|Start date||Site||Identified by||Title||Description|
|05000373/FIN-2015009-01||30 September 2015 23:59:59||LaSalle||NRC identified||Use of an Analytical Method to Determine the Core Operating Limits without Prior NRC Approva||The inspectors identified a Severity Level IV NCV of Technical Specification (TS) Section 5.6.5, for using an analytical method that was not previously reviewed and approved by the NRC. Specifically in 2013, the licensee used TRACG04P code to determine the Oscillation Power Range Monitor setpoints prior to NRC approval. The TRACG04P code was reviewed and approved in April 24, 2015. TS Section 5.6.5.b stated, in part that the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the TS. The licensee entered this finding into their Corrective Action Program (CAP) as IR 02528609 and IR 02528612 to correct the issue. The inspectors determined that this issue was a performance deficiency and because the issue had the potential to affect the NRCs ability to perform its regulatory function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using the Enforcement Manual, the inspectors characterized the violation as Severity Level IV because the underlying analytical method required NRC approval prior to use. The inspectors did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Section 07.03.c.|
|05000374/FIN-2015002-03||30 June 2015 23:59:59||LaSalle||NRC identified||Inadequate 10 CFR 50.59 Evaluation for Jet Pump Plugs Affecting Fuel Bundle Cooling||The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.59, Changes, Tests, and Experiments, having very low safety significance (Green), for the licensees failure to provide a written safety evaluation supporting the determination that a license amendment was not required for operation with jet pump seal plugs (lost in the reactor vessel in February 2015 during a refueling outage) that could negatively impact fuel bundle cooling during an anticipated operational occurrence (misplaced fuel bundle). The licensee entered this issue into the corrective action program (CAP) as action report (AR) 02486215 and considered the core operable because additional testing demonstrated that with sufficient time (approximately 11 days) at operating temperature, the rubber plugs would degrade and pass through the affected flow orifices. The finding was determined to be more than minor because the inspectors could not reasonably determine that the activity, to operate with jet pump plugs blocking peripheral fuel bundle flow, would not have required prior NRC approval. Specifically, if the licensee operated a peripheral blocked fuel bundle coincident with a misplaced fuel bundle, the minimum critical power ratio limits/margins may not have been assured. Additionally, this finding was more than minor because the underlying technical issue adversely affected the Barrier Integrity Cornerstone objective of design control and cladding performance. The finding involves the potential for a misplaced fuel bundle concurrent with complete flow blockage to a fuel assembly. Given standard refueling practices, an error that results in plant operation with a misplaced fuel bundle is very unlikely due to strict procedural controls and multiple verifications of fuel assembly placement. In addition, the misplaced fuel assembly would have to be located at a peripheral core location to be susceptible to a jet pump plug that could possibly block bundle cooling and this was very unlikely. Further, the inspectors considered the relatively short duration of time where the plug material parameters were sufficient to cause plugging of an orifice coincident with plant power levels that could challenge the fuel integrity limits. Given these factors, the inspectors determined that the likelihood of a misplaced fuel assembly combined with a blocked orifice that could result in fuel clad damage was very low. Given the very low likelihood of the event scenario to occur and the low consequences if it were to occur, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors identified a cross-cutting aspect associated with this finding in the area of Human Performance, Conservative Bias, because the licensee staff did not use a decision-making practice that emphasized prudent choices over those that are simply allowable (H.14).|
|05000374/FIN-2015002-04||30 June 2015 23:59:59||LaSalle||NRC identified||Failure to Include Limiting Conditions for Operation in the Technical Specifications||The inspectors identified a Severity Level IV NCV of 10 CFR 50.36, Technical Specifications, having very low safety significance (Green), for the licensees failure to ensure that limiting conditions for operation (LCOs) were contained in the stations Technical Specifications (TSs). Specifically, as of March 15, 2015, through the Unit 2 Core Operating Limits Report (COLR), Cycle 16, Revisions 1 and 2, the licensee introduced new Operating Limits for Lost Jet Pump Plug Seals Mitigation Strategy, that created new LCOs as defined by 50.36(c)(2) but did not incorporate these LCOs into the TSs. The licensee incorrectly believed that because the COLR was revised via the 50.59 process and the special content that accounted for the existence of the plugs was developed using NRC-approved methodologies, the change was acceptable and no change to the TSs was obtained from the NRC. This finding was considered more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the Unit 2 COLR was revised in a manner that created new LCOs, and further, could have resulted in the operation of Unit 2 outside of its approved TSs and license. Operating the unit in accordance with its NRC-approved TSs could have resulted in the plant operating in an unanalyzed condition that could have resulted in fuel failure. The finding involves the potential for a failed safety/relief valve (SRV) or turbine bypass valve concurrent with complete flow blockage to a peripheral fuel assembly, with a simultaneous breakdown of control room operator knowledge of the special steps required by the COLR revision. Given standard operating practices and the significant amount of extra attention and sensitivity placed on the jet pump plugs and their potential effect, an error that results in licensed operators failing to comply with the restrictive limits of the COLR would be very unlikely. Additionally, a read-and-sign was required of all Unit 2 control room operators and supervisors delineating the special compensatory measures to be taken in the event that a COLR base case component, such as an SRV, were to fail. Further, the inspectors considered the relatively short duration of time (March 1523, 2015) where the plug material parameters were sufficient to cause plugging of an orifice coincident with plant power levels that could challenge the fuel integrity limits. Given these factors, the inspectors determined that the likelihood of a failed COLR base case component, combined with the operation of the unit in an unanalyzed condition in accordance with the NRC-approved TSs, combined with a blocked orifice that could result in fuel clad damage was very low. Given the very low likelihood of the event scenario to occur and the low consequences if it were to occur, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Change Management, because the licensee leaders did not ensure the use of a systematic process for evaluating and implementing change so that nuclear safety remained the overriding priority (H.3).|
|05000373/FIN-2015001-06||31 March 2015 23:59:59||LaSalle||Licensee-identified||Licensee-Identified Violation||Title 10 CFR 50.72(b)(3)(xiii) states, in part, a licensee shall report (notify the NRC as soon as practical, and in all cases within 8 hours of the occurrence) any event that results in a major loss of emergency assessment capability. Contrary to this requirement, on March 24, 2015, the licensee identified a failure to submit a report for the loss of emergency assessment capability when the site declared seismic monitoring instrumentation inoperable. Specifically, on January 28, 2015, the Instrument Maintenance Department discovered the seismic monitoring program on the seismic laptop computer in the auxiliary electrical equipment room was not running; thereby, preventing the seismic monitoring instrumentation from providing indications required for emergency assessment of a potential seismic event. The system degradation would have adversely impacted the sites ability to declare an ALERT Emergency Action Level in accordance with EPAA1005, Radiological Emergency Plan Annex for LaSalle Station, in the event of an earthquake of sufficient magnitude. The licensee entered the issue into the CAP as AR 02473472, Need to Assess Seismic Monitor Reportability, and conducted an extent of condition review for the prior 3-year period. The licensee identified a total of six times in which the seismic monitoring system experienced this degradation, and the licensee failed to submit an event report at the time, as required by 10 CFR 50.72(b)(3)xiii). Upon completion of the extent of condition review, the licensee initiated AR 02474658, Emergency Notification System Notification Required for Past Seismic Monitor Inoperative, and submitted the required notification to the NRC on March 26, 2015, to restore compliance (Event Number 50926, Seismic Monitor Not Available for Emergency Plan Assessment ). The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the generic communications input that 10 CFR 50.72 reports serve. Since the issue impacted the regulatory process, it was dispositioned through the Traditional Enforcement Process. The inspectors determined that this issue was a Severity Level IV violation based upon Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, Example d.9 in the NRC Enforcement Policy. Example d.9 specifically states, The licensee fails to make a report requirement by 10 CFR 50.72, or 10 CFR 50.73. Because the issue was entered into the licensees CAP (as AR 02473472 and AR 02474658), the violation is being treated as an NCV consistent with Section 2.3.2 of the NRC Enforcement Policy.|
|05000373/FIN-2013009-02||30 June 2013 23:59:59||LaSalle||NRC identified||Failure to Satisfy 10 CFR 50.72 Reporting Requirements||The inspectors identified a finding of very low safety significance with an associated Severity Level IV non-cited violation of the NRCs reporting requirements in 10 CFR 50.72(a)(1), Immediate Notification Requirements for Operating Nuclear Power Reactors. The licensee failed to make a required 8-hour non-emergency notification call to the NRC Operations Center after discovery of a condition that could have prevented fulfillment of the safety function of the low pressure core spray (LPCS) system. The licensee entered this issue into its corrective action program (CAP) for evaluation and made an appropriate notification call to the NRC Operations Center. This finding was of more than minor significance because the NRC relies on licensees to identify and report conditions or events meeting the criteria specified in the Technical Specifications and the regulations in order to perform its regulatory function. Because this issue affected the NRC\'s ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process and assessed the significance of the underlying issue using the Significance Determination Process (SDP). The underlying technical issue (i.e., an inoperable LPCS system) was determined to be of very low safety significance using the SDP. Consistent with the guidance in Section 6.9, Paragraph d.9, of the NRC Enforcement Policy, the violation associated with this finding was determined to be a Severity Level IV Violation. This finding affected the cross-cutting area of problem identification and resolution. Specifically, the licensee did not implement and institutionalize operating experience from a similar event reported at another licensees facility while evaluating the reportability of the inoperable single train safety system with respect to the 10 CFR 50.72 reporting requirements.|
|05000373/FIN-2012012-01||31 December 2012 23:59:59||LaSalle||Self-revealing||Failure to provide complete and accurate decommissioning status reports||During an NRC investigation completed on November 22, 2011, and a supplemental investigation completed on October 10, 2012, a violation of NRC requirements was identified. In accordance with the NRC Enforcement Policy, the violation is listed below: 10 CFR 50.75(a) establishes requirements for indicating to the NRC how a licensee will provide reasonable assurance that funds will be available for the decommissioning process and states that for power reactor licensees, reasonable assurance consists of a series of steps as provided in paragraphs (b), (c), (e), and (f) of 10 CFR 50.75. 10 CFR 50.75(f)(2) states, in part, that power reactor licensees shall report at least every 2 years on the status of its decommissioning funding for each reactor or part of a reactor that it owns; and, that the information in this report must include, at a minimum, the amount of decommissioning funds estimated to be required pursuant to 10 CFR 50.75(b) and (c). 10 CFR 50.75(b)(1) states, in part, that for a holder of an operating license under 10 CFR Part 50, financial assurance for decommissioning shall be provided in an amount which may be more, but not less, than the amount stated in the table in paragraph (c)(1) adjusted using a rate at least equal to that stated in paragraph (c)(2). 10 CFR 50.75(c)(1) states the minimum amount required to demonstrate reasonable assurance of funds for decommissioning by reactor type and power level. 10 CFR 50.75(c)(2) requires, in part, that an adjustment factor be applied, which is based on escalation factors for labor and energy, and waste burial. 10 CFR 50.9(a) states, in part, that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on March 31, 2005, March 31, 2006, March 31, 2007, and March 31,2009, Exelon Generation Company, LLC (Exelon) provided information on the status of its decommissioning funding that was not complete and accurate in all material respects, when it submitted the decommissioning funding status (DFS) reports pursuant to 10 CFR 50.75. Specifically, the March 31, 2005, March 31, 2007, March 31, 2006, and March 31, 2009, DFS reports stated that the decommissioning funds estimated to be required for each of the reactors, as listed in the report, were determined in accordance with 10 CFR 50.75(b) and the applicable formulas of 10 CFR 50.75(c). However, in multiple instances, the amount reported was a discounted value that was less than the minimum required amount specified by 10 CFR 50.75(b) and (c). This is a Severity Level IV violation.|
|05000373/FIN-2011005-02||31 December 2011 23:59:59||LaSalle||NRC identified||Failure to Perform an Adequate 10 CFR 50.59 Screening for the Use of Racklife Spent Fuel Pool Monitoring Computer Model||A finding of very low safety significance and associated SL-IV NCV of 10 CFR 50.59(c)(2) was identified by the inspectors for the licensees failure to perform an adequate 10 CFR 50.59 screening when evaluating if the implementation of Racklife to monitor Unit 2 spent fuel pool (SFP) rack degradation was a departure from a method of evaluation described in the Updated Final Safety Analysis Report (UFSAR). Specifically, when evaluating in 2005, gif the proposed activity involved the use of an alternative evaluation methodology that is used in establishing the design bases or used in the safety analyses, the licensee dismissed the screening question as not applicable to the circumstances. As a result, the inspectors could not reasonably determine that the changes would not have ultimately required prior NRC approval. The licensee entered this issue into its CAP as AR 1294090. Since the licensee recently completed the installation of neutron absorbing inserts in the entire Unit 2 SFP, as referenced in License Amendment No.186, the use of Racklife to monitor its degradation will no longer be necessary. The inspectors determined that the performance deficiency is greater than minor because it was associated with the Barrier Integrity Cornerstone attribute of configuration controls (reactivity control) and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in IMC 0609, and the finding screened as Green because all the questions in the Barrier Integrity Cornerstone column of IMC 0609s Table 4a were answered no. Because violations of 10 CFR 50.59 can affect the NRCs ability to perform its regulatory function, they are disposition using the traditional enforcement process. The inspectors used the NRCs Enforcement Policy to determine that the violation was a SL-IV violation because the resulting changes were evaluated by the SDP as having very low safety significance. The inspectors did not identify a cross-cutting aspect associated with the underlying finding because the finding was not representative of current performance.|
|05000373/FIN-2011003-01||30 June 2011 23:59:59||LaSalle||NRC identified||Failure to Make a Non-Emergency Event Notification to the NRC and Submit Required LER Following a Loss of Shutdown Cooling Safety Function on Unit 1||A Severity Level IV NCV of 10 CFR 50.72(b)(3)(v) and 10 CFR 50.73(a)(2)(v) was identified by the inspectors for the licensees failure to report an event or condition that could have prevented the fulfillment of the residual heat removal shutdown cooling safety function, which is relied upon to remove residual heat from the reactor. Specifically, when attempting to place the Unit 1 shutdown cooling system in service, the common suction valve unexpectedly closed and caused a complete isolation of the system. The licensee entered this issue into its Corrective Action Program (CAP) as Issue Report (IR) 1244457, and, at the time of this report, was in the process of conducting an apparent cause evaluation to determine the causes of the occurrence and to develop corrective actions. The inspectors determined that the finding should be evaluated using the traditional enforcement process, since the failure to make a required report to the NRC had the potential to impact the agencys ability to perform its regulatory function. The finding was considered to be Severity Level IV, in accordance with the NRC Enforcement Policy. Because this violation did not affect the Reactor Oversight Process cornerstones, a cross-cutting aspect was not assigned.|
|05000373/FIN-2011002-03||31 March 2011 23:59:59||LaSalle||NRC identified||Failure to Make a Required 10 CFR 50.72 Report for an Inoperable Secondary Containment||A Severity Level IV NCV of 10 CFR 50.72(b)(3)(v) was identified by the inspectors for the licensees failure to report an event or condition that could have prevented the fulfillment of the secondary containments safety functions, which are relied upon to control the release of radioactive material. Specifically, the licensee had not properly controlled the opening of two airlock doors that served as a boundary to maintain the ventilation envelope of the reactor building. The licensee entered this issue into its CAP as ARs 1182255 and 1195987, and, at the time of this report, was in the process of conducting an apparent cause evaluation to determine the causes of the occurrence and to develop corrective actions. Violations of 10 CFR 50.72 are considered to be violations that potentially impact the regulatory process and are dispositioned using the traditional enforcement process instead of the Reactor Oversight Process SDP. As such, a cross-cutting aspect was not assigned to this violation.|
|05000373/FIN-2010005-04||31 December 2010 23:59:59||LaSalle||NRC identified||Failure to Design the ISFSI Pad to Adequately Support the Static and Dynamic Loads of Stored Casks||The inspectors identified an NCV of 10 CFR 72.212 (b)(2)(i)(B), Conditions of a General License Issued Under 72.210, for the licensees failure to perform adequate evaluations of the ISFSI pad. Specifically, the inspectors identified five examples where the licensee failed to design the ISFSI pad to adequately support the static and dynamic loads of the stored casks, considering potential amplification of earthquakes through soil-structure interaction. The licensee documented the conditions in IRs 900610, 966506 and 1102633. As an interim corrective action, the licensee provided a technical paper containing justification for partial loading of the pad with 10 casks. Because this violation was related to an ISFSI license, it was dispositioned using the traditional enforcement process in accordance with Section 2.2 of the Enforcement Policy. The inspectors determined that the deficiency was of more than minor significance because, if left uncorrected, a failure of the ISFSI pad could lead to a more significant safety concern. The inspectors determined that the violation could be screened using Section 6.5.d.1 of the NRC Enforcement Policy as a Severity Level IV Violation.|
|05000373/FIN-2010005-05||31 December 2010 23:59:59||LaSalle||NRC identified||Failure to Perform Adequate Evaluations to Ensure Compliance with 10 CFR 72.212(b)(3) and 10 CFR 72.122(b)(i)||The inspectors identified an NCV of 10 CFR 72.146, Design Control, for the licensees failure to perform adequate evaluations to ensure compliance with 10 CFR 72.212(b)(3) and 10 CFR 72.122 (b)(2)(i). Specifically, the inspectors identified that the licensee failed to evaluate that the reactor site parameters including analyses of tornado effects were enveloped by the cask design basis, and perform additional analysis to ensure compliance with 10 CFR 72.122(b)(2)(i). The licensee documented the condition in IR 1137279 and initiated a new calculation to demonstrate compliance. Because this violation was related to an ISFSI license, it was dispositioned using the traditional enforcement process in accordance with Section 2.2 of the Enforcement Policy. The violation was determined to be of more than minor significance because the licensee failed to have an evaluation to assure transfer cask (HI-TRAC) integrity during a tornado event and an additional calculation was required. The licensees new calculation determined that overturning and sliding of the HI-TRAC on the refuel floor would not occur during a tornado. Therefore, the violation screened as having very low safety significance (Severity Level IV).|
|05000373/FIN-2007009-04||30 September 2007 23:59:59||LaSalle||NRC identified||Lake Level Instrumentation Removed from Service without 10 CFR 50.59 Evaluation||The inspectors identified an NCV of 10 CFR Part 50.59, Changes, Tests, and Experiments, which had very low safety significance. Specifically, the licensee failed to complete a 50.59 evaluation for removing main control room lake level instrumentation from service. Although the UFSAR stated that the lake level was continuously monitored in the main control room, the level instrument had not functioned reliably for several years and was removed from the plant maintenance schedule in December 2005. At the time of the inspection, control room monitoring of the lake level was not available. The licensee entered the issue into their corrective action program and initiated more frequent operator rounds as a compensatory measure. The finding was more than minor because the inspectors could not reasonably determine that this change would not have ultimately required prior approval from the NRC. This finding was categorized as Severity Level IV because the underlying technical issue for the finding was determined to be of very low safety significance based on a Phase 1 screening in accordance with IMC 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situation. The inspectors concluded that this finding was cross-cutting in the area of Human Performance, Resources, because the licensee failed to effectively address a long standing equipment issue (H.2(a)).|
|05000373/FIN-2005008-01||30 September 2005 23:59:59||LaSalle||NRC identified||Failure to Perform 10 CFR 50.59 Evaluation for an Adverse Change to the UFSAR|
A finding of very low safety significance was identified by the inspectors associated with a Non-Cited Violation of 10 CFR 50.59, "Changes, Tests, and Experiments," where the licensee failed to complete a full evaluation in accordance with 10 CFR 50.59 for an adverse change to the nitrogen supply header description in the updated final safety analysis report. This issue was entered into the licensee's corrective action system.
This finding was more than minor because the screening was adverse and there was insufficient information to reasonably conclude that prior NRC approval was not necessary. This finding was categorized as Severity Level IV because the underlying technical issue for the finding was determined to be of very low safety significance using the Phase 1 worksheet.