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	Quarter	Site	Title	Description
05000247/FIN-2018003-04	2018Q3	Indian Point	Inadequate Procedure for Turbine Startup Caused a Reactor Trip	A self-revealing Green NCV of TS 5.4.1, Procedures, was identified because Entergy did not provide adequate guidance in 2-SOP-26.4, Turbine Generator Startup, Synchronization, Voltage Control, and Shutdown. Specifically, Entergy did not provide adequate procedural direction to ensure the main turbine control oil stop valve Z was in the correct position. As a result, the steam generator water level exceeded the trip setpoint for the main boiler feed pumps which led the operators to insert a manual reactor trip.
05000247/FIN-2018003-03	2018Q3	Indian Point	Containment Fan Cooler 24 Through-Wall Service Water Leak Caused by Inadequate Application of Epoxy Coating Resulting in Corrosion and a Safety System Functional Failure of Containment	A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified when Entergy did not ensure that activities affecting quality were prescribed by documented instructions or procedures, of a type appropriate to the circumstances, and that these activities were accomplished in accordance with these instructions, procedures or drawings. Furthermore, Entergy did not ensure that the instructions or procedures included appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, Entergy did not ensure that the maintenance procedure for applying the internal EneconTM epoxy coating to the 24 fan cooler main cooler supply line elbow was adequate to ensure proper epoxy coating adherence, and Entergy did not adequately verify the coating adherence prior to placing the elbow in service. This resulted in a through-wall leak and a safety system functional failure of containment.
05000247/FIN-2018003-02	2018Q3	Indian Point	Containment Fan Coolers 21 and 24 Motor Cooler Elbow Through-Wall Leaks Due to Excessive Service Water Flow Rates and Safety System Functional Failures of Containment	A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified when Entergy did not ensure that measures were established for the selection and review for suitability of application of materials, parts, equipment, and processes that are essential to the safety-related functions of the structures, systems, and components. Specifically, in 1998, when the former license-holder for Unit 2 decided to replace the original-construction large-radius, butt-welded elbow joints in the service water motor cooler return lines from the Unit 2 FCUs with a new design (a short radius, socket-weld fitting), these elbow joints were not properly evaluated for suitability of application. The service water flow velocity through the modified FCU return piping was in excess of the vendor-allowable flow velocity limit, which resulted in the gradual erosion of the motor cooler elbow joints, eventually leading to through-wall leaks on an ASME class III piping system inside containment, leading to breaches of containment integrity and safety system functional failures.
05000247/FIN-2018003-01	2018Q3	Indian Point	Inadequate Procedural Guidance for Spent Fuel Movement and Storage Requirements	The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Procedures, when Entergy did not have appropriate documented instructions or written procedures for spent fuel movement and storage requirements adjacent to potentially degraded Boraflex panels. Specifically, configuration restrictions were not addressed in some cases and, therefore, did not comply with controls to meet the criticality analysis of record (CAOR) in 2016; and the resultant revised guidance did not accurately reflect the modeled supporting analysis
05000336/FIN-2018003-01	2018Q3	Millstone	Failure to Assure that Safety-Related Service Water Piping Conformed to the Procurement Documents	The inspectors identified a Green finding and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion VII, Control of Purchased Material, Equipment, and Services, when the licensee failed to identify that a replacement service water pipe spool (JGD-1-25) was not in conformance with the American National Standards Institute (ANSI) B31.1 code, a condition of the purchase order, and was installed in the plant.
05000336/FIN-2018011-01	2018Q3	Millstone	Reviews of Incoming Industry Operation Experience Not Completed	The inspectors identified that Millstone could not demonstrate that incoming industry operational experience reports (ICES) since 2015 had been properly reviewed for applicability to Millstone and for those items that were applicable, were evaluated and corrective actions developed as necessary as required by program guidance. A population of over 1600 ICES reports were identified where it could not be determined if required reviews were complete. Because there are parallel processes which may have reviewed these items, additional review is necessary to determine whether this issue represents a performance deficiency that is of more than minor significance. Therefore, this item is characterized as an unresolved item (URI). The purpose of the operational experience program is to identify conditions adverse to quality (CAQs) found at other plants, evaluate whether the concern is applicable to either Millstone unit, and evaluate and develop corrective actions for those CAQs when necessary. The inspectors noted that a performance improvement report (PIR) is automatically created for the Dominion fleet whenever an OPEX report is received (regardless of its source). Once the corporate PIR is generated, each site is required to check a box that it was received and also disposition it. The PIR remains opened until each site has completed this action. Prior to 2015, the corporate Operating Experience Coordinator would perform an applicability review and assign the remaining items to the site for further evaluation. When the corporate organization was reorganized, the headquarters review of OPEX became mostly administrative and the individual sites were expected to fully disposition the report. Since 2015, more than 1600 OPEX records were discovered that required disposition for Millstone. These records were still open and no records exist to show whether reviews were completed. Therefore it is uncertain if all applicable ICES reports were reviewed. Planned Closure Actions: The NRC will conduct a problem identification and resolution annual sample using NRC IP 71152 once Dominion has notified the NRC that they have completed their review of the 1600 ICES reports. Licensee Actions: Dominion wrote Condition Report (CR) 1105042 to capture the issue, conducted an investigation, and developed a plan to review the 1600 ICES reports which have no documented reviews. Dominion anticipates this review will be completed by the end of the first quarter of 2019.Corrective Action Reference: CR 1105042NRC Tracking Number: 05000336 & 05000423/2018-011-01
05000286/FIN-2018002-01	2018Q2	Indian Point	Reactor Pressure Boundary Leakage Due to Weld Failure in Reactor Vessel Head Penetration #3	A self-revealing Severity Level IV NCV of Technical Specification (TS) 3.4.13.a, Reactor Coolant System Operational Leakage, was identified when Entergy operated the reactor in Mode 1 with pressure boundary leakage for a period of time longer than the allowable limiting condition of operation. Specifically, a leak in the J-weld around reactor pressure vessel (RPV) head penetration #3 occurred during the last operating cycle and was not identified until after the reactor was shutdown for a refueling outage.
05000286/FIN-2018001-02	2018Q1	Indian Point	Inadequate Procedure for Placing Chemical and Volume Control System Demineralizer In Service	A self-revealing Green NCV of Technical Specification 5.4.1, Procedures, was identified because Entergy failed to provide adequate guidance in 3-SOP-CVCS-004, Placing the CVCS Demineralizers In or Out of Service. Specifically, Entergy did not provide adequate procedural direction to prevent exceeding the reactor coolant filter differential pressure while placing the demineralizers in service. As a result, the pressurizer water level technical specification limit was exceeded and the CVCS piping upstream of the filter was over-pressurized resulting in diaphram ruptures on valves CH-305 and CH-352 thereby spreading contamination throughout the Primary Auxiliary Building.
05000247/FIN-2018001-01	2018Q1	Indian Point	Failure to Incorporate Adequate Test Controls for Quarterly Stroke Close Testing of the Steam Supply Valves to Turbine-Driven Auxiliary Feedwater Pump	The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when Entergy did not assure that surveillance tests required to demonstrate that structures, systems, and components will perform satisfactorily in service are identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, during quarterly stroke testing of the steam isolation valves to the 22 turbine-driven auxiliary feedwater pump, PCV-1310A and PCV-1310B, Entergy did not ensure that these valves traveled to the closed position as required to verify that the safety function was met.
05000387/FIN-2017003-02	2017Q3	Susquehanna	RBCCW PCIV Design Control Issue	The inspectors identified a finding of very low safety significance (Green), an associated NCV of 10CFR50 Appendix B, Criterion III, Design Control, and a resultant violation of technical specification (TS) 3.6.1.3, Primary Containment Isolation Valves (PCIVs), when the reactor building closed cooling water (RBCCW) outboard isolation supply valve, HV21314, was found with a pull apart terminal block unseated within the motor control center (MCC), resulting in the loss of function for the valve to close given an initiation signal. Based on questions from inspectors, it was discovered that the terminal block was not installed in accordance w ith its dynamic qualification report. Immediate corrective actions included correctly seating the terminal block and performing an engineering evaluation to validate that the configuration conformed to the dynamic qualification report. The finding was more than minor because it was associated with the design control attribute of the reactor safety barrier integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the RBCCW outboard PCIV was inoperable for more than four years. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 3 of IMC 0609, Appendix A, The SDP for Findings At -Power, dated June 19, 2012, inspectors determined the significance to be of very low safety significance (Green) since the finding did not represent an actual open pathway in the containment isolation system and was not associated with hydrogen recombiners. The finding had a cross -cutting aspect in the area of Human Performance, Documentation, because Susquehanna did not maintain complete, accurate, and up -to-date documentation. Specifically, Susquehanna was not able to make a clear determination of the acceptability of the as -left configuration of the terminal block until the issue was discussed with the vendor to determine that the configuration was not in accordance with the dynamic qualification of the 480VAC MCC buckets. (H.7)
05000352/FIN-2017003-02	2017Q3	Limerick	Licensee-Identified Violation	LGS Unit 1 Renewed Facility Operating License, NPF- 39, and LGS Unit 2 Renewed Facility Operating License, NPF- 85, License Condition 2.C.(3) requires , in part, that Exelon Generation Company shall implement and maintain all provisions of the approved Fire Protection Program as described in the UFSAR. LGS Unit 1 and Unit 2 UFSAR Chapter 9A requires compliance with Branch Technical Position, Chemical Engineering Branch 9.5- 1, guideline C.5.b(1), to limit fire damage so that one train of systems necessary to achieve and maintain cold shutdown conditions from either the control room or emergency control station can be repaired within 72 hours. Contrary to the above, from July 2014 to December 2016, an unanalyzed condition existed in which an abnormal ESW system alignment placed two Fire Areas in noncompliance with the FSSD analysis described in the UFSAR. Specifically, in July 2014, ESW to RHRSW flow return valve, HV -011 -015A was de- energized and tagged closed following ESW system testing. With on ly one RHRSW return path available to the A ESW loop, a postulated fire in Fire Area 12 or Fire Area 18 could cause a single spurious valve operation of either spray pond bypass valves HV -012- 031A or HV -012 -031C, when the ESW system is aligned in the spray pond winter bypass mode. This condition would result in no return flow path for the A loop of ESW, which would in turn result in loss of cooling water to EDGs aligned to the A ESW cooling loop. The affected EDGs would be inoperable until the ESW system could be realigned to provide cooling water flow. This condition coupled with a loss of offsite power assumed in FSSD analysis would result in a loss of power to SRVs needed to transition both LGS units from hot shutdown conditions to cold shutdown conditions. Following the depletion of station batteries after 4 hours, until offsite power is assumed to be restored after 72 hours, direct current power would be lost to SRVs that are necessary to reduce plant pressure low enough to place the shutdown cooling system into service and establish cold shutdown plant temperatures. The failure to have a cold shutdown repair that could be implemented within 72 hours in accordance with the FSSD analysis described in the UFSAR, was a performance deficiency. 24 The performance deficiency was more than minor because it was associated with the protection against external factors (fire) attribute of the mitigating systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green ), based on IMC 0609, Appendix F, Fire Protection Significance Determination Process , Attachment 1, Part 1: Fire Protection Significance Determination Process Phase 1 Worksheet, dated September 2013. The finding screened to Green based upon task 1.3.1 screening question A, since the inspectors determined that for conditions evaluated by Appendix F the reactors were able to reach and maintain hot shutdown. Specifically, LGS Units 1 and 2 would have been able to achieve and maintain hot shutdown during the period the unanalyzed condition existed. This would have been accomplished by using HPCI and SRVs for pressure and level control. Both units would have been capable of maintaining hot shutdown conditions with postulated fire damage until offsite power could be restored. Because this issue was of very low safety significance (Green) and Exelon entered the issue into the corrective action program as IR 3955705, this finding is being treated as a licensee identified NCV , consistent with Section 2.3.2.a of the Enforcement Policy.
05000277/FIN-2017003-02	2017Q3	Peach Bottom	Licensee-Identified Violation	10 CFR 55.25 states, in part, that if an operator develops a permanent physical or mental condition that causes the operator to fail to meet the requirements of 10 CFR 55.21, the facility licensee shall notify the Commission within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c),which states,that the regional administrator shall be notified if a licensed operator develops a permanent disability or illness. Contrary to these requirements, as the result of Exelons medical examination audit completed September 26, 2017, Exelon identified a change in a licensed operators medical condition that was not communicated to the NRC within the required 30 days. The results of the medical examination audit were documented in IR 4054146 and subsequent notifications were made to the NRC.This violation is subject to traditional enforcement because of the potential impact upon the regulatory process for issuing restrictions to operators licenses. The inspectors determined that this issue meets the criteria for a Severity Level IV violation using example 6.4.d.1(a) from the NRC Enforcement Policy because no incorrect regulatory decision was made as the result of the failure of the licensee to report within 30 days. This is of very low safety significance because after NRC review of the subsequent notifications, no changes to license restrictions were required.
05000387/FIN-2017003-01	2017Q3	Susquehanna	Failure to Prepare Work Packages with Necessary Detail Results in Automatic Reactor Scram	The inspectors identified a self -revealing finding of very low safety significance (Green) because Susquehanna did not ensure that a work package was prepared to the detail necessary based on task difficulty in accordance with administrative procedure, NDAP- QA -0502, Revision 51. Specifically, on June 8, 2017, maintenance workers inadvertently shorted the Unit 1 main electro -hydraulic control (EHC) logic power supply to ground while working in a cabinet with little space to manipulate tools, resulting in a reactor scram. 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, Susquehanna did not ensure measures were in place to prevent an adverse impact on the EHC control system during power supply voltage adjustment. This resulted in a rapid rise in reactor pressure and neutron flux, and subsequent automatic reactor scram. In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP 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 the station failed to recognize and plan for the possibility of mistakes and inherent risk even while expecting successful outcomes. Specifically, individuals at various organizational levels did not ensure measures were in place to prepare maintenance technicians to perform a task on the EHC system that involved manipulating tools in a small space with tight clearances.
05000353/FIN-2017003-01	2017Q3	Limerick	Operational Condition Mode Change from Startup to Run was Made with RCIC Inoperable	The inspectors identified a Green NCV of Unit 2 technical specification (TS) 3.0.4, when Exelon changed the operating condition of Unit 2 from mode 2 (startup) to mode 1 (run) with reactor core isolation cooling ( RCIC ) inoperable for surveillance testing. Specifically, the TS 3.7.3 limiting condition for operation (LCO) for RCIC was not met, a mode change from startup to run was made, and none of the allowances, TS 3.0.4.a, TS 3.0.4.b, or TS 3.0.4.c, were met to allow the mode change in that condition. Exelon entered this issue into the corrective action program with issue report (IR) 4057128. The inspectors determined that the change in operating condition of LGS Unit 2 from startup to run with RCIC inoperable was reasonably within Exelons ability to foresee and correct and should have been prevented and therefore was a performance deficiency. This finding is more than minor because it adversely affected the configuration control attribute of the mitigating systems cornerstone to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, RCIC was inoperable during the time it was required to be operable, i.e. the mode change from startup to run. Additionally, this finding was similar to example 2.g of IMC 0612, Appendix E, in that a mode change was made without all required equipment being operable. Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the finding did not represent a loss of function and did not represent the loss of a single train for greater than technical specification allowed outage times or greater than 24 hours. The inspectors determined that this finding has a cross - cutting aspect in the area of Human Performance, Documentation, because with respect to TS 3/4.7.3 Exelon did not create and maintain complete and accurate documentation of the correct usage of TS 3.0.4 that was more fully explained in the applicable safety evaluation. (H.7)
05000278/FIN-2017003-01	2017Q3	Peach Bottom	Instructions Not Followed for Replacement of HPSW Ventilation Switch Block	A self-revealing NCV of Technical Specification (TS) 5.4.1, Procedures,of very low safety significance (Green) was identified for Exelonnot implementing procedural instructions for the replacement of the HS-3-40H-3AV060 switch block associated with the 3AV060 high pressure service water (HPSW) ventilation fan. Exelon did not ensure that electrical connections were free of loose wire strands per their procedural standard E-1317,Wire and Cable Notes and Details, Power, Control, and Instrumentation, Revision 55, and from the vendor manual instructions. As a result,on July 10, 2017, the 3AV060 HPSW ventilation fan failed its surveillance test(ST)and rendered one subsystem of Unit 3 HPSW inoperable. Exelon entered this issue into their corrective action program (CAP) asissue reports(IR)4030367 and 4044444, straightened out the remaining loose strands, and specified additional electrical panels for an extent of condition (EOC) review.Thisfinding ismore than minor because it isassociated with the equipment performance attribute of the Mitigating Systemscornerstoneand affected the cornerstones objective to ensure the reliability, availability, and capability of systems to respond to initiating events to prevent undesirable consequences (i.e. core damage).By not implementing theE-1317 procedural instructions, the 3AV060 fan failed and affected the reliability of one HPSW subsystem.The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, SDP for Findings At-Power and determined the finding was of very low safety significance (Green) because it did notrepresent a loss of system function or represent an actual loss of function of at least a single train for longer than itsTSallowed outage time. The inspectors determined no cross-cutting aspect applied because the PD occurred in 2010 and was not indicative of current performance.
05000387/FIN-2017002-01	2017Q2	Susquehanna	Inadequate Assessment of Fire Brigade Performance during an Unannounced Drill	The inspectors identified a Green NCV of Susquehanna Unit 1 and 2 Operating License Condition 2.C.6, Fire Protection, because Susquehanna did not adequately assess an unannounced fire brigade drill, as required by the fire protection program. Susquehannaentered this issue into the corrective action program (CAP) for resolution as condition report(CR) CR-2017-10767 and is conducting an apparent cause evaluation to determine the most appropriate corrective actions.The performance deficiency (PD) was more than minor since the deficiency was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and impacted its objective of ensuring 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 (Green) in accordance with D.1 of IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. Because the finding involved fire brigade training requirements, the fire brigade demonstrated the ability to meet the required times for fire extinguishment for the fire drill scenarios, and the finding did not significantly affect the fire brigades ability to respond to a fire, the finding screened as Green. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Self and Independent Assessments, because Susquehanna did not conduct assessments of their activities to assess performance and identify areas of improvement. Specifically, the Susquehanna self-evaluation of fire brigade performance was not of sufficient depth, appropriately objective, or self-critical. (P.6)
05000388/FIN-2017002-02	2017Q2	Susquehanna	Failure to Assess and Manage Risk Associated with Emergent Work	The inspectors identified a Green, self-revealing, NCV of 10 Code of Federal Regulations (CFR) 50.65 (a)(4) because Susquehanna failed to assess and manage the increase in risk for emergent work on the Unit 1 A 125 voltage direct current (VDC) battery charger. Susquehanna entered this issue into the CAP as CR-2017-09589. Corrective actions include conducting training on the emergent risk assessment process and reinforcing the expectation that control room staff is notified prior to releasing work. The PD was more than minor because it adversely impacted the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and the related attribute of equipment performance involving availability and reliability. In addition, it is similar to Example 7.e from IMC 0612, Appendix E, Examples of Minor Issues, which states that the failure to perform an adequate risk assessment when required to do so is more than minor if the overall elevated plant risk would put the plant into a high licensee-established risk category and would require risk management actions under licensee procedures. The inspectors evaluated the significance using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management SDP and determined that this PD was of very low safety significance (Green). Specifically the PD was associated with risk management actions only and the incremental core damage probability (ICDP) was 2E-7 (<1E-6) for charger 1D613 out of service for approximately one hour.This finding had a cross-cutting aspect in the area of Human Performance, Consistent Process because individuals did not implement systematic approach to make decisions to commence work, and did not incorporate appropriate risk insights. (H.13)
05000278/FIN-2017002-01	2017Q2	Peach Bottom	Corrective Action Not Implemented Correctly for Replacement of RCIC RCR Contacts	A self-revealing non-cited violation (NCV) of 10 Code of Federal Regulation(CFR)Part 50, Appendix B, Criterion XVI, Corrective Actions, of very low safety significance (Green) was identified for Exelon not correcting a condition adverse to quality concerning reverse control relay (RCR) contacts for valves associated with the reactor core isolation cooling (RCIC) system. Specifically, Exelon specified a corrective action (CA) from an October 18, 2013, Unit 3 RCIC equipment apparent cause evaluation (EACE) to replace RCR contacts after 12 years of service, however, the CA was not correctly implemented. As a result, on January 12, 2017, an RCR contact associated with the Unit 3 RCIC suppression pool suction valve remained in service for 15 years, exhibited a high resistance failure during a surveillance which resulted in Unit 3 RCIC being inoperable. Following the failure, Exelon initiated issue reports (IRs) 03962563 and 03977949, implemented corrective actions to replace the RCR contact, restored Unit 3 RCIC operability, and risk-informed their corrective maintenance schedule for replacing all RCR contacts that currently exceeded the recommended 12-year service life.Exelons failure to recognize and correct a condition adverse to quality associated with certain RCR contacts in their Unit 3 RCIC system that had exceeded their 12-year service life, was a performance deficiency (PD) that was within their ability to foresee and correct and should have been prevented. The finding was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstones objective to ensure the reliability of systems to respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, not recognizing that existing RCR contacts were installed in safety-related equipment beyond their 12-year service life, resulted in the failure of the Unit 3 RCIC suppression pool suction valve. The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, SDP for Findings At-Power, and determined the finding was of very low safety significance (Green) because it did not represent a loss of system function or represent an actual loss of function of at least a single train for longer than its technical specification (TS) allowed outage time of 14 days. The inspectors determined that the finding has a cross-cutting aspect in Human Performance, Procedure Adherence, because Exelon did not validate that the correct revision of procedure WC-AA-120, Attachment 2, Preventive Maintenance (PM) Change Review Form, was used when creating a new PM to replace RCR contacts. (H.8)
05000353/FIN-2017002-01	2017Q2	Limerick	Inadequate Design Control of the Drywell Unit Cooler Condensate Flow Rate Monitoring System	Green . A self -revealing Green NCV of 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control, occurred when Exelon failed to verify or check the adequacy of design of a new Unit 2 drywell unit cooler condensate flow rate monitoring system. Specifically, the design did not identify that the low conductivity of the drain fluid affected the ability of the flow elements to accurately detect drain flow. In addition to this, LGS staff did not assure adequate post modification acceptance test ing in accordance with CC- AA- 107- 1001, Post Modification A cceptance Testing. This inadequately designed and tested modification also resulted in a violation of technical specification (TS) 3.4.3.1, Leakage Detection Systems , because the system was inoperable and unavailable to perform its function following t he Unit 2 April 2015 refueling outage, and the TS 3.4.3.1 action statement was not met until the system was decl ared inoperable on December10, 2015. In response to this issue, Exelon initiated a condition report, IR 2598308, performed an apparent cause investigation, and replaced the Rosemount drywell unit cooler condensate flow rate monitoring system with a modified ver sion of the previously used system. The inspectors determined that the failure to verify the adequacy of the newly installed Rosemount dr ywell unit cooler condensate flow rate monitoring was within Exelons ability to foresee and correct and should have been prevented and therefore w as a performance deficiency . This issue is more than minor because it adversely affected the design control attribute of the barrier integrity cornerstone to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the Unit 2 drywell unit cooler condensate flow rate monitoring system was inoperable and unavailable to perform its function as part of the reactor coolant leakage detection system following the Unit 2 April 2015 refueling outage . This issue was evaluated in accordance with IMC 0609, Appendix A, "Significance Determination Process for Findings At-Power, using Exhibit 3, Barrier Integrity Screening Questions, Section B, Reactor Containment . The finding was determined to be of very low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of the reactor containment and did not involve an actual reduction in function of hydrogen ig niters in the reactor containment. The inspectors determined that this finding has a cross -cutting aspect in the area of Human Performance, Conservative Bias , because LGS staff ma de inappropriate decisions based on informal vendor input and a successful implementation of the modification at another facility . (H.1 4)
05000352/FIN-2017003-00	2017Q2	Limerick	Licensee-Identified Violation	LER 05000352/2017- 003 -00 Condition Prohibited by Technical Specifications Due to an Inoperable Rod Position Indication System . TS 3.1.3.7 requires, in part, with one or more control rod position indicators inoperable, within 1 hour, determine the position of the control rod by using an alternate method, or otherwise, be in at least hot shutdown within the next 12 hours. Contrary to the above, on March 16, 2017, a power supply for the Unit 1 rod position indication system rendered position indication for 83 control rods inoperable for approximately 19.5 hours until the power supply was replaced. Exelon incorrectly used the full core display to verify control rod position for 81 of the 83 rods. The power supply failure rendered the full core display incapable of updating in response to a rod position change and was, therefore, not a valid means to determine rod position. Exelon initiated condition report IR 3988302 to document the TS violation. The inspectors evaluated the significance of this findi ng using IMC 0609 Appendix A , Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the issue did not affect a single reactor protection system trip signal or the function of the ot her redundant trips or diverse methods of reactor shutdown, did not involve addition of positive reactivity, and did not result in mismanagement of reactivity by operators. Because this issue was of very low safety significance (Green) and Exelon entered the issue into the corrective action program (IR 3988302), this finding is being treated as a non- cited violation, consistent with Section 2.3.2 .a of the NRC Enforcement Policy.
05000277/FIN-2017002-02	2017Q2	Peach Bottom	EDG Exhaust Stacks Nonconforming Design for Tornado Missile Protection	On January 9, 2017, it was determined that PB's EDGs do not conform with the licensing basis for protection against tornado-generated missiles. The exhaust stacks for the four on-site EDGs extend approximately seven feet above the roof of the EDG building. In the event of a tornado, debris generated from the tornado could strike the exhaust stacks and, if at a sufficient mass and velocity, could crimp the exhaust stacks in a manner that would affect EDG operation.As a result of the non-conforming condition, on January 9, 2017, at 1530, all four EDGswere declared inoperable. Compensatory measures were put in place and, in accordance with NRC guidance contained in Enforcement Guidance Memorandum (EGM) 15-002, the EDGs were returned to an operable but non-conforming status.There are no actual consequences as a result of the non-conforming condition. This LER is closed.b. FindingsDescription. 10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that measures shall be established to assure that the applicable regulatory requirements and the design basis for SSCs are correctly translated into specifications, drawing, procedures, and instructions. Contrary to the above, Exelon failed to correctly translate the design basis for protection against tornado-generated missiles into their specifications and procedures. Specifically, Exelon did not adequately protect Unit 2 and Unit 3s EDG exhaust stacks from tornado-generated missiles.Exelon documented the condition adverse to quality in their CAP under IR 3961028 and took immediate compensatory actions. The inspectors evaluated Exelons immediate compensatory measures, which included verifying that procedures are in place, equipment was appropriately staged, and training is current for performing actions in response to a tornado to preserve EDG operability. Enforcement. Because this violation was identified during the discretion period covered by EGM 15-002, Revision 1, Enforcement Discretion for Tornado Generated Missile Protection Non-Compliance, (ML16355A286) and because Exelon has implemented compensatory measures, the NRC is exercising enforcement discretion, is not issuing enforcement action, and is allowing continued reactor operation.
05000352/FIN-2017002-02	2017Q2	Limerick	Follow -Up of Events and Notices of Enforcement Discretion	Inspection Scope On March 20, 2016, Limerick Unit 1 was performing a planned shutdown to support a refueling outage. The drywell leak inspection team identified a 0.5 gallons per minute reactor coolant system (RCS) pressure boundary leak on the shutdown cooling equalizing line. The apparent cause evaluation determined that the 34 inch A RHR shutdown cooling return check valve equalizing line developed a crack at the toe of the weld due to high cyclic fatigue induced by vibration from the reactor recirculation system. This check valve was previously replaced in 2006, and the equalizing line came pre - fabricated to the valve body. The affected section of the piping was replaced with a new socket weld with a 2x1 overlay to improve the pipe stability and minimize stresses. The Unit 1 B RHR shutdown cooling return check valve equalizing line weld was also reworked using the 2x1 weld method during the Unit 1 refueling out age in April 2016. The similar Unit 2 welds on the equalizing lines were examined and reinforced during the May 2017 refueling outage. The LER and associated evaluations and follow -up actions were reviewed for accuracy, the appropriateness of corrective actions, violations of requirements, and potential generic issues. This LER is closed. b. Findings Description. On March 20, 2016, Limerick Unit 1 was performing a planned shutdown to support a refueling outage. The drywell leak inspection team identified a 0.5 gallons per minute RCS pressure boundary leak on the shutdown cooling equalizing line. Additionally, Exelon determined that this leakage constituted a violation of the Unit 1, TS 3.4.3.2. Operational Leakage that requires the RCS leakage to be limited to no pressure boundary leakage. The condition was reported in event notification 51809 as required by 10 CFR 50.72(b)(3)(ii)(A ) because it represented a degradation of a principal safety barrier. Exelon evaluated the flaw and determined the cause of the RCS pressure boundary leakage was that the 34 inch A RHR shutdown cooling return check valve equalizing line developed a crack at the toe of the weld due to high cyclic fatigue induced by vibration from the reactor recirculation system. The inspectors reviewed the LER and Exelons apparent cause evaluation of the event. The inspectors reviewed the event information and leakage data over the previous cycle and concluded that reactor pressure boundary leakage reasonably began on an unknown date that was more than 36 hours before March 20, 2016. However, the inspectors determined that the existence of R CS pressure boundary leakage was not within Exelons ability to foresee and correct and therefore was not a performance deficiency. In particular, the RHR shutdown cooling return check valve was replaced on the recommended periodicity, and the equalizing line that developed the crack came pre- fabricated to the valve body when replaced in 2006. For information, the inspectors screened the significance of the condition using IMC 0609, Appendix A, The Significance Determination Process For Findings At -Power , and determined that the condition represented very low safety significance (Green) because it would not result in exceeding the RCS leak rate for a small LOCA and would not have likely affected other systems used to mitigate a LOCA. 19 Enforcement. TS 3.4.3.2 requires, in part, that RCS operational leakage shall be limited to no pressure boundary leakage. If pressure boundary leakage exists, the TS 3.4.3.2 limiting condition for operation action statement requires Unit 1 to be in at least hot shutdown within 12 hours and in cold shutdown within the next 24 hours. Contrary to the above, for a period that began on an unknown date that was very likely more than 36 hours before March 20, 2016, and ending on March 20, 2016, RCS pressure boundary leakage existed, and Exelon did not place Unit 1 in at least hot shutdown within 12 hours and in cold shutdown within the next 24 hours. This issue is considered within the traditional enforcement process because there was no performance deficiency associated with the violation of NRC requirements. Inspection Manual Chapter 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- 076) 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 NRC determined that the issue was not within 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.
05000387/FIN-2017002-03	2017Q2	Susquehanna	Follow -Up of Events and Notices of Enforcement Discretion	Inspection Scope For the plant event listed below, the inspectors reviewed and/or observed plant parameters, reviewed personnel performance, and evaluated performance of mitigating systems. The inspectors communicated the plant events to appropriate regional personnel, and compared the event details with criteria contained in IMC 0309, Reactive Inspection Decision Basis for Reactors, for consideration of potential reactive inspection activities. As applicable, the inspectors verified that Susquehanna made appropriate emergency classification assessments and properly reported the event in accordance with 10 CFR Parts 50.72. The inspectors reviewed Susquehannas follow - up actions related to the events to assure that Susquehanna implemented appropriate corrective actions commensurate with their safety significance. Unit 1, reactor scram due to transient initiated by an inadvertent loss of main turbine electrohydraulic control system control power due to a maintenance error . b. Findings No findings were identified.
05000352/FIN-2017002-03	2017Q2	Limerick	Licensee-Identified Violation	LER 05000352/2017-003-00 Condition Prohibited by Technical Specifications Due to an Inoperable Rod Position Indication System. TS 3.1.3.7 requires, in part, with one or more control rod position indicators inoperable, within 1 hour, determine the position of the control rod by using an alternate method, or otherwise, be in at least hot shutdown within the next 12 hours. Contrary to the above, on March 16, 2017, a power supply for the Unit 1 rod position indication system rendered position indication for 83 control rods inoperable for approximately 19.5 hours until the power supply was replaced. Exelon incorrectly used the full core display to verify control rod position for 81 of the 83 rods. The power supply failure rendered the full core display incapable of updating in response to a rod position change and was, therefore, not a valid means to determine rod position. Exelon initiated condition report IR 3988302 to document the TS violation. The inspectors evaluated the significance of this finding using IMC 0609 Appendix A, Significance Determination Process for Findings at Power, Exhibit 2, Mitigating Systems Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the issue did not affect a single reactor protection system trip signal or the function of the other redundant trips or diverse methods of reactor shutdown, did not involve addition of positive reactivity, and did not result in mismanagement of reactivity by operators. Because this issue was of very low safety significance (Green) and Exelon entered the issue into the corrective action program (IR 3988302), this finding is being treated as a non-cited violation, consistent with Section 2.3.2.a of the NRC Enforcement Policy.
05000278/FIN-2017008-02	2017Q1	Peach Bottom	Untimely Corrective Actions to Address Elevated Primary Containment Isolation Valve Leakage	Green. The inspectors identified a self-revealing non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, because Exelon did not promptly implement corrective actions to address a condition adverse to quality on two containment isolation valves. Specifically, drywell air sampling valves SV-3-7D-3671A and SV-3-7D-3671D failed to perform their primary containment isolation function on March 15 and September 26, 2016, respectively, as a result of untimely corrective actions to address elevated leakage. The valve internals were repaired, declared operable, and the issue was entered into the corrective action program (IR 3990490). The finding was more than minor, because it was associated with the barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstones objective to provide reasonable assurance that the containment design barrier protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined this finding was of very low safety significance, because the finding did not result in an actual open pathway in the physical integrity of the reactor containment or involve an actual reduction in the function of hydrogen igniters in the reactor containment. The inspectors determined this finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Exelon did not perform effective corrective actions in a timely manner commensurate with the safety significance of the issue. Specifically, corrective actions to address a CAQ on SV-3-7D-3671A and SV-3-7D-3671D were delayed which resulted in the valves failing their LLRT and being declared inoperable. (P.3)
05000277/FIN-2017008-01	2017Q1	Peach Bottom	Untimely Corrective Actions to Address 2C Core Spray Motor Elevated Vibrations	Green. The inspectors identified a non-cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Action, because Exelon did not implement corrective actions in a timely manner to correct a condition adverse to quality on the 2C core spray motor. Specifically, Exelon did not perform appropriate corrective actions to evaluate and address an increasing motor bearing vibration trend that had existed for over ten years. Consequently, motor vibration reached the fault level established in Exelons vibration analysis procedure. The finding was more than minor, because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely impacted the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined this finding was of very low safety significance because the performance deficiency did not impact the design or qualification of the component, did not result in a loss of system function, did not result in the loss of function of a train greater than its Tech Spec allowed outage time, and did not represent an actual loss of function for a high safety significant component in accordance with Exelons maintenance rule program. The inspectors determined the finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Resolution, because Exelon did not take effective corrective actions in a timely manner commensurate with the safety significance of the issue. Specifically, corrective actions to address the elevated vibrations on the 2C core spray motor were not implemented before motor vibration reached the fault level and adversely impacted the long-term reliability of the motor. (P.3)
05000352/FIN-2017001-02	2017Q1	Limerick	Failure to Implement Human Performance Tools Results in Draining of Emergency Diesel Generator Jacket Water System	Green. The inspectors identified a Green self-revealing finding for the failure of Exelon personnel to follow procedures related to human performance tools which resulted in the inadvertent opening of a valve on the D13 emergency diesel generator (EDG). Specifically, Exelon personnel did not correctly identify and maintain a distance barrier from the diesel generator jacket water drain valve during a maintenance activity which resulted in the draining of the jacket water system and unplanned inoperability and unavailability of the D13 EDG. Exelon refilled the jacket water system, restored D13 EDG to an operable condition, and entered the issue into the corrective action program as IR 3986305. This finding is more than minor because it adversely affected the configuration control attribute of the mitigating systems cornerstone to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the valve mispositioning caused the D13 EDG to be inoperable and unavailable. Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the finding did not represent a loss of system or function and did not represent the loss of a single train for greater than technical specification allowed outage times or greater than 24 hours. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because Exelon personnel did not properly implement error reduction tools. (H.12)
05000353/FIN-2017001-01	2017Q1	Limerick	Inadequate Work Instructions for Staging of Equipment and Routing of Temporary Power Cables	Green. The inspectors identified a Green NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for Exelons failure to establish instructions appropriate to the circumstances to properly stage equipment and route temporary power cables. Specifically, during cell replacement of the Class 1E 2A2 125/250 volts direct current (Vdc) safeguards battery, a portable battery charger was staged adjacent to operable 2A1 battery cells and not restrained to prevent potential tipping and shorting of exposed battery cell terminals and a non-safety related extension cord was routed in near contact with exposed safety related cables in an open cable tray. Exelon moved the portable battery charger, removed and rerouted extension cords, and entered the issues into the corrective action program as issue report (IR) 3980217; IR 3980203; and IR 3983203. This finding is more than minor because it adversely affected the configuration control attribute of the mitigating systems cornerstone to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the portable battery charger was adjacent to the 2A1 battery rack and oriented such that it was susceptible to tipping over and causing electrical shorting, and a non-safety related temporary power cable connected to a non-safety related power source was routed in near contact with safety related cables in an open cable tray which introduced a potential to damage and disable safety related equipment. Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the finding did not represent a loss of system or function and did not represent the loss of a single train for greater than technical specification allowed outage times or greater than 24 hours. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Training, because Exelon did not provide sufficient training to maintain a knowledgeable workforce and instill nuclear safety values associated with the staging of material and equipment. (H.9)
05000387/FIN-2017001-01	2017Q1	Susquehanna	Human Performance Error Results in Loss of Safety Secondary Containment Function	Green. A self-revealing finding of very low safety significance (Green) and associated NCV of TS 5.4.1, Procedures, was identified for failure to implement procedures that resulted in a secondary containment fan trip and associated loss of safety function. Susquehannas immediate corrective actions included restoring the secondary containment system to an operable configuration, and entering the issue into their corrective action program (CAP). Inspectors determined that the finding was more than minor because it was associated with the Human Performance attribute (Routine OPS/Maintenance Performance) of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (Secondary Containment) protect the public from radionuclide releases caused by accidents or events. The failure to adequately implement procedures for operation and maintenance of the secondary containment resulted in the inoperability of Zone 3 secondary containment and an associated loss of safety function. In accordance with IMC 0609.04, Initial Characterization of Findings, dated October 7, 2016, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency only impacted the radiological barrier function of secondary containment. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety was maintained. Specifically, personnel did not conduct a re-brief of the team after the plan deviated from what was originally briefed, and the team did not adequately respond to challenges from workers in the field about whether it was appropriate to commence load center restoration with work still in progress. (H.4)
05000352/FIN-2016004-01	2016Q4	Limerick	Failure to Demonstrate Effective Preventive Maintenance Under 50.65(a)(2) for the Instrument Air System	Green. The inspectors identified a Green NCV of 10 Code of Federal Regulations (CFR) 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants," because Exelon did not demonstrate that the performance of the Unit 1 instrument air system had been effectively controlled through the performance of appropriate preventive maintenance and did not monitor against licensee-established goals in accordance with 10 CFR 50.65(a)(1). Specifically, the inspectors identified that the instrument air system reliability performance monitoring did not properly account for instrument air compressor failures such that the system exceeded the performance criteria established by Exelons procedures. Exelon entered the issue into the corrective action program (CAP) as IR 3961244. This issue is more than minor because it adversely affected the equipment performance attribute of the mitigating systems cornerstone to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the instrument air system reliability performance monitoring did not accurately account for multiple functional failures that resulted in the system exceeding the performance criteria established by Exelons procedures. Additionally, this finding was similar to example 7.d of IMC 0612, Appendix E, in that appropriate preventive maintenance under 10 CFR 50.65 (a)(2) was not demonstrated. Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the finding did not represent a loss of system or function and did not represent the loss of a single train for greater than technical specification allowed outage times or greater than 24 hours. The inspectors determined that the finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelons staff did not adequately implement the procedures for reliability performance criteria evaluation. Specifically, Exelon did not verify that the established performance criteria for train reliability accurately monitored the scope of the function and demonstrated the effectiveness of maintenance when performing functional failure determinations and the periodic 10 CFR 50.65(a)(3) assessment. (H.8)
05000352/FIN-2016004-02	2016Q4	Limerick	Control Structure Chiller Unit Trip Caused by Failure to Properly Implement Procedures	Green. A self-revealing Green NCV of LGS Units 1 and 2 technical specification 6.8.1 was identified when Exelon did not properly implement a surveillance procedure. Specifically, operators secured cooling water to the operating A control structure chilled water system (CSCWS) chiller unit which resulted in the unit automatically tripping to prevent damage. Operators subsequently restored cooling water flow in accordance with procedures. Exelon entered the issue into the corrective action program as IR 2720374. This finding is more than minor because it is associated with the human performance attribute of the mitigating systems cornerstone to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the loss of cooling water to the A CSCWS chiller unit resulted in a trip of the unit on high condenser pressure and rendered the chiller unavailable. Using IMC 0609, Appendix A, Exhibit 2, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the finding did not represent a loss of system or function and did not represent the loss of a single train for greater than technical specification allowed outage times or greater than 24 hours. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because operators did not recognize and plan for the possibility of mistakes and inherent risk and did not use appropriate error reduction tools. (H.12)
05000387/FIN-2016004-01	2016Q4	Susquehanna	Failure Rates Exceed (20%) Twenty Percent for Biennial Requalification Exam	Green. A self-revealing finding was identified associated with inadequate licensed operator performance during the annual licensed operator requalification operating test and biennial written examination. Specifically, 17 of 71 operators (23.9%) failed at least one portion of the requalification examinations. This finding is more than minor because it is associated with the Mitigating Systems cornerstone attribute of human performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, 17 of 71 licensed operators failed to demonstrate a satisfactory understanding of the required knowledge and abilities required to safely operate the facility under normal, abnormal, and emergency conditions. The inspectors evaluated this performance deficiency using IMC 0609, SDP, Appendix I, Licensed Operator Requalification SDP. This finding is of very low safety significance (Green) because the finding is related to requalification exam results, did not result in a failure rate of greater than 40 percent and all 17 operators were remediated and successfully retested prior to returning to licensed duties. This finding has a cross-cutting aspect in the area of Human Performance, Training, because Susquehanna did not provide adequate operator requalification training to maintain a knowledgeable, technically competent workforce. (H.7)
05000387/FIN-2016004-02	2016Q4	Susquehanna	Failure to Promptly Correct a Condition Adverse to Quality with LPCI Swing Bus Automatic Transfer Switches	Green. A finding of very low safety significance (Green) and associated NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion XVI, Corrective Action, was self-revealed when Susquehanna failed to assure that conditions adverse to quality were promptly identified and corrected on two separate occasions. Both examples resulted in the failures of safety-related automatic transfer switches (ATSs) associated with the low pressure coolant injection (LPCI) swing buses. Corrective actions included enhancing the work instructions for all applicable ATSs based off original equipment manufacturer (OEM) input and scheduling the enhanced work instructions to be performed on the four swing bus ATSs during their next scheduled bus outages. Inspectors determined that the finding was more than minor because it was associated with the Equipment Performance attribute of the Reactor Safety Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). In both examples, the failure to correct conditions adverse to quality resulted in the loss of power to the LPCI swing bus and inoperability of the respective division of LPCI. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, inspectors and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, inspectors determined that the finding was of very low safety significance (Green). Specifically, though a single train was inoperable for greater than its technical specification (TS) allowed outage time, in consultation with regional senior reactor analysts, inspectors determined it did not represent an actual loss of function. The finding is related to the cross-cutting area of Problem Identification and Resolution, Evaluation, because Susquehanna did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Susquehanna either failed to evaluate deficiencies encountered during maintenance or failed to ensure that corrective actions aligned with and corrected the identified causes. (P.2)
05000388/FIN-2016004-03	2016Q4	Susquehanna	Refuel Floor Radiation Monitor Inoperable Due to being Improperly Calibrated	Green. A finding of very low safety significance (Green) and NCV of TS 5.4.1, Procedures was self-revealed when Susquehanna incorrectly calibrated the Unit 1 B refuel floor high exhaust duct high radiation monitor on November 15, 2014. This impacted the initiation capability of secondary containment isolation and control room emergency outside air supply system (CREOASS) and resulted in Susquehanna exceeding the allowed outage time for TSs 3.3.6.2, Secondary Containment Isolation, and 3.3.7.1, CREOASS Instrumentation. Upon identification of the issue, Susquehanna properly calibrated the radiation monitor to restore its operability. This finding is more than minor because it is associated with the Human Performance (Routine OPS/Maintenance Performance) attribute of the Barrier Integrity cornerstone and affected the cornerstone objective of providing reasonable assurance that physical design barriers (Secondary Containment and Control Room Ventilation) protect the public from radionuclide releases caused by accidents or events. Specifically, incorrectly calibrating the radiation monitor resulted in both systems being inoperable for almost two years. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The SDP for Findings At-Power, both dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency was only associated with the radiological barrier function of the Control Room and Secondary Containment. This finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency because Susquehanna did not recognize and plan for the possibility of mistakes, latent problems, or inherent risk, even while expecting successful outcomes. Specifically, Susquehanna personnel did not consider the potential undesired consequences of their actions before performing work and implement appropriate error-reduction tools (e.g. self-check, peer-check). (H.12)
05000387/FIN-2016004-04	2016Q4	Susquehanna	Auxiliary Bus Load Shed when a Daisy Chained Neutral was Interrupted during Maintenance	Green. A finding of very low safety significance (Green) for failure to develop an adequate work plan for replacement of a voltage potential indicating light on a breaker on the Unit 2 B auxiliary bus was self-revealed when the Unit 2 B reactor recirculation pump (RRP) tripped, along with other non-safety related loads on November 14, 2016, resulting in a rapid unplanned power change and transition to single loop operation. Specifically, operations and maintenance personnel did not recognize that disconnecting the neutral wires from the light socket would interrupt power to all of the degraded voltage relays for the auxiliary bus. Therefore, the relays de-energized when the maintenance was performed, tripping all the breakers on the bus. Susquehannas immediate corrective actions included stabilizing the plant, entering single loop operations, and entering the issue into their corrective action program (CAP). Additionally, Susquehanna performed a maintenance department stand down to communicate immediate lessons learned from the event while a more thorough causal analysis was conducted. The performance deficiency was more than minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone and affected its objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, implementation of work instructions resulted in the trip of the Unit 2 B RRP, B and D circulating water (CW) pumps, B and D condensate pumps, and the B service water (SW) pump, which caused an automatic trip of the C reactor feed pump and runback of the A RRP, resulting in a rapid power reduction to 32 percent rated thermal power (RTP). The inspectors evaluated the finding in accordance with IMC 0609, Appendix A "The SDP for Findings At-Power," dated June 19, 2012, Exhibit 1 for the Initiating Events cornerstone and determined the finding was of very low safety significance (Green) because it did not cause a reactor trip. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Work Management because Susquehanna did not implement a process of planning work activities such that nuclear safety is the overriding priority, including the identification and management of risk commensurate with the work. Specifically, Susquehanna did not recognize the risk of interrupting a daisy chained neutral when planning a minor maintenance work order and did not recognize the impact of the work activity in the field. (H.5)
05000387/FIN-2016004-05	2016Q4	Susquehanna	LERs Associated with Reactor Coolant Pressure Boundary Leakage	Enforcement. TS 3.4.4, "RCS" requires RCS leakage be limited to no pressure boundary leakage in Mode 1. Contrary to this, pressure boundary leakage from a LPRM instrument housing and from socket weld #8 occurred between plant start-up in December 2015 and plant shutdown on June 6, 2016, and existed while in Mode 1. The inspectors determined that these violations of TS 3.4.4 are more than minor, but not the result of performance deficiencies. Specifically, for the first event, though leakage likely existed during the previous refueling outage when personnel were performing unrelated maintenance and inspection activities, it was likely too small to reasonably identify and correct. Similarly, for the 2016 leak identified in weld #8, the leakage causes were not within Susquehannas ability to foresee as they had replaced the weld with the industry recommended 2 x 1 taper configuration and used qualified procedures and personnel. The Susquehanna staff had also measured the susceptibility of the attached piping for vibrational inputs. In accordance with the NRC Enforcement Policy guidance and IMC 0612, these violations are being treated under the traditional enforcement process and best characterized as a Severity Level (SL) IV (very low safety significance) violation, similar to example d.1 in NRC Enforcement Policy, Section 6.1, Reactor Operations. Although a performance deficiency was not identified, to verify that the issue was of very low safety significance, the inspectors considered risk insights obtained by using IMC 0609, SDP, Appendix A, Exhibit 1, Initiating Events Screening Questions. The inspectors determined that these TS violations would screen to Green (very low safety significance) because the boundary leakage would not have exceeded the leak rate for a small loss of coolant accident (LOCA) and would not affect any LOCA accident mitigating systems or components. Therefore, the inspectors considered that the SL IV characterization was appropriate. The licensee entered these issues into the Susquehannas CAP as CR-2016-14544 and CR-2016-14366. Because these issues are of very low safety significance, it has been determined that it was not reasonable for Susquehanna to be able to foresee and prevent, and as such no performance deficiencies exist. The NRC has decided to exercise enforcement discretion in accordance with Sections 2.2.4 and 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation of TS (EA-16-283). Further, because Susquehanna's actions did not contribute to this violation, it will not be considered in the assessment process or the NRC's Action Matrix.
05000387/FIN-2016003-02	2016Q3	Susquehanna	Risk Management Actions Not Adequately Implemented	The inspectors identified a Green NCV of 10 CFR 50.65(a)(4) because Susquehanna did not assess and manage the increase in risk from online maintenance activities. From September 11 to 16, 2016, there were multiple affected areas that the fire protection engineer or designee did not walk down to inspect for fire impairments resulting in deficiencies not being corrected prior to releasing work and no fire watch was established for the impairments. Susquehanna removed the combustible materials from the areas or stationed a fire watch, and entered these issues into their CAP as CR-2016-21125, CR-2016-21423, CR-2016-21616, and CR-2016-21741. This finding is more than minor because it adversely impacted the protection against external factors attribute of the Mitigating Systems cornerstone objective to ensure the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, not implementing the required risk management actions (RMAs) for the only available safe shutdown pathway placed the station in a much higher risk condition in the event of an internal fire. The inspectors evaluated the finding in accordance with IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. Since the performance deficiency was related to maintenance activities affecting structure, system, and components needed for fire mitigation, Appendix K directed the significance to be determined by an internal NRC management review using risk insights. IMC 0609, Appendix F, Attachment 1 Fire Protection Significance Determination Process Phase 1 Worksheet, was used to develop this risk insight. Based on the nature and quantity of combustible materials in the areas, combined with the relatively short duration of which the fire risk was unmitigated, inspectors determined that it was of very low safety significance (Green). The finding was determined to have a cross-cutting aspect in the area of Human Performance, Avoid Complacency, in that, individuals did not plan for latent issues and inherent risk, even while expecting successful outcomes. Specifically, combustible materials were not appropriately controlled as required by OI-013-002, Fire Risk Management, Revision 10, because in some cases they were assumed to be exempt from the program requirement or staff did not tour the areas because they assumed there were no combustible materials present based on past experience.
05000388/FIN-2016003-01	2016Q3	Susquehanna	Inadequate Work Instructions for Breaching Internal Flood Barrier	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Susquehanna did not ensure that work instructions to breach a flood barrier appropriately incorporated design requirements for internal flooding so that equipment necessary to achieve and maintain safe shutdown would not be impacted. From August 30, 2016 to September 2, 2016, work instructions directed a breach of a flood barrier that was credited to provide assurance that equipment necessary for safe shutdown of the plant was protected against the effects of medium energy line breaks and, therefore, were not appropriate to the circumstances. Susquehanna entered this issue into their corrective action program (CAP) as condition report CR-2016-20472 and CR-2016-20859 and revised the work instructions to require a worker remain in the vicinity of the penetration to ensure that flooding could be secured prior to impacting equipment necessary to reach and maintain safe shutdown. This finding is more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, had the breach been completed, it could have allowed a medium energy line break in one flooding area to communicate with another area, potentially impacting equipment necessary to achieve and maintain safe shutdown. The inspectors evaluated the finding using IMC 0609, Appendix A, Exhibit 2, "Mitigating System Screening Questions," and determined the finding to be of very low safety significance (Green) because the PD 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 allowed outage time, and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding has a cross-cutting aspect of Human Performance, Work Management because Susquehanna did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Implementation of Susquehannas work planning process did not ensure that the maintenance incorporated all requirements for protection against internal flooding and did not ensure that job site conditions were consistent with assumptions in engineering analyses.
05000388/FIN-2016008-03	2016Q3	Susquehanna	Failure to Implement or Develop Timely Interim or Final Corrective Actions for a Degraded Condition	The inspectors documented a self-revealing finding of very low safety significance (Green) against Susquehanna procedures LS-125 Revision 4, Corrective Action Program (CAP), and OI-AD-096 Revision 18, Operator Challenges, for the failure to correct and establish appropriate corrective actions for a known degraded condition for an uninterruptable power supply (UPS) for vital 120 VAC load centers. Specifically, Susquehanna did not correct nor establish compensatory actions for the transfer switch for a UPS which was failed for over one year. The degraded condition subsequently complicated operator response to the loss of a vital 480 VAC switchboard and resulted in an unplanned manual reactor scram and valid emergency core cooling system (ECCS) actuation on May 13, 2016. Susquehanna entered this issue into their CAP, conducted an apparent cause evaluation, and repaired the UPS transfer switch. The finding was more than minor because it was associated with the Equipment Performance 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, the long standing degraded condition of UPS 2D14212/2B246082 was not corrected or compensated for and did not function as designed, as a result operators had to manually scram the reactor following the loss of a vital bus on May 13, 2016. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not cause both a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. Specifically, while this performance deficiency resulted in a reactor scram, it was not the cause of the loss of mitigation equipment credited in the Susquehanna safety analysis. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution Resolution because the organization did not take effective corrective actions to address issues in a timely manner commensurate with its safety significance. Specifically, failing to establish appropriate compensatory actions for this known degraded condition, prevented the operators from responding appropriately to a loss of a vital 480 VAC switchboard initiating event. (P.3). (Section 4OA2.1.c(3))
05000277/FIN-2016003-01	2016Q3	Peach Bottom	Reactor Feed Pump Controller Power Supply Shelf Life Not Maintained	A self-revealing finding of very low safety significance (Green) was identified for Exelons failure to maintain the Unit 2 C reactor feed pump (RFP) Woodward controller secondary power supply in accordance with PES-S-002, Exelon Shelf Life Program. Specifically, on May 27, 2016, the Unit 2 C RFP experienced speed oscillations due to an age-related failure of the Woodward controller secondary power supply, which resulted in an automatic recirculation runback to 53 percent rated thermal power (RTP). The power supply contained an electrolytic capacitor that had exceeded its shelf life per PES-S-002. This issue was entered into Exelons corrective action program (CAP) under issue report (IR) 02691322. Exelons corrective actions included replacement of the faulted power supply and an extent of condition (EOC) review of proper expiration date entry for shelf life program components. The finding was more than minor because it was associated with the equipment performance attribute of the Initiating Events cornerstone and affected the cornerstones objective of limiting the likelihood of events that upset plant stability during power operations. The inspectors evaluated the finding in accordance with Exhibit 1 of Inspection Manual Chapter (IMC) 0609, Appendix A, SDP for Findings At-Power, and determined the finding was of very low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that no cross-cutting aspect was applicable to this finding because the performance deficiency (PD) was not indicative of current performance. The PD occurred between 1997 and 1999 when the power supply expiration date was incorrectly coded in Exelons work management process in accordance with PES-S-002.
05000387/FIN-2016008-01	2016Q3	Susquehanna	Failure to Write a Condition Report for Degraded Conditions Which Challenged Operability of Safety Related Equipment	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for Susquehanna failing to identify and correct conditions adverse to quality in a timely manner. Specifically, between April 16, 2016 and April 22, 2016, condition reports for potential or suspected degraded or non-conforming conditions related to the High Pressure Coolant Injection System (HPCI) and Reactor Core Isolation Cooling System (RCIC) were not written and operability determinations performed. In both cases, the equipment was subsequently declared inoperable due to the conditions. The issues were entered into the CAP and the equipment was taken out of service, repaired, and retested satisfactorily. The inspectors determined that there were two examples of the same performance deficiency and violation. In accordance with NRC Enforcement Manual Section 1.3.4, Documenting Multiple Examples of a Violation, multiple examples of a single violation are allowed to be documented as a single violation bounded by the characterization of the most significant example. The RCIC example is considered the most significant due to the longer exposure time in a required mode and number of mode changes that occurred during the exposure period. The finding was more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the associated cornerstone objective to ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to identify and correct degraded conditions associated with a RCIC system lube oil leak which rendered that system inoperable. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, the inspectors determined that this finding screened to Green because the safety function was not lost, and the finding did not represent an actual loss of function of at least a single train for greater than its Tech Spec Allowed Outage Time or two separate safety systems out-of-service for greater than its Tech Spec Allowed Outage Time. This finding had a cross-cutting aspect in the area of Human Performance, Teamwork, because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, in both examples, individuals were aware of potential degraded conditions but actions were not taken to communicate the activity to other groups, such as the control room operators, to allow for the issues to be evaluated for operability and determine if proposed actions were timely and/or appropriate. (H.4) (Section 4OA2.1.c(1))
05000387/FIN-2016008-02	2016Q3	Susquehanna	Failure to Implement and Maintain Quality Procedure Results in Control Room Chiller Inoperability	The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to implement and maintain a quality procedure, MT-GE-021, Chiller Maintenance and Inspection. This resulted in the safety related 0K112A chiller being operated outside of its design specifications and being declared inoperable. Specifically, on January 9, 2014, a system engineer directed the maintenance personnel to overcharge 0K112A with R-114a refrigerant, which led to higher power consumption by the chillers compressor motor, and the failure of the next biennial surveillance test on December 10, 2015 due to excessive compressor motor current. Susquehanna entered the issue into the CAP, conducted testing to establish the proper refrigerant charge, removed the excess refrigerant, and revised the procedure. The finding was determined to be more than minor because it was associated with the Mitigating System cornerstone attribute of Equipment Performance and adversely affected the associated cornerstone objective to ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The refrigerant overcharge condition resulted in the 0K112A chiller being inoperable and unable to fulfil its safety function to cool safety related switchgear and equipment during accident conditions for a period of 23 months. 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 a detailed risk evaluation would be required because the finding involved an actual loss of function of at least a single Train for greater than its Technical Specification allowed outage time of 30 days. A detailed risk assessment was performed by a Region 1 Senior Reactor Analyst (SRA). The SRA determined the finding to be of very low safety significance (Green.) This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence because individuals did not follow processes, procedures, and work instructions. Specifically, for many years maintenance and engineering personnel relied upon informal work practices vice referring to the procedure when charging the chillers with refrigerant. (H.8) (Section 4OA2.1.c(2))
05000387/FIN-2016008-04	2016Q3	Susquehanna	Failure to Promptly Identify and Correct a Condition Adverse to Quality on Vital 480 VAC MCCs	The inspectors documented a self-revealing Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to identify and correct a condition adverse to quality. Specifically, in October and December 2006 and July 2009, Susquehanna did not identify a non-conforming condition with the design and performance requirements of several 480 volt motor control center (MCC) breaker assemblies during receipt inspections. These non-conforming breaker assemblies were installed in vital 480 VAC applications and subsequently led to a phase to ground short and loss of a 480 volt safety-related motor control center on May 12, 2016. Susquehanna entered this issue into their CAP, conducted an apparent cause evaluation, replaced the damaged breaker assembly, and is conducting an extent of cause review for other susceptible breaker assemblies. The finding was more than minor because it was associated with the Design Control attribute of the Initiating Events cornerstone and adversely affected the associated cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, on May 12, 2016, an electrical transient on vital AC bus 2B246 occurred as a result of a phase to ground fault in breaker cubicle 2B24609, which resulted in a loss of bus 2B246 and associated safety related loads. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not cause both a reactor trip and loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding did not have a crosscutting aspect because the performance deficiency was a historical issue with the actions taken in 2005, 2006, and 2009, and is not indicative of current licensee performance. (Section 4OA2.1.c(4))
05000353/FIN-2016003-01	2016Q3	Limerick	Inadequate Design Control of Plant Processing Computer Modification	A self-revealing finding of very low safety significance (Green) was identified when Exelon did not implement their engineering design control procedures during the plant processing computer (PPC) modification. Specifically, Exelon did not fully address effects of the modification on other plant systems and did not establish a testing boundary that encompassed all components whose operation was altered by the modification. As a result, the PPC modification had a wiring design error that resulted in the trip of both reactor recirculation pumps (RRPs) which required a manual reactor trip of Unit 2. In response to this issue, Exelon initiated IR 2676712, investigated the cause of the trip, fixed the wiring design error, performed a root cause evaluation, and performed an extent of condition review. This issue is more than minor because it adversely affected the design control attribute of the initiating events cornerstone to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the PPC modification process had a wiring design error that resulted in the trip of both RRPs which required a manual reactor trip of Unit 2. The issue was evaluated in accordance with IMC 0609, Appendix A, "Significance Determination Process for Findings At-Power, using Exhibit 1, "Initiating Events Screening Questions, Section B, Transient initiators. The finding was determined to be of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Challenge the Unknown, because LGS staff did not stop when faced with uncertain conditions, and risks were not evaluated and managed before proceeding. Specifically, Exelon did not stop and reevaluate the risks and effects on plant systems when changes were made to the PPC design modification package. (H.11)
05000352/FIN-2016003-02	2016Q3	Limerick	Licensee-Identified Violation	The following violation of very low safety significance (Green) was identified by Exelon and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation. 10 CFR 50.54(q)(2), Emergency Plans, requires, in part, that a holder of a licensee under this part shall follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to this part, and for nuclear power reactor licensees, the planning standards of 50.47(b). 10 CFR 50.47(b)(4) requires that a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee. Contrary to the above, from April 25, 2016, until August 3, 2016, the spent fuel pool level emergency action level (EAL) RG2/RS2 threshold of Limericks Emergency Plan for a General Emergency and Site Area Emergency did not meet the requirements of Appendix E and the planning standards of 10 CFR 50.47(b). Specifically, Exelon identified that the spent fuel pool level for RG2/RS2 threshold was 0.08 feet, and the correct threshold value was 0.8 feet. The spent fuel pool EAL threshold values for a lowering water level for an Alert and Unusual Event were correct at 10.20 feet and less than 22 feet, respectively. The normal spent fuel pool water level is over 23 feet. The inspectors evaluated this finding using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.4-1. This Table indicates, in part, that the following should be assessed as low safety significance (White): an EAL has been rendered ineffective such that any General Emergency would not be declared for a particular off-normal event, but because of other EALs, an appropriate declaration could be made in a degraded manner (e.g. delayed), and, an EAL that has been rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event. However, the inspectors confirmed that the spent fuel pool level instrumentation at LGS goes off scale at approximately 0.635 feet, and the Limerick Emergency Plan, in Addendum 3, directs any Emergency Director to assume the EAL threshold has been exceeded if the associated parameter goes off scale. In addition, the NEI recommended and NRC endorsed value for this EAL threshold would have been at nominally 0.0 feet, the level at which the fuel remains covered and actions to implement make-up water addition should no longer be deferred. Although the LGS threshold for declaration at 0.8 feet would have been exceeded, the inspectors concluded that the event would have been classified when the SFP level dropped below 0.635 feet, sufficiently above the NEI recommended level. Because the event would have been declared with margin to the actual water level needed for protection of the public, i.e. the spent fuel would still be fully covered by water at the time of the EAL declaration(s), the inspectors concluded that this performance deficiency was most similar to the Table 5.4-1 branches representing very low safety significance (Green). Exelons corrective actions included revising EP-AA-1008, Addendum 3, with the correct spent fuel pool level EAL RG2/RS2 threshold of 0.8 feet. Because this issue was of very low safety significance (Green) and Exelon entered the issue into the corrective action program (IR 2700440), this finding is being treated as a non-cited violation, consistent with Section 2.3.2.a of the Enforcement Policy.
05000277/FIN-2016002-03	2016Q2	Peach Bottom	Human Performance Event Results in Emergent Downpower	A self-revealing finding of very low safety significance (Green) was identified for the failure of Exelon operators to use human performance error reduction tools during equipment manipulation in accordance with HU-AA-101, Human Performance Tools and Verification Practices. Specifically, on March 28, 2016, an equipment operator failed to use self-check (STAR) while removing a circuit breaker from service and incorrectly tripped the E-124 480 volt supply breaker which required a rapid manual power reduction to 80 percent rated thermal power (RTP) due to lowering main condenser vacuum and a partial loss of feedwater heating. Exelon entered the issue into their corrective action program (CAP) under issue report (IR) 2646772 and performed a root cause which identified corrective actions to address the adverse human performance behaviors at the station. The finding was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown, as well as power operations. Specifically, an equipment operator failed to adequately use human performance error reduction tools and opened an incorrect breaker which required a rapid downpower. The inspectors evaluated the finding in accordance with Exhibit 1 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, and determined the finding was of very low safety significance (Green) because it did not result in a reactor trip and the loss of mitigation equipment relied upon for transition to a stable shutdown condition. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Field Presence, because Exelon did not ensure that deviations from standards and expectations, which were identified by leaders, were corrected promptly. Specifically, Exelon identified that adverse human performance behaviors existed with certain equipment operators, however, those observations were not appropriately input into their performance management system, such that the behaviors could be addressed. Thus, these adverse behaviors were a primary contributor to this human performance error.
05000277/FIN-2016002-01	2016Q2	Peach Bottom	Improperly Stored Material in Reactor Building	The NRC identified a very low safety significance (Green) NCV of Technical Specification (TS) 5.4.1 for Exelons failure to adequately implement procedure requirements governing the storage of material in a safety-related structure. Specifically, on April 26, 2016, Exelon technicians stored ladders vertically without them being adequately tied off to prevent the ladders from falling over in accordance with MA-AA-716-026, Station Housekeeping / Material Condition Program. The inspectors identified that the ladders were stored in the PB Unit 2 reactor building (RB), such that they could fall over and impact safety-related equipment. The inspectors promptly notified Exelon, the ladders were immediately removed, and the condition was documented under IR 2661309. This finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the reliability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The SDP for Findings At-Power, Exhibit 2. The inspectors determined this finding to be of very low safety significance (Green) because the degraded condition was not a design deficiency that affected system operability; did not represent an actual loss of function of a system; did not represent an actual loss of function of a single train or two separate trains for greater than its TS allowed outage time; and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety significant. The finding was determined to have a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon technicians did not store ladders in safety-related buildings in accordance with station procedures, such that they could not fall over and damage safety-related equipment.
05000388/FIN-2016002-07	2016Q2	Susquehanna	HPCI Overridden Prior to Manual Reactor Scram	An NRC-identified finding of very low safety significance (Green) and associated NCV of TS 5.4.1.a, Procedures, was identified when Susquehanna failed to implement procedures for controlling the high pressure coolant injection (HPCI) system. Specifically, operators overrode automatic initiation of the system prior to inserting a manual scram, contrary to the requirements of OP-252-001, HPCI System, and OP-AD-300, Administration of Operations. This was entered into the CAP as CRs 2016-12854 and 2016-13118 and 2016-13136, the operators involved in the event were remediated, and lessons learned communicated to other station personnel. The finding was more than minor because it was associated with the Human Performance attribute of the Mitigating Systems Cornerstone and affected the objective to ensure the availability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, overriding the HPCI system prior to initiating a plant scram rendered the system unavailable to respond to a level transient or failure of the non-safety related feedwater system. The inspectors evaluated the finding in accordance with Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012 and determined that it required a detailed risk assessment because it represented a loss of the single train systems function. The Region 1 SRA performed a detailed risk evaluation using the Susquehanna Unit 2 standardized plant analysis risk (SPAR) Model, version 8.23. The issue was conservatively modeled with a HPCI failure to start due to the system automatic start signal being overridden. The change in core damage frequency per year was determined to be in the E-10 range due to the very short duration the system auto start feature was defeated. Therefore the issue was determined to be of very low safety significance (Green). The finding is related to the cross-cutting area of Human Performance, Procedure Adherence because Susquehanna did not follow processes, procedures and work instructions. Specifically, operators did not ensure that their actions were appropriately authorized by procedures when taking action to override a key safety system prior to a plant transient.
05000387/FIN-2016002-01	2016Q2	Susquehanna	Failure to Promptly Identify a Condition Adverse to Quality Associated with Primary Containment Isolation Valves	A self-revealing Green finding and associated violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, and TS 3.6.1.3, Primary Containment Isolation Valves (PCIVs), was identified when Susquehanna did not promptly identify a condition adverse to quality. Despite observing abnormal behavior during local leak rate testing following replacement in May 2014, Susquehanna did not take any action to ensure that certain Reactor Water Cleanup (RWCU) system PCIVs passed their subsequent testing. Consequently, these valves failed their in-service and local leak rate test in March 2016 when they failed to close upon securing system flow. The failure was caused by an internal interference between the check valve hinge and body. Following the failures in March 2016, Susquehanna repaired the valves and successfully performed local leak rate testing, restoring operability of the PCIVs. The repeat failure was entered into the CAP as CRs 2016-06960 and 2016-09940. The finding was determined 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 cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to identify a condition adverse to quality during post-maintenance testing resulted in two PCIVs being rendered inoperable for longer than the TS allowed outage time. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency did not involve the hydrogen recombiners and did not result in an actual open pathway in the physical integrity of reactor containment. Specifically, the redundant valve for each penetration remained operable during the period in which these two valves were inoperable. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because Susquehanna did not use decision making practices that emphasized prudent choices over those that are simply allowable. Specifically, Susquehanna decided to accept elevated seat leakage for two new PCIVs, assuming that they could be declassified as PCIVs.
05000387/FIN-2016002-02	2016Q2	Susquehanna	Failure to Promptly Correct a Condition Adverse to Quality with A EDG MOC Switch	A self-revealing finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for failure to correct a condition adverse to quality. Specifically, on March 23, 2016, the A emergency diesel generator (EDG) failed its technical specification (TS) surveillance test in that the emergency switchgear room cooler, 1V222A, started immediately when the EDG loaded onto the emergency bus following a simulated loss of off-site power (LOOP) and simulated Emergency Core Cooling System (ECCS) Initiation, rather than sequencing onto the bus as intended by design. Susquehanna identified the direct cause of the failure was due to a misadjustment of the mechanism-operated cell (MOC) linkage switch (S1) in the A EDG output breaker to the 1A 4 kilovolt (kV) bus, which provides the electrical logic to the 1V222A load timer. The repeat failure was entered into the corrective action program (CAP) as CR-2016-08643, the MOC linkage was realigned, and the functions satisfactorily tested. The finding was determined to be more than minor because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to correct the degraded condition rendered the A EDG inoperable for longer than the TS allowed outage time. In accordance with IMC 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the inspectors determined that this finding required a detailed risk assessment because the finding represents an actual loss of function of a single train for greater than the TS allowed outage time. Specifically, the A EDG was inoperable from July 19, 2010 until April 2, 2016, because TS requires functioning of the sequencing timers for the EDG to be operable. In coordination with a Region 1 Senior Risk Analyst, the issue was qualitatively screened as Green (very low safety significance) based on the low initiating event frequency associated with a loss of coolant accident (LOCA) co-incident with a LOOP event, and observed successful EDG function during multiple LOOP/LOCA tests over the period in question. This would result in a delta core damage frequency substantially less than E-6. Additionally, it was reasonable to conclude that the A EDG remained available to perform its function given the minimal increased load on the machine as evidenced during the performance of the LOOP-LOCA surveillance testing in 2012, 2014, and 2016. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Susquehanna did not thoroughly evaluate the issue to ensure that the resolution addressed the cause and extent of conditions commensurate with their safety significance. Specifically, Susquehanna corrected a suspected condition without appropriate troubleshooting until the third identical failure of the 1V222A load timer.

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