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05000482/FIN-2018010-042018Q3Wolf CreekFailure to Identify 125 VDC Equalizing Voltage Exceeded Design RequirementsThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify or check the adequacy of design calculation NK-E-001, 125 Volt Direct-Current (VDC) Class 1E Battery System Sizing, Voltage Drop and Short Circuit Studies, Revision 4. The licensee failed to recognize that the actual 125 VDC Class 1E bus voltages had exceeded the maximum design limit voltages for downstream equipment identified in the calculation, and they had not placed any limits on voltages which could exceed the design limit of 140 VDC on the Class 1E System components.
05000482/FIN-2018010-032018Q3Wolf CreekFailure to Correct Reoccurring Problems with Time Critical/Sensitive Action ActivitiesThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to correct reoccurring problems with completing Time Critical/Time Sensitive Action issues.
05000482/FIN-2018010-022018Q3Wolf CreekFailure to Establish an Adequate Procedure for Operator Time Critical Actions ValidationThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to have an adequate Procedure. Procedure AI 21-016, Operator Time Critical Actions Validation, Revision 14, Attachment B Time Sensitive Action List, does not have unique identifiers for cross referencing the records to the procedure.
05000482/FIN-2018010-012018Q3Wolf CreekFailure to Follow ProceduresThe team identified two examples of a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for failure to follow procedures.
05000445/FIN-2018010-022018Q3Comanche PeakFailure to Provide Procedural Guidance for the Failure of a Component Cooling Water Surge Tank Makeup ValveThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to provide procedural guidance for the failure of a component cooling water surge tank makeup valve.
05000445/FIN-2018010-012018Q3Comanche PeakFailure to Establish Test Program to Verify Residual Heat Removal Suction Valve CapabilityThe inspectors identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to establish a test program to ensure that residual heat removal suction isolation valves would perform adequately in service.
05000482/FIN-2018010-052018Q3Wolf CreekLicensee-Identified ViolationThis violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and non-conformances are promptly identified and corrected.Contrary to the above, prior to 2015, the licensee failed to promptly identify and correct a repetitive deficiency or non-conformance. Specifically, the licensee had identified a leaking flange on the residual heat removal heat exchanger since 1997. Prior to 1997 a different data base had been used to record boric acid leakage, and the data was not available during the inspection.Over the years since plant startup, the licensee had been diligent in completing boric acid evaluations on the leaking residual heat removal heat exchanger flange, indicating minimal wastage of the flange closure studs and nuts that had been subjected to boric acid. Corrective actions included cleaning up the boric acid leakage, and checking or re-torqueing the closure nuts. These measures did not correct the problem of the leaking heat exchanger flange. In 2015 the licensee completed an in-depth engineering evaluation of the leaking flange, including discussions with the heat exchanger manufacturer. New corrective measures included changing the torque values on the closure studs and nuts. The licensee is still evaluating the results of the corrective actions taken to preclude further leakage.
05000298/FIN-2018011-042018Q2CooperIncorrect Classification of Potential Safety-Related ComponentsAn NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the inspectors identified three examples of the licensees failure to properly classify potential safety-related components in the emergency diesel generator ventilation system and RHR service water booster pump room cooling systems.
05000298/FIN-2018011-032018Q2CooperInadequate Design Basis Calculation for the EDG Rooms Temperature DistributionAn NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for the licensees failure to ensure design control measures provide for verifying or checking the adequacy of design of the emergency diesel generator room ventilation system by use of alternate or simplified calculation methods, or by a suitable testing program. Specifically, the licensee incorrectly extrapolated the results of the test program, which led to an incorrect room temperature profile. Additionally, the design calculation did not assume potential failures of the CO2 dampers.
05000298/FIN-2018011-022018Q2CooperFailure to Ensure Adequate Design Control Measures are in Place Associated with RHR Service Water Booster Pump Room CoolingAn NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, occurred for failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to incorporate malfunctions of the residual heat removal (RHR) service water booster pump (SWBP) room cooling temperature switch, which could cause environmental changes leading to functional degradation of system performance, into the design basis to verify the necessary protection system action be retained.
05000298/FIN-2018011-012018Q2CooperFailure to Correct Extent of Condition of Surge Suppression Varistor FailuresAn NRC-identified, Green, Non-cited Violation of Title 10, Code of Federal Regulations Part 50, Appendix B, Criterion XVI, Corrective Action, occurred when the licensee failed to correct conditions adverse to quality associated with the corrective actions identified in Condition Report RCR 2002-1665 to verify that installed surge suppressor varistors were appropriately sized and that design information was correctly reflected in controlled drawings for the reactor protection system, diesel generator control circuits, and high pressure coolant injection control circuits.
05000416/FIN-2017007-072017Q4Grand GulfLicensee-Identified ViolationThe following violations of very low safety significance (Green) were identified by the licensee and are violations of NRC requirements which meet the criteria of the NRC Enforcement Policy for being dispositioned as non- cited violations. Technical Specification 5.4.1(a) requires written procedures to be established, implemented, and maintained as recommended by Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Section 4.e recommends, in part, instructions for startup of shutdown cooling and reactor vessel head spray system be prepared. Contrary to the above, from about 2004 until September 1, 2017, the 04-1-01-E12-2 instruction failed to provide instruction for placing the alternate decay heat removal system in service. Specifically, Step 4.9.2a.7(d) instructs an operator to, Manually control component cooling water temperature by throttling P44-F010A(B)(C), PSW inlet to CCW HXs. However, the purpose of that step is to throttle plant service water flow through the alternate decay heat removal system and component cooling water system to ensure both systems have plant service water flow, which is not accomplished by the instruction step. The licensee identified this procedural violation before the system was credited for availability during an inservice demonstration on September 1, 2017, and entered it in the corrective action program as Condition Report CR-GGN-2017-08643. The violation is of very low safety significance (Green) because, although the procedure did delay placing the system in service due to the procedure error, the system was capable of performing its design function, consistent with Inspection Manual Chapter 0609, Appendix G, Attachment 1, Exhibit 3 screening.
05000416/FIN-2017007-082017Q4Grand GulfLicensee-Identified Violation10 CFR 50 Appendix B, Criterion III, requires in part, That measures shall be established to assure that the design bases are correctly translated into specifications, drawings procedures, and instructions. Contrary to the above, from original plant construction until June 22, 2016, GGNS failed to ensure the design basis tornado and differential pressures associated with it, would not cause a spurious trip of the Division I and II standby diesel generators. Specifically, a design basis tornado, includes a differential pressure of 3.0 psig, whereas an active diesel generator trip on high crankcase pressure actuates at 1.5 psig. The licensee identified this issue using an effective operating experience program and entered it in the corrective action program as Condition Report CR -GGN -2016- 04919. The violation is of very low safety significance (Green), for the same reason as NCV -05000416/2017007 -05, discussed in Section 1R21.4.5 of this report.
05000416/FIN-2017007-062017Q4Grand GulfFailure to Ensure Adequate Design Control Measures Are in Place Associated with Leakage Control SystemsThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis...for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, prior to October 17, 2017, the licensee failed to provide adequate procedures or training to licensed operators to ensure the main steam isolation valve-leakage control system and feedwater leakage control system are manually started consistent with the licensees design basis assumptions. In response to this issue the licensee has provided specific guidance and training to the operators. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2017-09112. The team determined that the failure to ensure adequate design control measures are translated into procedures and training is a performance deficiency. The performance deficiency is more than minor, and therefore a finding, because it was associated with the barrier 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 caused by accidents or events. Specifically, the plant was operated at power for an extended period of time without adequate procedures and training for licensed operators to ensure that the system would be placed in service in a manner that ensured radiological leakage across main steam isolation valves and through feedwater pi ping is addressed during a postulated accident. In accordance with Manual Chapter 0609, Significance Determination Process, Attachment 4 (effective date October 7, 2016); and the corresponding Appendix A, The Significance Determination Process (SDP) f or Findings At-Power, Exhibit 3, Barrier Integrity Screening Questions (issue date June 19, 2012); the issue was evaluated using Appendix H, Containment Integrity Significance Determination Process (issue date May 6, 2004). Because the opportunities to ensure the design control measures were correctly captured in procedures and instructions for the main steam isolation valve-leakage control system and feedwater leakage control system were in 2001 and 1987, respectively; and the licensee instituted a time-critical operator action program within the last year to prevent such issues from occurring, the issue was determined to have very low safety significance (Green). The performance deficiency was not indicative of current performance. Therefore, no cross-cutting aspect is being assigned.
05000416/FIN-2017007-052017Q4Grand GulfFailure to Update a Calculation and Procedure to Address Standby Service Water Passive FailureThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Specifically, since September 25, 2013, the licensee failed to include a design basis standby service water system (SSWS) piping crack in the appropriate design calculation and procedure. In response to this issue the licensee performed an operability determination to ensure that the ultimate heat sink basins would still have sufficient capacity to meet the 30 -day mission time. This finding was entered into the licensees corrective action program as Condition Report CR-GGN-2017-10192. The team determined that the failure to update a design calculation and a procedure to address a postulated standby service water passive failure was a performance deficiency. The finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000416/FIN-2017007-042017Q4Grand GulfFailure to Update the Final Safety Analys is ReportThe team identified three examples of a Severity Level IV, non-cited violation of 10 CFR 50.71, Maintenance of Records, Making of Reports, Section (e), which states, in part, Each person licensed to operate a nuclear power reactor shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. This submittal shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the applicant or licensee, or prepared by the applicant or licensee pursuant to Commission requirement since the submittal of the original or the last update to the final safety analysis report. Specific ally, prior to September 29, 2017, the licensee failed to ensure the final safety analysis report reflected the current plant configuration. In response to this issue, the licensee created a corrective action to update the final safety analysis report. The finding was entered into the licensees corrective action program as Condition Reports CR -GGN -2017- 09154, CR- HQN- 2017- 01356, and CR -GGN -2017 -09747. 5 The team determined that the failure to update the final safety analysis report in accordance with 10 CFR 50.71(e) was a performance deficiency. Following the Reactor Oversight Process (ROPs) significance determination process, the team determined this violation was associated with a minor performance deficiency. The ROPs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to also address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non- compliance. Assessing the performance deficiency in accordance with the NRC Enforcement Policy, the team determined it to be a Severity Level IV violation because the lack of up- to-date information in the final safety analysis report has not resulted in any unacceptable change to the facility or procedures. This finding did not have an assigned cross-cutting aspect because cross-cutting aspects are not assigned to traditional enforcement violations
05000416/FIN-2017007-032017Q4Grand GulfFailure to Establish a Preventive Maintenance Procedure for Safety -Related EquipmentThe team identified a Green, non -cited violation of Technical Specification (TS) 5.4.1, which states , in part , Written procedures shall be established, implemented, and maintained covering the following activities , referenced in Regulatory Guide (RG) 1.33, Revision 2, dated February 1978, Appendix A.9 , Procedures for Performing Maintenance, which requires that maintenance that can affect the performance of safety -related equipment should be properly pre- planned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstance. Specifically, prior to September 29, 2017, the licensee did not have a procedure to implement maintenance as recommended by the vendor in Vendor Document VM460000161, ELMA Cast Coil Power Transformers Installation, Maintenance, Operating, and Storage Instructions. In response to this issue, the licensee performed testing to ensure that the transformer will perform its design function and is developing an improved maintenance procedure. The finding was entered into the corrective action program as Condition Report CR- GGN -2017- 09390. The team determined that the failure to implement vendor recommended preventive maintenance is a performance deficiency. The finding was more than minor because it was associated with the equipment performance 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, the failure to have procedures that implement vendor recommended maintenance resulted in a question regarding the functionality of the transformer at elevated temperature. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At -Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non- technical specification equipment; and did not screen as potentially risk -significant due to seismic, flooding, or severe weather. This finding had a cross -cutting aspect in the area of human performance associated with conservative bias because the licensee failed to ensure that maintenance implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority (H.5).
05000416/FIN-2017007-022017Q4Grand GulfFailure to Correct Standby Diesel Generator TripThe team identified a Green, self -revealed, non -cited violation of 10 CFR 50 Appendix B, Criterion XVI, Corrective Actions, which states, in part, Conditions adverse to quality are promptly identified and corrected. On June 22, 2016, the station identified a condition adverse to quality affecting the standby diesel generators, but did not promptly correct the issue until September 22, 2017. Specifically, the actions described in Standing Order 17 -0011 were not appropriate for restoring full capability during a design basis tornado event, which could affect the capability of the Division I and II standby diesel generators. In response to this issue the licensee revised the standing order to have the operator press the diesel generator manual start button while the diesel is running to eliminate the associated non- safety trips. The finding was entered into the corrective action program as Condition Report CR -GGN -2017 -09751. The team determined that the failure to promptly correct a condition adverse to quality, regarding diesel generator capability during a design basis tornado is 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. Specifically, the failure to correct an identified condition adverse to quality resulted in a prolonged design challenge to the Division I and II standby diesel generator capability, which adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In accordance with Inspection Manual Chapter 0609, Appendix A, Exhibit 2, dated June 19, 2012, the team determined that the finding required a detailed risk evaluation, per Exhibit 4 screening question number 1, for external event mitigation systems. According to the tornado analysis database prepared by the Office of Reactor Research, the frequency of an F -2 tornado or stronger at Grand Gulf Nuclear Station (GGNS) is 4.71E -4/year. The design basis tornado at GGNS has a maximum wind velocity of 360 mph which correlates to a strong F- 5 tornado. The design basis tornado generates a differential pressure of 3 psi. The pressure of concern is the 1.5 psi that could affect operation of the Division I and II diesel generators. Given that pressure is proportional to the square of the velocity, the wind speed affecting the diesel generators would be approximately 250 mph. 250 mph is in the range of an F- 4 tornado. Using generic distributions of the frequency of varying tornado strengths, the analyst estimated that the frequency of an F- 4 tornado or stronger at GGNS is 3.93E -6/year. Using the site- specific SPAR model, the analyst quantified the conditional core damage probability for a tornado- induced loss of offsite power with the failure of both Division I and II diesel generators. The conditional core damage probability was 6.35E -2. Therefore, the incremental conditional core damage probability of the performance deficiency, using the bounding assumption that all F -4 or stronger tornados striking the site would fail both diesel generators, was 2.50E -7. Qualitatively, given this bounding assumption, and the potential to recover the diesels after failure, the analyst determined that the CDF was less than 1E -7. This results in a finding of very low safety significance (Green). This finding had a cross -cutting aspect in the area of human performance associated with procedure adherence because the licensee failed to follow the operability evaluation process to properly determine operability (H.8).
05000313/FIN-2017007-012017Q3Arkansas NuclearFailure to Promptly Identify and Correct an Inadequate Design Bases CalculationThe inspectors identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, requires, in part, that measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformance are promptly identified and corrected. Specifically, from 1996 until August 10, 2017, the licensee failed to properly resolve the environmental conditions in room 38 following a high-energy line break, even when challenged during a self-assessment by members of the quality assurance group in June 29, 2015. In response to this issue, the licensee determined that in the event of a break in the letdown line, an engineered safety feature signal automatic closure of both the inside and outside reactor building isolation valves occurs in approximately 40 seconds, preventing room 38 from going harsh. This finding was entered into the licensees corrective action program as Condition Report CR-ANO-1-2017-02441. The inspectors determined that the licensees failure to adequately evaluate and take prompt corrective actions to resolve an identified condition adverse to quality related to the high energy line break analysis for room 38 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 affected the associated objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, when the licensee identified that the environmental conditions in room 38 of the auxiliary building were harsh, as determined by Design Bases Calculation CALC-01-EQ-1002-02, they failed to properly resolve the condition adverse to quality. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, dated July 1, 2012, the inspectors determined that 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 function of at least a single train for longer than its allowed outage time, or two separate safety systems out-of-service for longer than their technical specification allowed outage time; and (4) does not represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours in accordance with the licensees maintenance rule program. The finding has a cross-cutting aspect in the area of human resources, training, because the organization failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, there was a lack of understanding of the current licensing bases for the plant displayed by engineering, operations, and management (H.9).
05000482/FIN-2017007-022017Q2Wolf CreekLicensee-Identified ViolationTechnical Specification 3.3.3, Post Accident Monitoring Instrumentation, required two channels of containment area radiation (high range) detectors to be operable when the unit is in Modes 1, 2, or 3. It also required, for one or more functions with two required channels inoperable, that one of the required channels be restored to Operable within 7 days or initiate action in accordance with Technical Specification 5.6.6. Specification 5.6.6 required that a Post Accident Monitoring Instrumentation Report be submitted within 14 days that outlined the preplanned alternate method of monitoring, the cause of inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to Operable status. Contrary to the above, from 1997 to March 2017, the licensee failed to restore at least one channel of containment high range radiation monitors to operable status, initiate preplanned alternate methods of monitoring the appropriate parameter, or prepare and submit a Post Accident Monitoring Instrumentation Report within 14 days pursuant to Technical Specification 5.6.6. The violation was more than minor because it was associated with the Facilities and Equipment attribute of the Emergency Preparedness Cornerstone and adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors determined the significance using Inspection Manual Chapter 0609, Attachment 04, Initial Characterization of Findings, and Inspection Manual Chapter0609, Appendix B, Emergency Preparedness Significance Determination Process, Section 5.4 for failure to comply with risk significant planning standard 10 CFR 50.47(b)(4). The finding was determined to be of very low safety significance (Green) because (1) emergency action level schemes were rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event, but because of other emergency action levels, an appropriate declaration could be made in a degraded manner; and, (2) the emergency action level classification process would result in an over-classification causing an unnecessary emergency declaration. The violation was entered into the licensees corrective action program as Condition Reports CR-111440, CR-111536, and CR-113217.
05000482/FIN-2017007-012017Q2Wolf CreekFailure to Maintain Effectiveness of the Emergency Plan upon Loss of Containment High Radiation MonitoringThe inspectors identified a Green non-cited violation of 10 CFR 50.54(q)(2) which requires that a holder of a nuclear power plant operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in Appendix E of this part and the risk significant planning standards of 10 CFR 50.47(b). Specifically, from March 7, 2017, to July 12, 2017, Wolf Creek Generating Stations response to the inoperability of containment high radiation monitors failed to restore capability to classify emergency action levels during a loss-of-coolant accident or main-steam-line-break accident. In response to this issue, the licensee provided additional radiation survey monitoring measures and correlations to monitor radiation in the containment building. This finding was entered into the licensees corrective action program as Condition Report CR-114274. The inspectors determined that the failure to maintain the effectiveness of the emergency action level schemes by providing adequate preplanned methods and compensatory measures for the loss of the containment high range radiation monitors in accordance with 50.54 (q)(2) was a performance deficiency. This finding was determined to be more thanminor because it was associated with emergency response organization performance attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective. Specifically, the failure to maintain the effectiveness using appropriate compensatory measures adversely affected the objective of ensuring the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding was determined to be of very low safety significance (Green) because (1) emergency action level schemes were rendered ineffective such that any Site Area Emergency would not be declared for a particular off-normal event, but because of other emergency action levels, an appropriate declaration could be made in a degraded manner; and, (2) the emergency action level classification process would result in an over-classification causing an unnecessary emergency declaration. This finding had a cross-cutting aspect in the area of human performance associated with conservative bias because the licensee failed use decision making-practices that emphasized prudent choices over those that are simply allowable. (H.14)(Section 1R21N)
05000528/FIN-2017007-012017Q1Palo VerdeFailure to Analyze Shutdown Cooling and Feedwater Lines for High-Energy Line Break Pipe Whip EffectsGreen. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, Measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2 and as specified in the license application, for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures, and instructions. Specifically, from August 11, 1982, to March 3, 2017, the licensee did not analyze dynamic pipe whip effects of a main feedwater line for a high-energy line break of a shutdown cooling line. In response to this issue, the licensee performed immediate and prompt operability evaluations and determined that the piping systems remained operable and could withstand the effects of a high-energy line break. This finding was entered into the licensees corrective action program as Condition Report CR-17-02815. The team determined that the failure to perform an adequate analysis for shutdown cooling and feedwater lines for high-energy line break pipe whip effects was a performance deficiency. This finding was more-than-minor because it was associated with the design control attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to analyze the main feedwater piping for high-energy line break effects called the operability of the piping system into question. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as hav ing very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross- cutting aspect because the most significant contributor to the performance deficiency did 3 not reflect current licensee performance. Specifically, the licensee performed the calculation in 1982 and revised it in 1991; therefore, the performance deficiency occurred outside of the nominal three-year period for present performance.
05000416/FIN-2016007-022016Q4Grand GulfFailure to Obtain NRC Approval For Changes to the Reactor Protection SystemThe team identified a Severity Level IV non-cited violation of 10 CFR 50.59(c)(2), Changes, Tests, and Experiments, for the licensees failure to obtain a license amendment prior to implementing a proposed change, test, or experiment that would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report. Specifically, from June 24, 2014, until November 3, 2016, the licensee modified its reactor protection system to remove turbine first stage pressure instrumentation to measure reactor power, which resulted in a more than minimal increase of the likelihood of a malfunction. The failure to obtain a license amendment prior to implementing a change that resulted in a more than a minimal increase in the likelihood of occurrence of a malfunction of a system important to safety was a performance deficiency. In response to this issue, the licensee implemented compensatory actions to ensure the reactor protection system trips would be enabled when required, will either prepare a new evaluation under current regulatory guidelines, or submit a license amendment request to the NRC, and documented the condition in its corrective action program as Condition Report CR-GGN-2016-08298. This performance deficiency was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the elimination of the turbine first stage pressure instruments increased the likelihood of a malfunction of the reactor protection system. Additionally, the violation was similar to the more-than-minor examples in the NRC Enforcement Manual Appendix E, Minor Violations Examples, dated September 9, 2013. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Since the violation was determined to be Green in the significance determination process, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000416/FIN-2016007-032016Q4Grand GulfFailure to Obtain NRC Approval For Changes to Diesel Generator Trips and Flood Mitigation StrategyThe team identified two examples of a Severity Level IV non-cited violation of 10 CFR 50.59(c)(2), Changes, Tests, and Experiments, for the licensees failure to conclude that modifications to the Division 3 diesel generator trip logic circuits and flood mitigation strategy would have required a license amendment. Specifically, from October 7 to November 3, 2016, the licensee removed the automatic high crankcase diesel generator trip and from March 5, 2013, to November 3, 2016, used an unapproved method for mitigating design basis flooding. The licensees failure to obtain a license amendment prior to implementing a change that resulted in a more than a minimal increase in the likelihood of occurrence of a malfunction of a system important to safety was a performance deficiency. In response to these issues, the licensee entered the issues into the corrective action program as Condition Reports CR-GGN-2016-08328 and CR-GGN-2016-08329 and will either prepare new evaluations under current regulatory guidelines, or submit a license amendment request to the NRC. The first example of a performance deficiency for the change to the Division 3 diesel generator trip logic was more-than-minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the elimination of the diesel generator automatic trips increased the likelihood of a malfunction of systems important to safety. The second example of a performance deficiency for a change to the flood mitigation strategy to rely on the construction of temporary sandbag barriers was more-than-minor because it was associated with the protection against external hazards attribute of the Mitigating Systems Cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the violation was similar to the more-than-minor example of a change in requirements in the NRC Enforcement Manual Appendix E, Minor Violations Examples, dated September 9, 2013. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Since the violation was determined to be Green in the significance determination process, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. Traditional enforcement violations are not assessed for crosscutting aspects.
05000416/FIN-2016007-042016Q4Grand GulfFailure to Evaluate Delaying Inspection of Diesel Fuel Oil Storage TankThe team identified a Severity Level IV non-cited violation of 10 CFR 50.59(d)(1), Changes, Tests, and Experiments, for the licensees failure to provide a written evaluation describing the basis for determining that a change to how often the Division 3 diesel fuel oil storage tank is cleaned and inspected did not require a license amendment. The failure to perform an evaluation prior to implementing a change that resulted in a more than a minimal increase in the likelihood of occurrence of a malfunction of a system important to safety as required by 10 CFR 50.59 was a performance deficiency. In response to this issue, the licensee declared the Division 3 diesel generator inoperable until it performed the cleaning and inspections required by Regulatory Guide 1.137. After the inspection was successfully completed without issues, the licensee declared the Division 3 diesel generator to operable. This issue was entered the issue into the corrective action program as Condition Report CR-GGN-2016-08327. This performance deficiency was more-than-minor because if left uncorrected, the issue would the performance deficiency have the potential to lead to a more significant safety concern. Specifically, the failure to clean and inspect the Division 3 fuel oil storage tank could result in the failure of the Division 3 diesel system. Additionally, the violation was similar to the more-than-minor example of changes to requirements in the NRC Enforcement Manual Appendix E, Minor Violations Examples, dated September 9, 2013. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. Since the violation was determined to be Green in the significance determination process, the traditional enforcement violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d(2) of the NRC Enforcement Policy. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000416/FIN-2016007-052016Q4Grand GulfFailure to Establish Adequate Procedures for Building Sandbag BarriersThe team identified a Green, non-cited violation of Technical Specification 5.4.1(a), Procedures, for failure to establish adequate procedures for severe weather operations. Specifically, the licensee failed to establish adequate severe weather procedures to ensure the control building, diesel building, and standby service water pump houses would be adequately protected from flooding. The failure to establish adequate procedures for severe weather operations to ensure compliance with Technical Specification 5.4.1(a), Procedures, and with the Regulatory Guide 1.33, Appendix A, Section 6.w, Acts of Nature, was a performance deficiency. In response to this issue, the licensee calculated the maximum allowable leakage of the sandbag barriers that would adequately protect any structure, system, or components important to safety from flooding. Additionally, the licensee performed a mock-up of the sandbag barriers and determined that the expected leakage through the sandbag barriers during a probable maximum precipitation event would be less than the maximum leakage allowed by the calculation. This finding was entered into the licensees corrective action program as Condition Reports CR-GGN-2016-08294 and CR-GGN-1-2016-08912. This performance deficiency 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 licensee failed to establish adequate procedures to ensure the sandbag barriers offer adequate flood protection during a probable maximum precipitation event that no structures, systems, or components important to safety are affected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. The team determined the finding had a crosscutting aspect of Avoiding Complacency within the area of Human Performance because the licensee failed to recognize and plan for the possibility of mistakes, latent issues, and inherent risk in building the sandbag barriers, even while expecting successful outcomes (H.12).
05000313/FIN-2016008-022016Q4Arkansas NuclearFailure to Incorporate NRC Safety Guide 9 Criteria into Surveillance ProceduresGreen. The team identified Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, A test program shall be established to assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Additionally, Test results shall be documented and evaluated to assure that test requirements have been satisfied. Specifically, as of December 2, 2016, Units 1 and 2 emergency diesel generator surveillance procedures failed to incorporate the applicable voltage and frequency limits of NRC Safety Guide 9, and did not consistently document or evaluate results to assure test requirements have been satisfied. In response to this issue, the licensee provided the team test results which demonstrated that an immediate safety concern was not present. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-1-2016-4785 and CR-ANO-2-2016-4257. The team determined that the failure to incorporate the acceptance limits of NRC Safety Guide 9 into surveillance test procedures for emergency diesel generators and assure that test requirements have been satisfied in accordance with 10 CFR Part 50, Appendix B, Criterion XI, Test Control, was a performance deficiency. The performance deficiency was determined to be more than minor because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences, and would have the potential to lead to a more significant safety concern. Specifically, the failure to incorporate appropriate acceptance criteria in test procedures and assure that the criteria have been satisfied had the potential to lead to a worse condition, if left uncorrected. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016008-032016Q4Arkansas NuclearFailure to Monitor Startup Transformers 1, 2, and 3 Voltage Regulator/Tap Changer FunctionGreen. The team identified a Green finding for the failure to meet the surveillance standards of IEEE 308-1971, Criteria for Class 1E Electric Systems for Nuclear Power Generating Stations, Section 5.2.3, Preferred Power Supply. Specifically, from 2001 to December 2, 2016, the licensee failed to monitor the operation of the voltage regulator/load tap changer functions on startup transformers 1, 2, and 3. In response to this issue, the licensee provided reasonable assurance that the voltage regulator/load tap changer was operating properly based on review of plant computer voltage plot data following an Arkansas Nuclear One, Unit 1 trip that occurred on December 14, 2015. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-4777, CR-ANO-C-2016-4879, and CR-ANO-C-2016-5015. The team determined that the failure to monitor startup transformers 1, 2, and 3 voltage regulator/load tap changers to the extent that they are shown to be ready to perform their intended function, in accordance with IEEE Standard 308-1971, was a performance deficiency. The finding was determined to be more than minor because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to monitor the adequacy of the voltage supplied from startup transformers 1, 2, and 3 voltage regulator/load tap changer did not ensure that offsite power would be available to perform its necessary functions to provide power to the safety-related mitigation equipment. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016008-042016Q4Arkansas NuclearFailure to Perform an Adequate Emergency Feedwater Pump Suction Transfer Design Calculation or Testing (EA 2017-017)Green. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part that, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to December 22, 2016, the licensee failed to verify the adequacy of the emergency feedwater suction transfer procedure by determining if the qualified condensate storage tank will be completely empty of water, possibly causing an air ingestion failure of the Unit 1 emergency feedwater pumps, prior to transferring to the credited safety-related alternate suction source. In response to this issue, the licensee resolved the immediate safety concern by revising the emergency feedwater pump operating procedure, removing the steps that were the cause of the concern. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-1-2016-5166, CR-ANO-1-2016-5725, and CR-ANO-1-2017-0040. The team determined that the failure to verify the adequacy of the design of the Unit 1 emergency feedwater suction from the qualified condensate storage tank to alternate sources of water by performance of design review, by use of calculational methods, or by performance of a suitable testing program in accordance with 10 CFR Part 50, Appendix B, Criterion III, Design Control, was a performance deficiency. This finding was more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the reliability, availability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee failed to have adequate measures in place to ensure an acceptable design analysis or a suitable test program would verify that the process of transferring emergency feedwater suction from the qualified storage tank to the alternate sources ensures the capability of the Unit 1 emergency feedwater system to perform its safety function. In accordance with Inspection Manual Chapter 0609, Attachment 4, Initial Characterization of Findings, dated June 19, 2012, the team determined this finding affected the secondary short term heat removal function of the Mitigating Systems Cornerstone. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the finding represented a loss of the emergency feedwater system and function. Therefore, a detailed risk evaluation was necessary. The senior reactor analyst determined that the change in core damage frequency of this finding was 7 x 10-7 per year, therefore the significance was of very low safety significance (Green). This finding did not have a cross-cutting aspect because the performance deficiency did not reflect current licensee performance.
05000313/FIN-2016008-052016Q4Arkansas NuclearFailure to Ensure Safety Systems Would Survive Sustained Degraded Voltage ConditionsGreen. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, from December 17, 1979, to December 2, 2016, the licensee did not verify that the design of the protective devices for the loads required at the beginning of a loss-of-coolant accident were adequate to prevent tripping these devices under degraded voltage conditions, which would render the affected loads non-functional. In response to this issue, the licensee performed a preliminary analysis to determine that the protective overload devices would not cause safety equipment to fail at degraded voltages allowed by technical specifications. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-5027 and CR-ANO-C-2016-5191. The team determined that the failure to ensure that safety-related electrical components would not fail during the allowable time duration of a degraded voltage condition (in accordance with NRC Multi-Plant Action B-23, Position 1.C) was a performance deficiency. The finding was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure that the protective devices for the loads required at the beginning of a Loss of Control Accident would not fail under degraded voltage conditions did not ensure that these loads would be available to perform their mitigating functions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000416/FIN-2016007-012016Q4Grand GulfInadequate Technical Specification Surveillance Requirements for Reactor Protection SystemThe team identified a Green non-cited violation of 10 CFR 50.36, Technical Specifications, which requires that surveillance requirements are requirements relating to test, calibration, or inspection to assure that the necessary quality of systems and components is maintained, that facility operation will be within safety limits, and that the limiting conditions for operation will be met. Contrary to this requirement, from June 24, 2014, until November 3, 2016, the licensee failed to include in its technical specification a surveillance requirement to assure that the facility operation will be within safety limits. Specifically, after modifying its reactor protection system to remove turbine first stage pressure instrumentation, the licensee failed to adjust the interval at which it calibrates the average power range monitor channels during surveillance tests to ensure the signals were accurately indicating the true core average power and that reactor protection system trips were enabled when required to assure the facility will be within safety limits. The licensees failure to ensure surveillance requirements relating to calibration to ... assure that ... facility operation will be within safety limits, and that the limiting conditions for operation will be met was a performance deficiency. In response to this issue, the licensee implemented compensatory actions to ensure the reactor protection system trips would be enabled when required, and documented the condition in its corrective action program as Condition Report CR-GGN-2016-08297. This performance deficiency was more-than-minor because it was associated with the thermal limit design control attribute of the Barrier Integrity Cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the surveillance requirements did not assure calibration of the average power range monitors to ensure an accurate measurement of reactor power such that the reactor protection system trips were enabled at 35.4 percent power. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with change management because the licensee failed to use a systematic process for evaluating and implementing changes to the reactor protection system so that nuclear safety remains the overriding priority (H.3).
05000313/FIN-2016008-012016Q4Arkansas NuclearFailure to Verify the Adequacy of Motor Operated Valve Thermal Overload DevicesGreen. The team identified a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to December 2, 2016, the licensee failed to use appropriate assumptions in thermal overload device calculations and failed to establish a suitable periodic test program for safety-related Unit 1 motor operated valve thermal overload device trip setpoints, as discussed in Regulatory Guide 1.106, Regulatory Position C.2. In response to this issue, the licensee demonstrated reasonable assurance of operability by using the results of the 18-month high pressure injection system valve testing which required multiple stroking of block valves to obtain various flows without tripping the thermal overload devices. This finding was entered into the licensees corrective action program as Condition Reports CR-ANO-C-2016-5017 and CR-ANO-1-2016-5130. The team determined that the failure to meet the intent of Regulatory Guide 1.106, Regulatory Position C.2 was a performance deficiency. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify the adequacy of the design and perform suitable testing for thermal overload device setpoint drift did not ensure that the safety-related motor operated valves would be available to throttle the associated system flows during a design basis accident. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluations because the licensee failed to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to thoroughly evaluate Condition Report CR-ANO-1-2016-0778 which documented NRC inspector concerns associated with design and testing of motor operated valve thermal overload devices (P.2).
05000313/FIN-2016008-062016Q4Arkansas NuclearReadiness to Cope with External FloodingGreen. The team identified three examples of a Green non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part that, activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances. Specifically, prior to December 2, 2016, Unit 1 Operating Procedure OP 1203.025, Natural Emergencies, Revision 60 and Unit 2 Operating Procedure OP 2203.008 Natural Emergencies, Revision 42 failed to ensure all actions required to establish external flood protection, as specified by flood protection design basis engineering report CALC-ANOC-CS-00003, Revision 00 were implemented. This issue was entered into the licensees corrective action program as Condition Report CR-ANO-2-2016-4265. The licensees failure to prescribe procedures appropriate to the circumstances for combating emergencies or other significant acts of nature such as flooding 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 of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences, and would have the potential to lead to a more significant safety concern. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it does not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding, or severe weather initiating event. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with identification because the licensee failed to identify issues, completely, accurately, and in a timely manner in accordance with the corrective action program. Specifically, the licensee failed to identify these deficiencies during a review of these same procedures as part of actions to close significant performance deficiencies as documented in Arkansas Nuclear One Area Action Plan FP-6 (P.1).
05000285/FIN-2016008-012016Q3Fort CalhounFailure to Obtain Prior NRC Approval for a Change When RequiredThe inspectors identified a Severity Level IV, Green, non-cited violation of 10 CFR 50.59, Changes, Tests, and Experiments, Section (c)(1), which states, in part, that a licensee may make changes in the facility as described in the final safety analysis report (as updated) without obtaining a license amendment pursuant to 10 CFR 50.90 only if: (i) a change to the technical specifications incorporated in the license is not required, and (ii) the change, test, or experiment does not meet any of the criteria in paragraph (c)(2). Title 10 CFR 50.59, Section (c)(2), states, in part, that a licensee shall obtain a license amendment pursuant to Section 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report (as updated). Specifically, from June 9, 2015 through August 11, 2016, the licensee implemented a change to Operating Instruction OI-VA-2, Auxiliary Building Ventilation System Normal Operation, Attachment 11, Revision 47, after incorrectly concluding that the opening of certain high-energy line break barriers and selected fire barrier doors to allow supplemental cooling of both safety-related switchgear rooms did not increase the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the final safety analysis report. In response to this issue, the licensee revised entry conditions to Operating Instruction OI-VA-2, Attachment II, to ensure that high energy line break barriers are not impaired prematurely. This finding was entered into the licensees corrective action program as Condition Report CR-2016-06667. The licensees failure to implement the requirements of 10 CFR 50.59 and adequately evaluate the disabling of certain high-energy line break barriers to facilitate supplemental cooling of both safety-related switchgear rooms was a performance deficiency. This finding was evaluated using traditional enforcement because it had the potential for impacting the NRCs ability to perform its regulatory function. In accordance with Section 2.1.3.E.6 of the NRC Enforcement Manual, the inspectors used Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process For Findings At-Power, dated June 19, 2012, to determine that this performance deficiency was of 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-ofservice for longer than its 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 significance 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. As described in Inspection Manual Chapter 0612, Power Reactor Inspection Reports, no cross-cutting aspect was assigned to this violation because traditional enforcement violations are not assessed for cross-cutting aspects.
05000445/FIN-2016007-012016Q3Comanche PeakFailure to Update Final Safety Analysis Report Section 8.3.1.1.11The inspectors identified a Severity Level IV violation of 10 CFR 50.71(e) which states, in part, that the licensee shall update periodically the final safety analysis report originally submitted as part of the application for the license, to assure that the information included in the report contains the latest information developed. The submittal shall include the effects of all changes made in the facility or procedures as described in the final safety analysis report, or all safety analyses and evaluation performed by the licensee either in support of approved license amendments or in support of conclusions that changes did not require a license amendment in accordance with 10 CFR 50.59 (c)(2). Specifically, from October 9, 2012, to September 29, 2016, the licensee did not include the effects of changes to the K300 voltage relay setpoint or the safety evaluation in submittals to the Final Safety Analysis Report, Section 8.3.1.1.11, that supported the conclusion that the changes did not require a license amendment. In response to this issue, the licensee planned a corrective action to initiate a licensing document change request to update the final safety analysis report. This finding was entered into the licensees corrective action program as Condition Report CR-2016-008177. The inspectors determined that the licensees failure to initiate a Licensing Document Change Request, in accordance with Procedure STA-116, Maintenance of CPNPP Licensing Basis Documents, Operating License conditions and Technical Specifications, Revision 14, Instruction 6.1, to update the Final Safety Analysis Report, Section 8.3.1.1.11, for the setpoint revision of the K300 voltage relays was a performance deficiency. In accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening, dated September 7, 2012, this was determined to be a minor performance deficiency. This violation was evaluated using the traditional enforcement process because it had the potential for impacting the NRCs ability to perform its regulatory oversight function. The reactor oversight processs significance determination process does not consider violations that impact the NRCs regulatory oversight function. This violation was determined to be a Severity Level IV violation, consistent with the example in paragraph 6.1.d.3 of the NRC Enforcement Policy, dated August 1, 2016. Specifically, the licensee failed to update the final safety analysis report as required by 10 CFR 50.71(e), but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. The inspectors determined that this violation did not have a cross-cutting aspect because traditional enforcement violations are not assessed for cross-cutting aspects.
05000482/FIN-2016007-042016Q2Wolf CreekFailure to Promptly Correct Deficiencies With Operator Time Critical ActionsThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, prior to April 28, 2016, the licensee failed to correct deficiencies identified in 2012 for operator time critical actions associated with control room habitability; in 2013 after revising training materials for control room habitability time critical actions; in a 2014 condition report documenting the failure to validate scenarios in the time critical action program; and again in a 2015 self-assessment of the time critical action program. During the inspection, five out of six operators in a test crew failed to complete the control room habitability scenario within the required two minutes. In response to this finding, the licensee performed just-in-time training to remediate the crews and ensure time critical actions can be met. After re-training, each crew successfully performed the control room habitability time critical action within the two-minute requirement. This finding was entered into the licensee's corrective action program as Condition Reports 103910, 103915, and 103658. The team determined the failure to correct the deficiencies with the control room habitability time critical action was a performance deficiency. The performance deficiency was morethan- minor, and therefore a finding, because it related to the human performance 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, operators failed to meet the time critical action for the control room habitability scenario within the required two minutes. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with training because the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values (H.9).
05000482/FIN-2016007-032016Q2Wolf CreekInadequate Analysis of Essential Service Water PipingThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to April 28, 2016, the licensee failed to verify the design of the essential service water piping because the analyses assumed that the essential service water piping upstream of the containment air coolers was full of water after a loss of offsite power. However, the essential service water pump check valve was never tested to ensure water would not drain from the essential service water piping. In response to this issue, the licensee conducted a preliminary evaluation using data from the last surveillance test and inspection of the check valve, and concluded that the worst-case expected leakage through the check valve was not large enough to cause a water hammer event in the piping that exceeded operability criteria. This finding was entered into the licensee's corrective action program as Condition Reports 104222 and 104184. The team determined that the failure to verify the adequacy of the design of the essential service water piping was a performance deficiency. The performance deficiency was morethan- minor, and therefore a finding, because it related to the design 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, the failure to account for check valve leakage in the essential service water system led to a non-conservative assumption that the piping upstream of the containment air coolers would not drain after a loss of offsite power, which contributes to water hammer events that could challenge the integrity of the piping. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding was assigned a cross cutting aspect in the area of problem identification and resolution, specifically operating experience, because the water hammer issue was previously documented in several NRC inspection reports, the licensee made recent modifications to the system, and a companion check valve in the normal service water system was installed and correctly categorized in the inservice testing basis document. The operating experience cross-cutting aspect requires that the licensee systematically and effectively collects, evaluates, and implements relevant internal and external operating experience in a timely manner (P.5).
05000482/FIN-2016007-022016Q2Wolf CreekFailure to Verify the Adequacy of Design of the Control Circuitry of the Fuel Oil Transfer PumpsThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to April 28, 2016, the licensee failed to verify the adequacy of the design of the fuel oil transfer pump control circuitry to ensure that the thermal overloads associated with the fuel oil transfer pump would not activate, trip the pump, and render the emergency diesel generator inoperable in the case of excessive cycling. In response to this issue, the licensee conducted a preliminary evaluation of the fuel oil transfer pump and confirmed there is not any significant active leakage on the day tank which would lead to excessive cycling, and that starting currents are sufficiently below the thermal overload trip settings and are unlikely to trip the pump. Additionally, the licensee planned to initiate a program to determine fuel oil leakage from the day tank and require operators to initiate interim corrective actions until final corrective actions can be determined. This finding was entered into the licensee's corrective action program as Condition Report 104066. The team determined the failure to evaluate the effects of cyclical fuel oil transfer pump operation was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design 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, the design of the fuel oil transfer pump control circuity does not prevent activation of the pump thermal overloads that would trip the pump and render the emergency diesel generator inoperable in the event of cyclical operation of the fuel oil transfer pump. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000482/FIN-2016007-012016Q2Wolf CreekInadequate Degraded Voltage Analyses of Class 1E SystemsThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to April 28, 2016, the licensee failed to verify the adequacy of the design of the Class 1E electrical equipment, because it failed to perform adequate analyses demonstrating 1) that the degraded voltage relay setpoints specified in technical specifications would ensure adequate voltage to safety-related equipment, 2) adequate voltage would be available to the safety-related loads during transient voltage conditions caused by load sequencing, and 3) that the degraded voltage relay-associated time delays provide timely separation from offsite power and transfer to the emergency diesel generator to ensure that the Class 1E safety-related loads can achieve their safety function without protective device tripping. In response to these issues, the licensee performed preliminary analyses to demonstrate that the Class 1E electrical equipment would function at degraded voltages and was operable. This finding was entered into the corrective action program as Condition Reports 47791, 104253, 104098, 104389, and 104390. The team determined the licensees failure to ensure the adequacy of the design of the Class 1E electrical equipment was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design 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, the licensees electrical analyses failed to verify degraded voltage relay setpoints specified in technical specifications would ensure adequate voltage to safety-related equipment, that adequate voltage would be available to the safety-related loads during transient voltage conditions caused by load sequencing, and that degraded voltage relay time delays would provide timely separation from offsite power and transfer to the emergency diesel generator to ensure that the Class 1E safety-related loads can achieve their safety function without protective device tripping. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not to result in loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding did not have a cross-cutting aspect because the most significant contributor did not reflect current licensee performance.
05000275/FIN-2016007-032016Q1Diablo CanyonFailure to Ensure Safety-Related Alternating Current and Direct Current Equipment Functionality at Maximum Allowable VoltagesThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to February 10, 2016, the licensee failed to verify the design of (1) equipment on the nominally 125 Vdc system at the maximum voltage specified in Procedure OP J-9:IV, Performing a Battery Equalizing Charge, and (2) equipment on 480 Vac and 120 Vac vital buses at maximum voltages specified in Procedure OP J-2:VIII, Guidelines for Reliable Transmission Service for DCPP, by the use of alternate or simplified calculational methods, to ensure equipment functionality. In response to this finding, the licensee conducted a preliminary evaluation of the affected equipment and concluded that any past exposure to voltages above their maximum rating would not have caused a loss of functionality. This finding was entered into the licensee's corrective action program as Notifications 50834558, 50835906, 50835394, 50835945, 50835949, 50836376, 50836439, 50836638, 50836872, and 50836995. The team determined the failure to evaluate operation of 125 Vdc and 480 and 120 Vac equipment at maximum allowable voltages was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the equipment performance 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, operation of equipment outside of its rated or analyzed maximum allowable voltages adversely affects the reliability and capability of that equipment required to perform safety-related functions. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with design margins because the licensee failed to ensure that the organization operated and maintained equipment within design margins and that margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000275/FIN-2016007-042016Q1Diablo CanyonFailure to Evaluate the Extent of Condition for a Degraded Condition on a Nonsafety-Related 4160 Vac BreakerThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Specifically, in October of 2015, the licensee failed to evaluate the extent of condition of a cracked holding pawl on a nonsafety-related 4160 Vac SF6 breaker, which was procured as safety-related, in accordance with Procedure OM7.ID1, Problem Identification and Resolution, when the failure of the component could adversely impact safety-related breakers of the same make and model. In response to this finding, the licensee is performing a procedure review to include steps to perform an extent of condition analysis for unplanned nonsafety-related equipment issues that may also affect similar safety-related equipment. This finding was entered into the licensee's corrective action program as Notifications 50836859 and 50836689. The team determined the failure to evaluate the impact of a cracked holding pawl identified on a nonsafety-related 4160 Vac SF6 breaker on additional safety-related 4160 Vac SF6 breakers was a performance deficiency. The performance deficiency was more-thanminor, and therefore a finding, because it related to the equipment performance 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, the 4160 Vac breaker with the cracked holding pawl was procured as safetyrelated; therefore, the condition extends to safety-related 4160 Vac breakers of the same make and model and potentially adversely affects the ability to perform their safety function. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with conservative bias because the licensee failed to ensure that individuals used decision-making practices that emphasized prudent choices (H.14).
05000498/FIN-2016007-032016Q1South TexasFailure to Include Applicable Safety System Criteria in the Final Safety Analysis ReportThe team identified a Severity Level IV, non-cited violation of 10 CFR 50.34(b)(2), Final Safety Analysis Report which requires, in part, that the final safety analysis report shall include a description and analysis of the structures, systems, and components of the facility, with emphasis upon performance requirements, the bases, with technical justification therefor, upon which such requirements have been established, and the evaluations required to show that safety functions will be accomplished. The description shall be sufficient to permit understanding of the system designs and their relationship to safety evaluations. Specifically, since March 22, 1988, the licensee failed to include, in the final safety analysis report, the safety system criteria specified by IEEE 603-1980 and IEEE 7.4-3-2 for the Eagle 21 control system, which described the facility, presented the design bases, and the limits on its operation. This violation does not represent an immediate safety concern. In response to this issue, the licensee created corrective actions to determine the appropriate information to include in the next update to the updated final safety analysis report. This violation was entered into the licensees corrective action program as Condition Report CR 16-1281. The team determined that the failure to revise the final safety analysis report with the supplemental information that presented the design bases of the qualified display processing system was a violation of 10 CFR 50.34(b)(2). The violation was more than minor because the design basis information affected certain safety system functions (i.e., the auxiliary feedwater system control valves), which had a material impact on safety. Because the issue affected the NRCs ability to perform its regulatory function, the inspectors evaluated this violation using the traditional enforcement process. The inspectors used the NRC Enforcement Policy, Subsection 6.1, Reactor Operations, dated February 4, 2015, to evaluate the significance of this violation. This violation is similar to example 6.1.d.3 in the Enforcement Policy. Therefore, this was a Severity Level IV violation because the violation represented a failure to update the final safety analysis report as required by 10 CFR 50.34(b)(2), but the lack of up-to-date information has not resulted in any unacceptable change to the facility or procedures. The team determined there was no cross-cutting aspect because cross-cutting aspects are not assigned to traditional enforcement violations.
05000498/FIN-2016007-092016Q1South TexasFailure to Ensure Adequate Design Control Measures in Place to Mitigate a Loss of Normal Feedwater Flow EventThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, that Measures shall be established to assure that applicable regulatory requirements and the design basis...for those structures, systems, and components to which this appendix applies are correctly translated into specifications, drawings, procedures and instructions. Specifically, since August 1, 2001, the licensee failed to translate into procedures that a loss of normal feedwater flow event would be mitigated consistent with the licensees design basis assumptions. In response to this issue, the licensee initiated actions to establish interim emergency operating procedure directions for the licensed operators to ensure that credited safety-related equipment is used with priority in the event if this were to occur at the plant. The emergency operating procedure is being revised to ensure permanent corrective action is taken. This finding was entered into the licensee's corrective action program as Condition Report CR 16-1694. The team determined that the failure to establish measures to assure that the design bases was correctly translated into procedures and instructions was a performance deficiency. The performance deficiency was determined to be more than minor, and therefore a finding, because it was associated with the Mitigating Systems cornerstone attribute of procedure quality, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. In addition, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if the licensee used the procedure to mitigate a loss of normal feedwater flow event, the licensee may place the plant in an unanalyzed condition. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality. The team determined that this finding did not have a cross-cutting aspect because the most significant contributor did not reflect present licensee performance.
05000498/FIN-2016007-072016Q1South TexasFailure to Implement Administrative Controls for a Nonconservative Technical Specification of Standby Diesel Generator Frequency VariationThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, since 1997, the licensee failed to correct a condition adverse to quality by imposing administrative controls in response to a nonconservative Technical Specification. In response to this issue, the licensee performed an operability determination regarding past performance on the auxiliary feedwater motor-driven pumps and concluded that they have always retained their safety function. This violation was entered into the licensees corrective action program as Condition Report CR 16-2176. The team determined that the failure to impose administrative limits in surveillance procedures to promptly correct a condition adverse to quality was a performance deficiency. The performance deficiency was determined to be 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, if left uncorrected, the performance deficiency would have the potential to become a more significant safety concern. Specifically, operation of the motor driven auxiliary feedwater pumps with a diesel generator frequency acceptance criteria of up to 2 percent would allow operation in a regime where the pumps would not perform their safety function when called upon. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality. This finding had a cross-cutting aspect in the area of human performance associated with change management because the licensee failed to use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority. Specifically, the licensee did not properly evaluate the need to take appropriate interim corrective actions before the appropriate guidance was endorsed (H.3).
05000275/FIN-2016007-022016Q1Diablo CanyonFailure to Promptly Correct the Lack of Design Verification of 460 Vac Motors at Maximum Allowable FrequencyThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, prior to March 16, 2016, the licensee failed to assure that the lack of design verification of 460 Vac motors, which could be overloaded at the maximum allowable diesel generator frequency, was promptly corrected after having been identified in a 2013 apparent cause evaluation and again in a 2015 self-assessment as documented in Notifications 50572850 and 50826105, respectively. In response to this finding, the licensee performed a preliminary evaluation of the affected 460 Vac motors and concluded that operation at maximum emergency diesel generator frequency would not cause them to overheat or trip on overcurrent. This finding was entered into the licensee's corrective action program as Notifications 50835699 and 50838988. The team determined the failure to correct the lack of design verification of 460 Vac motors at maximum allowable frequency when powered from the emergency diesel generators was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design 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, operation of 460 Vac motors above their rated or analyzed maximum allowable frequencies could result in motor overheating or a trip of the thermal overload relays. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of problem identification and resolution associated with evaluation because the licensee failed to ensure that the organization thoroughly evaluated issues to ensure that resolutions address causes and extent of conditions (P.2).
05000275/FIN-2016007-052016Q1Diablo CanyonFailure to Evaluate the Voltage Effects of Limiting Design Basis Events on the 230 kV Offsite Power CircuitThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, which states, in part, The design control measures shall provide for verifying or checking the adequacy of design, such as by the performance of design reviews, by the use of alternate or simplified calculational methods, or by the performance of a suitable testing program. Specifically, prior to January 30, 2014, the licensee failed to verify the design of the 230 kV preferred offsite power source, such as by the performance of design reviews or use of alternate or simplified calculational methods, by assuming in calculation 359-DC, Determination of 230 kV Grid Capability Limits as DCPP Offsite Power Source, that the reactor trip and engineered safety features actuation system signals are coincident in time for all postulated design basis events. However, the plant is designed such that, during some events, the signals are separate in time and would result in a greater vital bus voltage depression than analyzed. In response to this finding, the licensee conducted a preliminary evaluation and concluded that the current transmission grid conditions were such that the calculation criteria would be met in the event of a design basis event involving non-coincident reactor trip and engineered safety features actuation system signals. This finding was entered into the licensee's corrective action program as Notification 50839137. The team determined the failure to evaluate the voltage effects of a limiting design basis event with non-coincident reactor trip and engineered safety features actuation system signals on the 230 kV offsite power circuit was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design 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, the failure to ensure adequate bus voltages as a result of a design basis event with non-coincident reactor trip and engineered safety features actuation system signals would result in a trip of the undervoltage relays and the loss of the preferred offsite power circuit. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with design margins because the licensee failed to ensure that the organization operated and maintained equipment within design margins and that margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000275/FIN-2016007-062016Q1Diablo CanyonFailure to Translate Appropriate Load Tap Changer Timing Acceptance Criteria into Periodic TestsThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, which states, in part, Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished. Specifically, prior to November 25, 2015, the licensee failed to include appropriate quantitative acceptance criteria in Procedure MP E-62.3, Tap Changer Functional Test for Standby-Startup Transformer 11, to ensure that the load tap changer speed for standbystartup transformer 11 was adequate to restore vital bus voltages to the required level during design basis events. In response to this finding, the licensee performed a preliminary evaluation of the condition and concluded that the most recently measured speed of the load tap changer was adequate to ensure that it would restore vital bus voltage within the required time. This finding was entered into the licensee's corrective action program as Notification 50839333. The team determined the failure to translate appropriate load tap changer timing acceptance criteria into functional tests to ensure that design assumptions were being maintained was a performance deficiency. The performance deficiency was more-than-minor, and therefore a finding, because it related to the design 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, the load tap changer could meet its functional test acceptance criterion, but not operate fast enough to restore vital bus voltages within the required time during design basis events, which would result in an undervoltage trip and loss of the preferred offsite power circuit. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated July 19, 2012, the finding screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of nontechnical specification equipment; and did not screen as potentially risk-significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with design margins because the licensee failed to ensure that the organization operated and maintained equipment within design margins and that margins were carefully guarded and changed only through a systematic and rigorous process (H.6).
05000498/FIN-2016007-012016Q1South TexasFailure to Perform Adequate Periodic Testing of Molded Case Circuit BreakersThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, which states, in part, a test program shall assure that all testing required to demonstrate that structures, systems, and components will perform satisfactorily in service is identified and performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents. Specifically, since March 22, 1988, the licensee failed to assure that all testing required to demonstrate that the safety-related molded case circuit breakers would perform satisfactorily in service was performed in accordance with the acceptance limits contained in Institute of Electrical and Electronics Engineers (IEEE) 308-1974. In response to this issue, the licensee determined that the molded case circuit breakers will remain operable while implementing corrective actions to ensure the appropriate testing requirements of the molded case circuit breaker were included in the test programs. This violation was entered into the licensees corrective action program as Condition Report CR 16-2166. The team determined that the failure to detect deterioration and demonstrate operability of molded case circuit breakers through appropriate testing was a performance deficiency. The performance deficiency was determined to be 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, inadequate periodic testing to detect deterioration and to demonstrate continued operability was a significant programmatic deficiency that would adversely affect the reliability of Class 1E molded case circuit breakers to perform satisfactorily in service. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the team determined the finding to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a structure, system, or component, and the structure, system, or component maintained its operability or functionality. This finding had a cross-cutting aspect in the area of human performance associated with consistent practices because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensee did not use a consistent approach to determine which molded case circuit breakers would or would not be tested (H.13).
05000498/FIN-2016007-062016Q1South TexasFailure to Correct Conditions Adverse to Quality Associated with the Eagle 21 SystemThe team identified a Green, non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, which states, in part, Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Specifically, since September 24, 2014, the licensee failed to establish measures to assure that deficiencies, deviations, defective material and equipment, and nonconformances that were responsible for malfunctions in the Class 1E Eagle 21 system were corrected. In response to this issue, the licensee performed an operability determination which determined the system was operable but in a degraded condition. This violation was entered into the licensees corrective action program as Condition Report CR 16-2220. The team determined that the failure to correct conditions adverse to quality in the Class 1E Eagle 21 system that were nonconformances with requirements was a performance deficiency. The performance deficiency was determined to be 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 of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to correct conditions adverse to quality in the Class 1E Eagle 21 system adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of the protective action implemented by the qualified display processing system. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with conservative bias because the licensee individuals failed to use decision making practices that emphasize prudent choices over those that are simply allowable (H.14).
05000498/FIN-2016007-082016Q1South TexasFailure to Ensure Sufficient Capacity and Capability of Mitigating Systems during a Station Blackout EventThe team identified a Green, non-cited violation of 10 CFR 50.63(a)(2) which states, in part, The reactor core and associated coolant, control, and protection systems, including station batteries and any other necessary support systems, must provide sufficient capacity and capability to ensure that the core is cooled and appropriate containment integrity is maintained in the event of a station blackout for the specified duration. Specifically, since September 12, 2013, the battery sizing and load profile calculations of the channel I (A train) direct current battery bus failed to include proper design data for expected loads and possible worst case load currents. In response to these issues, the licensee determined the battery bus was operable and the licensee initiated actions to analyze the effects of the change in calculation methodology, as well as to account for the additional loads. This finding was entered into the licensee's corrective action program as Condition Reports CR 16-1794, CR 16-2197, and CR 16-2236. The team determined that the failure to ensure the capacity and capability of protection systems to provide support for core cooling and containment integrity maintenance in the event of a station blackout 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. In addition, if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. Specifically, if the channel I emergency safety features direct current bus were required to support loads for the four hour coping period, the licensee may subject components used to ensure core cooling and containment integrity to conditions that were not assumed in their station blackout analysis. In accordance with Inspection Manual Chapter 0609, Appendix A, The Significance Determination Process (SDP) for Findings At- Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, the issue screened as having very low safety significance (Green) because it was a design or qualification deficiency that did not represent a loss of operability or functionality; did not represent an actual loss of safety function of the system or train; did not result in the loss of one or more trains of non-technical specification equipment; and did not screen as potentially risk significant due to seismic, flooding, or severe weather. This finding had a cross-cutting aspect in the area of human performance associated with procedure adherence because the licensee failed to follow process, procedures, and work instructions. Specifically, the licensee did not follow the calculation change process procedures to complete an impact review of pertinent licensing information associated with station blackout when the battery load assumptions were revised in the station blackout coping calculation (H.8).
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