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05000244/FIN-2018002-012018Q2GinnaIncorrect Scaling Factors in Reactor Vessel Level Monitoring System Instrumentation Uncertainty CalculationThe inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, when Exelon failed to ensure that adequate design control measures existed to verify the adequacy of the Reactor Vessel Level Monitoring System (RVLMS) uncertainty calculation. Specifically, Exelon failed to identify errors in the RVLMS uncertainty calculation which resulted in a reasonable doubt of operability for the system after a temporary modification was implemented.
05000244/FIN-2017004-012017Q4GinnaInadequate Component Monitoring Relating to Online Risk Management and AssessmentThe inspectors identified a finding because Exelon personnel did not follow Procedure WC-AA-101-1006, On-Line Risk Management and Assessment, Revision 2 to sufficiently monitor components such that the latest information was used to evaluate plant risk. Specifically, on December 27, 2017, Exelon failed to sufficiently monitor the diesel driven air compressor, commensurate with its operating history, such that a failure would be assessed and updated in the current plant risk assessment. Exelon entered this issue into the corrective action program (CAP) for resolution as action request (AR) 0487519. Corrective actions included declaring the diesel driven air compressor non-functional, transitioned to Yellow online plant risk, and completed restoration of the C Instrument Air Compressor.This finding is more than minor because it is associated with the configuration control attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, this issue is similar to Example 7.f of IMC 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009, because the overall elevated plant risk placed the plant into a higher licensee-established risk category. The inspectors evaluated this finding using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, to determine the significance of the finding. The inspectors used Flowchart 2, Assessment of (risk management actions) RMAs, to analyze the finding and calculated the incremental large early release probability using PARAGON, Exelons risk assessment tool, and found the increase in incremental large early release probability was less than 1E-7. The inspectors determined that if this condition existed for the full duration of the maintenance period, the large early release probability would have been 2.22E-7. Because the increase in incremental large early release probability, was less than 1E-7, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Exelon did not recognize and plan for the possibility of mistakes, latent issues and inherent risk, even while expecting successful outcomes. Specifically, Exelon did not ensure a component used to manage and assess risk was monitored at a frequency commensurate with its past performance. (H.12)
05000244/FIN-2017001-012017Q1GinnaLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non- cited violation (NC V). Ginna TS Table 3.3.1- 1 requires the function of under frequency Bus 11A and 11B be tested to be greater than or equal to 57.5 hertz in accordance with surveillance requirement 3.3.1.10. Surveillance requirement 3.3.1.10 requires this testing to be completed in accordance with the Surveillance Frequency Control Program. The Surveillance Frequency Control Program requires the function of under frequency Bus 11A and 11B be tested every 24 months. Contrary to the above, on February 6, 2017, Ginna engineering personnel determined that the Bus 11A under frequency function had not been tested within the interval specified frequency ; the function had last been tested on May 1, 2014 . Upon identification, Exelon conducted a risk evaluation and completed the surveillance requirement at the next available opportunity i n accordance with surveillance requirement 3.0.3 for a missed surveillance. Exelon entered this issue into the CAP as AR 03970849 and completed the testing on March 11, 2017. Additional evaluation was required to demonstrate operability since the acceptance criteria of greater than or equal to 57.5 Hz was not met. The inspectors determined the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, The SDP for Findings at Power, Exhibit 1, Initiating Events Screening Questions, issued June 19, 2012, because the transient initiator did not cause a reactor trip and the loss of mitigating equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition.
05000334/FIN-2016004-012016Q4Beaver ValleyFailure to Follow Procedure Results in an Inoperable A River Water TrainA self-revealing NCV of Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for FENOCs failure to assure that activities affecting quality were accomplished in accordance with procedures. Specifically, FENOC failed to follow NOP-OP-1001, Clearance/Tagging Program, and clearance 1W11-30-MNM-002 when removing the clearance for the A bay of the main intake structure. This resulted in disabling the automatic start capability of the standby C river water pump and made the A river water train inoperable and unavailable. FENOCs immediate corrective action was to rack the breaker for the A river water pump to the disconnect position, which cleared the annunciator and restored operability to the A train of river water. FENOC entered this issue into their corrective action program (CAP) as condition report (CR) 2016-14253. The performance deficiency is more-than-minor because it is associated with the Human Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, FENOC incorrectly racked the A river water pump breaker onto the 1AE 4160 volts alternating current (VAC) safety bus while the C river water pump was already racked onto the bus. This caused the A train of river water to be inoperable and unavailable because the automatic start capability of the C pump was disabled. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent a loss of system and/or function, an actual loss of function of a single train for greater than its technical specification allowed outage time, or an actual loss of function of one non-technical specification train designated as high safety significance. This finding has a cross-cutting aspect in Human Performance, Avoid Complacency, because the operators did not plan for the possibility of mistakes and did not implement appropriate error-reduction tools (H.12).
05000334/FIN-2016004-022016Q4Beaver ValleyLicensee-Identified ViolationThe following licensee-identified violation of NRC requirements was determined to be of very low safety significance and meets the NRC Enforcement Policy criteria for being dispositioned as a NCV. Radioactive material shipment B-4655, was made from Beaver Valley on May 5, 2016, to ResinSolutions in Erwin, TN. During a self-assessment performed by the FENOC staff on November 3, 2016, it was identified that the scaling factors used to determine the hard-to-detect nuclides listed on the manifest (NRC Form 540) for shipment B-4655 were incorrect. The scaling factors used to manifest the shipment were not for the waste stream shipped. Recalculation of the isotopic values using the correct waste stream scaling factors resulted in different numeric values for multiple radionuclides in the shipment, but did not cause a change in the proper shipping name, packaging, or labeling. 10 CFR 71.5 requires, in part, that radioactive materials be transported with an accurate shipment manifest. Contrary to the above, on May 5, 2016, FENOC transported radioactive materials with a shipment manifest that incorrectly stated that the radiological activity of the package was higher than the actual activity. FENOC documented this issue in CR 2016-13071, and provided a corrected shipment manifest to the recipient of the material. 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 FENOC had an issue involving transportation of radioactive material, but it did not involve a radiation limit that was exceeded, a breach of package during transport, a certificate of compliance issue, a low level burial ground nonconformance, or a failure to make notifications or provide emergency information.
05000286/FIN-2016009-012016Q4Indian Point345 kV Insulator Failure Causes Reactor TripA green self-revealing finding of ENN-EP-G-004, Switchyard and Large Power Transformer Preventive Maintenance Guidelines, occurred in that preventive maintenance (PM) was not performed as required on the W96 345kV line insulators. Specifically, the semi-annual corona surveys to identify degradation of insulators were not performed for line W96, which led to an insulator failure and resulted in an automatic trip of the reactor. Entergy replaced the damaged insulator and added the W96 line to the corona survey PM work order. Inspectors determined that Entergy did not perform PMs in accordance with ENN-EP-G-004, Switchyard and Large Power Transformer Preventive Maintenance Guidelines, on the 345kV insulators, which is a performance deficiency that was reasonably within Entergys ability to foresee and correct and should have been prevented. Specifically, the lack of PMs on the insulators allowed the insulators to degrade to a point where the condition of the insulator combined with environmental conditions led to a flashover event and a reactor trip. This finding is more than minor because it is associated with the Initiating Events cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure of the insulator led to a reactor trip. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the finding had a cross-cutting aspect in Human Performance, Avoid Complacency, because Entergy did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk. Specifically, Entergy performed a site review ensuring appropriate PMs were in place, and did not identify that the PM for the insulator was not being performed.
05000336/FIN-2016003-012016Q3MillstoneFailure to Review Standing OrdersThe inspectors identified a Green NCV of Technical Specification (TS) 6.8.1.a, for Dominions failure to implement procedures as required by Regulatory Guide 1.33, Revision 2, Appendix A.1, Administrative Procedures, during the performance of watch turnover. This resulted in multiple operators across multiple crews in both Unit 2 and 3 standing watch without performing a review of the applicable standing orders for up to 4 months from March to July 2016. Dominion entered the condition in their corrective action program (CAP) as condition report (CR)1042287. The inspectors determined that the finding was more than minor because if left uncorrected the performance deficiency could lead to a more significant event. Specifically, the operators did not review TS amendments, emergency action level classifications, emergency operating procedures, and plant computer issues impacting the plant prior to taking watch. Without reviewing the standing orders to understand the information contained within, operators could potentially take improper actions to control the plant during evolutions and abnormal conditions. The finding was determined to be of very low safety significance (Green) because it did not affect design or qualification of a mitigating structure, system, and component (SSC), did not represent a loss of system function, and did not involve external event mitigation systems. The inspectors determined that the finding has a cross-cutting aspect in the Human Performance cross-cutting area associated with Field Presence, where leaders are commonly seen in the work areas of the plant observing, coaching, and reinforcing standards and expectations. Specifically, Dominion leadership observations in the control room or management review of monthly standing order audits could have discovered the deviation from standards and expectations. (H.2)
05000423/FIN-2016003-022016Q3MillstoneFailure to Scope Safety Related Acoustic Valve Monitoring System into the Maintenance RuleThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(b)(1), for Dominions failure to include the safety-related Unit 2 Pressurizer Safety Valve, Acoustic Valve Monitoring System (AVMS) SSC within the scope of the maintenance rule program. Specifically, Dominion removed the Millstone Unit 2 AVMS, which is required to remain functional during and following a design bases event to provide indication to operators in the control room of significant abnormal degradation of the reactor coolant pressure boundary and monitor for loss of coolant due to an open safety relief valve, from the scope of the maintenance rule monitoring program. Dominion has documented this condition in their CAP as CR1049493. The inspectors determined that the finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, Dominions removal of AVMS from maintenance rule performance and condition monitoring and the failures observed have resulted in the complete loss of availability and reliability of each channel of AVMS such that they cannot perform their intended function. The finding was determined to be of very low safety significance (Green) because the conditions associated with the most applicable design basis event are bound by the small break loss of coolant accident (LOCA) analysis and did not affect other systems used to mitigate a LOCA. This finding has a crosscutting aspect in the Human Performance cross-cutting area associated with Procedure Adherence, in that Millstone Maintenance Rule Expert Panel (MREP) members did not follow the Dominion maintenance rule program implementing procedure, ER-AA-MRL-100, which provides guidance for scoping systems into the maintenance rule. (H.8)
05000333/FIN-2016002-012016Q2FitzPatrickFailure to Determine Dose Rates Prior to Entering a High Radiation AreaThe inspectors identified a self-revealing Green NCV of Technical Specification (TS) 5.7.1, High Radiation Area. Specifically, on January 24 and 25, 2016, operations personnel failed to notify the Radiation Protection (RP) department and non-licensed operators in the field when operating plant equipment that created high radiation areas (HRAs). These areas therefore were not surveyed by RP to determine dose rates prior to non-licensed operators entering the areas. Personnel entry into HRAs without knowledge of the current dose rates is a performance deficiency. In both instances, RP evaluated the operators dose, validated the dosimeter alarms, surveyed both areas in response to the dose rate alarms, and reposted the areas as HRAs. Entergy documented the events in condition reports (CR)-JAF-2016-00269 and CR-JAF-2016-00369 The finding was more than minor because it resulted in the unintended exposure of two workers and affected the Occupational Radiation Safety cornerstone attribute of program and process associated with exposure/contamination controls and if left uncorrected could result in more significant exposures. The finding was determined to be of very low safety significance (Green) because it was not related to as low as is reasonably achievable (ALARA), did not result in an overexposure or a substantial potential for overexposure, and did not compromise the licensee's ability to assess dose. A cross-cutting aspect of Human Performance, Teamwork, was associated with this finding. Specifically, licensed operators did not communicate to RP or non-licensed operators in the field when operating plant equipment that caused plant radiological conditions to change. (H.4)
05000333/FIN-2016002-022016Q2FitzPatrickFailure to Conduct Operations to Minimize the Introduction of Residual Radioactivity to the SiteThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 20.1406(c) due to Entergy not conducting operations to minimize the introduction of residual radioactivity into the site. For at least the past four years, Entergy allowed leakage of the solid radwaste processing system to occur, resulting in spilled radioactive waste that accumulated and remained on the floor of the filter sludge tank room in the radwaste building. The failure to control spilled radioactive wastes is a performance deficiency. Entergy entered this issue into their corrective action program (CAP) as CR-JAF-2016-01784 with actions to characterize the introduction of residual radioactivity and evaluate cleanup actions. This issue is more than minor because it is associated with the program and process attribute of the Public Radiation Safety cornerstone and affected the cornerstone objective to ensure the licensees ability to prevent inadvertent release and/or loss of control of licensed material. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, in that the condition was known to exist for over four years, impacted the radwaste system effectiveness to process solid radwaste, and had not been corrected. (P.3)
05000333/FIN-2016002-032016Q2FitzPatrickLicensee-Identified ViolationTS 3.6.3.2, Containment Atmosphere Dilution System, requires that, if one CAD subsystem is inoperable, then restore the subsystem to operable within 30 days or be in mode 3 within 12 hours. Contrary to the above, from June 17, 2015, to July 31, 2015, a period of 34 days, A CAD subsystem was inoperable without the plant being placed in mode 3 within 30 days and 12 hours of becoming inoperable. Also, contrary to the above, from August 5, 2015, to November 11, 2015, a period of 99 days, B CAD subsystem was inoperable without the plant being placed in mode 3 within 30 days and 12 hours of becoming inoperable. FitzPatrick staff entered this issue into their CAP as CR-JAF-2015-05453. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its TS allowed outage time (because operator action could be taken to restore system function if the subject temperature transmitter failed), and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event.
05000352/FIN-2016001-072016Q1LimerickLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a NCV. From 2010 to 2014, Exelon made a total of 16 shipments of radioactive material which contained category 2 quantities of radioactive material. Exelon did not implement a transportation security plan for any of these shipments, which is contrary to the requirements of 49 CFR 172, Subpart I, Safety and Security Plans. This performance deficiency adversely affected the Public Radiation Safety cornerstone attribute of Program and Process based on inadequate procedures associated with the transportation of radioactive materials. The finding was determined to be of very low safety significance (Green) because the transportation of radioactive material issue did not involve: (1) a radiation limit that was exceeded; (2) a breach of package during transport; (3) a certificate of compliance issue; (4) a low level burial ground nonconformance; or (5) a failure to make notifications or provide emergency information. This issue was documented in Exelons corrective action program as IR 2490592. Corrective actions included contracting with a vendor to receive regular, prompt notifications of potentially applicable rule changes in the Federal Register.
05000333/FIN-2016001-012016Q1FitzPatrickUnintended HPCI Pump Suction Transfer during Pressure Control Mode OperationThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to maintain a condition specified in an emergency operating procedure. Specifically, while operating the high pressure coolant injection (HPCI) system in the pressure control mode, operators failed to override automatic transfer of the HPCI pump suction from the condensate storage tank (CST) to the suppression pool prior to the transfer actually occurring. As a result, operators had to revert to using the safety/relief valves (S/RVs) for pressure control, which introduced additional, unnecessary plant challenges. As immediate corrective action, operators secured HPCI, overrode the automatic HPCI pump suction transfer, realigned the pump suction to the CST, and restarted HPCI in the pressure control mode. The issue was entered into the corrective action program (CAP) as condition report (CR)-JAF-2016-00765. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the operators failure to timely override automatic transfer of the HPCI suction to the suppression pool resulted in an additional, avoidable post-scram pressure and level transient being placed on the reactor pressure vessel (RPV) and unnecessarily reduced the thermal capacity of the suppression pool. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its technical specification (TS) allowed outage time, and did not screen as potentially risksignificant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because operators did not follow guidance of EOP-2 for the HPCI pump suction to be aligned to the CST by bypassing the HPCI pump suction swap to the suppression pool in a timely manner, such that the swap actually occurred (H.8).
05000333/FIN-2016001-022016Q1FitzPatrickUncontrolled RPV Level Increase after Initiation of RHR Shutdown CoolingThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to take actions specified in the procedure for initiation of shutdown cooling. Specifically, prior to placing the A loop of the residual heat removal (RHR) system into shutdown cooling, an operator was not stationed to close the condensate transfer system cross-connect valve to the A RHR loop (10RHR-274), nor was the valve immediately closed after initiation of shutdown cooling, as specified by the operating procedure. This resulted in a significant loss of operational control, in that RPV level increased to the point of putting water down the main steam lines. As immediate corrective action, operators closed 10RHR-274, thus stopping the RPV inventory increase. The issue was entered into the CAP as CR-JAF-2016-00273. The finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the resultant loss of RPV level control represented a significant loss of operational control that could have affected the operability of the HPCI and reactor core isolation cooling (RCIC) systems, as well as the S/RVs, had their use again been required in the near term. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of a safety function of a single train for greater than its TS allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because operators did not stop when faced with uncertain conditions. Specifically, without otherwise having maintained status control on the condensate transfer system cross-connect valve to the A RHR loop, operators did not stop to positively establish the condition of the valve when it appeared in a conditional step in the procedure (that is, if 10RHR-274 is open, then station an operator at 10RHR-274) (H.11).
05000333/FIN-2016001-032016Q1FitzPatrickInadequate Post-Maintenance Testing of the Reactor Building Ventilation System Resulted in Short-Term Inoperability of Secondary ContainmentThe inspectors identified a self-revealing NCV of TS 5.4, Procedures, for FitzPatrick staffs failure to perform adequate post-maintenance testing (PMT) following maintenance on a limit switch in the reactor building ventilation system in August 2014, that, along with another unrelated component failure in the reactor building ventilation system, resulted in secondary containment pressure, relative to the outside pressure, exceeding the TS limit of 0.25 inches of vacuum water gauge. As immediate corrective action, operators started both trains of the standby gas treatment system (SBGTS), which restored secondary containment pressure to within the TS limit. Operators subsequently secured the A refuel floor exhaust train and placed the B train in service. The issue was entered into the CAP as CR-JAF-2015-04166. The finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, as a result of this event, secondary containment was not preserved, in that secondary containment pressure exceeded the limit of TS surveillance requirement (SR) 3.6.4.1.1. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, did not involve an actual reduction in function of hydrogen igniters in the reactor containment, and only represented a degradation of the radiological barrier function provided by the reactor building and SBGTS. The finding had a cross-cutting aspect in the area of Human Performance, Resources, because FitzPatrick staff did not ensure that procedures for PMT of the reactor building refuel floor exhaust damper limit switch following maintenance performed in August 2014, were adequate to support the nuclear safety function of the secondary containment (H.1).
05000353/FIN-2016001-062016Q1LimerickFailure to Implement Procedures for Control of Potentially Contaminated Clean SystemsThe inspectors identified a Green NCV of technical specification 6.8.1 because Exelon failed to implement procedure CY-AA-170-210, Potentially Contaminated System Control Program, for the evaluation and control of potentially cross-contaminated systems. Specifically, Exelon did not implement CY-AA-170-210 for the evaluation and control of a potentially cross-contaminated system when samples collected from the Unit 2 service air system, a non-contaminated system, indicated the potential presence of contamination on June 16, 2015. Exelon entered this issue into the corrective action program (IR 2556568), restricted use of the service air system, conducted a 10 CFR 50.59 screening and radiological evaluation of the system, conducted bounding radiation dose analyses for both occupational workers and members of the public, conducted an extent of condition review, decontaminated the system, and subsequently modified operation of the service air system to preclude re-contamination. This finding is more-than-minor because it is associated with the program and process attributes of the occupational and public radiation safety cornerstones and adversely affected both cornerstone objectives to ensure adequate protection of worker and public health and safety from exposure to radioactive material. Specifically, during the time the service air system was contaminated but not recognized as such and not restricted in use, the potential existed to inadvertently contaminate workers and release radioactive material to the environment. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not involve an as low as is reasonably achievable (ALARA) issue, was not an overexposure, did not result in a substantial potential for an overexposure, and did not compromise the ability to assess dose. In addition, using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, the inspectors determined that the issue did not involve a substantial failure to implement the effluent release program and did not result in public doses exceeding 10 CFR 50, Appendix I or 10 CFR 20.1301 (e) and thus was of very low safety significance (Green). The inspectors determined this finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Exelon did not take effective corrective actions when service air system issues were identified.
05000352/FIN-2016001-012016Q1LimerickReactor Enclosure Recirculation System Design Change not EvaluatedA self-revealing Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50 (10 CFR 50), Appendix B, Criterion III, Design Control, was identified because Exelon did not properly maintain the design of the LGS Unit 1 reactor enclosure recirculation system (RERS). Specifically, Exelon replaced the Unit 1 1A RERS flow straightener assembly using thinner material than was originally qualified and did not evaluate the change in design. Exelon initiated IR 2563872 and implemented a temporary configuration change that removed the flow straightener assembly from the system and restored Unit 1 RERS to operability on October 5, 2015. Exelon also initiated corrective actions to install a new flow straightener assembly with correctly sized honeycomb material. This finding is more than minor because it adversely affected the design control attribute of the barrier integrity cornerstone to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate 1A RERS flow straightener assembly installed in 2012 resulted in degraded performance and then unplanned unavailability of 1A RERS from October 1 to 5, 2015. Using IMC 0609, Appendix A, Exhibit 3, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the degraded 1A RERS performance and associated unavailability only represented a degradation of the radiological barrier function provided for the standby gas treatment system and screened to Green. The inspectors determined that the finding did not have cross-cutting aspect because the performance deficiency did not occur within the last three years, and the inspectors did not conclude that the primary cause of the performance deficiency represented present Exelon performance.
05000353/FIN-2016001-052016Q1LimerickMain Turbine Digital Electrohydraulic Control System Modification Failed to Revise the Plant Startup ProcedureA self-revealing Green NCV of LGS Unit 2 technical specification 6.8.1 was identified because Exelon failed to maintain a plant startup procedure. Specifically, the implementing procedure for normal plant startup from hot shutdown or cold shutdown to rated power was not maintained when a modification to the Unit 2 turbine electrohydraulic control system was performed and required changes to the plant startup procedure were not identified and implemented. Exelon initiated issue report (IR) 2602637, revised the startup procedure to properly incorporate the software changes made at the factory acceptance test, validated the software changes that were made were technically correct, trained all operators on the new procedural changes, and reviewed operating procedures for extent of condition. This finding is more than minor because it is associated with the procedure quality attribute of the initiating events cornerstone and affected the objective to limit the likelihood of events that upset plant stability during power operations. Specifically, the procedure directed actions intended in the software for rapid reactor depressurization that resulted in a reactor trip. Using IMC 0609, Significance Determination Process, Appendix A, Exhibit 1, Initiating Events Screening Questions, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, although the finding caused a Level 8 trip of the feedwater pumps followed by a reactor trip, the rate of water injection from the condensate pumps was sufficient when the reactor was tripped to safely shutdown and operators were able to reset the feedwater pumps. The inspectors determined that this finding has a cross-cutting in the area of Human Performance, Change Management, because leaders did not use a systematic process for implementing the modification so that nuclear safety remained the overriding priority.
05000352/FIN-2016001-042016Q1LimerickEntry into a High Radiation Area without Radiological Briefing and Complying with the RWPA self-revealing Green NCV of LGS Unit 1 technical specification 6.12.1 was identified involving improper entry of two workers into the Unit 1 reactor drywell on March 22, 2016. Specifically, the workers entered the drywell, an area controlled as a Locked High Radiation Area, without obtaining the required access radiological conditions briefing. Further, one of the two workers entered under the control of an RWP that did not authorize access into High Radiation Areas. Exelon initiated IR 2644005, restricted the workers from further radiological controlled area access, re-configured the access area, conducted an extent of condition and human performance review, issued a site communication, and performed a staff stand down. This finding is more than minor because it is associated with the programs and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure adequate protection of workers from radiation exposure. In addition, this example is similar to example 6.h of IMC 0612, Appendix E. Specifically, the workers did not receive a brief and did not review surveys prior to entering a work area with radiation levels that exceeded 100 mrem/hr at 30 cm. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the inspectors determined the finding was of very low safety significance (Green) because: 1) it was not an as low as is reasonably achievable (ALARA) finding, 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 that this finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because the individuals failed to follow verbal work instructions. (H.8)
05000333/FIN-2016001-042016Q1FitzPatrickUntimely 10 CFR 50.72 Notification of Inoperable Secondary ContainmentThe inspectors identified a SL IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because unplanned inoperability of the secondary containment system was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition That Could Have Prevented Fulfillment of a Safety Function. Specifically, following reasonable resolution of questions regarding the reliability of secondary containment differential pressure (d/p) instrumentation indications, FitzPatrick staff did not promptly report that, during a transfer from normal reactor building ventilation in service to the reactor building being isolated with the SBGTS in service, reactor building d/p briefly dropped below the TS required minimum value of 0.25 inches of vacuum water gauge and therefore caused the secondary containment system to be inoperable. As immediate corrective action, the event was reported to the NRC in accordance with 10 CFR 50.72(b)(3)(v). The issue was entered into the CAP as CR-JAF-2015-05244 and CR-JAF-2015-05265. The inspectors determined that the failure to inform the NRC of the secondary containment system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a safety system functional failure was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was a SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects.
05000352/FIN-2016001-032016Q1LimerickInadequate Work Staging and Housekeeping Walkdowns During PreOutage PreparationsThe inspectors identified a Green NCV of technical specification 6.8.1 for Exelons failure to properly control, store, and stage material in accordance with station procedures within Class I buildings during refueling outage preparation. Specifically, Exelon personnel did not secure numerous rolling carts staged in both units, did not secure welding blankets in the common pipe tunnel to prevent blocking floor drains, and did not properly build scaffolds to include engineering approval for scaffold procedure deviations. In addition, Exelons housekeeping and material condition program did not identify and resolve these conditions through the corrective action process during a time of increased activities in the plant. Exelon restrained the carts and other rolling equipment, removed the weld blankets, and removed, reworked, and evaluated scaffolding. This finding is more than minor because it adversely affected the protection against external factors (flood and seismic hazards) attribute of the mitigating systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the loose unattended welding blankets would have blocked the pipe tunnel floor drains during an analyzed internal flooding event which would result in structural failures if not identified and corrected by operations personnel; the unrestrained carts would translate and rotate during a seismic event which could potentially impact safety related equipment and challenge the function or barrier; and the scaffold clearance and attachment issues could potentially cause impact with ductwork, cable trays, hangers, and structural supports during a seismic event. In addition, the performance deficiency is similar to the more-than-minor example described in IMC 0612, Appendix E, example 4.A, in that Exelon routinely failed to perform engineering evaluations on similar issues. Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the finding is a deficiency affecting the design or qualification of mitigating structures, systems, and components, and the actual functions of the structures, systems, and components were maintained. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Training, because the organization did not provide sufficient training to maintain a knowledgeable workforce with respect to proper material handling and storage, awareness of flood hazards and floor drains, and scaffolding requirements. (H.9)
05000352/FIN-2016001-022016Q1LimerickSeismic Qualification of Safety Related Battery not MaintainedThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, and technical specification 3.8.2, D.C. Sources, because Exelon failed to ensure the design control measures for field changes impacting the seismic support of station batteries were commensurate with those applied to the original design requirements. Specifically, during cell replacement of the Class 1E 1A1 125/250 volts direct current (Vdc) safeguards battery, removal of adjacent cells and restraint barriers left the battery in a state in which the seismic qualification was not maintained. Exelon initiated IR 2624349, stopped the battery cell replacement work, and performed a technical evaluation to determine the requirements to maintain the seismic qualification during the cell replacement process. This finding is more than minor because it adversely affected the protection against external factors (seismic) attribute of the mitigating systems cornerstone to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, during cell replacement of the Class 1E 1A1 125/250 Vdc safeguards battery, removal of adjacent cells and restraint barriers left the battery in a state in which the seismic qualification was not maintained. In accordance with IMC 0609, Appendix A, Exhibit 4, External Event Screening Questions, the inspectors determined that a detailed risk evaluation was required because the loss of this equipment by itself during the seismic event it was intended to mitigate would degrade one or more trains of a system that supports a risk significant function. The Region I Senior Reactor Analyst referenced the Limerick External Events Notebook to assess the potential increase in plant risk associated with this condition. As referenced in the Notebook, the initiating event frequency for the safe shutdown earthquake (SSE) is approximately 5E-4/year. Based upon the inspectors review of operations logs, the five battery replacement activities that occurred over the past 12 months ranged in duration from between one to six days. Assuming the seismic qualification was compromised the entire duration of these maintenance activities, the consequential increase in risk for any single event would be in the low to mid E-9 delta core damage frequency range. The dominant core damage sequences involve an SSE that results in a loss of offsite power and the subsequent failure to remove heat from containment (via the multi-train residual heat removal system and associated service water cooling trains). This estimated small increase in core damage frequency represents a condition of very low safety significance (Green). The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Exelon did not recognize and plan for the possibility of latent issues associated with the battery replacement process. (H.12)
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