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05000293/FIN-2018003-01Pilgrim2018Q3Failure to Identify an Adverse Condition Associated with Elevated Standby Gas Treatment System Accumulator LeakageThe inspectors identified a Green non-cited violation (NCV) of Technical Specifications 3.7.B.1.c because Entergy exceeded the TS allowed outage time for the standby gas treatment system (SBGT) when the station did not identify an adverse condition associated with elevated air accumulator leakage in the system.
05000293/FIN-2018002-04Pilgrim2018Q2Licensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions appropriate to the circumstances and shall be accomplished in accordance with the instructions. Contrary to the above, from January 1994 to June 2017, Entergy modified site surveillance procedure 8.M.3-18, Standby Gas Treatment System Exhaust Fan Logic Test and Instrument Calibration, without prescribing adequate documented instructions for the condition caused by the testing. Specifically, Entergy failed to identify that the procedurally prescribed lineup of the standby gas treatment system resulted in secondary containment being inoperable due to the large opening introduced into the system. Significance/Severity: The inspectors evaluated this finding using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance. Corrective Action Reference: CR-PNP-2017-11714 The disposition of this violation closes Licensee Event Reports 05000293/2017-013-00 and 05000293/2017-013-01.
05000410/FIN-2018002-01Nine Mile Point2018Q2Failure to Ensure Proper Control of the Standby Gas Treatment System Damper Valve, 2GTS*V2000B, Within Procedures, Materials, and Design Control MeasuresThe inspectors identified a Green finding and associated NCVof 10 CFRPart 50, Appendix B, Criterion III, Design Control, when Exelon failed to ensure proper control of the SGTS damper valve 2GTS*V2000B within procedures, materials, and design control measures. Specifically, on April 15, 2018 operators attempted to run B SGTS for containment purge; however, no flow was observed and the system was secured. Operators discovered the 2GTS*V2000B closed due to the failure of the operating mechanism to maintain control of the valve position.
05000416/FIN-2017011-05Grand Gulf2018Q1Failure to Correct Control Room Boundary Door Resulted in Loss of Safety FunctionThe inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Criterion XVI, Corrective Action, for the licensees failure to appropriately correct a condition adverse to quality. Specifically, the control room envelope door had been documented in several condition reports for not consistently working properly. Subsequent to these condition reports, on July 9, 2017, the door was opened and did not close automatically, and therefore the door was left in an unsecured position. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2017-06705. The licensee restored compliance by securing the door and replacing the hinge bushings to ensure the door would close properly. The failure to correct a condition adverse to quality for a control room envelope boundary door was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the structures, systems, and components and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (functionality of the control room) protect the public from radionuclide releases caused by accidents or events. Specifically, on July 9, 2017, since the licensee had not corrected the adverse conditions identified on the control room envelope door, the door was left in an unsecured position and resulted in the station declaring both trains of standby fresh air inoperable. Using Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, and Nuclear Regulatory Commission Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system, and did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. The period of the barrier in the open position was of short duration, approximately 1 minute, and therefore did not result in significant risk input. This finding had a cross-cutting aspect in the area of problem identification and resolution, resolution, because the licensee did not take corrective actions in a timely manner commensurate with their safety significance. Specifically, the licensee did not ensure proper priority of corrective actions on the degraded control room envelope boundary door (P.3).
05000298/FIN-2017003-02Cooper2017Q3Failure to Account for Instrument Uncertainty in Safety-Related Ventilation Surveillance ProceduresThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for multiple examples of the licensees failure to assure that required testing was performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, on July 12, 2017, the inspectors identified that Surveillance Procedure 6.1SGT.501, Standby Gas Treatment A Carbon Sample, Carbon Adsorber and HEPA Filter In-place Leak Test, and Components Leak Test, Revision 16, failed to account for test instrument uncertainty in the surveillance acceptance criteria. In response to the inspectors question, the licensee discovered that instrument uncertainty was not accounted for in several standby gas treatment system surveillance procedures, as well as surveillance procedures for the control room emergency filter system; diesel generator ventilation system; control building essential ventilation system; emergency core cooling essential ventilation systems; and several emergency preparedness ventilation systems. Corrective actions to restore compliance included incorporation of instrument uncertainty into procedure changes for the affected surveillance procedures and verification that the new acceptance criteria did not challenge past operability for the affected systems. The licensee entered this issue into the corrective action program as Condition Report CR-CNS-2017-04229.The inspectors determined that the licensees failure to assure surveillance test procedures for safety-related ventilation systems incorporated test instrument uncertainty into acceptance criteria was a performance deficiency. Because the systems involved in this performance deficiency were systems that mitigate the consequences of accidents, the inspectors evaluated the finding under the Mitigating Systems Cornerstone. In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined that the performance deficiency was more than minor, and therefore a finding, because it was a programmatic deficiency which adversely impacted the procedure quality 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 acceptance criteria for the licensees safety-related ventilation systems did not assure the availability of these systems to respond to accident conditions, as required by the technical specifications. The inspectors assessed the significance of this finding in accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, and determined this finding was of very low safety significance (Green) because it: was not a design deficiency; did not represent a loss of system and/or function; did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time; and did not result in the loss of a high safety-significant, nontechnical specification train. The finding had a cross-cutting aspect in the area of human performance associated with documentation because the licensee failed to ensure that the organization created and maintained complete, accurate, and up-to-date documentation (H.7).
05000321/FIN-2017002-04Hatch2017Q2Licensee-Identified ViolationTS 3.6.4.1 requires secondary containment be operable in Mode 1 and during movements of irradiated fuel assemblies in the secondary containment. Contrary to the above, on February 8 at 1035, with Unit 1 operating at 100 percent RTP and Unit 2 conducting refueling operations, secondary containment was made inoperable when Unit 2 reactor building containment was breached for a scheduled refueling outage and a configuration control error on the Unit 2 standby gas treatment system provided a uncontrolled opening into the secondary containment for the Unit 1 reactor building and the common refueling floor. A temporary blind flange had been incorrectly installed on the upstream side vice downstream side of the Unit 2 standby gas treatment inlet isolation valve when the valve had been removed from the system for testing. This configuration rendered secondary containment for the Unit 1 reactor building and the common refueling floor inoperable. A senior reactor operator performing a plant tour noted the incorrect flange configuration and at 2017 on February 17, the blind flange was moved to the downstream side of the Unit 2 standby gas treatment inlet isolation valve to restore compliance. Inspectors screened the finding in accordance with IMC 609 Appendix A The Significance Determination Process (SDP) for Findings at-Power. The finding screened as very low safety significance (Green) because the questions in Appendix A Exhibit 3 for Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building, were answered no. This issue was documented in the licensees corrective action program as CR 10332592.
05000293/FIN-2017002-05Pilgrim2017Q2Damper 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)
05000341/FIN-2017009-01Fermi2017Q1Failure to Maintain the Effectiveness of the Sites Emergency PlanPreliminary White. An NRC identified finding preliminarily determined to be of low to moderate safety significance (White), and an associated apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) and 10 CFR 50.47(b)(9) was identified for the licensees failure to maintain the effectiveness of its emergency plan and use adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency. Specifically, the licensee failed to maintain the ability to accurately declare an Emergency Action Level (EAL) classification, RG-1.1, and develop and issue accurate protective action recommendations (PARs) during the implementation of the sites Emergency Plan in response to a rapidly progressing accident. The licensee inaccurately analyzed the effect of increasing background radiation on the sites Standby Gas Treatment System accident range radiation monitor (AXM) indications based on the installed configuration of the AXM. As configured, the AXM could provide inaccurate indications of radioactive releases that are used as the licensees basis for determining EAL classification and development of PARs. The licensee documented the issue in the corrective action program as CR-16-29230, and actions were completed to restore the accuracy of the indications provided by the AXM. The inspectors determined that the licensees failure to maintain the effectiveness of its emergency plan and use adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency was a performance deficiency; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors determined the issue was more than minor because it adversely affected the emergency preparedness cornerstone objective to ensure 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 finding would result in the potential over classification of an emergency event and the potential issuance of unnecessary or early PARs. 3 The inspectors applied Inspection Manual Chapter (IMC) 0609, Appendix B, Section 5.9. to screen this finding, and determined the licensee failed to maintain the risk significant planning standard (RSPS) identified in 10 CFR 50.47(b)(9) by ensuring adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use. Using Table 5.9-1, the inspectors determined the sites dose assessment process was incapable of providing technically adequate estimates of radioactive material releases to the environment or projected offsite doses in some cases (specifically a rapidly progressing accident scenario). This significance example corresponds to a Degraded RSPS Function, which is a finding of low to moderate safety significance (White). The inspectors determined no cross-cutting aspects were associated with the performance deficiency.
05000341/FIN-2016004-04Fermi2016Q4Inadequate Testing of SGTS FiltersThe inspectors identified a finding of very low safety significance with an associated NCV of TS 5.5.7, Ventilation Filter Testing Program. The licensee failed to perform testing of the standby gas treatment system (SGTS) high-efficiency particulate air (HEPA) filters that demonstrated a penetration and system bypass of less than 0.05 percent. The licensee entered this violation into its CAP as CARD 1628812. The licensee declared the Division 1 SGTS subsystem inoperable until testing was performed satisfactorily and evaluated the extent of condition on the control room filtration system. This performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute for the control room and auxiliary building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not adequately testing the SGTS HEPA filters, the ability of the SGTS to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured. The finding was determined to be of very low safety significance because it involved only a degradation of the radiological barrier function provided by the SGTS. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
05000298/FIN-2016004-03Cooper2016Q4Failure to Maintain Service Water Pump Maintenance ProcedureThe inspectors reviewed a self-revealed, non-cited violation of Technical Specification 3.6.4.2, Secondary Containment Isolation Valves, for the licensees failure to maintain secondary containment isolation valve HV-AOV-265 operable as a result of erecting scaffolding that interfered with valve operation. Specifically, between June 29, 2016, and September 14, 2016, the licensee erected scaffolding in close proximity of valve HV-AOV-265, such that, during valve stroking, the scaffolding would pinch the actuator air line and prevent the valve from closing, rendering the valve inoperable for approximately 10 weeks. This resulted in the licensees need to reduce power to approximately 50 percent in order to comply with technical specifications upon discovery. Immediate corrective actions included removal of the scaffolding, replacement of the pinched air line, and restoration of the valve to operable status. The licensee entered this deficiency into the corrective action program as Condition Report CR-CNS-2016-05608 and initiated a root cause evaluation to investigate this condition. The licensees failure to implement Procedure 7.0.7, Scaffolding Construction and Control, Revision 34, to ensure scaffolding did not adversely affect plant equipment, in violation of Technical Specification 3.6.4.2, was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the structure, system, and component and barrier performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the improperly erected scaffolding prevented the operation of a secondary containment isolation valve, rendering it inoperable for approximately 10 weeks. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, the inspectors determined that the finding had very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room, reactor building, spent fuel pool building, or standby gas treatment system. The finding had a cross-cutting aspect in the area of human performance associated with resources. Specifically, the licensee failed to ensure that personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety (H.1).
05000410/FIN-2016001-02Nine Mile Point2016Q150.65(a)(4) Risk Evaluation Not Properly Performed Prior to Residual Heat Removal Heat Exchanger TestingThe inspectors identified a Green non-cited (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for Exelons failure to take risk management actions (RMAs) as required by procedure OP-AA-108-117, Protected Equipment Program, Revision 004, during a Unit 2, Division III, emergency switchgear electrical maintenance window on January 27, 2016. Specifically contrary to procedure OP-AA-108-117, during planned maintenance, Exelon failed to post the unit coolers in the A and B RHR pump and HX rooms, the C RHR pump room, and their associated breakers as protected equipment although their inoperability would have resulted in both trains of the standby gas treatment system (SBGT) being inoperable which would require entry into Technical Specification (TS) Limiting Condition for Operation (LCO) 3.0.3 and a short term shutdown action statement. Upon identification, Exelon generated IR 02617915 to document this issue. Corrective actions included creating an action item to evaluate Attachment 3 of N2-OP-52 and to determine the relevance of the TS LCO 3.0.3 entry requirement. The inspectors determined the performance deficiency to be more than minor because it was associated with the structure, system, and component (SSC) and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, contrary to OP-AA-108-117, Exelon personnel failed to include the unit coolers for the Unit 2 RHR pump and HX rooms and their associated breakers, whose unavailability would have resulted in the inoperability of both trains of SBGT and necessitated entry into LCO 3.0.3. Additionally, Examples 7.e, 7.f, and 7.g from IMC 0612, Appendix E, Examples of Minor Issues, provided similar scenarios to this issue. Example 7.e details that a performance deficiency is more than minor if a failure to include accurate TS requirements in a risk assessment and if done properly, would have required RMAs, or additional RMAs under applicable plant procedures. The inspectors evaluated the finding using Phase 1, Initial Screening and Characterization worksheet in Attachment 4 to IMC 0609, Significance Determination Process. For findings within the Initiating Events, Mitigating Systems, and Barrier Integrity cornerstones, Attachment 4, Table 3, Paragraph 5.C, directs that if the finding affects the licensees assessment and management of risk associated with performing maintenance activities under all plant operating or shutdown conditions in accordance with Baseline Inspection Procedure 71111.13, Maintenance Risk Assessment and Emergent Work Control, the inspectors shall use IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, to determine the significance of the finding. The inspectors used Flowchart 2, Assessment of RMAs, to analyze the finding and calculated incremental core damage probability using EOOS, Exelons risk assessment tool, and found the result to be less than 1E-6. The inspectors determined that had this condition existed for the full duration of the TS LCO, the incremental core damage probability would have been 6.8E-7. Because the incremental core damage probability deficit was less than 1E-6 and the incremental large early release probability was less than 1E-7, this finding was determined to be of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Exelon failed to follow processes, procedures and work instructions. Specifically, Exelon failed to follow procedure OP-AA-108-117, which led to the failure to protect the unit coolers for the RHR pump rooms, HX rooms, and associated breakers which could have led to a TS LCO 3.0.3 entry.
05000352/FIN-2016001-01Limerick2016Q1Reactor Enclosure Recirculation System Design Change not EvaluatedA self-revealing Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50 (10 CFR 50), Appendix B, Criterion III, Design Control, was identified because Exelon did not properly maintain the design of the LGS Unit 1 reactor enclosure recirculation system (RERS). Specifically, Exelon replaced the Unit 1 1A RERS flow straightener assembly using thinner material than was originally qualified and did not evaluate the change in design. Exelon initiated IR 2563872 and implemented a temporary configuration change that removed the flow straightener assembly from the system and restored Unit 1 RERS to operability on October 5, 2015. Exelon also initiated corrective actions to install a new flow straightener assembly with correctly sized honeycomb material. This finding is more than minor because it adversely affected the design control attribute of the barrier integrity cornerstone to provide reasonable assurance that physical design barriers (secondary containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate 1A RERS flow straightener assembly installed in 2012 resulted in degraded performance and then unplanned unavailability of 1A RERS from October 1 to 5, 2015. Using IMC 0609, Appendix A, Exhibit 3, the inspectors determined that this finding was of very low safety significance (Green). Specifically, the degraded 1A RERS performance and associated unavailability only represented a degradation of the radiological barrier function provided for the standby gas treatment system and screened to Green. The inspectors determined that the finding did not have cross-cutting aspect because the performance deficiency did not occur within the last three years, and the inspectors did not conclude that the primary cause of the performance deficiency represented present Exelon performance.
05000333/FIN-2016001-03FitzPatrick2016Q1Inadequate Post-Maintenance Testing of the Reactor Building Ventilation System Resulted in Short-Term Inoperability of Secondary ContainmentThe inspectors identified a self-revealing NCV of TS 5.4, Procedures, for FitzPatrick staffs failure to perform adequate post-maintenance testing (PMT) following maintenance on a limit switch in the reactor building ventilation system in August 2014, that, along with another unrelated component failure in the reactor building ventilation system, resulted in secondary containment pressure, relative to the outside pressure, exceeding the TS limit of 0.25 inches of vacuum water gauge. As immediate corrective action, operators started both trains of the standby gas treatment system (SBGTS), which restored secondary containment pressure to within the TS limit. Operators subsequently secured the A refuel floor exhaust train and placed the B train in service. The issue was entered into the CAP as CR-JAF-2015-04166. The finding was more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, as a result of this event, secondary containment was not preserved, in that secondary containment pressure exceeded the limit of TS surveillance requirement (SR) 3.6.4.1.1. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, did not involve an actual reduction in function of hydrogen igniters in the reactor containment, and only represented a degradation of the radiological barrier function provided by the reactor building and SBGTS. The finding had a cross-cutting aspect in the area of Human Performance, Resources, because FitzPatrick staff did not ensure that procedures for PMT of the reactor building refuel floor exhaust damper limit switch following maintenance performed in August 2014, were adequate to support the nuclear safety function of the secondary containment (H.1).
05000341/FIN-2016001-07Fermi2016Q1Inadequate Test Criteria in SGTS Flow/Heater Operability Surveillance TestThe inspectors identified a finding of very low safety significance with an associated NCV of 10 CFR 50, Appendix B, Criteria V, Instructions, Procedures, and Drawings. Specifically, the licensee failed to include appropriate quantitative or qualitative acceptance criteria in its surveillance test procedures for fulfilling the monthly Technical Specification surveillance requirement to demonstrate operability of the standby gas treatment system (SGTS). The licensee entered this violation into its corrective action program to evaluate the issue and identify appropriate corrective actions. No immediate operability concern was identified. The performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute for the control room and auxiliary building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not providing appropriate acceptance criteria by which the operability of the SGTS trains could be assessed, the ability of the SGTS to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured. The finding was determined to be of very low safety significance because it involved only a degradation of the radiological barrier function provided by the SGTS. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
05000352/FIN-2015004-02Limerick2015Q4Licensee-Identified ViolationTechnical Specification 3.6.5.3, Standby Gas Treatment System Common System, requires with one SGTS subsystem, restore the inoperable subsystem to operable status within 7 days, or be in at least hot shutdown within the next 12 hours and in cold shutdown within the following 24 hours. Contrary to Technical Specification 3.6.5.3, SGTS subsystem B was inoperable for Unit 1 from August 27, 2015, to September 4, 2015, for a time of 8 days 18 hours, and Exelon did not place Unit 1 in hot shutdown or cold shutdown. Exelon entered this issue into the corrective action program as IR 2517538. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. 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 of the SGTS. In addition, the inoperable condition would have resulted in a flowrate exceeding the analyzed 2500 cfm with a differential pressure greater than the minimum 0.25 inches of vacuum water gauge. However, the condition did not represent a larger pathway through secondary containment and SGTS retained radiological filtering capability.
05000333/FIN-2015003-02FitzPatrick2015Q3Inadequate Instructions for Reactor Building Roof Replacement Result in Inadvertent Loss of Secondary ContainmentThe inspectors identified a self-revealing violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because FitzPatrick staff failed to provide instructions appropriate to the reactor building roof replacement project. Specifically, inadequate instructions were provided to ensure that roofing material removal would be performed in slow, deliberate manner, such that its effect on secondary containment could be assessed and operability maintained. As a result, this activity caused secondary containment to be inoperable for a period in excess of its four hour technical specification (TS) allowed outage time. As immediate corrective action, roofing material removal was stopped and the new roofing materials were installed to reseal the affected area of the reactor building roof. Secondary containment vacuum was restored to greater than the TS-required minimum after a period of 92 minutes and secondary containment was declared operable after a period of five hours and 26 minutes. The issue was entered into the corrective action program (CAP) as CRJAF- 2015-03260. The finding was more than minor because it is associated with the procedure quality attribute of the Barrier Integrity cornerstone, and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system (RCS), and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the work order (WO) did not provide adequate instruction to ensure that roofing material removal would be performed in slow, deliberate manner, coordinated between operations and maintenance personnel, and allowing adequate time after actions that could impact secondary containment such that their effect on secondary containment could be assessed and operability maintained. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency was not a pressurized thermal shock issue, did not represent an actual open pathway in the physical integrity of the reactor containment, did not involve an actual reduction in function of hydrogen igniters in the reactor containment, and only represented a degradation of the radiological barrier function provided by the reactor building and standby gas treatment system. The finding had a cross-cutting aspect in the area of Human Performance, Avoid Complacency, because FitzPatrick staff did not adequately plan for the possibility of latent issues and inherent risk associated with the reactor building roof replacement project, such that the commencement of work resulted in a loss of secondary containment.
05000341/FIN-2015003-04Fermi2015Q3Failure to Perform Preventive Maintenance on SafetyRelated Auxiliary Trip Unit Relays for the Spent Fuel Pool Ventilation Exhaust Radiation MonitorsA finding of very low safety significance with an associated Non-Cited Violation of Technical Specification 5.4, "Procedures," was self-revealed on May 16, 2015, when the failure of an auxiliary trip unit relay for the Division 2 spent fuel pool ventilation exhaust radiation monitor caused an invalid actuation of primary and secondary containment isolation valve logic for numerous valves in the drywell and suppression pool ventilation and nitrogen inerting systems, and an invalid engineered safety features system actuation of the standby gas treatment system and control center heating, ventilation, and air conditioning system. The licensee failed to perform any replacement preventive maintenance for the component throughout the history of plant operation. The licensee subsequently replaced the failed relay and returned the Division 2 spent fuel pool ventilation exhaust radiation monitor to service. In addition, the licensee initiated a corrective action to create preventive maintenance activities to replace all potentially age-degraded auxiliary trip unit relays and to create new preventive maintenance strategies for relays not currently within the scope of its preventive maintenance template. The finding was of more than minor safety significance because it was associated with the Equipment Performance 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 (i.e., core damage). Specifically, the age-related auxiliary trip unit relay failure unnecessarily challenged actuation of engineered safety features and resulted in inoperable safety-related equipment until maintenance was completed to replace the failed relay. The finding was determined to be a licensee performance deficiency of very low safety significance. Although the issue affected the design or qualification of a mitigating system or component, failure of the auxiliary trip unit relay did not result in the loss of safety function of any safety-related structure, system, or component but instead resulted in invalid actuation of safety features. The inspectors concluded this finding affected the cross-cutting area of problem identification and resolution and the cross-cutting aspect of operating experience (IMC 0310, P.5). Specifically, the licensee did not appropriately evaluate and implement relevant internal and external operating experience to appropriately adjust its preventive maintenance program to replace auxiliary trip unit relays.
05000220/FIN-2015009-01Nine Mile Point2015Q3Failure to Identify and Correct a Condition Adverse to Quality Associated with Secondary Containment LeakageThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion XVI, Corrective Actions, because between 2007 and 2015, Exelon staff did not promptly identify and correct a deficiency associated with Unit 2 reactor building service water pipe penetration W-3177-C. Specifically, on August 20, 2015, during Exelon staffs investigation of an inspector concern associated with the service water pipe penetration into secondary containment, a leakage path into secondary containment was discovered and was not previously identified and evaluated for impact on operability of Unit 2 secondary containment. Exelon generated issue report (IR) 2544831 to document the newly identified condition. The assessment included a review of previously identified leakage paths that were being tracked in accordance with procedure, previously performed secondary containment drawdown leakage testing, and a comparison to the maximum allowable flow rate leakage area. The assessment concluded that based on the gap that was identified at secondary containment penetration W-3177-C, there was a new total of 1.783 square inches of surface area allowing leakage into the Unit 2 secondary containment. Exelon determined this to be acceptable because calculations for secondary containment drawdown testing allows for up to 33.6 square inches of surface area causing in-leakage into secondary containment. Given 1.783 square inches of total identified leakage being less than the allowable 33.6 square inches, there was reasonable assurance that standby gas treatment system will be able to perform its drawdown function and maintain secondary containment vacuum 0.25 inches of vacuum water gauge in accordance with Technical Specification (TS) 3.6.4.1, Secondary Containment. This performance deficiency was more than minor because it impacted the design control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, Exelons staff failed to identify the degraded penetration seal that impacted the reasonable assurance of Unit 2 secondary containment operability. 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 this finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, or auxiliary, spent fuel pool, or standby gas treatment system (i.e., secondary containment). This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon staff failed to properly evaluate the condition identified in multiple IRs to determine the extent of condition associated with secondary containment water in-leakage. Specifically, between 2007 and 2015, three IRs were generated and a 2012 structural monitoring review documented the service water penetration water in-leakage and the issue was not appropriately evaluated for the potential for a service water pipe through-wall leak or the potential impact on secondary containment.
05000461/FIN-2015003-03Clinton2015Q3Failure to Obtain a License Amendment prior to Making Modifications to Secondary ContainmentThe inspectors identified a Severity Level IV NCV of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, on December 18, 2014, for the licensees failure to provide a written evaluation showing that a change to the secondary containment did not require a license amendment. Specifically, the licensee eliminated the tornado wind and tornado missile loading conditions from the fuel building railroad airlock enclosure walls, roof and associated outer door Seismic Category I design requirements. However, the licensee failed to provide a written evaluation describing why the change would not result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system or component important to safety. The licensee documented the issue in the CAP as AR 02534694. Corrective actions included complying with TS anytime the inner railroad bay door is opened by entering the applicable action statements, evaluating weather conditions and impact to plant risk and establishing the necessary mitigating actions required prior to opening the door. The inspectors determined that the licensees failure to provide a written evaluation that documented the basis for determining that the change to the secondary containment did not require a license amendment was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Barrier Integrity cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system and containment) protect the public from radionuclide releases caused by accidents or events. In addition, the associated violation was determined to be more than minor because the inspectors could not reasonably determine if the changes to secondary containment would have required prior NRC approval. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process instead of the significance determination process (SDP) because they are considered to be violations that potentially impede or impact the regulatory process. However, if possible, the underlying technical issue is evaluated under the SDP to determine the severity of the violation. In this case, the finding was screened against the Barrier Integrity cornerstone, and determined to be of very low safety significance because the finding did not represent a degradation only of the radiological barrier function for the standby gas treatment system nor did it represent a degradation of the function of the control room against smoke or toxic atmosphere. The inspectors determined this finding has a cross-cutting aspect of procedure adherence in the area of human performance because the licensee failed to follow the 50.59 regulatory process as defined in procedure LS-AA-104-1000, 50.59 Resource Manual, Revision 9. (H.8)
05000263/FIN-2015002-03Monticello2015Q2Failure to Maintain Secondary Containment and Standby Gas Treatment System Operable During OPDRV ActivitiesThe inspectors identified a finding of very low safety significance and an associated NCV of TS 3.6.4.1, Secondary Containment and TS 3.6.4.3, Standby Gas Treatment System (SBGT) because the licensee did not maintain secondary containment and the SBGT system operable as required during activities considered OPDRVs. Specifically, on April 14, 2015, and again on May 13, 2015, the licensee failed to classify activities associated with draining reactor inventory as OPDRVs while relying on an automatic isolation function for the drain path, and as a result failed to maintain required equipment operable during these activities. Once questioned by the inspectors, the licensee took action to control other outage related draining activities as OPDRVs and placed this issue into its CAP (CAP 1479284). The inspectors determined that the failure to maintain secondary containment and SBGT operable while an OPDRV was in progress was a performance deficiency. The performance deficiency was more than minor because it was 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, RCS, and containment) protect the public from radionuclide releases caused by accidents or events because the secondary containment boundary and the SBGT were not maintained operable during an OPDRV activity. The inspectors evaluated the finding using IMC 0609, Attachment 4, Phase 1 Initial Screening and Characterization of Findings, which required an analysis using IMC 0609 Appendix G, the Shutdown Operations SDP since the reactor was shut down. The finding was assessed in accordance with IMC 0609 Appendix G, Attachment 1, Exhibit 4 and Appendix H for containment integrity findings. Using Appendix H, the inspectors concluded the finding had very low safety significance (Green) because decay heat was low and containment was deinerted. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Documentation aspect because of the failure of the licensee to create and maintain complete, accurate and up-to-date documentation (H.7).
05000219/FIN-2015001-04Oyster Creek2015Q1Licensee-Identified ViolationTechnical Specification 3.5.B, Secondary Containment, requires in part, that secondary containment integrity be maintained at all times when the reactor vessel head and the drywell head are in not in place. Technical Specification 1.14, Secondary Containment Integrity, requires in part, that the standby gas treatment system is operable. Technical Specification 4.5.G.3 specifies that with the trunnion room door open and the trunnion room is isolated from secondary containment in support of outage activities, testing of the standby gas treatment system to be performed to demonstrate the capability to maintain 14 inch of water vacuum under calm wind conditions and a standby gas treatment system filter train flow rate of not more than 4000 cfm. Contrary to Technical Specification 3.5.B, on September 20, 2014, with the reactor vessel head and drywell head removed for the refueling outage, Exelon determined that they did not have secondary containment integrity when performing testing to demonstrate standby gas treatment system capability in accordance with Technical Specification 4.5.G.3 and subsequently found that the outer railroad air lock personnel access hatch had not been closed properly, which prevented a proper vacuum from being achieved. Exelon entered this issue into the corrective action program as IR 2383852. Using guidance in IMC 0609, Appendices G and H, the inspectors determined that this finding was of very low safety significance (Green) because the decay heat values were low and the reactor water level inventory was above that required to move irradiated fuel.
05000387/FIN-2014005-06Susquehanna2014Q4Licensee-Identified ViolationSecondary Containment Door Found Ajar On February 12, 2014, PPL identified a secondary containment door (Door 612) between the HVAC room and central railroad bay wedged open by a door sign. In order for secondary containment to be operable in the as-found mode of operation, Door 612 had to be secured. PPL immediately secured the door, entered the condition into their CAP (2014-04709), and reported the condition under LER 50-387; 388/2014-002. Contrary to TS 5.4.1a, PPL did not secure the secondary containment door and maintain system operability in accordance with OP-134-002, RB HVAC Zones 1 and 3 after realignment of the secondary containment. The finding was more than minor because it adversely impacted the barrier performance attribute of barrier integrity and was determined to be of very low safety significance (Green) in accordance with IMC 0609, Appendix A, since the finding only represented a degradation of the radiological barrier function provided by standby gas treatment system.
05000387/FIN-2014004-03Susquehanna2014Q3Adequacy of Secondary Containment and Standby Gas Treatment System TestingThe NRC communicated the results of NRRs TIA on this subject by a memorandum dated May 6, 2014 (ADAMS ML14085A411). Subsequently, PPL initiated CR-1256036 to evaluate the need to change the affected TS SRs. In addition, PPL implemented the guidance provided in NRC Administrative Letter (AL) 98-10, Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety, which includes taking actions (e.g., compensatory measures, appropriate SGTS testing) during the interim. As stated in NRC Inspection 05000387&388/2012007, PPL implemented several actions to provide reasonable assurance that the secondary containment and the SGTS were capable of performing their intended functions including a temporary system modification to align all three zones of ventilation to the SGTS if either unit received an accident signal. PPLs interim actions ensured that the safety function of secondary containment would be maintained. The inspectors determined there was no performance deficiency because the issue of concern was not reasonably within PPLs ability to foresee and correct in that the NRC had explicitly approved the secondary containment testing methodology, which the station had been applying in compliance with TSs. Accordingly, NRC IMC 0612, Appendix B, Issue Screening, directs disposition of this issue using traditional enforcement in accordance with the Enforcement Policy. The inspectors used Enforcement Policy, Section 6.1.d.1, Reactor Operations, to evaluate the significance of this violation, and concluded that the violation was more than minor and best characterized as a Severity Level IV violation in that the issue was associated with allowances for surveillance requirements in Section 3.0 of TSs. In reaching this conclusion, the inspectors considered that the underlying technical finding would have been evaluated as having very low safety significance (i.e., Green) under the Reactor Oversight Process using NRC IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, dated June 19, 2012, since the issue was only associated with the radiological barrier function of the auxiliary building and SGTS. 10 CFR 50.36(c)(3), Surveillance Requirements, requires, in part, that surveillance requirements related to testing assure that the necessary quality of systems and components are maintained and that the limiting conditions for operation will be met. Contrary to the above, from March 1984 to the present, the existing text for TS SRs 3.6.4.1.4 and 3.6.4.1.5 did not adequately demonstrate: (1) the quality of the secondary containment: and (2) that the LCOs were met.
05000410/FIN-2014004-01Nine Mile Point2014Q3Loss of Secondary Containment due to Loss of Auxiliary Boiler SystemThe inspectors identified a Green finding (FIN) of CNG-PR-1.01-1005, Control of Technical Procedure Format and Content, Revision 00500, because Exelon Generation Company, LLC (Exelon) provided Unit 2 operators with an inadequate auxiliary boiler system operating procedure. Specifically, N2-OP-48, Auxiliary Boiler System, Revision 01100.00, did not provide operators adequate detail to properly establish chemistry requirements for water conductivity of the auxiliary boiler system. On March 23, 2014, when Unit 2 experienced a trip of the auxiliary boiler system due to inadequate water conductivity, operators became challenged with system restoration which caused an unplanned loss of secondary containment and entry into Technical Specification (TS) 3.6.4.1, Secondary Containment. Exelon generated condition report (CR)-2014-002281 regarding this issue. Immediate corrective actions included updating chemistry requirements associated with auxiliary boiler procedures, implementing new preventive maintenance (PM) strategies for significant components associated with the auxiliary boilers, and implementing new performance monitoring plans. This finding is more than minor because it affected the procedure quality attribute of the Barrier Integrity cornerstone and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, over the past 2 years, the auxiliary boilers have experienced trips as a result of insufficient procedural guidance. On March 23, 2014, the inadequate procedural guidance resulted in a trip and subsequent loss of reactor building (RB) differential pressure (DP). This caused an unplanned entry into the secondary containment emergency operating procedure and an unplanned entry into TS 3.6.4.1, which presented unnecessary challenges and distractions to operators during a planned downpower. In accordance with IMC 0609.04, Initial Characterization of Findings, the inspectors used IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, because secondary containment was declared inoperable following a loss of building heating. Using Appendix A, Exhibit 3, Barrier Integrity Screening Questions, Section C, Control Room, Auxiliary, Reactor, or Spent Fuel Pool Building, the inspectors determined that this finding is of very low safety significance (Green) because although the performance deficiency resulted in a trip of the auxiliary boiler system and a loss of secondary containment, the RB DP was restored to greater than 0.25 inches of water, within the allowable limiting condition for operation time, and did not result in a failure of the ability for secondary containment to maintain isolation or impact the ability for standby gas treatment system to maintain secondary containment. This finding has a cross-cutting aspect in the area of Human Performance, Resources, because Exelon did not ensure personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. Specifically, the inadequate management oversight of the auxiliary boilers resulted in numerous failures of the auxiliary boilers due to inadequate knowledge transfer, inaccurate classifications of maintenance rule functional failures for the system, inadequate procedures for boiler operation, and inadequate procedures for the prompt restoration of secondary containment when the auxiliary boiler system is not available (H.1).
05000271/FIN-2014004-04Vermont Yankee2014Q3Licensee-Identified Violation10 CFR 50, Appendix B, Criterion III, Design Control, requires, in part, that the design basis is correctly translated into specifications. Contrary to the above, the design basis was not correctly translated into specifications in that the specified current rating for electrical switches supplying electric heaters in the standby gas treatment system were less than the designed circuit amperage from original plant construction, February 28, 1973, until November 13, 2013. Entergy identified that the standby gas treatment auxiliary switches supplying the charcoal bed heaters in both A and B subsystems were rated for a continuous current of 3 A when the continuous current was approximately 8 A. Entergy entered this issue into the corrective action program as CR-VTY-2013-06257. The inspectors determined that the finding was of very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, because the finding only represented a degradation of the radiological barrier function provided for the standby gas treatment system.
05000397/FIN-2014003-06Columbia2014Q2Failure to Properly Pre-Plan Maintenance on Reactor Building Ventilation Differential Pressure ControllersThe inspectors reviewed a Green non-cited violation for the licensee's failure t properly pre-plan calibrations of differential pressure controllers used to maintain secondary containment pressure. Specifically, the licensee failed to establish and maintain the appropriate gain settings for the reactor building normal ventilation system differential pressure controllers in accordance with procedure DES-2-19, Instrument Master Data Sheets, Revision 0. As a corrective action, the licensee properly adjusted the gain settings for the affected controllers. The licensee also entered this issue into their corrective action program as AR 300787. This performance deficiency was more than minor because it affected the equipment performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to establish and maintain configuration control of reactor building ventilation differential pressure controllers resulted in multiple instances of unplanned inoperability of secondary containment. The finding is of very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for by the standby gas treatment system. This finding had a cross-cutting aspect in the area of problem identification and resolution because the licensee failed to thoroughly evaluate concerns related to the operation of the normal reactor building differential pressure controller such that the resolution addressed the causes of the observed sluggish response (P.2). (Section 4OA3)
05000321/FIN-2014003-01Hatch2014Q2Failure to Prove Operability Following the Failure of the Secondary Containment Surveillance TestThe inspectors identified a Green non-cited violation of 10 CFR 50, Appendix B, Criterion V, Procedures, Instructions, and Drawings, for the licensees failure to prove operability following a failure of a surveillance test as required by Hatch procedure 90ACOAM- 001-0, Test and Surveillance Control, Ver. 1.0, on May 12, 2014. To restore compliance, the licensee isolated the refueling floor dampers and re-performed Surveillance Requirement 3.6.4.1.3 with satisfactory results later that day on May 12, 2014. This violation was entered into the licensees corrective action program as condition report (CR) 819563. Failure to prove operability following failure of a surveillance test as required by Hatch procedure 90AC-OAM-001-0, Test and Surveillance Control, Ver. 1.0, on May 12, 2014, was a performance deficiency. The performance deficiency affected the barrier integrity cornerstone and was more-than-minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, declaring equipment operable following a failed surveillance test would have the potential for the facility to operate outside of technical specification requirements. The inspectors screened this finding using IMC 0609, Appendix A, The Significant Determination Process (SDP) For Findings At-Power , dated June 19, 2012. The finding screened as Green per Section C of Exhibit 3, Barrier Integrity Screening Questions, because the finding only represented a degradation of the radiological barrier function provided by the standby gas treatment system. The inspectors determined the finding had a cross-cutting aspect of training in the human performance area, because the licensee did not ensure knowledge transfer of Surveillance Requirement 3.0.1 requirements to maintain a knowledgeable, technically competent workface and instill nuclear safety values.
05000324/FIN-2014002-02Brunswick2014Q1Unit 2 Secondary Containment Inoperable Due to Door GapsTS 3.6.4.1, Secondary Containment, requires secondary containment to be operable during Modes 1, 2, and 3. Contrary to the above, from September 18, 2013, to September 24, 2013, and from October 2, 2013, to October 15, 2013, Unit 2 Secondary Containment had door gaps exceeding the allowable leakage criteria. During periods in which the degraded doors were relied upon for Secondary Containment pressure boundary, the Secondary Containment was inoperable. This violation screened to IMC 0609, Appendix A, The SDP for Findings at Power, Exhibit 3 Barrier Integrity Screening Questions. This violation was determined to be very low afety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the spent fuel pool and standby gas treatment system. The licensee entered this issue into their CAP as NCR 641834. The license took action to repair the degraded doors, translate the design basis requirements for secondary containment into acceptance criteria for the doors in Engineering Change 93918, Determine Reactor Building Pressure Boundary Door Seal Inspection Acceptance Criteria, and revised Procedure 0PT-34.2.2.1, Fire Door, Pressure Boundary Door, Alternate Safe Shutdown Access/Egress Door, and Severe Weather/Flood Control Door Inspections, to incorporate the criteria for Reactor Building pressure boundary doors.
05000397/FIN-2014002-03Columbia2014Q1Inadequate Operability Evaluation of Degraded Reactor Core Isolation Cooling ValveThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, for the failure of the licensee to perform a detailed examination of a degraded condition associated with the reactor core isolation cooling system in accordance with the station operability determination Procedure PPM 1.3.66, Operability and Functionality Determinations, Revision 29. For an immediate corrective action, the licensee reassessed the condition for operability. The licensee entered this issue into their corrective action program as Action Request 303216. The performance deficiency was more than minor because it affected the equipment performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors performed an initial screening of the finding in accordance with NRC Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. Using NRC Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions, the inspectors determined this finding is of very low safety significance (Green) because the finding represents only a degradation of the radiological barrier function provided for by the standby gas treatment system. This finding has a cross-cutting aspect in the area of human performance because the licensee rationalized the unexpected plant response when performing reactor core isolation cooling system surveillance testing and relied on previous, unrelated evaluations as justification of system operability instead of challenging the unknown (H.11).
05000387/FIN-2013005-03Susquehanna2013Q4Missed Technical Specification Surveillance for Secondary Containment Drawdown TestingThe inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XI, Test Control, because PPL did not ensure all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service was identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, PPLs procedure used to implement the requirements of TS Surveillance Requirements (SR) 3.6.4.1.4 and 3.6.4.1.5 did not ensure that secondary containment integrity was tested in all required configurations. PPLs immediate corrective actions included entering the issue into their CAP as CR-2013-03891 and applied a status control tag to the railroad access bay door-101 as an administrative control until corrective actions can be completed and the configuration tested satisfactorily. The finding is more than minor because it is associated with the procedure quality attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate surveillance procedure resulted in missed surveillances for SRs 3.6.4.1.4 and 3.6.4.1.5. Additionally, it was similar to example 3.d in IMC 0612 Appendix E, Examples of Minor Issues, in that the failure to implement the TS SR as required is not minor if the surveillance had not been conducted. In this case, the surveillance requirement had not been completed for all configurations of secondary containment. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The SDP for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the performance deficiency only represented a degradation of the radiological barrier function provided for the Standby Gas Treatment system. This finding was determined to have a cross-cutting aspect in the area of Human Performance Resources area because the licensee failed to ensure that personnel, equipment, procedures, and other resources are available and adequate to assure nuclear safety. Specifically, those necessary for: complete, accurate and up-to-date design documentation, procedures, and work packages, and correct labeling of components (H.2(c)).
05000354/FIN-2013004-01Hope Creek2013Q3Failure to Follow PMT Procedure Prior to Returning the B FRVS Recirculation Fan to ServiceA self-revealing finding of very low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulation (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified for PSEGs failure to properly test the B Filtration, Recirculation and Ventilation System (FRVS) recirculation fan following maintenance in accordance with site procedures. Specifically, on June 3, 2013, PSEG did not perform the required post-maintenance test (PMT) prior to returning the system to service. Consequently, when the fan failed during its surveillance on June 24, 2013, there was no reasonable assurance that the fan was operable since the last time maintenance was performed on it. Corrective actions included adding this event to the Licensed Operator Requalification training program to improve knowledge regarding PMT requirements. The performance deficiency (PD) was determined to be more than minor because it is associated with the system, structure, or component (SSC) and Barrier Performance attribute of the Barrier Integrity cornerstone, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using IMC 0609, Appendix A, The SDP for Findings At-Power, dated June 19, 2012, the finding was determined to be of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the standby gas treatment system. This finding had a cross-cutting aspect in the area of Human Performance, Decision-Making, because PSEGs decisions did not demonstrate that nuclear safety was the overriding priority. Specifically, PSEG did not use conservative assumptions in decision-making when determining the proper PMT for the B FRVS recirculation fan prior to returning it to service.
05000397/FIN-2013003-02Columbia2013Q2Failure to Follow Corrective Action Program ProceduresThe inspectors identified two examples of a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, associated with the licensees failure to follow the corrective action program procedure by promptly entering conditions adverse to quality into the corrective action program. The first example occurred on March 16, 2013, when the reactor building exhaust air experienced a step reduction in flow due to a stack access door being inadvertently left open. The step change in reactor building exhaust air was not entered into the corrective action program until March 26, 2013. The second example occurred on May 20, 2013, during licensee inspections of reactor vessel internal components. During these licensee inspections, ultrasonic examinations identified cracking on the weld of the core shroud. The inspectors reviewed these inspections on June 3, 2013, and found that no condition reports had been initiated for the identified cracks. Procedurally, station personnel are required to initiate an action request condition report for any actual or suspected conditions adverse to quality no later than the end of shift. Following discussion with the inspectors, engineering personnel initiated action requests to address the indications found on core shroud welds. The licensee initiated Action Requests AR 286688 and AR 287423 to address the timeliness issues involving condition report initiation. The performance deficiency was more than minor, because if left uncorrected, the failure to follow procedures associated with the corrective action program could lead to a more significant safety concern. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) For Findings At-Power, the inspectors determined that the finding was associated with the Barrier Integrity cornerstone and was of very low safety significance because (1) the finding did not involve reactor coolant system pressurized thermal shock issues; (2) the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system or heat removal components; (3) the finding did not involve an actual reduction in function of hydrogen igniters in the reactor containment; and (4) the finding represented a degradation of the standby gas treatment system only in its radiological barrier function for secondary containment. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with the corrective action program component, in that, the licensee failed to implement their program at a sufficiently low threshold. Consequently, the licensee failed to ensure the timely entry of conditions adverse to quality into the corrective action program as required by station procedures.
05000321/FIN-2013002-03Hatch2013Q1Licensee-Identified ViolationA licensee-identified violation of Hatch Unit 2 Technical Specification 3.6.4.3, Standby Gas Treatment System, was discovered by site engineering on December 10, 2012. Technical Specification 3.6.4.3 requires in part that Unit 1 and Unit 2 standby gas treatment subsystems required to support LCO 3.6.4.1, Secondary Containment, shall be OPERABLE in Mode 1. Contrary to this Technical Specification requirement on November 6, 2012, while Unit 2 was operating in Mode 1, the licensee removed the Unit 2 refueling floor hatch which caused the 2B standby gas treatment train flow to exceed 4000 cubic feet per minute. Hatch Unit 2 Surveillance Requirement 3.6.4.1.4 lists the operable flow limit for each standby gas treatment subsystem to be less than or equal to 4000 cubic feet per minute. From November 6, 2012 through November 17, 2012, the 2B standby gas treatment train flow exceeded Surveillance Requirement 3.6.4.1.4 requirements and was therefore inoperable. The licensee entered this issue into their corrective action program as CR 576864. The inspectors screened this violation as Green per IMC 0609, Appendix A, Exhibit 3, question C.1, because this violation represented only a degradation of the radiological barrier function provided for the standby gas treatment system.
05000458/FIN-2012005-03River Bend2012Q4Failure to Implement Effective Corrective Actions for Defects in Masterpact BreakersThe inspectors reviewed a self-revealing, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, for the failure to promptly correct a condition adverse to quality. Specifically, station personnel failed to implement repairs to the mechanism-operated contact linkages for safety-related breakers, ultimately resulting in the failure of standby gas treatment filtration train 1B to start on demand. This issue was entered into the licensees corrective action program as Condition Report CR-RBS-2012-005894. The failure to correct a condition adverse to quality is a performance deficiency. This performance deficiency is more-than-minor because it is associated with the systems, structures, and components and barrier performance attributes 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) protect the public from radionuclide releases caused by accidents or events. Specifically, the standby gas treatment exhaust filter train failed to start during a surveillance test because of a nonconforming mechanical linkage in the feeder breaker resulting in unavailability for standby gas train 1B. In accordance with NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process for Findings At Power, Exhibit 3, Section C, question 1, the finding screened as very low safety significance (Green), because the finding represented only a degradation of the radiological barrier function provided by the standby gas treatment system. No cross-cutting aspect was assigned to this finding because the NRC concluded the finding did not reflect current licensee performance
05000387/FIN-2012007-01Susquehanna2012Q4Adequacy of Secondary Containment and Standby Gas Treatment System TestingAn unresolved item (URI) was identified because additional NRC review and evaluation is needed to determine whether the technical specifications (TS) and associated bases documents were adequate to provide reasonable assurance of operability of the secondary containment boundary and the standby gas treatment system (SGTS). During a review of a modification to the secondary containment boundary, the team identified a potential deficiency related to the adequacy of testing performed to comply with TS surveillance requirements (SR) 3.6.4.1.4 and 3.6.4.1.5, which were associated with verifying secondary containment and SGTS operability. The team questioned whether the allowed configuration for secondary containment during testing could mask potential leakage between the secondary containment zones (through a common boundary) and could potentially make secondary containment inoperable. At Susquehanna, the secondary containments are divided into three zones. Zone 1 is associated with Unit 1, Zone 2 is associated with Unit 2, and Zone 3 encompasses the spent fuel area. These zones are normally maintained at > 0.25 water vacuum via the normal non-safety related ventilation lineup, and the ventilation flow is discharged to the atmosphere via an unfiltered but monitored release path. During a postulated design basis accident, the secondary containment ventilation system is designed to isolate the affected units zone (either Zone 1 or Zone 2) and the spent fuel pool zone (Zone 3). These two zones are automatically placed on recirculation and are discharged through the SGTS via common ventilation ductwork. The unaffected units zone does not isolate and remains on the normal ventilation lineup. During a postulated design basis accident combined with a loss-of-offsite power, all three zones would automatically isolate and be placed on recirculation with discharge via the SGTS. The team reviewed the TS SRs used to verify the operability of secondary containment boundary and the SGTS, as well as the surveillance procedure used to meet TS SRs 3.6.4.1.4 and 3.6.4.1.5. The team found that any of three ventilation line-ups were allowed in the procedure. The acceptable configurations were; 1. All three zones on recirculation connected to the SGTS. 2. Zone 1 (2) and Zone 3 on the recirculation system connected to the SGTS with Zone 2 (1) operable (> 0.25 water column vacuum) and on normal ventilation. 3. Zone 1 (2) and Zone 3 on the recirculation system connected to the SGTS with Zone 2 (1) inoperable (atmospheric pressure). The TS SR acceptance criteria specified that if the SGTS can maintain the tested zone at a vacuum > 0.25 of water in the required time (3.6.4.1.4) and the flow rate through the SGTS is less than the established limit (3.6.4.1.5), then the SRs are considered met. The team found that, as a matter of routine, PPL tested the SGTS as per Items 1 and 2 above to provide reasonable assurance of operability of secondary containment and the SGTS system, and that the test results met the TS surveillance acceptance criteria. The team determined that when the all three zones configuration is used (Item 1 above), the boundaries between zones are not subject to any testing (i.e., the SGTS maintains all three zones at about the same pressure). This configuration does, however, test the secondary containment exterior boundaries and the SGTS for a design basis event with a loss-of-offsite power. Similar to the Item 1 configuration, for either of the two zone configurations (i.e., with the unaffected units non-safety related system in operation, Item 2 above), the common boundary between the zones is not tested. Specifically, the teams review of the test results for this two zone configuration identified that the unaffected units vacuum was better than the units vacuum that was being established by the SGTS. As a result, if there was an opening between the zones (at the common boundary), the normal ventilation lineup would be assisting the SGTS, thus invalidating the test. The team determined that the configuration with one zone inoperable (Item 3 above) represented a valid surveillance test in that it would ensure that potential leakage between the zones would be small enough such that the SGTS could maintain sufficient secondary containment vacuum. However, because this is not the normal configuration, the potential for leakage at the common boundary between the affected and nonaffected zones could exist since normal ventilation creates a better vacuum on the nonaffected unit. This potential leakage would be discharged without filtration to the environment via the normal ventilation system, and manual operator action would be necessary to stop this discharge. In response to the teams concerns, PPL conducted field walkdowns to visually confirm the absence of openings at the internal boundaries. In addition, PPL reviewed the results of recently completed surveillance test data, which indicated the normal ventilation system that was in-service on the zone that was not being tested, was not significantly more negative than the zone tested by the SGTS. In addition, PPL confirmed that operators would receive a control room alarm in the event that the normal ventilation system experiences a high radiation condition in its effluent discharge path, and that the operators would respond and manually place the SGTS in service on that unit. The team determined that the results of PPLs review demonstrated reasonable assurance that the SGTS was capable of performing its intended function. The team will coordinate with the NRCs Office of Nuclear Reactor Regulation to review the adequacy of PPLs SGTS testing methodology to ensure secondary containment boundary integrity. Pending resolution of this issue and determination of any potential enforcement actions, this item is an Unresolved Item.
05000461/FIN-2012005-01Clinton2012Q4Failure to Perform Preventive Maintenance on Standby Gas Treatment System Relay 0UAY-VG506DA finding of very low safety significance with an associated Non-Cited Violation of TS 5.4.1.a. was self-revealed when the age-related failure of Standby Gas Treatment (VG) system relay 0UAY-VG506D caused the removal of VG Train A electric heater 0VG04AA from operation, an entry into TS 3.6.4.3 due to the inoperability of VG Train A, and an unplanned on-line plant risk condition increase from Green to Yellow. The relay failure occurred due to the licensees failure to perform any replacement preventive maintenance on the component throughout the history of plant operation. During two separate independent reviews performed by the licensee on July 15, 2011, and on August 24, 2011, the licensee failed to correctly classify the component in accordance with its preventive maintenance procedure. This resulted in no replacement maintenance activity ever being performed for the relay and its eventual failure on August 22, 2012. The licensee initiated corrective actions to replace the relay and put in place the appropriate preventive maintenance actions. The finding was of more than minor safety significance because it was sufficiently similar to several examples in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, Appendix E, Examples of Minor Issues, wherein licensees failed to adequately implement procedural requirements and the consequences had some safety impact. The performance deficiency was also associated with the SSC (Systems, Structures, and Components) and Barrier Performance attribute and adversely affected the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the age-related failure of 0UAY-VG506D on August 22, 2012, rendered VG Train A inoperable and caused an unplanned increase in the plants on-line risk condition from Green to Yellow. The finding was a licensee performance deficiency of very low safety significance because it only represented a degradation of the radiological barrier function provided for the Auxiliary Building and the Fuel Building and was not a complete loss of the barrier function provided by the VG system since VG Train B remained operable. This finding affected the cross-cutting area of human performance. Specifically, in the area of work control, the licensee did not appropriately coordinate work activities by incorporating actions to plan work activities to support long-term equipment reliability by scheduling maintenance as more preventive than reactive.
05000416/FIN-2012005-01Grand Gulf2012Q4Failure to Make Timely Corrective Actions to Repair the Degraded Auxiliary Building Water Intrusion BarrierThe inspectors identified a non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, involving the failure to correct a condition adverse to quality in a timely manner. Specifically, the licensee failed to correct multiple degraded conditions associated with the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-10314. Corrective actions included generating Work Order 318398 and delegating funds to repair the water intrusion barrier at the next available opportunity. The finding is more than minor because if left uncorrected, the condition of a degraded auxiliary building water intrusion barrier could lead to a more significant safety concern. Specifically, continued degradation of the water intrusion barrier could lead to the auxiliary building (secondary containment) being degraded such that the standby gas treatment system would not be able to achieve and maintain the design negative pressure of 14 inch water column within 120 seconds. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding had very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that the apparent cause of this finding was that the licensee had failed to classify the degraded water intrusion barrier as a condition adverse to quality that warranted correction in a timely manner. Therefore, the finding has a cross-cutting aspect in the problem identification and resolution area, corrective action program component because the licensee failed to properly classify conditions adverse to quality
05000416/FIN-2012005-02Grand Gulf2012Q4Failure to Adequately Monitor the Condition of the Auxiliary Building Water Intrusion BarrierThe inspectors identified a non-cited violation of 10 CFR 50.65(a)(2), for the failure to monitor the performance of the auxiliary building water intrusion barrier. The licensee entered this issue into their corrective action program as Condition Report CR-GGN-2012-11740. Corrective actions included initiating Condition Report CR-GGN-2012-12286, in which the licensee concluded the degraded water intrusion barrier had experienced a Maintenance Rule Functional Failure and required further evaluation to determine if the barrier should be classified in 10 CFR 50.65 (a)(1). The finding is more than minor because if left uncorrected, the failure to monitor the performance of the auxiliary building water intrusion barrier in accordance with the maintenance rule program could lead to a more significant safety concern. Specifically, continued unmonitored degradation of the water intrusion barrier could compromise the integrity of the secondary containment function of the auxiliary building. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, the inspectors determined that the finding affected the Barrier Integrity Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings at Power, the inspectors determined that the finding had a very low safety significance (Green) because the finding only represents a degradation of the radiological barrier function provided for the auxiliary building and standby gas treatment system. The inspectors determined that the apparent cause of this finding was the licensee failed to recognize that the auxiliary building water intrusion barrier was scoped into their Maintenance Rule program with the monitoring criteria of zero occurrences of water intrusion barrier degradation. Therefore, the finding had a cross-cutting aspect in the human performance area, work practices component because the licensee failed to follow maintenance rule program procedures.
05000397/FIN-2012004-04Columbia2012Q3Failure to Provide Adequate Work InstructionsThe inspectors reviewed a self-revealing Green non-cited violation of 10 CFR Part 50 Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to provide work instructions appropriate for performing maintenance on the standby gas treatment system. Specifically, the licensee failed to provide work instructions that would have precluded a trip of the in-service reactor building ventilation system during calibration of the standby gas treatment system. The licensee updated similar work orders to provide provisions to swap to redundant trains to preclude future trips of running equipment. The licensee entered this issue into the corrective action program as Action Request AR 267373. This performance deficiency was more than minor because it affected the configuration control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, the inspectors determined the finding to be of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for by the standby gas treatment system. The inspectors determined the finding had a cross-cutting aspect in the area of human performance associated with the work control component in that the licensee failed to appropriately coordinate work activities to address the operational impact to the reactor building ventilation system when calibrating the standby gas treatment control flow transmitter.
05000298/FIN-2012004-13Cooper2012Q3Recirculation Pump Motor Generator A10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, that, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to the above, from October 27, 2004 to June 11, 2012, the licensee failed to assure adequate corrective actions were implemented to correct an identified nonconformance associated with the oiler position for solenoid operated valves. This resulted in HV-AO-263AV, reactor motor generator set 1A ventilation supply outboard isolation valve operator, not meeting its technical specification required closing stroke time. The performance deficiency was determined to be more than minor because it was associated with the barrier performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The inspectors evaluated the finding using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At- Power. The inspectors determined that the finding is of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier provided for by the standby gas treatment system.
05000387/FIN-2012004-02Susquehanna2012Q3Inadequate Troubleshooting Results in Loss of Secondary Containment and Protected EquipmentA self-revealing Green finding against PPL procedure NDAP-QA-0510, Troubleshooting Plant Equipment, was identified when inadequate troubleshooting caused repeated inoperability of secondary containment, an associated unplanned Unit 2 entry into a 4-hour limiting condition for operation (LCO) action statement, and a loss of the 1C fuel pool cooling (FPC) pump during equipment restoration. The FPC pump had been designated as protected equipment as a risk management action. The failure to perform adequate troubleshooting activities to identify and correct equipment problems prior to restoration was a performance deficiency that was within PPLs ability to foresee and prevent. PPL entered this issue into their CAP as CR 1628250. The inspectors determined that the finding was more than minor because it was associated with the configuration control 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. Specifically, the event resulted in the inoperability of secondary containment and loss of a FPC pump. The finding was evaluated in accordance with IMC 0609, Attachment 4, and Appendix A - Exhibit 3, and was determined to be of very low safety significance (Green) because the finding did not only represent a degradation of the radiological barrier function provided for the standby gas treatment system and it did not: a) cause the spent fuel pool to exceed a maximum temperature limit; b) cause mechanical fuel damage and detectable release of radionuclides; c) result in the loss of spent fuel pool water inventory; or d) affect spent fuel shutdown margin. This finding is related to the cross-cutting area of Human Performance Decision-Making because PPL did not make safety-significant or risk-significant decisions using a systematic process, especially when faced with uncertain or unexpected plant conditions, to ensure safety is maintained. Specifically, PPL failed to restore equipment in a systematic manner, given the intermittent nature of heater faults, to preclude a repeated loss of protected equipment and secondary containment.
05000387/FIN-2012003-04Susquehanna2012Q2Improperly Performed Maintenance Impacts Secondary ContainmentA self-revealing Green NCV of TS 5.4.1, Procedures, was identified regarding PPLs conduct of maintenance during a Unit 1 refueling outage which impacted the operating unit, Unit 2. Specifically, improperly performed maintenance on a Unit 1 main stop valve (MSV) and outboard main steam isolation valve (MSIV) affected safety-related equipment to include the standby gas treatment system (SGTS) and Unit 2 secondary containment in an unplanned manner. PPL entered this issue in their CAP via CRs 1558764, 1558718, and 1560235 and performed a root cause analysis (RCA) on this. Improperly performed MSIV and MSV maintenance was a performance deficiency within PPLs ability to foresee and correct. This finding was considered more than minor because it was similar to IMC 0612, Appendix E, Examples 3.j and 3.k, in that a physical plant condition and subsequent engineering calculation resulted in a condition where there was reasonable doubt on the operability of a system or component, in this case secondary containment. Further, the performance deficiency affected the procedure quality and SSC and barrier performance attributes of the Barrier Integrity cornerstone and its objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In this case, lack of coordination resulted in a loss of reasonable assurance that secondary containment was operable. The issue screened to Green via IMC 0609, Attachment 4, since it did not represent a degradation of the barrier function of the control room, did not represent an actual open pathway in the physical integrity of reactor containment, and did not involve the actual reduction in function of hydrogen igniters in containment. The issue was determined to have a cross-cutting aspect in the area of Human Performance to plan and coordinate work activities, consistent with nuclear safety. In this case, the MSV and MSIV work activities were not coordinated amongst various departments to address the operational impact of sequence changes on plant configuration.
05000354/FIN-2012003-01Hope Creek2012Q2Preconditioning of the Reactor Building to Torus Vacuum Relief ValvesThe inspectors identified a non-cited violation (NCV) of very low safety significance of 10 CFR 50, Appendix B, Criterion XI, Test Control, because PSEG conducted unacceptable preconditioning of the reactor building to torus vacuum relief valve. Specifically, PSEGs surveillance test procedure for these valves cycled the valve (H1GS- 1GSPSV-5032) prior to recording the as-found opening setpoint required to meet Technical Specification (TS) Surveillance Requirement (SR) 4.6.4.2.b.2.a. PSEGs immediate corrective actions included revising the surveillance test procedure to record the as-found setpoint before cycling the valve manually. The violation was entered into the corrective action program (CAP) as notification 20554080. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the Barrier Integrity Cornerstone and 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, preconditioning of the reactor building to torus vacuum relief opening setpoint could mask its actual as-found condition and result in an inability to verify its operability and potentially make it difficult to determine whether the vacuum breaker would perform its intended safety function during an event. The inspectors evaluated the finding using IMC 0609, Attachment 4, Initial Screening and Characterization of Findings, and determined the finding was of very low safety significance (Green) because it was not a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or standby gas treatment system, did not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere, did not represent an actual open pathway in the physical integrity of reactor containment and heat removal components, and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The finding had a cross-cutting aspect in the area of problem identification and resolution, corrective action component, because PSEG did not thoroughly evaluate a prior problem such that the problem resolution addressed the extent of condition. Specifically, PSEGs extent of condition for notification 20370021, Potential Preconditioning BJHV-F004, did not go beyond operations procedures and review maintenance procedures for unacceptable preconditioning. Therefore, PSEG did not identify the unacceptable preconditioning of the reactor building to torus vacuum relief valve opening setpoint because the surveillance test was in a maintenance procedure.
05000373/FIN-2012003-01LaSalle2012Q2Failure to Perform Surveillance Test Procedure StepA finding of very low safety significance was identified by the inspectors for the licensees failure to implement a station-required procedure step during surveillance testing of the standby gas treatment (SBGT) system. Specifically, the licensee failed to perform the step in LaSalle procedure LOS-VG-M1, Standby Gas Treatment System Operability and Inservice Test , which directs the SBGT manual initiation pushbuttons be tested every three years. Since the particular function of the pushbuttons is not required by regulation, and the procedure step was created only as a self-imposed station requirement, no violation of regulatory requirements occurred. Upon notification by the inspectors of the discrepancy, the licensee promptly entered the issue into its corrective action program (CAP) for evaluation and resolution. The finding was determined to be more than minor because the performance deficiency of failing to meet procedure requirements, if left uncorrected, could have the potential to lead to a more significant safety concern. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Attachment 0609.04, Phase 1 - Initial Screening and Characterization of Findings, Table 4a, for the Mitigating Systems Cornerstone, dated January 10, 2008. The finding was determined to be of very low safety significance because all questions in the Mitigating Systems column were answered No. This finding has a cross-cutting aspect in the area of human performance, work control, for failing to appropriately coordinate work activities and keep personnel apprised of work status. Specifically, because there was no predefine in the work management system, operators performing the surveillance test were not aware of the status of the triennial requirement.
05000416/FIN-2012002-06Grand Gulf2012Q1Licensee-Identified ViolationTechnical Specifications section 3.6.1.3 requires that main steam isolation valves have a closure time of less than or equal to 5 seconds. Contrary to this, on February 20, 2012, during refueling outage 18, the licensee performed Surveillance Procedure 06-OP-1B21-V-0001, Revision 114, MSIV Operability Test , on main steam isolation valve 1B21F028A, and the valve closure time was 6.7 seconds. The licensee entered this issue into their corrective action program in condition report CR-GGN-2012-01848 and initiated work order 306292 to repair the valve prior to the end of the refueling outage. The finding was determined to be of very low safety significance (Green) because it did not represent a degradation of the radiological barrier function provided for the control room, or auxiliary building, or spent fuel pool, or standby gas treatment system, a degradation of the barrier function of the control room against smoke or a toxic atmosphere, an actual open pathway in the physical integrity of reactor containment, and the finding did not involve an actual reduction in function of hydrogen igniters in the reactor containment.
05000321/FIN-2012002-04Hatch2012Q1Licensee-Identified ViolationA licensee-identified violation of 10 CFR 50 Appendix B, Criterion V, Procedures, was discovered on September 22, 2011, when the 2A standby gas treatment train discharge damper failed to fully open during surveillance testing. 10 CFR 50 Appendix B, Criterion V, requires in part that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions. Contrary to this requirement on January 14, 2010, the licensee failed to install a required cotter pin in the 2A standby gas treatment train discharge damper which led to the failure of the damper on September 22, 2011. The inspectors screened the significance of this violation using IMC 0609 Attachment 4, Table 4a where this finding screened as Green because the finding only represented a degradation of the radiological barrier function provided for the standby gas treatment system. This issue is captured in the licensees corrective action program as CR 353886
05000397/FIN-2011005-03Columbia2011Q4Missed Procedural Step Results in Secondary Containment Pressure ExcursionThe inspectors reviewed a self-revealing non-cited violation of Technical Specification 5.4.1.a, for the licensees failure to follow procedures. Specifically, on November 2, 2011, operators failed to follow Procedure SOP-HVAC/RB-START, Reactor Building Ventilation Start, Revision 2, by skipping a required step for restoration of reactor building ventilation to the normal alignment following testing of secondary containment isolation valves. As a result, when the reactor building ventilation fans were started, secondary containment pressure increased rapidly to a peak positive pressure of approximately 0.29 inch of water, while secondary containment is normally maintained at 0.6 inch of water vacuum to meet its design basis function. When operators completed of the surveillance test of the secondary containment isolation valves, operators entered Procedure SOP-HVAC/RB-START at Step 5.1.5 which started the fans. The operators should have entered the procedure at Step 5.1.1 which would have placed pressure controller REA-DPIC-1B in manual. This step was necessary since the response time of the controller was not rapid enough to compensate for the rapid changes in air flows associated with a fan start. An event investigation concluded that the missed procedural step was caused by poor planning and preparation and less than adequate self and peer checks. This issue was entered into the licensees corrective action program as Action Request AR 00251613. The finding was more than minor because it affected the human performance attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Using Inspection Manual Chapter 0609.04, Phase 1 Initial Screening and Characterization of Findings, the inspectors determined this finding to be of very low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for by the standby gas treatment system. The inspectors determined that this finding had a cross-cutting aspect in the area of human performance associated with the work practices component because the licensee failed to use human error prevention techniques such as self and peer checking (H.4(a))
05000324/FIN-2011004-05Brunswick2011Q3Licensee-Identified ViolationTechnical Specification (TS) 3.3.6.1, Primary Containment Isolation Instrumentation, requires that the RWCU high differential flow instrumentation be operable in modes 1, 2, or 3. If the instrumentation is not operable, then TS 3.3.6.1 requires that the RWCU penetration flow path be isolated within 1 hour. Contrary to the above, the licensee identified that the RWCU high differential flow instrumentation was not operable and the penetration flow path was not isolated when the unit entered mode 1 on April 16, 2011 until August 2, 2011, because the RWCU inlet flow sensing element was installed backwards, causing the flow sensing element to be inaccurate. The resulting inaccuracy caused the instrumentation to be unable to isolate within the required TS limit of less than or equal to 73 gallons per minute differential flow. The finding was determined to be of very low safety significance per Appendix A of Inspection Manual Chapter 0609, Significance Determination Process, because the finding: 1) did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or the standby gas treatment system, 2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, and 3) did not represent an actual open pathway in the physical integrity of reactor containment. Upon discovery of the condition, the licensee isolated the affected penetration flow path and installed the flow sensing element correctly. The issue is in the licensees CAP as NCR #479248.
05000324/FIN-2011004-03Brunswick2011Q3Inadequate Maintenance Results in Containment Isolation Valve FailureA self-revealing Green finding was identified for inadequate maintenance on the overload relay of the unit 2 reactor water cleanup (RWCU) system inlet isolation valve 2-G31-F001. As a result of the inadequate maintenance, the overload relay actuated during operation of the valve under normal conditions, and the valve failed to shut. This was revealed while operators were attempting to isolate the RWCU system on August 2, 2011. After the valve failed to fully shut on August 2, 2011, the licensee shut the valve in series with 2-G31-F001 (2-G31-F004), repaired the overload relay for the 2-G31-F001 valve by installing the correct fasteners, returned the 2-G31-F001 valve to service, and entered the issue into their corrective action program (AR #480063). The inadequate maintenance on the 2-G31-F001 valve overload relay was a performance deficiency. The finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of structure, system, and component (SSC) and Barrier Performance, and it affected the associated cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the finding prevented a primary containment isolation valve from shutting. This finding was evaluated using Inspection Manual Chapter 0609, Significance Determination Process, Phase 1 Worksheet for Containment Barriers. The finding was determined to be of very low safety significance (Green) because the finding: 1) did not only represent a degradation of the radiological barrier function provided for the control room, auxiliary building, spent fuel pool, or the standby gas treatment system, 2) did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere, and 3) did not represent an actual open pathway in the physical integrity of reactor containment. The cause of this finding has no cross-cutting aspect because the maintenance took place in 1992 and is not indicative of current licensee performance.
05000458/FIN-2011004-01River Bend2011Q3Inadequate Standby Gas Treatment Electric Heater Power Output CalculationThe inspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III Design Control, for an inadequate calculation methodology used in determining standby gas treatment system operability. The inspectors found that the calculation neither considered instrument uncertainty nor applied a proper voltage drop from the breaker to the standby gas treatment system filter train heater. The licensee entered this issue into the licensees corrective action program as Condition Report CR-RBS-2011-07332. The finding was more than minor because it was associated with the design control attribute of the Barrier Integrity Cornerstone to maintain radiological barrier functionality of standby gas treatment trains, and affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, operating the standby gas system filter train heaters without sufficient output power is detrimental to the charcoal filters ability to retain radioactive iodine. This could result in a greater amount of radiation release to the environment in the event of an accident. In accordance with Inspection manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the Phase 1 significance determination process screening determined the finding to be only of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the standby gas treatment system. The apparent cause of this finding was the decision to develop an engineering evaluation that did not include instrument uncertainly and did not validate the correct voltage drop between the filter train heater feeder breaker and the heater elements. The finding has a crosscutting aspect in the area of human performance associated with the decision-making component because station personnel failed to use conservative assumptions when developing the modified output power methodology for operation of the standby gas treatment system filter heaters with only 8 of 9 heater elements installed.
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