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05000266/FIN-2015003-022015Q3Point BeachPotential Failure of Multiple Safety-Related Trains During Flooding EventsThe inspectors identified a finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, "Design Control," for the licensees failure to ensure that a non-Category I (seismic) component failure, that results in flooding, would not adversely affect safety-related equipment needed to get the plant to safe shutdown (SSD) or to limit the consequences of an accident. Specifically, the design of Point Beach did not ensure that the Residual Heat Removal (RHR) pumps would be protected from all credible non-Category I (seismic) system failures. The licensees corrective actions included an extensive internal flooding design review, which will result in an updated Final Safety Analysis Report (FSAR) with a more detailed description of the stations flooding licensing basis; modifications to multiple flood barriers to bring them into compliance with the licensees flooding licensing basis; installation of additional flood level alarms where necessary, and evaluation or modification of service water (SW) piping to properly qualify it as seismic. The inspectors determined that the finding was more than minor because it was associated with the Design Control attribute of the Mitigating System cornerstone and affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the inadequate design resulted in an unanalyzed condition and loss of safety function of the RHR system while the plants were in Modes 4, 5, and 6, when relying on the RHR system for decay heat removal. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors answered yes to question 2 of the screening questions because the finding represented a loss of safety function. Thus the inspectors consulted the Region III Senior Risk Analysts (SRAs) who performed a detailed risk evaluation and determined that the finding was of very low safety significance (Green). The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000266/FIN-2015003-032015Q3Point BeachFailure to Perform a Written Safety Evaluation for FSAR ChangesThe inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very low safety significance for the licensees failure to perform a safety evaluation to demonstrate that the removal of statements from the FSAR did not require a license amendment. Specifically, the licensee failed to perform a safety evaluation to determine whether removing an FSAR statement, which defined the RHR pump cubicle design flood height as seven feet, could be performed without a license amendment. The licensee entered the deficiency in their CAP as Action Request (AR) 02069425 by which the licensee intends on re-evaluating the 1996 FSAR change. The inspectors determined that the finding was more than minor because the finding, if left uncorrected, would become a more significant safety concern. Specifically, inappropriately removing the information from the FSAR allowed the licensee to decrease the design basis flood protection height of the RHR compartments and significantly reduced the available time to isolate the leaking RHR pump seal. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the SDP because they are considered to be violations that potentially impede or impact the regulatory process. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance (Green), because the inspectors answered No to the Mitigating Systems screening questions. The inspectors determined that the associated finding did not have a cross-cutting aspect because the finding was not reflective of current performance.
05000266/FIN-2015003-042015Q3Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 3.8.9; Distribution SystemsOperating, Condition A, which required the licensee to immediately declare associated supported features inoperable for the 4.16 kV safeguards busses. Failure to implement this action subsequently required the licensee to place both units in mode 5 within 36 hours. Contrary to the above, the licensee discovered that numerous occasions existed over the past three years where safetyrelated 4.16kV switchgear associated with B Train EDGs was inoperable due to the inoperability of the W-185A and W-185B, 1A-06 and 2A-06 Switchgear room fans, which were required support systems for the EDGs and associated switchgear. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, and determined that the finding required a detailed risk evaluation which was performed by Region III SRAs. The SRAs gathered data from licensee GOTHIC model calculations, licensee engineering evaluations associated with the POR of the condition and the NRCs Standard Plant Analysis Risk model. Based on the SSCs being available for their respective 24-hour mission time(s), the SRAs determined that the increase in CDF for this issue was negligible and the delta risk is of very low safety significance (i.e., Green). The licensee reported this condition in LER 2015-004-00, which was closed in Section 4OA3 of this report. The licensees corrective actions included improving administrative and procedural controls for removing these fans from service and used lessons learned from this condition to implement corrective actions to improve procedural guidance for similar activities where ventilation systems may cause support system inoperabilities.
05000266/FIN-2015004-012015Q4Point BeachFailure to Follow Fire Protection Program Requirements for Care, Use and Maintenance of Fire HoseThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of license condition 4.F for the licensees failure to have procedures or instructions to prevent firefighting booster hoses from being kinked and/or twisted on hose reels. Specifically, booster hoses were installed on hose reels in both units containments and in the turbine building (TB), which were twisted and kinked. The licensees corrective actions included rewinding hoses in the Unit 2 containment, four hoses in the TB, and creating compensatory measures for hose reels for the Unit 1 containment. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee failed to ensure that activities such as inspection, testing, and maintenance of fire protection systems were prescribed and accomplished in accordance with documented instructions, procedures, and drawings. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue to Green under the Phase 1 Screening Question 1.3.1A, because the inspectors determined that the impact of a fire would be limited to one train/division of equipment for the affected fire areas and at least one credited safe shutdown path would be unaffected. This finding has a cross-cutting aspect of Training (H.9), in the area of human performance, because the licensee did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce, and instill nuclear safety values. Specifically, the inspectors determined that operations personnel were not adequately trained to recognize deficiencies associated with firefighting equipment standards, such as kinked and twisted hoses on hose reels, and subsequently failed to initiate actions to remedy such conditions.
05000266/FIN-2015004-032015Q4Point BeachLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an NCV of TS 5.4.1, Procedures for the failure to maintain the emergency operating procedures (EOPs). The licensees TS 5.4.1 required, in part, that written procedures shall be maintained including the EOPs required to implement the requirements of NUREG-0737 and to NUREG-0737, Supplement 1, as stated in Generic Letter 82-33. During design reviews, the licensee discovered that following a 2012 calculation update, the licensee inconsistently applied pre and post-modification uncertainties that had resulted from a 2010 modification associated with the sensitivity and calibration of both units Subcooling Margin Monitors. Ultimately the calculative errors resulted in 19 EOP Subcooling setpoints being incorrectly calculated. These Subcooling setpoints are used throughout the licensees EOPs network to provide operators with discrete indications for key EOP decision making. Contrary to the above, from April 12, 2012 through November 5, 2015, the licensees EOP network of procedures for both Unit 1 and 2, contained the incorrect setpoints for decision points with respect to subcooling. The licensee entered this issue into the CAP as AR 02089011 and AR 02099152. The inspectors consulted the Region III Senior Reactor Analysts and determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated July 1, 2012 and IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, dated July 1, 2012. The inspectors determined that the issue was a design or qualification deficiency confirmed not to result in a loss of operability; therefore, answered yes to question A.1 in Exhibit 2, Section A, Mitigating SSCs and Functionality. This resulted in the finding screening as Green.
05000280/FIN-2018010-012018Q1SurryFailure to implement the 10 CFR, Part 50, Appendix R, III.G.3 requirements consistent with fire protection license condition 3I.The NRC identified a Green finding and associated non-cited violation (NCV) of the requirements consistent with license condition 3.I, Surry Units 1 and Unit 2. Specifically, the licensee failed to adequately protect fiberglass pipe that is susceptible to fire damage and required for safe shutdown. By not protecting the pipe, the licensee did not ensure the alternative shutdown methodology was implemented with the independence as defined by the 10 CFR 50 Appendix R section III.G.3 requirements.
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.
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-2014009-022014Q3Fort 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-2014009-032014Q3Fort 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-2014009-042014Q3Fort 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-2014009-052014Q3Fort 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-2014009-082014Q3Fort 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-2014009-092014Q3Fort 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-2014009-122014Q3Fort 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-2014009-132014Q3Fort 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-2014009-142014Q3Fort 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-2014009-172014Q3Fort 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-2014009-182014Q3Fort 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-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-2014009-202014Q3Fort 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-2014009-212014Q3Fort 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-2014009-222014Q3Fort 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-2014009-232014Q3Fort 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-2014009-242014Q3Fort 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-2014009-252014Q3Fort 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.
05000298/FIN-2014005-012014Q4CooperFailure to Follow Procedure for Post Maintenance TestingThe inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to follow Special Procedure GEH-TP-116, Procedure for the Operation and Maintenance of the REM*TAKE-2/D-100 Modified REM*TAKE 2, Revision 3, for postmaintenance testing following corrective maintenance. Specifically, the licensee did not follow post-maintenance testing requirements associated with the calibration of the bleeder valve for the REM*TAKE-2/D-100 tool following corrective maintenance to address water intrusion. This resulted in the bleeder valve being misadjusted and nullifying the fail-safe feature of the REM*TAKE-2/D-100 tool. With the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when the supplemental employee inadvertently pressed the disengage button. No reactor fuel was damaged as indicated by normal radiation levels and air samples on the refuel floor and reactor water coolant samples. The licensees immediate corrective actions for the event was to suspended all in-vessel maintenance activities and remove REM*Take-2/D-100 grapple from service and determined functionality of the tool. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2014-06809. The licensees failure to follow the post-maintenance testing requirements in Special Procedure GEH-TP-116 was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and affected the associated objective of maintaining functionality of fuel cladding. Specifically, with the fail-safe nullified, Control Rod Blade 30-47 became disengaged from the REM*TAKE-2/D-100 tool and dropped onto the reactor core top guide when a supplemental employee inadvertently pressed the disengage button. Using Inspection Manual Chapter 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, dated May 09, 2014, inspectors determined that the finding was of very low safety significance (Green) because the finding did not impact the fuel barrier because it: (1) does not increase the potential for failure of the freeze seal or if unmitigated have the potential to cause a disruption of residual heat removal/decay heat removal or a loss of inventory event; (2) does not involve two or more adjacent control rods with the potential to, or actually, add postive reactivity; and (3) does not degrade the ability to isolate a drain down or leakage path. The finding has a cross-cutting aspect in the area of human performance associated with the field presence component because the licensee failed to ensure supervisory and management oversight of work activities including contractors and supplemental personnel (H.2).
05000298/FIN-2014005-022014Q4CooperImplementation of Enforcement Guidance Memorandum 11-003, Revision 2, Causes Conditions Prohibited by Technical SpecificationsDuring Refueling Outage 28, Cooper Nuclear Station performed Operations with a Potential for Draining the Reactor Vessel (OPDRV) activities while in Mode 5 without an operable secondary containment. An OPDRV is an activity that could result in the draining or siphoning of the reactor pressure vessel water level below the top of fuel, without crediting the use of mitigating measure to terminate the uncovering of fuel. Secondary containment is required by TS 3.6.4.1 to be operable during OPDRV. The required action for this specification is to suspend OPDRV operations. Therefore, entering the OPDRV without establishing secondary containment integrity was considered a condition prohibited by TS as defined by 10 CFR 50.73(a)(2)(i)(B). The NRC issued Enforcement Guidance Memorandum (EGM) 11-003, Revision 2, on December 13, 2013, to provide guidance on how to disposition boiling water reactor licensee noncompliances with TS containment requirements during OPDRV operations. The NRC considers enforcement discretion related to secondary containment operability during Mode 5 OPDRV activities appropriate because the associated interim actions necessary to receive the discretion ensure an adequate level of safety by requiring licensees immediate actions to: (1) adhere to the NRC plain language meaning of OPDRV activities, (2) meet the requirements which specify the minimum makeup flow rate and water inventory based on OPDRV activities with long drain down times, (3) ensure that adequate defense in depth is maintained to minimize the potential for the release of fission products with secondary containment not operable by (a) monitoring RPV level to identify the onset of a loss of inventory event, (b) maintaining the capability to isolate the potential leakage paths, (c) prohibiting Mode 4 (cold shutdown) OPDRV activities, and (d) prohibiting movement of irradiated fuel with the spent fuel storage pool gates removed in Mode 5, and (4) ensure that licensees follow all other Mode 5 TS requirements for OPDRV activities. The inspectors reviewed this Licensee Event Report for potential performance deficiencies and/or violations of regulatory requirements. The inspectors reviewed the stations implementation of the Enforcement Guidance Memorandum 11-003, Revision 2, during operations with a potential for draining the reactor vessel. Specific observations included: 1. The inspectors observed that the operations with a potential for draining the reactor vessel activities were logged in the control room narrative logs, and that the log entry appropriately recorded the standby source of makeup designated for the evolutions. 2. The inspectors noted that the reactor vessel water level was maintained at least greater than 21 feet above the top of the reactor pressure vessel flange as required by Technical Specification 3.9.6. The inspectors also verified that at least one safety-related pump was the standby source of makeup designed in the control room narrative logs for the evolutions. The inspectors confirmed that the worst case estimated time to drain the reactor cavity to the reactor pressure vessel flange was greater than 24 hours. 3. The inspectors verified that the operations with a potential for draining the reactor vessels were not conducted in Mode 4 and that the licensee did not move irradiated fuel during the operations with a potential for draining the reactor vessels. The inspectors verified that two independent means of measuring reactor pressure vessel water level were available for identifying the onset of loss of inventory events. Technical Specification 3.6.4.1 requires, in part, that secondary containment shall be operable during operations with a potential for draining the reactor vessel. Technical Specification 3.6.4.1, Condition C, requires the licensee to initiate actions to suspend operations with a potential for draining the reactor vessel immediately when secondary containment is inoperable. Contrary to the above, from October 3, 2014 to October 22, 2014, Cooper Nuclear Station performed operations with a potential for draining the reactor vessel activities while in Mode 5 without an operable secondary containment. Specifically, the station conducted the following seven operations with a potential for draining the reactor vessel activities without an operable secondary containment: Draining reactor recirculation pump without the jet pump plugs fully installed Control rod drive maintenance Removal of jet pump plugs associated with reactor recirculation pump B maintenance Venting the control rod drives Defeating the scram function for two control rod drives and support IVVI inspections Reactor recirculation pump A seal maintenance Control rod drive freeze seal These conditions were reported as conditions prohibited by Technical Specifications. The licensee entered this issue into its corrective action program as Condition Reports CR-CNS-2014-06293. Since this violation occurred during the discretion period described in EGM 11-003, Revision 2, the NRC is exercising enforcement discretion in accordance with Section 3.5, Violations Involving Special Circumstances, of the NRC Enforcement Policy, and, therefore, will not issue enforcement action for this violation. In accordance with EGM 11-003, Revision 2, each licensee that receives discretion must submit a license amendment request within 4 months of the NRC staffs publication in the Federal Register of the notice of availability for a generic change to the standard TS to provide more clarity to the term OPDRV. The Licensee Event Report is closed.
05000298/FIN-2015001-012015Q1CooperInadequate Operations ProcedureThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the inadequate Operations Procedure 2.2.7, Condensate Storage and Transfer System, Revision 56. Specifically, the procedure did not require that the affected system, either the high pressure coolant injection system or the reactor core isolation cooling system, be declared inoperable when one or more of the high pressure coolant injection or reactor core isolation cooling test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, or RCIC-MOV-33, were moved off of their closed (passive safety function position) seats. The license entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-00274. The failure to establish and maintain a correct filling procedure required by Technical Specification 5.4.1.a. was a performance deficiency and resulted in the licensees failure to declare the high pressure coolant injection and reactor core isolation cooling systems inoperable when required to do so. The performance deficiency is more than minor, and therefore a finding, because it is associated with the procedural 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. Specifically, the high pressure coolant injection and reactor core isolation cooling systems were not declared inoperable when their test return line isolation valves, HPCI-MOV-21, HPCI-MOV-24, RCIC-MOV-30, and RCIC-MOV-33, were taken off their normally closed (passive safety function position) seats. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design 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-ofservice 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. The finding has a cross-cutting aspect in the area of human performance associated with Avoid Complacency: Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Individuals implement appropriate error reduction techniques. Specifically, licensee personnel fell into a pattern of acceptance regarding Procedure 2.2.7. This resulted in a failure to question the lack of an operability caution statement, even though there was other guidance in the inservice inspection program to that effect (H.12).
05000301/FIN-2015004-022015Q4Point BeachInadequate Evaluation of Non-Conforming Auxiliary Feedwater System Pipe DefectsThe inspectors identified a finding of very low safety significance and associated Non-Cited Violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to maintain a Unit 2 auxiliary feedwater system (AFW) pipe segment containing linear defects in accordance with the design and material specifications. As a corrective action, the licensee performed light filing to remove the defects from this pipe segment. The licensee entered the failure to maintain the AFW pipe segment in accordance with the design into the corrective action program (CAP) as action request (AR) 02084077, and was evaluating additional corrective actions. This finding was determined to be more than minor in accordance with IMC 0612, Appendix B, because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the licensees failure to maintain the Unit 2 AFW pipe segment containing linear defects in accordance with the design and material specifications could result in an increase in the possibility of pipe leakage or failure. In addition, the failure to maintain the AFW pipe segment containing linear defects in accordance with the design and material specification adversely affected the Mitigating System Cornerstone attribute of Equipment Performance because it could result in failure of AFW piping which would reduce the availability and reliability of the this mitigating system. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors answered Yes to screening question A.1 of Exhibit 2. Although this finding adversely affected the design or qualification of the AFW pipe segments, the finding screened as very low safety significance (Green), because it did not result in the loss of operability or functionality of the affected pipe segment. This finding has a cross-cutting aspect in the Teamwork (H.4) component of the human performance cross-cutting area. Specifically, the licensees Projects Team responsible for the AFW modifications did not effectively communicate and coordinate with the licensees Programs Engineering Group for resolution of the AFW pipe nonconforming conditions to ensure nuclear safety was maintained.
05000339/FIN-2018011-012018Q2North AnnaFailure to ensure compliance with the Technical Specification (TS) 5.4.1.a requirement relevant to procedures for plant firesThe NRC identified a Green finding and associated non-cited violation (NCV) of the TS 5.4.1.a requirement to establish and maintain fire contingency action procedures based upon the licensees failure to effectively perform reviews during the revisions of the procedures in accordance with procedure VPAP-0502, Procedure Process Control. The failure led to undetected errors and was a performance deficiency that was determined to be more than minor because, if left uncorrected, it could potentially lead to a more significant safety concern during Appendix R fire events.
05000395/FIN-2010005-012010Q4SummerFailure to Correct Condition Adverse to Quality Involving Inadequate EDG Engine Driven Pump Preventative MaintenanceThe inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify and correct a condition adverse to quality following the February 10, 2010, failure of the A Emergency Diesel Generator (EDG) jacket water pump mechanical seal. Specifically, the licensee failed to identify and correct inadequacies in their EDG preventive maintenance program for monitoring engine driven pump seal leakage in accordance with vendor recommendations, leading to subsequent A EDG jacket water seal leakage going unidentified from approximately June 1, 2010, until October 20, 2010. The licensee initiated condition report (CR)-10-03861 and implemented requirements and operator training to conduct proper seal leakage monitoring during subsequent EDG operations. The inspectors determined that the licensees failure to take adequate corrective actions to identify and correct inadequacies in the EDG PM program for monitoring EDG engine driven pump seal leakage in accordance with vendor recommendations was a performance deficiency that was within the licensees ability to foresee and correct. This finding is more than minor because if left uncorrected, the issue would become a more significant safety concern, in that, the potential exists for unidentified engine driven pump seal leakage that could lead to EDG failure. This issue is associated with the equipment performance attribute of the Mitigating System cornerstone and adversely affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to take adequate corrective actions to identify inadequacies in the EDG preventive maintenance program for monitoring EDG engine driven pump seal leakage in accordance with vendor recommendations could adversely affect the reliability of the EDGs. This finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for mitigating systems. The finding was determined to be of very low safety significance (Green) because it did not actually result in the loss of the EDG system safety function or the loss of function of a single EDG. The cause of this finding was directly related to the problem evaluation cross-cutting aspect in the corrective action program component of the Problem Identification and Resolution area because the licensee did not thoroughly evaluate the February 10, 2010, jacket water pump mechanical seal failure event and identify nonconformance with the vendor recommended visual inspections of engine driven pump seals during EDG operations (P.1(c))
05000395/FIN-2010005-022010Q4SummerLicensee-Identified ViolationTS 3.0.4 requires, in part, that when a Limiting Condition for Operation is not met, entry into a mode or other specified condition shall only be made when the associated actions to be entered permit continued operation in the mode for an unlimited period of time. TS 3.7.1.6 requires that each FWIV be operable in Modes 1-3 and the associated ACTIONS for Mode 1 operation do not allow operation for an unlimited period of time. Contrary to the above, on September 25, 2010, due to inadequate procedural guidance, operators did not ensure that one of the three FWIVs air accumulator pressure was above the minimum for operability when Mode 1 was entered from Mode 2. This condition existed for seven minutes until the minimum pressure for operability was attained in the FWIV air accumulator. The violation was determined to be of very low safety significance because of the short duration that pressure was below the minimum for operability and subsequent FWIV testing determined that the FWIV remained capable of performing its design function at the reduced air accumulator pressure. The licensee planned to revise GOP-4A, Power Operation Mode 1 Ascending, to ensure verification that all FWIV actuator pressures are above the minimum pressure for operability prior to allowing entry into Mode 1. The licensee identified and addressed this issue in their corrective action program as CR-10-03766
05000395/FIN-2011002-012011Q1SummerNone10CFR 55.25 states If, during the term of the license, the licensee develops a permanent physical or mental condition that causes the licensee to fail to meet the requirements of 55.21 of this part, the facility licensee shall notify the Commission, within 30 days of learning of the diagnosis, in accordance with 50.74(c). For conditions for which a conditional license (as described in 55.33(b) of this part) is requested, the facility licensee shall provide medical certification on Form NRC 396 to the Commission (as described in 55.23 of this part). Contrary to this, the licensee identified that they did not notify the Commission within 30 days after a licensed operator was diagnosed with a permanent physical medical condition as required by 10 CFR 55.25. This was identified in the licensees CAP as CR 11- 00304 and 11-00031. This finding was of very low safety significance because the licensed operator performs non-licensed shift technical advisor duties and has been inactive during the time in question.
05000395/FIN-2011003-012011Q2SummerFailure to Adequately Assess and Manage Risk of Switchyard Maintenance Activities During Lowered RCS Inventory ConditionsThe inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the failure to perform an adequate risk assessment and implement approved high risk management contingency plans for work in the stations electrical switchyard. Specifically, on April 21, 2011, operations work control personnel failed to adequately assess the impact of work activities in the switchyard involving the use of vehicles, resulting in outage high risk management actions that prohibited the movement of vehicles during lowered reactor coolant system (RCS) inventory conditions from being implemented. Following the inspectors identification of this issue, the licensee adequately assessed and managed the increase in risk for the activities. The issue was entered into the licensees corrective action program as condition report CR-11-01908. The failure to perform an adequate risk assessment and implement high risk evolution contingency plans for work in the stations switchyard was a performance deficiency within the licensees ability to foresee and correct. This finding was associated with the Initiating Events Cornerstone and affected the cornerstone objective for limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown, such as, loss of offsite power (LOOP) due to trucks damaging critical electrical components in the switchyard. The inspectors determined that the finding is more than minor because it was similar to both the more than minor examples 7.e and 7.g in NRC Inspection Manual Chapter (IMC) 0612, Appendix E, Examples of Minor Issues, because the risk assessment for the switchyard work activity failed to consider the impact of vehicle movements resulting in outage high risk management actions that prohibited the movement of vehicles during lowered RCS inventory conditions from being implemented. A Significance Determination Process (SDP), Phase 1 screening determined that the performance deficiency represented an increase in the likelihood of a LOOP during shutdown and therefore the risk was estimated using NRC IMC 0609, Appendix G, Shutdown Operations Significance Determination Process. A Phase 2 SDP risk evaluation was done by a regional senior risk analyst using IMC 0609, Appendix G, Attachment 2. The major assumptions of the analysis were that the plant was in plant operating state (POS-2) in Mode 6, with the RCS vented and the residual heat removal (RHR) system in service for decay heat removal. Time to boil was estimated at 35 minutes with an estimated time to core damage of 8.8 hours. The exposure period was approximately 2.5 hours. The LOOP initiating event likelihood was increased by one order of magnitude due to the impact of the performance deficiency. Multiple (i.e., three) qualified sources of offsite power and both onsite emergency diesel generators were available when the vehicles were moved into the switchyard. Recovery credit for restoration of offsite power was included. The dominant sequence was a LOOP with failure of emergency power sources causing a loss of RHR and failure to recover offsite power or emergency power prior to core damage ensuing. The risk was mitigated by the short exposure period and the availability of mitigating system equipment. The result of the analysis was a core damage frequency risk increase of <1E-6/year, a finding of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, because personnel did not appropriately plan and coordinate switchyard work activities consistent with nuclear safety by incorporating appropriate outage risk insights and risk management contingency plans
05000395/FIN-2011003-022011Q2SummerFailure to Perform ISI General Visual Examinations of Containment Moisture Barrier Associated with Containment Liner Leak Chase Test Connection Threaded Pipe PlugsThe inspectors identified a Green NCV of 10 CFR Part 50.55a, Codes and Standards, involving the licensees failure to properly apply Subsection IWE of ASME Section XI for conducting general visual examinations of the metal-to-metal pipe plugs installed in the containment liner channel weld leak chase test connections that provide a moisture barrier to the containment liner seam welds. Following the inspectors identification of this issue, the licensee conducted the visual examinations and found missing pipe plugs and water in four of the leak chase test connection zones. The licensee adequately assessed and corrected the deficiencies prior to entering Mode 4 (Hot Shutdown) to ensure the integrity of containment was maintained. The issue was entered into the licensees corrective action program as condition report CR-11-02834. The failure to conduct a general visual examination of 100 percent of the moisture barriers intended to prevent intrusion of moisture against inaccessible areas of the containment liner at metal-to-metal interfaces which are not seal welded, was a performance deficiency that was within the licensees ability to foresee and correct. This finding was of more than minor significance because the failure to conduct required visual examinations and identify the degraded moisture barriers which allowed the intrusion of water into the four liner leak chase channels, if left uncorrected, could have resulted in more significant corrosion degradation of the containment liner or associated liner welds. The finding was associated with the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, visual examinations of the containment metal liner provide assurance that the liner remains capable of performing its intended safety function. The inspectors used IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 Initial Screening and Characterization of Findings, and determined that the finding was of low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of the reactor containment. A cross-cutting aspect was not identified because the finding does not represent current licensee performance.
05000395/FIN-2011003-032011Q2SummerFailure to Conduct Adequate Testing of Appendix R Fire SwitchesThe NRC identified an apparent violation (AV) of the Virgil C. Summer Nuclear Station Operating License Condition 2.C.(18), Fire Protection System, related to the licensee\\\'s failure to implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report (FSAR). Specifically, the licensee failed to adequately test the isolation function of all 10 CFR 50 Appendix R isolation local control transfer switches ( fire switches ), including the B EDG fire switch, designed to assure isolation of safe shutdown equipment from the control room in the event of a control room evacuation due to a fire. This resulted in the licensee not identifying a wiring discrepancy that had existed in the B EDG fire switch circuitry since original plant startup until its discovery on April 29, 2010, that would have defeated the Appendix R isolation function during a design basis fire event requiring evacuation from the Control Room. The issue was entered into the licensees corrective action program as condition report CR-10-01814. The failure to demonstrate proper Appendix R isolation capability of safe shutdown equipment controlled from remote shutdown locations during surveillance testing of Appendix R fire switches is a performance deficiency that was within the licensees ability to foresee and correct. The inspectors determined that the finding is more than minor because it was associated with both the procedure quality and protection against external events (i.e., fire) attributes of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to adequately test Appendix R isolation contacts associated with fire switches contributed to not identifying a wiring discrepancy in the B EDG fire switch circuitry that defeated its Appendix R isolation function. This condition could have led to the improper operation of the switch or prevented the B EDG output breaker from automatically closing during certain fire scenarios due to fire damage of the electrical circuitry. In accordance with NRC IMC 0609, Significance Determination Process, the inspectors performed a Phase 1 screening analysis and determined that since the finding affected the fire protection defense-in-depth strategies involving post fire safe shutdown systems, the finding required a significance evaluation under IMC 0609, Appendix F, Fire Protection Significance Determination Process. Using Appendix F, Attachment 1, Fire Protection SDP Phase 1 Worksheet, the inspectors determined that the category of post fire safe shutdown was affected and the finding required a Phase 2 analysis by a senior reactor analyst. The significance of this finding is to be determined pending completion of the Phase 2 analysis. A cross-cutting aspect was not identified because the finding does not represent current licensee performance.
05000395/FIN-2011003-042011Q2SummerInadvertent Safety Injection in Mode 3 Due to Opening C Main Steam Isolation ValveThe inspectors confirmed that the operator response to the event was appropriate and consistent with emergency and abnormal operating procedure requirements. Following the SI actuation, all plant systems functioned as designed and emergency core cooling system water was injected into the RCS. The operators were successful in timely termination of unnecessary injection flows and preventing potential pressurizer overfill and adverse RCS overpressure conditions. The plant was effectively stabilized in Mode 3. The licensee correctly determined that no emergency action level entry condition was reached; however, the event was determined to be reportable to the NRC under the 4 hour non-emergency requirement of 10 CFR 50.72(b)(2) for an emergency core cooling system discharge to the RCS. The licensee reported the notification in a timely manner. Based on interviews with the operators following completion of plant recovery actions, the inspectors noted that the operators had failed to recognize that the main steam line header downstream of the MSIVs had been depressurized when the MSIVs and their bypass valves were closed earlier in the shift. In addition, procedural guidance for stroking the MSIV, such as the stroke test procedure (i.e., STP-130.004D, Rev. 1, Main Steam Isolation Valve Full Stroke Test ), was not formally utilized when the valve was opened at the request of I&C. This procedure contained a signoff action to verify current plant conditions will permit performance of the stroke test, and could have provided an opportunity for the operators to have recognized that the main steam line header was depressuried, had this procedure been utilized. The licensee documented this event in their CAP as CR-11-03001 and planned to submit a LER within 60 days of the event date. At the end of the inspection period, the inspectors were awaiting the completion of the licensees root cause evaluation results to understand and properly characterize the potential performance deficiencies associated with this event. This issue is unresolved pending inspector review of the licensees evaluation, proposed corrective actions, and review of the licensees LER in order to characterize the potential performance deficiencies associated with this event. This unresolved item (URI) is identified as 05000395/2011003-04, Inadvertent Safety Injection in Mode 3 Due to Opening C Main Steam Isolation Valve.
05000395/FIN-2011004-012011Q3SummerFailure to Implement a Procedure for Manipulation of the C Main Steam Isolation ValveA self-revealing, non-cited violation was identified for the failure to comply with Technical Specification 6.8.1 to adequately implement a main steam operating procedure during manipulation of the C main steam isolation valve (MSIV) resulting in excessive steam generator line differential pressure and subsequent safety injection. The issue was entered into the licensees corrective action program as condition report CR-11-03001. The failure to implement a procedure for manipulation of the C MSIV was a performance deficiency (PD). The PD was more than minor and therefore a finding because it impacted the initiating events cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown and the related attribute of human performance because the licensee failed to properly implement a procedure controlling the manipulation of a MSIV. In accordance with Inspector Manual Chapter 0609, Significant Determination Process, the inspectors performed a Phase 1 analysis and determined the finding was of very low safety significance or Green because the finding did not contribute to both the likelihood of both a reactor trip and the unavailability of mitigation equipment and associated functions. This finding involved the cross-cutting area of human performance, the component of the resources, and the aspect of procedure use and adherence, H.4(b), because the licensee failed to adequately follow procedures.
05000395/FIN-2011004-022011Q3SummerLicensee-Identified ViolationContrary to requirements, the licensee failed to implement and maintain in effect all provisions of the approved fire protection program as described in the FSAR for the facility, in that, the Appendix R fire switch test program did not adequately verify that the switches were capable of performing their required isolation function. This finding has been entered into the licensees corrective action program as condition report CR-10-01814. The finding affected safe shutdown and was judged to represent moderate degradation. Because the finding involved main control room (MCR) fire scenarios and scenarios in multiple fire areas, a phase 3 SDP analysis was performed by a regional senior reactor analyst. The finding was determined to have existed since 1983 when a modification temporarily installed the jumper wire; therefore a one year exposure period was utilized for the analysis. Only fires which could lead to MCR abandonment requiring use of the B EDG isolation switch and which also would damage the B EDG output breaker control circuit would contribute to the risk of the performance deficiency. Only fire scenarios in the MCR (within the main control board (MCB)) and in the cable spreading room which impacted the cable tray and main termination cabinet associated with the B EDG represented credible fire scenarios which could lead to risk from the PD. Factors which mitigated the risk from the PD included: the few credible fire ignition sources, use of thermoset cables, the low cable loading in the specific MCB section housing the EDG and offsite power circuit breaker control switches, detection in the main termination cabinet, and the proceduralized actions for local operation of the B EDG breaker. The dominant sequence was a fire in either the MCR or cable spreading room damaging the EDG and offsite power breaker controls requiring MCR abandonment coupled with failure of the B EDG breaker to operate due to the PD and failure of the operator to locally close the B EDG breaker resulting in core damage from inadequate core cooling. The SDP phase 3 evaluation determined that the risk of the finding was an increase in core damage frequency of <1E-6/year, a Green finding of low safety significance.
05000395/FIN-2011005-012011Q4SummerFire Protection Program Requirements for Procurement and Use of Fire HoseAn unresolved item (URI) was identified by the inspectors relating to the procurement and use of 1.5 non-collapsible rubber hose used throughout the plant for fire protection. On November 3, 2011, the inspectors identified that the licensee was using 1.5 Thermoid Mexacon General Purpose (GP) 250 PSI hose at a majority of the Fire Protection Evaluation Report (FPER) and non-FPER interior hose reel stations. The inspectors determined that HBD Thermoids specification for the hose notes a minimum bend radius of 10.5 , conflicting with the 5 radius of the respective hose reels. On November 18, 2011, inspectors reviewed the engineering evaluation for the Thermoid hose. The engineering evaluation failed to include other fire hoses as part of an extent of condition review. Inspectors questioned the licensee on why 1.5 Gates Duro Flex hose was also being used as non-collapsible hose as it has a minimum bend radius of 12 . The licensee initiated CR 11-05852 and found that the Gates Duro Flex hose was installed on fire hose reels with a 5 radius. The licensee took immediate compensatory actions including staging collapsible fire hose at the two affected FPER hose reels. On November 21, 2011, an engineering evaluation for the Gates Duro Flex hose determined that two FPER hoses and four non-FPER hoses were not compatible with the hose reels due to exceeding the minimum bend radius of the hose. Additional information is required for evaluation and finalization of the performance deficiency. The issue is identified as URI 05000395/2011005-01, Fire Protection Program Requirements for Procurement and Use of Fire Hose.
05000395/FIN-2011005-022011Q4SummerLightning Induced Trips of Safety-Related ChillersA URI was identified by the inspectors for safety-related chiller trips due to lightning. On October 13, 2011, following a lightning strike at the station the A train safety-related chiller tripped on overcurrent when the 250 amp limit for circuit 1 of the two circuit chiller was exceeded. The B train of chilled water system was also inoperable which required the licensee to enter TS 3.0.3. The B train chiller was returned to an operable status to allow exiting TS 3.0.3. The licensee entered the problem into their CAP as CR-11-05225 and performed an operability evaluation. The evaluation also referenced CR-11-03187 that documented a similar, previous trip of the C chiller while aligned to A train power on June 5, 2011, during a lightning storm. The inspectors continue to evaluate the regulatory aspects of these events. This issue is identified as URI 05000395/2011005-02, Lightning Induced Trips of Safety-Related Chillers.
05000395/FIN-2011005-032011Q4SummerLicensee-Identified Violation10 CFR 50, Appendix B, Criterion III, Design Control states, in part, that measures are established to ensure applicable regulatory requirements and design basis are correctly translated into procedures. Contrary to this, October 13, 2009, the licensee identified that they failed to have adequate measures in place to ensure that correct design bases were translated into procedures for RHR system operation. This issue is more than minor because it impacted the mitigating systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and the attribute of procedure quality. The SDP screening determined that the PD affected both short term and long term core decay heat removal during shutdown and was evaluated with NRC Inspection Manual Chapter procedure 0609 Appendix G Shutdown Operations Significance Determination Process . Since the PD affected both trains of the Residual Heat Removal (RHR) System, the finding increased the likelihood that a loss of decay heat removal would occur during shutdown and a phase 3 SDP evaluation was performed by a regional SRA. A bounding analysis was performed assuming a conditional core damage probability of 1.0 for any loss of coolant (LOCA) or steam generator tube rupture (SGTR) initiators occurring during the exposure period. The analysis assumed a reduction in LOCA and SGTR pipe rupture frequencies of 10% of nominal for the exposure at shutdown conditions due to the reduced pressures. The exposure period was 4.8 hours over a three year period. No recovery was assumed in the bounding analysis. The dominant sequence would be a LOCA or SGTR at Modes 3/4 with a loss of both RHR pumps due to flashing at the suction leading to core damage due to a loss of core heat removal. The risk was mitigated by the short exposure period. The result of the phase 3 SDP analysis was an increase in core damage frequency < 1E-6 a Green finding of very low safety significance. This issue is in the licensees CAP as CR-09-03980.
05000395/FIN-2011501-012011Q3SummerAdequacy of Procedures to Assess and Monitor Emergency Radiological ReleasesWhile reviewing procedures used during the biennial emergency preparedness exercise, inspectors questioned the adequacy of licensee procedure EPP- 7, Environmental Monitoring, Rev. 11, used to direct deployment of field monitoring teams to effectively locate radioactive release plume boundaries. The licensee initiated corrective action CR-11-04977 to evaluate this issue. Inspectors concluded that further review of information related to the procedures adequacy, and its interrelationship with other licensee radiological assessment procedures is necessary to determine if the issue constitutes a violation of regulatory requirements. This issue is identified as URI 05000395/2011501-01, Adequacy of Procedures to Assess and Monitor Emergency Radiological Releases.
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