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05000482/FIN-2018003-012018Q3Wolf CreekFailure to Correct Degraded Performance of a Safety-Related Tornado DamperThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Criterion XVI, Corrective Action, for the licensees failure to promptly correct a condition adverse to quality associated with a safety-related tornado damper. Specifically, damper GTD0002 failed tests in 2012 and 2015, and following maintenance on the damper in 2017, again failed its next as-found test on February 8, 2018. As a result, this safety-related tornado dampers ability to close during a design basis tornado event was adversely impacted.
05000482/FIN-2018003-022018Q3Wolf CreekFailure to Submit a Licensee Event Report for a Condition Prohibited by Technical SpecificationsThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.73(a)(2)(i)(B), because the licensee did not provide a written licensee event report (LER) to the NRC within 60 days. Specifically, the licensee did not provide a written LER to the NRC within 60 days of identifying a condition prohibited by the plants Technical Specifications associated with inoperability of control room emergency ventilation system train B for longer than its Technical Specification allowed outage time. As a result, the NRCs ability to regulate was impacted.
05000528/FIN-2018002-022018Q2Palo VerdeFailure to Implement and Maintain Procedures Regarding Breathing Air QualityThe inspectors identified a Green, non-cited violation of 10 CFR 20.1703 for failing to implement and maintain written procedures to ensure that respiratory protection equipment (air compressors and bubble hood suites) supplied respirable air of grade D quality or better to radiation workers.
05000530/FIN-2018002-032018Q2Palo VerdeFailure to Assess the Operability of a Degraded or Nonconforming Structure, System, or ComponentThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to evaluate conditions adverse to quality for impacts on the operability of the essential spray ponds.
05000528/FIN-2018002-012018Q2Palo VerdeFailure to Re-baseline Valve Stroke Times as Required by ASME OM CodeThe inspectors identified a Green, non-cited violation of Palo Verde Technical Specification 5.5.8, Inservice Testing Program, which requires inservice testing of ASME Code Class 1, 2, and 3 components in accordance with the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code). On October 22, 2017, the licensee failed to establish new stroke time reference values for Unit 1 safety injection (SI) valve 660 following maintenance which could affect the valves performance
05000482/FIN-2018002-032018Q2Wolf CreekFailure to Adequately Implement Instrumentation and Controls Surveillance ProceduresA self-revealed Green NCV of 10 CFR Part 50, Criterion V, Instructions, Procedures, and Drawings, was identified when the licensee failed to adequately implement surveillance procedures that affected safety-related equipment and plant stability. Specifically, the licensee failed to adequately implement testing and calibration procedures for pressurizer level instrumentation. This resulted in two letdown isolation signals, securing of pressurizer heaters, and a pressurizer level transient on March 29, 2018.
05000482/FIN-2018002-022018Q2Wolf CreekFailure to Maintain Adequate Pressurization of the Control Room EnvelopeA self-revealed Green NCV of 10 CFR Part 50, Criterion III, Design Control, was identified when the licensee failed to adequately recognize that the cable spreading room floor was a control building ventilation isolation boundary. Specifically, the licensee cut openings in the floor/ceiling between the 2,032 foot and 2,016 foot elevations of the control building and the impact on the control room envelopes ability to pressurize was not recognized. This was a primary contributor to the train B control room emergency ventilation system being unable to maintain the appropriate pressure in the control room envelope.
05000482/FIN-2018002-012018Q2Wolf CreekAnnouncement of an NRC Inspectors Presence by Station PersonnelThe inspectors identified a Severity Level IV non-cited violation (NCV) of 10 CFR 50.70(b)(4), Inspections, associated with the licensees failure to ensure the arrival and presence of NRC Inspectors, who had been properly authorized facility access as described in 10 CFR 50.70(b)(3), were not announced or otherwise communicated by its employees or contractors to other persons at the facility without a specific request by the NRC inspector. Specifically, a contract radiation protection technician entered the spent fuel pool building where the resident inspector was present and observing core offload activities, and the technician informed members of a work crew of the whereabouts of an NRC radiation protection inspection team without being requested to do so; this impacts the NRCs ability to regulate and perform unannounced inspections.
05000483/FIN-2018001-012018Q1CallawayFailure to Maintain Emergency Operating ProceduresThe inspectors identified a Green, non-cited violation of Technical Specification 5.4.1.a, "Procedures," for the licensee's failure to maintain emergency operating procedures for aligning auxiliary feedwater suction sources. Specifically, the licensee added continuous action steps to emergency operating procedures that placed both motor-driven auxiliary feedwater pumps in pull-to-lock and isolated their associated recirculation lines after depleting the two non-safety-related suction sources. These actions cause two of the three safety-related auxiliary feedwater pumps to be rendered inoperable prior to aligning the safety-related suction source of essential service water which is credited in accident analysis.
05000483/FIN-2017002-022017Q2CallawayFailure to Analyze the Effect of Changes to Maintaining the Gaitronics SystemSeverity Level IV. The inspectors identified a Severity Level IV non- cited violation for the licensees failure to perform an analysis of a change to processes supporting the emergency preparedness program that demonstrated the change did not reduce the effectiveness of the emergency plan in accordance with the requirements of 10 CFR 50.54(q)(3). There were no immediate safety concerns associated with this violation because less than 10 percent of the public address speakers were determined to be degraded or non- functional. This issue has been placed in the licensees corrective action system as Condition Report 201702343. The failure to perform an analysis of the effect of changes in processes supporting emergency preparedness is a performance deficiency within the licensees ability to foresee and correct. The finding was more than minor because the finding was associated with the Facilities and Equipment Cornerstone attribute and adversely affected the Emergency Preparedness Cornerstone objective. The finding was assessed using traditional enforcement because the licensees failure to perform a required analysis impacted the regulatory process . The finding was evaluated using the NRCs Enforcement Policy, dated November 1, 2016, Section 6.6(d) , and was determined to be a Severity Level IV violation because the violation did not affect radiological assessment or offsite notification. Traditional enforcement violations are not assessed for cross -cutting aspects.
05000483/FIN-2017002-012017Q2CallawayFailure to Follow Motor Control Center ProcedureGreen . The inspectors reviewed a self -revealed, non- cited violation of Technical Specification 5.4.1.a, Procedures, for the licensees failure to follow Procedure MPE-ZZ-QS001, Cleaning and Inspection of Motor Control Centers, Revision 34. On May 2, 2017, the licensee failed to ensure contactors operated freely per step 7.6.8 during reassembly of motor control center NG08F for the essential service water cooling tower by pass valve EFHV0066. As a result, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. As immediate corrective actions, the licensee replaced the starter assembly under Job 17001973, completed testing including electrically cycling valve EFHV0066, and restored the system to operable status on May 4, 2017. The licensee entered this issue into the corrective action program under Condition Report 201702418. The failure to follow Procedure MPE-ZZ-QS001 was a performance deficiency. This performance deficiency was more than minor, and therefore a finding, because it adversely affected the configuration control attribute of the Mitigating Systems Cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, one train of the essential service water system was rendered inoperable for a total of 57 hours, of which 17 hours was unplanned, and the issue was only discovered when valve EFHV0066 failed to operate during a periodic surveillance test on May 3, 2017. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings At -Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012, the inspectors determined the finding was of very low safety significance (Green) because (1) the finding was not a deficiency affecting the design or qualification of a mitigating system; (2) the finding did not represent a loss of system and/or function; (3) the finding did not represent an actual loss of function of a single train for greater than its technical specification allowed outage time; and (4) the finding does not represent an actual loss of function of one or more non- technical specification trains of equipment designated as high safety -significant in accordance with the licensees maintenance rule program for greater than 24 hours. Specifically, the total duration of 3 inoperability was approximately 57 hours which is less than the allowed completion time of 72 hours for this system. The finding had a cross-cutting aspect in the area of human performance associated with challenge the unknown because the licensee failed to stop when faced with uncertain conditions. Specifically, the maintenance technician encountered resistance when manually operating the contactors, signed off the step as complete, and later rationalized the decision with the supervisor aft er completing the work (H.11 ).
05000298/FIN-2016002-012016Q2CooperFailure to Meet Technical Specification Requirements for Traversing In-Core Probe B Ball Valve (The inspectors identified a non-cited violation of Technical Specification 3.6.1.3, Primary Containment Isolation Valves, for the licensees failure to maintain traversing incore probe B ball valve, a primary containment isolation valve, operable for its containment isolation function. Specifically, on May 5, 2016, from 5:20 a.m. until 1:08 p.m., the licensee failed to maintain the traversing in-core probe B ball valve operable or isolate its flow path within 4 hours of indications that the mechanical in-shield limit switch had failed. This failure prevented the ball valve from performing its containment isolation function. The licensee took immediate corrective actions upon discovery to restore compliance with Technical Specification 3.6.1.3 by de-energizing the ball valves solenoid operating valve, causing it to close. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2016-03665. The licensees failure to maintain the traversing in-core probe B ball valve, a primary containment isolation valve, operable for its containment isolation function, in violation of Technical Specification 3.6.1.3, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases and that the radiological barrier functionality of containment is maintained. Specifically, the traversing in-core probe B ball valve was unable to perform its primary containment isolation function with a failed mechanical inshield limit switch. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because it did not represent an actual open pathway in the physical integrity of reactor containment (valves, airlocks, etc.), containment isolation system (logic and instrumentation), 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 human performance associated with conservative bias because the licensee failed to use decision making practices that emphasized prudent choices over those that were simply allowable and failed to ensure proposed actions were determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee failed to validate the assumption that the traversing in-core probe B ball valve would fulfill its containment isolation function with a failed mechanical in-shield limit switch, and failed to validate the degraded condition prior to exceeding the 4-hour completion time of Technical Specification 3.6.1.3 (Section 1R12). (H.14)
05000298/FIN-2016002-022016Q2CooperFailure to Follow Work Instructions for Post-Maintenance Testing of Safety-Related Ventilation SystemsThe inspectors identified two examples of a non-cited violation of Technical Specification 5.4.1.a, associated with the licensees failure to perform required postmaintenance testing for safety-related ventilation systems in accordance with documented instructions, prior to system restoration. Specifically, the licensee failed to follow work order instructions contained in Work Orders 5062878 and 5065112 for (1) performing surveillance testing to measure the airflow of emergency diesel generator supply fan coil unit HV-DG-1C following maintenance, and (2) performing leak testing of a newly created control room ventilation boundary penetration. Corrective actions included performing the required surveillance test for the diesel generator ventilation unit, retesting the control room penetration in accordance with the procedure, and initiating site-wide communications discussing the errors and reemphasizing procedural adherence. The licensee entered these deficiencies into their corrective action program for resolution as Condition Reports CR-CNS-2016-02207 and CR-CNS-2016-02232. The licensees failure to perform required post-maintenance testing for safety-related ventilation systems, in accordance with documented instructions, was a performance deficiency. This performance deficiency was associated with multiple cornerstones. The first example of the performance deficiency was more than minor, and therefore a finding, because it was associated with the human performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to measure supply fan coil unit HV-DG-1C airflow resulted in delayed identification that the maintenance had resulted in degraded flow through the ventilation unit. The second example of the performance deficiency was more than minor, and therefore a finding, because it was associated with the human 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 and that the radiological barrier functionality of the control room is maintained. Specifically, the licensees failure to follow post-maintenance testing instructions resulted in a challenge to the operability of the newly created control room boundary penetration seal. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because it did not represent a design or qualification deficiency; did not represent a loss of safety function; did not represent a loss of a single train for greater than its technical specification allowed outage time; did not screen as potentially risk significant due to a seismic, flooding, or severe weather initiating events; did not represent an actual open containment pathway; and did not involve a reduction in function of hydrogen igniters. The finding had a crosscutting aspect in the area of human performance associated with work management, because the licensee failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority, including the need for coordination with different work groups or job activities. Specifically, the licensee failed to control, execute, and coordinate safety-related ventilation work activities to ensure all required post-maintenance testing was completed satisfactorily prior to declaring the associated equipment operable (Section 1R19). (H.5)
05000298/FIN-2016002-032016Q2CooperFailure to Maintain Design Control for High Pressure Coolant Injection System Electrical CircuitThe inspectors reviewed a self-revealed, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to verify the adequacy of design of the high pressure coolant injection auxiliary lube oil pump 125 Vdc starter circuit. Specifically, in 1984, the licensee modified the design of the starter circuit and eliminated a resistor that served to protect the circuit from shorting due to indication light bulb failures. As a result, on April 26, 2016, a shorted light bulb resulted in the loss of power to the auxiliary lube oil pump, rendering the high pressure coolant injection system inoperable and unavailable. Immediate corrective actions included replacing the light socket and blown fuse and changing out the nonessential light bulb with an essential bulb. This event was entered into the licensees corrective action program as Condition Report CR-CNS-2016-02318, and the licensee initiated a root cause evaluation to investigate the failure. The licensees failure to verify the adequacy of design of the high pressure coolant injection auxiliary lube oil pump starter circuit in accordance with 10 CFR Part 50, Appendix B, Criterion III, was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, at the time the modification was installed, the licensee had not taken sufficient actions to ensure that the electrical circuit was protected from light bulb shorting failures, resulting in the high pressure coolant injection system ultimately being rendered inoperable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings AtPower, dated June 19, 2012, inspectors determined that the finding required a detailed risk evaluation because it represented a loss of the system and function of high pressure coolant injection. The inspectors determined that the finding was of very low safety significance (Green) through performing a detailed risk evaluation. A cross-cutting aspect was not assigned to this finding because the performance deficiency occurred in 1984, and therefore, is not indicative of current licensee performance (Section 4OA3).
05000483/FIN-2016001-022016Q1CallawayInadequate Operability Evaluation for Degraded Flood Mitigation Capability in Piping Penetration RoomThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to perform an adequate operability determination for safety related components located in the 1988 foot auxiliary building train B piping penetration room (room 1203) based on degraded internal flooding drain capability. Specifically, the immediate operability determination included incorrect assumptions that were not verified to support the operability determination as required by Procedure ODP-ZZ-00001, Addendum 15, Operability and Functionality Determinations, Revision 8. The immediate corrective action was to implement a compensatory measure to support operability of the equipment in room 1203. The issue was placed in the corrective action program as Callaway Action Request 201601412. The licensees failure to verify assumptions used in the immediate operability determination and ensure a sound basis for operability exists per plant procedures was a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it is similar to examples 3.j and 3.k in Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, and if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, failure to perform adequate operability evaluations by verifying assumptions and ensuring a sound basis for operability exists may result in the failure to enter the appropriate limiting conditions of operation for technical specification equipment. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding involved the degradation of equipment specifically designed to mitigate a flooding initiating event, therefore, Exhibit 4, External Events Screening Questions, was used to complete the screening. The finding was determined to need a detailed risk evaluation because if the equipment (i.e., floor drain lines) is assumed to be completely failed or unavailable, it would degrade one or more trains of a system that supports a risk significant system or function. In consultation with the Senior Reactor Analyst, the finding was determined to be of very low safety significance because, based on the actual condition of the drains and the extent of the clogging in room 1203, an evaluation by the licensee showed that the maximum internal flooding water level in the room would not challenge the operability of any equipment needed for safe shutdown or to mitigate an accident. This finding has a team work cross-cutting aspect in the human performance cross-cutting area because individuals and work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, inadequate communication between engineering and operations personnel led to the belief that a passageway existed between rooms 1203 and 1204 when it did not (H.4).
05000285/FIN-2016001-012016Q1Fort CalhounImplementing a Procedure Change for Alternative Shutdown Cooling that would have Required NRC ApprovalThe inspectors identified a Severity Level IV non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, for the failure to recognize that a change to the facility as described in the Updated Safety Analysis Report would require prior NRC review and approval. Specifically, the 10 CFR 50.59 evaluation revised a site procedure, without NRC approval, to substitute automatic flow control of shutdown cooling flow and temperature with manual control using the low pressure safety injection loop injection valves. The licensees corrective actions included revising the affected procedure to reflect the original automatic flow control. The licensee entered this issue in the corrective action program as Condition Report 2013-15342. The licensees failure to implement the requirements of 10 CFR 50.59 and adequately evaluate changes to determine if prior NRC approval is required was a performance deficiency. Because this violation had the potential to impact the NRCs ability to perform its regulatory function, the inspectors evaluated the violation using traditional enforcement. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the team evaluated this finding using the significance determination process to assess its significance. The inspectors performed an initial screening of the finding in accordance with NRC Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 1, 2012. Using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the finding was determined to have very low safety significance (Green) because it: (1) was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of function of at least a single train for longer than its technical specification allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant in accordance with the licensees Maintenance Rule program. Therefore, in accordance with Section 6.1.d.2 of the NRC Enforcement Policy, the inspectors characterized this performance deficiency as a Severity Level IV violation. The inspectors determined that a cross-cutting aspect was not applicable because the issue involving the failure to perform an adequate 10 CFR 50.59 evaluation was strictly associated with a traditional enforcement violation.
05000285/FIN-2016001-022016Q1Fort CalhounLicensee-Identified ViolationTechnical Specification 2.6(1) requires containment integrity to be maintained unless the reactor is in a cold or refueling shutdown condition. If containment integrity is not maintained and the reactor does not meet these cold or refueling shutdown conditions, then containment integrity must be restored within one hour or the reactor is required to be in hot shutdown within the next six hours. From November 22, 2013, through June 27, 2014, a test connection cap was left off of a containment penetration which constituted a loss of containment integrity. Upon discovery of this condition on June 27, 2014, the licensee entered Technical Specification 2.6(1) and Abnormal Operating Procedure 12 for loss of containment integrity. The cap was re-installed and containment integrity was restored within one hour. The violation is more than minor because it is associated with the configuration control attribute of the Barrier Integrity Cornerstone. Failure to install the containment penetration cap following local leak rate testing on November 22, 2013, resulted in a loss of containment integrity until it was discovered missing on June 27, 2014. This adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (i.e., containment) protect the public from radionuclide releases caused by accidents or events. The violation was reviewed by a Senior Reactor Analyst and was determined to be of very low safety significance because the test connection fitting was a 14-inch diameter opening. Inspection Manual Chapter 0609, Significance Determination Process, Appendix H, identifies that small lines (less than 1 to 2 inches in diameter) would not generally contribute to large early release frequency. Therefore, this finding screens to Green. The licensee entered the issue into their corrective action program as Condition Report 2014-07958.
05000483/FIN-2016001-012016Q1CallawayPossible Incorrect Screening of the Spent Fuel Pool Decay Heat Removal Key Safety FunctionThe inspectors identified an unresolved item associated with the National Fire Protection Association (NFPA) Standard 805, Performance-Based Standard for Fire Protection for Light-Water Reactor Electric Generating Plants, non-power operations assessment. Specifically, the inspectors developed an issue of concern in that the licensee screened the potential loss of spent fuel pool cooling from further consideration for any fire event based on adequate procedural guidance and time when the procedures would not maintain the fuel in a safe and stable condition. On January 13, 2014, the licensee transitioned their fire protection program to a risk-informed, performance-based program based on NFPA Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants. Paragraph 1.3.1 of NFPA Standard 805 requires licensees to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition. Paragraph 1.5.1 of NFPA Standard 805 lists five nuclear safety performance criteria. These criteria provide requirements to demonstrate that fire protection features are capable of providing reasonable assurance that the plant is not placed in an unrecoverable condition in the event of a fire. For the decay heat removal nuclear safety performance criterion, the standard requires that decay heat removal shall be capable of removing sufficient heat from the reactor core or spent fuel such that fuel is maintained in a safe and stable condition. Paragraph 1.6.56 of NFPA Standard 805 defines safe and stable conditions: For fuel in the reactor vessel, head on and tensioned, safe and stable conditions are defined as the ability to maintain Keff <0.99, with a reactor coolant temperature at or below the requirements for hot shutdown for a boiling water reactor and hot standby for a pressurized water reactor. For all other configurations, safe and stable conditions are defined as maintaining Keff <0.99 and fuel coolant temperature below boiling. The licensee described how they satisfied the nuclear safety performance criteria in Calculation KC-26, Nuclear Safety Capability Assessment, Revision 1. The Nuclear Safety Capability Assessment applied to both power and non-power operations. For non-power operations, the licensee evaluated the spent fuel pool decay heat removal key safety function and determined that the spent fuel pool decay heat removal key safety function did not require a detailed review since adequate time was available, and procedural guidance was provided, for operators to respond to and mitigate a loss of spent fuel pool decay heat removal, even under full hot core offload conditions. The licensee stated that the shortest time to boil, under worst case conditions for a normal plant shutdown, was two hours. In addition, the licensee stated that all of the analyses to address a loss of spent fuel pool decay heat removal utilized a success criterion of no boiling. The licensee implemented the process outlined in Frequently Asked Question (FAQ) 07-0040, Non-Power Operations Clarifications, Revision 4, for the non-power operations assessment. This FAQ stated that licensees should conservatively assume the entire contents of a fire area are lost and document the loss of success paths. This FAQ also stated that licensees should specifically identify those areas (pinch points) that cause the loss of all success paths for a key safety function. The inspectors noted that the licensee did not perform these actions for the spent fuel pool decay heat removal key safety function because this key safety function was screened out from further consideration. If the licensee had evaluated the spent fuel pool decay heat removal key safety function using the process outlined in this FAQ, then the licensee would have assumed that both trains of spent fuel pool cooling are lost during a fire in the fuel handling building because both trains are located within the same fire area and were unprotected. This FAQ also stated that fire modeling may be used to determine if postulated fires in a fire area are expected to damage equipment (and cabling), thereby eliminating a pinch point. However, the licensee stated that no fire modeling was used to eliminate the identification of pinch point fire areas as part of the non-power operations assessment performed using the process in FAQ 07-0040. In the event that a fire in the fuel handling building disabled both trains of spent fuel pool cooling, operators were expected to enter Procedure OTO-EC-00002, Spent Fuel Pool High Temperature, Revision 9, due to the increasing temperature of the spent fuel pool. This procedure provided directions for operators to restore one or both trains of spent fuel pool cooling. Since both trains of spent fuel pool cooling were assumed lost due to the fire, the operators would be unable to restore spent fuel pool cooling using this procedure. After a period of time, the spent fuel pool would begin boiling and the level would begin lowering. At this time, operators were expected to enter Procedure OTO-EC-00001, Loss of SPF/Refuel Pool Level, Revision 13. Procedure OTO-EC-00001 directed the operators to open two normally locked essential service water valves to restore and maintain spent fuel pool level. The licensees procedures allowed the spent fuel pool to reach boiling conditions prior to restoring and maintaining level. Since NFPA Standard 805 defined safe and stable conditions, in part, as fuel coolant temperature below boiling, the procedures did not maintain the fuel in a safe and stable condition. The inspectors identified an issue of concern in that the licensee screened the potential loss of spent fuel pool cooling from further consideration for any fire event based on adequate procedural guidance and time when the procedures would not maintain the fuel in a safe and stable condition. The inspectors determined that additional information is required to determine if a performance deficiency exists. Specifically, the inspectors need to determine if this scenario should have been addressed as part of the current FAQ 07-0040 guidance, or if new guidance is needed to address this type of scenario where the full core has been offloaded to the spent fuel pool. On March 31, 2016, additional guidance was requested from the Office of Nuclear Reactor Regulation via a request to review and update FAQ 07-0040. This memorandum is documented in ADAMS as Accession Number ML16091A152. The licensee entered this issue of concern into the corrective action program as Callaway Action Request 201600726. This issue of concern is being treated as Unresolved Item 05000483/2016001-01, Possible Incorrect Screening of the Spent Fuel Pool Decay Heat Removal Key Safety Function.
05000285/FIN-2015004-012015Q4Fort CalhounLicensee-Identified ViolationTitle 10 CFR 72.174 requires that each licensee maintain sufficient records to furnish evidence of activities affecting quality. Records pertaining to the design, fabrication, erection, testing, maintenance, and use of structures, systems, and components important to safety must be maintained by or under the control of the licensee until the NRC terminates the license. Contrary to the above, as of June 21, 2013, Fort Calhoun failed to maintain sufficient records to furnish evidence of activities affecting quality. Specifically, the licensee did not maintain records for loading activities associated with DFS-HSM-06 that was placed on the ISFSI pad in July of 2009. This violation was identified by FCS and placed in their corrective action program (CR 2013-12884). The fuel assembly data was reconstituted based on records from the Reactor Engineering group, and the canister helium leak-test data was reconstituted based on the helium leak-test technician's field notes. The remaining canister records associated with the canister processing, sealing, and transportation to the ISFSI, including several TS requirements, were not found. Fort Calhoun reconstituted the fuel and helium leak test data and conducted interviews with cask loading personnel to conclude that there was no evidence to suggest that loading activities did not comply with the licensee's procedures and the licensed Technical Specifications. This violation did not have any safety impact because all fuel assemblies met the requirements for burn-up, decay heat, and cooling time and the licensee demonstrated that the canister integrity was intact based on the reconstituted helium leak test records. All the fuel inside the canister and the cask remain in a safe condition. This finding was reviewed by NRC Headquarters Division of Spent Fuel Managements Spent Fuel Licensing Branch. Based on the reconstituted records and interviews with the dry fuel loading staff, the NRC found no evidence to demonstrate that the canister did not meet the required license conditions and as such, found the canister acceptable for continued storage under FCSs general Part 72 license. However, though the canister is acceptable for storage, the licensee must track this issue to identify that further analyses may be required for this canister to meet all applicable Part 71 requirements to be acceptable for transportation. In accordance with the NRC Enforcement Policy Section 2.2 and IMC 0612 Section 03.23, Part 72 ISFSI inspection findings follow the traditional enforcement process and are not dispositioned through the Reactor Oversight Process or the Significance Determination Process. The violation screened as having very low safety significance, Severity Level IV, and is being treated as a non-cited violation, consistent with Section 2.3.2.a. of the Enforcement Policy. The violation was determined to be more than minor since the licensee failed to establish, maintain, or implement adequate controls over procurement, construction, examination, or testing processes that are important to safety. The violation was entered into the licensees corrective action program as CR 2013-12884. Following identification of the issue the licensee performed an assessment that showed the cask would continue to perform its design function. Corrective actions for this issue included performing an extent of condition review, performing an apparent cause analysis report, reconstitution of the missing documents, conducting interviews with the dry cask loading personnel, providing training to the staff involved, and changing processes and responsibilities within FCS Records Management Group.
05000285/FIN-2015004-022015Q4Fort CalhounLicensee-Identified ViolationTechnical Specification (TS) 2.4(1)a.iv requires that all valves, piping, and interlocks associated with the components of the containment cooling system required to function during accident conditions be operable. In the event that any of these components, required to function during accident conditions become inoperable, the reactor shall be placed in a hot shutdown condition within 12 hours. The containment spray pumps and the associated piping are part of the containment cooling system. Prior to making modifications to containment spray piping in 2015, the operability of this piping would have been challenged by a main steam line break or a loss of coolant accident due to thermal stresses induced in the piping after a rise in containment temperature after the postulated event. Operation prior to the implementation of the modifications was a violation of the technical specification requirements to maintain operability of containment cooling systems. The violation is more than minor because it is associated with the design control attribute of the mitigating systems cornerstone because the failure to anticipate the rise in containment spray piping temperature dates back to the original design of the plant. This adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The violation was of very low safety significance because although the subject piping was inoperable due to exceeding code specified stress limits, analysis showed that the piping would have been able to perform its safety function to deliver adequate containment spray flow in the event of an accident. The licensee entered the issue into their corrective action program as Condition Report 2015-04578.
05000285/FIN-2015004-032015Q4Fort CalhounLicensee-Identified ViolationTechnical Specification (TS) 2.5(1) requires two trains of auxiliary feedwater (AFW) to be operable when cold leg temperature is above 300F. In the event that both trains become inoperable, immediate action is required to restore one AFW train to operable status. Technical Specification 2.0.1 and all TS actions requiring mode changes are suspended until one AFW train is restored to operable status. Operation with the main and auxiliary feedwater cross-tied was a violation of the technical specification requirements to maintain operability of AFW systems. The violation is more than minor because it is associated with the configuration control attribute of the mitigating systems cornerstone because the failure to prevent cross-tying these systems resulted in unrecognized inoperability of both trains of AFW. This adversely affects the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The violation was of very low safety significance because although MFW and AFW were momentarily cross-tied, this condition existed for only a brief period of time as operators restored system line-ups following system testing. In addition, a Senior Reactor Analyst evaluated the postulated main feedwater line break frequency and exposure time of the condition and determined the likelihood of this event during the exposure time is less than the Green/White threshold and of very low safety significance. The licensee entered the issue into their corrective action program as Condition Report 2015 03698.
05000530/FIN-2015004-012015Q4Palo VerdeLicensee-Identified ViolationTechnical Specification 3.0.4 requires, in part, that when an LCO is not met, entry into a mode or other specified condition in the applicability shall only be made when the associated actions in the mode permit continued operation; a risk assessment is performed and accepted for the inoperable components; or when an allowance is stated. Technical Specification 3.7.4, Atmospheric Dump Valves, requires that four ADV lines shall be operable in Modes 1, 2, 3, and 4 when the steam generator is relied upon for heat removal. Contrary to the above, on May 1, 2015, Unit 3 operators entered a mode with an LCO not met. Specifically, one atmospheric dump valve line was not operable as required by Technical Specification 3.7.4 prior to entering Mode 3. The licensees investigation concluded that the valve failure was a result of inadequate reassembly following maintenance. The licensee reported this condition in Licensee Event Report 05000530/2015-002-00 as a condition prohibited by Technical Specifications due to entering a mode in the applicability of LCO 3.7.4 while the LCO was not met. The inspectors concluded that the finding is of very low safety-significance (Green) because it was not a design or qualification deficiency, did not result in a loss of safety function, did not result in a loss of function of a train of safety equipment out greater than its allowed outage time, or a loss of function of high importance maintenance rule equipment greater than 24 hours. The licensee has entered the issue in the corrective action program as CRDR 4654422.
05000528/FIN-2015004-022015Q4Palo VerdeLicensee-Identified ViolationTitle 10 CFR 55.49, Integrity of examinations and tests, requires, in part, that facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Contrary to the above, during the week of November 9, 2015, the licensee caused a compromise of examination integrity when two licensed operators, who had previously validated portions of the 2015 annual operating test and had signed the examination security agreement, administered emergency preparedness (EP) job performance measures (JPMs) to a total of three licensed operators who had not yet taken their annual operating test. Specifically, the two licensed operators validated and/or approved simulator scenarios and EP JPMs for the annual operating test and then subsequently administered JPMs to three other licensed operators for the purpose of supporting EP program indicators. If not for detection, this activity could have affected the equitable and consistent administration of the annual operating examination. The failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV violation because it involved a nonwillful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because it had no actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Requalification. The licensee entered this issue into their corrective action program as Condition Report 15-10910.
05000298/FIN-2015008-022015Q2CooperFailure to Adequately Torque Fasteners on Emergency Diesel Generator Number 2The team reviewed a self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which occurred when the licensee failed to include specific instructions in work orders with respect to the use of lubrication during emergency diesel generator (EDG) fastener torquing. The failure to include specific lubrication instructions in work orders resulted in the inadequate torquing of bolting on the number 2 EDG and contributed to a lube oil leak during a surveillance run of the affected diesel. Procedures in effect during the time the fasteners were torqued required planners to include specific lubrication instructions in work orders for the EDGs. The licensee corrected the current issue by properly lubricating and torquing the fasteners for the right bank camshaft and restored the EDG 2 to operable status. The licensee entered this issue into the corrective action program as condition report CR-CNS-2014-06885. The failure to specify lubricants in EDG work order instructions involving fastener torquing, in violation of Procedure 7.2.53.12, Cooper Bessemer Bolting and Torque Program, is a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Additionally, if left uncorrected, it has the potential to lead to a more significant safety concerns, in that the failure to include these instructions in work orders has resulted in, and could continue to result in loose fasteners on the emergency diesel generator. Using Inspection Manual Chapter 0609, Appendix A, issued June 19, 2012, the Significance Determination Process for Findings At Power; the inspectors determined the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of a function of a single train for greater than the technical specification (TS) allowed outage time, and (4) did not represent an actual loss of a function of one or more non-TS trains of equipment. The finding has a cross-cutting aspect in the problem identification and resolution area due to the organizations failure to take effective corrective actions to address the deficiency after it was identified in a 2010 root cause evaluation and failure to recognize the ineffectiveness of the previous corrective actions until after the lube oil leak in 2014 (P.3).
05000298/FIN-2015008-032015Q2CooperMain Steam Isolation Valve Scram Closure condition Prohibited By Technical SpecificationsThe team identified two examples of a non-cited violation of Technical Specification 3.3.1.1, Reactor Protection System Instrumentation, required Action A, for the licensees failure to place inoperable main steam isolation valve closure scram channels in trip within 12 hours when Surveillance Requirement 3.3.1.1.9 to perform channel functional testing was not met. Specifically, on January 31 and May 16, 2015, the licensee tested inboard main steam isolation valves MS-AOV-80A and MS-AOV-80B limit switches associated with main steam isolation valve closure scram channel multiple times prior to declaring them operable. The licensee did not evaluate for pre-conditioning of the limit switches to determine if the actual as-found condition was masked, and did not ensure the discrepancy was corrected, before repeating the surveillance test. This resulted in repetitive testing to achieve acceptable results that led to declaring the limit switches operable. The station did enter the required action statements for Technical Specification 3.3.1.1 for MS-AOV-80A limit switch A on May 16, 2015, and MS-AOV-80B limit switch A on May 19, 2015. All inboard main steam isolation valve limit switches in question were replaced during Planned Outage 2015-01 conducted from May 30 to June 1, 2015. The licensee entered this issue into the corrective action program as condition reports CR-CNS-2015-03456, CR-CNS-2015-03483, and CR-CNS-2015-03484. The licensees failure to adequately assess operability during multiple performances of channel functional surveillance testing for reactor protection system main steam isolation valve closure scram function in violation of Technical Specification 3.3.1.1, Reactor Protection System Instrumentation, is a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Specifically, the licensee did not evaluate for pre-conditioning of the limit switches to determine if the actual as-found condition was masked, and ensure the discrepancies were corrected, before repeating the surveillance test. This resulted in repetitive testing to achieve acceptable results that led to declaring the limit switches operable. Using Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Finding At-Power, dated June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding: (1) did not affect a single reactor protection system trip signal to initiate a reactor scram and the function of other redundant trips or diverse methods of reactor shutdown (e.g. other automatic reactor protection system trips, alternate rod insertion, or manual reactor trip capacity); (2) did not involve control manipulations that unintentionally added positive reactivity (e.g., cold-water injection, inadvertent control rod movement, recirculation pumps speed control); and (3) did not result in a mismanagement of reactivity by the operator(s) (e.g., reactor power exceeding the licensed power limit, inability to anticipate and control changes in reactivity during crew operations). The finding has a cross-cutting aspect in the area of human performance associated with procedural adherence because individuals did not follow processes, procedures, and work instructions (H.8).
05000530/FIN-2015002-042015Q2Palo VerdeNotice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System Operating Conditions B and C(Open) Unresolved Item 05000530/2015002-04, TAC Number MF6276 - NOED Number 15-4-01. Notice of Enforcement Discretion of Technical Specification 3.5.3 Emergency Core Cooling System - Operating Conditions B and C On May 27, 2015, the licensee removed Unit 3 high pressure safety injection train A for planned maintenance. The following morning, during the maintenance, the licensee noted lube oil contamination, and determined that an outboard motor bearing had apparently failed during the last run following maintenance during the last refueling outage which involved disassembling and reassembling the bearing. The licensee identified procedural guidance inadequacies in the reassembly procedure that were the likely cause of the failure. The licensee could not perform required repairs in a controlled manner within the remaining action statement completion time, so on May 29, 2015, the licensee requested a Notice of Enforcement Discretion for a one-time action statement extension of 24 hours to allow time to reassemble and test the replacement bearings prior to restoring operability. The NRC granted that request as NOED 15-4-01. The licensee completed maintenance, testing, and restoration approximately 11 hours into the 24-hour extension window. In accordance with Inspection Manual Chapter 0410, Unresolved Item (URI) 05000530/2015002-04 is opened for NOED 15-4-01, and remains open pending further inspection and disposition in a future inspection report.
05000298/FIN-2015002-012015Q2CooperFailure to Prevent Reactor Thermal Power from Exceeding 2419 MWt for Preplanned ActivityThe inspectors identified a non-cited violation of Technical Specification 5.4.1.a, associated with the licensees failure to appropriately implement General Operating Procedure 2.1.10, Station Power Changes, Revision 107. Specifically, the procedure required in Step 10.3 that the licensee, Ensure any pre-planned evolution (e.g., pressure change, flow change, etc.) will not result in operation greater than 2419 MWt. On May 8, 2015, the licensee failed to implement Step 10.3 of General Operating Procedure 2.1.10, when they failed to reduce power to ensure that reactor power did not exceed 2419 MWt as the reactor recirculation motor generator B scoop tube was unlocked. As a result of this failure to reduce power for this planned evolution, reactor power increased to 2422 MWt. The licensee entered this deficiency into their corrective action program for resolution as Condition Report CR-CNS-2015-04259. The performance deficiency is more than minor, and therefore a finding, because it is associated with the human performance attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown as well as power operations. Specifically, the licensee did not know the condition of the reactor recirculation motor generator set B potentiometer prior to unlocking it and failed to reduce power such that when the scoop tube was unlocked, the resulting power increase would not exceed 2419 MWt. The inspectors screened the finding using Inspection Manual Chapter 0609, Appendix A, Exhibit 2, Section C, Reactivity Control Systems, which resulted in a Yes answer to Question 2 since the finding involved control manipulations that unintentionally added positive reactivity. This referred the inspectors to Inspection Manual Chapter 0609, Appendix M, Significance Determination Using Qualitative Criteria. A Senior Reactor Analyst performed a bounding qualitative evaluation and determined that the finding was of very low safety significance (Green) because of the relatively small magnitude of the overpower event, the prompt operator actions to return power to below the licensed limit upon discovery, and the fact that the overpower event did not result in any failure of the fuel cladding. This finding has a cross-cutting aspect in the area of human performance associated with conservative bias. Specifically, the affected evolution was known in advance to have the possibility of a positive reactivity impact; however, operators did not take appropriate actions to reduce power sufficiently prior to unlocking the reactor recirculation motor generator set B scoop tube in order to prevent the reactor from exceeding 2419 MWt (H.14).
05000298/FIN-2015008-012015Q2CooperFailure to Evaluate a Valve Degraded Condition before Returning the Valve to ServiceThe team identified a non-cited violation of Technical Specification 5.4.1.a regarding implementation of maintenance procedures for work on safety-related motor-operated valves (MOVs). Specifically, a degraded component within the actuator was not evaluated as acceptable to use as is before returning the valve to service. The Division 2 low-pressure coolant injection (LPCI) Throttle valve, RHR-MOV-MO27B, failed in the closed position during a surveillance test. The licensees investigation revealed that the helical motor pinion gear in the Limitorque valve actuator broke in three parts. This failed pinion gear additionally caused damage to part of the motor shaft where the setscrew engaged the shaft to attach the pinion gear. The licensees corrective action was to drill the setscrew hole slightly deeper, and reuse the motor shaft when reassembling the Limitorque motor actuator and returning the valve to an operable status. The licensee failed to document this process through an engineering evaluation to accept the setscrew and motor shaft repair use-as-is per their engineering change procedure. The evaluation was performed after the valve was returned to service and determined that the setscrew configuration was acceptable. The licensee entered this issue into the corrective action program as Condition Report CR-CNS-2015-00880 The licensees failure to perform an evaluation for a degraded condition when performing safety-related MOV maintenance in violation of Procedure 3-EN-DC-115, Engineering Change Process, is a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the human performance attribute of the mitigating systems cornerstone and affects the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events. Specifically, the performance deficiency resulted in the reuse of the motor shaft in the actuator to Valve RHR-MOV-MO27B, as acceptable to use-as-is even though a degraded condition existed, returning the valve to operable status without performing the required engineering evaluation. Using Inspection Manual Chapter 0609, Appendix A, issued June 19, 2012, the Significance Determination Process for Findings At Power, the inspectors determined the finding was of very low safety significance (Green) because the finding: (1) was not a deficiency affecting the design and qualification of a mitigating structure, system, or component, and did not result in a loss of operability or functionality; (2) did not represent a loss of system and/or function; (3) did not represent an actual loss of a function of a single train for greater than the technical specification (TS) allowed outage time; and (4) did not represent an actual loss of a function of one or more non-TS trains of equipment. The finding has a cross-cutting aspect in the area of human performance associated with Teamwork: Individuals and work groups communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the licensee failed to perform an evaluation of the setscrew location to ensure that that location was properly drilled and tapped. This was due to a lack of coordination between the maintenance and engineering groups (H.4)
05000368/FIN-2015001-022015Q1Arkansas NuclearFailure to Accurately Report Unplanned Scrams per 7000 Critical Hours Performance IndicatorThe inspectors identified a non-cited violation of 10 CFR 50.9, Completeness and Accuracy of Information, for the licensees failure to provide information to the NRC that was complete and accurate in all material respects. Specifically, the Unit 2 unplanned scrams per 7000 critical hours performance indicator data submitted to the NRC for the second and third quarters of 2014 was inaccurate. The performance indicator data submitted did not include a Unit 2 plant scram that occurred on April 27, 2014. When the second quarter and third quarter 2014 data was corrected and submitted to the NRC on March 4, 2015, the unplanned scrams per 7000 critical hours performance indicator transitioned from Green to White. The issue was entered into the licensees corrective action program as Condition Report CR-ANO-2-2015-00362. The licensee failed to provided information to the NRC that was complete and accurate in all material respects, as required by 10 CFR 50.9. The NRCs significance determination process (SDP) is not designed to assess the significance of violations that impact or impede the regulatory process. Therefore, the issue of two quarterly submittals of discrepant unplanned scrams performance indicator data was assessed using the traditional enforcement process in accordance with the Enforcement Policy. The inspectors determined the violation to be at Severity Level IV, because the licensee submitted inaccurate performance indicator data to the NRC that would have caused the performance indicator to change from Green to White (Enforcement Policy example 6.9.d.11). Traditional enforcement violations are not assigned a cross-cutting aspect.
05000368/FIN-2015001-012015Q1Arkansas NuclearFailure to Protect Safety Equipment From Potential High Energy Line BreaksThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," for the licensees failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions and that design changes were subject to design control measures commensurate with those applied to the original design. Specifically, the Unit 2 radwaste supply fans, 2VSF-7A and B, plenum doors and turbine building fire door 447 were maintained open, which provided a potential path for steam to enter the auxiliary building and impact both safety-related dc power trains during a high energy line break event in the turbine building. On February 12, 2014, the licensee suspended the modification and corrected the procedure. The licensee documented the concern in Condition Report CR-ANO-2-2014-00345. The licensees failure to maintain separation of safety related systems and high energy piping systems in accordance with design, as stated in the Safety Analysis Report, was a performance deficiency. The performance deficiency was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated July 1, 2012 and Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, dated July 1, 2012, the inspectors determined that the finding required a detailed risk evaluation because the finding represented a potential loss of system and/or function of the safety-related dc motor control centers, battery chargers and inverters. A senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was less than 4.8E-7/year (Green). The dominant core damage sequences included losses of the plants DC electrical systems. The initiating event likelihood of a rupture of the specific section of piping needed to initiate core damage sequences was extremely low. The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current licensee performance.
05000313/FIN-2015001-032015Q1Arkansas NuclearLicensee-Identified ViolationThe following violation of very low safety significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as a non-cited violation. Title 10 CFR 71.5, Section (a), Transportation of Licensed Material, requires each licensee who transports licensed material outside the site of usage, shall comply with the applicable requirements of the DOT regulations in 49 CFR. 49 CFR Part 172.800(b) requires, in part, that the licensee must develop and adhere to a transportation security plan. The licensee implemented Procedure EN-RW-106, Integrated Transportation Security Plan, to adhere to these requirements. Contrary to the above, on December 18, 2014, the licensee identified that they failed to follow their Transportation Security Plan (TSP). Specifically, licensee personnel shipped a radioactive quantity of Category 2, RAM-QC, on the public highways to a waste processor without acknowledging the shipment as a RAM-QC shipment or making appropriate notifications as required by Procedure EN-RW-106. Six shipments were identified as being shipped in violation of the TSP requirements because they failed to identify the material as RAM-QC due to inadequate Category 2 threshold values. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix D, Public Radiation Safety Significance Determination Process, dated February 12, 2008, the inspectors determined the finding has very low safety significance (Green) because the licensee had an issue involving transportation of radioactive waste, but it did not involve: (1) a radiation limit being exceeded, (2) a breach of package during transport,(3) a certificate of compliance issue, (4) a low level burial ground nonconformance, or (5) a failure to provide emergency information. The licensee documented this issue in their corrective action program as Condition Report CR-ANO-C-2014-03341 and made corrections to Procedure EN-RW-106 to prevent this issue from reoccurring. There is no cross-cutting aspect with this violation due to it being licensee-identified.
05000313/FIN-2014005-032014Q4Arkansas NuclearFailure to Correct Weaknesses During Drills and ExercisesThe inspectors identified a non-cited violation of 10 CFR Part 50.47(b)(14) for the failure to correct a deficiency identified in a 2013 simulator drill. Specifically, control room operators did not implement the procedure that describes how the site will maintain continuous communication with threat notification sources during a drill conducted August 7, 2013, and also during the September 16, 2014, biennial exercise. The inspectors determined that the licensees corrective actions for this issue were incomplete and did not address the extent of condition. The licensee has entered the issue into the corrective action program in corrective action documents WT-WTANO-2014-00189 and Condition Report CR-ANO-C-2014-02478. The failure to correct weaknesses occurring in drills and exercises is a performanc deficiency within the licensees ability to foresee and correct. The performance deficiency is more than minor because it is associated with the emergency response organization performance attribute of the Emergency Preparedness Cornerstone and it adversely impacted the cornerstone objective. The licensees ability to implement adequate measures to protect the health and safety of the public in the event of hostile action and a radiological emergency is degraded when it fails to correct performance that precludes the effective implementation of the emergency plan. This finding was evaluated using Manua Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process (SDP), Attachment 2, dated February 24, 2012, and was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements, was not associated with a risk-significant planning standard, and was not a loss of planning standard function. The finding was not a loss of function because the deficiency that was identified was not associated with classification, notifications to state and local agencies, or the development of protective action recommendations. The finding was assigned a cross-cutting aspect in the area of problem identification and resolution, associated with the resolution of issues because the licensee failed to evaluate the initial performance issues to ensure that resolutions adequately addressed the extent of condition commensurate with their safety significance. The licensee failed to recognize in August 2013 that continuous communications with threat notification sources is required by regulation and that performance issues with the implementing procedure should be communicated to the entire control room staff population (P.2).
05000368/FIN-2014005-012014Q4Arkansas NuclearFailure to Develop Adequate Guidance for Extreme Damage MitigationThe inspectors identified a noncited violation of 10 CFR 50.54(hh)(2) for the failure to develop mitigating strategy guidance that would successfully maintain or restore Unit 2 core cooling after the loss of large areas of the plant. Specifically, the guidance did not ensure the capability of the mitigating strategy because an unisolated flow diversion could have prevented water from reaching the steam generators and cooling the core. The issue was documented in Condition Report CR-ANO-2-2014-03277, and the procedure was revised to correct the condition. The licensees failure to develop mitigating strategy guidance that would successfully maintain or restore Unit 2 core cooling after loss of large areas of the plant, as required by 10 CFR 50.54(hh)(2), was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems. Specifically, the guidance did not ensure the capability of the mitigating strategy because an unisolated flow diversion could have prevented water from reaching the steam generators and cooling the core. Using NRC Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and NRC Manual Chapter 0609, Appendix L, B.5.b Significance Determination Process, dated December 24, 2009, Table 1, SDP Screening Worksheet for B.5.b, the finding was determined to be of very low safety significance because the performance deficiency represented the unrecoverable unavailability of an individual mitigating strategy; other core cooling mitigating strategies were available. This finding has a human performance crosscutting aspect associated with avoid complacency, in that the licensee failed to recognize and plan for the possibility of latent issues, even while expecting successful outcomes (H.12).
05000313/FIN-2014005-022014Q4Arkansas NuclearFailure to Provide Flow Protection For Auxiliary Feedwater Pump in Emergency Operating ProceduresInspectors identified a noncited violation of Unit 1 Technical Specification 5.4, Procedures, for the licensees failure to establish adequate emergency operating procedures. Specifically, the licensees emergency operating procedures failed to establish minimum flow protection for the Unit 1 auxiliary feedwater pump, which could result in catastrophic failure of the pump. The issue was documented in Condition Report CR-ANO-1-2014-00286 and the procedures were revised to correct the condition. The failure to establish minimum flow protection for the Unit 1 auxiliary feedwater pump in emergency and abnormal operating procedures in accordance with the emergency operating procedure writers guide was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with 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, inadequat emergency and abnormal operating procedures could have resulted in failure of the auxiliary feedwater pump, a mitigating system for a loss of main and emergency feedwater. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding required a detailed risk evaluation because the finding represented a loss of system function. A Region IV senior reactor analyst performed the detailed risk evaluation and determined that the change to the core damage frequency was less than 4.2E-7/year (Green). The dominant core damage sequences included losses of one of the safety related 4160 volt electrical buses, steam generator tube ruptures, and plant transients. The equipment that helped mitigate the risk included the high pressure injection system (for feed and bleed) and the main and emergency feedwater systems. This finding did not have a cross-cutting aspect because the most significant contributing cause was not indicative of current performance. Specifically, the emergency and abnormal operating procedures for operating auxiliary feedwater had not changed for at least 2 years.
05000368/FIN-2014005-042014Q4Arkansas NuclearFailure to Verify Ventilation Design for Vital SwitchgearInspectors identified a noncited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to maintain design control of the Unit 2 vital switchgear ventilation system. Specifically, in 2002, the licensee failed to ensure that the ventilation system was capable of cooling the switchgear under design basis conditions. The licensee documented the issue in Condition Report CR-ANO-2-2014-00352 and conducted an evaluation to verify the capability of the ventilation system. Failure to ensure that the ventilation system was capable of cooling the switchgear under design basis conditions was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined that the finding was of very low safety significance (Green) because the finding was a deficiency affecting the design of a mitigating system, and the system maintained its functionality. The inspector determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current licensee performance.
05000368/FIN-2014005-052014Q4Arkansas NuclearUnit 2 Unplanned Scrams Performance IndicatorThe inspectors identified an unresolved item associated with the Unit 2 unplanned scrams per 7,000 critical hours performance indicator related to a reactor trip. On April 27, 2014, Unit 2 experienced an Axial Shape Index (ASI) trip when performing a rapid downpower at the request of the transmission grid operator due to severe weather affecting the grid. This unplanned reactor trip was caused by exceeding the Core Protection Calculator ASI limits. As noted in Licensee Event Report 05000368/2014-003-00, and NRC Inspection Report 2014004, the ASI limits were exceeded, due in part to plant operators not following the downpower reactivity plan. The automatic trip occurred at approximately 50 percent power and was uncomplicated. The unplanned scrams per 7000 critical hours performance indicator measures the rate of scrams per year of operation at power and provides an indication of initiating event frequency. The licensee did not include this scram as an input into the unplanned scram performance indicator and submitted a frequently asked question to the NRC Reactor Oversight Process Working Group. The frequently asked question is currently under review to decide whether the above event should be captured as an unplanned scram. The licensee noted that anticipatory plant shutdowns to reduce the impact of externa events are excluded from this performance indicator. The licensee believed the intent of the exclusion was met because the shutdown being performed at the time the reactor trip occurred had been requested by the transmission grid operator due to the impacts of weather conditions. The inspectors noted that Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 7, guidance states that an unplanned scram is a scram that is not an intentional part of a planned evolution or test as directed by a normal operating or test procedure. This includes scrams that occurred during the execution of procedures or evolutions in which there was a high chance of a scram occurring but the scram was neither planned nor intended. The inspectors noted that the April 27, 2014 reactor trip was an automatic trip, which was not intended as part of the rapid downpower evolution that was being performed. The inspectors also noted that had the licensees reactivity plan been followed, the severity of the ASI transient would likely have been managed and a trip avoided. The inspectors concluded that additional inspection was required to assess whether the scram should have been reported in the unplanned scrams per 7,000 critical hours performance indicator for Unit 2. This issue was identified as Unresolved Item URI 05000368/2014005-05, Unit 2 Unplanned Scrams Performance Indicator.
05000368/FIN-2014004-022014Q3Arkansas NuclearFailure to Establish Preventative Maintenance on Unit 2 Main Steam Isolation ValvesInspectors documented a Green self-revealing non-cited violation of Technical Specification 6.4.1.a for the licensees failure to establish procedures recommended by Regulatory Guide 1.33, Revision 2, Appendix A, Section 9, February 1978. Specifically, the licensee failed to establish preventative maintenance procedures for valve internal inspection and testing of the Unit 2 main steam isolation valves. On December 23, 2013, the train A main steam isolation valve (2CV-1010-1) was declared Inoperable due to the valve sticking at fifteen percent open on multiple stroke attempts. The licensees cause evaluation identified that mechanical binding and corrosion of the valve internals were results of a lack of preventive maintenance. The licensee repaired the 2CV-1010-1 valve and performed subsequent testing to demonstrate Operability. The issue was documented in Condition Report CR-ANO-2-2013-02502. The inspectors determined that the failure to establish preventative maintenance procedures for valve internal inspection and testing of the Unit 2 main steam isolation valves was a performance deficiency. The performance deficiency is more than minor because it was associated with 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, and is therefore a finding. Specifically, the lack of preventative maintenance adversely affected the reliability of the main steam isolation valve 2CV-1010-1 to close within the time assumed in the accident analysis. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 2, the inspectors determined the finding to be of very low safety significance (Green) because the finding did not represent the loss of a system safety function and did not represent an actual loss of safety function of at least one train for greater than its technical specification allowed outage time. The finding was determined to have a cross-cutting aspect in the area of problem identification and resolution, in that the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes commensurate with their safety significance. Specifically, during a previous stroke test of the 2CV-1010-1 valve in 2011, the licensee identified that the valve experienced a sluggish or jerky motion and took longer than normal to open. The licensee entered this issue into the corrective action program but did not fully evaluate and troubleshoot the condition adverse to quality to ensure resolution of the cause.
05000313/FIN-2014004-042014Q3Arkansas NuclearLicensee-Identified ViolationThe following violation of very low safety significance (Green) and Severity Level IV wa identified by the licensee and is a violation of NRC requirements which meets the criteria of th NRC Enforcement Policy, for being dispositioned as a Non-Cited Violation. Title 10 CFR 55.49, Integrity of Examinations, requires, in part, that facility licensees shall not engage in any activity that compromises the integrity of any application, test, or examination required by this part. Contrary to the above, on June 24, 2014, the licensee caused a compromise to examination integrity by violating an examination security agreement to not divulge information about examination content to unauthorized individuals. The failure to meet 10 CFR 55.49 was evaluated through the traditional enforcement process because it impacted the ability of the NRC to perform its regulatory oversight function. This resulted in assignment of a Severity Level IV violation because it involved a non-willful compromise of examination integrity and is consistent with Section 6.4.d of the NRC Enforcement Policy. The associated performance deficiency was screened as Green because there was not an actual effect on the equitable and consistent administration of any examination required by 10 CFR 55.59, Requalification. The licensee entered this issue into their corrective action program as Condition Report CR-ANO-1-2014-01062.
05000313/FIN-2014004-012014Q3Arkansas NuclearImproper Maintenance on Circuit Breaker Caused Loss of Unit 1 Decay Heat Removal PumpInspectors documented a Green self-revealing non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to ensure activities affecting quality were accomplished in accordance with documented instructions. Specifically, the licensee failed to follow Job Order JO-00968863 for replacement of a prop spring in circuit breaker MA137. As a result, the wrong prop spring was replaced, reducing the reliability of the Unit 1 train B decay heat removal pump P-34B and ultimately causing a failure of the pump to start. The licensee corrected the condition by replacing the breaker and returning the pump to service. The issue was documented in Condition Report CR-ANO-1-2013-00701. The inspectors determined that the failure to follow Job Order JO-00968863 in 1998 for replacement of a prop spring in circuit breaker MA137 was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences, and was therefore a finding. Specifically, the failure to replace the appropriate prop spring in 1998 adversely affected the availability and reliability of Unit 1 decay heat removal pump P-34B and caused a failure to start in 2013. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 3, Mitigating Systems Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding did not represent a loss of system safety function and did not represent an actual loss of safety function of at least one train for greater than its technical specification allowed outage time. The inspectors determined that there was no cross-cutting aspect associated with this finding because the cause of the performance deficiency occurred more than three years ago, and was not representative of current licensee performance.
05000445/FIN-2014004-022014Q3Comanche PeakInadequate Acceptance Criteria for Inservice Testing of Auxiliary Feedwater Discharge Check ValvesThe inspectors identified a non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to incorporate adequate acceptance limits in a written procedure to demonstrate components will perform satisfactorily. The licensee used acceptance criteria for inservice testing that did not demonstrate successful performance of the test. Specifically, the licensee failed to use appropriate acceptance limits which would have identified a failed check valve when testing auxiliary feedwater discharge check valves. The licensee revised the inadequate test procedure. The licensee entered the finding into the corrective action program as Condition Report CR-2014-010082. The licensees failure to incorporate adequate acceptance limits in a written procedure to demonstrate components perform satisfactorily was a performance deficiency. The finding was more than minor because it was associated with 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. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating Systems Screening Questions, the finding was determined to be of very low safety significance (Green) because although the finding was a deficiency affecting the design or qualification of a mitigating system, the system maintained its operability and functionality. The inspectors determined that the finding was not representative of current licensee performance and no cross-cutting aspect was assigned.
05000445/FIN-2014004-012014Q3Comanche PeakFailure to Establish Goals and Monitor the Performance of the Uninterruptible Power Supply Air Conditioning SystemThe inspectors identified a non-cited violation of 10 CFR 50.65(a)(1) for the failure to establish performance goals and perform monitoring to ensure the uninterruptible power supply air conditioning unit X-01 was capable of performing its intended function. Specifically, the licensee failed to include unavailability hours that caused the equipment to exceed the performance criteria. The licensee planned to establish goals for the system. The licensee entered the finding into the corrective action program as Condition Report CR-2014-010188. The failure to establish goals and monitor the performance of the uninterruptible power supply air conditioning system was a performance deficiency. The performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability and reliability of systems that respond to initiating events to prevent undesirable consequences. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 2, Mitigating System Screening Questions, the finding was determined to be of very low safety significance (Green) because the finding did not represent an actual loss of safety function of a system and did not represent an actual loss of a technical specification train for greater than its allowed outage time. The finding has a human performance cross-cutting aspect associated with procedure adherence because the engineer failed to use human error reduction techniques when following procedure (H.8).
05000313/FIN-2014009-012014Q3Arkansas NuclearInadequate Flood Protection for Auxiliary and Emergency Diesel Fuel Storage BuildingsApparent Violation. The inspectors identified a finding of preliminary substantial safety significance (Yellow) for the failure to design, construct, and maintain the Units 1 and 2 auxiliary and emergency diesel fuel storage buildings in accordance with the safety analysis reports' description of internal and external flood barriers so that they could protect safety-related equipment from flooding. Two apparent violations were associated with this finding: Contrary to 10 CFR Part 50, Appendix B, Criterion Ill, "Design Control," the licensee failed to assure that regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions, and that design changes were subjected to design control measures commensurate with those applied to the original design. Contrary to 10 CFR Part 50, Appendix B, Criterion V, "Instructions, Procedures, and Drawings," the licensee failed to prescribe documented instructions for activities affecting quality and accomplish activities affecting quality in accordance with drawings. The licensee entered these issues into the corrective action program as Condition Reports CR-ANO-C-2013-01304 and CR-ANO-C-2014-00259. The licensee resolved the safety concern by replacing the degraded seals or parts, installing penetration seals, implementing compensatory measures, and/or incorporating instructions into procedures. The inspectors determined that the finding was more than minor because it was associated with the protection against external factors attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in the vulnerability to flooding of safety-related equipment necessary to maintain core cooling in the auxiliary and emergency diesel fuel storage buildings. The inspectors used Inspection Manual Chapter 0609, Attachment 0609.04, "Initial Characterization of Findings," dated June 19, 2012, and Appendix A, "The Significance Determination Process (SOP) for Findings At-Power," dated June 19, 2012, to evaluate the significance of the finding. In accordance with Appendix A, Exhibit 4, the inspectors determined that a detailed risk evaluation was necessary because, if the flood barriers were assumed to be completely failed, two or more trains of a multi-train system would be degraded during an external flood. The NRC risk analysts determined that the finding should be evaluated in accordance with NRC Inspection Manual Chapter 0609, Appendix M, "Significance Determination Process Using Qualitative Criteria," April 12, 2012. Appropriate quantitative significance determination process tools did not exist to provide a reasonable estimate of the significance because a plant-specific flood hazard analysis did not exist and was not expected to be available until sometime in 2015. The risk analysts used NRC Inspection Manual Chapter 0609, Appendix M, Table 4.1, "Qualitative Decision-Making Attributes for NRC Management Review," to determine the preliminary safety significance of the finding. The following were the dominant considerations in reaching a preliminary risk determination conclusion: With respect to the auxiliary and emergency diesel fuel storage buildings, there were more than 100 unknown ingress pathways for a flooding event, therefore if an external flood above grade level were to occur, the buildings would flood. The unexpected rate of flooding would likely be beyond the licensee's capability to prevent or mitigate as equipment and connections associated with alternative mitigating strategies, could be submerged. All reactor core cooling and makeup could fail due to significant flooding of the auxiliary and emergency diesel fuel storage buildings. The change in core damage frequency was quantitatively bounded below 2 x 1 o-3 and qualitatively determined to likely be less than 1 x 1 o-4. The bounding and qualitative results are based on the frequency of the probable maximum flood event and a loss of all equipment needed for core cooling and makeup. This finding was preliminarily determined to be of substantial safety significance (Yellow) for Unit 1 and Unit 2, as determined by a Significance and Enforcement Review Panel. This finding had a cross-cutting aspect in the area of human performance related to maintaining design margins. Specifically, the licensee did not design, construct, and/or maintain over 100 flood barriers to ensure design margins were sustained (H.6).
05000445/FIN-2014004-032014Q3Comanche PeakMaintenance of a Standard Action Level Scheme for Main Steam Line MonitorsThe inspectors identified a non-cited violation of 10 CFR 50.54(q)(2) for the failure to follow and maintain the effectiveness of an emergency plan that meets the requirements of planning standard 50.47(b)(4), which requires that a standard emergency classification and action level scheme is in use by the licensee. Specifically, several main steam line monitors were out of service for extended periods of time without apparent contingency actions in place in order to be able to declare an emergency. The licensee entered the finding into the corrective action program as Condition Report CR-2014-005874. The failure to maintain a standard emergency classification and action level scheme for the initiating condition requiring the main steam line monitors was a performance deficiency. The performance deficiency was more than minor because it affected the licensees ability to implement adequate measures to protect the health and safety of the public. Using Inspection Manual Chapter 0609, Appendix B, Emergency Preparedness Significance Determination Process, and Table 5.4-1, Significance Examples 50.47(b)(4), the finding was determined to be of very low safety significance (Green) because it was a failure to comply with NRC requirements and was not a degraded risk significant planning standard function. The planning standard function was not degraded because of other emergency action levels; an appropriate declaration could be made in an accurate and timely manner. This finding has a problem identification and resolution cross-cutting aspect associated with evaluation because the licensee failed to thoroughly evaluate the extent of condition of the inoperable monitors on the emergency plan and scheme for declaring emergencies (P.2).
05000445/FIN-2014004-042014Q3Comanche PeakFailure to Install an Insulated Bushing on a Generator Current Transformer Circuit Results in an Automatic Reactor TripThe inspectors reviewed a self-revealing finding for the licensees failure to follow an electrical installation specification and install an insulated bushing on the end of a flexible conduit. As a result, a generator current transformer conductor shorted to ground causing a generator trip and ultimately an automatic reactor trip. The licensee repaired the conductor and returned the unit to service. The licensee entered the finding into the corrective action program as Condition Report CR-2014-000579. The failure to follow an electrical installation specification and install an insulated bushing on the end of a flexible conduit was a performance deficiency. The performance deficiency was more than minor because it was associated with the human performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Appendix A, Exhibit 1, Initiating Event Screening Questions, the finding was determined to be of very low safety significance (Green) because although the finding caused a reactor trip, it did not involve the loss of mitigation equipment. The inspectors determined that the finding was not representative of current licensee performance and no cross-cutting aspect was assigned.
05000368/FIN-2014004-032014Q3Arkansas NuclearFailure to Implement Procedural Requirements for Axial Shape Index during a Rapid Power ReductionThe inspectors documented a Green self-revealing non-cited violation of Technical Specification 6.4.1.a for the failure to implement procedures for changing load recommended by Regulatory Guide 1.33, Revision 2, Appendix A, Section 2.f, dated February 1978. Specifically, the licensee did not maintain axial shape index within the limits of the core operating limits report during a rapid power reduction at the end of core life, resulting in an automatic reactor trip. The issue was documented in Condition Report CR-ANO-C-2014-01142. The inspectors determined that the failure to maintain axial shape index within the limits of the core operating limits report during a rapid power reduction was a performance deficiency. The performance deficiency is more than minor because it is associated with the human performance attribute of the Initiating Events Cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge the critical safety functions during shutdown as well as power operations. Specifically, the failure to maintain axial shape index caused an automatic reactor trip. Using Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, Exhibit 1, Initiating Events Screening Questions, the inspectors determined the finding to be of very low safety significance (Green) because the finding did cause a reactor trip but did not cause a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The finding has a cross-cutting aspect in the area of human performance associated with training because the organization did not provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce. Specifically, the operators were not trained to understand the effects of the axial shape index during rapid power reductions with a core at an End-of-Life condition.
05000313/FIN-2014003-072014Q2Arkansas NuclearInadequate Filling and Venting of High Pressure Injection PumpThe inspectors identified a non-cited violation of Unit 1 Technical Specification 5.4 Procedures, for the licensees failure to establish adequate instructions for filling an venting the emergency core cooling system. Specifically, an inadequate fill and vent could have allowed gas voids to enter the suction of an operable high pressure injection pump. As immediate corrective actions, the licensee revised the filling and venting instructions. The issue was documented in Condition Report CR-ANO-1-2014-00295 The failure to establish adequate fill and vent instructions for a drained high pressur injection pump was a performance deficiency. The performance deficiency was more than minor because it was associated with 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, and was therefore a finding. Specifically, the inadequate fill and vent instructions caused a high pressure injection pump to become inoperable for the standby emergency core cooling function. Using Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the inspectors determined this finding was of very low safety significance (Green) because the degraded condition was not a design or qualification deficiency; did not represent an actual loss of function or a system; did not represent an actual loss of function of a single train or two separate trains for greater than its technical specification allowed outage time; did not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant; and did not screen as potentially risk-significant due to a seismic, flooding, or severe weather initiating event. The finding had a cross-cutting aspect in the area of problem identification and resolution for the licensees failure to effectively evaluate and implement external operating experience. Specifically, the licensee failed to effectively evaluate and implement gas voiding operating experience when establishing Unit 1 fill and vent instructions (P.5). (Section 1R19)
05000382/FIN-2014003-022014Q2WaterfordFailure to Maintain Adequate Public Address System to Implement Onsite Protective ActionsThe inspectors identified a non-cited violation of 10 CFR Part 50.54(q)(2) for a failure to maintain the effectiveness of an emergency plan that meets the planning standards of 10 CFR Part 50.47(b). Specifically, the licensee failed to maintain the public address system in a manner that could provide prompt protective action notifications via voice or emergency alarms to all areas and buildings on the plant site. The capability to implement onsite protective actions for its workers is required by 10 CFR Part 50.47(b)(10). The licensee implemented compensatory measures while the system was being restored. Based on communications from the licensee on January 14, 2014, signs have been placed on entrances to areas affected by the non-functional public address speakers detailing alternate radio communications protocols that must be used while in the areas. In addition, public address speaker communications were sent out via group pagers and plant radio systems as well to enhance the ability to reach all workers. These compensatory measures have been communicated to their operations staff via written instructions in their daily turnover documentation. The licensee entered the issue into the corrective action program as Condition Report CR-WF3-2013-05860. The failure to maintain the effectiveness of the means to warn or advise onsite individuals of the range of protective measures consistent with the licensees emergency plan was a performance deficiency. The performance deficiency is more than minor because it is associated with the facilities and equipment attribute of the emergency preparedness cornerstone and it adversely impacted the objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. In addition, if left uncorrected, continued degradation of the public address system could lead to workers not receiving emergency instructions in a manner timely enough to ensure their safety. Using NRC Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings; and the corresponding Appendix B, Emergency Preparedness Significance Determination Process (SDP), the finding was determined to have very low safety significance (Green) because it did not result in a loss of risk-significant planning standard function, a risk-significant planning standard degraded function, or a loss of planning standard function. The finding had a cross-cutting aspect in the evaluation area of problem identification and resolution, associated with thoroughly evaluating issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. From August 2011 to December 4, 2013, as documented by multiple condition reports, there have been many instances of speaker and system component failures that have resulted in fixing failed components only without addressing the underlying conditions causing those failures. None of the failures caused the licensee to question whether they fully understood the reasons for the repetitive failures and whether alternative actions were necessary to correct the causes (P.2).
05000445/FIN-2014003-032014Q2Comanche PeakMaintenance of a Standard Action Level Scheme for Main Steam Line MonitorsIntroduction. The inspectors identified an unresolved item related to maintaining the effectiveness of the licensees emergency plan that meets planning standard 50.47(b)(4), which requires, in part, that a standard emergency classification and action level scheme is in use by the licensee. Specifically, several main steam line monitors were out of service for extended periods of time without apparent contingency actions in place to ensure the correct emergency action level would be implemented. Description. On November 20, 2013, the licensee initiated Condition Report CR-2013-011914 identifying that the main steam line radiation monitors had a trend of being out-of-service for significant time periods. Monitor 1-RE-2328 was out of service for 110 and 210 days on two separate occasions. Monitor 1-RE-2326 was out of service for 77 and 157 days on two separate occasions. Monitor 1-RE-2325 was out of service for 61 days. Four other monitors from the two units had been out of service, some more than once, for periods of five days or less. There are four online main steam line monitors for each unit. The licensee addressed the trend by trouble-shooting, repairing, and replacing detectors. The main steam line radiation monitors are important to emergency preparedness because they are inputs into the emergency action levels and define the initiating conditions related to abnormal radiation releases/radiation effluent emergency declarations. The inspectors determined that the licensee had taken appropriate action to initiate corrective action and repair. The licensee also tracked the out of service time of the monitors as operational focus items and in the station tactical equipment issues list. All eight main steam line monitors are currently in service with zero out of service days in 2014. However, there was no evidence that contingency actions were implemented to maintain the approved emergency action level scheme when the monitors were out of service. Title 10 CFR 50.54(q) requires licensees to maintain the effectiveness of an emergency plan that meets the requirements in the planning standards of 50.47(b). Title 10 CFR 50.47(b)(4) requires a standard emergency classification and action level scheme, the bases of which include facility system and effluent parameters, is in use by the nuclear facility licensee. This issue was identified as an unresolved item because the NRC has not determined whether the licensee has adequately implemented planning standard 10 CFR 50.47(b)(4). Specifically, the NRC has not determined whether the emergency action level initiating condition was rendered ineffective, such that, any general emergency would not be declared for a particular off-normal event in an accurate and timely manner or in a degraded manner. The licensee has entered this issue into the corrective action program as Condition Report CR-2014-005874. This issue is identified as unresolved item URI 05000445/2014003-03; 05000446/2014003-03, Maintenance of a Standard Action Level Scheme for Main Steam Line Monitors.
05000382/FIN-2014003-012014Q2WaterfordFailure to Control Entry into a High Radiation AreaThe inspectors reviewed a self-revealing, non-cited violation of Technical Specification 6.12.1 because a worker entered a high radiation area, but was not on a radiation work permit that authorized entry and was not knowledgeable of the dose rates in the area. Specifically, on April 14, 2014, a worker entered shutdown heat exchanger room B, a posted high radiation area during crud burst operations, and received an unanticipated electronic dose rate alarm of 107 millirem per hour. Radiation protection personnel counseled the worker, revoked his access to radiological controlled areas, and documented the occurrence in the corrective action program as Condition Report CR-WF3-2014-01638. The entry into a high radiation area while not on a radiation work permit that allows entry into high radiation areas and without knowledge of the dose rates in the area is a performance deficiency. The performance deficiency is more than minor and a violation of Technical Specification 6.12.1 because it impacted the program and process attribute (exposure control) of the occupational radiation safety cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Using Inspection Manual Chapter 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, the inspectors determined the violation has very low safety significance because: (1) it was not as low as is reasonably achievable (ALARA) finding, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. This violation has a cross-cutting aspect in the human performance area, associated with an individuals failure to implement appropriate error reduction tools necessary for avoiding complacency by recognizing and planning for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes (H.12).
05000368/FIN-2014003-092014Q2Arkansas NuclearReporting of Unit 2 Events as Unplanned Scrams with ComplicationsThe inspectors identified an unresolved item associated with not reporting two events in the unplanned scrams with complication performance indicator for Unit 2. Description. On March 31, 2013, and December 9, 2013, Unit 2 experienced a loss of condenser vacuum due to the transfer of electrical busses to startup transformer 2. By design, the lockout of the preferred offsite power source startup transformer 3 in these events resulted in the loss of non-vital circulating water pumps and the subsequent loss of condenser vacuum. The loss of condenser vacuum ultimately resulted in the loss of main feedwater pump capability. Both main feedwater pumps are steam driven a Unit 2. Neither of these concerns were reported under this performance indicator Unit 2 has a non-vital electric-driven feedwater pump, 2P-75, which remained availabl and capable of supplying sufficient feedwater flow to remove decay heat up throug about 4 percent reactor power. The non-vital 2P-75 pump, which can be supplie directly from the condensate storage tanks, does not rely on condenser vacuum o portions of the main feedwater system, and supplies feedwater for plant cooldown heatup, hot standby conditions, and startup. The NRC stated that the intent of the guidance in Nuclear Energy Institute Document 99-02, Regulatory Assessment Performance Indicator Guidance, for unplanned scrams with complications was to have the main feedwater available or recoverable within 30 minutes after a trip or scram assuming a loss of emergency feedwater to ensure safe shutdown of the plant The licensee submitted a frequently asked question to the Nuclear Energy Institut working group because the licensee considered the trip to be uncomplicated because a least one or more electric-driven feedwater pumps remained available as backup to th emergency feedwater system. The Nuclear Energy Institute guidance refers to a electric-driven main feedwater source; however, the intent was to provide backu feedwater capability should emergency feedwater be lost, which would be met b the 2P-75 pump. The frequently asked question is currently under review by NR headquarters and the Nuclear Energy Institute working group to decide whether o not the above events should be captured as unplanned scrams with complications The inspectors concluded that an additional inspection was required to assess whethe or not the events should have been included in the unplanned scrams with complications performance indicator for Unit 2. This issue was identified as an Unresolved Item URI 05000368/2014003-09, Reporting of Unit 2 Events as Unplanned Scrams with Complications.
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