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05000285/FIN-2013008-412013Q2Fort CalhounInappropriate Modification of Turbine Driven Auxiliary Feedwater Pump Back Pressure Protection TripThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with an inappropriate modification of the auxiliary feedwater system. Specifically, from April 2011 through February 2013, measures established by the licensee did not assure that the modification to remove the turbine driven auxiliary feedwater pumps exhaust back pressure trip, properly considered and addressed the open configuration of the pumps exhaust piping to prevent blockage of the exhaust piping. This issue has been entered into the corrective action program as Condition Report CR 2013-05026, and an immediate operability determination was performed. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected, the continued practice of modifying the facility without evaluating for adverse impacts had the potential to lead to a more significant safety concern. Specifically, unevaluated modifications to the facility could introduce adverse changes that result in systems not able to perform their intended safety function which would not be recognized. This finding was associated with the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-312013Q2Fort CalhounMultiple Examples of Operability Determinations That Lacked Adequate Technical JustificationThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, involving multiple examples of the licensees failure to perform an adequate operability determination as required by Procedure NOD-QP-31, Operability Determination Process. In each example, the team identified that the operability determination lacked adequate technical justification for why the structure, system, or component was operable with the degraded or nonconforming condition. Specifically, on January 24, 2012, June 6, 2012, December 27, 2012, January 22, 2013, and February 5, 2013, the operability determinations for Condition Reports CR 2012-00580, CR 2012-04973, CR 2012-20806, CR 2013-00907, and CR 2013-02260 were not performed in accordance with Procedure NOD-QP-31, Revision 49-53, Step 4.1.3 J. This issue has been entered into the corrective action program as Condition Reports CR 2013-08343, CR 2013-05596, CR 2013-08590, CR 2013-04163, and CR 2013-05353. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involving inadequate operability determinations occurred while in a shutdown condition, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective action program component. Specifically, the team identified that the licensee failed provide an adequate technical discussion such that a reasonable expectation of operability was demonstrated for several degraded or nonconforming conditions
05000285/FIN-2013008-322013Q2Fort CalhounMultiple Examples of Inadequate RISK-BASED Operability DeterminationsThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, involving multiple examples of the licensees use of probability or probabilistic risk assessment when performing operability determinations. The use of probability or probabilistic risk assessment when determining operability is contrary to Procedure NOD-QP-31, Operability Determination Process, Revision 49-53. Specifically, on January 26, 2012 and twice on February 21, 2013, the operability determinations performed for Condition Reports CR 2012-00626, CR 2013-03839, and CR 2013-03842 used probability and/or probabilistic risk assessment to justify the operability of structures, systems, and components. This issue has been entered into the corrective action program as Condition Reports CR 2013-05590, CR 2013-05466, and CR 2013-05597. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since the finding involved inadequate operability determinations that occurred while in a shutdown condition and involved plant equipment needed during shutdown conditions, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee failed to use conservative assumptions in decision making when performing operability determinations. Specifically, the licensee proposed that a degraded/nonconforming condition was safe by relying on a non-conservative assumption that an event such as a tornado generated missile or external flooding at the site were not likely to occur
05000285/FIN-2013008-332013Q2Fort CalhounInadequate Operability Determination Due to Failure to Establish Component Cooling Water System Leakage CriteriaThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, involving the licensees failure to follow procedures when evaluating the impact of component cooling water system leakage on the containment air coolers. Specifically, on October 6, 2010, and December 29, 2010, the operability determinations for Condition Reports CR 2010-04955 and CR 2010-06905 were not performed in accordance with Procedure NOD-QP-31, Operability Determination Process, Revision 43-44, Step 4.1.3 J, and consequently, failed to evaluate the impact of component cooling water system leakage on containment air coolers operability. This issue has been entered into the corrective action program as Condition Report CR 2013-05630. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with corrective action program component. Specifically, the team identified that the licensee failed provide an adequate technical discussion such that a reasonable expectation of operability was demonstrated for containment air coolers with known leakage in the component cooling water system
05000285/FIN-2013008-342013Q2Fort CalhounFailure to Follow ASME Code Requirements When Establishing New Pump Reference Values as Corrective ActionsThe team identified a non-cited violation of 10 CFR 50.55a, Codes and Standards, for the failure of the licensee to follow the ASME Code when establishing new reference curves as corrective action to address the performance of component cooling water pump AC-3A within the low required action range of the in-service testing program. Specifically, on July 29, 2011, the licensee failed to follow ASME Code, Subsection ISTB 6200(c), in that, the new reference curves were established without performing an analysis which included verification of the pumps operational readiness at a pump level and a system level, without determining the cause of the change in pump performance, and without an evaluation of all trends indicated by available data. The team confirmed that while the pump was inoperable from an in-service testing perspective during this period, required surveillance testing showed that pump flows and differential pressures were still sufficient to meet the assumptions used in the Fort Calhoun Station safety analysis. This issue has been entered into the corrective action program as Condition Report CR 2013-04010. This performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Mitigating Systems Cornerstone, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Since this finding was discovered during plant shutdown and involved plant equipment needed during shutdown conditions, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to fully evaluate the degraded performance of component cooling water pump AC-3A to ensure that resolutions correctly addressed causes of the degraded performance and the cumulative impact on system operational readiness
05000285/FIN-2013008-352013Q2Fort CalhounFailure to Correct Condition Adverse to Quality Associated with Corrective Action Program Procedures and the Operability ProcessThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to implement corrective actions to address inadequate procedures involving the degraded/nonconforming condition evaluation and operability determination process. Specifically, prior to March 1, 2013, the licensee failed to correct the procedural inadequacies associated with Procedure FCSG-24-3, Condition Report Screening, Revision 3, as identified in the root cause analysis for Condition Report CR 2012-09494. This issue has been entered into the corrective action program as Condition Report CR 2013-04380. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected, inadequate corrective action program procedures could become a more significant safety concern. This finding is associated with the Mitigating Systems Cornerstone. Since the finding was discovered while in a shutdown condition, the team used Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, and determined the finding to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to implement a corrective action program with a sufficiently low threshold. Specifically, although the licensee identified significant flaws in Fort Calhoun Station procedures while performing the root cause analysis for Condition Report CR 2012-09494, the licensee failed to initiate the appropriate corrective action documents to drive the necessary procedure changes
05000285/FIN-2013008-362013Q2Fort CalhounDeficient Evaluation of NRC Bulletin 88-04, Strong Pump Weak Pump Due to Failure to Consider the Effect of AFW Pumps Discharge Check Valves LeakageThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to properly evaluate NRC Bulletin 88-04, Potential Safety-Related Pump Loss, regarding the auxiliary feedwater pumps. Specifically, from November 28, 2010, through February 2013, the licensee failed to properly evaluate NRC Bulletin 88-04, for strong pump, weak pump, interaction regarding auxiliary feedwater pumps FW-6 and FW-10. The evaluation failed to consider pump-to-pump interaction that may result due to pump discharge check valve leakage. In addition, the licensee failed to re-evaluate the condition after surveillance testing performed on November 28, 2010, and September 1, 2012, identified leakage past both pump discharge check valves. This issue has been entered into the corrective action program as Condition Reports CR 2013-04680 and CR 2013-04806. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that appropriate corrective actions were promptly implemented
05000285/FIN-2013008-372013Q2Fort CalhounImproper Storage of the RAW Water to Auxiliary Feedwater Emergency Tank Fill HoseThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to properly store the raw water to emergency feedwater storage tank fill hose. Specifically, from July 1996 to February 27, 2013, the licensee failed to provide adequate instructions or procedures to ensure proper storage and temperature qualification of the auxiliary feedwater emergency fill hose. This issue has been entered into the corrective action program as Condition Report CR 2013-52276. This performance deficiency is more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-382013Q2Fort CalhounDeficient Evaluation for Known Degraded Conditions - AFW Pumps Discharge Check Valve Leakage and Potential Overpressure of AFW Pump Suction PipingThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to properly evaluate a known degraded condition regarding the auxiliary feedwater pump discharge check valve leakage and potential over-pressurization of the pumps suction piping. Specifically, from October 10, 2012, to March 15, 2013, the licensee failed to properly evaluate concerns regarding the auxiliary feedwater pump discharge check valves which resulted in the failure to implement adequate corrective actions to verify leak tightness of the check valves and prevent potential over pressurization of the pumps suction piping. This issue has been entered into the corrective action program as Condition Reports CR 2013-04806 and CR 2013-05018. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate it is unsafe in order to disapprove the action
05000285/FIN-2013008-392013Q2Fort CalhounFailure to Properly Implement Applicable ASME OM Code RequirementsThe team identified two examples of a non-cited violation of 10 CFR 50.55a.(f)(4)(ii), Codes and Standards, associated with the licensees failure to properly implement applicable code requirements for in-service testing of safety-related pumps and check valves. Specifically, prior to March 11, 2013, the licensee failed to ensure that the testing of safety-related pumps and valves met the requirements of the American Society of Mechanical Engineers Operation and Maintenance Code. The applicable Code for the current in-service test program is the 1998 Edition through the 2000 Addenda. This issue has been entered into the corrective action program as Condition Reports CR 2013-04680, CR 2013-05018, CR 2013-05514, and CR 2013-05569. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions addressed the causes
05000285/FIN-2013008-402013Q2Fort CalhounFailure to Obtain Prior NRC Approval for a Facility ChangeThe team identified a non-cited violation of 10 CFR 50.54(q)(2) for the licensees failure to maintain the effectiveness of an emergency plan. Specifically, since May 14, 2009, the licensee failed to maintain a proper value for low river level associated with the declaration of an emergency at the ALERT classification level. The licensee did not maintain a standard emergency action level scheme in accordance with the requirements of 10 CFR 50.47(b)(4), which states in part, that a standard emergency classification and action level scheme is in use by the nuclear facility licensee. The emergency action level scheme was not maintained because emergency action levels HU1 and HA1, Natural or destructive phenomena affecting the Protected Area, contained an inaccurate river level of 973 feet 9 inches. The river level was inaccurate because the basis document, Procedure EPIP-OSC-1, Emergency Classification, Revision 46, stated the emergency action level was based on the minimum elevation of the raw water pump suction. Based on available plant data, the minimum elevation of the raw water pump suction was above the Alert declaration point of 973 feet 9 inches. This issue has been entered into the corrective action program as Condition Reports CR2013-04198 and CR 2013-04169. This performance deficiency is more than minor, and therefore a finding, because it is associated with emergency response organization performance attribute of the Emergency Preparedness Cornerstone and affected the associated cornerstone objective to ensure 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. Specifically, inaccurate emergency action levels degrade the licensees ability to implement adequate measures to protect public health and safety. The finding was evaluated using the Emergency Preparedness Significance Determination Process, and was determined to be of very low safety significance (Green) because the finding was not a lost or degraded risk significant planning function. The planning standard function was not degraded because the Notification of Unusual Event and Alert emergency classifications would have been declared although potentially in a delayed manner. This finding was not assigned a cross-cutting aspect because the performance deficiency is not reflective of current performance
05000285/FIN-2013008-032013Q2Fort CalhounLack of SAFETY-RELATED Equipment for Design Basis LOW River LevelThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to have safety-related equipment to ensure safe operations down to the design basis low river level. Specifically, from initial plant operations, the licensee failed to ensure that raw water cooling was provided down to the design basis low river level by ensuring the associated specifications and procedures supported raw water pump operation. This issue has been entered into the corrective action program as Condition Reports CR 2013-04169 and CR 2013-06436. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions
05000285/FIN-2013008-422013Q2Fort CalhounFailure to Make Timely Event Notifications for Unanalyzed ConditionsThe team identified four examples of a non-cited violation of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, for the licensees failure to make required event notifications within 8 hours following discovery of an event requiring a report. Specifically, on April 12, 2012, February 7, 2013, February 25, 2013, and February 27, 2013, the licensee failed to notify the NRC within 8 hours of the occurrence an event or condition that resulted in the nuclear power plant being in an unanalyzed condition that significantly degraded plant safety. This issue has been entered into the corrective action program as Condition Report CR 2013-05070. The violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy, because the failure to required event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. The team determined that a cross-cutting aspect was not applicable to this performance deficiency because the failure to make a required report was strictly associated with a traditional enforcement violation
05000285/FIN-2013008-432013Q2Fort CalhounRepetitive Issues Involving Untimely Submittal of Required Licensee Event ReportsThe team identified nine examples of a non-cited violation of 10 CFR 50.73, Immediate Notification Requirements for Operating Nuclear Power Reactors, for the licensees failure to make required licensee event reports within 60 days following discovery of an event requiring a report. Specifically, on nine occurrences between May 9, 2011, and August 30, 2012, the licensee failed to submit a licensee event report for an event meeting the requirements for reporting specified in 10 CFR 50.73. This issue has been entered into the corrective action program as Condition Report CR 2012-03796. The violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy, because the failure to submit a required licensee event report may impact the ability of the NRC to perform its regulatory oversight function. As a result, this violation was evaluated using traditional enforcement. In accordance with Section 6.9 of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV non-cited violation. The team determined that a cross-cutting aspect was not applicable to this performance deficiency because the failure to make a required report was strictly associated with a traditional enforcement violation
05000285/FIN-2013008-142013Q2Fort CalhounFailure to Promptly Identify and Correct a Condition Adverse to QualityThe team identified a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, associated with the licensees failure to promptly identify and correct a condition adverse to quality. Specifically, from 1991 to present, the licensee failed to properly evaluate a 4160 Vac/480 Vac transformer fault or a 480 Vac load center bus fault and the potential effect on system operability. This issue has been entered into the corrective action program as Condition Report CR 2013-05631. This finding is related, and potentially contributed, to the Red finding issued on April 10, 2012, regarding a significant internal fire event in the 480 Vac safety-related switchgear. The performance deficiency is more than minor, and therefore a finding, because it was associated with both the design control and protection against external factors attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The significance of this finding is was potentially greater than Green significance but does not exceed the final significance of the Red finding involving a fire in the 480 Vac safety-related switchgear in June 2011 (NRC Inspection Report 05000285/2012010). Since the identified finding significance is bounded by the Red finding a final significance color is not assigned to this finding. The team determined that although the performance deficiency occurred in 1991, this finding is indicative of current plant performance because the performance characteristic has not been corrected or eliminated. Specifically, the licensee continued to display the same behaviors with regard to decision-making. Therefore, this finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate it is unsafe in order to disapprove the action
05000285/FIN-2013008-012013Q2Fort CalhounInadequate Procedure for Combating Frazil IceThe team identified a URI for a potentially inadequate procedure for combating frazil icing of the travelling screens. Description. The team performed a walkdown with a licensee senior reactor operator to determine the adequacy of the licensees preparations for the occurrence of a frazil ice event. The team noted three issues with the adequacy of the licensees procedure for such conditions, which was in Section V, Degraded River Conditions, of Procedure AOP-01, Acts of Nature, Revision 33. First, the procedure did not give direction to open a valve that would need to be opened to direct auxiliary steam to the travelling screens to melt the ice formed on the screens. Step 15 of Section V, Degraded River Conditions, of Procedure AOP-01, Acts of Nature, instructed operators to direct auxiliary steam to the travelling screens. The procedural step directs hooking up a high temperature hose to a fitting and throttling open valve AS-823. During the walkdown, the licensee and the team noted that the procedure did not include direction to open valve AS-811, a normally closed valve. Valve AS-811 would need to be opened in order to allow steam to the hose used to melt the ice formed on the travelling screens. Valve AS-811 was easily identified as being in the steam flowpath as it was adjacent to valve AS-823. Second, the team noted that the procedure lacked any personnel safety warnings associated with handling a hose with steam. The team did note, however, that gloves and a face shield were staged with the hose to protect personnel, however the procedure did not warn of the danger of handling live steam which is a condition rarely encountered by operators. Finally, the team noted that the procedure did not direct the operators where to direct the steam on the travelling screens, instead Procedure AOP-01, Step 15, Section V, simply instructed operators to initiate steam to the travelling screens. Also, the procedure gave no guidance for a strategy for preferential de-icing of one or more screens, such that an adequate river water flow path would be assured. This issue was entered into the CAP as Condition Report CR 2013-04309. The licensee informed the team that the susceptibility of frazil icing of the intake components was extremely low due to the low flow velocity of river water into the intake structure. The licensee provided Army Corps of Engineer reports which described the phenomenon. The team required additional inspection to determine if the flow into the intake structure was sufficiently low to preclude frazil ice formation. As a result, the team identified Unresolved Item URI 05000285/2013008-01, Inadequate Procedure for Combating Frazil Ice.
05000285/FIN-2013008-072013Q2Fort CalhounAdministrative Controls for a Technical Specification for LOW River LevelThe team identified an unresolved item associated with the technical specification for low river level and its use of administrative controls. The team reviewed the Technical Specification 2.16, River Level, as part of their review of the extent of condition review for the Yellow flooding finding. The team noted that the technical specification was applicable to the Missouri River level at the intake structure. The objective of the technical specification was to specify the maximum and minimum Missouri River levels which must be present to assure safe reactor operation. The team noted during their review of low river level procedures that the design basis low river level for FCS was 976 feet 9 inches. The team also noted that plant procedures for low river level have operators secure the raw water pumps at an intake cell level of 976 feet 9 inches as prescribed by the raw water pump vendor manual The two levels, river level and cell level, represent two physically different water levels which are not the same. Because of the licensees use of trash racks and travelling screens between the intake cell and the river, cell level will be lower than river level. Calculation FC 07384 described the relationship of intake cell level and river level and highlighted that intake cell level will be at least 1 inch lower than river level due to the differential pressure across the racks and screens. As a result, raw water pumps would have to be secured by procedure prior to lowering river level reaching its lower technical specification value and the raw water pumps would not be available for heat removal. The team questioned how the technical specification required action to place the plant in cold shutdown at a river level 976 feet 9 inches could be carried out with the raw water pumps secured by procedure. The team also questioned whether basing action in the technical specification on a river level of 976 feet 9 inches was conservative, or whether a higher river level would be more appropriate to assure plant safety. When questioned by the team, the licensee produced Condition Report CR 2006-03381 which previously evaluated this issue. The condition report corrective actions initially were focused towards initiating a license amendment, but later the corrective actions focused on revising the technical specification bases. A plant memorandum reviewed by the team summarized the final actions as follows: It was decided that the current technical specification minimum river level of 976 feet 9 inches is valid as the lower limit of operability but that additional information was needed to define higher operational limits that would be dependent on plant conditions. Revisions to Technical Specification 2.4 and 2.16 Basis Sections were made (TSBC 07-002-0) to clarify that the minimum submergence water level of 976 feet 9 inches required for operability of raw water pumps applies to the cell level at the pump suction, which can be lower than river level due to head losses across the trash racks, traveling screens and other components that result from debris and/or icing. The team considered these actions to be administrative controls, and therefore, Administrative Letter 98-10, Dispositioning of Technical Specifications That Are Insufficient to Assure Plant Safety, would be applicable. Administrative Letter 98-10 states, Imposing administrative controls in response to an improper or inadequate technical specification is considered an acceptable short-term corrective action. The staff expects that, following the imposition of administrative controls, an amendment to the technical specification, with appropriate justification and schedule, will be submitted in a timely fashion. This issue was entered into the CAP as Condition Report CR 2013- 04733. Additional NRC inspection is necessary to determine if this case was an instance of untimely corrective action for a deficient technical specification. The team considered this to be an unresolved item, URI 05000285/2013008-07, Administrative Controls for a Low River Level Technical Specification.
05000285/FIN-2013008-222013Q2Fort CalhounFort Calhouns Station\'S Ability to Classify Components as SAFETY-RELATEDThe team found it difficult to determine how the licensee identified SSCs important to safety as required by 10 CFR Part 50, Appendix B, Criterion II. The team noted that several recent performance deficiencies have been identified by the NRC for not classifying SSCs as safety-related based on the safety function being performed. The team determined the licensees causal analysis to date for these issues was inadequate. Consequently, during the on-site inspection, the team questioned the classification of various SSCs which appear to meet the definition of safety-related, but have not been designated as such by the licensee. 10 CFR Part 50.2, Definitions defines safety-related SSCs as, ...those structures, systems, and components that are relied upon to remain functional during and following design basis events to assure: (1) the integrity of the reactor coolant pressure boundary; (2) the capability to shut down the reactor and maintain it in a safe shutdown condition; or (3) the capability to prevent or mitigate the consequences of accidents which could result in potential offsite exposures comparable to the applicable guideline exposures set forth in 10 CFR Part 50.34 (a)(1) or 10 CFR Part 50.100.11 of this chapter, as applicable. The team determined that the licensee may not have identified all the SSCs important to safety to ensure the controls of 10 CFR Part 50, Appendix B are applied. Additional review and follow up will be required to determine if this issue represents a performance deficiency or constitutes a violation of NRC requirements. This issue is identified as URI 05000285/2013008-22, Fort Calhoun Stations Ability to Classify Components as Safety-Related.
05000285/FIN-2013008-232013Q2Fort CalhounCode of Record for SAFETY-RELATED Piping SystemsThe team identified a URI associated with the licensees use of piping codes. Specifically, the team is concerned that the licensee is no longer maintaining the construction code of record for safety-related piping systems as described in the USAR. Fort Calhoun Station original code of record for safety-related piping is USAS B31.7, Nuclear Power Piping, 1968 Draft Edition. The licensee reclassified a number of systems in the early 1990s. The licensee then reconciled the code of construction to newer ASME Section III code. The codes were not able to be reconciled in all instances. Specifically, the analysis stated that the material allowable values had to be evaluated on a case by case basis to ensure the conservative values were chosen for analysis. The team has identified numerous examples where the licensee has used the wrong material allowable as required by the reconciliation analysis. The team was concerned that using the incorrect material allowable in piping analyses could be non-conservative and affects the operability of structures, systems, or components required to ensure safe operation. Additional review and follow up will be required to determine if this issue represents a performance deficiency or constitutes a violation of NRC requirements. This issue is identified as URI 05000285/2013008-23, Code of Record for Safety-Related Piping Systems.
05000285/FIN-2013008-272013Q2Fort CalhounContinuous Monitoring Capability of Post Accident Main Steam Radiation Monitor RM-064The team identified an unresolved item associated with post accident radiation monitor RM-064. Specifically, the team is concerned about the capability of the monitor to provide representative measurements due to the system configuration, and this could represent a failure to ensure continuous effluent monitoring of the main steam lines following a steam generator tube rupture accident. Following the Three Mile Island Accident in March 25, 1979, licensees were required to ensure all potential effluent release points from nuclear power plants were equipped with high range radiation monitors. In particular, NUREG-0737, Section II.F.1.1 requires in part; that for pressurized water reactors such as FCS, Unit 1, steam release points be monitored for noble gases, and that indication of the activity must be monitored and recorded continuously. In addition Section II.F.1.1 requires the monitors shall be capable of functioning both during and following an accident. System designs shall accommodate a design-basis release and then be capable of following decreasing concentrations of noble gasses. In addition, the monitoring system shall be capable of obtaining readings at least every 15 minutes during and following an accident. . The team identified an unresolved item associated with post accident radiation monitor RM-064. Specifically, the team is concerned about the capability of the monitor to provide representative measurements due to the system configuration, and this could represent a failure to ensure continuous effluent monitoring of the main steam lines following a steam generator tube rupture accident. By application dated March 9, 1984, the licensee requested an amendment to the stations technical specifications in response to the Commissions Generic Letter 83-37, NUREG-0737 Technical Specifications. The generic letter, which was issued in November 1, 1983, advised licensees to submit new technical specifications for NUREG-0737 items, including Section II.F.1.1, Noble Gas Effluent Monitors (II.F.1.1). The stations potential post-accident steam release points include the main steam relief valves, the atmospheric dump valve, and the steam driven AFW pumps steam turbine. To comply with the high range radiation monitoring requirements, the licensee installed noble gas effluent monitors including, radiation monitor RM-064. Per USAR Section 11.2.3.11, RM-064, the post-accident main steam line monitor, is an off-line monitor designed to measure the steam activity by sampling steam from the two steam headers via two isolation valves HCV-921 and HVC-922. The monitor is placed in service in the event of a steam generator tube rupture. The monitor is capable of sampling steam from both steam headers and the recorded data from this monitor can then be utilized to quantify effluents released through the main atmospheric dump valve, the main steam safety valves, and the AFW pump turbine. Radiation monitor RM-064 is located in the turbine building next to Room 81. - 223 - The team noted that the design basis accident analysis contained in USAR, Section 14.14, Steam Generator Tube Rupture Accident, required the licensee to assume a coincident reactor trip and a loss of off-site power. Due to the assumed simultaneous loss of off-site power with the reactor trip, the reactor is cooled down by releasing steam via the main steam safety valves and atmospheric dump valve, creating a direct release path to the environment. In addition, due to the loss of off-site power, the normal condenser off-gas radiation monitor becomes un-available due to the loss of condenser vacuum. This leaves radiation monitor RM-064 as the only monitor available to measure radioactivity in the main steam lines. The analysis assumes all activity released from the faulted steam generator ceases when it is isolated by plant operators 2 hours after the event. The design of the FCS main steam line monitor is provided in MR-FC-79-190C, Post Accident Main Steam Line High Range Radiation Monitor RM-064, Revision 0, dated June 4, 1982. The station has two 28 inch diameter headers leading to the main turbine. Each main steam line is provided with six main steam safety valves each having different lift set-points. The pipe connecting these valves is 2.5 inches in diameter. The pipe connecting to the atmospheric dump valve is 3 inches in diameter. The sample line to radiation monitor RM-064 is 3/8 inch in diameter. This line is located upstream of the main steam isolation valves, in Room 81 of the auxiliary building. The distance from the main steam header to the actual location of radiation monitor RM-064 (outside Room 81) is over sixty feet long, while the main steam safeties and steam dump valve, are within 12 feet away from the main steam headers. The team reviewed the USAR, main steam drawings, applicable calculations, and interviewed engineers and operators to identify the design basis requirements for radiation monitor RM-064 and to verify it was capable of performing its intended functions. On The team also determined that for the B steam generator header the location of the 3/8 inch sample line leading to radiation monitor RM-064 was installed downstream of three of the main steam safety valves, including the lowest lift set-point valve. For the A steam generator header, the 3/8 inch sample line was located downstream of two of the safety valves but upstream of the lowest lift set point relief. Due to the location of the sample lines being downstream of the safety valves, the difference in pipe sizing between the lines to the monitor (3/8 inch), the main steam safety valves (2.5 inch), and the atmospheric dump valve (3 inch) and the distance from the main steam header to the monitor, the team questioned how the licensee assured a representative measurement would be obtained during and after a steam generator tube rupture accident. The team informed the licensee of their concerns and the licensee initiated Condition Reports CR 2013-04442, 2013-05515, and 2013-06267, to capture these concerns in the CAP. February 27, 2013, the team performed a walkdown of radiation monitor RM-064 and the steam lines. Because radiation monitor RM-064 is normally isolated, the team questioned how long it would take operators to put the monitor in service, and how the licensee met the requirement of continuous monitoring. During subsequent evaluations the licensee determined that there was not an established time requirement for operators to put radiation monitor RM-064 in service. - 224 - The licensee performed a simulator dry run with licensed operators to estimate the time required to place the monitor in service. During this simulated event, it took operators approximately 23 minutes to put the monitor in service, thus indicating that there could be an unmonitored release to the environment for at least 23 minutes following a steam generator tube rupture accident. Regarding the representative sample concern, engineers determined that without a sophisticated computer model it could not be definitely shown that the degree of turbulent mixing in the steam lines is sufficient to equalize the concentrations of radioactive gasses and entrained particulates downstream of the main steam safety valves where the lines connecting to radiation monitor RM-064 were located. The licensee issued Condition Report CR 2013-10507 requesting a detailed calculation to address this concern. The team determined this condition has existed since the time radiation monitor RM-064 was installed in February 1983, until February 27, 2013, when the issue was identified by the team. An engineering technical evaluation was then performed under Condition Report CR 2013-04442, based on existing radiological analysis Calculation FC06820 used for the steam generator accident analysis (USAR 14.14). This technical evaluation removed many of the conservative assumptions included in Calculation FC06820. Based on this basic evaluation and using engineering judgment, the licensee determined that there would be sufficient mixing and adequate concentration to provide a representative radiation measurement. The team concluded that further review is necessary in order to properly evaluate and disposition this issue. This issue is identified as URI 05000285/2013008-27, Continuous Monitoring Capability of Post Accident Main Steam Radiation Monitor RM- 064.
05000285/FIN-2013008-292013Q2Fort CalhounUse of Alternate Seismic Evaluation CriteriaThe team identified an unresolved item associated with the licensees use of alternate seismic criteria when evaluating site structures. The alternate seismic criteria were approved for use by the NRC for piping and HVAC systems but were not explicitly approved for structures. The team identified that the licensee applied an alternate seismic evaluation method called Alternate Seismic Criteria Methodology in place of the original USAR seismic criteria delineated in Appendix F. The use of ASCM was originally proposed to the NRC on December 2, 1988, as alternate design criteria for new designs, modifications, and reanalysis of piping and pipe supports, electrical raceways, HVAC ducting, and anchor bolts. The team reviewed licensing documents and correspondence, and concluded that the NRC approved the use of ASCM for piping and HVAC systems and that the licensee had assumed the NRC had tacitly approved ASCM for structural calculations because the NRC staff did not specifically deny its use in those areas. The licensee informed the team that ASCM was used in several structural seismic evaluations including the auxiliary building and intake structure. At the close of the inspection, other potentially affected structures were being evaluated by the licensee. The team is concerned that the licensee inappropriately used ASCM without NRC approval and that ASCM may be non-conservative with respect to the seismic evaluation criteria specified in the USAR, Appendix F. The effect of the alternative seismic analysis is not known because the seismic spectra are different and are too complex for a qualitative analysis. Additional NRC review and follow up will be required to determine if this issue represents a performance deficiency or constitutes a violation of NRC requirements. This issue is identified as URI 05000285/2013008-29, Use of Alternate Seismic Evaluation Criteria.
05000285/FIN-2013008-282013Q2Fort CalhounFailure to Perform an Evaluation for a Change to Component Cooling Water MAKE-UPThe team identified a Severity Level IV non-cited violation of 10 CFR Part 50.59, with an associated Green finding, because the licensee failed to perform an evaluation for a design change that may have required NRC review and approval. Specifically, from June 2008, the licensee did not evaluate a change that would permanently substitute manual actions for an automatic action to add water and nitrogen gas to the component cooling water surge tank, which is an updated safety analysis report described design function for the component cooling water system. The licensee entered this condition into their corrective action program and planned to perform an evaluation to determine if prior NRC review and approval is needed for this design change. This issue has been entered into the corrective action program as Condition Report CR 2013-04417. The team determined that it was reasonable for the licensee to be able to foresee and prevent the occurrence of this deficiency. The team evaluated this performance deficiency as both a traditional enforcement violation, and a reactor oversight process finding. The violation of 10 CFR Part 50.59 was more than minor because it involved a change to an updated safety analysis report design function in that there was a reasonable likelihood that the change would require NRC review and approval. This finding is associated with the Mitigating Systems Cornerstone. The team used the NRC Enforcement Manual and Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, to evaluate this issue. The finding is determined to have very low safety significance (Green) because it was a design deficiency confirmed not to result in the loss of operability or functionality. The violation of 10 CFR 50.59 impacted the ability of the NRC to perform its regulatory oversight function and was determined to be Severity Level IV because the resulting changes were evaluated by the significance determination process as having very low safety significance, in accordance with the NRC Enforcement Policy. The NRC concluded that the finding did not reflect current licensee performance
05000285/FIN-2013008-022013Q2Fort CalhounFrazil Ice Monitor Not OperationalThe team identified a finding for the licensee\'s failure to maintain their frazil ice detector operational. The detector was sampling a non-representative water temperature which would not have warned operators of the presence of conditions favorable for the formation of frazil ice on intake structure components. The licensee entered the issue into the corrective action program as Condition Report CR 2013-04310 and switched the points they monitored for potential frazil ice formation. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, the finding is determined to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not take appropriate corrective actions to address a similar condition during the winter of 2011-2012 in a timely manner, commensurate with the safety significance and complexity
05000285/FIN-2013008-042013Q2Fort CalhounNON-CONSERVATIVE Value for Declaring an Alert on LOW River LevelThe team identified a non-cited violation of 10 CFR 50.54(q)(2) for the licensees failure to maintain the effectiveness of an emergency plan. Specifically, since May 14, 2009, the licensee failed to maintain a proper value for low river level associated with the declaration of an emergency at the ALERT classification level. The licensee did not maintain a standard emergency action level scheme in accordance with the requirements of 10 CFR 50.47(b)(4), which states in part, that a standard emergency classification and action level scheme is in use by the nuclear facility licensee. The emergency action level scheme was not maintained because emergency action levels HU1 and HA1, Natural or destructive phenomena affecting the Protected Area, contained an inaccurate river level of 973 feet 9 inches. The river level was inaccurate because the basis document, Procedure EPIP-OSC-1, Emergency Classification, Revision 46, stated the emergency action level was based on the minimum elevation of the raw water pump suction. Based on available plant data, the minimum elevation of the raw water pump suction was above the Alert declaration point of 973 feet 9 inches. This issue has been entered into the corrective action program as Condition Reports CR2013-04198 and CR 2013-04169. This performance deficiency is more than minor, and therefore a finding, because it is associated with emergency response organization performance attribute of the Emergency Preparedness Cornerstone and affected the associated cornerstone objective to ensure 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. Specifically, inaccurate emergency action levels degrade the licensees ability to implement adequate measures to protect public health and safety. The finding was evaluated using the Emergency Preparedness Significance Determination Process, and was determined to be of very low safety significance (Green) because the finding was not a lost or degraded risk significant planning function. The planning standard function was not degraded because the Notification of Unusual Event and Alert emergency classifications would have been declared although potentially in a delayed manner. This finding was not assigned a cross-cutting aspect because the performance deficiency is not reflective of current performance
05000285/FIN-2013008-052013Q2Fort CalhounInadequate Procedure for Combating Loss of RAW WaterThe team identified a non-cited violation of Technical Specification 5.8.1, Procedures, for the licensee\'s failure to maintain an adequate procedure for the loss of raw water cooling. Specifically, since April 2011, the licensee failed to maintain Procedure AOP-18, Loss of Raw Water, to adequately align the component cooling water system for the feed and bleed mode. This issue has been entered into the corrective action program as Condition Report CR 2013-04417. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events - 4 - to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, the finding is determined to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions
05000285/FIN-2013008-062013Q2Fort CalhounFailure to Account for Worst Case Conditions in Fuel Oil Inventory CalculationThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to account for design basis conditions in their fuel oil consumption calculation. Specifically, since June 2011, the licensee failed to translate the worst case design emergency diesel generator frequency that could impact the consumption of fuel oil into the applicable design documentation. To address the deficiency, this issue has been entered into the corrective action program as Condition Reports CR 2013-04311 and CR 2013-04470. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions
05000285/FIN-2013008-082013Q2Fort CalhounSluice Gate Leakage Not Periodically VerifiedThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to ensure that a critical parameter in the design calculation for intake cell level control (sluice gate leakage) was periodically measured to ensure the plant stayed within the parameters of the design calculation. Specifically, since April 2011, the licensee failed to assure that the assumed leakage of the sluice gates was translated into a procedure to periodically measure leakage against acceptance criteria to ensure the leakage was low enough to support the intake structure design calculation. This issue has been entered into the corrective action program as Condition Report CR 2013-04315. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, the finding is determined to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions
05000285/FIN-2013008-102013Q2Fort CalhounFailure to Accurately Model RAW Water Flow Into the Intake StructureThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to accurately model the traveling screens and trash racks in the flow calculation for cell level control. Specifically, since April 2011, the licensee failed to translate the actual plant configuration for flow of water into the intake structure during flooding into the applicable design calculation. This issue has been entered into the corrective action program as Condition Reports CRs 2013-04468 and CR 2013-04310. This performance deficiency is more than minor, and therefore a finding, because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not thoroughly evaluate problems such that the resolutions address causes and extent of conditions
05000285/FIN-2013008-122013Q2Fort CalhounInadequate Root Cause for a Significant Condition Adverse to QualityThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensees failure to promptly identify, correct, and prevent recurrence of a significant condition adverse to quality. Specifically, from November 2009 to present, measures established by the licensee failed to assure that the cause of an identified significant condition adverse to quality was corrected and corrective actions taken would preclude repetition involving the failure to identify nonconforming quality equipment before it is installed and relied upon to perform specified safety functions. Specifically, in this instance, the licensee failed to identify that a 480 Volt replacement breaker has a jumper installed inappropriately resulting in the failure of the breaker to trip during a faulted condition. This issue has been entered into the corrective action program as Condition Report CR 2013-04037. The performance deficiency is more than minor, and therefore a finding, because it is associated with the protection against external factors attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the licensees root cause analysis will not provide assurance that effective corrective actions are taken to preclude recurrence of a breaker trip failure. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, which contained the initial screening for pressurized water reactors that are shutdown with a time to boil of greater than 2 hours. Technical Specification 2.7, Electrical Systems, states that the reactor shall not be heated up or maintained at temperatures above 300 degrees Fahrenheit unless the electrical systems listed in that section (includes the 480 V busses) are operable. Because the plant was maintained below 300 degrees during the exposure period, the team determined that power availability technical specifications were being met as discussed in Checklist 4. Because the finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. Therefore, the finding is determined to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of accountability associated with the other safety culture components because the licensee failed to demonstrate a proper safety focus and reinforce safety principles among their peers. Specifically, the licensee focused on sending a message about the vendor rather than the licensees failures to establish accountability for the vendors products and services
05000285/FIN-2013008-132013Q2Fort CalhounFailure to Establish and Document Basis for Test Acceptance CriteriaThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to assure that applicable design basis information, as defined in 10 CFR 50.2, for breaker testing was correctly translated into specifications, drawings, procedures, and instructions. Specifically, from July 2011, to the present the licensee failed to incorporate the basis for the acceptance limits of the digital low resistance ohmmeter values into specifications and procedures. Without a basis for the acceptance values the licensee cannot show that the breakers will perform satisfactorily in service, and incorrect acceptance values could allow high resistance connections to go unnoticed. This issue has been entered into the corrective action program as Condition Report CR 2013-04032. This performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, the team determined that because this finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. Therefore, the finding is determined to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the work practices component because licensee personnel failed to follow procedures. Specifically, Fort Calhoun Station personnel failed to follow the requirements specified in Procedure PED-GEI-7, Specification of Post-Modification Test Criteria
05000285/FIN-2013008-152013Q2Fort CalhounFailure to Correct Conditions Adverse to Quality Involving Frequency Compatibility Issues in the 120VAC SystemThe team reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to address frequency compatibility issues in the 120 Vac electrical distribution system. Specifically, between June 5, 2008, and February 22, 2013, the licensee failed to correct known frequency compatibility issues in the 120 Vac instrument system that resulted in voltage transients and damage to instrumentation supplied by the 120 Vac instrument inverters. This issue has been entered into the corrective action program as Condition Report CR 2013-03866. At the close of the inspection, the licensee was still completing causal analysis and identification of corrective actions necessary to address frequency compatibility issues in the 120 Vac electrical distribution system. This performance deficiency is more than minor, and therefore a finding, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, the finding is determined to have very low safety significance (Green) because the finding did not increase the likelihood of a loss of reactor coolant system inventory, the finding did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory when needed, and the finding did not degrade the licensees ability to recover decay heat removal once it was lost. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component. Specifically, the team identified that the licensee failed to adequately evaluate repeated low voltage/ground alarm associated with the 120 Vac distribution system
05000285/FIN-2013008-162013Q2Fort CalhounFailure to Account for Additional Diesel Loading from NON-SAFETY LoadsThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criteria III, Design Control, for the licensees failure to update calculations to account for non safety-related loads supplied by the emergency diesel generator through non-qualified isolation devices and the cumulative impact on diesel fuel oil consumption. Specifically, prior to April 1, 2013, Calculation EA-FC-92-072, Diesel Generator Transient Loading Analysis Using EDSA Design Base 3.0, Revision 6, failed to account for the additional diesel fuel oil consumption that would occur due to the loads that would be supplied from the emergency diesel generators through non-CQE isolation devices. The licensee modified Calculation EA-FC-92-072 to address the teams concerns. This issue has been entered into the corrective action program as Condition Report CR 2013-09817. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Because this performance deficiency affected the calculation used to determine the required diesel fuel oil inventory for an accident or a loss of offsite power occurring from at power conditions, the team used Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined the finding to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and - 11 - resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate the condition identified in Condition Report CR 2013-04594 to determine its impact to emergency diesel generator fuel oil consumption
05000285/FIN-2013008-172013Q2Fort CalhounFailure to Adequately Implement the Maintenance RuleThe team identified a non-cited violation of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, associated with the licensees failure to adequately monitor the performance of structures, systems, and components, against established goals in a manner sufficient to provide reasonable assurance that these structures, systems, and components are capable of fulfilling their intended functions. Specifically, from June 7, 2011, to the present, the licensee failed to monitor the performance of the 480 Vac busses in a manner sufficient to provide reasonable assurance that they are capable of fulfilling their intended safety functions. This issue has been entered into the corrective action program as Condition Report CR 2013-04352. This performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, which contained the initial screening for pressurized water reactors that are shutdown with a time to boil of greater than 2 hours. Technical Specification 2.7, Electrical Systems, stated that the reactor shall not be heated up or maintained at temperatures above 300 degrees Fahrenheit unless the electrical systems listed in that section (includes the 480 V busses) are operable. Because the plant was maintained below 300 degrees during the exposure period, the team determined that power availability Technical Specifications were being met as discussed in Checklist 4. Because the finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. Therefore, the finding is determined to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the decision-making component because the licensee failed to use conservative assumptions in decision-making and adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate it is unsafe in order to disapprove the action
05000285/FIN-2013008-182013Q2Fort CalhounFailure to Establish Adequate Instructions for Restoring Temporary ModificationsThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to establish adequate instructions for restoring temporary modifications. Specifically, from January 17, 2013, to the present, the licensees temporary modification control procedure did not include appropriate criteria for determining that control room and operations control center references reflect current plant configuration and were updated in a timely manner. The licensee initiated Condition Report CR 2013-04286, which stated that the licensees transition to a new procedure will help ensure that control room and operations control center documents were updated in a timely manner and that the licensee is determining whether any near-term action is necessary to address the issue until then. This performance deficiency is more than minor, and therefore a finding, because it is associated with the procedure quality attribute of the Mitigating Systems Cornerstone and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the procedure inadequacy could become a more significant issue because it could allow operators to continue to reference material that does not reflect current plant configuration. Using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, Checklist 4, PWR Refueling Operation: RCS level > 23\' OR PWR Shutdown Operation with Time to Boil > 2 hours And Inventory in the Pressurizer, the team determined that because this finding did not increase the likelihood of a loss of reactor coolant system inventory; did not degrade the licensees ability to terminate a leak path or add reactor coolant system inventory; and did not degrade the licensees ability to recover decay heat removal, this finding did not require a Phase 2 or 3 analysis as stated in Checklist 4. Therefore, the finding is determined to have very low safety significance (Green). This finding has a cross-cutting aspect in the area of human performance associated with the work control component because the licensee failed to appropriately coordinate work activities by incorporating actions to address the need to keep personnel apprised of work status, the operational impact of work activities, and plant conditions that may affect work activities. Specifically, the licensee did not incorporate actions into the procedure that would address the impact of out-of-date control room references on operator performance
05000285/FIN-2013008-192013Q2Fort CalhounFailure to Initiate Condition Reports in Accordance with the Corrective Action Program ProceduresThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to initiate condition reports when problems or conditions adverse to quality were identified in accordance with Procedure FCSG-24-1, Condition Report Initiation, Revision 3. Specifically, between July 2012 and March 2013, the team identified 11 instances where licensee staff failed to initiate a condition report after identifying a deficiency or a condition adverse to quality. In some instances, licensee personnel had to be prompted by the team to initiate a condition report. As a result, the corrective actions taken to address the conditions could have been potentially untimely. This issue has been entered into the corrective action program as Condition Report CR 2013-06991. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, if the licensee does not enter conditions adverse to quality into the corrective action program, the conditions adverse to quality may not be evaluated and corrected in a timely manner. This finding is associated with Mitigating Systems Cornerstone. The team determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, and conducted a Phase 1 characterization and initial screening. Using Phase 1, Table 3, SDP Appendix Router, the team answered yes to the following question: Does the finding pertain to operations, and event, or a degraded condition while the plant was shutdown? As a result, the team used IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, the finding is determined to have very low safety significance (Green) since it did not need a quantitative assessment. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues
05000285/FIN-2013008-202013Q2Fort CalhounFailure to Identify Conditions Adverse to QualityThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct conditions adverse to quality. Specifically, between July 2012 and March 2013, the team identified 6 instances where the licensee failed to identify a deficiency or a condition adverse to quality and to enter them into the corrective action program. As a result, conditions adverse to quality may not be corrected in a timely manner commensurate with the safety significance. This issue has been entered into the corrective action program as Condition Report CR 2013-07959. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to identify conditions adverse to quality and enter them into the corrective action program, has the potential to lead to a failure to correct conditions adverse to quality in a timely manner commensurate with the safety significance. This finding was associated with the Mitigating Systems Cornerstone. The team determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, and conducted a Phase 1 characterization and initial screening. Using Phase 1, Table 3, SDP Appendix Router, the team answered yes to the following question: Does the finding pertain to operations, and event, or a degraded condition while the plant was shutdown? As a result, the team used IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, the finding is determined to have very low safety significance (Green) since it did not need a quantitative assessment. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues and did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance
05000285/FIN-2013008-212013Q2Fort CalhounFailure to Ensure That Design Requirements Associated with the Containment Electrical Penetration Assemblies Were Correctly Translated Into Installed Plant EquipmentThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to translate applicable regulatory requirements and the design basis into specifications, drawings, procedures, and instructions. Specifically, from initial construction to present, the licensee did not perform adequate analysis and/or post-accident condition functional testing of the teflon insulated and teflon sealed Conax electrical penetration assemblies to determine if they were suitable for expected post accident conditions. The licensee has decided to replace or cap all Teflon-insulated containment electrical penetration assemblies prior to returning to power operations. This issue has been entered into the corrective action program as Condition Report CR 2013-03571. This performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Barrier Integrity Cornerstone and affected the associated cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to implement a corrective action program with a low threshold for identifying issues and identify such issues completely, accurately, and in a timely manner commensurate with their safety significance
05000285/FIN-2013008-242013Q2Fort CalhounFailure to Effectively Monitor the Performance of Penetration SealsThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct conditions adverse to quality. Specifically, between July 2012 and March 2013, the team identified 6 instances where the licensee failed to identify a deficiency or a condition adverse to quality and to enter them into the corrective action program. As a result, conditions adverse to quality may not be corrected in a timely manner commensurate with the safety significance. This issue has been entered into the corrective action program as Condition Report CR 2013-07959. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the failure to identify conditions adverse to quality and enter them into the corrective action program, has the potential to lead to a failure to correct conditions adverse to quality in a timely manner commensurate with the safety significance. This finding was associated with the Mitigating Systems Cornerstone. The team determined that the finding could be evaluated using the significance determination process in accordance with IMC 0609, Significance Determination Process, and conducted a Phase 1 characterization and initial screening. Using Phase 1, Table 3, SDP Appendix Router, the team answered yes to the following question: Does the finding pertain to operations, and event, or a degraded condition while the plant was shutdown? As a result, the team used IMC 0609 Appendix G, Shutdown Operations Significance Determination Process. Using Appendix G, the finding is determined to have very low safety significance (Green) since it did not need a quantitative assessment. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee did not implement a corrective action program with a low threshold for identifying issues and did not identify issues completely, accurately, and in a timely manner commensurate with their safety significance
05000285/FIN-2013008-252013Q2Fort CalhounDeficient Evaluation for Known Degraded Conditions: SAFETY-RELATED Air Operated Valve Elastomers Not Qualified for Helb/Loca TemeraturesThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, associated with the licensee\'s failure to properly evaluate a known degraded condition regarding safety-related air operated valve elastomers that were not qualified for high energy line break or loss of coolant accident temperatures. Specifically, from January 11 through January 18, 2013, due to a an improper application of the single failure criteria, the licensee failed to properly evaluate and correct a known degraded condition associated with safety-related air operated valve elastomers that were not qualified for high energy line break or loss of coolant accident temperatures. This issue has been entered into the corrective action program as Condition Reports CR 2013-01396 and CR 2013-02611. This performance deficiency is more than minor, and therefore a finding, because if left uncorrected, the failure to correct the degraded condition had the potential to lead to a more significant safety concern. Specifically, the affected air operated valves would have been in a condition where they would not have been qualified to perform their intended safety function. This issue was associated with the Mitigating Systems Cornerstone. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-262013Q2Fort CalhounFailure to Properly Inspect, Maintain, and Test Emergency Feedwater Tank EquipmentThe team identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, associated with the licensees failure to ensure proper inspection, maintenance, and testing of equipment associated with emergency feedwater tank FW-19. Specifically, from initial construction until February 27, 2013, the licensee failed to ensure proper inspection, maintenance, and testing was performed on the emergency feedwater storage tanks sight glass ball check isolation valves, to prevent draining of the tank following failure of the sight glass. The licensee performed an analysis and concluded that operators have adequate time to respond to such a loss of tank FW-19 inventory. This issue has been entered into the corrective action program as Condition Reports CRs 2012-15687, CR 2013-03974, and CR 2013-06170. This performance deficiency is more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding is determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; (4) did not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significance in accordance with the licensees maintenance rule program; and (5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. This finding has a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component because the licensee failed to thoroughly evaluate problems such that the resolutions address the causes
05000285/FIN-2013008-302013Q2Fort CalhounEvaluation of Change to Alternate Shutdown Cooling FlowpathThe team identified an unresolved item related to engineering change modifications that changed a procedure to include the replacement of automatic actions with manual actions. Specifically, the10 CFR 50.59 evaluation proposed a change to substitute automatic flow control of shutdown cooling flow and temperature with manual control using the low pressure safety injection loop injection valves alternate shutdown cooling flow control. During a review of Engineering Change Modification 54058, Procedure Change to Allow Closing of HCV-335 while on Alternate Shutdown Cooling, the team identified that the licensee changed a procedure to include the replacement of an automatic action with a manual action. Specifically, the engineering change proposed to close both shutdown cooling heat exchanger isolation valve HCV-335 and flow control valve FCV-326 while pinning open valve HCV-341 and manually throttling low pressure safety injection loop injection valves to maintain the desired RCS temperature and flow rate. The team questioned whether the licensee required prior NRC review and approval to make this change since flow control valve FCV-326 normally controls temperature and flow automatically as described in Section 9.3.4.3 of the USAR. The licensee entered this issue of concern into the CAP. Additional NRC review and follow up will be required to determine if this issue represents a performance deficiency associated with meeting the 10 CFR 50.59 requirement of more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component important to safety previously evaluated in the USAR. This item is unresolved pending review of the licensees evaluation. This issue is identified as URI 05000285/2013008- 30, Evaluation of Change to Alternate Shutdown Cooling Flowpath.
05000285/FIN-2011014-022012Q1Fort CalhounFailure to Perform Adequate 10 CFR 50.59 ReviewThe team identified an unresolved item related to the licensees implementation of the requirements in 10 CFR 50.59 for modification EC 33464, Replace AK-50 480 V Main and Bus-Tie Breakers With Molded Case Type or Equivalent, Revision 0, and the adequacy of the design review and screening performed to support the modification. In November 2009, the licensee implemented a modification to replace 12 General Electric AK-50 low voltage power circuit breakers with Nuclear Logistics Incorporated/Square-D Masterpact circuit breaker/cradle assemblies and digital trip devices. This modification was developed to address obsolescence issues and maintenance problems with the older AK-50 circuit breakers. Fort Calhoun Station used General Electric AKD-5 Powermaster Low Voltage Drawout Switchgear, with a welded aluminum bus bar structure that transitioned to copper bus stabs in each breaker cell. The original AK-50 circuit breakers connected directly to the silver-plated areas on the line and load stabs. The new Nuclear Logistics Incorporated /Square-D circuit breaker design was an integrated unit consisting of a circuit breaker and cradle assembly. The cradle assembly converted the internal vertical breaker connectors to top and bottom spring-loaded horizontal finger assemblies which connected to the switchgear bus stabs. The team reviewed the licensees implementation of the requirements in 10 CFR 50.59, Changes, Tests and Experiments, for the modification. The team noted that the screening process did not recognize the potential for adverse effects on the design basis function of the 480 Vac electrical distribution system because of the introduction of a cradle assembly that had different connections to the switchgear bus bars. The original breakers did not require the cradle adapter and had connector assemblies which mounted to the silver-plated portion of the bus bars. The new assembly introduced the cradle as an adapter between the breaker and switchgear, and introduced additional resistances in the circuits. The licensees screening documents stated that the electrical connections on the cradle matched the existing switchgear General Electric breakers; however, the licensees root cause analysis showed that the cradle finger assemblies were not the same size as the original breaker connections. The team also noted that the licensees screening process did not recognize the potential for high resistance connections to exist, and did not analyze additional failure mechanisms that may have been created by the addition of the cradle assemblies. Condition Report CR 2011-6319 was written, after the fire, for the discovery of the improper engagement of cradle fingers to silver plating on the stabs. Further inspection is required to determine if the licensees implementation of the requirements of 10 CFR 50.59 were appropriate.
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