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	Quarter	Site	Title	Description
05000275/FIN-2015004-01	2015Q4	Diablo Canyon	Failure to Properly Evaluate for Aggregate Impact of Fire Impairments	The inspectors identified a non-cited violation of Technical Specification 5.4.1.d, Procedures, for the failure to follow approved fire protection program procedures to review the fire impairments list to assess the aggregate impact on the fire protection design and safe shutdown analysis. Specifically, from August 31 to September 2, 2015, the licensee failed to evaluate the aggregate impact of having three fire doors simultaneously blocked open in adjacent Unit 1 vital battery charger rooms. The licensee implemented immediate corrective actions by assigning a continuous fire watch to the area and documented the issue in the corrective action program as Notification 50826793. The failure to follow approved fire protection program procedures to review the fire impairments list to assess the aggregate impact on the fire protection design and safe shutdown analysis was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it was associated with the Initiating Events cornerstone attribute of Protection against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during plant operations. Specifically, the failure to evaluate the aggregate impact of multiple fire system impairments affected the licensee ability to limit the impact of a potential fire. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1Initial Screening and Characterization of Findings. Because the finding involved fire protection, the inspectors transitioned to IMC 0609, Appendix F Fire Protection Significance Determination Process. The inspectors characterized the finding using IMC 0609, Appendix F, Attachment 1, "Fire Protection SDP Phase 1 Worksheet," dated September 20, 2013. The finding screened as very low safety significance (Green), per Attachment 1, Question 1.4.3-A since the fire finding category was determined to be fire confinement, due to the fire doors being propped open, and the combustion loading on both sides of the door was determined to be a duration of 30 minutes as documented in licensee calculation M-824, Controlled Combustion Loading Tracking. In addition, the inspectors determined this finding had a cross-cutting aspect in human performance associated with the teamwork component because the licensees work groups did not properly communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, the work planners did not properly communicate to the fire protection department that all three fire doors would be open at the same time during battery charger load testing. (H.4)
05000275/FIN-2015004-02	2015Q4	Diablo Canyon	Failure to Identify a Cause and Implement Actions to Prevent Recurrence of a Significant Condition Adverse to Quality	The inspectors reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI Corrective Action, for the failure to identify the cause and take corrective action to prevent recurrence of a significant condition adverse to quality impacting both trains of the Unit 1 safety-related residual heat removal (RHR) system. Specifically, the licensee failed to identify a definitive cause and implement corrective actions to prevent recurrent failures of the socket weld for relief valve RHR-1-RV-8708 for both trains of the RHR system. As immediate corrective actions, the licensee installed additional piping supports to mitigate the vibrations at the socket weld and documented this issue in the corrective action program as Notification 50680750. The failure to identify the cause of the RHR vibration-induced problems and to take adequate corrective actions to prevent recurrence of the weld failures was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it could lead to a more significant safety concern. Specifically, no additional supports were installed and no actions were taken to reduce or eliminate the vibrations to prevent recurring weld failures, which could affect the availability of the RHR system. The lack of corrective actions to prevent recurrence could leave RHR components and other components physically connected to the system susceptible to future failures. Using Inspection Manual Chapter 0609, Appendix A, the inspectors determined the issue to have very low safety significance (Green) because the performance deficiency, which affected the mitigating systems cornerstone, did not result in a loss of safety function and did not result in an actual loss of function for greater than the technical specification allowed outage time. The licensee entered this into their corrective action program as Notification 50680750. In addition, this finding has a cross-cutting aspect in the human performance area associated with conservative bias decision making component because individuals failed to use decision making practices that emphasize prudent choices over those that are simply allowable. Specifically, the licensee chose to only install a fatigue resistance weld rather than install additional pipe supports as were in the Unit 2 system (H.14).
05000275/FIN-2015004-03	2015Q4	Diablo Canyon	Failure to Design the Emergency Diesel Generators to operate under Worst Case Environmental Conditions	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for the failure to implement design control measures to verify the adequacy of the Unit 1 emergency diesel generators (EDGs) cooling system design to ensure operation of the EDGs under worst-case environmental conditions. Specifically, since initial licensed operations began in 1984, the licensee failed to ensure the Unit 1 EDGs were designed and built to operate under worst-case high wind and temperature conditions. As a result, sustained high winds from specific directions could have impacted EDG radiator performance resulting in the unavailability of the Unit 1 EDGs. Immediate corrective actions included issuing shift orders to the reactor operators to monitor for specific weather conditions (high air temperature, high wind speed and direction) and provide additional room cooling using established procedures, as necessary. The licensee documented the issue in the corrective action program as Notification 50599190. The failure to implement design control measures to ensure the emergency diesel generators could perform their design basis function was a performance deficiency. The performance deficiency was more than minor, and is therefore a finding, because it was associated with the design control attribute of the mitigating system cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in a condition where sustained high winds from specific directions could have impacted EDG radiator performance resulting in the unavailability of the Unit 1 EDGs. The inspectors evaluated the finding using Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, dated June 19, 2012. The inspectors determined that a detailed risk evaluation by an NRC senior reactor analyst was required since the finding was associated with a loss of EDG function. The regional senior reactor analyst performed a Phase 3 SDP analysis for the finding. The results of analysis established the incremental conditional core damage probability (ICCDP) was 2.74E-07, less than 1 x 10-6, and therefore the analyst determined that the subject finding was of very low safety significance (Green). A cross-cutting aspect was not assigned to the finding since the finding did not represent current licensee performance. The condition existed since original construction of the plant.
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.
05000285/FIN-2010007-01	2010Q2	Fort Calhoun	Failure to Maintain External Flood Procedures	The inspectors identified an apparent violation of Technical Specification 5.8.1.a, Procedures, or failure to establish and maintain procedures that protect the intake structure and auxiliary building during external flooding events. The inspectors determined that the procedural guidance of GMRR- AE-1002,Flood Control Preparedness for Sandbagging, as inadequate because stacking and draping sandbags at a height of four feet over the top of floodgates would be insufficient to protect the vital facilities to 1014 feet mean sea level, as described in the Updated Safety Analysis Report and station procedures. The licensee has entered this condition into their corrective action program as Condition Report 2010-2387. As result of this violation, the licensee has implemented a corrective action plan to correct identified deficiencies and ensure site readiness. This performance deficiency is more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of external events and affected the cornerstone objective of ensuring the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding resulted in the degradation of equipment and functions specifically designed to mitigate a flooding initiating event. In addition, an external flood event would degrade two or more trains of a multi-train safety system. Therefore, the finding was potentially risk significant to flood initiators and a Phase 3 analysis was required. The preliminary change in core damage frequency was calculated to be 3.1 E-5/year indicating that the finding was of substantial safety significance (Yellow). The finding was determined to have a crosscutting aspect in the area of problem identification and resolution, corrective action program, for failure to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and compleXity. Specifically, from 2003 to 2008, the licensee failed to initiate appropriate corrective actions to ensure regulatory compiiance of the external flooding design basis was maintained. (P.1 (d)) (Section 40A5.1)
05000285/FIN-2011006-02	2011Q4	Fort Calhoun	Inadequate Corrective Actions to Ensure the Reliability of the RAW Water Pump Power Cables	The NRC identified a cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to take effective corrective action following the initial discovery of water intrusion in cable vault manholes MH-5 and MH-31 in 1998, 2005, 2009, and 2011. Specifically, the licensee failed to take effective corrective action to establish an appropriate monitoring frequency, which took into account variable environmental conditions to mitigate potential common mode failure of raw water 4160 V motor cables in underground ducts and manholes identified during the Component Design Basis Inspection performed in 2009. The violation is being cited because the licensee had failed to restore compliance in a reasonable period following documentation of the issue as a non-cited violation issued December 30, 2009. The failure to take effective corrective action to ensure the reliability and capability of the safety-related cables powering the raw water pump motors was a performance deficiency. Furthermore, the finding was within the licensee\\\'s ability to foresee and correct because the licensee had multiple opportunities to correct the continuing challenge to the safety-related cables and raceways for the raw water system over an extended period. The finding was more than minor because it adversely affected the Mitigating Systems Cornerstone attribute of design control for ensuring the availability, reliability, and capability of systems that respond to Initiating Events to prevent undesirable consequences. The finding is of very low safety significance because it was a design deficiency that did not result in loss of operability or functionality. This finding has a crosscutting aspect in the decision-making program component of the human performance area because the licensee failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that it was unsafe in order to disapprove the action. Specifically, from 2005 until 2011, the licensee chose to postpone installation of proposed water level control corrective actions and failed to appropriately monitor water intrusion into underground ducts and manholes MH-5 and MH-31 for raw water 4160 V motor cables multiple times
05000285/FIN-2012002-01	2012Q1	Fort Calhoun	Inadequate Procedures to Mitigate a Design Basis Flood	The inspectors identified four examples of a violation of Technical Specification 5.8.1.a, Procedures, for failure to establish and maintain procedures to mitigate an external flooding event. The procedural guidance for flooding was inadequate to mitigate the consequences of external flooding. This finding, and its corrective actions, will be managed by the Manual Chapter 0350 Oversight Panel. This finding was more than minor because it adversely impacted the procedure quality, human performance and protection against external events attributes of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding is bounded by the significance of a related Yellow finding regarding the ability to mitigate an external flooding event (Inspection Report 05000285/2010008). This finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, for failure to thoroughly evaluate problems such that the resolutions address causes and extent of conditions. This also includes, for significant problems, conducting effectiveness reviews of corrective actions to ensure that the problems are resolved
05000285/FIN-2012002-02	2012Q1	Fort Calhoun	Failure to Classify Intake Structure Sluice Gates as Safety Class III	The inspectors identified a violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure of the licensee to classify the six intake structure exterior sluice gates and their motor operators as Safety Class III. This finding, and its corrective actions, will be managed by the Manual Chapter 0350 Oversight Panel. This finding was more than minor because it adversely impacted the protection against external events attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding is bounded by the significance of a related Yellow finding regarding the ability to mitigate an external flooding event (Inspection Report 05000285/2010008). This finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, for failure to thoroughly evaluate problems such that the resolutions address causes and extent of conditions. This also includes, for significant problems, conducting effectiveness reviews of corrective actions to ensure that the problems are resolved
05000285/FIN-2012002-03	2012Q1	Fort Calhoun	Failure to Meet Design Basis Requirements for Design Basis Flood Event	The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for failure to meet design basis requirements for protection of the safety related raw water system during a design basis flood for flood levels between 1,010-1,014 feet mean sea level as identified in Updated Safety Analysis Report, Section 9.8, Raw Water System. Specifically, the design basis states that water level inside the intake cells can be controlled during a design basis flood by positioning the exterior sluice gates to restrict the inflow into the cells. This finding, and its corrective actions, will be managed by the Manual Chapter 0350 Oversight Panel. This finding was more than minor because it adversely impacted the equipment performance and protection against external events attributes of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding is bounded by the significance of a related Yellow finding regarding the ability to mitigate an external flooding event (Inspection Report 05000285/2010008). This finding has a cross-cutting aspect in the area of problem identification and resolution, corrective action program, for failure to thoroughly evaluate problems such that the resolutions address causes and extent of conditions
05000285/FIN-2013008-01	2013Q2	Fort Calhoun	Inadequate Procedure for Combating Frazil Ice	The team identified a URI for a potentially inadequate procedure for combating frazil icing of the travelling screens. Description. The team performed a walkdown with a licensee senior reactor operator to determine the adequacy of the licensees preparations for the occurrence of a frazil ice event. The team noted three issues with the adequacy of the licensees procedure for such conditions, which was in Section V, Degraded River Conditions, of Procedure AOP-01, Acts of Nature, Revision 33. First, the procedure did not give direction to open a valve that would need to be opened to direct auxiliary steam to the travelling screens to melt the ice formed on the screens. Step 15 of Section V, Degraded River Conditions, of Procedure AOP-01, Acts of Nature, instructed operators to direct auxiliary steam to the travelling screens. The procedural step directs hooking up a high temperature hose to a fitting and throttling open valve AS-823. During the walkdown, the licensee and the team noted that the procedure did not include direction to open valve AS-811, a normally closed valve. Valve AS-811 would need to be opened in order to allow steam to the hose used to melt the ice formed on the travelling screens. Valve AS-811 was easily identified as being in the steam flowpath as it was adjacent to valve AS-823. Second, the team noted that the procedure lacked any personnel safety warnings associated with handling a hose with steam. The team did note, however, that gloves and a face shield were staged with the hose to protect personnel, however the procedure did not warn of the danger of handling live steam which is a condition rarely encountered by operators. Finally, the team noted that the procedure did not direct the operators where to direct the steam on the travelling screens, instead Procedure AOP-01, Step 15, Section V, simply instructed operators to initiate steam to the travelling screens. Also, the procedure gave no guidance for a strategy for preferential de-icing of one or more screens, such that an adequate river water flow path would be assured. This issue was entered into the CAP as Condition Report CR 2013-04309. The licensee informed the team that the susceptibility of frazil icing of the intake components was extremely low due to the low flow velocity of river water into the intake structure. The licensee provided Army Corps of Engineer reports which described the phenomenon. The team required additional inspection to determine if the flow into the intake structure was sufficiently low to preclude frazil ice formation. As a result, the team identified Unresolved Item URI 05000285/2013008-01, Inadequate Procedure for Combating Frazil Ice.
05000285/FIN-2013019-01	2013Q4	Fort Calhoun	Failure to Correct Deficiencies in Operations Support Center Functions	A Green noncited violation was identified for the failure of the licensee to correct deficiencies identified as a result of four exercises conducted between March 27, 2012, and May 7, 2013, as required by 10 CFR 50.47(b)(14). Specifically, the licensee failed to correct deficiencies associated with team briefing and tracking in the Operations Support Center (OSC) identified as a result of exercises conducted March 27, 2012; July 17, 2012; March 5, 2013; and May 7, 2013. The inspectors determined that the licensees failure to correct deficiencies identified by licensee evaluators is a performance deficiency within the licensees control. This finding is more than minor because it affected the emergency preparedness cornerstone objective and the Emergency Response Organization Performance cornerstone attribute. This finding was evaluated using the Emergency Preparedness Significance Determination Process and was determined to be of very low safety significance because it was a failure to comply with NRC requirements, was not a risk significant planning standard function, and was not a loss of planning standard function. The finding was not a loss of planning standard function because the licensee adequately corrected some deficiencies identified in exercises conducted in 2012 and 2013. The finding was entered into the licensees corrective action system as Condition Report 2013-22495. The finding was assigned a cross-cutting aspect of Problem Identification and Resolution because the finding was reflective of current performance and the licensee did not take appropriate corrective action to address safety issues and adverse trends.
05000285/FIN-2013019-02	2013Q4	Fort Calhoun	Licensee-Identified Violation	Title 10 CFR 50, Appendix B, Criterion III, Design Control requires that measures shall be established to assure that the design basis for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Contrary to this requirement, the licensee failed to assure that the design basis for safety related instrument racks inside containment were correctly translated into specifications, drawings, procedures, and instructions. The licensee initially identified and documented this violation in CR 2012-03100 and CR 2013-10935. This violation was of very low safety significance because it did not result in the loss of operability or functionality of any system or train.
05000285/FIN-2015003-01	2015Q3	Fort Calhoun	Failure to Maintain Safety Injection Tank Boron Concentration within Technical Specification Limits	A Green, self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI Corrective Action was identified because the licensee failed to identify and evaluate an adverse trend related to boron concentration in Safety Injection Tank (SIT) SI-6A and to take corrective actions to prevent boron concentration from going below the minimum concentration required by Technical Specifications. The licensees immediate corrective actions included documenting this condition in their corrective action program in Condition Report (CR) 2015-10181, declared SI-6A inoperable, and raised SI-6A boron concentration. The finding is more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone, in that this finding resulted in the SIT becoming inoperable when boron concentration fell below TS limits for approximately 8.5 days prior to August 20, 2015. Analysis conducted by a Senior Reactor Analyst determined the finding to be of very low safety significance (Green), primarily because the SIT function is needed only for mitigation of a postulated large-break loss of coolant accident, and the initiating-event frequency for such accidents is 2.5 x 10-6/year. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution and the Evaluation aspect, because the licensee did not thoroughly evaluate the issue and ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance.
05000285/FIN-2015003-02	2015Q3	Fort Calhoun	Failure to Maintain Fire Watch and Fire Watch Logs	Inspectors identified a Green, Severity Level IV, non-cited violation of 10 CFR 50.9(a), Completeness and Accuracy of Information, for the licensees failure to maintain the required fire watch logs complete and accurate in all material respects. The licensee entered this into their corrective action program as Condition Reports (CR) 2014-06416 and 2014-06680. This finding is more than minor because it adversely affected the human performance attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding has very low safety significance (Green) because it did not impact the ability to achieve safe shutdown. This findings severity level is based on an example in the Enforcement Policy, Section 6.1.d.2, which states, in part, that Severity Level IV violations involve violations of 10 CFR 50.59 (which) result in conditions evaluated as having very low safety significance.
05000298/FIN-2014003-01	2014Q2	Cooper	Failure to Correctly Translate Design Requirements into Installed Plant Configuration	Inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to assure that the applicable design basis for applicable structures, systems, and components were correctly translated into specifications, procedures, and instructions. Specifically, the licensee failed to correctly translate design requirements associated with high energy line breaks into the as-built facility for the service water pump room, diesel generator rooms 1 and 2, cable spreading room, and 4160 Vac vital switch gear room G. This does not represent an immediate safety concern because the licensee performed operability assessments for the affected areas, which established a reasonable expectation for operability pending resolution of the identified issue. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2014-01828. The failure to ensure that design requirements were correctly translated into installed plant equipment was a performance deficiency. This 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 associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to translate the design requirements into installed plant equipment resulted in a condition where structures, systems and components necessary to mitigate the effects of a high energy pipe break may not have functioned as required. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, 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) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program. Inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor of this finding occurred in 2003, and does not reflect current licensee performance.
05000298/FIN-2014003-02	2014Q2	Cooper	Failure to Report Conditions Prohibited by Technical Specifications	Inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73, Licensee Event Report, associated with the licensees failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria. Specifically, a condition prohibited by technical specifications existed for trip and throttle valve RCIC-MOV-14 for a period of time longer than the allowed outage time. This does not represent an immediate safety concern because this issue is only associated with reporting requirements. The licensee entered this deficiency into their corrective action program for resolution as Condition Reports CR-CNS-2014-03387 and CR-CNS-2014-03457. The licensees failure to submit a licensee event report within 60 days following discovery of an event meeting the reportability criteria was a performance deficiency. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, inspectors evaluated the performance deficiency using traditional enforcement. The violation was evaluated using Section 2.3.11 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. In accordance with Section 6.9, Example 9, of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. Inspectors determined that a cross-cutting aspect was not applicable to this performance deficiency because the failure to make a required report was strictly associated with a traditional enforcement violation
05000298/FIN-2014003-03	2014Q2	Cooper	Failure to Follow Seismic Housekeeping Requirements for Scaffolding	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to follow the requirements of Station Procedure 0.41, Seismic Housekeeping, Revision 10. Specifically, the licensee stored a rolling scaffold in the vicinity of Division II service water booster pumps and failed to properly restrain it. The licensee restrained the rolling scaffold in accordance with Station Procedure 0.41 and assessed operability of the service water booster pumps. The licensee determined that during the time the rolling scaffold was unrestrained one of the Division II service water booster pumps was inoperable. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2014-03000. The licensees failure to follow Station Procedure 0.41 seismic housekeeping requirements for a rolling scaffold in the vicinity of Division II service water booster pumps was a performance deficiency. This 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 associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the unrestrained scaffolding resulted in a condition where during a seismic event a service water booster pump may not have been able to perform its specified safety function. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, 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 a least a single train for longer than its 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-significance in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human performance associated with training because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values (H.9).
05000298/FIN-2014005-01	2014Q4	Cooper	Failure to Follow Procedure for Post Maintenance Testing	The inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to follow Special Procedure GEH-TP-116, Procedure for the Operation and Maintenance of the REMTAKE-2/D-100 Modified REMTAKE 2, Revision 3, for postmaintenance testing following corrective maintenance. Specifically, the licensee did not follow post-maintenance testing requirements associated with the calibration of the bleeder valve for the REMTAKE-2/D-100 tool following corrective maintenance to address water intrusion. This resulted in the bleeder valve being misadjusted and nullifying the fail-safe feature of the REMTAKE-2/D-100 tool. With the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REMTAKE-2/D-100 tool and dropped onto the reactor core top guide when the supplemental employee inadvertently pressed the disengage button. No reactor fuel was damaged as indicated by normal radiation levels and air samples on the refuel floor and reactor water coolant samples. The licensees immediate corrective actions for the event was to suspended all in-vessel maintenance activities and remove REMTake-2/D-100 grapple from service and determined functionality of the tool. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2014-06809. The licensees failure to follow the post-maintenance testing requirements in Special Procedure GEH-TP-116 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the associated objective of maintaining functionality of fuel cladding. Specifically, with the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when a supplemental employee inadvertently pressed the disengage button. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 09, 2014, inspectors determined that the finding was of very low safety significance (Green) because the finding did not impact the fuel barrier because it: (1) does not increase the potential for failure of the freeze seal or if unmitigated have the potential to cause a disruption of residual heat removal/decay heat removal or a loss of inventory event; (2) does not involve two or more adjacent control rods with the potential to, or actually, add postive reactivity; and (3) does not degrade the ability to isolate a drain down or leakage path. The finding has a cross-cutting aspect in the area of human performance associated with the field presence component because the licensee failed to ensure supervisory and management oversight of work activities including contractors and supplemental personnel (H.2).
05000298/FIN-2014005-02	2014Q4	Cooper	Implementation of Enforcement Guidance Memorandum 11-003, Revision 2, Causes Conditions Prohibited by Technical Specifications	During Refueling Outage 28, Cooper Nuclear Station performed Operations with a Potential for Draining the Reactor Vessel (OPDRV) activities while in Mode 5 without an operable secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measure to terminate the uncovering of fuel. Secondary containment is required by TS 3.6.4.1 to be operable during OPDRV. The required action for this specification is to suspend OPDRV operations. Therefore, entering the OPDRV without establishing secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). The NRC issued Enforcement Guidance Memorandum (EGM) 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliances with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to: (1) adhere to the NRC plain language meaning of OPDRV activities, (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times, (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5, and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities. The inspectors reviewed this Licensee Event Report for potential performance deficiencies and/or violations of regulatory requirements. The inspectors reviewed the stations implementation of the Enforcement Guidance Memorandum 11-003, Revision 2, during operations with a potential for draining the reactor vessel. Specific observations included: 1. The inspectors observed that the operations with a potential for draining the reactor vessel activities were logged in the control room narrative logs, and that the log entry appropriately recorded the standby source of makeup designated for the evolutions. 2. The inspectors noted that the reactor vessel water level was maintained at least greater than 21 feet above the top of the reactor pressure vessel flange as required by Technical Specification 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designed in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours. 3. The inspectors verified that the operations with a potential for draining the reactor vessels were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the operations with a potential for draining the reactor vessels. The inspectors verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events. Technical Specification 3.6.4.1 requires, in part, that secondary containment shall be operable during operations with a potential for draining the reactor vessel. Technical Specification 3.6.4.1, Condition C, requires the licensee to initiate actions to suspend operations with a potential for draining the reactor vessel immediately when secondary containment is inoperable. Contrary to the above, from October 3, 2014 to October 22, 2014, Cooper Nuclear Station performed operations with a potential for draining the reactor vessel activities while in Mode 5 without an operable secondary containment. Specifically, the station conducted the following seven operations with a potential for draining the reactor vessel activities without an operable secondary containment: Draining reactor recirculation pump without the jet pump plugs fully installed Control rod drive maintenance Removal of jet pump plugs associated with reactor recirculation pump B maintenance Venting the control rod drives Defeating the scram function for two control rod drives and support IVVI inspections Reactor recirculation pump A seal maintenance Control rod drive freeze seal These conditions were reported as conditions prohibited by Technical Specifications. The licensee entered this issue into its corrective action program as Condition Reports CR-CNS-2014-06293. Since this violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy, and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request within 4 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The Licensee Event Report is closed.
05000313/FIN-2009006-01	2009Q1	Arkansas Nuclear	Licensee-Identified Violation	The following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section VI of the NRC Enforcement Policy, NUREG-1600, for being dispositioned as a noncited violation. Title 10 CFR Part 50, Appendix B, Criterion XV, Nonconforming Materials, Parts, or components, requires, in part, that measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their inadvertent use or installation. These measures shall include, as appropriate, procedures for identification, documentation, segregation, disposition, and notification to affected organizations. Contrary to the above requirement, the licensee failed to ensure that a solenoid valve design that had been determined to be inadequate in January 2006 was controlled and not issued for use or installation. This resulted in a subsequent failure of decay heat cooler A bypass valve (CV-1433) because of the inadequate solenoid valve. This finding was determined to have very low safety significance because the condition did not result in the actual loss of any component, train, or system. This issue was entered into the licensees corrective action program as condition reports ANO-1-2008-2525, ANO-1-2008-2578 and ANO-1-2008-2625.
05000313/FIN-2012005-01	2012Q4	Arkansas Nuclear	Failure to Implement Adequate Design Change Controls for Permanent Removal of Service Water Check Valves SW-604A and SW-604B	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that design changes, including field changes, shall be subject to design control measures commensurate with those applied to the original design. Specifically, from October 4, 2012, to November 8, 2012, the licensee failed to ensure that the design change, which directed the permanent removal of check valves SW-604A and SW-604B from the service water return lines of safety-related auxiliary building electrical rooms emergency chillers VCH-4A and VCH-4B, included the requisite evaluation of the initial design basis and mitigating safety system functions of these components. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-1-2012-1681. The failure to ensure that safety-related system modifications were subject to design control measures commensurate with those applied to the original design for the removal of check valves SW-604A and SW-604B and replacement of these components with spool pieces was a performance deficiency. The finding is more than minor because it is associated with the design control attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to have very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating component that did not affect the operability or functionality of the system. The inspectors determined that the finding had a cross-cutting aspect in the area of human performance associated with the component of decision making because the licensee failed to use conservative assumptions and adopt a requirement to demonstrate that the proposed action was safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action. Specifically, the licensee assumed that the check valves had no safety function without determining the actual design basis and mitigating safety system functions of these components.
05000313/FIN-2012005-02	2012Q4	Arkansas Nuclear	Failure to Perform Required Examinations of Reactor Vessel Flange Seal Leak-Off Lines	The inspectors identified a non-cited violation, with two examples, of Title 10 CFR 50.55a(g)(4), which requires that components classified as ASME Code Class 1, Class 2, and Class 3 meet the requirements set forth in Section XI of the applicable editions of the ASME Boiler and Pressure Vessel Code and Addenda. Title 10 CFR 50.55(a)(g)(4)(ii) requires that inservice examination of components be conducted during successive 120-month inspection intervals and comply with the requirements of the latest edition and addenda of the Code applicable to the specific interval. Section XI (of prior and current applicable editions of the Code), Articles IWC-5221 and IWD-5221 require that, for Class 2 and Class 3 components, a system leakage test be performed at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function. Contrary to the above, prior to September 17, 2012, for the Class 2 and Class 3 reactor vessel flange leak-off lines for both Units 1 and 2, the licensee failed to perform leakage tests at the system pressure obtained while the system was performing its normal operating function. The licensee has entered this issue into the corrective action program as Condition Report CR-ANO-C-2012-02672. The inspectors determined that the failure to perform the examinations required by 10 CFR 50.55a(g)(4) on the Units 1 and 2 reactor vessel flange seal leak-off lines is a performance deficiency. The performance deficiency is more than minor because it is associated with the Barrier Integrity Cornerstone attribute of structures, systems, and components and barrier performance and adversely affects the cornerstone objective to provide a reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the finding could not result in exceeding the reactor coolant system leak rate for a small loss-of-coolant accident, nor could the finding have likely affected other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. This issue did not have a cross-cutting aspect associated with it because it is not indicative of current performance.
05000313/FIN-2012005-03	2012Q4	Arkansas Nuclear	Failure to Adequately Evaluate Discolored Boric Acid on ReactorMake-up Water Pipe	The inspectors identified a finding for the failure to perform an adequate boric acid evaluation on the reactor make-up water pipe located in the overhead of the train B charging pump room. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-C-2012-03119 The inspectors determined that the failure to adequately evaluate the reddish to brownish discoloration near a reactor make-up water pipe fillet weld and demonstrate that the structural integrity of the weld or the pipe was not adversely impacted was a performance deficiency. This finding was more than minor because it is associated with the Initiating Events Cornerstone attribute of performance and affects the cornerstone objective to limit the likelihood of those events that upset plant stability and that challenge critical safety functions during power operations and if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if licensee personnel continue to perform boric acid evaluations under the assumption that reddish to brownish discoloration on stainless steel pipe at low temperature is not indicative of localized corrosion, a more significant instance of corrosion on stainless steel pipe may not be appropriately evaluated and corrected. Using Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the finding was determined to be of very low safety significance (Green) because the finding could not result in exceeding the reactor coolant system leak rate for a small loss-ofcoolant accident, nor could the finding have likely affected other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. This finding had a cross-cutting aspect in the area of human performance, associated with the decision-making component, because the licensee failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disprove the action. Specifically, the licensee inappropriately assumed that the discoloration on the reactor make-up water line was staining by migrating particulate without fully evaluating other possible causes of the discoloration.
05000313/FIN-2012005-04	2012Q4	Arkansas Nuclear	Failure to Maintain Licensed Operator Examination Integrity	The inspectors identified a Severity Level IV traditional enforcement violation with an associated Green non-cited violation of 10 CFR 55.49, Integrity of Examinations and Tests, for the failure of the licensee to ensure the integrity of Unit 2 licensed operator biennial written examinations. During the 2012 biennial written examination cycle, the exams were administered in a classroom that lacked positive controls to ensure that no one could observe the exam material being administered. Three of the six written exams administered in this room had repeat exam questions and references compared to other weeks test, and the references used on the exam, were accessed using computer terminals whose screens were viewable if a curtain was not fully closed. Having the ability to view into the room while exam material was being displayed on the computer screens during exam administration is considered an exam integrity compromise. However, an evaluation of the written exam results and interviews with the licensed operators signed in on an exam security agreement and consistent administration of the examination. The licensee has entered this issue into the corrective action program as Condition Report CR-ANO-C-2012-01834. The failure of the licensees training staff to maintain the integrity of examinations administered to licensed operations personnel was a performance deficiency. The finding was more than minor because it adversely affected the Human Performance attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the performance deficiency could have become more significant in that allowing licensed operators to return to the control room without valid demonstration of appropriate knowledge on the biennial written examinations could be a precursor to a more significant event. Using NRC Inspection Manual Chapter 0609, Significance Determination Process, Attachment 4, Tables 1 and 2 worksheets; and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process, the finding was determined to have very low safety significance (Green). Although the 2012 finding resulted in a compromise of the integrity of biennial written examinations, with no compensatory actions immediately taken when the compromise should have been discovered, the equitable and consistent administration of the biennial written examination was not actually affected by this compromise. In addition, the failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process, which resulted in its association with a Severity Level IV violation consistent with Sections 2.2.4 and 6.4d of the NRC Enforcement Policy. This finding has a crosscutting aspect in the area of resources associated with ensuring that procedures are adequate to ensure nuclear safety. A combination of a NRC procedure review and discussion with the licensee revealed that there were inadequate standardized criteria on site for what minimum actions have to be taken to ensure the subject room is secure prior to and during administration of licensed operator exams.
05000313/FIN-2012005-05	2012Q4	Arkansas Nuclear	Radiation Workers Failed to Follow High Radiation Area Procedure	The inspectors documented a self-revealing non-cited violation of Unit 2 Technical Specification 6.4.1.a for the failure to follow station procedure EN-RP-100, Radiation Worker Expectations, Revision 7. Specifically, when working in a posted high radiation area, a worker received several electronic dose rate alarms and failed to immediately exit the area, notify others in the work area and notify radiological protection personnel of the dose rate alarm. The licensee has entered this issue into the corrective action program as Condition Report CR-ANO-2-2012-2830. Radiation workers failing to follow the station procedure for high radiation areas is a performance deficiency. This finding is more than minor because it affected the human performance attribute of the Occupational Radiation Cornerstone and adversely affected the cornerstone objective to ensure the adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Using Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to have very low safety significance (Green) because the finding: (1) was not related to as-low-as-reasonably-achievable planning or work control, or exposure control, (2) did not involve an overexposure, (3) did not constitute a substantial potential for overexposure, and (4) did not compromise the licensees ability to assess dose. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with work practices component, self and peer checking. Specifically, at multiple points while entering and working in the high radiation area the radiation workers failed to perform self and peer checks commensurate with the risk of the assigned task to perform the work safely.
05000313/FIN-2012005-06	2012Q4	Arkansas Nuclear	Failure to Provide an Accurate Maintenance Tagout Results in Loss of Reactor Coolant System Inventory	The inspectors documented a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, for the failure to provide an appropriate maintenance tagout. Specifically, maintenance tagout LPSI-013-A-2SI-14C for maintenance on the Unit 2 low pressure safety injection header C incorrectly specified that the reactor coolant system level should be less than 70 inches to work on the low pressure safety injection header components which resulted in a loss of reactor coolant system inventory. The licensee has entered this issue into the corrective action program as Condition Report CR-ANO-2-2012-2645. The failure to provide an appropriate tagout for maintenance on the low pressure safety injection header was a performance deficiency. Specifically, the reactor coolant system level should have been lowered to less than 65 inches rather than less than 70 inches, as stated in the tagout, to prevent the loss of reactor coolant inventory. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations, and is therefore a finding. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix G, Shutdown Operation, Attachments 1 and 2, it was determined that because this finding increased the likelihood of a loss of RCS inventory, especially during reduced inventory condition, a Phase 2 analysis was required. The senior reactor analyst determined the finding to have very low safety significance (Green) because even without operator action residual heat removal would not have been lost and there were no other complicating issues. The finding has a cross-cutting aspect in the area of human performance, associated with the resources component, in that the licensee failed to ensure that personnel equipment, procedures, and other resources are available and adequate to assure nuclear safety. Specifically, station Procedure OP-2103.011, Draining the Reactor Coolant System, Revision 48, was not up to date and accurate for determining the appropriate reactor vessel level for the development of the maintenance tagout.
05000313/FIN-2012005-07	2012Q4	Arkansas Nuclear	Failure to Correct a Degraded Condition Associated with the Unit 2Condenser Vacuum Pump Solenoid Valves Results in Reactor Trip	The inspectors documented a self-revealing finding for the failure to identify the cause and take appropriate corrective actions to address the degraded performance of the Unit 2 condenser vacuum pump solenoid valves. Specifically, from 2008 through 2012, Unit 2 operations staff identified the degraded performance of several solenoid valves associated with the condenser vacuum pumps. These performance issues were entered into the corrective action program a number of times during this period. On August 8, 2012, while switching condenser vacuum pumps for oil checks, two solenoid valves failed to close resulting in a turbine trip and an automatic trip of the reactor. The licensee has entered this issue into the corrective action program as Condition Report CR-ANO-2-2012-1429. The failure to identify the cause and take appropriate corrective actions to address the degraded performance of the Unit 2 condenser vacuum pump solenoid valves is determined to be a performance deficiency. The performance deficiency is determined to be more than minor 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 challenged critical safety functions during power operations, and is therefore a finding. Using Manual Chapter 0609, Attachment 4, Initial Characterizations of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, the finding was determined to have very low safety significance (Green) because, although it resulted in a reactor trip, it did not result in the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding does not have a cross-cutting aspect because none were determined to be appropriate.
05000313/FIN-2012005-08	2012Q4	Arkansas Nuclear	Incorrectly Positioned Manual Jack Sleeve Results in Feedwater Regulating Valve not Fully Closing	The inspectors identified a finding associated with a failure to provide sufficient work instructions for a maintenance activity on 2CV-0748 main feedwater regulating valve. Specifically, contrary to station procedure EN-WM-105, Planning , Revision 10, the work instructions generated to repair 2CV-0748 main feed regulating valve, incorrectly positioned the manual jack sleeve following repairs. In addition, the work instructions did not provide an adequate post maintenance test that would verify that no new problems were created by the maintenance activity. This resulted in the main feedwater regulating valve not fully closing following the Unit 2 trip on August 8, 2012. The licensee entered this issue into the corrective action program as Condition Report CR-ANO-2-2012-1432. The failure to provide sufficient work instructions to correctly position the manual jack sleeve after repairs and to provide a sufficient post maintenance test that would verify no new problems were created by the maintenance activity is a performance deficiency. The performance deficiency was more than minor 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 and is therefore a finding. Using Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process (SDP) for Findings at Power, the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding did not: (1) result in an actual loss or operability of functionality, (2) represent a loss of system and/or function, (3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, (4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safetysignificant for greater than 24 hours and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The finding was determined to have a cross-cutting aspect in the area of human performance, associated with work control component, in that the licensee failed to plan and coordinate work activities consistent with nuclear safety. Specifically, the licensee failed to appropriately coordinate work activities by using instructions that incorrectly positioned the manual jacking sleeve fully upward.
05000313/FIN-2012005-09	2012Q4	Arkansas Nuclear	Licensee-Identified Violation	Unit 1 Technical Specification 6.4.1.a states, in part, that written procedures shall be implemented in accordance with Regulatory Guide 1.33, Revision 2, Appendix A, Section 7.e, Radiation Protection Procedures, which requires procedures for access control to radiation areas. Station Procedure EN-RP-100 Radiation Worker Expectations, Revision 7, step 5.3 (16) stated, in part, To enter a high radiation area, the radiation worker must be provided with a radiation monitoring device which continuously indicates the radiation dose rate in the area. Contrary to the above, a radiation worker entered a posted high radiation area without a radiation monitoring device which continuously indicates dose rate. Using Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to have very low safety significance (Green) because the finding: (1) was not related to as-low-as-reasonably-achievable planning or work control, or exposure control, (2) did not involve an overexposure, (3) did not constitute a substantial potential for overexposure, and (4) did not compromise the licensees ability to assess dose. The issue was placed into the corrective action program as Condition Report CR-ANO-1-2012-1599.
05000313/FIN-2014005-02	2014Q4	Arkansas Nuclear	Failure to Provide Flow Protection For Auxiliary Feedwater Pump in Emergency Operating Procedures	Inspectors identified a noncited violation of Unit 1 Technical Specification 5.4, Procedures, for the licensees failure to establish adequate emergency operating procedures. Specifically, the licensees emergency operating procedures failed to establish minimum flow protection for the Unit 1 auxiliary feedwater pump, which could result in catastrophic failure of the pump. The issue was documented in Condition Report CR-ANO-1-2014-00286 and the procedures were revised to correct the condition. The failure to establish minimum flow protection for the Unit 1 auxiliary feedwater pump in emergency and abnormal operating procedures in accordance with the emergency operating procedure writers guide was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was 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. Specifically, inadequat emergency and abnormal operating procedures could have resulted in failure of the auxiliary feedwater pump, a mitigating system for a loss of main and emergency feedwater. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation because the finding represented a loss of system function. A Region IV senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was less than 4.2E-7/year (Green). The dominant core damage sequences included losses of one of the safety related 4160 volt electrical buses, steam generator tube ruptures, and plant transients. The equipment that helped mitigate the risk included the high pressure injection system (for feed and bleed) and the main and emergency feedwater systems. This finding did not have a cross-cutting aspect because the most significant contributing cause was not indicative of current performance. Specifically, the emergency and abnormal operating procedures for operating auxiliary feedwater had not changed for at least 2 years.
05000313/FIN-2014005-03	2014Q4	Arkansas Nuclear	Failure to Correct Weaknesses During Drills and Exercises	The inspectors identified a non-cited violation of 10 CFR Part 50.47(b)(14) for the failure to correct a deficiency identified in a 2013 simulator drill. Specifically, control room operators did not implement the procedure that describes how the site will maintain continuous communication with threat notification sources during a drill conducted August 7, 2013, and also during the September 16, 2014, biennial exercise. The inspectors determined that the licensees corrective actions for this issue were incomplete and did not address the extent of condition. The licensee has entered the issue into the corrective action program in corrective action documents WT-WTANO-2014-00189 and Condition Report CR-ANO-C-2014-02478. The failure to correct weaknesses occurring in drills and exercises is a performanc deficiency within the licensees ability to foresee and correct. The performance deficiency is more than minor because it is associated with the emergency response organization performance attribute of the Emergency Preparedness Cornerstone and it adversely impacted the cornerstone objective. The licensees ability to implement adequate measures to protect the health and safety of the public in the event of hostile action and a radiological emergency is degraded when it fails to correct performance that precludes the effective implementation of the emergency plan. This finding was evaluated using Manua Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), Attachment 2, dated February 24, 2012, and was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements, was not associated with a risk-significant planning standard, and was not a loss of planning standard function. The finding was not a loss of function because the deficiency that was identified was not associated with classification, notifications to state and local agencies, or the development of protective action recommendations. The finding was assigned a cross-cutting aspect in the area of problem identification and resolution, associated with the resolution of issues because the licensee failed to evaluate the initial performance issues to ensure that resolutions adequately addressed the extent of condition commensurate with their safety significance. The licensee failed to recognize in August 2013 that continuous communications with threat notification sources is required by regulation and that performance issues with the implementing procedure should be communicated to the entire control room staff population (P.2).
05000313/FIN-2015002-01	2015Q2	Arkansas Nuclear	Inadequate Procedure for Severe Weather Preparation	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, & Drawings, for the failure to establish appropriate procedures for preparations for severe weather. Specifically, inspectors observed that the licensee failed to ensure that all outside areas were inspected in order to secure material prior to severe weather, to reduce the probability of light material missile damage on plant equipment. The licensee concluded that the assignment of responsibilities was unclear in Procedure EN-FAP-EP-010, Severe Weather Response, Revision 1, leading to confusion among the two operating crews. This issue was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2015-00854 and CR-ANO-C-2015-00859. The failure to have a procedure to ensure that all outside areas would be inspected in order to secure loose material prior to the arrival of severe weather, to reduce the probability of light material missile damage on plant equipment was a performance deficiency. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, during severe weather, unsecured material could become a missile that impacts equipment and upsets plant stability. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that the finding had very low safety significance (Green) because it did not represent an actual reactor trip and the loss of mitigation equipment. This finding has a human performance crosscutting aspect associated with work management, in that the organization failed to implement a process of planning, controlling, and executing work activities, including coordination with different groups or job activities. Specifically, only one crew performed the required inspections when severe weather had been forecast since the procedure in use did not clearly assign responsibilities to both operating crews (H.5).
05000313/FIN-2015002-04	2015Q2	Arkansas Nuclear	Failure to Perform Testing of Diesel Fuel Oil Transfer Piping	The inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to establish and maintain an adequate testing program for the fuel oil transfer piping for Units 1 and 2. Specifically, the licensee did not establish inservice testing to detect degradation of the fuel oil piping between the fuel oil storage tanks and the emergency diesel generator day tanks. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-2-2015-01092. The failure to perform the required testing of the fuel oil piping is a performance deficiency. The performance deficiency is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems Cornerstone, and affects the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequence. Specifically, the licensee failed to perform examinations required to provide reasonable assurance that the piping could perform its intended function during design basis seismic events, and therefore maintain the ability to supply fuel to the emergency diesel generators. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems, the inspectors determined the finding is 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 initiating event. The finding has a cross-cutting aspect in the area of human performance, associated with conservative bias, because the licensee did not use decision-making practices that emphasized prudent choices over those that were simply allowable. Specifically, during the buried piping initiative inspections that were completed in August 2013, the licensee failed to identify that the condition of the safety-related piping had never been evaluated and was being treated as a run to failure component (H.14).
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.
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).
05000368/FIN-2014003-09	2014Q2	Arkansas Nuclear	Reporting of Unit 2 Events as Unplanned Scrams with Complications	The inspectors identified an unresolved item associated with not reporting two events in the unplanned scrams with complication performance indicator for Unit 2. Description. On March 31, 2013, and December 9, 2013, Unit 2 experienced a loss of condenser vacuum due to the transfer of electrical busses to startup transformer 2. By design, the lockout of the preferred offsite power source startup transformer 3 in these events resulted in the loss of non-vital circulating water pumps and the subsequent loss of condenser vacuum. The loss of condenser vacuum ultimately resulted in the loss of main feedwater pump capability. Both main feedwater pumps are steam driven a Unit 2. Neither of these concerns were reported under this performance indicator Unit 2 has a non-vital electric-driven feedwater pump, 2P-75, which remained availabl and capable of supplying sufficient feedwater flow to remove decay heat up throug about 4 percent reactor power. The non-vital 2P-75 pump, which can be supplie directly from the condensate storage tanks, does not rely on condenser vacuum o portions of the main feedwater system, and supplies feedwater for plant cooldown heatup, hot standby conditions, and startup. The NRC stated that the intent of the guidance in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guidance, for unplanned scrams with complications was to have the main feedwater available or recoverable within 30 minutes after a trip or scram assuming a loss of emergency feedwater to ensure safe shutdown of the plant The licensee submitted a frequently asked question to the Nuclear Energy Institut working group because the licensee considered the trip to be uncomplicated because a least one or more electric-driven feedwater pumps remained available as backup to th emergency feedwater system. The Nuclear Energy Institute guidance refers to a electric-driven main feedwater source; however, the intent was to provide backu feedwater capability should emergency feedwater be lost, which would be met b the 2P-75 pump. The frequently asked question is currently under review by NR headquarters and the Nuclear Energy Institute working group to decide whether o not the above events should be captured as unplanned scrams with complications The inspectors concluded that an additional inspection was required to assess whethe or not the events should have been included in the unplanned scrams with complications performance indicator for Unit 2. This issue was identified as an Unresolved Item URI 05000368/2014003-09, Reporting of Unit 2 Events as Unplanned Scrams with Complications.
05000368/FIN-2014005-01	2014Q4	Arkansas Nuclear	Failure to Develop Adequate Guidance for Extreme Damage Mitigation	The inspectors identified a noncited violation of 10 CFR 50.54(hh)(2) for the failure to develop mitigating strategy guidance that would successfully maintain or restore Unit 2 core cooling after the loss of large areas of the plant. Specifically, the guidance did not ensure the capability of the mitigating strategy because an unisolated flow diversion could have prevented water from reaching the steam generators and cooling the core. The issue was documented in Condition Report CR-ANO-2-2014-03277, and the procedure was revised to correct the condition. The licensees failure to develop mitigating strategy guidance that would successfully maintain or restore Unit 2 core cooling after loss of large areas of the plant, as required by 10 CFR 50.54(hh)(2), was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was 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. Specifically, the guidance did not ensure the capability of the mitigating strategy because an unisolated flow diversion could have prevented water from reaching the steam generators and cooling the core. Using NRC Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and NRC Manual Chapter 0609, Appendix L, B.5.b Significance Determination Process, dated December 24, 2009, Table 1, SDP Screening Worksheet for B.5.b, the finding was determined to be of very low safety significance because the performance deficiency represented the unrecoverable unavailability of an individual mitigating strategy; other core cooling mitigating strategies were available. This finding has a human performance crosscutting aspect associated with avoid complacency, in that the licensee failed to recognize and plan for the possibility of latent issues, even while expecting successful outcomes (H.12).
05000368/FIN-2014005-04	2014Q4	Arkansas Nuclear	Failure to Verify Ventilation Design for Vital Switchgear	Inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to maintain design control of the Unit 2 vital switchgear ventilation system. Specifically, in 2002, the licensee failed to ensure that the ventilation system was capable of cooling the switchgear under design basis conditions. The licensee documented the issue in Condition Report CR-ANO-2-2014-00352 and conducted an evaluation to verify the capability of the ventilation system. Failure to ensure that the ventilation system was capable of cooling the switchgear under design basis conditions was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating system, and the system maintained its functionality. The inspector determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current licensee performance.
05000368/FIN-2014005-05	2014Q4	Arkansas Nuclear	Unit 2 Unplanned Scrams Performance Indicator	The inspectors identified an unresolved item associated with the Unit 2 unplanned scrams per 7,000 critical hours performance indicator related to a reactor trip. On April 27, 2014, Unit 2 experienced an Axial Shape Index (ASI) trip when performing a rapid downpower at the request of the transmission grid operator due to severe weather affecting the grid. This unplanned reactor trip was caused by exceeding the Core Protection Calculator ASI limits. As noted in Licensee Event Report 05000368/2014-003-00, and NRC Inspection Report 2014004, the ASI limits were exceeded, due in part to plant operators not following the downpower reactivity plan. The automatic trip occurred at approximately 50 percent power and was uncomplicated. The unplanned scrams per 7000 critical hours performance indicator measures the rate of scrams per year of operation at power and provides an indication of initiating event frequency. The licensee did not include this scram as an input into the unplanned scram performance indicator and submitted a frequently asked question to the NRC Reactor Oversight Process Working Group. The frequently asked question is currently under review to decide whether the above event should be captured as an unplanned scram. The licensee noted that anticipatory plant shutdowns to reduce the impact of externa events are excluded from this performance indicator. The licensee believed the intent of the exclusion was met because the shutdown being performed at the time the reactor trip occurred had been requested by the transmission grid operator due to the impacts of weather conditions. The inspectors noted that Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, guidance states that an unplanned scram is a scram that is not an intentional part of a planned evolution or test as directed by a normal operating or test procedure. This includes scrams that occurred during the execution of procedures or evolutions in which there was a high chance of a scram occurring but the scram was neither planned nor intended. The inspectors noted that the April 27, 2014 reactor trip was an automatic trip, which was not intended as part of the rapid downpower evolution that was being performed. The inspectors also noted that had the licensees reactivity plan been followed, the severity of the ASI transient would likely have been managed and a trip avoided. The inspectors concluded that additional inspection was required to assess whether the scram should have been reported in the unplanned scrams per 7,000 critical hours performance indicator for Unit 2. This issue was identified as Unresolved Item URI 05000368/2014005-05, Unit 2 Unplanned Scrams Performance Indicator.
05000368/FIN-2015002-02	2015Q2	Arkansas Nuclear	Failure to Protect Motor Control Center from Potential Pipe Spray	The inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to select and review equipment for suitability of application that is essential to the safety-related function of Unit 2 motor control center (MCC) 2B-52. Specifically, the licensee failed to ensure that the safety-related electrical equipment inside the MCC was adequately protected from water spray in the event of a failure of overhead non-seismic category 1 pipes, in accordance with the safety analysis report. Inspectors identified that the installed spray curtain only protected the front of the cabinet, while a cooling water pipe that could break during a seismic event was located directly above the length of the MCC. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-C-2015-01342. The failure to protect Unit 2 MCC 2B-52 from possible spray of overhead non-seismic category 1 pipes by installing a spray shield in accordance with the safety analysis report 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 adversely affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency could result in failure of one train of essential safety features during a seismic event, such as exhaust fans for the emergency diesel generators, containment spray isolation valves, and high pressure safety injection isolation valves. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined to require a detailed risk evaluation because the finding involved degradation of equipment specifically designed to mitigate a seismic event and could degrade one train of a system that supports a risk significant function. A senior reactor analyst performed the detailed risk evaluation and estimated the change to the core damage frequency was 3.8E-8/year (Green). The dominant core damage sequences included seismically induced losses of offsite power. This finding did not have a cross-cutting aspect associated with it because the most significant contributing cause was not indicative of present performance. Specifically, the condition had existed since plant construction, with no recent substantial opportunities to identify the issue.
05000368/FIN-2015002-03	2015Q2	Arkansas Nuclear	Failure to Verify Material Properties Prior to Installation	The inspectors reviewed a self-revealing finding involving failure to verify that the proper material was installed in the plant during initial construction of the Unit 2 reactor coolant system (RCS) sample system. Specifically, failure to use the correct material resulted in two through-wall leaks in the supply line to the 2E30 cooler for the RCS sample system. The licensee removed the components with the incorrect material and installed components of the correct material. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-C-2014-01800. The failure to verify the correct materials were installed in the plant is a performance deficiency. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Initiating Events Cornerstone and affects the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as during power operations. Specifically, failure to install the correct material resulted in failure of the RCS sample system and the inability to meet technical specification requirements for determining dose equivalent Xenon-133. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions, the inspectors determined the finding is of very low safety significance (Green) because the transient initiator did not cause a reactor trip and the loss of mitigating equipment. This finding has not been assigned a cross cutting aspect because the incorrect material was installed during initial construction, and is not indicative of current plant performance.
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
05000382/FIN-2003007-05	2003Q4	Waterford	Failure to establish appropriate instructions and accomplish those instructions	A self-revealing apparent violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the failure to establish appropriate instructions and accomplish those instructions for installation of a fuel line for Train A emergency diesel generator in May 2003. This failure resulted in uneven and excessive scoring of the tubing that ultimately led to a complete 360 degree failure of the fuel supply line on September 29, 2003, during a monthly surveillance test. This finding is unresolved pending completion of a significance determination. The finding was greater than minor because it directly impacted the availability and reliability of an emergency diesel generator which is used to mitigate the loss of AC power to the respective safety related bus. The finding was determined to have a potential safety significance greater than very low significance because the failure resulted in an actual loss of the safety function of the Train A emergency diesel generator for an extended period of time (Section 4OA3).
05000382/FIN-2011009-01	2011Q3	Waterford	Failure To Use Effective Engineering Controls As Part Of PreJob Planning To Reduce Contamination and Subsequent Exposure	The inspectors identified an apparent White finding because the licensee failed to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure. The primary reason for the dose overage was the licensee\\\'s failure to prevent radioactive water from leaking into work areas and raising radiation dose rates. As corrective action, the licensee installed a trough system to collect and route the radioactive water away from the work area and to the reactor containment floor drain system. This issue was placed in the corrective action program as Condition Report CR-WF3-2011-05672. The failure to use effective engineering controls as part of pre-job planning to reduce contamination and subsequent exposure is a performance deficiency. The finding is more than minor because it was similar to (the more than minor) Example 6.i in Inspection Manual Chapter 0612, Appendix E, Example of Minor Issues, in that the actual collective dose exceeded 5 person-rem and exceeded the planned, intended dose by more than 50 percent. Additionally, the finding is associated with the program and process attribute of the Occupational Radiation Safety cornerstone and affected the cornerstone objective in that it increased collective radiation dose. The inspectors used Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, to analyze the significance of the finding. The finding was preliminarily determined to be White (low to moderate safety significance) because it involved ALARA planning or work controls; the average collective dose at the time the finding was identified was greater than 135 person-rem; and the actual dose associated with a work activity was greater than 25 person-rem. Alternately, there were greater than four occurrences in which the actual collective dose exceeded 5 person-rem and the estimated/planned dose by more than 50 percent. The finai significance of this finding is to be determined. The finding had a crosscutting aspect in the area of problem identification and resolution, associated with the operating experience component, because the iicensee did not institutionalize operating experience concerning the effects of reactor coolant pump leakage on work area dose rates.
05000397/FIN-2014005-01	2014Q4	Columbia	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established for the selection and review for suitability of application o materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Contrary to the above, from April 2010 to November 4, 2014, the licensee failed to establish measures to review the suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee failed to review the suitability of parts for the safety related control room emergency filtration system resulting in non-qualified sealant and rivets used for the systems air handling units. The finding was of very low safety significance because the finding only represented a degradation of the radiological barrier function provided for the control room. This issue was entered into the licensees corrective action program as AR 316847, AR 317173 and AR 317184.
05000397/FIN-2014005-02	2014Q4	Columbia	Licensee-Identified Violation	Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established for the selection and review for suitability of application o materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Licensee procedure EES-5, General Fuse Selection Criteria and the Electrical Protection of 460 VAC and 125-250 VDC Motors, Revisions 0-7, is a procedure used to select and review the suitability of fuses in safety related applications. Contrary to the above, prior to November 4, 2014, the licensee failed to establish measures to review the suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems and components. Specifically, the licensee failed to review the suitability of fuses used in safety related applications because procedure EES-5 did not specify that the actual full load amperage rating (as opposed to nameplate rating) of components be considered when selecting fuses. The finding was of very low safety significance because it was a design or qualification deficiency that did not result in a loss of operability or functionality. This issue was entered into the licensees corrective action program as AR 317288, 314956 and 315170.
05000397/FIN-2015002-01	2015Q2	Columbia	Failure to Follow Procedures Associated with Operation of the Fuel Pool Cooling System	The inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow procedures associated with operation of the fuel pool cooling system. Specifically, on May 12, 2015, the licensee failed to follow operating procedures for the fuel pool cooling system resulting in a trip of the running fuel pool-cooling pump and subsequent lifting of a relief valve in the fuel pool cooling system. The standby fuel pool cooling pump automatically started to maintain fuel pool cooling. No significant change in refueling cavity level occurred since the plant was in the refueling mode of operation with the refueling cavity flooded approximately 23 feet above the reactor vessel flange. The licensee initiated Action Request 327593 to document the transient on the fuel pool cooling system and took immediate corrective action to disqualify the reactor operator pending remediation to address the human performance error. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the configuration control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using NRC Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined the finding was of very low safety significance because (1) it did not adversely affect decay heat removal capabilities from the spent fuel pool causing the pool temperature to exceed the maximum analyzed temperature limit specified in the site-specific licensing basis, (2) it did not result from fuel handling errors, dropped fuel assembly, dropped storage cask, or crane operations over the spent fuel pool, (3) it did not result in a loss of spent fuel pool water inventory decreasing below the minimum analyzed level limit specified in the sitespecific licensing basis, and (4) it did not involve spent fuel pool neutron absorber or a fuel bundle misplacement. This finding had a cross-cutting aspect in the area of human performance, avoid complacency, in that the reactor operator failed to consider potential undesired consequences of his actions before performing work and failed to implement appropriate error-reduction tools such as self and peer checking (H.12)
05000397/FIN-2015002-02	2015Q2	Columbia	Failure to Barricade a High Radiation Area	The inspectors identified a non-cited violation of Technical Specification 5.7.1.a for the failure to barricade a high radiation area. Specifically, the high radiation area entry gate in the 572-foot A RHR Heat Exchanger Room was found in the open position and access to the area was unimpeded. The licensee took immediate corrective action to restore the boundary gate to the closed position, impeding access to the high radiation area. This issue was documented in the licensees corrective action program as Action Request 328310. The failure to barricade a high radiation area was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation was of very low safety significance (Green) because: (1) it was not an as low as reasonably achievable (ALARA) 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 finding had an avoid complacency cross-cutting aspect, in the area of human performance, because radiation workers failed to obtain radiation protection support to reposition the high radiation area boundary and/or restore the entry gate to the closed position when the area was exited (H.12).

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