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05000285/FIN-2016004-0331 December 2016 23:59:59Fort CalhounLicensee-identifiedLicensee-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, though the licensee identified a potential vulnerability to raw water pumps from a missile hazard striking diesel driven fire pump FP-1B or associated piping during review of missile hazards during the 2013 tornado missile project, the licensee failed to evaluate this condition or specify a modification to the plant to protect the raw water pumps at that time. This was discovered on August 25, 2016, by the licensee during a design review. This finding is of very low safety significance (Green) considering compensatory measures that were put in place to disable pump FP-1B and isolate associated piping when severe weather is forecast and the very low probability of the postulated event. This issue was entered into the licensees corrective action program as CR 2016-06972.
05000285/FIN-2016004-0431 December 2016 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTitle 10 of the Code of Federal Regulations, Part 50.9(a), requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on December 26, 2014, Fort Calhoun Station provided information to the Commission which was not complete and accurate in all material respects. Specifically, a license amendment request (ML14365A123) to adopt a scheme of emergency action levels based on Nuclear Energy Institute Document 99-02, Revision 6, contained inaccurate information about the characteristics of the cask used in the licensees Independent Spent Fuel Storage Installation and, as a result, incorrect external radiation levels were incorporated into emergency action level E-HU1. Subsequently, while preparing another emergency action level submittal, the emergency preparedness staff discovered the incorrect information that had previously been submitted. The issue was determined to be a Severity Level IV violation of NRC requirements, in accordance with Section 6.9 of the Enforcement Policy, dated November 1, 2016, because the inaccurate information would not have caused the NRC to reconsider a regulatory position or undertake substantial further inquiry. The issue was documented in the licensees corrective action program as Condition Report CR-2016-08400. Because the Severity Level IV violation has been entered into the licensees corrective action program, this violation is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000285/FIN-2016004-0131 December 2016 23:59:59Fort CalhounSelf-revealingFailure to Provide Training on Changes to Protective Action Recommendation ProceduresThe inspector reviewed a self-revealed non-cited violation associated with Fort Calhoun Stations failure to provide radiological emergency response training to those who may be called upon to assist in an emergency, as required by 10 CFR 50.47(b)(15). Specifically, in December 2014, 10 shift managers and 6 Technical Support Center and Emergency Operations Facility staff, responsible for making and reviewing protective action recommendations, were not trained on Procedure EPIP-EOF-7, Protective Action Recommendations, Revision 26, and flowchart EP-FC-111-AD-F-02, before they were implemented on December 23, 2014. As immediate corrective actions, the licensee issued a reading package covering the new protective action recommendation process to the 16 individuals who had not been trained. The issue was entered into the licensees corrective action program as Condition Report CR-2015-08951. The failure to provide radiological emergency response training to those who may be called upon to assist in an emergency is a performance deficiency within the licensees ability to foresee and correct. The performance deficiency was more than minor because it was associated with the procedure quality attribute of Emergency Prepardness Cornerstone and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was evaluated using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated September 22, 2015, and was determined to be of very low safety significance (Green) because it was a failure to comply, was not a risk significant planning standard function, was not a loss of the planning standard function, and was a degraded planning standard function. This finding had a cross-cutting aspect in the area of human performance associated with change management because the emergency preparedness department failed to identify all of the emergency response organization staff who required training on revisions to the process for making protective action recommendations (H.3).
05000285/FIN-2016004-0231 December 2016 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTechnical Specification 2.0.1 requires the unit to be shut down within 6 hours in the event a limiting condition for operation and/or associated action requirement cannot be satisfied because of circumstances in excess of those addressed in the specification. Contrary to the above, the licensee failed to enter Technical Specification 2.0.1 and take the prescribed actions on several occasions when shutdown cooling heat exchanger valves were opened which impacted component cooling water (CCW) flow to the containment air cooling units under certain accident conditions. On May 10, 2016, an unanalyzed condition was discovered during scheduled maintenance on the shutdown cooling heat exchanger valves. As part of the maintenance, HCV-484, Shutdown Heat Exchanger AC-4A Component Cooling Water Outlet Valve, and HCV-481, Shutdown Cooling Heat Exchanger AC-4B Component Cooling Water Inlet Valve, were failed open which rendered both valves inoperable. Under these conditions, with the assumed single failure loss of DC control power during a loss of coolant accident (LOCA), CCW would be allowed to flow through both shutdown cooling heat exchangers, effectively reducing CCW system flow to the containment air cooling units. These conditions are not assumed under plant design basis calculations and placed the plant in an unanalyzed condition. It has not been demonstrated that the CCW system would adequately perform its design function of providing a cooling medium for the containment atmosphere under LOCA conditions with CCW flow diverted through the shutdown cooling heat exchangers. With two containment air cooling units inoperable, Technical Specification 2.4, does not provide an associated action; therefore, Technical Specification 2.0.1 applies. Upon completion of the maintenance activity, both valves were returned to service which eliminated the condition. The licensee conducted an extent of condition review and identified that they had created this unanalyzed condition six times within the last 3 years and had exceeded the Technical Specification 2.0.1 6-hour shutdown action statement on March 8, 2016; April 21, 2016; and May 10, 2016. In addition, the licensee determined this condition was first identified on February 3, 2015, in Condition Report 2015-01388. Procedure TDB-VIII, Equipment Applicability Guidance, Revision 64, incorrectly stated the valves had a required safety function in the open direction. The licensee initiated procedure change EC-68088 on September 26, 2015, to correct the procedure; however, the proposed change did not accurately reflect the safety function of the valves to remain closed for all LOCA conditions. This procedure change was still under review on May 10, 2016. The failure to promptly correct Procedure TDB-VIII was a contributing cause of the violation. The violation is more than minor because it is associated with the configuration control attribute of the Mitigating Systems Cornerstone. On March 8, 2016; April 21, 2016; and May 10, 2016, the plant was placed in a condition prohibited by technical specifications and exceeded the Technical Specification 2.0.1, 6-hour shutdown action statement. This adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. A senior reactor analyst qualitatively determined that this finding was of very low safety significance (Green) for increases in core damage frequency and large early release frequency because of the short exposure time of less than 3 days and because of the low frequency of events where a LOCA with an independent and coincidental loss of DC control power would occur. Therefore, this finding screens to Green. The licensee entered the issue into their corrective action program as Condition Reports 2016-05340 and 2016-04468.
05000285/FIN-2016008-0130 September 2016 23:59:59Fort CalhounNRC identifiedFailure to Obtain Prior NRC Approval for a Change When RequiredThe inspectors identified a Severity Level IV, Green, non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section (c)(1), which states, in part, that a licensee may make changes in the facility as described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to 10 CFR 50.90 only if: (i) a change to the technical specifications incorporated in the license is not required, and (ii) the change, test, or experiment does not meet any of the criteria in paragraph (c)(2). Title 10 CFR 50.59, Section (c)(2), states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report (as updated). Specifically, from June 9, 2015 through August 11, 2016, the licensee implemented a change to Operating Instruction OI-VA-2, Auxiliary Building Ventilation System Normal Operation, Attachment 11, Revision 47, after incorrectly concluding that the opening of certain high-energy line break barriers and selected fire barrier doors to allow supplemental cooling of both safety-related switchgear rooms did not increase the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report. In response to this issue, the licensee revised entry conditions to Operating Instruction OI-VA-2, Attachment II, to ensure that high energy line break barriers are not impaired prematurely. This finding was entered into the licensees corrective action program as Condition Report CR-2016-06667. The licensees failure to implement the requirements of 10 CFR 50.59 and adequately evaluate the disabling of certain high-energy line break barriers to facilitate supplemental cooling of both safety-related switchgear rooms was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the inspectors used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, dated June 19, 2012, to determine that this performance deficiency 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-ofservice for longer than its 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 significance in accordance with the licensees maintenance rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy, the inspectors characterized this performance deficiency as a Severity Level IV violation. As described in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, no cross-cutting aspect was assigned to this violation because traditional enforcement violations are not assessed for cross-cutting aspects.
05000285/FIN-2016405-0130 June 2016 23:59:59Fort CalhounNRC identifiedSecurity
05000285/FIN-2016405-0230 June 2016 23:59:59Fort CalhounNRC identifiedSecurity
05000285/FIN-2016405-0330 June 2016 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified Violation
05000285/FIN-2016201-0130 June 2016 23:59:59Fort CalhounNRC identifiedSecurity
05000285/FIN-2016403-0130 June 2016 23:59:59Fort CalhounNRC identifiedSecurity
05000285/FIN-2016002-0230 June 2016 23:59:59Fort CalhounSelf-revealingFailure to Develop Adequate Procedures for Post Modification TestingThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 5.8.1., for failure to establish, implement, and maintain a procedure recommended in Regulatory Guide 1.33, Revision 2, Appendix A. Specifically, the licensee failed to develop adequate procedures for testing equipment important to safety. The licensee failed to identify and mitigate all possible turbine logic trip signals when testing Distributed Control System logic. Following the logic modification of the Turbine Control System, post modification testing inserted two Emergency Trip System test signals which caused an automatic turbine trip resulting in an automatic reactor protective system scram actuation. Failure to establish, implement, and maintain procedures as required by technical specifications is a performance deficiency. The performance deficiency is more than minor because it adversely affected the procedure quality attribute of the initiating event cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to mitigate all possible turbine trip signals while testing the Distributed Control System which caused the turbine to trip and thereby caused a loss of load reactor trip. Using NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, the finding screened as having very low safety significance (Green) because although the deficiency resulted in a reactor trip, the trip was uncomplicated and mitigating equipment remained unaffected. This finding has a cross-cutting aspect in the teamwork component of the human performance cross-cutting area because the licensee did not ensure that individuals and work groups communicate across organizational boundaries to ensure nuclear safety is maintained. Specifically, the Distributed Control System expert did not review the post modification testing procedure prior to implementation (H.4).
05000285/FIN-2016201-0230 June 2016 23:59:59Fort CalhounNRC identifiedSecurity
05000285/FIN-2016002-0130 June 2016 23:59:59Fort CalhounNRC identifiedFailure to Perform an Adequate Evaluation of Service Life for Component Cooling Water Pump MotorsThe inspectors identified a Green, non-cited violation of 10 CFR 50, Appendix B, Criterion III, design control, associated with the licensees failure to perform an adequate evaluation of the service life of component cooling water pump motors. Specifically, the licensee operated component cooling water pump motors beyond the vendor recommended horsepower, temperature and voltage limits for the pumps which resulted in the potential for early winding failure of the motors. The licensees existing calculation determined a component cooling water pump motor life of 16.9 years. During the inspection, the licensee re-evaluated component cooling water pump motor life and determined the expected motor life was actually between 6.8 (if degraded voltage is considered) and 7.2 years. Actual in-service life of the longest operating component cooling water pump was approximately 6.13 years. The licensee entered this issue into the corrective action program as Condition Report 2016-04319. The inspectors determined that the failure to adequately evaluate the service life of the component cooling water pump motors is a performance deficiency. This performance deficiency was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. The inspectors determined the finding was of very low safety significance in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, because although the finding was a deficiency affecting the design of a mitigating system, it did not result in a loss of operability or functionality. Specifically, although there was a significant reduction in the calculated service life of the component cooling water pump motors, the actual in-service life of the longest operating component cooling water pump was approximately 6.13 years, which is still encompassed by the revised service life calculation. The finding does not have a cross-cutting aspect because the failure to perform an adequate service life evaluation for component cooling water pump motors is not indicative of current licensee performance. The licensees current design process requires reviews of in-service temperature effects on equipment service life including pump motors.
05000285/FIN-2016001-0131 March 2016 23:59:59Fort CalhounNRC identifiedImplementing a Procedure Change for Alternative Shutdown Cooling that would have Required NRC ApprovalThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, for the failure to recognize that a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the 10 CFR 50.59 evaluation revised a site procedure, without NRC approval, to substitute automatic flow control of shutdown cooling flow and temperature with manual control using the low pressure safety injection loop injection valves. The licensees corrective actions included revising the affected procedure to reflect the original automatic flow control. The licensee entered this issue in the corrective action program as Condition Report 2013-15342. The licensees failure to implement the requirements of 10 CFR 50.59 and adequately evaluate changes to determine if prior NRC approval is required was a performance deficiency. Because this violation had the potential to impact the NRCs ability to perform its regulatory function, the inspectors evaluated the violation 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 inspectors performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 1, 2012. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; 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, the inspectors characterized this performance deficiency as a Severity Level IV violation. The inspectors 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-2016001-0231 March 2016 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTechnical Specification 2.6(1) requires containment integrity to be maintained unless the reactor is in a cold or refueling shutdown condition. If containment integrity is not maintained and the reactor does not meet these cold or refueling shutdown conditions, then containment integrity must be restored within one hour or the reactor is required to be in hot shutdown within the next six hours. From November 22, 2013, through June 27, 2014, a test connection cap was left off of a containment penetration which constituted a loss of containment integrity. Upon discovery of this condition on June 27, 2014, the licensee entered Technical Specification 2.6(1) and Abnormal Operating Procedure 12 for loss of containment integrity. The cap was re-installed and containment integrity was restored within one hour. The violation is more than minor because it is associated with the configuration control attribute of the Barrier Integrity Cornerstone. Failure to install the containment penetration cap following local leak rate testing on November 22, 2013, resulted in a loss of containment integrity until it was discovered missing on June 27, 2014. This adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (i.e., containment) protect the public from radionuclide releases caused by accidents or events. The violation was reviewed by a Senior Reactor Analyst and was determined to be of very low safety significance because the test connection fitting was a 14-inch diameter opening. Inspection Manual Chapter 0609, Significance Determination Process, Appendix H, identifies that small lines (less than 1 to 2 inches in diameter) would not generally contribute to large early release frequency. Therefore, this finding screens to Green. The licensee entered the issue into their corrective action program as Condition Report 2014-07958.
05000285/FIN-2016007-0131 March 2016 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTechnical Specification 5.8.1 requires in part, that procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A of Regulatory Guide 1.33, Revision 2. That appendix states, in part, that maintenance that can affect the performance of safety-related equipment should be performed in accordance with written procedures appropriate to the circumstances. Contrary to the above, maintenance that can affect the performance of safety-related equipment was not performed in accordance with written procedures appropriate to the circumstances. Specifically, maintenance that can affect the performance of safetyrelated valves was not performed in accordance with a procedure that required the licensee to review all diagnostic test results for compliance with the setpoint criteria for all diagnostic tests performed. As described in CR-2012-01601-017, the licensee restored compliance by writing and implementing procedure ER-FC- 410-AD-SETPOINT, Air-Operated Valve Setpoint Control, Revision 0. This procedure requires, in part, that the licensee review all diagnostic test results for compliance with the setpoint criteria for all diagnostic tests performed. The licensees failure to complete maintenance that can affect the performance of safety-related valves in accordance with written procedures appropriate to the circumstances was a performance deficiency that is more-than-minor because it adversely affected the Procedure Quality attribute of the Mitigating Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, this performance deficiency resulted in valve HCV-2987, High Pressure Safety Injection Alternate Header Isolation, being not able to fulfill its design safety function from February, 2013, through July, 2013. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that the finding should be processed through Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 1, 2012. Using Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was not a design or qualification deficiency but represented a loss of train function for greater than the outage time allowed by Technical Specifications. Therefore, a Region IV senior reactor analyst performed a detailed risk evaluation in accordance with Manual Chapter 0609, Appendix A, Section 6.0, Detailed Risk Evaluation. The analyst determined that the condition of valve HCV-2987 inoperability would affect only the plants response to a large-break loss-of-coolant accident followed by the failure of the instrument air system. The analyst calculated the initiating-event frequency to be 2.63 x 10-10 /year. Also, the analyst determined that the finding did not affect external initiator risk and would not involve a significant increase in the risk of a large, early release of radiation. Therefore, this violation has very low (Green) safety significance.
05000285/FIN-2015009-0231 December 2015 23:59:59Fort CalhounSelf-revealingFailure to Revise Procedures and Perform Additional TrainingThe team evaluated a self-revealing NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, which states, in part, that Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies... are promptly identified and corrected. Specifically, prior to September 30, 2015, the licensee failed to revise procedures, and perform additional operator training, to prevent the inadvertent opening of steam bypass and steam dump valves during plant startup, and any subsequent plant impacts. In response to this issue, the licensee initiated a condition report to document these corrective actions. This finding was entered into the licensees corrective action program as Condition Report CR-FCS-2015-13718. The team determined that the failure to take timely corrective actions to revise procedures and complete additional training to correct a condition adverse to quality, was a performance deficiency. This finding was more than minor because it was associated with the initiating events cornerstone objective of configuration control to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the licensee failed to take recommended corrective actions to revise procedures and perform additional operator training to ensure proper alignment of the steam dump and bypass valves controller during startup. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 1, Initiating Events Screening Questions, the team determined that the finding was determined to have very low safety significance (Green) since the transient did not result in a reactor trip or loss of mitigation equipment. The finding has a problem identification and resolution cross-cutting aspect in the area of Operating Experience, because the licensee failed to systematically and effectively collect, evaluate, and implement relevant internal operating experience in a timely manner (P.5).
05000285/FIN-2015004-0331 December 2015 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTechnical Specification (TS) 2.5(1) requires two trains of auxiliary feedwater (AFW) to be operable when cold leg temperature is above 300F. In the event that both trains become inoperable, immediate action is required to restore one AFW train to operable status. Technical Specification 2.0.1 and all TS actions requiring mode changes are suspended until one AFW train is restored to operable status. Operation with the main and auxiliary feedwater cross-tied was a violation of the technical specification requirements to maintain operability of AFW systems. The violation is more than minor because it is associated with the configuration control attribute of the mitigating systems cornerstone because the failure to prevent cross-tying these systems resulted in unrecognized inoperability of both trains of AFW. This adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The violation was of very low safety significance because although MFW and AFW were momentarily cross-tied, this condition existed for only a brief period of time as operators restored system line-ups following system testing. In addition, a Senior Reactor Analyst evaluated the postulated main feedwater line break frequency and exposure time of the condition and determined the likelihood of this event during the exposure time is less than the Green/White threshold and of very low safety significance. The licensee entered the issue into their corrective action program as Condition Report 2015 03698.
05000285/FIN-2015012-0131 December 2015 23:59:59Fort CalhounNRC identifiedFailure to Adequately Implement and Maintain Required NFPA 805 Implementation ItemsThe inspectors identified two examples of a non-cited violation of License Condition 3.D, Fire Protection Program, for the failure to adequately implement required National Fire Protection Association Standard 805 implementation items in accordance with the approved fire protection program. Specifically, the licensee did not implement two items listed in Table S-3, Implementation Items, of Omaha Public Power District letter LIC-14-0042 by June 15, 2015. There was no immediate safety concern with either example and the licensee entered this violation into the corrective action program as Condition Reports 2015-2620 and 2015-2683. The failure to implement a requirement of a license condition within the allowed implementation period was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute 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. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and determined that the issue was of very low safety significance (Green). These findings had a cross-cutting aspect associated with change management within the human performance area since the leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the inspectors determined that the licensee did not have a process in place to ensure system level design basis documents were updated within the period required by a license condition and to assure plant-specific requirements were incorporated into the appropriate procedures (H.3).
05000285/FIN-2015012-0231 December 2015 23:59:59Fort CalhounNRC identifiedFailure to Provide Adequate Isolation for Pressurizer HeatersThe inspectors identified a non-cited violation of License Condition 3.D, Fire Protection Program, for the failure to ensure one success path necessary to achieve and maintain the nuclear safety performance criteria was maintained free of fire damage for all single fires. Specifically, the licensee failed to provide adequate isolation for the pressurizer heaters credited for achieving safe and stable plant conditions for fires that require shutdown from outside the control room. The licensee entered this issue into their corrective action program as Condition Report 2015-12195 and added this issue to their compensatory measures for the control room and cable spreading room. The failure to provide adequate isolation for equipment relied upon to achieve safe and stable plant conditions for a shutdown from outside of the control room was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute 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. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process. Because the finding affected the ability to reach and maintain safe shutdown conditions in case of a fire requiring evacuation of the control room, a senior reactor analyst performed a Phase 3 evaluation and determined that the issue was of very low safety significance (Green). This finding did not have a cross-cutting aspect since it was not indicative of present performance in that the performance deficiency occurred more than three years ago.
05000285/FIN-2015012-0331 December 2015 23:59:59Fort CalhounNRC identifiedFailure to Set Action Levels to Ensure that the Assumptions in the Engineering Analysis Remain ValidThe inspectors identified a non-cited violation of License Condition 3.D, Fire Protection Program, for the failure to establish an appropriate monitoring program in accordance with National Fire Protection Association Standard 805, Section 2.6. Specifically, the licensee failed to set the action level for the availability of the raw water system pumps to ensure that the assumptions in the engineering analysis remained valid. There was no immediate safety concern since the raw water pumps availability remained above the value assumed in the analysis and the licensee entered this violation into the corrective action program as Condition Report 2015-12612. The failure to set the action level for the availability of the raw water system pumps to ensure that the assumptions in the engineering analysis remained valid was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external events (fire) attribute 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. The inspectors evaluated this finding using Inspection Manual Chapter 0609, Appendix F, Fire Protection Significance Determination Process, and determined that the issue was of very low safety significance (Green). This finding had a cross-cutting aspect associated with change management within the human performance area since the leaders failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the inspectors determined that the licensee did not use the process that was in place to ensure that the appropriate fire risk assessment monitoring action levels were incorporated into the maintenance rule program and monitored (H.3).
05000285/FIN-2015009-0131 December 2015 23:59:59Fort CalhounNRC identifiedFailure to Take Adequate Corrective Action to Preclude Repetition of a Significant Condition Adverse to Quality Associated with Emergency Diesel Generator Room Water IntrusionsThe team identified an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the licensees failure to take corrective actions to prevent repetition of a significant condition adverse to quality. Specifically, since February 2009, the licensee failed to take corrective actions to prevent repetitive water intrusions from the Auxiliary Building HVAC room (Room 82) into the number one Emergency Diesel Generator room (Room 63). The inspectors determined that the licensees failure to implement corrective actions to preclude repetitive water intrusions into Room 63 was a performance deficiency. The performance deficiency was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone. Specifically, water intrusion events from Room 82 into Room 63 could challenge the reliability of the emergency diesel generator when relied upon during a loss of offsite power. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Question, inspectors determined that the finding was of very low safety significance (Green). The finding has a problem identification and resolution cross-cutting aspect within the area of Resolution, because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance (P.3).
05000285/FIN-2015004-0131 December 2015 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 72.174 requires that each licensee maintain sufficient records to furnish evidence of activities affecting quality. Records pertaining to the design, fabrication, erection, testing, maintenance, and use of structures, systems, and components important to safety must be maintained by or under the control of the licensee until the NRC terminates the license. Contrary to the above, as of June 21, 2013, Fort Calhoun failed to maintain sufficient records to furnish evidence of activities affecting quality. Specifically, the licensee did not maintain records for loading activities associated with DFS-HSM-06 that was placed on the ISFSI pad in July of 2009. This violation was identified by FCS and placed in their corrective action program (CR 2013-12884). The fuel assembly data was reconstituted based on records from the Reactor Engineering group, and the canister helium leak-test data was reconstituted based on the helium leak-test technician's field notes. The remaining canister records associated with the canister processing, sealing, and transportation to the ISFSI, including several TS requirements, were not found. Fort Calhoun reconstituted the fuel and helium leak test data and conducted interviews with cask loading personnel to conclude that there was no evidence to suggest that loading activities did not comply with the licensee's procedures and the licensed Technical Specifications. This violation did not have any safety impact because all fuel assemblies met the requirements for burn-up, decay heat, and cooling time and the licensee demonstrated that the canister integrity was intact based on the reconstituted helium leak test records. All the fuel inside the canister and the cask remain in a safe condition. This finding was reviewed by NRC Headquarters Division of Spent Fuel Managements Spent Fuel Licensing Branch. Based on the reconstituted records and interviews with the dry fuel loading staff, the NRC found no evidence to demonstrate that the canister did not meet the required license conditions and as such, found the canister acceptable for continued storage under FCSs general Part 72 license. However, though the canister is acceptable for storage, the licensee must track this issue to identify that further analyses may be required for this canister to meet all applicable Part 71 requirements to be acceptable for transportation. In accordance with the NRC Enforcement Policy Section 2.2 and IMC 0612 Section 03.23, Part 72 ISFSI inspection findings follow the traditional enforcement process and are not dispositioned through the Reactor Oversight Process or the Significance Determination Process. The violation screened as having very low safety significance, Severity Level IV, and is being treated as a non-cited violation, consistent with Section 2.3.2.a. of the Enforcement Policy. The violation was determined to be more than minor since the licensee failed to establish, maintain, or implement adequate controls over procurement, construction, examination, or testing processes that are important to safety. The violation was entered into the licensees corrective action program as CR 2013-12884. Following identification of the issue the licensee performed an assessment that showed the cask would continue to perform its design function. Corrective actions for this issue included performing an extent of condition review, performing an apparent cause analysis report, reconstitution of the missing documents, conducting interviews with the dry cask loading personnel, providing training to the staff involved, and changing processes and responsibilities within FCS Records Management Group.
05000285/FIN-2015004-0231 December 2015 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTechnical Specification (TS) 2.4(1)a.iv requires that all valves, piping, and interlocks associated with the components of the containment cooling system required to function during accident conditions be operable. In the event that any of these components, required to function during accident conditions become inoperable, the reactor shall be placed in a hot shutdown condition within 12 hours. The containment spray pumps and the associated piping are part of the containment cooling system. Prior to making modifications to containment spray piping in 2015, the operability of this piping would have been challenged by a main steam line break or a loss of coolant accident due to thermal stresses induced in the piping after a rise in containment temperature after the postulated event. Operation prior to the implementation of the modifications was a violation of the technical specification requirements to maintain operability of containment cooling systems. The violation is more than minor because it is associated with the design control attribute of the mitigating systems cornerstone because the failure to anticipate the rise in containment spray piping temperature dates back to the original design of the plant. This adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The violation was of very low safety significance because although the subject piping was inoperable due to exceeding code specified stress limits, analysis showed that the piping would have been able to perform its safety function to deliver adequate containment spray flow in the event of an accident. The licensee entered the issue into their corrective action program as Condition Report 2015-04578.
05000285/FIN-2015011-0230 September 2015 23:59:59Fort CalhounNRC identifiedFailure to Establish a Technical Basis for Operability of the Auxiliary Feedwater SystemThe 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 the operability determination procedure. Specifically, the licensee failed to establish a valid technical basis for operability of auxiliary feed containment isolation valves HCV-1107A and HCV-1108A. Following the valves failure on June 5, the licensee replaced the failed valve elastomers with new PTFE seals and nitrile O-rings. The licensee then performed an operability evaluation that considered the effect of high temperatures from a main steam line break on the valve elastomers. The inspectors found that the evaluation was not sufficient because it did not determine that the new O-rings would function under all potential temperature conditions and did not consider the function of the other valve components. The licensee entered these issues in their corrective action program as Condition Report CR-2015-08362 and revised their operability evaluation. The licensees failure to follow the operability determination procedure was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the licensee failed to sufficiently address the capability of the steam generator auxiliary feed containment isolation valves HCV-1107A and HCV-1108A to perform their safety function, requiring significant further analysis to demonstrate operability. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because although the finding was a deficiency affecting design or qualification, but the mitigating structure, system or component maintained its operability. The finding has a consistent process cross-cutting aspect in the human performance cross-cutting area since the organization did not use a consistent, systematic approach to make decisions and incorporate risk insights appropriately. Specifically, the licensee failed to re-evaluate the operability decision when new information on the conditions and susceptibility affecting valves HCV-1107A and HCV-1108A during normal operations was available (H.13).
05000285/FIN-2015011-0330 September 2015 23:59:59Fort CalhounNRC identifiedFailure to Correct a Non-Conforming Condition Associated with Auxiliary Feedwater ValvesThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct a condition adverse to quality. Specifically, the licensee failed to take corrective actions after identifying that the steam generator auxiliary feed containment isolation valves were not rated for the maximum temperature they would experience in service. The inspectors determined that on February 2, 2015, an NRC inspector questioned the licensee whether valves HCV-1107A and HCV-1108A were adequately designed for containment temperatures. The licensee determined that the design specification for the valves was 180F, and the containment temperature following a main steam line break was evaluated to be 374F. The fact that the valve was not designed for the most limiting conditions was a non-conforming condition of a safety related component, and was a condition adverse to quality. However, the licensee did not initiate a condition report to resolve and correct the condition. Additionally, the inspectors determined that in 2002, the licensee initiated Condition Report CR-2002-02124 after identifying elevated temperatures in the auxiliary feedwater piping. This condition report documented that the design specification for the two valves was 180F and had been exceeded in service. Although the condition report description recommended modifying the design of the valves, the licensee did not take actions to correct the condition. In both of these instances, the licensee recognized that the valve design temperature was not adequate for its application, but did not take action to resolve the discrepancy. The inspectors determined that although the inadequate design was a non-conforming condition, the valves were not inoperable until the licensee installed inappropriate elastomer material during the 2015 refueling outage as a result of inadequate design control. The licensee entered the failure to identify and correct the non-conforming design in their corrective action program as Condition Report CR-2015-08523. The licensees failure to take corrective action for a non-conforming condition was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the design control attribute of the Mitigating Systems Cornerstone and 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). Specifically, the licensee failed to take corrective actions to ensure an adequate design for the steam generator auxiliary feed containment isolation valves HCV-1107A and HCV-1108A. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because although the finding was a deficiency affecting the design or qualification of a mitigating system, structure, or component, the system, structure, or component maintained its operability. The finding has a basis for decisions cross-cutting aspect in the human performance cross-cutting area since leaders and individuals did not verify their understanding or question the basis of decisions. Specifically, the licensee failed to understand the potential significance of the non-conforming design of the valves and the basis for not taking corrective actions (H.10). (Section 4OA5.6)
05000285/FIN-2015003-0130 September 2015 23:59:59Fort CalhounSelf-revealingFailure to Maintain Safety Injection Tank Boron Concentration within Technical Specification LimitsA Green, self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI Corrective Action was identified because the licensee failed to identify and evaluate an adverse trend related to boron concentration in Safety Injection Tank (SIT) SI-6A and to take corrective actions to prevent boron concentration from going below the minimum concentration required by Technical Specifications. The licensees immediate corrective actions included documenting this condition in their corrective action program in Condition Report (CR) 2015-10181, declared SI-6A inoperable, and raised SI-6A boron concentration. The finding is more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone, in that this finding resulted in the SIT becoming inoperable when boron concentration fell below TS limits for approximately 8.5 days prior to August 20, 2015. Analysis conducted by a Senior Reactor Analyst determined the finding to be of very low safety significance (Green), primarily because the SIT function is needed only for mitigation of a postulated large-break loss of coolant accident, and the initiating-event frequency for such accidents is 2.5 x 10-6/year. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution and the Evaluation aspect, because the licensee did not thoroughly evaluate the issue and ensure that resolutions addressed causes and extent of conditions commensurate with their safety significance.
05000285/FIN-2015003-0230 September 2015 23:59:59Fort CalhounNRC identifiedFailure to Maintain Fire Watch and Fire Watch LogsInspectors identified a Green, Severity Level IV, non-cited violation of 10 CFR 50.9(a), Completeness and Accuracy of Information, for the licensees failure to maintain the required fire watch logs complete and accurate in all material respects. The licensee entered this into their corrective action program as Condition Reports (CR) 2014-06416 and 2014-06680. This finding is more than minor because it adversely affected the human performance 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. This finding has very low safety significance (Green) because it did not impact the ability to achieve safe shutdown. This findings severity level is based on an example in the Enforcement Policy, Section 6.1.d.2, which states, in part, that Severity Level IV violations involve violations of 10 CFR 50.59 (which) result in conditions evaluated as having very low safety significance.
05000285/FIN-2015011-0130 September 2015 23:59:59Fort CalhounSelf-revealingFailure to Ensure the Suitability of Replacement Materials during the Design Review ProcessThe inspectors reviewed a Green, self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the suitability of materials utilized during the design review process. Specifically, the licensee failed to identify during the design review process that replacement valve internal seal materials for the steam generator auxiliary feed containment isolation valves would not be suitable for high temperature conditions that the valves would experience in service, and as a result, caused both trains of the safety-related auxiliary feedwater system to become inoperable during hot standby conditions. The licensee entered this issue into their corrective action program as Condition Report CR-2015-07564 and replaced the valve internals with material that had been previously installed in valves HCV-1107A and HCV-1108A before the modification. The inspectors determined that the licensees failure to evaluate the suitability of the materials used during the design review process for the steam generator auxiliary feed containment isolation valves was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the licensees failure to properly evaluate the suitability of CTFE for use in the steam generator auxiliary feed containment isolation valves led to the failure of HCV-1107A and HCV-1108A and rendered both safety-related trains of auxiliary feedwater inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation since the finding represented a loss of system and/or function. A Region IV senior reactor analyst performed the detailed risk evaluation in accordance with Appendix A, Section 6.0, Detailed Risk Evaluation. The detailed risk evaluation result is a finding of very low safety significance (Green). The calculated change in core damage frequency of 2.3 x 10-7 was dominated by a loss of offsite power; common cause failure of the auxiliary feedwater discharge air-operated valves; failure of diesel-driven auxiliary feedwater pump FW-54; failure of the feed and bleed operation; and failure of operators to manually override a steam generator isolation signal and establish a flowpath for the main feedwater system. The analyst determined that the finding did not involve a significant impact to external initiators because of the short exposure time, or a significant increase in the risk of a large, early release of radiation. The finding has an operating experience cross-cutting aspect in the problem identification and resolution cross-cutting area since the organization did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, readily available internal operating experience on the high temperature conditions that valves HCV-1107A and HCV-1108A experienced during normal operations was not utilized during the design change process (P.5).
05000285/FIN-2015002-0730 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Perform Functionality Assessments for the Spent Fuel Pool Cooling SystemThe inspectors identified a finding associated with the failure of operations personnel to follow procedures used to perform functionality assessments. Specifically, operations personnel failed to provide sufficient technical justification for the reasonable assurance of functionality of the spent fuel pool cooling system when boric acid leaks were identified on discharge header vent valve AC-898. Vent valve AC-898 was replaced and the issue was entered into the licensees corrective action program as Condition Report 2015-05856. The failure of operations personnel to follow station procedures to perform functionality assessments was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not adversely affect decay heat removal capabilities from the spent fuel pool causing the pool temperature to exceed the maximum analyzed temperature limit specified in the site-specific licensing basis, did not result from fuel handling errors, dropped fuel assembly, dropped storage cask, or crane operations over the SFP that caused mechanical damage to fuel clad and a detectible release of radionuclides, did not result in a loss of spent fuel pool water inventory decreasing below the minimum analyzed level limit specified in the site-specific licensing basis, and did not affect the SFP neutron absorber, fuel bundle misplacement or soluble boron concentration. The inspectors determined that the finding had a training cross-cutting aspect 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. (H.9)
05000285/FIN-2015002-0130 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Include a Class 1 Component in the Reactor Vessel Pressure Boundary Integrity TestThe inspectors identified a non-cited violation of 10 CFR 50.55a(g)(4), involving the failure to adequately perform periodic reactor coolant system (RCS) integrity inspections as required by ASME Code Section XI. Specifically, Procedure OP-ST-RC-3007, Periodic Reactor Coolant System Integrity Test, required testing of all ASME Class 1 pressure boundary components of the reactor vessel pressure boundary but failed to include reactor vessel head vent line RC-2501R. As a result, the requirements of ASME Code Section XI were not met. This issue was entered into the licensees corrective action program as Condition Report 2015-05858. The inspectors concluded that the failure to include reactor vessel head vent line RC-2501R within the reactor vessel pressure boundary in the periodic RCS integrity inspection was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the procedure quality attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. The inspectors determined that the finding had a conservative bias cross-cutting aspect in the area of human performance because the licensee failed to use decision making-practices that emphasized prudent choices over those that are simply allowable. (H.14
05000285/FIN-2015002-0330 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Establish Adequate Work Instructions to Clean and Inspect the Reactor Vessel HeadThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to establish adequate work instructions to clean and inspect the reactor vessel head. Specifically, the work instructions for the required visual examination of the reactor vessel head failed to specify what constituted a relevant condition as defined by ASME Code Case N-729-1, Alternative Enclosure Examination Requirements for PWR Reactor Vessel Upper Head with Nozzles Having Pressure Retaining Partial Penetration Welds. As a result, the licensee failed to identify several relevant conditions that required additional inspections to adequately assure that the structural integrity of the reactor vessel head was not compromised. This issue was entered into the licensees corrective action program as Condition Report 2015-05995. The failure to establish adequate work instructions to clean and inspect the reactor vessel head was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the procedure quality attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small lossof-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. The inspectors determined that the finding had a teamwork cross-cutting aspect in the area of human performance because individuals and work groups failed to communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. (H.4)
05000285/FIN-2015002-0830 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Submit Summaries of the Impact of Changes to the Emergency Plan and Implementing ProceduresThe inspector identified a non-cited violation of 10 CFR 50.54(q)(5) for the licensees failure to submit reports of its analysis of the impact of changes to the emergency plan and implementing procedures on the emergency plan. Specifically, the inspector identified three examples between February 21 and June 18, 2015, of the licensee submitting changes to the emergency plan and implementing procedures without the required summaries. The issue was entered into the licensees corrective action program as Condition Report CR 2015-04934. The failure to submit a summary of the analysis of the effect of changes to emergency plan implementing procedures on the site emergency plan is a performance deficiency within the licensees ability to foresee and correct. The issue is more than minor because the licensees failure to submit the required summary affects the NRCs ability to perform its regulatory function, and the licensee has not incorporated this requirement into its program. The inspectors evaluated the issue using Section 6.6.d of the NRC Enforcement Policy, dated July 12, 2011, and determined it to be a Severity Level IV violation because the issue involved the licensees ability to implement a regulatory requirement not related to assessment or notification. Traditional enforcement violations are not assigned a crosscutting aspect.
05000285/FIN-2015002-0930 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Follow Instructions and Procedures Related to Snubber ActivitiesThe inspectors identified a non-cited violation of very low safety significance of 10 CFR Part 50, Appendix B, Criterion V Instructions, Procedures, and Drawings, because activities affecting quality were not accomplished in accordance with instructions and procedures established by the licensee. Specifically, the licensee failed to document a degraded condition associated with a safety related seismic snubber affecting the auxiliary feedwater system, did not notify operations of the degraded condition, and did not assess the risk of the inoperable snubber in accordance with licensee instructions and procedures. The licensee entered this violation into their corrective action program. Immediate actions taken to address this violation included a review of all other snubber inspections that were rejected to ensure that other degraded conditions were reported to the control room, a review of all planned snubber maintenance with respect to online risk, and the issuance of interim guidance to all Shift Managers on the subject of snubber operability and risk. The inspectors determined that the licensees failure to follow instructions and procedures associated with safety related snubbers was a performance deficiency. The finding is more than minor because if left uncorrected, the performance deficiency could have led to a more significant safety concern. Specifically, the failure to follow instructions and procedures associated with safety related snubbers could result in unacceptable risk configurations that are not analyzed under technical specifications and could challenge the reliability of safety related equipment during a seismic event. Using NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2 Mitigating System Screening Questions Part B, dated July 1, 2012, the inspectors determined the finding to be of very low safety significance (Green) since the finding did not result in the loss of equipment specifically designed to mitigate a seismic initiating event. The finding has a cross-cutting aspect in the area of Human Performance, the Work Management aspect, since the licensee did not implement a work process that ensured the identification and management of risk commensurate to the work. (H.5)
05000285/FIN-2015002-0430 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Incorporate Vendor Manual Recommendations for Conducting Preventative Maintenance on the Reactor Vessel Head Vent ValveThe inspectors identified a non-cited violation of Technical Specification 5.8.1.a associated with the failure to establish a preventative maintenance schedule for the reactor vessel head vent manual isolation valve, RC-100. Specifically, engineering personnel failed to consider vendor recommended maintenance activities/schedules, and determined that the valve could be run to failure. As a result, when the valve packing failed during operation, boric acid leaked onto the reactor vessel head. The licensee replaced the valve internals during refueling outage RFO 27 under Work Order 551054. This issue was entered into the licensees corrective action program as Condition Report 2015-05432. The failure of engineering personnel to establish a preventative maintenance schedule for the reactor vessel head vent manual isolation valve was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the equipment performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. No cross-cutting aspect was assigned because the inspectors determined that the finding was not indicative of current performance.
05000285/FIN-2015002-0530 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Promptly Identify and Correct a Condition Adverse to Quality Involving a Spent Fuel Pool Cooling Vent Valve LeakThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to promptly identify and correct a condition adverse to quality. Specifically, the licensee failed to take corrective action to replace spent fuel pool cooling system discharge header vent valve AC-898 after a leak was identified. A work order for the condition was opened in 2009 but was never implemented. Subsequently, a pressure boundary leak was identified in 2013 and misidentified in 2014 but was never addressed. The licensee replaced vent valve AC-898 and repaired the affected weld in April 2015. This issue was entered into the licensees corrective action program as Condition Report 2015-05038. The failure to promptly identify and correct a condition adverse to quality was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not adversely affect decay heat removal capabilities from the spent fuel pool causing the pool temperature to exceed the maximum analyzed temperature limit specified in the site-specific licensing basis, did not result from fuel handling errors, dropped fuel assembly, dropped storage cask, or crane operations over the SFP that caused mechanical damage to fuel clad and a detectible release of radionuclides, did not result in a loss of spent fuel pool water inventory decreasing below the minimum analyzed level limit specified in the site-specific licensing basis, and did not affect the SFP neutron absorber, fuel bundle misplacement or soluble boron concentration. The inspectors determined that the finding had a basis for decision cross-cutting aspect in the area of human performance because leaders failed to ensure that the bases for operational and organizational decisions were communicated during multiple instances where the leak in valve AC-898 could have been repaired. (H.10)
05000285/FIN-2015002-0630 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Identify and Correct Loose Incore Instrument Nozzle ConnectionThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the failure to identify and correct a condition adverse to quality. Specifically, maintenance personnel failed to document a loose connection on incore instrument port 44 in the corrective action program. As a result, the connection was not tightened and boric acid leaked onto the reactor vessel head during operation. This issue was entered into the licensees corrective action program as Condition Report 2015-05864. The failure of maintenance personnel to document a loose connection on incore instrument port 44 in the corrective action program was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the equipment performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not result in exceeding the RCS leak rate for a small loss-of-coolant accident, and did not affect other systems used to mitigate a loss-of-coolant accident resulting in a total loss of their function. The inspectors determined that the finding had a field presence cross-cutting aspect in the area of human performance because the licensee did not ensure supervisory and management oversight of work activities, including contractors and supplemental personnel. (H.2)
05000285/FIN-2015002-0230 June 2015 23:59:59Fort CalhounNRC identifiedFailure to Perform a Valid 40-Month Inservice TestThe inspectors identified a non-cited violation of 10 CFR 50.55a(g)(4) for the failure to perform a valid 40-month inservice test of the spent fuel pool cooling system. Specifically, the licensee failed to identify an existing through-wall leak on discharge header vent valve AC-898 that invalidated the test. The licensee replaced vent valve AC-898 and repaired the affected weld in April 2015. This issue was entered into the licensees corrective action program as Condition Report 2015-05038. The failure to perform a valid 40-month inservice test of the spent fuel pool cooling system was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it adversely affected the design control attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions, the issue screened as having very low safety significance (Green) because the finding did not adversely affect decay heat removal capabilities from the spent fuel pool causing the pool temperature to exceed the maximum analyzed temperature limit specified in the site-specific licensing basis, did not result from fuel handling errors, dropped fuel assembly, dropped storage cask, or crane operations over the SFP that caused mechanical damage to fuel clad and a detectible release of radionuclides, did not result in a loss of spent fuel pool water inventory decreasing below the minimum analyzed level limit specified in the site-specific licensing basis, and did not affect the SFP neutron absorber, fuel bundle misplacement or soluble boron concentration. The inspectors determined that the finding had a conservative bias cross-cutting aspect in the area of human performance because individuals failed to use decision making-practices that emphasized prudent choices over those that are simply allowable. Although the licensee had previously identified the leak in valve AC-898 and determined that the leak had compromised the structural integrity of the system, the licensee failed to fix the leak. (H.14)
05000285/FIN-2015007-1331 March 2015 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis...are correctly translated into specifications, drawings, procedures, and instructions. Contrary to the above, in December 2014, the licensee identified a failure to ensure that the design basis for station blackout was correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to include a diesel fuel oil pump load in the design basis battery load profiles for station blackout, design basis accident, and safe shutdown. This finding was determined to be of very low safety significance because the licensee performed an operability determination and concluded that Battery EE-8A was operable but non-conforming; by reducing the aging factor from 1.25 to 1.2. The licensee entered this issue into their corrective action program as Condition Report CR 2014-14857.
05000285/FIN-2015008-0331 March 2015 23:59:59Fort CalhounSelf-revealingFailure to Promptly Identify and Correct a Condition Adverse to QualityThe team reviewed a self-revealing Green non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify a condition adverse to quality. On October 27, 2014, a condition report was written to investigate dry boric acid on the high pressure safety injection Pump SI-2B vent valve piping. The initial investigation concluded that no degraded or nonconforming condition existed. On October 29, 2014, the Boric Acid Corrosion Control Program engineer conducted a review of the dry boric acid residue. The engineer identified the boric acid appeared to originate from a weld and needed to be cleaned and repaired; however, the engineer failed to initiate a condition report documenting this condition adverse to qualilty. The failure to promptly identify and correct a condition adverse to quality in accordance with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was a performance deficiency. Specifically, the licensee failed to write a condition report when there was evidence of a boric acid leak on the high pressure safety injection pump casing. This performance deficiency was of more-than-minor safety significance because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609 Appendix A, Exhibit 2, the finding was of very low safety significance (Green) because all questions in Exhibit 2 could be answered in the negative. The finding had a cross-cutting aspect in the procedure adherence component of the human performance cross-cutting area because the individual failed to write a condition report as required by procedure after identifying a condition adverse to quality.
05000285/FIN-2015001-0131 March 2015 23:59:59Fort CalhounNRC identifiedFailure to Conduct and Evaluate Simulator Testing In Accordance with ANSI/ANS-3.5-2009The inspectors identified a Green finding with four examples for failing to conduct and evaluate simulator performance testing in accordance with the standards of ANSI/ANS-3.5- 2009. Specifically, the licensee failed to do the following: - Set initial reactor power at 15 percent in accordance with plant design for all performances between 1990 and 2014 of Transient (6), "Main Turbine Trip from Maximum Power Level That Does Not Result in Immediate Reactor Trip" - Set the instantaneous main turbine load reduction to 1 0 percent as supported by design basis data in the 2014 performance of Transient (11), "Maximum Design Load Rejection" - Evaluate the results of the 100 percent power Steady-State Performance Test using the correct acceptance criteria in accordance with the standard, Appendix 8, Section 8.1.1 - Evaluate all transient test results versus acceptance criteria 4.1.4(1) in accordance with the standard, Appendix 8, Section 8.1.2 After NRC identification of the transient test issues, licensee evaluation revealed that the initial conditions for Transients (5) and (1 0) were in error as well. The licensee initiated corrective action documented in condition reports 2014-14190, 2014-14208, and 2015-02547. The licensee's failure to conduct and evaluate performance testing in accordance with the ANSI/ANS-3.5-2009 standard as endorsed by Regulatory Guide 1.149, Revision 4, was the performance deficiency. Per licensee Procedure TQ-AA-306, "Simulator Management," the licensee uses ANSI/ANS-3.5-2009 as the standard for their simulator testing. The performance deficiency is more than minor because if left uncorrected, the performance deficiency could have become more significant in that not completing the required simulator testing correctly can lead to not detecting and correcting errors in the simulator so it actually models the plant correctly. This can both leave the potential for negative training of licensed operators and call into question the ability to conduct valid licensing examinations with the simulator. Using Inspection Manual Chapter (IMC) 0609, "Significance Determination Process," Attachment 4, Tables 1 and 2 worksheets, and the corresponding Appendix I, "Licensed Operator Requalification Significance Determination Process (SOP)," Flowchart Block No.14, the finding was determined to have very low safety significance (Green) because it dealt with deficiencies associated with simulator testing, modification, and maintenance and there was no evidence that the plant-referenced simulator does not demonstrate the expected plant response or have uncorrected modeling and hardware deficiencies. This finding has a cross-cutting aspect in the change management area of human performance, associated with leaders using a systematic process for evaluating and implementing change so that nuclear safety remains their overriding priority. There were efforts on-site to change to the 2009 version of the standard as early as 2011, but the efforts were rescinded by plant management in December 2011 for unknown reasons. When they officially switched from the 1985 to the 2009 version of the standard (on March 1, 20 13), there is no evidence an effective change management plan was implemented. Efforts to transition between the testing and maintenance requirement differences were complicated by lack of allocating necessary resources to support this effort. There was minimal simulator staffing during the extended plant outage (April2011 to December 2013), and no effective plan to deal with knowledge management to compensate for simulator employee turnover. Internal audits in May 2014 and October 2014 found numerous issues with their simulator testing and configuration management program, many of which could have been averted or addressed earlier with an effective transition plan in place (H.3).
05000285/FIN-2015007-0631 March 2015 23:59:59Fort CalhounNRC identifiedFailure to Perform an Adequate Auxiliary Feedwater Pump Runout Design CalculationThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, 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 March 13, 2015, the licensee did not verify the adequacy of the design calculation or a suitable testing program to ensure the required net positive suction head was available for the turbine-driven auxiliary feedwater pump. In response to this issue, the licensee performed an operability determination; revised several calculational errors, including removing conservatisms which resulted in a gain of net positive suction head; and contacted the original equipment manufacturer who provided a testing summary that determined the turbine-driven pump could operate for a period of time below the required net positive suction head. This provided the licensee with the basis for an operable but non-conforming condition. This finding was entered into the licensees corrective action program as Condition Report CR 2015-02414. The team determined that the failure to verify the adequacy of the auxiliary feedwater system design through calculational analysis and a suitable test program was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to have adequate measures in place to ensure an acceptable design analysis and a suitable test program to verify the design inputs and ensure the capability of the auxiliary feedwater system to perform its safety function. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of human performance associated with conservative bias because individuals failed to use decision making practices that emphasize prudent choices over those that are simply allowed.
05000285/FIN-2015007-0731 March 2015 23:59:59Fort CalhounNRC identifiedFailure to Perform an Adequate Evaluation for the Intake Structure Crane Trolley and Bridge RailThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, 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 March 13, 2015, the licensee failed to perform an adequate design review to ensure the intake structure crane trolley and bridge rail were constructed to seismic class II over I standards. The licensee failed to ensure the intake structure crane trolley rail, trolley rail clip, trolley clip connection, crane rail, crane rail clip and crane clip connection were evaluated for loads due to the safe shutdown earthquake loading concurrent with a lifted load. In response to this issue, the licensee performed an operability determination and concluded that the crane was operable but nonconforming based on a load test that was performed at 1.25 times the rated capacity. This finding was entered into the licensees corrective action program as Condition Report CR 2015-02353. The team determined that the failure to perform an adequate design review to ensure the intake structure crane trolley and bridge rail were constructed to seismic class II over I standards was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to events to prevent undesirable consequences. Specifically, the licensee failed to comply with seismic class II over I requirements to ensure the intake structure crane structural integrity when subjected to safe shutdown earthquake loads concurrent with a lifted load; for safe load handling of heavy loads near the safetyrelated raw water system. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened to Exhibit 4, External Events Screening Questions, because it was a function specifically design to mitigate a seismic event. Per Exhibit 4 the issue screened to a more detailed risk evaluation because if the seismic function were assumed to be completely failed and a load were dropped it would impact the safety function of the raw water system. Therefore, the Region IV senior reactor analyst performed a more detailed risk evaluation. Given that the frequency of the initiating event is less than 1 x 10-6, the analyst determined that the finding was of very low safety significance (Green). This finding had a crosscutting aspect in the area of human performance associated with documentation because the licensee failed to maintain complete, accurate, and up-to-date documentation.
05000285/FIN-2015008-0231 March 2015 23:59:59Fort CalhounNRC identifiedUnresolved Item associated with the weld repair of SI-339 vent pipe leakageOn January 23, 2015, during a surveillance test of HPSI Pump 2B, a water leak was discovered at the seal weld between the pump casing and the halfinch ASME Class 2 pipe connected to discharge vent Valve SI-339. The leak rate was approximately 2 drips per minute. The licensee declared the pump inoperable and proceeded to repair the weld. The licensee performed visual and dye penetrant testing to identify a pinhole flaw in the seal weld. The testing did not reveal any flaws on the pump casing or the vent pipe. The pressure boundary for the safety injection system at this location is the threaded pipe connection to the pump casing. The seal weld is applied for leak tightness of the system. The licensee removed half of the seal weld around the pipe and did not observe any additional flaws on the pipe or casing. The licensee did not remove the pipe or perform additional testing on the pipe threads to evaluate if degradation of the threads led to the leakage or if the structural integrity of threaded connection remained intact. The seal weld was repaired and the HPSI pump was restored to operable status following postmaintenance testing. The NRC will review the actions taken by the licensee to determine if the structural integrity of the threaded pipe connection should have been further evaluated through additional testing or analyses. Until completion of this review, this issue will be tracked as unresolved item.
05000285/FIN-2015007-0131 March 2015 23:59:59Fort CalhounNRC identifiedFailure to Perform an Adequate Battery Sizing and Load Profile CalculationThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis...are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to March 13, 2015, the licensee failed to ensure that battery sizing and load profile calculations included proper design data for inrush currents, a random load, and possible worst case load currents. In response to these issues, the licensee updated the design values to account for the missed loads to ensure the batteries maintained adequate available margin. This finding was entered into the licensees corrective action program as Condition Report CR 2014-14857. The team determined that the failure to adequately perform a battery sizing and load profile calculation, to ensure proper battery size and margin was maintained, was a performance deficiency. This finding was more than minor because it was associated with the design control 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 licensee failed to account for inrush currents, random loads, and worst case load currents during load profile and battery sizing calculations. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a crosscutting aspect because the most significant contributor did not reflect current licensee performance.
05000285/FIN-2015001-0231 March 2015 23:59:59Fort CalhounNRC identifiedFailure to Implement Risk Management Actions for Planned Maintenance ActivitiesThe inspectors identified a non-cited violation of very low safety significance of 10 CFR 50.65 paragraph (a)(4) "Requirements for Monitoring the Effectiveness of Maintenance of Nuclear Power Plants," because the licensee did not effectively manage the increase in risk that resulted from maintenance activities. Specifically, the licensee failed to implement key risk management actions outlined in site risk assessment and management guidance for diesel driven auxiliary feedwater (AFW) pump maintenance that resulted in a "Yellow" risk configuration. This violation was entered into the licensee's corrective action program and actions taken for this violation included verifying that all remaining online work prior to the scheduled refueling outage was properly screened and assessed in accordance with site risk management procedures. In addition, the licensee conducted training on risk management guidance that had been recently implemented during corporate alignment for personnel involved with scheduling and operations. The inspectors determined that the licensee's failure to implement key risk management actions outlined in site risk assessment and management guidance for diesel driven AFW pump maintenance was a performance deficiency within the licensee's ability to foresee and correct and should have been prevented. The finding is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to perform maintenance on a continuous work schedule as required by site procedures resulted in a longer unavailability time of the equipment and an extended "Yellow" risk condition. Using NRC IMC 0609, Appendix K, "Maintenance Risk Assessment and Risk Management Significance Determination Process", dated May 19, 2005, Flowchart 2, "Assessment of (Risk Management Actions)", the inspectors determined the incremental core damage probability (ICDP) associated with the maintenance activity to be approximately 1 E-7, and therefore was determined to have a very low safety significance (Green), since the calculated I COP was less than 1 E-6. Because the licensee did not use a systematic process to ensure that nuclear safety remained the overriding priority while they implemented a corporate alignment, the finding has a cross-cutting aspect in the area of Human Performance, Change Management (H.3).
05000285/FIN-2015007-0831 March 2015 23:59:59Fort CalhounNRC identifiedFailure to Obtain Prior NRC Approval for a Change in Seismic Analysis DampingThe team identified two examples of a Severity Level IV, Green, non-cited violation, of 10 CFR 50.59, Changes, Tests and Experiments, for the licensees failure to obtain a license amendment prior to implementing a change if the change would result in a departure from a method of evaluation described in the updated safety analysis report. Specifically, on February 23, 2015, and March 10, 2015, the licensee changed the facility to incorporate increased seismic damping for use in the intake structure crane and intake superstructure seismic analysis and seismic design; and in the raw water piping seismic analysis, respectively. In response to this issue, the licensee declared the intake structure as operable but non-conforming pending resolution of a license amendment request to permit the use of the increased damping value; and declared the raw water system as operable but non-conforming pending completion of the corrective actions to determine what actions are necessary to restore compliance to the licensing basis. This finding was entered into the licensees corrective action program as Condition Reports CR 2015-02224 and CR 2015-02842. The team determined that the failure to identify that the proposed change to incorporate increased seismic damping for use in the intake structure crane and intake superstructure seismic analysis and seismic design; and in the raw water piping seismic analysis, was a performance deficiency. This finding was also evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences; and there was a reasonable likelihood that the change would have required NRC review and approval prior to implementation. Specifically, the licensee failed to determine that the proposed updated safety analysis report change, and associated design calculations, did involve a change to a structure, systems, or components such that it did adversely affect an updated safety analysis report described design function; less conservative seismic damping values, which required an evaluation to be performed. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Since the 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. This finding had a crosscutting aspect in the area of human performance associated with design margins because individuals failed to ensure margins were carefully guarded and changed only through a systematic and rigorous process.
05000285/FIN-2015007-0931 March 2015 23:59:59Fort CalhounNRC identifiedIntake Structure Design RequirementsThe following issues were discussed during the inspection, but the licensee was unable to provide the required information to be able to disposition the issue in accordance with Inspection Manual Chapter 0612 as either minor or more than minor. Additionally, the aforementioned calculations were performed by third-party vendors, so the licensee staff did not have specific knowledge of the calculations because they did not prepare them. To close the unresolved item the NRC needs additional information to be able to address the following concerns: Hydrodynamic forces and levee construction: The team found that the intake structure licensing basis design requirements are unclear for the natural phenomenon expected during a flood, specifically hydrodynamic effects. The team questioned the rationale that the design of the intake structure only included the consideration of hydrostatic forces during a design basis flood of the Missouri River. The expected design would include hydrodynamic forces (as the river is moving at a velocity of approximately 10 feet per second) and the potential impact loads from debris floating downstream. Historical correspondence between the Atomic Energy Commission (predecessor to the NRC) and the licensee related to this question revealed that the original flood mitigation strategy included construction of temporary earthen levees (EA10-032, Revision 0; Preliminary Safety Analysis Report, Supplement 12; and Final Safety Analysis Report, Supplement 15, February 1972). Designing the intake structure to only include hydrostatic loads from the Missouri River during flood stages is not intuitive; however, it can be reasoned that the licensees original proposed solution of constructing earth levees to protect the plant would be sufficient to isolate the flowing river water and debris such that hydrodynamic flooding effects would not need to be considered. The licensee does not construct temporary earthen levees to protect structures during flood events, nor does it have plans to do so in the future. Code limits for flooding conditions: Calculation FC07802, Intake Structure Project Design Manual, states that the NRC accepted Calculation FC01414, Exterior Wall Design, Diesel Generator Room, Revision 1, for the exterior wall designs of the auxiliary building and that the design code limits established in that calculation would be used in the reconstitution of the intake structure. Calculation FC01414, Revision 1, was not retrievable during the inspection. Revision 0, dated March 23, 1968, was found and provided to the team, which showed the building was designed using ultimate strength design limits of the American Concrete Institute ACI 318 code for a flood to elevation 1014 MSL. However, the final safety analysis report requires the intake structure and auxiliary building to be designed to the working stress design limits of the American Concrete Institute ACI 318 code. In addition, Revision 0 identifies an area that is overstressed during a flood event and would require larger diameter reinforcement bars than analyzed. The licensee was unable to provide the team with documentation explaining how the ultimate stress design versus working stress design and the additional reinforcement bars were resolved. Required loading combinations: A design memo (WIP file 019070, dated July 7, 1970) between the architect engineer who designed the intake structure and the licensee states that the intake structure has been designed as Class I structure up to El. 1007-6 but that the perimeter wall between elevations 1007-6 and 1014-6 is not so designed and will not resist simultaneous flood and earthquake loading... Final Safety Analysis Report, Section 5.11.3, requires the intake structure to be able to resist a simultaneous flood and earthquake loading combination. The licensee was unable to provide the team with documentation explaining why the intake structure perimeter walls were not designed to resist the required loading combination of an earthquake and flood. Building, foundation, and soil stiffness: Calculations FC07495, Intake Structure Seismic Analysis and Pile Design, and FC07803, lntake Structure, SubStructure Analysis, derived new stiffness values for both the steel pile foundation that supports the intake structure and the soil stiffness interaction values. The use of these values affects the behavior of the structure during an earthquake and barge impact. In addition, there are discrepancies in the natural frequency of the building that are not addressed. The licensee was unable to provide the team with documentation explaining the use of the new stiffness values or the lack of addressing the natural frequency of the building. Sloshing: Calculation FC07495 assumes that the opposing walls of the intake structure are relative short distances so the water acts as a rigid body (no water sloshing effects are accounted for). This assumption did not have any technical justification. Specifically, a representative raw water cell is 20 feet deep by 10 feet long by 7 feet wide, and the total estimated water weight for each cell is over 1.1 million pounds. Consequently sloshing effects could comprise a large additional load on the structure. The licensee was unable to provide the team with documentation explaining the lack of including water sloshing effects on the walls of the intake structure. Barge impact loading: A barge crashing into the intake structure is a required design load for the building. Previous NRC Inspection Reports (05000285/2005011 and 05000285/2009006) documented the lack of design calculations for this required loading condition. Calculation FC07803 analyzes the barge impact loads, however the team raised the following concerns which the licensee was unable to provide documentation explaining the discrepancies: o A barge impact is only assumed to occur at average river levels (986.5 to 1001.3 feet MSL) and is neglected at low or high river levels (979.5 to 1014 feet MSL), because of the assumption that the Army Corps of Engineers would impose barge restrictions or raise/lower water level. No formal agreement or justification for this assumption has been provided. o The barge impact calculation arbitrarily selected a 40 percent increase in material strengths of the intake structure. This increased factor was based on engineering experience without any technical justification or data to support the use of higher strength materials than specified in the final safety analysis report. o The barge impact assumes a barge velocity of 2.2 feet per second; however the actual river velocity is 10 feet per second. The calculation assumes the maximum barge speed is 10 feet per second, despite the fact the barge has propulsion and could travel at a speed equal to the river velocity plus the barge propulsion speed. The use of a lower velocity increases the duration of the loading event; however it significantly underestimates the energy of the barge impact. If the barge is assumed to impact the intake structure at 10 feet per second, the calculation states that: ...the force would entirely overwhelm the flexural capacity of the nosing wall sections, which confirms that these structural elements would fail under Barge Impact loading conditions. However, the extremely short duration of the postulated loading condition (0.075 seconds) may not be structurally significant on the remainder of the Intake Structure, nor on the safety-related SSCs. The team questioned why the lower velocity was used, because the effective force of a 10 feet per second impact (distributed over a short duration) would be significantly greater than a 2.2 feet per second impact distributed over a larger time duration.
05000285/FIN-2015007-0231 March 2015 23:59:59Fort CalhounNRC identifiedFailure to Establish Correct Acceptance Criteria Values for Battery Intercell Resistance MeasurementsThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis...are correctly translated into specifications, drawings, procedures, and instructions. Specifically, since 2009, the licensee failed to update battery maintenance procedures with the current maximum intercell resistance values. In response to this issue, the licensee performed a visual inspection of the battery intercell connections, performed a review of the latest intercell resistance measurements to identify any values that exceeded the correct acceptance criteria value, and performed an immediate operability determination. This finding was entered into the licensees corrective action program as Condition Report CR 2015-02129. The team determined that the failure to establish the correct acceptance criteria values for battery intercell resistance measurements was a performance deficiency. This finding was more than minor because it was associated with the procedure quality 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 licensee had incorrect acceptance criteria for maximum intercell connection resistance measurements, and failed to identify an intercell connection that should have been disassembled, cleaned, reassembled, and remeasured. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a crosscutting aspect in the area of human performance associated with documentation because the licensee failed to maintain complete, accurate, and up-to-date documentation.
05000285/FIN-2015001-0431 March 2015 23:59:59Fort CalhounLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50.74, "Notification of Change in Operator or Senior Operator Status," requires that "each licensee shall notify the appropriate Regional Administrator within 30 days in regard to a licensed operator or senior operator: permanent disability or illness as described in Part 55.25." Contrary to the above, from January 27, 2014, to April23, 2014, the licensee failed to notify the Regional Administrator within 30 days of a permanent disability or illness of a licensed operator. Specifically, a licensed operator was prescribed medication for hypertension; however, this condition was not reported to the NRC as required by 10 CFR 55.25. The licensee documented the deficiency in Condition Report 2014-05185. The failure to report required information to the NRC is a violation. The violation was evaluated using the traditional enforcement process because it impacted the NRC's ability to perform its regulatory function. The violation was determined to be Severity Level IV because it fits the example of Enforcement Policy Section 6.4.d.1 (d), "Violation Examples: Licensed Reactor Operators." This section states, "SL IV violations involve, for example: ... an individual operator who met ANSI/ANS 3.4, Section 5, as certified on NRC Form 396, required by 1 0 CFR 55.23, but failed to report a condition that would have required a license restriction to establish or maintain medical qualification based on having the undisclosed medical condition."