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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000482/FIN-2018010-01 | 2018Q3 | Green | NRC identified | Failure to Follow Procedures | The team identified two examples of a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to follow procedures. | |
| 05000482/FIN-2018201-04 | 2018Q3 | Green | NRC identified | Security | ||
| 05000482/FIN-2018003-01 | 2018Q3 | Green | H.14 | NRC identified | 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. |
| 05000482/FIN-2018003-02 | 2018Q3 | Severity level IV | NRC identified | 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. | |
| 05000482/FIN-2018010-02 | 2018Q3 | Green | NRC identified | Failure to Establish an Adequate Procedure for Operator Time Critical Actions Validation | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to have an adequate Procedure. Procedure AI 21-016, Operator Time Critical Actions Validation, Revision 14, Attachment B Time Sensitive Action List, does not have unique identifiers for cross referencing the records to the procedure. | |
| 05000482/FIN-2018010-03 | 2018Q3 | Green | P.4 | NRC identified | Failure to Correct Reoccurring Problems with Time Critical/Sensitive Action Activities | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct reoccurring problems with completing Time Critical/Time Sensitive Action issues. |
| 05000482/FIN-2018010-04 | 2018Q3 | Green | H.13 | NRC identified | Failure to Identify 125 VDC Equalizing Voltage Exceeded Design Requirements | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify or check the adequacy of design calculation NK-E-001, 125 Volt Direct-Current (VDC) Class 1E Battery System Sizing, Voltage Drop and Short Circuit Studies, Revision 4. The licensee failed to recognize that the actual 125 VDC Class 1E bus voltages had exceeded the maximum design limit voltages for downstream equipment identified in the calculation, and they had not placed any limits on voltages which could exceed the design limit of 140 VDC on the Class 1E System components. |
| 05000482/FIN-2018010-05 | 2018Q3 | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance 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, Appendix B, Criterion XVI, Corrective Action, states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected.Contrary to the above, prior to 2015, the licensee failed to promptly identify and correct a repetitive deficiency or non-conformance. Specifically, the licensee had identified a leaking flange on the residual heat removal heat exchanger since 1997. Prior to 1997 a different data base had been used to record boric acid leakage, and the data was not available during the inspection.Over the years since plant startup, the licensee had been diligent in completing boric acid evaluations on the leaking residual heat removal heat exchanger flange, indicating minimal wastage of the flange closure studs and nuts that had been subjected to boric acid. Corrective actions included cleaning up the boric acid leakage, and checking or re-torqueing the closure nuts. These measures did not correct the problem of the leaking heat exchanger flange. In 2015 the licensee completed an in-depth engineering evaluation of the leaking flange, including discussions with the heat exchanger manufacturer. New corrective measures included changing the torque values on the closure studs and nuts. The licensee is still evaluating the results of the corrective actions taken to preclude further leakage. | |
| 05000482/FIN-2018201-01 | 2018Q3 | Green | NRC identified | Security | ||
| 05000482/FIN-2018201-02 | 2018Q3 | Green | NRC identified | Security | ||
| 05000482/FIN-2018201-03 | 2018Q3 | Green | NRC identified | Security | ||
| 05000482/FIN-2018007-02 | 2018Q2 | Severity level Minor | NRC identified | Minor Violation | Performance Deficiency: Failure to promptly identify and correct known-defective switches in inservice safety-related breakers, or to control nonconforming breakers accepted into warehouse stores, as required by 10 CFR 50 Appendix B Criteria XV and XVI. In February 2008, the licensee received a notification from GE Hitachi of reduced reliability of some safety-related circuit breakers due to defective cutoff switches internal to the breakers. The licensee incorrectly screened this information as not applicable to the Wolf Creek Generating Station. In August 2011, after licensee engineers received the information again from industry peers, the licensee screened the information as applicable. The licensee then added steps to its overhaul and pre-install test procedures to check for the defective subcomponent. These steps were performed during subsequent regularly scheduled overhaul or pre-install tests, with the last affected switches being replaced in June 2014 and the last potentially susceptible safety-related breaker being inspected in March 2015. The team determined that because the station had information on the defect in February 2008, but did not correct the condition until 2014 and did not confirm that it was corrected until 2015, the licensee had failed to promptly identify and correct a condition adverse to quality. Further, the licensee failed to inspect or place administrative controls on potentially affected spare breakers that had been accepted into warehouse stores, though the added steps in the pre-install procedure likely would have prevented a defective component from being installed. However, by failing to segregate the potentially affected components until they were inspected, the licensee failed to comply with quality assurance requirements for control of nonconforming components. On June 26, 2018, the licensee put a hold on four potentially affected breakers that were in warehouse stores. The licensee documented this performance deficiency in CR 124693. Screening: The performance deficiency was minor because the licensee did not experience an inservice failure as a result of the defect during the 6 years they remained in service and had a procedure in place that would likely have prevented a defective spare from being issued for installation. Therefore, there was no adverse effect on the mitigating systems cornerstone objective and there was no potential to create a more significant safety concern. Enforcement: This failure to comply with 10 CFR 50 Appendix B Criteria XV and XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. | |
| 05000482/FIN-2018007-01 | 2018Q2 | Green | P.2 | Self-revealing | Failure to Provide Adequate Work Instructions for Preventive Maintenance on Safety-Related Equipment | The team reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a to establish, implement, and maintain written procedures recommended by Regulatory Guide 1.33, Appendix A, Revision 2. Specifically, work instructions for the preventive maintenance for the train B Class 1E electrical equipment A/C unit SGK05B, lacked adequate guidance for preventive maintenance and calibration of the associated thermostat. This resulted in the loss of cooling failure of the A/C unit SGK05B,on February 12, 2018. |
| 05000482/FIN-2018002-03 | 2018Q2 | Green | H.2 | Self-revealing | Failure to Adequately Implement Instrumentation and Controls Surveillance Procedures | A self-revealed Green NCV of 10 CFR Part 50, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to adequately implement surveillance procedures that affected safety-related equipment and plant stability. Specifically, the licensee failed to adequately implement testing and calibration procedures for pressurizer level instrumentation. This resulted in two letdown isolation signals, securing of pressurizer heaters, and a pressurizer level transient on March 29, 2018. |
| 05000482/FIN-2018002-02 | 2018Q2 | Green | H.3 | Self-revealing | Failure to Maintain Adequate Pressurization of the Control Room Envelope | A self-revealed Green NCV of 10 CFR Part 50, Criterion III, Design Control, was identified when the licensee failed to adequately recognize that the cable spreading room floor was a control building ventilation isolation boundary. Specifically, the licensee cut openings in the floor/ceiling between the 2,032 foot and 2,016 foot elevations of the control building and the impact on the control room envelopes ability to pressurize was not recognized. This was a primary contributor to the train B control room emergency ventilation system being unable to maintain the appropriate pressure in the control room envelope. |
| 05000482/FIN-2018002-01 | 2018Q2 | Severity level IV | NRC identified | Announcement of an NRC Inspectors Presence by Station Personnel | The inspectors identified a Severity Level IV non-cited violation (NCV) of 10 CFR 50.70(b)(4), Inspections, associated with the licensees failure to ensure the arrival and presence of NRC Inspectors, who had been properly authorized facility access as described in 10 CFR 50.70(b)(3), were not announced or otherwise communicated by its employees or contractors to other persons at the facility without a specific request by the NRC inspector. Specifically, a contract radiation protection technician entered the spent fuel pool building where the resident inspector was present and observing core offload activities, and the technician informed members of a work crew of the whereabouts of an NRC radiation protection inspection team without being requested to do so; this impacts the NRCs ability to regulate and perform unannounced inspections. | |
| 05000482/FIN-2018001-01 | 2018Q1 | Green | NRC identified | Inadequate Functionality Assessment Associated with the Emergency Excess Letdown Flowpath | The inspectors identified a Green finding and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when the licensee failed to adequately implement the operability determination and functionality assessment procedure. Specifically, the licensee failed to document a functionality assessment of sufficient scope to address the capability of a safety-related excess letdown heat exchanger to pressurizer relief tank isolation valve and the excess letdown system to perform their specified safety functions, which resulted in the licensee failing to recognize that two independent Technical Requirements Manual required boration injection subsystems were not functional. | |
| 05000482/FIN-2017008-02 | 2017Q4 | Green | H.9 | NRC identified | Failure to Provide Adequate Emergency Lighting | The team identified a non-cited violation of License Condition 2.C.(5) for failure to provide emergency lighting along alternate routes plant operators are allowed to take during implementation of the procedure for control room evacuation due to fire.The failure to provide 8-hour emergency lights along alternate routes used by operators during control room evacuation due to fire is a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because it affected the ability to reach and maintain safe shutdown conditions in case of a fire. The team assigned the finding to the post-fire safe shutdown category since it impacted the alternate shutdown element. The issue screened to Green because the reactor would be able to achieve and maintain hot shutdown because the operators are required to carry flashlights. Specifically, the team had reasonable assurance that the operators would be able to complete the evacuation procedure using handheld flashlights to access safe shutdown equipment. The finding is assigned a cross-cutting aspect in the area of human performance, associated with training, because the operators are not being trained on the access and egress routes that are provided with 8-hour emergency lights during implementation of the control room evacuation procedure due to fire to ensure the time critical actions can be met (H.9). |
| 05000482/FIN-2017008-01 | 2017Q4 | Green | NRC identified | InadequateEvaluation of Spurious Valve Operation | The team identified a non-cited violation of License Condition 2.C.(5) for failure to implement and maintain in effect all provisions of the approved fire protection program. Specifically, the licensee failed to adequately evaluate the potential impacts on post-fire safe shutdown of two motor operated valves spuriously closing due to fire damage.The failure to adequately evaluate the impact of pressure operated relief valve block valves spuriously closing on post-fire safe shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire that led to control room evacuation and because the Phase 2 methodology of Inspection Manual Chapter 0609, Appendix F, was not appropriate for this finding, a senior reactor analyst performed a Phase 3 evaluation to determine the risk significance. The analyst determined this finding was of very low risk significance (Green). There is no cross-cutting aspect associated with this finding since the performance deficiency is not reflective of present performance (i.e., the performance deficiency occurred more than 3 years ago). | |
| 05000482/FIN-2017003-04 | 2017Q3 | Green | H.1 | NRC identified | Failure to Verify Equipment or Systems are Capable of Performing Their Intended Design Function Following Maintenance | The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a for the licensees failure to ensure that maintenance that can affect the performance of safety-related equipment was properly pre-planned and performed in accordance with written procedures, documented, instructions, or drawings appropriate to the circumstances. Specifically, the licensee failed to verify that the wiring in the transformer 7 primary differential protective relay was landed on the correct termination point, and as a result, the station experienced an unplanned loss of normal offsite power to bus NB01, the train A Class 1E electrical bus.Description. On November 16, 2016, at approximately 9:09 p.m., a fault occurred that isolated the east switchyard bus from the train A safety-related 4160 volt alternating current bus NB01, while the Wolf Creek Nuclear Generating Station was in Mode 5 with the reactor coolant system filled and a bubble in the pressurizer. During refueling outage 21, a modification to transformer 7 allowed the offsite power through transformer 7 to bus NB01 to be fed from either the east or west switchyard busses through two different breakers (345-80 or 345-90). After the loss of the east switchyard bus, the second breaker unexpectedly tripped, which resulted in a loss of offsite power to NB01. An undervoltage condition was detected on bus NB01, which caused the train A emergency diesel generator to start and power bus NB01 as designed. All other systems functioned as expected. Westar, the substation owner, determined that the initial fault was caused by a mouse on the 13-4 circuit at Wolf Creek. The 13-4 relay and breaker cleared the fault and coordinated with all upstream devices. Approximately 5.5 seconds after the initial fault, a second fault occurred in transformer 6. The transformer 7 digital differential relay scheme provides a standard configuration with primary and secondary protective relays, each with the capability of isolating transformer 7. Troubleshooting activities focused on the reason why the primary relay tripped and the secondary relay did not trip. Westar technicians identified a jumper on the transformer 7 primary differential relay current transformer circuit that had been improperly landed. The jumper was designed to run from the neutral circuit of one current transformer to the neutral circuit of the other. However, Westar Energy technicians had incorrectly landed the jumper from the neutral of the first current transformer onto the C phase of the other. This allowed current from the transformer 6 fault event to be detected in the transformer 7 primary differential relay circuit.The inspectors reviewed the cause evaluation completed by the licensee, whichdetermined that the direct cause of this event was the wiring in the transformer 7 primary differential protective relay was landed on the incorrect termination point. This cause is supported by the fact that this incorrect termination allowed additional current to be introduced onto the C phase relay circuit, which initiated the trip circuit actuation.The inspectors also reviewed corrective actions associated with the root cause evaluation for the unplanned plant shutdown, loss of offsite power, and Notification of Unusual Event declaration that occurred on January 13, 2012. An Augmented Inspection Team was chartered to review the circumstances surrounding the loss of offsite power event and Notification of Unusual Event declarationan issue of Yellow safety significance was identified. The event from January 13, 2012, involved equipment owned by Wolf Creek (startup transformer XMR01), with work being performed by Wolf Creek contractors. The November 16, 2016, event involved equipment owned by Westar (transformer 7). While inspectors acknowledge that the two events from January 13, 2012, and November 16, 2016, are not exactly the same, the inspectors noted that they are similar in that they both involved the modification of current transformer wiring associated with transformers that provide power to train A and B engineered safety function transformers (XNB01 and XNB02, respectively), which supply train A and B Class 1E electrical busses NB01 and NB02, respectively. The inspectors did not determine that the 2012 event actions were causal to the 2016 event; however, the inspectors noted similarities between the identified causes. Procedure AP 21C-001, Wolf Creek Substation, establishes responsibilities and defines necessary interfaces and communications for the operational control, coordination and maintenance necessary to ensure Wolf Creek Substation protection, safety and reliability. The inspectors reviewed the licensees assessment associated with the 2016 event and concluded that the substation work control process requirements in procedure AP 21C-001 were not adequately met. Specifically, step 6.2.5.1 states, in part, that following preventive or corrective maintenance work, appropriate post-maintenance inspections, checks, and/or testing shall be performed to verify that affected equipment or systems (primary and secondary differential relay circuitry) are capable of performing their intended design function.The wiring error on the primary differential protective relay was corrected and its functionality was verified. The secondary differential protective relay wiring was also verified to be correct. The east switchyard bus, transformer 7, and its differential relays were all restored to service. The licensee documented the event in LER 2016-002-00 and Condition Reports 109467 and 116849. The licensee also updated procedure AP 21C-001 to include additional detail and steps that require work instructions for post maintenance testing of current transformer wiring to ensure independent verification of wiring terminations.Analysis. The licensees failure to verify that the primary and secondary differential relay circuitry is capable of performing its intended design function following maintenance was a performance deficiency. The performance deficiency was more than minor because it affected the design control attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to verify that the wiring terminations for the primary differential protective relay for transformer 7 were installed correctly, leading to the isolation of transformer 7, resulting in an unplanned loss of offsite power to NB01, the train A Class 1E electrical bus. The inspectors evaluated the finding using Exhibit 3, "Mitigating SystemsScreening Questions," of Inspection Manual Chapter 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase I Initial Screening and Characterization of Finding," and Appendix G, "Shutdown Operations Significance Determination Process," both issued May 9, 2014. The inspectors determined this finding is a deficiency affecting the design or qualification of a mitigating SSC, and the SSC maintained its operability or functionality. Therefore, the inspectors determined the fi nding was of very low safety significance (Green). The inspectors determined that the finding has a human performance cross-cutting aspect in the area of resources because leaders did not ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, leaders did not ensure adequate procedures were available to support successful work performance including necessary standards for verifying wiring circuitry terminations such that the loss of power to the NB01 Class 1E electrical bus would not have occurred. This issue is indicative of current performance because the issue occurred in the last three years (H.1). |
| 05000482/FIN-2017003-03 | 2017Q3 | Green | H.1 | NRC identified | Failure to Verify Equipment or Systems are Capable of Performing Their Intended Design Function Following Maintenance | The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a for the licensees failure to ensure that maintenance that can affect the performance of safety-related equipment was properly pre-planned and performed in accordance with written procedures, documented, instructions, or drawings appropriate to the circumstances. Specifically, the licensee failed to verify that the wiring in the transformer 7 primary differential protective relay was landed on the correct termination point, and as a result, the station experienced an unplanned loss of normal offsite power to bus NB01, the train A Class 1E electrical bus. The licensee took the immediate corrective actions of working with Westar to ensure the protective relay wiring termination issue for transformer 7 was identified and corrected, and that transformer 7 was returned to service. The licensee also updated procedure AP 21C-001 to include additional detail and steps that require work instructions for post maintenance testing of current transformer wiring to ensure independent verification of wiring terminations. The licensee entered the issue into the corrective action program as Condition Reports 109467 and 116849. The licensees failure to verify that the primary and secondary differential relay circuitry is capable of performing its intended design function following maintenance was a performance deficiency. The performance deficiency was more than minor because it affected the design control attribute of the Mitigating Systems 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 3, "Mitigating Systems Screening Questions," of Inspection Manual Chapter 0609, Appendix G, Attachment 1, "Shutdown Operations Significance Determination Process Phase I Initial Screening and Characterization of Finding," and Appendix G, "Shutdown Operations Significance Determination Process." The inspectors 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 resources because leaders did not ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. This issue is indicative of current performance because the issue occurred in the last three years (H.1). |
| 05000482/FIN-2017003-02 | 2017Q3 | Green | P.2 | NRC identified | Failure to Ensure the Design Basis was Adequately Represented in the Technical Specification Bases | The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish adequate measures to ensure that the design bases are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee did not ensure the auxiliary feedwater system design basis was adequately represented in the Technical Specification Bases; as a result, the Technical Specification Bases and other station procedures allowed for one train of essential service water supply to the turbine-driven auxiliary feedwater pump to be removed from service without recognition that auxiliary feedwater operability was impacted. Immediate corrective actions included entering Condition Reports 113304 and 116852 into the corrective action program and incorporating a note on operations turnover documents to temporarily postpone applicable portions of the operations quarterly tasks.The licensee also completed a past operability review, and created actions to develop a license amendment request to add a specific Technical Specification condition and submit for NRC approval.The failure to ensure the auxiliary feedwater system design basis was adequately represented in the Technical Specification Bases was a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is 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). 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 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 evaluation because the organization did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. This issue is indicative of current performance because the evaluation of Condition Report 111808 in May 2017 was a reasonable opportunity for the licensee to identify that the Technical Specification Bases was inadequate (P.2). |
| 05000482/FIN-2017003-01 | 2017Q3 | Green | NRC identified | Programmatic Failure to Scope Floor Drain Function within the Maintenance Rule Monitoring Program | The inspectors identified a Green non-cited violation of 10 CFR 50.65(b)(2)(ii), because the licensee did not adequately include nonsafety-related SSC functions within the scope of the maintenance rule monitoring program. Specifically, the licensee failed to adequately include within the scope of the maintenance rule monitoring program the function of draining. This scoping issue has resulted in a failure to monitor floor drain degradation and to provide reasonable assurance that safety-related SSCs in an estimated 76 rooms are capable of fulfilling their intended functions. Immediate corrective actions included entering the condition into the corrective action program as Condition Report 116319 and later as Condition Report 116851. The inspectors determined that the licensees failure to meet the requirements of 10 CFR 50.65(b)(2)(ii) and appropriately place the function of draining, for nonsafety-related floor drains in up to 76 rooms containing safety-related SSCs, within the scope of the maintenance rule monitoring program was a performance deficiency. The performance deficiency was more than minor, because it is associated with the Protection Against External Factors 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 did not have a cross-cutting aspect because the issue was not indicative of current performance. | |
| 05000482/FIN-2017407-01 | 2017Q3 | Greater Than Green | H.1 | NRC identified | Security | |
| 05000482/FIN-2017007-01 | 2017Q2 | Green | H.14 | NRC identified | Failure to Maintain Effectiveness of the Emergency Plan upon Loss of Containment High Radiation Monitoring | The inspectors identified a Green non-cited violation of 10 CFR 50.54(q)(2) which requires that a holder of a nuclear power plant operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E of this part and the risk significant planning standards of 10 CFR 50.47(b). Specifically, from March 7, 2017, to July 12, 2017, Wolf Creek Generating Stations response to the inoperability of containment high radiation monitors failed to restore capability to classify emergency action levels during a loss-of-coolant accident or main-steam-line-break accident. In response to this issue, the licensee provided additional radiation survey monitoring measures and correlations to monitor radiation in the containment building. This finding was entered into the licensees corrective action program as Condition Report CR-114274. The inspectors determined that the failure to maintain the effectiveness of the emergency action level schemes by providing adequate preplanned methods and compensatory measures for the loss of the containment high range radiation monitors in accordance with 50.54 (q)(2) was a performance deficiency. This finding was determined to be more thanminor because it was associated with emergency response organization performance attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective. Specifically, the failure to maintain the effectiveness using appropriate compensatory measures adversely affected the objective of ensuring the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was determined to be of very low safety significance (Green) because (1) emergency action level schemes were rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event, but because of other emergency action levels, an appropriate declaration could be made in a degraded manner; and, (2) the emergency action level classification process would result in an over-classification causing an unnecessary emergency declaration. This finding had a cross-cutting aspect in the area of human performance associated with conservative bias because the licensee failed use decision making-practices that emphasized prudent choices over those that are simply allowable. (H.14)(Section 1R21N) |
| 05000482/FIN-2017002-01 | 2017Q2 | Green | P.1 | NRC identified | 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). |
| 05000482/FIN-2017007-02 | 2017Q2 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.3.3, Post Accident Monitoring Instrumentation, required two channels of containment area radiation (high range) detectors to be operable when the unit is in Modes 1, 2, or 3. It also required, for one or more functions with two required channels inoperable, that one of the required channels be restored to Operable within 7 days or initiate action in accordance with Technical Specification 5.6.6. Specification 5.6.6 required that a Post Accident Monitoring Instrumentation Report be submitted within 14 days that outlined the preplanned alternate method of monitoring, the cause of inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. Contrary to the above, from 1997 to March 2017, the licensee failed to restore at least one channel of containment high range radiation monitors to operable status, initiate preplanned alternate methods of monitoring the appropriate parameter, or prepare and submit a Post Accident Monitoring Instrumentation Report within 14 days pursuant to Technical Specification 5.6.6. The violation was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness Cornerstone 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. The inspectors determined the significance using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter0609, Appendix B, Emergency Preparedness Significance Determination Process, Section 5.4 for failure to comply with risk significant planning standard 10 CFR 50.47(b)(4). The finding was determined to be of very low safety significance (Green) because (1) emergency action level schemes were rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event, but because of other emergency action levels, an appropriate declaration could be made in a degraded manner; and, (2) the emergency action level classification process would result in an over-classification causing an unnecessary emergency declaration. The violation was entered into the licensees corrective action program as Condition Reports CR-111440, CR-111536, and CR-113217. | |
| 05000482/FIN-2017002-03 | 2017Q2 | Severity level Enforcement Discretion | NRC identified | 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. | |
| 05000482/FIN-2017002-02 | 2017Q2 | Green | H.11 | NRC identified | 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). |
| 05000482/FIN-2017404-01 | 2017Q2 | Green | H.8 | Self-revealing | Security | |
| 05000482/FIN-2017001-01 | 2017Q1 | Green | P.3 | Self-revealing | Failure to Provide Adequate Work Instructions for Preventive Maintenance | Green. The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a and Regulatory Guide 1.33 for the licensees failure to provide adequate work instructions for preventive maintenance on safety-related equipment. Specifically, work instructions to inspect and clean the condensate drain lines on the class 1E air conditioner air handling units lacked guidance for adequately cleaning the drain line. This caused the unit to become non-functional. The licensee took the immediate corrective action to clear the clogged condensate drain line on SGK05B, and entered the issue in the corrective action program as Condition Report 106416. The failure to provide adequate work instructions for preventive maintenance on safety-related equipment is a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined this finding screened to Green. The inspectors determined that the finding has a problem identification and resolution cross-cutting aspect of resolution because the organization did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. This issue is indicative of current performance because neither the preventive maintenance change process was substantively changed nor were human performance errors associated with the preventive maintenance change corrected, and the same resolution inadequacies that resulted in the inadequate preventive maintenance instructions would be expected to occur (P.3). |
| 05000482/FIN-2016410-01 | 2016Q4 | Green | H.9 | NRC identified | Security | |
| 05000482/FIN-2016004-01 | 2016Q4 | Green | P.5 | NRC identified | Failure to Adequately Establish and Adjust Preventive Maintenance for Emergency Diesel Generator Excitation System Diodes | Green. The inspectors identified a non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to adequately develop and adjust preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, Revision 5. Specifically, the licensee did not create a preventive maintenance task for emergency diesel generator excitation system diodes, which resulted in degradation of the excitation system diodes in emergency diesel generator B. The licensee restored compliance by establishing preventive maintenance tasks 49286, 49287, 49288, and 49289, which refurbish the power rectifier assemblies and replace the diodes on a 12-year replacement frequency. The licensee entered this issue into the corrective action program as Condition Report 88665. The failure to adequately develop and adjust emergency diesel generator excitation system diode preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, was a performance deficiency. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, this finding was not a deficiency affecting the design or qualification of a mitigating structure, system, or component that maintained its operability or functionality; the finding did not represent a loss of system and/or function; the finding did not represent an actual loss of function of at least a single train for greater than its Technical Specification allowed outage time; and the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, because the organization did not systematically and effectively evaluate relevant internal and external operating experience in a timely manner. This issue is indicative of current performance because the station did not take any formal corrective actions to address the stations failure to adequately consider operating experience (P.5) |
| 05000482/FIN-2016410-02 | 2016Q4 | Green | NRC identified | Security | ||
| 05000482/FIN-2016411-01 | 2016Q4 | Green | H.7 | NRC identified | Security | |
| 05000482/FIN-2016411-02 | 2016Q4 | Green | NRC identified | Security | ||
| 05000482/FIN-2016408-04 | 2016Q3 | Severity level IV | NRC identified | Security | ||
| 05000482/FIN-2016008-02 | 2016Q2 | Green | H.14 | NRC identified | Failure to Promptly Identify and Correct a Condition Adverse to Quality Associated with the Emergency Diesel Generator B Excitation System Diodes | The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure that conditions adverse to quality, such as failures, malfunctions, and deficiencies are promptly identified and corrected. Specifically, the licensee failed to promptly identify and correct a failed rectifier bridge diode after smoke was observed coming from the three power potential transformers in the emergency diesel generator exciter cabinet NE106 on June 11, 2014, which contributed to the emergency diesel generator B being declared inoperable and unavailable when it tripped during a 24-hour surveillance test on October 6, 2014. To address the failure to take adequate corrective actions Wolf Creek entered this issue into its corrective action program as Condition Report 105480 and plans to implement a modification to install overcurrent detection for each emergency diesel generators power potential transformer. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct the failed emergency diesel generator excitation system diode contributed to the emergency diesel generator B failure on October 6, 2014. The inspectors evaluated the finding using Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. The inspectors determined this finding is not a deficiency affecting the design or qualification of a mitigating structure, system, or component that maintained its operability or functionality, the finding does not represent a loss of system and/or function, the finding does not represent an actual loss of function of at least a single train for greater than its Technical Specification allowed outage time, and the finding does not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the inspectors determined the finding was of very low safety significance (Green). The inspectors determined that in accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, conservative bias, because when smoke was identified coming from the power potential transformers on multiple occasions, licensee personnel did not use decision making-practices that emphasize prudent choices over those that are simply allowable, and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. As a result, the licensee missed an opportunity to identify and correct the condition of the failed diode in the static exciter (H.14). |
| 05000482/FIN-2016002-06 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.6.3, Containment Isolation Valves, requires each containment isolation valve to be operable in modes 1, 2, 3, and 4. To be operable, containment isolation valves GTHZ0007 and GTHZ0009, which are Category 3 valves, must be closed with the motive force removed. Technical Specification 3.6.3, Condition A, Required Action A.1, requires, in part, that the affected penetration flow path for any inoperable Category 3 containment isolation valve be isolated within 12 hours. Additionally, Required Action A.2, requires, in part, that the licensee verify the affected penetration flow path is isolated prior to entering mode 4 from mode 5. Contrary to the above, from April 28, 2015, through May 5, 2015, the licensee failed to verify the affected penetration flow path was isolated prior to entering mode 4 from mode 5 on April 28, 2015. As a result, Technical specification 3.6.3, Condition A, was not met On May 5, 2015, the licensee discovered that the motive force for valves GTHZ0007 and GTHZ0009 was not removed and the air supply valves had not been locked closed, and the affected penetration flow paths were not isolated prior to entering mode 4 from mode 5 on April 28, 2015. The inspectors noted that although the motive force was not removed for valves GTHZ0007 and GTHZ0009, the valves were in their closed safeguards positions and redundant valves in series were closed with the motive force removed, which ensured each penetration flow path had one operable valve closed with its motive force removed. Using Exhibit 3, Barrier Integrity Screening Questions, of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, the inspectors determined the finding did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), or heat removal components, and the finding did not involve an actual reduction in function of hydrogen igniters in the reactor containment. Therefore, the inspectors determined that this finding is of very low safety significance (Green). | |
| 05000482/FIN-2016002-01 | 2016Q2 | Green | H.3 | NRC identified | Failure to Adequately Establish Control Room Air Conditioning System Testing Flow Rate Acceptance Criteria | The inspectors identified a Green non-cited violation of Technical Specification Limiting Condition for Operation 3.7.11 and 3.0.3 for the licensees failure to place the unit in mode 3 within 7 hours, mode 4 within 13 hours, and mode 5 within 37 hours with two trains (SGK04A and SGK04B) of the control room air conditioning system (CRACS) inoperable. Specifically, the licensee failed to adequately establish CRACS testing flow rate acceptance criteria, which resulted in train A of the safety-related CRACS being inoperable from October 11, 2005, to August 13, 2013; and train B being inoperable from October 3, 2002, to July 18, 2013. The licensees immediate corrective actions included corrective maintenance on the CRACS to increase the airflow to meet acceptance criteria limits. Condition Report 105208 was initiated by the licensee for any necessary process changes and extent of condition actions. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated 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 utilized Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 2 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined this finding was not a deficiency affecting the design or qualification of a mitigating SSC that maintained its operability or functionality, the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of function of at least a single train for greater than its Technical Specification allowed outage time, and the finding did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the inspectors determined the finding was of very low safety significance (Green). The finding has a cross-cutting aspect in the area of human performance, change management, because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, there is not currently a formal process for procedure writers to consider measurement uncertainty when establishing and changing testing acceptance criteria, which resulted in extended inoperability of both the SGK04A and SGK04B units following significant changes to Technical Specifications that included adding surveillance requirements for the SGK04A and SGK04B units in 1999. This issue is indicative of current performance because the same issue would be expected to occur today (H.3). |
| 05000482/FIN-2016002-05 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.4.15, (Reactor Coolant System) Leakage Detection Instrumentation, states, in part, that reactor coolant system leakage detection instrumentation shall be operable, including the containment sump level and flow monitoring system. Required Action A of Technical Specification 3.4.15, states, in part, that with the required containment sump level and flow monitoring system inoperable, restore the required containment sump level and flow monitoring system to operable status within 30 daysif the required action and associated completion time are not met, Condition E requires the reactor to be in mode 3 within 6 hours and in mode 5 within 36 hours. Contrary to the above, from the period of July 13, 2013, to November 20, 2013, with the containment sump level and flow monitoring system inoperable for greater than 30 days, the reactor was not placed in mode 3 within 6 hours or mode 5 within 36 hours. Specifically, the instrument tunnel sump level indication was inoperable because of erratic indication, but the licensee did not take the required action of Technical Specification 3.4.15. The licensee placed this issue in the corrective action program as Condition Report 84690. Using Manual Chapter 0609, Appendix A, Significance Determination Process, for Findings at Power, dated June 19, 2012, this issue screened to Green because it did not result in reactor coolant system leakage or degrade the licensees ability to detect and mitigate a small break loss of coolant accident. | |
| 05000482/FIN-2016007-02 | 2016Q2 | Green | NRC identified | Failure to Verify the Adequacy of Design of the Control Circuitry of the Fuel Oil Transfer Pumps | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to April 28, 2016, the licensee failed to verify the adequacy of the design of the fuel oil transfer pump control circuitry to ensure that the thermal overloads associated with the fuel oil transfer pump would not activate, trip the pump, and render the emergency diesel generator inoperable in the case of excessive cycling. In response to this issue, the licensee conducted a preliminary evaluation of the fuel oil transfer pump and confirmed there is not any significant active leakage on the day tank which would lead to excessive cycling, and that starting currents are sufficiently below the thermal overload trip settings and are unlikely to trip the pump. Additionally, the licensee planned to initiate a program to determine fuel oil leakage from the day tank and require operators to initiate interim corrective actions until final corrective actions can be determined. This finding was entered into the licensee's corrective action program as Condition Report 104066. The team determined the failure to evaluate the effects of cyclical fuel oil transfer pump operation was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design of the fuel oil transfer pump control circuity does not prevent activation of the pump thermal overloads that would trip the pump and render the emergency diesel generator inoperable in the event of cyclical operation of the fuel oil transfer pump. 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 did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. | |
| 05000482/FIN-2016002-02 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 5.7.2 states, in part, that high radiation areas with dose rates greater than 1.0 rem per hour at 30 centimeters shall be conspicuously posted as a high radiation area and shall be provided with a locked or continuously guarded door or gate to prevent unauthorized entry. Contrary to the above, on January 27, 2016, room 7406 on the 2013 foot elevation of the radwaste building areas had dose rates greater than 1.0 rem per hour and was not conspicuously posted as a high radiation area nor provided with a locked or continuously guarded door or gate to prevent unauthorized entry. This issue was identified by radiation protection technicians performing radiological surveys in the area. The licensee documented this issue in the corrective action program as Condition Report 102344. The finding was determined to be of very low safety significance (Green) because it was not an as-low-as-reasonably-achievable planning issue, there was no overexposure or potential for overexposure, and the licensees ability to assess dose was not compromised. | |
| 05000482/FIN-2016407-01 | 2016Q2 | Green | NRC identified | Security | ||
| 05000482/FIN-2016007-01 | 2016Q2 | Green | NRC identified | Inadequate Degraded Voltage Analyses of Class 1E Systems | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to April 28, 2016, the licensee failed to verify the adequacy of the design of the Class 1E electrical equipment, because it failed to perform adequate analyses demonstrating 1) that the degraded voltage relay setpoints specified in technical specifications would ensure adequate voltage to safety-related equipment, 2) adequate voltage would be available to the safety-related loads during transient voltage conditions caused by load sequencing, and 3) that the degraded voltage relay-associated time delays provide timely separation from offsite power and transfer to the emergency diesel generator to ensure that the Class 1E safety-related loads can achieve their safety function without protective device tripping. In response to these issues, the licensee performed preliminary analyses to demonstrate that the Class 1E electrical equipment would function at degraded voltages and was operable. This finding was entered into the corrective action program as Condition Reports 47791, 104253, 104098, 104389, and 104390. The team determined the licensees failure to ensure the adequacy of the design of the Class 1E electrical equipment was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees electrical analyses failed to verify degraded voltage relay setpoints specified in technical specifications would ensure adequate voltage to safety-related equipment, that adequate voltage would be available to the safety-related loads during transient voltage conditions caused by load sequencing, and that degraded voltage relay time delays would provide timely separation from offsite power and transfer to the emergency diesel generator to ensure that the Class 1E safety-related loads can achieve their safety function without protective device tripping. 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 to result in 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 did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance. | |
| 05000482/FIN-2016002-04 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | Technical Specification 3.4.3, (Reactor Coolant System) Pressure and Temperature Limits, states, in part, that reactor coolant system pressure, reactor coolant system temperature, and reactor coolant system heatup and cooldown rates shall be maintained within the limits specified in the Pressure and Temperature Limits Report (PTLR). Section 2.1.2 of the PTLR specifies that the reactor coolant system shall be maintained within the parameters of Figure 2.1-1 of the PTLR, which specifies a minimum pressure of 0 psig. Required Action C.1 of Technical Specification 3.4.3 specifies that with the reactor coolant system parameters outside the limits of the PTLR, restore the parameters to within the limits immediately. Contrary to the above, on May 8, 2011, and March 30, 2013, with the reactor coolant system parameters outside the limits of the PTLR, parameters were not restored to within the limits immediately. Specifically, the licensee drew a vacuum on the reactor coolant system to less than 0 psig to support filling operations but did not take action to immediately restore the reactor coolant system pressure to greater than or equal to 0 psig, as specified in the PTLR. The licensee placed this issue in the corrective action program as Condition Report 78920. The licensee performed Engineering Evaluation EER 92-BB-02 and determined that drawing a vacuum on the reactor coolant system would not result in excessive stresses for reactor coolant system structures, systems and components. Using Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, dated May 9, 2014, this issue screened to Green because it did not result in a loss of reactor coolant system barrier integrity. | |
| 05000482/FIN-2016405-01 | 2016Q2 | Licensee-identified | Security | |||
| 05000482/FIN-2016009-02 | 2016Q2 | Green | P.4 | NRC identified | Failure to Identify and Correct Negative Trend in Breach Permit Corrective Actions | The team identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to identify and correct a condition adverse to quality. Specifically, between October 2012 and June 2016, the licensee identified 15 instances of individuals failing to properly issue or use breach permits per the licensees procedures, yet failed to identify or address this site-wide adverse trend. The failure to identify and correct a negative trend in the proper issuance and use of breach permits was a performance deficiency. This performance deficiency was more than minor because it negatively affected the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Using Inspection Manual Chapter 0609, Appendix A, the team determined that this finding was of very low safety significance (Green) because it did not result in the loss of operability or functionality of the system, structure, or component. This finding has a trending cross-cutting aspect in the problem identification and resolution cross-cutting area because the licensee failed to periodically analyze information from the corrective action program and other assessments in the aggregate to identify programmatic and common cause issues (P.4). Specifically, the licensee did not address numerous site-wide breach permit issues concerning the corrective actions, only addressing the individuals involved. |
| 05000482/FIN-2016008-01 | 2016Q2 | White | P.5 | NRC identified | Failure to Adequately Establish and Adjust Preventive Maintenance for Emergency Diesel Generator Excitation System Diodes | The inspectors identified a preliminary White finding associated with an apparent violation of Technical Specification 5.4.1.a, for the licensees failure to adequately develop and adjust preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, Revision 5. Specifically, the licensee did not create a preventive maintenance replacement task or schedule for emergency diesel generator excitation system diodes, which resulted in emergency diesel generator B being declared inoperable and unavailable when it tripped during a 24-hour surveillance test. The licensee entered this condition into its corrective action program as Condition Report 88665. The licensee restored compliance by establishing preventive maintenance tasks 49286, 49287, 49288, and 49289, which refurbish the power rectifier assemblies and replace the diodes on a 12-year replacement frequency. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with one emergency diesel generator excitation system diode failed as a result of thermal degradation, emergency diesel generator B was not operable or available to perform its safety function. The inspectors evaluated the finding using Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation because it represented an actual loss of function of the emergency diesel generator B for greater than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation in accordance with Appendix A, Section 6.0, Detailed Risk Evaluation. The calculated change in core damage frequency was dominated by a loss of offsite power initiator leading to station blackout with failures of the turbine-driven and non-safety-related auxiliary feedwater pumps. The analyst did not evaluate the large early release frequency because this performance deficiency would not have challenged the containment. The NRC preliminarily determined that the incremental conditional core damage probability for internal and external initiators was 1.54E-06, in the low to moderate risk significance range (White). This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, because the organization did not systematically and effectively evaluate relevant internal and external operating experience in a timely manner. Specifically, Condition Report 55103 documented industry operating experience regarding emergency diesel generator excitation system diodes failing at an increased rate, and the operating experience was not effectively implemented and institutionalized through changes to station processes, procedures, equipment, and training programs, and at least one emergency diesel generator excitation system diode failed due to aging (P.5). |
| 05000482/FIN-2016007-03 | 2016Q2 | Green | P.5 | NRC identified | Inadequate Analysis of Essential Service Water Piping | The team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to April 28, 2016, the licensee failed to verify the design of the essential service water piping because the analyses assumed that the essential service water piping upstream of the containment air coolers was full of water after a loss of offsite power. However, the essential service water pump check valve was never tested to ensure water would not drain from the essential service water piping. In response to this issue, the licensee conducted a preliminary evaluation using data from the last surveillance test and inspection of the check valve, and concluded that the worst-case expected leakage through the check valve was not large enough to cause a water hammer event in the piping that exceeded operability criteria. This finding was entered into the licensee's corrective action program as Condition Reports 104222 and 104184. The team determined that the failure to verify the adequacy of the design of the essential service water piping was a performance deficiency. The performance deficiency was morethan- minor, and therefore a finding, because it related to the design control attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to account for check valve leakage in the essential service water system led to a non-conservative assumption that the piping upstream of the containment air coolers would not drain after a loss of offsite power, which contributes to water hammer events that could challenge the integrity of the piping. 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 was assigned a cross cutting aspect in the area of problem identification and resolution, specifically operating experience, because the water hammer issue was previously documented in several NRC inspection reports, the licensee made recent modifications to the system, and a companion check valve in the normal service water system was installed and correctly categorized in the inservice testing basis document. The operating experience cross-cutting aspect requires that the licensee systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner (P.5). |