pump was operable after racking its feeder breaker into the switchgear cubicle was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. 4 The inspectors performed the initial significance determination for the performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012 , and concluded that it required a detailed risk evaluation because it involved the loss of a single train of mitigating equipment for longer than the technical specification allowed outage time. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The estimate in the increase in core damage frequency is 4.4 E-8 per year, or of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because the licensee failed to ensure that individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to verify that the pump was operable after its breaker was rein stalled, even though an infrequently-used method was employed (H.12). |
| 05000482/FIN-2017002-03 | 2017Q2 | Wolf Creek | Enforcement Action EA-17-064, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliances | Title 10 CFR Part 50, Appendix A, General Design Criteria for Nuclear Power Plants, Criterion 2, Design Bases for Protection Against Natural Phenomena, states, in part, that SSCs important to safety shall be designed to withstand the effects of natural phenomena, such as tornadoes. Criterion 4, Environmental and Dynamic Effects Design Basis, states, in part, that SSCs important to safety shall be appropriately protected against dynamic effects including missiles that may result from events and conditions outside the nuclear power unit. Section 9.5.4.1.1, Safety Design Bases, of the Updated Safety Analysis Report describes Safety Design Basis One for the emergency diesel engine fuel storage tank system, (It) is protected from the effects of natural phenomena, such as earthquakes, tornadoes, hurricanes, floods, and external missiles ((General Design Criteria)-2). On April 5, 2017, the licensee reevaluated operating experience that was initially entered into the corrective action program and evaluated on March 14, 2017, concerning a low-probability scenario where one or more tornado-generated missiles could impact the emergency fuel oil truck connection lines on the south wall of the diesel generator building. The two non-safety-related connection lines are each connected to the safety-related normal fuel oil transfer lines via a tee connection and a normally closed isolation valve. Direct impact by a tornado-generated missile to either trains truck connection line could impart a load that has not been evaluated on the tee connection to the fuel oil transfer line. Failure of the tee connection could result in the associated emergency diesel generator being incapable of performing its safety function.The licensee concluded that a potential unanalyzed condition prohibited by Technical Specifications existed for emergency diesel generator fuel transfer line connections, as described in Condition Report 112131 and in LER 2017-002-00, Tornado Missile Vulnerabilities Result in Condition Prohibited by Technical Specifications. On February 7, 2017, the NRC issued Enforcement Guidance Memorandum (EGM) 15-002, Enforcement Discretion for Tornado-Generated Missile Protection Noncompliance, Revision 1 (ADAMS Accession Number ML16355A286). The EGM referenced a bounding generic risk analysis performed by the NRC staff that concluded that tornado missile vulnerabilities pose a low risk significance to operating nuclear plants. Because of this, the EGM described the conditions under which the NRC staff may exercise enforcement discretion for noncompliance with the current licensing basis for tornado-generated missile protection. Specifically, if the licensee could not meet the technical specification required actions within the required completion time, the EGM allows the staff to exercise enforcement discretion provided the licensee implements initial compensatory measures prior to the expiration of the time allowed by the limiting condition for operation. The compensatory actions should provide additional protection such that the likelihood of tornado missile effects are lessened. The EGM then requires the licensee to implement more comprehensive compensatory measures within approximately 60 days of issue discovery. The compensatory measures must remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Because EGM 15-002 listed Wolf Creek as a Group A plant, enforcement discretion will expire on June 10, 2018. The licensee declared both diesel generators inoperable, complied with the applicable technical specification action statements, initiated condition report 112131, invoked the enforcement discretion guidance, implemented prompt compensatory measures, and returned the SSCs to an operable-degraded/non-conforming status. The licensee instituted compensatory measures intended to reduce the likelihood of tornado missile effects. These included verifying that guidance was in place for severe weather procedures, abnormal and emergency operating procedures, and FLEX support guidelines, that training on these procedures was current, and that a heightened level of awareness of the vulnerability was established.Enforcement. Technical Specification 3.8.1 requires, in part, that two diesel generators capable of supplying the onsite Class 1E power distribution subsystem(s) shall be operable and one of the two out of service diesel generators be restored to operable status within 2 hours, or the reactor must be in MODE 3 in an additional 6 hours. Contrary to the above, prior to April 5, 2017, two diesel generators capable of supplying the onsite Class 1E power distribution subsystem(s) were not operable and neither one of the two out of service diesel generators was restored to operable status within 2 hoursnor the reactor placed in MODE 3 in an additional 6 hours. Specifically, the emergency diesel generator fuel oil transfer lines were not designed to withstand the effects of natural phenomena, such as tornadoes. Licensee Event Report 2017-002-00 described the licensees corrective actions, including eliminating the tornado missile vulnerability by completing Design Change Package 15264, which cut, plugged, and covered the emergency fuel oil truck connection lines with 7/8 inch thick carbon steel plates. The inspectors verified through inspection sampling that the EGM 15-002 criteria were metand that the issue was documented in Condition Reports 111624, 111625, and 112131. Therefore, the NRC exercised enforcement discretion (Enforcement Action (EA)-17-064) in accordance with Section 3.2 of the Enforcement Policy because the violation involves an old design issue that was identified by the licensee as a result of a voluntary initiative, was corrected, and was unlikely to be identified by efforts such as normal surveillances or routinely scheduled quality assurance activities. |
| 05000313/FIN-2017002-04 | 2017Q2 | Arkansas Nuclear | Licensee-Identified Violation | Title 10 CFR 50.55a(g)4, Inservice Inspection Standards Requirement for Operating Plants, states in part, Throughout the service life of a pressurized water -cooled nuclear power facility, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the ASME Code. The ASME Section XI, Article IWA - 2610, requires that all welds and components subject to a surface or volumetric examination be included in the licensees inservice inspection program. This includes identifying system supports in the inservice inspection plan, per ASME Section XI, Article IWA -1310. Contrary to the above, prior to March 9, 2017, the licensee did not ensure that all welds and components subject to a surface or volumetric examination were included in the licensees inservice inspection. Specifically, the licensee did not apply the applicable inservice inspection requirements for surface or volumetric examination to all portions of the Unit 2 emergency feedwater system within the system ASME Code Class 3 boundary. The licensee identified that they failed to include the emergency feed pump supports in their inservice inspection program. The licensee entered this issue into their corrective action program as Condition Report CR- ANO -2-2016 -01023 and reasonably determined the emergency feedwater system remained operable. The licensee restored compliance by inspecting the supports, with no degradation identified, and entering the emergency feedwater pump supports into the ASME Section XI program. The finding was of very low safety significance (Green) because the finding did not 34 represent an actual loss of safety function of a system or train and did not result in the loss of a single train for greater than technical specification allowed outage time. This issue was entered into the licensees corrective action program as Condition Report CR- ANO -2-2016- 01023. |
| 05000368/FIN-2017002-01 | 2017Q2 | Arkansas Nuclear | Failure to Follow Fire Protection Program Procedures | Green . The inspectors identified a finding and associated non -cited violation of License Conditions 2.C.( 3)(b), Fire Protection, for Arkansas Nuclear One Unit 2, associated with the failure to adequately implement the fire protection program. Specifically, the licensee failed to follow the requirements for control of flammable liquid lockers and compressed hydrogen gas cylinders. The licensee immediately removed the hydrogen cylinders and stored them in an approved location and began processing the flammable liquid lockers through the design change process. The licensee entered these issues into their corrective action program as Condition Reports CR -ANO -2-2017- 01525 and CR -ANO -C-2017 -01508 . The failure to properly control transient combustible material in accordance with the approved fire protection program was a performance deficiency. The finding was considered more than minor because storing unanalyzed flammable material could result in the potential to exceed combustible material limits , and is associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to follow procedures resulted in conditions that increased the risk of fire which could upset plant stability and challenge critical safety functions. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and assigned the finding to the Fire Prevention and Administrative Controls category; because it affected the licensees combustible materials control. The finding was determined to be Green, or very low safety significance, in accordance with Inspection Manual Chapter 0609, Appendix F, Question 1.3.1, because the reactor would have been able to reach and maintain safe shutdown since the postulated fires would not have affected both trains of safe shutdown equipment . This finding had a cross -cutting aspect associated with teamwork within the human performance area since multiple groups in the licensee staff were involved in the decisions that resulted in the improper introduction of the flammable liquids lockers and the improper storage of the hydrogen cylinders (H.4). |
| 05000368/FIN-2017002-02 | 2017Q2 | Arkansas Nuclear | Failure to Install Set Screw Leads to Breaker Failure | Green . The inspectors documented a Green self -revealing finding and associated non- cited violation of Unit 2 Technical Specification 6.4.1.a, for failure to properly pre-plan and perform maintenance on the Unit 2 containment spray pump B breaker in accordance with written procedures. Specifically, the licensee failed to install a cam shaft set screw during the breakers last overhaul. The cam eventually became displaced on the shaft, and the breaker failed to close. To correct the issue, the licensee replaced the breaker and installed a cam shaft set screw in the failed breaker. The licensee also inspected all other similar breakers to verify the cams were properly secured. The licensee entered the issue in to their corrective action program as Condition Report CR -ANO -2-2017- 03168. The failure to install a cam shaft set screw during the overhaul of the Unit 2 containment spray pump B breaker is a performance deficiency. The performance deficiency is more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the failure of a Unit 2 containment spray pump breaker. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was not a design or qualification deficiency; did not represent a loss of system; did not result in the actual loss of function of a train of technical specification equipment for greater than its allowed outage time; and did not screen as potentially risk significant due to seismic, flooding, or severe weather events. The inspectors determined this finding did not have a cross -cutting aspect because the most significant contributor did not reflect current licensee performance. Specifically, the error occurred during the breakers last overhaul, which occurred in 2011 |
| 05000298/FIN-2017002-01 | 2017Q2 | Cooper | Failure to Assess Operability of Technical Specification System Functions during Surveillance Testing | Green . The inspectors identified a non- cited violation of Technical Specification 5.4.1.a, for the licensees fail ure to follow Station Procedure 0.26, Surveillance Program, Revision 70, and to assess the operability of alternate shutdown reactor pressure instrumentation during surveillance testing. Specifically, the licensee failed to assess the operability of the hig h pressure coolant injection turbine steam inlet pressure instrument that provides indications of reactor pressure for the alternate shutdown panel when the instrument was isolated during surveillance testing. As a result, operations personnel failed to r ecognize that the instrument was inoperable and failed to enter the appropriate technical specification action statements . As immediate corrective actions, the licensee validated that the alternate shutdown reactor pressure function was inoperable and that Technical Specification 3.3.3.2, Altern ate Shutdown System, Condition A, should have been entered, and generated a procedure change request to ensure T echnical Specification 3.3.3.2 would be entered during future surveillances . The licensee entered this deficiency into the corrective action program as Condition Report CR -CNS -2017- 02280. The licensees failure to assess the operability of alternate shutdown reactor pressure instrument ation when the high pressure coolant injection turbine inlet steam pr essure instrument was isolated for surveillance testing, in violation of Station Procedure 0.26, was a performance deficiency. The performance deficiency was determined to be more than minor , and therefore a finding, because it was associated with the hum an 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. Specifically, the alternate shutdown reactor pressure instrument was inoperable when the high pressure coolant injection turbine inlet pressure instrument was isolated for surveillance testing, and the appropriate technical specification action statement was not entered. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it: was not a design deficiency; did not repr esent a loss of system and/or function; did not represent an 3 actual loss of function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety -significant nontechnical specification train. The finding had a cross -cutting aspect in the area of human performance associated with work management. Specifically, the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the identification and management of risk commensurate with the isolation of the high pressure coolant injection turbine inlet pressure instrument during surveillance testing (H.5). |
| 05000298/FIN-2017002-02 | 2017Q2 | Cooper | Loss of Control Room Ventilation Due to Ineffective Preventive Maintenance Strategy | Green . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a , for the licensees failure to maintain work order instructions for control room supply fan maintenance resulting in the loss of the control room emergency filtration system. Specifically, prior to October 23, 2016, work order instructions for periodic preventive maintenance on the SF- C-1A supply fan failed to include industry recommended checks to ensure that the bearings were adequately engaged with the fan shaft, and failed to include proper work sequencing to ensure vibration data trending was meaningful. The ineffective preventive maintenance strategy resulted in the failure of the control room supply fan i nboard bearing during operation and a loss of the control room emergency filtration system function. Corrective actions to restore compliance included repair of the s upply fan and changes to improve the effectiveness of the fans preventive maintenance strategy. The licensee entered this deficiency into the corrective action program as Condition Report CR- CNS -2016- 07426. The licensees failure to maintain work order instructions for control room supply fan maintenance , in violation of Technical Specification 5.4.1.a , was a performance deficiency. The performance deficiency was more than minor , and therefore a finding, because it was associated with the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers ( control room envelope) protect the public fro m radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Process Phase 1 Initial Screening and Characterization of Findings, dated May 9, 2014, the inspectors determined that the finding had very low safety significance (Green) because the inspectors answered no to all of the Barrier Integrity screening questions. The finding had a cross -cutting aspect in the area of human performance associated with resourc es, because the licensee failed to ensure that personnel, equipment, procedures, and other resources we re available and adequate to support nuclear safety (H.1). |
| 05000298/FIN-2017002-03 | 2017Q2 | Cooper | Loss of Control Room Ventilation Due to Improper Switch Manipulation | The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a , for the licensees f ailure to implement System Operating Procedure 2.2.38, HVAC Control Building, Revision 43, during control building ventilation testing. Specifically, on December 7, 2016, when directed to turn off control building ventilation recirculation fan, RF- C-1A, operations personnel instead inadvertently turned off the operating control room emergency filtration system supply fan, 1 -SF -C-1A, resulting in the loss of the control room emergency filtration system function. Corrective actions to restore compliance included restoration of the control room emergency filtration supply fan and procedure changes to require peer checks for this surveillance test and similar 4 activities. The licensee entered this deficiency into the corrective action program as Condition Report CR -CNS -2016- 08744. The licensees failure to implement System Operating Procedure 2.2.38 , in violation of Technical Specification 5.4.1.a , was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers ( control room envelope) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The finding had a cross -cutting aspect in the area of human performance associated with challenge the unknown, because the licensee did not stop when faced with uncertain conditions, and did not ensure that risks we re evaluated and managed before proceeding. Specifically, despite noting several a bnormalities with the switch being manipulated, operations personnel did not stop to evaluate the uncertain conditions nor did they evaluate the risks associated with proceeding (H.11). |
| 05000382/FIN-2017002-01 | 2017Q2 | Waterford | Failure to Prepare the Site for Impending Adverse Weather | The inspectors identified multiple examples of a non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, for the licensees failure to follow Licensee Procedure OP-901-521, Severe Weather and Flooding, Revision 323. Specifically, on three occasions, the licensee did not close exterior doors when required by the procedure due to potential severe weather conditions. As a result, plant equipment was at an increased failure risk due to severe weather at the site. The licensee entered this condition into their corrective action program as Condition Reports CR-WF3-2017-03961 and CR-WF3-2017-04944. The licensee is planning corrective actions to ensure doors do not remain blocked open during conditions that require their closure.The performance deficiency was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to maintain all of the doors required by Licensee Procedure OP-901-521 with all fuel offloaded to the spent fuel pool threatened the licensees ability to maintain the functionality of the spent fuel pool cooling system. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process, and determined that a qualitative analysis by a senior reactor analyst was required. The senior reactor analyst determined that the finding was of very low safety significance (Green). Using Inspection Manual Chapter 0609, Appendix M, Signifiance Determination Process Using Qualitative Criteria, the senior reactor analyst performed a bounding analysis indicated that the total increase in core damage frequency from the failure to close the doors during severe weather was less than 1E-6. The finding had a work management cross-cutting aspect in the area of human performance because the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority and the work process did not include the identification and management of risk commensurate to the work and the need for coordination with different groups of job activities. Specifically, during the planning and executing of work activities associated with Refueling Outage 21, the licensee did not consider the nuclear safety implications of blocking open exterior watertight and tornado doors and the work process did not include the identification and management of the risk associated with the blocked-open doors (H.5). |
| 05000298/FIN-2017002-04 | 2017Q2 | Cooper | Licensee-Identified Violation | Technical Specification 5.7.1 states, in part, that high radiation areas w ith dose rates greater than 0.1 rem/hr at 30 centimeters shall be barricaded and conspicuously posted as a high radiation area. Contrary to the above, on November 2, 2016, a high radiation area with does rates greater than 0.1 rem/hr at 30 centimeters was not barricaded and conspicuously posted as a high radiation area. Specifically, a radiation protection technician (RPT) identified an unposted high radiation area at the control rod drive (CRD) A pump filter area on r eactor building 881 feet southea st quadrant. D ose rates of 120 mrem/hr at 30 centimeters from the CRD filter were identified. This issue was identified as a result of a RPTs deliberate and focused observations during the course of performing their normal duties of performing radiological surveys. The licensee documented this issue in the corrective action program as Condition Report CR- CNS -2016 -00788. The finding was determined to be of very low safety significance (Green) because it was not an ALARA planning issue, there was no overexposure or potential for overexposure, and the licensees ability to assess dose was not compromised. |
| 05000382/FIN-2017002-02 | 2017Q2 | Waterford | Failure to Ensure Containment Equipment Hatch Closure Prior to RCS Time to Boil | The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, which occurred because the licensee did not implement instructions for maintaining containment integrity. Specifically, on April 18, 2017, the licensee did not ensure that the containment equipment hatch could be closed within the calculated reactor coolant system time to boil as required by Licensee Procedure OP-010-006, Outage Operations, Revision 330. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-02541. The licensees corrective actions included exiting the applicable condition, re-performing the equipment hatch closure drill to show the equipment hatch could be closed prior to the reactor coolant system time to boil, and performing repairs to the containment equipment hatch. The performance deficiency was more than minor because it was associated with thehuman performance attribute of the Barrier Integrity Cornerstone and adversely affected its objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the licensee must close containment penetrations prior to the reactor coolant system time to boil in order to minimize radionuclide releases under accident conditions. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, instructed the inspectors to use Appendix H, Containment Integrity Significance Determination Process, the inspectors determined the finding to be of very low safety significance (Green) because licensee maintained in-depth shutdown capability and because the duration of the performance deficiency was less than 8 hours. The inspectors concluded that the finding had a teamwork cross-cutting aspect in the area of human performance because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, personnel performed work resulting in a short calculated reactor coolant system time to boil without first communicating their actions to operations or the outage control center, resulting in an unexpected plant condition (H.4). |
| 05000382/FIN-2017002-03 | 2017Q2 | Waterford | Failure to Ensure Appropriate Testing of TSP Baskets Inside Containment | The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to assure that testing required to demonstrate that structures, systems, and components will perform satisfactorily while in service was identified and performed in accordance with written test procedures incorporating the requirements and acceptance limits contained in the applicable design documents. Specifically, prior to performing Licensee Procedure OP-903-027, Inspection of Containment, Attachment 10.3, Trisodium Phosphate Storage Basket Inspection, the licensee routinely performed a preliminary check to fill the trisodium phosphate storagebaskets, thereby ensuring the successful completion of the technical specification-required surveillance. As a result, following unsatisfactory preliminary checks, the trisodium phosphate storage baskets were not evaluated for past operability. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05108. The licensees corrective actions will include performing the surveillance procedure as an as-found check and evaluating failed surveillances for past operability.The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, conducting preliminary checks of the trisodium phosphate storage baskets and refilling them prior to performing the technical specification surveillance can mask the as-found condition of the test and preclude an evaluation of past operability if the levels are below the technical specification-required values. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, Attachment 1, Exhibit 3, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; (2) did not represent a loss of system safety function; (3) did not represent an actual loss of safety function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; (4) with the cavity flooded, it did not represent an actual loss of safety function of one or more nontechnicalspecification trains of equipment during shutdown designated as risk-significant, for greater than 24 hours; (5) did not degrade the reactor coolant system level indication and/or core exit thermal couples when the cavity was not flooded; (6) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event; (7) did not involve fire brigade training and qualification requirements, or brigade staffing; (8) did not involve the response time of the fire brigade to a fire, and; (9) did not involve fire extinguishers, fire hoses, or fire hose stations. The finding had a change management cross-cutting aspect in the area of human performance because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, when the licensee implemented the preliminary check practice in 2012, they did not evaluate the unintended consequences of how that practice would impact the results of the technicalspecification surveillance. Additionally, the licensee performed the preliminary check during each successive refueling outage between 2012 and 2017 giving the licensee an opportunity to identify the improper practice. As a result, the inspectors concluded this performance deficiency was indicative of current performance (H.3). |
| 05000382/FIN-2017002-04 | 2017Q2 | Waterford | Failure to Perform a Post Maintenance Test on a Main Steam Isolation Valve Solenoid Valve | The inspectors identified a non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, for the licensees failure to perform operability testing on a safety-related component. Specifically, following the coil replacement of main steam isolation valve 2 solenoid valve, a safety-related component, the licensee did not perform a retest of the solenoid valve. As a result, main steam isolation valve 2 was returned to service without the assurance that no new deficiencies had been introduced, calling into question its operability. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-05507. The licensees corrective action was to perform a voltage check of the solenoid valve to ensure it would energize in the event that a main steam isolation valve 2 closure was needed.The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee restored main steam isolation valve 2 to an operable status without ensuring that its solenoid valve, which is a main steam isolation valve support system, was properly retested following maintenance.The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix A, Significance Determination Process for Findings At-Power. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with licensees maintenance rule program for greater than 24 hours.The finding had a conservative bias cross-cutting aspect in the area of human performance because individuals did not use decision making-practices that emphasized prudent choices over those that were simply allowable. Specifically, the licensee did not make a conservative decision when determining whether the main steam isolation valve or its solenoid valve should be tested prior to proceeding with plant startup (H.14). |
| 05000382/FIN-2017002-05 | 2017Q2 | Waterford | Failure to Perform Maintenance on the Correct Safety-Related Component | The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 6.8, Procedures and Programs, and Regulatory Guide 1.33, Quality Assurance Program Requirements, which occurred due to the licensees failure to perform field work on reactor coolant loop 2 shutdown cooling warm-up valve, SI-135A. Specifically, mechanical maintenance technicians, who were assigned work on safety injection train A, erroneously performed work on safety injection train B on reactor coolant loop 1 shutdown cooling warm-up valve, SI-135B. As a result, both trains of emergency core cooling systems were simultaneously inoperable, which placed the plant in a 1-hour technical specification shutdown action statement. The licensee entered this condition into their corrective action program as Condition Report CR-WF3-2017-01433. The licensees corrective actions included a revision of the model work order to require concurrent verification for component identification, and adding the valves to the protected equipment list for when the opposite train is inoperable.The performance deficiency was more than minor because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely affected its objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the mechanics worked on valve SI-135B instead of valve SI-135A, they simultaneously made both trains of emergency core cooling systems inoperable. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, instructed the inspectors to use Appendix A, Significance Determination Process for Findings At-Power. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, and component; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems out-of-service for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with licensees maintenance rule program for greater than 24 hours.The finding had an avoid complacency cross-cutting aspect in the area of human performance because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, and did notimplement appropriate error reduction tools. Specifically, maintenance technicians repeatedly visited the incorrect work location and didnt properly verify the valve number to ensure they would work on the correct component (H.12). |
| 05000382/FIN-2017002-06 | 2017Q2 | Waterford | Licensee-Identified Violation | Licensee Audit LO-WLO-2016-00037, Bioassay Program, dated November 21, 2016, identified that during Refueling Outage 20, staff reviewing air sample and lapel air sampler results had not been identifying positive results. The audit revealed that two positive lapel air samples from Refueling Outage 20 had not been identified nor had estimated personnel exposures been calculated. In addition, the audit identified seven positive air sample results which had no documented estimated exposures. As a result, dose was not assigned to individuals exposed to airborne radioactivity. As a result of the audit findings, the licensee retroactively assigned dose to three individuals working the October 25, 2015, cavity drain job in the amount of 36 mrem committed effective dose equivalent (CEDE) and 700 mrem committed dose equivalent (CDE) to bone surfaces and to one individual working on a November 8, 2015, decontamination job in theamount of 33 mrem CEDE and 661 mrem CDE to bone surfaces.Title 10 CFR 20.1703 states, in part, the licensee shall implement and maintain a respiratory protection program that includes: (1) air sampling sufficient to identify the potential hazard and estimate doses, and (2) surveys and bioassays, as necessary, to evaluate actual intakes.Contrary to the above, on November 21, 2016, the licensee failed to implement and maintain their respiratory protection program to include air sampling sufficient to identify the potential hazard and estimate doses, and surveys and bioassays, as necessary to evaluate actual intakes. Specifically, for two jobs and four individuals, the licensee failed to identify positive air sample results and assign internal dose to the subject individuals.In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, the inspectors determined that the performance deficiency was more than minor. The finding adversely affected the Occupational Radiation Safety Cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the failure to adequately assess internal exposure affects the licensees ability to control and limit radiation exposure to the worker. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green) because it did not involve: (1) as low as reasonably achievable (ALARA) planning and controls; (2) a radiological overexposure; (3) a substantial potential for an exposure; or (4) a compromised ability to assess the dose.The licensees immediate corrective action was to coach all technicians on surveying airborne areas, ensure all air sample and lapel results were discussed with management, and count all air and lapel samples for alpha and beta to evaluate any potential internal radiation exposure. The licensee entered this issue into their corrective action program as Condition Report CR-WF3-2016-07300. |
| 05000482/FIN-2017002-01 | 2017Q2 | Wolf Creek | Failure to Ensure Safety-Related Valves were Adequately Protected from Internal Flooding Hazards | The inspectors identified a Green non-cited violation of 10 Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish adequate measures to ensure that safety-related components remained capable of performing their functions. Specifically, the licensee did not have adequate preventive maintenance or testing tasks established to provide reasonable assurance that floor drains would not become clogged and impact the ability of train A safety-related components to perform their expected functions. As a result, a containment isolation valve was not adequately protected. The stations immediate corrective actions included entering the condition into the corrective action program, declaring the subject valves inoperable, and cleaning the debris from the clogged floor drains. The licensee created Work Order 17-429068-000 to evaluate and establish new preventive maintenance tasks for floor drains, and the licensee is continuing with, but had not yet completed, the remainder of the floor drain inspections for other safety-related areas.The failure to establish adequate measures to ensure that floor drains in safety-related areas remained free of debris and safety-related components remained capable of performing their function is a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using Exhibit 3, Barrier Integrity Screening Questions, of Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, and determined this finding was of very low safety significance (Green). The inspectors determined that the finding has a problem identification and resolution cross-cutting aspect in the area of identification because individuals did not identify issues completely, accurately, and in a timely manner in accordance with the program. Condition Report 90879, documented in January 2015, was an opportunity for the licensee to identify the inadequacy of the floor drain preventive maintenance and testing strategy and reflects current performance (P.1). |
| 05000397/FIN-2016009-04 | 2017Q1 | Columbia | Failure to Provide an Accurate Shipping Manifest | Green. The team identified a non-cited violation of 10 CFR 20.2006(b) for the licensees failure to ship radwaste with an accurate shipping manifest. Specifically, the licensee failed to provide the correct identification number and proper shipping name, radionuclide activity, net waste volume, surface radiation level, and waste classification. The incorrect surface radiation levels resulted in rejection of the package and the licensees immediate suspension from usage of the land disposal site at US Ecology. This issue was entered into the corrective action program as Action Requests 357593 and 359498. The licensees failure to ship radwaste intended for ultimate disposal with an accurate shipping manifest was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the program and process attribute of the Public Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive material released in the public domain. Specifically, inaccurate information on a shipping manifest could result in inappropriate handling of radioactive material while in the public domain. Using NRC Inspection Manual Chapter 0609, Appendix D, Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because: (1) radiation limits were not exceeded, (2) there was no breach of a package during transit, (3) it did not involve a certificate of compliance issue, (4) it was not a low-level burial ground nonconformance, and (5) it did not involve a failure to make notifications or provide emergency information. The finding has a cross-cutting aspect in the area of human performance, associated with avoiding complacency, because licensee personnel failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes, by not implementing appropriate error reduction tools. Due to the lack of appropriate error prevention tools, inaccurate survey data was provided to the vendor and errors in the waste characterization and shipping manifest were not identified in a timely fashion (H.12). |