Semantic search

Jump to navigation Jump to search
 QuarterSiteTitleDescription
05000352/FIN-2018003-022018Q3LimerickFailure to Correct Adverse Environmental Conditions Impacting Low Pressure Coolant Injection Outboard Primary Containment Isolation ValveA self-revealed Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI was identified when Exelon failed to correct adverse environmental conditions affecting the Unit 1 LPCI outboard PCIV actuator that resulted in long term water intrusion, corrosion, and failure of the valve to stroke closed.
05000352/FIN-2018003-012018Q3LimerickFailure to Assess and Manage Risk Associated with Fuel Oil Storage Tank MaintenanceAn NRC-identified Green NCV of 10 CFR 50.65(a)(4) was identified when Exelon failed to assess and manage risk associated with fuel oil storage tank maintenance by not properly evaluating and establishing compensatory actions for maintaining availability of associated EDGs
05000333/FIN-2018002-012018Q2FitzPatrickLicensee-Identified Violation

This violation of very low safety significance was identified by Exelon and has been entered into Exelons CAP and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy 

Violation: 10 CFR 71.5 requires that licensees who transport licensed material comply with the applicable requirements of the Department of Transportation (49 CFR). 49 CFR 172.202(a)(1) and (a)(2) require that the shipping description on the shipping paper include the proper shipping name and identification number for the material. 49 CFR 172.302(a) requires that shipments in bulk packages be marked with the identification number. Contrary to the above, on July 12, 2016, the shipping description on the shipping paper for shipment JAF-2016-1613 from FitzPatrick to Tennessee did not include the proper shipping name and identification number for the material. Exelon identified the error during a subsequent review of the shipping paperwork. Significance/Severity Level: No examples of transportation issues are presented in IMC 0612, Appendix E (Examples of Minor Issues). IMC 0609, Appendix D, Section VII.C.e.1 lists examples of Green findings that include documentation deficiencies including failure to properly document compliance with 49 CFR requirements such as shipping papers. Corrective Action Reference: Exelon placed this issue into its CAP as CR-JAF-2016-02857. Corrective actions included providing a corrected shipping paper to the facility in Tennessee that had received the package.
05000286/FIN-2018001-022018Q1Indian PointInadequate Procedure for Placing Chemical and Volume Control System Demineralizer In ServiceA self-revealing Green NCV of Technical Specification 5.4.1, Procedures, was identified because Entergy failed to provide adequate guidance in 3-SOP-CVCS-004, Placing the CVCS Demineralizers In or Out of Service. Specifically, Entergy did not provide adequate procedural direction to prevent exceeding the reactor coolant filter differential pressure while placing the demineralizers in service. As a result, the pressurizer water level technical specification limit was exceeded and the CVCS piping upstream of the filter was over-pressurized resulting in diaphram ruptures on valves CH-305 and CH-352 thereby spreading contamination throughout the Primary Auxiliary Building.
05000247/FIN-2018001-012018Q1Indian PointFailure to Incorporate Adequate Test Controls for Quarterly Stroke Close Testing of the Steam Supply Valves to Turbine-Driven Auxiliary Feedwater PumpThe inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, when Entergy did not assure that surveillance tests required to demonstrate that structures, systems, and components will perform satisfactorily in service are identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, during quarterly stroke testing of the steam isolation valves to the 22 turbine-driven auxiliary feedwater pump, PCV-1310A and PCV-1310B, Entergy did not ensure that these valves traveled to the closed position as required to verify that the safety function was met.
05000293/FIN-2017002-022017Q2PilgrimReporting of Unplanned Scrams with Complications Performance Indicator for Feedwater Regulating Valve ScramThe inspectors identified an unresolved item (URI) associated with Entergys reporting of Unplanned Scrams with Complications PI data for the third quarter of 2016. Description. On September 6, 2016, PNPS operators initiated a manual reactor scram based on oscillating feed flow as a result of a malfunction with feedwater regulating valve (FRV) A. As a result of high reactor vessel water level, all of the reactor feed pumps tripped, the HPCI and RCIC systems isolated, and a Group 1 isolation signal was present, initiating closure of the MSIVs. In order to maintain pressure control of the reactor, SRV 3B was manually cycled. This event was reported under Licensee Event Report (LER) 05000293/2016-007-00. During the scram response, PNPS operators were required to use an SRV to maintain reactor pressure control, but Entergys submittal of PI data for the third quarter of 2016 does not count the scram as an Unplanned Scram with Complications, which is required by EN-LI-114, Regulatory Performance Indicator Process. This URI is being opened to determine if a performance deficiency exists pending resolution of the differing interpretation of guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guidance, Revision 7, at the next scheduled Reactor Oversight Process Working Group Meeting. (URI 05000293/2017002-02, Reporting of Unplanned Scrams with Complications Performance Indicator for Feedwater Regulating Valve Scram)
05000293/FIN-2017002-082017Q2PilgrimLicensee-Identified Violation10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires in part, that activities affecting quality shall be accomplished in accordance with documented procedures. Entergy Procedure EN-OP-104, Operability Determination Process, requires that operators have a reasonable expectation of operability when determining the operability of a component. On April 15, 2017, operators did not have a reasonable expectation of operability, as required by EN-OP-104, and incorrectly declared the B SRM operable without reasonable assurance. This resulted in a violation of TS 3.10.B, Core Alterations, which requires, during core alterations, when fuel is in the vessel, at least 2 SRMs shall be operable, one in the quadrant where fuel or control rods are being moved and one in an adjacent quadrant. Entergy entered this issue into the CAP as CRs 2017-3541, 2017-3952, 2017-5294, and 2017-6724. Entergy repaired the B SRM, and performed a causal evaluation on the equipment failure that includes the late inoperability determination by the operators. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix G, Attachment 1, Exhibit 3, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a system, and did not represent a loss of safety function of a train or system, and did not degrade a functional auto-isolation of RHR on low reactor vessel level.
05000293/FIN-2017002-062017Q2PilgrimSecondary Containment Testing not performed per Technical SpecificationsAn NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, and TS 4.7.C, Containment Systems Secondary Containment, was identified when Entergy performed a surveillance test requiring a refueling outage while online. Specifically, Entergy performed Procedure 8.7.3, Secondary Containment Leak Rate Test, TS Surveillance Requirement (SR) 4.7.C from February 27, 1997, to April 5, 2017. As corrective actions, Entergy re-performed the test during the April 2017 refueling outage prior to refueling. This issue was entered into the CAP as CR 2017-2900. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protects the public from radionuclide releases caused by accidents or events. Specifically, Entergy intentionally removed the safety function of standby gas and secondary containment for operational convenience and did not comply with the requirements of TS SR 4.7.C which requires the test to be performed during a refueling outage before refueling. In accordance with IMC 0609.04, Initial Characterization of Findings, issued October 7, 2016, 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 was of very low safety significance (Green), because the finding only represented a degradation of the radiological barrier function provided for the SBGTS. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance - Conservative Bias, in that Entergy personnel did not use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, operators did not refer to the TSs to understand the required conditions for a secondary containment surveillance test. Operators followed an inadequate site procedure for the plant conditions at the time and did not question why removal of a safety function for operational convenience was acceptable. (H.14)
05000293/FIN-2017002-052017Q2PilgrimDamper Failure Causes Loss of Secondary ContainmentA self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and TS 3.7.C.2, Containment Systems Secondary Containment, was identified because Entergy did not establish an appropriate interval to overhaul the secondary containment isolation dampers. As a result, the refueling floor supply isolation dampers were operated beyond the recommended overhaul interval and subsequently failed. Entergys corrective actions included cleaning, lubricating, and post-work testing the failed refueling floor supply isolation dampers. This issue was entered into the CAP as CR 2017-0494. The performance deficiency is more than minor because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, Entergys preventative maintenance (PM) for the refueling floor supply isolation dampers was inadequate to ensure the availability and reliability of SSCs required to maintain secondary containment operable. 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, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency only represented a degradation of the radiological barrier function provided by the reactor building and standby gas treatment system (SBGTS). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution - Resolution, in that Entergy personnel did not take effective corrective actions to address issues in a timely manner. Specifically, in 2016, Entergy personnel identified there were deficiencies in the PM program with technical justifications for deferring PMs. Entergy reasonably had the opportunity to identify which PMs were not performed within recommended guidelines and make appropriate changes as needed. (P.3)
05000293/FIN-2017002-042017Q2PilgrimImproper System Restoration Results in Suppression Pool InoperabilityA self-revealing Green NCV of TS 5.4.1.a, Procedures, was identified on March 31, 2017, when operators did not follow procedures and caused an inadvertent increase in the suppression pool water level. The inspectors determined that the operators did not restore the core spray system valve line-up as prescribed in Attachment 11 of Entergy Procedure 2.2.20, Core Spray, and the maintenance safety tag clearance sheet. Operator implementation of these documents is directed by Entergy Procedure EN-OP-102, Protective Caution Tagging, section 5.19(4)(b). As corrective actions, Entergy performed additional management oversight of control room operations and performed a root cause evaluation (RCE). This issue was entered into the CAP as CR-2017-2785. The performance deficiency is more than minor because it is associated with the equipment reliability attribute of the Mitigating Systems cornerstone objective 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 out of specification conditions on March 31, 2017, impacted suppression pool reliability because the suppression pool was not maintained within parameters required to ensure operability. Additionally, significant analysis was necessary to show the suppression pool and associated supports remained functional when TS requirements were not met. Using IMC 0609, Appendix A, Exhibit 2, issued June 19, 2012, The Significance Determination Process for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system, or component (SSC), the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time (AOT), and the finding did not represent an actual loss of a function of one or more non-TS trains of equipment. Specifically, the suppression pool, including downcomers and supports, remained functional following the influx of water. The finding has a cross-cutting aspect in the area of Human Performance - Procedure Adherence, because Entergy personnel did not follow processes, procedures, and work instructions. Specifically, Entergy personnel did not follow procedures and work instructions during the restoration of the core spray system. (H.8)
05000293/FIN-2017002-032017Q2PilgrimInaccurate Suppression Pool Water Level Instrument not Identified during Post-event Prompt InvestigationAn NRC-identified Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy staff did not identify and correct a condition adverse to quality related to suppression pool water level indication when the A suppression pool wide range instrument provided inaccurate level indication during the inadvertent suppression pool water level increase event on March 31, 2017. As corrective actions, Entergy entered Technical Specification (TS) 3.2.F, Protective Instrumentation - Surveillance Information Readouts, and repaired the instrument. This issue was entered into Entergys corrective action program (CAP) as condition report (CR) 2017-2965. The performance deficiency is more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, inaccurate level indication during off-normal changing level conditions in the suppression pool could result in operator actions not warranted by plant conditions. The finding is also associated with the Initiating Events cornerstone. Using IMC 0609, Appendix A, Exhibit 1, issued June 19, 2012, The Significance Determination Process for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not cause a reactor trip and a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution - Identification, because the Entergy organization did not demonstrate an appropriately low threshold for entering problems into their CAP. Specifically, Entergys prompt investigation of the inadvertent suppression pool level increase event did not identify that the A suppression pool wide range level instrument was not indicating properly and required corrective maintenance. (P.1)
05000293/FIN-2017002-012017Q2PilgrimFailure to Follow Procedure Requirements for the Control of a Flood Protection BarrierAn NRC-identified Green finding was identified because Entergy personnel did not follow Procedure 1.3.135, Control of Doors, to adequately control a condenser bay flood protection door. Specifically, on May 22, 2017, Entergy personnel failed to control door 25A, which is designed to mitigate condenser bay flooding to preclude adversely impacting the important to safety instrument air system. Entergys short-term corrective actions included closing the door and providing additional operator training. This issue was entered into the CAP as CR 2017-5746. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using IMC 0609, Appendix A, Exhibit 4, External Events Screening Questions, issued June 19, 2012, with respect to the degraded safety function of the flood barrier door. The finding was determined to be of very low safety significance (Green) because the failure of the flood door was determined to not degrade the instrument air system ability to support the feedwater injection function or the alternate injection through the control rod drive system. This is because the backup diesel driven compressor was available to be started locally and supply the instrument air headers. The finding also did not involve the total loss of any safety function. The finding has a cross-cutting aspect in the area of Human Performance - Procedure Adherence, because Entergy personnel did not follow processes, procedures, and work instructions. Specifically, Entergy personnel did not follow procedural requirements to adequately control flood protection door 25A. (H.8)
05000293/FIN-2017002-072017Q2PilgrimUntimely 10 CFR 50.72 Notification of a Secondary Containment System Functional FailureAn NRC-identified SL IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, was identified because both trains of the SBGTS were made inoperable during surveillance testing, and the condition was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, on April 5, 2017, while performing TS SR 4.7.C, trains A and B of the SBGTS were made inoperable leading to the inoperability of the Secondary Containment System (SCS). As a corrective action, Entergy personnel performed a causal evaluation. This issue was entered into the CAP as CR 2017-7446. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a condition that could have prevented a safety function was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy (the failure of a licensee to make a report as required by 10 CFR 50.72 or 10 CFR 50.73), the inspectors determined that the violation was a SL IV violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, inspectors did not assign a cross-cutting aspect, in accordance with IMC 0612, Appendix B.
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.
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-2016003-032016Q3PilgrimFailure 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.
05000293/FIN-2015004-032015Q4PilgrimFailure to Identify the Cause of a Significant Condition Adverse to QualityThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, when Entergy did not determine the cause of a significant condition adverse to quality (SCAQ). Specifically, a causal evaluation was not performed for a failed safety-related relay that ensured the automatic operation of the low pressure coolant injection (LPCI) system injection valves in a degraded voltage condition. Entergy replaced the failed relay and restored LPCI to an operable status on May 10, 2015. Entergy entered the issue into the CAP as CR 2015-9762. This finding is more than minor because it is associated with the Mitigating System cornerstone attribute of equipment performance and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The failure to identify the cause and extent of condition of the relay failure as directed by site procedures could result in repeat events which adversely affect safety system availability. In accordance with IMC 0609.04 and Exhibit 2 of IMC 0609, Appendix A, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not involve the design of a mitigating structure, system, or component (SSC) or a loss of function of a train or system for greater than the technical specification (TS) allowed outage time. The inspectors determined this finding has a cross-cutting aspect in Human Performance, Procedure Adherence, because individuals did not recatergorize the CR to a higher level requiring a causal evaluation, as required by EN-LI-102 when a licensee event report (LER) was issued. The site also did not retain the failed safety-related part, as required by EN-MA-101-02.
05000293/FIN-2015004-012015Q4PilgrimInadequate Implementation of Corrective Action following Winter Storm JunoThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, because Entergy did not adequately implement corrective actions for an identified condition adverse to quality. Specifically, Entergy did not implement all of the procedure changes needed to ensure shutdown cooling was placed in service in a timely manner after plant shutdown in preparation for or during a severe winter storm. Entergy entered this issue into the CAP as CR 2016-0120 and updated procedure 2.1.42 to meet the requirements of the corrective actions in CR 2015-0558. Inspectors verified that the new procedure revision included the required actions. The inspectors determined this performance deficiency is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity cornerstone, and adversely affected its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The inspectors determined that this finding is of very low safety significance (Green) in accordance with IMC 0609, Attachment 4 and Exhibit 3 of Appendix A, because it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, and heat removal components. The inspectors determined that this finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Entergy staff did not ensure procedure revisions were made in accordance with the requirements of EN-LI-102, Corrective Action Program.
05000293/FIN-2015004-042015Q4PilgrimInadequate Design Control of MSIV Nitrogen Supply Line Support leads to ScramA self-revealing Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, was identified because Entergy did not use the correct work planning and design controls to repair the support for the nitrogen supply line for the 1C inboard main steam isolation valve (MSIV). Specifically, inadequate design controls led to a failed horizontal unistrut support for the nitrogen supply line to the 1C MSIV, resulting in the header resting on the main steam line. This caused vibration-induced cyclic failure of the nitrogen supply line, closure of 1C MSIV, and a plant scram. The damaged line was modified and repaired using an additional unistrut for support as determined by the engineering change process. Entergy entered the issue into the corrective action program (CAP) under condition report (CR) 2015- 07285. This finding is more than minor because it is associated with the Initiating Events cornerstone attribute of equipment performance and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure of the pneumatic supply header support resulted in a plant scram due to the vibration induced cyclic failure of the nitrogen supply line and subsequent closure of 1C MSIV. In accordance with IMC 0609.04 and Exhibit 1 of IMC 0609, Appendix A, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not involve the complete or partial loss of a support system that contributes to the likelihood of, or cause, an initiating event and affect mitigation equipment. The inspectors determined this finding does not have a cross-cutting aspect because the performance deficiency occurred in 2001 and is not indicative of current performance.
05000293/FIN-2015004-022015Q4PilgrimFailure to Properly Implement Procedure Changes in accordance with TS 5.4.1aThe inspectors identified an NCV of TS 5.4.1, Procedures, because Entergy was not adequately maintaining procedures listed in Regulatory Guide (RG) 1.33, Revision 2, Appendix A, February 1978. Specifically, the inspectors identified several examples where Entergy staff inappropriately used Entergy procedure EN-OP-112, Night and Standing Orders, to implement procedure changes instead of PNPS quality assurance procedure NOP98A1, Procedure Process. Entergy entered the issue into the CAP as CR 2015-09233. The performance deficiency was determined to be more than minor because if left uncorrected it has the potential to lead to a more significant safety concern. Specifically, the inspectors determined the issue was similar to Example 4.a of IMC 0612, Appendix E, which states that an insignificant procedure error would be more than minor if the licensee routinely failed to adhere to the applicable procedure. The inspectors evaluated the finding using IMC 0609, Attachment 4 and Appendix A. Using Exhibit 2 of Appendix A, the inspectors determined this finding was of very low safety significance (Green) because it did not involve a design or qualification deficiency, it would not lead to a potential or actual loss of system or safety functions, it did not involve the loss or degradation of equipment or a function specifically designed to mitigate a seismic, flooding, or severe weather initiating event, and it did not involve the total loss of any safety function as identified in Exhibit 4. The inspectors determined that the finding had a cross-cutting aspect in Problem Identification and Resolution, Resolution, because, Pilgrim did not adhere to the CAP evaluation and corrective action program timeliness requirements that would have likely led them to use the appropriate procedure change process.
05000293/FIN-2015007-072015Q1PilgrimFailure to Report a Major Loss of Emergency Assessment CapabilityAn NRC-identified SL IV NCV of 10 CFR Part 50.72(b)(3)(xiii) was identified when Entergy failed to make a required event notification within eight hours for a major loss of assessment capability. Specifically, an unplanned loss occurred of all EAL instrumentation associated with Sea Water Bay level that resulted in an inability to evaluate all EALs for an abnormal water level condition. Entergy entered the issue into the CAP as CR-PNP-2015-00949. Compliance was restored on February 5, 2015, when Entergy reported the major loss of assessment capability under Event Notification (EN) 50790. The inspectors determined that Entergys failure to submit an event notification in accordance with 10 CFR 50.72 within the required time was a performance deficiency that was reasonably within Entergys ability to foresee and correct, and should have been prevented. Since the failure to submit a required event report impacts the regulatory process, the violation was evaluated using Section 2.2.4 of the NRCs Enforcement Policy, dated July 9, 2013, instead of the SDP. Using the example listed in Section 6.9.d.9, A licensee fails to make a report required by 10 CFR 50.72 or 10 CFR 50.73, the issue was evaluated and determined to be a SL IV violation. The inspectors reviewed the condition for reactor oversight process significance. Because this NRC-identified violation involves the traditional enforcement process and does not have an underlying technical violation that would be considered more-than-minor, the inspectors did not assign a cross-cutting aspect to this violation in accordance with IMC 0612.
05000293/FIN-2015007-082015Q1PilgrimInadequate Testing of the Diesel-Driven Air CompressorA self-revealing Green finding was identified for Entergys failure to verify that the diesel-driven air compressor (K-117) was available for service prior to the January 27, 2015 winter storm. Specifically, although K-117 was tested prior to the winter storm, the test methodology did not reveal that the capacity of the starting battery was inadequate. The failure to verify that the diesel-driven air compressor (K-117) was available for service prior to the January 27, 2015 winter storm is a performance deficiency that was within Entergys ability to foresee and correct. This resulted in a loss of instrument air during the plant trip which complicated the event response. Entergy entered the issue into the corrective action program (CAP) as condition report (CR)-PNP-2015-00559 and initiated actions to supply instrument air with a temporary air compressor. Entergy also revised the operability test for K-117 air compressor to remove the alternating current (AC) power source prior to starting the air compressor. This self-revealing issue was more than minor because it is associated with the procedure quality and design control attributes of the Initiating Events cornerstone and adversely impacted the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, failure of K-117 resulted in loss of instrument air, which adversely impacted the plant response during the January 27, 2015 winter storm. Additionally, this issue is also associated with the procedure quality and design control attributes of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating event to prevent undesirable consequences. The inspectors screened the issue under the Initiating Events cornerstone using Attachment 4 and Exhibit 1 of Appendix A to IMC 0609, Significance Determination Process, because that cornerstone was determined to be more impacted by the finding than the Mitigating Systems cornerstone. The inspectors concluded that a detailed risk evaluation would be required because the finding involved the complete loss of a support system (instrument air) that contributes to the likelihood of an initiating event and affects mitigation equipment. A senior reactor analyst performed a detailed risk evaluation of this issue. The NRC model for PNPS was adjusted to account for a loss of the instrument air compressor on a LOOP. The change in core damage frequency was very low. A review of the dominant accident sequences indicated the contribution from a large early release and from external risk contributors to be very small. Therefore, the issue was determined to be of very low risk significance (Green). The finding had a cross-cutting aspect in the area of Human Performance, Design Margins, because Entergy failed to ensure that the K-117 battery was designed with adequate margin. This finding is reflective of current performance because the inadequate design margin of the battery should have been discovered through proper testing.
05000293/FIN-2015007-062015Q1PilgrimFailure to Implement Compensatory Measures for Out-of-Service EAL InstrumentationThe inspectors identified a Green NCV of 10 CFR 50.54(q)(2) for failing to follow and maintain an emergency plan that meets the requirements of planning standards 10 CFR 50.47(b) and Appendix E. Specifically, on January 27, 2015, following a loss of instrument air, the indications in the Control Room for Sea Water Bay level were lost, and Entergy did not implement compensatory measures, as directed by an Emergency Plan Implementing Procedure, to determine whether a Sea Water Bay level emergency action level (EAL) threshold had been exceeded. Entergy entered this issue into the CAP as CR-PNP-2015- 00948 and initiated corrective actions to identify alternative means for assessing this EAL in the event of a loss of Sea Water Bay level instruments. The inspectors determined that Entergys failure to implement compensatory measures for out-of-service EAL instrumentation was a performance deficiency that was within Entergys ability to foresee and correct and should have been prevented. Specifically, Entergy did not implement the compensatory measure listed in Attachment 9.2 of EP-IP-100.1, Emergency Action Levels, Revision 10. The inspectors determined that following a loss of instrument air, the indications for Sea Water Bay level EAL were lost, rendering those EALs ineffective such that Entergy was not able to determine whether a Sea Water Bay level EAL threshold had been exceeded and to declare an emergency based on the Sea Water Bay level. This NRC-identified performance deficiency was more than minor because it was associated with the emergency response organization performance (program elements not meeting 50.47(b) planning standards) attribute of the Emergency Preparedness cornerstone and 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. Specifically, the out-of-service Sea Water Bay level instrumentation could have led to an emergency not being declared in a timely manner. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Characterization of Findings, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, issued September 23, 2014, when the finding is in the licensees Emergency Preparedness cornerstone. The inspectors determined the finding was associated with risk significant planning standard 10 CFR 50.47(b)(4), Emergency Classification System, and corresponded to the following Green Finding example in Table 5.4-1: an EAL has been rendered ineffective such that any Alert or Unusual Event would not be declared, or declared in a degraded manner for a particular off-normal event. Therefore, using Figure 5.4-1, Significance Determination for Ineffective EALs and Overclassification, and the example in Table 5.4-1, the inspectors determined the finding was of very low safety significance (Green). The finding had a cross-cutting aspect in the area of Human Performance, Documentation, because Entergy did not maintain complete and accurate documentation. Specifically, compensatory measures associated with out-of-service EAL instrumentation are not governed by comprehensive and high-quality programs, processes, and procedures.
05000293/FIN-2015007-052015Q1PilgrimFailure to Identify Condition Adverse to Quality Associated with CS Discharge Header VoidingThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because PNPS staff failed to identify and correct conditions adverse to quality associated with the partial voiding of the A core spray (CS) discharge header on January 27, 2015, following the loss of the keepfill system due to a LOOP. PNPS entered the issue into the CAP as CR-PNP-2015-01406 and planned procedural changes that would provide guidance to operate the affected pumps in order to prevent pump discharge piping from voiding if keepfill pressure is lost. The failure to identify, evaluate, and correct the A CS discharge header partial voiding following loss of keepfill on January 27, 2015, is a performance deficiency that was within Entergys ability to foresee and correct. Because the issue was not entered into the CAP, the condition was neither evaluated nor was corrective action taken or planned. This NRCidentified issue is more than minor because it is associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, to IMC 0609, Significance Determination Process. This finding was determined to be of very low The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, because PNPS staff failed to identify and correct conditions adverse to quality associated with the partial voiding of the A core spray (CS) discharge header on January 27, 2015, following the loss of the keepfill system due to a LOOP. PNPS entered the issue into the CAP as CR-PNP-2015-01406 and planned procedural changes that would provide guidance to operate the affected pumps in order to prevent pump discharge piping from voiding if keepfill pressure is lost. The failure to identify, evaluate, and correct the A CS discharge header partial voiding following loss of keepfill on January 27, 2015, is a performance deficiency that was within Entergys ability to foresee and correct. Because the issue was not entered into the CAP, the condition was neither evaluated nor was corrective action taken or planned. This NRCidentified issue is more than minor because it is associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Appendix A,The Significance Determination Process for Findings At-Power, to IMC 0609, Significance Determination Process. This finding was determined to be of very low.
05000293/FIN-2015007-042015Q1PilgrimFailure to Follow RCIC System Manual Restart ProcedureA self-revealing Green NCV of TS 5.4.1, Procedures, was identified because the operating crew failed to implement a procedure step to open the reactor core isolation cooling (RCIC) system cooling water supply valve during a manual startup of the system. As a result, the RCIC system was operated for over 2 12 hours with no cooling water being supplied to the lubricating oil cooler or to the barometric condenser. Entergy entered the issue into the CAP as CR-PNP-2015-0566, CR-PNP-2015-0570, and CR-PNP-2015-0952 and conducted a human performance review of the Control Room operators involved with the issue. The inspectors determined that the failure to implement Procedure 5.3.35.1, Attachment 29, RCIC Injection Manual Alignment Checklist, and the Vacuum Tank Pressure Hi Alarm, C904L-F3, alarm response procedure was a performance deficiency and was reasonably within the ability of Entergy personnel to foresee and prevent. This self-revealing finding was more than minor because it was associated with the human performance attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. Specifically, on January 27, 2015, reactor operators failed to open MO-1301-62, cooling water supply valve, during a manual restart of the RCIC system in accordance with procedure 5.3.35.1, RCIC Injection Manual Alignment Checklist. Additionally, the operating crew failed to identify the valve was out of position even after the Vacuum Tank Pressure Hi Alarm, C904L-F3, was received two minutes after the system was re-started and the alarm response procedure identified Improper Valve Lineup as a probable cause. The team evaluated the finding using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The team determined this finding was not a design or qualification deficiency and was not a potential or actual loss of system or safety function, and is therefore of very low safety significance (Green). During the period when the RCIC system was operated in this condition, no temperature limits were exceeded. The inspectors noted that in the event of a RCIC system automatic start, the cooling water supply valve would have opened automatically. This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Entergy licensed personnel did not implement procedure 5.3.35.1, RCIC Injection Manual Alignment Checklist , to open MO-1301-62. Additionally, Entergy licensed personnel did not implement the Vacuum Tank Pressure Hi Alarm, C904L-F3, response procedure to check for an improper valve line-up.
05000293/FIN-2015007-032015Q1PilgrimInadequate Loss of Instrument Air Abnormal Operating ProcedureA self-revealing Green NCV of TS 5.4.1, Procedures, was identified because Entergy failed to include appropriate operator actions to both recognize the effects of and recover systems and components important to safety within Procedure 5.3.8, Loss of Instrument Air, abnormal operating procedure. Entergy entered this issue into the CAP as PNP-CR-2015 0888 and issued a revision to Procedure 5.3.8 to provide additional guidance to operators during a loss of instrument air. The inspectors determined that the level of detail in Procedure 5.3.8, Loss of Instrument Air, Revision 39, was inadequate to provide appropriate operator guidance to identify and mitigate key events of January 27, 2015. This self-revealing performance deficiency was reasonably within the ability of Entergy personnel to foresee and the issue should have been prevented. The finding was more than minor because it was associated with the procedure quality attribute of the Mitigating System cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesired consequences. The lack of adequate instructions in the procedure adversely affected several operator actions and plant equipment on January 27, 2015, during the LOOP and loss of instrument air. The team evaluated the finding using IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The team determined this finding was of very low safety significance (Green) because it was not a design or qualification deficiency, did not result in a loss of function of a TS required system, and did not represent an actual loss of function of one or more non-TS trains of equipment designated as a high safety-significant system. This finding had a cross-cutting aspect in the area of Human Performance, Resources, because Entergy leaders did not ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety.
05000293/FIN-2015007-012015Q1PilgrimInadequate Past Operability Assessment of C Safety Relief ValveThe team identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, when Entergy staff performed an inadequate past operability determination that assessed performance of the C safety/relief valve (SRV), which did not open as expected when called upon to function. Specifically, following the January 27, 2015 reactor scram, operators placed an open demand for the C SRV twice during post-scram recovery operations, but the valve did not respond as expected and did not perform its pressure reduction function on both occasions. Entergys subsequent past operability assessment for the valves operation incorrectly concluded that the valve was fully capable of performing its required functions during its installed service. In response to the teams past operability concerns, Entergy subsequently re-evaluated the past operability of C SRV and concluded that it was inoperable and placed the issue into the corrective action program (CAP) as CR-PNP-2015- 02051. The team determined the failure to adequately assess past operability of the C SRV was a performance deficiency that was reasonably within Entergys ability to foresee and correct. This NRC-identified performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affects the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent core damage. The team evaluated the finding using IMC 0609, Appendix 0609.04, Initial Characterization of Findings, which directed the use of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. Using Exhibit 2, Mitigating Systems Screening Questions, of IMC 0609, Appendix A, the team determined this finding was not a design or qualification deficiency and was not a potential or actual loss of system or safety function, and was therefore of very low safety significance (Green). The finding had a cross-cutting aspect in Human Performance, Conservative Bias, because Entergy did not use decision making practices that emphasized prudent choices over those that are simply allowable. Specifically, Entergy did not appropriately evaluate unexpected and unsatisfactory performance of the C SRV in consideration of the entire pressure range that the SRV, including its automatic depressurization system (ADS) function, was required to be operable.
05000293/FIN-2015007-022015Q1PilgrimFailure to Identify, Evaluate, and Correct A SRV Failure to Open Upon Manual ActuationA self-revealing preliminary White finding and AV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, and Technical Specification (TS) 3.5.E, Automatic Depressurization System, was identified for the failure to identify, evaluate, and correct a significant condition adverse to quality associated with the A SRV. Specifically, Entergy failed to identify, evaluate, and correct the A SRVs failure to open upon manual actuation during a plant cooldown on February 9, 2013. In addition, the failure to take actions to preclude repetition resulted in the C SRV failing to open due to a similar cause following the January 27, 2015, LOOP event. Entergy entered this issue in to the CAP as CR-PNP-2015-01983, CR-PNP-2015-00561, and CR-PNP-2015-01520. Immediate corrective actions included replacing the A and C SRVs and completing a detailed operability analysis of the installed SRVs which concluded that a reasonable assurance of operability existed. Entergys failure to identify, evaluate, and correct the condition of the A SRVs failure to open upon manual actuation during a plant cooldown on February 9, 2013, was a performance deficiency. In addition, the failure to take actions to preclude repetition resulted in the C SRV failing to open due to a similar cause following the January 27, 2015 LOOP event. The self-revealing finding was within Entergys ability to foresee and correct because indications were available to determine that the A SRV valve did not open upon manual actuation. This was discovered as a result of an extent of condition review of the C SRV failing to open upon manual actuation following the January 27, 2015 LOOP event. This performance deficiency is more than minor because it could reasonably be viewed as a precursor to a significant event if two of the four SRVs failed to open when demanded to depressurize the reactor, following the failure of high pressure injection systems or torus cooling, to allow low pressure injection systems to maintain reactor coolant system inventory following certain initiating events. In addition, it is associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors screened this issue for safety significance in accordance with IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, issued June 19, 2012. The screening determined that a detailed risk evaluation was required because it was assumed that for a year period, two of the four SRVs were in a degraded state such that they potentially would not have functioned to open at some pressure lower than rated pressure and would not fulfill their safety function for greater than the TS allowed outage time. Specifically, the assumptions of failures to open were based on: a failed actual opening demand at 200 psig reactor pressure on January 27, 2015, for the C SRV; examination of the valve internals at the testing vendor (National Technical Systems); and a previous failed actual opening demand at 114 psig reactor pressure on February 9, 2013, for the A SRV. The staff determined that there wasnt an existing SDP risk tool that is suitable to assess the significance of this finding with high confidence, mainly because of the uncertainties associated with: the degradation mechanism and its rate; the range of reactor pressure at which the degraded valves could be assumed to fully function; any potential benefit from an SRV lifting at rated pressure, such that the degradation would be less likely to occur and, therefore, prevent a subsequent failure at low pressure in the near-term; the time based nature of plant transient response relative to when high pressure injection sources fail and the associated impact of reduced decay heat on the SRV depressurization success criteria; and the ability to credit other high pressure sources of water. Based on the considerations above, the risk evaluation was performed using IMC 0609, Appendix M, Significance Determination Process Using Qualitative Criteria, issued April 12, 2012. The NRC made a preliminary determination that the finding was of low to moderate safety significance (White) based on quantitative and qualitative evaluations. The detailed risk evaluation is contained in Attachment 4 to this report. This finding does not present a current safety concern because the A and C SRVs were replaced during the outage following the January 27, 2015 LOOP and reactor trip event. Also, Entergy performed a detailed operability analysis of the installed SRVs which concluded that a reasonable assurance of operability existed. This finding had a cross-cutting aspect in Problem Identification and Resolution, Evaluation, because Entergy did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, Entergy staff did not thoroughly evaluate the operation of the A SRV during the February 9, 2015 plant cooldown and should have reasonably identified that the A SRV did not open upon three manual actuation demands.
05000293/FIN-2014005-012014Q4PilgrimModification to the Spent Fuel Pool Cask Area without Prior NRC ApprovalSeverity Level lV. The inspectors identified a Severity Level lV NCV of Title 10 of the Cod of Federal Regulations (10 CFR) 50.59 in that Entergy did not obtain a license amendment prior to implementing a change to the plant that required a change to technical specification (TS). Specifically, Entergy removed the energy absorbing pad described in TS 4.3.4.b, Design Features, and Updated Final Safety Analysis Report (UFSAR) section 10.3.6, Consequences of a Dropped Fuel Cask, without receiving prior NRC approval. Entergy submitted a License Amendment Request (LAR) supplement to the NRC on September 11, 2014, to remove the energy absorbing pad language from TS, and performed an extent of condition review on previous engineering changes and prohibited placing a cask in the spent fuel pool (SFP) until receiving NRC approval for a change to TS 4.3.4.b. The inspectors determined that Entergy did not perform an adequate 10 CFR 50.59 evaluation and obtain a license amendment prior to removing the SFP energy absorbing pad. The inspectors determined this was a performance deficiency that was within Entergys ability to foresee and correct and should have been prevented. Because the issue had the potential to affect the NRCs ability to perform its regulatory function, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using the Enforcement Manual, the inspectors determined that the violation was a Severity Level IV (a 10 CFR 50.59 violation that resulted in conditions that required NRC approval before implementation) violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation that would be considered more-than-minor, inspectors did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Appendix B.
05000220/FIN-2014005-022014Q4Nine Mile PointFailure to Make Timely Reports of Changes in Licensed Operator Medical Status Which Impacted Issuance of Initial and Renewal LicensesExelon Generation Company, LLC (Exelon) identified two AVs: (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, during an internal audit in July 2014, Exelon identified that between September 2002 and February 2012, NMPNS staff submitted certified copies of an NRC reactor operator and/or senior operator license applications for seven applicants that did not specify that the applicants required a restriction in order to maintain medical qualifications. The NRC issued the reactor operator and senior operator initial and renewed licenses for the seven applicants, but without the necessary medical restrictions (AV #1). From June 2002 through August 2014, Exelon had numerous additional opportunities to identify these potentially disqualifying medical conditions and that license conditions were required during the biennial licensed operator requalification program reviews and medical examinations. On September 25, 2014, a period that exceeded 30 days from when the conditions were identified, the facility notified the NRC of these medical conditions via a letter requesting amendment to the seven operators licenses to include the appropriate restrictions (AV #2). The NRC issued the license amendment with the new restrictions. The NRC inspectors also identified an additional example of both AVs which had not been reported by Exelon to the NRC in the September 25, 2014 letter. On November 5, 2014, Exelon requested termination of the license for that operator. This issue was entered into Exelons corrective action program (CAP) The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the reactor operator and senior operator license applications and to notify the NRC of a change in a reactor operator or senior operators status for a condition which was known by Exelon were performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC requires Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued reactor operator and senior operator licenses to the applicants based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
05000410/FIN-2014005-032014Q4Nine Mile PointMissed Surveillance Test of Alternate Decay Heat Removal Secondary Containment Isolation ValvesThe inspectors identified a Green NCV of Unit 2 Technical Specification (TS) 5.4, Procedures, for Exelons failure to properly perform procedure N2-OSP-GTS-R001, Secondary Containment Integrity Test, Revision 01100. Specifically, Exelon staff failed to ensure spectacle flanges associated with alternate decay heat (ADH) secondary containment isolation were properly installed. As a result, surveillance testing associated with ADH check valves 2ADH*V21A/B and 2ADH*V22A/B was not performed to ensure secondary containment integrity as required by N2-OSP-GTS-R001. Exelon immediately entered this issue into their CAP as issue report (IR) 2403311. Exelon entered TS Surveillance Requirement (SR) 3.0.3, Limiting Condition for Operability Applicability, which is used when a licensee discovers that a surveillance test requirement has not been performed. As required by the TS, Exelon completed a risk evaluation of the missed surveillance and determined large early release frequency remained low without ADH secondary containment isolation. Exelon also performed extent-of-condition inspections for other systems which may not have proper alignment to ensure they are meeting TS requirements. On December 4, Exelon rotated the spectacle flanges to the no flow isolation position to ensure secondary containment integrity was maintained The finding is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance tha physical design barriers protect the public from radionuclide releases caused by accident or events. Specifically, by performing N2-OSP-GTS-R001 in 2012 and 2014 without first ensuring the spectacle flanges were properly installed, Exelon did not verify the secondar containment requirements of TS SR 3.4.6.1 were maintained. Additionally, this issue wa similar to Example 3.d in IMC 0612, Appendix E, Examples of Minor Issues, in that th failure to implement the TS SR as required was not minor if the surveillance had not bee conducted. By not correctly testing the secondary containment in 2012 and 2014, the SR o TS 3.4.6.1 was not met. In accordance with IMC 0609.04, Initial Characterization o Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Proces for Findings At-Power, the inspectors determined this finding is of very low safet significance (Green) because the finding only represents a degradation of the radiologica barrier function provided for the control room, or auxiliary, spent fuel pool (SFP), or standb gas treatment system (boiling water reactor). This finding has a cross-cutting aspect in th area of Human Performance, Avoid Complacency, because Exelon staff did not implemen appropriate error reduction tools. Specifically, operators did not use error reduction tools t ensure the spectacle flanges were installed in the no flow position and as a result, the failed to leak test the ADH check valves in the secondary containment drawdown test a required by N2-OSP-GTS-R001 (H.12).
05000352/FIN-2014005-012014Q4LimerickUnplanned Manual Power Reduction to 90% on Unit 1A self-revealing, Green non-cited violation (NCV) of Technical Specification (TS) 6.8.1.b, Administrative Controls, was identified for LGS failure to properly implement station procedure MA-AA-716-100, Maintenance Alterations Process, during troubleshooting and calibration associated with the Unit 1 condensate filter (CF) system. As a result, on September 9, 2014, one of two Instrument Maintenance (IM) technicians inadvertently mispositioned the air supply valve to the 1G CF flow transmitter causing an unplanned plant transient. The inspectors determined that the failure to properly implement station procedure MA-AA-716-100, Maintenance Alterations Process, during troubleshooting of CF system instrumentation, was a performance deficiency. LGS promptly performed an investigation, verified the plant alignment and safely returned the Unit 1 reactor to 100 percent power. LGS entered the issue into their corrective action program (CAP) as issue report (IR) 2116233 This self-revealing finding is more than minor because it affected the human performance attribute of the Initiating Events Cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions. This resulted in elevated main steam line radiation levels which required operators to reduce reactor power in accordance with abnormal operating procedures. The inspectors evaluated the finding using inspection manual chapter (IMC) 0609, Appendix A, The Significance Determination Process for Findings At-Power, to IMC 0609, Significance Determination Process. This finding was determined to be of very low safety significance (Green) because it was associated with a transient initiator, but didnt cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of Human Performance, because LGS maintenance management did not ensure supervisory and management oversight of work activities (H.2).
05000220/FIN-2014005-012014Q4Nine Mile PointIncomplete and Inaccurate Medical Information Provided by Exelon Which Impacted Issuance of Initial and Renewal LicensesExelon Generation Company, LLC (Exelon) identified two AVs: (1) An AV of Title 10 of the Code of Federal Regulations (10 CFR) 50.9, Completeness and Accuracy of Information; and (2) An AV of 10 CFR 50.74, Notification of Change in Operator or Senior Operator Status. Specifically, during an internal audit in July 2014, Exelon identified that between September 2002 and February 2012, NMPNS staff submitted certified copies of an NRC reactor operator and/or senior operator license applications for seven applicants that did not specify that the applicants required a restriction in order to maintain medical qualifications. The NRC issued the reactor operator and senior operator initial and renewed licenses for the seven applicants, but without the necessary medical restrictions (AV #1). From June 2002 through August 2014, Exelon had numerous additional opportunities to identify these potentially disqualifying medical conditions and that license conditions were required during the biennial licensed operator requalification program reviews and medical examinations. On September 25, 2014, a period that exceeded 30 days from when the conditions were identified, the facility notified the NRC of these medical conditions via a letter requesting amendment to the seven operators licenses to include the appropriate restrictions (AV #2). The NRC issued the license amendment with the new restrictions. The NRC inspectors also identified an additional example of both AVs which had not been reported by Exelon to the NRC in the September 25, 2014 letter. On November 5, 2014, Exelon requested termination of the license for that operator. This issue was entered into Exelons corrective action program (CAP) The inspectors determined that Exelons failure to provide complete and accurate information to the NRC in the reactor operator and senior operator license applications and to notify the NRC of a change in a reactor operator or senior operators status for a condition which was known by Exelon were performance deficiencies that were within their ability to foresee and correct and should have been prevented. The inspectors determined that traditional enforcement applies, as the issue affected the NRCs ability to perform its regulatory function. Namely, the NRC requires Exelon to ensure all licensed operators meet the medical conditions of their licenses. If, during the term of the individual operator license, an operator develops a permanent physical or mental disability that causes the operator to fail to meet the requirements of 10 CFR 55.21, Medical Examination, the licensee shall notify the NRC within 30 days of learning of the diagnosis, in accordance with 10 CFR 50.74(c). Additionally, the NRC issued reactor operator and senior operator licenses to the applicants based on information that was not complete and accurate in all material aspects. The performance deficiencies were screened against the Reactor Oversight Process per the guidance of IMC 0612, Appendix B, Issue Screening. No associated Reactor Oversight Process finding was identified and no cross-cutting aspect was assigned. These issues constitute AVs in accordance with the NRCs Enforcement Policy, and their final significance will be dispositioned in separate future correspondence. (Section 1R11)
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BInspector::Thomas-20Dunn-5D-5D/-3FQuarter/-3FSite/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3Dstart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20List-5D"