Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000255/FIN-2016001-032016Q1PalisadesFailure to Meet the Minimum Staffing Requirements of the Fire BrigadeAn NRC-identified finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Section 48(c) and the National Fire Protection Association (NFPA) Standard 805 Section 3.4.1 was identified for the failure to meet the minimum staffing requirements for the Fire Brigade on January 4 and 5, 2016. Specifically, two nuclear plant operators (NPOs) who had their Fire Brigade qualifications suspended, stood watch as Fire Brigade members during day shift on January 4, 2016 and approximately one half of day shift on January 5, 2016. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PLP-2016-00198, performed an apparent cause evaluation, successfully performed a fire drill to requalify the Fire Brigade members with suspended qualifications on January 6, 2016, and planned to update the tracking method used to validate drill completion for Fire Brigade qualifications. The performance deficiency was determined to be more than minor because it was associated with the Protection Against External Factors 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). The finding screened as having very low safety significance based on using qualitative criteria located in IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria. The finding had a cross-cutting aspect of Documentation in the Human Performance cross-cutting area because the licensee informally tracked drill completion and this information was not accessible to each individual Fire Brigade member to validate their qualifications (H.7).
05000255/FIN-2016001-042016Q1PalisadesLicensee-Identified ViolationTitle 10 CFR 50.54(m)(2)(iii), Condition of Licenses, states that when a nuclear power unit is in an operational mode other than cold shutdown or refueling, as defined by the units technical specifications, each licensee shall have a person holding a senior operator license for the nuclear power unit in the control room at all times. TS 5.2.1 states in part, that during any absence of the Shift Supervisor from the control room while the plant is in Mode 1, an individual with an active Senior Reactor Operator (SRO) license shall be designated to assume the control room command function. Contrary to the above, at approximately 2:00 a.m. on September 2, 2015, with the unit in Mode 1, the Command SRO left the control room without another SRO being present in the control room and without turning over the command function. A few minutes prior to the event, the shift Command SRO turned over to the Shift Technical Advisor (STA) the Command SRO function of the control room so that the shift Command SRO could take a break outside the control room boundary. A minute or so after the STA (who had the Unit Command SRO function at the time) left the control room, a control room reactor operator observed that there were no SROs in the control room and summoned the Shift Manager from an office across the hall to the control room. The Shift Manager then assumed the Command SRO function and the STA was called back to the control room. This issue was identified by the licensee on September 2, 2015, and documented in CRPLP201503637, The SRO with Command and Control Momentarily Left the Control Room. There were no risk-significant plant evolutions in progress and no adverse reactor plant operations occurred during the SROs absence. The STA was relieved from shift responsibilities until corrective actions were taken. The inspectors screened the issue using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power. The inspectors reviewed the screening questions under all three Cornerstones and all of the logic questions did not apply, therefore the finding screened as having a very low safety significance (Green).
05000255/FIN-2016001-052016Q1PalisadesLicensee-Identified ViolationTS Limiting Condition for Operation (LCO) 3.0.6 states, in part, that when a supported system LCO is not met solely due to a support system LCO not being met, the Conditions and Required Actions associated with this supported system are not required to be entered; only the support system LCO actions are required to be entered. TS LCO 3.0.6 further specifies that an evaluation shall be performed in accordance with TS 5.5.13, Safety Function Determination Program. Palisades Administrative Procedure 4.11, Safety Function Determination Program, step 5.4.3 requires documentation of entry into TS LCO 3.0.6 for the inoperable supported system in the Operations Log. Contrary to the above, on January 19, 2016, the licensee failed to document entry into TS LCO 3.0.6 in the operations log when work was commenced on breaker 521214, Motor Control Center (MCC) 22 and MCC24 480 Volt feeder breaker. The licensee identified this issue when a similar condition was entered on January 22, 2016 and documented the missed entry into TS LCO 3.0.6 in CRPLP201600413, Operations Failed to Log Entry into LCO 3.8.1B and LCO 3.5.2B or LCO 3.0.6. The licensee provided coaching to the individuals involved. The inspectors screened the issue using IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, Mitigating System Screening Questions, and answered No to all the questions. Therefore, the finding screened as having very low safety significance (Green).
05000255/FIN-2016001-012016Q1PalisadesDesign Review of Modification to Track Alley Wall for Dry Fuel Storage ActivitiesThe inspectors identified a unresolved item (URI) associated with the design review of a modification to the Track Alley wall for dry fuel storage (DFS) campaign activities. Specifically, the licensee is currently revising the process applicability determination (50.59 and 72.48 screenings), and reviewing any necessary actions, associated with altering the newly modified wall in support of upcoming DFS campaign activities. The wall, a protective barrier with safety functions per the UFSAR, in its newly modified condition, will be altered when the steel plate covering the opening cut into it will be raised to accommodate the DFS transporter. The DFS campaign is currently on hold pending resolution of other issues. In January 2016, the licensee began work on an engineering change to permanently modify the west wall of Track Alley in order to accommodate the new transporter used for moving the casks associated with the dry fuel storage campaign. This modification removed a section of the reinforced concrete wall by cutting out an opening approximately 9 feet wide by 4 feet high by 18 inches deep into the existing wall. A three inch thick steel plate was mounted onto vertical rails which can slide down to cover the window cut into the wall and raised to open the window for when the transporter is brought into Track Alley. The west wall of Track Alley is also the east wall of the Technical Support Center (TSC). This wall is designed to withstand seismic, high wind, and tornado missile loads. It also serves as a radiation protection barrier for personnel in the TSC during emergency situations. The permanent modification of cutting the opening in the wall and installing the steel plate, to provide equivalent protection of the 18 inches of concrete that were cut out, was evaluated in Engineering Change 59170 and calculation EAEC5917001. The inspectors reviewed these documents, the supporting process applicability determination (50.59 screening), and risk assessment of implementing the design change. During this review, the inspectors identified that the licensee did not assess the alteration of the wall, a protective barrier with safety functions per the UFSAR, when the steel plate covering the window would need to be raised to accommodate the DFS transporter. The inspectors questioned this condition and the licensee subsequently completed a process applicability determination (PAD) form (72.48 and 50.59 screening). When reviewing the PAD, the inspectors questioned the licensees underlying assumption that moving the steel plate to uncover the window was considered to be in support of a maintenance activity and, hence, screened out of the 50.59 process, including not requiring certain compensatory actions for the walls safety functions during the period of time in which the opening was exposed. At the end of the inspection period the licensee was reviewing their assessment. Once their review is completed, including any changes that may be made, the inspectors will re-assess their evaluation and determine what actions, if any, will need to be accomplished in support of the DFS campaign. Since the campaign is on hold, a URI is being opened to track resolution of this issue.
05000255/FIN-2016001-022016Q1PalisadesMovement of Radioactive Material Results in an Unposted and Un-Barricaded High-Radiation AreaA self-revealed finding of very low safety significance and an associated NCV of Technical Specification 5.7.1 was identified when movement of a bag of radioactive material caused an area to become a high radiation area without the proper posting and barricades. The licensee immediately moved this bag of radioactive material to a posted locked high-radiation area and entered this issue into their CAP as CRPLP201505019. The performance deficiency was determined to be more than minor because it was associated with the Program and Process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the movement of the bag from an area that was a high-radiation area to an area that was not posted and barricaded as a high-radiation area removed a barrier that was intended to prevent workers from receiving unexpected dose. The finding was determined to be of very low safety significance in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The violation was of very low safety significance because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The finding had a cross-cutting aspect of Teamwork in the Human Performance cross-cutting area because the individuals and work groups involved did not communicate or coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained (H.4).
05000456/FIN-2015001-012015Q1BraidwoodFailure to Ensure that Temporary Structures Did Not Adversely Impact Safety during Postulated Probable Maximum Precipitation EventThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, when licensee personnel failed to establish adequate measures to ensure that temporary equipment and structures stored at the station did not create an unanalyzed condition during a probable maximum precipitation (PMP) event. Specifically, the licensees processes did not prevent the placement and storage of temporary equipment in a manner that could result in a condition not bounded by the stations plant design that prevents rainwater from impacting safety-related equipment. This issue was entered into the licensees Corrective Action Program (CAP) as Issue Report (IR) 2473324. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Design Control 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, the failure to ensure that credited rainwater runoff flow paths were not impeded by the storage of temporary structures resulted in the licensee not ensuring the availability, reliability, and capability of systems that would be needed to respond to an initiating event. This assessment was based upon the inspectors review of current flood barrier margins, assumed turbine building below-grade flooding levels, the number of safety-related or risk-significant systems that could be adversely affected, and the absence of an abnormal operating procedure or any other similar procedure that could create additional margin. The inspectors determined that because the finding did not involve a confirmed loss or degradation of equipment or function specifically designed to mitigate a PMP external flooding event, the issue was of very low safety significance. The inspectors determined that the finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current performance.
05000456/FIN-2015001-022015Q1BraidwoodFailure to Adequately Evaluate Operability of a Degraded Control Room ChillerThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when licensee personnel failed to adhere to the operability determination process after identifying a degraded condition on the 0B control room chiller. This issue was entered into the licensees CAP as IR 2435363. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, 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, the licensee did not provide an adequate basis to support 0B control room chiller availability, reliability, and capability to respond to an initiating event. The inspectors determined that the finding was of very low safety significance because all questions related to structures, systems, and components (SSCs) and functionality in the associated significance determination process (SDP) were answered "No." The finding had a cross-cutting aspect in the Design Margins component of the Human Performance cross-cutting area because the licensee failed to adequately evaluate whether the degraded oil return line in the 0B control room chiller had sufficient margin to assure operability (H.6).
05000456/FIN-2015001-042015Q1BraidwoodLicensee-Identified ViolationOn February 19, 2014, the licensee identified that Braidwood Station had not complied with TS 3.4.3, RCS Pressure and Temperature Limits, between March 2011 and October 2013, during startup of the plant following plant refueling outages. Braidwood TS 3.4.3 stated, RCS pressure, RCS temperature, and RCS heat up and cooldown rates shall be maintained within the limits specified in the PTLR (Pressure Temperature Limits Report.) The PTLR is generated by Westinghouse and contains graphs depicting the acceptable operating ranges of RCS pressure and temperature supported by the analysis. The lower bound of these graphs was 0 pounds per square inch gauge (psig). Braidwood Procedure BwOP RC-9 was used by the station to fill the loops. This procedure allowed RCS piping pressure to go as low as 28 inches of mercury (or about14 psig) which was below the lower limit of the PTLR acceptable region. At the licensees request, Westinghouse performed the additional analysis needed to expand the lower value of the curves and determined that the lower bounding parameter could be revised to14.7 psig with no impact to RCS barriers. The analysis was subsequently revised and the PTLR was revised to designate the lower boundary accordingly. Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by procedures appropriate to the circumstances. Contrary to the above, from March 2011 through October 2013, BwOP RC-9 allowed RCS pressures to be lower than the analyzed bound of the parameter inputs of the PTLR graphs and, as a result, was not appropriate to the circumstances. The finding was more than minor because it impacted the Procedural Quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that the RCS design barrier would function to protect the public from radionuclide release caused by accidents or events. Given the analytical conclusions that the condition was acceptable with the new lower bounding parameter, the inspectors determined that the issue was of very low safety significance (Green). The licensee entered this issue into their CAP as IR 1625970 and corrective actions consisted of updating the PTLR.
05000456/FIN-2015001-032015Q1BraidwoodFailure to Activate the ERO During an Actual EventA self-revealed finding of very low safety significance and an associated NCV of 10 CFR 50.54(q)(2) and 10 CFR 50.47(b)(2) was identified on July 23, 2014, when after a Notice of Unusual Event was declared and the Shift Manager activated the Emergency Response Organization (ERO), several of the ERO members failed to respond as required. This issue was entered into the licensee's CAP as IR 2469494. The inspectors determined that the performance deficiency was more than minor because it was associated with the Emergency Response Organization Readiness attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Since the finding involved a failure to comply with emergency preparedness requirements, the inspectors reviewed IMC 0609, Appendix B, Attachment 2, and determined that the finding was of very low safety significance because it involved a degraded planning standard function. The finding had a cross-cutting aspect in the Change Management component of the Human Performance cross-cutting area because the licensee did not appropriately evaluate and implement changes when the new ERO Augmentation System was implemented (H.3).
05000456/FIN-2014005-012014Q4BraidwoodFailure to Adequately Evaluate Operability Following the Discovery of an Unanalyzed Condition Involving the Probable Maximum Precipitation EventThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when licensee personnel failed to adhere to Operability Determination Process standards after identifying an unanalyzed condition that had the potential to adversely impact numerous safety-related systems during a probable maximum precipitation (PMP) event. The issue was entered into the Corrective Action Program (CAP) as Issue Report (IR) 2396124. Corrective actions for this issue included performing an operability evaluation. The performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening because the issue was associated with the Protection Against External Factors attribute of the Mitigating System 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, the licensee evaluated an unanalyzed condition utilizing another power plants licensing basis in a manner that was not accurate and was not adequate. The finding was of very low safety significance (Green) because the potentially impacted systems remained operable. The finding had a cross-cutting aspect of Avoid Complacency in the Human Performance area. Specifically, the licensee failed to recognize and plan for the possibility of mistakes and plant specific differences between Braidwood and Byron while using Byrons current licensing basis to evaluate a Braidwood condition not previously analyzed (H.12).
05000282/FIN-2014005-022014Q4Prairie IslandUnqualified Reactor Vessel Examination ProceduresThe inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion IX, Control of Special Processes, on October 21, 2014, due to the licensees failure to perform the reactor vessel weld ultrasonic examinations with procedures qualified in accordance with the American Society of Mechanical Engineers (ASME) Code. Corrective actions for this issue included entering the issue into the corrective action program (CAP) and considering the available options to restore compliance with the ASME Code. The inspectors determined that this issue was more than minor because if left uncorrected, this deficiency had the potential to lead to a more significant safety concern. Specifically, the failure to properly qualify ultrasonic examination procedures prior to examining the Unit 1 reactor vessel welds could result in the failure to detect weld flaws. In turn, the undetected weld flaws could increase the risk of a loss of coolant accident. The inspectors concluded that this issue was of very low safety significance because Questions 1 and 2 provided in IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, were answered No. In this case, the ultrasonic examination intended to detect weld degradation had not yet affected the ability of the reactor vessel to perform its design functions. This finding was cross-cutting in the Human Performance, Resources area because the licensee did not have adequate supervisory and management oversight of work activities to ensure that the procedures used during the ultrasonic examination of reactor vessel welds were properly qualified in accordance with the applicable ASME Code (H.2).
05000456/FIN-2014005-032014Q4BraidwoodFailure to Evaluation Impact of PMP Event On Turbine Building Flooding and Associated Safety-Related SSCsThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to assess the impact of plant modifications on the PMP event analysis in the plan design basis. Specifically, the licensee failed to determine if modifications to plant grading that caused higher water levels during a PMP event would adversely affect safety-related equipment. The licensee entered this issue into the CAP as IR 2413941. Corrective actions included performing an operability determination to ensure safety unti a formal quality design review can be completed at a later date. The performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue was associated with the Protection Against External Factors attribute of the Mitigating System cornerstone and adversel 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, the licensee failed to evaluate the design to ensure that the consequences of the licensing basis PMP would be acceptable with respect to NRC regulations. The finding was of very low safety significance (Green) because it did not result in the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect of Design Margins in the Human Performance area. Specifically, the licensee did not carefully guard design margins when making station grade modifications that could adversely affect safety-related equipment during a heavy rainfall event. This issue was determined to be indicative of recent performance based upon two recent major revisions to station calculation WRBRPF10, Local PMP Analysis, which evaluated the acceptability of recent grade modifications at the station (H.6).
05000456/FIN-2014005-022014Q4BraidwoodFailure to Correct Undersize Essential Service Water Pump Bearing Casing Drain Line Resulted in System InoperabilityA finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control was self-revealed following the licensees failure to design the 1B essential service water (SX) pump inboard bearing casing drain line in a manner that ensured pump operability. Specifically, the licensee had re-designed the 1B SX pump inboard bearing drain line by replacing a hard pipe drain with a flexible hose drain line consisting of fittings of a smaller diameter when compared to the previous hard pipe drain line. This design change resulted in unplanned 1B SX pump inoperability and required operator action to secure the pump to preclude pump damage. The licensee entered this issue into the CAP as IR 2413941. Corrective actions included restoring adequate drain flow by replacing the flexible hose drain line with a hard pipe of a larger diameter. The performance deficiency was of more than minor safety significance because the issue 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, the failure t adequately design the 1B SX pump inboard bearing housing drain line resulted in an inoperable 1B SX pump. The finding was of very low safety significance (Green) because the inspector answered No to all of the associated Mitigating Systems screening questions within IMC 0609, Attachment 4, Initial Characterization of Findings. The finding is associated with the cross-cutting area of Problem Identification and Resolution with an aspect of Evaluation because the licensee did not thoroughl evaluate plant design in a manner commensurate with the safety significance. Specifically, the licensee inappropriately evaluated the design of the 1B SX pump inboard bearing housing drain line after identifying that the drain line size was the contributing cause for a loss of oil inventory in December 2013 (P.2).
05000282/FIN-2014005-042014Q4Prairie IslandLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, requires, in part, that activities affecting quality be prescribed by documented instructions, procedures and drawings of a type appropriate to the circumstance and be accomplished in accordance with these instructions, procedures and drawings. Contrary to the above, on December 30, 2013, the licensee identified that plant personnel performed Unit 2 SG hot gap checks, an activity affecting quality, without having documented instructions, procedures and drawings appropriate to the circumstance. Specifically, the design drawings failed to include information indicating that only one set of steam generator upper lateral support shims and bumpers was to be removed at a time during the hot gap clearance checks. Due to this deficiency, the workers removed all of the steam generator shims and bumpers which resulted in making the Unit 2 RCS inoperable and placing the unit in an unanalyzed condition. The inspectors determined that the licensees failure to have drawings appropriate to the circumstance for performing the hot gap clearance checks was a performance deficiency. The performance deficiency was evaluated in accordance with IMC 0612, Appendix B, Issue Screening. The inspectors determined that the performance deficiency did not involve a violation that impeded the regulatory process or contributed to actual safety consequences. The inspectors determined that the finding was more than minor because it impacted the design control and configuration control attributes of the initiating events cornerstone. In addition, the finding impacted the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Table 3SDP APPENDIX ROUTER. The inspectors answered No to all the questions listed in Table 3; therefore, the risk evaluation continued with IMC 0609 Appendix A, The Significance Determination Process For Findings At-Power. Under the Initiating Events Cornerstone for Exhibit 1, the inspectors answered Yes to the question, After a reasonable assessment of degradation, could the finding result in exceeding the RCS leak rate for a small break LOCA? Therefore, the inspectors contacted the Region III Senior Reactor Analysts (SRAs) for a detailed risk evaluation. The SRAs performed a detailed risk evaluation for the effect of the missing SG shims on Unit 2 with regard to the change in core damage frequency (CDF). The evaluation assumed certain initiating events (such as seismic events) would impart a significant load upon the steam generators causing their displacement, which in turn could lead to breaks in the reactor coolant primary piping or secondary system piping. The SRAs reviewed the licensees risk evaluation for this issue as documented in: PRA Document No. V.SPA.14.010, Unit 2 SG Shim SDP Calculation, Rev. 1; PRA Document No. V.SPA.14.011, U2 SG Secondary Break Size Threshold, Rev. 1; and PRA Document No. V.SPA.14.009, Unit 2 SG Shim SDPMSLB and MFLB Initiating Event Frequency Development, Rev. 1. Based on review of the above licensee documents and input from the inspectors, the SRAs determined that the following initiating events would result in a change in CDF: Seismic Events; Large Loss of Coolant Accident Events; Main Steamline Breaks in Containment; and Main Feedwater Line Breaks in Containment. Other initiating events, such as steam generator tube ruptures and medium and small break LOCAs were determined not to result in a change in risk due to the missing shims. Following a steam generator tube rupture, imparted loads were small enough such that the connected primary and secondary piping were expected to remain intact and functional. Similarly, following medium and small break LOCAs the steam generators, secondary piping, and the unaffected primary loop were expected to remain intact and functional. According to the licensees Unit 2 SG Shim SDP Calculation referenced above, the duration when all shims on each SG were removed to the time when all of the shims were re-installed was about 20.5 hours. The calculation did not state the time when the first shim was removed; therefore, the SRAs doubled this exposure time and assumed 41 hour duration for the exposure time for this finding, which was a conservative assumption. Also, the SRAs conservatively assumed that the initiating events subject to this analysis proceed directly to core damage without credit for mitigation or recovery (i.e., conditional core damage probability (CCDP) is 1.0). Seismic Events - The licensee screened out seismic events of magnitude greater than the design basis earthquake of level 0.12g for contributing to a change in CDF. Their analysis showed the primary and secondary side piping would remain intact and functional during seismic events less than 0.12g. The inspectors and SRAs accepted this assumption. The Risk Assessment of Operational Events Handbook listed the frequency of 0.08g and greater seismic events, and 0.15g and greater seismic events, as 1.907E04/yr and 7.272E05/yr respectively. Interpolating these values on a logarithmic scale resulted in a frequency occurrence for seismic events greater than 0.12g to be 1.024E04/yr. Assuming a conditional core damage probability of 1.0, the seismic contribution to the risk increase was taken to be frequency of the seismic event during the exposure time or 4.79E07/yr as calculated below: CDFseismic = (1.024E04/yr) * (41/8760) = 4.79E07/yr. Large Loss of Coolant Accidents - The SRAs used the Prairie Island Standardized Plant Analysis Risk (SPAR) Model Version 8.19 to obtain the frequency of a large loss of coolant accident (LLOCA). The SPAR model lists the frequency of LLOCAs as 2.50E06/yr. Assuming a conditional core damage probability of 1.0, the LLOCA contribution to the risk increase was taken to be frequency of the LLOCA during the exposure time or 1.17E08/yr as calculated below: CDFlloca = (2.50E06/yr) * (41/8760) = 1.17E08/yr. Main Steamline Breaks in Containment - For main steamline (MSL) breaks in containment, the licensee performed an evaluation that determined that only pipe breaks of certain sections of 5.5-inch diameter pipe and larger could result in loadings large enough to impact the change in CDF. The inspectors and SRAs accepted this assumption. The licensee calculated a MSL break initiating event frequency for 4inch equivalent diameter piping for this analysis using EPRI Report 3002000079, Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments (PRAs), Revision 3, and other plant documents and drawings. The licensee calculated a MSL break frequency of 4.45E05/yr. Assuming a conditional core damage probability of 1.0, the MSL break contribution to the risk increase was taken to be frequency of the MSL break during the exposure time, or 2.08E07/yr: CDFMSLB = (4.45E05/yr) * (41/8760) = 2.08E07/yr. Main Feedwater Line Breaks in Containment - For main feedwater line (MFL) breaks in containment, the licensee performed an evaluation that determined that only pipe breaks of certain sections of 5.5inch diameter pipe and larger could result in loadings large enough to impact the change in CDF. The inspectors and SRAs accepted this assumption. The licensee calculated a MFL break initiating event frequency for 4inch equivalent diameter piping for this analysis using EPRI Report 3002000079 and other plant documents and drawings. The licensee calculated a MFL break frequency of 3.53E06/yr. Assuming a conditional core damage probability of 1.0, the MFL break contribution to the risk increase was taken to be frequency of the MFL break during the exposure time, or 1.65E08/yr: CDFMFLB = (3.53E06/yr) * (41/8760) = 1.65E08/yr. Results - The total change in CDF (i.e., CDF) represents the sum of the individual CDF values above, or 7.16E07/yr. In regards to Large Early Release Frequency (LERF), IMC 0609 Appendix H, Containment Integrity Significance Determination Process, was used to determine the potential risk contribution due to LERF. Prairie Island is a two loop Westinghouse pressurized water reactor with a large dry containment. Sequences important to LERF include steam generator tube rupture events and inter-system LOCA events. These were not the dominant core damage sequences for this finding. Therefore, the risk significance due to the change in CDF and LERF was determined to be of very low safety significance (GREEN). The licensee documented this issue in the corrective action program as CAP 1412886. Corrective actions included re-installing the shims and bumpers, performing additional hot gap adjustments, and revising the upper lateral support drawings to specifically state that only one shim and bumper package can be removed at a time.
05000282/FIN-2014005-032014Q4Prairie IslandFailure to Follow Procedures during EDG 24 Hour Load TestThe inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, on September 29, 2014, due to the licensees failure to follow procedure during the performance of SP 1335, D2 Diesel Generator 18 Month 24 Hour Load Test. Specifically, operations personnel failed to comply with steps within SP 1335 which directed that the emergency diesel generators (EDGs) kVAR loading be adjusted until a power factor of less than or equal to 0.85 was achieved or Bus 16 voltage was between 4350 and 4375 volts. An extent of condition review determined that operations personnel failed to comply with a similar procedure step during the 24 hour load test of the D1 EDG performed in May 2013. As a result, the licensee had to re-perform the tests, which resulted in additional EDG inoperability and unavailability. Corrective actions for this issue included training the operators on the need to maintain the power factor or bus voltage within limits during testing, requiring all data collected by the operations department during Technical Specification (TS) surveillance testing to be independently verified, and requiring all TS surveillance requirement results to be reviewed and approved by two senior reactor operators. The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and impacted the cornerstones objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operations personnel were required to declare the D1 and D2 EDGs inoperable and unavailable to perform their safety functions while the 24 hour load testing was re-performed. The inspectors concluded that this issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, was answered No. This finding was cross-cutting in the Human Performance, Avoid Complacency area because operations personnel failed to implement appropriate error reduction tools to ensure that the power factor or bus voltage requirements were met during the surveillance test (H.12).
05000282/FIN-2014005-012014Q4Prairie IslandFailure to Implement the Winter Plant Operation ProcedureThe inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, on December 4, 2014, due to the licensees failure to follow procedure during the performance of test procedure (TP) 1637, Winter Plant Operation. Specifically, maintenance personnel failed to comply with a step within TP 1637 which directed that a tent and heater be installed around the Unit 2 cooling water (CL) discharge to grade header to prevent ice buildup and subsequent blockage during freezing conditions. Consequently, the inspectors identified ice buildup on the CL header discharge orifice which if left uncorrected, could result in header blockage and subsequent inoperability. Corrective actions for this issue included removing the ice buildup on the cooling water discharge header, installing a tent and heater in accordance with TP 1637, revising the associated procedures and performing an apparent cause evaluation. The inspectors determined that this issue impacted the Mitigating Systems cornerstone and was more than minor because if left uncorrected, this issue could become a more significant safety concern. Specifically, with freezing conditions present coupled with the existence of leakage and resultant ice buildup on 20CL61, the potential existed for subsequent ice blockage if left uncorrected and resultant inoperability of the cooling water system. This issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, was answered No. The inspectors concluded that this finding was associated with a conservative bias cross cutting aspect in the human performance cross cutting area. Specifically, operations and maintenance personnel did not utilize prudent decision making practices to ensure the cooling water header was adequately protected against freezing conditions (H.14).
05000456/FIN-2014004-062014Q3BraidwoodLicensee-Identified Violation10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that applicable regulatory requirements and the design basis are appropriately translated into specifications, drawings, procedures, and instructions. Contrary to the above, as of June 25, 2014, the licensee had failed to translate the design basis of the UHS into procedures and instructions. Specifically, procedure 0BwOAENV3, Braidwood Cooling Lake Low Level Unit 0, did not reflect the assumptions in the UHS analysis in that non-essential service water pumps were not directed to be secured to prevent loss of inventory in the UHS. This issue was entered into the CAP as IR1674557 and the procedure was corrected. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue was associated with the Design Control 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, based on the analysis of record, at the time of discovery, there was reasonable doubt that the UHS could meet its mission time of 30 days. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors determined that the finding affected the design of the UHS, but did not result in a loss of operability, and therefore screened the finding as having very low safety significance (Green).
05000456/FIN-2014004-052014Q3BraidwoodInadequate Evacuation Time Estimate SubmittalsThe NRC identified a finding of very low safety significance and an associated NCV of 10 CFR 50.54(q)(2) related to 10 CFR 50.47(b)(10) and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Braidwood Station Emergency Plan as a result of failing to provide the station Evacuation Time Estimate (ETE) to the responsible offsite response organizations by the required due date. Exelon submitted the Braidwood Station ETE to the NRC on December 12, 2012, prior to the required due date of December 22, 2012. However, an NRC review found the ETE to be incomplete due to Exelon fleet common and site-specific deficiencies, thereby preventing Exelon from providing the ETE to responsible offsite response organizations and from updating site-specific protective action strategies as necessary. The NRC discussed its concerns regarding the completeness of the ETE in a teleconference with Exelon on June 10, 2013, and on September 5, 2013, Exelon resubmitted the ETEs for its sites. Subsequently, the NRC again found the ETE to be incomplete. Exelons failure to submit a complete updated ETE for Braidwood Station by December 22, 2012, was a licensee performance deficiency because the issue was a failure to comply with a regulatory requirement and the issue was reasonably within the licensees ability to foresee and correct, and therefore should have been prevented. The inspectors determined the performance deficiency was more than minor because it was associated with the Emergency Preparedness cornerstone attribute of Procedure Quality and adversely affected the cornerstone objective of ensuring that the licensee was capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was of very low safety significance because it was a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The licensee entered this issue into their CAP and re-submitted a new revision of the Braidwood Station ETE to the NRC on May 2, 2014. The inspectors concluded that this finding had a cross-cutting aspect in the Documentation component of the Human Performance cross-cutting area (H.7).
05000456/FIN-2014004-012014Q3BraidwoodAdverse Impact of Floor Drain Design on Flooding AnalysisThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III Design Control, when licensee personnel failed to verify the design of bag-strainers in the floor drains of the auxiliary building and their impact on the associated flooding analysis. Specifically, when Calculation 3C80686002, Auxiliary Building Flood Level Calculation, was revised on May 16, 2013, the licensee credited the use of floor drains, which had bag-type strainers that were designed in such a way that they increased the potential for blockage, and therefore adversely impacted the analysis of record for internal flooding. This issue was entered into the licensees Corrective Action Program (CAP) as Issue Report (IR) 2385204, NRC Questions on Aux (Auxiliary) Building Flood Evaluation. Corrective actions for this issue included instituting Standing Order 14005 to prevent the interim removal of flood seals, and a plan to revise Calculation 3C80685002 to resolve the identified non-conformances. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue was associated with the Design Control 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, the floor drain strainer bags were inadequately designed in such a manner that instead of ensuring that the floor drains would be able to function properly to remove flood water, they would act to increase the possibility that the floor drains would become plugged and unable to perform this function adequately. The inspectors concluded that the finding was of very low safety significance in accordance with IMC 0609, Appendix A, Exhibit 2 and Exhibit 4. The inspectors determined that the finding had a cross-cutting aspect in the Evaluation component of the Problem Identification and Resolution (PI&R) cross-cutting area because the licensee failed to thoroughly evaluate the issue to ensure that the resolution addressed the causes. Specifically, when the licensee made a major revision to Calculation 3C80685002 in 2013 to, in part, incorporate minor revisions and address non-conservatisms in the calculation, the licensee failed to adequately consider a previous minor revision that had removed credit for 3 the drain system due to problems with its design that were previously identified by the NRC (P.2).
05000456/FIN-2014004-032014Q3BraidwoodStation Diesel-Driven Fire Pump Restored to Service Non-Functional Due to Incorrect Stop Push Button Switch ReplacementA finding of very low safety significance and an associated NCV of Braidwood Operating License Condition 2.E, Fire Protection Program, was self-revealed during the performance of a scheduled diesel-driven fire pump (DDFP) sequential start surveillance when the DDFP was observed by operators to start, but then cycle on and off. The DDFP was declared non-functional and a subsequent causal evaluation determined that an incorrectly designed DDFP stop pushbutton switch had been installed several months prior to the identification of the issue. The licensee entered this issue into their CAP as IR 1649515, Incorrect Stop Pushbutton Installed on 0B Fire Pump. Corrective actions included replacing the switch with a switch of a correct design. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue 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, the performance deficiency resulted in a non-functional DDFP. The finding was determined to be of very low safety significance by a NRC Senior Reactor Analyst. The inspectors concluded that this finding had a cross-cutting aspect in the Avoid Complacency component of the Human Performance cross-cutting area because the licensee did not adequately recognize and plan for the possibility that the DDFP stop pushbutton replacement switch design could have been different than plant-specific design requirements (H.12).
05000457/FIN-2014004-042014Q3BraidwoodUnit 2 Pressurizer Pressure Transmitter 458 Returned to Service with Instrument IsolatedA finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on May 21, 2014, when licensee personnel failed to use a quality instruction to reposition Unit 2 safety-related pressurizer pressure transmitter isolation valve 2PT458. Specifically, although the licensee identified that safety-related 2PT458 had been isolated from service and was not in service during a plant startup, as anticipated, the licensee could not locate the work instruction that isolated the instrument from service. The licensee entered this issue into their CAP as IR 1663588, Level 3 CCE2PT0458 Found Isolated. Corrective actions included restoring the pressure transmitter to service by opening a shut isolation valve and performing a causal evaluation. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because the issue 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, as a result of the performance deficiency, the automatic function of pressurizer power-operated relief valve (PORV) 2RY455A was not available for a number of days to perform its design function to mitigate an Anticipated Transient Without Scram (ATWS) event. In addition, IMC 0612, Appendix E, Examples of More than Minor Inspection Findings, Example 7e, informed this more-than-minor bases. Specifically, the issue was more than minor because it resulted in overall plant risk being in a higher risk category (i.e., Yellow vs. Green). The inspectors determined that the issue was of very low safety significance in accordance with IMC 0609, Attachment 4, Initial Characterization of Findings. In particular, Table 3, SDP Appendix Router, directed that the finding be screened using IMC 0609, Appendix A, The Significance Determination Process for At-Power Findings. The inspector answered No to all of the associated Mitigating Systems screening questions. This finding did not have an assigned cross-cutting aspect because the cause of the performance deficiency was indeterminate.
05000255/FIN-2014003-012014Q2PalisadesWritten NRC Biennial Written Examinations Did Not Meet Qualitative StandardsThe inspectors identified a finding of very low safety significance associated with 10 CFR 55.59, Requalification, based on a determination that greater than 20 percent of the biennial requalification written exam questions administered to licensed operators during weeks three and five of the 2012 examination cycle were flawed. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PNP-2014-02521, Written Exam Quality, dated April 10, 2014. The inspectors determined that the finding was more than minor because it was associated with the Human 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, the finding adversely affected the quality and level of difficulty of biennial written exams, which potentially impacted Palisades ability to appropriately evaluate licensed operators. The risk importance of this issue was evaluated using IMC 0609, Appendix l, Licensed Operator Requalification Significance Determination Process (SDP). The inspectors considered the number of written exam questions that did not meet the qualitative standard for written exam questions. The qualitative standards used by the inspectors are defined in NUREG-1021, Revision 9, ES-602, Attachment 1, Guidelines for Developing Open-Reference Examinations, and Appendix B, Written Examination Guidelines. Because more than 30 percent of the questions reviewed did not satisfy the guidance, Block 4 of Appendix I applied. Based on the screening criteria, the finding was characterized by the SDP as having very low safety significance (Green) because greater than 20 percent, but less than 40 percent, of the reviewed written exam questions were flawed. A review of the cross-cutting aspects was performed and no associated cross-cutting aspect was identified.
05000255/FIN-2014003-032014Q2PalisadesFailure to Notify the NRC Within 30 Days of Discovering Changes in Medical ConditionsA Severity Level IV non-cited violation of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status, was identified by the inspectors during a review of licensed operator medical records. Specifically, Palisades did not notify the NRC within 30 days of discovering a change in medical condition for a licensed operator. Subsequently, the licensee submitted the required notification for the operator on April 11, 2014, and entered the issue into their CAP as CR-PLP-2014-02518, NRC Informed the Palisades Training Department that an NRC Form 396 was Not Submitted, dated April 10, 2014. The inspectors determined that Traditional Enforcement applied because a failure to make a required report impacted the regulatory process. Specifically, the licensee had not notified the NRC within 30 days of learning of a change in medical condition for a licensed operator for which a license condition was required. Based on Example 6.9.d.1 of the NRCs Enforcement Policy, the inspectors determined that the issue represented a Severity Level IV violation. No associated Reactor Oversight Process finding was identified, thus there was no associated cross-cutting aspect.
05000255/FIN-2014003-082014Q2PalisadesLicensee-Identified ViolationTechnical Specification 5.7.2, High Radiation Areas with Dose Rates Greater than 1.0 Rem/Hour at 30 Centimeters from the Radiation Source or from Any Surface Penetrated by the Radiation, But Less than 500 Rads/Hour at 1 Meter from the Radiation Source or from any Surface Penetrated by the Radiation, requires, in part, that each entryway to such an area shall be barricaded and conspicuously posted as a high radiation area. Such barricades may be opened as necessary to permit entry or exit of personnel or equipment. Contrary to the above, on March 12, 2014, radwaste operators found that the south east steam generator bio-wall cage door, a locked high radiation area, was open and not locked. The licensee documented this issue as CR-PLP-2014-02083, Radwaste Operators Found a Locked High Radiation Area Gate Left Open, dated March 13, 2014. The finding was determined to be of very low safety significance (Green) because it was not an ALARA planning issue; there was no overexposure, nor substantial potential for an overexposure; and the licensees ability to assess dose was not compromised.
05000255/FIN-2014003-042014Q2PalisadesFailure to Evaluate Long-Term Scaffolds in Accordance with ProceduresThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when licensee personnel failed to adequately implement procedure EN-MA-133, Control of Scaffolding. Specifically, multiple examples were identified of scaffolds installed in the plant for greater than 90 days that had not undergone process applicability determinations, were not appropriately documented in the scaffold control log, and/or did not contain proper tags. The licensee documented the issue in their CAP as CR-PLP-2014-2646, Two Scaffolds Near Safety-Related Equipment Not Being Controlled as Long-Term, dated April 17, 2014; conducted an extent-of-condition review of the entire scaffold log and identified additional discrepancies; completed the required process applicability determinations; and re-inspected scaffolds that had been categorized as long-term. The inspectors determined that the performance deficiency was more than minor because it was similar to Example 4.a) of IMC 0612, Appendix E, Examples of Minor Issues. This example described an engineering evaluation that was not performed for scaffolding erected near safety-related equipment and stated that it would be a more than minor issue if the licensee routinely failed to perform the engineering evaluations. For this specific finding, there were multiple examples of process applicability determinations not being performed within the procedurally required timeframe. The finding was determined to be of very low safety significance (Green) because it did not affect the operability/functionality of structures, systems and components (SSCs) and all required safety functions were maintained. This finding was associated with the cross-cutting aspect of Teamwork in the Human Performance area. Specifically, licensee and supplemental individuals and work groups did not sufficiently communicate and coordinate work activities associated with maintaining the scaffold control log or documentation related to scaffolding installed in the plant. The workers also did not understand how to account for time during refueling and forced outages when determining the long-term status of scaffolds, which could have been resolved with input from other work groups.
05000255/FIN-2014003-022014Q2PalisadesExam Security IssuesThe inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 55.49, Integrity of Examinations and Tests, which stated, Applicants, licensees, and facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Specifically, Palisades placed personnel in the simulator operating booth that were not identified in the security agreement, placed the scenario turnover sheet for a second scenario in the simulator during the first scenario, and left a job performance measure turnover sheet in the simulator after the applicant left the simulator and brought the next applicant into the simulator. This issue was entered into the licensees CAP as CR-PLP-2014-02533, Issues Were Identified During the Annual Exam Administered on April 10, 2014, dated April 10, 2014. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to become a more significant safety concern. Specifically, the failure to properly control operational examination material in a manner in which applicants were not prematurely exposed to the material provided opportunities to compromise the examination. The finding was screened as one of very low safety significance (Green) in accordance with IMC 0609, Appendix I, Licensed Operator Requalification SDP. This finding was associated with the cross-cutting aspect of Procedure Adherence in the Human Performance area.
05000456/FIN-2014002-032014Q1BraidwoodFailure to Identify Fire Doors that Did Not Conform to NFPA Codes and StandardsThe inspectors identified a finding of very low safety significance and an associated NCV of Braidwood Operating License Condition 2.E, Fire Protection Program, when licensee personnel failed to identify fire doors that did not conform to the current licensing basis standard within the National Fire Protection Agency (NFPA)-80 Code that required fire doors to automatically shut and latch without assistance. Specifically, station personnel were not adequately performing a daily fire door testing procedure and, as a result, failed to identify a number of fire doors that were not conforming to the standard. As a result, IRs were not generated when degraded conditions existed. The licensee entered this issue into their CAP as IR 1629689, Unclear Direction in 0BwOS FP.7.2.D-1. Corrective actions included training plant operators on the expectations regarding generation of IRs for any abnormal condition in the plant, and requiring the use of a copy of the surveillance procedure in the field while completing the daily fire door surveillance. The inspectors determined that the performance deficiency was more than minor because it is associated with the Protection Against External Factors 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, licensee personnel did not identify a number of fire doors that were not capable of closing and latching without assistance, which impacted the doors ability to perform its design function. Using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet, the inspectors determined that the finding category was Fire Confinement, and that the finding did not impact the ability of the plant to achieve safe shutdown. As a result, the finding screened as having very low safety significance (Green). This finding had a cross-cutting aspect in the Procedure Adherence component of the Human Performance cross-cutting area because licensee personnel did not follow procedures, processes and work instructions. Specifically, the licensee did not have the fire door testing procedure in hand while performing the surveillance and did not follow the procedure steps (H.8).
05000456/FIN-2013008-012013Q4BraidwoodInaccurate/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-2013008-022013Q4BraidwoodInaccurate/Incomplete Information For Exemption Request From 10 CFR 50.60The inspectors identified a finding of very low safety significance (Green) and an associated Severity Level IV Violation of 10 CFR 50.9 Completeness and Accuracy of Information, for the licensees failure to provide information to the NRC that was complete and accurate in all material respects. Specifically, in Letter RS-05-103 License Amendment Request Regarding Reactor Coolant System Pressure and Temperature Limits Report and Request for Exemption from 10 CFR 50.60, the licensee stated that WCAP-16143 provides a valid basis for changing the reactor pressure vessel (RPV) 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. However, the analysis documented in WCAP-16143 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. Therefore, this analysis did not provide a valid basis for changing the Unit 2 RPV closure head flange limits in 10 CFR Part 50, Appendix G. The licensee entered this issue into the Corrective Action Program (AR 01558067), performed an operability evaluation, and was evaluating several options for corrective measures. The corrective actions under consideration by the licensee included: completing a calculation to validate the Westinghouse Electric vendor letter, revision to WCAP-16143, installation of a 54th head stud, submittal of a license amendment request with a revised WCAP-16143, or negate the existing exemption methodology and return to the pressure temperature limit curves based upon 10 CFR Part 50, Appendix G requirements. The inspectors determined that this issue was more than minor because it adversely affected the Barrier Integrity Cornerstone attribute of Design Control. The inspectors also answered yes to the More-than-Minor screening question, If left uncorrected, would the performance deficiency have the potential to lead to a more significant safety concern? Specifically, the inspectors determined that this issue was more than minor because, if left uncorrected, the failure to provide complete and accurate information for the Unit 2 vessel head stud configuration could have resulted in non-conservative pressure temperature limit curves that allowed operation in an unacceptable region that would increase the possibility of vessel failure during a pressurized thermal shock event. The inspectors performed a Phase I SDP screening using IMC 0609, Attachment 0609 Appendix A, Exhibit 3-Barrier Integrity Screening Questions, and selected the box under the Reactor Coolant System Boundary (e.g., pressurized thermal shock issues), which required a detailed risk-evaluation. A Region III Senior Reactor Analyst performed a detailed risk-evaluation of this finding. A potential increase in the probability for vessel failure would exist if the plant was operated in the unacceptable pressure temperature regions and a pressurized thermal shock event occurred. Based on the licensee and supporting vendor assessments which concluded that no substantial increase in vessel stresses will occur due to operation with 53 head studs, the driving force for crack propagation (e.g., K1) will remain essentially unchanged. However, to bound the delta risk-evaluation, it was assumed that the initiating event frequency for a reactor vessel failure increased by 10 percent. From the Braidwood Standardized Plant Analysis Risk Model Version 8.21, the initiating event frequency for reactor vessel failure from any cause was 1E-7/yr. Core damage is expected to occur if reactor vessel failure occurs. The exposure time for the finding was the maximum of one year. Thus, a bounding risk-assessment yields a delta risk of 1E-8/yr. Therefore, based on the detailed risk-evaluation, this finding is of very low risk significance (Green). Because the failure to provide complete and accurate information to the NRC had the potential to impede or impact the regulatory process, the finding was also evaluated in accordance with NRC Enforcement Policy for traditional enforcement. This violation was similar to an example of a Severity Level III violation identified in Section 6.9.c.1 of the NRC Enforcement Policy. However, after consideration by NRC management, and with the approval of the Director of the Office of Enforcement, it was determined that a Severity Level IV Cited Violation was appropriate. This decision was based upon the very low safety significance (Green) of the associated finding. The inspectors concluded that no cross-cutting aspect was applicable as the performance deficiency was not reflective of current performance because the issue was in excess of three years old.
05000255/FIN-2013005-072013Q4PalisadesPeriodic Design Basis Testing of Safety-Related Electrical ComponentsThe licensees Quality Assurance Program Manual stated they were committed to Regulatory Guide 1.30, Quality Assurance Requirements for the Installation, Inspection, and Testing of Instrumentation and Electric Equipment. This Regulatory Guide endorsed Institute of Electrical and Electronics Engineers (IEEE) Standard 336-1971 (also known as American National Standards Institute (ANSI) N45.2.4-1972) as adequate for demonstrating compliance with the pertinent quality assurance requirements of 10 CFR Part 50, Appendix B. In addition, Section C.3 of Regulatory Guide 1.30 stated, Although Subdivision 1.1 of ANSI N45.2.4-1972 states the requirements promulgated apply during the construction phase of a nuclear power plant, these requirements are also to be considered applicable for the installation, inspection, and testing of instrumentation and electric equipment during the operation phase of a nuclear power plant. IEEE Standard 336-1971, Section 3.3, Procedures and Instructions, required the licensee to produce documents that shall be kept current by controlled supervision so that installation, inspections, and tests are performed in accordance with the latest approved design and manufacturers instructions. However, while reviewing the licensees management of component design life, the inspectors noted the licensee did not periodically test safety-related electrical components to the design requirements. The licensee interpreted the intent of Section C.3 of Regulatory Guide 1.30 as to apply IEEE 336 requirements only to modifications and activities that were similar to initial construction activities. This issue is a URI pending further review, including consultation with the Office of Nuclear Reactor Regulation, and determination of further NRC actions to resolve the Issue.
05000255/FIN-2011014-092011Q4PalisadesPotential Loss of Preferred AC Sources in Harsh EnvironmentOn September 25, 2011, a fault occurred on Panel D11-2, which resulted in reactor and turbine trip, and de-energiziation of Bus D-10. Breaker 72-37, which supplied DC power to Inverter D-06, was found tripped. According to the manufacturer, the inverters were capable of reverse-feeding DC short circuits for short durations and this could have caused Breaker 72-37 to trip. This was possible because the inverter had four 7700 microFarad parallel capacitors on the DC side of the inverter. During a DC short circuit, the capacitors would rapidly discharge and feed the fault. Breaker 72-37 had a rating of 100 Amps for the thermal setting and 700 Amps for the magnetic setting. According to the manufacturer an approximation for an inverter DC fault current contribution was about 1100 Amps per capacitor; therefore, this was approximately a total of 4400 Amps for Inverter D-06. This exceeded the magnetic rating of the breaker and explained why the breaker tripped during the fault condition. The PCP motor DC oil lift Pumps P-81A and P-81C were nonsafety-related loads, which received power from Bus D-10 via safety-related Breakers 72-13 and 72-14, respectively. The PCP motor DC oil lift Pumps P-81B and P-81D were also nonsafety-related loads that received power from D-20 via safety-related Breakers 72-23 and 72-24, respectively. The cabling for these loads was not environmentally qualified and was routed through containment, which could be susceptible to failure due to a harsh environment. The inspectors were concerned that if all four nonsafety-related cables for these pumps faulted due to a harsh environment during a design basis event, this could result in the loss of all preferred AC power busses due to the internal capacitors contributing to the fault as seen by each DC bus. However, without further analysis of the design and licensing basis, the inspectors could not determine if a postulated harsh environment affecting all four cables during a design basis event was a credible event. Therefore, the inspectors initial conclusion, based on the available information was that this event may not be credible; however, further analysis was required. In addition, all four PCP motor DC oil lift pump breakers were opened as one of the compensatory measures for the operability of the 125-Volt DC system. Therefore, this is not a current safety concern. Title 10 CFR 50.49, Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, Section b(2), requires nonsafety-related electric equipment to be environmentally qualified if the failure of the nonsafety-related electric equipment under postulated environmental conditions could prevent satisfactory accomplishment of safety functions specified in subparagraphs (b)(1)(i) (A) through (C) of paragraph (b)(1) of this section by the safety-related equipment. The inspectors were concerned that the cables associated with the PCP motor DC oil lift Pumps P-81A, P-81B, P-81C, and P-81D were not evaluated for the effect on the safety-related equipment specifically the safety-related inverters and their associated preferred AC sources. The licensee entered this issue into their CAP as CR-PLP-2011-6210. This issue is a URI pending the licensee evaluation, and the inspectors review of the licensees design and licensing basis, and evaluation to determine if a performance deficiency existed (URI 05000255/2011014-09; Potential Loss of Preferred AC Sources in Harsh Environment).
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Brent-20Boston-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"