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	Quarter	Site	Title	Description
05000219/FIN-2014004-01	2014Q3	Oyster Creek	Inadequate Evacuation Time Estimate Submittals	The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2), 10 CFR 50.47(b)(10), and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Oyster Creek emergency plan as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date. Exelon entered this issue into its corrective action program as issue reports 1525923 and 1578649. Additionally, Exelon resubmitted a new revision of the Oyster Creek ETE to the NRC on April 4, 2014, and the NRCs review of that ETE is documented in Section 1EP4 of this report. The performance deficiency is more than minor because it is associated with the Emergency Preparedness cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs have the potential to reduce the effectiveness of public protective actions implemented by the OROs. The finding is determined to be of very low safety significance (Green) because it is a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The cause of the finding is related to a cross-cutting aspect of Human Performance, Documentation, because Exelon did not appropriately create and maintain complete, accurate, and up-to-date documentation.
05000219/FIN-2014005-01	2014Q4	Oyster Creek	Reactor Head Cooling Spray Piping Flange Misalignment	The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly correct a condition adverse to quality associated with the reactor head cooling (RHC) spray line 2-inch upper flange which was installed in a configuration that exceeded the allowable acceptance criteria. Specifically, Exelon staff identified a misaligned flange condition in IR 845395 but did not correct the deficiency by evaluation, repair or replacement during the 1R22 refueling outage in 2008 or subsequently during the 1R23 and 1R24 refueling outages. Exelon staff completed corrective actions to replace the flange during the 1R25 refueling outage after the NRC inspector questioned the acceptability of this condition. Exelon staff entered this issue into their corrective action program as IR 2385501. The finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, misalignment of the RHC spray line flange was greater than that provided in Oyster Creek pipe specifications and resulted in additional stresses in the flange weld. This condition was identified by Exelon staff as a possible contributor to the occurrence of a through wall crack and leak in the N7B upper flange socket weld joint that was identified and repaired in November 2012, but the misalignment was not corrected at that time. The inspectors screened this issue using IMC 0609.04, Phase 1 Initial Screening and Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and determined this finding was of very low safety significance (Green). The inspectors determined that this finding had a Problem Identification and Resolution cross-cutting aspect because Exelon did not evaluate and take timely corrective actions to address the long-standing repetitive flange alignment issue of the reactor head cooling spray piping flange connection to reactor pressure vessel head N7B nozzle.
05000219/FIN-2014005-02	2014Q4	Oyster Creek	Inadequate Review of Change in Maintenance Process Results in Inoperable Emergency Diesel Generator	The inspectors identified a preliminary White finding and an associated apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control, because Exelon staff did not review the suitability of the application of a different maintenance process at Oyster Creek that was essential to a safety-related function of the emergency diesel generators (EDG). Specifically, in May 2005, Exelon staff changed the method for tensioning the cooling fan belt on the EDG from measuring belt deflection to belt frequency and did not verify the adequacy of the acceptance criteria stated for the new method. As a result, Exelon staff did not identify that the specified belt frequency imposed a stress above the fatigue endurance limit of the shaft material, making the EDG cooling fan shaft susceptible to fatigue and subsequent failure on July 28, 2014. As a consequence, Exelon also violated Technical Specification 3.7.C, because the EDG No. 2 was determined to be inoperable for greater than the technical specification allowed outage time. Exelons immediate corrective actions included entering the issue into their corrective action program as issue report (IR) 1686101, replacing the EDG No. 2 fan shaft, examining the EDG No.1 fan shaft for extent of condition, and performing a failure analysis to determine the causes of the broken shaft. This finding is more than minor because it is associated with the equipment 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 (i.e., core damage). 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, issued June 19, 2012, the inspectors screened the finding for safety significance and determined that a detailed risk evaluation was required because the finding represented an actual loss of function of a single train for greater than its technical specification allowed outage time. The detailed risk evaluation concluded that the increase in core damage frequency was 5.1E-6, or White (low to moderate safety significance). This finding does not have an associated cross-cutting aspect because the performance deficiency occurred in 2005 and is not reflective of present performance.
05000219/FIN-2014005-03	2014Q4	Oyster Creek	Plant Shutdown Procedure was Inadequate for Soft Shutdown	The inspectors identified a non-cited violation (NCV) of Technical Specification 6.8.1, Procedures and Programs, because Exelon did not adequately establish and maintain the plant shutdown procedure. Specifically, the procedure was not adequate in that it did not contain precautions concerning rod insertion when reactor power is below the point of adding heat; operational limitations on plant cooldown when power is below the point of adding heat; and contingency actions for re-criticality during shutdown. Exelon entered this issue into their corrective action program as IR 2412093 and conducted a root cause analysis. This finding is more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability and capability of systems that respond to initiating events. Specifically, the plant shutdown procedure did not contain precautions to continuously insert control rods when reactor power is less than the point of adding heat, did not define operational considerations for limiting reactor cooldown, and did not contain contingency actions for return to criticality during shutdown. The inspectors screened this issue using IMC 0609.04, Initial Characterization of Findings, Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria. Inspectors determined this finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Exelon did not ensure that the shutdown procedure contained adequate controls for soft shutdown.
05000219/FIN-2014005-04	2014Q4	Oyster Creek	Procedures Not Implemented During Plant Shutdown	The inspectors identified an NCV of Technical Specification 6.8.1, Procedures and Programs, because Oyster Creek operators did not adequately implement procedures when performing a plant shutdown. Specifically, the operators did not ensure that all personnel on shift had received Just-in-Time-Training for their role in the shutdown; operators did not perform a reactivity Heightened Level Awareness brief for the shutdown, and did not insert source range monitors (SRMs) in accordance with procedure. These performance deficiencies contributed to two unanticipated criticalities during the shutdown. Exelon entered this issue into their corrective action program as IR 2412093 and conducted a root cause analysis. This finding is more than minor because it affected the procedure quality attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability and capability of systems that respond to initiating events. Specifically, Exelon did not implement procedures during the plant shutdown which contributed to two unanticipated returns to criticality which required operator action to mitigate. The inspectors screened this issue using IMC 0609.04, Initial Characterization of Findings, Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, and IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria. Inspectors determined this finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because licensed operators did not implement processes, procedures and work instructions during the plant shutdown.
05000219/FIN-2014009-01	2014Q4	Oyster Creek	Inadequate Application of Materials, Parts, Equipment, and Processes Associated with the Electromatic Relief Valves	The NRC identified a preliminary Yellow finding and associated apparent violation of 10 CFR 50, Appendix B, Criterion III, Design Control, and Technical Specification 3.4.B, Automatic Depressurization System, because the station did not establish adequate measures for selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the electromatic relief valves (EMRVs). The violation was also preliminarily determined to meet the IMC 0305, Section 11.05, criteria for treatment as an old design issue. Specifically, on June 20, 2014, during refurbishment of EMRVs that were removed from the plant during the 2012 refueling outage, Exelon personnel identified deficiencies with the B and D EMRVs. As part of the planned EMRV actuator testing and refurbishment activities, Exelon personnel conducted bench testing on June 26, 2014. Both valves did not stroke satisfactorily and resulted in two inoperable EMRVs for greater than the Technical Specification allowed outage time of 24 hours. Exelons immediate corrective actions included placing this issue into the corrective action program as issue report 1679428 and redesigning the EMRV actuators to ensure the spring is on the outside of the guide bushing, therefore removing the possibility of the spring entering the guide bushing area and subsequently jamming the actuator causing valve failure. All of the actuators were replaced with redesigned actuators during the refueling outage in October 2014. In addition, Exelon issued a 10 CFR Part 21 report to inform the industry of the deficient EMRV actuator design. This finding is more than minor because it adversely affected the design control quality attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the design deficiency of the EMRVs and the inadequate maintenance process led to the inability of the B and D EMRVs to perform their safety function. The inspectors screened this issue for safety significance in accordance with IMC 0609, Appendix A, Exhibit 2, and determined a detailed risk evaluation was required because the EMRVs were potentially failed or unreliable for greater than the Technical Specification allowed outage time. As described in Attachment 3 to this report, a detailed risk evaluation concluded that the increase in core damage frequency (CDF) related to failure of the B and D EMRVs is in the mid E-5 range; therefore, this finding was preliminarily determined to have a substantial safety significance (Yellow). Due to the nature of the failures, no recovery credit was assigned. The dominant sequences included loss of main feedwater with failures of the isolation condensers, and failure to depressurize. This finding does not represent an immediate safety concern because Exelon replaced all of the actuators with the redesigned actuators during the refueling outage in October 2014. Further, the NRC is considering treatment of this finding as an old design issue because the condition existed since the original installation of the EMRVs, and is not indicative of current licensee performance. Additional details are discussed in Attachment 1. The inspectors determined that this finding did not have a cross-cutting aspect because the most significant contributor to the performance deficiency was not reflective of current licensee performance. Specifically, the inspectors determined that the performance deficiency existed since original installation of the EMRVs. Although an opportunity to identify this issue following original installation occurred in 2006 when Quad Cities changed the EMRV actuator design due to similar issues, the inspectors could not conclude that the issue would have likely been identified during that period since a Part 21 Report was not issued to inform the industry and NRC of the design change and industry operating experience focused on plants that completed or were scheduled to complete an extended power uprate.
05000219/FIN-2014010-01	2014Q4	Oyster Creek	Failure to Evaluate a Temporary Configuration Change	A self-revealing finding (FIN) of very low safety significance was identified for Exelons failure to implement the temporary configuration change program when a temporary repair was performed on condenser bellows expansion joint Y-1-26. The temporary repair impacted the design function of Y-1-26 and led to failure of the downstream side of the bellows, causing a loss of condenser vacuum and manual reactor scram on July 11, 2014. Exelon replaced both the expansion joint Y-1-26 and the 2nd stage reheater steam supply relief valve V-1-132 on July 11, 2014, during forced outage 1F35. Exelon entered this issue into the corrective action program (IR 2422831). This finding was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that this finding was of very low safety significance (Green) using Exhibit 1 of NRC IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, 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 (e.g. loss of condenser, loss of feed water). The inspectors determined that this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Exelon did not systematically and effectively evaluate relevant internal operating experience related to a similar condenser bellows expansion joint failure in 1986.
05000219/FIN-2015001-01	2015Q1	Oyster Creek	Post Maintenance Test Results Were Not Evaluated to Assure that Technical Specifications Requirements Were Satisfied	The inspectors identified an NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when Exelon did not document and adequately evaluate test results to assure that test requirements had been satisfied. Specifically, Exelon did not perform the proper post maintenance test procedure to assure that the requirements of Technical Specification 4.5.G.3 were satisfied following installation of a temporary modification to secondary containment. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 2440643. Corrective actions include revising the process to perform the correct post maintenance test to ensure Technical Specification 4.5.G.3 is met. This finding is more than minor because it is associated with the configuration control (Standby Gas Trains) attribute of the Barrier Integrity cornerstone 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. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, and IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process: Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014. Because the finding degraded the ability to close or isolate secondary containment, the inspectors were required to further assess the finding using IMC 0609, Appendix H, Containment Integrity Significance Determination Process, issued May 6, 2004. The inspectors determined that this finding is of very low safety significance (Green) because the decay heat values were low, given that the unit had been shut down for approximately three days, and reactor water level was greater than that required for movement of irradiated fuel assemblies within the reactor pressure vessel. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not perform the post maintenance test specified by the work order.
05000219/FIN-2015001-02	2015Q1	Oyster Creek	Inadequate Post Maintenance Testing for Emergency Service Water Pump Breaker	The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings for Exelons failure to develop an adequate post maintenance test to determine operability of the A emergency service water pump breaker. Specifically, the corrective maintenance work performed on April 16, 2013, did not correct the cause of the failure and Exelon did not perform an adequate post maintenance test to verify conditions had been corrected. As a result, the emergency service water system was returned to service even though it did not meet all the requirements for operability. The issue was not identified and resolved until a subsequent surveillance test on April 17, 2013, which identified a failed breaker. Exelon entered this issue into their corrective action program (IR 2471069). Planned corrective actions include revising work order activities to specify the correct post maintenance test. This performance deficiency is more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected its objective to ensure the availability and reliability of the systems that respond to initiating events. Specifically, the inadequate post maintenance test for A emergency service water pump breaker on April 16, 2013, led to the A emergency service water pump failing to perform its function during the subsequent surveillance testing on April 17, 2013. The inspectors assessed this finding in accordance with the IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors concluded that this finding did not represent an actual loss of function of the emergency service water system for greater than its technical specification allowed outage time (15 days). Therefore, the inspectors determined that this finding is of very low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Work Management, in that Exelons work planning and executing of work activities did not include documented instructions for performing an adequate post maintenance test.
05000219/FIN-2015001-03	2015Q1	Oyster Creek	Incomplete 50.72 and 50.73 Reports Associated with Secondary Containment Integrity	The inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a) in that Exelon did not provide complete information in reports submitted per 10 CFR 50.72 and 10 CFR 50.73. Specifically, a licensee event report (LER) submitted on November 18, 2014, did not discuss a separate, partially opened secondary containment door that was discovered during the same time frame, which could have prevented the fulfillment of the safety function of secondary containment, and therefore was required to be discussed in the original LER. Exelon entered this issue into their corrective action program as IR 2440641. Planned corrective actions include revising the original LER to add a discussion of the partially opened secondary containment door. The inspectors determined that not providing a complete report in accordance with 10 CFR 50.9(a) is a performance deficiency that was reasonably within Exelons ability to foresee and correct and should have been prevented. Because the issue had the potential to affect the NRCs ability to perform its regulatory oversight function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. In accordance with Section 2.2.2.d of the NRC Enforcement Policy, the inspectors determined that the performance deficiency identified with the reporting aspect of the event is a Severity Level IV violation because it is of more than minor concern with relatively inappreciable potential safety significance and is related to findings that were determined to be more than minor issues. In accordance with IMC 0612, Appendix B, this issue was not assigned a cross-cutting aspect.
05000219/FIN-2015001-04	2015Q1	Oyster Creek	Licensee-Identified Violation	Technical Specification 3.5.B, Secondary Containment, requires in part, that secondary containment integrity be maintained at all times when the reactor vessel head and the drywell head are in not in place. Technical Specification 1.14, Secondary Containment Integrity, requires in part, that the standby gas treatment system is operable. Technical Specification 4.5.G.3 specifies that with the trunnion room door open and the trunnion room is isolated from secondary containment in support of outage activities, testing of the standby gas treatment system to be performed to demonstrate the capability to maintain 14 inch of water vacuum under calm wind conditions and a standby gas treatment system filter train flow rate of not more than 4000 cfm. Contrary to Technical Specification 3.5.B, on September 20, 2014, with the reactor vessel head and drywell head removed for the refueling outage, Exelon determined that they did not have secondary containment integrity when performing testing to demonstrate standby gas treatment system capability in accordance with Technical Specification 4.5.G.3 and subsequently found that the outer railroad air lock personnel access hatch had not been closed properly, which prevented a proper vacuum from being achieved. Exelon entered this issue into the corrective action program as IR 2383852. Using guidance in IMC 0609, Appendices G and H, the inspectors determined that this finding was of very low safety significance (Green) because the decay heat values were low and the reactor water level inventory was above that required to move irradiated fuel.
05000219/FIN-2015001-05	2015Q1	Oyster Creek	Licensee-Identified Violation	Technical Specification 6.8.1b states that, Written procedures shall be established, implemented, and maintained covering surveillance and test activities of equipment that affects nuclear safety and radioactive waste management equipment. Contrary to the above, from August 2012 through September 2013, Exelon took no action following receipt of ten lubricating oil analysis report results taken from two emergency diesel generator No. 2 sample locations which indicated silver content at 1.0 ppm, which exceeded procedural action levels. Specifically, Exelon maintenance procedure MA-AA-716-230-1001, Oil Analysis Interpretation Guideline, Section 3 governs safety system oil analyses and describes actions to be taken when equipment wear metals exceed specific thresholds, as obtained through monthly oil analysis. Section 3 of procedure MA-AA-716-230-1001 lists potential actions to be taken when oil analysis results indicate silver content above 0.3 and 0.7 ppm respectively. These actions include resampling immediately to verify abnormal results, performing confirmatory testing using more accurate methods if required, reviewing all vibration and thermography data immediately for adverse trends, and contacting the equipment manufacturer for additional assistance. Exelon identified this issue on October 21, 2013, during the performance of a 24-month lubricating oil system inspection on the emergency diesel generator No. 2 when silver metal shavings were found in the main lubricating oil filter housing and in the sump below cylinder #15. The inspectors determined that the failure to identify an out-of-specification lubricating oil sample result on numerous occasions was a performance deficiency that was within Exelons ability to foresee and correct. The inspectors determined that the issue adversely impacted the reliability of the safety-related emergency diesel generator in that the wrist pin bearing was degraded and had partially failed. The inspectors determined that the issue was of very low safety significance (Green) because it did not: affect design or qualification; represent a loss of system or function; exceed technical speciation allowed outage times; and involve external events. Exelon entered this issue into the corrective action program as IR 1575045.
05000219/FIN-2015002-01	2015Q2	Oyster Creek	Inadequate Assessment of 4k Emergency Switchgear Roll-Up Door Degraded Floor Gasket	The inspectors identified a finding associated with Exelon procedure, OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess a degraded floor gasket for the D emergency 4 kilovolt (kV) switchgear roll-up door. Specifically, Exelon did not adequately assess the flood and fire functionality of the degraded gasket, which is credited to provide protection to safety-related D emergency 4kV switchgear during a postulated internal flood event and to contain the carbon dioxide (CO2) gaseous suppression system during a postulated fire within the D switchgear room. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing the operability determination procedure and enhancing operator training in fire and flood functionality of gaskets. Additional corrective actions included repairing the gasket and performing a detailed analysis of the ability of degraded gasket to meet its flooding and fire function. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone, and affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the degraded floor gasket could have resulted in increased water level in the D emergency 4kV switchgear room during a postulated internal flood due to a fire water pipe rupture, therefore affecting the reliability of the D emergency 4k switchgear to perform its safety function. In addition, the degraded floor gasket could have resulted in CO2 leakage out of the D emergency 4k switchgear room during a postulated fire in that room, therefore affecting the reliability of the D emergency 4k switchgear gaseous suppression system to perform its safety function. The inspectors determined that this finding is of very low safety significance (Green) because it is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate issues to ensure that resolutions address the causes and extent of conditions commensurate with their safety significance. Specifically, Exelon staff did not thoroughly evaluate the issue associated with the degraded floor gasket for fire and flood functionality.
05000219/FIN-2015002-02	2015Q2	Oyster Creek	Failure Rates Exceed Twenty Percent for Annual Requalification Exam	A self-revealing finding was identified associated with inadequate licensed operator performance during licensed operator requalification exams in accordance with TQ-AA-150, Operator Training Program. Specifically, two of seven crews failed the simulator scenario portion of the requalification examinations. As an immediate corrective action, the crews that failed were restricted from licensed duties. Exelon entered this issue into the corrective action program, and facility training staff remediated the crews (the crews were retrained and successfully retested), and those crews were returned to licensed duties. This finding is more than minor because it is associated with the human performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, two of seven crews failed to demonstrate a satisfactory understanding of the knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors determined the finding to be of very low safety significance (Green) because it is related to requalification exam results, did not result in a failure rate of greater than forty percent, and the two crews were remediated (i.e., the crews were retrained and successfully retested) prior to returning to shift. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Exelon staff did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce.
05000219/FIN-2015002-03	2015Q2	Oyster Creek	Reactor Water Cleanup Procedure Not Followed Resulting in a Level Transient	A self-revealing NCV of Technical Specification 6.8.1(a), Procedures and Programs, was identified because Exelon did not follow procedure 303, Reactor Cleanup Demineralizer System, during the system restoration on March 26, 2015. Specifically, during startup from a forced outage (1F36), Exelon did not follow procedure 303, which required correct valve lineups for system restoration of reactor water cleanup (RWCU) after system isolation. This resulted in decreasing reactor water level, which was automatically terminated by a second RWCU isolation. Exelon entered this issue into the corrective action program. Planned corrective actions include enhancing operator training in system knowledge and procedure compliance and revising startup procedures. This finding is determined to be more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone 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. Specifically, Exelon did not properly lineup the RWCU system after isolation, which resulted in a water level transient and challenging the critical safety function of inventory control. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, the operators did not stop and fully communicate plant condition after the initial RWCU isolation. Consequently, operators opened the RWCU system inlet valve due to the increasing water level without following procedure guidance.
05000219/FIN-2015002-04	2015Q2	Oyster Creek	Reset of the Automatic Voltage Regulator Controller Led to an Automatic Reactor Scram	A self-revealing finding was identified because Exelon did not properly screen work in accordance with MA-AA-716-010, Maintenance Planning. Specifically, on September 12, 2014, Exelon did not screen the automatic voltage regulators (AVR) human machine interface (HMI) post-maintenance test per the maintenance planning procedure. As a result, on October 12, 2014, Exelon personnel performing the post-maintenance test did not have a work order, which would have included plant configurations and limitations associated with the test. This led to an automatic reactor scram. Exelon entered this issue into the corrective action program. Planned corrective actions include reinforcing with work planners that a work order is required for similar work activities. This finding was determined to be more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during plant operation. Specifically, resetting the three AVR controllers caused an automatic plant scram. This finding is determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not recognize and plan for the possibility of mistakes, or implement appropriate error reduction tools. Specifically, on October 12, 2014, Exelon personnel did not stop when faced with the uncertain situation of the HMI screen that did not respond as expected.
05000219/FIN-2015003-01	2015Q3	Oyster Creek	Non-Conservative Temperature Input in the Electromatic Relief Valve Voltage Drop Calculation	The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, in that Exelons measures for verifying the adequacy of design of the electromatic relief valve (EMRV) voltage drop calculation were inadequate. Specifically, non-conservative temperature inputs were used for the safety related EMRV direct current voltage drop calculation, which reduced the margin of available voltage to the EMRV solenoids. Exelon entered this issue into the corrective action program for resolution as issue report 2522756, and corrective actions included revising the calculation to include the correct temperature values and conduct an extent of condition of other voltage drop calculations that could have similar temperature values. The performance deficiency is more than minor because it is associated with the design control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, lower voltage to the EMRV solenoid at higher temperatures could affect the reliability and capability of the EMRV to perform its design function. In addition, the performance deficiency is determined to be more than minor because it is similar to example 3.j of NRC IMC 0612, Appendix E, Example of Minor Issues, in that as a result of the calculation errors and the magnitude of the decrease of margin, there was a reasonable doubt on the operability of the component. The inspectors evaluated the finding using 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 2, Mitigating System Screening Questions. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding is not assigned a cross-cutting aspect because it is not reflective of current performance. Specifically, the last time Exelon had an opportunity to evaluate this issue was in 2010 when Exelon identified that the EMRV solenoid voltage had low margin.
05000219/FIN-2015004-01	2015Q4	Oyster Creek	Preconditioning of the Standby Liquid Control Relief Valves	The inspectors identified an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XI, Test Control, because Exelon conducted unacceptable preconditioning of the standby liquid control (SLC) relief valves prior to American Society of Mechanical Engineers (ASME) code testing. Specifically, Exelon performed a SLC system functional test prior to performing the SLC relief valve as-found testing. Exelons immediate corrective actions included completing the as-found test prior to the functional test. Exelon entered this issue into their corrective action program (CAP) as issue report 2566036 to track the resolution of the issue. The performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, if left uncorrected, the performance deficiency could have the potential to lead to a more significant safety concern. Specifically, completion of the functional test prior to the replacement of the SLC relief valves masks the actual as-found condition by solidifying the valve internals. As a result, the as-found condition of the SLC relief valves have not been conducted and in the worst case scenario, could open below the design setpoint, which would divert flow back to the liquid poison tank instead of into the vessel to shut down the reactor during an anticipated transient without scram (ATWS) condition. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because the structure, system or component (SSC) maintained its operability. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation because Exelon did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Exelon did not evaluate the effect of performing the SLC system functional test prior to conducting the ASME code as-found test on the SLC relief valves.
05000219/FIN-2015004-02	2015Q4	Oyster Creek	Inadequate Problem Identification and Resolution Leading to Degradation of EPR Causing a Reactor Scram	A self-revealing finding was identified because Exelon did not adequately identify and correct conditions, per LS-AA-120, Issue Identification and Screening Process, that led to degradation of the electric pressure regulator (EPR) wiring, which resulted in an uncontrolled rise in reactor pressure and subsequent reactor scram on average power range monitor (APRM) Hi-Hi Flux. Specifically, Exelon failed to generate issue reports to document degraded EPR wiring that was previously identified, and therefore did not take corrective action prior to a reactor scram. Planned corrective actions include reinforcing with station personnel that an issue report is required when issues are identified. This finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and adversely impacted its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions. In accordance with IMC 0609, Attachment 4 and Exhibit 1 of Appendix A, the inspectors determined that this finding is 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 there is no cross-cutting aspect associated with this finding since it is not representative of current Exelon performance. Specifically, in accordance IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and considered not representative of present performance.
05000219/FIN-2016001-01	2016Q1	Oyster Creek	Failure to Identify a Slower than Normal Scram Time of a Control Rod Drive	The inspectors identified an NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly identify and correct a condition adverse to quality. Specifically, Exelon did not identify that the scram time test result for control rod drive 18-47 was beyond the analyzed scram time, which resulted in a degraded control rod drive. Exelon entered this issue into their corrective action program. Immediate corrective actions included fully inserting the control rod drive and developing a casual analysis to determine the degraded condition. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency affected the reliability of control rod drive 18-47 to perform its safety function due to a slower than normal scram time. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings. The inspectors determined that this finding is a deficiency that affected the design or qualification of a mitigating structure, system, or component (SSC), when the SSC maintained its operability or functionality. Therefore, the inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Identification, because Exelon did not identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, Exelon did not identify that the actual scram time of control rod drive 18-47 was beyond the analyzed scram time, resulting in a degraded control rod drive.
05000219/FIN-2016001-02	2016Q1	Oyster Creek	Failure to Use Respiratory Protection as Required in RWP/ALARA Plan for Drywell Head Reassembly	A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs was identified for Exelons failure to use respiratory protection, as required in the radiation work permit (RWP)/as low as reasonably achievable (ALARA) plan 14-406 for drywell head reassembly work on October 2, 2014. The radiation protection (RP) supervisor overseeing this work removed the respiratory protection requirement for this work contrary to the RWP/ALARA requirement and without engineering approval. As a result, two workers received an unplanned intake of radioactive material that resulted in unintended internal dose. Upon identification of the intake, Exelon stopped work on this task and subsequently reinstituted the respiratory protection requirements to complete the remaining work and entered this event into their corrective action program as issue report 2390111. This finding is more than minor because it is associated with the Occupational Radiation Safety cornerstone to ensure adequate protection of the worker from radiation exposure. Specifically, without the use of respiratory protection two workers received unintended internal dose. The inspectors evaluated the finding using inspection manual chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process. The inspectors determined that this finding is of very low safety significance (Green), because it did not result in an overexposure as defined by 10 CFR 20.1201, there was no substantial potential for an overexposure, and the ability to assess dose was not compromised. This finding has a cross-cutting aspect in Human Performance, Procedural Adherence, because Exelon did not follow procedures and work instructions. Specifically, RP supervision instructed the workers that respiratory protection was not required contrary to the applicable RWP/ALARA plan.
05000219/FIN-2016001-03	2016Q1	Oyster Creek	Inadequate Instructions for the Flexible Coupling Hose Preventative Maintenance Resulting in an Inoperable Emergency Diesel Generator	The inspectors identified a preliminary White finding and associated apparent violation of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Exelon did not appropriately prescribe instructions or procedures for maintenance on the emergency diesel generator (EDG) No. 1 cooling water system to ensure the EDG cooling flexible coupling hose was maintained to support the EDG safety function. Specifically, Exelon did not have appropriate work instructions to replace the EDG cooling flexible coupling hoses every 12 years as specified by Exelons procedure and vendor information. As a result, the flexible coupling hose was in service for approximately 22 years and subjected to thermal degradation and aging that eventually led to the failure of EDG No. 1 during operation on January 4, 2016. As a consequence of this inappropriate work instruction issue, Exelon violated Technical Specification 3.7.C because EDG No. 1 was determined to be inoperable for greater than the technical specification allowed outage time of seven days. Exelons immediate corrective actions included entering the issue into their corrective action program (issue reports 2607247 and 2610027), replacing of the EDG No. 1 and No. 2 flexible coupling hoses, and initiating a failure analysis to determine the causes of the failed flexible coupling hose. This finding is more than minor because it is associated with the equipment 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 (i.e., core damage). Specifically, the ruptured flexible coupling hose caused the failure of EDG No. 1 to perform its safety function. 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, this finding required a detailed risk evaluation (DRE) because EDG No. 1 was inoperable for greater than the technical specification allowed outage time. The DRE estimated the increase in core damage frequency was 7E-6, or White (low to moderate safety significance) for this finding. This finding does not have an associated cross-cutting aspect because the performance deficiency occurred in 2005 and is not reflective of present performance.
05000219/FIN-2016001-04	2016Q1	Oyster Creek	Licensee-Identified Violation	From 2010 to 2014, Oyster Creek made a total of four shipments of radioactive material which contained category 2 quantities of radioactive material. Oyster Creek 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 the Exelons corrective action program as IR 2484646. Corrective actions included contracting with a vendor to receive regular, prompt notifications of potentially applicable rule changes in the Federal Register.
05000219/FIN-2016002-01	2016Q2	Oyster Creek	Inadequate Maintenance Procedure associated with Reactor Recirculation Pump Seal	A self-revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the reactor recirculation pump (RRP) reassembly maintenance procedures as required by NRC Regulatory Guide 1.33, Appendix A, Section 9, Procedures for Performing Maintenance. Specifically, the RRP reassembly procedure, 2400-SMM-3226.03, Reactor Recirculation Pump Mechanical Seal Rebuild Using CAN-2A Parts, did not provide critical dimensional checks for the locking plate and seal adjusting cap. This led to the incorrect reassembly of the D RRP. Exelon entered this issue into their corrective action program as issue report 2663436. The corrective actions included repairing the D RRP and revising RRP maintenance procedures to include critical dimensional information. 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 those events that upset plant stability and challenge critical safety functions during shutdown and power operation. Specifically, the incorrect reassembly of the D RRP created a leakage path, which led to an unexpected increase in reactor coolant system (RCS) unidentified leakage. As a result, the operators inserted a manual scram on April 30, 2016. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined that this finding is a transient initiator that did not contribute to both the likelihood of 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, and therefore was of very low safety significance (Green). The inspectors determined that there was no cross-cutting aspect associated with this finding since it was not representative of current Exelon performance. Specifically, in accordance with IMC 0612, the causal factors associated with this finding occurred outside the nominal three-year period of consideration and were not considered representative of present performance.
05000219/FIN-2016004-01	2016Q4	Oyster Creek	E EMRV Failureto Stroke Due to Incorrect Reassembly	The NRC identified a preliminary White finding and associated apparent violation of Technical Specification 6.8.1, Procedures and Programs, and Technical Specification 3.4.B, Automatic Depressurization System, because Exelon failed to implement a procedure related to the maintenance of safety related equipment. Specifically, Exelon personnel did not follow electromatic relief valve (EMRV) reassembly instructions that required personnel to reinstall previously removed lock washers from the E EMRV cut-out switch lever. The incorrect reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, which led to the E EMRVs failure to perform its safety function. This resulted in one inoperable EMRV for greater than the Technical Specification allowed outage time. The issue was entered into the corrective action program as issue report 2722109, and Exelons immediate corrective actions include installing new cut-out switch lever plates with increased clearances, replacing star lock washers with split ring lock washers for additional clearance, and verifying the five EMRV solenoid actuators being installed into the drywell following the most recent refueling outage were correctly assembled. The finding is more than minor because it adversely affects the human performance 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 (i.e., core damage). Specifically, the missing lock washers due to the incorrect EMRV lever plate reassembly caused excessive friction between the solenoid frame and the cut-out switch lever, causing the cut-out switch lever to become bound in the energized position. This led to the E EMRVs failure to perform its safety function. The inspectors screened this issue for safety significance in accordance with Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and determined a detailed risk evaluation was required because the E EMRV had potentially failed or was unreliable for greater than the Technical Specification allowed outage time. A detailed risk evaluation concluded that the increase in core damage frequency (CDF) related to the failure of the E EMRV is 5.4E-6/year; therefore, this finding was preliminary determined to have a low to moderate safety significance (White). Due to the nature of the failure, no recovery credit was assigned. The dominant core damage sequences involve loss of main feedwater events with operator errors resulting in failure to make-up to the 4 isolation condensers or otherwise maintain reactor vessel level and the loss of reactor pressure vessel depressurization capability (due to common cause failure of the remaining four EMRVs). The finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon personnel did not follow station processes. Specifically, Exelon did not follow written instructions when reassembling the E EMRV. The missing lock washers resulted in excessive friction between the solenoid frame and cut-out switch lever, causing the cut-out switch lever to become bound in the energized position, which led to the E EMRVs failure to perform its safety function. (H.8)
05000219/FIN-2017002-01	2017Q2	Oyster Creek	Inadequate Assessment of Degraded Fuel Oil Filter Impact to Emergency Diesel Generator Operability	The inspectors identified a finding associated with Exelon procedure OP-AA-108-115, Operability Determinations, because Exelon did not adequately assess the No. 2 emergency diesel generator operability with a degraded fuel oil filter. Specifically, Exelon did not adequately assess the capability of the emergency diesel generator to perform its function during its credited duration time of 72 hours. Exelon entered this issue into the corrective action program for resolution as issue report (IR) 3999576 and IR 3990799 and subsequently replaced the fuel oil filter. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. This issue was also similar to Example 3j of IMC 0612, Appendix E, Examples of Minor Issues, because the condition resulted in reasonable doubt of the operability of the No. 2 emergency diesel generator and additional analysis was necessary to verify operability. The inspectors evaluated the finding using Exhibit 2, Mitigating System Screening Questions, in Appendix A to IMC 0609, Significance Determination Process. The inspectors determined that this finding was a deficiency affecting the design or qualification of a mitigating structure, system, or component (SSC), where the SSC maintained its operability or functionality. Therefore, inspectors determined the finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon did not thoroughly evaluate the issue associated with the degraded fuel oil filter and its impact to the No. 2 emergency diesel generator operability (P.2).
05000219/FIN-2017003-01	2017Q3	Oyster Creek	Inadequate Augmented Offgas System Procedure Resulted in a Manual Scram	A self -revealing NCV of Technical Specification 6.8.1, Procedures and Programs, was identified because Exelon did not adequately establish and maintain the augmented offgas (AOG) system operation procedure as required by NRC Regulatory Guide 1.33, Quality Assurance Requirements (Operation), Appendix A, Section 7, Procedures for Control of Radioactivity. Specifically, Exelon procedure 350.1, Augmented Offgas System Operation, did not include adequate guidance for placing the AOG system into a recycle or shutdown configuration following a system trip. Without this guidance, Operations personnel failed to ensure the correct configuration of the AOG system following a partial trip of the system which resulted in degraded main condenser vacuum and a subsequent manual reactor scram on July 3, 2017. This issue was entered into the corrective action program as issue report 4028402. The corrective actions included placing the AOG system in the correct configuration and revising the AOG system operation procedure to provide guidance for verifying proper alignment of the AOG system when the system is in recycle or shutdown. The inspectors determined the performance deficiency was more than minor because it was associated with the Initiating Events cornerstone attribute of Procedure Quality and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to establish an adequate procedure for verifying proper alignment of the AOG system following a full or partial trip of the system resulted in the AOG inlet valve being left in the open position, which allowed demineralized water to be siphoned from the flame arrestor tank and slowly fill the offgas hold- up pipe. This caused a degradation of main condenser vacuum and resulted in operators inserting a manual reactor scram on July 3, 2017. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and IMC 0609, Appendix A, Exhibit 1, Initiating Event Screening Questions. The inspectors determined the finding was a transient initiator that did not contribute to both the likelihood of a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition, and therefore was of very low safety significance (Green). The finding had a cross- cutting aspect in the area of Human Performance, Avoid Complacency , because Exelon failed to recognize and plan for the possibility of mistakes or latent errors and implement appropriate error reduction tools by verifying the AOG system was properly aligned following a system trip ; instead , Operations personnel relied upon using a procedure that did not contain adequate guidance to place the AOG system in the correct configuration following a system trip (H. 12)
05000220/FIN-2012005-01	2012Q4	Nine Mile Point	Failure to Develop Adequate Inspection Requirements for Main Transformer Modification Results in Reactor Scram	A self-revealing Green finding (FIN) was identified for Nine Mile Point Nuclear Station, LLC. (NMPNSs) failure to develop adequate inspection requirements for the Unit 1 main transformer replacement. As a result, improper configuration of the main transformer current transformers (CT) 11 and 12 bus bars went undetected. On October 29, 2012, the improper configuration of the CT bus bars combined with an electrical transient due to a lightning arrestor collapse in the 345kV switchyard resulted in a reactor scram. Following the scram, an investigation revealed the improper configuration of the CT bus bars. NMPNS took immediate corrective actions to correct the configuration of the CT 11 and 12 bus bars. NMPNS entered the issue into their corrective action program (CAP) as condition report (CR)-2012-009820. This finding is more than minor because it adversely affected the design control attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The finding was evaluated 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, issued June 19, 2012. The inspectors determined that this finding is of very low safety significance (Green) because while the performance deficiency caused a reactor scram, it did not result in the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance, work practices, because NMPNS did not ensure proper supervisory or management oversight of the Unit 1 main transformer replacement. Specifically, NMPNS failed to ensure proper oversight of the main transformer modification by not developing adequate inspection requirements, as required by NEP-DES-09, Engineering Specification.
05000220/FIN-2012005-02	2012Q4	Nine Mile Point	Inadequate Post Maintenance Test Results in Subsequent Failure of 11 CREVS Fan	A self-revealing Green NCV of TS 6.4.1 occurred because NMPNS failed to develop an adequate post maintenance test (PMT) to determine operability of the 11 control room emergency ventilation system. Specifically, troubleshooting on December 2 failed to identify a cause of the failure and an inadequate PMT was performed to determine operability. As a result the degraded system was returned to service even though it did not meet all the requirements for operability. The limiting condition for operation (LCO) was exited incorrectly, and the issue was not identified and resolved until subsequent surveillance testing. Following subsequent surveillance testing, the degraded circuit was repaired and a successful PMT was performed. The issue was entered into NMPNS CAP as CR-2012- 011027. This finding is more than minor because it adversely affected the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity Cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the operators in the control room from radionuclide releases caused by accidents or events. The finding was evaluated in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A. The inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency only represented a degradation of the radiological barrier function provided for the control room. This finding has a cross-cutting aspect in the area of problem identification and resolution, because NMPNS failed to thoroughly evaluate the problem such that the resolution addressed the cause. Specifically, if NMPNS would have identified the cause of the problem and performed an adequate PMT, the system would not have been restored with a degraded condition
05000220/FIN-2012005-03	2012Q4	Nine Mile Point	Assessment of Containment Leakage Due to Containment Isolation Valve Failure	On December 3, 2012, at 11:31 a.m., Unit 1 established primary containment integrity and commenced a reactor startup from an unplanned outage. The following day at 2:40 a.m., NMPNS commenced injecting nitrogen into the primary containment as part of a planned activity to reduce primary containment oxygen concentration to less than four percent as required by TS 3.3.1, Oxygen Concentration . This activity was completed 10:55 a.m., on December 4. Once an appropriate nitrogen concentration has been achieved in containment, additional makeup is generally not required. However, from December 6 - 8, on three occasions, operators added additional nitrogen to the containment to maintain pressure within procedural limits. This issue was documented in CR 2012-011157, Adverse Trend in Unit 1 Nitrogen Usage. This issue was initially classified as a priority 2 work item and NMPNS commenced initial troubleshooting activities, which included examining systems/components that were possible sources of nitrogen leakage. However, a definitive source for the leakage was not identified. On December 12, following a fourth addition of nitrogen that occurred on December 11, NMPNS increased the importance of the issue to Priority 1, formed an incident response team and staffed the outage coordination center. As part of the investigation process, NMPNS cycled several containment isolation valves in the nitrogen purge and vent system, and attempted to quantify the amount of seat leakage through the valves by opening test fittings located between isolation valves. In parallel with the troubleshooting efforts, NMPNS and vendor personnel began to develop analytical tools that could be used to quantify the amount of containment leakage. On December 13, at 6:47 p.m., after observing a decrease in containment pressure following a fifth nitrogen addition, and receiving preliminary data that a containment isolation valve local leak rate test (LLRT) between reactor containment inert gas purge and fill drywell cooling system isolation valves IV-201-31 and IV-201-32 may fail, NMPNS commenced a plant shutdown because primary integrity as required in TS 3.3.3 could not be assured. The plant reached cold shutdown on December 13 at 11:33 p.m. Subsequent NMPNS testing of containment isolation valves revealed that three valves in the reactor containment inert gas purge and fill drywell cooling system IV-201-10, IV- 201-31 and IV-201-32 had unacceptable seat leak rates. These conditions were documented in several CRs including 2012-011210 and 2012-011288. When the valves were disassembled and examined, NMPNS identified that iron oxide buildup on the valve resilient seats had prevented the valves from closing tightly and adversely impacted seat leakage performance. The reactor containment inert gas purge and fill drywell cooling system is a carbon steel system and the internal piping surface adjacent to the valves had visible signs of iron oxide degradation (rust). NMPNS corrective action included removing the loose surface rust, installing new seats on the valves, and successfully performing as-left LLRTs on the subject valves. Additional corrective actions are outlined in CR 2012-011157. This issue will be tracked as a URI pending NMPNS quantification of the drywell leakage that existed from December 3 - 13, 2012 and NRC review of the NMPNS evaluation to determine whether the issue is more than minor and whether a violation exists. NMPNS intends to complete the evaluation by January 31, 2013.
05000220/FIN-2012005-04	2012Q4	Nine Mile Point	Licensee-Identified Violation	10 CFR Part 55.53(i) requires as a condition of a license, that the licensee (licensed operator) shall have a biennial medical examination. Contrary to the above, for approximately three hours on December 14, 2012, a licensed Unit 2 Senior Reactor Operator (SRO) filled the dayshift control room supervisor role without having a fully completed biennial medical examination. The SROs previous medical examination was completed November 3, 2010 and his latest medical examination should, therefore, have been completed in November 2012. Although the SRO had successfully completed the physical testing portion of the medical examination on September 19, 2012, the examination was not complete in that it still required review and approval of the licensed medical practitioner, who was not available on that day. The licensed operator did not realize his physical was incomplete and the qualification matrix, used to track whether operators meet conditions of their licenses, identified the operator as meeting all requirements to assume licensed duties. Accordingly, the SRO did not identify this deficiency prior to assuming the shift on December 14, 2012. Approximately three hours into the shift, other station personnel performing a paperwork verification of annual examination completion identified that the SRO had not completed his required biennial medical examination. The SRO was immediately relieved of watch standing duties, and his physical was subsequently completed on December 17, 2012. No disqualifying medical conditions were identified. NMPNS promptly entered the issue into its corrective action process as CR-2012- 011258 and CR-2012-011261 and initiated a root cause investigation. An extent of condition review determined that medical physical examinations for all other Nine Mile Point licensed operators were completed within the required periodicity.
05000244/FIN-2009008-01	2009Q3	Ginna	Failure to Preclude Recurrence of a Significant Condition Adverse to Quality Associated with the Turbine Driven Auxiliary Feedwater Pump Governor Control Valve	A self-revealing apparent violation (AV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified for the failure to preclude recurrence of a significant condition adverse to quality (SCAQ) associated with the Turbine Driven Auxiliary Feedwater (TDAFW) pump governor control valve. Specifically, after identifying corrosion of the governor control valve stem in April 2005, Ginna did not take adequate corrective actions to preclude the recurrence of corrosion which led to the binding of the governor control valve and failure of the TDAFW pump on July 2, 2009.In addition, the inspectors concluded that governor control valve stem binding was the likely cause of the failure of the TDAFW pump on May 26, 2009. The overspeed trip of the TDAFW pump on May 26, 2009, was originally determined by Ginna to be failure of the governor control system relay valve. Governor control valve stem corrosion is a SCAQ because corrosion of the stem can lead to governor control valve stem binding and failure of the TDAFW pump as discussed in NRC Information Notice (IN) 94-66: Overspeed of Turbine-Driven Pumps Caused by Governor Valve Stem Binding and other related industry operating experience documents. Immediate corrective actions included entering this condition in the corrective action program (CAP), conducting a root cause analysis (RCA), replacing the governor control valve stem, and conducting weekly monitoring of the governor control valve during surveillance testing to identify any potential for stem binding. In addition, corrective actions included a follow-up inspection of the governor control valve during the fall 2009 refueling outage. Ginna will continue to monitor the governor control valve under an enhanced TDAFW surveillance program to ensure TDAFW pump operability. The finding is more than minor because it is associated with the equipment performance attribute of the Mitigating System cornerstone and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, stem corrosion caused binding of the governor control valve and led to the failure of the TDAFW pump. This finding was assessed using IMC 0609 and preliminarily determined to be White (low to moderate safety significance) based on a Phase 3 analysis with a total (internal and external contributions) calculated conditional core damage frequency (CCDF) of 8.6E-6.This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Ginna did not implement a corrective action program with a low threshold for identifying issues completely, accurately, and in a timely manner commensurate with their safety significance P.1(a) per IMC 0305.Specifically, Ginna did not identify issues associated with corrosion of the governor control valve within the corrective action program
05000271/FIN-2009004-01	2009Q3	Vermont Yankee	Failure to initiate corrective action condition reports for all deficient items identified during cooling tower inspections	The inspectors identified a Green NCVof 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and DraWings, in that Entergy did not initiate corrective action condition reports (CRs)\'for all deficient items identified during Cooling Tower (CT) inspections. Entergy entered this issue into their corrective action program (CAP) and performed an operability assessment which determined that the safety related function of the CTs was always available. The inspectors determined that the finding was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, deficiencies might not be tracked to resolution, management attention or other independent reviews would not be appropriately applied, and the need for operability determinations may be missed. The finding was determined to be of very low safety significance (Green) because the finding did not involve a design or qualification deficiency resulting in loss of operability or functionality, did not result in a loss of system safety function, and did not screen as potentially risk significant due to external initiating events. This finding had a cross-cutting aspect in the Work Practices component of the Human Performance cross-cutting area because Entergy did not follow procedures and initiate CRs to identify cooling tower deficiencies as required by operating procedure (OP) 52114. IH.4(b)
05000271/FIN-2009005-01	2009Q4	Vermont Yankee	Inadequate Risk Assessment Associated with the Low Pressure Coolant Injection Subsystem	The inspectors identified a non-cited violation (NCV) of 10 CFR 50.65 paragraph (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, because Entergy did not assess and manage the increase in risk that resulted from maintenance activities that impacted the availability of the low pressure coolant injection subsystem (LPCI). On December 4, 2009, Entergy conducted a test of the high pressure coolant injection (HPCI) system as a retest following maintenance activities. Operations placed both trains of the residual heat removal (RHR) system in the torus cooling mode. This alignment impacted the ability of the LPCI subsystem to automatically perform its function in some design basis accident scenarios. However, the inspectors noted that the LPCI subsystem was not included as part of the risk assessment, and that subsystem was not maintained as available in accordance with Entergy procedures. Entergy entered this issue into the corrective action program (CAP), and initiated a preliminary investigation to review the effectiveness of Maintenance Rule accounting for LPCI unavailability while in the torus cooling mode. The finding is more than minor because Entergy\'s risk assessment did not consider risk significant structures, systems, and components (SSCs) (i.e. LPCI subsystem) that were unavailable during the maintenance activity. The finding is associated with the Configuration Control attribute of the Mitigating Systems cornerstone, and affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding is of very low safety significance because the incremental core damage probability deficit was less than 1.0E-6. This finding has a cross-cutting aspect in the Human Performance cross-cutting area,Work Control component, because Entergy did not appropriately plan and incorporate risk insights. in work activities that impacted the availability of the LPCI subsystem. (H.3(a))
05000271/FIN-2009005-02	2009Q4	Vermont Yankee	Troubleshooting Activities on Inoperable vacuum Breakers	The inspectors identified an Unresolved Item (URI) associated with troubleshooting activities on inoperable vacuum breakers. On May 14, 2009, and August 14, 2009, Entergy declared both the V1619- 5E and V16-19-5F torus-to-drywell vacuum breakers inoperable, respectively. The vacuum breakers were declared inoperable when it was identified that their breakaway force exceeded the maximum allowable TS value of 0.5 psid. Entergy entered TS 3.7.A.6.b, which stated that up to two out of ten torus-to-drywell vacuum breakers may be determined to be inoperable provided that they are secured, or known to be, in the closed position. On September 29,2009, and October 8,2009, Entergy conducted troubleshooting activities on both inoperable vacuum breakers. The troubleshooting activities involved opening and closing the inoperable vacuum breakers to obtain breakaway force data and to possibly repair the vacuum breakers. The inspectors noted TS 3.7.A.6 did not have a condition that allowed the opening of an inoperable vacuum breaker once it is secured in the closed position in accordance with the requirements of TS 3.7.A.6.b. Furthermore, TS 3.7.A.8 stated that if TS 3.7.A.6 cannot be met, an orderly shutdown shall be initiated immediately, and the reactor shall be in a cold shutdown within 24 hours. The inspectors noted that once the inoperable vacuum breakers are secured in the closed position, TS 3.7.A.6 can be met, and that opening the inoperable vacuum breakers is a potential violation of TS 3.7.A.6.b. Vacuum breakers can be opened for surveillance testing, however, there are no TS requirements to conduct surveillances on inoperable equipment. This issue remains unresolved pending a review by the NRC Office of Nuclear Reactor Regulation to determine if this issue constitutes a violation of Entergy Vermont Yankee TS. (URI 05000271/2009005-02, Troubleshooting Activities on Inoperable Vacuum Breakers
05000289/FIN-2014004-01	2014Q3	Three Mile Island	Inadequate Evacuation Time Estimate Submittals	The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2), 10 CFR 50.47(b)(10), and 10 CFR 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Three Mile Island Nuclear Station (TMI) emergency plan as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date. Upon identification, Exelon entered this issue into its corrective action program (CAP) as issue reports (IRs) 1525923 and 1578649. Exelon submitted a third ETE for TMI on April 4, 2014, and the NRCs review of that ETE is documented in section 1EP4 of this report. The finding is more than minor because it is associated with the Emergency Preparedness cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs had the potential to reduce the effectiveness of public protective actions implemented by the OROs. The finding is determined to be of very low safety significance (Green) because it is a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The cause of the finding is related to cross-cutting aspect of Human Performance, Documentation, because Exelon did not appropriately create and maintain complete, accurate and, up-to date documentation (H.7).
05000289/FIN-2015001-01	2015Q1	Three Mile Island	Licensee-Identified Violation	LER 05000289/2014-001-00 describes an unanalyzed condition in which Exelon identified DC motor control circuits were unfused. Specifically, Exelon did not provide overcurrent protection for wiring associated with 250VDC full-voltage control circuits for four non-safety emergency bearing oil pumps in the turbine building to prevent wires from overheating due to fire-induced faults and excessive currents flowing through the cable. With enough current flowing through the cable, the potential exists that the overloaded motor control wiring could damage adjacent control circuit wiring for both instrument air compressors (IA-P-1A/B), which are needed to achieve and maintain post-fire safe shutdown for a fire in the cable spreading room. This condition could result in a loss of the associated safe shutdown components or a secondary fire in another fire area. The failure to protect safe shutdown cables from the effect of postulated fires was a performance deficiency. This performance deficiency was a violation of TMI Operating License Condition 2.C.(4), which requires, in part, post-fire safe shutdown cables remain free of the effects of fire-induced cable faults during postulated fires. Contrary to the above, Exelon identified they failed to meet this requirement and the condition existed since initial construction. The issue was more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase 2 screening criteria. The finding screened to Green based upon task number 2.3.5, and because no credible fire ignition source was determined to adversely affect the motor control circuits of concern as determined. Additionally, a fire area of concern (cable spreading area) is an alternate shutdown fire area protected by detection and an automatic suppression system. The cables in the other fire area of concern (turbine building) are Institute of Electrical and Electronics Engineers 383 (thermoset) construction with steel armor and tied to station ground which decreases the likelihood of inter-cable and intra-cable interactions. Because this finding is of very low safety significance and had been entered into Exelons corrective action program (IRs 1651702, 1658837, 1658842), this violation is being treated as a Green, licensee-identified NCV consistent with the NRCs Enforcement Policy.
05000289/FIN-2015002-01	2015Q2	Three Mile Island	Failure to Maintain Turbine Bypass Valve Simulator Modeling	A self-revealing NCV of 10 CFR Part 55.46(c), Plant-Referenced Simulators, was identified for Exelons failure to ensure that the plant-referenced simulator demonstrated expected plant response to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, Exelon failed to ensure simulator modeling of once through steam generator (OTSG) turbine bypass valve (TBV) operation was consistent with the actual plant which introduced negative operator training and challenged orderly unit shutdown on May 7, 2015. The licensee documented their corrective actions for this issue in TMI issue reports (IR) 02496279 and 2497542, which included software changes to the simulator to reflect actual system design, crew remediation, and procedure changes. The performance deficiency is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and 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 simulator difference introduced negative operator training and, as a result, challenged orderly shutdown of the unit on May 7, 2015. The inspectors evaluated the finding in accordance with NRC Manual Chapter 0609, Significance Determination Process, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process. The finding was determined to have very low safety significance (Green) because the impact on operator performance was not during a reportable event. This finding has no cross-cutting aspect assigned because the cause was not representative of current licensee performance. Specifically, the difference in TBV modeling existed since initial simulator certification on June 28, 1990.
05000289/FIN-2015003-01	2015Q3	Three Mile Island	Internal Flooding Licensing Basis Commitment Not Met	The inspectors identified a finding because Exelon failed to meet a commitment made during original licensing to mitigate an internal flooding event. Specifically, Exelon committed to making changes to the fire water supply system to mitigate the impact of a pipe rupture in the auxiliary building. The inspectors identified that the commitment actions were not completed and no changes to the commitment were identified. The inspectors determined that the failure to perform the modifications to the fire service system, as committed to the NRC in a letter dated November 10, 1972, was a performance deficiency that was reasonably within its ability to foresee and correct. Exelon documented the issue in issue report 2544387, performed an immediate operability evaluation, and developed corrective actions to restore compliance with the commitment. The inspectors determined that the performance deficiency is associated with the Mitigating Systems cornerstone attribute of protection against external factors (internal flood hazard) and is more than minor because it 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 performance deficiency adversely impacted the operators ability to detect and mitigate a fire service system pipe rupture in the safety related auxiliary building. The inspectors utilized IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, to determine the significance of the performance deficiency. The inspectors determined the finding to be of very low safety significance (Green) because the finding is not a design or qualification deficiency, does not represent a loss of system safety function or loss of a single train for greater than its allowed technical specification time, does not result in the loss of a high safety-significant maintenance rule train and does not involve the loss of function to mitigate internal flooding events. The finding is not assigned a cross-cutting aspect because the performance deficiency occurred during original plant construction and is not indicative of current plant performance.
05000289/FIN-2015004-01	2015Q4	Three Mile Island	Failure to Trend Vibration Data for Safety Related River Water Pump	The inspectors identified a finding of very low safety significance involving an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B Criterion XVI, Corrective Action Program, because Exelon did not identify and correct a condition adverse to quality on the B nuclear river water pump (NR-P-1B). Specifically, Exelon did not properly evaluate an adverse vibration trend on NR-P-1B, which resulted in exceeding its in-service test (IST) required action level and declared inoperable on October 10, 2015. Exelon entered the condition into their corrective action program (CAP) as issue report 2568763 and emergently replaced the pump, engaged the vendor for short and long term design and material changes to correct the vibration, and created process and peer check corrective actions to ensure all vibration data is reviewed timely and trends are addressed commensurate with their safety significance. The performance deficiency is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems 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. Specifically, the elevated vibrations reduced the reliability and capability of NR-P-1B to perform its safety function. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, and determined this finding to be of very low safety significance (Green) because the degraded condition was not a design deficiency that affected system operability; did not represent an actual loss of function of a system; did not represent an actual loss of function of a single train or two separate trains for greater than its technical specification (TS) allowed outage time and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the station did not thoroughly evaluate the elevated vibration data such that the issue was addressed before NR-P-1B became inoperable (P.2).
05000289/FIN-2016001-01	2016Q1	Three Mile Island	Deficient Design Control of ECCS Level Transmitter Instrument Line Heat Trace Causes Freezing and Inoperability	A self-revealing NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, was identified for failure to establish and implement adequate design control measures to assure that the borated water storage tank (BWST) was capable of performing its design function to mitigate a design basis loss of coolant accident (LOCA) event. Specifically, Exelon made a modification to the BWST level indicator safety grade heat trace circuit that placed the circuit in an unapproved electrical configuration, which failed to prevent instrument line freezing during cold weather periods, contrary to its safety-function to maintain BWST level indication operable in cold weather. This adversely impacted the availability of a BWST level indication necessary for operators to reliably perform a critical design basis manual action. Exelon documented these issues in issue reports 2609417 and 2611119. Immediate corrective actions included replacement of the affected heat trace and completion of a compatible modification to its electrical configuration. This performance deficiency was more than minor because it was associated with the design control attributes 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. Additionally, the finding was similar to example 2.f in Appendix E of IMC 0612, in that failure to properly maintain cold weather protection equipment for the BWST level transmitters resulted in DH-LT-809 becoming inoperable. The finding was of very low safety significance (Green) because it did not affect design or qualification, did not represent a loss of system function, did not cause at least one train of BWST level instrumentation to be inoperable for greater than its Technical Specification limiting condition of operation (LCO) allowed outage time, and did not involve external event mitigation systems. The finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because station personnel did not follow the heat trace procedure, which did not allow the two types of heat trace to be spliced together.
05000289/FIN-2016001-02	2016Q1	Three Mile Island	Licensee-Identified Violation	On February 6, 2016, while making preparations to perform procedure 1303-11.45, PORV Setpoint Check, a senior operator identified that the assigned risk for this planned maintenance activity was inaccurate. Specifically, the risk for the maintenance activity was Yellow, not Green, as originally determined. The reason for the inaccurate risk was due to not previously recognizing the pressurizers block valve (RC-RV-2) would be rendered inoperable during the maintenance activity. This condition could result in failure to operate the pressurizers power operated relief valve. The failure to accurately assess the risk of the power operated relief valve setpoint check was a performance deficiency that was within the licensees ability to identify and correct. The inspectors noted that this maintenance activity had an inaccurate risk assessment for at least the past three years. This performance deficiency was a violation 10 CFR Part 50.65(a)(4), which requires, in part, the licensee to assess and manage the increase in risk that may result from the proposed maintenance activity. Contrary to the above, Exelon failed to accurately assess the risk for the power operated relief valve setpoint check over the past three years. The issue was more than minor because it was associated with the configuration control attribute of the initiating systems cornerstone and it 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. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, screening criteria. The finding screened to Green because the incremental core damage probability of failing to operate RC-RV-2 is less than 1.00 10-6 per year during the short period which the valve is rendered inoperable during each performance of this maintenance activity. Exelon has entered this issue into its corrective action program (issue report 2622859) and revised the risk assigned to this maintenance activity. Because this finding is of very low safety significance and had been entered into Exelons corrective action program, this violation is being treated as a Green, licensee-identified NCV, consistent with section 2.3.2 of the NRCs Enforcement Policy.
05000289/FIN-2016003-01	2016Q3	Three Mile Island	Emergency Diesel Generator Internal Flooding Risk Not Evaluated	The inspectors identified an NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, in that Exelon did not ensure the availability of the emergency diesel generator (EDG) following a seismic event. The inspectors reviewed the TMI licensing basis for internal flooding, associated evaluations and conditions reports, and walked down safety-related structures system and components (SSCs). During this review the inspectors determined that non-seismic piping failures in the EDG room were not properly evaluated. Specifically, the inspectors determined that pressurized fire water pipes in both EDG rooms were not classified as safety-related or seismically qualified. The inspectors reviewed Exelons evaluation of the potential failure of the pipe, as assumed in the TMI design and licensing basis, and determined that operator actions were credited to mitigate the pipe failure in order to prevent water from affecting the operation of the EDGs. The inspectors determined that these operator actions could not be performed prior to water from the pipe break impacting the operation of the EDGs. Following identification of the issue, Exelon entered this issue into their corrective action program and performed an analysis on the structural loading on the fire water piping during a safe shutdown earthquake and concluded that the piping would not break during the design basis event and, therefore, the EDGs remained operable. The inspectors reviewed the analysis and found it reasonable. The inspectors determined the failure to adequately evaluate the effects of a pipe failure in the EDG room in accordance with the design and licensing basis was a performance deficiency. The performance deficiency is considered more than minor because it is associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, the performance deficiency is considered more than minor in accordance with Manual Chapter 0612, Appendix E - Question 3K, in that there was a reasonable doubt of operability for the EDGs requiring engineering calculations and analysis to resolve. 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 the finding to be of very low safety significance (Green) because the finding was determined to be a design or qualification deficiency that did not result in an inoperability. No cross-cutting attribute is assigned to this finding because the performance deficiency was not indicative of Exelons current performance. Specifically, this issue was last identified and reviewed by Exelon in issue report 1201424 in 2010.
05000289/FIN-2016004-01	2016Q4	Three Mile Island	Licensee-Identified Violation	Technical specification 3.2.12.1, "LTOP Protection", requires when the reactor vessel head is installed and indicated reactor coolant system temperature is 313F, high pressure injection pump breakers shall not be racked in unless injection valves (MU-V16A/B/C/D and MU-V217) are closed with their associated breakers open and that pressurizer level is maintained 100 inches, or restore pressurizer level to 100 inches within 1 hour. Contrary to technical specification 3.2.12.1, during reactor coolant system filling with the vessel head installed and temperature < 313F, high pressure injection pump breakers were racked in while pressurizer level was >100 inches for greater than 1 hour. The condition existed for 2 hours and 49 minutes until recognized by the operating crew when questioned by a senior reactor operator trainee, at which time the crew took immediate actions to reduce pressurizer level <100 inches within 1 hour. Additional corrective actions included crew remediation, additional main control room supervisory oversight, and procedure changes. Exelon entered this issue into the corrective action program as issue report 3949713. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix G, Shutdown Operations, Attachment 1, Exhibit 4, since the finding did not represent an inadvertent safety injection and did not render the power-operated relief valve (LTOP Protection) unavailable or degraded.
05000289/FIN-2017003-01	2017Q3	Three Mile Island	Licensee-Identified Violation	The 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.Technical specification 4.1.4, Operational Safety Review, requires each remote shutdown system function shown in Table 3.5-4 shall be demonstrated operable by the performance of the following check, test, and calibration. The technical specification surveillance requirement 4.1.4.b states that the licensee shall verify each required control circuit and transfer switch is capable of performing the intended function in accordance with the licensees surveillance frequency control program, in this caseevery refueling interval. Contrary to SR 4.1.4.b, from January, 1987, until September 2017, Exelon did not verify that each required control circuit on the Unit 1 remote shutdown panel was capable of performing the intended function. Specifically, Exelon did not test four of the required six relays for the B EDG either by operation of the components or by performance of a continuity check. Exelons corrective action included entering this issue into the CAP as issue reports 4020064 and 4047426, developing a remote shutdown system testing procedure for the B EDG system, and the completion of a risk evaluation as required by surveillance requirement 4.0.2. The inspectors determined that the finding was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. It is of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix F, Fire Protection Significance Determination Process, since the missed surveillance did not impact the ability to reach safe shutdown.
05000289/FIN-2017004-01	2017Q4	Three Mile Island	Failure to correct degraded control rod connections	The inspectors documented a self-revealing finding involving the failure to follow LS-AA-125, Corrective Action Program, Revision 14. Specifically, the licensee failed to take appropriate corrective actions to correct degraded control rod drive mechanism cable connections identified during a 2010 stuck rod event. This resulted in a rod drop event on October 10, 2017, that caused a turbine runback to 55 percent and required a plant shutdown to repair. As an immediate corrective action, the licensee replaced the Bendix 7-pin electrical connector for the control rod drive mechanism (CRDM) and performed extent of condition visual and resistance checks on the other CRDM cables. The issue was entered into their corrective action program (CAP) as issue report (IR) 04061160.The performance deficiency is 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 power operations. Specifically, a transient resulting from a dropped rod challenged the critical safety function of reactivity control. The inspectors determined that this finding was of very low safety significance (Green) since it did not cause both a reactor trip and the loss of mitigation equipment relied upon to transition the plant to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because despite indications of degradation during inspections in 2013 and 2015, the site failed to ensure that a resolution addressed the cause commensurate with its safety significance (P.2).
05000289/FIN-2017008-01	2017Q1	Three Mile Island	Failure to Correct Deficiency in Implementing Controls for Pre-Staging Material in the Reactor Building	Green. The inspectors identified a finding of very low safety significance involving a non- cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Action Program," because Exelon did not effectively correct a condition adverse to quality regarding the implementation of controls for pre-staging of materials in the reactor building. Specifically, Exelon did not effectively implement corrective actions regarding the control of pre-staging materials in the reactor building during power operations, which resulted in unsecured prohibited material in a location that had the potential, during a large break loss of coolant accident (LOCA), to be transported to and impact the emergency core cooling system (ECCS) sump. Exelon documented this finding in issue reports 2608560 and 2578255. Corrective actions include Exelon to establish a focus team, led by the maintenance manager, to ensure pre-outage loading of the reactor building is conducted in accordance with requirements and directly supervised by Exelon personnel. The performance deficiency is rnore than minor because, if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, without proper controls implemented, materials may be pre-staged in the reactor building in a quantity or configuration that may render the ECCS sump inoperable. The inspectors evaluated the finding against the Mitigating System Cornerstone using Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," and Appendix A, "The Significance Determination Process for Findings At-Power," Exhibit 2, and determined this finding to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Human Performance, Field Presence, because Exelon senior managers did not ensure the oversight of work activities by supplemental personnel (H.2).
05000289/FIN-2017008-02	2017Q1	Three Mile Island	Licensee-Identified Violation	The 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 Non-Cited Violation. Technical Specifications 6.8., "Procedures and Programs," requires, in part, that written procedures be established, implemented, and maintained covering the applicable procedures recommended in Appendix 'A' of Regulatory Guide 1.33, Revision 2, 1978. Regulatory Guide 1.33, Revision 2, "Quality Assurance Program Requirements," Appendix A, requires administrative procedures for access to containment. Exelon Administrative Procedure 1015, Revision 7, "Equipment Storage Inside Class I Building," requires that no equipment shall be stored, placed, or staged inside a Class I Building without an approved Equipment Storage Data Sheet (ESDS). It further states, in part, that within the reactor building materials such as plastic sheeting must be fastened/secured in such a way as to prevent them from being washed into the reactor building sump post-LOCA. Contrary to the above, between October 27, 2015, and October 28, 2015, Exelon did not properly implement a procedure related to the staging of equipment in preparation for a Three Mile Island, Unit 1 refueling outage. Specifically, on October 28, 2015, Exelon performed a reactor building loading walkdown to review the equipment staged for the upcoming outage. During the walkdown, Exelon noted that several items staged in the reactor building were not in accordance with TMI Procedure AP 1 015. Items inappropriately stored included loose plastic, light stands, light bulbs, a Knaack locker box, and bolt cutters. Exelon immediately removed the prohibited items from the reactor building and documented the condition in IR 2575255. The finding is more than minor because it was associated with the availability and reliability 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 loose plastic had the potential to adversely impact the ECCS by compromising the recirculation suction flow path due to blockage of the suction strainer. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Attachment 4, "Initial Characterization of Findings," and Appendix A, "The Significance Determination Process for Findings At- Power," Exhibit 2, and determined this finding screened as very low safety significance (Green) because, based on inspector review of a technical debris evaluation (ACIT 4 2578255-08) by Exelon, the finding did not represent an actual loss of function of a system.
05000317/FIN-2007003-06	2007Q2	Calvert Cliffs	Degraded fire barrier between two fire areas	During a fire protection walkdown of the Unit 1 and Unit 2 4kV switchgear rooms, on May 14, 2007, the inspectors identified a potentially degraded fire barrier between two fire areas. The inspectors noted that the fire barrier penetration was missing a retaining angle around the perimeter of a ventilation duct such that there was an open pathway between the two switchgear rooms. The inspectors also noted that the ventilation duct installation was not consistent with the inspection criteria in the penetration surveillance test procedure, STP-F-592, Penetration Fire Barrier Inspection. The inspectors provided this information to Constellation personnel. On June 6, 2007, Constellation determined that the fire barriers between the 27' and 45' 4kV switchgear rooms for both Unit 1 and Unit 2 were inoperable. Constellation entered a Technical Requirement Manual (TRM) action statement and established hourly fire watches for the inoperable barriers. Constellation also conducted a functional assessment and established compensatory action to control transient combustibles and hotwork in the affected fire areas. Additionally, an extent of condition review discovered that there were additional degraded fire barriers due to the improperly installed fire dampers, which were located in different fire areas. The inspectors identified a noncompliance to the Calvert Cliffs Renewed Facility Operating License Numbers DPR-53 and DPR-54, License Condition 2.E, because the sites fire dampers were not installed in accordance with vendor instructions as required by the National Fire Protection Association (NFPA) Standard 90A, Air Conditioning and Ventilating Systems. Specifically, License Condition 2.E, requires, in part, that Constellation is required to implement and maintain in effect all provisions of the approved fire protection program as described in the UFSAR for the facility. Section 9.8.3 of the UFSAR states, in part, that the work, equipment, and materials conform to the requirements and recommendations of the NFPA Code, Pamphlet 90A. NFPA 90A states that ventilation containing fire dampers shall be installed in accordance with the vendors instructions. Contrary to the above, Constellation did not install fire dampers in accordance with vendor instructions. The inspectors determined that the above noncompliance of License Condition 2.E met the enforcement discretion criteria specified in the NRC Enforcement Policy. The NRC Enforcement Policy, Interim Enforcement Policies, Interim Enforcement Policy Regarding Enforcement Discretion for Certain Fire Protection Issues (10 CFR 50.48), states, in part, that enforcement discretion may be exercised if a noncompliance is identified during the transition period to NFPA 805 and it is not associated with a finding of high safety significance (Red). Specifically, although the NRC identified the concern, it is likely that Constellation would have identified and corrected this issue as part of their transition to NFPA 805. Constellation entered the issue into their CAP, implemented appropriate compensatory measures, determined the violation was not of high safety significance, and would not likely have identified the issue by routine licensee efforts. The NRC determined there was no willful violation. Therefore, the NRC will not take any enforcement actions for this noncompliance because the conditions for this noncompliance meet the enforcement discretion criteria specified in the NRC Enforcement Policy.
05000317/FIN-2007004-04	2007Q3	Calvert Cliffs	Unit 2 Pressurizer Safety Valve LOW Lift Setting	On November 16, 2006, Unit 2 automatically tripped due to high pressurizer pressure during the performance of a clearance order to support scheduled maintenance. As a result of the trip, one of two pressurizer code safety valve simmered (approximately three minutes) based on acoustic monitoring indications. Constellation sent the code safety valve to Wyle labs for analysis. The as-found setpoint was 2414 psia versus a required TS value of 2475 psia. Constellation determined that the low set point was due to degraded threads on the bonnet and compression screw. The inspectors identified concerns with the ACE. Constellation wrote condition report IRE-025-193 and plan on revising the ACE. This is issue is unresolved pending the inspectors review of the revised ACE to determine if there is a performance deficiency associated with the setting of the pressurizer safety valves

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