Semantic search

Jump to navigation Jump to search
 Start dateSiteIdentified byTitleDescription
05000456/FIN-2018411-0130 September 2018 23:59:59BraidwoodLicensee-identifiedLicensee-Identified Violation
05000456/FIN-2018003-0130 September 2018 23:59:59BraidwoodSelf-revealingInadequate Detail in Maintenance Procedure for Emergency Diesel Generator 2-Year Inspection Contributed to 1A Emergency Diesel GeneratorFuel Rack BindingA self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to include adequate detail within their maintenance procedures to enable proper lubrication of the emergency diesel generator (EDG) fuel rack control linkage. Specifically, the preventative maintenance template for the fuel rack control linkage required that the manual fuel trip lever and associated linkage be lubricated every 2 years. However, the licensees implementing 2year maintenance procedure failed to include specific instructions to disassemble the lever assembly for lubrication. This lack of lubrication contributed to the mechanical binding of the emergency diesel generator fuel rack and failure of the 1A EDG during surveillance testing on April 22,2018.
05000456/FIN-2018003-0230 September 2018 23:59:59BraidwoodNRC identifiedMinor ViolationAll Braidwood Station EDG governors were replaced during the late 1990s. During design testing, the licensee noted that the historical EDG frequency response had changed slightly due to installation of new electronic governors. Prior to these governor replacements, EDG frequency was always above 57 hertz (Hz) during load sequencing. However, with the newly installed electronic governors, 1A and 2A EDG frequency was observed to dip below the 57 Hz under frequency relay setpoint following start of the 1A and 2A motor-driven AF pumps. (Note that because the 1B and 2B AF pumps are diesel-driven, there is no corresponding impact on the 1B or 2B EDGs.) As a result, an external 2-second time delay, provided by an Agastat time delay relay, was incorporated into the under frequency trip logic for the 1A and 2A EDGs to provide an additional margin for frequency recovery following motor-driven AF pump load starts. The Braidwood governor modification was installed in 1998, with the external time delay added to the 1A and 2A EDGs as part of the design changes to prevent inadvertent actuations of the under frequency logic.During the licensees investigation into the issue discussed in the subject LER, it was identified that the external Agastat time delay was installed incorrectly on the 1A EDG. Specifically, the original trip logic wiring had not been properly removed, which permitted the actuation of the under frequency trip after the original 0.5 second internal time delay through the bypassing of the additional 2.0 second external time delay. The wiring error was introduced during the original modification installation in October 1998. Screening: The inspectors determined that the error was of minor safety significance. Absent the mechanical binding of the manual fuel trip lever and associated linkage, as discussed in NCV 05000456/201800301 in this report, the 1A EDG had performed reliably and satisfactorily during surveillance testing prior to the Unit 1 refueling outage testing in April of 2018. Additionally, the inspectors determined that the error, having occurred some 20 years ago, was not indicative of current licensee performance.Violation: This failure to comply with the requirements of 10 CFR Part 50, Appendix B, Criterion III , Design Control, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000457/FIN-2018002-0130 June 2018 23:59:59BraidwoodSelf-revealingInadequate Detail in Maintenance Work Instructions Resulted in Failed Gearbox Oil Cooler Head Gasket and Inoperable 2B Auxiliary Feedwater PumpA self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to have adequate detail within their maintenance work instructions to enable proper reassembly of the 2B auxiliary feedwater (AF)pump gearbox oil cooler. Specifically, during the licensees 19th Unit 2 refueling outage in April 2017, the gearbox oil cooler closure head was reassembled following scheduled maintenance using an excessive amount of room temperature vulcanizing silicone (RTV) on the joint and an insufficient amount of torque on the closure head bolting. As a result, on March 16, 2018, the closure head joint failed causing several hours of unplanned inoperability and unavailability for the 2B AFPump.
05000456/FIN-2018002-0230 June 2018 23:59:59BraidwoodSelf-revealingWork Instruction Error Results in Reactor Coolant System Pressure TransientA self-revealed finding of very low safety significance (i.e., Green) was identified due to the licensees failure to follow work instructions while performing a digital upgrade to plant control systems. Specifically, while performing maintenance on the volume control tank (VCT) level transmitter on April 10, 2018, maintenance personnel failed to properly track the steps being performed while simultaneously working on multiple packages. This resulted in the Unit 1 reactor coolant system (RCS) experiencing a pressure transient and the actuation of a VCT relief valve.
05000456/FIN-2018002-0330 June 2018 23:59:59BraidwoodSelf-revealingInadequate Test Activity Coordination Results in Unintended Valve Actuation and Reactor Coolant System Pressure DropA self-revealed finding of very low safety significance (i.e., Green) and an associated NCV of Technical Specification 5.4, Procedures, was identified for the licensees failure to have properly coordinated testing activities associated with redundant Unit 1 pressurizer pressure instruments in accordance with the stations procedural requirements governing such testing. Specifically, during the licensees 20th Unit 1 refueling outage, on April 23, 2018, redundant pressurizer pressure instrumentation channels were inadvertently subjected to simultaneous testing activities. This resulted in the coincidence logic for both of the units pressurizer power-operated relief valves (PORVs) being satisfied and the PORVs opening to depressurize the RCS from approximately 345 pounds per square inch gauge (psig) to approximately 320 psig
05000456/FIN-2017008-0131 December 2017 23:59:59BraidwoodNRC identifiedFailure to Prevent Secondary Missiles Following a Postulated HELBThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design basis for the main steam safety valve (MSSV) room maintenance hatches was maintained. Specifically, the high energy line break ( HELB) analysis performed for the MSSV rooms and steam tunnels prior to initial construction concluded that no secondary missiles were generated as a result of a HELB although maintenance hatches in the ceiling of the MSSV rooms were identified to become secondary missiles following a HELB in the MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their corrective action program (CAP) as AR 4075641 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control 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). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC (Structure, System, and Component) , does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross- cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-0231 December 2017 23:59:59BraidwoodNRC identifiedInadequate Blow Out Panel Design ControlThe 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 the licensee originall y designed the MSSV blow out panels in a manner that prevented them 3 from functioning properly. The licensee entered this issue into their CAP as A R 4075641 and corrected the design issue in March of 2009 . The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control 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). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, E xhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-0331 December 2017 23:59:59BraidwoodNRC identifiedFailure to Properly Correct Errors in Design Analysis for Main Steam Line Break in Main Steam TunnelThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly correct errors in the design analysis for a main steam line break in the main steam tunnel. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and completed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control 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). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-0531 December 2017 23:59:59BraidwoodNRC identifiedInaccurate Analysis of RecordThe 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 the licensee failed to maintain an accurate and up- to-date analysis of record for a postulated HELB in th e MSSV rooms and steam tunnels. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control 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). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not resul t in a loss of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017008-0431 December 2017 23:59:59BraidwoodNRC identifiedUntimely Corrective Action for Secondary MissilesThe inspectors identified a finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, when the licensee failed to promptly address the identification of secondary missiles following a HELB event. As part of their immediate corrective actions, the licensee entered this issue into their CAP as A R 4075641 and performed an operability evaluation. The finding was determined to be more than minor because it was associated with t he Mitigating Systems cornerstone attribute of Design Control 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). In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 2, Mitigating 4 Systems Sc reening Questions, the inspectors answered Yes to Question 1, If the finding is a deficiency affecting the design or qualification of a mitigating SSC, does the SSC maintain its operability or functionality? because the finding did not result in a los s of operability or functionality. Therefore, this finding was of very low safety significance. No cross -cutting aspect was assigned to this finding as it was not reflective of current performance.
05000456/FIN-2017003-0130 September 2017 23:59:59BraidwoodNRC identifiedFailure to Implement Adequate Radiological Controls for Treated Liquid Radioactive Effluents Containing TritiumThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1406(c), when the licensee failed to conduct operations to minimize the introduction of residual radioactivity onto the site. Specifically, the licensee failed to identify and evaluate the environmental risk and control work practices with a credible mechanism to prevent spills and leaks from reaching groundwater at the circulating water blowdown (CWBD) area, a radiologically unrestricted area in the licensees owner controlled area. Specifically, tritium contaminated sump water was intermittently pumped to the environs. The licensee documented this finding in their corrective action program (CAP) as Issue Report (IR) 4020644. The failure to conduct operations and control work practices with a credible mechanism to prevent spills and leaks to reach groundwater and minimize residual radioactivity onto the site represented a licensee performance deficiency. The performance deficiency was of more than minor significance because it was associated with the Program and Process attribute of the Public Radiation Safety cornerstone and adversely affected the cornerstone objective of ensuring the adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. In accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because the issue involved a radioactive effluent release, but did not: (1) represent a substantial failure to implement the radioactive effluent release program; or (2) result in public exposure that exceeded the dose values in Appendix I to 10 CFR Part 50 and/or 10 CFR 20.1301(e) limits. The inspectors determined that this finding had a cross-cutting component in the area of Human Performance, in the aspect of Challenging the Unknown, because licensee personnel did not stop when faced with uncertain conditions or evaluate and manage risk before proceeding.
05000456/FIN-2017002-0130 June 2017 23:59:59BraidwoodNRC identifiedFailure to Adequately Implement Surveillance Frequency Program for the Deferral of a Technical Specification SurveillanceGreen. A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.5.19.b, Surveillance Frequency Program, were identified by the inspectors for the licensees failure to implement the requirements contained in the surveillance frequency control program when making a change to the specified frequency of TS Surveillance Requirement (SR) 3.3.1.11. On May 3, 2017, the licensee improperly deferred a TS required surveillance through the preventive maintenance deferral process due to a belief that it was a preventive maintenance activity and not an activity supporting a TS SR. The licensee entered this issue into their corrective action program (CAP) as Issue Report (IR) 4009050 with an action to re-establish the surveillance at an 18-month frequency and to perform it before the end of the Unit 2 refueling outage (RFO) A2R19. The performance deficiency was determined to be more than minor because if left uncorrected it could lead to a more significant safety concern. The finding screened as being of very low safety significance (Green) because it did not result in the loss of operability or functionality of any system, structure, or component (SSC). The inspectors determined that this finding had a cross-cutting component in the area of human performance, work management aspect, because the licensee failed to utilize a work process that included proper coordination with different groups or job activities. Specifically, licensee personnel conducting the deferral did not coordinate the activity with personnel in either the operations or regulatory assurance departments. Knowledgeable personnel in either of these station organizations could have identified that the wrong process for deferral was being utilized. (H.5)
05000456/FIN-2017002-0230 June 2017 23:59:59BraidwoodNRC identifiedFailure to Adequately Implement Technical Specification Surveillance Frequency Requirements into Implementing ProceduresGreen. A finding of very low safety significance and an associated NCV of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were identified by the inspectors for the licensees failure to have appropriate implementing procedures for TS SR 3.9.3.2. Specifically, procedure BwIS NR203, Post Accident Neutron Monitoring System Discriminator Adjustment, 3 did not provide for determining and checking the discriminator voltage for the system at an 18-month frequency, as specified by TS SR 3.9.3.2. The licensee entered this issue into their CAP as IR 4010147 with an action to revise the surveillance frequency to every 18 months for each channel. The performance deficiency was determined to be more than minor because it was associated with the Procedure Quality attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective. The finding screened as having very low safety significance (Green) because it did not result in the loss of operability or functionality of any SSC. The licensee performed a review of the records associated with the last three years of operation and did not find any instances in which the post-accident neutron monitors (PANMs) were used to satisfy TS 3.9.3, Nuclear Instrumentation, requirements. No cross-cutting aspect was associated with this finding because it was confirmed not to be reflective of current licensee performance due to the age of the performance deficiency.
05000456/FIN-2017002-0330 June 2017 23:59:59BraidwoodNRC identifiedFailure to Adequately Implement and Maintain the Radiological Environmental Monitoring Program by Collecting Representative Samples from the Principal Food PathwaysGreen. A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix I, Section IV(B), were identified by the inspectors for the licensees failure to establish an appropriate surveillance and monitoring program in order to provide data on measurable levels of radiation and radioactive materials in the environment to evaluate the relationship between quantities of radioactive material released in effluents and resultant radiation doses to individuals from principal pathways of exposure. This was an NRC-identified finding for the failure to implement and maintain the licensees radiological environmental monitoring program (REMP) by collecting representative samples from the highest deposition coefficient (D/Q) quadrant locations during annual REMP sampling and collections of food products in 2015. On May 25, 2016, during a review of the stations annual radiological environmental operating report for 2015, the inspectors noted that the licensee documented missed samples in three out of four quadrants where the principal food pathways were grown within the 10 kilometers from the station and missed milk samples. The licensee captured this issue in their CAP as IR 4002540. Licensee corrective actions included, but were not limited to, revising the applicable REMP procedures and investigating the possibility of growing the principal food pathways on the licensees owner controlled area or other approved licensee property within the 10 kilometer site radius. The performance deficiency was determined to be more than minor because it was associated with the Program and Process attribute of the Public Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of the public from radiation. Specifically, the licensee failed to implement effective sample collection from sample locations for food products from three of the major quadrants during annual REMP sampling and collections in 2015. The licensees Offsite Dose Calculation Manual (ODCM), as written, did not meet 10 CFR Part 50, Appendix I, which requires the licensee to establish and provide data on measurable levels of radiation and radioactive materials in the site environs. The finding was determined to be of very low safety significance in accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, because it only involved the licensees REMP. The inspectors determined that this finding had a 4 cross-cutting component in the area of human performance, change management aspect, because the licensee did not use a systematic process for evaluating and implementing changes in their REMP sampling and collection program. (H.3)
05000456/FIN-2016004-0131 December 2016 23:59:59BraidwoodLicensee-identifiedInadequate Control of Welding During FW System Pipe ReplacementA finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors for the licensees failure to assure that thermo couple (TC) attachment welding was controlled and accomplished by qualified personnel using qualified procedures and to assure that the post-TC attachment weld removal non-destructive examination (NDE) was incorporated into Work Order (WO) 01836557 that provided instructions to replace a pipe segment in the safety-related portion of the feedwater (FW) system. The licensee corrective actions for this finding included documenting this issue as a potential violation of NRC requirements in Issue Report (IR) 02728742, removal of the unqualified welds, and issuing revisions to WO 01836557 that included licensee-approved weld procedures and surface examinations of FW pipe affected by unqualified TC welds. This finding was determined to be of more than minor significance because it affected the Reactor Safety Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. In particular, if left uncorrected this issue would have the potential to lead to a more significant safety concern because it increased the likelihood of an operational challenge to the plant caused by a FW system line break induced by cracking initiated from unqualified welds. 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 1, Initiating Events Screening Questions. Under Part B, Transient Initiators, of the Exhibit 1 questions, the inspectors answered No because the finding did not result in a reactor trip and/or loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Therefore, this finding was screened as having very low safety significance (Green). This finding had a cross-cutting aspect of Field Presence in the cross-cutting area of Human Performance since licensee managers failed to provide adequate oversight of site and vendor personnel to assure that the TC attachment welding was controlled and accomplished by qualified personnel using qualified procedures and to assure that the post-TC attachment weld removal NDE was incorporated into WO 01836557. (H.2)
05000456/FIN-2016007-0230 September 2016 23:59:59BraidwoodSelf-revealingOperation of SX System Valves Results in Cavitation Damage and Pipe LeakageA finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to prescribe essential service water (SX) system operating and/or surveillance procedures appropriate to the circumstances. Specifically, the licensee failed to provide SX operating procedure guidance to limit the closure position of valves 1SX007, 2SX007 and 0SX007, such that cavitation-induced damage/failure of components did not occur or to establish a procedure to monitor and correct cavitation-induced damage prior to component failure associated with the operation of these valves. Consequently, a through-wall leak occurred downstream of valve 1SX007 that was caused by cavitation-induced wall loss at the neck of the pipe flange supporting this valve. The licensee replaced the damaged valve and piping and entered this issue into their CAP as Issue Report (IR) 02697962. The team determined that the performance deficiency was 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. Specifically, continued operation of the SX007 valves without monitoring or correcting cavitation-induced damage could result in a more significant failure resulting in the loss of an SX train and/or an internal flooding event. The team determined that this finding was of very low safety significance because although it was determined to be a deficiency affecting the design or qualification of a mitigating structure, system, and component (SSC), the operability or functionality of the component was not affected. The team did not identify a cross-cutting aspect for this finding because the finding did not reflect current licensee performance.
05000456/FIN-2016007-0130 September 2016 23:59:59BraidwoodNRC identifiedIdentification of SCAQs in Accordance with the QATRThe team identified an Unresolved Item (URI) regarding the identification of significant conditions adverse to quality (SCAQs) in the CAP. Specifically, the team determined that the CAP, as implemented by PI-AA-125,Corrective Action Program, and PI-AA-120, Issue Identification and Resolution, appeared to not ensure that SCAQs were appropriately identified and corrected to prevent recurrence. Chapter 16 of the Braidwood Quality Assurance Topical Report (QATR) describes the licensees program to identify and correct conditions adverse to quality. Procedure PI-AA-125 implemented the requirements established in the QATR. During this inspection, the team reviewed the CAP procedure to determine how it ensured that SCAQs were identified and resolved. As part of this review, the team requested a copy of identified SCAQs over the last two years and were subsequently informed that none had been identified. Issue #1 - The team reviewed the QATR and noted that the following requirements applied: Section 2.1 stated that measures are required to assure that the cause of any significant condition adverse to quality is determined and that corrective actions to prevent recurrence (CAPRs) are implemented. Section 2.2.1, Significant Conditions Adverse to Quality, stated that in cases of significant conditions adverse to quality the cause of the condition must be determined and documented, the resolution determined and documented, and the corrective actions taken and documented to prevent recurrence. Step 2.116 of Appendix D of the QATR defined a significant condition adverse to quality as, a condition, which if left uncorrected, could have a serious effect on safety or operability. The team reviewed procedure PI-AA-125 and PI-AA-120, which delineated the process for the identification and screening of issues, and identified that these procedures did not include a provision to classify an identified issue as a SCAQ. The team also noted that the definition of a SCAQ was not being used to determine whether a RCE was needed; therefore, a CAPR did not appear to be directly associated with a SCAQ. Based on the above, the team questioned whether CAP procedure PI-AA-125 prescribed a process through which SCAQs were identified and documented, and corrective actions taken and documented to prevent recurrence as required by the QATR. The team discussed this issue with the licensee. The licensee stated that since the terms SCAQ and condition adverse to quality (CAQ) were not explicitly defined in NRC regulations, that they had created a graded approach of significance level and likelihood (which included risk and uncertainty) to ensure that items were properly dispositioned and the level of resources and rigor applied appropriately followed the CAP governance. The licensee further stated that the graded approach, along with a well-trained management team that has nuclear safety and conservative decision-making as their primary focus, provided for an effective CAP. Finally, the licensee stated that even if a CAPR was not issued, that CAs would prevent recurrence of the events entered into the CAP. The team questioned whether a CAPR and a CA would be equally effective as corrective actions to prevent the recurrence of issues dispositioned in the CAP. The licensee agreed that the two types of CAs were treated differently. For example, 1) the MRC was required to assess changes to the intent of a CAPR, which was not required for a CA, 2) an effectiveness review may not necessarily be assigned if an issue was corrected using only a CA, and 3) if there was a desire to suspend or modify a previously implemented CAPR, then a risk analysis and MRC concurrence was necessary; which was not the case for a CA. At the end of the inspection it was not clear how procedures PI-AA-120 and PI-AA-125 ensured that SCAQs were identified and documented, and corrective actions taken and documented to prevent recurrence. Additionally, it was not clear if the licensees process implemented the requirements in the QATR. Resolution of this issue will be based on additional NRC review to determine if a violation of NRC requirements occurred. Issue #2 - The team identified an example of a potential SCAQ for which the licensee implemented CAs that failed to prevent the issue from recurring. Specifically, for a December 30, 2013 oil leak on the inboard bearing housing of the Unit 1 Train B (1B) SX pump, the licensees CAs restored operability, but were not adequate to prevent recurrence and consequently an oil leak recurred on November 18, 2014. Both of these oil leaks resulted in the licensee declaring the 1B SX pump inoperable and required entry into Technical Specification (TS) Limited Condition for Operation (LCO) 3.7.8 (reference Non-Cited Violation (NCV) 05000456/201400502; Failure to Correct Undersized Essential Service Water Pump Bearing Casing Drain Line Resulted in System Inoperability). The team questioned whether the oil leaks on the inboard pump bearing housing of the 1B SX pump should have been categorized as a SCAQ as defined in the licensees QATR. Specifically, QATR Section 2.116, Definitions, defined a SCAQ as, A condition, which if left uncorrected, could have a serious effect on safety or operability. In this case, although the oil leakage at the inboard pump bearing housing first identified in 2013 was specifically addressed through repairs, the CAs were not adequate to prevent recurrence and a second oil leak occurred in 2014 that caused a serious effect on the operability of the 1B SX pump (i.e. rendered the 1B SX pump inoperable). Additionally, the team considered this issue to have a potentially serious effect on operability, because if left uncorrected the oil leakage would have depleted the oil supply reservoir resulting in a loss of lubrication to the pump shaft bearings that could damage the pump shaft and require substantial repairs to return the pump to operation. The team discussed this issue with the licensee. The licensees response was that because there was no potential for common cause failure, and there was no significant change to plant risk after removing the 1B SX pump from service, the events discussed above were appropriately screened as Significance Level 3 issues. The licensee also stated that a SCAQ would typically be assigned for a Significance Level 1 or 2 issue, but even if an issue was assigned this level of significance, it would not necessarily be categorized as a SCAQ. At the end of the inspection it was not clear how the definition of SCAQ in the QATR was utilized in the CAP. Resolution of this issue will be based upon additional NRC review and a determination of whether the failure of the 1B SX pump constituted a SCAQ as defined in the QATR.
05000456/FIN-2016003-0130 September 2016 23:59:59BraidwoodNRC identifiedFailure to Erect Scaffolding in Accordance with Station ProceduresThe 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, for the failure to follow Revision 7 of NESMS04.1, Seismic Prequalified Scaffolds. Specifically, the licensee erected four scaffolds within 3 inches of safety-related equipment and failed to account for seismic movements of safety-related equipment in close proximity to scaffolds in accordance with NESMS04.1. As part of their corrective actions, the licensee performed walk downs of installed scaffolds to ensure that they were in compliance with NESMS04.1. Additionally, the licensee performed refresher training for all personnel involved in erecting and inspecting scaffolds. This issue was entered into the licensees CAP as IRs 2703650, 2703895, 2703967, and 2705092. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors and adversely impacted 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, scaffolds built in close proximity to or in contact with safety-related equipment could adversely affect the ability of those systems to perform their intended safety function during a seismic event. The inspectors determined that this finding was of very low safety significance because it did not result in the loss of operability or functionality of a mitigating system. Specifically, an engineering evaluation reasonably determined that the failure to build the scaffolds in accordance with NESMS04.1 did not result in a loss of operability to safety-related equipment. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance area of Teamwork. Specifically, there were multiple points in the scaffold erection process to engage other workgroups to ensure the seismic qualification of scaffolds, and in every example there was no coordination with other groups to ensure nuclear safety was maintained (H.4).
05000456/FIN-2016003-0230 September 2016 23:59:59BraidwoodNRC identifiedFailure to Follow Inservice Testing Requirements for the 2A Essential Service Water Pump Leads to an Invalid TestThe 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, for the failure to follow Revision 9 of Procedure 2BwOSR 5.5.8.SX6A, Comprehensive Inservice Testing (IST) Requirements for 2A Essential Service Water Pump (2SX01PA). Specifically, on September 7, 2016, the licensee failed to establish flow as close as possible to the reference point of 24,000 gallons per minute (gpm), as specified in Step 1.17 of the procedure, which ultimately led to an invalid test. The planned corrective actions included re-performing the comprehensive test on September 26, 2017, and an action to revise affected procedures to specify that the flow should be established as close as possible to the reference value, and to not throttle flow to below the reference value to obtain acceptable testing results. This issue was entered into the licensees CAP as IRs 2644532 and 2660824. The inspectors determined that the performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely impacted 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 follow the requirements established by the American Society of Mechanical Engineers (ASME) for comprehensive testing led to an invalid test of the pump on September 7, 2016. The inspectors determined that this finding was of very low safety significance because it did not result in the loss of operability or functionality of a mitigating system. Specifically, when the test was re-performed on September 26, 2016, it was confirmed that the 2A essential service water pump was operable. The inspectors determined that this finding had a cross-cutting aspect in the Human Performance area of Training. Specifically, licensee staff in Operations and Engineering were under the impression that they did not need to establish flow as close as possible to the reference value of 24,000 gpm. Instead, their belief was that the flow band in the surveillance procedure allowed them to set flow at any point in the band; therefore, when faced with results that fell within the Required Action Range, licensee staff believed that it was acceptable to lower flow to obtain more favorable results provided the system flow remained within the flow band (H.9).
05000456/FIN-2016201-0130 September 2016 23:59:59BraidwoodNRC identifiedSecurity
05000456/FIN-2016002-0130 June 2016 23:59:59BraidwoodNRC identifiedMultiple Failure to Follow Procedures Leads to Inadequate Monitoring of Gas Susceptible LocationsThe 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, for the failure to follow Revision 3 of procedure ERAA2009, Managing Gas Accumulation. Specifically, 36 gas-susceptible safety-related piping locations were not being monitored in accordance with the procedure. The planned corrective actions included an action to revise the Surveillance Frequency Control Program surveillance frequencies of accessible locations from 18 months to 6 months to align with procedural requirements, and an action to address the monitoring of locations inside the missile barrier (non-accessible locations at power). This issue was entered into the licensees Corrective Action Program (CAP) as Issue Reports (IRs) 2644532 and 2660824. The inspectors determined the performance deficiency was more than minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to adequately monitor for gas accumulation in piping did not ensure the availability and reliability of systems required to perform accident mitigating functions because a potential adverse void would not be detected and assessed for operability impact. The inspectors determined that this finding was of very low safety significance because it did not result in the loss of operability or functionality of mitigating systems. Specifically, an engineering evaluation reasonably determined that the non-conforming condition did not result in a loss of operability. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risks, even while expecting successful outcomes. Specifically, the licensee had multiple recent opportunities to discover the non-compliance, but failed to do so because the licensee assumed that the surveillance frequencies were established correctly (H.12)
05000456/FIN-2016002-0430 June 2016 23:59:59BraidwoodNRC identifiedLicensee Implementation of Enforcement Guidance Memorandum 15002, Enforcement Discretion for Tornado-Generated Missile Protection NoncomplianceOn June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection, focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced enforcement guidance memorandum (EGM) 15002, which was also issued on June 10, 2015. The EGM provided guidance to allow the NRC staff to exercise enforcement discretion when an operating power plant licensee did not in comply with the current license basis for tornado-generated missile protection. Specifically, the discretion would be applied to structure system and components (SSCs) declared inoperable resulting in TS LCOs that would require a reactor shutdown or mode change if the licensee could not meet the required actions within the TS completion time. The discretion allowed the licensee to reestablish operability through compensatory measures and established criteria for continued operation of the facility as longer term corrective actions were implemented. This allows the licensee to continue operating until final corrective actions are taken in the timelines established in the EGM. The EGM stated that the bounding risk analysis performed for this issue concluded that this issue was of low risk significance and, in Braidwoods case, provided for enforcement discretion of up to 3 years from the date of issue of the EGM. However, the EGM does not provide the licensees enforcement discretion for any related underlying technical violations; and moreover, the EGM specifically requires that any associated underlying technical violation be assessed through the enforcement process. Appendix A to 10 CFR Part 50, General Design Criteria for Nuclear Power Plants (GDC), Criterion 4, Environmental and Dynamic Effects Design Basis, stated in part that SSCs important to safety shall be adequately protected against dynamic effects including missiles. On May 25, 2016, the licensee initiated IR 02673854, identifying a nonconforming condition of Criterion 4. Specifically, multiple locations were identified in the refueling water storage tank (RWST) roof hatches and in the L-line wall above the 451 elevation (separating the turbine building from the Class I auxiliary building) where SSCs were not adequately protected from tornado-generated missiles. The licensee declared multiple SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The inspectors reviewed the licensees compensatory measures that included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados/high winds, and potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The condition was reported to the NRC as Event Notice 51959 as an unanalyzed condition and potential loss of safety function. The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); 19 TS 3.3.7, Control Room Ventilation (VC) Filtration System Actuation Instrumentation; TS 3.5.2, ECCS Operating; TS 3.5.4, Refueling Water Storage Tank (RWST); TS 3.6.6, Containment Spray and Cooling Systems; TS 3.7.5, Auxiliary Feedwater System; TS 3.7.10, Control Room Ventilation (VC) Filtration System; TS 3.7.11, Control Room Ventilation (VC) Temperature Control System; TS 3.8.4, DC Sources Operating; TS 3.8.7, Inverters Operating; and TS 3.8.9, Distribution Systems Operating. The inspectors review addressed the material issues in the plant, and whether the measures were implemented in accordance with the guidance documentation for the EGM. The inspectors also evaluated whether the measures as implemented would function as intended and were properly controlled. The licensee implemented actions to track the more comprehensive actions to resolve the nonconforming conditions within the required 60 days. These comprehensive actions were to remain in place until permanent repairs were completed, which for Braidwood were required to be completed in 3 years, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC such that discretion was no longer needed. The inspectors did not review the underlying circumstances that resulted in the TS violations. As stated in the EGM guidance, violations of other requirements, including 10 CFR 50 Appendix A Criterion 4, that may have contributed to the TS violations would be evaluated independently of the EGM implementation. This operability inspection constituted a partial sample as defined in IP 71111.1505 since all the corrective actions to support continued operability and resolution of the nonconforming conditions had not been identified. These actions and any underlying technical violations will be addressed with the completion of this inspection sample.
05000456/FIN-2016002-0330 June 2016 23:59:59BraidwoodNRC identifiedMissed Radiological Environmental Monitoring Program SamplingThe 2015 Braidwood Annual Radiological Environmental Operating Report for 2015 identified missed food samples in three out of four quadrants where food products were required by the licensees ODCM. The inspectors also noted that this issue was identified in the licensees CAP as Action Request 0214924, dated January 20, 2016, but it did not appear that any action was taken in 2015 to identify suitable alternative sampling locations. Discussion: The assessment of the issue could not be completed within this inspection period. In particular, the licensees ODCM is a site-specific document that included the radioactive effluent controls and the associated radiological environmental monitoring activities used to validate those controls. At the end of this inspection, the inspectors had not had the opportunity to review the bases documents for the ODCM to better understand the site-specific dose pathways for airborne and liquid effluent receptors and to assess the impact of these missed samples. Specifically, it was not clear whether the intended food product samples were designed to validate the airborne effluent control program or the liquid effluent control pathway. Each pathway has a different or unique requirement for validating the effluent controls. For example, validation of the airborne effluent control program would frequently measure, throughout the growing season, the radioactive material deposited on the sample surface and validation of the liquid effluent control pathway would measure, at the time of harvest, the radioactive material incorporated into the food product through irrigation. The issue remains under review by the NRC to determine the adequacy of ODCM performance and whether any violation of regulatory requirements occurred. This issue is categorized as an Unresolved Item (URI) pending completion of this review. (URI 05000456/201600203; 05000457/201600203; Missed Radiological Environmental Monitoring Program Sampling)
05000457/FIN-2016002-0230 June 2016 23:59:59BraidwoodNRC identifiedFailure to Manage Gas Accumulation in the 2A SI TrainThe 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, for the failure to manage gas accumulation in the safety injection (SI) system in accordance with procedure ERAA2009, Managing Gas Accumulation. Specifically, following identification of a void in the 2A SI train, the licensee failed to increase the monitoring frequency and account for the potential for the void to grow due to active gas mechanisms or planned evolutions, as required by the procedure. This ultimately led to a previously identified void growing beyond the pre-established limit by the next scheduled surveillance. Corrective actions for this issue included a planned action to establish an increased monitoring frequency for the affected line, and an action to remove the void in the upcoming Unit 2 Outage (Spring 2017). This issue was entered into the licensees CAP as IR 2640751. The inspectors determined the performance deficiency was more than minor because, it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to monitor the gas accumulation for the 2A train of SI at the appropriate frequency did not ensure the availability and reliability of the SI system to perform its accident mitigating function. Additionally, this failure led to the 2A SI train exceeding the associated operability limits as established by evaluation BW150100M during the next scheduled surveillance. The inspectors determined that this finding was of very low safety significance because it did not result in the loss of operability or functionality of mitigating systems. Specifically, an engineering evaluation reasonably determined that the non-conforming condition did not result in a loss of operability. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance because the licensee did not stop when faced with uncertain conditions. Specifically, the licensee did not reassess the gas accumulation monitoring plan to consider the potential for void growth due to active gas mechanisms or planned evolutions when accepting an unexpected void condition that differed with the initial conditions assumed by the monitoring plan. Ultimately, this led to a monitoring plan not being implemented as required (H.11).
05000456/FIN-2016008-0131 March 2016 23:59:59BraidwoodNRC identifiedFailure to Verify the Tripping Characteristic of Molded Case Circuit Breakers (MCCBs) Used as Isolation Devices for the 120 Vac Instrument Power SystemThe inspectors identified a finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to test the 120 Vac molded case circuit breakers (MCCBs) used as isolation devices on the instrument power system. Specifically, although the licensee had committed to test circuit breakers used as isolation devices in response to Final Safety Analysis Report Question 40.73 in 1982, there was no evidence that these MCCBs had ever been tested. The licensee subsequently entered the issue into its Corrective Action Program. The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and affected the cornerstone objective of ensuring the availability of the safety-related instrument power system. Specifically, the licensee did not assure, by periodically verifying the time-current characteristic of the MCCBs, that the isolation devices would perform their safety function to isolate the nonsafety-related instrument bus from the safety-related instrument power bus before the safety bus could be affected by a fault on the nonsafety-related load. The inspectors determined that the finding was of very-low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that there was no cross-cutting aspect associated with this finding because the finding was not indicative of the licensees current performance.
05000456/FIN-2016008-0231 March 2016 23:59:59BraidwoodNRC identifiedFailure to Verify Air Intake for Diesel Driven Auxiliary Feedwater Pump was Adequately Protected from a High Energy Line BreakThe inspectors identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the adequacy of the diesel driven Auxiliary Feedwater (AFW) pump design. Specifically, the licensee failed to verify the diesel driven AFW pump could perform its safe shutdown function following a high energy line break (HELB) in the Turbine Building. Since the diesels air intake was located in the Turbine Building, it would be impacted by a HELB. The licensee entered this issue into its Corrective Action Program and took immediate corrective actions by declaring the diesel driven AFW pump inoperable and then implementing a temporary plant modification to relocate the diesel air intake to the Auxiliary Building where it is not susceptible to a HELB to restore operability of the pump. The licensees planned corrective actions are to complete a permanent plant modification to relocate the air intake to a location that is not susceptible to a HELB. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify that the diesel driven AFW pump could perform its safety function following a HELB event in the Turbine Building did not ensure its availability, reliability, and capability to respond to the initiating event. Since the finding did represent an actual loss of function of at least a single Train for greater than its Technical Specification Allowed Outage Time, a Detailed Risk Evaluation was performed which concluded that the estimated change in core damage frequency was approximately 3.4E-7/yr., which represents a finding of very-low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not indicative of the licensees current performance.
05000456/FIN-2016001-0431 March 2016 23:59:59BraidwoodNRC identifiedQuestions Regarding the Implementation of the Gas Accumulation ProgramQuestions Regarding the Implementation of the Gas Accumulation Progra The inspectors identified an URI regarding the implementation of the Gas Accumulation Program at Braidwood. Specifically, the inspectors were concerned with whether a number of surveillance frequencies that were contained in the Surveillance Frequency Program meet the requirements as specified in procedure ERAA2009, Managing Gas Accumulation. Additionally, the inspectors were concerned with the basis for not increasing the frequency of the UT examinations following the discovery of a void on October 20, 2015. At the end of the inspection period, the licensees investigation on the cause of an unexpected void growth, and the potential surveillance frequency discrepancies was ongoing. Resolution of this issue will be based on the inspectors review of the licensees completed investigation. On March 15, 2016, while performing a semi-annual gas monitoring surveillance on Unit 2 under 2BwOSR 3.2.22, ECCS and Containment Spray Venting and Valve Alignment/UT Verification Surveillance, a gas void was found along line 2SI03BA, which is a SI line that feeds the A and D SI hot leg injection lines. The ultrasonic examination revealed that a 0.960 cubic foot void was present. A void had been previously identified in the same location on October 20, 2015, which had a volume of 0.25 cubic feet. Calculation BRW150100M was performed in October 2015 to justify operability of the SI system. The calculation produced a void size acceptance criteria of 0.389 cubic feet. Upon identification of the void in March 2016, the licensee declared the 2A SI train inoperable due to the previously established acceptance criteria of 0.389 cubic feet not being met, and entered LCO 3.5.2, ECCS Operating, Condition A, which required that the affected train be restored to an operable status within 7 days. The licensee exited the LCO on March 16, 2016 upon completion of a revision to calculation BRW150100M, which documented a revised acceptance criteria of 1.5 cubic feet. During this inspection period, the inspectors reviewed the licensees revision to the aforementioned calculation, and the requirements contained in procedure ERAA2009. Based on their review, the inspectors questioned the basis for not increasing the frequency of the UT examinations following the discovery of the void on October 20, 2015. Additionally, the inspectors were concerned with the frequency of inspection of a number of locations outside the missile barrier (17 for Unit 1 and 19 for Unit 2), which appeared to conflict with what was specified in the procedure. Specifically, the locations in question were examined at an 18 month frequency, although the procedure stated that frequency of once per refueling outage shall be used only for locations that are inaccessible due to actual (not just posted) high radiation conditions. Finally, the inspectors had a concern regarding the means by which gas accumulation was managed for locations inside the missile barrier, since the prescribed locations were only monitored once upon Mode ascension from an outage. The licensee entered the inspectors concerns into their CAP as IR 2644532 and IR 2640751. At the conclusion of the inspection, two work group evaluations were in progress to: 1) address the void growth observed since October 2015, and 2) evaluate the compliance with the program document procedure, ERAA2009. This URI will remain open until the evaluations are completed and the inspectors review the evaluations to determine whether a performance deficiency exists. (URI 05000456/20160104; 05000457/201600104; Questions Regarding the Implementation of the Gas Accumulation Program)
05000456/FIN-2016001-0131 March 2016 23:59:59BraidwoodNRC identifiedFailure to Follow Fire Prevention for Hot Work ProcedureThe inspectors identified a finding of very low safety significance and an associated NCV of License Condition 2.E when licensee personnel failed to follow the requirements of the Fire Prevention for Hot Work procedure on two separate occasions. Specifically, (Issue 1) on February 2, 2016, a very small fire occurred during a planned hot work activity that involved pipe grinding on a small waste gas decay tank pressure line because the licensee failed to recognize the potential for hydrogen within the line. Additionally, (Issue 2) on February 25, 2016, the inspectors identified that a hot work permit was inadequate prior to the licensee performing a piping weld repair activity associated with the Unit 2 main generator stator cooling water system because the permit referenced the wrong work location and did not require appropriate controls. These issues were entered into the licensees Corrective Action Program (CAP) as Issue Reports (IRs) 2620772 and 2632182. The inspectors determined that the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown and power operations. Specifically, for Issue 1, the performance deficiency resulted in the occurrence of a small hydrogen fire in the auxiliary building. For Issue 2, the performance deficiency increased the likelihood of a fire occurring during an emergent weld repair in the turbine building. The inspectors determined that this finding was of very low safety significance (Green) because the fire (Issue 1) and increased likelihood of a fire occurring (Issue 2) was limited to equipment which was not important to safety. The inspectors determined that the finding had a Work Management cross-cutting aspect in the Human Performance area. Specifically, a significant contributor to the performance deficiency was related to the organization not implementing a process for planning, controlling, and executing work activities such that nuclear safety is the overriding priority (H.5).
05000457/FIN-2016001-0231 March 2016 23:59:59BraidwoodSelf-revealingFailure to Have Adequate Work Instructions and Procedures Leads to a Loss of Inventory From the Volume Control TankA finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on February 1, 2016, when licensee personnel failed to have appropriate work instructions for performing planned motor-operated valve (MOV) 2SI8807A diagnostic testing. Specifically, the work order (WO) used did not provide appropriate instructions to ensure that the proper equipment line-up for the test was established prior to stroking the valve. Ultimately, this led to an unplanned transfer of about 304 gallons of water from the volume control tank (VCT) to the refueling water storage tank (RWST). This issue was entered into the licensees CAP as IR 2620523. The inspectors determined that the performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical functions during shutdown and power operations. Specifically, the failure to have an appropriate procedure for a maintenance activity led to 304 gallons of inventory being diverted to the RWST. The finding screened as having very low safety significance (Green) because it was determined that the reactor coolant system (RCS) leak rate for a small loss of coolant accident was not exceeded, and it did not result in a loss of a mitigating systems ability to perform an intended safety function. The inspectors determined that the finding had a Work Management cross-cutting aspect in the Human Performance area because the licensee did not implement a process of planning, controlling and executing work activities such that nuclear safety is an overriding priority. Specifically, proper work planning and coordination between maintenance and operations would have ensured that the WO being utilized established the proper system line-up prior to the start of the maintenance (H.5).
05000456/FIN-2016001-0331 March 2016 23:59:59BraidwoodSelf-revealingFailure to Correct a Condition Adverse to Quality Leads to Loss of One Train of Shutdown Cooling in Mode 6A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, was self-revealed when the licensee failed to ensure that a condition adverse to quality was promptly identified and corrected. Specifically, on October 8, 2015, valve 2RH606 failed to open and caused a loss of one train of shutdown cooling in Mode 6 and an unplanned orange risk condition. The reason for the failure was improper use of a lower strength carbon steel valve key instead of the specified high strength hardened steel valve key, which had been the subject of a vendor Part 21, Reports of Defects and Non Compliance, Report. This issue was entered into the licensees CAP as IR 2567811. The inspectors determined that the performance deficiency was 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. Specifically, the failure to correct a condition adverse to quality in the form of the improper use of a lower strength carbon steel key instead of the specified high strength hardened steel key in a safety-related valve ultimately led to a loss of one train of shutdown cooling in Mode 6. The inspectors determined that the finding was of very low safety significance based upon a detailed risk evaluation. The inspectors did not identify a cross-cutting aspect associated with this finding because the performance deficiency was greater than three years old and therefore was not indicative of current performance.
05000456/FIN-2016001-0531 March 2016 23:59:59BraidwoodNRC identifiedFailure to Ensure Unit 2 Startup Feedwater Pump AvailabilityThe inspectors identified a finding of very low safety significance when licensee personnel failed to ensure that the Unit 2 startup feedwater pump (SUFWP) was available during an 18 month operating cycle. Specifically, the licensee had failed to ensure that the pump oil pressure regulator was properly adjusted, and had failed to perform a post-maintenance test following on-line work in a manner to ensure that no new deficiency was introduced. The license entered this issue into their CAP as IR 2565442. Corrective actions consisted of updating the station SUFWP model work orders (WOs) to ensure that interlock continuity checks were performed as a part of the post-maintenance testing when necessary, and to include procedural steps to verify lube oil pressure when starting a SUFWP. The inspectors determined that the performance deficiency was more than minor because the issue was associated with the Procedural Quality 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. Specifically, the Unit 2 SUFWP is a backup method of decay heat removal following a reactor trip, and is utilized in plant startup and shutdown procedures. A detail risk evaluation was performed and the performance deficiency was determined to be of very low safety significance based upon an evaluation bounding the risk to a Delta Core Damage Frequency (CDF) of 2.9E7/year. No cross-cutting aspect was identified because the cause of the failure were probable causes and not confirmed to be the actual cause.
05000456/FIN-2015004-0431 December 2015 23:59:59BraidwoodNRC identifiedFailure to Establish a Written Procedure for a Loss of Feedwater Event in Mode 3A finding of very low safety significance and an associated NCV of Technical Specification 5.4.1, Procedures, was self-revealed on October 5, 2015, due to the licensees failure to establish a written procedure for combating emergencies and other significant events, as required by Regulatory Guide 1.33, Quality Assurance Program Requirements. Specifically, upon a loss of feedwater in Mode 3 (Hot Standby), which is an expected design and licensing basis event, the licensee did not have a written procedure as established by the Regulatory Guide. This issue was entered into the licensees CAP as IRs 2566239 and 2565513. The inspectors determined the finding to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, it was associated with the Mitigating Systems cornerstone Procedural Quality attribute, and adversely impacted the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the absence of a procedure(s) complicated the operator response to the loss of feedwater event in Mode 3. The inspectors determined the finding to be of very low safety significance in accordance with IMC 0609, Appendix A, The SDP for Findings at Power, dated September 7, 2012, Exhibit 2, since the inspectors answered "No" to the Mitigating Systems questions under Section A, Mitigating Systems, Structures, and Components and Functionality. The inspectors did not identify a cross-cutting aspect associated with this finding, because it was confirmed not to be reflective of current performance due to the age of the performance deficiency.
05000457/FIN-2015004-0131 December 2015 23:59:59BraidwoodNRC identifiedLoss of Shutdown Cooling Train During Refueling Cavity Fill and Associated Reduced Inventory OperationsOn October 8, 2015, the inspectors identified an Unresolved Item (URI) regarding the failure of valve 2RH606, which is the 2A RHR heat exchanger flow control valve. The valves failure to open caused a loss of one train of shutdown cooling, and an unplanned Orange risk configuration with Unit 2 in Mode 6, and the reactor refueling cavity level less than 23 feet above the vessel flange. At the closure of the inspection period, the licensees investigation on the cause of the failure was ongoing. Resolution of this issue will be based on the inspectors review of the licensees completed investigation. A function of the RHR system in Mode 6 is to remove decay heat and sensible heat from the reactor coolant system (RCS). Heat is removed from the RCS by circulating reactor coolant through the RHR heat exchangers where the heat is transferred to the component cooling water system. The coolant is then returned to the RCS via the RCS cold legs. On October 8, 2015, valve 2RH606 became mechanically bound while in the process of filling the Unit 2 reactor refueling cavity to greater than 23 feet. This was identified when the operators attempted to open the valve from the control room. The failure of the valve to open caused Unit 2 shutdown risk to change from a planned Yellow configuration to unplanned Orange condition. Additionally, the licensee entered Limiting Condition for Operation 3.9.6, Residual Heat Removal and Coolant Recirculation-Low Water Level, Condition A, for one train of RHR cooling inoperable. This action required the licensee to initiate actions immediately to either restore the affected RHR loop to operable status or to initiate actions to establish greater than or equal to 23 feet of water above the reactor vessel flange. The licensee accomplished this action by raising water level in the cavity to greater than 23 feet. Troubleshooting of the failed valve revealed that a shaft key sheared, which prevented the valve from opening. The valve had been previously manipulated during the outage without an issue. The malfunctioning part was sent offsite for failure analysis. The valve was repaired. At the conclusion of the inspection, an apparent cause investigation was in process. This URI will remain open until the investigation is complete and the inspectors review the report to determine whether a performance deficiency exists.
05000456/FIN-2015004-0331 December 2015 23:59:59BraidwoodNRC identifiedFailure to Establish Adequate Feedwater Pump Operational Guidance During a Normal Plant ShutdownA finding of very low safety significance and an associated NCV of Technical Specification 5.4.1, Procedures, was self-revealed on October 5, 2015, due to the licensees failure to establish and maintain adequate guidance for operating the Unit 1 and Unit 2 motor driven main feedwater pump (MDFWP) during plant shutdown conditions. Specifically, on October 4, 2015, during a Unit 2 plant shutdown, the Unit 2 MDFWP was placed in service at low forward feedwater flow conditions and was manually tripped when the pumps main journal bearing temperature exceeded the procedural limit. Subsequent review, determined that the procedural limit was too low as previously recognized by historic station specific operating experience. This issue was entered into the licensees corrective action program (CAP) as Issue Report (IR) 2565486. The inspectors determined that the performance deficiency was more than minor because the issue was associated with the Procedural Quality attribute of the Initiating Event cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the performance deficiency contributed to a loss of main feedwater event that upset plant stability and challenged the critical safety function of removing decay heat via the steam generators in Mode 3. For Unit 1, the increased potential for a loss of main feedwater event existed under similar conditions. The inspectors determined that the finding was of very low safety significance based upon a detailed risk evaluation. The inspectors concluded that this finding did not have a cross-cutting aspect because the performance deficiency was greater than 3 years old and, therefore, not indicative of recent performance.
05000456/FIN-2015407-0131 December 2015 23:59:59BraidwoodLicensee-identifiedLicensee-Identified Violation
05000457/FIN-2015004-0231 December 2015 23:59:59BraidwoodNRC identifiedFailure of Startup Feedwater Pump to Start During Plant ShutdownThe inspectors identified an URI based upon the startup feedwater pumps (SUFWPs) failure to start during a plant shutdown. In addition to being used in plant startups and shutdowns, the SUFWP is also credited in the licensees emergency operating procedure as a means to add water to the steam generators for decay heat removal if the safety-related auxiliary feedwater systems failed to function properly during an event. On October 4, 2015, operations attempted to start the Unit 2 SUFWP at low power in Mode 1 during plant shutdown activities for a refueling outage. Upon start, the SUFWP automatically tripped. The licensee completed an apparent cause evaluation to determine the reason why the pump did not start and run. At the end of the inspection period, the inspectors were awaiting additional information to complete their review to determine if this issue of concern constituted a performance deficiency. This URI will remain open pending this review.
05000456/FIN-2015007-0130 September 2015 23:59:59BraidwoodNRC identifiedFailure to Ensure that Circuits Associated with Pressurizer PORVs and Block Valves Were Free of Fire DamageThe inspectors identified a finding of very low safety significance, and an associated NCV of the Braidwood Station facility operating license condition 2.E associated with the Fire Protection Program for the licensees failure to ensure that the safe shutdown capability was independent of the fire area and thus free of fire damage. Specifically, in the event of a fire in the control room, cable spreading rooms, or electrical cable penetration areas the circuits associated with the Pressurizer Power Operated Relief Valve (PORV) block valves, which are relied upon to safely shutdown the plant, could be affected and may not be available due to fire-induced failures. The licensee entered this issue into their Corrective Action Program, established fire watches, and intended to perform plant modifications to correct the issue. The inspectors determined that the issue was more than minor because fire-induced circuit failures could impair the operation of the PORV block valves and complicate shutdown of the plant in the event of a fire in the control room, cable spreading rooms, or electrical cable penetration areas. The finding affected the Mitigating Systems Cornerstone. The finding was determined to be of very low safety significance based on a detailed risk-evaluation by a Region III Senior Reactor Analyst. This finding was not associated with a cross-cutting aspect because the finding was not representative of the licensees current performance.
05000456/FIN-2015003-0130 September 2015 23:59:59BraidwoodLicensee-identifiedLicensee-Identified ViolationTitle 10 of the Code of Federal Regulations, 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 October 10, 2014, the licensee failed to translate the design basis essential service cooling pond berm height into procedures and instructions. Specifically, procedure BwVSR 3.7.9.3, "Braidwood Cooling Lake Hydrographic Survey," did not ensure that the height of the essential service cooling pond berm was being verified. This issue was entered into the licensees CAP as IR 2400960; The UHS EL. At Top of the East Slope Found Less than 590 ft, dated October 24, 2014, 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, following a seismic event that drains the non-essential main Braidwood cooling pond, the essential cooling pond (i.e., UHS) would have a decrease in available inventory at the start of a design basis event. This could reduce the available net positive suction head for the service water pumps that take suction from the UHS, as well as potentially resulting in the UHS design temperature of 100 degrees Fahrenheit being exceeded. The inspectors determined the finding could be evaluated using the Significance Determination Process 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-2015003-0230 September 2015 23:59:59BraidwoodLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions. Procedures, and Drawings, requires, in part, that activities affecting quality shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Contrary to the above, from April 23, 2015, to June 24, 2015, the licensee failed to translate specific acceptance criteria into procedures and instructions. Specifically, when the licensee modified the DG fuel oil system in a manner that reduced the DG fuel oil system train separation from two isolation points to one isolation point, the licensee failed to establish quantifiable acceptance criteria in the post-maintenance test, and failed to establish performance monitoring with acceptance criteria, as specified in the design change. This issue was entered into the CAP as IR 2519208 with immediate corrective actions of re-establishing the dual isolation point. 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). 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 was of very low safety significance (Green), because the issue did not prevent the 2B DG from operating for its specified probable risk assessment mission time of 24 hours.
05000456/FIN-2015002-0230 June 2015 23:59:59BraidwoodNRC identifiedMechanic Joint Leakage Accepted for Continued Service Without Code Corrective ActionsThe 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, for the licensees failure to follow a procedure for completing an American Society of Mechanical Engineers (ASME) Section XI Code pressure test. Specifically, the licensee failed to implement the required corrective actions or evaluations for evidence of leakage (boric acid deposits) identified on a containment spray (CS) system valve bolted connection prior to returning this component to service. The licensee entered this issue into their CAP and initiated actions to clarify procedures to ensure the ASME Code Section XI, Paragraph IWB-3522, requirements were implemented, and components with Code relevant conditions were corrected or evaluated prior to returning them to service. The performance deficiency was determined to be more than minor in accordance with IMC 0612, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the failure to adhere to procedure ER AA-330-001 was based upon the licensees decision to return a component exhibiting evidence of boric acid leakage to service without Code corrective measures or evaluation. Additionally, this type of error could result in inservice failure of equipment. Therefore, this finding affected the Mitigating Systems Cornerstone attribute of Equipment Performance 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 (Green), because the licensees failure to adhere to procedure ER AA-330-001 and remove valve 1CS011B from service with a Code relevant condition did not result in operation of the plant with an inoperable system or component. Therefore, the inspectors answered Yes to Question A.1 of Exhibit 2, Mitigating Systems Screening Questions, identified in Appendix A of IMC 0609, and the finding screened as having very low safety significance. The inspectors identified a cross-cutting aspect associated with this finding in the area of Human Performance, Conservative Bias because the licensee staff did not use a decision-making practice that emphasized prudent choices over those that were simply allowable. Specifically, the failure to remove valve 1CS011B from service with a relevant condition was based upon the licensees decision that this was an allowable option because the ASME Code Section XI paragraph was not clear. (H14)
05000456/FIN-2015002-0330 June 2015 23:59:59BraidwoodSelf-revealingControl Room Chiller Inoperability Due to High Oil Content in the RefrigerantA finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on April 28, 2015, when licensee personnel failed to establish adequate procedural controls related to how much oil could be added or removed from the control room chillers without affecting its functionality. Specifically, the 0A control room ventilation (VC) chiller was declared inoperable due to high oil content in the refrigerant, which caused reduced cooling efficiency to the point of non-functionality. The licensee entered this issue in their CAP, restored the 0A VC chiller to operable status on May 1, 2015, and performed an evaluation to establish the acceptable level of oil migration to retain functionality of the VC chiller. The performance deficiency was determined to be more than minor in accordance with IMC 0612, because, it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences (i.e. core damage). The finding screened as having very low safety significance (Green), because it did not result in the loss of safety function, and did not result in an actual loss of function of at least a single train for greater than its technical specification allowed outage time. The inspectors determined that the associated finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because the licensee staff did not implement effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, following three documented issues with VC chiller performance, Engineering determined that the issues were related to high oil content in the chiller refrigerant. Based on this information, corrective actions related to optimizing refrigerant/oil levels in the chiller were recommended to the Plant Health Committee, which were approved for immediate implementation. However, the actions were not appropriately incorporated into the work control process or the CAP, which led to them not being implemented in a timely manner. (P.3)
05000456/FIN-2015404-0130 June 2015 23:59:59BraidwoodNRC identifiedSecurity
05000456/FIN-2015002-0130 June 2015 23:59:59BraidwoodNRC identifiedFailure to Update the Final Safety Analysis Report-Thimble Tube Inspection ProgramThe inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.71(e), Periodic Update of the Updated Final Safety Analysis Report (UFSAR), and an associated Green finding for the licensees failure to update the UFSAR with a description of the Thimble Tube Inspection Program to reflect information submitted to the NRC in response to NRC Bulletin 88-09. Specifically, the licensee did not update Section 5.2.4, Inservice Inspection and Testing of Reactor Coolant Pressure Boundary, of the UFSAR to include the Incore Thimble Tube Inspection Program, which provided the basis for leakage integrity for this portion of the reactor coolant pressure boundary. The licensee entered this issue into their Corrective Action Program (CAP) and identified a recommended action to incorporate the Incore Thimble Tube Inspection Program into the UFSAR. The performance deficiency was determined to be more than minor in accordance with IMC 0612, "Power Reactor Inspection Reports," Appendix B, "Issue Screening," dated September 7, 2012, because, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, the failure to update the UFSAR with the Thimble Tube Inspection Program could result in reductions or elimination of the program without seeking prior NRC approval and insufficient thimble tube inspections could also result in the failure to detect thimble tube wear prior to an un-isolable leak in the reactor coolant pressure boundary. Additionally, the failure to update the UFSAR was more than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and adversely affected the Cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions. The finding screened as having very low safety significance (Green), because the licensees failure to update the UFSAR with a description of the Thimble Tube Inspection Program had not resulted in degradation of a thimble tube such that the reactor coolant system leak rate for a small break loss of coolant accident was exceeded and did not affect systems used to mitigate a loss of coolant accident. Therefore, the inspectors answered No to Questions A.1 and A.2, of Exhibit 1, Initiating Events Screening Questions, identified in Appendix A of IMC 0609 and the finding screened as having very low safety significance. Violations of 10 CFR 50.71(e) are dispositioned using the traditional enforcement process, because they are considered to be violations that potentially impede or impact the regulatory process. In accordance with Sections 6.1.c.7 and 6.1.d.3 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensees failure to update the UFSAR as required by 10 CFR 50.71(e) had not yet resulted in an unacceptable change to the facility (e.g. thimble tube structural integrity was maintained) or procedures and the associated finding was of very low risk significance. The finding was the result of an error made in excess of 10 years ago, and thus was not indicative of current licensee performance. Therefore, no cross-cutting aspect was identified.
05000456/FIN-2015001-0231 March 2015 23:59:59BraidwoodNRC identifiedFailure 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-2015001-0431 March 2015 23:59:59BraidwoodLicensee-identifiedLicensee-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-2015001-0331 March 2015 23:59:59BraidwoodSelf-revealingFailure 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-2015001-0131 March 2015 23:59:59BraidwoodNRC identifiedFailure 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-2014005-0331 December 2014 23:59:59BraidwoodNRC identifiedFailure 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-2014005-0231 December 2014 23:59:59BraidwoodSelf-revealingFailure 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).