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05000454/FIN-2017009-0130 June 2017 23:59:59ByronNRC identifiedFailure to Perform 10 CFR 50.59 Evaluation for UFSAR ChangeThe inspectors identified a Severity Level IV, Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section(d)(1) and an associated finding of very low safety significance (Green) for the licensees failure to provide a written evaluation which provided the basis for the determination that a change did not require a license amendment. Specifically, the licensee failed to provide a basis for why a change to the surveillance frequencies of emergency diesel generators described in the Updated Final Safety Analysis Report did not require prior NRC approval.The inspectors determined that the performance deficiency was more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The associated finding screened to Green (very low safety significance) because it did not result in the loss of operability or functionality. The diesel generators passed their most recent surveillances. As a result the violation is categorized as Severity Level IV in accordance with section 6.1.d of the NRC Enforcement Policy. The issue did not have a cross-cutting aspect because it was not reflective of current performance.
05000454/FIN-2015008-0330 September 2015 23:59:59ByronNRC identifiedFailure to Evaluate the Adverse Effects of Sharing the RWSTs of Both Reactor UnitsThe team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to perform a written safety evaluation that provided the bases for the determination that a change which resulted in the sharing of the refueling water storage tanks (RWSTs) of both reactor units did not require a license amendment. Specifically, the licensee did not evaluate the adverse effect of reducing reactor unit independence. The licensee captured this issue into their CAP with a proposed action to revise the associated calculation to remove the dependence on the opposite unit, and/or review the implications of crediting the opposite unit RWST under their 10 CFR 50.59 process. The performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, it was associated with the Barrier Integrity cornerstone attribute of design control, and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In addition, the associated traditional enforcement violation was more than minor because the team could not reasonably determine that the changes would not have ultimately required NRC prior approval. The finding screened as very-low safety significance (Green) because it did not result in the loss of operability or functionality, and it did not represent an actual open pathway in the physical integrity of the reactor containment. Specifically, the licensee reviewed the affected calculation and reasonably determined that enough conservatism existed such that adequate net positive suction head (NPSH) could be maintained without sharing the RWSTs of both reactor units. The team did not identify a cross-cutting aspect associated with this finding because it was confirmed not to be reflective of current performance due to the age of the performance deficiency.
05000454/FIN-2015008-0630 September 2015 23:59:59ByronNRC identifiedFailure to Evaluate the Adverse Effects of Changing the SXCT Tornado Analysis as Described in the UFSARThe team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to perform a written evaluation that provided the bases for the determination that the changes to the emergency service water cooling tower (SXCT) tornado analysis as described in the UFSAR did not require a license amendment. Specifically, the associated 10 CFR 50.59 Evaluation did not address the introduction of a new failure mode, the resulting loss of heat removal capacity during worst postulated conditions, and addition of operator actions that have not been demonstrated can be completed within the required time to restore the required SXCT heat removal capacity during worst case conditions. The licensee captured this issue in their CAP with a proposed action to revise the 10 CFR 50.59 Evaluation and submit a Licensee Amendment Request. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, the associated tradition enforcement violation was determined to be more than minor because the team could not reasonably determine that the changes would not have ultimately required prior NRC approval. The finding screened as of very-low safety significance (Green) using a detailed evaluation because a loss of SXCT during a tornado event would degrade one or more trains of a system that supports a risk-significant system or function. The bounding change to the core damage frequency was less than 5.4E-8/year. The team did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance due to the age of the performance deficiency.
05000454/FIN-2015008-0730 September 2015 23:59:59ByronNRC identifiedFailure to Evaluate the Adverse Effects of Changing the SXCT Tornado Analysis as Described in the UFSARThe team identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to perform a written evaluation that provided the bases for the determination that the changes to the emergency service water cooling tower (SXCT) tornado analysis as described in the UFSAR did not require a license amendment. Specifically, the associated 10 CFR 50.59 Evaluation did not address the introduction of a new failure mode, the resulting loss of heat removal capacity during worst postulated conditions, and addition of operator actions that have not been demonstrated can be completed within the required time to restore the required SXCT heat removal capacity during worst case conditions. The licensee captured this issue in their CAP with a proposed action to revise the 10 CFR 50.59 Evaluation and submit a Licensee Amendment Request. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events, and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, the associated tradition enforcement violation was determined to be more than minor because the team could not reasonably determine that the changes would not have ultimately required prior NRC approval. The finding screened as of very-low safety significance (Green) using a detailed evaluation because a loss of SXCT during a tornado event would degrade one or more trains of a system that supports a risk-significant system or function. The bounding change to the core damage frequency was less than 5.4E-8/year. The team did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance due to the age of the performance deficiency.
05000454/FIN-2015001-0231 March 2015 23:59:59ByronNRC identifiedFailure to Report Loss of Emergency Assessment Capability Following Seismic Monitoring System FailureThe inspectors identified a Severity Level IV NCV of 10 CFR 50.72(b)(3)(xiii) when licensee personnel failed to notify the NRC of a major loss of emergency assessment capability within 8 hours of the failure of the onsite seismic monitor used to classify an emergency action level. Upon recognizing that the event was reportable, the licensee made the notification to the NRC and entered the issue into their CAP as IR 2464734. 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 on 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 stated, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. Because a more-than-minor Reactor Oversight Process finding was not identified, there was no cross-cutting aspect associated with this violation.
05000454/FIN-2012012-0131 December 2012 23:59:59ByronNRC identifiedFailure to provide complete and accurate decommissioning status reportsDuring 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.
05000454/FIN-2012005-0331 December 2012 23:59:59ByronNRC identifiedFailure to Submit a 10 CFR 50.73(a)(2)(v) Report for Inoperable Containment Area Radiation MonitorsThe inspectors identified a Severity Level IV NCV of 10 CFR 50.73(a)(2)(v) when licensee personnel failed to report a condition that resulted in a loss of safety function when both containment area radiation monitors were declared inoperable. Specifically, on May 24, 2011, the licensee identified that when reducing reactor power with the isolation setpoints for containment area radiation monitors 1/2AR11J and 1/2AR12J constant and background radiation levels decreasing, the TS setpoint limit for containment area radiation monitors were exceeded and could have prevented the fulfillment of a safety function to automatically isolate containment. The inspectors determined that although this condition represented a loss of safety function in accordance with the 10 CFR 50.73 reporting requirements and NUREG-1022, Event Reporting Guidelines: 10 CFR 50.72 and 10 CFR 50.73, Revision 2, the condition was not reported as required. This issue was entered into the licensees CAP as IR 1463675. Corrective actions included an action to report this event in accordance with NRC requirements. The inspectors determined that the failure to submit a Licensee Event Report (LER) required by 10 CFR 50.73 for a loss of safety function after both containment area radiation monitors were declared inoperable was a performance deficiency. This violation had the potential to impact the regulatory process based, in part, on the generic communications that 10 CFR 50.72 and 10 CFR 50.73 reports serve, the required Reactor Oversight Process (ROP) inspection reviews that the NRC performs on all LERs, and the potential impact on licensee performance assessment. The inspectors determined that this issue was a Severity Level IV violation based on Example 9, The licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73, and Example 10, A failure to identify all applicable reporting codes on a Licensee Event Report that may impact the completeness or accuracy of other information (e.g., performance indicator data) submitted to the NRC, discussed in Section 6.9 of the NRC Enforcement Policy. Because cross-cutting aspects do not apply to traditional enforcement issues, no cross-cutting aspect was assigned to this Severity Level IV violation.
05000454/FIN-2011004-0230 September 2011 23:59:59ByronNRC identifiedModification of the Auxiliary Feedwater System Without Prior NRC ApprovalThe inspectors identified a finding of very low safety significance (Green) and an associated Severity Level IV NCV of 10 CFR 50.59,Changes, Tests, and Experiments, when licensee personnel failed to obtain a license amendment prior to implementing a proposed change to the plant that resulted in a more than minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety previously evaluated in the Updated Final Safety Analysis Report (UFSAR). Specifically, the licensee performed a modification to the facility that permitted the Unit 1 and Unit 2 A Auxiliary Feedwater (AF) trains to be shared between units and the 10 CFR 50.59 evaluation that was performed reached the erroneous conclusion that prior NRC approval was not required. The licensee issued a Standing Order to modify the Emergency Operating Procedure which governed the use of the modification and planned to submit a License Amendment Request (LAR) to the NRC for this design change. The issue was entered into the licensees corrective action program as IR 1257908. The violation was determined to be more than minor because the inspectors determined that the change required prior NRC approval. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated through the SDP to determine the severity of the violation. In this case, the inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone. Specifically, the inspectors answered Yes to Question 1 of the Mitigating Systems Cornerstone column of the Phase 1 worksheet because the inspectors concluded that this was a change confirmed not to result in the loss of operability. Based upon this Phase 1 screening, the inspectors concluded that the issue was of very low safety significance (Green). In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very low safety significance. This finding had a cross-cutting aspect in the Operating Experience component of the Problem Identification and Resolution (PI&R) cross-cutting area (P.2.(b)) because the licensee failed to make adequate use of known industry operating experience in the screening of a modification prior to installation.
05000454/FIN-2011013-0130 June 2011 23:59:59ByronNRC identifiedRestoring Compliance With Respect to Single FailuresOn January 19, 2011, the NRC issued IR 05000454/2011010; 05000455/2011010, and notified the licensee of the agencys decision to issue a compliance backfit in order to address the technical issue. The report listed the licensees initial corrective actions and requested the licensee to provide a written response within 30 days of their assessment of the issue and a description of their intended actions to address the non-compliance, including a proposed schedule to complete those actions and an assessment of the extent of condition of this issue The licensee responded by letter dated February 18, 2011, and committed to the following 1. The power supplies to the Steam Generators PORVs will be modified with a safety-related battery backup. The committed date for performing this action was no later than Unit 1s October 2012 refueling outage and Unit 2s May 2013 refueling outage 2. The licensee will issue a supplement to their February 18, 2011, response letter, in order to communicate any revisions to the modification installation schedule based on the online/outage determination (i.e., whether the modification could be installed online or require an outage). The committed date for performing this action is October 14, 2011 3. An extent of condition review will be conducted of other transients and accidents outlined in Chapter 15 of the Byron Station Updated Final Safety Analysis Report to identify similar discrepancies with respect to the inappropriate reliance or assumption of single active failure. The identified discrepancies, if any, would be resolved within the Corrective Action Program and communicated to the NRC Region III Regional Administrator. The committed date for performing this action is August 4, 2011 The NRC staff considered these actions and timeframe adequate for complying with the agencys requirements In accordance with the Reactor Oversight Process (ROP) Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, the staff determined that this issue did not meet the definition for a performance deficiency since it was not reasonably within the licensees ability to foresee and correct. In accordance with Inspection Manual Chapter 0612, the inspectors determined whether this issue of concern involved a more than minor violation. In order to assess the significance of the underlying technical issue associated with the violation, the inspectors used the guidance in IMC 0612, Appendix B, Issue Screening, and IMC 0609 Significance Determination Process. Based on these, the underlying technical issue was a Non-Finding (i.e., no performance deficiency), was considered a more-than-minor violation, and from a risk perspective, screened as having very low safety significance. The inspectors determined that no example from either Section 6.1 or 6.9 of the Enforcement Policy adequately applied or described the situation. Therefore, Regional NRC management considered the risk significance of the underlying technical issue and, in consultation with the Office of Enforcement, determined the issue was best represented as a Severity Level IV violation This violation is not considered a finding; therefore, in accordance with IMC 0305, no cross-cutting aspect is assigned to the violation.
05000454/FIN-2010005-0231 December 2010 23:59:59ByronNRC identifiedInadequate Procedures for Implementing FSAR Required Annulus CoolingThe inspectors identified a Severity Level IV NCV of very low safety significance of 10 CFR 72.150, Instruction, Procedures, and Drawings. Specifically, the licensee failed to have procedures in place to ensure that the design basis peak fuel cladding limit would not be exceeded during canister loading operations. The licensee entered this issue into their corrective action program and revised the procedure to provide monitoring criteria. The violation was determined to be of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, Example 2c, in that the procedures failed to incorporate thermal acceptance criteria established by the Holtec Final Safety Analysis Report and that the failure to incorporate thermal acceptance criteria was repetitive. Although the violation contributed to the likelihood of peak fuel cladding temperatures exceeding the safety limit, subsequent analysis by the licensee determined that fuel cladding temperature limits were not exceeded. The violation screened as having very low safety significance.
05000454/FIN-2010005-0331 December 2010 23:59:59ByronNRC identifiedInadequate Procedural Guidance for Heavy Loads OperationsThe inspectors identified a Severity Level IV NCV of very low safety significance of 10 CFR 72.150, Instructions, Procedures, and Drawings. Specifically, the licensee failed to have procedures in place to ensure that heavy loads were operated safely in the Fuel Handling Building. The licensee entered this issue into their corrective action program and revised the procedure to provide monitoring criteria. The violation was determined to be of more than minor significance because if left uncorrected, it could lead to a more significant safety concern. Consistent with the guidance in Section 2.6.D of the NRC Enforcement Manual, if a violation does not fit an example in the Enforcement Policy Violation Examples, it should be assigned a severity level: (1) commensurate with its safety significance; and (2) informed by similar violations addressed in the Violation Examples. The violation screened as having very low safety significance.
05000454/FIN-2010007-0130 September 2010 23:59:59ByronNRC identifiedInadequate Procedures for Implementing FSAR Required Annulus CoolingThe inspectors identified a Severity Level IV NCV of 10 CFR 72.150, Instructions, Procedures, and Drawings. Specifically, the licensee failed to have procedures in place to ensure that the design basis peak fuel cladding temperature limit would not be exceeded during vacuum drying operations. The licensee entered this issue into its corrective action program and revised the procedure to provide monitoring criteria. The violation was determined to be of more than minor significance because, if left uncorrected, it could lead to a more safety significant event. Although the violation contributed to the likelihood of peak fuel cladding temperatures exceeding the safety limit, subsequent analysis by the licensee and the NRC determined that fuel cladding temperature limits were not exceeded during this event; therefore, the violation screened as having very low safety significance.
05000454/FIN-2010003-0630 June 2010 23:59:59ByronNRC identifiedFailure to Report an Automatic RPS and Auxiliary Feedwater Actuation While Shut DownA Severity Level IV, NCV of 10 CFR 50.72(b)(3)(iv)(A) was identified by the inspectors for the licensees failure to recognize that a valid Unit 2 automatic Reactor Protection System (RPS) and Auxiliary Feedwater (AF) actuation while shut down were reportable conditions. Consequently, the licensee failed to make an 8 hour report as required by 10 CFR 50.72. This issue was documented in the licensees CAP as IR 1060177 and the licensee subsequently reported the event. This finding was evaluated under Traditional Enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. However, this violation was of very low safety significance because immediate NRC follow-up action was not required. The NRC has characterized this violation as a Severity Level IV NCV in accordance with Section IV.A.3 and Supplement 1 of the NRC Enforcement Policy. The cause of this finding was directly related to the cross-cutting area of Problem Identification and Resolution (P.1(c)) because the licensee did not thoroughly evaluate and classify a condition adverse to quality for reportability.
05000454/FIN-2008009-0131 December 2008 23:59:59ByronNRC identifiedFailure to Update the Boron Recycle and RHR System Descriptions in the UFSARThe inspectors identified a Severity Level IV Non-Cited Violation (NCV), having very low safety significance of 10 CFR 50.71, Maintenance of Records, Making of Reports, for the licensees failure to adequately update the Byron Station Updated Final Safety Analysis Report. Specifically, the description of: (1) the boron recycle system did not identify if the system was designed or capable of handling discharges from the safety injection and residual heat removal relief valves; (2) the residual heat removal system did not identify deviations from the system design standard with respect to the suction pipe relief valve single failure analysis and collection of relief valve discharges outside containment. The licensee entered this issue into the corrective action system. Because this finding affected the NRCs ability to perform its regulatory function, this issue was evaluated using the traditional enforcement process. The finding was determined to be more than minor because the inspectors could not reasonably determine that a change to correct the Final Safety Analysis Report to reflect actual design would not have ultimately required NRC prior approval. The finding was determined to be of very low safety significance because the design deviations associated with the residual heat removal system and boron recycle system did not impact system operability. The inspectors determined that the finding did not have a cross-cutting aspect. (Section 4OA2.b.1
05000454/FIN-2008006-0131 March 2008 23:59:59ByronNRC identifiedFailure to Perform 10 CFR 50.59 Evaluations for Changes in Assumed Operator TimesThe inspectors identified a Severity Level IV NCV, having very low safety significance, of 10 CFR 50.59, Changes, Tests, and Experiments, from the licensees failure to provide a documented basis for determining that changes in how operator response times for postulated steam generator tube ruptures were credited in accident analyses did not require prior NRC approval. This issue was entered into the licensees corrective action program. Because the issue affected the NRCs ability to perform its regulatory function, this issue was evaluated using the traditional enforcement process. The finding was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The finding was determined to be of very low safety significance because there was evidence that operator actions could be performed in time to prevent overfill of a steam generator. This finding had a cross-cutting aspect in the area of Human Performance because the licensee failed to ensure that training was adequate to assure nuclear safety (H.2(b)) in that training for 10 CFR 50.59 safety evaluations and screenings was deficient.
05000454/FIN-2007004-0230 September 2007 23:59:59ByronNRC identifiedInadequate Basis in 10 CFR 50.59 Evaluation Associated with a Special Test ProcedureThe inspectors identified a NCV of 10 CFR 50.59(d)(1) for the licensees failure to document an evaluation that provided a basis for the determination that the change, test, or experiment did not require a license amendment. Specifically, for Special Test Procedure SPP-07-002, Test of 1B DG Voltage Regulator Following Maintenance Via SX Pump Start, the licensee failed to provide an evaluation as to why disconnecting the offsite electrical power feed to the emergency bus during power operation with an inoperable diesel generator did not present 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 Updated Final Safety Analysis Report. The licensee entered the appropriate limiting condition of operation for the offsite power circuit during the test, entered this issue into the corrective action program, and initiated actions to complete a 10 CFR 50.59 evaluation to determine if these procedure changes were acceptable without a license amendment. The primary cause of this issue was related to the cross-cutting area of Human Performance for failure to use conservative assumptions in decision making and to adopt a requirement that demonstrates the proposed action is safe in order to proceed (H.1(b)). Because the issue potentially impacted the NRCs ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be more than minor because the inspectors could not reasonably determine that the special test procedure, that affected the Updated Final Safety Analysis Report described design function of equipment important to safety, would not have ultimately required NRC prior approval. Based upon the Phase 1 screening, the inspectors concluded that the issue was of very low safety significance
05000454/FIN-2005011-0331 December 2005 23:59:59ByronNRC identifiedFailure to Obtain Prior NRC Approval for Change to an EAL That Decreased the Effectiveness of the Emergency PlanThe inspectors identified that the licensee had changed its standard emergency action level (EAL) scheme by revising one EALs criteria for an Unusual Event declaration that addressed an unplanned radiological release in excess of effluent radiation monitor readings unless the release could be determined to be below Offsite Dose Calculation Manual limits within 15 minutes for releases that could not be terminated in 60 minutes or less. The inspectors determined that this EAL change decreased the effectiveness of the emergency plan, and that the licensee did not obtain prior NRC approval for this change, contrary to the requirements of 10 CFR 50.54(q). The licensee is evaluating the options to correct the EAL This finding was more than minor because extending the time period required for the appropriate emergency classification of a radiological release could adversely affect the performance of both onsite and offsite emergency actions. Because the issue affected the NRCs ability to perform its regulatory function, it was evaluated with the traditional enforcement process as specified in Section IV.A.3 of the Enforcement Policy. According to Supplement VIII of the Enforcement Policy, this finding was determined to be a Severity Level IV because it involved a failure to meet a requirement not directly related to assessment and notification. Further, this problem was isolated to one EAL and was not indicative of a functional problem with the EAL scheme. Additionally, because the violation was a Severity Level IV and the licensee entered this issue into its corrective action program this finding is being treated as a Severity Level IV Non-Cited Violation of 10 CFR 50.54(q).
05000454/FIN-2003007-0231 December 2003 23:59:59ByronSelf-revealingFailure to Update the Updated Final Safety Analysis Report in a Timely Manner.

A finding of very low safety significance was self-revealed when the licensee discovered that an update to the Updated Final Safety Analysis Report was not accomplished for a period of almost 6 years following a design change. Between June and September of 1996, the licensee made a revision to the reactor water storage tank level set-point calculation to clarify design basis information with respect to emergency core cooling system and containment spray system operation and re-evaluated the time available to complete switchover to recirculation. The licensee did not include this update until the December 2002 revision to the Updated Final Safety Analysis Report.

Because this issue potentially impacted the NRC's ability to perform its regulatory function, this finding was evaluated using the traditional enforcement process. The finding was determined to be of very low safety signficance because it did not actually impede or influence any regulatory actions. This was determined to be a Severity Level IV NCV of 10 CFR 50.71.