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	Quarter	Site	Title	Description
05000483/FIN-2018003-02	2018Q3	Callaway	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 6 of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for combating emergencies and other significant events. The licensee established Emergency Operating Procedure (EOP) ES-0.2, Natural Circulation Cooldown, Revision 9, in part, to meet the regulatory requirement. Figure 1 of ES-0.2 allowed cooldown rates that exceeded the values used in the license basis for radiological consequence analyses and exceeded the values used in the design of the nitrogen accumulators for atmospheric steam dumps and turbine-driven auxiliary feedwater system actuations. This issue was discussed in Licensee Event Report 2018-002-00, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown Contrary to the above, from April 29, 2008 through May 7, 2018, the licensee failed to maintain procedures for combating emergencies and other significant events. Specifically, the licensee failed to maintain EOPs for natural circulation cooldown. This performance deficiency resulted in atmospheric steam dumps and turbine-driven auxiliary feedwater systems being rendered inoperable due to depletion of the safety-related actuation nitrogen.
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.
05000483/FIN-2018003-01	2018Q3	Callaway	Failure to perform 10 CFR 50.59 evaluation for compensatory measures associated with stagnant, inactive loop	The inspectors identified an unresolved item related to implementation of 10 CFR 50.59, Evaluations Changes, Tests and Experiments, for the licensees failure to perform an adequate evaluation for compensatory measures for a stagnant, inactive loop. The inspectors identified an unresolved item related to implementation of 10 CFR 50.59, Evaluations Changes, Tests and Experiments, for the licensees failure to perform an adequate evaluation for compensatory measures for a stagnant, inactive loop. The licensee enacted compensatory measures to support atmospheric dump valve/turbine-driven AFW pump operability due to an issue identified for natural circulation cooldown with a faulted steam generator (i.e., inactive loop). A reduction in the Technical Specification 3.4.16 dose equivalent iodine (DEI) limit (from 1Ci/gm to 0.4Ci/gm) was imposed without a 10 CFR 50.59 evaluation and/or license amendment. Specifically, the licensee did not consider the compensatory measure of reducing Technical Specification 3.4.16 limits on DEI-131 as a change to technical specifications.The licensee considered this a temporary action that did not meet the intent of 10 CFR 50.90 for a technical specification change.
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.
05000498/FIN-2018002-01	2018Q2	South Texas	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee, has been entered into the licensees corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification 6.8.1.a requires that, Written procedures shall be established, implemented, and maintained covering the activities referenced below: The applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Section 9.a, Procedures for Performing Maintenance, states, in part, that Maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. The licensee established Procedure COM-0001, Conduct of Maintenance, to guide maintenance craft on what to do if a condition or issue arises during a maintenance activity. Specifically, Section 1.4 Supervisor Responsibilities, states, in part, that, If we cannot find the problem with the component or piece of equipment, the issue must be raised to the Division Manager/General Supervisor BEFORE we close the work control document AND return the equipment to operations. Contrary to the above, on March 10, 2017, Unit 1 E1B undervoltage relay was found outside the technical specification acceptance criteria, and was retested until the relay it was back in tolerance and placed back into service (declared operable) instead of raising the issue up to the division manager for further evaluation. The issue was discussed with the electrical maintenance supervisor and the findings were documented in Condition Report 17-12616. The relay was declared operable and placed back into service. Subsequently, after review of the condition report, approximately 99 hours after the relay was declared inoperable, the relay was replaced, and the system declared operable. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the equipment performance 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 undervoltage relay was outside its tolerance and placed back into service without correcting the cause of being outside its tolerance. The inspectors assessed the significance of the finding using Exhibit 2, Mitigating Systems Screening Questions, of Inspection Manual Chapter 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and determined this finding is not a deficiency affecting the design or qualification of a mitigating structure, system, and 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). Corrective Action Reference: Condition Report 17-12616
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.
05000482/FIN-2018002-03	2018Q2	Wolf Creek	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	Wolf Creek	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	Wolf Creek	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.
05000498/FIN-2018001-04	2018Q1	South Texas	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Title 10 CFR 50.9, Completeness and accuracy of information, requires, in part, that information required by the Commissions regulations, orders, or license conditions to be maintained by the licensee shall be complete and accurate in all material respects. STP Nuclear Operating Company, Unit 2 Renewed Facility Operating License Condition 2.E. Fire Protection states, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 62. Updated Final Safety Analysis Report Subsection 9.5.1.6.1 Administrative Controls states, in part that the operability/functional capability of the fire protection systems required to protect safe shutdown capability is assured through the implementation of an administrative program. This program includes compensatory actions for systems out-of-service.Procedure 0PGP03-ZF-0001, Fire Protection Program, Revision 31, Step 7.3 requires, in part, that completed fire watch logs Form 4 or their equivalent shall be retained for 3 years.Contrary to the above, the licensee failed to maintained information required by the Commissions regulations, orders, or license conditions that was complete and accurate in all material respects as evidenced by the following two examples:1. On May 25, 2016, the written fire watch log, documented on Form 4, for Unit 2 Fire Watch 10118, for Room 105, for the hours of 1928 and 2015, indicated that the hourly fire watches were conducted by passing through the areas covered by the fire watch. However, the fire watch never entered Room 105 for these 2 hours. The hourly fire watch patrol data is material to the NRC in that it provides sufficient evidence of compliance with regulatory requirements.2. On May 25-26, 2016, the electronic fire watch scanned logs for Unit 2 Fire Watch 10118, for Room 105 between the hours of 2105 on May 25, 2016, to 0504 on May 26, 2016, show that the 9 hourly fire watches were conducted by passing through the areas covered by the fire watch. However, a temporary scan point was placed at the base of the ladder in Room 002 to scan for Room 105. The hourly fire watch individuals never entered Room 105. The hourly fire watch patrol data is material to the NRC in that it provides sufficient evidence of compliance with regulatory requirements.Significance/Severity Level: Although this violation is willful, it was brought to the NRCs attention by the licensee, it involved isolated acts of low-level individuals, and it was addressed by appropriate remedial action. Therefore, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy. Corrective Action Reference: Condition Report 18-0948
05000498/FIN-2018001-03	2018Q1	South Texas	Licensee-Identified Violation	This violation of very low safety significancewas identified by the licensee and has beenentered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Title 10 CFR50.48, Fire Protection, requires, in part, that licensees have a fire protection plan that outlines the plans for fire protection, fire detection, suppression capability, and limitation of damage.STP Nuclear Operating Company, Unit 2 Renewed Facility Operating License Condition 2.E. Fire Protection states, in part, that the licensee shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 62. Updated Final Safety Analysis Report Subsection 9.5.1.6.1 Administrative Controls states, in part, that the operability/functional capability of the fire protection systems required to protect safe shutdown capability is assured through the implementation of an administrative program. This program includes compensatory actions for systems out-of-service.Procedure 0PGP03-ZF-0001, Fire Protection Program, Revision 31, Step 4.7.2.14, requires, in part, hourly fire watch personnel must pass through the areas covered by the fire watch and then sign and enter the time on the fire watch log, using the bar code reader, at least once every clock hour.Contrary to the above, on May 25 and 26, 2016, the licensees fire watch personnel failed to pass through the areas covered by the fire watch and then sign and enter the time on the fire watch log, using the bar code reader, at least once every clock hour. Specifically, two fire watch individuals documented conducting Unit 2 Fire Watch 10118 for Room 105 starting on May 25, 2016, at 1928 and finishing on May 26, 2016, at 0504 when in fact the individuals did not enter Room 105.Significance/Severity Level: Although this violation is willful, it was brought to the NRCs attention by the licensee, it involved isolated acts of low-level individuals, and it was addressed by appropriate remedial action. Therefore, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy. Corrective Action References: Condition Reports 16-7305, 16-7089, 16-7344, and 16-9077
05000499/FIN-2018001-02	2018Q1	South Texas	Failure to Promptly Identify and Correct a Condition Adverse to Quality Associated with Reactor Containment Fan Coolers	The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI for the failure to promptly identify and correct a condition adverse to quality. Specifically, the backdraft damper of the Unit 2, train B reactor containment fan cooler failed to close, as designed, due to a failed closing spring. The backdraft damper had undergone a preventative maintenance activity one month prior to the failure, but the closing spring degradation was not identified.
05000498/FIN-2018001-01	2018Q1	South Texas	Failure to Perform a Maintenance Risk Assessment Prior to Conducting Maintenance	The inspectors identified a Green, non-cited violation of 10 CFR 50.65(a)(4) for the failure to perform a maintenance risk assessment prior to performing maintenance that could have resulted in a reactor shutdown. Specifically, maintenance performed to install bird netting on the Unit 1 deaerator structure above the balance of plant 13.8 kV transformers was not evaluated or identified as being a threat to stable power operations.
05000483/FIN-2018001-01	2018Q1	Callaway	Failure to Maintain Emergency Operating Procedures	The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, "Procedures," for the licensee's failure to maintain emergency operating procedures for aligning auxiliary feedwater suction sources. Specifically, the licensee added continuous action steps to emergency operating procedures that placed both motor-driven auxiliary feedwater pumps in pull-to-lock and isolated their associated recirculation lines after depleting the two non-safety-related suction sources. These actions cause two of the three safety-related auxiliary feedwater pumps to be rendered inoperable prior to aligning the safety-related suction source of essential service water which is credited in accident analysis.
05000482/FIN-2018001-01	2018Q1	Wolf Creek	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.
05000483/FIN-2017003-01	2017Q3	Callaway	Spurious Containment Spray Pump Start	The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to implement Preventative Maintenance Basis document IC-LSELS, Load Shed and Emergency Load Sequencer (LSELS), Revision 0. Specifically, the licensee failed to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23, a Consolidated Controls 6N232 relay driver card, within the scheduled periodicity. On June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours, of which all 44 hours w ere unplanned. As immediate corrective actions, the licensee replaced the circuit card under Job 17002747, completed post -maintenance testing, and restored the system to operable status on June 30, 2017. The licensee entered this issue into the corrective action program under Condition Report 20170 3433. The failure to replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 within the scheduled periodicity was a performance deficiency. This performance deficiency was more than minor , and therefore a finding, because it adversely affected the equipment performance attribute of the Mitigating Systems Cornerstone and its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, o n June 28, 2017, containment spray train A pump , PEN01A, spuriously started due to the cards failure. As a result, one train of the containment spray system was rendered inoperable for a total of 44 hours . Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At - Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; ( 3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) 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 in accordance with the licensees maintenance rule program for greater than 24 hours. Specifically, the total duration of inoperability was 44 hours which is less 3 than the technical specification allowed completion time of 72 hours for this system. The finding had a cross -cutting aspect in the area of problem identification and resolution associated with resolution because the licensee failed to take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee did not replace load shed and emergency load sequencer relay driver Card NF039AR06SL23 prior to failure although this issue was documented in corrective actions ranging from April 2008 to January 2017 (P.3).
05000482/FIN-2017003-04	2017Q3	Wolf Creek	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	Wolf Creek	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	Wolf Creek	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	Wolf Creek	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-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.
05000483/FIN-2017002-02	2017Q2	Callaway	Failure to Analyze the Effect of Changes to Maintaining the Gaitronics System	Severity Level IV. The inspectors identified a Severity Level IV non- cited violation for the licensees failure to perform an analysis of a change to processes supporting the emergency preparedness program that demonstrated the change did not reduce the effectiveness of the emergency plan in accordance with the requirements of 10 CFR 50.54(q)(3). There were no immediate safety concerns associated with this violation because less than 10 percent of the public address speakers were determined to be degraded or non- functional. This issue has been placed in the licensees corrective action system as Condition Report 201702343. The failure to perform an analysis of the effect of changes in processes supporting emergency preparedness is a performance deficiency within the licensees ability to foresee and correct. The finding was more than minor because the finding was associated with the Facilities and Equipment Cornerstone attribute and adversely affected the Emergency Preparedness Cornerstone objective. The finding was assessed using traditional enforcement because the licensees failure to perform a required analysis impacted the regulatory process . The finding was evaluated using the NRCs Enforcement Policy, dated November 1, 2016, Section 6.6(d) , and was determined to be a Severity Level IV violation because the violation did not affect radiological assessment or offsite notification. Traditional enforcement violations are not assessed for cross -cutting aspects.
05000483/FIN-2017002-01	2017Q2	Callaway	Failure to Follow Motor Control Center Procedure	Green . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow Procedure MPE-ZZ-QS001, Cleaning and Inspection of Motor Control Centers, Revision 34. On May 2, 2017, the licensee failed to ensure contactors operated freely per step 7.6.8 during reassembly of motor control center NG08F for the essential service water cooling tower by pass valve EFHV0066. As a result, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. As immediate corrective actions, the licensee replaced the starter assembly under Job 17001973, completed testing including electrically cycling valve EFHV0066, and restored the system to operable status on May 4, 2017. The licensee entered this issue into the corrective action program under Condition Report 201702418. The failure to follow Procedure MPE-ZZ-QS001 was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it adversely affected the configuration 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, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) 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 in accordance with the licensees maintenance rule program for greater than 24 hours. Specifically, the total duration of 3 inoperability was approximately 57 hours which is less than the allowed completion time of 72 hours for this system. The finding had a cross-cutting aspect in the area of human performance associated with challenge the unknown because the licensee failed to stop when faced with uncertain conditions. Specifically, the maintenance technician encountered resistance when manually operating the contactors, signed off the step as complete, and later rationalized the decision with the supervisor aft er completing the work (H.11 ).
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).
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).
05000483/FIN-2016004-01	2016Q4	Callaway	Licensee-Identified Violation	Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, Section 9.a, requires, in part, that maintenance should be properly pre-planned and performed in accordance with documented instructions appropriate to the circumstances. Contrary to the above, on October 19, 2016, the licensee failed to properly pre-plan and perform a post-maintenance test in accordance with documented instructions appropriate to the circumstances. Specifically, the post-maintenance test for work performed on valve EFHV0066, the essential service water to ultimate heat sink cooling tower train B bypass valve, did not include a seat leak test, which would be necessary for the work performed. As a result, on November 17, 2016, operators discovered this valve leaking by at approximately 3900 gallons per minute. The licensee subsequently determined that the safety function of the ultimate heat sink would not be adversely affected with leakage up to 4100 gallons per minute. The inspectors evaluated the significance of the issue under the Significance Determination Process, as defined in Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and 0609 Appendix A, The Significance Determination Process (SDP) for Findings at-Power, dated June 19, 2012. The inspectors concluded the finding was of very low safety significance (Green) because all questions in Exhibit 2 could be answered no. The licensee entered this issue into their corrective action program as Condition Report 201608791.
05000482/FIN-2016004-01	2016Q4	Wolf Creek	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)
05000498/FIN-2016002-02	2016Q2	South Texas	Failure to Control Steam Generator Water Levels at Low Power	The inspectors documented a self-revealed, non-cited violation of Technical Specification 6.8.1.a, Procedures, for failure to implement procedures for power operation as described in Regulatory Guide 1.33, Revision 2, Appendix A, Section 2.g, dated February 1978. Specifically, the procedure the licensee used for low power operation failed to include adequate instructions for the control of steam generator water levels, which resulted in a plant cooldown, a letdown isolation, a pressurizer power-operated relief valve lift, and unplanned entry into two technical specification action statements. The licensee entered this issue into the corrective action program as Condition Report 2015-26657. The inspectors determined that the failure to control steam generator water levels due to an inadequate procedure during lower power operations was a performance deficiency. The performance deficiency is more than minor because it is associated with the procedure quality 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 failure to control steam generator water levels resulted in a plant cooldown, a reactor coolant system letdown isolation, a pressurizer power-operated relief valve to lift, and unplanned entry into two technical specification action statements. The inspectors screened this finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012. The finding screened as Green per Section B. of Exhibit 1, Initiating Events Screening Questions, because the finding did not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident, did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function, and did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Inspectors determined the finding had a cross-cutting aspect of training in the human performance area because the organization failed to provide training and ensure knowledge was transferred to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, because the licensee provided start-up training and simulator based training, skill of the craft vice detailed procedures was thought to be adequate for controlling steam generator water levels at low power (H.9).
05000483/FIN-2016002-01	2016Q2	Callaway	Failure to Account for Water Hammer Stresses in Essential Service Water System Calculations	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for the essential service water pipe stresses caused by pressure fluctuations of the known column closure water hammer phenomenon. The licensee failed to properly account for essential service water piping membrane stress and impact loads as required by the 1974 ASME Code, Section III, paragraphs ND-3112.4 and ND-3111. Specifically, the licensees design calculations for the essential service water system did not account for the pressure fluctuations caused by a known column closure water hammer phenomenon that occurs during a loss of off-site power or load sequencer testing. The licensee completed a prompt operability determination assuring the system was operable under the current conditions and was completing engineering evaluations of the data collected to demonstrate the operability of the system under design conditions. The licensee entered this issued into the corrective action program as Callaway Action Requests 201603472 and 201603819. The inspectors determined that the licensees failure to account for the pressure fluctuations caused by a known column closure water hammer phenomenon in the design calculations for the essential service water system was a performance deficiency. The 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 associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (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 longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, when the licensee recognized that the column separation water hammer phenomenon was occurring in the essential service water system, they only applied the forces to the containment coolers, not the entire system (H.14).
05000483/FIN-2016002-02	2016Q2	Callaway	Failure to Meet Applicable ASME Code Requirements for Repairs to Components in the Essential Service Water System	The inspectors identified a non-cited violation of 10 CFR 50.55a, Codes and Standards, for the licensees failure to repair various ASME Code Class 3 components in accordance with ASME Code, Section XI requirements. Specifically, the licensee did not follow the applicable ASME Code requirements when making repairs to various components in the ASME Code Class 3 essential service water system. The licensee reasonably determined the essential service water system remained operable, and completed the necessary repairs and testing to restore compliance with ASME Code. The licensee entered this issue into their corrective action program as Callaway Action Requests 201603640 and 201604282. The inspectors determined that the programmatic failure to repair various ASME Code Class 3 components in the essential service water system in accordance with ASME Code was a performance deficiency. The 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 associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (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 longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours in accordance with the licensees maintenance rule program. Specifically, the licensee performed a historical system health review and reasonably determined the essential service water system remained operable because periodic system walkdowns by the system owner and shiftly rounds by operations had not identified significant system leaks, and the appropriate repairs and testing were completed on the affected components. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee failed to ensure training of the personnel was adequate to recognize that the repair of the leaks constituted repairs in accordance with ASME Code, Section XI and thus failed to include the necessary ASME testing requirements in the work performance packages to ensure adequate performance of an activity which affected testing of a safety-related modification/repair to risk-significant systems, and thereby ensure nuclear safety (H.9).
05000483/FIN-2016002-03	2016Q2	Callaway	Failure to Adequately Evaluate Operability for a Degraded Condition	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an adequate operability assessment when a degraded or nonconforming condition was identified. Specifically, after the licensee identified that a severe water hammer transient would occur following a loss of off-site power, the licensee generated an operability evaluation that relied on judgement and inaccurate information which failed to establish a reasonable expectation of operability. Following questions from inspectors the licensee determined that this judgement was not correct and performed a new evaluation to ensure operability of the essential service water system. The licensee entered this issue into their corrective action program as Callaway Action Request 201605488. The licensees failure to properly assess and document the basis for operability when a severe water hammer occurred in the essential service water system was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, severe water hammer transients in the essential service water system due to a loss of off-site power, result in a condition where structures, systems, and components necessary to mitigate the effects of accidents may not have functioned as required. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event, and (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (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 longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensees use of unsupported judgement and incorrect data resulted in an evaluation that failed to demonstrate a reasonable expectation of operability (H.14).
05000483/FIN-2016002-04	2016Q2	Callaway	Failure to Promptly Correct Conditions Adverse to Quality	During an NRC inspection conducted June 6-30, 2016, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that conditions adverse to quality are promptly identified and corrected. Contrary to the above, from November 2010 through June 2016, the licensee failed to promptly correct a condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues which were previously identified by the NRC as non-cited violation 05000483/2010006-01. The failure to resolve these issues resulted in subsequent safety-related equipment failures. This violation is associated with a Green Significance Determination Process finding The inspectors identified a cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to take timely corrective action for a previously identified condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues that were previously identified by the NRC as non-cited violation 05000483/2010006-01 and the failure to resolve these issues resulted in subsequent safety-related equipment failures. The licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The licensee entered this issue into their corrective action program as Callaway Action Request 201604440. The licensees failure to take timely and adequate corrective actions to correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct water hammer and corrosion issue resulted in the licensee declaring safety-related room coolers and chillers inoperable until an analysis of system operability was completed. This affected their capability to respond to initiating events to prevent undesirable consequences Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (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 longer than its allowed outage time, or two separate safety systems out-ofservice for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of resources in the human performance area because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, by failing to address water hammer and corrosion issues, station management failed to ensure that the essential service water system was available and adequately maintained to respond during a loss of off-site power event (H.1).
05000483/FIN-2016002-05	2016Q2	Callaway	Failure to Follow Plant Foreign Material Exclusion Procedure	The inspectors reviewed a self-revealed finding for the licensees failure to follow the plant procedure for foreign material exclusion. Specifically, after finding foreign material (broken cable ties) within the main generator excitation transformer, established as a foreign material exclusion Level 2 area, the licensee failed to determine the reason for the foreign material and enter the issue into the corrective action program for resolution as required by Procedure APA-ZZ-00801, Foreign Material Exclusion, Revision 32. The licensees failure to follow the plant procedure for foreign material exclusion was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, after identifying several broken cable ties on the floor inside a foreign material exclusion Level 2 area the licensee did not determine the reason for the foreign material nor enter the condition into the corrective action program as required by Procedure APA-ZZ-00801. Because the licensee failed to understand what caused the cable tie degradation, a subsequent cable tie failure resulted in a plant trip. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the finding was determined to be of very low safety significance because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, several groups within the licensees organization were unaware the excitation transformer cabinet was classified as a foreign material exclusion Level 2 area nor the requirements if foreign material is found within the foreign material exclusion area (H.9).
05000498/FIN-2016002-01	2016Q2	South Texas	Inadequate Scaffold Procedure to Ensure Safety-Related Equipment Not Impacted	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide an adequate scaffold procedure to ensure that safety-related equipment would not be impacted. Specifically, Procedure 0PGP03-ZM-0028, Erection and Use of Temporary Scaffolding, Revision 20, did not give scaffold clearance parameters when constructing scaffold around safety-related mechanical and structural components, nor did it direct an engineering evaluation if scaffold is in contact with safety-related components or when clearances cannot be met. The licensee entered this issue into the corrective action program as Condition Report 16-5503. The failure to have adequate procedural guidance for erecting temporary scaffold in the vicinity of safety-related components was a performance deficiency. Specifically, Procedure 0PGP03-ZM-0028, Erection and Use of Temporary Scaffolding, Revision 20, only described scaffold clearance around safety-related electrical equipment, but not safety-related mechanical and structural components. The performance deficiency is more than minor, and therefore a finding, because if left uncorrected could become a more safety significant safety issue following a seismic event. Specifically, the continued practice of building scaffolding in contact with safety-related equipment and without an engineering evaluation could lead to damage, inoperability, or unavailability during system perturbations or following a seismic event. The inspectors evaluated this finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Screening Questions. The inspectors determined the finding was of very low safety significance (Green) because the finding did not: 1) affect the design or qualification of a mitigating structure, system, and component; 2) represent a loss of system and/or function; 3) represent an actual loos of function of at least a single train for greater than its technical specification allowed outage time or two separate safety systems for greater than its technical specification allowed outage time; or 4) represent an actual loss of function of one or more technical specification trains of equipment designated as high safety significance in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined that the finding has a cross-cutting aspect of self-assessment in the problem identification and resolution area, because the licensee had not recently conducted a periodic and critical review of the temporary scaffold program and procedures (P.6).
05000482/FIN-2016002-01	2016Q2	Wolf Creek	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-2016002-02	2016Q2	Wolf Creek	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-2016002-03	2016Q2	Wolf Creek	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be accomplished in accordance with documented instructions, procedures, or drawings of a type appropriate to the circumstances. Licensee Procedure AP 26C-004, Operability Determination and Functionality Assessment, Revision 32, an Appendix B quality related procedure, provides instructions for determining whether equipment is operable when oil leakage is identified. Procedure AP 26C-004, Step 6.2.1.1, states in part, that if operability of a system/component is being questioned due to system leakage that the leak rate has been quantified and total identified leakage for the affected system has been determined and compared to the limits of Attachment F, Allowable Oil Leakage for Successful Mission. Contrary to the above, from May 28, 2016, until May 31, 2016, operability of a system/component was being questioned due to system leakage and the leak rate had not been quantified and the total identified leakage for the affected system was not determined and compared to the limits of Attachment F, Allowable Oil Leakage for Successful Mission. Specifically, operability of the B component cooling water pump was questioned due to system leakage as documented in Condition Report 104910, and the leak rate had not been quantified and the total identified leakage for the affected system was not determined, which resulted in the immediate operability determination being incorrect and the immediate operability determination requiring revision. Immediate corrective actions included revising the immediate operability determination for the B component cooling water pump from operable to inoperable, generating a required reading for senior reactor operators, and documenting Condition Report 104959. Using Exhibit 2, Mitigating Systems Screening Questions, of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Finding At-Power, dated June 19, 2012, the inspectors 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 it 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).
05000482/FIN-2016002-04	2016Q2	Wolf Creek	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-2016002-05	2016Q2	Wolf Creek	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-2016002-06	2016Q2	Wolf Creek	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-2016001-01	2016Q1	Wolf Creek	Failure to Adequately Establish and Adjust Preventive Maintenance Activities for Control Room Air Conditioning Unit SGK04A Sensing lines and Fittings	The inspectors identified a Green 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 adequately develop a preventive maintenance replacement task and schedule for control room air conditioning unit SGK04A refrigerant sensing lines and fittings. The licensees immediate actions included securing and declaring the SGK04A system inoperable, completing corrective maintenance to eliminate the refrigerant leak, and confirming that the impacted preventive maintenance frequency was adequately established. The licensee entered this condition into the corrective action program as Condition Reports 101862 and 101867. 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 utilized Inspection Manual Chapter 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 structures, systems, and components (SSC) 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 it 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 finding has a cross-cutting aspect in the area of human performance, 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 procedures and other resource materials were available to support successful work performance when setting preventive maintenance activity base dates, which resulted in the licensee failing to adequately develop and adjust preventive maintenance activities associated with control room air conditioning unit SGK04A refrigerant sensing lines and fittings (H.1).
05000498/FIN-2016001-01	2016Q1	South Texas	Failure to Identify and Correct Faulty NI-36 Channel	Inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the licensees failure to identify a condition adverse to quality. Specifically, the licensee failed to identify that a faulty logarithmic amplifier was producing inaccurate intermediate range nuclear instrument channel NI-36 indications. This resulted in multiple instances of delays in the change of state of reactor trip instrumentation permissive P-6 when shutting down the reactor. The licensee replaced NI-136s log current amplifier using approved procedures and returned the channel to service. This issue was entered into the corrective action program as Condition Report 16-1227. The licensees failure to identify a condition adverse to quality regarding intermediate range nuclear instrument channel NI-36 was a performance deficiency. Specifically, the licensee failed to identify a faulty log current amplifier in intermediate range nuclear instrument channel NI-36, which led to multiple instances of inaccurate indication and delays in the change of state of reactor trip instrumentation permissive P-6, when shutting down the reactor that required operator action and unplanned technical specification entries. This performance deficiency is more than minor and, therefore, a finding because it impacts 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 screened this finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated June 19, 2012. The finding screened as Green per Section A of Exhibit 2, Mitigating Systems Screening Questions, because the finding did not affect the design or qualification of a mitigating structure, system, or component; the finding did not represent a loss of the 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 in accordance with the licensees maintenance rule for more than 24 hours. Inspectors determined the finding had a cross-cutting aspect of conservative bias in the human performance area because leaders did not take a conservative approach to decision making, particularly when information is incomplete or conditions are unusual. Specifically, the licensee made the decision not to enter their procedure for preventing recurring equipment problems process, even though entry criteria to do so was met, because of a false confidence that the correct cause had already been identified (H.14).
05000483/FIN-2016001-02	2016Q1	Callaway	Inadequate Operability Evaluation for Degraded Flood Mitigation Capability in Piping Penetration Room	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an adequate operability determination for safety related components located in the 1988 foot auxiliary building train B piping penetration room (room 1203) based on degraded internal flooding drain capability. Specifically, the immediate operability determination included incorrect assumptions that were not verified to support the operability determination as required by Procedure ODP-ZZ-00001, Addendum 15, Operability and Functionality Determinations, Revision 8. The immediate corrective action was to implement a compensatory measure to support operability of the equipment in room 1203. The issue was placed in the corrective action program as Callaway Action Request 201601412. The licensees failure to verify assumptions used in the immediate operability determination and ensure a sound basis for operability exists per plant procedures was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is similar to examples 3.j and 3.k in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, and if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, failure to perform adequate operability evaluations by verifying assumptions and ensuring a sound basis for operability exists may result in the failure to enter the appropriate limiting conditions of operation for technical specification equipment. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding involved the degradation of equipment specifically designed to mitigate a flooding initiating event, therefore, Exhibit 4, External Events Screening Questions, was used to complete the screening. The finding was determined to need a detailed risk evaluation because if the equipment (i.e., floor drain lines) is assumed to be completely failed or unavailable, it would degrade one or more trains of a system that supports a risk significant system or function. In consultation with the Senior Reactor Analyst, the finding was determined to be of very low safety significance because, based on the actual condition of the drains and the extent of the clogging in room 1203, an evaluation by the licensee showed that the maximum internal flooding water level in the room would not challenge the operability of any equipment needed for safe shutdown or to mitigate an accident. This finding has a team work cross-cutting aspect in the human performance cross-cutting area because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, inadequate communication between engineering and operations personnel led to the belief that a passageway existed between rooms 1203 and 1204 when it did not (H.4).
05000498/FIN-2016001-02	2016Q1	South Texas	Licensee-Identified Violation	Technical Specification 6.8.1.a states, in part, written procedures shall be established, implemented, and maintained covering applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2, February 1978. Section 3.a of appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for the startup, operation, and shutdown or the RCS, and Section 9.c requires procedures for repair or replacement of major equipment that is expected to be repaired or replaced during the life of the plant. Contrary to the above, the licensee failed to have procedures established for the operation of the RCS and for the repair of major equipment that is expected to be repaired during the life of the plant. Specifically, on November 2, 2015, without procedural guidance, the Unit 1 reactor coolant pump 1C was recoupled with the RCS at approximately 66 feet in the cavity. Coupling the pump to the motor in this condition introduced unfiltered RCS water into the seal cartridge area. On November 11, 2015, operations placed reactor coolant pump 1C into service and immediately noted a higher than normal leak off from the number 1 seal. Several attempts were made to adjust the seal and reduce the leakage, but on November 13, 2015, the decision was made to depressurize and cool down the RCS to repair the seal. The licensee discovered that foreign material from the unfiltered RCS had contaminated the seal. The licensee determined that this occurred during pump recoupling while at 66 feet in the reactor cavity. This finding has a very low safety significance (Green) because the finding did not result in an RCS leak rate that exceeded that of a small LOCA or have likely affected other systems that are used to mitigate a LOCA resulting in a total loss of their function. This issue was entered into the licensees corrective action program as Condition Report 15-24818.
05000498/FIN-2016001-03	2016Q1	South Texas	Licensee-Identified Violation	Technical Specification 6.8.1.a. states, in part, written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a requires, in part, that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstance. The licensee established procedure 0PMP04-ZG-0022, Hills McCanna/Rockwell/Edwards Ball Valve Maintenance, Revision 24, to meet the Regulatory Guide 1.33 requirement for rebuilding chemical and volume control system (CVCS) mixed bed demineralizer drain valve CV-123A, a safety-related valve. Step 5.10 of this procedure directs stem seals to be installed during bonnet reassembly. Contrary to the above, on October 26, 2015, the licensee failed to follow Step 5.10 that directs stem seals to be installed during bonnet reassembly. Specifically, the stem seals were installed in the wrong locations and, on November 13, 2015, resulted in a 12-15 gpm RCS leak rate when the CVCS mixed bed demineralizer 1A was placed in service. A search for the leak determined that CV-123A was leaking by due to the lower stem seals being improperly installed. The licensee restored compliance by correctly rebuilding valve CV-123A, demineralizer 1A drain valve, in accordance with the approved procedure. The finding was of very low safety significance because the finding did not affect other systems used to mitigate a LOCA resulting in a loss of their function. This issue was entered into the licensees corrective action program as Condition Report 15-25192.
05000483/FIN-2016001-01	2016Q1	Callaway	Possible Incorrect Screening of the Spent Fuel Pool Decay Heat Removal Key Safety Function	The inspectors identified an unresolved item associated with the National Fire Protection Association (NFPA) Standard 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants, non-power operations assessment. Specifically, the inspectors developed an issue of concern in that the licensee screened the potential loss of spent fuel pool cooling from further consideration for any fire event based on adequate procedural guidance and time when the procedures would not maintain the fuel in a safe and stable condition. On January 13, 2014, the licensee transitioned their fire protection program to a risk-informed, performance-based program based on NFPA Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. Paragraph 1.3.1 of NFPA Standard 805 requires licensees to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. Paragraph 1.5.1 of NFPA Standard 805 lists five nuclear safety performance criteria. These criteria provide requirements to demonstrate that fire protection features are capable of providing reasonable assurance that the plant is not placed in an unrecoverable condition in the event of a fire. For the decay heat removal nuclear safety performance criterion, the standard requires that decay heat removal shall be capable of removing sufficient heat from the reactor core or spent fuel such that fuel is maintained in a safe and stable condition. Paragraph 1.6.56 of NFPA Standard 805 defines safe and stable conditions: For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff <0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff <0.99 and fuel coolant temperature below boiling. The licensee described how they satisfied the nuclear safety performance criteria in Calculation KC-26, Nuclear Safety Capability Assessment, Revision 1. The Nuclear Safety Capability Assessment applied to both power and non-power operations. For non-power operations, the licensee evaluated the spent fuel pool decay heat removal key safety function and determined that the spent fuel pool decay heat removal key safety function did not require a detailed review since adequate time was available, and procedural guidance was provided, for operators to respond to and mitigate a loss of spent fuel pool decay heat removal, even under full hot core offload conditions. The licensee stated that the shortest time to boil, under worst case conditions for a normal plant shutdown, was two hours. In addition, the licensee stated that all of the analyses to address a loss of spent fuel pool decay heat removal utilized a success criterion of no boiling. The licensee implemented the process outlined in Frequently Asked Question (FAQ) 07-0040, Non-Power Operations Clarifications, Revision 4, for the non-power operations assessment. This FAQ stated that licensees should conservatively assume the entire contents of a fire area are lost and document the loss of success paths. This FAQ also stated that licensees should specifically identify those areas (pinch points) that cause the loss of all success paths for a key safety function. The inspectors noted that the licensee did not perform these actions for the spent fuel pool decay heat removal key safety function because this key safety function was screened out from further consideration. If the licensee had evaluated the spent fuel pool decay heat removal key safety function using the process outlined in this FAQ, then the licensee would have assumed that both trains of spent fuel pool cooling are lost during a fire in the fuel handling building because both trains are located within the same fire area and were unprotected. This FAQ also stated that fire modeling may be used to determine if postulated fires in a fire area are expected to damage equipment (and cabling), thereby eliminating a pinch point. However, the licensee stated that no fire modeling was used to eliminate the identification of pinch point fire areas as part of the non-power operations assessment performed using the process in FAQ 07-0040. In the event that a fire in the fuel handling building disabled both trains of spent fuel pool cooling, operators were expected to enter Procedure OTO-EC-00002, Spent Fuel Pool High Temperature, Revision 9, due to the increasing temperature of the spent fuel pool. This procedure provided directions for operators to restore one or both trains of spent fuel pool cooling. Since both trains of spent fuel pool cooling were assumed lost due to the fire, the operators would be unable to restore spent fuel pool cooling using this procedure. After a period of time, the spent fuel pool would begin boiling and the level would begin lowering. At this time, operators were expected to enter Procedure OTO-EC-00001, Loss of SPF/Refuel Pool Level, Revision 13. Procedure OTO-EC-00001 directed the operators to open two normally locked essential service water valves to restore and maintain spent fuel pool level. The licensees procedures allowed the spent fuel pool to reach boiling conditions prior to restoring and maintaining level. Since NFPA Standard 805 defined safe and stable conditions, in part, as fuel coolant temperature below boiling, the procedures did not maintain the fuel in a safe and stable condition. The inspectors identified an issue of concern in that the licensee screened the potential loss of spent fuel pool cooling from further consideration for any fire event based on adequate procedural guidance and time when the procedures would not maintain the fuel in a safe and stable condition. The inspectors determined that additional information is required to determine if a performance deficiency exists. Specifically, the inspectors need to determine if this scenario should have been addressed as part of the current FAQ 07-0040 guidance, or if new guidance is needed to address this type of scenario where the full core has been offloaded to the spent fuel pool. On March 31, 2016, additional guidance was requested from the Office of Nuclear Reactor Regulation via a request to review and update FAQ 07-0040. This memorandum is documented in ADAMS as Accession Number ML16091A152. The licensee entered this issue of concern into the corrective action program as Callaway Action Request 201600726. This issue of concern is being treated as Unresolved Item 05000483/2016001-01, Possible Incorrect Screening of the Spent Fuel Pool Decay Heat Removal Key Safety Function.

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