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05000263/FIN-2015008-042015Q3MonticelloLicensee-Identified ViolationTitle 10 CFR 72.158, Control of special processes, requires, in part, that licensees establish measures to ensure that special processes, including welding... and nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. Contrary to the above, the licensee identified on May 10, 2014, the licensee failed to perform verifications of a calibrated leak test instrument used on DSC lid to shell welds in accordance with Procedure TN 61BT-61BTH-HSMLD, Helium Leak Testing for NUHOMS Systems, Revision 1. Procedure TN 61BT-61BTHHSMLD, Revision 1, performs helium leak tests to demonstrate compliance with TS 1.2.4.a, 61BTH DSC Helium Leak rate of Inner Seal Weld. Additionally, contrary to 10 CFR 72.158, on April 2, 2014, the licensee failed to ensure enough filler material was deposited to achieve the minimum depth of the shell to outer top cover plate weld on DSC 16 in accordance with Procedure 12751-MNGP-OPS, Spent Fuel Cask Welding: 61BT/BTH NUHOMS Canisters, Revision 0. Instructions for welding operations are provided in Procedure 12751 MNGP-OPS, Revision 0, to ensure in field fabrication is performed in accordance with the Final Safety Analysis Report design basis drawings. During a nuclear oversight review of 2013 dry cask storage loading operations, the licensee identified that the helium mass spectrometer leak detection, calibrated leak instrument verification stabilizations, were not performed in accordance with TN 61BT-61BTH-HSMLD, Revision 1. Specifically TN 61BT-61BTH-HSMLD, Revision 1, Steps 8.3 and 8.4, require the user to determine the final instrument indicated leakage rate with the calibrated standard open and closed. The procedure step requires the user to ensure the system stabilizes while determining these results. TN 61BT-61BTH-HSMLD, Revision 1, Note 2, defines a stable signal as no more than a 1.0 E-8 std cm3/sec deviation in the indicated signal in 60 seconds. The licensee determined that for DSC 11, 12, 14, 15 and 16, stabilization times were less than 60 seconds. Specifically for DSC 12, stabilization times with the calibrated standard open were performed in 24 seconds, and stabilization times with the calibrated standard closed were performed in 22 seconds.
05000263/FIN-2015008-012015Q3MonticelloFailure to Perform Penetrant Tests in Accordance with Procedural RequirementsThe NRC staff identified an AV that is being processed through the traditional enforcement process because it appears to involve willfulness and is associated with an Independent Spent Fuel Storage Installation (ISFSI). The AV involves Title 10 of the Code of Federal Regulations (CFR) 72.158, Control of Special Processes, which required, in part, that the licensee establish measures to ensure that special processes, including nondestructive testing, are controlled and accomplished by qualified personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. The licensee established TriVis Procedure 12751 QP-9.202, Color Contrast Liquid Penetrant Examination Using the Solvent-Removable Method, Revision 1, as the qualified procedure for use in Dry Shielded Canister (DSC) NDE PT. However, from approximately September 5 to October 17, 2013, the NRC determined that licensee contractors apparently willfully failed to follow the TriVis procedure for developer dwell times, while performing PT on 66 of 66 DSC closure welds examined. The NRC also determined that the licensee contractors apparently failed to follow other parts of the TriVis procedure.
05000263/FIN-2015008-022015Q3MonticelloInaccurate and Incomplete Information Documented on VT/PT Report FormsThe NRC staff identified an AV that is being processed through the traditional enforcement process because it appears to involve willfulness, impacts the regulatory process, and is associated with an ISFSI. The AV involves 10 CFR 72.11, Completeness and accuracy of information, which required, in part, that information required by Commission regulations be maintained by the licensee to be complete and accurate in all material respects. However, from approximately September 5 to October 17, 2013, the NRC determined that licensee contractors apparently willfully completed PT examination forms, a quality assurance record, with inaccurate developer dwell times. The NRC also determined that the licensee contractors apparently completed PT examination forms, a quality assurance record, with other inaccurate information. This information was determined to be material to the NRC because it had the potential to mislead the NRC and the licensee as to the suitability for service of the DSCs.
05000263/FIN-2015008-032015Q3MonticelloFailure to Assess Contractor Control of QualityThe NRC staff identified an AV that is being processed through the traditional enforcement process because it appears to be associated with an ISFSI, which falls under traditional enforcement. The AV involves 10 CFR 72.154(c), Control of purchased material, equipment, and services which required, in part, that licensees assess the effectiveness of the control of quality by contractors and subcontractors at intervals consistent with the importance, complexity, and quantity of the product or services. However, from approximately September 5 to October 17, 2013, the NRC determined that the licensee apparently did not assess the effectiveness of the control of quality by contractors in that the licensee apparently did not monitor the work of contractors performing PT testing on DSCs Number 11 through #16.
05000461/FIN-2016008-012015Q1ClintonFailure to Perform and Adequate Equipment Apparent Cause EvaluationThe inspectors identified a finding of very-low safety significance (Green), and an associated Non-Cited Violation of Title 10, Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow Step 4.3.4 of procedure PI-AA-125, Corrective Action Program Procedure. Specifically, the licensee failed to perform Class B Equipment Apparent Cause Evaluation (EACE) 2381871, 1SX01PC Failed to Start for Testing, in accordance with PI-AA-125-1003, Apparent Cause Evaluation Manual, because they: (1) failed to analyze each causal factor to determine contributing causes as required by Step 4.4.1.2; and (2) failed to assign an effectiveness review for the EACE as required by Step 4.4.9.1. The licensee entered this finding into their Corrective Action Program and revised their EACE to: (1) include three contributing causes; (2) upgrade a corrective action to a corrective action to prevent recurrence; and (3) assign an effectiveness review to determine the effectiveness of the corrective action to prevent recurrence. The performance deficiency was determined to be more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, an effectiveness review is required to provide assurance that the Division 3 SX pump design change is successful in preventing recurrence of pump failure before another pump failure occurs, which would be a more significant safety concern. The finding impacted the Mitigating Systems Cornerstone and screened as having very-low safety significance (Green) because although the finding is a deficiency ultimately affecting the design or qualification of the Division 3 SX pump, the pump still maintains its operability. The inspectors determined this finding had an associated cross-cutting aspect in the area of Human Performance (Conservative Bias) because although a B Apparent Cause Evaluation may have been allowable for investigating the failure of the Division 3 SX pump, had an A Root Cause Analysis been performed, a more rigorous investigation process would have been used to identify contributing causes, assign corrective actions, and identify effectiveness reviews for the failure of the Division 3 SX pump.
05000461/FIN-2015001-032015Q1ClintonFailure of the Division 3 Shutdown Service Water Pump due to an inadequate Bushing DesignA self-revealed finding, preliminarily determined to be of low to moderate safety significance (White) and an associated apparent violation (AV) of 10 CFR Part 50 Appendix B, Criterion III, Design Control was identified for the failure to verify the suitability of the replacement pump design for the Division 3 Shutdown Service Water system. Specifically, the licensee failed to verify the design of the suction bell bushing for the replacement pump would pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water, resulting in the failure of the pump. This finding was self-revealed on September 16, 2014, during a surveillance test to ensure operability of the Division 3 shutdown cooling water pump, after the pump failed to start due to a damaged bushing rendering the pump inoperable. This finding does not represent an immediate safety concern because the licensee replaced the pump in September of 2014 with a pump of similar design in combination with additional monitoring of pump performance and provided adequate documentation that assures the pump will remain operable until a different design can be installed by June of 2016. The inspectors determined that the licensees failure to verify the suitability of the design for the Division 3 SX replacement pump for conditions under which it was to be used, as required by 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency. Specifically, the licensee failed to verify the design of the suction bell bushing for the replacement pump would pass sufficient cooling water flow to the pump internals without being affected by mud and silt from the lake water, resulting in the failure of the pump. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, 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. A Significance and Enforcement Review Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, dated June 19, 2012, preliminarily determined the finding to be of low to moderate safety significance (White). The performance deficiency associated with this finding did not reflect current licensee performance; therefore, no cross-cutting aspect was identified with this finding. (Section 4OA3)
05000461/FIN-2013002-012013Q1ClintonIncomplete ED Dose Rate Alarm EvaluationThe inspectors identified a finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 20.1501(a) for the failure to perform surveys to ensure compliance with 10 CFR 20.1201 shallow-dose equivalent (SDE) limits for five individuals during the fourth quarter 2011 due to contamination build-up on the workers gloves. This issue was entered into the licensees corrective action program as AR 01335298 and AR 01454976. Corrective actions include performing an apparent cause evaluation and performing dose assessments for the individuals involved. The performance deficiency was determined to be of more than minor safety significance in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Program And Process Attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that not performing an adequate SDE assessment affected the licensees ability to monitor, control, and limit radiation exposures. The inspectors also reviewed the guidance in IMC 0612, Appendix E, Examples of Minor Issues, and did not find any similar examples. In accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that the finding had very low safety significance because the finding did not involve: (1) ALARA planning and controls, (2) a radiological overexposure, (3) a substantial potential for an overexposure, or (4) a compromised ability to assess dose. The primary cause of this finding was related to the cross-cutting aspect of human performance with the component of work practices. The specific aspect was that the licensee ensures supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported.
05000254/FIN-2013002-012013Q1Quad CitiesFailure to Follow Clearance Order InstructionsA finding of very low safety significance and associated non-citied violation of Technical Specifications 5.4.1.a, Procedures, was self-revealed on March 13, 2013, when operators placing a clearance on the Unit 1 analog trip system de-energized the Unit 2 analog trip system resulting in a Unit 2 half-scram. The operators that opened the wrong breaker did not follow the instructions in the clearance order brief as required by OP-AA-109-101, Clearance and Tagging, and misidentified the inverter on the south wall of the cable spreading room as the Unit 1 analog trip system inverter when it was actually the Unit 2 inverter. The operators did not use the concurrent verification techniques specified in the pre-job briefing for ensuring that the inverter was the correct component to be manipulated, and did not implement the clearance order as written. Immediate actions taken were removal of the implementing operators qualifications and briefing to all operating personnel. Inspectors determined that the issue was more than minor because it adversely affected the Reactor Safety Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. The performance deficiency challenged the configuration control attribute of the objective for operating equipment lineups. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Appendix A, The Significance Determination Process For Findings At-Power. The inspectors answered all questions of Exhibit 1, Initiating Events Screening Questions, for transient initiators and support system initiators. Questions in both categories were answered No, and the finding screened as very low safety significance, or Green. Inspectors determined that a significant contributor to this finding was the failure of the operator performing breaker manipulation to verify the component label matched the clearance checklist and card in accordance with the site standard, HU-AA-101, Human Performance Tools and Verification Practices. As a result, inspectors identified that this issue had a cross-cutting aspect in the area of Human Performance - Work Practices for failure to use the human performance techniques to ensure that the work tasks are performed safely and individuals do not proceed in the face of uncertainty
05000461/FIN-2013002-022013Q1ClintonFailure to Perform Adequate Past Operability EvaluationThe inspectors identified a finding of very low safety significance associated with the licensees failure to correctly evaluate the past operability of safety-related motor operated valve 1E51-F031, reactor core isolation cooling system suppression pool suction valve, which failed quarterly surveillance testing on October 29, 2012. No violation of regulatory requirements was identified. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the past operability evaluation. The finding was of more than minor significance since the failure to correctly evaluate a degraded/nonconforming condition potentially affecting the operability of structures, systems, and components (SSC) required to be operable by Technical Specifications (TS) would become a more significant safety concern, if left uncorrected, because it could reasonably result in an unrecognized condition of an SSC failing to fulfill a safety-related function. The finding was a licensee performance deficiency of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee failed to use conservative assumptions in decision making while evaluating past operability of the valve by assuming that the time of inoperability was the same as the time of discovery for a time dependent failure mechanism (i.e., hardened grease) since no firm evidence to support operability was obtained by testing.
05000254/FIN-2013002-022013Q1Quad CitiesDiesel Generator Cooling Water Pump Aligned to Wrong UnitA finding of very low safety significance and associated non-citied violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on March 1, 2013, during restoration from the 1B core spray logic test, when the 1/2 diesel generator cooling water pump (DGCWP) was discovered to have been lined up to Unit 2 emergency core cooling system room coolers instead of Unit 1 coolers as expected. The operators that had performed the initial valve manipulations on February 28, 2013, did not complete the alignment as required by QCOP 6600-15, 1/2 Diesel Generator Cooling Water Pump Cross Connect Alignment. Specifically, the operators executing QCOP 6600-15 did not follow the procedure for aligning the Unit 1/2 DGCWP to supply the Unit 1 emergency core cooling system room coolers. The issue was entered into the licensees CAP as Issue Report 1486754, and the licensee restored operability of the Unit 1 DGCW pump to restore compliance. Standdown briefings were conducted for all station operators to discuss the event lesson learned, and performance management actions were implemented for the operators involved in the event. This issue was more than minor because it adversely affected the Reactor Safety Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences in that failure to align cooling water per the procedure adversely impacted the cornerstone attribute of Configuration Control for operating plant equipment lineups. Specifically, the as-left equipment lineup was different than that reported to the main control room when the activity was completed. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609 Appendix A, The Significance Determination Process For Findings At-Power. The inspectors answered all questions of Exhibit 2, Mitigating Systems Screening Questions, Section A - Mitigating SSCs and Functionality (Except Reactivity Control Systems) No, and therefore, the finding screened as Green or very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance - Work Practices because the licensee personnel did not use human performance tools and techniques to ensure proper execution of the task
05000461/FIN-2013002-032013Q1ClintonFailure to Perform Adequate MOV Preventive Maintenance Resulted in Inoperable RCIC SystemA finding of very low safety significance with an associated Non-Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings was self-revealed when safety-related motor operated valve 1E51-F031, reactor core isolation cooling (RCIC) system suppression pool suction valve, failed to fully close during surveillance testing on October 29, 2012. The valve failure occurred due to the licensees failure to establish an adequate procedure to perform preventive maintenance on it. Specifically, the maintenance procedure did not contain a requirement to stroke a motor operated valve during the performance of periodic stem lubrication activities. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the maintenance procedure. The finding was of more than minor significance since it was associated with the Procedure Quality attribute and adversely affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the valve failure rendered the RCIC system inoperable. This finding is of very low safety significance because it: (1) was not a design or qualification deficiency; (2) did not represent an actual loss of function of a system; (3) did not represent an actual loss of function of a single train or two separate trains for greater than its Technical Specification (TS) allowed outage time; (4) did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant; and (5) did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating event The inspectors concluded that this finding affected the cross-cutting area of human performance since adequate licensee resources involving personnel and procedures did not support successful human performance. Specifically, the maintenance procedure was not appropriate to the circumstances because it did not contain adequate instructions to ensure that motor operated valve stems were adequately lubricated.
05000305/FIN-2013007-022013Q1KewauneeFailure to Perform Announced Fire Drills In Accordance With Fire Protection Program PlanAs discussed in the factual summary in Enclosure 1, the NRC staff identified an AV that is being processed through the traditional enforcement process because it involves willfulness. The technical aspects of the AV were evaluated through the SDP as described in IMC 0612. The apparent violation involves License Condition 2.C.(3), Fire Protection, which requires, in part, that the licensee shall implement and maintain in effect all provisions of the approved Fire Protection Program as described in the licensees Fire Protection Program Plan, and as referenced in the Updated Safety Analysis Report. Specifically, from at least August 2009 through December 2011, the licensee failed to perform the required quarterly announced fire drills in that the announced fire drills were instead conducted as training sessions. Furthermore, the licensees staff also failed to comply with a procedural requirement to have five members of a brigade team present during an announced drill, conducted as a training session, on four occasions. The AV is discussed in Enclosure 1 and in the Enforcement portion of Section 1R05A.
05000254/FIN-2013002-032013Q1Quad CitiesLicensee-Identified ViolationTitle 10 CFR 50, Appendix B, Criterion V states, in part, that activities affecting quality shall be prescribed by documented instructions and procedures appropriate to the circumstances and shall be accomplished in accordance with these instructions and procedures. Contrary to the above, on November 5, 2012 and November 7, 2012, the licensee failed to incorporate ASME Code required VT-2 examination requirements into work instructions for work to be performed on a safety-related component. Specifically, WO 1453938-30 and WO 1453938-36, which required cutting and welding activities to be performed on the RHRSW system during installation of branched sub-assemblies, did not contain EC 385181 CC-AA-103, Attachment C. The EC 385181 CC-AA-103, Attachment C is the document that contained the requirement to perform a VT-2 examination immediately after sub-assembly installation as required by the ASME Code Section XI. On November 12, 2012, Project Management initiated IR 1439104, VT-2 Inspections Not Performed. Issue Report 1439104 identified that the required VT-2 had not been performed on the installed sub-assemblies under WOs 1453938-30 and 145398-36. This issue is more than minor because the performance deficiency, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, not performing the VT-2 system pressure/leakage test following hot tapping the RHRSW line would call into question the quality of welds created during hot tapping of the RHRSW line and could potentially undermine the integrity of the RCS pressure boundary because that RHRSW line was in operation during the hot tapping process. This finding was determined to be of very low safety significance or Green in accordance with IMC 0609, Appendix G because at least one RHR loop (train A) was operable during the time period from when the hot tapping was performed to when the ASME Code required VT-2 was finally performed on the RHRSW Train B line. In addition, the sub-assemblies weld locations off the RHRSW train B line did not exhibit any system leakage during the VT-2 examination that was conducted on November 12, 2012.
05000461/FIN-2013002-042013Q1ClintonUN-FUSED DC Ammeter Circuits Results in Unanalyzed ConditionClinton Power Station Unit 1 Operating Licensee Condition 2.F required, in part, that the licensee implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Final Safety Analysis Report, as amended, and as approved in the Safety Evaluation Report (NUREG-0853), dated February 1982 and Supplemented Numbers 1 through 8. The Clinton Power Station UFSAR, Appendix F, Safe Shutdown Analysis, Section 1.5, Associated Circuits, stated, in part, that at Clinton Power Station, there were no associated non-safe shutdown cables that were not electrically protected and shared a common enclosure with safe shutdown cables. Contrary to the above, on December 8, 2011, the licensee identified that non-safe shutdown cables that shared a common enclosure with safe shutdown cables were not electrically protected. Specifically, the licensee identified that the battery ammeter circuits routed from the DC motor control centers to the ammeters located in the Control Room were not fused. These cables were routed in trays and installed in panels with other safe shutdown cables. During a fire event in the Control Room, fire-induced failures could have damaged the ammeter circuit and could have resulted in damaging other safe shutdown cables that are in direct physical contact with these cables in different fire zones. The licensee completed a fire risk evaluation for worst case fire scenarios in the Control Room and concluded that the delta core damage frequency (ACDF) from a fire due to the un-fused ammeter circuit was 8.26E-8 per year. The evaluation also concluded that for all postulated fire scenarios in the Control Room; at least one safe shutdown division would not be impacted by the fire. The inspectors reviewed the licensees risk evaluation and based on the small calculated ACDF and at least one safe shutdown train would not be affected by all postulated fire scenarios in the Control Room, the inspectors determined that the issue was of very low safety-significance. The licensee entered this violation into its corrective action program as AR 01299460. The licensee submitted LER 05000461/2011-007-00 to report this issue. Refer to Section 4OA3.2 of this inspection report for the review and closure of the LER.
05000331/FIN-2013007-012013Q1Duane ArnoldFuel Pool Radiation Monitor Corrective ActionsThe inspectors identified a finding of very low safety significance (Green) and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for two examples of the failure to follow procedures associated to the troubleshooting and repair of RIS4131A Refuel Floor Exhaust Radiation Monitor. The licensee initiated WO 40190702-01, RIS4131A Refuel Floor Exhaust Rad Mon Upscale and Group 3, to troubleshoot and repair the power supplies. The licensee was still evaluating planned corrective actions for the failure to follow the work order instructions. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, RIS4131A failing to operate in conjunction with a single additional failure (RIS4131B) could allow the release of radioactive contamination due to preventing an automatic secondary containment isolation (Group 3). The finding screened as having very low safety significance (Green) because the inspectors answered Yes to question C.1 of IMC 0609, Appendix A, Exhibit 3. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Work Control components that support long-term equipment reliability by performing maintenance that is more preventive than reactive.
05000254/FIN-2012005-012012Q4Quad CitiesDiesel Generator Technical Specification Frequency and Voltage Variation Not Considered in Loading CalculationsThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify and ensure that operating the emergency diesel generators (EDGs) at the limits of voltage and frequency, allowed by Technical Specification (TS) 3.8.1.2, would not affect the safety-related components. Specifically, the license failed to ensure the EDGs, operating under any combination of allowed voltage and frequency, would not be loaded in excess of the licensed limit and would not cause supplied components to become inoperable. The licensee entered the issue into the corrective action program (CAP) as Issue Report (IR) 01288784, CDBI Technical Specification Limits for EDG, and restricted EDG operation to near the midpoint of the allowed TS range during any potential event until the licensee demonstrates operability over the full TS range. The finding was more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design control attribute was adversely affected because the licensee failed to ensure the TS- allowed operating band for EDG frequency and voltage could not affect the operability and reliability of mitigating system components. Based on a Phase 3 internal events SDP evaluation performed by a regional senior reactor analyst, the inspectors determined the finding was of very low safety significance (Green). No cross-cutting aspect was assigned since the analysis was last performed in May of 2007 and is not necessarily reflective of current performance.
05000254/FIN-2012005-022012Q4Quad CitiesFailure to Follow Surveillance ProcedureA self-revealed finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedure, and Drawings, were identified on October 25, 2012, when the operator performing the Unit 2 EDG surveillance test failed to follow procedural direction when applying load to the machine resulting in the Unit 2 diesel generator being inoperable for approximately seven hours while troubleshooting activities were conducted. The operator did not perform the diesel loading in accordance with the procedure in that real load was applied in a manner that changed reactive load significantly in the opposite polarity from real load and resulted in a loss of field trip of the diesel generator output breaker. After troubleshooting, the surveillance was completed to ensure no impact to the voltage regulating circuit and restore operability for prior work activities. This issue was entered into the licensees CAP as IR 1431240. Immediate corrective actions included revision of procedures that operated the diesel generator in parallel with another source to include information reminding operators that the Unit 2 EDG responded differently to load adjustments, and care should be used when making adjustments to prevent a loss of field trip. The finding was more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The deficiency impacted the Equipment Performance attribute for reliability in that the performance deficiency challenged the voltage regulator protective feature and could have damaged the excitation circuit for the diesel generator. Inspectors performed the Phase 1 screening of the finding using the SDP and determined that the issue was of very low safety significance, or Green. The questions in IMC 0609, Attachment 4, Appendix A, Exhibit 2, Section A were answered No by inspectors because the diesel was quickly made available for emergency response following the breaker trip, and the remaining diesel generator and both offsite power sources were operable. Inspectors determined this finding to be cross-cutting in Human Performance-Resources in that the licensee ensures that appropriate training is provided to assure nuclear safety (H.2(b)) because a contributor to this finding was that a post-maintenance change in voltage regulator performance was not systematically communicated to the operating staff through training.
05000254/FIN-2012005-032012Q4Quad CitiesBoth Unit 1 Core Spray Subsystems InoperableA self-revealed finding of very low safety significance (Green) and an associated NCV of TS 3.5.1.K were identified for two core spray systems inoperable due to degraded flood barriers on August 6, 2012. The failure of the 1B core spray and Unit 2 reactor core isolation cooling/2B core spray floor drain ball valves was caused by wear related degradation that occurred at the valve-to-actuator coupling that allowed the valve to not be fully seated despite the actuator indicating fully closed. Since the surveillance tested the floor drain ball valves in the as-found condition, the condition existed prior to discovery. Therefore, both Unit 1 core spray subsystems were inoperable due to degraded flood barriers. This condition would have required immediate entry into Limiting Condition for Operation 3.0.3 to commence a shutdown within 1 hour. This issue was entered into the licensees CAP as IR 1397306. Corrective actions for this issue included repairs to the floor drain ball valves, extent of condition inspection of all reactor building floor drain ball valves and shortening the surveillance interval from 4 years to 2 years. The finding was more than minor because it affected the Mitigating Systems Cornerstone objective to ensure the availability of systems to respond to initiating events to prevent undesirable consequences. In this case, the Cornerstone attribute of protection against external factors (internal flood) was impacted. The inspectors performed an SDP Phase 1 screening for the finding using IMC 0609, Attachment 04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and answered the first four questions No. Therefore, the finding screened as very low safety significance, or Green. The inspectors identified that this issue had a cross-cutting aspect in the area of Problem Identification and Resolution - Identification (P.1(a)). A contributor to this finding was that the Operations and Engineering Departments were aware that the reach rod operators for the floor drain ball valves were difficult to operate. However, an issue report was never entered into the corrective action program to make the organization aware of this issue, assess for proper operation, trend the valve performance, identify potential failure mechanisms, or document conclusions.
05000254/FIN-2012005-042012Q4Quad CitiesLicensee-Identified ViolationA licensee-identified finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified on November 23, 2012, when the licensee identified that a temporary alteration to the plant had been in place for longer than allowed by procedure without a 50.59 review. During troubleshooting under WO 1527623 on April 30, 2012, the licensee changed the operating set point for differential pressure controller 1-5741-8557. The set-point change was part of an activity to improve responsiveness of the reactor building ventilation system to changes in differential pressure. Troubleshooting continued until June 1, 2012, when the active troubleshooting stopped, but the work activity remained active while other work was performed that impacted reactor building ventilation. On November 23, 2012, the licensee identified that the temporary alteration had been in place for more than 90 days with no 50.59 review having been performed as required by Step 4.1.3 of MA-AA-716-004, Control of Troubleshooting. The finding is more than minor because it adversely affected the design control attribute of the Barrier Integrity Cornerstone objective to provide assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Failure to adhere to process controls for design control during plant modifications or alterations could result in a more significant challenge to plant barriers. The inspectors performed an SDP Phase 1 screening for the finding using IMC 0609, Attachment 04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, and answered the Reactor Containment questions No. Therefore, the finding screened as very low safety significance, or Green. Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings requires in part that activities affecting quality be accomplished in accordance with written procedures. MA-AA-716-004 governs troubleshooting activities and states in part in Step 4.1.3, 10 CFR 50.59 review is also required if a temporary alteration to the facility is established in direct support of the troubleshooting, and the temporary alteration will be installed for 90 days or greater while at power. Contrary to the above, the temporary alteration was in place for more than 90 days without a 50.59 review being performed. The licensee entered the issue into the corrective action program as IR 1430938 and performed the required 50.59 review. Additional reviews were performed to provide assurance that all other changes were being tracked and reviewed as required by procedure.
05000254/FIN-2012005-052012Q4Quad CitiesLicensee-Identified ViolationA licensee-identified finding of very low safety significance (Green) and associated NCV of TS 5.4.1.a was identified on May 1, 2012, when the licensee identified that a preventative maintenance (PM) task to adjust/repair/replace door latches was not completed within the specified timeframe. This yearly PM for the Unit 2 HRSS door was last completed in January 2010. The work package documentation was not closed so no subsequent task was scheduled. Failure to properly close the task and reschedule the PM was a finding. When the licensee identified the deficiency, a request was initiated to defer the two missed PMs and re-schedule the next PM for January 2013. While the extension was performed in accordance with the procedure and potential consequences of failure of the door (i.e. maintenance rule functional failure) were included in the evaluation, no evaluation of the material condition of the door or latch was performed before the PM was extended. On September 6, 2012, the worker entering the interlock recognized that the Unit 2 HRSS door opened as he entered the airlock and took action to shut the door and notify the control room. A review of the alarm history determined that the Unit 2 HRSS was open for 8 seconds. The finding was more than minor because it adversely affected the SSC and Barrier Performance attribute of the Barrier Integrity Cornerstone objective to provide assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The process for deferring PMs was changed in August 2012. Inspectors verified that the revised process would require an assessment of work history, component performance, and a technical justification when a maintenance rule function failure is identified as a potential consequence because the process would dictate that maintenance rule components have a high consequence failure. Given these changes to the service request program the licensee would have a different classification and prioritization for a similar issue undergoing deferral. Technical Specifications 5.4.1.a requires that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Quality Assurance Program Requirements (Operation). Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance, states in part that maintenance that can affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with written procedures, or documented instructions appropriate to the circumstances. Contrary to the above, in January 2010, the licensee failed to implement MA-AA-716-011, Work Execution and Closeout, because a work task was left open which prevented the PM task from being rescheduled.
05000331/FIN-2012005-052012Q4Duane ArnoldLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion XII, Control of Measuring and Test Equipment, on October 18, 2012, following a weekly audit of Measuring and Test Equipment (M&TE) storage areas. The licensee identified 16 M&TE items that were either missing or not properly checked out from tool cribs in accordance with Administrative Control Procedure (ACP) 1408.8, Control of Measuring and Test Equipment. These concerns were identified following a recent revision to ACP 1408.8 to improve the M&TE checkout process due, in part, as a response to a licensee-identified NCV in NRC Inspection Report 05000331/2012004; and, recent site-wide communications to reinforce expectations for M&TE controls. The licensee performed an Apparent Cause Evaluation (CR 01814517) to determine why prior corrective actions to ensure proper control of M&TE were not effective. Corrective actions included establishing more robust barriers to the M&TE tool cribs to ensure M&TE would be checked out properly. The failure to establish measures to assure controls of M&TE was a performance deficiency. The performance deficiency was determined to be more than minor and a finding because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, failing to properly control M&TE would have the potential to impact the quality of maintenance or the results of testing of safety-related equipment. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and then proceeded to IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2 Questions. Since the inspectors answered No to Question 4, the finding screened as very low safety significance (Green).
05000331/FIN-2012005-042012Q4Duane ArnoldFailure to Accomplish Safety/Relief Valve Test InstructionsA finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on October 24, 2012, for the licensees failure to accomplish instructions for functional testing of the main steam line safety/relief valve PSV-4402 pilot valve. Specifically, a model work order to perform testing of the pilot valve required the main steam lines to be drained; however, the decision was made to allow performance of the testing following removal of the main steam line plugs. Due to a minor leak of the closed safety/relief valve nitrogen accumulator isolation valve, the testing and the resultant brief opening of the pilot valves solenoid valve caused nitrogen to reposition the pilot valve disc of the safety/relief valve. This then resulted in momentary opening of PSV-4402 and discharge of approximately 700 gallons of reactor cavity water into the drained suppression pool. The licensee entered the issue into the Corrective Action Program (CAP) as Condition Report (CR) 01816385. The licensee revised the model work orders for safety/relief valve pilot valve functional testing and was in the process of creating separate return-toservice tasks to ensure that testing of the pilot valves could not be performed unless the main steam lines were drained. The inspectors determined that testing of PSV-4402 without the main steam line plugs installed represented a performance deficiency because it was the result of the licensees failure to meet a regulatory requirement, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented. The performance deficiency was determined to be more than minor and a finding because it was associated with the Initiating Events Cornerstone attributes of configuration control and human performance and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability during shutdown operations. The inspectors applied IMC 0609.04, Initial Characterization of Findings, to this finding. Because the finding pertained to an event while the plant was shutdown, Table 3 instructed reference of IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, and IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs. Because all attributes IMC 0609, Appendix G, Attachment 1, Checklist 7 BWR Refueling Operation with Reactor Coolant System (RCS) Level > 23, were met throughout the event, the finding did not require a quantitative analysis and screened as very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Decision-Making components, and involving the licensee using conservative assumptions in decision making and adopting a requirement to demonstrate that a proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove an action.
05000331/FIN-2012005-032012Q4Duane ArnoldFailure to Make Surveys to Evaluate the Potential Radiological Hazards in the TorusA finding of very low safety significance and associated NCV of 10 CFR 20.1501 was self-revealed on October 16, 2012, for the licensees failure to make surveys to evaluate the potential radiological hazards during work inside the torus. Specifically, ten workers were externally contaminated and nine workers were assigned low level unintended internal radiation doses after installing rigging and fall protection inside the torus proper. The issues were entered into the licensees CAP as CR 01813761. Immediate corrective actions included performance of radiological dose assessments on the individuals involved and performance management coaching of the individuals in accordance with station management protocols. The inspectors determined that failing to effectively maintain radiological control of work activities in the torus proper represented a performance deficiency because it was the result of the licensees failure to meet a regulatory requirement, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented. The performance deficiency was determined to be more than minor and a finding because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern (additional unplanned or more significant radiological exposures). The inspectors applied IMC 0609.04, Initial Characterization of Findings, to this finding. Per Table 3, because the finding was associated with a programmatic weakness in the licensees Occupational Radiation Safety Cornerstone, IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, was used. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not involve As-Low-As-Is-Reasonably- Achievable (ALARA) planning or work controls, there was no overexposure or substantial potential for an overexposure, nor was the licensee\'s ability to assess worker dose compromised. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Work Practices components, and involving the licensee defining and effectively communicating expectations regarding procedural compliance and personnel follow procedures.
05000331/FIN-2012005-022012Q4Duane ArnoldDegraded/Non-Conforming Conditions Not Properly EvaluatedA finding of very low safety significance and associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors on October 17, 2012, for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments, when degraded or non-conforming conditions were identified. Specifically, the duty Shift Manager approved Prompt Operability Determination (POD) 01812339 that was performed following the identification of submerged power and control cables associated with the A Standby Diesel Generator (SBDG). The POD did not discuss the fact that the applicable cables were not qualified for submergence, incorrectly concluded that the cables conformed to the Updated Final Safety Analysis Report (UFSAR), and did not evaluate whether compensatory actions were required. The POD conclusions were contrary to the requirements of procedure EN-AA-203-1001 that required all degraded or non-conforming conditions be evaluated for compensatory actions. The licensee documented the inspectors concerns in CR 01813800, revised POD 01812339, and assigned compensatory actions for the degraded and non-conforming conditions. The inspectors determined that failing to evaluate a degraded or non-conforming condition for compensatory actions represented a performance deficiency because it was the result of the licensees failure to meet a regulatory requirement, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented. The performance deficiency was determined to be more than minor and a finding because, if left uncorrected, failing to properly assess the operability of degraded or non-conforming conditions and evaluating the necessity for compensatory actions would have the potential to lead to a more significant safety concern. The inspectors applied IMC 0609.04, Initial Characterization of Findings, to this finding. Because the finding pertained to an event while the plant was shutdown, Table 3 instructed reference of IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. Because the finding did not require a quantitative assessment, the finding screened as very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Problem Identification and Resolution, having Operating Experience (OE) components, and involving the licensee implementing and institutionalizing OE through changes to station processes and procedures.
05000331/FIN-2012005-012012Q4Duane ArnoldLack of Procedure Leads to Over Filling Condensate Storage Overflow TankA finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1, Procedures, was identified by the inspectors for the licensees failure to establish a procedure for filling the condensate storage tanks (CSTs) from multiple sources. Specifically, the lack of procedural instructions or guidance for controlling the CST filling process resulted in over filling the CST overflow tank on October 8, 2012, and subsequent leakage past a missing CST pit penetration seal to the nearby soil. The licensee entered the inspectors concerns into the CAP as CR 01812345. The licensee repaired the penetration seal and revised the applicable Annunciator Response Procedures and Operating Instructions. The inspectors determined that failing to establish a written procedure for filling the CSTs represented a performance deficiency because it was the result of the licensees failure to meet a TS requirement, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented. The performance deficiency was determined to be more than minor and a finding because it was associated with the Public Radiation Safety Cornerstone attribute of programs and processes and adversely affected the cornerstone objective of ensuring the adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of civilian nuclear reactor operation. The inspectors applied IMC 0609.04, Initial Characterization of Findings, to this finding. Because the finding and associated programmatic weakness was in the licensees Public Radiation Safety Cornerstone, Table 3 instructed reference of IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process. Because the finding was related to the effluent release program, did not constitute a substantial failure to implement the effluent program, and did not result in any public dose, the finding screened as very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the crosscutting aspect of Human Performance, having Work Control components, and involving the licensee appropriately planning the work activity by incorporating the need for planned contingencies, compensatory actions, and abort criteria.
05000254/FIN-2012007-012012Q3Quad CitiesCAPR Not CompletedA finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program was identified by the inspectors when they determined that a licensee-specified corrective action to prevent recurrence (CAPR) of a significant event was not completed as required by a quality assurance program implementing procedure, LS-AA-125, Corrective Action Program (CAP) Procedure. Inspectors determined that the failure to complete the CAPR and install auxiliary contactors that had undergone enhanced testing (designated PQI testing in the licensees documentation) before installation was a performance deficiency entered into the licensees CAP as IR 1409378. Immediate corrective actions included performing a functional evaluation of installed components and quarantine of remaining spare parts. This finding was more than minor because the CAPR established criteria that should have prevented installation of the parts until testing was performed, but the parts were installed in the plant and the components were returned to service, thus impacting the reactor safety, initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Inspectors performed a SDP Phase 1 screening using IMC 0609 Attachment 4 and Appendix A Exhibit 1, Initiating Events Screening Questions, and answered all of the questions, No. Therefore, the finding screened as very low safety significance or Green. The inspectors identified that this finding has a cross-cutting aspect in the area of Human Performance Work Practices, in that, licensee personnel did not follow procedures (H.4(b)). Inspectors determined that the primary contributor to this finding was that procurement personnel did not follow procedure SM-AC-3019, Parts Quality Process, which stated in Attachment 6 that the station shall inform the test facility of any unique or special test requirements for the equipment. Otherwise, Exelon PowerLabs will apply standard PQI testing criteria for the item. Procurement personnel did not identify the enhanced PQI testing requirement to PowerLabs when the part was sent for testing.
05000237/FIN-2012007-012012Q2DresdenFailure to Identify and Remove Diesel Fire Pump Battery Terminal CorrosionThe inspectors identified a finding of very low safety significance (Green) and associated NCV of Technical Specifications for the licensees failure to adequately implement the diesel fire pump (DFP) battery surveillance procedure. Specifically, the licensee failed to identify and remove corrosion on the DFP battery terminals, which was contrary to the surveillance procedure that implemented the fire protection program. A similar NCV was previously cited by the NRC on October 17, 2011, and documented in inspection report 05000237/2011008; 05000249/2011008, Failure to Identify Diesel Fire Pump Battery Terminal Corrosion. The licensee entered the issue into their corrective action program and planned to clean the battery terminals. In addition, the licensee planned to replace the 2/3 DFP batteries in July 2012. The inspectors determined that the finding was more than minor because, if left uncorrected, the presence of corrosion in conjunction with identified voltage issues for two battery cells could affect the reliability of the diesel fire pump. This finding was of very low safety significance because the DFP had started as part of a recent routine surveillance. This finding has a cross-cutting aspect in the area of PI&R because the licensee failed to identify the battery corrosion accurately and in a timely manner commensurate with their safety significance.
05000254/FIN-2012003-012012Q2Quad CitiesBus Energized With Grounding Device InstalledA finding of very low safety significance with an associated NCV of TS 5.4.1.a, Procedures, was self-revealed on March 24, 2012, when operators energized an electrical bus in the switchyard with a grounding device still installed on that bus. OP-AA-109-101, Clearance and Tagging, and have operations place a danger tag on a grounding strap installed on 345 kV Bus 9 resulted in a significant voltage perturbation and operating transient on Unit 1. The licensee entered the issue in the CAP as IR 1345302 and immediate actions included clearing the fault and restoring plant equipment. Individual qualifications were removed for parties involved in the event, and a root cause evaluation was performed. The finding was determined to be more than minor because it impacted the Human Performance attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the Human Performance attribute was challenged because the human error resulted in a voltage transient that produced an operational transient on Unit 1 and could have resulted in a more severe challenge to both units. The inspectors performed a SDP Phase 1 screening for the finding using IMC 0609, Table 4a, for the Initiation Events Transient Initiators and determined that the finding did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions will not be available. The duration of the event, separation of divisional and emergency power supplies, and redundancy of equipment supplying safety functions were considered for this determination. Therefore, the finding screened as Green, or very low safety significance. The inspectors identified that this finding has a cross-cutting aspect in the area of Human Performance - Decision Making because both the station supervisor overseeing the electrical bus realignment and the clearance holder took action based on non-conservative assumptions that could easily have been validated before placing the electrical system at risk.
05000254/FIN-2012003-022012Q2Quad CitiesProcedure Noncompliance Impacting Reliability of HPCIAn NRC-identified finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criterion V, Procedures, was identified on March 27, 2012, when station employees did not adhere to station seismic controls while performing Unit 2 outage work. Failure to implement the requirements of the seismic procedure was a performance deficiency. In placing the stacked Unit 2 high pressure coolant injection turbine steam chest too close to the Unit 1 high pressure coolant injection pump and piping, technicians circumvented administrative controls in place to prevent unrestrained equipment from impacting safety-related equipment during a seismic event. The licensee subsequently secured the assembly per the procedure. This issue was entered into the licensees corrective action program as IR 1358458. The finding was determined to be more than minor because it adversely affected the equipment reliability attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems responding to initiating events to prevent undesirable consequences (i.e., core damage). This finding was assessed using the Phase 1 screening worksheets of IMC 0609 and determined to be of very low safety significance (Green). The finding did not result in an actual loss of safety function of a single train for greater than the TS allowed outage time. The finding did not involve a total loss of any safety function, as identified by the licensee through a Probabilistic Risk Assessment, Individual Plan Examination of External Events, or similar analysis, contributing to external event-initiated core damage accident sequences (i.e., initiated by a seismic, flooding, or severe weather event). The inspectors identified a cross-cutting aspect in the area of Human Performance - Resources because the licensee did not ensure that the work package included sufficient information to ensure that the cribbing used for the activity met the requirements specified by engineering in the analyzed load movement plan.
05000254/FIN-2012003-032012Q2Quad CitiesFailure to Identify Design Deficiency in Vendor ProductA self-revealed finding of very low safety significance with an associated NCV of Technical Specification (TS) 3.7.7, Main Turbine Bypass Valves System, was identified on April 18, 2012, when an unplanned reactor scram occurred during generator voltage regulator testing. Inspectors subsequently determined the licensee had failed to identify elimination of a time delay that changed how the system responded to a load reject with no turbine trip during vendor design documentation review for the digital electro-hydraulic control (DEHC) system modification implemented in 2006. Failure to perform the review with the rigor required by CC-AA-103-1003, Owners Acceptance Review of External Engineering Technical Products, is a performance deficiency entered into the licensees corrective action program (CAP) as Issue Report (IR) 1355763. This finding resulted in exceeding the allowed out-of-service time for TS 3.7.7, Main Turbine Bypass System, on at least eleven occasions between the two units since the modifications were installed. The finding was determined to be more than minor because the performance deficiency adversely affected the Reactor Safety - Initiating Events Cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions. In this circumstance, the Design Control attribute of the cornerstone was adversely impacted when unintended consequences were introduced during a modification. Using IMC 0609, Attachment 4, Table 4a, Initiating Events Cornerstone, Transient Initiators, inspectors determined that the performance deficiency did not contribute to the likelihood of both a reactor trip and unavailability of mitigation equipment since the main steam safety and relief valves are the credited pressure mitigation equipment and were unaffected by the event. Therefore, this finding screens as Green, or very low safety significance. The inspectors did not identify a cross-cutting aspect for this performance deficiency since it occurred during the DEHC modification review in 2006 and was considered a legacy issue.
05000254/FIN-2012003-042012Q2Quad CitiesLicensee-Identified ViolationThe licensee identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for unacceptable preconditioning of high pressure coolant injection (HPCI) valves due to performing a safety-related PM prior to prior the quarterly in-service testing of the same valves. A maintenance delay, which caused a shift in the scheduled performance of the quarterly testing of the HPCI system air operated. The licensee identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for unacceptable preconditioning of high pressure coolant injection (HPCI) valves due to performing a safety-related PM prior to prior the quarterly in-service testing of the same valves. A maintenance delay, which caused a shift in the scheduled performance of the quarterly testing of the HPCI system air operated
05000461/FIN-2012002-012012Q1ClintonEvaluation of Apparent Nonconforming Condition Affecting Circulating Water Pump Auto Stop Feature During a Flooding EventThe inspectors opened an Unresolved Item (URI) pending evaluation of an issue that affects the function of a flood protection design feature intended to automatically stop running CW pumps in the event of flooding in the Turbine Building Condenser Pit. As described in the UFSAR (Sections D3.6.4 and 10.4.5.5), each condenser cavity designed to contain flooding to Elevation 715 is equipped with a redundant system of level switches, which will alarm in the Control Room if the water level in the condenser cavity reaches an elevation of more than 1 foot (Elevation 710) above the condenser cavity floor at Elevation 709. These level switches will close a motor-operated valve in the floor drain piping between the condenser cavity and the Turbine Building floor drain sump to slow flooding of the Turbine Building. Isolating the condenser pit from the Turbine Building floor drain sump slows early flooding of the Turbine Building basement. A second system of redundant level switches will automatically stop the CW pumps if the flood water reaches an elevation of 714 within the condenser cavity. An additional foot, from Elevation 714 to Elevation 715 remains to contain the water flow due to coast down of the CW pumps after they are initially shut off. Neither set of level switches are safety related; however, they are important internal flooding mitigation features as described in the UFSAR. During review of Action Requests (ARs) 01192988 and 01197763, the inspectors noted that operators discussed a known design issue affecting the function of the CW pump auto stop feature. If CW Pump A has power removed then all CW pump tripping protection is lost during a flooding event. With power removed from CW Pump A, CW Pumps B and C would not auto stop during a flooding event because the tripping power comes through the CW Pump A tripping fuses. This appeared to the inspectors to be a nonconformance with the UFSAR description under the above specified circumstance. This design issue has apparently been known by operators for many years, yet no engineering evaluation was performed for the condition and no corrective action other than a procedure enhancement has been initiated by the licensee to address it. As stated in AR 01197763: Another lesson learned was to plan ahead for off service CW pumps and that if possible, do not have A CW Pump tagged out due to this removes the CW pump trip on high-high CW Pit level. The procedure for isolating a waterbox requires tagging out the off-service CW pump to preclude inadvertent start. Having this trip defeated while performing an evolution with the potential to cause CW Pit flooding is contrary to good industrial safety practices. The inspectors reviewed CPS 3113.01, Circulating Water (CW), Revision 37e, and noted that Step 4.11 described the condition. The High-High level (~714) in the Condenser Pit CW Pump Trip Circuitry is electrically powered by 1CW01PA DC (direct current) control power. If the A pump is de-energized, then no CW pump tripping will be received from condenser pit flooding. The 710 floor drain sump floats which give Main Control Room alarm and 1TF013 valve closure would be the only indication of flooding. The inspectors discussed this issue with the licensee and in response to the inspectors questions; the licensee initiated AR 01355130 to evaluate the design concern. This issue is considered to be an Unresolved Item pending additional review of the licensees evaluation by the inspectors
05000461/FIN-2012002-032012Q1ClintonIncomplete ED Dose Rate Alarm EvaluationThe inspectors identified an Unresolved Item concerning events that occurred on December 17, 2011, when three workers received electronic dosimeter (ED) dose rate alarms while working in the reactor cavity. The workers were lifting and setting the drywell head bolts in the reactor cavity, an area with high levels of radioactive contamination. The handling of these bolts caused contamination to build up on the workers gloves. The contamination levels on the gloves became high enough and caused the ED worn on the chest to alarm. Two of the alarms occurred when the worker handled the ED to read the accumulated dose. The third worker alarm occurred when his hands came close to the chest and the ED reported dose rates of 1195 mrem/hour deep dose and 1405 mrem/hour shallow dose. The initial investigation performed by the licensee evaluated the whole body dose to the workers. This investigation determined that the workers were briefed to receive dose rate alarms, therefore the events were not entered into the licensees corrective action program. At the time of the inspection, the licensee had not completed an evaluation of the radiological dose to the extremity (hands) from the build-up of contamination on the gloves worn by the workers that caused the ED alarms. Additionally, the licensee could not demonstrate the radiological controls that were in place when the alarms occurred. The issue is categorized as an Unresolved Item pending completion of a revised evaluation and the NRCs review of it
05000331/FIN-2012002-072012Q1Duane ArnoldLicensee-Identified ViolationThe licensee identified a Severity Level (SL) IV NCV of 10 CFR 50.9, Completeness and Accuracy of Information, on June 29, 2010, after it was discovered that six operator licenses did not have medical restrictions for use of Continuous Positive Airway Pressure (CPAP) devices. This issue was reviewed by the inspectors on September 17, 2010, and exited as an URI 05000331/2010004-05, pending further review by NRC and NRC doctors. Corrective actions included re-submittal of NRC Form 396 for the effected operators, documenting the issue in the corrective action program as CR 00580281, and performing an ACE. The licensee also developed a checklist of questions for licensed operators to answer semi-annually regarding the operators use of prescribed medication and reporting of medical conditions. On October 12, 2010, the NRC amended six reactor operator licenses to include the license condition, must use therapeutic devices as prescribed to maintain medical qualifications. The failure to include medical condition restrictions within six operator licenses was a performance deficiency. Because the performance deficiency is considered to potentially impede or impact the ability of the NRC to perform its regulatory oversight function, the performance deficiency was dispositioned using the traditional enforcement process. Per NRC Enforcement Policy, Section 6.4, failing to include medical condition restrictions within operator licenses, which did not adversely affect their ability to safely operate the facility was categorized as an example of a SL IV violation. Additionally, the control room operators performance was monitored and they continued to be evaluated as satisfactory during periodic testing and requalification testing. Because the violation was entered into the licensees CAP, compliance was restored in a reasonable period of time, and was not repetitive or willful; this violation is being treated as a non-cited SL IV violation, consistent with Section 2.3.2 of the NRC Enforcement Policy. The performance deficiency was not considered a finding using IMC 0612, Appendix B, Issue Screening, and did not impact the Reactor Oversight Program Cornerstones of Safety.
05000461/FIN-2012002-022012Q1ClintonUnacceptable Preconditioning of LOW Pressure Coolant Injection from Residual Heat Removal a Check Valve Prior to Leak Rate Test MeasurementThe inspectors identified a finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The licensee failed to establish an adequate procedure to perform required leak rate testing for the Low Pressure Coolant Injection from Residual Heat Removal A Check Valve. Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to a leak rate test measurement due to improper test sequencing. In addition, the licensee failed to correctly evaluate a failed leak rate test of the valve. The licensee entered this issue into its corrective action program for evaluation and initiated corrective actions to revise the test procedure and train engineering personnel. The finding was of more than minor significance since it was associated with the Procedure Quality attribute for the containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Additionally, the licensees failure to correctly evaluate the initial failed leak rate test would become a more significant safety concern if left uncorrected because it could reasonably result in an unrecognized condition with a check valve failing to fulfill a safety-related function. Therefore, this performance deficiency had a direct effect on the licensees ability to fully assess the past operability, as well as the ability to trend as-found data for the purpose of assessing the reliability of the check valve. The finding was a licensee performance deficiency of very low safety significance because it would not result in exceeding the Technical Specification limit for reactor coolant system leakage and would not have likely affected mitigation systems resulting in a loss of safety function. In addition, the finding did not represent an actual open pathway in the physical integrity of the reactor containment. Based on consultation and review with the Regional Senior Reactor Analyst, the inspectors concluded that the finding did not result in an increase in the likelihood of an initiating event such as an inter-system loss-of-coolant accident or a containment bypass event because the redundant isolation valve and closed loop system piping passed leak rate measurement testing during the refueling outage with considerable margin. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, the licensee did not have adequately trained and knowledgeable personnel available to correctly evaluate the cause of the initial failed leak rate measurement test and to ensure that appropriate actions to correct the test sequence in the procedure were identified
05000461/FIN-2012002-042012Q1ClintonFailure to Incorporate Operating Experience Into Preventive Maintenance ActivitesA self-revealed finding of very low safety significance was identified with an associated non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants. The licensee failed to incorporate operating experience into its preventive maintenance practices associated with steam bypass system control circuit cards. Specifically, during two operating experience driven initiatives performed by the licensee in 2001 and 2007, and once again on September 24, 2011, the licensee failed to implement any preventive maintenance activity for critical component circuit cards, which resulted in age-related failure and a reactor scram on November 29, 2011. The licensee initiated corrective actions to replace system circuit cards, perform periodic replacement/refurbishment maintenance activities, and trend circuit card performance during routine calibration. The finding was of more than minor significance because it was sufficiently similar to IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, Example 7(c), in that this violation of 10 CFR 50.65(a)(3) had a consequence ... such as equipment problems attributable to failure to take industry operating experience into account when practicable. The finding was a licensee performance deficiency of very low safety significance because it: (1) did not contribute to the likelihood of a loss-of-coolant accident initiator, (2) did not contribute to both the likelihood of a reactor scram AND the likelihood that mitigation equipment or functions would not be available, and (3) did not increase the likelihood of a fire or internal/external flooding event. The inspectors concluded that this finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support long-term equipment reliability by scheduling maintenance as more preventive than reactive
05000331/FIN-2012002-092012Q1Duane ArnoldLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the failure to adequately review the design of the RHR system to ensure vents were located at system high points and to ensure that design basis requirements were translated into procedures. Specifically, the licensees original design reviews in response to Generic Letter 2008-01 did not identify that vents were not located at the actual system high points, and design reviews failed to identify portions of the system which would not receive adequate flow during fill and vent operations following maintenance. Corrective actions for this issue included installing vents at the system high points (completed for the B RHR subsystem and planned for the A RHR subsystem). In addition, the licensee plans to review RHR, Core Spray, RCIC and HPCI system flow paths to ensure the systems are adequately vented during fill and vent procedures following maintenance. The issue was entered into the CAP as CR 01712033. The failure to adequately review the design of the RHR system to ensure vents were located at system high points and to ensure that design basis requirements were translated into procedures was a performance deficiency. The performance deficiency was determined to be more than minor and a finding because it was associated with the design control, procedure quality, and human performance attributes of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee did not identify the design inadequacy of the RHR system vent location during review of the system, and procedures used to vent the RHR system following maintenance did not ensure that all gas voids were vented. The finding screened as of very low safety significance (Green) because the finding involved a design or qualification deficiency that did not result in a loss of operability.
05000331/FIN-2012002-082012Q1Duane ArnoldLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, on December 29, 2011, when sequentiallyperformed functional tests of the LPCI recirculation riser differential pressure instruments failed. Section 4OA3.2 documents the chronology of the instrument conditions and apparent cause. Corrective actions included a revision to I.PDIS-I204-01 to include appropriate vendor manual information, performance of STP 3.3.5.1-22 at an increased frequency, and longer term actions to replace the instruments with a different design and/or pursuance of a TS amendment to broaden the trip setpoints to reduce cycling of the instruments and switches. The failure to assure that I.PDIS-I204-01 was appropriate to the circumstances was a performance deficiency. The performance deficiency was determined to be more than minor and 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 (i.e., core damage). Specifically, LPCI loop select availability and capability was not ensured for approximately 1.6 hours due to I.PDIS-I204-01 not including available vendor manual information. The inspectors evaluated the finding in accordance with 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a. In the Mitigating Systems Cornerstone, the inspectors answered Yes to the screening question Does the finding represent a loss of system safety function? Therefore, the finding required further risk evaluation. Since the SDP Phase 2 risk tools are being replaced with tools based on the plant-specific Standardized Plant Analysis Risk (SPAR) models, the Region III Senior Risk Analysts (SRAs) performed a Phase 3 risk evaluation using the SPAR model. The risk evaluation was performed using SAPHIRE Version 8.0.7.18 and the Duane Arnold SPAR model (Version 8.18). The increase in core damage frequency (CDF) was analyzed assuming the safety function of LPCI during design basis loss of coolant accidents (LOCAs) was lost. The exposure time was assumed to be 1.6 hours since this was the duration when the ability of the LPCI loop select logic scheme to select the correct loop for injection was nonfunctional. For the performance deficiency, CDF was only affected during LOCA scenarios. The SRAs used updated LOCA frequency data for large, medium, and small LOCAs from the 2010 reactor operational database maintained by Idaho National Laboratory. The SRAs performed a bounding assessment for the change in CDF by setting all four RHR/LPCI pump fail-to-run values to True (i.e., failed) and solving the LOCA scenarios. The resultant ACDF was 8.9E-8/yr for an exposure time of one year. The actual risk significance is considerably less considering the short exposure time of only 1.6 hours. Based on this Phase 3 analysis, the inspectors determined that the finding was of very low safety significance (Green).
05000461/FIN-2012002-052012Q1ClintonViolation of TS 3.4.6 and TS 3.6.1.3 Due to Missed Surveillance of an RCS PIVTS 3.4.6 requires, in part, that the leakage from each RCS PIV shall be within limits in Modes 1, 2, and 3. TS 3.4.6, Condition A, states that with one or more flow paths with leakage from one or more RCS PIVs not within the limit, isolate the high pressure portion of the affected system from the low pressure portion by use of one closed manual, deactivated automatic, or check valve within 4 hours. TS 3.4.6, Condition B, states that if the required action and associated completion time of Condition A is not met, be in Mode 3 within 12 hours and Mode 4 within 36 hours. TS 3.6.1.3 requires, in part, that each primary containment isolation valve be operable in Modes 1, 2, and 3. TS 3.6.1.3, Condition C, states, in part, that with one or more penetration flow paths with leakage rate not within limit, restore leakage rate to within the limit within 4 hours. TS 3.6.1.3, Condition E, states, in part, that if the required action and associated completion time of Condition C is not met, be in Mode 3 within 12 hours and Mode 4 within 36 hours. Contrary to the above, RCS PIV 1E12-F042C was found with leakage in excess of the limit during leak rate measurement surveillance testing on December 7, 2011. The licensee determined the cause for the failure was due to wear on the valve guide ribs and excessive disc to rib clearances, such that the disc became cocked when closing and it did not fully seat. In addition, testing of this valve during the previous operating cycle to satisfy TSSR 3.4.6.1 was not correctly performed, for which the licensee had taken action in accordance with TSSR 3.0.3. Based on the cause determination and the previous missed surveillance of 1E12-F042C, the inspectors concluded that this valve had not been capable of performing its specified safety function (and thus was inoperable) for a period of time before its discovery longer than allowed by TS 3.4.6 and TS 3.6.1.3. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings. In accordance with Table 4a, Characterization Worksheet for IE, MS, and BI Cornerstones, the inspectors determined that that this finding was a licensee performance deficiency of very low safety significance (Green) because the finding would not result in exceeding the TS limit for RCS leakage and would not have likely affected mitigation systems resulting in a loss of safety function. In addition, the finding did not represent an actual open pathway in the physical integrity of the reactor containment. Based on consultation and review with the Regional Senior Reactor Analyst, the inspectors concluded that the valve test failure did not result in an increase in the likelihood of an initiating event such as an intersystem LOCA or a containment bypass event because the redundant isolation valve and closed loop system piping passed leak rate measurement testing during the refueling outage with considerable margin. This violation of TS 3.4.6 and TS 3.6.1.3 is being treated as a non-cited violation consistent with Section 2.3.2 of the NRC Enforcement Policy. The licensee entered this violation into its corrective action program as AR 01305725. The licensee submitted LER 05000461/2011-006-00 to report this issue as a condition prohibited by TS. Refer to Section 4OA3.2 of this inspection report for the review and closure of the LER.
05000331/FIN-2012002-062012Q1Duane ArnoldInadequate Causal Evaluation and Corrective Actions for Loss of RHR System LPCI Safety Function due to Inoperable ECCS InstrumentationA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action was identified by the inspectors on March 7, 2012, following review of apparent cause evaluation (ACE) 01720033 associated with the loss of Low Pressure Coolant Injection (LPCI) loop select capability. Specifically, the inspectors identified several concerns with the implementation of the licensees corrective action program characterization of CR 01720033 that resulted in the inadequate evaluation of cause, extent of cause and condition; and incomplete corrective actions to prevent recurrence. The licensee entered the issue into the CAP as CR 01742201, and was in the process of revising the original causal evaluation and performing an additional ACE to investigate the CAP implementation issues. The inspectors determined that failing to properly determine the cause and take corrective actions to prevent recurrence for LPCI loop select instrument failures represented a performance deficiency. The performance deficiency was determined to be more than minor and a finding because, if left uncorrected, failing to properly determine the cause and take corrective actions to prevent recurrence for significant conditions adverse to quality would have the potential to lead to a more significant safety concern. The inspectors evaluated the finding in accordance with IMC 0609.04, Table 4a. Because the inspectors answered No to all five screening questions under the Mitigating Systems Cornerstone column, the finding screened as very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Decision Making components, and involving the licensee making safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained.
05000331/FIN-2012002-052012Q1Duane ArnoldSecondary Containment Airlock Door Interlock System Conditions Adverse to Quality Not Promptly CorrectedA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors on February 21, 2012, for the licensees failure to promptly correct secondary containment (SCT) airlock door interlock system conditions adverse to quality. Specifically, the inspectors identified several instances during 2010 and 2011 where the licensee did not adequately correct interlock system conditions resulting in simultaneous opening of SCT airlock doors. For each occurrence, the interlock system conditions resulted in unplanned inoperability of secondary containment and entries into short-term limiting condition for operation (LCO) action statements. The licensee entered the inspectors concerns into the CAP as CR 01716446 and CR 01737495, and was in the process of performing a condition evaluation and apparent cause evaluation. The inspectors determined that the licensees failure to promptly correct SCT airlock door interlock system conditions adverse to quality represented a performance deficiency. The performance deficiency was determined to be more than minor and a finding because, if left uncorrected, failing to promptly correct conditions adverse to quality would have the potential to lead to a more significant safety concern. The inspectors evaluated the finding in accordance with IMC 0609.04, Table 4a. Because the inspectors answered No to all questions under the Containment Barrier column, the finding screened as very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Decision Making components, and involving the licensee making safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained.
05000331/FIN-2012002-042012Q1Duane ArnoldBattery Conditions Adverse to Quality Not Promptly IdentifiedA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors on January 23, 2012, for the licensees failure to promptly identify and correct safety-related direct current (DC) battery system conditions adverse to quality. Specifically, several through-lid cracks on the 1D1 and 1D2 125 volts direct current (VDC), and 1D4 250 VDC batteries, that were considered degraded conditions, were not promptly identified by the licensee. The susceptibility and progression of lid cracking was a known condition; however, monitoring of the condition was not adequate to ensure correction of the conditions prior to impacting the qualification of the batteries. The licensee entered the inspectors issues into the CAP as CRs 01727026, 01727028 and 01727030. The licensee performed prompt operability determinations (PODs) that determined the affected DC electrical subsystems were operable, but degraded, pending restoration of the batteries to full qualification (epoxy repairs). The inspectors determined that failing to promptly identify and correct battery lid cracking that impacted qualification represented a performance deficiency because it was the result of the licensees failure to meet a regulatory requirement, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented. The performance deficiency was determined to be more than minor and a finding because, if left uncorrected, failing to promptly identify and evaluate the operability of a degraded condition would have the potential to lead to a more significant safety concern. The inspectors evaluated the finding in accordance with IMC 0609.04, Table 4a. Because the finding was a qualification deficiency confirmed not to result in loss of operability (Question 1 under the Mitigating Systems Cornerstone column), the finding screened as very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Resources components, and involving the licensee maintaining long term plant safety by maintenance of design margin and minimization of long-standing equipment issues.
05000331/FIN-2012002-032012Q1Duane ArnoldFailure to Maintain Type A Container Design TestsA finding of very low safety significance and associated NCV of 10 CFR 71.5 was identified by the inspectors on February 2, 2012, due to the licensees failure to maintain a licensed material shipment on file for at least one year after the latest shipment, and not providing on request, complete documentation of tests supporting the engineering evaluation or comparative data showing that the construction methods, packaging design, and materials of construction complied with the Type A specification. Specifically, the licensee maintained a container certificate from the owner of a container that stated the container complied with the specification testing of 49 CFR 173.465, but upon further review, the testing basis for the engineering evaluation could not be produced by the package owner for the use of the shipper and review by the NRC. The licensee entered this issue into the CAP as CR 01730713. The inspectors determined that the licensees failure to maintain a licensed material shipment on file for at least one year after the latest shipment, and not providing on request, complete documentation of tests supporting the engineering evaluation or comparative data showing that the construction methods, packaging design, and materials of construction comply with the Type A specification, was a performance deficiency. The inspectors determined that the performance deficiency was more than minor and a finding because, if left uncorrected, failing to maintain and provide licensed material shipment documentation would have the potential to lead to a more significant safety concern. The inspectors evaluated the finding in accordance with IMC 0609 Appendix D, Public Radiation Safety Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the performance deficiency did not involve exceeding a radiation limit, a breach of package during transit, a certificate of compliance, low level ground burial, or failure to make notification or provide emergency information. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Work Practices components, and involving the licensee not ensuring supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported.
05000331/FIN-2012002-022012Q1Duane ArnoldFailure to Properly Label and Maintain Labels on Containers in the Radioactive Waste FacilityA finding of very low safety significance and associated NCV of 10 CFR 20.1904(a) was identified by the inspectors on January 31, 2012, due to the licensees failure to label several containers holding radioactive material in the radioactive waste facility and two sea land containers inside the radiologically controlled area (RCA). In some cases, the licensee also failed to assure that labels were affixed and readable to support the function of providing information to radiation workers in the vicinity. The licensee entered the inspectors issues into the CAP as CR 01730867. The inspectors determined that the licensees failure to appropriately affix labels to containers storing radioactive material in the radioactive waste facility and perform periodic reviews of labeling conditions was a performance deficiency. The inspectors determined that the performance deficiency was more than minor and a finding because, if left uncorrected, failing to ensure labeling of radioactive material would have the potential to lead to a more significant safety concern. The inspectors evaluated the finding in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process. The finding was determined to be of very low safety significance (Green) because the performance deficiency did not affect As-Low-As Is-Reasonably-Achievable Planning or Work Controls, did not involve an overexposure, there was not a substantial potential for overexposure, and the ability to assess dose was not compromised. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Problem Identification and Resolution, having Self and Independent Assessment components, and involving the licensee not conducting self-assessments at an appropriate frequency and with sufficient depth, objectivity, and critical assessment.
05000331/FIN-2012002-012012Q1Duane ArnoldLack of Acceptance Criteria within Emergency Core Cooling System Surveillance ProcedureA finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified by the inspectors on February 2, 2012, for the licensees failure to prescribe a procedure appropriate to the circumstances and include appropriate acceptance criteria in Surveillance Test Procedure (STP) 3.5.1-15, RHR System Water Fill Test, Revision 1. Specifically, STP 3.5.1-15 did not provide guidance for quantifying the size of any voids within the system, such that the effect on system operability could not be readily evaluated, nor did the STP establish criteria for an acceptable as-found condition. The licensee entered this issue into the corrective action program (CAP) as condition report (CR) 1731106 and initiated procedure revisions to provide appropriate acceptance criteria. The inspectors determined that failing to establish appropriate acceptance criteria for a Technical Specification (TS) surveillance procedure was a performance deficiency. The performance deficiency was determined to be more than minor and 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 (i.e., core damage). The finding was evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Mitigating Systems Cornerstone. The finding screened as of very low safety significance (Green) because the finding was a qualification deficiency confirmed not to result in loss of operability or functionality. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, having Corrective Action Program components, such that issues potentially affecting nuclear safety are promptly identified (at a low threshold), fully evaluated, and that actions are taken to address safety issues in a timely manner.
05000254/FIN-2011005-012011Q4Quad CitiesLicensee-Identified ViolationThe licensee identified that 12 fire damper inspections in the main control room, performed as part of one work order, were not completed before the 25 percent grace period expired. This occurred because the inspections were delayed and then rescheduled past the drop dead date. In addition, departmental and organizational tools used to track the WO did not identify that the planned maintenance was past due since it was still in a working status. The issue is more than minor because, if left uncorrected, the programmatic breakdowns, which allowed the inspections to go beyond the designated frequency, could lead to a more significant safety concern with fire protection equipment or barriers potentially degraded and not evaluated. Technical Specification 5.4.1.c requires written procedures be established, implemented, and maintained to cover Fire Protection Program implementation. Contrary to this requirement, the preventative maintenance work order for control room fire damper inspections was not implemented as required by the program frequency. The issue was identified by the station nuclear oversight auditors and documented in the CAP as IR 1286382. The issue was determined to be of very low safety significance since the inspections were subsequently completed with no problems identified. Corrective actions included completing the inspections and changing organizational tracking tools to identify similar work orders to management when the task went into the grace period.
05000331/FIN-2011005-012011Q4Duane ArnoldFailure to Make Required Eight Hour Event Report per 10 CFR 50.72(b)(3)(v)(B)A Severity Level (SL) IV NCV of 10 CFR 50.72(b)(3)(v)(B) was identified by the inspectors for the licensees failure to report within eight hours a condition that, at the time of discovery, could have prevented the fulfillment of the Residual Heat Removal (RHR) system Low Pressure Coolant Injection (LPCI) safety function. Specifically, on December 2, 2011, a sizable void was identified in the B LPCI discharge injection line resulting in the LPCI mode of RHR being declared inoperable. The licensee documented the issue into their corrective action program (CAP), reported the condition to the NRC on December 8, 2011, and, was in the process of reviewing the cause of the issue to determine additional corrective actions.The inspectors determined that the issue of concern represented a performancedeficiency because it was the result of the licensees failure to meet a regulatory requirement, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented. Because the performance deficiency is considered to potentially impede or impact the ability of the NRC to perform its regulatory oversight function, the performance deficiency was dispositioned using the traditional enforcement process. Per NRC Enforcement Policy, Section 6.9.d.9, failing to make a report required by 10 CFR 50.72 is categorized as an example of a Severity Level IV violation. Additionally, because the violation was entered into the licensees CAP, compliance was restored in a reasonable period of time, and was not repetitive or willful; this violation is being treated as a non-cited SL IV violation, consistent with Section 2.3.2 of the NRC Enforcement Policy. Because the performance deficiency was not considered a finding using IMC 0612, Appendix B, Issue Screening, and did not impact the Reactor Oversight Process Cornerstones of Safety, a cross-cutting aspect was not assigned.
05000331/FIN-2011005-032011Q4Duane ArnoldProcedural Non-Compliance Results in Reactor Building Crane Colliding with ISFSI Inspection StanDA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on October 31, 2011, when operators failed to follow Operating Instruction 999, Reactor Building Crane. Specifically, this error resulted in the reactor building (RB) crane striking the Independent Spent Fuel Storage Installation (ISFSI) inspection stand. Immediate corrective actions included performing inspections of the dry storage container transfer cask, ISFSI inspection stand, and reactor building crane. The inspectors determined that attempting to move the crane over the ISFSI work platform while the hand rails were installed was contrary to the RB crane operating instruction and was an issue of concern. Failing to follow the RB crane operating instruction was a performance deficiency because it was the result of the licensees failure to meet a procedural requirement, and the cause was reasonably within the licensees ability to foresee and correct and should have been prevented. The performance deficiency was determined to be more than minor and a finding because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, not following the RB crane operating instructions could lead to a more significant event or cause damage to safety-related equipment. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a for the Barrier Integrity Cornerstone. Because the finding only affected the fuel barrier, the finding was determined to be of very low safety significance (Green). The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting aspect of Human Performance, having Work Control components, and involving appropriately coordinating work activities by incorporating actions to address the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities. Specifically, the licensee did not implement appropriate work controls to ensure the hand rails of the ISFSI inspection stand were removed prior to moving the crane for an activity that was not associated with the ISFSI project.
05000461/FIN-2011005-042011Q4ClintonUnacceptable Preconditioning of Reactor Core Isolation Cooling System Check Valve Prior to Leak Rate Test MeasurementThe inspectors identified a finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The licensee failed to establish an adequate procedure to perform required leak rate testing for the reactor core isolation cooling turbine exhaust check valve. Specifically, the surveillance test procedure resulted in unacceptable preconditioning of the valve prior to an as-found leak rate test measurement. The licensee entered this issue into its corrective action program for evaluation and initiated a corrective action to revise the test procedure. The finding was of more than minor significance since it was associated with the Procedure Quality attribute for the containment and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Because the preconditioning altered the as-found condition of the check valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Therefore, this performance deficiency had a direct effect on the licensees ability to fully assess the past operability, as well as the ability to trend as-found data for the purpose of assessing the reliability of the check valve. The finding was a licensee performance deficiency of very low safety significance because it did not involve an actual open pathway in the physical integrity of the reactor containment. The inspectors concluded that this finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee did not implement operating experience into station processes, procedures, and training in that the licensee did not update/revise the surveillance test procedure consistent with NRC guidance and the licensees corporate technical position to prevent unacceptable preconditioning of the check valve.
05000254/FIN-2011009-052011Q4Quad CitiesDiesel Generator Technical Specification Frequency and Voltage Variation not Considered in Loading CalculationsWhile reviewing actions in response to NRC Information Notice (IN) 2008- 002, the inspectors noted the licensee staff had evaluated the effects of lower than nominal frequency on pump flows and pressures and determined there were no adverse effects on pump flow and pressures and the upper frequency limit did not cause the EDG loading to exceed the 200 hour per year rating. The inspectors found that the current calculation (QDC-6700-E-1500, Rev 005A) determined diesel loading based on maximum loads during a design basis accident. The loading was based on nominal 60 hertz (Hz) operation of pumps and fans, and did not account for the +2 percent variations allowed by TS 3.8.1.2. Mechanical affinity laws show power demanded by centrifugal pumps and fans increases by the cube of the ratio of the speeds (1.02 cubed = 1.061). Since the EDG accident loading was comprised primarily of centrifugal loads, the inspectors determined this phenomenon should have been considered in loading calculations. In CR 591442, Effect of EDG Frequency on Loading and Pump Flows, dated February 14, 2007, the licensee had noted this increase in loading and had determined that it was still within the 200 hr per year rating (2973 kW) for the EDGs. The evaluation of decreased frequency loading showed all pumps developed adequate pressure and flow. The inspectors were concerned the maximum loading exceeded the 2000 hr rating of 2864 kW. The inspectors consulted with the Office of Nuclear Reactor Regulations (NRR) and were informed the loading should be compared to the licensed value for the EDGs. Additionally, the licensee should have demonstrated there were no detrimental effects on all systems, structures, and components (SSCs) over the full range of frequency and voltage allowed by TS. Specifically, the torque developed by a motor is directly proportional to the square of the voltage and inversely proportional to the square of the frequency. The inspectors compared the TS limits to the nominal values for voltage and frequency and determined that torque developed by the motors supplied by the EDGs could vary as much as 14 percent from the nominal torque. The inspectors informed the licensee of their concerns about the operability of the supplied motors being able to meet the design requirements, particularly for pumps and motoroperated valves under the worst torque conditions (minimum voltage and maximum frequency) and EDG loading for the best conditions. Additionally, the inspectors noted the UFSAR loading limit was the 2000 hr rating. The inspectors asked for any licensing document that could support using the 200 hr load limit. In response to the inspectors questions, the issue was entered into the licensees corrective action program as AR 01288784, CDBI Technical Specification Limits for EDG, dated November 10, 2011. A review of operating procedures provided reasonable assurance that the EDGs would be operated near the midpoint of the allowed TS range during a potential event until the licensee demonstrates operability over the full TS range. This will be an unresolved issue pending the results of the licensees evaluation of the effects of the full TS voltage and frequency ranges on all SSCs and verifying the licensed load limit for the EDGs.
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