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05000346/FIN-2017004-012017Q4Davis BesseFailure to Maintain Procedures Associated with Ventilation Air Monitoring Assessment ProgramThe inspectors identified a finding of very-low safety significance and an associated NCV of Technical Specification 5.4.1 for the failure to maintain procedures for station vent releases during planned scenarios. Specifically, the inspectors identified multiple procedures that were not updated when the station vent monitors were replaced in 2014. This issue has been entered into the licensees Corrective Action Program as CR201710817. Corrective actions taken included the issuance of a Standing Order for collecting samples during accident conditions, provided Just-In-Time training for chemistry technicians, and revision of the outdated procedures. The performance deficiency was determined to be more-than-minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening. Specifically, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern in that the failure to maintain procedures to collect station vent samples under all predicted conditions could result in the inability to measure the amount of gaseous radioactivity leaving the plant and to ensure 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. The finding was assessed using Inspection Manual Chapter 0609 Appendix D, Public Radiation Safety Significance Determination Process, and was determined to be of very-low safety significance because the issue involved radioactive effluent releases, 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, Part 20.1301(e) limits. The inspectors determined that the finding had a cross-cutting component in the area of Human Performance, in the aspect of Work Management: specifically, the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. (H.5)
05000346/FIN-2017004-022017Q4Davis BesseInterface Between New Accident Range Ventilation Monitors and the Emergency Preparedness Dose Assessment ProgramDuring inspection activities associated with the accident range station vent monitor, the inspectors identified an unresolved item (URI) associated with the interface between the monitor and the Dose Assessment Program used to project dose to members of the public during potential accident conditions. Description: The licensee replaced the accident range station vent monitors in 2014 using ECP 040006, Replace Kaman Radiation Monitors. The replacement monitors were manufactured by a different company than the original monitors, had different detection capabilities, different system calibration, and different computer hardware to convert detector output into usable information. The licensee could not immediately provide specifics regarding the interface between the new accident range monitors and the program used during accident conditions for providing dose projections and the resulting protective action recommendations. The inspectors focus of concern was how the new accident range monitors accounted for the potentially rapidly changing mixture of radioactive gases during the early phase of a postulated accident. Consequently, this issue remains under review by the NRC awaiting for additional information from the licensee to verify the new monitor interface to determine if it represents a performance deficiency and is categorized as a URI. (URI 05000346/201700403, Interface Between New Accident Range Ventilation Monitors and the Emergency Preparedness Dose Assessment Program)
05000346/FIN-2017004-032017Q4Davis BesseFailure to Prescribe Appropriate Work Instructions for an Activity Affecting QualityA self-revealed finding with an Apparent Violation (AV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and an associated violation of technical specification (TS) 3.7.5, Emergency Feedwater (EFW), was identified on September 13, 2017, due to the licensees apparent failure to prescribe appropriate work instructions for an activity affecting quality of the safety-related auxiliary feedwater (AFW) system. Specifically, the licensee apparently did not provide appropriate instructions to maintain an adequate amount of oil in the AFW turbine bearing oil sumps, resulting in the failure of AFW 1 on September 13, 2017. The licensee entered this issue into the CAP as CR201709443 and CR201709857, immediately replaced the damaged bearing, and updated the lubrication manual data sheets to include sight glass marking dimensions per vendor guidance. The apparent performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and potentially adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that respond to initiating events. Specifically, the apparent performance deficiency resulted in the failure of the AFW 1. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609 Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the mitigating systems cornerstone. The inspectors determined the finding represented an apparent actual loss of function of at least a single train for greater than its technical specification allowed outage time. Therefore, a detailed risk evaluation will be performed by a regional senior reactor analyst. Because the safety characterization of this finding is not yet finalized, it is being documented with a significance of to be determined (TBD). The inspectors determined this finding affected the cross-cutting aspect of challenge the unknown in the area of Human Performance, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, licensee personnel apparently did not stop when faced with uncertain conditions in the preventive maintenance procedure for replacing the AFPT sight glasses. Although the replacement of the AFPT 1 inboard bearing sight glass occurred in 1997, the licensee had the opportunity to challenge the lack of detail in the work instructions in late 2014 when the AFPT 2 outboard bearing sight glass was replaced. (H.11)
05000346/FIN-2017004-042017Q4Davis BesseFailure to Document a Degraded Condition on the AFPT 1 Outboard BearingThe inspectors identified a finding of very low safety significance for the licensees failure to document a degraded condition of a safety-related system in the corrective action program (CAP), as required by licensee procedure, NOPLP2001. Specifically, during planned maintenance on auxiliary feedwater pump turbine (AFPT) 1, the licensee identified scoring on the outboard turbine bearing and failed to generate a condition report detailing the issue. The licensee entered this issue into the CAP as condition report (CR) 201712487 for evaluation. The inspectors determined the performance deficiency was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, the failure to document a degraded condition in the CAP did not allow the organization to properly assess the issue. Therefore, the underlying cause may not have been appropriately addressed. Using IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014, the inspectors determined the finding to be of very low safety significance (Green) because the inspectors answered no to all questions in Exhibit 3 of Appendix G, Attachment 1. The inspectors determined this finding affected the cross-cutting aspect of identification in the area of Problem Identification and Resolution, where the organization implements a corrective action program with a low threshold for identifying issues and individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to completely identify the degraded condition, resulting in the failure to document the issue. (P.2)
05000461/FIN-2017009-012017Q3ClintonFailure to Evaluate Replacement Relay Dropout VoltagePreliminary White. A self-revealed finding preliminarily determined to be of low to moderate safety significance, and an associated apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, was identified on March 9, 2017, for the licensees failure to implement measures for the selection and review for suitability of application replacement relays for the Division 1 Emergency Diesel Generator (EDG) Room Vent Fan, which were components subject to the requirements of 10 CFR Part 50, Appendix B. Specifically, Engineering Changes 330624 and 366624 failed to evaluate the change in the actual drop out voltages for replacement relays on the associated fan circuitry, and instead, introduced new relays into the circuit that resulted in the failure of the fan to operate during an under voltage condition. This rendered the Division 1 EDG inoperable for a time longer than its technical specification allowed outage time, which was a violation of Technical Specification 3.8.1, AC SourcesOperating. The licensee entered this issue into the corrective action program as action request (AR) 03982792. Corrective actions for this issue included restoring the circuit to allow the ventilation fan to operate and returning the emergency diesel generator to an operable condition. The inspectors determined that the licensees failure to verify the suitability of the replacement relays for the Division 1 EDG room vent fan was contrary to the requirements of 10 CFR Part 50, Appendix B, Criterion III and a performance deficiency which was within the licensees ability to foresee and correct. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of the systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify the suitability of the replacement relays prior to installation in the Division 1 EDG room vent fan circuitry resulted in the inoperability and unavailability of the Division 1 EDG from May 18, 2016 to March 11, 2017, when one of the unsuitable relays was replaced. Using IMC 0609, Appendix A, Significance Determination Process for 3 Findings At-Power, dated June 19, 2012, a Significance and Enforcement Review Panel preliminarily determined the finding to be of low to moderate safety significance. The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of challenge the unknown, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, a questioning attitude was not used to understand the consequence of the differences in relay features resulting with installing a relay that was incompatible with the current design. (H.11)
05000461/FIN-2017003-012017Q3ClintonMSIV TS Leakage Limits Exceeded Due to Condition Based Maintenance ApproachThe inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS limiting condition for operation (LCO) 3.6.1.3, for the failure to follow station procedure ERAA200, Preventative Maintenance Program, Revision 3. Specifically, the licensee utilized a condition-based maintenance approach on the main steam isolation valves (MSIVs) that failed to monitor and trend equipment performance so that planned maintenance could be performed prior to the MSIVs exceeding the TS leakage limits. The licensee entered this issue into their CAP as AR 04009845. As corrective actions, the licensee repaired and tested the valves prior to returning the unit to the modes of applicability.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 it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the monitoring and trending of local leak rate tests on the MSIVs did not provide performance data that would allow planned maintenance to the valves prior to the valves failing resulting in exceeding TSleakage requirements for the MSIVs. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone Reactor Containment and did represent an actual open pathway in the physical integrity of reactor containment. The inspectors proceeded to Appendix H, Containment Integrity Significance Determination Process, and determined that it was a Type B finding that was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. The inspectors used Figure 6.1, Road Map for LERF based Risk Significance for Evaluation of Type-B Findings at Full Power and determined this finding is of very low safety significance (Green). The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of design margins, where the organization operates and maintains equipment within design margins. Special attention is placed on maintaining fission product barriers, defense-in-depth and safety related equipment. Specifically, the procedure for testing the MSIVs utilized an administrative limit that provided no margin to correct performance prior the valves becoming inoperable. (H.6)
05000461/FIN-2017003-022017Q3ClintonFailure to Adequately Control Access in Locked High Radiation AreaA finding of very low safety significance and an associated NCV of TS 5.4.1 was self-revealed when individuals failed to adequately control access in locked high radiation areas (LHRAs). Specifically, the failure to meet all of the requirements of Procedure RPAA460, Attachment 5, represented a failure to comply with Radiation Work Permit CL1700518, C1R17 (Drywell) DW Bioshield Inservice Inspection Activities. This resulted in four individuals entering a LHRA that they had not been specifically authorized to enter. These individuals entered the incorrect location and were inside the area for approximately 2-3 minutes before they noticed that they were in the incorrect area. The individuals knew that they were in the incorrect location when they could not find the nozzles that they planned on inspecting. The individuals exited the area and were simultaneously told to exit the area by the radiation protection technician (RPT) providing remote coverage which demonstrated that the four workers were not in the authorized work area. Immediate corrective actions taken by the licensee included immediately suspending the work that was scheduled to take place within the bioshield associated with this job. Electronic dosimeters and dosimeters were immediately collected from the individuals that entered the area so the dose that was received could be known. The licensee also interviewed all the individuals that were involved in this bioshield entry, and the RPT that performed the brief. These interviews were conducted to understand which parts of the process associated with entry into LHRAs failed and led to this event transpiring. The licensee entered this event into their CAP as AR 04012075. As corrective actions the licensee planned to observe high radiation area and locked high radiation area briefs, for both in house and traveling RPTs. The licensee also planned to modify the bioshield as-low-as-reasonably-achievable (ALARA) plan template to label all accessible bioshield doors with elevation and azimuth.The inspectors determined that the performance deficiency was more-than-minor in accordance with IMC 0612, Appendix B, because the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the workers entered an area that required the radiation dosimeter to be relocated to the workers knee, and the workers were wearing them on the head for the intended work location. The finding was determined to be of very-low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls; (2) there was no overexposure; (3) there was no substantial potential for an overexposure; and (4) the ability to assess dose was not compromised.The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources, where leaders ensure that personnel, 6 equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, radiation protection leadership failed to ensure that the RPT was capable of meeting the expectations for performing the LHRA briefing in accordance with station procedure RPAA460, Attachment 5. (H.1)
05000456/FIN-2017003-012017Q3BraidwoodFailure 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.
05000461/FIN-2017003-052017Q3ClintonFailure to Establish Secondary Containment Prior to Entering MODE 2The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS LCO 3.0.4, for the failure to follow station procedure CCAA201, Plant Barrier Control Program, Revision 11. Specifically, the licensee entered MODE 2 from MODE 4 without meeting the requirements of LCO 3.0.4 for entering a mode when an applicable LCO is not met. The licensee had not met LCO 3.6.4.1 because the doors to the B reactor water cleanup room were both opened instead of being closed to make secondary containment operable as required in MODE 2. The licensee entered this issue into their CAP as AR 04017613. As corrective actions, the licensee planned to conduct training for site personnel.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 it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to follow the station procedure by not identifying that the open doors required a plant barrier impairment (PBI) permit that would have identified the doors as a constraint to entering MODE 2 resulted in the unit transitioning to MODE 2 with the secondary containment inoperable. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone and determined 5 to be of very low safety significance because the finding only represented a degradation of a radiological barrier function provided for auxiliary building. The inspectors determined that this finding affected the cross-cutting are of human performance in the aspect of training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent work force and instill nuclear safety values. Specifically, station personnel did not know the process for routing a PBI permit and did not know when a PBI permit was required. (H.9)
05000461/FIN-2017003-042017Q3ClintonFlow Control Valves Not Locked Out Results in Reactor Recirculation Pump RunbackThe inspectors documented a self-revealed finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the licensees failure to establish sufficient instructions in station procedure Clinton Power Station (CPS) 3103.01, Feedwater (FW), Revision 31e, for changing modes of operation for the nuclear steam supply system. Specifically, the station procedure did not provide instructions requiring the locking out the flow control valves (FCVs) to prevent a reactor recirculation FCV runback while changing the feedwater pump lineup resulting in an unexpected plant transient and 9.2 percent change in reactor power. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 04007861. As corrective actions, the licensee revised their CPS 3103.01 procedure to require that the FCVs be locked out prior to shifting reactor feed water pumps. The performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012,because the finding was associated with the procedure quality 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 as well as power operations. Specifically, the failure to have adequate procedures for shifting feedwater pumps during a plant shutdown on May 7, 2017, resulted in an unexpected recirculation pump run back and a 9.2 percent change in reactor power. Using IMC 0609, Attachment 4, Initial Characterization of Findings, andAppendix A, The Significance Determination Process for Findings At-Power, issuedJune 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the event did not cause a reactor scram. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the procedure provided for the option to lockout the reactor 3 recirculation flow control valves if deemed necessary during a shift of the reactor feedwater pumps and the operations crew did not make the prudent choice of locking out the valves before determining that it was safe to proceed. (H.14)
05000461/FIN-2017003-032017Q3ClintonFailure to Perform Engineering Evaluation to Determine the Cause of Failure of SnubbersThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to demonstrate compliance with the requirement as prescribed in procedure ERCL330, CPS Snubber Program, Revisions 1 and 2. Specifically, the licensee failed to perform engineering evaluations to determine the cause of failure of snubbers that did not satisfy their functional testing acceptance criteria. The licensee entered this issue into their CAP as ARs 04015242 and 04041302. As corrective actions, the licensee evaluated the components affected by the failed snubber and determined that no operability issues existed. The performance deficiency was determined to be more-than-minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of Protection against External Factors and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability for mitigating systems to respond to initiating events. Specifically, compliance with ERCL330 would ensure the failed snubber wasevaluated for the cause of failure, to ensure the licensee identified other snubbers that may have been vulnerable to the same type of deficiency. This would ensure that any potential undesired loading on the piping system could be avoided and the affected safety-related residual heat removal and reactor water cleanup piping systems could continue to perform their design function of maintaining the pressure boundary and structural integrity following a postulated design basis seismic event. The inspectors determined the finding could be evaluated using the Significance Determination Processin accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, for the Mitigating Systems cornerstone and then Exhibit 4, External Events Screening Question. The finding screened as having very low safety significance because in each instance, the inspectors answered No to Questions 1 and 2 ofExhibit 4. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. Specifically, the licensee failed to establish a systematic approach to evaluating snubbers that did not meet the acceptance criteria to ensure all required aspects were addressed. (H.13)
05000461/FIN-2017002-042017Q2ClintonFailure to Provide Sufficient Work Instructions for Performing Maintenance on the Control Room Ventilation System Charcoal FilteGreen . The inspectors documented a self -revealed finding of very low safety significance and an associated non- cited violation of 10 of CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure of the licensee to provide sufficient work instructions for performing maintenance on the control room ventilation charcoal filter bed. Specifically, the work order used to change out the charcoal filter bed ( Work Order 01494189 ) contained only the minimum required amount of charcoal to place in the bed. Sometime after filling the bed April 6, 2015, the charcoal settled, resulting in the B control room ventilation system being declared inoperable after failing a surveillance test. The licensee entered this issue into their CAP as AR 03995612. As corrective actions, the licensee is revising the WO instructions and Clinton Power Station Procedure 9866.03 to require that charcoal be filled completely to the bottom of the deluge piping to allow for settling. The performance deficiency was determined to be more than minor because it impacted the Barrier Integrity cornerstone attribute of procedure quality and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to provide sufficient guidance in the work order regarding the quantity of charcoal to be installed resulted in the B control room ventilation system failing a surveillance test and being declared inoperable. The finding was screened against the Barrier Integrity cornerstone and determined to be of very low safety significance because the finding only represents a degradation of a radiological barrier function provided for the control room. The inspectors determined that this finding affected the cross -cutting area of human performance in the aspect of design margins, where the organization operates and maintains equipment within design margins. Special attention is placed on maintaining fission product barriers, defense in depth, and safety -related equipment. Specifically, when performing maintenance on the charcoal bed, the licensee failed to recognize that filling the charcoal to the minimum bed level provided no margin if settling occurred. (H.6)
05000461/FIN-2017002-052017Q2ClintonFailure to Properly Classify a Shipment per DOT RegulationsGreen . A finding of very low safety significance and an associated non -cited violation of Title 10 of CFR 71.5(a) and 49 CFR 173.421(b) was self -revealed when the licensee 6 failed to properly classify a shipment per Department of Transportation (DOT) regulations. The failure to properly classify the shipment per DOT regulations allowed the shipment to proceed in transit with dose rates that were greater than what was stated on the shipping manifest. When the discrepancy in dose rates was noticed by the receiving entity, the shipment was immediately isolated and the licensee was contacted about the survey results. The licensee then dispatched two radiation protection technicians to perform confirmatory surveys. The survey data was confirmed, and the licensee was able to determine that the misclassification of the shipment was caused by dust and debris contained inside of a dust collector shifting during transportation, which created the elevated dose rate. The site implemented immediate corrective actions which included all shipments classified as limited quantity to be approved by a senior manager in the Radiation Protection Department prior to shipping. Another immediate corrective action required that the first 4 shipments conducted by the site shipper after this event be under the direct observation of a fleet independent shipper and a senior manager in the Radiation Protection Department . The licensee entered this event into their CAP as AR 03961544. The inspectors determined that the performance deficiency was more than minor because the finding impacted the program and process attribute of the Public Radiation Safety cornerstone and adversely effected the cornerstone objective of ensuring adequate protection to public health and safety from exposure to radiation from routine civilian nuclear operations. Specifically, the misclassification of the shipment per DOT regulations could have led to individuals in the public domain being exposed to radiation dose that was greater than anticipated if conditions had been slightly altered. The finding was screened against the Public Radiation Safety cornerstone and determined to be of very low safety significance because: (1) the finding did not involve a certificate of compliance issue; (2) the failure to make emergency Notifications; (3) a lo w-level burial issue; or (4) a breach of the transportation package occurring during transit. The finding did involve a radioactive shipment above radiation limits. However, the shipment contained less than a Type A quantity of material (LSA I shipment), and dose rates were <2 millirem per hour on contact. The inspectors determined that this finding affected the cross- cutting area of human performance in the aspect of challenging the unknown, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, the risk associated with the content of the dust -collector shifting during transportation and creating an area that would lead to elevated dose rates was not evaluated by Clinton Power Station radiation protection staff. (H.11)
05000440/FIN-2017009-012017Q2PerryUnsuitable Application of Surge Suppression Diodes in Standby Diesel Generator Control Power CircuitryPreliminary White . The inspectors identified a finding preliminarily determined to be of low to moderate safety significance (White), and an associated apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) 50, Criterion III , Design Control, for the licensees failure to implement measures for the selection and review for suitability of application of voltage suppression diodes installed in the control circuitry for the Division 2 Standby Diesel Generator, which was a component subject to the requirements of 10 CFR Part 50, Appendix B. Specifically , Engineering Change Package 04 00049 failed to consider the effects of a shorted diode on the control circuitry for the Division 2 Standby Diesel Generator, and instead, introduced new components (diodes) into the control circuitry that resulted in the eventual failure of this safety -related equipment . This rendered the standby diesel generator inoperable and unable to start for longer than its technical specification allowed outage time , which was a violation of Technical Specification 3.8.1, AC Sources -Operating . The licensee documented the issue in CR 2016 13183, and subsequently replaced the failed component and then modified circuitry to remove the replacement diode and the remaining diodes from similar components. The inspectors determined that the licensees failure to evaluate the effects of voltage suppression diode failure on the Standby Diesel Generator control circuit was contrary to the requirements of 10 CFR Part 50, Appendix B , Criterion III and a performance deficiency which was within the licensees ability to foresee and prevent . The inspectors determined that the performance deficiency was of more than minor significance because it was associated with the design control attribute of the mitigating syst ems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the design of the Division 2 Standby Diesel Generator control circuit resulted in the inoperability and unavailability of the Division 2 Standby Diesel Generator from April 2, 2015, to November 8, 2016, when the failed diode was replaced. 3 A Significance and Enforcement Revi ew Panel, using IMC 0609, Appendix A, Significance Determination Process for Findings At -Power, dated June 19, 2012, preliminarily determined the finding to be of low -to-moderate safety significance. The inspectors did not identify any cross- cutting aspects associated with this finding because the condition had existed since at least 2007, when the diodes were originally installed in the DC control power circuits, and therefore, was not indicative of current plant performance
05000461/FIN-2017002-032017Q2ClintonUnexpected Start of the Division 3 Emergency Diesel GeneratoGreen . The inspectors documented a self -revealed finding o f very low safety significance and an associated non- cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to follow steps in Work Order (WO) 04640788 while performing troubleshooting on blown power transformer fuses in the division 3 emergency diesel start circuitry. Specifically, the electricians opened test switches in the wrong electrical cubicle resulting in the unexpected start of the division 3 emergency diesel generator and a loss of power to the 1C1 bus from an offsite source. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR ) 04012393. As corrective actions, the licensee performed a human performance review to identify the reasons the procedure was not followed and restored power to the 1C1 safety bus . The performance deficiency was determined to be more than minor because it impacted the Initiating Event s cornerstone at tribute of human performance 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 failure of the electrical maintenance technicians to follow their procedures resulted in a loss of power to the 1C1 electrical bus. T he finding was screened against the Initiating Event s cornerstone and determined to be of very low safety significance because the loss of power to the 1C1 bus occurred while Clinton was in a refueling outage when the high pressure core spray system was removed from service and not being relied upon for shutdown safety defense in depth. The loss of the 1C1 bus did not affect decay heat removal from the core, did not affect reactor coolant inventory, and the event occurred while the refuel cavity was flooded up for refueling operations. The inspectors determined that this finding affected the cross -cutting area of human performance in the aspect of avoid complacency where individuals implement 3 appropriate error reduction tools. Specifically, as documented in the licensees human performance review, the electricians performing the work did not utilize any human performance tools to flag the equipment to be operated and improperly performed the concurrent verification of the component to be manipulated. (H.12)
05000461/FIN-2017002-012017Q2ClintonFailure of Operators to Meet Time Critical Operator ActionsGreen . The inspectors identified a finding of very low safety significance and an associated non -cited violation of Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to assure that applicable regulatory requirements and the design basis was correctly translated into specifications, drawings, procedures, and instructions and that design control measures provided for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculation al methods, or by the performance of a suitable testing program . Specifically, the licensee failed to assure/validate operators were able to complete the standby liquid control time critical action for an anticipated transient without a scram specified in their licensing documents. The licensee entered this issue into their CAP as AR 03980202. As corrective actions, the licensee determined the scram choreography required to complete the time critical action in the specified time, initiated a standing order to inform the operating crews, processed a procedure change for the anticipated transient without scram choreography and performed an evaluation to determine the impact of initiating the standby liquid control system at 172 seconds. The performance deficiency was determined to be more than minor because the finding was associated with the procedure quality attribute of the Mitigating Systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, with the operators initiating standby liquid control at 172 seconds instead of 120 seconds, the accident analysis calculations were required to be re- performed to assure the accident analysis requirements were met. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors were able to answer all of the associated screening questions No. The inspectors determined that this finding is not indicative of current performance and therefore did not assign a cross -cutting aspect.
05000461/FIN-2017002-022017Q2ClintonFailure to Perform Adequate Evaluation of Crane Rail ClipsGreen . The inspectors identified a finding of very -low safety significance and an associated cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to properly verify the adequacy of design of the fuel building crane and crane support structure elements. Specifically, calculations involving the crane rail clips and clip bolts had multiple technical errors and failed to adequately demonstrate that the design met the design basis requirements. The licensee initiated corrective actions by documenting the deficiency in A R 4001089 and performed an evaluation demonstrating that the functionality of the crane was maintained. The finding was determined to be more -than -minor because it was associated with the design control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of maintaining the functionality of the spent fuel pool (SFP) cooling system. Specifically, crane rail clip bolts were required to ensure structural integrity of structures, systems, and components described in the Updated Safety Analysis Report, 5 when subjected to design loads as part of safe load handling of heavy loads near the SFP and to ensure integrity of the spent fuel cask. In accordance with IMC 0609, Significance Determination Process , Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Barrier Integrity cornerstone because it was associated with SFP/fuel handling activities . Based on answering No to questions A through F in Table 3, the inspectors determined the finding could be evaluated using Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 3, for the Barrier Integrity cornerstone screening questions. Based on the crane remaining functional, the inspectors answered No to Questions D.1 through D.4 because the finding did not adversely affect decay heat removal capabilities, did not result from fuel handling errors, did not result in loss of SFP inventory, and did not affect the SFP neutron absorber or fuel bundle misplacement ; therefore , the finding screened as having very -low safety significance. The finding was cross- cutting in the resolution aspect of the problem identification and resolution area because the licensee failed to take effective corrective actions in a timely manner to address issues identified earlier in the rail clip evaluations. (P.3)
05000461/FIN-2017002-062017Q2ClintonFailure to Perform Preventive Maintenance on a Safety - Related Breaker CubiclGreen . The inspectors identified a finding of very low safety significance for the licensees failure to perform maintenance on a safety -related motor control center cubicle. Specifically, the licensee failed to perform thermography on the division 1 shutdown service water pump room cooler breaker cubicle in accordance with the maintenance strategy/template without providing justification for differing from the template as required by MA AA 716 210, Performance Centered Maintenance Process , Revision 3. This resulted in the division 1 shutdown service water pump room cooler fan failing because of a high resistance connection that went undetected. The licensee entered this issue into their CAP as AR 02667822. As corrective actions, the licensee replaced the thermal overload relays and created a preventative maintenance action to perform thermography on this equipment on a periodic basis. This performance deficiency was determined to be more than minor because it impacted the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that responds to initiating events. Specifically , the room cooler fan failure directly impacted the operability of the division 1 shutdown service water pump and the 4 division 1 emergency diesel generator which are safety -related, risk significant systems. The finding was screened against the Mitigating Systems cornerstone and determined to be of very low safety significance because the inspectors were able to answer all of the associated screening questions No. The inspectors determined that this finding is not indicative of current plant performance and therefore did not assign a cross -cutting aspect.
05000461/FIN-2017002-072017Q2ClintonRoot Cause Evaluation Failed to Identify Corrective Action to Preclude RepetitionGreen . The inspectors identified a finding of very low safety significance and an associated non -cited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to implement a quality assurance program procedure. Specifically, the licensee failed to document a root cause and develop a corrective action to preclude repetition for the 1A bus transformer failure in accordance with quality assurance procedure PI AA 125 1001, Root Cause Analysis Manual. The licensee entered this issue into their CAP as AR 01594407. The corrective actions in response to this issue were to revise the root cause report with a root cause of insulation degradation of the phase windings over time and develop a corrective action to prevent recurrence by using Doble testing to ensure indication of transformer insulation degradation was discovered prior to failure. The performance deficiency was determined to be more than minor because if left uncorrected the performance deficiency had the potential to lead to a more significant safety concern. Specifically, the root cause and corrective actions to prevent recurrence were not identified until the licensee was prompted by the inspectors. As a result, additional transformer failures could have occurred. The finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of or cause an initiating event nor did it affect mitigation equipment. The inspectors determined this finding affected the cross -cutting area of human performance, in the aspect of resources, where leaders ensure that personnel, equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, the licensees station procedure did not provide guidance on when a corrective action to preclude repetition is required, regardless of whether a risk assessment was performed. (H.1)
05000341/FIN-2017001-012017Q1FermiInadequate Work Instructions for Maintenance on EDG 14Green . A finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations ( 10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self -revealed when plant operators discovered a thick white smoke plume coming from the emergency diesel generator (EDG) 14 engine exhaust manifold during surveillance testing. Consequently, operators shut down the engine and removed it from service. The licensee failed to have work instructions for maintenance on the safety -related EDG appropriate to ensure insulation blankets on the engines exhaust manifold were replaced with insulation blankets conforming to the approved engineering design. The licensee entered this violation into its corrective action program for evaluation and identification of appropriate corrective actions. The licensee replaced the insulation blankets with insulation blank ets conforming to the approved engineering design. The finding was of more than minor safety significance because it was related to the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective o f ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operators shutdown the engine after discovering a thick white smoke plume coming from the engines exhaust manifold , which resulted in unplanned inoperability and unavailability of this onsite emergency power source. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time nor did it represent a loss of function of a non -TS train designated as high safety significant in accordance with the licensees Maintenance Rule Program . The inspectors concluded this finding affected the cross -cutting area of human performance and the cross- cutting aspect of documentation. Plant activities are governed by comprehensive, high -quality, programs, processes and procedures. Design documentation, procedures, and work pack ages are complete, thorough, accurate, and current. In this case, the licensees process for implementing and maintaining engineering configuration control of the newly designed EDG exhaust manifold insulation blankets was inadequate because 3 it did not follow the licensees formal engineering configuration management process. (IMC 0310, H.7)
05000341/FIN-2017001-022017Q1FermiFailure to Maintain Adequate SLC Storage Tank Boron ConcentrationGreen . A finding of very low safety significance with an associated Non- Cited Violation of TS 3.1.7, Standby Liquid Control (SLC) System, was self -revealed when the licensee measured the boron concentration in the SLC storage tank and discovered the concentration was below the minimum requirement of 8.5 percent. Specifically, the licensee failed to adequately monitor and identify a decreasing trend in SLC storage tank sodium pentaborate concentration concurrent with known dilution of the SLC storage tank during pump and valve testing. The licensee entered this violation into its corrective action program for evaluation and identifi cation of appropriate corrective actions and restored the SLC sodium pentaborate concentration to within TS limits. The finding was of more than minor safety significance 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, a lower than allowable sodium pentaborate concentration affected the SLC systems ability to shut down the reactor during a design basis event. The finding was determined to be a licensee performance deficiency of very low safety significance during a detailed Significance Determination Process review since the delta core damage frequency ( CDF ) was determined to be less than 1.0E 6/year. The inspectors concluded this finding affected the cross -cutting area of human performance and the cross -cutting aspect of resources. Specifically, the licensee failed to ensure equipment and procedures were adequate to support nuclear safety . Th is issue would have been avoided if the system monitoring plan was trending tank level via a pressure indicator . Also, chemistry had no administrative limits in their procedure to add boron prior to the minimum TS limit was reached and the system engineer was not a reviewer on the routine surveillance procedure and was not trending the concentration as a backup. (IMC 0310, H.1 )
05000341/FIN-2017001-032017Q1FermiLicensee-Identified ViolationTS 3.3.5.1, ECCS Instrumentation, states the ECCS instrumentation for each function in Table 3.3.5.1 1 shall be operable. As specified in Table 3.3.5.1 1, Function 3b, HPCI System High Drywell Pressure (4 channels) and Function 3f, HPCI System Manual Initiation (1 channel) are required to be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. TS 3.3.5.1, Required Action A.1 states with one or more channel(s) inoperable, immediately enter the condition referenced in Table 3.3.5.1 1 for the channel. Table 3.3. 5.1 1, Function 3b, references Condition B for inoperable HPCI System High Drywell Pressure channels. Required Action B.2 states declare the HPCI system inoperable within 1 hour from discovery of loss of HPCI initiation capability and Required Action B.3 states place the affected channel(s) in trip within 24 hours. Table 3.3.5.1 1, Function 3f, references Condition C for an inoperable HPCI System Manual Initiation channel. Required Action C.2 states restore the channel to operable status within 24 hours . If the required actions and associated completion 42 times of Condition B or C are not met, Required Action G.1 states immediately declare the associated supported feature (i.e., HPCI system) inoperable. TS 3.5.1, ECCS Operating, states, in part, each ECCS injection subsystem shall be operable in Modes 1, 2, and 3, except HPCI is not required to be operable with reactor steam dome pressure less than or equal to 150 psig. With the HPCI system inoperable, Required Action E.1 states immediately verify by administrative means RCIC system is operable and Required Action E.2 states restore HPCI system to operable status in 14 days. If the required actions and associated completion times of Condition E are not met, Required Action I.1 states be in Mode 3 in 12 hours. LCO 3.0.4.b is not applicable to HPCI. TS 3.3.5.2, RCIC System Instrumentation, states the RCIC instrumentation for each function in Table 3.3.5.2 1 shall be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. As specified in Table 3.3.5.2 1, Function 4, RCIC System Manual Initiation (one channel per valve) is required to be operable. TS 3.3.5.2, Condition A states with one or more channels inoperable, immediately enter the condit ion referenced in Table 3.3.5. 21 for the channel. Table 3.3.5.2 1, Function 4, references Condition C for an inoperable RCIC System Manual Initiation channel. Required Action C.1 states restore the channel to operable status within 24 hours. If the required actions and associated c ompletion times of Condition C are not met, Required Action E.1 states immediately declare the RCIC system inoperable. TS 3.5.3, RCIC System, states the RCIC system shall be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. With the RCIC system inoperable, Required Action A.1 states immediately verify by administrative means HPCI system is operable and Required Action A.2 states restore RCIC system to operable status in 14 days. If the required actions and associated completion times of Condition A are not met, Required Action B.1 states be in Mode 3 in 12 hours. LCO 3.0.4.b is not applicable to RCIC. TS 3.0.4, Limiting Condition for Operation (LCO) Applicability, Paragraph (a) states, in part, when a LCO is not met , entry into an operational mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation in the operational mode or other specified condition in the applicability for an unli mited period of time. This specification shall not prevent changes in modes or other specified conditions in the applicability that are part of a shutdown of the unit. Contrary to the above: 1. On six occasions (February 10, 2014, April 16, 2014, March 19, 2015, September 13, 2015, May 3, 2016, and November 7, 2016 ), the licensee entered Mode 3 following plant shutdowns without declaring the HPCI system instrumentation functions of high drywell pressur e and manual initiation inoperable and entering LCO 3.3.5.1. During the shutdowns, Fermi 2 was in Mode 3 for up to fifteen hours with reactor steam dome pressure greater than 43 150 psig without the licensee satisfying TS 3.3.5.1, Required Actions A.1, B.2, and G.1. This is a violation of TS 3.3.5.1. With HPCI inoperable as specified by TS 3.3.5.1, Required Actions B.2 and G.1, the licensee did not satisfy TS 3.5.1, Required Action E.1. This is a violation of TS 3.5.1. 2. On six occasions (February 10, 2014, A pril 16, 2014, March 19, 2015, September 13, 2015, May 3, 2016, and November 7, 2016 ), the licensee entered Mode 3 following plant shutdowns without declaring the RCIC system instrumentation function of manual initiation inoperable and entering LCO 3.3.5.2 . During the shutdowns, Fermi 2 was in Mode 3 for up to fifteen hours with reactor steam dome pressure greater than 150 psig without the licensee satisfying TS 3.3.5.2, Required Action A.1. This is a violation of TS 3.3.5.2. 3. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without declaring the HPCI system instrumentation functions of high drywell pressure and manual initiation inoperable and entering LCO 3.3.5.1. For up to nineteen hours during this time, the licensee did not satisfy TS 3.3.5.1, Required Actions A.1, B.2, and G.1. This is a violation of TS 3.3.5.1. With HPCI in operable as specified by TS 3.3.5.1, Required Actions B.2 and G.1, the licensee did not satisfy TS 3.5.1, Required Action E.1. This is a violation of TS 3.5.1. 4. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without declaring the RCIC system instrumentation function of manual initiation inoperable and entering LCO 3.3.5.2. For up to nineteen hours during this time, the licensee did not satisfy TS 3.3.5.2, Required Action A.1. This is a violation of TS 3.3.5.2. 5. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without meeting the LCOs of TS 3.3.5.1 and TS 3.3.5.2 for HPCI and RCIC systems instrumentation functions of high drywell pressure (HPCI only) and m anual initiation ( both HPCI and RCIC) . This is a violation of TS 3.0.4. This violation was entered into the licensees corrective action program as CARD 16 26153. The violation was determined to be of very low safety significance (Green) during a detailed Significance Determination Process review since the CDF was determined to be less than 1.0E -7/year.
05000440/FIN-2017001-012017Q1PerryFailure to Implement Procedures for Combating a Loss of Shutdown CoolingGreen. A finding of very-low safety significance and associated NCV of TS 5.4, Procedures, was identified by the inspectors for the failure to implement procedures for combating a loss of shutdown cooling (SDC). Specifically, the licensee failed to implement its procedure for combating a loss of SDC resulting from emergency service water (ESW) inoperability and during high decay heat load. This finding was entered into the licensees Corrective Action Program to perform analyses for various conditions to identify available alternate methods of decay heat removal and provide associated procedural guidance. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as very-low safety significance (Green) because it was a design deficiency that did not impact the operability or Probabilistic Risk Assessment functionality of any mitigating structures, systems, and components. The inspectors did not identify a cross-cutting aspect associated with this finding because it did not reflect current performance due to the age of the performance deficiency
05000440/FIN-2017001-022017Q1PerryLicensee-Identified ViolationIn part, 10 CFR 20.1703 (c)(5) states, The licensee shall implement and maintain a respiratory protection program that includes Determination by a physician that the individual user is medically fit to use respiratory protection equipment. Contrary to the above, the licensee identified that an individual wore a powered air purifying respirator (PAPR) three times during the period of March 67, 2017 for the purpose of radiological protection without the required medical determination. This was entered into the licensees corrective action program, CR 201702957, Vessel Technician Wore PAPR Three Times without Being Qualified. The significance of this violation was determined in accordance with IMC 0609 Appendix C, Occupational Radiation Safety Significance Determination Process dated August 19, 2008. This violation was determined to be of very low safety significance (Green), because this violation was not associated with ALARA Planning or Work Controls, there was no overexposure nor substantial potential for overexposure and the ability to access dose was not compromised.
05000254/FIN-2016004-012016Q4Quad CitiesFailure to Implement Foreign Materials Exclusion ControlsA finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to implement foreign material exclusion (FME) controls during the implementation of modification Work Order 1649339, Modify the Target Rock to Increase the Volume per Engineering Change 394119, and was contrary to MAAA716008, Foreign Material Exclusion Program, Revision 9. The failure to implement FME controls during maintenance led to the failure of the Unit 2 Target Rock safety relief valve solenoid valve during surveillance testing on April 5, 2016. The licensees corrective actions included replacing the Target Rock safety relief valve solenoid valve. In addition, the licensee made procedure revisions to the standard template for welding activities to ensure that a FME plan is developed when performing butt welds or weld repairs. The licensee entered this issue into their corrective action program as Issue Report 2703233. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance. The inspectors determined the finding represented a potential loss of the valve function and, therefore, a detailed risk evaluation was required. A regional senior risk analyst performed a detailed risk evaluation and determined the finding was of very low safety significance. This finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety was an overriding priority. Specifically, during the implementation of Work Order 1649339 and subsequent revisions, the licensee failed to control and execute the work while following FME processes and procedures (H.5).
05000440/FIN-2016004-022016Q4PerryModifications to Underdrain and Gravity Discharge System Manhole Covers Without a 10 CFR 50.59 Safety EvaluationGreen-Severity Level IV. The inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Test, and Experiments, and an associated finding, for the licensees failure to perform a written evaluation which provided the bases for the determination that a change did not require a license amendment. Specifically, the licensee made a change pursuant to 10 CFR 50.59(c) with the installation of grated manhole covers, replacing the rubber gasket, watertight manhole covers for the underdrain and gravity discharge systems and did not provide a basis for the determination that this change would not result in a more than a minimal increase in the likelihood of occurrence of a malfunction of a system structure or component important to safety. The licensee entered this issue into the CAP as CR 201611864 and performed a prompt operability determination to show that the underdrain and gravity drain systems remained functional while the engineering change package was developed to support the change and bring the underdrain and gravity discharge systems into compliance with the design basis. The performance deficiency was determined to be more than minor in accordance with Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and adversely affected the cornerstone attribute of protection against external factors and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Per IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, the finding was screened against the Mitigating Systems Screening Questions and determined to be of very low safety significance (Green) because the finding did not cause the underdrain and gravity discharge systems to become inoperable or non-functional. Traditional enforcement applied to this finding because it involved a violation that impacted the regulatory process. The inspectors determined it to be of Severity Level IV because it resulted in a condition evaluated by the SDP as having very low safety significance (Enforcement Policy example 6.1.d.2). The inspectors determined that the finding had a cross-cutting aspect in the area of human performance, procedure adherence, in that individuals did not follow processes, procedures, and work instructions. Specifically, a design engineer authorized the permanent modification to be made without the required 50.59 evaluation being completed (H.8).
05000440/FIN-2016004-042016Q4PerryFailure to Notify the NRC within Eight Hours of a Non-Emergency Event that Could Have Prevented the Fulfillment of a Safety FunctionSeverity Level IV. The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.72(b)(3)(v)(A) and (D), for the licensees failure to report to the NRC within eight hours, an event or condition that could have prevented the fulfillment of a safety function. The licensees evaluation of this condition, where both trains of the standby liquid control (SLC) system had been inoperable simultaneously, determined that it was not a reportable event. However, the inspectors determined that as described in NUREG 1022, Event Reporting Guidelines 50.72 and 50.73, Revision 3, Section 3.2.7, the licensee had failed to make a non-emergency eight hour report as required by 10 CFR 50.72(b)(3)(v)(A) and (D). The licensee submitted the eight-hour report on December 30, 2016, and entered this issue into the corrective action program (CAP) as CR 201700098. The failure to make an applicable non-emergency eight-hour event notification report within the required time frame was determined to be a performance deficiency. The inspectors determined that traditional enforcement was applicable to this issue because it impacted the NRC's regulatory process. In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d.9 of the NRC Enforcement Policy, this violation was screened as a Severity Level IV violation that was more than minor. In accordance with IMC 0612, because this violation involved traditional enforcement and does not have an underlying technical violation that would be considered more-than-minor, a cross-cutting aspect was not assigned to this violation.
05000440/FIN-2016004-012016Q4PerryECC B Heat Exchanger Flow Root Valves Out of PositionGreen. A finding of very-low safety significance and associated NCV of TS 5.4.1, Procedures, was self-revealed for the licensees failure to follow valve lineup procedure restoration requirements after an emergency service water (ESW) pump B and valve operability test. Specifically, incorrect valve manipulations of the root valves for 1P42R043B and 1P42R043A flow indicators caused the emergency closed cooling (ECC) heat exchanger B flow to read zero with flow through the heat exchanger. The incorrect flow indication rendered the remote shutdown panel inoperable. The licensee subsequently re-positioned the root valves, 1P42R043B and 1P42R043A, and restored the remote shutdown panel to operable. The licensee entered this issue into the CAP as CR 201612935. The inspectors determined that the performance deficiency for failure to follow procedure was more than minor and thus a finding because it was associated with the Mitigating Systems cornerstone attribute of human performance. The performance deficiency adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding has a cross-cutting aspect in the area of human performance, avoid complacency because the licensee failed to ensure that individuals follow processes, procedures, and work instructions. Specifically the individual performing the surveillance did not utilize all the required human performance tools to prevent the error (H.12).
05000440/FIN-2016004-032016Q4PerryRCS Pressure Boundary Leakage Operation Prohibited by TSsGreen. A finding of very low safety significance and an associated non-cited violation (NCV) of Technical Specification (TS) 3.4.5, RCS Operational Leakage, was self-revealed when the licensee operated with reactor coolant system (RCS) pressure boundary leakage, as a result of the failure of the weld connecting the root appendage of the vent line on the recirculation loop A discharge valve, between January 19, 2016, and January 24, 2016, which is a condition prohibited by TS. The licensee entered this issue into the Corrective Action Program (CAP) as Condition Report (CR) 201601071 and performed a significant condition adverse to quality root cause evaluation due to a principal safety barrier being seriously degraded, replaced the vent line appendage on the recirculation loop A discharge valve with a more robust pipe and cap, and developed plans to replace ten additional vent and drain line appendages on the reactor recirculation loops prior to the end of the 1R17 refueling outage in 2019. The inspectors determined that the licensees operation with RCS pressure boundary leakage, a condition prohibited by TSs, was a performance deficiency requiring evaluation. The inspectors determined that the finding was more than minor because it adversely impacted the Initiating Events cornerstone attribute of equipment performance-barrier integrity, and 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. The inspectors determined this finding was of very low safety significance because the leak would not have exceeded the RCS leak rate for a small loss-of-coolant accident (LOCA) and would not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function. The inspectors concluded that this finding had no additional cross-cutting aspects than what was discussed in Inspection Report 0500440/2016001.
05000265/FIN-2016003-012016Q3Quad CitiesLicensee-Identified ViolationTechnical Specification 5.7.2, states, in part, that each high-radiation area, accessible to personnel with radiation levels > 1000 mrem/hr at 30 cm (12 in.) from the radiation source or from any surface which the radiation penetrates shall have doors that are locked to prevent unauthorized entry. Contrary to the above, on April 26, 2016, the licensee identified the locking mechanism for a door was non-functional and could not prevent unauthorized entry to the area. Specifically, a worker intentionally challenged the locking mechanism to the Unit 2 low pressure heater bay door when the latch opened. The individual left the area and later reported the issue to the radiation protection staff that promptly secured the area with an alternate locking mechanism and determined that the dose rates exceeded 1000 mrem/hr inside the area. The licensee documented this issue in Issue Report 2661096 and reported a PI occurrence under the Occupational Radiation Safety Cornerstone. The inspectors determined that this issue was of very low safety-significance (Green) after reviewing IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The inspectors determined that it was not an as-low-as-reasonably-achievable planning issue, there was neither overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. Therefore, the finding screened as Green (very-low safety significance).
05000341/FIN-2016002-012016Q2FermiFailure to Control Combustible MaterialsThe inspectors identified a finding of very low safety significance with an associated NCV of Technical Specification (TS) 5.4, Procedures. During fire protection walkdowns in safety-related and risk-significant areas of the plant, the inspectors identified multiple instances of the licensees failure to implement procedural requirements for implementing its fire protection program as required by TS 5.4.1.d, specifically for the controls of combustible materials. The licensee entered this violation into its corrective action program for evaluation and identification of appropriate corrective actions. As immediate corrective actions, the licensee rectified all of the inspector-identified issues, performed walkdowns inspecting all fire storage cabinets in the plant, and directed individual departments to examine all other storage cabinets for combustible materials. Any additional discrepancies found during these walkdowns were promptly corrected. The finding was of more than minor safety significance because it was related to the Initiating Events Cornerstone attribute of Protection Against External Factors (Fire) and adversely affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during plant operations. Specifically, the failure to properly control combustible materials in safety-related and risk-significant plant areas could increase the likelihood of a fire in these areas causing a plant transient and/or affecting plant equipment. The finding was determined to be a licensee performance deficiency of very low safety significance since redundant safe shutdown systems would have remained available during a postulated fire scenario in the affected locations. The inspectors concluded this finding affected the cross-cutting area of human performance and the cross-cutting aspect of avoid complacency. The licensees failure to follow its fire protection program implementing procedure requirements involved several work groups and had existed for a sufficient period of time, such that individuals were accustomed to and accepted the discrepancies between what was required by the licensees fire protection program and the actual condition of materials in the plant. (IMC 0310, H.12)
05000341/FIN-2016002-032016Q2FermiLoss of Power Instrumentation TS 3.3.8.1 Applicability Following Bus 64C Potential Transformer Fuse FailuresLoss of Power Instrumentation TS 3.3.8.1 Applicability Following Bus 64C Potential Transformer Fuse Failures Introduction. The inspectors opened an Unresolved Item to further evaluate the applicability of TS 3.3.8.1, Loss of Power (LOP) Instrumentation, following the failure of potential transformer fuses that caused half of the bus 64C LOP and degraded voltage relays to de-energize. Description. On April 24, 2016, a loss of output from the line to neutral potential transformers occurred on 4160 volt AC busses 64A and 64C. This resulted in the loss of bus indications, loss of automatic control of station transformer 64 load tap changer, and actuation of half of the LOP and degraded voltage relaying for safety-related bus 64C. The licensee subsequently discovered all six of the primary side fuses to the potential transformers had blown. Although the actual cause for the blown fuses was not conclusively determined, the most likely cause was attributed to an intermittent low energy transient on the secondary side of station transformer 64 or a transient on the 120 kilovolt electrical grid supplying the transformer. The inspectors noted the licensee did not consider the de-energized LOP and degraded voltage instrument channels to be inoperable, and therefore, did not enter the applicable action requirements of TS 3.3.8.1. The licensee had concluded the affected LOP and degraded voltage instrument channels remained operable since their safety function was believed to have been satisfied while they were de-energized and tripped. The inspectors raised several questions with the licensee concerning the operability of the affected LOP and degraded voltage instrument channels. The questions included whether the potential transformers were part of the LOP and degraded voltage instrumentation described in TS 3.3.8.1 and whether applicable surveillance requirements had been satisfied for the instrumentation prior to and during the event. The licensee entered this issue into its corrective action program as CARD 1625194 for further evaluation. This issue of concern is considered an Unresolved Item pending additional review by the inspectors to determine whether the licensee had correctly applied the TS limitations and satisfied applicable regulatory reporting requirements (URI 05000341/201600203, Loss of Power Instrumentation TS 3.3.8.1 Applicability Following Bus 64C Potential Transformer Fuse Failures).
05000341/FIN-2016002-022016Q2FermiInadequate Test Procedure Used for Measuring and Determining Average Silt Levels in the Service Water ReservoirThe inspectors identified a finding of very-low safety significance with an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have a procedure that prescribed instructions to determine the average silt level in the residual heat removal (RHR) reservoir to ensure the stand-alone document ensures silt levels in the reservoir are maintained below the proceduralized limit of 3 inches. Specifically, in 2014 and 2015, the licensee failed to include the documented reservoir surveys or a method to determine the average silt levels in the RHR reservoir. After discussing the issue with the responsible site staff for the 2014 inspection, the licensee was able to locate the reservoir survey map outside of the quality records system; the records for 2015 were not provided. The licensee entered this issue into its corrective action program, verified that additional margin existed, and confirmed the reservoirs were still able to maintain their required design volume with the silt accumulation. The performance deficiency was determined to be more-than-minor because if left uncorrected it would have the potential to lead to a more significant safety concern. Specifically, since licensee procedures failed to prescribe instructions for silt depth determination, and failed to prescribe how responsible site staff determines an average reservoir silt level based on diver inspection reports, both quality related activities, the potential exists for an unacceptable condition to go unnoticed, affecting service water systems operability. The finding was of very-low safety significance because the finding did not represent a loss of system operability and/or function. The inspectors did not assign a cross-cutting aspect because the finding was not indicative of current performance.
05000341/FIN-2016002-042016Q2FermiFailure to Control the Work Hours of Covered WorkersThe inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 26.205(c) and (d). The licensee failed to schedule and control the work hours of two maintenance craftsmen performing work covered under 10 CFR 26.4(a) by not ensuring the individuals had, at a minimum, a 34-hour break in any 9-day period as required by 26.205(d)(2)(ii). The licensee entered this violation into its corrective action program for evaluation and identification of appropriate corrective actions. The finding was of more than minor safety significance because a failure to schedule and control the work hours of workers performing covered work, if left uncorrected, would become a more significant safety concern since it could reasonably result in human performance errors due to fatigue that could result in plant transients and/or affect the function of safety-related systems or components. The finding was determined to be a licensee performance deficiency of very low safety significance based on a qualitative evaluation of the potential consequences of the performance issue since there were no human performance related incidents attributed to the two maintenance craftsmen while they were not in compliance with the work hour limits. The inspectors concluded this finding affected the cross-cutting area of problem identification and resolution and the cross-cutting aspect of evaluation. The licensee did not thoroughly evaluate the problem after it was identified and reached an incorrect conclusion because it failed to sufficiently understand the regulatory requirements and the basis for its decisions that contributed to the non-compliance with the 26.205 work hour requirements (P.2).
05000341/FIN-2016002-052016Q2FermiFailure to Use Correct Material in a Feedwater Heater Level Control Valve Resulted in a Loss of Feedwater Heater Drains and a Reactor Recirculation System RunbackA finding of very low safety significance was self-revealed when a reactor recirculation system runback occurred during power ascension due to a loss of feedwater heater drains caused by a feedwater heater level control valve malfunction. The control valve malfunction occurred because the licensee had failed to use correct material in the component during maintenance in October 2010. No violation of regulatory requirements was identified because the feedwater heating system is not safety-related and the applicable maintenance procedures were not covered under 10 CFR Part 50, Appendix B. The finding was of more than minor safety significance because it was related to 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 safety functions during shutdown as well as power operations. Specifically, the control valve malfunction resulted in a reactor recirculation system runback. In addition, the finding was sufficiently similar to IMC 0612, "Power Reactor Inspection Reports," Appendix E, "Examples of Minor Issues," Examples 4(b) and 4(f), to conclude it was not of minor significance because there was an adverse safety impact (i.e., a plant transient) due to the licensees failure to meet its technical requirements. The finding was determined to be a licensee performance deficiency of very low safety significance because it did not cause a reactor scram. The inspectors concluded that because the error occurred greater than three years ago, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
05000341/FIN-2016002-062016Q2FermiFailure to Implement Adequate Preventive Maintenance on Spare Terminals in Safety-Related Motor Control CentersA finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the Division 1 low pressure coolant injection (LPCI) outboard injection motoroperated valve failed to open during surveillance testing. The licensee failed to have preventive maintenance work instructions and procedures for safety-related moto control center (MCC) inspections appropriate to the circumstances, such that appropriate steps were incorporated to ensure spare terminal screws were maintained tight. The licensee entered this violation into its corrective action program for evaluation and identification of appropriate corrective actions. Corrective actions for the event included revising preventive maintenance work instructions and procedures to include instructions to check accessible spare terminal screws for tightness, personnel training, and inspection of all engineered safety feature MCC positions with relays susceptible to loose or missing screws and for susceptible contactor orientations. The finding was of more than minor safety significance because it was related to the Equipment Reliability 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 of the Division 1 LPCI outboard isolation valve to stroke open affected the LPCI loop select logic function to respond to a design basis event. The finding was determined to be of very low safety significance based on a detailed significance determination process review since the delta core damage frequency was determined to be less than 1.0E6/year. The inspectors concluded that because the inadequate procedures were in use for greater than three years, this issue would not be reflective of current licensee performance, and no cross-cutting aspect was identified.
05000341/FIN-2016002-072016Q2FermiLicensee-Identified ViolationTitle 10 CFR 26.205, Paragraph (a) requires, in part, Any individual who performs duties identified in Paragraphs 26.4(a)(1) through (a)(5) shall be subject to the requirements of this section. Title 10 CFR 26.4, Paragraph (a)(4) identifies individuals who are Performing maintenance or onsite directing of the maintenance of SSCs that a risk-informed evaluation process has shown to be significant to public health and safety. Title 10 CFR 26.205, Paragraph (c) requires, in part, Licensees shall schedule the work hours of individuals who are subject to this section consistent with the objective of preventing impairment from fatigue due to the duration, frequency, or sequencing of successive shifts. Title 10 CFR 26.205, Paragraph (d)(2)(ii) requires, in part, Licensees shall ensure that individuals have, at a minimum, a 34-hour break in any 9-day period. Contrary to the above, from March 26, 2016, through April 3, 2016, two individuals who performed duties identified in 26.4(a)(4) were not scheduled work hours as required by 26.205(c). Specifically, the individuals were inappropriately excluded from the work hour limits specified in 26.205(d)(2)(ii). As a result, the individuals were not provided a 34-hour break in any 9-day period. The violation was determined to be of very low safety significance based on a qualitative evaluation of the potential consequences of the performance issue since there were no human performance related incidents attributed to the two maintenance craftsmen while they were not in compliance with the work hour limits. The licensee entered this violation into its corrective action program as CARD 1622779.
05000456/FIN-2016001-012016Q1BraidwoodFailure 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-2016001-022016Q1BraidwoodFailure 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-2016001-032016Q1BraidwoodFailure 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-2016001-052016Q1BraidwoodFailure 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-2016001-042016Q1BraidwoodQuestions 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)
05000457/FIN-2015004-022015Q4BraidwoodFailure 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.
05000457/FIN-2015004-012015Q4BraidwoodLoss 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.
05000440/FIN-2015010-012015Q4PerryUnqualified Radiation Protection ManagerThe inspectors identified a finding of very low safety significance, and an associated violation of Technical Specification (TS) 5.3.1 when an unqualified individual was designated and performed the duties of the Radiation Protection Manager since early 2015. Specifically, the individual did not have the required experience and background necessary to provide sound judgement for safe and successful operation of the plant. This designation occurred after an April 29, 2015 report documented an internal review by the licensees Fleet Oversight group that concluded that the candidate did not meet qualifications of TS 5.3.1. The NRC determined that this violation did not meet the criteria to be treated as a Non-Cited Violation because this issue was not documented in the licensees Corrective Action Program. In addition, the licensees staff communicated to the inspector that no violation of TS had taken place The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612 because it was associated with the human performance attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that the lack of experience and background necessary to provide sound judgement for the Radiation Protection Program affects the licensees ability to control and limit radiation exposures. The finding was determined to be of very low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an as-low-as-reasonably-achievable planning issue, there was neither an overexposure nor a substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting aspect in the area of Human Performance, change management, because the licensee did not use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000263/FIN-2015003-042015Q3MonticelloDrywell to Torus Vacuum Breaker Past OperabilityDuring the cycle preceding the 2015 refueling outage, two evaluations associated with torus to drywell vacuum breaker operation were developed due to issues identified in the first quarter 2014. These included: CAP 1417977, Failure of drywell-torus vacuum breaker to close, which identified an occasion of dual indication during Procedure 0143 procedure. A second occurrence was observed several days later and was documented in CAP 1418471, AO-2382A Torus-to-DW vacuum breaker closed indication anomaly. CAP 1420318, DW-Torus vacuum breaker work performed with inadequate PMT, identified the PMT following shaft sealing component (O-ring) replacement during the 2013 outage was not performed as planned. The licensee evaluations for these CAP conditions concluded the Drywell to Torus vacuum breakers were operable. However, neither evaluation specifically considered the effect of an interference between the vacuum breaker test lever and vacuum breaker test actuator stem. Since this specific mechanism was not addressed in these two evaluations, past operability of the torus to drywell vacuum breakers was questioned. As a result, the licensee established a past operability evaluation be conducted via CAPs 1479198 and 1478212. The licensee completed its past operability evaluation on June 26, 2015. After review, the inspectors conveyed a number of questions to the licensees engineering staff in regard to the past operability evaluation. Although the licensee provided responses for the majority of these questions during the remainder inspection quarter, the licensee had requested external input in regard to one of the inspectors questions. Specifically, inspectors questioned whether it was possible for the bottom of the lever arm to be at an elevation above the top of the actuator stem at valve disc full open and if so, could the valve test lever arm have come to rest on top of the actuator stem, potentially impacting the ability of the vacuum breaker valve to close. Upon the close of this inspection period, that input had not yet been finalized and made available to the inspectors. As a result, this issue was considered to be an unresolved item pending a review of the licensees response and past operability for CAPs 1479198 and 1478212, including and the licensee response to open inspector questions.
05000263/FIN-2015003-012015Q3MonticelloInadequate Evaluation of Refueling Floor Structural Steel BeamsThe inspectors identified a finding of very low safety significance, and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, on September 3, 2008, licensee personnel failed to verify the adequacy of design when they failed to use correct section properties in their calculation of stresses on structural steel beams supporting the refueling floor for the increased spent fuel cask loading. Reevaluation of the beams using correct methodology resulted in the conclusion that the beams would not meet the design basis stress limits. Immediate corrective actions for this issue included initiation of a CAP, performance of a functionality assessment which concluded that the refueling floor remained functional but non-conforming, and creating compensatory measures which limited the refueling floor live load in the cask loading area (CAP 1492837). The inspectors determined that the licensees calculational methodology was contrary to the standard engineering principles applicable for determination of stresses in structural members, which resulted in a failure to meet Criterion III, Design Control, and was a performance deficiency. The finding was determined to be more than minor in accordance with IMC 0612 because it was associated with the Design Control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical barriers (reactor building) protect the public from radionuclide releases caused by accidents or events. Additionally, More than Minor Example 3.j of IMC 0612, Appendix E, Examples of Minor Issues, was used to inform the more than minor screening. The inspectors used IMC 0609, SDP, Attachment 4, Initial Characterization of Findings, and Appendix A of IMC 0609 to screen this finding. The inspectors answered No to questions C.1 and C.2 in Exhibit 3, Barrier Integrity Screening Questions. As a result, the inspectors concluded that the finding was of very low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance.
05000263/FIN-2015003-022015Q3MonticelloFailure to Perform High Radiation Area Portable Fire Extinguisher SurveillancesThe inspectors identified a finding of very low safety significance and an NCV of Technical Specification (TS) 5.4.1.d when the licensee failed to implement procedures associated with Fire Protection Program Implementation, to ensure that required refueling outage surveillances were performed for fire extinguishers located in high radiation areas (HRAs). Specifically, between March 2007 and May 2015, the licensee failed to implement steps 9 and 10 of 1123, Portable Fire Extinguishers, which required weighing and verifying adequate hydrostatic testing of the fire extinguishers in HRAs on a refueling outage frequency. Corrective actions included surveillance process changes and evaluation of the current status of the high radiation area fire extinguishers which resulted in the determination that outside of the surveillance process, a separate work activity had exchanged all the affected extinguishers with ones that were current on their surveillances in May 2015. This issue was entered into the licensees Corrective Action Program (CAP) 1484257 The inspectors determined that the failure to implement HRA fire extinguisher surveillances was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Work Management aspect because of the failure to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority and the failure to identify the need for coordination with different groups or job activities
05000263/FIN-2015003-032015Q3MonticelloFailure to Identify Safe Shutdown Equipment Impacts in Fire Strategy ProceduresThe inspectors identified a finding of very low safety significance and an NCV of TS 5.4.1.d when the licensee failed to maintain procedures associated with Fire Protection Program Implementation, consistent with the Updated Safety Analysis Report (USAR), to ensure that fire strategy procedures accurately indicated safe shutdown (SSD) equipment. Specifically, on June 25, 2015, the licensee failed to maintain A.3-12-C, Condenser Room Fire Strategy, to ensure SSD equipment was appropriately identified. In this case, fire strategy A.3-12-C failed to identify any SSD equipment in the room, despite the fact that SSD cabling ran through the room and was included in the USAR Fire Hazards Analysis. Corrective actions included performance of an extent of condition review which identified 40 other fire strategies where safe shutdown cabling was not identified, and initiation of procedure changes to include the appropriate SSD equipment. This issue was entered into the licensees CAP (CAP 1484142). The inspectors determined that the failure to maintain fire strategy procedures to ensure that SSD equipment was identified was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor in accordance with IMC 0612, Appendix B, because it was associated with the Mitigating Systems Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding using IMC 0609, Attachment 4, Initial Characterization of Findings," and IMC 0609, Appendix F, Fire Protection SDP, and determined that it had very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Self-Assessment aspect because of the licensees failure to conduct self-critical and objective assessments of its programs and practices.
05000263/FIN-2015003-052015Q3MonticelloFailure to Provide Complete and Accurate Information in LER 05000263/2015-002-00The inspectors identified a Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9 due to the licensees failure to provide information to the NRC that was complete and accurate in all material respects in accordance with the NRCs reporting requirements in 10 CFR 50.73(a)(1), Licensee Event Report (LER) System. Specifically, on June 29, 2015, the licensee failed to include an accurate assessment of the safety consequences and implications of a loss of shutdown cooling event when they issued LER 05000263/2015-002-00. This LER included an inaccurate assessment of safety implications, stating that engineering calculations show a potential worst case maximum temperature of 115 degrees Fahrenheit (F). The inspectors identified that engineering models actually showed potential worst case temperatures of 25-26 degrees F higher, which could have challenged or exceeded fuel pool cooling design specifications. Corrective actions included issuance of a revision to LER 2015-002-00 which contained the correct engineering modeling results and associated discussion of safety implications. The licensee entered this issue into its CAP (CAP 1484633). This issue was of more than minor significance under the Traditional Enforcement Process because the NRC relies on licensees to identify and correctly report conditions or events meeting the criteria specified in the regulations in order to perform its regulatory function. Because this issue affected the NRC's ability to perform its regulatory function, the inspectors evaluated it using the traditional enforcement process. The underlying technical issue (i.e., loss of shutdown cooling) was evaluated separately and determined to be a finding of very low safety significance as documented in the 2015 2nd Quarter Integrated Inspection Report (05000263/2015002-01). In accordance with Section 2.2.2.d, and consistent with the examples included in Section 6.9.d of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because it was of more than minor concern with relatively inappreciable potential safety significance and is related to a finding that was determined to be a more than minor issue. Consistent with Example 6.9.d.1, this represented an example where the licensee submitted inaccurate information in a required report, which resulted in expansion of the scope of the next regularly scheduled inspection and required LER revision. Because there was no finding evaluated with this violation, the inspectors did not assign a cross-cutting aspect to this issue.
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