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	Quarter	Site	Title	Description
05000244/FIN-2018002-01	2018Q2	Ginna	Incorrect Scaling Factors in Reactor Vessel Level Monitoring System Instrumentation Uncertainty Calculation	The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, when Exelon failed to ensure that adequate design control measures existed to verify the adequacy of the Reactor Vessel Level Monitoring System (RVLMS) uncertainty calculation. Specifically, Exelon failed to identify errors in the RVLMS uncertainty calculation which resulted in a reasonable doubt of operability for the system after a temporary modification was implemented.
05000333/FIN-2018002-01	2018Q2	FitzPatrick	Licensee-Identified Violation	This violation of very low safety significance was identified by Exelon and has been entered into Exelons CAP and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy Violation: 10 CFR 71.5 requires that licensees who transport licensed material comply with the applicable requirements of the Department of Transportation (49 CFR). 49 CFR 172.202(a)(1) and (a)(2) require that the shipping description on the shipping paper include the proper shipping name and identification number for the material. 49 CFR 172.302(a) requires that shipments in bulk packages be marked with the identification number. Contrary to the above, on July 12, 2016, the shipping description on the shipping paper for shipment JAF-2016-1613 from FitzPatrick to Tennessee did not include the proper shipping name and identification number for the material. Exelon identified the error during a subsequent review of the shipping paperwork. Significance/Severity Level: No examples of transportation issues are presented in IMC 0612, Appendix E (Examples of Minor Issues). IMC 0609, Appendix D, Section VII.C.e.1 lists examples of Green findings that include documentation deficiencies including failure to properly document compliance with 49 CFR requirements such as shipping papers. Corrective Action Reference: Exelon placed this issue into its CAP as CR-JAF-2016-02857. Corrective actions included providing a corrected shipping paper to the facility in Tennessee that had received the package.
05000333/FIN-2017004-01	2017Q4	FitzPatrick	Inadequate Design Control for Battery Sizing Calculation	The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, because Exelon did not verify the adequacy of the low pressure coolant injection (LPCI) motor operated valve (MOV) independent power supply (IPS) with respect to the 419 volt direct current (VDC) battery sizing calculation. Specifically, non-conservative design inputs were used for the safety-related battery sizing calculation which reduced the battery capacity margin. On November 22, 2017, Exelon performed an operability determination for the identified issue and determined that the batteries had sufficient capacity. This issue was entered into the corrective action program (CAP) as issue report (IR) 4079452. The performance deficiency was determined to be more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, based on the quantity and magnitude of the errors, there was reasonable doubt that the LPCI MOV batteries would have adequate capacity under all design conditions. 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 finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in a loss of operability. This finding does not have a cross-cutting aspect because the calculation was last revised in 2003 so the finding is not indicative of current performance.
05000333/FIN-2017004-02	2017Q4	FitzPatrick	Human Error Resulting in Unplanned HPCI Isolation	A self-revealing NCV of very low safety significance (Green) of Technical Specification (TS) 5.4, Procedures, was identified for a procedural error which resulted in the inadvertent isolation of the high pressure coolant injection (HPCI) system. Specifically, on April 4, 2017, an instrumentation and controls (I&C) technician did not correctly perform procedure ISP-175B1, Reactor and Containment Cooling Instrument Functional Test/Calibration, which caused the HPCI system to isolate. Exelons immediate response to the event included stopping the surveillance test, and developing and implementing a plan to restore the HPCI system to an operable status. The HPCI system was subsequently restored to service approximately five hours after the inadvertent isolation. Additional corrective actions included increased observations of peer checks and validation of I&C activities. This issue was entered into the CAP as IR 03993791. This performance defficiency 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 of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correctly implement procedure ISP-175B1 caused an isolation of the HPCI system and rendered it unavailable to respond to an initiating event. 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 determined that this finding required a detailed risk evaluation since the HPCI isolation resulted in a loss of safety function. Using the Standardized Plant Assessment Risk Model (SPAR), the Region I senior reactor analyst (SRA) determined this finding was of low safety significance (Green). The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because the I&C technician did not correctly implement error reduction tools and verify that the direct current voltage source was installed on the correct trip unit prior to performing the surveillance procedure. (H.12)
05000333/FIN-2017004-03	2017Q4	FitzPatrick	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 NCV. 10 CFR 50.65(a)(4) states, in part, that before performing maintenance activities (including but not limited to surveillance, post-maintenance testing, and corrective and preventive maintenance), the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. The scope of the assessment may be limited to structures, systems, and components that a risk-informed evaluation process has shown to be significant to public health and safety. Contrary to the above, on March 28, 2011, and April 16, 2015, before performing maintenance activities on the electric bay unit coolers, as discussed in Section 4OA2.4, Exelon did not assess and manage the increase in risk that resulted from the maintenance activities. This issue was documented in Condition Report JAF-2016-0838. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609 Appendix K, Flowchart 2, Assessment of RMAs. The inspectors determined that the violation was of very low safety significance (Green) because the incremental core damage probability was less than 1E-5, with three risk management actions taken during the maintenance activities.
05000333/FIN-2017003-01	2017Q3	FitzPatrick	Vent Line Socket Weld Failure	On January 14, 2017, during the initial drywell walkdown following shutdown for a refueling outage, Entergy personnel identified a through-wall leak on the vent line off of the bonnet of the motor operated gate valve on the suction side of the A reactor water recirculation pump. A three- to four-foot steam plume was observed. Entergy determined this constituted a violation of TS 3.4.4, RCS Operational Leakage, that requires RCS leakage to be limited to no pressure boundary leakage. Based on the unidentified leakage rate of 0.06 gallons per minute measured during plant operation and visual inspection of the leak area, the leak likely existed while the plant was online. The condition was reported in Event Notification 52490 as required by 10 CFR 50.72(b)(3)(ii)(A) because it represented a degradation of a principal safety barrier. The inspectors reviewed LER 05000333/2017-001, CR-JAF-2017-00245, and the associated apparent cause evaluation. Entergy determined that this leak was caused by the existing pipe support allowing for excessive lateral movement which led to higher stresses in the socket weld connection. Additionally, the recirculation pumps were operated at a reduced flow condition for an extended period during the previous cycle, which likely resulted in an increased number of vibration cycles. The inspectors also reviewed the leakage data over the previ ous cycle and Entergys operational decision making IR and determined that the existence of RCS pressure boundary leakage was not within Entergys ability to foresee and correct and therefore was not a performance deficiency. The inspectors screened the significance of the condition using IMC 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, and determined that the condition represented very low safety significance (Green) because it would not have resulted in exceeding the RCS leak rate for a small loss of coolant accident and would not have likely affected other systems used to mitigate a loss of cooling accident. Enforcement. TS 3.4.4 requires, in part, that RCS operational leakage shall be limited to no pressure boundary leakage. If pressure boundary leakage exists, the TS 3.4.4 limiting condition for operation action statement requires the unit be in at least hot shutdown within 12 hours and in cold shutdown within 36 hours. Contrary to the above, for a period that began on an unknown date that was likely more than 36 hours before January 14, 2017, and ending on January 14, 2017, RCS pressure boundary leakage existed, and the licensee did not place FitzPatrick in at least hot shutdown within 12 hours and in cold shutdown within 36 hours. This issue is considered within the traditional enforcement process because there was no performance deficiency associated with the violation of NRC requirements. IMC 0612, Power Reactor Inspection Reports, Section 03.22 states, in part, that traditional enforcement is used to disposition violations receiving enforcement discretion or violations without a performance deficiency. The NRC Enforcement Policy, Section 2.2.1 states, in part, that, whenever possible, the NRC uses risk information in assessing the safety significance of violations. Accordingly, after considering that the condition represented very low safety significance, the inspectors concluded that the violation would be best characterized as Severity Level IV under the traditional enforcement process. However, the NRC is exercising enforcement discretion (EA-17-121) in accordance with Section 3.10 of the NRC Enforcement Policy, which states that the NRC may exercise discretion for violations of NRC requirements by reactor licensees for which there are no associated performance deficiencies. In reaching this decision, the 19 NRC determined that the issue was not with in the licensees ability to foresee and correct, the licensees actions did not contribute to the degraded condition, and the actions taken were reasonable to identify and address the condition. Furthermore, because the licensees actions did not contribute to this violation, it will not be considered in the assessment process or the NRCs Action Matrix. This LER is closed.
05000333/FIN-2017002-01	2017Q2	FitzPatrick	A Control Room Ventilation Subsystems Inoperable Longer than Allowed by Technical Specifications	Green. A self-revealing Green NCV of Technical Specification (TS) 3.7.3, Control Room Emergency Ventilation Air Supply (CREVAS) System, and TS 3.7.4, Control Room Air Conditioning (AC) System, was identified for the failure to declare one subsystem of the control room AC and CREVAS systems inoperable. Specifically, on August 16, 2016, control room operators failed to declare the A CREVAS and A control room AC subsystems inoperable due to a degraded damper actuator. As a result, the A CREVAS and A control room AC subsystems were inoperable from August 16, 2016, until a compensatory measure to assist the dam per linkage by hand as needed was implemented on September 19, 2016, which exceeded the TS allowed outage time. On October 4, 2016, FitzPatrick personnel replaced the actuator. This issue was entered into the corrective action program (CAP) as JAF-CR-2016-3593. The performance deficiency is more than minor because it is associated with the structure, system, and component (SSC) and barrier performance 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. Specifically, this resulted in the A control room AC and A CREVAS subsystems being inoperable from August 16, 2016, to September 19, 2016, and the exceedance of the allowable TS out-of-service times. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 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 the performance deficiency did not represent a degradation of the radiological barrier function provided for the control room, and the finding did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere (i.e. the B train of both subsystems remained operable). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because FitzPatrick personnel failed to thoroughly evaluate the problem such that resolution addressed the cause. Specifically, FitzPatrick failed to fully evaluate the degraded condition during troubleshooting following the failed post-maintenance test (PMT) on August 16, 2016. Thorough testing and evaluation of the degraded actuator would have led to the identification of the need for replacement to restore the damper and its actuator to fully operable status. (P.2)
05000220/FIN-2017002-02	2017Q2	Nine Mile Point	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 an NCV.Technical Specification 6.4 Procedures, Section 6.4.1, states, in part, that, written procedures and administrative policies shall be established, implemented and maintained ... that cover the following activities: a. The applicable procedures recommended in Regulatory Guide (RG) 1.33, Appendix A, November 3, 1972.Appendix A of RG 1.33 lists typical safety-related activities which should be covered by written procedures. Section I.1 of RG 1.33 includes procedures for performing maintenance which can affect the performance of safety-related equipment and should be properly pre-planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Section 4.3.4 of MA-AA-796-024, Scaffold Installation, Inspection, and Removal, Revision 11 states to ensure an adequate inspection is performed upon completion of scaffold erection for planned maintenance. Contrary to the above, on June 29, 2017 it was identified by Exelon staff that a scaffold surrounding the 11 feedwater flow control valve, FCV-29-141, would have prevented manual operation as required in accordance with EOP-1, NMP1 EOP Support Procedure, Revision 01601 Attachment 26, Reactor Pressure Valve Level Control Through Feedwater Pumps 11 and 12 flow control valves, and other special operating procedures during the previous 45 days.The inspectors evaluated the finding using 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 determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. Because this violation was determined to be of very low safety significance and entered into the CAP as IR 4027382, it is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000410/FIN-2017002-01	2017Q2	Nine Mile Point	Inadequate Extent of Condition Results inUnplanned Yellow Risk Condition	The inspectors identified a Green non-cited violation (NCV) of Title 10 of the Code of Federal Regulation (10 CFR) 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, when Exelon did not assess and manage the increase in risk for online maintenance activities. Specifically, on May 24, 2017, the inspectors identified a planned surveillance activity which caused unavailability of the A residual heat removal (RHR) system minimum flow valve that was not recognized by the Exelon staff as a, Yellow, elevated risk activity in accordance with their EOOS (Equipment Out of Service) probabilistic risk assessment (PRA) model. Exelon staff generated issue report (IR) 04015294 to address the failure to recognize the Yellow, elevated risk activity and failure to review adequate extent of condition. Corrective actions include evaluating PRA to assess if risk can be reduced to Green with compensatory actions and providing training to operations to enhance PRA modeling of system availability. Following review of the PRA model, Exelon plans to evaluate all surveillance procedures as part of extent of condition that could impact availability of the A RHR minimum flow valves.This performance deficiency is more than minor because it affected the human 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, on May 24, 2017, the inspectors identified a planned activity that resulted in an unplanned Yellow risk activity during planned maintenance that resulted in unavailability of a component used to support the A RHR system. Additionally, this issue is similar to Example 7.f of IMC 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009, because the overall elevated plant risk placed the plant into a higher licensee-established risk category. The inspectors determined that this finding is of very low safety significance (Green). Because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability was less than 1E-7, this finding was determined to be of very low safety significance (Green). The cause of the finding has a cross-cutting aspect in the area of Human Performance, Teamwork, because Exelon staff did not effectively communicate internally to ensure that corrective actions were being addressed to resolve concerns with risk associated with A RHR minimum flow valve availability. Specifically, Exelon staff incorrectly believed that integrated risk management guidance corresponded to PRA availability. Thus, it was assumed risk would remain Green during surveillance and maintenance activities that resulted in the A RHR minimum flow valve being unavailable; and a failure to recognize future maintenance activities that resulted in risk being Yellow. (H.4)
05000220/FIN-2017001-01	2017Q1	Nine Mile Point	Deficient Design Control of Outboard MSIV Pilot Valve Instrument Air Supply	Green. The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, for Exelons failure to correctly translate the design basis into the NMPNS Unit 1 instrument air system to ensure the Unit 1 outboard main steam isolation valves (MSIVs) were capable of performing their design function. Specifically, the NMPNS Unit 1 Updated Final Safety Analysis Report (UFSAR) states, Reliable operation of instrument air end users and in-line components is dependent on the filtration and removal of particulates greater than 40 microns. Additional filtration for various components exists where the 40 micron limit is not satisfactory. The MSIV pilot valves at Unit 1 have a tighter clearance than the 40 micron limit. However, contrary to the UFSAR, NMPNS did not install additional filtration upstream of the pilot valves. As a result, during a surveillance test conducted on December 10, 2016, foreign material in the instrument air system potentially contributed to the failure of an outboard MSIV. Exelons immediate corrective actions included entering this issue into its corrective action program (CAP) as issue report (IR) 03959732, performing an air purge of the instrument air system to remove foreign material from the system, and replacing the current style pilot valves with new style valves with larger clearances during the spring 2017 refueling outage. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents for events. Specifically, Exelon failed to install additional filtration in the instrument air system upstream of the outboard MSIV pilot valve in accordance with the Unit 1 UFSAR even though the internal clearance of the pilot valve was significantly less than the 40 micron particulate limit. Additionally, example 3.j from IMC 0612, Appendix E, Examples of Minor Issues, provides a similar scenario to this issue. Example 3.j details that a performance deficiency is more than minor if the error results in a condition where there is a reasonable doubt of the operability of a system or component. This performance deficiency is more than minor because without the additional filtration defined in the UFSAR there 4 existed a reasonable doubt of operability for the Unit 1 outboard MSIVs. The finding was evaluated in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and determined to be of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Exelon failed to create and maintain complete, accurate, and up-to-date documentation pertaining to instrument air sampling for high particulate. Specifically, Exelon failed to develop and implement a surveillance testing program for the instrument air system that would alert personnel that particulate greater than 5 microns could jeopardize the operability of the outboard MSIVs. (H.7)
05000220/FIN-2017001-02	2017Q1	Nine Mile Point	Failure to Identify and Correct a Non- Conforming Condition in Safety-Related UPSs	Green. The inspectors documented a self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, for the failure to identify and correct a non-conformance (an inadequate capacitor) in safety-related uninterruptable power supplies (UPSs) 162 and 172. Between 2008 and 2017, this non-conformance led to multiple component failures, loss of vital power supplies, plant transients, and in one case, loss of the emergency condenser safety function. Specifically, in 2003, during a preventative maintenance activity, NMPNS installed a commercially dedicated capacitor (part number C-805) that was not rated for the normal service temperature for the application. This resulted in chronic overheating, reduction of service life, and in seven cases failures (internal shorts of C-805) which resulted in the loss of the associated safety-related UPS. Upon identification, Exelon entered each failure into the CAP conducted an apparent cause evaluation (ACE) following the 2016 and 2017 failures, and developed corrective actions to replace the underrated capacitors. The performance deficiency was determined to be more than minor because it affected the equipment performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge the critical safety functions during shutdown as well as power operations. Specifically, the underrated capacitors failure resulted in the loss of a vital alternating current (AC) bus, a support system and in one case the unplanned loss of a safety function required to bring and maintain the plant in safe shutdown. In accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, a detailed risk assessment was required. Using the NMPNS Unit 1 Standardized Plant Analysis Risk (SPAR) Model Version: 8.21, model date January 28, 2010, a Region I senior reactor analyst ran a zero maintenance condition assessment with basic events for emergency condenser (EC) motor operated valve (MOV) 39-09R and EC MOV 39-10R, normally closed condensate return isolation valves, failed for a duration of one hour. The results were a CDP of 1.37E-08. The dominant risk sequences involved loss of feedwater and loss of offsite power. As a result, the finding is of very low safety significance (Green). The performance deficiency for this finding occurred in 2008. Because the performance deficiency occurred greater than 3 years ago and is not indicative of current performance based upon the corrective actions taken following the 2016 failure, there is no cross-cutting aspect assigned to this finding.
05000220/FIN-2016003-01	2016Q3	Nine Mile Point	Licensee-Identified Violation	10 CFR 50.54q(2) requires, in part, that the license holder shall follow and maintain the effectiveness of an emergency plan that meets the requirements in appendix E and, for nuclear power reactor licensees, the planning standards of 50.47(b). 10 CFR 50.47(b)(14) requires, in part, periodic exercises be conducted to evaluate major portions of emergency response capabilities and develop and maintain key skills. Exelon procedure EP-AA-122-100, Drills and Exercise Planning and Scheduling, Revision 6, implements this planning standard and requires health physics drills be performed every 6 months. Contrary to the above, from December 28, 2015 to July 15, 2016 Exelon failed to appropriately implement its approved emergency plan by not meeting planning standard 10 CFR 50.47(b)(14). Specifically, Exelon failed to conduct and document the performance of a required health physics drill for the second half of 2015 as required by step 4.4 of Exelon procedure EP-AA-122-100. This performance deficiency was determined to be more than minor because it impacted the Emergency Preparedness cornerstone objective of ERO readiness to ensure that Exelon is capable of implementing adequate measures to protect the health and safety of the public and its workers in the event of a radiological emergency. The finding was evaluated using IMC 0609 Appendix B, Emergency Preparedness Significance Determination Process. The finding was determined to affect planning standard 10 CFR 50.47(b)(14) and matched an example of a degraded planning standard function. Therefore, the finding was determined to be of very low safety significance (Green). Exelon has entered this issue into its CAP as IR 02686128.
05000410/FIN-2016002-01	2016Q2	Nine Mile Point	Ineffective Corrective Action Results in Water Intrusion to Battery Switchgear Room	The inspectors identified a Green finding (FIN) of PI-AA-125, Corrective Action Program, Revision 3, when Exelon failed to implement adequate corrective actions in March 2003, to prevent water intrusion into the Unit 2 normal switchgear building area at elevation 237. Specifically, on May 4, 2016, the inspectors observed water leaking into the normal switchgear room through a wall on elevation 237. The leakage was through a section of the wall that contained electrical junction boxes that were not sealed. The water progressed under inverter 2BYS-SWG001B, which led to the potential for a reactor scram from an electrical fault associated with uninterruptible power supply battery breakers. Previously, a reactor scram had occurred at Unit 2 on March 4, 2014, when the inverter was lost because of an electrical fault, as such this was a known initiating event single point vulnerability . Corrective actions included entering the issue into the corrective action program (CAP) (IR 02664534), generating work order (WO) C93414574 to seal or repair the wall, and installing temporary barriers to redirect any water away from the switchboard. The WO is scheduled to be performed in October 2016 with an action to assess moving the work to the refueling outage if needed to remove the electrical junction boxes to apply coating to the wall. The finding is more than minor because it is associated with the Protection Against External Factors 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. Specifically, Exelon did not ensure the surface area behind the electrical junction boxes was coated to prevent water intrusion into the normal switchgear room at elevation 237. The water intrusion through this area of the wall had the potential to cause an electrical fault on 2BYS-SWG001B resulting in a reactor scram similar to the reactor scram in March 2014. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because it did not represent the potential for both a reactor scram and a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors did not assign a cross-cutting aspect to this finding because the performance deficiency occurred greater than three years ago; therefore, it is not considered to be indicative of current plant performance.
05000410/FIN-2016002-02	2016Q2	Nine Mile Point	Failure to Identify Wide Range Level Indication Impacts Operability of HPCS and RCIC	The inspectors identified a Green NCV of Unit 2 Technical Specification (TS) 3.5.1, Emergency Core Cooling (ECCS) Systems-Operating, and TS 3.5.3, Reactor Core Isolation Cooling (RCIC) System, for failure to ensure all necessary attendant instrumentation required for the systems to perform their specified safety functions were capable of performing their related support function in all require modes of applicability. Specifically, the inspectors identified the Unit 2 wide range level indication to be inaccurate during Mode 2 and at 200 pounds per square inch gauge (psig) reactor pressure, a mode of applicability requiring both high-pressure core spray (HPCS) and RCIC to be operable. This resulted in a high level trip signal being locked preventing HPCS or RCIC from auto initiating, rendering the systems inoperable. Upon identification, Exelon generated issue report (IR) 02667837 to address the inspectors concern regarding the wide range level indication. An action was created to evaluate the impact of the wide range level discrepancy with regard to its impact on safety-related functions to supply water in the TS Mode of Applicability. Exelon also plans to assess the need for a TS amendment. The performance deficiency was more than minor because it was 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. Specifically, Exelon failed to recognize that the wide range level indication did not provide accurate indication at low reactor pressures and temperatures, preventing automatic safety-related functions associated with high drywell pressure automatic initiation signals and manual start functions. This would require operators to manually open the HPCS and RCIC injection valves during these conditions should a loss of offsite power or loss-of-coolant accident occur. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of system safety function, and did not screen as potentially risk significant due to seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Identification. Exelon personnel had many opportunities, including during the reactor startup in May of 2016, to question operability of the instrumentation that provides input for automatic initiation and isolation signals. As a result, Exelon personnel failed to identify that the wide range level indication did not support operability of the HPCS and RCIC systems during reactor startup on May 5, 2016. (P.1)
05000410/FIN-2016002-03	2016Q2	Nine Mile Point	Failure to Understand Radiological Conditions Results in Unintended Exposure	A self-revealing NCV of TS 5.4.1 Procedures was identified when a worker performed a radiological work activity without notifying radiation protection personnel and, as a result, did not comply with procedure RP-AA-1008, Unescorted Access to and Conduct in Radiologically Controlled Areas, Revision 5, in being briefed on the necessary radiological work controls and conditions for performance of the Unit 2 reactor seal cleaning work activity. Specifically, on April 11, 2016, a worker entered the Unit 2 reactor cavity to perform inspection of the reactor seal that was highly contaminated. Although not previously discussed with radiation protection staff, the worker cleaned the highly contaminated reactor seal with rags and carried the highly contaminated rags (5 rem/hr) in his hand out of the reactor cavity, which resulted in unplanned radiation exposure to the workers hand. Exelons immediate corrective actions included reinforcing the need to properly communicate radiological work activities with radiation protection, and require workers to carry WOs with them to improve communications with radiation protection. Exelon entered the issue into the corrective action program (CAP) as IR 02654591. The failure of the worker to discuss the full scope of the radiological work activity with radiation protection staff, who were subsequently not effectively briefed on the expected radiological work conditions and requisite radiological controls needed for the work activity, is a performance deficiency that was reasonably within Exelons ability to foresee and correct. The finding was determined to be more than minor because it affected the human performance attribute of the Occupational Radiation Safety cornerstone objective. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because it did not involve: (1) as low as reasonably achievable (ALARA) occupational collective exposure planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The finding is self-revealing because Exelon was made aware of the situation when an air monitor alarmed. The finding had a cross-cutting aspect of Human Performance, Team Work, since individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, the worker did not adequately communicate to radiation protection staff, the reactor seal cleaning activity to be performed. As a result, radiation protection personnel did not prescribe sufficient radiological controls for this high-contamination work activity, and led to an unintended exposure to the workers hand.
05000410/FIN-2016001-01	2016Q1	Nine Mile Point	Inadequate Procedure Leading to Failure to Manage Elevated Risk during Preventive Maintenance	The inspectors identified a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, when Exelon did not assess and manage the increase in risk for online maintenance activities. Specifically on February 12, 2016, Exelon did not assess and manage risk during Unit 2 planned testing associated with the A residual heat removal (RHR) system heat exchanger (HX). The inspectors identified that although the testing would render the A RHR minimum flow valve 2RHSMOV4A unavailable, this was not considered as part of the planned maintenance window, which resulted in an increase in risk during the unavailability of 2RHSMOV4A. When properly calculated, plant risk should have been indicated as Yellow for the day and not Green. Exelon generated issue report (IR) 02625546 to document the inspectors concern regarding the status of the availability associated with the A RHR minimum flow valve during test setup for the A RHR HX. Exelon corrective actions included evaluating the risk management activities to be implemented when the minimum flow valves are subject to maintenance or testing activities to ensure future work is properly screened. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, Exelons failure to plan for the unavailability of the A RHR minimum flow valve resulted in Unit 2 being placed in an unplanned elevated risk category (i.e., Yellow) without ensuring adequate compensatory measures were established and briefed to ensure maximum availability, reliability, and capability of the system. This issue is similar to Example 7.f of IMC 0612, Appendix E, Examples of Minor Issues, because the overall elevated plant risk placed the plant into a higher licensee-established risk category. The inspectors evaluated the finding using Phase 1, Initial Screening and Characterization worksheet in Attachment 4 and IMC 0609, Significance Determination Process. For findings within the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones, Attachment 4, Table 3, Paragraph 5.C, directs that if the finding affects the licensees assessment and management of risk associated with performing maintenance activities under all plant operating or shutdown conditions in accordance with Baseline Inspection Procedure 71111.13, Maintenance Risk Assessment and Emergent Work Control, the inspectors shall use IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, to determine the significance of the finding. The inspectors used Flowchart 1, Assessment of Risk Deficit, to analyze the finding and calculated incremental core damage probability using Equipment Out Of Service (EOOS), Exelons risk assessment tool. The inspectors determined that had this condition existed for the full duration of the Technical Specification (TS) limiting condition for operation (LCO), the incremental conditional core damage probability would have been 3.46E-9. Because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability was less than 1E-7, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Management, because Exelon did not properly implement a process of planning, controlling, and executing the work activity such that nuclear safety was the overriding priority. Specifically, Exelon did not ensure risk was properly assessed during the planning process in accordance with WC-AA-101-1006, On-Line Risk Management and Assessment, Revision 001, prior to testing the A RHR HX, which caused unavailability of the A RHR minimum flow valve during certain periods of the test.
05000410/FIN-2016001-02	2016Q1	Nine Mile Point	50.65(a)(4) Risk Evaluation Not Properly Performed Prior to Residual Heat Removal Heat Exchanger Testing	The inspectors identified a Green non-cited (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for Exelons failure to take risk management actions (RMAs) as required by procedure OP-AA-108-117, Protected Equipment Program, Revision 004, during a Unit 2, Division III, emergency switchgear electrical maintenance window on January 27, 2016. Specifically contrary to procedure OP-AA-108-117, during planned maintenance, Exelon failed to post the unit coolers in the A and B RHR pump and HX rooms, the C RHR pump room, and their associated breakers as protected equipment although their inoperability would have resulted in both trains of the standby gas treatment system (SBGT) being inoperable which would require entry into Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 and a short term shutdown action statement. Upon identification, Exelon generated IR 02617915 to document this issue. Corrective actions included creating an action item to evaluate Attachment 3 of N2-OP-52 and to determine the relevance of the TS LCO 3.0.3 entry requirement. The inspectors determined the performance deficiency to be more than minor because it was associated with the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, contrary to OP-AA-108-117, Exelon personnel failed to include the unit coolers for the Unit 2 RHR pump and HX rooms and their associated breakers, whose unavailability would have resulted in the inoperability of both trains of SBGT and necessitated entry into LCO 3.0.3. Additionally, Examples 7.e, 7.f, and 7.g from IMC 0612, Appendix E, Examples of Minor Issues, provided similar scenarios to this issue. Example 7.e details that a performance deficiency is more than minor if a failure to include accurate TS requirements in a risk assessment and if done properly, would have required RMAs, or additional RMAs under applicable plant procedures. The inspectors evaluated the finding using Phase 1, Initial Screening and Characterization worksheet in Attachment 4 to IMC 0609, Significance Determination Process. For findings within the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones, Attachment 4, Table 3, Paragraph 5.C, directs that if the finding affects the licensees assessment and management of risk associated with performing maintenance activities under all plant operating or shutdown conditions in accordance with Baseline Inspection Procedure 71111.13, Maintenance Risk Assessment and Emergent Work Control, the inspectors shall use IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, to determine the significance of the finding. The inspectors used Flowchart 2, Assessment of RMAs, to analyze the finding and calculated incremental core damage probability using EOOS, Exelons risk assessment tool, and found the result to be less than 1E-6. The inspectors determined that had this condition existed for the full duration of the TS LCO, the incremental core damage probability would have been 6.8E-7. Because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability was less than 1E-7, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon failed to follow processes, procedures and work instructions. Specifically, Exelon failed to follow procedure OP-AA-108-117, which led to the failure to protect the unit coolers for the RHR pump rooms, HX rooms, and associated breakers which could have led to a TS LCO 3.0.3 entry.
05000220/FIN-2016001-03	2016Q1	Nine Mile Point	Inadequate Tagout Resulting in Reactor Building Closed-Loop Cooling Drain Down Event	A self-revealing Green non-cited violation (NCV) of Technical Specification (TS) 6.4.1, Procedures, was identified when a Unit 1 Exelon operator did not maintain proper configuration control of a plant system during a system tagout for planned maintenance. Specifically, on January 25, 2016, a Unit 1 non-licensed operator manipulated a reactor building closed-loop cooling (RBCLC) system drain valve out of sequence while performing a tagout for the #13 shutdown cooling (SDC) HX for planned maintenance. This resulted in unintentional draining of the operating RBCLC system, annunciation of multiple alarms in the main control room, and operators entering abnormal operating procedures to recover the RBCLC system. As part of corrective actions, proper configuration was promptly restored and the operator involved in the event was given a remediation plan for requalification and placed on an operations excellence plan. This finding is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences; and if left uncorrected, the event had potential to lead to a more significant safety concern. Specifically, the failure to quickly isolate the drain down of the RBCLC system would have required a manual reactor scram, a manual trip of all five reactor recirculation pumps (RRPs), a manual isolation of the reactor water cleanup system, a loss of cooling to the spent fuel pool (SFP) cooling system, instrument air compressors, and the control room emergency ventilation system. The inspectors evaluated the finding using IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency did not result in the loss of a support system, RBCLC, or affect mitigation equipment. This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because the non-licensed operator failed to follow Exelons procedures and the instructions he received at the pre job brief stop when manipulating the drain valve. Specifically, the non-licensed operator rationalized, without being the designated performer of the tagout, that it was acceptable to perform a valve manipulation out of sequence with the tagout plan.
05000410/FIN-2016001-04	2016Q1	Nine Mile Point	Licensee-Identified Violation	Eight-hour reports. If not reported under paragraphs (a), (b)(1), or (b)(2) of this section, the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the following: (v) Any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material. Contrary to the above, from April 2, 2014, until October 5, 2015, Exelon failed to submit an EN to the NRC within 8 hours upon discovery on a condition which could have prevented the safety function of a SSC needed to control the release of radioactivity on April 2, 2014, at 11:20 a.m. Specifically, secondary containment being declared inoperable due to both airlock doors being open at the same time in Mode 5 with an OPDRV in progress. The inspectors reviewed the violation using IMC 0612 Appendix B, Issue Screening, and the NRC Enforcement Policy. This violation impacted the regulatory process so traditional enforcement applies. Comparing this violation to the examples in the NRC Enforcement Policy Chapter 6, the violation matches Severity Level IV Example 6.9.d.9, a licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. The NRC did not rely upon the information to make any regulatory decisions and the error did not result in increased scope or effort of NRC inspections. Compliance was restored when Exelon submitted LER 05000410/2014-007-01, Secondary Containment Inoperable due to Simultaneous Opening of Airlock Doors, to correct the public record and inform the NRC. Exelon staff entered the issue into its CAP.
05000410/FIN-2016001-05	2016Q1	Nine Mile Point	Licensee-Identified Violation	The holder of an operating license under this part shall submit a Licensee Event Report (LER) for any event of the type described in this paragraph within 60 days after the discovery of the event. (v) Any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to: (C) Control the release of radioactive material. Contrary to the above from June 2, 2014, until October 5, 2015, Exelon failed to submit an LER notification to the NRC within 60 days after discovery of a condition which could have prevented the safety function of a SSC needed to control the release of radioactivity on April 2, 2014 at 11:20 a.m. Specifically, secondary containment being declared inoperable due to both airlock doors being open at the same time in Mode 5 with an OPDRV in progress. The inspectors reviewed the violation using IMC 0612, Appendix B and the NRC Enforcement Policy. This violation impacted the regulatory process so traditional enforcement applies. Comparing this violation to the examples in the NRC Enforcement Policy Chapter 6, the violation matches Severity Level IV Example 6.9.d.9, a licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73. The NRC did not rely upon the information to make any regulatory decisions, and the error did not result in increased scope or effort of NRC inspections. Compliance was restored when Exelon submitted LER 05000410/2014-007-01 to correct the public record and inform the NRC. Exelon staff entered the issue into its CAP.
05000440/FIN-2015004-01	2015Q4	Perry	Failure to Ensure Required 3 Hour Fire Barriers (gypsum board walls) Were In-Place	The inspectors identified a finding of very low safety significance and an associated NCV of Perry Operating License Condition 2.C(6), Fire Protection, for the licensees failure to maintain a three-hour fire barriers as required by the Updated Safety Analysis Report (USAR). Specifically, the inspectors identified a through-wall hole, approximately two feet wide and two feet tall in the common wall between the Unit 2, Division 1 and Division 2, direct current (DC) switchgear rooms and another hole, approximately one foot wide and one foot tall between the Unit 2, Division 2 DC switchgear room and the outside hallway. The two through-wall holes were determined to be a performance deficiency associated with compliance to the licensees fire protection program because the walls are described in the USAR as three-hour fire barriers for the rooms in question. The performance deficiency was more than minor; and thus a finding, because it was 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. The inspectors determined that the finding was of very low safety significance through analysis of the issue as a fire confinement problem and the fact that the reactor would still be able to reach and maintain safe shutdown despite the deficiency. The inspectors identified no cross-cutting issues associated with this finding because the condition has existed since at least July 2011, and therefore, is not indicative of current plant performance.
05000440/FIN-2015004-03	2015Q4	Perry	Failure to Properly Implement System Operating Instructions to Restore RHR B to Service	A finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, was self-revealed on November 4, 2015 when operators failed to follow procedures and caused an increase in level of the suppression pool. Specifically, during the process of recovering the B RHR system in accordance with system operating instruction SOIE12, Residual Heat Removal System, the operators failed to follow an If/Then statement and did not isolate the alternate keep-fill system prior to starting the RHR pump to sweep voids into the suppression pool. This resulted in the condensate transfer system remaining lined up to B RHR train and transfer of an estimated 15,000 gallons of condensate water to the suppression pool. The resultant increasing suppression pool level caused a suction swaps for both HPCS and RCIC to the suppression pool. The licensee took immediate actions to suspend the evolution, restored the suppression pool level to the middle of the acceptable band, and restored the suction sources for HPCS and RCIC to the condensate storage tank. A human performance event response investigation was conducted and the operating crew was remediated. The issue was entered into the licensees CAP as CR 201515089. The operators failure to follow the procedure was a performance deficiency that was determined to be more than minor; and thus a finding, because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significance in accordance with the licensees Maintenance Rule Program for greater than 24 hours. This finding has a cross-cutting aspect in the area of problem identification and resolution, problem resolution, because the licensee had not solved a similar issue in third quarter of 2015 that involved the same contributing factors of poor maintenance supervision, inadequate pre-job briefs and poor shift management oversight.
05000440/FIN-2015004-02	2015Q4	Perry	Liquid Effluent Calibration	The inspectors identified that the efficiency calibration for the liquid effluent radiation monitor, 0D17K0606, could not be located. The licensee performed a new efficiency determination on the monitor during digital modification upgrade on 2006 for all effluent monitors. According to the licensee, the calculated count rate using a new standard National Institute of Standards and Technology traceable sources indicated a close approximation to the liquid detector count rates data determined during the detector initial (primary) calibration. To date, the licensee was unable to provide the initial calibration paperwork indicating that the calibration count rates for the detector efficiency determinations were correlatable. The inspectors attempted to assess whether the original standard count rates for efficiency determination were correlatable to the initial calibration paperwork; however, this assessment could not be completed within this inspection period. The issue remains under review by the U.S. Nuclear Regulatory Commission (NRC) pending further information from the licensee, and is categorized as an Unresolved Item (URI) pending completion of that NRC review.
05000410/FIN-2015003-01	2015Q3	Nine Mile Point	Use of Incorrect Grounding Cart Results in Loss of Electrical Bus	The inspectors identified a self-revealing Green finding (FIN) for Exelon Generation Company, LLC (Exelon) personnels failure to stop when met with unexpected conditions as required by procedure HU-AA-101, Human Performance Tools and Verification Practices. On August 21, 2015, a Unit 2 division of normal switchgear was unintentionally deenergized which required an unplanned down power to 90 percent and special operating procedure entry. The loss of the switchgear was the result of installation of an incorrect sized grounding cart in the electric fire pump breaker cubicle during breaker maintenance. Use of the correct sized grounding cart was discussed during the pre-job brief. This resulted in the loss of the electric fire pump, half of the drywell coolers, a heater drain pump, and unplanned reactivity change. Exelon entered this issue into their corrective action program (CAP) for resolution and developed corrective actions which included developing procedures for the use of grounding carts and evaluating where other skill-of-the-craft work may pose the same risk. This finding 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 those events that upset plant stability and challenge critical safety functions during power operations. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green). The finding has a cross-cutting aspect in the area of Human Performance, challenge the unknown, because Exelon personnel failed to stop when faced with uncertain conditions. Specifically, after having been briefed on the different stab sizes for 1200 amp and 2000 amp grounding carts, Exelon personnel failed to stop and notify supervision when faced with unlabeled grounding carts stored in the same location, Exelon personnel failed to notify supervision or compare stab sizes to ensure the correct grounding cart was used.
05000220/FIN-2015002-01	2015Q2	Nine Mile Point	Failure to Notify of Changes to Work Scope	The inspectors identified a self-revealing NCV of Unit 1 Technical Specification (TS) 6.4.1, Procedures, for failure to follow the planned scaffold erection work scope that resulted in two personnel receiving unplanned internal exposures. Specifically, on January 6, 2015, workers erecting scaffolding changed the work scope that specified the use of new equipment and used unsurveyed highly contaminated scaffold parts instead, without notifying radiation protection staff of the change in work scope that resulted in two workers receiving unplanned, unintended internal radiation exposures. The failure to follow the planned work scope is a performance deficiency. The performance deficiency constitutes a finding that is more than minor because the performance deficiency was associated with the Occupational Radiation Safety attribute of program and process and adversely affected the cornerstone objective in the protection of workers from exposure to radioactive material. Specifically, failure to follow the planned work scope resulted in two personnel receiving unplanned internal exposures. The finding is not subject to traditional enforcement because it did not affect the regulatory process or result in actual safety consequences. Using IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, the finding was of very low safety significance (Green) because it did not involve: (1) as low as reasonably achievable (ALARA) occupational collective exposure planning and controls, (2) an overexposure, (3) a substantial potential for overexposure, or (4) an impaired ability to assess dose. The cause of the finding is related to the cross-cutting area of Human Performance, Challenge the Unknown, because when workers discovered potentially contaminated scaffold materials in the work area, they did not question whether or not it was appropriate to use the material in their job and did not raise the question to their supervisors or Exelon Generation Company, LLC (Exelon) radiation protection technicians prior to deviating from the planned and briefed work scope. As a result, the radiological risks were not evaluated before proceeding to utilize the unsurveyed highly contaminated components, which resulted in unintended internal radiation exposures to the workers.
05000220/FIN-2015002-02	2015Q2	Nine Mile Point	Licensee-Identified Violation	10 CFR 50.55a(g)(4)(ii) states, in part, inservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section 12 months before the start of the 120-month inspection interval. The second 10-year ISI interval program was based on the ASME B&PV Code, Section XI, 1989 Edition with no addenda and was applicable from April 5, 1998, thru April 4, 2008. ASME B&PV Code Section XI, 1989 states, in part, In instances where a location may be found at the time of the examination that does not meet >90 percent coverage, the process outlined in the EPRI TR will be followed. EPRI TR-112657, Section 6.4, Item 4 states A new relief request will be generated for any RI-ISI piping element selection for which greater than 90 percent examination coverage is not achieved. EPRI TR-112657, Section 6.4, also goes on to state Consistent with the requirements of Code Case N-460, an examination will be considered limited if less than or equal to 90 percent coverage is obtained. This relief request addresses piping element selections for the second ISI interval where less than 90 percent of the examination volume was obtained. Contrary to the above, from April 4, 2009, until February 16, 2015, Exelon failed to submit a relief request to the NRC for instances found at the time of the examination that did not meet greater than 90 percent coverage as required. This violation impacts the regulatory process; therefore, traditional enforcement applies. This violation is similar to Example 6.9.d.1 of the NRC Enforcement Policy dated February 4, 2015. This is an example of a Severity Level IV finding. Exelon identified the issue during a self assessment and entered the issue into their CAP as IR 01991177 and IR 02450858.
05000220/FIN-2015001-01	2015Q1	Nine Mile Point	Failure to Declare Notice of Unusual Event Following Sodium Bisulfite Spill in Unit 1 Screenhouse	The inspectors documented a Green NRC-identified NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) when Exelon failed to declare a Notice of Unusual Event Emergency Action Level (EAL) (HU3.1) when entry conditions were met. Specifically, on February 4, 2015, between 9:55 a.m. and 11:15 a.m., access to the Screenhouse was prohibited due to the release of a toxic gas that adversely affected normal plant operations following a spill of sodium bisulfite. Immediate corrective actions included Exelon entering the issue into their corrective action program (CAP) as issue report (IR) 02474142, formally evaluating the decision-making process used during the incident, and clarifying responsibilities for air sampling and the reporting of samples during incidents in the future. This finding is more than minor because it was associated with the Emergency Preparedness cornerstone attribute of Emergency Response Organization Performance, and affected the cornerstone objective of ensuring that a licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, between 9:55 a.m. and 11:15 a.m., access to the Unit 1 Screenhouse was prohibited due to the release of sodium bisulfite to the Screenhouse, affecting normal plant operations of the station. This finding was evaluated using IMC 0609, Appendix B, Emergency Preparedness SDP, Section 4, Failure to Implement. The performance deficiency is associated with the emergency classification planning standard and is considered a Risk-Significant Planning Standard (RSPS). The failure to declare a Notice of Unusual Event when directed by the EAL Matrix is considered a lost or degraded RSPS in accordance with Section 4 of IMC 0609. Section 4.3.c and Attachment 1 of IMC 0609, Appendix B, provide the significance determination for a Failure to Implement, and the performance deficiency was determined to be of very low safety significance (Green). The inspectors determined that the cross-cutting aspect that contributed most to the root cause is Human Performance, Challenge the Unknown: Individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, during the event, an unknown substance was released and at no point was atmospheric analysis used in the EAL declaration decision-making process. Furthermore, although spill response personnel were experiencing symptoms that were not consistent with exposure to a spill of sodium bisulfite, this unexpected condition was not fully assessed by NMPNS for significance and reportability in accordance with procedures.
05000410/FIN-2015001-02	2015Q1	Nine Mile Point	Failure to Perform an Adequate Review of Planned Work Activities Results in a Manual Reactor Scram	The inspectors documented a self-revealing Green finding (FIN) for Exelons failure to properly review a work package associated with the replacement of a reactor vessel level recorder as required by MA-AA-716-234, FIN Team Process, Revision 8. Specifically, on February 18, 2015, control room operators manually scrammed Unit 2 when reactor vessel water level unexpectedly rose above desired limits during a planned replacement of Unit 2 reactor vessel level recorder 2ISC-LR1608. The unplanned rise in reactor water level occurred when daisy chained leads associated with the level recorder were lifted, which caused an interruption in the feedwater level control circuit. The inspectors determined that Exelons failure to ensure measures were in place to address the impact on reactor vessel level prior to level recorder replacement in accordance MA-AA-716-234 was a performance deficiency that was reasonably within Exelons ability to foresee and correct and should have been prevented. This finding 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 those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Exelon did not ensure measures were in place to prevent an adverse impact on the feedwater level control system during level recorder replacement. This resulted in a rapid rise in reactor water level and subsequent manual reactor scram. 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, Avoid Complacency, because Exelon failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk even while expecting successful outcomes. Specifically, Exelon did not ensure measures were in place to address the impact of the level recorder replacement on the feedwater level control system.
05000244/FIN-2014005-03	2014Q4	Ginna	Licensee-Identified Violation	According to 10 CFR 55.21 and 33, licensed operators are required to have a physical examination every 2 years to ensure that their medical condition and general health will not adversely affect the performance of assigned operator job duties or cause operational errors endangering public health and safety. As a part of licensed operator medical evaluations, olfactory testing is required as specified in ANSI/ANS 3.4 1983. Olfactory testing in the standard states, Nose. Ability to detect odor of products of combustion and of tracer and marker gases. Contrary to this requirement, in CR-2014-003860, Exelon identified that Ginna medical staff had not been testing operators for the mercaptan marker used in natural gas. This violation is subject to traditional enforcement because of the potential impact upon the regulatory process since the operators medical conditions are reviewed by the NRC when issuing or renewing operator licenses. This issue meets the criteria for a Severity Level IV violation because upon subsequent olfactory testing, all operators were found to meet the health requirements for licensing.
05000244/FIN-2014005-04	2014Q4	Ginna	Licensee-Identified Violation	According to 10 CFR 50.74, each licensee shall notify the NRC within 30 days of a change in an operators or senior operators status including termination of any operator or senior operator. Contrary to this requirement, in AR 02120732, Exelon identified that Ginna staff did not notify the NRC of termination of two senior operators. The facility terminated the affected operators August 9, 2013, but did not notify the NRC of the change in status until September 10, 2014. This issue meets the criteria for a Severity Level IV violation because the September 10, 2014, notification did not result in increased inspection activities or cause the NRC to reconsider a regulatory position.
05000410/FIN-2014005-03	2014Q4	Nine Mile Point	Missed Surveillance Test of Alternate Decay Heat Removal Secondary Containment Isolation Valves	The inspectors identified a Green NCV of Unit 2 Technical Specification (TS) 5.4, Procedures, for Exelons failure to properly perform procedure N2-OSP-GTS-R001, Secondary Containment Integrity Test, Revision 01100. Specifically, Exelon staff failed to ensure spectacle flanges associated with alternate decay heat (ADH) secondary containment isolation were properly installed. As a result, surveillance testing associated with ADH check valves 2ADHV21A/B and 2ADHV22A/B was not performed to ensure secondary containment integrity as required by N2-OSP-GTS-R001. Exelon immediately entered this issue into their CAP as issue report (IR) 2403311. Exelon entered TS Surveillance Requirement (SR) 3.0.3, Limiting Condition for Operability Applicability, which is used when a licensee discovers that a surveillance test requirement has not been performed. As required by the TS, Exelon completed a risk evaluation of the missed surveillance and determined large early release frequency remained low without ADH secondary containment isolation. Exelon also performed extent-of-condition inspections for other systems which may not have proper alignment to ensure they are meeting TS requirements. On December 4, Exelon rotated the spectacle flanges to the no flow isolation position to ensure secondary containment integrity was maintained The finding is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance tha physical design barriers protect the public from radionuclide releases caused by accident or events. Specifically, by performing N2-OSP-GTS-R001 in 2012 and 2014 without first ensuring the spectacle flanges were properly installed, Exelon did not verify the secondar containment requirements of TS SR 3.4.6.1 were maintained. Additionally, this issue wa similar to Example 3.d in IMC 0612, Appendix E, Examples of Minor Issues, in that th failure to implement the TS SR as required was not minor if the surveillance had not bee conducted. By not correctly testing the secondary containment in 2012 and 2014, the SR o TS 3.4.6.1 was not met. In accordance with IMC 0609.04, Initial Characterization o Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Proces for Findings At-Power, the inspectors determined this finding is of very low safet significance (Green) because the finding only represents a degradation of the radiologica barrier function provided for the control room, or auxiliary, spent fuel pool (SFP), or standb gas treatment system (boiling water reactor). This finding has a cross-cutting aspect in th area of Human Performance, Avoid Complacency, because Exelon staff did not implemen appropriate error reduction tools. Specifically, operators did not use error reduction tools t ensure the spectacle flanges were installed in the no flow position and as a result, the failed to leak test the ADH check valves in the secondary containment drawdown test a required by N2-OSP-GTS-R001 (H.12).
05000244/FIN-2014005-01	2014Q4	Ginna	Incomplete and Inaccurate Medical Information Provided by Exelon Which Resulted in Issuance of an Initial Senior Operator License without a Required Medical Restriction	Exelon Generation Company, LLC (Exelon) identified two apparent violations (AVs): (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, on October 8, 2008, Ginna submitted certified copies of an NRC senior operator license application that did not specify that the applicant required a restriction (to take medication as prescribed for high blood pressure) in order to maintain medical qualifications. The NRC issued the senior operators initial license on December 5, 2008, but without the necessary medical restriction (AV #1). From October 8, 2008, until July 16, 2014, Ginna had several additional opportunities to identify that the blood pressure medication was required to compensate for a disqualifying medical condition and that a license condition was required during the licensees biennial licensed operator requalification program reviews and medical examinations. On July 16, 2014, a period that exceeded 30 days from when the condition was identified, the facility notified the NRC of the medical condition via a letter requesting amendment to the operators license to include the restriction (AV #2). On August 28, 2014, the NRC issued the license amendment with the new restriction. This issue was entered into Exelons corrective action program (CAP). The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the senior operator license application and to notify the NRC of a change in a senior operators status for a condition which was known by the licensee and were a performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC relies upon Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued a senior operator license to the applicant based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000220/FIN-2014005-01	2014Q4	Nine Mile Point	Incomplete and Inaccurate Medical Information Provided by Exelon Which Impacted Issuance of Initial and Renewal Licenses	Exelon Generation Company, LLC (Exelon) identified two AVs: (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, during an internal audit in July 2014, Exelon identified that between September 2002 and February 2012, NMPNS staff submitted certified copies of an NRC reactor operator and/or senior operator license applications for seven applicants that did not specify that the applicants required a restriction in order to maintain medical qualifications. The NRC issued the reactor operator and senior operator initial and renewed licenses for the seven applicants, but without the necessary medical restrictions (AV #1). From June 2002 through August 2014, Exelon had numerous additional opportunities to identify these potentially disqualifying medical conditions and that license conditions were required during the biennial licensed operator requalification program reviews and medical examinations. On September 25, 2014, a period that exceeded 30 days from when the conditions were identified, the facility notified the NRC of these medical conditions via a letter requesting amendment to the seven operators licenses to include the appropriate restrictions (AV #2). The NRC issued the license amendment with the new restrictions. The NRC inspectors also identified an additional example of both AVs which had not been reported by Exelon to the NRC in the September 25, 2014 letter. On November 5, 2014, Exelon requested termination of the license for that operator. This issue was entered into Exelons corrective action program (CAP) The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the reactor operator and senior operator license applications and to notify the NRC of a change in a reactor operator or senior operators status for a condition which was known by Exelon were performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC requires Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued reactor operator and senior operator licenses to the applicants based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000244/FIN-2014005-02	2014Q4	Ginna	Failure to Report a Permanent Change in a Licensed Operator's Medical Status and Request a Condition be Placed on the Operator's License	Exelon Generation Company, LLC (Exelon) identified two apparent violations (AVs): (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, on October 8, 2008, Ginna submitted certified copies of an NRC senior operator license application that did not specify that the applicant required a restriction (to take medication as prescribed for high blood pressure) in order to maintain medical qualifications. The NRC issued the senior operators initial license on December 5, 2008, but without the necessary medical restriction (AV #1). From October 8, 2008, until July 16, 2014, Ginna had several additional opportunities to identify that the blood pressure medication was required to compensate for a disqualifying medical condition and that a license condition was required during the licensees biennial licensed operator requalification program reviews and medical examinations. On July 16, 2014, a period that exceeded 30 days from when the condition was identified, the facility notified the NRC of the medical condition via a letter requesting amendment to the operators license to include the restriction (AV #2). On August 28, 2014, the NRC issued the license amendment with the new restriction. This issue was entered into Exelons corrective action program (CAP). The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the senior operator license application and to notify the NRC of a change in a senior operators status for a condition which was known by the licensee and were a performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC relies upon Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued a senior operator license to the applicant based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000410/FIN-2014005-04	2014Q4	Nine Mile Point	Assessment of UPS3B Failure Which Resulted in a Reactor Scram	Introduction. A URI was identified pending Exelons revision and approval of their root cause report associated with the failure of UPS3B that caused a Unit 2 reactor scram on March 4, 2014 Description. Unit 2 is equipped with two 10-kVA UPSs (2VBB-UPS3A and 2VBBUPS3B) that feed RPS logic trip channel loads and main steam line isolation valves control solenoids through their associated distribution panels. 2VBB-UPS3B feeds the RPS trip system B. The loads are normally energized from 600 volts alternating current (VAC) non-safety-related power. In the case of the loss of normal supply power, an inverter allows the loads to receive power from its backup direct current source. In the case of an inverter failure, the UPS can be fed from an alternate non-safety-related 600 VAC source. Each UPS is connected to its associated distribution panel through two redundant electric protective assemblies connected in series. The electric protective assemblies provide redundant protection to the RPS system and other associated essential circuits against overvoltage, undervoltage, and under frequency conditions in the non-safety-related power sources. On March 4, 2014, 2VBB-UPS3B experienced a capacitor failure on an associated circuit card. This failure prevented the UPS from transferring to its alternate source of power causing the electrical protective assemblies to trip, a loss of cooling water to the reactor recirculation pumps, and a subsequent reactor trip. Exelon staff documented the issue in CR-2014-001725 and performed a root cause analysis. Using investigative root cause techniques outlined in procedure CNG-CA-1.01-1004, Root Cause Analysis, Revision 00801, Exelon staff determined the root cause to be a lack of vendor and industry guidance and internal/external operating experience resulting in lack of PM task to preclude backplane failure. The corrective actions to prevent recurrence involved revising the PM strategy in the IQ Review and Maximo database to include replacement of all single-point vulnerable components in 2VBA*UPS2A/2B and 2VBB-UPS3A/3B. During inspection of Unit 2 LER 2014-003-00, Uninterruptible Power Supply Failure and Subsequent Manual Scram, the inspectors reviewed the root cause report associated with this event. The inspectors discovered that, although the root cause postulated that warping/cracking of the backplane contributed to UPS3B failure, when new information regarding the backplane that contradicted this root cause was discovered, Exelo personnel did not properly enter this new information into the CAP or elevate the concern to Exelon plant management. Specifically, the engineering staff and a vendor representative had examined the UPS3B backplane during the Unit 2 refueling outage and found no indication of cracking or warping. This examination occurred following management review committee approval of the root cause. This information, along with other testing performed on the UPS3B during the refueling outage, showed that the theory for potential backplane warping/cracking likely was not the actual root cause and that the corrective actions developed for backplane replacement may not prevent recurrence of the UPS failure. Exelon documented the inspectors observation in IR 2416757 and plans to evaluate the issue further and to reopen and update the root cause report. This issue will be opened as a URI pending Exelon revision of the root cause report; and NRC review of the root cause report to determine whether the issue contains performance deficiency, whether or not that performance deficiency is more than minor, and whether a violation exists. Exelon is tracking this issue through their CAP database with a date to determine root cause revision requirements by December 19, 2014. (URI 05000410/2014005-04, Assessment of UPS3B Failure Which Resulted in a Reactor Scram)
05000244/FIN-2014005-05	2014Q4	Ginna	A' Emergency Diesel Generator Output Breaker Fails to Close during Routine Surveillance Testing Resulting in a Condition Prohibited by Technical Specifications and a Potential Inability to Fulfill a Safety Function	On September 10, 2014, during performance of a routine scheduled surveillance test, STP-O-12.1, Emergency Diesel Generator A, Revision 01600, the output supply breaker to safeguards bus 14 failed to close on demand. Initial troubleshooting revealed no obvious issues with the breaker, and the output supply breaker functioned as required during a second test. A spare breaker was installed and tested satisfactorily on Enclosure 22 September 11, and the A EDG was restored to operable. Exelon concluded that the A EDG had been inoperable since the last successful performance of STP-O-12.1 on August 13, 2014. This 29 day period exceeded the TS allowable outage time of 7 days. Exelons subsequent troubleshooting revealed no electrical issues with the circuit breaker, and the failure modes and effects analysis concluded that the most likely cause of the circuit breaker failing to close was the breaker did not properly reset after performance of the surveillance test on August 13, 2014. The breaker could not be verified to be reset without an internal inspection. The original equipment manufacturer was also requested by Exelon to investigate the cause of the breaker failure. The original equipment manufacturer concluded that the lack of free movement of the operating mechanism trip shaft was the cause of the breaker not resetting and closing. The trip shaft did not move freely due to lack of end-to-end play. Exelons apparent cause evaluation associated with this issue and AR 02178745 noted that these circuit breakers undergo full PM every 4 years, and all PMs on both EDG output breakers have been done in accordance with the PM frequency. The last performance of the PM for the bus 14 breaker was on November 14, 2011. The procedure for the PM has the technicians check for free movement of the trip of the trip shaft, but not end-to-end play movement or clearances to allow end-to-end play. Additionally, the vendor manual does not direct measuring clearances or verifying end-to-end play; this is called out as a vendor task. Therefore, the inspectors concluded that no performance deficiency existed since it was not reasonable for Exelon to foresee and prevent this issue. The inspectors reviewed LER 2014-003-00 and determined that traditional enforcement applies in accordance with IMC 0612, Sections 0612-09 and 0612-13, and NRC Enforcement Policy, Section 2.2.4.d, because a violation of NRC requirements existed without an associated Reactor Oversight Process performance deficiency. The inspectors determined that the maintenance completed on the bus breaker was in accordance with vendor recommendations. This issue was considered to be a Severity Level IV violation of TS 3.8.1 in accordance with Enforcement Policy Section 6.1.d. In addition, IMC 0612, Appendix B, Figures 1 and 2, Issue Screening, were referenced in documenting this Severity Level IV self-revealing violation. This issue was entered into Exelons CAP as AR 02178745. Because it was not reasonable for Exelon to have been able to foresee and prevent the breaker failure, the NRC determined no performance deficiency existed. Thus, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation (EA-15- 004). Further, because Exelons action and/or inaction did not contribute to this violation, it will not be considered in the assessment process or the NRCs action matrix. This LER is closed.
05000220/FIN-2014005-02	2014Q4	Nine Mile Point	Failure to Make Timely Reports of Changes in Licensed Operator Medical Status Which Impacted Issuance of Initial and Renewal Licenses	Exelon Generation Company, LLC (Exelon) identified two AVs: (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, during an internal audit in July 2014, Exelon identified that between September 2002 and February 2012, NMPNS staff submitted certified copies of an NRC reactor operator and/or senior operator license applications for seven applicants that did not specify that the applicants required a restriction in order to maintain medical qualifications. The NRC issued the reactor operator and senior operator initial and renewed licenses for the seven applicants, but without the necessary medical restrictions (AV #1). From June 2002 through August 2014, Exelon had numerous additional opportunities to identify these potentially disqualifying medical conditions and that license conditions were required during the biennial licensed operator requalification program reviews and medical examinations. On September 25, 2014, a period that exceeded 30 days from when the conditions were identified, the facility notified the NRC of these medical conditions via a letter requesting amendment to the seven operators licenses to include the appropriate restrictions (AV #2). The NRC issued the license amendment with the new restrictions. The NRC inspectors also identified an additional example of both AVs which had not been reported by Exelon to the NRC in the September 25, 2014 letter. On November 5, 2014, Exelon requested termination of the license for that operator. This issue was entered into Exelons corrective action program (CAP) The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the reactor operator and senior operator license applications and to notify the NRC of a change in a reactor operator or senior operators status for a condition which was known by Exelon were performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC requires Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued reactor operator and senior operator licenses to the applicants based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000410/FIN-2014004-01	2014Q3	Nine Mile Point	Loss of Secondary Containment due to Loss of Auxiliary Boiler System	The inspectors identified a Green finding (FIN) of CNG-PR-1.01-1005, Control of Technical Procedure Format and Content, Revision 00500, because Exelon Generation Company, LLC (Exelon) provided Unit 2 operators with an inadequate auxiliary boiler system operating procedure. Specifically, N2-OP-48, Auxiliary Boiler System, Revision 01100.00, did not provide operators adequate detail to properly establish chemistry requirements for water conductivity of the auxiliary boiler system. On March 23, 2014, when Unit 2 experienced a trip of the auxiliary boiler system due to inadequate water conductivity, operators became challenged with system restoration which caused an unplanned loss of secondary containment and entry into Technical Specification (TS) 3.6.4.1, Secondary Containment. Exelon generated condition report (CR)-2014-002281 regarding this issue. Immediate corrective actions included updating chemistry requirements associated with auxiliary boiler procedures, implementing new preventive maintenance (PM) strategies for significant components associated with the auxiliary boilers, and implementing new performance monitoring plans. This finding is more than minor because it affected the procedure quality attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, over the past 2 years, the auxiliary boilers have experienced trips as a result of insufficient procedural guidance. On March 23, 2014, the inadequate procedural guidance resulted in a trip and subsequent loss of reactor building (RB) differential pressure (DP). This caused an unplanned entry into the secondary containment emergency operating procedure and an unplanned entry into TS 3.6.4.1, which presented unnecessary challenges and distractions to operators during a planned downpower. In accordance with IMC 0609.04, Initial Characterization of Findings, the inspectors used IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, because secondary containment was declared inoperable following a loss of building heating. Using Appendix A, Exhibit 3, Barrier Integrity Screening Questions, Section C, Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building, the inspectors determined that this finding is of very low safety significance (Green) because although the performance deficiency resulted in a trip of the auxiliary boiler system and a loss of secondary containment, the RB DP was restored to greater than 0.25 inches of water, within the allowable limiting condition for operation time, and did not result in a failure of the ability for secondary containment to maintain isolation or impact the ability for standby gas treatment system to maintain secondary containment. This finding has a cross-cutting aspect in the area of Human Performance, Resources, because Exelon did not ensure personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, the inadequate management oversight of the auxiliary boilers resulted in numerous failures of the auxiliary boilers due to inadequate knowledge transfer, inaccurate classifications of maintenance rule functional failures for the system, inadequate procedures for boiler operation, and inadequate procedures for the prompt restoration of secondary containment when the auxiliary boiler system is not available (H.1).
05000410/FIN-2014003-01	2014Q2	Nine Mile Point	Inadequate Surveillance Testing of Reactor Core Isolation Cooling during 165 psig Reactor Pressure Test for Surveillance Requirement 3.5.3.4	The inspectors identified a NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, due to Exelon staffs procedures for meeting Unit 2 Technical Specification (TS) Surveillance Requirement (SR) 3.5.3.4 being inadequate since they did not test all required functions over the pressure range they were required since the start of plant operation. Specifically, inspectors identified that reactor core isolation cooling (RCIC) was being started with the flow controller in manual during the 165 pounds per square inch gauge (psig) reactor pressure test as opposed to automatic, which is its normal lineup. As a result, the RCIC system has not been adequately tested to develop flow at low reactor pressures to ensure that the surveillance had been met and that the RCIC system met its design basis. This finding is more than minor because it affected the equipment performance attribute of the Mitigating Systems cornerstone and its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This finding is similar to Example 3.d in IMC 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues. Specifically, the inadequate testing of the RCIC system with reactor pressure 165 psig has led to uncertainty in the reliability and capability of the system to perform at low reactor pressures. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the deficiency affects only the design or qualification of a mitigating SSC; and the design or qualification issue is not currently impacting its operability. The inspectors did not assign a cross-cutting aspect to this finding because the performance deficiency is not indicative of present performance because Exelons incorrect interpretation for conducting TS SR 3.5.3.4 did not occur within the last 3 years.
05000220/FIN-2014003-02	2014Q2	Nine Mile Point	Failure to Correct a Significant Condition Adverse to Quality in a Timely Manner	The inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for failure to ensure that corrective actions to preclude repetition for a significant condition adverse to quality were implemented in a timely manner. Specifically, corrective actions to preclude repetition for the April 16, 2013, loss of shutdown cooling event to revise two inadequate Unit 1 procedures had not been completed over a year later. If left uncorrected, the inspectors determined there was the potential for 10 different pumps and breakers to unexpectedly trip upon restoration of a direct current (DC) bus. The loss of several of these pumps and loads would result in an unexpected plant transient or require a manual reactor trip. Exelon wrote condition report (CR)-2014-005693 in response to the inspectors questions and determined that inadequate resources were assigned to this corrective action to preclude repetition. Procedures N1-OP-47A, 125 (volts direct current) VDC Power System, and N1-SOP- 47A.1, Loss of DC, were subsequently revised and issued on June 12, 2014. This finding is more than minor because it impacted the procedure quality attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, if left uncorrected, there was the potential for 10 different pumps and breakers to unexpectedly trip upon restoration of a DC bus. Several of these pumps and loads would result in an unexpected plant transient or require a manual reactor trip. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event and affected mitigation equipment. This finding has a cross-cutting aspect of in the area of Problem Identification and Resolution, Resolution, because Exelon did not take effective corrective actions to address an issue in a timely manner commensurate with its safety significance. Specifically, Exelon failed to implement corrective actions to prevent recurrence (CA#1 from CR-2013-002926), to revise procedures N1-SOP-47A.1 and N1-OP- 47A to contain adequate guidance to ensure recovery from a loss of a DC bus would not result in an unexpected plant transient a year after the event occurred.
05000410/FIN-2014003-03	2014Q2	Nine Mile Point	Loss of Spent Fuel Cooling Pump due to Inadequate Procedures	A self-revealing NCV of TS 5.4.1, Procedures, was identified at Unit 2 for Exelons failure to provide procedures to override valve 2SFCAOV154, filters inlet isolation valve, prior to loss of offsite power (LOOP) testing. Specifically, procedures N2-OSP-EGS-R004, Operating Cycle Diesel Generator Simulated Loss of Offsite Power With ECCS Division I & II, Revision 01200, and N2-VLU-01, Valve Lineup and Valve Operations, Revision 00001, did not contain adequate guidance to differentiate between overriding the valve (open position) and repositioning the valve to its non-failure position (closed position). As a result, on March 28, 2014, while implementing N2-OSP-EGS-R004, the spent fuel cooling pump A tripped off and had to be restored. Exelon entered the loss of spent fuel pool (SFP) cooling into the corrective action program (CAP) as CR-2014-002507. Corrective actions included coaching the individuals involved and reinforcing Exelons expectations regarding what information should be discussed during pre-job briefs. This finding is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, if N2-OSP-EGS-R004 and Figure 3, Operation of Bettis Actuators, of N2-VLU-01 are not revised, there is the potential for plant operators to incorrectly assume that Section 3.0 of N2-VLU-01 Figure 3 is the valve lineup required by the N2-OSP-EGS-R004, closing 2SFCAOV154 and causing a subsequent pump trip where the loss of cooling may have more significant consequence leading to an increase in the temperature of the spent fuel cooling pump. In accordance with IMC 0609.04, Initial Characterization of Findings, issued June 19, 2012, the inspectors used IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, because the SFP was still isolated from the reactor core at the time of the finding. Using Appendix A Exhibit 3 Barrier Integrity Screening Questions Section D, Spent Fuel Pool (SFP), the inspectors determined that this finding is of very low safety significance (Green) because although the performance deficiency adversely affected decay heat removal capabilities from the SFP, the pool temperature did not exceed the maximum analyzed temperature limit specified in the sitespecific licensing basis, the performance deficiency did not involve a fuel handling error, did not affect the SFP neutron absorber, and did not result in a loss of SFP water inventory. This finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because Exelon did not ensure that individuals stopped in the face of uncertain conditions. Specifically, after recognizing that N2-VLU-01 Figure 3 contained multiple sections that were not clearly labeled, plant operators did not stop and consult the senior reactor operator but continued on with an erroneous assumption as to which section to use.
05000410/FIN-2014003-04	2014Q2	Nine Mile Point	Failure to Identify Single-Point Vulnerabilities Results in a Manual Reactor Scram	A self-revealing Green Finding (FIN) was identified at Unit 2 against procedure CNG-AM-1.01-2000, Scoping and Identification of Critical Components, Revision 00200. Specifically, Exelon staff performed an inadequate AP-913 evaluation in 2006. This evaluation failed to identify that reactor recirculation pump (RRP) switches S101A and S101B were single-point vulnerable components, so mitigating strategies to ensure proper operation to minimize plant risk were not developed. As a result, on December 2, 2013, both RRPs failed to properly shift from fast to slow speed resulting in a loss of all recirculation flow through the core and requiring operators to insert a manual reactor scram in accordance with plant procedures. Exelon generated CR-2013-009735, performed a root cause analysis (RCA), and developed corrective actions which included revising procedure N2-OP-29, Reactor Recirculation System, Revision 01801, to direct operators to manually start the low frequency motor generator sets, implementing a preventive maintenance activity for these switches, and developing plans to replace the switches during the next refueling outage (RFO). This finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and adversely impacted the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Additionally, the performance deficiency is similar to Example 4b of IMC 0612, Appendix E, Examples of Minor Issues, in that the error resulted in a plant trip. Specifically, the failure to identify switches S101A and S101B as single-point vulnerabilities and develop appropriate mitigating strategies resulted in the failure of the switches and a manual reactor scram on December 2, 2013. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green); the performance deficiency 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 feedwater). The inspectors did not assign a cross-cutting aspect to this finding because the performance deficiency was determined to have occurred in 2006, and the guidance in the current revision of CNG-AM-1.01-2000, Appendix A, was sufficient for Exelons root cause team to determine the switches should have been screened in. Therefore, this finding is not indicative of current licensee performance and no cross-cutting issue was assigned.
05000293/FIN-2014003-01	2014Q2	Pilgrim	Failure to Manage a Yellow Risk Condition for Unavailable Torus Vent Valve	The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.65 paragraph (a)(4) because Entergy did not identify and manage risk for emergent maintenance on primary containment isolation valves (PCIVs). Specifically, an incorrect risk assessment resulted in Entergy not recognizing an increase in risk to a Yellow condition, and therefore no risk management actions were taken. Entergy has captured this issue in condition report (CR)-PNP-2014-2007, has corrected the inadequate risk assessment, and has initiated an apparent cause evaluation (ACE) to determine causes and appropriate corrective actions. The performance deficiency was more than minor because if left uncorrected the failure to recognize risk and take appropriate risk management actions has the potential to lead to more significant safety concerns. Moreover, a review of IMC 0612, Appendix E, Minor Examples, identified that Section 7, Maintenance Rule, Example e, reflected a similar more than minor example, in that the outcome of the overall elevated plant risk put the plant into a higher risk management category and thereby required additional risk management actions. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, issued May 19, 2005, the inspectors determined that this finding is of very low safety significance (Green) because the Incremental Core Damage Probability Deficit for the duration of the activity was less than 1.0 E-6 per year (approximately 1.29 E-7 per year). This finding has a cross-cutting aspect in the area of Human Performance, Consistent Process, because when faced with the requirement to perform emergent, unscheduled maintenance, Entergy did not use a consistent, systematic approach to make decisions, and did not incorporate appropriate risk insights. Specifically, while Entergy had the tools and processes in place to assess risk for emergent conditions, individuals did not consistently use this process, and therefore did not recognize the elevated risk condition.
05000293/FIN-2014003-02	2014Q2	Pilgrim	Failure to Comply with TS Required Actions for Inoperable PCIV	The inspectors identified a Green NCV of Technical Specification (TS) 3.7.A, Primary Containment, because Entergy failed to comply with the TS-required actions for inoperable PCIVs. Specifically, while maintenance was being performed on an inoperable automatic PCIV, Entergy failed to either isolate and deactivate at least one containment isolation valve in the same line, or to complete an orderly shutdown to the Cold Shutdown condition within 24 hours. Entergy has captured this issue in CR-PNP-2014-2008, and has assigned corrective actions to update the Pilgrim TS bases document to provide additional guidance on acceptable methods of PCIV isolation. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone, and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers (i.e. containment) protect the public from radionuclide releases caused by accidents or events. Specifically, Entergys failure to close and deactivate a valve in the same line as the inoperable PCIV as required by TS did not ensure the operability of the primary containment. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, effective July 1, 2012, the inspectors determined that this finding is of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components. This finding has a cross-cutting aspect in the area of Human Performance, Conservative Bias, because Entergy did not use decision-making practices that emphasize prudent choices over those that are simply allowable, or in this case, those that are perceived to be allowable. Specifically, Entergys reliance on the design characteristics of the PCIVs to meet the TS requirement, while refraining to take additional measures to ensure the valves remained closed in the case of personnel error or equipment malfunction, was not conservative.
05000293/FIN-2014003-03	2014Q2	Pilgrim	Failure to Follow Licensed Operator Medical Requirements	(Open/Closed) NCV 05000293/2014003-003: NRC Letter, dated February 26, 2014 (ML1405A584), documented an NRC Office of Investigation review to determine whether a contract medical assistant deliberately failed to conduct required tactile and/or olfactory testing during annual physical examinations of three licensed operators on January 10, 2013 (NRC Investigation Report Number 1-2013-010). The NRC concluded that the medical assistants actions caused Entergy to violate NRC requirements in 10 CFR 55.27 and 10 CFR 55.9. This is being treated as a Severity Level IV NCV. In order to facilitate entering this issue into the NRCs Plant Issues Matrix and assessment process, this issue is identified as NCV 05000293/2014003-03, Failure to Follow Licensed Operator Medical Requirements
05000220/FIN-2014002-01	2014Q1	Nine Mile Point	Inadequate Design Control Measures Employed During Control Room HVAC Modification	The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control, because CENG did not implement adequate design controls to ensure piping in the Reactor Building Closed Loop Cooling (RBCLC) system remained operable while implementing a modification to the Unit 1 control room heating and ventilation system. Specifically, while implementing the modification, CENG personnel removed permanent plant supports and piping for the safetyrelated RBCLC system and did not fully assess how this change could impact the operability of the system with respect to a hydraulic shock or seismic acceleration event. In response to this observation, CENG initiated CR-2014-001676 and evaluated the condition for operability. Existing temporary supports were enhanced to provide additional margin by bracing the structure for horizontal loads. An extent of condition walkdown was performed and no additional issues of concern were identified. Subsequently, CENGs operability review determined the RBCLC system had remained operable. This finding was more than minor because it was 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, while implementing the modification, CENG removed permanent plant supports and piping for the safety-related RBCLC system and did not fully assess how this change could impact the operability of the system if a hydraulic shock or seismic acceleration occurred. This finding is also similar to examples 3.j and 4.k in IMC 0612, Appendix E, Examples of Minor Issues, where a temporary modification was installed without adequate design information and adequate design controls were not implemented leading to a reasonable doubt of operability of plant components. 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 determined this finding is of very low safety significance (Green) because the performance deficiency was a design or qualification deficiency that did not result in the inoperability of the RBCLC system. The finding has a cross-cutting aspect in the area of Human Performance, Work Management, because CENG failed to implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, CENG failed to ensure that the installed temporary supports were adequate to ensure the RBCLC piping would not be stressed above code allowable values in the event of a seismic acceleration or hydraulic shock event prior to removing the permanently installed seismic supports.
05000410/FIN-2014002-02	2014Q1	Nine Mile Point	Invalid Low Reactor Water Level Results in Unit 2 Automatic Reactor Scram	Inspectors documented a self-revealing Green NCV of Technical Specification (TS) 5.4, Procedures, for CENGs failure to ensure proper communication of a change in work scope prior to implementation. Specifically, on March 10, 2014, valve label replacements at Unit 2 commenced in a trip sensitive area while the plant was on-line, although the work was previously scheduled to be conducted when the reactor was shut down. This change in work scope was not properly reviewed and communicated to the supporting work group prior to implementation. As a result, a reactor scram occurred when an instrumentation and control (I&C) technician inadvertently contacted an instrument rack located in a trip sensitive area while performing a valve label replacement. CENG generated condition report (CR)-2014- 001963 to document the Unit 2 reactor scram due to the technician contacting the instrument line while cutting the valve label. Immediate corrective actions included developing site communications to enhance awareness of trip sensitive equipment and to provide additional flagging barriers to ensure trip sensitive components are not inadvertently contacted. This finding 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 those events that upset plant stability and challenge critical safety functions during power operations. Specifically, CENG staff did not properly ensure that the scope change was properly reviewed and communicated to the supporting work group prior to implementation. This resulted in work being performed while Unit 2 was online and a subsequent automatic reactor scram when an instrument rack was inadvertently contacted. 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, Conservative Bias, because CENG failed to use proper decision makingpractices that emphasize prudent choices over those that are simply all.
05000220/FIN-2013005-01	2013Q4	Nine Mile Point	Failure to Perform Surveillance Test for Unit 1 Smoke Removal Dampers	The inspectors identified a Green NCV of Unit 1 license condition DPR-63, Section 2.D(7), Fire Protection, because CENG staff did not perform visual inspections of fire dampers associated with Unit 1 between 2002 and 2013 in accordance with the Fire Protection Program and Updated Final Safety Analysis Report (UFSAR) Section 10A.2.4.1.10.1.A. As a result, CENG staff determined 25 dampers were non-functional due to the surveillance test not being performed. CENG staffs planned corrective actions include revising the UFSAR to state that performance-based testing requirements apply only to non-smoke removal dampers. Further, the 25 smoke removal dampers will remain nonfunctional until visual inspections can be performed as planned in work order (WO) C92482273. This issue was entered into CENGs CAP as CR-2013-009208. This finding is more than minor because it is associated with 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 Significance Determination Process for Findings At-Power, issued June 19, 2012, and the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency only represented a degradation of the smoke removal and radiological barrier function provided for the control room. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because CENG staff failed to identify smoke removal damper visual inspections were not being performed. Specifically, UFSAR section 10A.2.4.1.10.1.A, as part of license condition DPR-63 2.D(7) and the Fire Protection Program, requires CENG staff to perform visual inspections of smoke removal dampers, which was not being performed between 2002 and 2013, resulting in the control room envelope not being operable and 25 smoke removal dampers being declared non-functional. CENG performed an evaluation to determine if the control room habitability requirements contained in TS 3.4.5.f for the control envelope were met. CENG staff subsequently determined that Unit 1 control room habitability requirements of TS 3.4.5.f were met based on previous successful surveillance testing for control room operability testing under N1-ST-C9, Control Room Emergency Ventilation System Testing, Revision 01502.
05000410/FIN-2013005-02	2013Q4	Nine Mile Point	Failure to Implement Procedural Requirements for Evaluating Control Room Deficiencies as Operator Workarounds	The inspectors identified a Green finding (FIN) for CENG staffs failure to properly classify operator workarounds, operator burdens, or control room deficiencies in accordance with CNG-OP-1.01-2010, Operator Workaround/Challenge Control, Revision 0. Specifically, the failure to properly classify operator workarouonds resulted in an operator error when control room operators did not recognize a meter was degraded, used that meter during the performance of a surveillance test, and overexcited the Division II emergency diesel generator (EDG) on July 30, 2013. CENG staff entered this inspector identified issue into the corrective action program (CAP) as condition report (CR)-2013-009004. Corrective actions included reviewing, classifying, and adding the inspector identified operator burdens to each of the respective Units shift turnover checklist. 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. Specifically, the failure to properly classify the Unit 2 Division II EDG degraded volt amperes reactive (VAR) meter as an operator burden resulted in an operator using the degraded meter during a surveillance test and inadvertently overexciting the diesel generator for 1.5 hours. 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 determined that this finding is of very low safety significance (Green) because the performance deficiency was not a design or qualification deficiency, did not involve an actual loss of safety function, did not represent actual loss of a safety function of a single train for greater than its technical specification (TS) allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, in that CENG staff did not ensure control room deficiencies were evaluated properly in accordance with CNG-OP-1.01-2010. Specifically, CENG staff failed to classify the known degraded Unit 2 Division II EDG VARs meter as an operator burden; which resulted in the EDG being overloaded during a surveillance test.
05000220/FIN-2013005-03	2013Q4	Nine Mile Point	Inadequate DSC Welding Procedure to Control and Monitor Hydrogen Concentrations	A self-revealing Severity Level IV NCV of Title 10 of the Code of Federal Regulations (10 CFR) 72.150, Instructions, Procedures, and Drawings, was identified when CENG personnel did not ensure that hydrogen concentrations were being properly monitored and maintained during welding on dry shielded container (DSC) #12 on August 14, 2013. Specifically, site procedure S-MMP-ISFSI-004, DSC Sealing Operation, Revision 00201, provided inadequate direction for the control of purging and hydrogen monitoring calibration, set-up, and operation. This caused an undetected loss of DSC purge and a failure of the hydrogen monitor, ultimately resulting in a hydrogen deflagration in DSC #12. CENG staff generated CR-2013-006840 to address the hydrogen deflagration. Corrective actions included: (1) reducing water level in the DSC by 1100 gallons during welding operations to reduce the amount of hydrogen generation; (2) installed dual hydrogen monitors off the vent line to provide redundant indication; (3) required the performance of local hydrogen monitoring at the weld joint prior to commencing welding; (4) reconfigured the location of the hydrogen monitors; (5) ensured hydrogen monitors were properly configured, including the use of the low flow differential pressure switch setting in a helium environment; and (6) adjusted the alarm settings on the hydrogen monitors. The inspectors determined that CENG personnels failure to provide adequate instructions, procedures, and drawings to ensure that hydrogen concentrations were being properly monitored and maintained in accordance with 10 CFR 72.150, Instructions, Procedures, and Drawings, during welding of DSC #12 on August 14, 2013, was a performance deficiency that was reasonably within CENG staffs ability to foresee and correct, and should have been prevented. As a result, a hydrogen deflagration occurred. The failure to properly monitor and maintain hydrogen concentrations had the potential to damage the DSC and spent fuel within the DSC. Because the issue involved independent spent fuel storage installation (ISFSI) operations, consistent with the guidance in Section 2.2 of the NRC Enforcement Policy, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using Example 6.3.d. from the NRC Enforcement Policy, the inspectors determined that the violation was a Severity Level IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation. The hydrogen deflagration ultimately did not result in the damage to fuel; however, the failure to properly monitor and maintain hydrogen concentrations had the potential to damage the DSC and spent fuel within the DSC. Because the violation involved the traditional enforcement process and was not associated with ISFSI support programs conducted under a 10 CFR 50 license, the inspector did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Appendix B.

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