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	Quarter	Site	Title	Description
05000483/FIN-2018003-01	2018Q3	Callaway	Failure to perform 10 CFR 50.59 evaluation for compensatory measures associated with stagnant, inactive loop	The inspectors identified an unresolved item related to implementation of 10 CFR 50.59, Evaluations Changes, Tests and Experiments, for the licensees failure to perform an adequate evaluation for compensatory measures for a stagnant, inactive loop. The inspectors identified an unresolved item related to implementation of 10 CFR 50.59, Evaluations Changes, Tests and Experiments, for the licensees failure to perform an adequate evaluation for compensatory measures for a stagnant, inactive loop. The licensee enacted compensatory measures to support atmospheric dump valve/turbine-driven AFW pump operability due to an issue identified for natural circulation cooldown with a faulted steam generator (i.e., inactive loop). A reduction in the Technical Specification 3.4.16 dose equivalent iodine (DEI) limit (from 1Ci/gm to 0.4Ci/gm) was imposed without a 10 CFR 50.59 evaluation and/or license amendment. Specifically, the licensee did not consider the compensatory measure of reducing Technical Specification 3.4.16 limits on DEI-131 as a change to technical specifications.The licensee considered this a temporary action that did not meet the intent of 10 CFR 50.90 for a technical specification change.
05000483/FIN-2018003-02	2018Q3	Callaway	Licensee-Identified Violation	This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy.Violation: Technical Specification 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. Section 6 of Appendix A to Regulatory Guide 1.33, Revision 2, requires procedures for combating emergencies and other significant events. The licensee established Emergency Operating Procedure (EOP) ES-0.2, Natural Circulation Cooldown, Revision 9, in part, to meet the regulatory requirement. Figure 1 of ES-0.2 allowed cooldown rates that exceeded the values used in the license basis for radiological consequence analyses and exceeded the values used in the design of the nitrogen accumulators for atmospheric steam dumps and turbine-driven auxiliary feedwater system actuations. This issue was discussed in Licensee Event Report 2018-002-00, Inadequate EOP Guidance for Asymmetric Natural Circulation Cooldown Contrary to the above, from April 29, 2008 through May 7, 2018, the licensee failed to maintain procedures for combating emergencies and other significant events. Specifically, the licensee failed to maintain EOPs for natural circulation cooldown. This performance deficiency resulted in atmospheric steam dumps and turbine-driven auxiliary feedwater systems being rendered inoperable due to depletion of the safety-related actuation nitrogen.
05000458/FIN-2018012-04	2018Q2	River Bend	Failure to Submit a Licensee Event Report for a Manual Scram	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report System, for the licensees failure to submit a required licensee event report (LER). Specifically, on February 1, 2018, after an unexpected trip of the recirculation pump B, the licensee initiated a manual scram of the reactor that was not part of a preplanned sequence and failed to submit an LER within 60 days.
05000458/FIN-2018012-01	2018Q2	River Bend	Failure to Conduct Adequate Transient Snap Shot Assessment Following Recirculation Pump Trip	The inspectors identified a finding for the licensees failure to adequately validate simulator response during a transient snap shot assessment following an unexpected trip of reactor recirculation pump A on December 19, 2012.
05000458/FIN-2018012-03	2018Q2	River Bend	Failure to Establish Procedural Guidance for Determining Core Flow During Unanticipated Single Loop Operations	The inspectors reviewed a self-revealed,non-cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish appropriate instructions in the abnormal operating procedure for thermal hydraulic instabilities. Specifically, the procedural step for determining core flow when in single loop operations at low power did not provide appropriate instructions to operators. As a result, station personnel could not conclusively determine core flow and inserted a manual reactor scram.
05000458/FIN-2018012-05	2018Q2	River Bend	Failure to Develop an Adequate Operational Decision-Making Issue for Compensatory Measures Related to a Degraded Condition of the Feedwater System Sparger Nozzles	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to develop an adequate Operational Decision-Making Issue (ODMI) document per Procedure EN-OP-111, Operational Decision-Making Issue Process. Specifically, the licensee failed to develop an ODMI that provided adequate guidance to the operators for safely operating the plant with degraded feedwater sparger nozzles.
05000458/FIN-2018012-06	2018Q2	River Bend	Failure to Provide Adequate Procedures for Post-Scram Recovery	The inspectors reviewed a self-revealed, non-cited violation of Technical Specification 5.4.1.a for the licensees failure to establish, implement and maintain a procedure required by Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, Procedure OSP-0053, Emergency and Transient Response Support Procedure, Revision 22, which is required by Regulatory Guide 1.33, inappropriately directed operations personnel to establish feedwater flow to the reactor pressure vessel using the main feedwater regulating valve as part of the post-scram actions. This resulted in the main feedwater regulating valves being operated outside their design limits. This resulted in catastrophic failure of the main feedwater regulating valve variseals and subsequent damage to multiple fuel assemblies.
05000458/FIN-2018012-02	2018Q2	River Bend	Failure to Identify and Correct a Broken Feedwater Chemistry Probe	Two examples of a self-revealed non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were identified for the licensees failure to identify that a broken chemistry probe in the feedwater system had the potential to cause an adverse impact on plant safety, and promptly implement appropriate measures to address that condition.
05000458/FIN-2018012-07	2018Q2	River Bend	Failure to Perform 10 CFR 50.59 Evaluation for Main Feedwater System Sparger Nozzle Damage	The inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59 , Changes, Tests, and Experiments, for the licensees failure to provide a written safety evaluation for the determination that operation with compensatory measures for damaged feedwater sparger nozzles did not require a license amendment pursuant to 10 CFR 50.90, Application for amendment of license, construction permit, or early site permit. Specifically, the licensee failed to recognize that compensatory measures prohibiting operation in single loop conditions required technical specification changes, and as such required prior NRC approval.
05000458/FIN-2018001-02	2018Q1	River Bend	Installation of an Incorrectly Specified Relay Causes Plant Transient and Reactor Scram	The inspectors reviewed two examples of a self-revealed finding for the licensees installation of an incorrectly specified relay in 1) the control circuitry for the feedwater level control systemand 2) the turbine generator voltage regulator circuitry. In each instance, the incorrectly specifiedrelay failed in service, causing a plant transient and automatic reactor scram
05000458/FIN-2018001-01	2018Q1	River Bend	Failure to Implement Procedure for Storage of Material in the Pools	The inspectors identified a non-cited violation of Technical Specification 5.4.1.a for the licensees failure to implement written procedures for activities referenced in Appendix A of Regulatory Guide 1.33, Revision 2, dated February 1978. Specifically, the licensee failed to implement radioactive material control Procedure ADM-0071, Fuel Pools Material Control, Revision 8, for the storage and movement of spent Tri-Nuke filters
05000313/FIN-2017007-01	2017Q3	Arkansas Nuclear	Failure to Promptly Identify and Correct an Inadequate Design Bases Calculation	The inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, from 1996 until August 10, 2017, the licensee failed to properly resolve the environmental conditions in room 38 following a high-energy line break, even when challenged during a self-assessment by members of the quality assurance group in June 29, 2015. In response to this issue, the licensee determined that in the event of a break in the letdown line, an engineered safety feature signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 40 seconds, preventing room 38 from going harsh. This finding was entered into the licensees corrective action program as Condition Report CR-ANO-1-2017-02441. The inspectors determined that the licensees failure to adequately evaluate and take prompt corrective actions to resolve an identified condition adverse to quality related to the high energy line break analysis for room 38 was a performance deficiency. 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 objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the licensee identified that the environmental conditions in room 38 of the auxiliary building were harsh, as determined by Design Bases Calculation CALC-01-EQ-1002-02, they failed to properly resolve the condition adverse to quality. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human resources, training, because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, there was a lack of understanding of the current licensing bases for the plant displayed by engineering, operations, and management (H.9).
05000313/FIN-2017002-03	2017Q2	Arkansas Nuclear	Failure to Comply with ECCS Technical Speci fications	Green . The inspectors reviewed a Green self -revealing finding and associated non -cited violation of Unit 1 Technical Specification 3.5.2, Emergency Core Cooling System (ECCS) Operating, for the licensees failure to ensure the operability of the P36A high pressure injection pump after reinstalling its feeder breaker during a unit outage. A violation of Unit 1 Technical Specification 3.0.4 was also identified for making a mode change without meeting the requirements to do so. Following unit restart, the pump failed to start during routine equipment rotation, resulting in one train of emergency core cooling system being inoperable for long er than allowed by Unit 1 Technical Specifications. The licensee subsequently identified that the feeder breaker had not been fully racked into position. Inspectors also noted that the breaker had been racked in manually rather than using the normal electric racking tool, and no special precautions had been taken to ensure this infrequently -used method was successful. When the breaker was correctly racked in, the pump was satisfactorily tested. The licensee subsequently verified that all similar breakers were correctly racked into position. The licensee entered this issue into their corrective action program as Condition Report CR- ANO -1-2017- 01764. The inspectors determined that the failure to verify that the P36A high pressure injection pump was operable after racking its feeder breaker into the switchgear cubicle was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. 4 The inspectors performed the initial significance determination for the performance deficiency using NRC Inspection Manual 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012 , and concluded that it required a detailed risk evaluation because it involved the loss of a single train of mitigating equipment for longer than the technical specification allowed outage time. Therefore, a Region IV senior reactor analyst performed a bounding detailed risk evaluation. The estimate in the increase in core damage frequency is 4.4 E-8 per year, or of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because the licensee failed to ensure that individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to verify that the pump was operable after its breaker was rein stalled, even though an infrequently-used method was employed (H.12).
05000313/FIN-2017002-04	2017Q2	Arkansas Nuclear	Licensee-Identified Violation	Title 10 CFR 50.55a(g)4, Inservice Inspection Standards Requirement for Operating Plants, states in part, Throughout the service life of a pressurized water -cooled nuclear power facility, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the ASME Code. The ASME Section XI, Article IWA - 2610, requires that all welds and components subject to a surface or volumetric examination be included in the licensees inservice inspection program. This includes identifying system supports in the inservice inspection plan, per ASME Section XI, Article IWA -1310. Contrary to the above, prior to March 9, 2017, the licensee did not ensure that all welds and components subject to a surface or volumetric examination were included in the licensees inservice inspection. Specifically, the licensee did not apply the applicable inservice inspection requirements for surface or volumetric examination to all portions of the Unit 2 emergency feedwater system within the system ASME Code Class 3 boundary. The licensee identified that they failed to include the emergency feed pump supports in their inservice inspection program. The licensee entered this issue into their corrective action program as Condition Report CR- ANO -2-2016 -01023 and reasonably determined the emergency feedwater system remained operable. The licensee restored compliance by inspecting the supports, with no degradation identified, and entering the emergency feedwater pump supports into the ASME Section XI program. The finding was of very low safety significance (Green) because the finding did not 34 represent an actual loss of safety function of a system or train and did not result in the loss of a single train for greater than technical specification allowed outage time. This issue was entered into the licensees corrective action program as Condition Report CR- ANO -2-2016- 01023.
05000368/FIN-2017002-01	2017Q2	Arkansas Nuclear	Failure to Follow Fire Protection Program Procedures	Green . The inspectors identified a finding and associated non -cited violation of License Conditions 2.C.( 3)(b), Fire Protection, for Arkansas Nuclear One Unit 2, associated with the failure to adequately implement the fire protection program. Specifically, the licensee failed to follow the requirements for control of flammable liquid lockers and compressed hydrogen gas cylinders. The licensee immediately removed the hydrogen cylinders and stored them in an approved location and began processing the flammable liquid lockers through the design change process. The licensee entered these issues into their corrective action program as Condition Reports CR -ANO -2-2017- 01525 and CR -ANO -C-2017 -01508 . The failure to properly control transient combustible material in accordance with the approved fire protection program was a performance deficiency. The finding was considered more than minor because storing unanalyzed flammable material could result in the potential to exceed combustible material limits , and is associated with the protection against external factors attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. Specifically, the failure to follow procedures resulted in conditions that increased the risk of fire which could upset plant stability and challenge critical safety functions. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and assigned the finding to the Fire Prevention and Administrative Controls category; because it affected the licensees combustible materials control. The finding was determined to be Green, or very low safety significance, in accordance with Inspection Manual Chapter 0609, Appendix F, Question 1.3.1, because the reactor would have been able to reach and maintain safe shutdown since the postulated fires would not have affected both trains of safe shutdown equipment . This finding had a cross -cutting aspect associated with teamwork within the human performance area since multiple groups in the licensee staff were involved in the decisions that resulted in the improper introduction of the flammable liquids lockers and the improper storage of the hydrogen cylinders (H.4).
05000368/FIN-2017002-02	2017Q2	Arkansas Nuclear	Failure to Install Set Screw Leads to Breaker Failure	Green . The inspectors documented a Green self -revealing finding and associated non- cited violation of Unit 2 Technical Specification 6.4.1.a, for failure to properly pre-plan and perform maintenance on the Unit 2 containment spray pump B breaker in accordance with written procedures. Specifically, the licensee failed to install a cam shaft set screw during the breakers last overhaul. The cam eventually became displaced on the shaft, and the breaker failed to close. To correct the issue, the licensee replaced the breaker and installed a cam shaft set screw in the failed breaker. The licensee also inspected all other similar breakers to verify the cams were properly secured. The licensee entered the issue in to their corrective action program as Condition Report CR -ANO -2-2017- 03168. The failure to install a cam shaft set screw during the overhaul of the Unit 2 containment spray pump B breaker is a performance deficiency. The performance deficiency is more than minor because it was associated with the equipment performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the failure of a Unit 2 containment spray pump breaker. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was not a design or qualification deficiency; did not represent a loss of system; did not result in the actual loss of function of a train of technical specification equipment for greater than its allowed outage time; and did not screen as potentially risk significant due to seismic, flooding, or severe weather events. The inspectors determined this finding did not have a cross -cutting aspect because the most significant contributor did not reflect current licensee performance. Specifically, the error occurred during the breakers last overhaul, which occurred in 2011
05000397/FIN-2017002-01	2017Q2	Columbia	Mechanism Operated Cell Switch Failure	Green . The inspectors reviewed a self -revealed finding for the licensees failure to follow plant Procedure SWP -CA P-01, Corrective Action Program, that ensures corrective actions are timely. As a corrective action for failures associated with mechanism operated cell switches for nonsafety 4160 VAC circuit breakers in 2013 and 2015, the licensee assigned modifications to the mechanism operated cell switches but failed to implement t hem in a timely manner. Consequently, on July 20, 2016, circuit breaker E -CB -S/3 mechanism operated cell switches failed to change state resulting in a loss of a main feed pump and an unplanned runback to 70 percent reactor power. As corrective action, the licensee declared the startup transformer inoperable, modified the mechanism operated cell assembly for circuit breaker E -CB -S/3 to remove one switch, and performed post -maintenance testing. The licensee also initiated Action Request 352504 to perform an apparent cause review and address long -term corrective actions. The failure to follow plant Procedure SWP -CAP -01, Corrective Action Program, that ensures corrective actions are timely was a performance deficiency. The performance deficiency was more than minor because it affected the equipment performance attribute of the Initiating Event Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the loss of major loads on E -SM -3 upset plant stability by causing a loss of feed and reactor runback transient. The inspector performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, the licensee remained at power and maintained diverse feed and condensate pumps. This finding had a cross -cutting aspect in the area of human performance, consistent process, in that the licensee failed to use a systematic approach to make decisions including incorporating risk insights. Specifically, circuit breaker E -CB -S/3 is utilized at least monthly 3 for emergency diesel generator surveillance testing and a failure could render the startup transformer inoperable. The mechanism operated cell assembly modification, recommended in 2013 and assigned for action in 2015, was not planned or scheduled as a work order at the time of the failure in 2016 (H.13).
05000397/FIN-2017002-02	2017Q2	Columbia	Failure to Conduct Adequate Surveys of Spent Filters Moved from the Spent Fuel Pool	Green . The inspectors reviewed a self -revealed, non- cited violation of 10 CFR 20.1501 resulting from the licensee's failure to conduct radiation surveys necessary to establish appropriate controls to support movement of spent filters from the spent fuel pool to a shipping cask. This issue was entered into the licensee's corrective action program as Action Requests 356390 and 358265. The licensees failure to perform surveys necessary to establish appropriate controls to support the movement of filters from the spent fuel pool to a shipping cask was a performance deficiency. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process and adversely affected the associated cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation. Specifically, the inadequate radiation surveys resulted in inadequate controls being implemented causing unplanned and unintended personnel dos e. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding was of very low safety significance (Green), because it did not involve: (1) ALARA planning and controls; (2) an overexposure; (3) a substantial potential for overexposure; or (4) an impaired ability to assess dose. The finding had a cross- cutting aspect in the area of human performance, associated with work management, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensees organization and work processes failed to include the identification and management of radiological risk commensurate with the spent fuel pool filter project and the need for strict coordination with different groups or job activities (H.5).
05000397/FIN-2017002-03	2017Q2	Columbia	Inadequate Corrective Actions Causes Failure of HPCS Room Normal Supply Fan	Green . The inspectors reviewed a self -revealed, non- cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to promptly identify and correct a condition adverse to quality. Specifically, since 2012, the licensee failed t o implement prompt corrective actions to correct an adverse condition related to the use of a contactor coil for a motor starter in the high pressure core spray room normal supply fan. As an immediate corrective action, the licensee replaced the contactor for the high pressure core spray room normal supply fan. The licensee entered this issue into the corrective action program as Action Request 360595. The failure to correct an adverse condition related to the use of a contactor coil for a motor starter in the HPCS room normal supply fan, though the licensee had an opportunity and plan to do so, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected 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. Specifically, the licensees failure to correct the use of a contactor coil for a motor starter in the high pressure core spray room normal supply fan resulted in an inoperable fan, high pressure core spray bus 4160 VAC switchgear, and high pressure core spray pump during the January 25, 2017, event when smoke was observed from the motor control center. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety -significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined that this finding did not have a cross -cutting aspect as the decision to not replace the contactor occ urred in 2014 and was not reflective of current performance.
05000397/FIN-2017002-04	2017Q2	Columbia	Licensee-Identified Violation	Title 10 CFR 50.55a(g)4, Inservice Inspection Standards Requirement For Operating Plants , requires , in part, that thro ughout the service life of a boiling water -cooled nuclear power facility, components that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements set forth in Section XI of the ASME Code. The ASME Code, Section XI, Article IWA -2610, requires that all welds and components subject to a surface or volumetric examination be included in the licensees inservice inspection program. This includes identifying each system support that is subject to Section XI requirements. Contrary to the above, prior to March 9, 2017, the licensee did not apply the applicable inservice inspection requirements to all system pressure boundaries within ASME Code Class 1, 2, and 3 boundaries. Specifically, the licensee failed to include the control rod d rive housing welds, as well as portions of the residual heat removal and high pressure core spray systems in their inservice inspection program. The licensee entered this issue into their corrective action program as AR 00343761 and reasonably determined the affected components and system remained operable. The licensee restored compliance by entering the components and systems into the ASME Section XI program. The finding was of very low safety significance (Green) because the finding did not represent an actual loss of safety function of a system or train, and did not result in the loss of a single train for greater than technical specification allowed outage time.
05000397/FIN-2017002-05	2017Q2	Columbia	Licensee-Identified Violation	Title 10 CFR 50.55a(g)(5 )(i) , ISI Program Update: Applicable ISI Code Editions and Addenda, requires , in part, that the inservice inspection program for a boiling water - cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)(4) of this section. Paragraph (g)4 (ii ), Applicable ISI Code: Successive 120- Month Intervals, requires, in part, that inservice examination of components and system pressure tests conducted during successive 120- month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section, 12 months before the start of the 120- month inspection interval. Contrary to these requirements, the licensee failed to issue the inspection plan for the fourth 10- year inservice inspection interval in a timely manner . Specifically, the licensee failed to issue the inservice inspection plan until January 27, 2016, even though the third 10- year inservice inspection interval had ended on December 13, 2015. A relief request to allow emergent repairs to be completed under the third 10 -year inservice inspection plan was requested by the licensee on December 16, 2015, and was approved by the NRC ; however, no repairs needed to be completed. The finding was of very low safety significance (Green) because the finding did not represent an actual loss of safety function of a system or train and did not result in the loss of a single train for greater than technical specification 31 allowed outage time. This issue was entered into the licensees corrective action program as A R 00341506.
05000397/FIN-2017002-06	2017Q2	Columbia	Licensee-Identified Violation	On October 13, 2016, several ARMs unexpectedly alarmed when six filters were simultaneously lifted from the SFP to be placed into a radioactive waste liner . The radiation work permit (RWP) governing performance of the job, RWP 3003788, Revision 00, dated September 7, 2016, had the following , Hold Point , requirements in the event that unexpected radiological conditions occurred during the movement of spent filters: Stop work immediately and notify RP personnel if an unanticipated ARM alarms. If a reading greater than 10 rem/hour contact or 800 millirem/hour at 30 centimeters was detected, but not expected, place the filter back into the SFP . The six filters that had been raised from the SFP had radiation levels as high as 14,000 rem/hour on contact and over 300 rem/hour at almost 30 centimeters. However, the filters were placed in the liner rather than back into the SFP , as specified in the RWP and instructed by RP staff during the evolution. Technical Specification 5.4.1.a requires, in part, that procedures be written, implemented, and established for those areas recommended in Regulatory Guide 1.33, Appendix A, Revision 2, 1978. Section 7(e) of Appendix A recommends written procedures for RWP systems to control access to radioactive materials and limit personnel exposure. Radiation Work Permit 3003788 stated, in part, in the event of unexpected radiological conditions during movement of spent filters, stop work immediately if an unanticipated area radiation monitor alarms, and if a reading greater than 10 rem/hour contact was detected but not expected, place the filter back into the SFP. Contrary to the above, on October 13, 2016 , the licensee f ailed to stop work immediately when several area radiation monitors unexpectedly alarmed and failed to place the filters back into the SFP when readings greater than 10 rem/hour contact were detected but not expected . Subsequently, 16 workers received an addition al 63.5 millirem when the instructions of the RWP and RP staff were not followed. The finding was of very low safety significance (Green) because it did not involve: (1) as-low- as-reasonably achievable (ALARA) planning and controls ; (2) a radiological overexposure; (3) a substantial potential for an exposure; or (4) a compromised ability to assess the dose. This issue was entered into the licensees corrective action program as ARs 356390 and 358265.
05000498/FIN-2017007-01	2017Q1	South Texas	Failure to Provide 8-HOUR Emergency Lighting in All Areas Where Operators Perform Manual Actions Required During an Alternative Shutdown	Green. The team identified a non-cited violation of License Condition 2.E for the failure to provide 8-hour emergency lighting in all areas where operators perform manual actions required during an alternative shutdown. As a compensatory measure, the licensee added flashlights to the procedure box in the essential cooling water intake structure. The team noted that operators were also required to carry a flashlight while on shift. The licensee entered this issue into their corrective action program as Condition Report 17-1741. The failure to provide 8-hour emergency lighting in all areas where operators perform manual actions required during an alternativ e shutdown was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) a ttribute of the Mitigating Systems Cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to provide 8-hour emergency lighting could adversely affect the ability of operators to perform the manual actions required for an alternative shutdown. The team determined this finding affected the Mitigating Systems Cornerstone. The team evaluated this finding using Ins pection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, dated September 20, 2013, because it affected the ability to reach and maintain safe shutdown conditions in case of a fire. The team determined this finding was of very low safety significance (Green) in Task 1.3.1 because it had a low degradation rating. The finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago. Specifically, the team determined that the per formance deficiency existed since original construction.
05000528/FIN-2017007-01	2017Q1	Palo Verde	Failure to Analyze Shutdown Cooling and Feedwater Lines for High-Energy Line Break Pipe Whip Effects	Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, from August 11, 1982, to March 3, 2017, the licensee did not analyze dynamic pipe whip effects of a main feedwater line for a high-energy line break of a shutdown cooling line. In response to this issue, the licensee performed immediate and prompt operability evaluations and determined that the piping systems remained operable and could withstand the effects of a high-energy line break. This finding was entered into the licensees corrective action program as Condition Report CR-17-02815. The team determined that the failure to perform an adequate analysis for shutdown cooling and feedwater lines for high-energy line break pipe whip effects was a performance deficiency. This finding was more-than-minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to analyze the main feedwater piping for high-energy line break effects called the operability of the piping system into question. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as hav ing very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross- cutting aspect because the most significant contributor to the performance deficiency did 3 not reflect current licensee performance. Specifically, the licensee performed the calculation in 1982 and revised it in 1991; therefore, the performance deficiency occurred outside of the nominal three-year period for present performance.
05000458/FIN-2017001-01	2017Q1	River Bend	Failure to Follow Station Guidance on Control of Scaffolding	Green . The inspectors identified a non-cited violation of Technical Specification 5.4, Procedures, for the licensees failure to follow station maintenance procedures related to the control of scaffolding in the reactor building. Specifically, the licensee installed scaffolding less than two inches from safety -related containment unit cooler HVR -UC1 B without completing an engineering evaluation. The licensee entered this issue into their corrective action program as Condition Report CR- RBS -2016- 07963 . Corrective actions included removing the scaffolding. The licensees installation of scaffolding within two inches of a safety -related containment unit cooler , without completing an engineering evaluation, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, containment unit cooler HVR- UC1B was rendered inoperable by the incorrectly installed scaffolding and remained inoperable until the scaffolding was removed. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance ( Green) because the finding did not represent an actual loss of function of one or more trains of safety-related equipment for greater than its technical specification allowed outage time. This finding has a cross -cutting aspect in the area of human performance, avoid complacency , because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risks, even while expecting successful outcomes . Specifically, the station failed to implement appropriate error reduction tools when it did not perform and document Procedure EN -MA -133, Control of Scaffolding, Attachments 9.5 and 9.6 , which could have prevented the scaffolding construction error (H. 12).
05000458/FIN-2017001-02	2017Q1	River Bend	Failure to Properly Pre - Plan and Perform Maintenance on the Control Building Chilled Water System	Green . The inspectors identified a non- cited violation of Technical Specification 5.4, Procedures, for the licensees failure to properly pre-plan and perform maintenance on safety -related components in accordance with documented instructions appropriate to the circumstances. Specifically, the licensee used work order instructions that did not contain sufficient detail for the reassembly of SWP -PVY32C, a safety -related valve in the control building ventilation system . As a result, SWP -PVY32C developed a refrigerant leak, and on November 17, 2015 , the valve failed. This in turn caused the control building ventilation system to fail , and the high pressure core spray system was consequently declared inoperable. The licensee entered this condition into their corrective action program as Condition Report CR- RBS -2017- 02364. Corrective actions included incorporating the torque values into the model work order instructions for future maintenance and reassembly . The failure to properly pre-plan and perform maintenance on safety -related components in accordance with documented instructions was a performance deficiency. The performance deficiency was more than minor , and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the control building ventilation system failed, it impact ed the operability of the high pressure core spray system. The inspectors screened the finding in accordance with NRC Inspection Manual Chapter 0609, Significance Determination Process. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, Exhibit 2 Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because it did not affect the design or qualification of a mitigating structure, system, or component (and the structure, system, or component maintained its operability), it did not represent a loss of safety function, it did not represent an actual loss of function of at least a single train for greater than its technical specification outage time, and it did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant in accordance with the licensees Maintenance Rule program for greater than 24 hours. This finding has a cross- cutting aspect in the area of human performance, challenge the unknown, because individuals did not stop when faced with uncertain conditions. Specifically, workers proceeded with assembling the valve when the torque values or torqueing sequence were not specified (H.11)
05000458/FIN-2017001-03	2017Q1	River Bend	Failure to Enter Applicable Technical Specification Action Statements When Control Building Chillers Were O ut of Service	Green . The inspectors identified a non- cited violation of Technical Specifications 3.8.4, DC Sources - Operating, 3.8.7, Inverters Operating, and 3.8.9, Distribution Systems Operating, for the licensees failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours. Specifically, electrical power systems required by the above limiting condition s for operation were inoperable due to the associated division of the control building chilled water system chillers being out of service and therefore unavailable to provide the technical specification support function of attendant cooling that is needed for the associated electrical systems to perform their specified safet y functions. As a result of this deficiency, the station reduced the reliability and availability of systems cooled by control building chilled water system chillers by allowing configurations that did not conform to the single failure criterion. The lic ensee entered this issue into their corrective action program as Condition Report CR- RBS -2015 -02525 . Corrective actions included entering the appropriate limiting conditions for operation of affected safety -related systems when the non -safety related support system were non -functional. 4 The failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours wa s a performance deficiency . Specifically, electrical power systems required by the above limiting condition s for operation were inoperable due to the associated division of the control building chilled water system chillers being out of service and therefore unavailable to provide the technical specification support function of attendant cooling that is needed for the associated electrical systems to perform their specified safety functions. The performance deficiency was more than minor, and therefore a finding, because it wa s associated with the configuration control attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that res pond to initiating events to prevent undesirable consequences. As a result of this deficiency, the station reduced the reliability and availability of systems cooled by control building chilled water system chillers by allowing configurations that did not conform to the single failure criterion. The inspectors performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to require a detailed risk evaluation because it represented a loss of system and/or function. A senior reactor analyst performed a det ailed risk evaluation for a previously identified performance deficiency associated with the licensees failure to account for a loss of all control building chilled water system cooling scenario, either quantitatively or qualitatively, which resulted in uncompensated impairment to all systems associated with the main control room (Agencywide Documents Access and Management System (ADAMS) Accession N o. ML16132A144). This previously performed detailed risk evaluation bounds the risk associated with the finding dispositioned in this write- up: the failure to either restore inoperable electrical power subsystems, inverters, and distribution subsystems to operable status within the applicable completion times, or be in Mode 3 in 12 hours and Mode 4 in 36 hours. Therefore, the finding was determined to be of very low safety significance (Green). No cross -cutting aspect was assigned as the performance deficiency is not indicative of current licensee performance
05000275/FIN-2017001-01	2017Q1	Diablo Canyon	Licensee-Identified Violation	Technical Specification 3.3.3 Post Accident Monitoring (PAM) Instrumentation, requires at least two channels of both wide range) hot leg reactor coolant system (RCS) temperature and wide range cold leg RCS temperature RTDs to be in service. If this action is not met, TS 3.3.3 requires the restoration of all but one channel to operable status within 7 days. If this action cannot be met, TS 3.3.3 requires the plant to be shutdown to Mode 3 within 6 hours and Mode 4 within 12 hours. Contrary to the above, in October 2015 during performance of an apparent cause evaluation investigating failing wide range RCS RTDs, PG&E discovered that the plant had been operating with all channels of hot leg and cold leg wide range RCS temperature monitoring inoperable for greater than the allowed TS 3.3.3 outage time without complying with the requirement to shut down the plant. Pacific Gas and Electric identified an incorrect insulation configuration, installed in 2010, on the thermal extension piping that houses the wires for the wide range RCS RTDs as the direct cause of the failures. The insulation, as installed, trapped heat inside of the thermal extension piping and overheated the associated wires. Pacific Gas and Electric determined that eight wide range RCS RTDs had ether failed or operated outside of the environmental qualification temperature range, however the required channels remained functional. Pacific Gas and Electric determined the cause of the incorrect installation to be insufficient guidance in the associated work package instructions. The inspectors determined that PG&Es failure to develop adequate work guidance to properly install wide range RCS RTD insulation was a performance deficiency that was within PG&Es ability to foresee and correct. Pacific Gas and Electric entered this issue into their corrective action program (CAP) as Notification 50808493, replaced the eight wide range RTDs, restored the insulation per design requirements, revised the drawings for Unit 1 wide range RTDs to provide adequate level of detail, and revised the work order to include the correct drawing and level of details for proper installation of all wide range RTDs. This performance deficiency is considered more than minor, and considered a finding, because it is associated with the Mitigating Systems cornerstone attribute of equipment performance and 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 in accordance with IMC 0609, Appendix A, Determining the Significance of Reactor Inspection Findings for At-Power Situations. The inspectors determined that the finding was of very low safety significance (Green) because the deficiency did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time.
05000529/FIN-2017001-01	2017Q1	Palo Verde	Failure to establish station procedure instructions for denial work authorizations	The inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, Procedures, for the failure to establish procedure instructions for work authorization denials or deferrals. Specifically, this led to a 60 day extended unavailability of the diverse auxiliary feedwater actuation system when corrective maintenance was inappropriately deferred by the operations department. Failure to provide adequate procedural guidance in the event of a denied work authorization, a circumstance anticipated to occur, is a performance deficiency. The performance deficiency is more than minor, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective to ensure the availability and reliability of equipment that responds to an initiating event. Specifically, because the corrective maintenance was not performed in a timely manner, both trains of the diverse auxiliary feedwater actuation system remained in bypass for an additional 60 days whereby the system was not capable of performing its required safety function. The inspectors evaluated the significance of the finding using Inspection Manual Chapter 0609, Appendix A, Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions, Section A, Question 2, which required a detailed risk evaluation because the finding involved a loss of system safety function. A Region IV senior reactor analyst performed a detailed risk assessment of the finding and determined that the finding was of very low safety significance (Green). The inspectors determined that the finding had a cross-cutting aspect in the human performance area of Work Management. The work process includes the identification and management of risk commensurate to the work and the need for coordination with different groups or job activities. Specifically, the Unit Operations Managers decision to deny the work authorization was based on conservative but faulty assumptions, and if other work groups with greater specific technical knowledge had been involved, the corrective maintenance likely would have proceeded (H.5)
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).
05000529/FIN-2016004-01	2016Q4	Palo Verde	Inadequate monitoring of MSIV nitrogen pre-charge pressure	The inspectors reviewed a self-revealing non-cited violation of Technical Specification 3.7.2 for exceeding the Condition A completion time for an inoperable main steam isolation valve (MSIV) single actuator train and not immediately declaring the affected main steam isolation valve inoperable in accordance with Condition E. Specifically, the Unit 2 main steam isolation valve 171 actuator A was inoperable from July 30, 2016, to August 9, 2016, when a known nitrogen leak was not adequately monitored. The licensees inadequate monitoring allowed the nitrogen pre-charge pressure in the actuator to decrease to below the minimum acceptable limit for operability. The licensee restored the pre-charge pressure and entered this issue into their corrective action program as Condition Report 16-12740. The failure to perform adequate monitoring for a degraded condition as required by procedure 40DP-9OP26, Operations Condition Reporting Process and Operability Determination/Functional Assessment, was a performance deficiency. The performance deficiency was more-than-minor and therefore a finding because it affected the equipment performance attribute of the Mitigating Systems Cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically the failure to adequately monitor a known nitrogen leak resulted in depressurizing one of two hydraulic accumulators thereby reducing the reliability of the system to initiate a fast closure of MSIV 171 upon receipt of a main steam isolation signal. The inspectors performed the initial significance determination using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, Issue Date: June 9, 2012. The finding required a detailed risk evaluation since it represented a loss of function for a single train for greater than the Technical Specification allowed outage time. A Region IV senior reactor analyst determined the finding was of very low safety significance (Green) since the MSIV remained capable of performing its safety function with the alternate actuator. The finding has a cross-cutting aspect in the area of human performance associated with the teamwork component. Specifically, the licensee failed to coordinate activities across organizational boundaries in that the operations personnel did not obtain engineering input to ensure that additional monitoring requirements for the nitrogen pre-charge leak were adequate to verify continued MSIV 171 operability (H.4).
05000416/FIN-2016004-01	2016Q4	Grand Gulf	Failure to Incorporate Design Requirements for Switchgear Room Cooling	Green. The inspector identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, involving the failure to implement appropriate design control measures associated with a safety-related service water flow calculation. Specifically, several unverified and potentially nonconservative inputs were identified associated with Calculation MC-QIP41-97020, Revision 11, Determination of Minimum Allowable SSW Flows (LOCA Lineup) to Safety Related Heat Exchangers, used to analyze minimum service water flow to the vital switchgear room coolers. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2016-07597, initiated action to update Calculation MC-QIP41-97020, and initiated actions to analyze the ability of vital switchgear room cooling to meet its specified safety function. This performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone, and it adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not assure that the vital switchgear ventilation system was capable of maintaining the rooms temperature below design requirements under all conditions. The NRC performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding had very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding had a cross-cutting aspect in the documentation aspect of the human performance cross-cutting area because the licensee failed to maintain complete, accurate, and up-to-date documentation of the design temperature limits for safety-related equipment. Specifically, the licensee failed to document and evaluate a change to temperature limits related to switchgear cooling to ensure that its use as a design parameter was consistent with original design specifications of the equipment (H.7).
05000416/FIN-2016004-02	2016Q4	Grand Gulf	Failure to Use Procedures and Engineering Controls to Maintain Occupational Doses ALARA	Green. The inspectors identified a non-cited violation of 10 CFR 20.1101(b) for the licensees failure to implement radiation exposure reduction procedures and engineering controls to minimize unplanned and unintended radiation dose to workers and to maintain occupational doses as low as is reasonably achievable (ALARA). Several radiological work permits exceeded initial dose estimates with minimal or no actions taken to evaluate the basis for the dose overrides and to develop mitigating strategies. The primary contributor to the unplanned exposures was elevated dose rates from increased cobalt-60 activity associated with a failure to properly plan and execute spent fuel pool and reactor cavity cleanup operations. In addition, the licensee failed to observe radiological work permit hold points, to initiate ALARA Management Committee meetings, and to perform radiological assessments of radiological work permit dose estimates as procedurally required. As immediate corrective actions, the licensee reviewed the work activity, documented lessons learned, and generated Condition Reports CR-GGN-2016-03151 and CR-GGN-2016-08543 to address these programmatic weaknesses for future outages. The failure to implement procedures and engineering controls to minimize unplanned and unintended radiation dose and to maintain occupational doses as low as is reasonably achievable was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (ALARA planning) and adversely affected the cornerstone objective to ensure the adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, inadequate ALARA planning and radiological controls resulted in unplanned, unintended dose for a number of work activities in which the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined this finding to be of very low safety significance (Green) because the finding involved ALARA planning and controls, and because the licensees latest 3-year rolling average did not exceed 240 person-rem per unit for boiling water reactors. The finding had a cross-cutting aspect in the area of problem identification and resolution, associated with operating experience, in that, the licensees organization failed to systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, the licensee failed to implement and incorporate relevant internal operating experience from Refueling Outage 18, which was of similar radiological circumstances, to mitigate the effects of cobalt-60 activity in the reactor cavity and unplanned spent fuel pool cleanup outages (P.5).
05000483/FIN-2016002-03	2016Q2	Callaway	Failure to Adequately Evaluate Operability for a Degraded Condition	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an adequate operability assessment when a degraded or nonconforming condition was identified. Specifically, after the licensee identified that a severe water hammer transient would occur following a loss of off-site power, the licensee generated an operability evaluation that relied on judgement and inaccurate information which failed to establish a reasonable expectation of operability. Following questions from inspectors the licensee determined that this judgement was not correct and performed a new evaluation to ensure operability of the essential service water system. The licensee entered this issue into their corrective action program as Callaway Action Request 201605488. The licensees failure to properly assess and document the basis for operability when a severe water hammer occurred in the essential service water system was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, severe water hammer transients in the essential service water system due to a loss of off-site power, result in a condition where structures, systems, and components necessary to mitigate the effects of accidents may not have functioned as required. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic event, and (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensees use of unsupported judgement and incorrect data resulted in an evaluation that failed to demonstrate a reasonable expectation of operability (H.14).
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-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-2016002-04	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-2016002-03	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-2016002-02	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.
05000445/FIN-2016002-02	2016Q2	Comanche Peak	Failure to Determine Dose Rates Prior to Allowing Entry into a High Radiation Area	The inspectors reviewed a self-revealed non-cited violation of Technical Specification 5.7.1.e associated with the licensee allowing a worker access into the 2-077-B penetration valve room, a high radiation area, without an adequate knowledge of the radiological conditions. Specifically, the licensee briefed the worker on the conditions with outdated radiation survey information even though the 2-077-B penetration valve room was subject to changing radiological conditions. As a result, an individual entered areas with general area dose rates of 210 mrem per hour rather than the briefed dose rates of less than 50 mrem per hour. This issue was entered into the licensees corrective action program as Condition Report CR-2015-010211. Corrective actions included performing follow-up radiation surveys and implementing improvements to the high radiation area access control program. The inspectors determined that allowing a worker access into a high radiation without an adequate knowledge of the radiological conditions was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it affected the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, entry into a high radiation area without adequate knowledge of the radiological conditions placed the individual at risk for unnecessary exposure. The finding was assessed using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, issued August 19, 2008, and was determined to be of very low safety significance (Green) because the performance deficiency was not an ALARA planning issue, there was not an overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The finding has a human performance cross-cutting aspect associated with work management, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority (H.5).
05000528/FIN-2016002-03	2016Q2	Palo Verde	Inadequate Engineering and Radiological Controls Resulting in a Unit 1 Containment Building Airborne Radioactivity Event with Unplanned Intakes	The inspectors identified a non-cited violation of 10 CFR 20.1701 due to the licensees failure to implement adequate processes and engineering controls necessary to reduce airborne radioactivity and prevent internal dose to workers in Unit 1. On April 20, 2016, inspectors identified that procedures and instructions for monitoring high efficiency particulate air ventilation filter unit to prevent worker exposures to radiation and airborne radioactivity were being inadequately implemented. On April 21, 2016, the licensees inadequate engineering and radiological controls during a high efficiency particulate air operations caused an airborne radioactivity event in containment, resulting in the evacuation of 41 potentially contaminated workers of whom 8 had measurable intakes of radioactive material. The licensees immediate corrective actions included stopping work in the Unit 1 containment, evacuating workers in containment, assessing workers for external and internal contamination, and investigating the cause and source of the contamination event. This matter was placed in the licensees corrective action program as Condition Reports16-06499 and 16-06578 and the licensee initiated a root cause investigation. The inspectors determined that the failures to implement adequate engineering and radiological controls to reduce airborne radioactivity during a high efficiency particulate air unit operations in accordance with 10 CFR 20.1701 and radiation protection procedures were performance deficiencies. The performance deficiencies were more than minor because they were associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. This was evident by the Unit 1 containment airborne radioactivity event on April 21, 2016, that resulted in at least eight workers with unplanned intakes. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the finding had very low safety significance (Green) because: (1) it was not an as low as is reasonably achievable planning and controls finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The inspectors concluded that the finding has a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee leaders failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, procedures and radiation exposure permits failed to have adequate instructions for ensuring a high efficiency particulate air filter loading and dose rates were monitored to prevent overloading, and safe handling of loaded a high efficiency particulate air filters (H.1).
05000483/FIN-2016002-01	2016Q2	Callaway	Failure to Account for Water Hammer Stresses in Essential Service Water System Calculations	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to account for the essential service water pipe stresses caused by pressure fluctuations of the known column closure water hammer phenomenon. The licensee failed to properly account for essential service water piping membrane stress and impact loads as required by the 1974 ASME Code, Section III, paragraphs ND-3112.4 and ND-3111. Specifically, the licensees design calculations for the essential service water system did not account for the pressure fluctuations caused by a known column closure water hammer phenomenon that occurs during a loss of off-site power or load sequencer testing. The licensee completed a prompt operability determination assuring the system was operable under the current conditions and was completing engineering evaluations of the data collected to demonstrate the operability of the system under design conditions. The licensee entered this issued into the corrective action program as Callaway Action Requests 201603472 and 201603819. The inspectors determined that the licensees failure to account for the pressure fluctuations caused by a known column closure water hammer phenomenon in the design calculations for the essential service water system was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of conservative bias in the human performance area because the licensee failed to demonstrate that a proposed action was safe in order to proceed, rather than unsafe in order to stop. Specifically, when the licensee recognized that the column separation water hammer phenomenon was occurring in the essential service water system, they only applied the forces to the containment coolers, not the entire system (H.14).
05000483/FIN-2016002-02	2016Q2	Callaway	Failure to Meet Applicable ASME Code Requirements for Repairs to Components in the Essential Service Water System	The inspectors identified a non-cited violation of 10 CFR 50.55a, Codes and Standards, for the licensees failure to repair various ASME Code Class 3 components in accordance with ASME Code, Section XI requirements. Specifically, the licensee did not follow the applicable ASME Code requirements when making repairs to various components in the ASME Code Class 3 essential service water system. The licensee reasonably determined the essential service water system remained operable, and completed the necessary repairs and testing to restore compliance with ASME Code. The licensee entered this issue into their corrective action program as Callaway Action Requests 201603640 and 201604282. The inspectors determined that the programmatic failure to repair various ASME Code Class 3 components in the essential service water system in accordance with ASME Code was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant for greater than 24 hours in accordance with the licensees maintenance rule program. Specifically, the licensee performed a historical system health review and reasonably determined the essential service water system remained operable because periodic system walkdowns by the system owner and shiftly rounds by operations had not identified significant system leaks, and the appropriate repairs and testing were completed on the affected components. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee failed to ensure training of the personnel was adequate to recognize that the repair of the leaks constituted repairs in accordance with ASME Code, Section XI and thus failed to include the necessary ASME testing requirements in the work performance packages to ensure adequate performance of an activity which affected testing of a safety-related modification/repair to risk-significant systems, and thereby ensure nuclear safety (H.9).
05000528/FIN-2016002-02	2016Q2	Palo Verde	Failure to Implement High Radiation Area Controls in an Area with a Dose Rates Greater Than 1 rem per Hour	The inspectors reviewed a Green, self-revealing, non-cited violation of Technical Specification 5.7.2, which was caused by the licensees failure to control a high radiation area with radiation levels greater than 1 rem per hour in the Unit 1 containment. A radiation protection technician received an unexpected dose rate alarm while conducting surveys on piping in the 87-foot elevation of the 2B reactor coolant pump bay area near a high efficiency particulate air unit in containment. Licensee personnel corrected the error by guarding the area, posting the area, and changing the pre-filters in the adjacent portable a high efficiency particulate air units to reduce the dose rates. This issue was entered into the licensees corrective action program as Condition Reports 16-06515 and 16-07479. The inspectors determined that the failure to identify a locked high radiation area through timely surveys and adequate a high efficiency particulate air maintenance procedures that could have revealed changing radiological conditions was a performance deficiency. The performance deficiency was more than minor because it was associated with the Occupational Radiation Safety Cornerstone attribute of program and process (exposure control) and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation because licensee personnel did not implement barriers intended to prevent workers from receiving unexpected dose. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation had very low safety significance (Green) because: (1) it was not an as low as is reasonably achievable finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This finding has a cross-cutting aspect in the human performance area, associated with the resources component, because the licensee leaders failed to ensure that personnel, equipment, and procedures were available and adequate to support nuclear safety. Specifically, the licensee failed to ensure that procedures were adequate to ensure radiation levels around portable high efficiency particulate air units were monitored to evaluate changing radiological conditions in a timely manner such that hazards were appropriately controlled (H.1).
05000483/FIN-2016002-05	2016Q2	Callaway	Failure to Follow Plant Foreign Material Exclusion Procedure	The inspectors reviewed a self-revealed finding for the licensees failure to follow the plant procedure for foreign material exclusion. Specifically, after finding foreign material (broken cable ties) within the main generator excitation transformer, established as a foreign material exclusion Level 2 area, the licensee failed to determine the reason for the foreign material and enter the issue into the corrective action program for resolution as required by Procedure APA-ZZ-00801, Foreign Material Exclusion, Revision 32. The licensees failure to follow the plant procedure for foreign material exclusion was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, after identifying several broken cable ties on the floor inside a foreign material exclusion Level 2 area the licensee did not determine the reason for the foreign material nor enter the condition into the corrective action program as required by Procedure APA-ZZ-00801. Because the licensee failed to understand what caused the cable tie degradation, a subsequent cable tie failure resulted in a plant trip. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the finding was determined to be of very low safety significance because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect of training in the human performance area because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, several groups within the licensees organization were unaware the excitation transformer cabinet was classified as a foreign material exclusion Level 2 area nor the requirements if foreign material is found within the foreign material exclusion area (H.9).
05000446/FIN-2016002-01	2016Q2	Comanche Peak	Failure to Correct Conditions Adverse to Quality	The inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to correct a condition adverse to quality in safety-related equipment. Specifically, following an in-service testing failure of auxiliary feedwater check valve 2FW-091 in November 2012, the licensee performed an operability evaluation of the auxiliary feedwater system. However, the inspectors identified that the licensee failed to take corrective action to address the condition adverse to quality that resulted in the valve failing to seat properly. Consequently, the same valve failed a subsequent inservice test in November 2015. Following discovery of this issue, the licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The licensee entered this issue into corrective action program as CR-2015-10961. The licensees failure to correct a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, 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, the licensee failed to correct auxiliary feedwater check valve 2FW-0191 failure to seat in November 2012 resulting in an additional failure in November 2015. Using Inspection Manual Chapter (IMC) 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time, and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The finding has a problem identification and resolution cross-cutting aspect associated with evaluation, in that, the licensee failed to thoroughly evaluate issues to ensure that resolutions address extent of conditions. Specifically, the licensee failed to appropriately classify the issue of the check valve not seating and recognize this as a degraded condition (P.2).
05000528/FIN-2016002-01	2016Q2	Palo Verde	Leakage From Reactor Coolant Pump 2B Discharge Pipe Instrument Nozzle	The inspectors identified an unresolved item for pressure boundary leakage from reactor coolant pump 2B discharge pipe instrument nozzle. On April 10, 2016, during the Unit 1 Refueling Outage 19, the licensee discovered reactor coolant system pressure boundary leakage at instrument nozzle 1JRCETW0121Y on the 2B reactor coolant pump discharge piping. The leakage was discovered during a planned visual inspection of Unit 1 hot and cold leg nozzles. The leak was not detectable by either the reactor coolant system leak rate procedure or the containment radiation monitor trend reviews while the unit was operating. Additionally, the leak had not been visually detected during the previous refueling outage. The leakage was consistent with a small leak characterized by moderate boric acid accumulation at the leakage site. The licensee determined that the cause of the leakage was primary water stress corrosion cracking of the Alloy 600 instrument nozzle. The licensee corrected the leakage using a mechanical nozzle seal assembly repair method utilizing ASME Code Case N-733, Mitigation of Flaws in NPS 2 (DN 50) and Smaller Nozzles and Nozzle Partial Penetration Welds in Vessels and Piping by Use of a Mechanical Connection Modification, Section XI, Division 1. The evaluation of the 2B cold leg RTD nozzle leakage is being evaluated by the licensee as part of Palo Verde Action Request 15-01640-012. The inspectors reviewed the circumstances surrounding the discovery of the leak and observed portions of the repair activity during the refueling outage. Once the licensee completes their evaluation, the inspectors will review and complete an inspection to determine if a performance deficiency exists as a result of the nozzle failure.
05000483/FIN-2016002-04	2016Q2	Callaway	Failure to Promptly Correct Conditions Adverse to Quality	During an NRC inspection conducted June 6-30, 2016, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part, that conditions adverse to quality are promptly identified and corrected. Contrary to the above, from November 2010 through June 2016, the licensee failed to promptly correct a condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues which were previously identified by the NRC as non-cited violation 05000483/2010006-01. The failure to resolve these issues resulted in subsequent safety-related equipment failures. This violation is associated with a Green Significance Determination Process finding The inspectors identified a cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to take timely corrective action for a previously identified condition adverse to quality. Specifically, the licensee failed to adequately resolve water hammer and corrosion issues that were previously identified by the NRC as non-cited violation 05000483/2010006-01 and the failure to resolve these issues resulted in subsequent safety-related equipment failures. The licensee performed an operability determination that established a reasonable expectation of operability pending implementation of corrective actions. The licensee entered this issue into their corrective action program as Callaway Action Request 201604440. The licensees failure to take timely and adequate corrective actions to correct a condition adverse to quality was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct water hammer and corrosion issue resulted in the licensee declaring safety-related room coolers and chillers inoperable until an analysis of system operability was completed. This affected their capability to respond to initiating events to prevent undesirable consequences Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, inspectors determined that this finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality, (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-ofservice for longer than their technical specification allowed outage time, and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect of resources in the human performance area because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, by failing to address water hammer and corrosion issues, station management failed to ensure that the essential service water system was available and adequately maintained to respond during a loss of off-site power event (H.1).
05000445/FIN-2016002-03	2016Q2	Comanche Peak	Licensee-Identified Violation	Comanche Peak Unit 1, Operating License NPF-87, Condition 2.G, Fire Protection, requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 78 and as approved in the Safety Evaluation Report and its supplements through Supplement 24. Comanche Peak Unit 2, Operating License NPF-89, Condition 2.G, Fire Protection, requires, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report through Amendment 87 and as approved in the Safety Evaluation Report and its supplements through Supplement 27. The stations approved fire protection program includes Fire Protection Report, Revision 29, Section 5.3.8, Fire Area EO Control Room, includes Deviation 3c-1, Control Room Missile Door, which requires, in part, that since the control room missile door in the west wall is not a three hour rated fire door, the area of the turbine deck within 100 feet of the door is to be void of combustibles. Contrary to the above, on May 5, 2016, the licensee stored combustible materials within 100 feet of the control room missile door in the west wall. Specifically, licensee personnel identified that contractors had stored combustibles within the combustible free zone, and that no compensatory measures had been implemented for the deviation from the Fire Protection Report. The licensee implemented a periodic roving fire watch to compensate for the reduction in fire protection. The violation is more than minor because if left uncorrected, it could lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, the inspector determined that the violation is of very low safety significance (Green) because the finding did not affect the ability of either unit to achieve safe shutdown. The violation was entered into the licensees corrective action program as CR-2016-004167.
05000446/FIN-2016002-04	2016Q2	Comanche Peak	Licensee-Identified Violation	Technical Specification 5.4.1.a states, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Appendix A, Section 9.a., identifies procedures for maintenance as required procedures. Work order 4831032 is a procedure established by the licensee for performing maintenance on diesel generator 2-02. The work order provided instructions for installation of the magnetic speed pickup sensor cable. Contrary to the above, from October 1996 through March 2, 2016, the licensee failed to install the unit 2 diesel generator 2-02 magnetic speed pickup sensor cable in accordance with the approved instructions. Specifically, the speed sensor cable conduit was not fully threaded onto the cable plug. This inadequate installation was present until 2016, when the conduit threaded connection was physically impacted at an undetermined time. The impact caused the conduit connection to break and the conduit to separate from the plug, leaving the cable leads exposed but intact. A licensee technician identified the broken connection during a system walk down on March 2, 2016. The licensee declared the diesel generator inoperable and restored the cable to its design configuration. The licensee analyzed the apparent thread engagement, and determined that, prior to the break in the conduit connection, the cable would have maintained its function in a seismic event, but after the break, the cable function could not be assured. The licensee determined that a failure of the cable would result in the diesel generator exceeding its allowed frequency, but would not result in a diesel generator failure to run. Because the time that the break occurred could not be determined, the diesel generator was assumed to be inoperable at the time of discovery. The violation is more than minor because it affected the configuration control attribute of the Mitigating Systems cornerstone and impacted the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspector determined that the violation is of very low safety significance (Green) because the finding did not represent a loss of system or function, and did not represent a loss of function of a single train for greater than its technical specification allowed outage time. The violation was entered into the licensees corrective action program as CR-2016-001941.

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