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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000282/FIN-2018002-01 | 2018Q2 | Prairie Island | Results of ISFSICask Array Dose Calculation Not Incorporated into FSAR | Prairie Island ISFSI FSAR, as updated, Revision 18, Section A7A.7 evaluates off-site dose rates for an array of ISFSI casks. In this dose rate calculation, explicit modeling credit is given to the earthen berm that surrounds the Prairie Island ISFSI as discussed in Section A7A.7.1. The earthen berm provides radiation shielding for the ISFSI. This calculation allows the licensee to demonstrate, in part, compliance with Title 10 of the Code of Federal Regulations (CFR) 72.104(a) which requires, in part, that, During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ. Calculation TN40HT0502, TN40HT Far Field Shielding Calculations, Revision 0, was performed by the licensee in support of a License Amendment Request (LAR) to modify the Prairie Island ISFSI TN40 cask design (designated as TN40HT casks). The TN40HT LAR was submitted to the NRC by the licensee on March 28, 2008. This dose rate calculation does not credit the earthen berm and, in part, also allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a). The licensee also provided this calculation directly to the NRC in a February 29, 2012, letter in response to a Request for Supplemental Information (RSI) from the NRC associated with the license renewal application for the Prairie Island ISFSI. Although the results from calculation TN40HT0502 for a single cask was incorporated into the Prairie Island ISFSI FSAR, Revision 18, in Tables A7A.22 and A7A.61, the results from TN40HT0502 for an array of casks which, in part, allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a), has not been incorporated into the ISFSI FSAR, Revision 18.Title 10 CFR 72.70, Safety analysis report updating requires, in part, that (a) Each specific licensee for an ISFSI shall update periodically, as provided in paragraphs (b) and (c) of this section, the FSAR to assure that the information included in the report contains the latest information developed (b) Each update shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last update to the FSAR under this section. The update shall include the effects of: (2) All safety analyses and evaluations performed by the licensee in support of approved license amendments.This Unresolved Item is being opened to determine whether or not the licensee is required to update the ISFSI FSAR with the results of calculation TN40HT0502 for an array of casks in accordance with 10 CFR 72.70.Planned Closure Action: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Material Safety and Safeguards to determine whether or not calculation TN40HT0502 is subject to the FSAR updating requirements of 10 CFR 72.70 for the Prairie Island ISFSI. |
| 05000285/FIN-2013008-30 | 2013Q2 | Fort Calhoun | Evaluation of Change to Alternate Shutdown Cooling Flowpath | The team identified an unresolved item related to engineering change modifications that changed a procedure to include the replacement of automatic actions with manual actions. Specifically, the10 CFR 50.59 evaluation proposed a change to substitute automatic flow control of shutdown cooling flow and temperature with manual control using the low pressure safety injection loop injection valves alternate shutdown cooling flow control. During a review of Engineering Change Modification 54058, Procedure Change to Allow Closing of HCV-335 while on Alternate Shutdown Cooling, the team identified that the licensee changed a procedure to include the replacement of an automatic action with a manual action. Specifically, the engineering change proposed to close both shutdown cooling heat exchanger isolation valve HCV-335 and flow control valve FCV-326 while pinning open valve HCV-341 and manually throttling low pressure safety injection loop injection valves to maintain the desired RCS temperature and flow rate. The team questioned whether the licensee required prior NRC review and approval to make this change since flow control valve FCV-326 normally controls temperature and flow automatically as described in Section 9.3.4.3 of the USAR. The licensee entered this issue of concern into the CAP. Additional NRC review and follow up will be required to determine if this issue represents a performance deficiency associated with meeting the 10 CFR 50.59 requirement of more than a minimal increase in the likelihood of occurrence of a malfunction of a system, structure, or component important to safety previously evaluated in the USAR. This item is unresolved pending review of the licensees evaluation. This issue is identified as URI 05000285/2013008- 30, Evaluation of Change to Alternate Shutdown Cooling Flowpath. |
| 05000285/FIN-2014005-01 | 2014Q4 | Fort Calhoun | Failure to establish Appropriate Preventive Maintenance and Failure to Identify Raw Water SSC Maintenance Rule Performance Criteria Exceeded and thereby establish Monitoring Requirements for the SSC | The inspectors identified an NCV of very low safety significance of 10 CFR 50.65 paragraph (a)(2) Requirements for Monitoring the Effectiveness of Maintenance of Nuclear Power Plants, because the licensee did not demonstrate that performance of a component was being effectively controlled through appropriate preventive maintenance, and did not monitor the performance of the component against licensee-established goals to provide reasonable assurance that the component was capable of fulfilling its intended function. Specifically, the licensee failed to demonstrate that the performance of raw water system valve HCV-2875A was being effectively controlled through appropriate preventive maintenance and failed to monitor the valves performance against licensee established goals when performance criteria were exceeded. Corrective actions taken for this violation included revising the Maintenance Rule performance criteria assessment for this component, classifying the component as 10 CFR 50.65 (a)(1), and specifying goals, corrective actions, and additional monitoring for the component. The licensees failure to demonstrate component performance through appropriate preventive maintenance, and the failure to identify that system performance criteria had been exceeded, and as a result, the failure to perform an evaluation of the system for 50.65 (a)(1) goals, corrective actions, and monitoring, was a performance deficiency within the licensees ability to foresee and correct. 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 reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to identify that valve HCV-2875A performance criteria had been exceeded resulted in a delayed assessment of this component and additional failures occurred in the intervening timeframe which adversely affected the overall reliability of the raw water system. The inspectors screened the finding in accordance with NRC IMC 0609, Appendix A, the Significance Determination Process (SDP) for Findings at Power. Using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, this finding is of very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating system (2) did not represent a loss of system and/or function, (3) did not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its TS allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution and the Evaluation aspect because the licensee failed to appropriately evaluate the preventive maintenance for valve HCV-2875A to demonstrate component performance and failed to correctly evaluate a functional failure against system performance criteria to ensure system goals, corrective actions, and monitoring were identified. |
| 05000285/FIN-2014005-02 | 2014Q4 | Fort Calhoun | Failure to determine the availability of local population data for use in estimating changes in the EPZ population | The NRC identified a Green non-cited violation for the licensees failure to determine the availability of year 2013 state and local population data in estimating annual changes in the plume exposure emergency planning zone population. The failure to determine whether State and/or local population data was available for 2013 was a performance deficiency within the licensees ability to forsee and correct. Appendix E to 10 CFR Part 50, Section IV.5, states, in part, that during the years between decennial censuses, nuclear power reactor licensees shall estimate emergency planning zone permanent resident population changes once a year using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. Contrary to the above, Fort Calhoun Station failed in 2013 to estimate emergency planning zone permanent resident population changes using the most recent U.S. Census Bureau annual resident population estimate and State/local government population data, if available. Specifically, Fort Calhoun Station failed to determine whether State and local government population data was available prior to performing the analysis. The issue was entered into the licensees corrective action system as Condition Report 2014-12474. This finding is more than minor because the issue is associated with procedure quality and offsite Emergency Preparedness cornerstone attributes and adversely affected the Emergency Preparedness cornerstone objective. The finding was evaluated using Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, dated February 24, 2014, and was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements, was not a loss of planning standard function, and was not a degraded planning standard function. The planning standard function was not degraded because including state and local 2013 data would not have required the current emergency planning zone time estimate to be updated. There are no immediate safety or security concerns associated with this finding. This finding was assigned a cross-cutting aspect in the area of human performance associated with work management because the licensee failed to understand the scope of work performed by a contractor on their behalf, and failed to ensure the contractor fully complied with regulatory requirements. |
| 05000285/FIN-2016001-01 | 2016Q1 | Fort Calhoun | Implementing a Procedure Change for Alternative Shutdown Cooling that would have Required NRC Approval | The 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-2016001-02 | 2016Q1 | Fort Calhoun | Licensee-Identified Violation | Technical 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. |