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05000285/FIN-2014002-012014Q1Fort CalhounFailure to Make Required 10 CFR 50.46 Report Within Required TimeA Severity Level IV non-cited violation of 10 CFR 50.46, Acceptance criteria for emergency core cooling systems (ECCS) for light-water nuclear power reactors, was identified involving the failure to submit a report within 30 days of discovery of a significant change in the application of the ECCS model that affected the peak cladding temperature. The licensee submitted the required 10 CFR 50.46 report late on September 20, 2013 (ML13266A108). This report was subsequently reviewed by the NRC staff date October 2, 2013, and determined to be acceptable. The NRC staff determined that while the configuration change to the HPSI system resulted in a higher peak cladding temperature, it is within the regulatory requirements of 10 CFR 50.46(b)(1). The licensee initiated CRs-2014-00674 and 2014-01356 to address issuance of the late report. This performance deficiency was determined to be subject to traditional enforcement because it impeded the regulatory process, in that the failure to submit a timely report of significant ECCS analytical changes prevented the NRC technical staff from independently evaluating the potential safety implications of reductions in safety injection flow into the reactor during an accident. This violation was determined to be a Severity Level IV violation because it is consistent with the examples in Paragraph 6.9.d of the NRC Enforcement Policy. Because this violation is subject to traditional enforcement, no cross-cutting aspects have been assigned.
05000285/FIN-2014002-072014Q1Fort CalhounInadequate 10 CFR 50.59 Screening for Containment Spray Design ChangeA cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified involving the failure to take timely corrective action for a condition adverse to quality. Specifically, the licensee failed to restore compliance following NRC identification of the licensees failure to correct a runout condition of the containment spray system (CS) documented in NCV 05000285/2008003-05, in August 2008. Licensee corrective actions to correct the issue included completion of an analysis of containment spray pump operation during the main steam line break (MSLB) event; revision of CS design documentation; analysis of motor performance by an electrical vendor; and completion of a temporary modification to throttle the CS pump discharge valves to provide additional system resistance preventing pump runout. Future corrective actions include a permanent design change to prevent CS pump runout. The licensee initiated CR 2014-02242 on February 19, 2014, to document this failure to restore compliance. This finding was more than minor because it adversely impacted the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. The inspectors reviewed NRC IMC 0609, Attachment 4, Initial Characterization of Findings , Table 3 SDP Appendix Router. While this issue was identified during a refueling outage, the inspectors determined that the majority of the exposure time for this violation occurred with the reactor at power and should be evaluated using the Significance Determination Process in accordance with IMC 0609, The Significance Determination Process (SDP) for Findings at- Power, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding did not represent an actual open pathway in containment or containment isolation logic, nor did the finding represent an actual reduction in the function of containment hydrogen igniters. Based on the guidance in the Exhibit 3 checklist the inspectors determined that the finding was of very low safety significance. The inspectors determined that the finding had a cross-cutting aspect of avoiding complacency in the human performance area, because the licensees staff failed to recognize latent issues even while expecting successful outcomes.
05000285/FIN-2014007-062014Q1Fort CalhounFailure to check the adequacy of the design for the Reactor Vessel Head structural elementsThe inspectors identified a green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure the design of the reactor vessel head stand met current licensing basis requirements. Specifically the design of the reactor vessel head stand did not meet the requirements as defined in the Updated Safety Analysis Report. Subsequently, the licensee evaluated the non-conformances to provide a reasonable assurance of operability, and planned corrective actions to restore the structures to design basis requirements. The failure to ensure the design of structures, systems, or components meet their current licensing basis is a performance deficiency. The licensee documented the finding in the corrective action program as Condition Report 2014-04218. The performance deficiency was determined to be more than minor because it adversely affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the safety injection system and the shutdown cooling system. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix G, Shutdown Operations Significance Determination Process (SDP), Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Operational Checklists for Both PWRs and BWRs, dated May 25, 2004, and determined that the finding is of very low safety significance (Green) because the finding did not require quantitative assessment. The finding has a crosscutting aspect in the area human performance because the licensee did not ensure the CIS at elevation 1045 ft. for storage of the reactor vessel head maintained adequate design margin.
05000285/FIN-2014007-052014Q1Fort CalhounFailure to correct conditions adverse to quality in the containment internal structure and auxiliary buildingThe inspectors identified multiple examples of a green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify and correct conditions adverse to quality. Specifically, the licensee failed to take appropriate corrective action since 1997 when it was identified that the containment internal structure and auxiliary building had discrepant documentation between the size of structural beams and columns shown in drawings versus calculations. Subsequently, the licensee evaluated the non-conformances to provide a reasonable assurance of operability, and planned corrective actions to restore the structures to design basis requirements. The failure to correct conditions adverse to quality is a performance deficiency. The licensee documented the finding in the corrective action program as Condition Report 2014-04219. The performance deficiency was determined to be more than minor because it adversely affected the Mitigating Systems cornerstone objective of ensuring the availability, reliability, and capability of the safety injection system and the shutdown cooling system. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, and determined that the finding is of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating SSC that did not affect operability or functionality. The finding does not have a cross-cutting aspect because it is not reflective of current plant performance.
05000285/FIN-2014007-042014Q1Fort CalhounFailure to perform an immediate operability determinationThe inspectors identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an operability determination as required by NOD-QP-31, Operability Determinations Process (ODP). Specifically, following the failure of an auxiliary building ventilation damper to open the licensee failed to evaluate the operability of equipment potentially impacted. Subsequently, the licensee performed an evaluation that provided reasonable assurance of operability. The licensee documented the finding in the corrective action program as Condition Report 2014-00211. The performance deficiency is more than minor, and therefore a finding, because if left uncorrected the failure to determine the ability of a structure, system, or component to perform its current licensing basis function in accordance with station procedures could lead to a more significant safety concern. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, and determined that the finding is of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, represent a loss of system function or loss of function of single or multiple trains of equipment. The finding has a cross-cutting aspect in the area 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.
05000285/FIN-2014007-032014Q1Fort CalhounFailure to implement an adequate PMT procedureThe inspectors identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, involving the failure to establish and implement an adequate procedure for Post Maintenance Testing (PMT). Specifically, following maintenance on a raw water strainer the licensees PMT failed to verify the flow capacity through the system required to determine operability. The failure to establish an adequate procedure to determine PMT is a performance deficiency. Subsequently, the licensee performed an adequate PMT verifying system flows were adequate and documented the deficiency in the corrective action program as Condition Report 2014-03084. The performance deficiency is more-than-minor and therefore a finding because inadequate PMT following maintenance activities could adversely affect the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, and determined that the finding is of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating SSC, represent a loss of system function or loss of function of single or multiple trains of equipment. The finding has a cross-cutting aspect in the area of problem identification and resolution because the licensee did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance.
05000285/FIN-2014007-022014Q1Fort CalhounFailure to follow an immediate operability determination procedureThe inspectors identified a green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow an operability determination procedure. Upon identifying that that a relief valve had not been tested within the required frequency the licensee failed to adequately address how this deficiency could affect the safety function of the component. Specifically, the licensee concluded the valve was operable based only on the consideration that it was not leaking. Subsequently, the licensee performed an evaluation providing adequate reasonable assurance of operability. The licensee documented the finding in the corrective action program as Condition Report 2014-03055. The performance deficiency is more than minor, and therefore a finding, because if left uncorrected the failure to determine the ability of a structure, system, or component to perform its current licensing basis function in accordance with station procedures could lead to a more significant safety concern. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, and determined that the finding is of very low safety significance (Green) because it did not affect the design or qualification of a mitigating SSC, represent a loss of system function or loss of function of single or multiple trains of equipment. The finding has a cross-cutting aspect in the human performance area because the licensee did not create and maintain complete, accurate, and up-to-date documentation.
05000285/FIN-2014007-012014Q1Fort CalhounFailure to Follow Procedures for Classifying Component FailuresThe inspectors identified a Green finding for the licensees failure to follow a procedure for classifying component failures. Specifically, the licensees failure to follow Procedure FCSG-69-5, Failure Identification and Reporting, is a performance deficiency. As a result, the failure of the Turbine-Driven Auxiliary Feedwater Pump, FW-10, to start on demand was not identified as a functional failure. Subsequently, the licensee properly evaluated the system performance taking into consideration the functional failure. The licensee documented the finding in the corrective action program as Condition Report 2014-04217. The performance deficiency is more than minor, and therefore a finding, because if left uncorrected the performance deficiency could have the potential to lead to a more significant safety concern. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The finding is of very low safety significance (Green) because it did not affect the design or qualification of a mitigating system, structure, or component (SSC), represent a loss of system function, or loss of function of single or multiple trains of equipment. The finding had a human performance cross-cutting aspect associated with training because the licensee failed to provide adequate training to the engineering staff.
05000285/FIN-2014002-092014Q1Fort CalhounFailure to Adequately Implement Design Requirements for Containment Air Cooler Pipe SupportsA non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified involving the failure to ensure the adequacy of the U-bolts for containment air cooler pipe supports VAS-1 and VAS-2. Specifically the U-bolt design was non-conservative with respect to the design basis requirements. The licensee entered these issues into the corrective action program as CR 2013-03722. The licensee revised the calculation to support operability. In addition, the licensee generated engineering change EC59570 to fix the degraded VAS-1 and VAS-2 supports. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of several safety injection tank valves. Specifically, the one-directional U-bolts for VAS-1 and VAS-2 are not designed to withstand two-directional loading and the condensate drain piping line has the potential to adversely impact the safety injection tank discharge isolation valves HCV-2934 and HCV-2974 during a design basis event. The licensee updated calculation FC05918 and provided an operability evaluation to address the degraded condition. The inspectors reviewed the information and did not find any issues. Using Inspection Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process (SDP) for findings at-power, both dated June 19, 2012, the inspectors determined performance deficiency affected the mitigating systems cornerstone and screened to Green because the finding affected the design and qualification of a mitigating SSC but remained operable. The inspectors used the at-power SDP because the condition existed since construction and while the plant was predominantly at power. The inspectors determined there was no cross-cutting aspect associated with this finding because the calculation was from the 1980s, and therefore was not reflective of current performance.
05000285/FIN-2014002-082014Q1Fort CalhounFailure to Adequately Design Anchorage for Containment Spray and Raw Water System Pipe SupportsDuring a previous inspection, the NRC reviewed multiple calculations for pipe supports on the raw water and containment spray systems and found that the calculations had several errors related to the design requirements for anchorage. The NRC issued an apparent violation AV 05000285/2013012-08, Failure to adequately design anchorage for containment spray and raw water system pipe supports in NRC Inspection Report 05000285/2013-012 (ML 13144A772). The licensee performed an operability determination for the affected calculations and found that the anchorage for the raw water and containment spray piping supports were operable. The NRC reviewed the evaluations and concluded that reasonable assurance of operability existed for the affected components. The inspectors determined that the failure to ensure adequacy of the anchorage of the aforementioned Containment Spray Pipe Supports and Raw Water Pipe Supports was not in accordance with design basis requirements and was a performance deficiency. The performance deficiency was determined to be more than minor because it required calculations to be re-performed to prove the system was operable, and it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of the containment spray system and raw water system. Using Inspection Manual Chapter 0609, Attachment 4 Initial Characterization of Findings, and Appendix A The Significance Determination Process (SDP) for findings at-power, both dated 6/19/12, the inspectors determined the performance deficiency affected the mitigating systems cornerstone and screened to Green because the finding affected the design and qualification of a mitigating SSC but remained operable. The inspectors used the at-power SDP because the condition existed since construction and while the plant was predominantly at power. The inspectors determined there was no cross-cutting aspect associated with this finding because the calculations were from the 1980s and therefore were not reflective of current performance. Title 10 CFR Part 50, Appendix B, Criterion III, Design Control states, in part, that the design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Contrary to this requirement the inspectors identified that calculations FC00607, FC01785, FC01786, FC01791, FC01864, FC01691, FC01902, FC02409, FC02412, FC04228, FC02433, FC02436, and FC02425 for the raw water and containment spray systems failed to ensure adequacy of the design. Specifically, these anchorage calculations did not conform to applicable design requirements from approximately 1980 until June 2013. The licensee entered these issues into the corrective action program as CR 2013-05304 and performed an operability determination as immediate actions. Long term actions to resolve the errors in the calculations are also implemented by the referenced CR. This violation is being treated as an NCV, consistent with Section 2.3.2.a of the Enforcement Policy.
05000285/FIN-2014002-042014Q1Fort CalhounFailure to Request a License Amendment for Required Change to Technical SpecificationsA Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, and an associated Green finding was identified involving the failure to request a license amendment for a facility change that required a change to the Technical Specifications. This issue is also associated with a Green finding related to the licensees failure to follow Procedure NOD-QP-3, 10 CFR 50.59 and 10 CFR 72.48 Reviews, and Procedure FCSG-23, 10 CFR 50.59 Resource Manual, both of which require submittal of a license amendment request prior to making a facility change that requires a change to Technical Specifications. The licensee initiated CR 2014-01029 on January 23, 2014, to document this violation and track corrective actions. This performance deficiency was considered to be of more than minor safety significance because it was associated with the procedure quality attribute of the mitigating systems cornerstone and it adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to follow station procedures for the 10 CFR 50.59 process caused the Technical Specifications to become insufficient to ensure that the limiting conditions for operation will be met. Using Inspection Manual Chapter 0609 Appendix G, Checklist 4, the inspectors determined that the finding did not result in the loss of any accident mitigation capability and did not require a quantitative risk assessment. This finding was determined to be of very low risk significance. This performance deficiency was also determined to be subject to traditional enforcement because it impeded the regulatory process, in that the failure to submit a license amendment and add required surveillance testing was in violation of 10 CFR 50.59(c)(1)(i) and caused the NRC-approved Technical Specifications to be out of alignment with the safety analysis for the facility. This violation is associated with a finding that has been evaluated by the SDP and communicated with an SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider the regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the safety significance of the associated finding. This violation was determined to be a Severity Level IV violation, because it is consistent with the examples in Paragraph 6.1.d of the NRC Enforcement Policy. The finding had a cross-cutting aspect in the training aspect of the human performance cross-cutting area because the licensees staff failed to understand and misapplied NRC generic guidance related to discovery of inadequate Technical Specifications
05000285/FIN-2014002-032014Q1Fort CalhounFailure to Maintain Design Control of HPSI Injection ValveTwo examples of a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, were identified. The first example involved the failure to establish procedures or Technical Specifications to accomplish required HPSI injection flow balancing. The second example involved the failure to provide controls or testing to ensure that replacement parts for HPSI injection valves were suitable for the application and were capable of supporting the safety-related functions of the HPSI system. The licensee has since implemented Engineering Change 59874 which included throttling of the HPSI loop injection valves. This change was completed on August 20, 2013, restoring the original plant design and overcoming the configuration control errors introduced on three of the eight injection valves. Post-work testing for the completed modification included flow balance testing for the HPSI loop injection lines. The inspectors reviewed the results of this testing and determined that the UFSAR assumptions regarding balanced loop flows were adequately addressed by licensee corrective actions. This finding was more than minor because it adversely impacted the design control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors reviewed NRC IMC 0609, Attachment 4, Initial Characterization of Findings, Table 3 SDP Appendix Router. While this issue was identified during a refueling outage, the inspectors determined that the majority of the exposure time for this violation occurred with the reactor at power. As such, the inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, The SDP for Findings at-Power, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors answered yes to the question of Does the finding represent a loss of system and/or function? The inspectors determined the finding required a detailed risk evaluation per IMC 0609 Paragraph 6.0, because the operability of the high pressure safety injection system (both trains) was in question. A Region IV senior reactor analyst performed a detailed risk evaluation and determined the flow imbalance did not result in a loss of safety function. Since the high pressure safety injection system was capable of meeting the functional success criteria, there was no quantifiable change to the core damage frequency and therefore was determined to be of very low safety significance (Green). The inspectors determined there was no cross-cutting aspect associated with this finding because events related to identification of needed procedures and specifications occurred in the 1970s and are not indicative of current performance. Additionally, the errant replacement of parts of three HPSI injection valves occurred between 1993 and 2006, and are also not indicative of current performance.
05000285/FIN-2014002-022014Q1Fort CalhounFailure to Translate HPSI Pump Design Requirements to Design Documents (Section 4OA3.2)A non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, was identified involving the failure to translate the High Pressure Safety Injection (HPSI) pump design and runout characteristics to design documents such as the Updated Safety Analysis Report or design calculations. On June 21, 2013, the licensee completed Engineering Change 59874, which permanently installed flow-limiting orifices in the discharge line of each pump, effectively preventing HPSI runout conditions from occurring for all plant conditions. This finding was more than minor because it adversely impacted the design control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors reviewed NRC IMC 0609, Attachment 4, Initial Characterization of Findings, Table 3 SDP Appendix Router. While this issue was identified during a refueling outage, the inspectors determined that the majority of the exposure time for this violation occurred with the reactor at power. As such, the inspectors determined the finding should be evaluated using the SDP in accordance with IMC 0609, The Significance Determination Process (SDP) for Findings at-Power, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The finding required a detailed risk evaluation because the high pressure safety injection system was inoperable for some of the large break loss of coolant accident scenarios (at reactor pressures less than 100 psi). A Region IV senior reactor analyst performed a bounding detailed risk evaluation. The change to the core damage frequency was 8E-8/year and, therefore, determined to be of very low safety significance (Green). The dominant core damage sequences included loss of coolant accidents where the high and low pressure safety injection systems failed during recirculation. The non-degraded low pressure safety injection system contributed to minimize the risk. The inspectors determined there was no cross-cutting aspect associated with this finding because events related to identification of needed procedures and specifications occurred in the 1970s and are not indicative of current performance.
05000397/FIN-2013007-082013Q3ColumbiaFailure to Correct a Condition Adverse to Quality with Emergency Operating ProceduresThe team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, from August 16, 2013, to August 21, 2013, the licensee failed to implement a prompt compensatory corrective action to correct an adverse condition in emergency operating procedures that would have led to the loss of emergency core cooling pumps due to inadequate available net positive suction head. This violation was entered into the corrective action program as Action Request 292437. On August 21, 2013, the licensee implemented a night order giving guidance to monitor the pumps for cavitation and take actions to prevent degraded operation until the procedures are revised. The team determined that the failure to implement an interim compensatory corrective action to promptly correct an adverse condition in emergency operating procedures in accordance with 10 CFR 50, Appendix B, Criterion XVI was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating System 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. Specifically, the finding was more than minor because the procedures were in a condition that would adversely affect the licensees response to an emergency. Using the Manual Chapter 0609, Appendix A, Exhibit 2, the team determined the finding represented a loss of safety system function; therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the bounding change to the core damage frequency was less than 1.8E-8 per year (Green). Since the change in core damage frequency was less than 1E-7 per year, the finding was not significant to the larger early release frequency. The team determined that this finding has a crosscutting aspect in the area of problem identification and resolution, corrective action program, because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
05000397/FIN-2013007-012013Q3ColumbiaFailure to Support the Ability to Restore Class 1E Diesel Generator Standby Power and Recover from Station Blackout (SBO) ConditionsThe team identified a Green non-cited violation of 10 CFR 50.63, Loss of All Alternating Current Power, which states, in part, that each light-water-cooled nuclear power plant licensed to operate under this part must be able to withstand for a specified duration and recover from a station blackout as defined in 50.2. Specifically, from June 8, 2013, to August 6, 2013, the licensee failed to demonstrate the ability to restore alternating current power and recover from a station blackout event when the licensee determined that the station battery voltage would be below the vendor minimum rated voltage to operate the diesel generator output breaker close coil. This violation was entered into the licensees corrective action program as Action Request 291162. Subsequently, the licensee tested a spare 4160 Vac breaker, similar to the diesel generator output breaker, to provide reasonable assurance that the diesel generator breaker would close after the 4-hour coping period. The test results determined that the breaker would close reliably with less than the manufacturers rated voltage and within the capability of the battery. The team determined that the failure to demonstrate the ability to restore emergency alternating current power to recover from a station blackout in accordance with 10 CFR 50.63 was a performance deficiency. 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 adversely affected cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding represented a reasonable question of functionality of the use of emergency diesel generators to recover from a station blackout. Using the NRC Manual Chapter 0609, Appendix A, Exhibit 2, the team determined the finding was of very low safety significance (Green), because the finding was confirmed to be a qualification deficiency that did not affect the functionality of the emergency diesel generators. The team determined that this finding had a cross-cutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to implement a corrective action program with a low threshold for identifying issues.
05000397/FIN-2013007-072013Q3ColumbiaFailure to Include ECCS Pumps NPSH Limits in the Emergency Operating ProceduresThe team identified a Green non-cited violation of Technical Specification 5.4.1(b) which states, in part, Written procedures shall be established, implemented, and maintained covering the following activities: The emergency operating procedures required to implement the requirements of NUREG-0737 and NUREG- 0737, Supplement 1, as stated in Generic Letter 82-33. Specifically, from 1997 to August 21, 2013, the licensee failed to revise emergency operating procedures for reactor pressure vessel control and primary containment control when it was determined that the required net positive suction head for the emergency core cooling pump were no longer bounded by the pumps vortex limits. This violation was entered into the corrective action program as Action Request 292153. On August 21, 2013, the licensee implemented a night order giving guidance to monitor the pumps for cavitation and take actions to prevent degraded operation until the procedures were revised. The team determined that the failure to maintain emergency operating procedures which included appropriate net positive suction head limits in accordance with Technical Specification 5.4.1(b) was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating System 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. Specifically, the finding was more than minor because the procedures were in a condition that would adversely affect the licensees response to an emergency. Using the Manual Chapter 0609, Appendix A, Exhibit 2, the team determined the finding represented a loss of safety system function; therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the bounding change to the core damage frequency was less than 1.8E-8 per year (Green). Since the change in core damage frequency was less than 1E-7 per year, the finding was not significant to the larger early release frequency. This finding did not have a crosscutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
05000397/FIN-2013007-062013Q3ColumbiaFailure to Analyze the Effect of System, Test Source, and Transient Harmonics on Proper Operation of Undervoltage RelaysThe team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which states in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculation methods, or by the performance of a suitable testing program. Specifically, prior to August 22, 2013, the licensee failed to assess the cumulative effects of the 4160 Vac system, test source, and transient harmonics on the secondary level undervoltage relays. This violation was entered into the licensees corrective action program as Action Requests 291665 and 292405. The violation did not present an immediate safety concern. The licensees failure to analyze the cumulative effect of electrical system, test source, and transient harmonics on the secondary level undervoltage relays was a performance deficiency. The 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 adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to analyze the cumulative effect of electrical harmonics on the secondary level undervoltage relays would have the potential to cause the relays to fail to actuate at the setpoints specified in technical specifications. Using the Manual Chapter 0609, Appendix A, Exhibit 2, the team determined the finding is of very low safety significance (Green), because the finding was confirmed to be a qualification deficiency that did not affect the functionality of the undervoltage relays. This finding did not have a crosscutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
05000397/FIN-2013007-052013Q3ColumbiaFailure to Provide Technical Basis for Assuming Turbulent Mixing of Diesel Combustion AirThe team identified a Green non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, from June 6, 1992, to August 20, 2013, the licensee failed to verify or check the adequacy of the design, by the use of alternate or simplified calculational methods, the technical basis that justified the dispersion of nitrogen in a tornado event to prevent loss of function of the emergency diesel generators. This violation was entered into the licensees corrective action program as Action Request 292322. The violation did not represent an immediate safety concern. The team determined that the failure to verify the adequacy of the technical basis that justified the dispersion of nitrogen in a tornado event was a performance deficiency. The 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 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 functionality of the diesel generators was called into question for the failure to provide a technical basis for the effects of nitrogen leakage on the combustion air system. Using the Manual Chapter 0609, Appendix A, Exhibit 4, the inspectors determined a detailed risk evaluation was necessary because, during an external initiating event, the finding would degrade one or more trains of a system that supports a risk significant system or function; therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the change in core damage frequency was 1.2E-8 per year (Green). Since the change in core damage frequency was less than 1E-7 per year, the finding was not significant to the larger early release frequency. This finding did not have a crosscutting aspect because the most significant contributor to the performance deficiency did not reflect current licensee performance.
05000397/FIN-2013007-042013Q3ColumbiaFailure to Update the FSAR for the Cleaning and Inspection Frequency of the Diesel Fuel Oil Storage TanksThe team identified a Severity Level IV non-cited violation of 10 CFR 50.71(e), which states, in part, that each person licensed to operate a nuclear power reactor shall update, periodically, the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information. Specifically, from May 2009 to August 22, 2013, the team identified that the diesel engine fuel oil storage tanks cleaning and inspection frequency was not updated in the final safety analysis report to include the latest information developed. This violation was entered into the licensees corrective action program as Action Request 292360. The violation did not represent an immediate safety concern. The licensees failure to update the final safety analysis report to reflect the cleaning and inspection frequency of the diesel engine fuel oil storage tanks in Section 9.5.4.4 Testing and Inspection Requirements was a violation of the NRC requirements. The inspectors determined that this violation was also a performance deficiency. However, the inspectors determined that the performance deficiency was minor. The inspectors considered this issue to be within the traditional enforcement process because it had the potential to impact the NRC\\\'s ability to perform its regulatory oversight function. The inspectors used the NRC Enforcement Policy to evaluate the significance of this violation. The inspectors determined that the violation was a Severity Level IV because it was similar to an example provided in Section 6.1 of the NRC Enforcement Policy. The inspectors did not assign a cross-cutting aspect to this non-cited violation because there was no finding associated with this traditional enforcement violation.
05000397/FIN-2013007-032013Q3ColumbiaFailure to Establish Emergency Procedures for Filling and Venting Diesel Fuel Oil Tanks after Tornado DamageThe team identified a Green non-cited violation of Technical Specification 5.4.1(a), Procedures which requires, Written procedures shall be established, implemented, and maintained covering the following activities: (a) The applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. From April 4, 1984, to August 7, 2013, the licensee failed to establish procedures for filling and venting the emergency diesel generator fuel oil tanks after potential tornado damage. This violation was entered into the licensees corrective action program as Action Request 291543. Subsequently, the licensee implemented Night Order 1477 to provide interim procedural guidance to operators prior to developing a formal emergency procedure. The team determined that failure to establish procedures for filling and venting diesel engine fuel oil storage tanks after tornado damage in accordance with Technical Specification 5.4.1(a) was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the procedures attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using the NRC Manual Chapter 0609, Appendix A, Exhibit 4, the inspectors determined a detailed risk evaluation was necessary because, during an external initiating event, the finding would degrade one or more trains of a system that supports a risk significant system or function; therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined that the change in core damage frequency was 1.2E-8 per year (Green). Since the change in core damage frequency was less than 1E-7 per year, the finding was not significant to the larger early release frequency. This finding had a crosscutting aspect in the area of problem identification and resolution, corrective action program component, because the licensee failed to take appropriate corrective actions to address safety issues and adverse trends in a timely manner, commensurate with their safety significance and complexity.
05000397/FIN-2013007-022013Q3ColumbiaFailure to Demonstrate Independence Requirements of IEEE 308-1974 for Divisions 1 and 2 Vital Instrumentation and Control Power SystemsThe team identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which states in part, that design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to August 22, 2013, the licensee failed to verify by either an analysis or test that the Class 1E inverters would continue to operate reliably when subjected to the effects of electrical faults that could be postulated to occur at non-Class 1E loads, due to a lack of seismic qualification of the loads, during and after a design basis loss-of-offsite power (LOOP) and seismic event. This violation was entered into the corrective actions program as Action Requests 291144 and 291248. Once identified, the licensee performed preliminary short circuit and coordination calculations during the inspection to provide reasonable assurance that the Class 1E fuses in the distribution to the non-Class 1E loads would operate within the first cycle of fault current. The team determined that the failure to demonstrate conformance to the independence requirements of IEEE 308-1974 was a performance deficiency. 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 adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the finding resulted in a condition where there was a reasonable doubt on the operability of the system. Using the NRC Manual Chapter 0609, Appendix A, Exhibit 4, the team determined a detailed risk evaluation was necessary because the finding involved the total loss of safety function that contributed to an external event initiated core damage accident sequence. Therefore, the senior reactor analyst performed a bounding detailed risk evaluation. The analyst determined, qualitatively, that the change to the core damage frequency would be less than 1E-7 per year (Green). Since the change to the core damage frequency was less than 1E-7 per year, the finding was not significant to the larger early release frequency. This finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000285/FIN-2012011-012012Q4Fort CalhounInadequate operability determination for containment internal structuresThe NRC identified a non-cited violation of Title 10 CFR Part 50, Appendix B, Criterion V, Procedures, for the failure to perform an adequate operability determination as required by FCS Procedure NOD-QP-31, Operability Determination Process. Specifically, the licensees operability determination for non-conforming containment internal structures failed to address that a section of the containment internal structures exceeded the allowable working stress criteria. The licensee entered this issue into its corrective action program for evaluation and review. Inspectors found that the failure to perform an adequate operability determination to specifically evaluate that the containment internal structures did not meet the design code of record was a performance deficiency. This violation is more than minor because it is associated with the design control attribute of the barrier integrity cornerstone and has the potential to adversely affect the cornerstone objective. The inspectors used Inspection Manual Chapter 0609, Appendix G Shutdown Operations Significance Determination Process , to determine that the issue screened as very low safety significance (green) because it did not require a quantitative assessment per Checklist 4. This violation was determined to have a crosscutting aspect in the area of human performance associated with decision making (H.1.b). Specifically, the licensee did not use conservative assumptions in decision making and did not adopt a requirement to demonstrate that the proposed action is safe in order to proceed rather than a requirement to demonstrate that it is unsafe in order to disapprove the action.
05000285/FIN-2012011-022012Q4Fort CalhounFailure to Follow Radiation Work Permit RequirementsInspectors reviewed a self-revealing Green noncited violation of Technical Specification 5.8.1.a for the failure to follow procedure requirements related to radiation work permit requirements. Specifically, workers unexpectedly created a high radiation area when working with tri nuke filter hosing while on a radiation work permit that did not allow access into a high radiation area. Both workers received alarms on their dosimeters. The licensee entered the issue into its corrective action program for evaluation and review. The failure to follow a procedure was a performance deficiency. The finding was more than minor because it negatively impacted the Occupational Radiation Safety cornerstones attribute of program and process, in that not following the requirements of the radiation work permit led to workers unplanned, unintended dose. Using NRC Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance because: (1) it was not associated with as low as is reasonably achievable (ALARA) planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding has a problem identification and resolution crosscutting component associated with operating experience because the licensee didnt implement operating experience through changes to station procedures. Specifically, there was operating experience which could have prevented the issue if it had been discussed at the pre-job brief.
05000285/FIN-2012011-032012Q4Fort CalhounFailure to Ensure that Adequate Equipment was Available to Measure River Level Locally to be Able to Comply with an Abnormal Operating ProcedureThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that adequate equipment was available to measure river level locally to comply with an abnormal operating procedure. Specifically, the length of the weighted tape measure used to measure river level locally was inadequate to ensure that the entire range of river levels needed for operation of the plant would be covered. The licensee entered the issue into its corrective action program for evaluation and review. The performance deficiency was determined to be more than minor because it is associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and it adversely affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was screened as very low safety significance (Green) because the licensee maintained an adequate mitigation capability and it would not be characterized as a loss of control. The inspectors determined the finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee falied to thoroughly evaluate problems such that resolutions address the causes and extent of condition specifically associated with deficiencies involving the Acts of Nature procedural guidance.
05000285/FIN-2012011-042012Q4Fort CalhounInadequate Design Basis DocumentationThe NRC identified a non-cited violation of 10 CFR 50 Appendix B, Criterion V, Procedures, for failing to follow a quality procedure. Specifically; PED-QP-13 Design Basis Document Control, requires FCS to update and maintain their Design Bases Documents. The license has failed to maintain these design documents. Some examples include PLDBD-51 Seismic Criteria where the configuration of the Steam Generator supports were not accurately described, and PLDBD-ME-10 Pipe Stress and Supports where the piping design code classification for Main Steam is incorrect. The licensee entered the issue into its corrective action program for evaluation and review. The performance deficiency is more than minor because if left uncorrected it would have the potential to lead to a more significant safety concern. The finding was determined to affect the Initiating Events, Mitigation Systems, and Barrier Cornerstones using Inspection Manual Chapter 0609.04, Initial Characterization of Findings. The finding was characterized as having very low safety significance (i.e., Green) using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, because all logic questions for the applicable cornerstones were answered in the negative. The finding is assigned a cross-cutting aspect in the area of Human Performance, in the component of Resources because the licensee failed to ensure that personnel, equipment, procedures, and other resources, specifically those necessary for complete, accurate and up-to-date design documentation, were available and adequate to assure nuclear safety.
05000285/FIN-2012011-052012Q4Fort CalhounFailure to Properly Scope All the Pertinent External Flood Protection Features into the Walkdown List in Accordance with Industry Guidance NEI 12-07The inspectors identified a finding of very low safety significance (Green) for the licensees failure to generate a complete inspection list, with all the external flood protection features credited in the current licensing basis documents for flooding events, to comply with NRC endorsed NEI 12-07, Guidelines for Performing Walkdowns of Plant Flood Protection Features. These walkdowns were being performed in response to a March 12, 2012, letter from the NRC to licensees, entitled, Request for Information Pursuant to Title 10 of the Code of Federal Regulations 50.54(f) Regarding Recommendations 2.1, 2.3, and 9.3, of the Near-Term Task Force Review of Insights from the Fukushima Dai-Ichi Accident. Specifically, the scoping list did not include several active components, which are an essential part of Fort Calhouns design basis flood mitigation strategy. The licensee entered the issue into the corrective action program and revised the scoping list accordingly. The performance deficiency was determined to be more than minor because it is associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factors (Flood Hazard) and it adversely affects the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, in addition to not scoping the sluice gates into the Flooding Features Walkdown List, fourteen additional active components would not have been scoped into the walkdown list. This would have prevented the licensee from identifying that preventive maintenance tasks needed to be created, and some active components that are an essential part of the flood mitigating strategy would not have been inspected and tested. The finding was screened as very low safety significance (Green) because the finding did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The inspectors determined the finding had a cross-cutting aspect in the area of human performance because licensee personnel did not properly apply human error prevention techniques such as peer checking and proper documentation of activities.
05000285/FIN-2012005-012012Q3Fort CalhounFailure to Update the Safety Analysis Report Solid WasteThe inspectors identified a cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, for the failure to update the Updated Safety Analysis Report with a detailed description of the Original Steam Generator Storage Facility. Specifically, since December 2006, the licensee stored a significant source of radioactivity in the Original Steam Generator Storage Facility, but failed to describe the volume of waste, the principal sources of radioactivity, the total quantity of radioactivity, and the estimated dose rate at the site boundary per curie of radioactivity in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2012-05725. This issue was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. This issue is being characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy. Cross-cutting aspects are not assigned to traditional enforcement violations. NOV summary: The inspectors identified a cited violation of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, for the failure to update the Updated Safety Analysis Report with a detailed description of the Original Steam Generator Storage Facility. Specifically, since December 2006, the licensee stored a significant source of radioactivity in the Original Steam Generator Storage Facility, but failed to describe the volume of waste, the principal sources of radioactivity, the total quantity of radioactivity, and the estimated dose rate at the site boundary per curie of radioactivity in the Updated Safety Analysis Report. The licensee has entered this violation into their corrective action program as Condition Report 2012-05725. This issue was evaluated using traditional enforcement because it has the potential to impact the NRCs ability to perform its regulatory function. This issue is being characterized as a Severity Level IV violation in accordance with Section 6.1.d.3 of the NRC Enforcement Policy. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000285/FIN-2012005-022012Q3Fort CalhounUntimely Corrective Actions for 480 VAC Breaker IssuesThe NRC identified a noncited violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to take timely corrective actions with respect to nonconforming conditions in several circuit breakers. These conditions were determined to have been the cause of the 1B4A bus bar failure that initiated a fire on June 7, 2011. These conditions were not corrected in a timely manner and the licensee continued to operate with a degraded breaker for nine months after the breaker tripped unexpectedly during the June 7, 2011, fire event. The licensee entered this issue into their corrective action program as CRs 2012-01884 and 2011-5414. The violation was determined to be more than minor because it affected the Initiating Events Cornerstone attribute of protection against external events (i.e., fire). The issue adversely affected the associated cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations because the condition that contributed to the fire event was left uncorrected. The finding screened to Green in accordance with IMC 0609, Appendix G because RCS makeup capability was not degraded. The inspectors determined that the issue had a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program.
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