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05000285/FIN-2014009-032014Q3Fort CalhounFailure to Adequately Perform an Operability Evaluation and a 50.59 EvaluationA non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified involving the failure to evaluate and implement adequate compensatory measures for a degraded condition associated with raw water pump AC-10C. Specifically, the licensees operability determination established a compensatory measure to place pump AC-10C in pull-to-lock, contrary to the system single failure analysis design criteria described in the Updated Safety Analysis Report. The licensee entered this issue into its corrective action program as Condition Reports 2014-09104 and 2014-08515 and performed an operability evaluation and associated 10 CFR 50.59 evaluation that used an acceptable compensatory measure to pump water from affected manholes prior to affecting the degraded power feeder cable for raw water pump AC-10C. The NRC evaluated this performance deficiency as both a reactor oversight process finding and a traditional enforcement violation. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect in the area of problem identification and resolution with an aspect of evaluation because the licensee failed to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2). In addition, because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function in that the failure to obtain a license amendment for a change that could result in a malfunction of a structure, system or component with a different result than previously evaluated in the Updated Safety Analysis Report is in violation of 10 CFR 50.59(c)(2)(vi), the NRC also evaluated the violation using traditional enforcement. Since this violation is associated with a Green reactor oversight process violation, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy.
05000285/FIN-2014009-242014Q3Fort CalhounLicensee-Identified ViolationTitle 10 CFR 50.73(a)(1) requires, in part, that licensees shall submit a licensee event report for any event of the type described in this paragraph within 60 days after the discovery of the event. Contrary to the above, on February 5, 2012, November 15, 2011, and February 19, 2013, the licensee failed to submit a licensee event report for an event meeting the requirements for reporting specified in 10 CFR 50.73. Specifically, the licensee submitted Licensee Event Reports 2012-013, 2012-015 and 2013-001 greater than 60 days following discovery of a reportable event. 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 licensee entered this issue into their corrective action program as CR 2014-02792.
05000285/FIN-2014009-132014Q3Fort CalhounFailure to Perform Evaluation for Design ChangeA cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, was identified involving the failure to evaluate if a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the licensee did not evaluate a change that would permanently substitute a manual action for an automatic action to add water and nitrogen gas to the component cooling water surge tank. The licensee entered this issue into its corrective action program as Condition Report 2014-09080 and initiated action to evaluate the change to the component cooling water system. The NRC determined that the licensees failure to perform an evaluation prior to implementing a proposed change described in the Updated Safety Analysis Report was a violation of 10 CFR 50.59. Because this performance deficiency had the potential to impact the NRCs ability to perform its regulatory function, the NRC evaluated the performance deficiency using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the team evaluated this finding using the significance determination process to assess its significance. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy this performance deficiency is being characterized as a Severity Level IV violation. The team determined that a cross-cutting aspect was not applicable to this finding because the issue was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-082014Q3Fort CalhounFailure to Report Loss of Environmental Qualification of Safety Related Limit Switches within Required Time LimitsA non-cited violation of 10 CFR 50.73(a)(1), Licensee Event Report System, was identified involving the failure to submit a required licensee event report. Specifically, the licensee failed to report within 60 days the discovery that NamcoTM Type EA 180 limit switches were not environmentally qualified as required due to inadequate maintenance procedures, a condition that resulted in operation prohibited by the plants technical specifications. The licensee restored compliance by submitting Licensee Event Report 05000285/2014-004 on June 20, 2014. The licensee entered this issue into its corrective action program as Condition Report 2014-08454. The NRC determined that the failure to submit a licensee event report within the time limits specified in regulations was a violation of 10 CFR 50.73. This 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 NRC determined that a cross-cutting aspect was not applicable because the issue was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-042014Q3Fort CalhounFailure to Perform an Evaluation for a New Operator Manual Action to Refill Component Cooling Water System During Post- Accident ConditionsA non-cited violation of 10 CFR 50.59, Changes, Test, and Experiments, was identified involving the failure to evaluate if a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the licensee failed to evaluate if a change implemented under Engineering Change 59252 that credited the non-safety related demineralized water system as a make-up source to the component cooling water system during post-accident conditions represented an adverse change to the Updated Safety Analysis Report described design function. The licensee entered this deficiency into its corrective action program for resolution as Condition Report 2014-09151 and established action items to update Engineering Change 59252. The NRC determined that the licensees failure to perform an evaluation prior to implementing a proposed change described in the Updated Safety Analysis Report was a violation of 10 CFR 50.59. Because this violation had the potential to impact the NRCs ability to perform its regulatory function, the NRC evaluated the violation using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the NRC evaluated this finding using the significance determination process to assess its significance. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this performance deficiency is characterized as a Severity Level IV violation. The team determined that a cross-cutting aspect was not applicable because the issue involving the failure to perform an adequate 10 CFR 50.59 evaluation was strictly associated with a traditional enforcement violation.
05000285/FIN-2014009-122014Q3Fort CalhounFailure to Maintain Effectiveness of an Emergency PlanA cited violation of 10 CFR 50.54(q)(2), Conditions of License, was identified involving the failure to maintain the effectiveness of the sites emergency plan. Specifically, the licensee established an Alert low river level emergency classification criteria that was below the raw water pumps minimum suction requirements, contrary to the standard emergency action level scheme. The licensee entered this issue into its corrective action program as Condition Report 2014-08757 which included actions to re-evaluate the capability of the raw water pumps to operate at low river levels. This finding was more than minor, and therefore a finding, because it was associated with the 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 actions 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 emergency classifications would have been declared although potentially in a delayed manner. This finding has a cross-cutting aspect in the area of human performance in that the licensee did not ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee deferred funding for a vendor analysis of the capabilities of the raw water pumps at the design low river level.
05000285/FIN-2014009-012014Q3Fort CalhounFailure to Initiate Condition Reports for Gaps Identified in Resolving NRC Non-Cited ViolationsA non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, was identified involving the failure to follow procedures to initiate condition reports to enter conditions adverse to quality into the corrective action program. Specifically, the licensee failed to initiate condition reports in accordance with Procedure FCSG 24-1, Condition Report Initiation, Step 4.1.1.G, when deficiencies related to the stations corrective actions implemented for NRC violations were identified. The licensee entered this issue into its corrective action program as Condition Report 2014-09063 and initiated action to write condition reports for identified gaps related to previous NRC violations. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it would have the potential to lead to a more significant safety concern. The team performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609 Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding was of very low safety significance (Green) because it did not involve a loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of human performance because the licensee elected to use an informal system to resolve these issues rather than the corrective action program.
05000285/FIN-2014009-252014Q3Fort CalhounLicensee-Identified ViolationTechnical Specification 5.8.1.a, requires, in part, that written procedures be established, implemented, and maintained as recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Regulatory Guide 1.33, Paragraph 9.a, requires that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to the above, the licensee failed to establish procedures for maintenance that can affect the performance of safety related equipment as recommended in Regulatory Guide 1.33, Revision 2, Appendix A, dated February 1978. Specifically, prior to May 3, 2013, the licensees maintenance procedure for NamcoTM Type EA 180 limit switches did not specify the correct torque values for the switch top cover to maintain the components environmental qualifications. This finding was determined to be of very low safety significance because the affected limits switches only affected the radiological barrier provided for by the control room. This issue was entered into the licensees corrective action program as CR 2012-03651.
05000285/FIN-2014009-232014Q3Fort CalhounLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires, in part that measures shall be established to assure that conditions adverse to quality are promptly identified and corrected. Contrary to the above, on June 2, 2008, the licensee completed flow scan valve testing for the high pressure safety injection alternate header isolation valve (HCV-2987) that showed a much higher stem friction value than previously analyzed, but failed to promptly identify and correct the condition adverse to quality until CR 2012-01601 was initiated on February 29, 2012. This finding is of very low safety significance (Green) because valve HCV-2987s failure did not represent an actual loss of safety function of a single train for greater than the technical specification allowed outage time in that EOP/AOP Attachments, Revision 13, dated November 19, 2002, requires operators to also close downstream valves that would back up the closure function of valve HCV-2987. This issue was entered into the licensees corrective action program as CR 2012-01601.
05000285/FIN-2014009-222014Q3Fort CalhounLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis for those structures, systems, and components are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, from initial construction until January 13, 2013, the licensee failed to establish measures to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to control the design inputs to ensure that piping in the chemical and volume control system would perform acceptably during a seismic event. This finding is of very low safety significance (Green) because a chemical and volume control system piping failure event is enveloped by the small break loss of coolant accident as described in Updated Safety Analysis Report Section 14.5.5. This issue was entered into the licensees corrective action program as CR 2013-01796.
05000285/FIN-2014009-212014Q3Fort CalhounFailure to Take Timely Corrective Actions for an Unsealed Raw Water System Control PanelA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take corrective actions to address a design deficiency affecting the control panel for raw water strainer AC-12B. Consequently, the panel experienced a water intrusion event on August 3, 2014, resulting in an unplanned inoperability of the raw water system. Following identification of this issue, the licensee implemented corrective actions to seal conduits leading to control panel AI-348 to prevent future water intrusion. The licensee entered this issue into its corrective action program as Condition Report 2014-09572. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of problem identification and resolution in that the licensee failed to adequately review and provide timely responses to past operating experience that demonstrated that panel AI-348 was susceptible to water intrusion.
05000285/FIN-2014009-202014Q3Fort CalhounFailure to Correct Conditions Adverse to Quality in the Diesel Generator Stating Air SystemA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take timely corrective actions to address service life related degradation of the emergency diesel generator starting air system. As a result, diesel generator 1 failed to roll during planned surveillance testing due to a degraded diesel starting air valve. The licensee replaced the faulty starting air valve and implemented corrective actions to develop preventative maintenance strategies for the starting air system. The licensee entered this issue into the corrective action program as Condition Report 2014-09424. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings , Exhibit 3, Mitigating Systems Screening Questions, dated May 9, 2014, the finding was of very low safety significance (Green) because the finding does not represent a loss of system safety function and the finding does not represent an actual loss of safety function of a single train for greater than its technical specification allowed outage time. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed to recognize and plan for the possibility of latent issues, and inherent risk, even while expecting successful outcomes when determining the repair schedule for starting air valve SA-148.
05000285/FIN-2014009-192014Q3Fort CalhounFailure to Maintain B.5.b Equipment in a State of Readiness to Support Mitigation StrategiesA non-cited violation of 10 CFR 50.54(hh)(2), Conditions of License, was identified involving the failure to maintain available equipment needed to implement mitigating strategies to maintain or restore core, containment, and spent fuel pool cooling capabilities following large fires or explosions. Specifically, the licensee failed to maintain available a flexible suction hose related to the reactor coolant system heat removal mitigating strategy. The licensee initiated Condition Report 2014-08876 to address this deficiency and initiated action to procure and replace the missing flexible suction hose. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The NRC determined that this finding was of very low safety significance (Green) using NRC Manual Chapter IMC 0609, Appendix L, B.5.b Significance Determination Process, because it resulted in an unrecoverable unavailability of an individual mitigating strategy but did not result in multiple unavailable mitigating strategies such that reactor coolant system heat removal could not occur. This finding has a crosscutting aspect in the area of human performance in that the licensees inadequate B.5.b inventory procedure contributed to the lack of recognition that the degraded flexible suction hose was required to implement mitigating strategies.
05000285/FIN-2014009-172014Q3Fort CalhounInadequate Corrective Actions to Properly Implement Applicable ASME OM Code RequirementsA non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to correct a condition adverse to quality associated with classification of check valves in the auxiliary feedwater system. Specifically, the licensee failed to update the in-service testing program to classify auxiliary feedwater discharge check valves as Category A/C valves and include required seat leakage testing. The licensee entered this issue into its corrective action program as Condition Report 2014-08452 and initiated actions to re-assess the current in-service testing methodology of check valves in the auxiliary feedwater system. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone, and affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to evaluate the function of discharge check valves FW-173 and FW-174 when developing the in-service testing program and addressing previous condition reports.
05000285/FIN-2014009-022014Q3Fort CalhounMultiple Examples of Failure to Evaluate Operability of Degraded or Non-Conforming ConditionMultiple examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified involving the failure to follow Procedure OP-FC-108-115, Operability Determinations, Revision 0a. In each example, the team identified that the licensee failed to make an immediate determination of operability for a degraded or non-conforming condition or failed to make an immediate determination of operability based on a detailed examination of the deficiency. The licensee took immediate corrective actions to update the incomplete or inaccurate operability determinations and entered the collective failures to follow station operability procedures into their corrective action program as Condition Report 2014-09163. This performance deficiency was more than minor, and therefore a finding, because it affected the equipment performance attribute of the Mitigating Systems Cornerstone objective of ensuring the reliability of systems that respond to initiating events. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to use decision-making practices that demonstrate that a proposed action is to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee made non-conservative decisions related to the impact of degraded or non-conforming conditions.
05000285/FIN-2014009-052014Q3Fort CalhounInadequate Design Inputs into Safety Injection Piping Stress CalculationA non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to implement appropriate design control measures associated with a safety-related pipe stress calculation. Specifically, several unverified and potentially non-conservative inputs were identified associated with Calculation FC07240 used to analyze stresses on a pipe reduction tee in the safety injection system. The licensee entered this issue into the corrective action program as Condition Report 2014-09098 and initiated action to update Calculation FC07240. This performance deficiency was more than minor, and therefore a finding, because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of components that respond to initiating events. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more nontechnical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed to apply the appropriate rigor when evaluating the overstressed pipe union tee.
05000285/FIN-2014009-062014Q3Fort CalhounFailure to Maintain Design Control of Raw Water Strainer Control PanelA self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to maintain design control of the raw water strainer AC-12B control panel AI-348. Specifically, the licensee failed to adequately design control panel AI-348 to protect it from the effects of spraying and wetting as required by the plants licensing and design basis. The licensee entered this issue into its corrective action program as Condition Reports 2013-03301 and 2014-06974 and initiated action to encase control panel AI-348 to protect it against the effects of spraying and wetting. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, control panel AI-348 was not designed to prevent water intrusion that resulted in a loss of power to raw water strainer AC-12B. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding was of 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 non-technical specification trains of equipment designated as high safety-significant 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 crosscutting aspect in the area of problem identification and resolution associated with the organization thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000285/FIN-2014009-072014Q3Fort CalhounFailure to Accurately Model Flow Path for External Flood MitigationA non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to accurately model cell level control of river water during external flooding events. Specifically, the licensee failed to account for losses due to the physical obstructions of trash racks for inflowing river water, the decreased withdrawal rate of the raw water pumps due to fouling across the traveling screens, and a bounding in leakage rate for the sluice gates when the river level is at maximum level of 1014 mean sea level and the intake cell levels are at minimum level of 9769 . The licensee entered this issue into its corrective action program as Condition Report 2014-09155, performed an operability determination, and initiated action to update station calculations related to intake cell level control. This performance deficiency was more than minor, and therefore a finding, because if left uncorrected, the finding would have the potential to lead to a more significant safety concern. Specifically, the failure to accurately model flow in and out of the cells could adversely affect the external flooding mitigation strategy beyond previously identified equipment capacities and operator actions. This finding was associated with the Mitigating Systems Cornerstone. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program; 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, operating experience, in that the licensee failed to incorporate relevant internal operating experience related to previous NRC inspection into Calculation FC08081.
05000285/FIN-2014009-092014Q3Fort CalhounFailure to Incorporate Design Requirements for Switchgear Room CoolingA non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to translate applicable design requirements into the specifications for plant systems. Specifically, inadequate design control inputs were used for analyzing the ability of the vital switchgear room cooling system to perform its safety function under all conditions. The licensee entered this issue into its corrective action program as Condition Report 2014-08317 and initiated actions to analyze the ability of vital switchgear room cooling to meet its specified safety function. This performance deficiency was more than minor, and therefore a finding, because it affected the design control attribute of the Mitigating Systems Cornerstone, and it directly affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the evaluation component of the problem identification and resolution cross-cutting area because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to analyze and evaluate a 1998 loss of switchgear cooling event to ensure that its use as a design assumption bound the worst design basis event.
05000285/FIN-2014009-102014Q3Fort CalhounDeficient Evaluation of NRC Bulletin 88-04, Strong Pump Weak Pump Due to Failure to Consider the Effect of Auxiliary Feedwater Pumps Discharge Check Valves LeakageA cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to assure that applicable regulatory requirements and design bases were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to properly evaluate NRC Bulletin 88-04, Potential Safety- Related Pump Loss, 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. The licensee entered this issue into its corrective action program as Condition Report 2014-08381 and initiated actions to re-evaluate NRC Bulletin 88-04. This performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment 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 NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect in the area of human performance because the licensee failed to demonstrate a conservative bias in decision making-practices. Specifically, the licensees determination that the event is not credible failed to consider documented check valve leakage in the auxiliary feedwater system.
05000285/FIN-2014009-112014Q3Fort CalhounFailure to Ensure Safe Operations at Design Basis Low River LevelA cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to ensure that the safety-related raw water pumps are available for safe plant operations down to the design basis low river level. Specifically, station analysis and abnormal operating procedures would not allow operation of the raw water pumps to the design basis low river water level. The licensee entered this issue into its corrective action program as Condition Report 2014-09159 which included actions to reevaluate the capability of the raw water pumps to operate at low river levels. This finding was 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. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of 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; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect in the area of human performance in that the licensee did not ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensee deferred funding for a vendor analysis of the capabilities of the raw water pumps at the design low river level.
05000285/FIN-2014009-142014Q3Fort CalhounFailure to Account for Worst Case Diesel Frequency in Fuel Oil Consumption CalculaA cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to account for design basis conditions in station calculations. Specifically, the licensee failed to account for worst-case electrical frequency when analyzing diesel fuel oil consumption and storage requirements. The licensee entered this issue into its corrective action program as Condition Report 2014-09157 and initiated action to update station calculations. This performance deficiency was more than minor, and therefore a finding, because it affected the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of components that respond to initiating events. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding is of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of problem identification and resolution in that the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000285/FIN-2014009-152014Q3Fort CalhounFailure to Promptly Identify and Correct a Condition Adverse to QualityA non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take corrective actions for a condition adverse to quality. Specifically, the licensee failed to take corrective actions to address multiple issues involving gas voiding of the component cooling water system. As immediate corrective action the licensee placed a maintenance hold on the component cooling water system until adequate fill and vent procedures were established. The licensee initiated corrective actions to analyze the effects of gas accumulation on the component cooling water system and entered this issue into the corrective action program as Condition Reports 2014-08892, 2014-09011 and 2014-09034. This performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and affected the associated objective to ensure availability, reliability, and capability of systems that responds to initiating events to prevent undesirable consequences. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed to operate the component cooling water system within design margins and failed to place special attention on minimizing longstanding equipment issues related to gas voiding in that system.
05000285/FIN-2014009-162014Q3Fort CalhounFailure to Correct Longstanding Software Classification IssuesA non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take timely corrective actions to ensure the proper control and use of software products used in safety related applications. Specifically, the team identified multiple instances of uncontrolled software products in use at the licensees facility following identification of similar deficiencies in 2009 and 2011. The licensee entered this issue into their corrective action program as Condition Report 2014-09162 and initiated action to strengthen their software control program. The performance deficiency was more than minor, and therefore a finding, because if left uncorrected, it could lead to a more significant safety concern. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance in that the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the apparent cause report for Condition Report 2009-04715 stated that a contributing cause was first and foremost (there is) a lack of knowledge associated with the procedural requirements for software control at FCS.
05000285/FIN-2014009-182014Q3Fort CalhounFailure to Complete Corrective Actions in a Timely MannerA non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take timely corrective actions to address deficiencies in station calculations. Specifically, the licensee failed to update station calculations to incorporate actual test data for sluice gate leakage to ensure design basis flood levels do not adversely affect equipment important to safety. The licensee entered this issue into its corrective action program as Condition Report 2014-09156 and initiated actions to update station calculations. This finding was more than minor, and therefore a finding, because if left uncorrected, the finding would have the potential to lead to a more significant safety concern. Specifically, failure to complete accurate calculations that support engineering modifications for mitigating the consequences of an external flooding event could lead to unanalyzed conditions adversely affecting safety related systems or components. The NRC performed an initial screening of the finding in accordance with NRC Manual Chapter IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because: (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program; 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 in that the licensee failed prioritize an update to Calculation FC08081 following completion of the May 2013 in-leakage test.
05000285/FIN-2014002-052014Q1Fort CalhounUntimely Submittal of Required Licensee Event ReportsTwo examples of a cited Severity Level IV violation of 10 CFR 50.73, Immediate Notification Requirements for Operating Nuclear Power Reactors, were identified involving the failure to submit a required licensee event report (LER) within 60 days following discovery of an event requiring a report. In the first example, LER 2013-010-0 was submitted on July 2, 2013, seventy-nine days after the flow imbalance was observed by the licensees staff. In the second example, LER 2013-017-0 was submitted to the NRC on December 27, 2013, 62 days after the event date on the licensees reportability evaluation and sixty-six days after a condition report documented the reportable condition. The licensee initiated CR 2014-01358 on January 29, 2014 to document this repetitive violation. The violation was evaluated using Section 2.2.4 of the NRC Enforcement Policy, because the failure to submit a required LER 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(d)(9) of the NRC Enforcement Policy, this violation was determined to be a Severity Level IV violation. The inspetors 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.
05000324/FIN-2007006-012007Q4BrunswickCapability of Emergency Diesel Generators to Meet Design and Licensing RequirementsThe team identified an Unresolved Item (URI) for failure to translate a key analytical assumption related to operation of the back draft and check dampers into specifications and ultimately into the installed hardware. The teams review of design calculation 0VA- 0033, Tornado Analysis of Diesel Generating Building, Rev. 1 identified the following concerns: nAssumption 7 of this calculation stated that the back draft/check dampers are assumed open in the normal outward flow direction during an atmospheric depressurization event associated with a tornado. The assumption also states that during the subsequent atmospheric repressurization associated with a tornado, the back draft/check dampers would open in the reverse direction to allow reverse inward flow when the Differential Pressure (dP) across damper exceeds 80 psf. There was no surveillance to verify the ability of the back draft/check dampers to open in reverse direction. Furthermore, the teams review of specification 226-002, Sheet Metal Work, Rev. 12; drawings SHW-D-10490, SHW P 2230 Backdraft Damper w/Extra Deep Frame, Rev. C and 9527-01-4283, Technocheck Valve, Rev. 0 (check damper); walk down observations; and interviews with plant personnel identified that the installed back draft and check dampers could not open in the reverse direction. Therefore, conclusions of the calculation 0VA-0033 about the maximum dPs for the structures and the ductwork were not valid or conservative. nAdditionally, this calculation and the design specification failed to address the effects of the high dP predicted in this calculation on the EDG building HVAC hardware other than the ductwork (e.g., in-line fans, recirculation dampers, etc). Progress Energy had not verified that the specified assumptions were translated into the procured and installed HVAC hardware for the EDG building. The analysis for the EDG building HVAC system did not assure that the critical analytical assumption were implemented; thus making it undetermined that the EDG building HVAC system, that serviced all four (4) EDGs for both units of Brunswick Nuclear Plant, would function during and following the design basis tornado. This condition has existed since original plant licensing. This finding was entered into the licensees corrective action program as NCR 00259088 with actions to evaluate the ability of the EDGs actual installed equipment to satisfy the intended safety function during and following the design basis tornado event. This issue is identified as URI 05000325/2007006, 05000324/2007006- 01, Capability of Emergency Diesel Generators to Meet Design and Licensing Requirements. This item is unresolved pending NRC review of the of the licensees analysis of the effects of the as-built configuration on the EDG building HVACs ability to satisfy the intended safety function during and following the design basis tornado event.
05000348/FIN-2007010-012007Q3FarleyUnavailability of CCW System to Automatically Actuate Due to Breaker FailuresThe team concluded that for a period of seven hours and 14 minutes neither train of the CCW system would have automatically started if called upon during a load shed event. For the A-Train, the 1C CCW pump was inoperable after a system operator manipulated the foot pedal on the breaker at 1700 on September 4, placing it in a trip free condition. It remained in this condition until the breaker was replaced and retested at 0014 on September 5. Concurrently, for the B-Train, the 1A CCW pump, although running, would not have automatically restarted from a load shed. This was demonstrated by its failure to manually start from the control room on September 5. (Note, the 1B swing pump was tagged out during this entire period and unavailable.) However, the inspectors did determine that following an actual load shed event in which the 1A CCW pump would have initially failed to start, control room operators would have manually recovered the B-Train of CCW by attempting another start of the 1A CCW pump. The licensee demonstrated that the 1A CCW pump would have restarted on a second attempt. Additionally, operator actions to manually start the 1A CCW pump that failed to start automatically were consistent with licensed operator continuing training and procedure requirements for Emergency Operating Procedures implementation. A detailed time line of the significant events described above is provided in Attachment 1. This item is left unresolved pending evaluation of the operational effects of the CCW system being unable to automatically actuate. It is identified as an Unresolved Item (URI), and is tracked under URI 05000348(364)/2007010-01, Unavailability of CCW System to Automatically Actuate Due to Breaker Failures.
05000348/FIN-2007010-022007Q3FarleyUse of NON-CONFORMING Components in SAFETY-RELATED ApplicationsThe licensee had a number of historical breaker failures for which no cause was identified, and others for which no root cause was performed. Assigning lower significance to the Condition Reports (CRs) associated with these failures resulted in lower tier reviews of cause and less thorough documentation of corrective measures. Had more rigorous evaluation of these failures been conducted, additional trending information may have been obtained. Review of the two databases revealed no significant insight into any but the most recent of breaker failures. The licensee, despite the previous awareness and sensitivity to the gap dimension, placed an Allis-Chalmers breaker which had been removed from the safety-related RHR system for excessive gap clearance (0.063\") into the DG04 cubicle from which the ECH breaker had been removed after its September 4th failure. It was not until two days later that the system engineer detected the discrepancy during paperwork reviews. This item is left unresolved pending investigation of the reasons for the non-conforming breaker being placed in service in a safety-related system. It is identified as an Unresolved Item and is tracked under URI 05000348(364)/2007010-02, Use of Non-Conforming Components in Safety Related Applications.
05000348/FIN-2007010-032007Q3FarleyAdequacy of Root Cause Analysis of the Failed BreakersThe circuit breaker that failed to close on demand in the 1A CCW pump breaker DG04 cubicle on September 5, 2007 was an Allis-Chalmers Type 5kV, 1200 Amp MA-350 air circuit breaker. The licensee had developed a fault tree and initially determined the probable cause for failure to be an excessive gap between the first knuckle of the fourbar and the trip latch. Manufacturer\'s specification for this gap dimension is 0.015 - 0.047 inches. When the gap is within these specifications, it reduces the severity of the impact on the latch when the roll begins its closure stroke. If this gap is large, the closing mechanism can build up more speed, hitting the trip latch harder and introducing more recoil, rapid shaking, and flex. The 1A CCW pump breaker was placed in service with a gap setting of 0.045 inches. In the licensees post-failure investigation, the electrical maintenance technician reported the gap to be measured at 0.063 inches. The NRC inspectors viewed the quarantined breaker on September 14. At this time, the stop bolt gap was observed to be between 0.045 -0.047 inches which was approximately the value that had been reported upon its original return-to-service. Although the breaker had been disassembled, there were no adjustments made and the licensee could not account for the discrepancy between the measurement observed by the inspectors and that reported immediately after the failure. The NRC inspector observed that the measured back plate bending would not account for a stop gap growth of 0.018 inches. The inspector also noted that the breaker exhibited a bent roll-pin through its close latch assembly (which works upon the second knuckle of the four-bar mechanism to release the stored energy of the closing springs), with impact markings consistent with markings on the closing mechanism assembly. The licensee had not previously made note of the bent roll-pin. The licensees root cause strategy was logical. However, their evaluation was not thorough enough to identify the bent roll-pin absent questioning by the NRC inspector; the licensee was focused on the out-of-tolerance stop-gap measurement as the cause. Subsequent to the week in which the inspectors were on site, the licensee reconsidered the root cause for this breaker failure to be the interference-based phenomenon related to the close latch assembly. Review of the licensees formal root cause report revealed that though this was the identified root cause, the fault tree which was the only formal analysis tool supporting the conclusion did not reflect this failure mechanism. Based on observations of the licensees root cause analyses for the 1A CCW pump breaker and other breaker failures discussed section 4.OA5.11, the inspectors raised concerns with the thoroughness of the licensees root cause evaluations. Pending further review of the effectiveness of the licensees root cause evaluations in determining adequate corrective actions, this concern is carried forward as an Unresolved Item and is tracked under URI 05000348,364/2007010-003, Adequacy of Root Cause Analysis of the Failed Breakers.
05000348/FIN-2007010-042007Q3FarleyQuality Control of Replacement Breakers During Manufacturing/DedicationOn October 26, the 1C Charging Pump failed to close when demanded from the control room. The cause was that the latch check switch did not make up electrically. In this instance, the angle of attack by the cantilever arm was not the issue, but rather the throw imposed on the switch was inadequate to cause the switch to make up. Interviews with Farley technicians indicated that they had made adjustments to this particular mechanism. They had either created or failed to correct this condition. Since the pump was not required to be operable in this mode, there were no consequences. As illustrated above, the licensee was challenged with several quality issues concerning the breakers upon receipt from Areva. Pending review of the dedication process and quality control at Areva, the breaker vendor, and their compliance with Part 21 guidance for reporting non-conformance issues, this concern is carried forward as an Unresolved Item and is tracked under URI 05000348,364/2007010-004, Quality Control of Replacement Breakers During Manufacturing/Dedication.
05000369/FIN-2007003-012007Q2Mcguire
McGuire		Debris in Unit 1 ECCS SumpWhile reviewing PIP M-07-1609, the inspectors discovered that on March 17, 2007, the licensee found fire wrap/blanket in the Unit 1 Train B ECCS sump. The blanket was folded over multiple times and partially stuffed into the annular area between the ECCS suction pipe penetration bellows and the bellows guard pipe. The licensee performed an extent of condition inspection for train A, and found a similar fire wrap/blanket in the same respective location. In addition, the PIP indicated that there was other additional material found inside the screened sump structure, behind the suction piping supports, which included non-transportable debris (i.e., two 16P nails, 12\\\" drill bit, 3\\\" cutting wheel, 12\\\" nut, and a 4\\\" partial welding rod stick) and transportable debris (i.e., 3\\\"x6\\\" paper tag dated 3/13/04, a cigarette butt, an empty cigarette package, and several small pieces (<2\\\"x3\\\") of aged, friable duct-tape). The licensee performed several evaluations with regard to this issue during the inspection period which were documented in a Materials Lab Report, dated April 30, 2007, and a Reportability Support Evaluation for PIP M-07-1609, dated May 21, 2007. The Materials Lab Report was included as Attachment 1 to the Reportability Support Evaluation, a thermal expansion analysis was included as Attachment 2, and Attachment 3 was a February 21, 2007, test on ECCS Throttle Valve Duct Tape Flow Testing, which was conducted as part of an evaluation for the Unit 2 duct tape issue documented in Unresolved Item (URI) 05000370/2006005-01. The licensee plans to conduct a more refined throttle valve test for the Unit 2 duct tape issue in the near future. The Unit 1 ECCS debris in the sump issue is greater than minor because if left uncorrected the transportable debris could have had a detrimental affect on the availability and reliability of both trains of the Unit 1 ECCS when called upon during an accident. Specifically, the debris had the potential to have detrimental effects on the high pressure and low pressure ECCS recirculation function. This issue is unresolved pending completion of the NRC review of the licensees reportability evaluation and the results of the more refined duct tape testing. It is identified as URI 05000369/2007003- 01, Debris in the Unit 1 ECCS Sump.
05000369/FIN-2007003-022007Q2Mcguire
McGuire		Reactor Vessel Head Lift Practices Related to Design and Licensing BasisBased on a review of the documents listed in the Attachment of this report related to heavy load lifts of the reactor vessel head and discussions with licensee personnel, the inspectors identified the following issues: The licensee could not demonstrate that a risk assessment had been performed for the increase in risk associated with the lifting and setting of the reactor vessel head. The licensee could not demonstrate that their reactor vessel head lifts, which lift the head to approximately 38 feet over the irradiated fuel in the reactor vessel, were bounded by any design calculations which evaluated the drop of the head through air onto the reactor vessel, upper internals, and irradiated fuel. The licensee could not demonstrate that their procedures for the reactor vessel head removal and installation, ever limited their head lifts to the bounds contained in an August 17, 1984 letter sent to the NRC concerning a load drop analysis for reactor vessel head lifts. The licensee could not demonstrate that their UFSAR had been adequately updated to reflect information and analyses provided to the NRC as the result of all generic communications relative to their resolution of heavy loads issues. The licensee issued PIPs M-07-3099, M-07-3410, and G-07-0449 to address the above issues. A complex maintenance plan was issued for the most recent head installation that occurred on May 18, 2007, to manage risk. A multi-site team has been formed to address the issues above and to work with vendors to determine whether an alternative design and licensing basis exists that bounds past practices. The issues identified above are greater than minor because they are associated with the design control attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown. The issues are also greater than minor because the failure to update the UFSAR could have an impact on safety and may require a license amendment for resolution. These issues are unresolved pending the completion of the licensees investigation into whether an alternative design and licensing basis exists and whether reactor vessel head lifts were ever performed within the bounds of that basis. They are identified as URI 05000369,370/2007003-02, Reactor Vessel Head Lift Practices Related to Design and Licensing Basis
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