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	Quarter	Site	Title	Description
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.
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.
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.
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.
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.
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).
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).
05000293/FIN-2016011-07	2017Q1	Pilgrim	Failure to Report Condition Prohibited by Technical Specifications and a Safety System Functional Failure	The NRC team identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, associated with Entergys failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. Specifically, on September 28, 2016, Entergy identified the A emergency diesel generator was inoperable. The NRC team determined that the condition was prohibited by technical specifications and the inoperability of the A emergency diesel generator existed for a period of time longer than allowed by Technical Specification 3.5.F, Core and Containment Cooling Systems. This was also reportable as a safety system functional failure. Entergy entered this issue into the corrective action program as CR-PNP-2016-09552. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC team evaluated the performance deficiency using traditional enforcement. The violation was evaluated using Section 2.3.11 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. In accordance with Section 6.9.d, Example 9, of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, the NRC team did not assign a cross-cutting aspect to this violation, in accordance with IMC 0612, Appendix B.
05000293/FIN-2016011-13	2017Q1	Pilgrim	Licensee-Identified Violation	10 CFR 50.54(q)(2) requires, in part, that the licensee follow and maintain the effectiveness of an emergency plan to meet the planning standard of 10 CFR 50.47(b)(4). Specifically, the licensee was to maintain the necessary equipment to support the effectiveness of EALs. Contrary to these requirements, PNPS identified in CR-PNP-2016-01491 that on three past occasions (March 15 through August 8, 2012; September 4 through October 14, 2012; and June 4 through June 14, 2015) both trains of the H2O2 monitors and the Post-Accident Sampling System were unavailable to ensure the effectiveness of EAL 24, Deflagration concentrations exist inside PC, for the potential loss of the containment barrier within the Fission Product Barrier category of the EALs. This issue meets the criteria for very low safety significance (Green) because, due to other EALs, an appropriate emergency declaration could have been made in an accurate and timely manner.
05000293/FIN-2016011-12	2017Q1	Pilgrim	Licensee-Identified Violation	10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, and shall be accomplished in accordance with those structures, procedures, and drawings. Entergy procedure EN-DC-148, Vendor Manuals and Vendor Re-Contact Process, Revision 6, requires, in part, that the station update vendor manuals every three years. Contrary to this, in July 2016, PNPS determined through a self-assessment that they had 13 vendor manuals that had not been evaluated for changes within 3 years. The NRC team determined that this finding did not affect the design or qualification of a mitigating structure, system or component; did not represent a loss of a system and/or function; did not result in loss of a train or two safety systems greater than any technical specification allowed outage time; did not result from an actual loss of safety function; and did not involve loss of any external event mitigating system. Consequently, the NRC team determined that this performance deficiency screened as having very low safety significance (Green). PNPS documented this issue in their corrective action program as CR-PNP-2016-05115.
05000293/FIN-2016011-11	2017Q1	Pilgrim	Failure to Adequately Develop and Implement Targeted Performance Improvement Plans	The NRC team identified a Green finding because Entergy did not adequately develop and implement a CAPR of a root cause related to a Category A CR, as required by Entergy Procedure EN-LI-102, Corrective Action Program. Specifically, Entergy did not adequately develop and implement the Targeted Performance Improvement Plans, which were designated as a CAPR for the root cause for the Nuclear Safety Culture Fundamental Problem. Entergy documented this issue in the corrective action program for further evaluation as CR-PNP-2017-00406. The performance deficiency was more than minor because if left uncorrected, it could lead to a more significant safety concern. Specifically, inadequate implementation of the Targeted Performance Improvement Plans could result in recurrence of a culture in which leaders are not holding themselves and their subordinates accountable to high standards of performance, resulting in continuing performance issues at the station. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Resources, Change Management, because leaders did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. In this case, PNPS leaders did not apply sufficient rigor in development and implementation of the Targeted Performance Improvement Plans such that they would be an adequate method to drive and sustain positive changes in the stations safety culture (H.3).
05000293/FIN-2016011-10	2017Q1	Pilgrim	Failure to Promptly Correct a Condition Adverse to Quality for the Residual Heat Removal System	The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not take timely corrective action for a previously identified condition adverse to quality. Specifically, Entergy failed to adequately resolve, through repair or adequate evaluation, gasket leakage on the B residual heat removal heat exchanger, which resulted in continued degradation and leakage for the heat exchanger gasket. Entergy did not consider this leakage as a degraded condition, with the potential to impact both the operability of the residual heat removal system, and PNPSs licensing basis with regards to leakage of a closed loop system outside of containment. After the NRC team raised the issue, Entergy performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. Entergy entered this issue into their corrective action program as CR-PNP-2016-09725. The performance deficiency was 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct identified gasket leakage resulted in continued degradation and leakage of the heat exchanger gasket. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in Human Performance, Conservative Bias, because Entergy failed to use decision making practices that emphasize prudent choices over those that are simply allowable (H.14).
05000293/FIN-2016011-09	2017Q1	Pilgrim	Ineffective Corrective Actions to Address Conditions Adverse to Quality Regarding Components in Contact with or Close Proximity to the Drywell Liner	The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with Entergys failure to correct a condition adverse to quality affecting safety-related equipment. Specifically, during a previous NRC inspection in August 2016, inspectors identified numerous locations in the drywell where non-seismic equipment was either in contact, or close proximity, with the drywell liner and had caused damage. Entergy initiated CRs and performed an operability evaluation for the identified issues. However, following a review of these CRs, the NRC team determined that Entergy failed to take corrective actions to address the condition adverse to quality. Entergy entered this issue into the corrective action program as CR-PNP-2016-09346 and CR-PNP-2016-09377 to perform an extent of condition review, secure the loose grating that had caused damage to the liner, and evaluate the need for a clearance criteria between components such as floor grating and support structures and the containment liner. The performance deficiency was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and 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. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the NRC team determined that this finding was of very low safety significance (Green) because 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), and heat removal components. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the engineering evaluation of the degraded condition identified by the inspectors did not thoroughly evaluate the containment liner issues to ensure that resolutions address causes and extents of condition commensurate with their safety significance (P.2).
05000293/FIN-2016011-08	2017Q1	Pilgrim	Failure to Adequately Monitor the Performance of Maintenance Rule Scoped Components	The NRC team identified a Green non-cited violation of 10 CFR 50.65(a)(2), Requirements for monitoring the effectiveness of maintenance at nuclear power plants. Specifically, Entergy did not demonstrate that the performance of 18 maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, and did not establish goals and monitoring in accordance with 10 CFR 50.65(a)(1). Entergys immediate corrective action was to initiate a CR to evaluate moving the affected systems to 10 CFR 50.65(a)(1) monitoring requirements. Entergy entered this issue in the corrective action program as CR-PNP-2017-00401. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Entergy failed to demonstrate that the performance of the 18 maintenance rule scoped components was being effectively controlled through the performance of appropriate preventive maintenance which adversely impacts the reliability of those systems. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, in that Entergy failed to thoroughly evaluate and ensure that resolution of the identified issue, maintenance not being performed on maintenance rule scoped components, included reclassifying the components as necessary. Specifically, Entergy failed to demonstrate that the performance of Maintenance rule scoped components was effectively controlled through the performance of appropriate preventive maintenance, or through performance goals and monitoring. (P.2).
05000293/FIN-2016011-05	2017Q1	Pilgrim	Failure to Establish Corrective Actions to Address Scope of Procedure Quality Issues	The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy implemented inadequate corrective actions to address the procedure quality issues identified in CR-PNP-2016-02058. Specifically, Entergy inappropriately limited their corrective actions to those procedures that increased integrated risk above normal, and did not include other types of safety-related procedures that did not meet their procedure quality standards and resulted in procedure quality being a problem area. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00400. The performance deficiency was more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Entergy limited corrective actions to procedures that increased integrated risk above normal or trip sensitive and failed to include other procedures associated with safety-related components that reflected the broader population reviewed during the collective evaluation. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that this finding had a cross-cutting aspect related to Human Performance, Resources, because the leaders failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically, based on available resources, Entergy chose to limit the scope of safety-related procedures being revised to only those that resulted in high integrated risk or were trip sensitive (H.1).
05000293/FIN-2016011-04	2017Q1	Pilgrim	Programmatic Issue with Implementation of the Operability Determination Process	The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Specifically, the NRC team identified a programmatic issue because in some cases, Entergy did not enter the operability determination process when appropriate, and, when the process was entered, did not adequately document the basis for operability, in accordance with Procedure ENOP-104, Operability Determination Process, Revision 11. In each of the examples discussed, though the basis for operability was not adequate, all components were determined to be operable upon further evaluation. Entergy entered this issue into their corrective action program as CR-PNP-2017-00626. The performance deficiency was more than minor because if left uncorrected, could lead to a more significant safety issue. Specifically, the failure to enter and document a basis for operability could lead to not recognizing inoperable safety-related equipment, and place the reactor at a higher risk of core damage in a design basis accident. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Teamwork. Specifically, the operations and engineering departments did not demonstrate a strong sense of collaboration and cooperation with respect to holding each other accountable when performing operability determinations to ensure nuclear safety is maintained (H.4).
05000293/FIN-2016011-03	2017Q1	Pilgrim	Failure to Issue Appropriate Corrective Actions to Preclude Repetition for the Causes of the September 2016 Scram	The NRC team identified a Green finding because Entergy did not issue appropriate CAPRs in accordance with Entergy procedure EN-LI-102, Corrective Action Process, Revision 28. Specifically, Entergy did not issue adequate CAPRs associated with Root Cause 1 of the feedwater regulating valve failure in September 2016 that resulted in a manual scram. As a result of the NRC teams questions, Entergy issued procedure 1.13.2, Vendor and Technical Information Reviews, Revision 0, as continuous use to ensure that planners will always have the checklist in-hand when planning work to ensure that appropriate vendor technical information is always included in applicable work instructions. Entergy entered the NRC teams concerns in the corrective action program as CR-PNP-2017-00687 and CR-PNP-2017-00936. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a significant condition adverse to quality, loss of control of feedwater regulating valve 642A and a manual scram. The NRC team evaluated the finding using Exhibit 1, Initiating Events Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not cause a reactor trip or the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8).
05000293/FIN-2016011-02	2017Q1	Pilgrim	Failure to Establish Corrective Actions to Preclude Repetition of a Significant Condition Adverse to Quality	The NRC team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not implement CAPRs for a significant condition adverse to quality identified in root cause evaluation CR-PNP-2016-00716, Implementation of the Corrective Action Program, Revision 2. Specifically, the team identified that CAPRs for Entergys continued weaknesses in the implementation of the corrective action program were inadequate. Entergy entered this issue into their corrective action program for further evaluation as CR-PNP-2017-00053, CR-PNP-2017-00410, and CR-PNP-2017-01134. The performance deficiency was more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to preclude repetition of this significant condition adverse to quality could result in continuing weaknesses in implementation of the corrective action program, which was designated as a fundamental problem, and thus a contributing factor for PNPS Column 4 performance. Additionally, weaknesses with corrective action program implementation could result in equipment issues where operability is not maintained. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specificationallowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because individuals did not follow processes, procedures, and work instructions. Specifically, Entergy did not follow procedure EN-LI-102, which provides the station standards for crafting a corrective action and states, in part, that the corrective action descriptions must be worded to ensure that the adverse condition or cause/factor is addressed (H.8).
05000293/FIN-2016011-01	2017Q1	Pilgrim	Failure to Identify All Root Causes of a Significant Condition Adverse to Quality	The NRC team identified a Green non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not adequately determine all root causes associated with a significant condition adverse to quality related to the failure to identify, evaluate, and correct the A SRVs failure to open upon manual actuation during a plant cooldown on February 9, 2013. Specifically, Entergy did not establish adequate measures to assure that the cause of a significant condition adverse to quality, inadequate shift manager operability determination rigor and its associated causes, were adequately determined and corrective action taken to preclude repetition. Entergys immediate corrective actions included planning to conduct operations management face-to-face conversations with shift manager qualified individuals to reinforce the shift managers responsibility for operability and functionality determination accuracy and rigor. Entergy entered this issue into the corrective action program as CRPNP-2017-00363 and CR-PNP-2017-00828. The performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if left uncorrected, the performance deficiency could have the potential to result in repetition of a failure to identify, evaluate, and correct an SRVs failure to open or a similar significant condition adverse to quality. The NRC team evaluated the finding using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, and determined this finding did not affect the design or qualification of a mitigating structure, system, or component; represent a loss of system and/or function; involve an actual loss of function of at least a single train or two separate safety systems for greater than its technical specification-allowed outage time; or represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant. Therefore, the NRC team determined the finding was of very low safety significance (Green). The NRC team determined that the finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, Entergy incorrectly assumed that CR-PNP-2013-00825 contained inadequate information to determine that the A SRV had not opened, and this assumption ultimately impacted the root cause results documented in CR-PNP-2016-01621 (H.12).
05000482/FIN-2017001-01	2017Q1	Wolf Creek	Failure to Provide Adequate Work Instructions for Preventive Maintenance	Green. The inspectors reviewed a Green, self-revealed non-cited violation of Technical Specification 5.4.1.a and Regulatory Guide 1.33 for the licensees failure to provide adequate work instructions for preventive maintenance on safety-related equipment. Specifically, work instructions to inspect and clean the condensate drain lines on the class 1E air conditioner air handling units lacked guidance for adequately cleaning the drain line. This caused the unit to become non-functional. The licensee took the immediate corrective action to clear the clogged condensate drain line on SGK05B, and entered the issue in the corrective action program as Condition Report 106416. The failure to provide adequate work instructions for preventive maintenance on safety-related equipment is a performance deficiency. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined this finding screened to Green. The inspectors determined that the finding has a problem identification and resolution cross-cutting aspect of resolution because the organization did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. This issue is indicative of current performance because neither the preventive maintenance change process was substantively changed nor were human performance errors associated with the preventive maintenance change corrected, and the same resolution inadequacies that resulted in the inadequate preventive maintenance instructions would be expected to occur (P.3).
05000293/FIN-2016011-06	2017Q1	Pilgrim	Design Change Not Appropriately Reviewed by Entergy	The NRC team identified a preliminary greater than Green finding and apparent violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with Entergys failure to ensure that design changes were subject to design control measures commensurate with those applied to the original design and were approved by the designated responsible organization. Specifically, Entergy received a new style right angle drive for the A emergency diesel generator radiator blower fan from a vendor but failed to adequately review the differences in the design of the drives to identify potential new failure mechanisms for the part or the need for related preventive measures. Entergy entered this issue into the corrective action program as CR-PNP-2016-07443. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone, and affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the team screened the finding for safety significance and determined that a detailed risk evaluation was required based on the A emergency diesel generator being inoperable for greater than the technical specification allowed outage time. Region I senior reactor analysts performed a detailed risk evaluation. The finding was preliminarily determined to be of greater than very low safety significance (greater than Green). The risk important sequences were dominated by external fire risk. Specifically, a postulated fire in the B 4 kilovolt (KV) switchgear room with a consequential loss of the unit auxiliary generator power supply, non-recoverable loss of off-site power (LOOP) to both safety buses A5 and A6, loss of the B emergency diesel generator with the conditional failure of the A emergency diesel generator, along with the loss of bus A8 feed (from the shutdown transformer or station blackout (SBO) diesel generator) to safety buses A5 and A6. The internal event risk was dominated by weather related LOOPs, failure of the A emergency diesel generator, with failure of the B emergency diesel generator and SBO diesel generator to run, along with failure to recover offsite power or the emergency diesel generators. See Attachment 1, A Emergency Diesel Generator Cooling Water System Degradation Detailed Risk Evaluation, for a detailed review of the quantitative criteria considered in the preliminary risk determination. The NRC team did not assign a cross-cutting aspect to this finding because the performance deficiency occurred in May 2000. Entergys program has undergone changes since May 2000, and the NRC team did not identify any recent examples of this performance deficiency. Other aspects of Entergys performance related to this issue are further discussed in Sections 5.10.3 and 6.3.4.
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)
05000482/FIN-2016008-02	2016Q2	Wolf Creek	Failure to Promptly Identify and Correct a Condition Adverse to Quality Associated with the Emergency Diesel Generator B Excitation System Diodes	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to assure that conditions adverse to quality, such as failures, malfunctions, and deficiencies are promptly identified and corrected. Specifically, the licensee failed to promptly identify and correct a failed rectifier bridge diode after smoke was observed coming from the three power potential transformers in the emergency diesel generator exciter cabinet NE106 on June 11, 2014, which contributed to the emergency diesel generator B being declared inoperable and unavailable when it tripped during a 24-hour surveillance test on October 6, 2014. To address the failure to take adequate corrective actions Wolf Creek entered this issue into its corrective action program as Condition Report 105480 and plans to implement a modification to install overcurrent detection for each emergency diesel generators power potential transformer. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify and correct the failed emergency diesel generator excitation system diode contributed to the emergency diesel generator B failure on October 6, 2014. The inspectors evaluated the finding using Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. The inspectors determined this finding is not a deficiency affecting the design or qualification of a mitigating structure, system, or component that maintained its operability or functionality, the finding does not represent a loss of system and/or function, the finding does not represent an actual loss of function of at least a single train for greater than its Technical Specification allowed outage time, and the finding does not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant. Therefore, the inspectors determined the finding was of very low safety significance (Green). The inspectors determined that in accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, conservative bias, because when smoke was identified coming from the power potential transformers on multiple occasions, licensee personnel did not use decision making-practices that emphasize prudent choices over those that are simply allowable, and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. As a result, the licensee missed an opportunity to identify and correct the condition of the failed diode in the static exciter (H.14).
05000482/FIN-2016008-01	2016Q2	Wolf Creek	Failure to Adequately Establish and Adjust Preventive Maintenance for Emergency Diesel Generator Excitation System Diodes	The inspectors identified a preliminary White finding associated with an apparent violation of Technical Specification 5.4.1.a, for the licensees failure to adequately develop and adjust preventive maintenance activities in accordance with Procedure AP 16B-003, Planning and Scheduling Preventive Maintenance, Revision 5. Specifically, the licensee did not create a preventive maintenance replacement task or schedule for emergency diesel generator excitation system diodes, which resulted in emergency diesel generator B being declared inoperable and unavailable when it tripped during a 24-hour surveillance test. The licensee entered this condition into its corrective action program as Condition Report 88665. The licensee restored compliance by establishing preventive maintenance tasks 49286, 49287, 49288, and 49289, which refurbish the power rectifier assemblies and replace the diodes on a 12-year replacement frequency. This finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with one emergency diesel generator excitation system diode failed as a result of thermal degradation, emergency diesel generator B was not operable or available to perform its safety function. The inspectors evaluated the finding using Attachment 0609.04, "Initial Characterization of Findings," worksheet to Inspection Manual Chapter (IMC) 0609, Significance Determination Process, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. In accordance with NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation because it represented an actual loss of function of the emergency diesel generator B for greater than its technical specification allowed outage time. A senior reactor analyst performed a detailed risk evaluation in accordance with Appendix A, Section 6.0, Detailed Risk Evaluation. The calculated change in core damage frequency was dominated by a loss of offsite power initiator leading to station blackout with failures of the turbine-driven and non-safety-related auxiliary feedwater pumps. The analyst did not evaluate the large early release frequency because this performance deficiency would not have challenged the containment. The NRC preliminarily determined that the incremental conditional core damage probability for internal and external initiators was 1.54E-06, in the low to moderate risk significance range (White). This finding has a cross-cutting aspect in the area of problem identification and resolution, operating experience, because the organization did not systematically and effectively evaluate relevant internal and external operating experience in a timely manner. Specifically, Condition Report 55103 documented industry operating experience regarding emergency diesel generator excitation system diodes failing at an increased rate, and the operating experience was not effectively implemented and institutionalized through changes to station processes, procedures, equipment, and training programs, and at least one emergency diesel generator excitation system diode failed due to aging (P.5).
05000482/FIN-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).
05000285/FIN-2016007-01	2016Q1	Fort Calhoun	Licensee-Identified Violation	Technical Specification 5.8.1 requires in part, that procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. That appendix states, in part, that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures appropriate to the circumstances. Contrary to the above, maintenance that can affect the performance of safety-related equipment was not performed in accordance with written procedures appropriate to the circumstances. Specifically, maintenance that can affect the performance of safetyrelated valves was not performed in accordance with a procedure that required the licensee to review all diagnostic test results for compliance with the setpoint criteria for all diagnostic tests performed. As described in CR-2012-01601-017, the licensee restored compliance by writing and implementing procedure ER-FC- 410-AD-SETPOINT, Air-Operated Valve Setpoint Control, Revision 0. This procedure requires, in part, that the licensee review all diagnostic test results for compliance with the setpoint criteria for all diagnostic tests performed. The licensees failure to complete maintenance that can affect the performance of safety-related valves in accordance with written procedures appropriate to the circumstances was a performance deficiency that is more-than-minor because it adversely affected the Procedure Quality attribute of the Mitigating Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this performance deficiency resulted in valve HCV-2987, High Pressure Safety Injection Alternate Header Isolation, being not able to fulfill its design safety function from February, 2013, through July, 2013. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that the finding should be processed through Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 1, 2012. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was not a design or qualification deficiency but represented a loss of train function for greater than the outage time allowed by Technical Specifications. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation in accordance with Manual Chapter 0609, Appendix A, Section 6.0, Detailed Risk Evaluation. The analyst determined that the condition of valve HCV-2987 inoperability would affect only the plants response to a large-break loss-of-coolant accident followed by the failure of the instrument air system. The analyst calculated the initiating-event frequency to be 2.63 x 10-10 /year. Also, the analyst determined that the finding did not affect external initiator risk and would not involve a significant increase in the risk of a large, early release of radiation. Therefore, this violation has very low (Green) safety significance.
05000482/FIN-2015004-03	2015Q4	Wolf Creek	Failure to Perform an Adequate Operability Determination and Consider Design Basis Events	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to accomplish activities affecting quality in accordance with Procedure AP 26C-004, Operability Determination and Functionality Assessment, Revision 31. Specifically, the licensee failed to document an operability determination of sufficient scope to address the capability of safety-related essential service water valves in the control building to perform their specified safety functions in the event of a design basis local intense precipitation event. Immediate corrective actions included completing a prompt operability determination and performing analyses that determined the valves remained operable. Additional corrective actions include accelerating three Fukushima project schedules that include a new sump pump in the turbine building area four cable vault, ground and surface water improvements for non-safety related electrical duct banks, and new sump pumps in electrical manholes near the turbine building. The violation was entered into the licensees corrective action program as Condition Report 100299. 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). Specifically, during design basis local intense precipitation events, the safety-related essential service water train A and B service water cross-connect motor-operated valves EFHV0023, EFHV0024, EFHV0025, and EFHV0026, and the essential service water train A and B to service water system valves EFHV0039, EFHV0040, EFHV0041, and EFHV0042 were susceptible to external flooding hazards, and there was a reasonable doubt on the operability of these essential service water valves; however, subsequent evaluation determined that the essential service water valves would not have been impacted in the event of a design basis local intense precipitation event, and the valves were determined to be operable. In accordance with 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, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects mitigating structures, systems, and components. The finding is not a deficiency affecting the design or qualification of mitigating structures, systems, and components; 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 or two separate safety systems out-of-service for greater than their allowed outage times; and the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, conservative bias, because Wolf Creek did not use decision making-practices that emphasize prudent choices over those that are simply allowable, and proposed action was not determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee did not consider long-term consequences or design basis events when determining how to resolve emergent concerns like the unexpected water in room 3101, which resulted in the licensees failure to thoroughly evaluate and assess impacts to the plant when Condition Report 96404 was entered into the corrective action program on May 17, 2015.
05000482/FIN-2015010-02	2015Q4	Wolf Creek	Failure to Report a Permanent Change in a Licensed Operators Medical Status and Request a Condition be Placed on the Operators License	Wolf Creek Nuclear Operating Corporation (Wolf Creek) identified an apparent violation of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. From May 31, 2006, until July 9, 2015, Wolf Creek had several additional opportunities to identify that medication was required to compensate for a disqualifying medical condition and that a license condition was required during the licensees biennial licensed operator requalification program reviews and medical examinations. On July 9, 2015, a period that exceeded 30 days from when the condition was identified, the facility notified the NRC of the medical condition via a letter requesting amendment to the operators license to include the restriction. On July 15, 2015, the NRC issued the license amendment with the new restriction. This issue was entered into Wolf Creeks corrective action program. The inspector determined that Wolf Creeks failure to provide complete and accurate information to the NRC in the operator license application and to notify the NRC of a change in a licensed operators status for a condition was a performance deficiency. This performance deficiency was known by the licensee and within its ability to foresee and correct and should have been prevented. The inspector determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC relies upon Wolf Creek to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued a renewed operator license to the applicant based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of Inspection Manual Chapter 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute apparent violations in accordance with the NRCs Enforcement Policy and their final significance will be dispositioned in separate future correspondence.
05000482/FIN-2015004-01	2015Q4	Wolf Creek	Inadequate Measures to Assure SGK05A Issues Were Promptly Corrected	The inspectors identified a Green cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees inadequate measures to assure that corrective action was taken to preclude repetition of a significant condition adverse to quality. Specifically, measures to correct train A Class 1E electrical equipment air-conditioning system (SGK05A) issues following two trips of the unit on October 18, 2013, failed to preclude repetition, which resulted in the SGK05A unit tripping twice on May 15, 2015; the train A safety-related batteries, inverters, and alternating and direct current buses being declared inoperable due to the loss of area cooling; two separate Technical Specification 3.0.3 entries; and separate technical specification required reactor power reductions to 93 and 94.7 percent. The licensees immediate corrective actions included troubleshooting to determine the direct cause of the compressor trips, stationing a dedicated operator following the second trip on May 15, 2015, and subsequently implementing Temporary Modification 15-013-GK-00, which restored compliance. Actions to prevent recurrence following the May 15, 2015, SGK05A trips, documented in apparent cause evaluation 96392, included conducting a seminar with station managers to review lessons learned from the event, completing a change package to replace the SGK05A compressor that has been the source of residual contamination that has led to numerous trips of the unit, and tracking of the timely replacement of the SGK05A compressor with a due date of December 15, 2016. Wolf Creek entered this issue into its corrective action program as Condition Reports 96392 and 96397. 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). Specifically, the train A safety-related batteries, inverters, and alternating and direct current buses became inoperable and their capability to respond to initiating events to prevent undesirable consequences was impacted as a result of the SGK05A unit tripping. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 3 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects a mitigating structure, system, and component. The performance deficiency does not affect the design or qualification of a mitigating structure, system, and component, and the structure, system, and component did not maintain its functionality. Additionally, 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 or two separate safety systems out-of-service for greater than their 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 in accordance with the licensees maintenance rule program for greater than 24 hours. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, senior managers did not ensure successful completion of the replacement of the SGK05A compressor in Refueling Outage 20, which was a missed opportunity that resulted in the SGK05A unit tripping twice on May 15, 2015, as a result of the same direct cause.
05000482/FIN-2015004-02	2015Q4	Wolf Creek	Failure to Ensure Essential Service Water Valves Were Adequately Protected from External Flooding Hazards	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 measures to assure that applicable regulatory requirements and the design basis, for applicable structures, systems, and components, are correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to ensure that safety-related essential service water valves in the control building were adequately protected from external flooding hazards in the event of a design basis local intense precipitation event, which resulted in a reasonable doubt on the operability of safety-related essential service water valves. The stations immediate corrective actions included entering the condition into the corrective action program and performing a prompt operability evaluation that showed the essential service water valves remained operable. Additional corrective actions include accelerating three Fukushima project schedules that include a new sump pump in the turbine building area four cable vault, ground and surface water improvements for non-safety related electrical duct banks, and new sump pumps in electrical manholes near the turbine building. The violation was entered into the licensees corrective action program as Condition Report 102250. This finding is more than minor because it is associated with the design control 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). Specifically, during design basis local intense precipitation events, the safety-related essential service water train A and B service water cross-connect motor-operated valves EFHV0023, EFHV0024, EFHV0025, and EFHV0026, and the essential service water train A and B to service water system valves EFHV0039, EFHV0040, EFHV0041, and EFHV0042 were susceptible to external flooding hazards, and there was a reasonable doubt on the operability of these essential service water valves; however, subsequent evaluation determined that the essential service water valves would not have been impacted in the event of a design basis local intense precipitation event, and the valves were determined to be operable. In accordance with 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, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency affects mitigating structures, systems, and components. The finding is a deficiency affecting the design or qualification of mitigating structures, systems, and components, and the structures, systems, and components maintained their operability and functionality. Therefore, the inspectors determined that this finding is of very low safety significance (Green). In accordance with Inspection Manual Chapter 0310, Aspects Within The Cross-Cutting Areas, issued December 4, 2014, the finding has a cross cutting aspect in the area of human performance, challenge the unknown, because Wolf Creek individuals did not stop when faced with uncertain conditions. Specifically, the licensee did not maintain a questioning attitude during flooding walk-downs performed in accordance with NEI 12-07 or during evaluation of Condition Report 59257 to identify and resolve unexpected conditions like the floor drain pathway from the communication corridor to the control building basement (room 3101), which was an opportunity for the station to identify the open pathway from the exterior of the plant.
05000482/FIN-2015003-01	2015Q3	Wolf Creek	Inadequate Implementation of the Breach Authorization Procedure	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees inadequate implementation of Procedure AP 10-104, Breach Authorization, Revision 34, which requires authorization to be obtained prior to breaching certain barriers and door 36042 to be within 0.25 inches of the closed position with door 36043 open. Specifically, control room door 36043, which is a fire, security, and control room ventilation isolation signal barrier, was fully opened prior to the breaching party obtaining required written authorization. Additionally, following the opening of normally closed control room door 36043, control room door 36042 was not adequately positioned, which resulted in a doubt on the operability of the control room envelope boundary. Wolf Creeks immediate corrective actions included closing door 36043 to restore the control room boundary, completing a breach authorization permit until repairs could be completed on door 36043, and entering the issue into its corrective action program as Condition Report 99097. This finding is more than minor because it is associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment), including the radiological barrier functionality of the control room, are maintained and protect the public from radionuclide releases caused by accidents or events. Specifically, while control room door 36043 was opened and control room door 36042 was being credited as the control room envelope boundary, door 36042 was not in the required position to maintain the control room envelope operable, and there was a reasonable doubt on the operability of the control room envelope because neither operations nor stationed security personnel verified and recognized the required position of door 36042. In accordance with Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Exhibit 3 of Inspection Manual Chapter 0609, Appendix A, The Significance Determination process (SDP) for Findings At-Power, issued June 19, 2012, and April 29, 2015, respectively, the performance deficiency was a deficiency affecting the control room, auxiliary, reactor, or spent fuel pool building. The performance deficiency represented a degradation of the radiological barrier function, but the deficiency did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. 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, challenge the unknown, because Wolf Creek did not stop when faced with uncertain conditions or evaluate and manage risks before proceeding. Specifically, Wolf Creeks immediate actions to open door 36043 without a breach authorization permit and its actions to inadequately position door 36042 were not adequately evaluated prior to implementation, which resulted in door 36042 being left in a position inconsistent with Procedure AP 10-104 and a reasonable doubt on the operability of the control room envelope (H.11).
05000244/FIN-2015002-03	2015Q2	Ginna	Inadequate Preferred Auxiliary Feedwater Protection from Potential Block Wall Failures	The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, because Exelon did not provide for verifying or checking the adequacy of design, such as by the performance of design reviews and calculations to ensure that masonry block wall failures in the intermediate building would not challenge preferred auxiliary feedwater (AFW) piping operability. Corrective actions included installation of a temporary modification which corrected the condition, and entering this issue into the CAP. Exelon is evaluating options for a permanent modification to correct the issue. This finding is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, if non-safety-related block wall 8-973-8I failed following a design basis seismic event, portions of the turbine-driven and B motor-driven AFW systems could be impacted by falling blocks. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibits 2 and 4 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the performance deficiency was a deficiency affecting external event mitigation systems (seismic/fire/flood/severe weather protection degraded). The performance deficiency did not involve the degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event (e.g., seismic snubbers, flooding barriers, tornado doors) only a reasonable doubt regarding the operability of the turbinedriven AFW system. Therefore, the inspectors determined that this finding is of very low safety significance (Green). Additionally, the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because Ginna did not implement a CAP with a low threshold for identifying issues and individuals did not identify issues completely, accurately, and in a timely manner in accordance with the program.
05000244/FIN-2015002-04	2015Q2	Ginna	Adequacy of Exelon's Protective Action Recommendation Strategy	Exelons Emergency Plan at Ginna was not in compliance with the regulation at the time the URI was opened. Specifically, as required by 10 CFR 50.47(b)(10), Exelon did not include bodies of water in their PAR scheme for the plume exposure pathway. In December 2014, Exelon worked with the local county OROs and revised the ERPAs at Ginna to include Lake Ontario. Exelon has amended Ginnas Emergency Plan (Revision 03900) and PAR scheme to reflect the change. The NRC concluded that it was not reasonable for Exelon to have been able to foresee and correct the violation caused by not having PARs for Lake Ontario. Specifically, in light of the NRC inspection reports, which approved the licensees Emergency Plan and did not identify this issue, the licensee reasonably concluded that it was in compliance with NRC requirements. Therefore, the NRC did not identify any performance deficiency associated with the violation. IMC 0612, Appendix B, Issue Screening, issued September 7, 2012, directs disposition of this issue in accordance with NRC Enforcement Policy because there was no performance deficiency. Therefore, in accordance with NRC Enforcement Policy, Section 2.2.4.d, which states that a violation involving no performance deficiency is considered an exception to using only the operating reactor assessment program, the inspectors dispositioned this violation using traditional enforcement. The inspectors used NRC Enforcement Policy, Section 6.6, Emergency Preparedness, to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor and best characterized as Severity Level III (low-to-moderate safety significance) because it is similar to Enforcement Policy Example Violation 6.6.c.2. Additionally, the inspectors compared this evaluation to the risk associated with the issue by using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, issued September 23, 2014. The inspectors screened the issue and evaluated it using Table 5.10-1, Significance Examples 50.47(b)(10), and Section 5.0.2.h, which describes the consideration of compensatory measures when screening for significance. The inspectors concluded that because Exelon provided the wind speed and direction on the notification form to the OROs, and because the counties take action to evacuate Lake Ontario at an Alert level, the planning standard function was not lost and would still be accomplished, albeit in a degraded manner. Based on these reviews, were it to be evaluated under the reactor oversight process, the issue would screen as low-to-moderate safety significance (White). This issue was entered into Exelons CAP as AR 1701509. Because the inspectors determined no performance deficiency existed, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation (EA-15-025). Further, because Exelons action and/or inaction did not contribute to this violation, it will not be considered in the assessment process, or the NRCs action matrix. This URI is closed.
05000244/FIN-2015002-01	2015Q2	Ginna	Failure to Perform 1-Hour Fire Tours as Required By the Technical Requirements Manual	The inspectors identified a Green NCV of Ginna Operating License Condition 2.C.(3), Fire Protection, because Exelon Generation Company, LLC (Exelon) failed to perform 1-hour compensatory fire tours as required by the Technical Requirements Manual (TRM). Specifically, while a fire barrier component was physically removed, the TRM required a 1-hour fire watch inspection of the affected fire zone; Exelon was performing a 6-hour fire watch. Corrective actions included performing 1-hour fire tours, reinstalling the fire barrier when the work requiring its removal was completed so that fire tours were no longer required, and entering the issue into the corrective action program (CAP). This finding is more than minor because it adversely affected the protection against external factors (i.e., fire) attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, issued on September 20, 2013, the inspectors determined that the finding is of very low significance (Green), because for localized cable protection (task 1.4.4), an automatic suppression system protected the area where the cable protection was affected by the fire finding. Additionally, the finding has a cross-cutting aspect in the area of Human Performance, Resources, because Exelon did not ensure that procedures were adequate to support nuclear safety.
05000244/FIN-2015002-02	2015Q2	Ginna	Inadequate Procedure Implementation Results in Inadvertent Entry into 72 Hour Technical Specification Action Statement	A self-revealing Green NCV of Technical Specification (TS) 5.4.1, Procedures, was identified for inadequate implementation of procedure M-71.4, Removal and/or Installation of Modules within Defeated or Out-of-Service Instrument Loops. Specifically, while performing maintenance procedures for the sodium hydroxide (NaOH) flow loop power supply replacement, Exelon inadvertently caused a short in electrical circuitry that resulted in an automatic switch of instrument bus C from inverter B to its backup power supply; this caused an entry into a 72-hour TS action statement and actuation of the control room emergency air treatment system (CREATS). Corrective actions included entering this issue into the CAP. This finding is more than minor because it is associated with the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, the inspectors determined that the finding is of very low significance (Green), because the finding was not a deficiency affecting the design or qualification of a mitigating structure, system, and component (SSC); did not represent a loss of system and/or function; and did not represent an actual loss of function of at least a single train. Additionally, the finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Exelon did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes.
05000482/FIN-2015001-01	2015Q1	Wolf Creek	Question Related to Ultrasonic Examination of Reactor Vessel Flange Stud Hole Threads	The inspector identified an unresolved item pertaining to 10 CFR 50 Appendix B, Criterion IX, Control of Special Processes, associated with the licensees method of performing ultrasonic examination of the reactor vessel flange stud hole threads in accordance with applicable American Society of Mechanical Engineers (ASME) Code requirements. The inspector identified several issues of concern while observing the licensees ultrasonic examination of the reactor vessel flange stud hole threads. The inspector questioned whether the licensee would be able to detect any reportable indication within the ASME Code examination zone using the technique employed. The inspector identified that in 2003 the licensee had modified the method used to perform the examination scanning, but never verified that the new methodology was capable of detecting relevant indications within the examination zone. The new method placed the one inch diameter zero angle transducer on a radial arm at the end of an approximately 30 foot pole. The pole is aligned on the handle of the protective cap that covers the stud hole in the flange. The inspector reviewed examination Procedure UT-11, Ultrasonic Stud Hole Threads, Revision 13, and Examination of Reactor Vessel Flange made note of the following: - The inner edge of the transducer is at a nominal distance of 3.875 inches from the center of the stud hole. - The protective cap has a nominal diameter of 7.25 inches or a radius of 3.625 inches while the stud hole diameter is 6.822 +0, -.01 inches. - This places the examination zone of inspection starting at a radius of approximately 3.411 inches and extending to a radius of 4.411 inches. The configuration of the transducer on the pole and the alignment mechanism results in the inside edge of the transducer being placed approximately 0.465 inches from the edge of the stud hole, which is the start of the one inch examination area. Because the technique employs a "zero" angle transducer and the examination area is not directly beneath the transducer, there is a concern with instrument signal coverage. The inspector also identified several procedural compliance issues while reviewing the licensees implementation of UT-11. The inspector questioned the following statements in the procedure: - Procedure UT-11, Section 11.1.1, states in part, The examination volume is a one inch annular band around each stud hole, extending to one stud diameter into the flange. - Procedure UT-11, Section 11.2.2, states in part that, Straight beam examination of ligaments shall be performed. - Procedure UT-11, Section 12.1.1, states in part, All indications which are found that are orientated on a plane normal to the axis of the stud that are equal to or exceed 0.2 in, as measured radially from the root of the thread, shall be reported to the LMT Site Supervisor and recorded on the Ultrasonic Examination report form. There is a concern that the technique currently being utilized by the licensee may not provide adequate coverage of the required examination area and may not be capable of detecting indications orientated on a plane normal to the axis of the stud that are equal to or exceed 0.2 inch, as measured radially from the root of the thread, as required by the licensee's procedure and Section XI of the ASME Code. Additional analysis and simulations need to be completed to determine if the licensee is meeting ASME Code requirements. This issue is being tracked as URI 05000482/2015001-01, Questions Related to Ultrasonic Examination of Reactor Vessel Flange Stud Hole Threads.
05000482/FIN-2015001-04	2015Q1	Wolf Creek	Failure to Station Boundary Watch for Opening Auxiliary Building Emergency Exhaust System Boundary Door	The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, Drawings, associated with the licensees failure to follow the requirements of Station Procedure AP 10-104, Breach Authorization, Revision 32. Specifically, the licensees failure initiate a breach permit and station a boundary watch when the auxiliary building emergency exhaust system boundary door 41015 was opened multiple times for transporting scaffolding from the turbine building to the auxiliary building. Opening this door without compensatory measures rendered the auxiliary building emergency exhaust system inoperable. The license entered this issue into their corrective action program for resolution as Condition Reports 92315 and 92630. The licensees failure to initiate a breach permit and implement required compensatory measures for when the auxiliary building emergency exhaust system boundary door 41015 was open was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the system, structure, and component and barrier performance attribute of the Barrier Integrity Cornerstone, and affected the associated cornerstone objective to ensure the radiological barrier functionality of the auxiliary building emergency exhaust system. Specifically, without a dedicated individual in constant communication with the control room, as required by AP 10-104, opening this door required entry of Technical Specification 3.7.13 Limited Condition of Operation Condition B. The longest period door 41015 was open was approximately one hour without the required compensatory measure. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding screened as having very low safety significance (Green) because the finding only involved a degradation of the radiological barrier function provided for the auxiliary building. The finding has a cross-cutting aspect in the area of human performance associated with work management. Specifically, the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate to the work (H.5).

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