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05000305/FIN-2013003-0230 June 2013 23:59:59KewauneeLicensee-identifiedLicensee-Identified Violation10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to the above, procedure OP-KW-NOP-FH-003, Reactor Cavity Draining With Fuel or Upper Internals Installed, prescribed instructions for draining of the refueling cavity, in order to refill the refueling water storage tank in preparation for making the emergency core cooling system operable. However, the procedure was not appropriate to the circumstances because it did not prescribe that if procedure NSP-SI-004, Monitoring SI System for Void After System Refill, was performed prior to the cavity draining to the RWST, that procedure NSP-SI-004 shall be re-performed to ensure the emergency core cooling system was free from voids, which could impact availability of the system. Consequently, on April 24, 2012, the licensee performed NSP-SI-004 and ensured the system was free of voids to ensure SI system availability and then completed draining of the refueling cavity to the refueling water storage tank in the emergency core cooling system. The system was not monitored for voids following the refueling water storage tank refill and on June 27, 2012, during the normally scheduled surveillance to perform testing to ensure the system was free of voids, the licensee discovered a 2.8 cubic-foot void near the common suction of the SI pumps. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of configuration control and affected the Cornerstone objective of ensuring the reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The risk associated with this issue was determined to be of very low safety significance (Green) because the finding was a deficiency affecting the qualification of a mitigating system and the system was operable but nonconforming. The licensee conducted full scale tests of the system, which demonstrated that the system was available with the gas void present for all accident conditions. The licensee entered this issue into the CAP as CR480150, Gas Void in SI Pump Suction Piping, conducted and apparent cause evaluation and shared the operating experience with the industry.
05000305/FIN-2013003-0130 June 2013 23:59:59KewauneeNRC identifiedInadequate Procedure for Testing of the Diesel Room Ventilation Damper Actuator Back Up Air SystemA finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors, for the failure to have procedures appropriate to the circumstances for an activity that affects quality. Specifically, Operating Surveillance Procedure OP-KW-OSP-TAV-001A (and B), Diesel Generator A (or B) Back Up Air Supply Leak Rate Test, allowed the performer to open safety related (SR) pressure boundary valves and install non-safety related (NSR) test equipment on both back up air bottle sets without declaring the respective Emergency Diesel Generator (EDG) inoperable. The licensee initiated a condition report and revised both procedures to prevent both bottle sets from being tested at the same time while maintaining the respective diesel operable. The finding was determined to be more than minor because the finding, if left uncorrected, had the potential to lead to a more significant safety concern. Specifically, the licensee concluded that procedures OP-KW-OSP-TAV-001A (and B) allowed unqualified test equipment to be relied upon as the SR pressure boundary for both back up air bottle sets without declaring the respective EDG inoperable. The inspectors concluded this finding was associated with the Mitigating Systems Cornerstone. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The licensee evaluated the installed test equipment and hose connections and concluded their pressure rating exceeded that necessary to function as a pressure boundary; therefore, the inspectors answered Yes to Mitigating Systems Screening question number 1, and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, Resources, because the licensee did not assure that procedures were adequate to assure nuclear safety.
05000305/FIN-2013007-0331 March 2013 23:59:59KewauneeNRC identifiedInaccurate and Incomplete Information Documented on Fire Drill Evaluation/Critique FormsIn addition, and also as discussed in Enclosure 1, the NRC staff identified a second AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9(a), Completeness and Accuracy of Information, which requires, in part, that information required by license conditions to be maintained by the licensee to be complete and accurate in all material respects. This violation is being processed through traditional enforcement because it involves willfulness and it impacts the regulatory process. There were no technical aspects to this AV. Specifically, from at least August 2009 through December 2011, a member of the licensees staff completed fire drill evaluation/ critique forms, a quality assurance record, inaccurately. This information was determined to be material to the NRC. The AV is discussed in Enclosure 1 and in the Enforcement portion of Section 1R05A.
05000305/FIN-2013002-0231 March 2013 23:59:59KewauneeNRC identifiedInadequate Risk Assessment for Degraded EquipmentThe inspectors identified a finding of very low safety significance and associated non-citied violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, Paragraph (a)(4), for the failure to properly assess the risk during scheduled testing of EDG B, on February 25. Specifically, the licensee failed to update the risk model after identifying that the manual bypass switch (MBS) for BRD-109, a risk-significant inverter, was not functioning properly. On February 13, 2013, the licensee identified that the MBS for the output of BRD-109 was broken, and BRD-109 was not powering BRD-115, a non-interruptible bus, even though the Normal position was selected. The licensee did not include the alternate switch position in the risk model until prompted by the NRC inspectors on February 26. This resulted in the risk from the EDG B testing being incorrectly assessed as green, when it was actually yellow. The licensee entered the issue into the corrective action program (CAP) and assigned an apparent cause evaluation (ACE) to determine the causes for the organizational failures that occurred. The ACE was not yet complete at the end of this inspection period, The inspectors determined that the failure to properly assess risk was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Human Performance, and adversely affected the Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors evaluated this finding using the worksheets from IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, dated May 19, 2005, because the finding is a maintenance risk assessment issue. Flowchart 1, Assessment of Risk Deficit, requires the inspectors to determine the risk deficit associated with this issue. This finding was determined to be of very low safety significance because the incremental core damage probability deficit was less than 1 x 10-6. This finding has a cross-cutting aspect in the area of problem identification, corrective action program, because the licensee did not ensure that issues potentially impacting nuclear safety are promptly identified, fully evaluated, and that actions were taken to address safety issues in a timely manner, commensurate with their significance.
05000305/FIN-2013002-0331 March 2013 23:59:59KewauneeNRC identifiedFailure to Provide Fire Detection and Fixed Fire Suppression in a III.G.3 AreaThe inspectors identified a finding of very low safety significance and associated non-citied violation of 10 CFR Part 50, Appendix R, Section III.G.3, Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979, for the licensees failure to meet the requirement for fire detection and a fixed fire suppression system in the control room heating, ventilation, and air conditioning (HVAC) room (fire zone AX-35). Specifically, the licensee failed to provide fire detection and a fixed fire suppression system in the HVAC equipment room. The licensee entered this into the CAP as corrective actions (CA)075268 and CA08365; and CRs CR108948 and CR463976. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factors (Fire), and adversely affected the Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the lack of area-wide detection and suppression systems could increase the response time and complicate fire fighting activities. That could result in a loss of both trains of safe shutdown equipment required for safe shutdown of the plant. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated July 19, 2012. Using Table 3, SDP Appendix Router, the inspectors answered Yes to Question E.2.(2), Does the finding involve fixed fire protection systems or the ability to confine a fire? because the finding involved failure to provide fire detection and fixed suppression in the control room HVAC room. As a result, the finding was evaluated using IMC 0609, Appendix F, Fire Protection Significance Determination Process, dated February 28, 2005. A detailed risk evaluation was performed by a Region III Senior Reactor Analyst (SRA), which concluded that the finding was of very low risk significance (green). The inspectors did not identify a cross-cutting aspect associated with this finding, because this was an original design issue, and the finding was not representative of current performance.
05000305/FIN-2013007-0131 March 2013 23:59:59KewauneeNRC identifiedFailure to Perform Announced Fire Drills In Accordance With Fire Protection Program PlanThe NRC staff identified a preliminary finding of low-to-moderate safety significance for the licensees failure to 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 required quarterly announced fire drills in that the announced fire drills were instead conducted as training sessions. The licensee documented this issue in the corrective action program as condition report (CR) 0456865 and the licensee continued to perform causal evaluations to determine the cause of the issue and develop long-term corrective actions. In January 2012, as a result of an apparent cause evaluation (ACE), the licensee took remedial corrective actions to ensure fire brigade drills were performed as required. In addition, several other corrective actions were implemented as a result of the ACE conducted in 2012. The finding 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 the finding was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factors for Fire and adversely affected the cornerstone objective of ensuring the reliability and capability of the fire brigade to respond to initiating events to prevent undesirable consequences. Specifically, the fire brigades ability to rapidly suppress fires and protect structures, systems and components, important to safety from potential fires that could prevent safe shutdown. The NRC staff determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012. Because the finding involved the Fire Brigade, Table 3, SDP Appendix Router, Section E.1, Fire Protection, directed NRC staff to use IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012. Exhibit 2, Mitigating Systems Screening Questions, Section D.1.a, Fire Brigade, was checked yes because the finding involved the Fire Brigade training and qualification requirements. Because none of the other questions in Section D were checked, the NRC staff continued to evaluate the finding using IMC 0609, Appendix M, The Significance Determination Process Using Qualitative Criteria, dated April 12, 2012. Based on the evaluation of the decision attributes contained in Appendix M, the NRC staff determined this finding to preliminarily have a low to moderate safety significance (White). This finding has a cross-cutting aspect in the area of human performance, work practices, because the licensee failed to ensure supervisory and management oversight of work activities such that nuclear safety is supported (H.4(c)). This cross-cutting aspect coincides with the licensees conclusions in an apparent cause evaluation that determined that there was inadequate management/oversight of the fire drill program within the licensees organization, which directly contributed to the issue.
05000305/FIN-2013002-0431 March 2013 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50.48(b)(2) requires, in part, that all nuclear power plants licensed to operate prior to January 1, 1979, must satisfy the applicable requirements of Appendix R to this part, including specifically the requirements of Sections Section III.G, III.J, and III.0. Section III.G.3 of 10 CFR, Part 50, Appendix R, requires, in part, that alternative of dedicated shutdown capability and its associated circuits, independent of cables, systems, or components in the area, room, or zone under consideration should be provided where the protection of systems whose function is required for hot shutdown does not satisfy the requirement of Paragraph G.2 of this section. Contrary to the above, prior to February 28, 2012, the licensee failed to ensure that alternative of dedicated shutdown capability and its associated circuits were independent of cables in the area. Specifically, the licensee failed to ensure that, in the event of a fire in the relay room or the 480V bus 51/52 area, fire damage to the cables and circuits for the PORV valve PR-2B, pressurizer valve PR-33A, and reactor head vent valves RC-45A and RC-46 would not cause spurious operation of the valves. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factors (Fire) and affected the Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The risk associated with this issue was determined to be of very low safety significance (Green) based on the initiating fire event frequencies and the probability of hot shorts occurring for the valves affected. The licensee entered this issue into the CAP as CR464393, Appendix R Fire Spurious Operation Concern for PR-33A, RC-45A, and RC-46, and expanded a roving hourly fire watch to include the 480V bus 51/52 area.
05000305/FIN-2013007-0231 March 2013 23:59:59KewauneeNRC identifiedFailure 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.
05000305/FIN-2013002-0131 March 2013 23:59:59KewauneeNRC identifiedIncorrect Ultimate Heat Sink Water LevelThe inspectors identified a finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to assure that appropriate quality standards were specified and included in design documents, and that deviations from such standards were controlled. Specifically, during implementation of Improved Technical Specifications (TSs), the licensee incorrectly specified the ultimate heat sink (UHS) level in the non-conservative direction by one foot. Therefore, since February 24, 2011, the TS 3.7.9.1 value was non-conservative. The licensee initiated a condition report (CR) and performed an operability determination, which included a review of NRC Administrative Letter 98-10 for non-conservative TS. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Design Control, and adversely affected the Cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee did not assign the appropriate UHS level to assure the safety-related (SR) service water (SW) pumps operability. Therefore, licensee surveillances were conducted with the incorrect UHS level. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined that the SR service water pumps remained operable but non-conforming for the time period in question as the UHS level had not approached 566 feet, 3 inches; therefore, the inspectors answered Yes to Mitigating Systems Screening Question 1, and screened the finding as having very low safety significance (Green). The inspectors concluded that there was no cross-cutting aspect with this finding because the failure occurred in 2008, and was not representative of current performance.
05000305/FIN-2012503-0131 December 2012 23:59:59KewauneeNRC identifiedDegraded Emergency Action Level SchemeA finding having a significance of preliminarily White with one AV of 10 CFR 50.54(q)(2) associated with risk-significant planning standard 10 CFR 50.47(b)(4) was identified by the NRC for the licensee failure to follow and maintain the effectiveness of its emergency plan. Specifically, from February 28, 2011, until March 30, 2011, KPSs unidentified loss of System Particulate, Iodine, and Noble Gas (SPING) indication on the Plant Process Computer System (PPCS) and Radserv stations precluded action to restore the capability to classify Emergency Action Levels (EALs) RG1.1, General Emergency, and RS1.1, Site Area Emergency. The NRC believes that the KPS staff had the opportunity to identify this condition at the time of the failures. On March 30, 2011, the system engineer identified the failure during a system walk down in CR 419976, and the server and computer point were subsequently restored to service. This finding was determined to be more than minor because it was associated with the Emergency Response Organization performance attribute of the Reactor Safety Emergency Preparedness Cornerstone. This finding adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated in accordance with IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process. As Appendix B was revised in February 2012, the finding was evaluated using both the version in effect at the time of the violation and the current version. Under both versions, other than changing the names of the involved Section and Sheet/Attachment, there was no effect on the final outcome. The issue was determined to be a Failure to Comply. The risk was evaluated using Section 4.0 of IMC 0609 and Sheet 1, Failure to Comply, in the previous revision, and Section 5.0 and Attachment 2, Failure to Comply Significance Logic, in the current revision, along with their associated narratives. With EALs, RG1.1 and RS1.1, ineffective, the inspectors considered mitigating factors, such as alternative EALs, within the same initiating condition and determined the alternative EALs were such that an accurate declaration of the initiating condition would have been made. Therefore, the inspectors determined that no loss of Risk-Significant Planning Standard (RSPS) function existed. However, the alternative EAL classifications would have been delayed, and, therefore, the event would have been declared in a degraded manner. The finding was preliminarily determined to be of low to moderate safety significance (White) in that ineffective EALs, RG1.1, and RS1.1 existed, degraded an RSPS function, and affected the ability of the licensee to properly classify events involving a radiological release. A cross-cutting aspect (P.1(d)) was identified within the Corrective Action Program component. The failure to identify the loss of SPING indication on the PPCS and Radserv stations prevented appropriate corrective actions to address the degradation of this risk significant planning standard function.
05000305/FIN-2012005-0431 December 2012 23:59:59KewauneeNRC identifiedRemoval of Refueling Procedure Requirement For Spent Fuel Pool Cooling Without Prior NRC ApprovalThe inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and associated finding of very low safety significance for the failure to document an evaluation that provided a basis for the determination that the changes implemented in Procedure RF-01.00, KNPP (Kewaunee Nuclear Power Plant) Refueling Procedure, Revision I, did not require a license amendment in 2001. Specifically, the licensee failed to include, in the written evaluation, a basis as to why the change did not create a possibility for a malfunction of an structure, system or component (SSC) important to safety with a different result than any previously evaluated in the Updated Safety Analysis Report (USAR) and did not result in a departure from a method of evaluation described in the USAR used in establishing the design bases. The changes to procedure RF-01.00 removed a requirement to place the residual heat removal (RHR) A heat exchanger in parallel with the spent fuel pool (SFP) heat exchanger, when a full core offload took place with service water temperatures greater than or equal to 60 degrees Fahrenheit (◦F). The licensee initiated condition report (CR) 493206, and at the end of the inspection period, was conducting a causal evaluation and developing corrective actions to restore the required actions to the refueling procedure. The violation 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 the performance deficiency adversely affected the Barrier Integrity Cornerstone objective to maintain functionality of the SFP cooling system. Specifically, the configuration control and design control attributes, related to the design margins for the SFP, increase in allowable fuel assemblies as documented in License Amendment No. 150. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012 and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered No to questions 1 through 4 in Section D, Spent Fuel Pool, and screened the finding as having very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation is categorized as SL IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross-cutting aspect associated with this finding.
05000305/FIN-2012009-0531 December 2012 23:59:59KewauneeNRC identifiedFailure to Ensure Correct Criterion Used for Time Critical Operator Actions to Prevent a Steam Generator OverfillThe inspectors identified a finding of very low safety significance and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure consistent criterion was used in OP-KW-ORT-MISC-014, Validation of Time Critical Operator Actions, Revision 2 (i.e., the Operations procedure) to validate the time-critical operator actions for Break Flow Termination. Specifically, the licensee failed to ensure the Break Flow Termination criterion (i.e., the time frame when the primary-to-secondary flow in the ruptured steam generator (SG) is required to be stopped) was consistent with the criterion contained in CN-CRA-03-16, Kewaunee Power Station Supplemental Steam Generator Tube Rupture Margin to Overfill Analysis, Revision 0 (i.e., the calculation). The licensee entered this finding into their Corrective Action Program as CR492485, 2012 CDBI TCA Validation Criteria Does Not Agree with Engineering Basis Document, dated October 19, 2012. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstones attribute of procedure quality and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, during the performance of the steam generator tube rupture (SGTR) design basis event scenario on the licensees simulator, the inspectors noted the operators did not terminate the primary-to-secondary break flow within the time frame assumed in design calculation. Therefore, the inspectors concluded there was reasonable doubt that adequate margin existed to ensure SG overfill would have been prevented during an actual SGTR design basis event. The inspectors determined the finding could be evaluated using the Mitigating Systems Screening questions in Exhibit 2 of IMC 0609 and concluded the finding screened as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000305/FIN-2012009-0131 December 2012 23:59:59KewauneeNRC identifiedFailure to Correct a Condition Adverse to Quality Associated with the Safety-related 125 Vdc Battery Service Test ProceduresThe inspectors identified a finding of very low safety significance and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish measures to assure that conditions adverse to quality were corrected. Specifically, the licensee failed to correct a previously identified finding concerning the safety-related 125 Volts direct current (Vdc) battery service test procedures, where the procedures failed to include the appropriate acceptance criteria for critical periods of the duty cycle, including the first minute. The licensee entered this finding into their Corrective Action Program as CR491149, 2012 CDBI Identified No Acceptance Criteria in the Battery Surveillance Procedure, dated October 10, 2012. The performance deficiency was determined to be more than minor because the licensee could not be assured that loads supplied by the 125 Vdc batteries would have adequate voltage to operate during critical periods of the duty cycle. Since the finding did not represent an actual loss of safety function, the finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not formally define the authority and roles for decisions affecting nuclear safety and as a result did not take the necessary steps to resolve an inadequate surveillance procedure in a timely manner. Specifically, the licensee delayed resolving the inadequate surveillance procedures until a major calculation revision was accomplished.
05000305/FIN-2012005-0631 December 2012 23:59:59KewauneeNRC identifiedC02 Detection Instrumentation Unavailable For Emergency Event ClassificationThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.54(q) Emergency Plans, for failing to follow and maintain an emergency plan that met the requirements of emergency planning standard 10 CFR 50.47(b)(4). Specifically, the licensee failed to maintain onsite atmospheric monitoring equipment capable of measuring carbon dioxide concentrations Immediately Dangerous to Life and Health (IDLH). The IDLH concentrations within specified plant areas are utilized to classify emergency conditions per the KPS Emergency Plan. The licensee entered this performance deficiency into their CAP as CR481430. The inspectors determined that the issue was a performance deficiency as it was within the licensees ability to foresee and correct. This finding was determined to be more than minor because it was associated with the ERO performance attribute of the Emergency Preparedness Cornerstone, and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated in accordance with IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, using Figure 5.4-1, Significance. Determination for Ineffective EALs and Overclassification , and resulted in a Green finding as an Alert event would not have been declared. A cross-cutting aspect was identified in the area of human performance, decision-making. The licensees risk-significant decision concerning this EAL-related equipment change did not use a systematic process to ensure that safety was maintained. A lack of formally defined authority and roles for decisions and communications precluded the appropriate interdisciplinary input and review of this equipment change.
05000305/FIN-2012009-0231 December 2012 23:59:59KewauneeNRC identifiedFailure to Ensure Four CC System Manual Valves Were in the Correct Position as Required by ITS, SR 3.7.7.1The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation (NCV) of Improved Technical Specifications (ITS), Surveillance Requirement (SR) 3.7.7.1 because the licensee failed to ensure four component cooling (CC) system manual valves in the flow path servicing the safety-related CC system pumps, that were not locked, sealed, or otherwise secured in position, were verified in the correct position (i.e., open) every 31 days. The licensee entered this finding into their Corrective Action Program as CR490316, 2012 CDBI CC The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish measures to assure that conditions adverse to quality were corrected. Specifically, the licensee failed to correct a previously identified finding concerning the safety-related 125 Volts direct current (Vdc) battery service test procedures, where the procedures failed to include the appropriate acceptance criteria for critical periods of the duty cycle, including the first minute. The licensee entered this finding into their Corrective Action Program as CR491149, 2012 CDBI Identified No Acceptance Criteria in the Battery Surveillance Procedure, dated October 10, 2012. The performance deficiency was determined to be more than minor because the licensee could not be assured that loads supplied by the 125 Vdc batteries would have adequate voltage to operate during critical periods of the duty cycle. Since the finding did not represent an actual loss of safety function, the finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, decision making because the licensee did not formally define the authority and roles for decisions affecting nuclear safety and as a result did not take the necessary steps to resolve an inadequate surveillance procedure in a timely manner. Specifically, the licensee delayed resolving the inadequate surveillance procedures until a major calculation revision was accomplished.
05000305/FIN-2012005-0731 December 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTechnical Specification SR 3.4.14.2 requires verification that the RHR system interlock prevents the valves from being opened with a simulated or actual pressure signal > 450 psig as implemented in the conversion to ITSs by License Amendment No. 207. Contrary to the above, from February 2011 until September 13, 2012, the licensee failed to recognize that the procedures that calibrated and tested the RHR system interlock would not assure that SR 3.4.14.2 was met. Specifically, during startup from the refueling outage in May 2012, after the 18-month SR was performed, the SR was not identified as not being met, and the required TS 3.4.14.C LCO entry was not made. The ITS conversion process failed to identify the +/- 15 psig (0.5 percent) setpoint tolerance for the 450 psig setpoint would allow opening the valve at a pressure greater than 450 psig; therefore, a change to the setpoint was needed to fulfill the SR requirement. The inspectors determined that the violation could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012. The inspectors answered No to all the screening questions and determined the finding screened as having very low significance (Green). The licensee documented this violation in CR488003 and conducted an apparent cause determination. The licensee took corrective actions to fulfill the TS LCO (closing valves with supply breakers off). Additionally, engineering was determining a lower setpoint and calibration tolerances that would fulfill surveillance requirements.
05000305/FIN-2012009-0331 December 2012 23:59:59KewauneeNRC identifiedNon-Conservative Voltage Calculations for Motor Operated ValvesThe inspectors identified a finding of very low safety significance and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that the design basis was correctly translated into specifications and procedures. Specifically, the licensee failed to provide design control measures to account for motor control center (MCC) voltage dips in MOV calculations that occur when large ESF motors are started at the onset of an accident. The licensee entered this finding into their Corrective Action Program as CR494297, 2012 CDBI: Potential MOV Stalls Not Proven by Calc to Not Occur, dated November 1, 2012. The performance deficiency was determined to be more than minor because the inspectors had reasonable doubt that adequate margin existed to ensure the MOVs required to start at the onset of an accident would have adequate voltage and/or torque. Upon further evaluation by the licensee, the inspectors determined the finding did not represent a loss of operability or functionality, therefore, the finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources because the licensee did not provide complete, accurate, and up-to-date design documentation, including calculations and procedures to assure nuclear safety. Specifically, the licensee failed to ensure that personnel had adequate procedural guidance to account for MCC voltage dips in MOVs that occur when large ESF motors are started at the onset of an accident.
05000305/FIN-2012005-0531 December 2012 23:59:59KewauneeNRC identifiedRelay Room Carbon Dioxide Fire Suppression System and Control Room Envelope Potentially Affected by HELBDuring the inspectors review of the relay room carbon dioxide (CO2) fire suppression system actuation failure, the inspectors identified that control cabinets for both the CO2 fire suppression actuation system and the relay room ventilation damper ACC-22, are mounted on the outside of the SR east wall of the relay room and exposed to the turbine building. The control cabinets could be exposed to steam from a nearby 30-inch steam header if a crack were to develop. The inspectors were concerned that the potential existed to actuate the relay room CO2 fire suppression system and/or cause one of the relay room ventilation dampers to open, complicating the control room response to the small steam break. At the conclusion of this inspection period, the licensee was evaluating the issue; the inspectors needed additional information to determine if a performance deficiency existed. As a result, this item was considered unresolved (URI 05000305/2012005-05, Relay Room Carbon Dioxide Fire Suppression System and Control Room Envelope Potentially Affected by HELB).
05000305/FIN-2012005-0331 December 2012 23:59:59KewauneeNRC identifiedRemoval of Refueling Procedure Requirement For Spent Fuel Pool Cooling Without Prior NRC ApprovalThe inspectors identified a Severity Level (SL) IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and associated finding of very low safety significance for the failure to document an evaluation that provided a basis for the determination that the changes implemented in Procedure RF-01.00, KNPP (Kewaunee Nuclear Power Plant) Refueling Procedure, Revision I, did not require a license amendment in 2001. Specifically, the licensee failed to include, in the written evaluation, a basis as to why the change did not create a possibility for a malfunction of an structure, system or component (SSC) important to safety with a different result than any previously evaluated in the Updated Safety Analysis Report (USAR) and did not result in a departure from a method of evaluation described in the USAR used in establishing the design bases. The changes to procedure RF-01.00 removed a requirement to place the residual heat removal (RHR) A heat exchanger in parallel with the spent fuel pool (SFP) heat exchanger, when a full core offload took place with service water temperatures greater than or equal to 60 degrees Fahrenheit (◦F). The licensee initiated condition report (CR) 493206, and at the end of the inspection period, was conducting a causal evaluation and developing corrective actions to restore the required actions to the refueling procedure. The violation 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 the performance deficiency adversely affected the Barrier Integrity Cornerstone objective to maintain functionality of the SFP cooling system. Specifically, the configuration control and design control attributes, related to the design margins for the SFP, increase in allowable fuel assemblies as documented in License Amendment No. 150. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012 and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012. The inspectors answered No to questions 1 through 4 in Section D, Spent Fuel Pool, and screened the finding as having very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation is categorized as SL IV because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors determined that this finding did not reflect present performance since the error was introduced in a design change that was greater than three years old; therefore, there was no cross-cutting aspect associated with this finding.
05000305/FIN-2012005-0831 December 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTechnical Specification 3.5.2, Emergency Core Cooling System (ECCS), Action, Limiting Condition for Operation, requires, in part, that two ECCS trains shall be Operable while in Mode 1. Contrary to the above, from December 7, 2011, through February 29, 2012, the B SI pump was not operable, because the associated SI pump breaker 1-606 installed on December 7, 2011, did not function properly due to a bent main charging shaft, while the reactor was in Mode 1. Upon discovery of the condition during the scheduled Technical Specification SR test, the licensee restored the system to an operable status through replacement of the SI pump B breaker. The inspectors determined the violation could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power Exhibit 2, Mitigating Systems Screening Questions, effective July 1, 2012. The inspectors answered Yes to the screening question 3 that stated the finding represented an actual loss of function of at least a single train for greater than its allowed Technical Specification allowed outage time, in this case 72 hours. Therefore, the finding required a detailed risk evaluation by the Region III Senior Reactor Analyst (SRA). The SRA determined that the failure would result in failure of the high pressure recirculation mode for SI pump B. The SRA performed the detailed risk evaluation using the Kewaunee SPAR Model Version 8.20, SAPHIRE Version 8.0.8.0. The exposure time was 84 days. The model was run by setting the basic event representing failure of motor-operated valve RHR-299B to True. Setting failure of this valve to True represents failure of the flowpath from the discharge of residual heat removal (RHR) pump B to the suction of SI pump B, thus failing the B train of the high pressure recirculation mode of SI with common cause potential for the A train as well. The resultant delta CDF was 4.13E-7/yr. This represents internal events risk with the dominant sequence involving a loss of offsite power initiating event, failure of secondary cooling, and failure of high pressure recirculation. Regarding external events risk, the SRA used the external event features of the Kewaunee SPAR model to determine the additional risk for the finding due to initiating events such as fires, floods, and seismic events. The SRA performed a similar analysis as that above for external events only. The resultant delta CDF was 1.4E-7 predominantly due to fire. The total delta CDF was the sum of the internal and external event risk, or 5.6E-7. Regarding the potential risk contribution due to Large Early Release Frequency (LERF), The SRA used IMC 0609 Appendix H, Containment Integrity Significance Determination Process. Kewaunee Station is a 2-loop Westinghouse PWR with a large dry containment. Sequences important to LERF include steam generator tube rupture events and inter-system LOCA events. These were not the dominant core damage sequences for this finding. The SRA determined that the risk due to internal and external events, and LERF to be of very low safety significance (GREEN). The licensee documented this violation in CR 0190063 and conducted an apparent cause determination. The licensee took immediate corrective actions upon discovery to restore the B SI train. At the end of the inspection period, the license continued to implement long term corrective actions from the causal evaluation.
05000305/FIN-2012009-0431 December 2012 23:59:59KewauneeNRC identifiedFailed to Consider Multiple Failures in ES-1.3, Transfer to Containment Sump Recirculation as Required by ITS, Section 5.4.1b)The inspectors identified a finding of very low safety significance and an associated Non-Cited Violation (NCV) of Improved Technical Specification (ITS), Section 5.4.1b because Procedure EOP ES-1.3, Transfer to Containment Sump Recirculation, Revision 36 did not establish the necessary actions as required. Specifically, the licensee failed to ensure Procedure EOP ES-1.3 contained the necessary actions for establishing containment sump recirculation for a large loss of coolant accident (LLOCA) with a concurrent failure of safety injection (SI) Pump A and the inability to establish containment sump recirculation using residual heat removal (RHR) Train B. The licensee entered this finding into their Corrective Action Program as CR491773, 2012 CDBI Identified Issue with EOP ES-1.3, dated October 15, 2012. In addition, the licensee initiated a procedure change to ES-1.3 to revise the procedure to allow transfer to containment sump recirculation without SI flow interruption to the RCS for a LLOCA with a failure of SI Pump A and with RHR Train B unable to be aligned for containment sump recirculation. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstones attribute of procedure quality and affected the cornerstones objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to ensure the procedure for establishing containment sump recirculation for a LLOCA contained the necessary actions for potential equipment failures. Since the finding resulted in the potential for a loss of the containment sump recirculation function during a LLOCA for certain equipment failures, the inspectors determined a Detailed Risk Evaluation was required. Based on the Detailed Risk Evaluation, the Senior Reactor Analysts determined that the delta core damage frequency for the finding was 1.0E-10/yr and was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000305/FIN-2012005-0231 December 2012 23:59:59KewauneeNRC identifiedEmergency Diesel Generator Starting Air System Quality Classification ErrorThe inspectors identified a finding of very low safety significance and associated non-cited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to assure that appropriate quality standards were specified and included in design documents, and that deviations from such standards were controlled. Specifically, during plant design change (DC) 10-011011, EDG Ventilation Air Supply Modification (Capital), the licensee failed to assign appropriate quality classifications to components that were relied upon to maintain a safety-related (SR) pressure boundary. As a short term corrective action, the licensee closed an upstream SR valve to restore the SR pressure boundary. The licensee entered the issue into their corrective action program (CAP) and assigned an apparent cause evaluation (ACE) to identify and asses the cause of the incorrect quality classification. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Design Control, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to assign the appropriate SR quality classification to components that were relied upon to maintain a SR pressure boundary. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions , dated June 19, 2012. The licensee confirmed that the emergency diesel generator (EDG) starting air system remained operable but non-conforming; therefore, the inspectors answered Yes to Mitigating Systems Screening question number 1, and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, work control, because the licensee did not maintain interfaces with offsite organizations, and communicate, coordinate, and cooperate with each other during activities in which interdepartmental coordination is necessary to assure plant and human performance. Specifically, during the licensees acceptance review of the offsite vendors work, the licensee questioned the new component qualifications; however, follow-up coordination and communications between the offsite vendor preparer and the licensee reviewer were not adequate to resolve the incorrect component classification.
05000305/FIN-2012005-0131 December 2012 23:59:59KewauneeNRC identifiedEAL Table Utilizes 1955 IGLD for Emergency Event ClassificationThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.54(q), Emergency Plans, for failing to follow and maintain the effectiveness of an emergency plan that met the requirements of emergency planning standard 10 CFR 50.47(b)(4). Specifically, the licensee failed to maintain emergency action levels (EALs), EAL bases, and abnormal operating procedures (AOPs) that ensured a declaration was made when Lake Michigan Levels reached low levels. Plant elevations for the Kewaunee Power Station (KPS) were based on Lake Michigan Levels in reference to the 1955 International Great Lakes Datum (IGLD). While the EALs, EAL bases, and AOPs listed the correct lake elevation in reference to the 1955 IGLD, all current alternate source information utilized by the KPS staff to declare an EAL was based on the 1985 IGLD Lake Michigan Level, a difference of 0.7 feet higher than the 1955 IGLD. Neither the EAL, EAL bases, nor AOP directed KPS staff to subtract 0.7 feet from the alternate source information used to declare an EAL; therefore, KPS staff would not have classified an emergency condition. The licensee entered this performance deficiency into their CAP as CR500145, and put in place interim corrective actions to ensure the correct value was obtained. The inspectors determined that from August 2001, to December 14, 2012, the licensees failure to ensure that EALs HU1 and HA1 were properly declared during an actual event was a performance deficiency. This finding was determined to be more than minor because it was associated with the emergency response organization (ERO) performance attribute of the Emergency Preparedness Cornerstone, and adversely affected the cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. This finding was evaluated in accordance with IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, using Figure 5.4-1, Significance Determination for Ineffective EALs and Overclassification, dated February 24, 2012, and resulted in a Green finding as an Unusual Event and Alert event would not have been declared. The inspectors concluded there was no cross-cutting aspect with this finding because the failure occurred in 2001, and was not representative of current performance.
05000305/FIN-2012004-0230 September 2012 23:59:59KewauneeNRC identifiedInadequate Degradation Evaluation of Leaking Refueling Water Storage TankThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to adequately evaluate a through-wall flaw in the refueling water storage tank (RWST). Specifically, the licensee failed to perform a degradation growth evaluation as required by ASME Code Case (CC) N-705, Section 2.5, and failed to use the appropriate acceptance criteria as required by CC N-705, Section 5, in their operability determination (OD) of the leaking RWST. The licensee subsequently took corrective actions to revise the RWST OD 492 to include the CC N-705, Section 2.5-required calculation of the degradation growth considering all the appropriate acceptance criteria described in CC N-705, Section 5. The licensees final revision to the POD adequately calculated the allowable time (Tallow) for RWST operation per CC N-705 to be ~25 months. This issue was entered into the licensees CAP as CR482703. The finding was determined to be more than minor, and determined to be a performance deficiency, because if left uncorrected, it could become a more significant safety concern. This finding was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance (reliability), and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012 and Appendix A, The Significance Determination Process (SDP) for Findings At-Power Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. Because the licensee intended to repair/replace the RWST weld during the next RFO, the inspectors answered No to all of the Mitigating Systems Screening questions, and screened the finding screened as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure that personnel, equipment, procedures, and other resources were available and adequate to assure nuclear safety, particularly those necessary for maintaining long-term plant safety by maintenance of design margins. Specifically, the licensee failed to provide adequate resources to verify the adequacy of the RWST degradation evaluation and supporting calculations. The inspectors determined the primary cause of this finding based upon discussions with the licensees engineering staff.
05000305/FIN-2012404-0130 September 2012 23:59:59KewauneeNRC identifiedSecurity
05000305/FIN-2012004-0430 September 2012 23:59:59KewauneeNRC identifiedEffective Dose Equivalent DeterminationThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR 20.1201(c) for the failure to accurately assess occupational dose specific to effective dose equivalent for external exposure (EDEex) determinations. The issue has been entered into the licensees CAP as CR487980. Corrective actions include procedural review and required revisions, a review of previous EDEex calculations and appropriate adjustments, and issuance of industry operating experience. The inspectors reviewed the guidance in IMC 0612, and determined that the finding was more than minor because it was associated with the program and process attribute of Occupational Radiation Safety Cornerstone and affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that inaccurate radiation monitoring affects the licensees ability to control and limit radiation exposures. Using IMC 0609, Attachment C, of the Occupational Radiation Safety SDP, the inspectors determined that the finding was of very low safetysignificance because the finding did not involve: (1) As-Low-As-Is-Reasonably- Achievable (ALARA) planning and controls; (2) a radiological overexposure; (3) a substantial potential for an overexposure; or (4) a compromised ability to assess dose. Because this finding was of very low safety significance, was not repetitive or willful, and was entered into the CAP, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. The primary cause of this finding was related to the cross-cutting aspect of human performance, resources. Specifically, the licensee failed to provide complete and accurate procedures to the radiation safety staff.
05000305/FIN-2012004-0530 September 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50.65(a)(4) states, in part, that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to this, from 2006 until February 28, 2012, licensee failed to adequately assess risk on multiple occasions because ventilation damper TAV-82 was modeled as open in the licensees risk model instead of the actual position of closed. The open cross-connect damper was credited as providing cooling to the opposite trains 480-Volt safeguards bus during maintenance on the room coolers. When the risk model was corrected, the daily risk changed from green to red for various maintenance activities that been previously performed. The inspectors and Region III SRAs assessed risk using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The licensee calculated the combined incremental core damage probability deficit to be 8.7E-7 and the combined incremental large early release probability deficit to be 1.1E-7. The licensee also credited one risk management action. The inspectors and Region III SRAs independently reviewed the calculation and determined that the finding was of very low risk significance (Green). The licensee documented this violation in CR464332 and CR479176. The licensee took immediate corrective actions to update damper position in the daily risk model.
05000305/FIN-2012004-0330 September 2012 23:59:59KewauneeNRC identifiedInadequate Commercial Grade Dedication Leads to Premature Bearing FailureThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, because the licensee failed to adequately dedicate a commercial-grade item for use in a safety-related (SR) application. Specifically, the licensee failed to properly dedicate two motors for use in SR fan coil units (FCUs). The finding was entered into the licensees CAP as CR 487269. The licensee assigned an ACE to determine the organizational failures that led to the violation. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone and adversely affected the design control attribute objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to properly dedicate a commercial-grade item, the FCU motor, caused the non-functionality of the TBFCU 1B and the inoperability of the train B AFW system. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power Exhibit 2, Mitigating Systems Screening Questions. The inspectors concluded that the TS equipment supported by TBFCU 1B, which consisted of buses 61, 62, and the train B AFW system, may have been inoperable for a time period greater than the respective TS action statements, and therefore answered Yes to Mitigating System Cornerstone Question 3, and determined that a detailed risk assessment was necessary. The inspectors worked with RIII senior reactor analysts (SRAs) and concluded that the finding is of very low safety significance (Green). Specifically, the licensee had abnormal operating procedures (AOPs) in place to prop open doors upon a complete loss of safeguards alley room cooling and supporting calculations that concluded the affected equipment would perform for the 24-hour probabilistic risk assessment (PRA) mission time without the TBFCU. The inspectors concluded there was no cross-cutting aspect with this finding because the failure to properly dedicate the motors occurred in 2005, and was not representative of current performance.
05000305/FIN-2012004-0730 September 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50, Appendix B, Criterion XVI, states in part that measures shall be established to assure that conditions adverse to quality such as failures or malfunctions are promptly identified and corrected. Contrary to the above, in April 2008, the licensee identified that the nuclear source and intermediate range detector cable for channels N31 and N35 exhibited a leak during pressure testing to ensure the integrity of the cable. At that time, without further investigation, the leak was attributed to a bad connector interface without any further action. On April 15, 2012, the same pressure testing was performed and identified the same leakrate; however, additional evaluation was performed that determined the leak was not from the connector but from the cable itself (this issue is discussed further in Section 4OA3.1 of this report). Therefore, the licensee determined that in 2008 they failed to promptly correct this condition adverse to quality. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power Exhibit 1, Initiating Events Screening Questions. The inspectors answered no to all the screening questions and determined the finding screened as very low significance (Green). The licensee documented this violation in CR470789 and conducted a causal evaluation.
05000305/FIN-2012004-0630 September 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTechnical Specification 5.4.1, Procedures, requires, in part, that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Section 2.0, General Plant Operating Procedures, requires, in part, that procedures for power operation and process monitoring be implemented. Procedure OP-AA-300, required, in part, that licensed reactor operators closely monitor parameters for reactor coolant temperature, reactor power, axial flux difference, rod insertion limits and turbine power and report trends to the crew during reactivity maneuvers. Contrary to the above, on May 10, 2012, during a planned load increase from 29 to 49 percent power following a reactor startup, reactor coolant system average temperature was allowed to increase 6.2 degrees Fahrenheit above the reference temperature resulting in a deviation alarm and temperature being 2.2 degress Fahrenheit above the critical parameter band. Immediate action was directed by the Shift Supervisor to ensure reactor coolant system temperature was brought back into the critical parameter band. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At-Power Exhibit 1, Initiating Events Screening Questions. The inspectors answered no to all the screening questions and determined the finding screened as very low significance (Green).
05000305/FIN-2012004-0130 September 2012 23:59:59KewauneeNRC identifiedInadequate Risk Assessment During Heavy Lift Over Service Water SystemThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, paragraph (a)(4), for the failure to properly assess and manage risk when the licensee lifted a 44,000-pound motor over the portion of the service water (SW) screenhouse containing the train B SW header. Specifically, the licensee should not have credited train B emergency diesel generator (EDG) in the power availability portion of their shutdown safety assessment (SSA). The licensee entered this into their corrective action program (CAP) as CR491721. The licensee assigned an apparent cause evaluation (ACE) to determine the organizational failures. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of protection against external factors, 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 licensee did not maintain a shutdown probabilistic risk analysis (PRA) model, so an incremental core damage probability could not be estimated for the specific plant conditions. For this reason, the inspectors determined that IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, could not be used. The inspectors used IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, dated April 12, 2012, which directs inspectors to perform a bounding analysis. The inspectors conservatively assumed that EDG B was actually lost and used Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Phase 1 Operational Checklists for Both PWRs (Pressurized-Water Reactors) and BWRs (Boiling-Water Reactors), dated May 25, 2004, to determine the risk. Specifically, the inspectors selected Checklist 4, PWR Refueling Operation, and found, for the power availability function, the licensee met the checklist objective to comply with TSs because only one EDG was required to be operable, and EDG A remained operable and protected. Therefore, the inspectors determined that the finding was determined to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, work control, because the licensee did not plan and coordinate work activities consistent with nuclear safety. Specifically, the licensee failed to incorporate risk insights and plant SSCs into their evaluation.
05000305/FIN-2012008-0130 September 2012 23:59:59KewauneeNRC identifiedBattery Rack Configuration Not In Accordance With Design BasisThe inspectors identified a finding of very low safety significance (Green) and an associated non-citied violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because the licensee failed to ensure that the configuration of the safeguards battery racks was in accordance with the design basis Seismic Category I qualification. Specifically, the Seismic Category I qualification specified that the battery rack end rails be snug against the battery. The inspectors found gaps greater than 1/8 inch and up to approximately 3/8 inch. The vendor instructions directed that the rails should be within 1/8 inch. The licensee entered this into the Corrective Action Program as CR489958 and CR487875 and took short term corrective actions to adjust the battery rack end gaps to within 1/8 inch, and assigned an apparent cause evaluation, which was not complete at the end of the inspection period. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external factors and affected the cornerstone objective to ensure the availability, reliability, and Corrective Action Programability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to ensure that the batteries were constrained from sliding along the rack to avoid over stressing the battery terminals, battery casing, or rack ends. The inspectors determined the finding could be evaluated using the Significance Determination Process in accordance with Inspection Manual Chapter 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions. The inspectors answered Yes to question 1, and screened the finding as having very low safety significance (Green). The inspectors did not assign a cross-cutting aspect because the installation of the battery racks occurred in 2008, and was not representative of current performance.
05000305/FIN-2012003-0130 June 2012 23:59:59KewauneeSelf-revealingFailure to Utilize Work Order for Temporary Weld Repair on ASME Code, Class 2 PipingA 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 for the failure to accomplish Temporary Modification (TMOD) 2012-11 in accordance with Work Order (WO) KW100894696 and the associated weld data sheet and map. Specifically, licensee personnel failed to utilize the WO instructions, weld data sheet and weld map when welding a temporary NRC-approved clamp on American Society of Mechanical Engineers (ASME) Code Class 2 residual heat removal (RHR) piping. The failure to use the required documentation to perform the work resulted in the worker creating a second through wall leak on the ASME Code, Class 2 RHR piping upstream of valve RHR-600. The licensee entered the issue into its corrective action program (CAP) as condition report (CR) 472915 and permanently corrected both through wall leaks on the RHR system piping following the approval of a second proposed alternative, without incident on May 5, 2012. At the end of the inspection period, the licensee continued to perform an apparent cause evaluation (ACE) to determine the causes for the organizational failures that occurred. The finding was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because the finding was associated with the Mitigating Systems Cornerstone attribute of human error (pre-event) and adversely affected the cornerstone objective to ensure the reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors determined that the finding could be evaluated in accordance with IMC 0609, Appendix G, Shut-down Operations SDP, dated February 28, 2005. The inspectors used Checklist 1, PWR Hot Shutdown Operation: Time to Core Boiling <2 Hours, contained in Attachment 1 and determined that the finding affected core heat removal guidelines I.B(1), Procedures, and I.C(2), Equipment. The inspectors screened the finding as very low safety significance (Green) because it did not degrade the licensees ability to establish an alternate core cooling path if decay heat removal could not be re-established and, therefore, did not require a phase 2 or phase 3 analysis. This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, the inspectors identified that the pre-job brief conducted by supervision and management for this work did not include a review of the WO, weld sheet, or weld map and did not convey accurate information regarding the significance of the activity, the type of weld to be performed and the system conditions where the weld was performed.
05000305/FIN-2012003-0230 June 2012 23:59:59KewauneeSelf-revealingLoose Cable Clamp Caused Loaded Spent Fuel Upender to Unintentionally LowerA finding of very low safety significance (Green) and associated NCV of Technical Specification (TS) 5.4.1, Procedures, was self-revealed because procedure MCM-FH-001, Repair of the Fuel Transfer System, was inadequate. Specifically, the procedure did not contain torque specifications for tightening the upender frame cable clamps and, on April 23, the cable for the spent fuel pool (SFP) upender slipped through the cable clamps and allowed the upender containing a fuel assembly to descend approximately 12 inches. The licensee confirmed that no damage occurred to the fuel assembly and placed procedure MCM-FH-001 on administrative hold to prevent its use until it could be updated with the appropriate torque specifications. At the end of this inspection period, the licensee was performing an ACE to determine the causes of the event, and develop corrective actions. The finding was determined to be more than minor because, if left uncorrected, the finding had the potential to lead to a more significant safety concern. Specifically, the upender containing the fuel assembly could have fallen from the near-full vertical position to the horizontal position. The inspectors evaluated the finding by applying the SFP questions in the Fuel Barrier column of Table 4a, located in IMC 0609, Attachment 4, dated January 10, 2008. The inspectors answered No to all three questions and determined that the finding was of very low safety significance (Green). The finding has a cross-cutting aspect in the areas of problem identification and resolution, operating experience (OE), because the licensee failed to communicate to affected internal stakeholders in a timely manner relevant external OE. Specifically, the licensee failed to discuss available and relevant OE related to the failure to appropriately torque cable clamps on an SFP upender.
05000305/FIN-2012003-0430 June 2012 23:59:59KewauneeNRC identifiedIncorrect Leakage Requirement Submitted for a Proposed Temporary Pipe ClampA Severity Level (SL) IV NCV of 10 CFR Part 50.9(a), Completeness and Accuracy of Information, was identified by the inspectors for the failure of the licensee to provide complete and accurate information in all material respects to the Commission in licensee Request RR-2-3, dated April 29, 2012 (ADAMS Accession No. ML12122A138). As part of a license amendment for a proposed temporary deviation from the requirements of 10 CFR 50.55a and ASME Code, Section XI, the licensee incorrectly stated the allowable leakage from the temporary clamp in transition from Mode 5 to 4 was governed by TS 5.5.2, Primary Coolant Sources Outside Containment, and proposed an allowable leakage value of 5.5 gallons per hour (gph). After licensee Request No. RR-2-3 was verbally approved by the NRC on April 30, 2012, the inspectors and NRC staff determined that the governing leakage requirement was no leakage in Mode 4 for the clamp as required by TS 3.4.13, Reactor Coolant System Operational Leakage. The performance deficiency was determined to be more than minor in accordance with the NRC Enforcement Policy and Enforcement Manual because the NRC identified the performance deficiency, the NRC relied on the information provided in a licensing decision, and the misinformation was identified after the NRC relied on the information in its licensing decision. Because violations of 10 CFR 50.9 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the ROP SDP. Because the performance deficiency, specifically a failure to submit complete and accurate information, was not an ROP finding per IMC 0612, Appendix B, Issue Screening, a cross-cutting aspect was not assigned to this violation. The severity of the violation was mitigated because of the facts surrounding the licensees implementation of Request No. RR-2-3.
05000305/FIN-2012003-0330 June 2012 23:59:59KewauneeNRC identifiedFailure to Provide Adequate Suppression in Cable Spreading AreaThe inspectors identified a finding of very low safety significance (Green) and associated NCV of 10 CFR Part 50, Appendix R, Section III.G.3, for the licensees failure to provide adequate fire suppression coverage for fire zone AX-32. Specifically, the licensee failed to provide required fire suppression coverage for safe shutdown functions of source range monitoring, isolation of a steam generator (SG) blowdown line, and pressurizer level instrumentation in the cable spreading area. The licensee entered the issue into the CAP, designated manual backup from hose stations, and implemented an hourly fire watch for the radiation protection office (RP) in fire zone AX-32. The inspectors determined that the finding was more than minor because the failure to provide suppression for redundant trains of safe shutdown equipment increased the likelihood that alternative shutdown methods would have to be used in the event of a fire. The finding was of very low safety significance based on a Phase 3 significance determination analysis. The finding has a cross-cutting aspect in the area of problem identification, corrective action program, because the licensee did not take appropriate corrective actions to address the inadequate suppression system in fire zone AX-32
05000305/FIN-2012002-0231 March 2012 23:59:59KewauneeNRC identifiedPotential Mobile Crane Heavy Load Risk Modeling ErrorDuring plant tours, the inspectors observed the licensees preparations for a heavy lift of circulating water pump motor B from the SW screen house through the screen house roof hatch to a motor stand outside. The inspectors had previously identified during daily plant status activities that EDG train A was tagged out for maintenance; and during discussions with the maintenance personnel performing the motor lift, found that the load path for the 44,000 lb motor went over both buried cables for the SW pumps train B, which were a needed support system for EDG train B. The inspectors questioned the licensee about the load path and found that the licensee had not recognized the location of the underground cables during the planning phase of the load lift, and had not evaluated a load drop over the cables and what affect the drop would have on the cables. The licensee suspended the lift until the next day, when EDG train A was operable and the licensee had an opportunity to perform a risk assessment for the heavy load travelling over the SW cables. The inspectors reviewed the risk assessment for the heavy lift and questioned whether the probability used for a load drop was conservative. At the conclusion of the inspection period, the inspectors needed additional information to determine if a performance deficiency occurred. As a result, this item was considered unresolved
05000305/FIN-2012002-0531 March 2012 23:59:59KewauneeNRC identifiedInadequate Procedure For Technical SpecificationThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have appropriate procedures to complete TS-required surveillances. Specifically, OSP-CCI-004, Containment Isolation Valve Verification, did not contain adequate steps to complete a TS-required airlock door check and the procedure did not include six manual containment isolation valves (CIVs) that should have been included in the procedure for position verification. The licensee corrected the procedure and entered the issue into its CAP as CRs 464355, 464494, and 467560, and planned to perform an ACE. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because the finding was associated with the Barrier Integrity Cornerstone attribute of procedure quality and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, inspectors found seven examples in OSP-CCI-004 where either the procedure steps were not adequate or CIVs were missing that should have been included in the procedure for position verification. 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, dated January 10, 2008. The inspectors answered No to the Containment Barrier questions and screened the finding as having very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety was supported. Specifically, OSP-CCI-004 did not get an approval review during the procedure review process and the supervisory review that was conducted did not identify the procedural errors.
05000305/FIN-2012002-0631 March 2012 23:59:59KewauneeNRC identifiedFailure to Submit LER Per 10 CFR 50.73The inspectors identified an SL IV NCV of 10 CFR 50.73(a)(2)(vii) for the failure of the licensee to report an event where a single cause or condition caused two independent trains to become inoperable in a single system designed to control the release of radioactive material. Specifically, the licensee failed to report that both trains of shield building ventilation (SBV) were inoperable due to a single cause, because both trains contained unqualified control card standoffs that were needed to maintain the seismic qualification and operability of the system. The licensee entered this into their CAP as CR429469, planned to perform an ACE, and was drafting an update to Licensee Event Report (LER) 05000305/2011-005. The inspectors determined that the failure to report the event in accordance with 10 CFR 50.73 was a performance deficiency. Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process (ROP) SDP. Because the performance deficiency, a failure to report, was not an ROP finding per IMC 0612, Appendix B,` Issue Screening,` a cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section The inspectors identified an SL IV NCV of 10 CFR 50.73(a)(2)(vii) for the failure of the licensee to report an event where a single cause or condition caused two independent trains to become inoperable in a single system designed to control the release of radioactive material. Specifically, the licensee failed to report that both trains of shield building ventilation (SBV) were inoperable due to a single cause, because both trains contained unqualified control card standoffs that were needed to maintain the seismic qualification and operability of the system. The licensee entered this into their CAP as CR429469, planned to perform an ACE, and was drafting an update to Licensee Event Report (LER) 05000305/2011-005. The inspectors determined that the failure to report the event in accordance with 10 CFR 50.73 was a performance deficiency. Because violations of 10 CFR 50.73 are considered to be violations that potentially impact the regulatory process, they are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process (ROP) SDP. Because the performance deficiency, a failure to report, was not an ROP finding per IMC 0612, Appendix B,` Issue Screening,` a cross-cutting aspect was not assigned to this violation. Per the NRC Enforcement Policy, Section 6.0,` Violation Examples,` a failure to submit a required LER is categorized as an SL IV violation.
05000305/FIN-2012002-0331 March 2012 23:59:59KewauneeNRC identifiedPotential Inadequate Commercial Grade DedicationOn January 21, the licensee discovered elevated noise on TBFCU motor 1B. Subsequent vibration data indicated that the outboard bearing was in imminent failure. The licensee declared the FCU inoperable, replaced the motor, and returned the FCU to a functional status at 7:41 a.m. on January 22. A failure analysis of the motor and bearing determined that the apparent cause of the motor failure was incorrect assembly of the motor after installation of sealed bearings at a local vendor. The licensee also considered an undetected original manufacturing defect of the motor that resulted in excessive thrust load on the bearing was also a potential apparent cause because it could not be ruled out. The motor was procured as NSR; the original bearings were replaced at a local vendor for double shielded bearings; and then the motor was commercially dedicated. Since the opposite train motor was procured at the same time using the same process, the licensee replaced that the motor and also sent it off for analysis. The inspectors were concerned that the incorrect assembly error was not discovered during the commercial dedication process and inspected the commercial dedication package for the motor. The inspectors found that the critical characteristics for the dedication process may have been inadequate to identify the incorrect assembly at the vendor; however, the licensee informed the inspectors near the end of the inspection period that the failure analysis for the second motor may have had different conclusions than the first. The inspectors were unable to evaluate the second failure analysis prior to the end of the inspection period, and needed additional time to determine if a performance deficiency existed. As a result, this item was considered to be unresolved pending a review of the second failure analysis.
05000305/FIN-2012002-0431 March 2012 23:59:59KewauneeNRC identifiedInadequate Post-Maintenance Test of Motor ReplacementsThe inspectors identified a finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1, Procedures, which required, in part, that written procedures shall be implemented covering the applicable procedures recommended in Regulatory Guide (RG) 1.33, Revision 2, Appendix A. Specifically, Procedure GNP-08.02.12, Post-Maintenance Testing/Operations Retest, stated, in part, that the post-maintenance tests (PMTs) were performed upon completion of maintenance activities, and demonstrated that the identified deficiency was repaired, and that no new deficiency was created. On July 4, 2011, the licensee replaced the spent fuel pool (SFP) pump motor B, and failed to conduct an adequate PMT, which demonstrated no new deficiency was created. The PMT only tested the replaced motor and failed to include testing of the pump to ensure that no new deficiency was created. The licensee entered the issue into its corrective action program (CAP) as condition reports (CRs) 464645, 466183, and 466215, and planned to perform an apparent cause evaluation (ACE) and take corrective actions. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because, if left uncorrected, the failure to perform adequate PMT on motor replacements would have the potential to lead to a more significant safety concern. The inspectors determined that 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, dated January 10, 2008. The inspectors answered No to the Reactor Coolant System or Fuel Barrier Questions related to Spent Fuel Pool Issues, and screened the finding as having very low safety significance (Green). The inspectors also determined that this finding had a cross-cutting aspect in the area of human performance, resources, because the licensee did not ensure the PMT procedure guidance related to motor replacements was adequate and accurate to assure nuclear safety
05000305/FIN-2012402-0131 March 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified Violation
05000305/FIN-2012002-0731 March 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances. Contrary to this, on November 28, 2011 and December 5, 2011, procedure ER-KW-STP-DGM-002A and -002B, Post-Tie-in Testing for DC KW-10-01101, Train A and Post-Tie-in Testing for DC KW-10-01101, Train B which were one time use special test procedures, were not appropriate to the circumstances. Specifically, the procedures determined the leak rate for the SR backup air supply for the EDG ventilation dampers and because of a calculation error, the procedure only required a test duration of 2 hours instead of a test duration of 24 hours. The licensee corrected the operations version of the procedure and, during the quarterly test of train A on February 19, 2012, found the leakage to be greater than expected, and determined the installed bottles would not have enough air to support operation of the dampers for 7 days without manual operator action. The licensee repaired the leaking components and entered this into the CAP as CRs 460168 and 463265. The air bottles contained enough air to operate the dampers for approximately 3.5 days, which exceeded the 24-hour PRA mission time of the EDG, and therefore, the inspectors screened this finding as having very low safety significance.
05000305/FIN-2012002-0831 March 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationAppendix R Fire Door Pinned Open and Unattended License condition 2.C(3) required the licensee to implement and maintain, in effect, all provisions of the approved FP program as described in the licensees fire plan, and as referenced in the USAR, and as approved through Safety Evaluation Reports dated November 25, 1977, and December 12, 1978, and supplement dated February 13, 1981. Appendix B of the KPS Fire Protection Program Plan lists the 1975 edition of NFPA-80, Fire Doors and Windows, as an applicable NFPA code. NFPA-80 states, in part, that a fire door shall be closed and latched at the time of fire. Contrary to the above, on May 15, 2010, the licensee failed to implement and maintain, in effect, all provisions of the approved FP program as described in the licensees Fire Plan. Specifically, the licensee found fire door 265 unattended and propped open with the doors bottom bolt engaged in the floor. The licensee closed fire door 265 and entered this into their CAP as CR381342. The inspectors walked down both sides of the fire door using guidance from IMC 0609, Appendix F, Fire Protection Significance Determination Process, and concluded that no creditable fire scenario existed that would allow fire to pass from either the alternate fire area to the dedicated fire area or from the dedicated fire area to the alternate fire area. The inspectors screened this finding as having very low safety significance.
05000305/FIN-2012002-0131 March 2012 23:59:59KewauneeNRC identifiedIncorrectly Modeled Ventilation Damper Changes Daily Risk ColorDuring the daily review of CRs, inspectors reviewed CR464332, which documented that cross connect damper TAV-82, which connected two trains of turbine building ventilation, was modeled in the licensees probabilistic risk assessment (PRA) tool as open, when the actual position of the damper was closed. When the licensees PRA tool was updated with the correct damper position, the daily risk calculated to comply with 10 CFR 50.65(a)(4) increased from green to orange on train B, and from green to red on train A. The licensee reevaluated the model taking credit for a nonsafety-related (NSR) general turbine building ventilation system that also provided air flow to the two rooms in question and the risk was reduced. The updated model for the removal of a TBFCU from service maintained green risk for train B and only increased train A to yellow risk. At the conclusion of the inspection period the inspectors needed additional information to determine if an error in the original inputs to the base PRA tool, versus an error in the daily application of the PRA tool, was a violation of 10 CFR 50.65(a)(4). As a result, this item was considered unresolved
05000305/FIN-2012002-0931 March 2012 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationFailure of Licensed Operator to Report a Medical Condition On December 21, 2005, a KPS employee was diagnosed with sleep apnea and was prescribed a CPAP (Continuous Positive Airway Pressure) device to aid in correcting sleep patterns. The employee was subsequently enrolled in the KPS initial operator license training to obtain an NRC operators license. The employee did not report the use of the CPAP device to the site nurse when she was preparing the applicants medical certification in conjunction with his application to the NRC to become a licensed operator. This prescribed device was used to treat sleep apnea and was a condition requiring notification of the NRC. The employee was unaware of the requirement to report the use of CPAP devices. The employee was issued an NRC operating license on March 2, 2011, without a requirement to use therapeutic devices as directed. The NRC issued the operators license without knowledge of the operators medical condition. If the NRC had been informed of this medical condition, the NRC would have required a medical restriction be included in the operators license. This was a potential violation of 10 CFR 50.9, Completeness and Accuracy of Information. On July 6, 2011, the operator notified the site nurse of his prescription for a CPAP device. On August 1, 2011, the licensee notified the NRC of the need to add this condition to the operators license. On October 24, 2011, the NRC amended the operators license to include the license condition, must use therapeutic devices as prescribed to maintain medical qualifications. Since the license had previously been submitted without the license condition, and the NRC doctor determined that the license needed to be revised; the original license submittal was incomplete/inaccurate. As such, this was a violation of 10 CFR 50.9. Because this issue impacted the ability of the NRC to perform its regulatory oversight function, the regulatory significance was determined using the traditional enforcement process. The inspectors determined that the operators medical condition did not adversely affect the operators ability to safely operate the facility even though the operators license was incorrect. The operators performance was monitored and evaluated as satisfactory during periodic testing and requalification testing. As such, the NRC determined this to be an SL IV violation, which may be dispositioned as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. Corrective actions included a resubmitted NRC Form 396 for the operator, documenting the issue in CR435966, and performing an ACE. Additionally, the licensee conducts annual training regarding operator license restrictions, including the use of prescribed medication, therapeutic devices, and reporting of medical conditions. The inspectors review of this issue was considered to be a part of the original inspection effort, and as such did not constitute any additional inspection samples.
05000305/FIN-2011005-0231 December 2011 23:59:59KewauneeNRC identifiedUnanalyzed Flood Source From Technical Support Center BuildingA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the failure to identify and analyze a potential flooding source that was within the Kewaunee licensing basis. Specifically, during the internal flood basis reconstitution in 2005, the licensee failed to realize and assess the potential for fire main piping in the technical support center (TSC) building to be ruptured during a tornado or seismic event. Water from a ruptured fire main had the potential to accumulate in the basement of the TSC building, flow into the attached auxiliary building, and potentially affect safety-related (SR) equipment. The licensee initiated a condition report (CR) and completed calculations and analyses to demonstrate the existing barriers, although not credited at the time, were adequate to support this internal flood scenario. In addition, the licensee performed an extent of condition analysis to determine if any additional internal flood scenarios were missed. The inspectors determined that the finding was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated December 24, 2009, because it was associated with the Mitigating Systems (MS) Cornerstone attribute of Equipment Performance, and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to analyze all potentially credible internal flood sources could affect the availability of SR systems. The inspectors determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the MS Cornerstone, dated January 10, 2008. The inspectors answered No to the MS questions and screened the finding as having very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the internal flood design basis reconstitution occurred in 2005 and the inspectors determined that there was not an opportunity to identify this deficiency in the past three years.
05000305/FIN-2011005-0831 December 2011 23:59:59KewauneeNRC identifiedProcedures for Reduced Inventory Operations were not Appropriate to Preclude Air EntrainmentThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to establish procedures for reduced inventory operations that were appropriate to manage gas accumulation. Specifically, the procedures did not preclude air entrainment into the residual heat removal (RHR) and reactor coolant systems (RCSs). This finding was entered into the licensees corrective action program (CAP). The licensee\\\'s immediate corrective actions included calculating the instrument inaccuracies for RHR flow and refueling level instrument loops, referencing the level inaccuracies based on inactive flow in RCS loops in the associated procedures, evaluating levels, and updating the procedures with a new graph. The performance deficiency was determined to be more than minor because it was associated with the Initiating Event Cornerstone attribute of procedure quality, and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shut-down operations. Specifically, the failure to establish procedures for reduced inventory operations that were appropriate to preclude air entrainment did not limit the likelihood of events that result from adverse air entrainment into the RHR and RCSs. The finding screened as having very low safety significance (Green) because the Region III Senior Reactor Analysts determined that it reasonably met the safety functions of core heat removal, RCS inventory control, power availability, containment control, and reactivity control; and there had been no actual air entrainment problems that had occurred using the procedures. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not thoroughly evaluate relevant operating experience. Specifically, the licensees evaluation of gas-related issues in response to NRC Generic Letter (GL) 2008-01 was deficient in that it did not consider vortexing during reduced inventory operations.
05000305/FIN-2011005-0331 December 2011 23:59:59KewauneeSelf-revealingTwo of Six Operating Crew Failures on the Simulator Operational Evaluation Portion of the 2011 Annual Requalification Operating TestA self-revealed finding associated with operating crew performance on the simulator during a licensee-administered requalification examination was identified. Two of the six crews evaluated during the annual operating tests failed to pass their simulator examinations. As immediate corrective action, the failed operating crews were remediated (i.e., the operating crews were re-trained and successfully re-tested) prior to returning to shift. The licensee entered this issue into the CAP as CR456328. The inspectors determined that the crew failures constituted a performance deficiency based on the fact that licensed operators are expected to operate the plant with acceptable standards of knowledge and abilities demonstrated through periodic testing as required by 10 CFR 55.59(a)(2). Two out of six crews of licensed operators failed to demonstrate a satisfactory understanding of the required actions and mitigating strategies required to safely operate the facility under normal, abnormal, and emergency conditions. The finding was more than minor because the performance deficiency potentially affects the Human Performance attribute of the MS Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding reflected the potential inability of the crews to take appropriate SR action in response to actual abnormal and emergency conditions. The perceived risk associated with the number of crews failing the annual operating test is provided in the Simulator Operational Evaluation matrix of IMC 0609, Appendix I, Licensed Operator Requalification SDP. The finding was of very low safety significance (Green) because only two of six of the operating crews failed; the failed operating crews were remediated (i.e., the operating crews were re-trained and successfully re-tested) prior to returning to shift; and because there was not a finding associated with operating crew failures during calendar year 2010. The cause of this finding was directly related to the cross-cutting aspect of personnel training and qualifications in the area of Human Performance Resources, in that the licensee failed to ensure the adequacy of the training provided to operators to assure nuclear safety.
05000305/FIN-2011005-0931 December 2011 23:59:59KewauneeLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states, in part, that activities affecting quality shall be prescribed and accomplished by documented instructions or procedures of a type appropriate to the circumstances. Contrary to this, prior to November 24, 2008, GMP-251, Common Electrical Preventative Maintenance Tasks (QA-1) did not contain documented instructions that were appropriate to the circumstances. Specifically, GMP-251 did not contain procedural guidance for adequately inspecting, cleaning, and lubricating all auxiliary contacts in SR breaker cubicles. On November 24, 2008, the licensee found that the control power fuses were blown on the train B screenhouse exhaust fan and the fan was not running. The licensees investigation found that higher than normal currents caused by the broken auxiliary contacts caused control power fuses to blow. The auxiliary contacts failed from long term wear and a lack of proper lubrication. The inspectors answered No to the MS Cornerstone questions and screened the finding as having very low safety significance (Green) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, Table 4a, dated January 10, 2008. The licensee documented this violation in CR119979. The licensee performed an apparent cause evaluation which determined that GMP-251 was inadequate and implemented corrective actions that included revision of the procedure.