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05000456/FIN-2013005-0131 December 2013 23:59:59BraidwoodNRC identifiedFailure to Maintain Accurate Operator LogsThe inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a), Completeness and Accuracy of Information, when licensee personnel failed to provide complete and accurate operator logs of record. Specifically, operator log entries of record on May 9, 2013, did not accurately document entry into and exit from Limiting Condition for Operation (LCO) 3.0.3. Initial corrective actions included additional late log entries and issuance of Operations Standing Order 13-10, Corrections to Electronic Log Entries, which provided interim guidance to operators regarding how to make revisions to electronic log entries. The Operations Director also initiated discussions with the fleet Operations Director peer group to determine how to incorporate guidance on revising electronic logs into procedure OP-AA-111-101, Operating Narrative Logs and Records. The licensee entered this issue into their Corrective Action Program (CAP) as Issue Report (IR) 1519660, Lack of Details in Log Entries. In consultation with regional enforcement staff, the inspectors determined that the issue was more than minor because operator logs of record are material documents to the NRC, in that inspection activities are planned and conducted based, in part, on the review of operator logs and the presumption of their accuracy. In determining the significance of the violation, the inspectors referenced the examples of violations in Section 6.9, Inaccurate and Incomplete Information or Failure to a Make a Required Report, of the NRC Enforcement Policy. Because the issue was determined to be more than minor, but did not meet the threshold of the examples of Severity Level I, II, or III violations, the inspectors determined this issue was a Severity Level IV violation. Because a more-than-minor Reactor Oversight Process finding was not identified, there was no cross-cutting aspect associated with this violation. (Section 1R22.2.b)
05000456/FIN-2013005-0331 December 2013 23:59:59BraidwoodNRC identifiedFailure to Submit Report Required by 10 cfr 50.72(b)(3)(xiii)The inspectors identified a Severity Level IV NCV of 10 CFR 50.72(b)(3)(xiii) when licensee personnel failed to submit a report required by 10 CFR 50.72 for a loss of emergency assessment capability when an unplanned degradation was identified associated with the Technical Support Center (TSC) ventilation filtered make-up train. Specifically, the discharge damper for the TSC ventilation filtered make-up fan was found unexpectedly closed, which adversely impacted the ability to supply filtered air to the TSC absent implementation of compensatory actions. Corrective actions included making the required Event Report on January 14, 2014. The licensee entered this issue into their CAP as IR 1598598, Wording Differences Between NUREG-1022 and Reportability Manual, and IR 1608133, ENS (Event Notification System) Call Made Due to TSC Ventilation Impact in October 2013. 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 Example 6.d.9 in the NRC Enforcement Policy. Example 6.d.9 specifically stated, The licensee fails to make a report requirement 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. (Section 4OA2.2b)
05000456/FIN-2013008-0131 December 2013 23:59:59BraidwoodNRC identifiedInaccurate/Incomplete Information For Exemption Request From 10 CFR 50.60Title 10 of the Code of Federal Regulations (10 CFR) 50.9(a), Completeness and Accuracy of Information, requires that Information provided to the Commission by an applicant for a license or by a licensee or information required by statute, or by the Commission\'s regulations, orders, or license conditions to be maintained by the applicant or the licensee shall be complete and accurate in all material respects. In Letter No. RS-05-103, License Amendment Request Regarding Reactor Coolant System Pressure, and Temperature Limits Report and Request for Exemption from 10 CFR 50.60, Acceptance Criteria for Fracture Prevention Measures for Lightwater Nuclear Power Reactors for Normal Operation, Attachment 4, Justification for Exemption from 10 CFR 50.60, the licensee (Exelon Generation Company, LLC) stated WCAP-16143 provides a valid basis for changing the RPV (Reactor Pressure Vessel) closure head flange limit and maintains the relative margin of safety commensurate with that which existed at the time the 10 CFR (Part) 50, Appendix G requirement was issued. Contrary to the above, on October 3, 2005, in Letter No. RS-05-103, the licensee (Exelon Generation Company, LLC) failed to provide information to the Commission that was complete and accurate in all material respects, in that, WCAP-16143, Reactor Vessel Closure Had/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2, did not provide a valid basis for changing the RPV closure head flange limit for Braidwood Unit 2. Specifically, WCAP-16143, Section 4, Flange Integrity, demonstrated adequate vessel margins based upon the original closure head flange configuration and did not represent the modified closure head configuration (53 head studs applicable to the Unit 2 reactor vessel). Operation of the Braidwood Unit 2 vessel with 53 closure head studs was not within the bounds and limitations of what the NRC had reviewed in Letter No. RS-05-103 and found to be an acceptable basis to grant the exemption request. Therefore, this information was considered material to the NRC.
05000456/FIN-2013004-0130 September 2013 23:59:59BraidwoodNRC identifiedFailure to Perform a Required 10 CFR 50.59 EvaluationThe inspectors identified a finding of very low safety significance and an associated Severity Level IV NCV of 10 CFR 50.59, Changes, Tests, and Experiments, when licensee personnel failed to perform and maintain a written evaluation to demonstrate that a procedure change did not require a license amendment. Specifically, the licensee implemented a change to procedures 1/2BwOA SEC-4, Loss of Instrument Air, Revision 3, that revised the actions to address a loss of component cooling water (CC) to the reactor coolant pump (RCP) thermal barrier heat exchange such that a complete loss of seal cooling could occur, which would result in damage to the RCP seals and a subsequent loss of coolant accident (LOCA). As part of the licensee corrective actions, procedures 1/2 BwOA SEC-4 were revised to address the issue. A revised 10 CFR 50.59 evaluation was also developed and approved. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it could be reasonably viewed as a precursor to a significant event. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 2, for the Initiating Events cornerstone. The inspectors then answered No to all of the screening questions in Table 3. The finding was further evaluated using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1. The inspectors answered No to all of the questions contained therein. Therefore, the inspectors concluded the finding was of very low safety significance (Green). Because the associated finding was determined to be of very low safety significance in accordance with the SDP, the traditional enforcement aspect of this issue was determined to be at the Severity Level IV level. The inspectors did not identify a cross-cutting aspect associated with this finding since it was not indicative of current performance.
05000456/FIN-2013002-0131 March 2013 23:59:59BraidwoodNRC identifiedFailure to Perform an Adequate 10 CFR 50.59 Evaluation Removing the Positive Displacement Pump from the Current Licensing BasisThe inspectors identified a finding of very low safety significance (Green) and an associated Severity Level IV NCV of 10 CFR 50.59 when licensee personnel failed to perform an adequate 10 CFR 50.59 safety evaluation that revised the Updated Final Safety Analysis Report (UFSAR) to permit the Chemical Volume Control System (CVCS) positive displacement pump (PDP) to be isolated and removed from service for an extended, but undefined, period of time. The licensee entered this issue into their Corrective Action Program (CAP) as Issue Report (IR) 1477923. As part of their corrective actions, the licensee planned to re-perform the 10 CFR 50.59 evaluation to include a review of the direct effects that this change had on the CVCS PDP functions that were important to safety. The finding was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, in 1997, the licensee failed to evaluate whether there was an increase in the probability of a malfunction for the PDP functions important to safety prior to isolating and removing the PDPs from service. The finding was evaluated using IMC 0609, Significance Determination Process. Using Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered No to Questions 1, 2, 3 and 4 and, as a result, determined the finding was of very low safety significance (Green). The finding was also determined to be a Severity Level IV NCV in accordance with Section 6.1.d.2 of the NRC Enforcement Policy because the resulting changes were evaluated by the SDP as having very low safety significance (Green). There was no cross-cutting aspect associated with the finding because it was not indicative of current licensee performance.
05000456/FIN-2012012-0131 December 2012 23:59:59BraidwoodNRC 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.
05000456/FIN-2012004-0530 September 2012 23:59:59BraidwoodNRC identifiedFailure to Submit a 10 CFR 50.72(b)(3)(v) and a 10 CFR 50.73(a)(2)(v) Report; Inoperable UHSThe inspectors identified a Severity Level IV NCV of 10 CFR 50.72(b)(3)(v) and 10 CFR 50.73(a)(2)(v) when licensee personnel failed to report a condition that resulted in a loss of safety function after the UHS was declared inoperable after exceeding the TS limit of 100 degrees Fahrenheit (F). Specifically, on July 7, 2012, the licensee had identified and entered TS 3.7.9, Ultimate Heat Sink, Condition (A), Ultimate Heat Sink Inoperable, after the UHS lake temperature exceeded the TS 3.7.9.2 Surveillance Requirement value of less than or equal to 100F. The inspectors determined that although this condition represented a loss of safety function in accordance with the 10 CFR 50.72 and 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 1422296. Corrective actions included an action to report this event in accordance with NRC requirements. The inspectors determined that the failure to submit a report required by 10 CFR 50.72 and a Licensee Event Report (LER) required by 10 CFR 50.73 for a loss of safety function after the UHS was declared inoperable on July 7, 2012, 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 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 similar examples referenced in Section 6.9 of the NRC Enforcement Policy. Specifically, 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. Because cross-cutting aspects do not apply to traditional enforcement issues, no cross-cutting aspect was assigned.
05000456/FIN-2011001-0231 December 2011 23:59:59BraidwoodNRC identifiedFailure to Follow Procedures to Ensure Multi Purpose Canister Design Basis Pressure is Not ExceededThe inspectors identified a Severity Level IV NCV of 10 CFR 72.150, Instructions, Procedures, and Drawings, when licensee personnel failed to adhere to procedures to ensure that the design pressure limit for the multi-purpose canister (MPC) would not be exceeded during loading operations. The licensee entered this issue into their CAP as IR 01279837, IR 01286670, and IR 01285354. The licensee imposed an ISFSI stand-down to reinforce and correct procedure use and adherence as a corrective action to restore compliance. The issue was determined to be of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, Example 2h, in that multiple examples of a failure to follow procedures were identified. Although the violation contributed to the likelihood of the canister design pressure being exceeded, it was verified that the canister was within its design pressure. Therefore, the inspectors determined that the issue represented a Severity Level IV violation. Reactor Oversight Process cross-cutting aspects do not apply to TE issues or licensee-identified ROP findings of very low safety significance, therefore, none was identified
05000456/FIN-2011005-0831 December 2011 23:59:59BraidwoodNRC identifiedFailure to Follow Procedures Adequate Evalations to Facilitate Independent Spent Fuel Storage Installation ActivitiesThe inspectors identified a Severity Level IV NCV of 10 CFR 72.150, Instructions, Procedures, and Drawings, when licensee personnel failed to adhere to procedures to ensure that the design pressure limit for the multi-purpose canister (MPC) would not be exceeded during loading operations. The licensee entered this issue into their CAP as IR 01279837, IR 01286670, and IR 01285354. The licensee imposed an ISFSI stand-down to reinforce and correct procedure use and adherence as a corrective action to restore compliance. The issue was determined to be of more than minor significance using IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, Example 2h, in that multiple examples of a failure to follow procedures were identified. Although the violation contributed to the likelihood of the canister design pressure being exceeded, it was verified that the canister was within its design pressure. Therefore, the inspectors determined that the issue represented a Severity Level IV violation. Reactor Oversight Process cross-cutting aspects do not apply to TE issues or licensee-identified ROP findings of very low safety significance, therefore, none was identified
05000456/FIN-2011005-0331 December 2011 23:59:59BraidwoodNRC identifiedFailure to Submit Licensee Event Report Per 10 CFR 50.73(a)(2)(vii)A Severity Level IV NCV of 10 CFR 50.73(a)(2)(vii) was identified by the inspectors when the licensee failed to submit a LER within 60 days after identifying instances in which a single cause (or condition) would cause two safety-related instrument channels to become inoperable in a single system designed to shutdown the reactor and maintain it in a safe shutdown condition. On March 8, 2011, the licensee identified a non-conservative assumption used in the stations Turbine Building High Energy Line Break (HELB) analysis of record in that the calculations were not updated for a historic power uprate. On March 14, 2011, the station completed an Operability Evaluation that examined how a HELB in the Turbine Building could affect safety-related equipment since many safety-related rooms had ventilation connections to the Turbine Building and shared a barrier wall with the Turbine Building. One of the errors identified by the licensee was associated with non-conservative HELB temperature and pressure parameters used in determining break flows. The licensee concluded that the higher break flows did not affect the operability of the safety-related rooms in a normal configuration, but concluded that the presence of open equipment rollup doors for the ESF Switchgear Rooms and MEERs could cause unacceptable temperatures. Therefore, the station instituted compensatory actions to control these rollup doors shut with equipment status tags when operating in Modes 1-4. On April 7, 2011, the inspectors identified that the station did not have any plans to review 10 CFR 50.73 reportability requirements regarding this issue. The inspectors notified the licensee that they had personally observed large breakers being moved through the rollup doors within the past 3 years. Based on these observations and conclusions reached in the operability determination, the inspectors specifically questioned if these conditions required the submittal of an LER. The station entered the inspectors observations into the CAP and created an assignment to review past reportability in IR 1199223. This LER evaluation assignment had an original due date of May 9, 2011, but the due date was later changed to July 29, 2011, based upon an on-going review to update the HELB model. The July 29, 2011 due date was also later changed to October 31, 2011 because the HELB model required substantially more work that originally believed. On October 31, 2011, the licensee completed the CAP assignment and concluded that with the Division 11 (or Division 21) Miscellaneous Electric Equipment Room (MEER) rollup door(s) open and the unit(s) operating in Modes 1-4, there was reasonable doubt that the safety-related instrument inverters inside of these rooms would have remained functional in the event of a concurrent HELB on the 451 elevation of the Turbine Building. This would have impacted operation of Instrument Bus 111 and 113 (or Instrument Bus 211 and 213). Based on this conclusion, and a formal review of when the associated rollup doors had been open within the past 3 years, the licensee concluded that four events were reportable under 10 CFR 50.73(a)(2)(vii). These events represented instances in which a Turbine Building HELB would have caused two safety-related instrument channels to become inoperable in a single system designed to shutdown the reactor and maintain it in a safe shutdown condition. Furthermore, the licensee concluded that the October 31, 2011, discovery marked the start of the 60-day reporting requirement specified in the regulations. The inspectors reviewed the reportability aspects associated with this issue from the March 8, 2011, date when the issue was first discovered and treated as an adverse condition affecting the operability of equipment through the time the LER was submitted to the NRC on December 22, 2011. The inspectors reviewed the NRCs event reporting guidelines contained in NUREG-1022, Revision 2, Event Reporting Guidelines 10 CFR 50.72 and 10 CFR 50.73, and discussed the report timeliness with NRC Office of Nuclear Reactor Regulation (NRR) experts. The inspectors concluded that the discovery date should have started when the licensee lost reasonable expectation that the equipment in question was not operable with the rollup doors open and the licensee understood that the doors had been opened within the past 3 years. Based on discussions with station staff, the licensee should have reported this event within 60 days of March 14, 2011. The licensee entered this issue into their CAP as IR 1299906. The inspectors determined that the failure to report this LER in accordance with NRC regulations was a performance deficiency. Specifically, the licensee should have created a CAP assignment to review the 10 CFR 50.73 reportability aspects of this issue without prompting from the inspectors and should have reported the issue in a timely manner. This violation had the potential to impact the regulatory process based upon the generic communication that LERs serve, the required Reactor Oversight Process (ROP) reviews that the NRC perform on all LERs, and the potential impact on licensee performance assessment. Since the issue impacted the regulatory process, it was dispositioned through the TS process. The inspectors determined that this issue was a Severity Level IV violation based on a similar example referenced in Supplement I, Example D.4, of the NRC Enforcement Policy. The inspectors evaluated the actual non-conforming technical condition through the ROP. The inspectors determined that the issue was licensee-identified and a violation of 10 CFR 50, Appendix B, Criterion III, Design Control. This issue is described in Section 4OA7 of this report. Reactor Oversight Process cross-cutting aspects do not apply to traditional enforcement issues or licensee-identified ROP findings of very low safety significance, therefore, none was identified. : Title 10 CFR 50.73(a), Reportable Events, required, in part, that The holder of an operating license under this part or a combined license under Part 52 of this chapter (after the Commission has made the finding under 52.103(g) of this chapter) for a nuclear power plant (licensee) shall submit a LER for any event of the type described in this paragraph within 60 days after the discovery of the event. In addition this section of the code requires that, Unless otherwise specified in this section, the licensee shall report an event if it occurred within 3 years of the date of discovery regardless of the plant mode or power level, and regardless of the significance of the structure, system, or component that initiated the event. Title 10 CFR 50.73(a)(2)(vii) described event(s) where a single cause or condition caused at least one independent train or channel to become inoperable in multiple systems or two independent trains or channels to become inoperable in a single system designed to: (A) Shutdown the reactor and maintain it in a safe shutdown condition; (B) Remove residual heat; (C) Control the releases of radioactive material; or (D) Mitigate the consequences of an accident. Contrary to the above, the licensee failed to report two Unit 1 and two Unit 2 conditions in which a Turbine Building HELB would have rendered two independent safety-related instrument channels inoperable in a single system designed to safely shutdown the reactor and maintain it in a safe shutdown condition within 60 days from the date when the condition was discovered. This information was known or available since March 14, 2011, but was not reported until December 22, 2011. Corrective actions included submitting an LER to the NRC on December 22, 2011. Because this violation was entered into the licensees CAP as IR 1299906, it is being treated as a Severity Level IV NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. Corrective actions included the issuance of LER 05000456/2011-004-00 on December 22, 2011.
05000456/FIN-2011001-0131 December 2011 23:59:59BraidwoodNRC identifiedFailure to Perform Adequate Evaluations to Facilitate Independent Spent Fuel Storage Installation ActivitiesThe inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (CFR) 72.146, Design Control, when licensee personnel failed to perform adequate evaluations of the Independent Spent Fuel Storage Installation (ISFSI) pad, ISFSI components, and the effects of ISFSI loading operations on the operating plant. Specifically, the inspectors identified three examples in which licensee personnel failed to perform adequate evaluations to ensure compliance with 10 CFR 72.212(b)(5)(ii); 10 CFR 72.212(b)(8); and the Safety Analysis Report (SAR) referenced in the Holtec Certificate of Compliance (CoC). The licensee entered these issues into their Corrective Action Program (CAP) as Issue Report (IR) 1280650, IR 1278520, and IR 1245756. Corrective actions included revisions to calculations. Because this violation was related to an ISFSI license, it was dispositioned using the Traditional Enforcement (TE) process in accordance with Section 2.2 of the Enforcement Policy. The inspectors determined that the deficiency was of more than minor significance because the licensees evaluation did not assure structural integrity of the affected components under the design basis loads, and required extensive revisions to the calculations. The inspectors determined that the issue represented a Severity Level IV violation. Reactor Oversight Process (ROP) cross-cutting aspects do not apply to TE issues or licensee-identified ROP findings of very low safety significance, therefore, none was identified
05000456/FIN-2011005-0731 December 2011 23:59:59BraidwoodNRC identifiedFailure to Perform Adequate Evaluations to Facilitate Independent Spent Fuel Storage Installation ActivitiesThe inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (CFR) 72.146, Design Control, when licensee personnel failed to perform adequate evaluations of the Independent Spent Fuel Storage Installation (ISFSI) pad, ISFSI components, and the effects of ISFSI loading operations on the operating plant. Specifically, the inspectors identified three examples in which licensee personnel failed to perform adequate evaluations to ensure compliance with 10 CFR 72.212(b)(5)(ii); 10 CFR 72.212(b)(8); and the Safety Analysis Report (SAR) referenced in the Holtec Certificate of Compliance (CoC). The licensee entered these issues into their Corrective Action Program (CAP) as Issue Report (IR) 1280650, IR 1278520, and IR 1245756. Corrective actions included revisions to calculations. Because this violation was related to an ISFSI license, it was dispositioned using the Traditional Enforcement (TE) process in accordance with Section 2.2 of the Enforcement Policy. The inspectors determined that the deficiency was of more than minor significance because the licensees evaluation did not assure structural integrity of the affected components under the design basis loads, and required extensive revisions to the calculations. The inspectors determined that the issue represented a Severity Level IV violation. Reactor Oversight Process (ROP) cross-cutting aspects do not apply to TE issues or licensee-identified ROP findings of very low safety significance, therefore, none was identified
05000456/FIN-2011004-0630 September 2011 23:59:59BraidwoodNRC identifiedModification of the Auxiliary Feedwater System Without Prior NRC ApprovalThe inspectors identified a finding of very low safety significance 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 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 entered this issue into the corrective action program as IR 1258017 and planned to submit a License Amendment Request (LAR) to the NRC for this design change. 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.
05000456/FIN-2011010-0130 June 2011 23:59:59BraidwoodNRC identifiedRestoring Compliance With Respect to Single FailuresTitle 10 CFR Part 50, Appendix B, Criterion III, Design Control, states, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions Title 10 CFR Part 50 Appendix A defines single failure as A single failure means an occurrence which results in the loss of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occurrence are considered to be a single failure. Fluid and electric systems are considered to be designed against an assumed single failure if neither: (1) a single failure of any active component (assuming passive components function properly); nor (2) a single failure of a passive component (assuming active components function properly), results in a loss of the capability of the system to perform its safety functions Contrary to the above, from February 1, 2011, the date when the licensee was informed of the issuance of a compliance backfit, until March 2, 2011, the date the licensee committed to the NRC, via letter, to restore compliance, the licensee failed to ensure the SG PORVs power supplies met the design bases. Specifically, the licensee failed to ensure the SG PORVs were capable of performing their safety function assuming a single failure as defined by 10 CFR Part 50 Appendix A General Design Criteria for Nuclear Power Plants in their SGTR analysis The NRC staff determined that this violation resulted from matters not reasonably within the licensees control; that is, the requirements could not be readily identified and therefore addressed. Enforcement Policy Section 3.5, Violations Involving Special Circumstances, states in part, the NRC may reduce or refrain from issuing a civil penalty or an NOV (Notice of Violation) for a Severity Level II, III, or IV violation based on the merits of the case after considering the guidance in this statement of policy and such factors as the age of the violation, the significance of the violation, the clarity of the requirement and associated guidance... In this case, the lack of clarity of the requirement influenced the licensees ability to comply. As stated above, an exemption to the backfit rule (a compliance backfit) was documented. Therefore, in accordance with the Enforcement Policy, and after consultation with the Director of the Office of Enforcement and the Region III Regional Administrator, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and to refrain from issuing enforcement action for the violation. In accordance with the NRCs Reactor Oversight Process, this condition will not be considered in the assessment process or the NRCs Action Matrix. The technical issue is considered open pending completion of the corrective actions.
05000456/FIN-2011002-0131 March 2011 23:59:59BraidwoodNRC identifiedFailure to Provide Complete and Accurate Information in an LERA Severity Level IV NCV of 10 CFR 50.9, Completeness and Accuracy of Information, was identified by the inspectors when licensee personnel failed to provide information to the NRC that was complete and accurate in all material respects in Licensee Event Report (LER) 05000457/2010-004-00, Unplanned Limiting Condition for Operation Entry Due to Low Header Pressure on the 2B Essential Service Water Pump. Specifically, the LER stated that an analysis had determined that both units and trains of the Essential Service Water (SX) system were capable of mitigating the effects of design basis events. However, the referenced analysis had not been performed at the time the LER was submitted. The inspectors determined that this issue was a Severity Level IV violation based on a similar example referenced in Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, of the NRC Enforcement Policy. In particular, example d.1 identified that a licensee failing to make a required report which, had it been submitted, would have resulted in, for instance, increasing the scope of the next regularly scheduled inspection, was a Severity Level IV violation. The inaccurate information was considered to be material to the NRC because it potentially affected an NRC assessment of whether a loss of safety function occurred and whether it should have been reported to the NRC. This issue was entered into the licenseei12s corrective action program and corrective actions included the station performing the analysis referenced in the LER. The inspectors had previously reviewed the Reactor Oversight Process (ROP) aspect of this finding and a self-revealed NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was documented in Section 1R22 of NRC Integrated Inspection Report 05000456/2010004; 05000457/2010004 for this issue.
05000456/FIN-2010503-0131 March 2011 23:59:59BraidwoodNRC identified(Traditional Enforcement) Changes to EAL basis Decreases the Effectiveness of the Plan without Prior NRC Approval (Tradiional EnforcementA Green finding involving a Severity Level IV, Cited Violation of 10 CFR 50.54(q) was identified by the inspector for the licensees change to the emergency plan which decreased the effectiveness of the plan without NRC approval. Specifically, the licensee modified the Emergency Action Level (EAL) Basis in EAL HU6, Revision 21, to delay the 15-minute classification time by the dispatching of personnel, reporting the notification of a fire from the field, and extinguishing the fire. As a result, this change indefinitely extends the start of the 15-minute emergency classification clock beyond a credible notification that a fire is occurring or indication of a valid fire detection system alarm. This change decreased the effectiveness of the emergency plan by reducing the capability to perform a risk significant planning function in a timely manner. The violation affected the NRCs ability to perform its regulatory function because it involved implementing a change that decreased the effectiveness of the emergency plan without NRC Commission approval. Therefore, this issue was evaluated using Traditional Enforcement. The NRC determined that a Severity Level IV violation was appropriate due to the reduction of the capability to perform a risk significant planning standard function in a timely manner. The violation is cited because no corrective action had been taken to restore compliance since the issue was entered in the licensees corrective action program in December 2009. The performance deficiency was more than minor and of very low safety-significance using Manual Chapter (MC) 0612 and MC 0609, Appendix B, because it is associated with the emergency preparedness cornerstone attribute of procedure quality for EAL and emergency plan changes, and it 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. Therefore, the performance deficiency was a finding. Using MC 0609, Appendix B, the inspector determined that the finding had a very low safety significance. The inspectors also determined that the finding had a cross-cutting aspect in the area of Human Performance, decision-making because the licensee did not recognize that the change made to the EAL basis document decreased the effectiveness of the emergency plan. (H.1.(b)) (Section 1EP4)
05000456/FIN-2010005-0231 December 2010 23:59:59BraidwoodNRC identifiedFailure to Submit a Licensee Event Report per 10 CFR 50.73(a)(2)(v)A Severity Level IV NCV of 10 CFR 50.73(a)(2)(v) was identified by the inspectors when licensee personnel failed to report known conditions that could have prevented the fulfillment of the Residual Heat Removal (RHR) system to perform its designed emergency core cooling safety function while operating in the shutdown cooling mode of operation, within 60 days of discovery. Specifically, upon receipt of Westinghouse Nuclear Safety Advisory Letter (NSAL) 0904, Presence of Vapor in Emergency Core Cooling System/Residual Heat Removal System in Modes 3 or 4 Loss-of-Coolant Accident Conditions, the licensee determined that a loss of RHR system safety function occurred when both trains of the RHR system were placed into the shutdown cooling mode of operation above 200 degrees Fahrenheit (F). The station identified four instances in which both trains of RHR were operated in the shutdown cooling mode of operation above 200F over the previous 3 year period. The licensee, however, failed to report to the NRC within 60 days that the RHR safety function had been lost. The station entered this issue into the CAP as IR 1155372. Corrective actions included the issuance of Licensee Event Report (LER) 05000456/457/2010-007-00 on January 18, 2010. The inspectors determined that the failure to report this LER in accordance with NRC regulations was a performance deficiency since this issue had the potential to impact the regulatory process. Therefore, this violation was dispositioned through the traditional enforcement process. The inspectors determined that this issue was a Severity Level IV violation based on a similar example referenced in NRC Enforcement Policy Supplement I, Example D.4. The inspectors evaluated this issue under the Reactor Oversight Process (ROP) and did not identify a performance deficiency that could be assessed under the SDP. (Section 4OA2.2)
05000456/FIN-2010002-0231 March 2010 23:59:59BraidwoodNRC identifiedFailure to Perform a 10 CFR 50.59 Evaluation of a Temporary Modification to the 2B RVLIS ProbeThe inspectors identified a finding of very low safety significance and an associated Severity Level IV Non-Cited Violation for the failure to perform an adequate 10 CFR 50.59 screening of a temporary modification. Specifically, the licensee failed to recognize the impact of a temporary modification on emergency operating procedures, which resulted in the failure to perform a full evaluation of the modification. The licensees corrective actions included reinforcing the current configuration of the 2B reactor vessel level indication system with operators and revising emergency operating procedures. In addition, the licensee plans to complete a full 10 CFR 50.59 evaluation to determine whether the modification required NRC approval prior to implementation. The inspectors concluded that the violation was more than minor because the inspectors could not reasonably conclude that the modification would not require prior NRC approval based on the 10 CFR 50.59 screening. The inspectors answered no to the Mitigating Systems cornerstone questions in Table 4 and, as a result, the issue screened as one of very low safety significance (Green). This finding is associated with the cross-cutting area component of decision-making in the human performance cross-cutting area. Specifically, when evaluating the operations impact of a new temporary modification on the 2B RVLIS probe, the licensee assumed the impact was unchanged from a prior temporary modification on the same equipment, which resulted in necessary procedure changes that were not identified (H.1(b))
05000456/FIN-2009003-0330 June 2009 23:59:59BraidwoodNRC identifiedFailure to Perform Appropriate 10 CFR 50.59 ReviewThe inspectors identified a Severity Level IV NCV of 10 CFR 50.59 following a review of changes made to TS required surveillance test procedures. These procedures allowed testing of Reactor Protection System (RPS) analog channels in the bypassed conditions by use of jumpers during surveillance test. This technique had been deemed unacceptable in NRC safety evaluation report for Westinghouse Topical Report WCAP 10271. This issue involves traditional enforcement because it involves a violation of 10 CFR 50.59 and is more than minor because there was a reasonable likelihood that the change would require NRC review and approval prior to its implementation. This issue did not represent an actual loss of safety function for greater than the TS allowed outage time; therefore it was of very low safety significance. Consequently, the finding is categorized as a Severity Level IV NCV in accordance with the NRC Enforcement Policy. There were no cross-cutting aspects identified by the inspectors. This finding was documented in the licenses corrective action program. Corrective actions included changing the method of reactor trip system testing.
05000456/FIN-2008004-0230 September 2008 23:59:59BraidwoodNRC identifiedFailure to Properly Evaluate Removal of Carbon Dioxide Fire Suppression for the Upper Cable spreading Room Carbon Dioxide Fire SuppressionA finding of very low safety significance and an associated NCV of Braidwood Operating License Condition 2.E was identified by the inspectors for the licensees failure to obtain NRC approval before making changes to the approved FPP. Specifically, the licensee permanently isolated the manual carbon dioxide (CO2) suppression system to the upper cable spreading rooms (UCSRs) without prior NRC approval. The licensee entered this issue in the corrective action program (CAP) and implemented compensatory actions to verify detection system operability and implement fire watches upon any single detector failure. Additionally, the licensee plans to submit a licensee change request associated with the removal of CO2 suppression from the UCSRs. The finding was determined to be more than minor because the inspectors could not reasonably determine that the isolation would not have ultimately required NRC prior approval. The inspectors determined this finding to be a Severity Level IV violation due to having very low safety significance (Green) based on the Phase 2 SDP evaluation. This finding is 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 with respect to reviewing the plant design and license basis. (H.1(b)) (Section 1R05.3)
05000456/FIN-2007006-0231 December 2007 23:59:59BraidwoodNRC identifiedDeficient Control of Plant Staff OvertimeThe inspectors identified a Severity Level IV Non-Cited Violation of Technical Specification 5.2.2.d for not properly implementing and maintaining procedures for controlling plant staff work hours of personnel performing safety-related activities. Procedure LS-AA-119, Overtime Controls, Revision 4, was deficient in that it permitted the plant manager to authorize work-hour deviations for routine refueling outage activities. This issue has a cross-cutting aspect in the area of Human Performance, Resources (Item H.2.(c) of IMC 0305), because Procedure LS-AA-119 did not provide adequate instructions to provide reasonable assurance that station management would properly control overtime for plant staff performing safety-related functions to assure nuclear safety as required by Technical Specification 5.2.2.d. The violation is more than minor because, if left uncorrected, the excessive work hours would increase the likelihood of human errors during refueling outage activities and response to plant events and would become a more significant safety concern. The finding is not suitable for Significance Determination Process evaluation, but has been reviewed by NRC management in accordance with IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The resulting increased likelihood of human error, would adversely affect the stations defense-in-depth. However, management determined the violation to be of very low significance, because no significant events or human performance issues were directly linked to personnel fatigue as a result of the hours worked. The licensee added this issue to their corrective action program to address correcting the procedure. In accordance with the NRC Enforcement Policy, Supplement I.D, the issue, being evaluated as having very low safety significance by the Significance Determination Process, is a Severity Level IV Violation.
05000456/FIN-2003006-0130 September 2003 23:59:59BraidwoodSelf-revealingFailure to Provide Accurate Performance Indicator Data to the NRCA self-revealing issue was identified when licensee engineers noted, during a review, that they had miscalculated and therefore misreported in July 2001, the fault exposure time for a 1B auxiliary feedwater pump failure. The issue was more than minor because it caused the performance indicator to cross the Green-to-White threshold during later quarters. The licensee submitted the corrected data to the NRC in a special mid-quarter data submittal in August 2003 and the issue was entered into the licensee's corrective action system. Because this issue affected the NRC's ability to perform its regulatory function, it was evaluated with the traditional enforcement process. The issue was determined to be a Severity Level IV Non-Cited Violation of 10 CFR 50.9. (4OA1.1)