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05000289/FIN-2018003-022018Q3Three Mile IslandMinor ViolationThis violation of minor significance was identified by the licensee and has been entered the licensee corrective action program and is being treated as a minor violation, consistent with the NRC Enforcement Policy. During TMIs 2015 refueling outage (T1R21) NRC and the licensee identified issues regarding reactor building pre-staging of materials were documented in NRC inspection report 05-289/2017008 (ADAMS Accession Number ML17191A697). Exelon evaluated and documented corrective actions in ACE report 2578255 which included an action to conduct an effectiveness review of those corrective actions. On October 18, 2017, after refueling outage T1R22, Exelon completed this effectiveness review. Exelon concluded that the implemented corrective actions were ineffective based on an adverse trend of licensee-identified reactor building pre-staging issues during the T1R22 refueling outage preparations. Exelon documented the results of the effectiveness review under assignment 21 of ACE 2578255 and the adverse trend in issue report 4051608. Primarily, direct oversight by Exelon staff during all phases of pre-staging, as approved by the management review committee, was not implemented and resulted in improper storage of materials in the reactor building during pre-staging activities. The improper storage was identified by Exelon during end-of-day walkdowns, from September 11 thru September 14, 2017, and documented in the corrective action program. All other corrective actions from ACE 2578255 were properly implemented. Screening: Exelons failure to implement the approved corrective actions is a performance deficiency. The inspector evaluated the significance in accordance with IMC 0612, Appendix B, Issue Screening. The inspector determined that this issue was of minor safety significance because non-compliant material configurations in the reactor building were corrected before being left unattended at the end of shift and that the corrective actions determined by ACE 2578255, except for direct Exelon supervision during pre-staging activities, were adequately implemented. Enforcement: Exelon identified this violation and documented the issue in report assignments 2578255-21 and 4051608-02. Exelon has initiated actions to include direct Exelon supervision to the current pre-staging corrective actions (AR 4051608-03) and will conduct an effectiveness review of pre-staging activities after the next outage (AR 2578255-22). This failure to comply with 10 CFR Part 50 Appendix B Criterion XVI constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000289/FIN-2018003-012018Q3Three Mile Island1A Emergency Diesel Generator Lube Oil Leak Inadequate Corrective ActionsA self-revealed Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified for failure to develop and implement adequate corrective actions to ensure the availability and reliability of the 1A emergency diesel generator.
05000334/FIN-2017007-012017Q4Beaver ValleyNon -Conservative Differential Pressure Value Used in Low Head Safety Injection Motor -Operated Valve Design AnalysisThe NRC team identified a finding of very low safety significance (Green) involving a non- cited violation of 10 CFR 50, Appendix B, Criterion III, Design Control, because FENOC staff did not establish measures to assure that the design bases were correctly translated into specifications, drawings, procedures, and instructions. Specifically, for the recirculation phase following a postulated small break loss -of-coolant accident, engineering staff determined the maximum differential pressure fo r motor- operated valves MOV -1SI -863A and MOV -1SI -863B to be the low head safety injection pump shutoff head, but the actual configuration could have resulted in a higher differential pressure at the valve due to allowable reactor coolant system leakage past downstream pressure isolation valves . In response, FENOC staff initiated corrective action program condition report s and assessed the deficiency , and concluded that affected motor -operated valves remained functional although with reduced valve thrust design margin . This finding was more than minor because it was associated with the Design Control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The SDP for Findings At -Power, the team determined that this finding was of very low safety significance (Green) because it was a design deficiency confirmed not to result in the loss of operability or functionality. This finding was not assigned a cross -cutting aspect because the issue did not reflect current licensee performance.
05000277/FIN-2017003-022017Q3Peach BottomLicensee-Identified Violation10 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.
05000278/FIN-2017003-012017Q3Peach BottomInstructions Not Followed for Replacement of HPSW Ventilation Switch BlockA 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.
05000443/FIN-2017008-022017Q3SeabrookFailure to Implement Test Program for Appendix R Emergency Lighting UnitsThe team identified a Green, non-cited violation of Seabrook License Condition 2.F, Fire Protection, because NextEra did not implement the fire protection test program to ensure that the emergency lighting units were in conformance with design requirements. Specifically, NextEra did not implement procedure LS0565.31, 8-Hour Emergency Light Inspections, to verify that the Appendix R emergency lighting units would meet the annual inspection requirements, as well as the 3-year preventive maintenance task for battery replacement and the 8-hour capacity test. Additionally, since the 3-year preventive maintenance task was coded incorrectly, there was no process to ensure that the LS0565.31 would be completed going forward. NextEra entered this issue into the corrective action program as AR 2214652. NextEras planned corrective actions included revising the classification of the emergency lighting unit preventive maintenance task in order to ensure that the task is performed at the appropriate frequency.The team determined that this issue was more than minor because if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, failure to conduct the annual inspection requirements and 3-year preventive maintenance activities could result in the emergency lighting units not meeting the 8-hour battery capacity requirement. The team evaluated this finding using Inspection Manual Chapter 0609, Fire Protection Significance Determination Process. Because safe shutdown conditions could be reached and maintained, this finding screened as having very low safety significance (Green). The team determined this finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the three-year preventive maintenance task to replace the batteries in the emergency lighting units was coded incorrectly in the work management system, which resulted in NextEra not completing the required testing and maintenance on the lighting units to ensure that they would perform their function during safe shutdown operations (H.5).
05000443/FIN-2017008-012017Q3SeabrookFailure to Correct Condition Adverse to Fire Protection Associated with Fire Safe ShutdownThe team identified a Green, non-cited violation of Seabrook Station Unit 1 Facility Operating License Condition 2.F, Fire Protection, for failure to implement and maintain in effect all provisions of the approved Fire Protection Program. Specifically, although NextEra identified that procedure OS1200.00 did not properly implement a mitigating action for a fire in the Switchgear Room A as prescribed in the Appendix R Safe Shutdown Analysis Report on August 30, 2010 (Action Requests (ARs) 576775 and 1638123), corrective actions were delayed due to higher priority work and were not timely commensurate with the potential safety significance. NextEra entered the issue into the corrective action program as AR 2214834 and planned to reprioritize the preparation and submittal of a license amendment request to resolve the issue.The issue was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors (fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, by failing to correct the condition in a timely manner, NextEra did not ensure that the associated fire safe shutdown procedure implemented actions to mitigate a fire in the Switchgear Room A as analyzed in the Appendix R Safe Shutdown Report. The team performed a Phase 1 screening in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process. The deficiency affected the post-fire safe shutdown category because NextEras fire response procedures were degraded. The finding was screened to very low safety significance (Green) because it was assigned a low degradation rating because the procedural deficiencies could be compensated by operator experience and system familiarity. This finding was determined to have a cross-cutting aspect in the area of Human Performance, Resources, in that, NextEra did not ensure that personnel, equipment, procedures, and other resources are available and adequate to support nuclear safety. Specifically,action to submit a license amendment request to support a deviation from the 10 CFR Part 50, Appendix R, III.G.2 requirements for cable separation had been rescheduled five times due to higher priority licensing work (H.1).
05000336/FIN-2017007-012017Q3MillstoneFailure to Replace Auxiliary Feedwater Solenoid Valves within the Required FrequencyThe inspection team identified a Green non-cited violation of Technical Specification 6.8.1.a, Procedures, because Dominion did not implement procedures as required by Regulatory Guide 1.33, Revision 2, Appendix A.9, Procedures for Performing Maintenance, to properly maintain the environmental qualification of safety-related auxiliary feedwater solenoid valves 2-FW-43AS and 2-FW-43BS. Specifically, Dominion failed to implement the recurring work event task and associated work order to ensure that these auxiliary feedwater solenoid valves were replaced prior to exceeding the qualified life of the solenoid coil and elastomer components. Dominion entered this issue into their corrective action program as condition report 1076005, planned replacement of the solenoid valves, and calculated an alternate ambient temperature for use in determining the qualified life of the solenoid valves. Dominion re-performed the qualified life calculation using this revised ambient temperature and extended the qualified life to support operability. The inspection team determined that this issue was more than minor because it adversely impacted the equipment performance attribute of the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. This issue is also similar to more- than-minor examples 3.j and 3.k presented in IMC 0612, Appendix E, Examples of Minor Issues. Specifically, this performance deficiency resulted in a condition where there was reasonable doubt as to the operability and reliability of the solenoid valves for both auxiliary feedwater regulating valves, and thus, both trains of auxiliary feedwater. As such, Dominion needed to conduct additional engineering evaluation to extend the service life of the solenoid valves, thus justifying that the valves would continue to perform their safety function. The inspection team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the reliability of a mitigating structure, system, or component, and the structure, system, or component maintained its operability or functionality. The inspection team determined that no cross-cutting aspect was applicable because the finding was not indicative of current performance.
05000387/FIN-2017002-032017Q2SusquehannaFollow -Up of Events and Notices of Enforcement DiscretionInspection 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.
05000388/FIN-2017002-022017Q2SusquehannaFailure to Assess and Manage Risk Associated with Emergent WorkThe 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)
05000387/FIN-2017002-012017Q2SusquehannaInadequate Assessment of Fire Brigade Performance during an Unannounced DrillThe 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)
05000289/FIN-2016004-012016Q4Three Mile IslandLicensee-Identified ViolationTechnical specification 3.2.12.1, "LTOP Protection", requires when the reactor vessel head is installed and indicated reactor coolant system temperature is 313F, high pressure injection pump breakers shall not be racked in unless injection valves (MU-V16A/B/C/D and MU-V217) are closed with their associated breakers open and that pressurizer level is maintained 100 inches, or restore pressurizer level to 100 inches within 1 hour. Contrary to technical specification 3.2.12.1, during reactor coolant system filling with the vessel head installed and temperature < 313F, high pressure injection pump breakers were racked in while pressurizer level was >100 inches for greater than 1 hour. The condition existed for 2 hours and 49 minutes until recognized by the operating crew when questioned by a senior reactor operator trainee, at which time the crew took immediate actions to reduce pressurizer level <100 inches within 1 hour. Additional corrective actions included crew remediation, additional main control room supervisory oversight, and procedure changes. Exelon entered this issue into the corrective action program as issue report 3949713. The inspectors determined that the finding was of very low safety significance (Green) in accordance with NRC IMC 0609, Appendix G, Shutdown Operations, Attachment 1, Exhibit 4, since the finding did not represent an inadvertent safety injection and did not render the power-operated relief valve (LTOP Protection) unavailable or degraded.
05000293/FIN-2016003-012016Q3PilgrimProcess Radiation Monitor Subsystems 10 CFR 50.65(a)(2) Not MetInspectors identified a Green NCV of 10 CFR 50.65 (a)(2), because Entergy did not adequately demonstrate that the performance of the process radiation monitors (PRMs) was effectively controlled through performance of appropriate preventive maintenance. Specifically, Entergy did not identify and properly account for functional failures of four PRM subsystems in July 2014 and February, April, and July 2015; and did not recognize that the subsystems had exceeded their performance criteria and required a Maintenance Rule (a)(1) evaluation. Entergy entered the issue into the CAP under CR-2016-05564. Entergy performed the Maintenance Rule (a)(1) evaluation, and placed them into (a)(1) where they will be monitored against specific goals. The finding is more than minor because it is associated with the Plant Facilities/Equipment and Instrumentation attribute of the Public Radiation Safety cornerstone and affects the cornerstone objective of ensuring the adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, following the failures of the Main Stack Normal Range subsystem in July 2014, the Reactor Building Closed Cooling Water (RBCCW) subsystem in February 2015, the Shared Components subsystem in April 2015, and the Torus Containment High Radiation Monitoring System (CHRMS) subsystem in July 2015, Entergy did not identify the failures as functional failures, and consequently, did not establish goals and monitoring criteria in accordance with 10 CFR 50.65(a)(1). The inspectors determined that the failures demonstrated that the performance of the subsystems was not being effectively controlled through appropriate preventive maintenance, because the incorrect screenings resulted in exceedance of the subsystems performance criteria and placement in (a)(1) status. The inspectors evaluated the significance of this finding using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process. The finding is of very low safety significance (Green) because the finding was in the Effluent Release Program, but did not result in a failure to implement the Effluent Release Program, and did not result in dose to the public in excess of 10 CFR 50, Appendix I criterion or 10 CFR 20.1301(e) limits. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, in that the organization did not thoroughly evaluate issues to ensure that resolution addressed causes and extent of conditions commensurate with their safety significance. Specifically, Entergy identified all of the failures of the PRM subsystems, however, Entergy did not thoroughly evaluate the failures as maintenance rule functional failures.
05000293/FIN-2016003-022016Q3PilgrimInadequate Operability Assessment on EDG BThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, in that Entergy did not perform an adequate operability evaluation in accordance with EN-OP-104, Operability Determination Process, Revision 10. Specifically, during an instrumented run of emergency diesel generator (EDG) B, the cabinet door was opened, resulting in a non-seismically qualified configuration of protective relays for EDG B. Inspectors determined that Entergy did not adequately assess the operability of EDG B as required by EN-OP-104, Operability Determination Process. Specifically, Entergy did not evaluate the operability of EDG B when opening a cabinet door containing relays that serve a safety function. Entergy entered this issue into the corrective action program (CAP) as condition report (CR)-2016-5779 and CR-2016-7877. Entergy has issued a standing order to assess operability of equipment tested with cabinet doors open prior to performing work or declare the equipment being tested inoperable. This is a performance deficiency that was within Entergys ability to foresee and correct. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, relays were no longer in a configuration known to operate as required during a seismic event with the cabinet door open. In accordance with IMC 0609.04, initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2013, the inspectors determined that this finding is of very low safety significance (Green) because the performance was not a design or qualification deficiency, did not involve an actual loss of safety function, and did not represent actual loss of safety function of a single train for greater than its technical specification (TS) allowed outage time. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, in that the organization did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, Entergy did not evaluate industry operating experience on control of cabinet doors containing safety-related equipment, which led to operability concerns.
05000289/FIN-2016003-012016Q3Three Mile IslandEmergency Diesel Generator Internal Flooding Risk Not EvaluatedThe inspectors identified an NCV of Title 10 Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, Design Control, in that Exelon did not ensure the availability of the emergency diesel generator (EDG) following a seismic event. The inspectors reviewed the TMI licensing basis for internal flooding, associated evaluations and conditions reports, and walked down safety-related structures system and components (SSCs). During this review the inspectors determined that non-seismic piping failures in the EDG room were not properly evaluated. Specifically, the inspectors determined that pressurized fire water pipes in both EDG rooms were not classified as safety-related or seismically qualified. The inspectors reviewed Exelons evaluation of the potential failure of the pipe, as assumed in the TMI design and licensing basis, and determined that operator actions were credited to mitigate the pipe failure in order to prevent water from affecting the operation of the EDGs. The inspectors determined that these operator actions could not be performed prior to water from the pipe break impacting the operation of the EDGs. Following identification of the issue, Exelon entered this issue into their corrective action program and performed an analysis on the structural loading on the fire water piping during a safe shutdown earthquake and concluded that the piping would not break during the design basis event and, therefore, the EDGs remained operable. The inspectors reviewed the analysis and found it reasonable. The inspectors determined the failure to adequately evaluate the effects of a pipe failure in the EDG room in accordance with the design and licensing basis was a performance deficiency. The performance deficiency is considered more than minor because it is associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Additionally, the performance deficiency is considered more than minor in accordance with Manual Chapter 0612, Appendix E - Question 3K, in that there was a reasonable doubt of operability for the EDGs requiring engineering calculations and analysis to resolve. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined the finding to be of very low safety significance (Green) because the finding was determined to be a design or qualification deficiency that did not result in an inoperability. No cross-cutting attribute is assigned to this finding because the performance deficiency was not indicative of Exelons current performance. Specifically, this issue was last identified and reviewed by Exelon in issue report 1201424 in 2010.
05000293/FIN-2016003-032016Q3PilgrimFailure to Properly Implement Agastat Control Relays Preventive Maintenance Procedure in Accordance with TS 5.4.1The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1, Procedures, because Entergy did not implement procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Entergy did not implement preventive maintenance procedural requirements to periodically replace six high critical, normally energized Agastat EGP relays every 10 years. Entergys immediate corrective actions included replacing all six relays and performing an equipment apparent cause evaluation. Entergy entered this issue into their CAP as CR-2016-04243. The performance deficiency was more than minor because it was associated with the structures, systems, and components (SSCs) 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. The failure to replace the relays in accordance with preventative maintenance requirements increased the likelihood of failure for safety systems that relied on these relays for operation. The inspectors determined that this finding is of very low safety significance (Green) in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2013, because the performance deficiency did not result in an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined that this finding had no cross-cutting aspect because the most significant causal factor, the failure to include the relays in the preventative maintenance program database, did not reflect current licensee performance. There was no indication that this specific performance deficiency occurred in the last three years.
05000293/FIN-2016003-042016Q3PilgrimFailure to Adequately Evaluate the Effect of Degraded Normally Energized Agastat relays on PCIVs OperabilityThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy did not perform an immediate operability determination and adequately evaluate the operability of primary containment isolation valves (PCIVs) in accordance with procedure EN-OP-104, Operability Determinations/Functionality Assessments, Revision 10. Entergys immediate corrective actions included electrically deactivating two relays, 16A-K17X11 and 16AK18X11. Subsequently, two PCIVs, CV-5065-91 and CV-5065-92, were closed until all six relays were replaced. Entergy entered this issue into the CAP as CR-2016-04753. The inspectors determined that the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Barrier Integrity cornerstone and adversely affected the objective of providing reasonable assurance that physical design barriers protect the public from postulated radionuclide releases caused by accidents or events. Specifically, Entergy did not perform a timely and adequate operability determination as required by procedure. It took Entergy 74 days and four different operability determinations upon discovery of the degraded relays to finally conclude that PCIVs CV-5065-91 and CV-5065-92 were operable. The inspectors determined that this finding is of very low safety significance (Green) in accordance with IMC 0609.04, initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2013, because it did not result in an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Entergy did not initially evaluate the operability of the Agastat relays thoroughly as prescribed in EN-OP-104. Furthermore, Entergy failed to adequately evaluate the effect of the aging Agastat relays pertaining to the PCIVs operability.
05000289/FIN-2016001-022016Q1Three Mile IslandLicensee-Identified ViolationOn February 6, 2016, while making preparations to perform procedure 1303-11.45, PORV Setpoint Check, a senior operator identified that the assigned risk for this planned maintenance activity was inaccurate. Specifically, the risk for the maintenance activity was Yellow, not Green, as originally determined. The reason for the inaccurate risk was due to not previously recognizing the pressurizers block valve (RC-RV-2) would be rendered inoperable during the maintenance activity. This condition could result in failure to operate the pressurizers power operated relief valve. The failure to accurately assess the risk of the power operated relief valve setpoint check was a performance deficiency that was within the licensees ability to identify and correct. The inspectors noted that this maintenance activity had an inaccurate risk assessment for at least the past three years. This performance deficiency was a violation 10 CFR Part 50.65(a)(4), which requires, in part, the licensee to assess and manage the increase in risk that may result from the proposed maintenance activity. Contrary to the above, Exelon failed to accurately assess the risk for the power operated relief valve setpoint check over the past three years. The issue was more than minor because it was associated with the configuration control attribute of the initiating systems cornerstone and it adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, screening criteria. The finding screened to Green because the incremental core damage probability of failing to operate RC-RV-2 is less than 1.00 10-6 per year during the short period which the valve is rendered inoperable during each performance of this maintenance activity. Exelon has entered this issue into its corrective action program (issue report 2622859) and revised the risk assigned to this maintenance activity. Because this finding is of very low safety significance and had been entered into Exelons corrective action program, this violation is being treated as a Green, licensee-identified NCV, consistent with section 2.3.2 of the NRCs Enforcement Policy.
05000289/FIN-2016001-012016Q1Three Mile IslandDeficient Design Control of ECCS Level Transmitter Instrument Line Heat Trace Causes Freezing and InoperabilityA self-revealing NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, was identified for failure to establish and implement adequate design control measures to assure that the borated water storage tank (BWST) was capable of performing its design function to mitigate a design basis loss of coolant accident (LOCA) event. Specifically, Exelon made a modification to the BWST level indicator safety grade heat trace circuit that placed the circuit in an unapproved electrical configuration, which failed to prevent instrument line freezing during cold weather periods, contrary to its safety-function to maintain BWST level indication operable in cold weather. This adversely impacted the availability of a BWST level indication necessary for operators to reliably perform a critical design basis manual action. Exelon documented these issues in issue reports 2609417 and 2611119. Immediate corrective actions included replacement of the affected heat trace and completion of a compatible modification to its electrical configuration. This performance deficiency was more than minor because it was associated with the design control attributes of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the finding was similar to example 2.f in Appendix E of IMC 0612, in that failure to properly maintain cold weather protection equipment for the BWST level transmitters resulted in DH-LT-809 becoming inoperable. The finding was of very low safety significance (Green) because it did not affect design or qualification, did not represent a loss of system function, did not cause at least one train of BWST level instrumentation to be inoperable for greater than its Technical Specification limiting condition of operation (LCO) allowed outage time, and did not involve external event mitigation systems. The finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because station personnel did not follow the heat trace procedure, which did not allow the two types of heat trace to be spliced together.
05000289/FIN-2015004-012015Q4Three Mile IslandFailure to Trend Vibration Data for Safety Related River Water PumpThe inspectors identified a finding of very low safety significance involving an NCV of 10 Code of Federal Regulations (CFR) 50, Appendix B Criterion XVI, Corrective Action Program, because Exelon did not identify and correct a condition adverse to quality on the B nuclear river water pump (NR-P-1B). Specifically, Exelon did not properly evaluate an adverse vibration trend on NR-P-1B, which resulted in exceeding its in-service test (IST) required action level and declared inoperable on October 10, 2015. Exelon entered the condition into their corrective action program (CAP) as issue report 2568763 and emergently replaced the pump, engaged the vendor for short and long term design and material changes to correct the vibration, and created process and peer check corrective actions to ensure all vibration data is reviewed timely and trends are addressed commensurate with their safety significance. The performance deficiency is more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the elevated vibrations reduced the reliability and capability of NR-P-1B to perform its safety function. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, and determined this finding to be of very low safety significance (Green) because the degraded condition was not a design deficiency that affected system operability; did not represent an actual loss of function of a system; did not represent an actual loss of function of a single train or two separate trains for greater than its technical specification (TS) allowed outage time and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the station did not thoroughly evaluate the elevated vibration data such that the issue was addressed before NR-P-1B became inoperable (P.2).
05000247/FIN-2015007-012015Q3Indian PointInadequate design verification that protective device settings do not allow connected Class 1E loads to become damaged or unavailable under normal and sustained degraded voltage conditions during a design basis event.The team identified a finding of very low safety significance involving a non-cited violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion III, Design Control. Specifically, Entergy failed to verify, in design basis calculations for Unit 2, that protective device settings do not allow connected Class 1E loads to become damaged or unavailable during a design basis event: (a) under normal voltage conditions; or (b) for a sustained degraded voltage and subsequent reconnection to the emergency diesel generator concurrent with: (1) a design basis event for the degraded voltage time delay of 8.4 - 11.4 seconds, and (2) a non-accident shutdown for the degraded voltage time delay of 153 - 207 seconds. Additionally, Entergy failed to periodically test the thermal overload relays protecting safety-related motor-operated valves (MOVs) to ensure that degradation or trip setpoint drift does not affect the reliability or availability of mitigating systems when called upon to operate. After identification, Entergy entered this issue into the corrective action program, performed several additional evaluations to verify operability, declared two low pressure injection valves inoperable, and replaced fuses to restore operability to these valves. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, and Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions, and concluded it required a detailed risk evaluation. The detailed risk evaluation was performed by a Region I senior reactor analyst (SRA) and concluded that the postulated inoperability of the two low pressure injections valves resulted in a change in core damage frequency of 1E-7/year, or very low safety significance. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely matter. Specifically, Entergy did not systematically and effectively evaluate NRC Regulatory Issue Summary 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages.
05000247/FIN-2015007-022015Q3Indian PointInadequate design verification that adequate voltages would be available to all Class 1E motors, MOVs, static loads, and MCC control circuits and contactors at the minimum DVR dropout settingThe team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, because Entergy did not verify the adequacy of their electrical design. Specifically, Entergy failed to verify, in design basis calculations and/or periodic testing, that adequate voltages would be available to all Class 1E motors, motor-operated valves (MOVs), static loads, and motor control center (MCC) control circuits and contactors powered from the 480 volt distribution system with the voltage at the 480 volt safety-related switchgear operating at the minimum degraded voltage dropout setting including tolerances. After identification, Entergy entered the issues into the corrective action program and performed several additional evaluations to verify adequate voltage to Class 1E motors, MOVs, static loads, and MCC control circuits. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, and Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, because Entergy did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely matter. Specifically, Entergy did not systematically and effectively evaluate NRC Regulatory Issue Summary 2011-12, Revision 1, Adequacy of Station Electric Distribution System Voltages.
05000286/FIN-2015007-032015Q3Indian PointFailure to account for elevated battery room temperature effects on battery service life.The team identified a finding of very low safety significance (Green) involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, because Entergy did not verify the adequacy of the safety-related battery test program. Specifically, Entergy did not adequately account for the effects of elevated temperature in the immediate vicinity of the No. 33 125 volts, direct current (Vdc) battery to ensure accurate and up-to-date determination of the batterys expected service life, in accordance with the vendor manual. After identification, Entergy entered this issue into the corrective action program and contacted the battery vendor for additional guidance. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a design deficiency confirmed not to result in a loss of operability. This finding was not assigned a cross-cutting aspect because it was a historical design issue not indicative of current performance. Specifically, the associated vendor technical manual guidance was not changed within the last 3 years and there was no recent operating experience that was directly applicable to the performance deficiency.
05000247/FIN-2015007-042015Q3Indian PointLess than adequate corrective actions associated with an evaluation of the seismic adequacy of a 138KV transmission tower located near the Unit 2 EDG building.The team identified a finding of very low safety significance involving a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, because Entergy did not promptly identify and correct a condition adverse to quality. Specifically, in April 2002, Entergy initiated a corrective action condition report (CR) to evaluate and document the seismic adequacy of a 138KV transmission tower, located in close proximity to the Unit 2 emergency diesel generator (EDG) building; however, Entergy staff closed the CR without adequately evaluating and documenting the seismic qualification concern. Entergys short-term corrective actions included initiating a corrective action CR and performing a seismic qualification evaluation. The team determined that the inadequate resolution of the condition adverse to quality is a performance deficiency that was within Entergys ability to foresee and correct. The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, and Appendix E, example 3j, because the engineering calculation error resulted in a condition where there was a reasonable doubt on the operability of a system. In addition, the performance deficiency was associated with the protection against external factors (seismic) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems (the EDGs, in particular) that respond to initiating events to prevent undesirable consequences. The team evaluated the finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2 Mitigating Systems Screening Questions. The finding was determined to be of very low safety significance because it was a qualification deficiency confirmed not to result in a loss of operability. The finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Entergy did not create and maintain complete, accurate, and up-to-date documentation. Specifically, Entergy did not create and maintain complete, accurate, and up-to-date design basis documentation to ensure that an adverse seismic II/I interaction would not result in the loss of the EDG safety function following a seismic event.
05000289/FIN-2015003-012015Q3Three Mile IslandInternal Flooding Licensing Basis Commitment Not MetThe inspectors identified a finding because Exelon failed to meet a commitment made during original licensing to mitigate an internal flooding event. Specifically, Exelon committed to making changes to the fire water supply system to mitigate the impact of a pipe rupture in the auxiliary building. The inspectors identified that the commitment actions were not completed and no changes to the commitment were identified. The inspectors determined that the failure to perform the modifications to the fire service system, as committed to the NRC in a letter dated November 10, 1972, was a performance deficiency that was reasonably within its ability to foresee and correct. Exelon documented the issue in issue report 2544387, performed an immediate operability evaluation, and developed corrective actions to restore compliance with the commitment. The inspectors determined that the performance deficiency is associated with the Mitigating Systems cornerstone attribute of protection against external factors (internal flood hazard) and is more than minor because it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency adversely impacted the operators ability to detect and mitigate a fire service system pipe rupture in the safety related auxiliary building. The inspectors utilized IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, to determine the significance of the performance deficiency. The inspectors determined the finding to be of very low safety significance (Green) because the finding is not a design or qualification deficiency, does not represent a loss of system safety function or loss of a single train for greater than its allowed technical specification time, does not result in the loss of a high safety-significant maintenance rule train and does not involve the loss of function to mitigate internal flooding events. The finding is not assigned a cross-cutting aspect because the performance deficiency occurred during original plant construction and is not indicative of current plant performance.
05000289/FIN-2015002-012015Q2Three Mile IslandFailure to Maintain Turbine Bypass Valve Simulator ModelingA self-revealing NCV of 10 CFR Part 55.46(c), Plant-Referenced Simulators, was identified for Exelons failure to ensure that the plant-referenced simulator demonstrated expected plant response to normal, transient, and accident conditions to which the simulator has been designed to respond. Specifically, Exelon failed to ensure simulator modeling of once through steam generator (OTSG) turbine bypass valve (TBV) operation was consistent with the actual plant which introduced negative operator training and challenged orderly unit shutdown on May 7, 2015. The licensee documented their corrective actions for this issue in TMI issue reports (IR) 02496279 and 2497542, which included software changes to the simulator to reflect actual system design, crew remediation, and procedure changes. The performance deficiency is more than minor because it is associated with the human performance attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the simulator difference introduced negative operator training and, as a result, challenged orderly shutdown of the unit on May 7, 2015. The inspectors evaluated the finding in accordance with NRC Manual Chapter 0609, Significance Determination Process, and the corresponding Appendix I, Licensed Operator Requalification Significance Determination Process. The finding was determined to have very low safety significance (Green) because the impact on operator performance was not during a reportable event. This finding has no cross-cutting aspect assigned because the cause was not representative of current licensee performance. Specifically, the difference in TBV modeling existed since initial simulator certification on June 28, 1990.
05000289/FIN-2015001-012015Q1Three Mile IslandLicensee-Identified ViolationLER 05000289/2014-001-00 describes an unanalyzed condition in which Exelon identified DC motor control circuits were unfused. Specifically, Exelon did not provide overcurrent protection for wiring associated with 250VDC full-voltage control circuits for four non-safety emergency bearing oil pumps in the turbine building to prevent wires from overheating due to fire-induced faults and excessive currents flowing through the cable. With enough current flowing through the cable, the potential exists that the overloaded motor control wiring could damage adjacent control circuit wiring for both instrument air compressors (IA-P-1A/B), which are needed to achieve and maintain post-fire safe shutdown for a fire in the cable spreading room. This condition could result in a loss of the associated safe shutdown components or a secondary fire in another fire area. The failure to protect safe shutdown cables from the effect of postulated fires was a performance deficiency. This performance deficiency was a violation of TMI Operating License Condition 2.C.(4), which requires, in part, post-fire safe shutdown cables remain free of the effects of fire-induced cable faults during postulated fires. Contrary to the above, Exelon identified they failed to meet this requirement and the condition existed since initial construction. The issue was more than minor because it was associated with the protection against external events (fire) attribute of the mitigating systems cornerstone and it adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green), based on IMC 0609, Appendix F, Fire Protection Significance Determination Process, Phase 2 screening criteria. The finding screened to Green based upon task number 2.3.5, and because no credible fire ignition source was determined to adversely affect the motor control circuits of concern as determined. Additionally, a fire area of concern (cable spreading area) is an alternate shutdown fire area protected by detection and an automatic suppression system. The cables in the other fire area of concern (turbine building) are Institute of Electrical and Electronics Engineers 383 (thermoset) construction with steel armor and tied to station ground which decreases the likelihood of inter-cable and intra-cable interactions. Because this finding is of very low safety significance and had been entered into Exelons corrective action program (IRs 1651702, 1658837, 1658842), this violation is being treated as a Green, licensee-identified NCV consistent with the NRCs Enforcement Policy.
05000412/FIN-2014004-012014Q3Beaver ValleyInadequate Plant Startup Procedure Led to Manual Reactor TripA self-revealing NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings was identified for FENOCs failure to have an adequate plant startup procedure. Specifically, 2OM-52.4A, Raising Power from 5% to Full Load Operation, did not adequately address plant startup with one condensate pump in operation. This led to an inability to adequately control steam generator (SG) level when the second condensate pump was started which required the operators to trip the reactor. FENOC is in the process of implementing corrective actions to revise procedure 2OM-52.4A and to address the human performance errors associated with this event. Additionally, FENOC entered the issue into their corrective action program as condition report (CR) 2014-09256. The finding is more than minor because it is associated with the procedure quality and human performance attributes of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the inadequate procedure led to SG level fluctuations that could not be adequately controlled when the second condensate pump was started, and required the operators to trip the reactor. The inspectors determined that this finding is of very low safety significance (Green), because while it did result in a reactor trip, it did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of a trip to a stable shutdown condition. The finding has a cross-cutting aspect in Human Performance, Challenge the Unknown, because FENOC operators did not stop when faced with uncertain conditions. Specifically, the adequacy of the procedure was not sufficiently questioned when the plant was not in the normal start up configuration of two running condensate pumps nor later when the condensate pump discharge header pressure low alarm occurred.
05000289/FIN-2014004-012014Q3Three Mile IslandInadequate Evacuation Time Estimate SubmittalsThe inspectors identified an NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2), 10 CFR 50.47(b)(10), and 10 CFR 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Three Mile Island Nuclear Station (TMI) emergency plan as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations (OROs) by the required date. Upon identification, Exelon entered this issue into its corrective action program (CAP) as issue reports (IRs) 1525923 and 1578649. Exelon submitted a third ETE for TMI on April 4, 2014, and the NRCs review of that ETE is documented in section 1EP4 of this report. The finding is more than minor because it is associated with the Emergency Preparedness cornerstone attribute of procedure quality and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The ETE is an input into the development of protective action strategies prior to an accident and to the protective action recommendation decision making process during an accident. Inadequate ETEs had the potential to reduce the effectiveness of public protective actions implemented by the OROs. The finding is determined to be of very low safety significance (Green) because it is a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The cause of the finding is related to cross-cutting aspect of Human Performance, Documentation, because Exelon did not appropriately create and maintain complete, accurate and, up-to date documentation (H.7).
05000289/FIN-2014003-012014Q2Three Mile IslandRisk Mitigation Actions Not Performed for Excavation of Nuclear River System Cable ConduitsThe inspectors identified a finding of very low safety significance (Green) involving a non-cited violation (NCV) of 10 CFR Part 50.65(a)(4), Requirements for monitoring the effectiveness of maintenance at nuclear power plants, because Exelon did not implement risk management actions (RMAs) to manage risk associated with the nuclear service river pump B (NR-P-1B) during excavation for fire service piping replacement. Specifically, the excavation exposed a cable conduit duct bank containing safety-related cables for nuclear service river valve 1B (NR-V-1B) without having adequate RMAs in place to ensure NR-V- 1B cabling would remain protected from a tornado generated missiles. Exelon entered the condition into their corrective action program as IR 1670876 and took immediate corrective actions to modify the work instructions to include RMAs for soil restoration over the conduit duct bank in the event of a tornado. The performance deficiency is more than minor because it is associated with the Protection Against External Factors attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstones objective to ensure the availability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the findings using IMC 0609.04, Initial Characterization of Findings. The finding involved the licensees management of risk in accordance with 10 CFR 50.65(a)(4) therefore, the inspectors evaluated the significance using IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process. The inspectors determine that this performance deficiency was of very low safety significance (Green) because the finding was associated with RMAs only and the incremental core damage probability (IDCP) was not >1E-6. This finding has a cross-cutting aspect in the area of Human Performance, Work Management; because Exelon did not manage risk associated with the underground piping replacement project and did not effectively communicate job activities between work groups to ensure the RMAs would be implemented as required.
05000289/FIN-2014003-022014Q2Three Mile IslandUFSAR Max Hypothetical Dose Not Updated, Consistent with Current Plant ConditionsThe inspectors identified a Severity Level lV (SL-lV) NCV of 10 CFR 50.71(e), Maintenance of Records, Making of Reports, because TMI personnel did not update the Updated Final Safety Analysis Report (UFSAR) with information consistent with plant conditions. Specifically, TMI personnel did not remove reference to or correct information in UFSAR Section 14.2.2.3.4.a, Environmental Analysis of Loss of Coolant Accidents - Consequences of LOCA Radioactive Releases to the Environment, to reflect current plant conditions with regard to maximum hypothetical accident doses at the main control room, exclusion area boundary, or low population zone. Exelon documented this in issue report 1662515 to address the UFSAR discrepancy. This issue was determined to be within the traditional enforcement process because it had the potential to impede or impact the NRC's ability to perform its regulatory functions. Specifically, the issue was determined to have a material impact on licensed activities and was considered more than minor using section 7.3.D of the NRC Enforcement Manual. Using example d.3 of section 6.1 of the NRC Enforcement Policy, the inspectors determined that the violation was a SL-IV violation because the erroneous information was not used to make an unacceptable change to the facility or procedure. In accordance with inspection manual chapter 0612, section 07.03c, this traditional enforcement violation was not assigned a cross-cutting aspect.
05000289/FIN-2014002-012014Q1Three Mile IslandFailure to Perform a 10 CFR 50.59 Evaluation for the BWST Seismic QualificationsThe inspectors identified a Severity Level IV (SL-IV), Non-Cited Violation of 10 CFR 50.59, Changes, Tests, and Experiments, and an associated finding of very low safety significance (Green) for Exelons failure to perform a 50.59 evaluation review to determine whether a license amendment was required to align the borated water storage tank (BWST) to non-seismic piping. Specifically, Exelon staffs 50.59 screening accepted the alignment of the seismically qualified BWST to a non-seismically qualified clean-up system. The inspectors determined the alignment would involve a change to the BWST that adversely affects its Updated Final Safety Analysis Report chapter 5.1.1, Classes of Structures and Systems for Seismic Design, described design function of being seismically qualified. Additionally, the inspectors determined that following the 50.59 review Exelon placed the line-up in service. The inspectors determined these two actions were performance deficiencies that were reasonably within Exelons ability to foresee and prevent. Furthermore, the 50.59 screening credited unapproved operator manual actions to ensure functionality of the BWST. Exelon documented this as issue report 1631468 and implemented interim corrective actions to isolate the BWST from the clean-up system until a permanent resolution is determined and implemented. The inspectors determined the 50.59 violation regarding the failure to perform an evaluation was more than minor because the inspectors could not reasonably determine that the alignment would not have ultimately required NRC prior approval, because the BWST alignment was not in accordance with the current licensing basis and the evaluation credited the use of unapproved operator manual actions. The inspectors also determined that the performance deficiency of accepting and aligning the adverse clean-up line-up, challenging the BWST seismic qualification, was more than minor because it adversely affected the Mitigating Systems cornerstone objective to ensure the availability, reliability, and capability 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 Significance Determination Process for Findings At-Power, and determined that this finding required a detailed risk evaluation. The detailed evaluation was performed which determined that the performance deficiency was a finding of very low safety significance (Green). Additionally, In accordance with Section 6.1.d.2 of the NRC Enforcement Policy, the 50.59 violation is categorized as a Severity Level IV. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, in that the station did not effectively evaluate and internalize relevant external operating experience (Information Notice (IN) 2012-01) regarding connections between safety-related seismic and non-seismic qualified piping and components.
05000289/FIN-2014002-022014Q1Three Mile IslandLoss of Air Intake Tunnel Sump Pump Function due to Inadequate Work ExecutionThe inspectors identified a finding of very low safety significant (Green) for Exelons failure to follow work order instructions in accordance with MA-AA-716-011, Work Execution and Close Out, during planned maintenance activities on the air intake tunnel (AIT) deluge sump pump (SD-P-7). Specifically, in May 2013, a maintenance worker applied epoxy to the sump pumps float switch contrary to work order instructions. Inspectors identified that the float switch was fixed in the OFF position, rendering the pump unavailable, during a system walkdown in March 2014. Exelon documented this as issue report 1628577 and performed prompt corrective actions to remove the epoxy coating from the float switch. In addition, corrective actions were performed to replace the float ball that likely was submerged and filled with water as a result of the float switch being stuck. Exelon successfully postmaintenance tested the float switch and pump on March 6, 2014, and returned it to service. The inspectors determined the performance deficiency associated with this finding involved Exelons failure to follow work order instructions in accordance with MA-AA-716-011, Work Execution and Close Out, during planned maintenance activities on SD-P-7 was more than minor because it was associated with mitigating systems cornerstone adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, in May 2013, a technician applied epoxy to SD-P-7s float switch, contrary to work order instructions, rendering the pump non-functional. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 4, External Events Screening Questions, and determined, based on operator response to an air intake tunnel deluge alarm, this finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance because the worker did not follow work order instructions and incorrectly applied epoxy to the SD-P-7 float switch assembly, rendering the pump non-functional and unavailable (H.8).
05000289/FIN-2014002-032014Q1Three Mile IslandFailure to Restore Station Blackout Diesel Generator Cooling Water Lineup following Maintenance and Testing ActivitiesA self-revealing non-cited violation (NCV) of 10 CFR 50.63, Loss of All Alternating Current Power, was identified for Exelons failure to properly restore the station blackout (SBO) diesel generator system following maintenance and testing activities, rendering the SBO diesel generator unable to be available in 10 minutes of and cope for 4 hours after a postulated SBO event. Specifically, during the restoration from SBO switchgear maintenance during the previous Fall 2013 refueling outage, operators failed to remove a blocking device (gag) from the SBO diesel generator fire service water cooling isolation valve (FS-V-646) as part of its restoration to an automatic, standby configuration. As a result the SBO diesel generator was not in the configuration required by 10 CFR 50.63 (c)(2), which describes acceptable capability standards for alternate AC power systems. Exelon entered this issue into their corrective action program as IR 1608625. Exelon restored the valve configuration and revised affected and related procedures. The inspectors determined this performance deficiency in that Exelon failed to remove the blocking devise from FS-V-646 prior to restoring the SBO diesel to service was more than minor because it is associated with the mitigating systems affecting the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, in the event of a station blackout, the SBO diesel generator was not able to be started and operated from the control room with no local operations required to allow the prompt restoration of electrical power to at least one vital bus as assumed in the TMI SBO analysis. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, and determined that this finding required a detailed risk evaluation because, with FS-V-646 gagged, the SBO diesel was not capable of performing its safety function. The detailed risk evaluation determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Documentation, because Exelons procedure for restoration from the maintenance and testing (OP-TM-731- 510, Rev. 5) was not adequate to specify actions to return the cooling water isolation valve (FS-V-646) to its normal automatic condition (H.7).
05000289/FIN-2013005-012013Q4Three Mile IslandImproper Storage of Material in Reactor BuildingThe inspectors identified a Green non-cited violation of Technical Specification 6.8.1 for Exelons failure to implement procedure requirements governing storage of equipment in Class 1 structures. Specifically, Exelon stored unsecured material, one (1) roll of plastic sheeting and three (3) plastic sheets, in the Reactor Building (RB) during power operations, contrary to Exelon Procedure 1015, Equipment Storage Inside Class 1 Buildings. This resulted in unsecured material in a location that had the potential, during a large break loss of coolant accident, to be transported to and adversely impact the performance of the emergency core cooling system (ECCS) suction sump. Exelon documented the issue in their corrective action program under issue report (IR) 1577437 and took immediate corrective actions to remove the unsecured plastic from the RB. This finding is more than minor because it is associated with the availability and reliability attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the unsecured plastic had the potential to impact the reliability and availability of the ECCS recirculation suction flow path, due to the potential increased debris loading. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 2, and determined this finding is very low safety significance (Green) because the degraded condition is a design deficiency that affects system operability, but did not represent an actual loss of function of a system; did not represent an actual loss of function of a single train or two separate trains for greater than its technical specification allowed outage time and did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety significant. The finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon did not take adequate corrective actions to address the cause of improperly staged material in the RB (IR 1577100), resulting in a subsequent recurrence of improper staging of additional material in the RB identified by the inspectors (IR 1577437).
05000289/FIN-2013005-022013Q4Three Mile IslandFailure to Perform Leak Rate Testing on Containment Boundary PipingThe inspectors identified a Green non-cited violation of 10 CFR 50, Appendix J, Primary Reactor Containment Leakage Testing for Water-Cooled Power Reactors, for Exelons failure to establish an adequate program that leak tested components penetrating the primary containment pressure boundary. Specifically, Exelon failed to implement leak rate testing of the reactor building (RB) normal closed loop cooling piping to verify piping integrity to support its containment isolation function. As a result, on November 10, 2013, engineering personnel identified an inoperable containment isolation boundary due to a degraded RB closed cooling piping condition. Exelon documented this issue in issue report (IR) 1598590 and took corrective actions to revise the Appendix J test program and address the missed leak rate surveillance test. This finding is more than minor because it is associated with the Barrier Performance attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical barriers, as designed, protect the public from radionuclide releases caused by accidents or events. Specifically, Exelon failed to perform leak rate testing of the RB normal closed loop cooling piping and failed to identify the degraded piping condition that impacted the containment isolation function. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding did not represent an actual open pathway in the physical integrity of the reactor containment isolation system nor did it involve an actual reduction in function of hydrogen recombiners for the reactor containment therefore, the finding was of very low safety significance (Green). The finding was not assigned a crosscutting aspect because the most significant causal factor of the finding was the failure to implement leak rate testing since 1991 and was not indicative of current plant performance.
05000289/FIN-2013005-032013Q4Three Mile IslandLicensee-Identified ViolationTMI License Condition 2.C.(4) and the Fire Hazards Analysis Report (FHAR) require administrative breaker position restrictions for Appendix R valves needed for safe shutdown, including reactor coolant pump #1 seal bypass valve (MU-V-38). TMI procedure AP-1038 is the implementing procedure for License Condition 2.C.(4) and the FHAR. On November 28, 2013, the licensee identified the breaker for MU-V-38 to be in the ON position contrary to the required OFF position. It was determined that the breaker was in the incorrect position for six (6) days of the seven (7) days allowed by the AP-1038 time clock. In that, compensatory measures and a risk assessment were not in place for this out-of-position breaker, in the event of a postulated fire and fire-induced spurious operation of MU-V-38, and the inspectors determined the issue was more than minor. The cause of the mispositioned breaker was determined by Exelon to be auxiliary operator distraction from multiple work activities and failure to restore the breaker to its expected position following post-maintenance testing during the fall refueling outage. Exelon entered this issue into their CAP as IR 1591314 and promptly positioned the breaker to the correct OFF position, including validation of the position of the remaining Appendix R breakers. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix F, Fire Protection Significance Determination Process, based upon this Fire Prevention and Administrative Controls issue having a low degradation category.
05000336/FIN-2013010-012013Q3MillstoneInadequate Alternative Shutdown ProcedureThe team identified an apparent violation of Millstone Unit 2 Operating License Condition 2.C. (3) for failure to implement and maintain all aspects of the approved Fire Protection Program (FPP). Specifically, Dominion had not adequately implemented an alternative shutdown procedure, as required by 10 CFR 50 Appendix R Section III.L.3 and the approved FPP. The procedure for a Unit 2 fire which could lead to control room abandonment did not ensure the electrical distribution system was correctly configured prior to re-energizing AC buses. As a result, an over-current condition could occur and trip the 4kV supply breaker complicating safe shutdown operations and delaying AC bus recovery. In response to this issue, Dominion promptly revised their fire safe shutdown operating procedure prior to the end of the inspection to correct this deficiency. This finding was more than minor because it was associated with the Protection Against External Factors (e.g., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The team performed a Phase 1 Significance Determination Process (SDP) screening in accordance with NRC Inspection Manual Chapter 0609, Appendix F, and Fire Protection Significance Determination Process. This finding affected the post-fire safe shutdown category, and was determined to have a high degradation rating because the alternative shutdown procedure lacked adequate instructions to ensure correct equipment alignment. Therefore, the team concluded that a more appropriate and accurate characterization of the risk significance of this issue would be obtained by performing a Phase 3 SDP analysis because the Phase 2 SDP analysis does not explicitly address alternative safe shutdown fire scenarios. The Phase 3 SDP analysis cannot be accurately calculated until additional cable routing and ignition source information is presented by Dominion and is necessary to develop the fire scenarios that would require the alternative shutdown procedure to be implemented. This finding did not have a cross-cutting aspect because it was a legacy issue and was considered to not be indicative of current licensee performance.
05000336/FIN-2013010-032013Q3MillstoneFailure to Maintain Cold Shutdown Material On-siteThe team identified a finding of very low safety significance, involving a non-cited violation of Millstone Unit 2 Operating License Condition 2.C. (3) and Unit 3 Operating License Condition 2.H for the failure to implement and maintain all aspects of the approved Fire Protection Program. Specifically, Dominion used large motors, pre-staged in the on-site warehouse for Appendix R cold shutdown (CSD) repairs, as spare parts to accomplish preventative maintenance tasks. As a result, Dominion could not have performed the designated CSD repairs and achieved CSD conditions within 72 hours as required for both Units 2 and 3 during the time period that the old motors were off-site for refurbishment. In addition, Dominion had not taken any compensatory measures to reduce the likelihood of a fire or its consequence, in lieu of not having required repair material on-site. Dominion entered these issues into its corrective action program as condition reports 522722, 522740, 522848, and 522850 and has planned corrective actions to ensure CSD repair material is never intentionally made unavailable or removed from the site. This finding was more than minor because it was associated with the Protection Against External Factors (e.g., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. The team performed a Significance Determination Process (SDP) screening, in accordance with NRC Inspection Manual Chapter 0609, Appendix F, and Fire Protection Significance Determination Process. This finding screened to very low safety significance in Phase 1 of the SDP because it only affected the ability to reach and maintain cold shutdown conditions. This finding did not have a cross-cutting aspect because it was a legacy issue and was considered to not be indicative of current licensee performance.
05000336/FIN-2013010-022013Q3MillstoneSpurious Operation of Pressurizer Spray Valves Not AnalyzedThe team identified a finding of very low safety significance involving a non-cited violation of Millstone Unit 2 Operating License Condition 2.C. (3) for failure to implement and maintain all aspects of the approved Fire Protection Program. Specifically, Dominion\'s safe shutdown methodology postulated spurious operation of the pressurizer spray valves, but had not analyzed the effect of the spurious operations and mitigation actions were not implemented to ensure operators could achieve safe shutdown if the spray valves spuriously opened. In response to this issue, Dominion revised their fire safe shutdown operating procedure prior to the end of the inspection to mitigate spurious opening of the spray valves. The finding was more than minor because it was similar to Example 3.k of NRC Inspection Manual Chapter (IMC) 0612, Appendix E, and was associated with the Protection Against External Factors (e.g., Fire) attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective to ensure the reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The team performed a Significance Determination Process (SDP) screening, in accordance with IMC 0609, Appendix F, and Fire Protection Significance Determination Process. This finding affected the post-fire safe shutdown category, and was determined to have a low degradation rating because a subsequent evaluation determined that that the performance requirements of Appendix R Section III.L.1 were satisfied. This finding did not have a cross-cutting aspect because it was a legacy issue and was considered to not be indicative of current licensee performance.
05000289/FIN-2013002-012013Q1Three Mile IslandFailure to Maintain Combustible Loading Near the B CST within Fhar LimitsThe inspectors identified a Green non-cited violation (NCV) of license condition DPR-50 section 2.C.(4), Fire Protection, for Exelons failure to maintain transient combustible loading within fire loading limits near the B condensate storage tank (CST). Specifically, on January 9, the inspectors identified a Portable On-Demand storage (POD) container staged within 50 feet of the B CST. The POD and its contents contained transient combustible materials in excess of the allowed fire loading in accordance with the fire hazards analysis report (FHAR). Exelon promptly removed the POD container and restored transient combustible loading within allowable limits. Exelon entered this issue into their corrective action program under issue report (IR) 1461029. Corrective actions included additional postings around the safety-related above-ground tanks, site-wide notifications and the performance of a root cause evaluation to address recent station fire protection issues. This performance deficiency is more than minor because it is associated with the Protection Against External Factors (Fire) attribute and adversely affected the Mitigating Systems cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, it was determined to be more than minor since it is similar to more than minor example 4.k of IMC 0612, Power Reactor Inspection Reports, Appendix E because the fire loading was not within the FHAR limits. In accordance with IMC 0609.04, Phase 1 Initial Screen and Characterization of Findings, the inspectors determined the finding affected the administrative controls for transient combustible materials. Therefore, the inspectors conducted a phase 1 SDP screening using IMC 0609, Appendix F, Fire Protection Significance Determination Process, and the inspectors determined that the finding affected the category of Fire Prevention and Administrative Controls in that combustible material was not being properly controlled, the finding had a low degradation rating, and the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon failed to thoroughly evaluate and take appropriate corrective actions for similar transient combustible loading issues such that the cause and extent of condition were fully addressed.
05000289/FIN-2012005-032012Q4Three Mile IslandFailure to Identify and Correct Missing Electrical Conduit Flood Seals in the Air Intake TunnelThe inspectors identified an apparent violation (AV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, was identified during the TI-187 flooding walkdowns for Exelons failure to identify and correct an external flood barrier deficiency. Specifically, Exelon failed to identify and correct, during external flood barrier walkdowns, that electrical cable conduits were not flood sealed in the Air Intake Tunnel (AIT), as designed, to maintain the integrity of the external flood barrier. The deficiency was entered into Exelons corrective action process and permanent corrective actions were taken to seal the electrical conduits and restore the external flood barrier integrity. The finding was determined to be more than minor because it is associated with the protection against external factors 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, Exelon failed, during multiple focused walkdowns, to identify the degraded external flood barrier in the Crouse-Hinds couplings in the AIT that challenged the external flood barrier operability. The significance of the degraded external flood barrier is to be determined and cannot accurately be calculated until additional testing and analysis of the as-found configuration is complete. Specifically, Exelon is performing additional testing on the capability of as-found foam fire sealant material, present in the conduits at the AIT/Aux Building interface, to mitigate flood water entry into the safety-related structures. These results will be an input into the licensees flood mitigation aggregate impact review. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon failed to review the external flood barrier with a low threshold for identifying issues which resulted in the failure to identify the unsealed electrical conduits in the AIT in a timely manner commensurate with its safety significance.
05000289/FIN-2012005-012012Q4Three Mile IslandAdequacy of Seismic Gap Flood SealThe inspectors identified a non-cited violation (NCV) of General Design Criterion 2, Performance Standards, because Exelon had not established measures to ensure that the seismic gap flood seal was adequate to remain watertight during a probable maximum flood (PMF) event, as required by the TMI design. Specifically, the design requirement for the seismic gap seal specified that it was to be watertight. However, the installed seal configuration had measurable leakage when tested. The inspectors determined that the failure to construct, maintain, and inspect the seismic gap flood seal consistent with its design (e.g., watertight) was a performance deficiency within Exelon\'s ability to foresee and prevent. Exelon entered this issue into their corrective action program, took appropriate interim corrective actions, and completed permanent modifications to restore the watertight function of the seismic gap barrier. This finding was more than minor because it was similar to the more than minor example 3.j in Inspection Manual Chapter (IMC) 0612 Appendix E, Examples of Minor Issues, in that the seal\'s as-built and maintained configuration resulted in a condition where there was reasonable doubt regarding the functionality of the seismic gap seal to remain watertight during a PMF event. Also, this finding was associated with the protection against external factors attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, the inspectors performed a bounding risk evaluation using an unavailability period of greater than one year for the watertight seal, and determined this finding was of very low safety significance (Green). This finding has a cross-cutting aspect, as described in IMC 0310, in the area of Human Performance, Decision Making, because Exelon failed to verify the validity of underlying assumptions or continued functionality of the seismic gap flood seal following an external flood re-analysis which revised the design basis PMF conditions.
05000289/FIN-2012005-022012Q4Three Mile IslandFailure to Identify and Correct Licensing Basis Flood Barrier and Support Equipment Deficiencies in Intake Screen and Pump HouseThe inspectors identified a non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, in that Exelon failed to identify and correct conditions adverse to quality regarding the licensing basis external flood barrier integrity. Specifically, Exelon failed to identify and correct 13 unsealed penetrations through the Intake Screen and Pump House (ISPH) flood barrier and multiple deficiencies that challenged the fulfillment of ISPH support equipment capability to maintain the integrity of the licensing basis flood barrier. The deficiencies were entered into the corrective action program and permanent corrective actions were taken to seal the penetrations to restore the external flood barrier integrity and restoration of the support equipment capability for flood protection. The finding was more than minor because it is associated with the protection against external factors 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, Exelon did not identify and correct 13 unsealed penetrations in a licensing basis external flood barrier and its associated support equipment deficiencies such that the barrier is fully capable of maintaining the ISPH free of flood water. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, Exhibit 2 Mitigating Systems Screening Questions and Exhibit 4 External Events Screening Questions and determined that a detailed risk evaluation was required based upon the assumed complete failure of the flood barrier would degrade two trains of Decay Heat Removal. A detailed risk evaluation modeled in SAPHIRE 8 using the TMI SPAR model version 8.18 determined the finding to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon failed to identify the unsealed penetrations through the flood barrier and multiple deficiencies in supporting equipment in a timely manner commensurate with its safety significance.
05000289/FIN-2012004-022012Q3Three Mile IslandRemote Shutdown Relay 69X1RR Contact FailureOn December 22, 2011, the reactor building emergency cooling water pump discharge valve B (RR-V-1B) failed to open during performance of an engineered safeguards actuation system (ESAS) quarterly surveillance test. TMI declared RR-V-1B inoperable and entered a 72 hour limited condition for operation (LCO) in accordance with technical specification (TS) 3.3.2. TMI performed troubleshooting and determined that the remote shutdown (RSD) transfer selector switch relay (69X1RR), which is in series with the ESAS signal, exhibited intermittent contact make-up. The function of the RSD selector switch and associated relay (69X1RR) is to transfer control of RR-V-1B from the main control room to the RSD panel. The relay (69X1RR) was found in the open-state thus, inhibiting the ESAS actuation signal. The selector switch relay was cycled until proper contact make-up was achieved. TMI applied administrative controls to ensure the transfer switch relay (69X1RR) contact closed properly if the RSD transfer switch was manipulated prior to the relay replacement. RR-V-1B was successfully tested and declared operable on December 22, 2011. The relay transfer switch was replaced and tested satisfactorily on January 6, 2012. The last successful RSD functional test of RR-V-1B had been completed on November 10, 2011 during refueling outage T1R19. TMI concluded that the relay (69X1RR) had most likely not fully re-closed at the completion of the test. Thus, TMI determined that RR-V-1B was inoperable from November 10, 2011 through December 22, 2011. This issue constituted two violations of NRC requirements. Namely, a) the licensee made the reactor critical on November 24, 2011 (while starting up from T1R19), without all engineered safeguards valves associated with the reactor building emergency cooling system being operable as required by TS 3.3.1; and, b) and that RR-V-1B was inoperable for more than 72 hours, and the unit wsa no paced in a hot shutdown condition within 6 hours, as required by TS 3.3.2. However, the NRC concluded that it was not reasonably within the licensees ability to foresee and correct the relay failure that caused these violations. Specifically, the failure analysis of the relay identified that an unforeseen manufacturing defect caused the failure and that the relay exhibited no visual abnormalities or indications to the licensee that a defect existed prior to its failure. In addition, the NRC identified that no significant plant or industry operating experience existed on this style relay that would have alerted the licensee to this potential issue. Therefore, the NRC did not identify any performance deficiency associated with the violations. Inspection Manual Chapter (IMC) 0612, Appendix B, Issue Screening, directs disposition of this issue in accordance with the Enforcement Policy because there was no performance deficiency associated with the violations. The inspectors used the enforcement policy, Section 6.1, Reactor Operations, to evaluate the significance of this violation. The inspectors concluded that the violation is more than minor and best characterized as Severity Level IV (very low safety significance) because it is similar to Enforcement Policy Section 6.1, example d.1. Additionally, the inspectors assessed the risk associated with the issue by using IMC 0609, Appendix A, SDP For Findings at Power. The inspectors screened the issue, and evaluated it using Exhibit 3 of IMC 0609, Appendix A. Evaluating the criteria under the Barrier Integrity cornerstone, the finding did not represent an actual open pathway in the physical integrity of the containment and did not involve a reduction in function of hydrogen igniters in the reactor containment. Based on these reviews, the issue would screen as very low safety significance (Green). Because it was not reasonable for TMI to have been able to foresee and prevent the relay failure, the NRC determined no performance deficiency existed. Thus, the NRC has decided to exercise enforcement discretion in accordance with Section 3.5 of the NRC Enforcement Policy and refrain from issuing enforcement action for the violation (EA-12-164). Further, because the licensees action and/or inaction did not contribute to this violation, it will not be considered in the assessment process or the NRCs Action Matrix. This LER is closed.
05000289/FIN-2012004-012012Q3Three Mile IslandFailure to Maintain Combustible Loading in the Bwst Tunnel within Fhar LimitsThe inspectors identified a Green non-cited violation (NCV) of license condition DPR- 50, section 2.C.(4), Fire Protection, for Exelon storing transient combustibles in excess of the fire loading allowed near the borated water storage tank (BWST). Specifically, on July 11, the inspectors identified eight bags of trash/transient combustible materials stored within 50 feet of the BWST which is in excess of the allowed fire loading in accordance with the Fire Hazards Analysis Report (FHAR) and transient combustible control program. The inspectors determined that the failure to maintain combustible loading in the BWST tunnel within the FHAR limits was a performance deficiency that was within Exelons ability to foresee and correct. Exelon promptly removed the improperly stored transient combustibles and entered the performance deficiency into their corrective action program as issue report 1388097. Corrective actions were implemented to alert technicians of the restrictions on transient combustible materials near the BWST. This finding was determined to be more than minor since it is similar to more than minor example 4.k of Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix E, because the fire loading was not within the FHAR limits. In accordance with Inspection Manual Chapter (IMC) 0609.04, Phase 1 Initial Screen and Characterization of Findings, the inspectors determined the finding affected the administrative controls for transient combustible materials. Additionally, the inspectors determined that this issue was more than minor because it affected the protection against external events attribute of the mitigating systems cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors conducted a phase 1 SDP screening using IMC 0609, Appendix F, Fire Protection Significance Determination Process, and the inspectors determined that the finding affected the category of Fire Prevention and Administrative Controls in that combustible material was not being properly controlled, the finding had a low degradation rating, and the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Resources, because Exelon failed to appropriately ensure interdepartmental coordination during the work activities such that the transient combustibles were promptly removed from the BWST tunnel.
05000289/FIN-2012002-012012Q1Three Mile IslandFailure to Implement Fire Protection Compensatory Actions for Out-of-Service Appendix R Heat ExchangerThe inspectors identified a non-cited violation of license condition DPR-50 section 2.C.(4), Fire Protection, for Exelons failure to implement compensatory actions during planned maintenance on the A nuclear service heat exchanger (NS-C-1A). Specifically, on May 10, 2010, Exelon failed to return Appendix R breakers to their correct position within the seven day allowed outage time and implement compensatory actions in accordance with administrative procedure (AP) 1038, Fire Protection Program. The inspectors determined Exelons failure to implement compensatory actions during planned maintenance on NS-C-1A in accordance with AP 1038 was a performance deficiency that was within Exelons ability to foresee and correct. Exelon performed an extent of condition review and created a requirement to review the fire hazard analysis report for applicability before removing equipment from service. Exelon has entered this issue in the corrective action program for resolution as IR 1347403. This finding is 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 availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with Inspection Manual Chapter 0609.04, Phase 1 Initial Screen and Characterization of Findings, the inspectors conducted a phase 1 SDP screening using Appendix F, Fire Protection Significance Determination Process, and determined that a detailed phase 2 analysis was required due to the elevated calculated delta core damage frequency. The inspectors performed a detailed walkdown of the control cables associated with the nuclear river system valves and identified no fire ignition sources and concluded that the finding was very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Resources, because Exelon failed to ensure complete, accurate, and up-to-date procedures were used to determine if compensatory actions were required for planned work activities
05000289/FIN-2012002-022012Q1Three Mile IslandInadequate Inspection of RCP FlangesThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, because Exelon did not specify, in writing, the exact inspection scope and criteria for boric acid inspections of the reactor coolant pump bolted flanges during plant refueling outages. Lack of specific procedural guidance contributed to the failure to detect reactor coolant system leakage from the thermal barrier flange of the B reactor coolant pump (RC-P-1B) prior to November 2011. Exelons failure to ensure that both the upper and lower RCP thermal barrier flanges were visually inspected for the complete 360 degrees for all RCPs is a performance deficiency within Exelons ability to foresee and prevent. Exelon completed a boric acid evaluation which showed there was reasonable assurance that the flange could safely operate until the next refueling outage. Additionally, Exelon prepared an adverse condition monitoring plan and is performing periodic remote monitoring of the affected flange for changes in leakage from the degraded gasket. Exelon entered this issue into the corrective action program as IR 01344561. The finding is more than minor because it is associated with the Equipment Performance attribute (a degraded RCP flange gasket) of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Also, this finding is similar to the more than minor example 4.a in Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix E. The inspectors completed IMC 0609.04, Phase 1- Initial Screening and Characterization of Findings, and screened the finding as very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Control, because Exelon did not ensure supervisory and management oversight of work activities, including contractors, such that nuclear safety is supported
05000289/FIN-2012002-032012Q1Three Mile IslandInadequate System Monitoring Results in Multiple IA-P-4 TripsA self-revealing finding was identified for inadequate performance monitoring of instrument air compressor number four (IA-P-4) in accordance with ER-AA-2003, System Performance Monitoring and Analysis. Specifically, performance monitoring action levels established for loaded and unloaded times in procedure 1104-25, Instrument and Control Air System, were not adequate to identify the adverse trend in performance and resulted in recurring drive-motor overload trips and unplanned accrued unavailability of IA-P-4 on September 28, October 8 and November 29, 2011. Maintenance technicians repaired the air leaks and subsequent IA-P-4 air loading decreased. Corrective actions were implemented to trend loaded and unloaded times of IA-P-4 in the system monitoring plan and implement acoustic monitoring for identification of system air leakage (IR 1295235). This finding is more than minor because it was associated with the equipment 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 shutdown as well as power operations. In accordance with Inspection Manual Chapter 0609.04, Phase 1 - Initial Screen and Characterization of Findings, the inspectors conducted a phase 1 SDP screening and determined that a detailed phase 2 evaluation was required to assess the safety significance because the finding contributed to both the likelihood of a reactor trip and the likelihood that mitigation equipment would not be available. The inspectors consulted a senior risk analyst (SRA) to perform a detailed phase 2 analysis. The SRA performed a bounding risk analysis using five days of IA-P-4 unavailability. The phase 2 analysis concluded that the significance of the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Exelon failed to thoroughly evaluate the cause of the IA-P-4 trips such that the resolution addressed the cause
05000289/FIN-2012002-042012Q1Three Mile IslandFailure to Perform an Adequate Maintenance Risk Evaluation for DH-V-3 Planned MaintenanceThe inspectors identified a non-cited violation of 10 CFR 50.65 (a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for Exelons failure to adequately assess and manage the impact to plant risk during a planned maintenance activity. Specifically, Exelon did not recognize an elevated online maintenance risk activity and implement appropriate risk management actions (RMAs) during maintenance on the decay heat removal (DHR) drop line valve (DH-V-3) on January 16, 2012. The inspectors determined that the failure to perform an adequate risk assessment and implement appropriate RMAs for the planned maintenance on DH-V-3 is a performance deficiency that was within Exelons ability to foresee and correct. Immediate corrective actions included operator and work planning training on risk evaluations and an extent of condition review to ensure planned maintenance activities that could impact DHR system operability were identified. Exelon entered this issue into the corrective action program for resolution as IR 1314551. This finding was determined to be more than minor since it is similar to more than minor example 7.e of Inspection Manual Chapter (IMC) 0612, Power Reactor Inspection Reports, Appendix E because the risk assessment, when adequately performed, resulted in an elevated station risk condition and required RMAs. The finding was evaluated in accordance with Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, of IMC 0609, Significance Determination Process . The inspectors, in consultation with a senior risk analyst, performed a phase 1 analysis and concluded that the incremental core damage probability deficit for DH-V-3 with an out-ofservice time of 8 hours was less than 1E-6. Therefore, the finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Control, because Exelon failed to incorporate appropriate risk insights into the planning and execution of the DH-V-3 maintenance activity
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