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05000331/FIN-2018003-012018Q3Duane ArnoldLicensee-Identified ViolationA violation of very low safety significance (Green)was identified by the licensee and has been entered into the corrective action program. This violation is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Title 10 of the Code of Federal Regulations (CFR) 50, Appendix B, Criterion III, 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. System Design Specification APEDA61019, Pressure Integrity of Piping and Equipment Pressure Parts Data Sheet, required in the applicable castings section T1.3.3.b, all accessible surfaces including machine surfaces shall be examined by either the magnetic particle or liquid penetrant method in either the furnished or finished condition. Contrary to the above, in October 2016, measures were not established to assure that applicable design basis requirements as defined in 10 CFR 50.2 were translated into work instructions repairing the B inboard main steam isolation valve, CV 4415, during RFO 25. Specifically, instructions to perform a NDE of machined surfaces following the valve repair were not included in the work package. As a result, the non-destructive examination was not performed prior to placing the valve into service.
05000331/FIN-2018003-022018Q3Duane ArnoldMinor ViolationDuring Mode 1 power operations on July 9, 2018, the licensee had both doors of a secondary containment airlock open simultaneously, and a minor violation of Technical Specification (TS) 3.6.4.1 Secondary Containment was self-revealed. During the time both doors were open, approximately 3 seconds, the allowable penetration opening area was exceeded and rendered the secondary containment inoperable. Technical Specification 3.6.4.1 requires secondary containment to be operable in Modes 1, 2 and 3. Technical Specification Surveillance Requirement 3.6.4.1.2 supports secondary containment operability by verifying that either the outer door(s) or the inner door(s) in each secondary containment access opening are closed. The posted instructions at each secondary containment airlock door stated, ATTENTION Push Button To Be Held In For 2 Seconds Prior To Opening Door, to be of a type appropriate for traversing the containment airlock. Contrary to the above, at approximately 1:34 p.m. on July 9, 2018, while operating in Mode 1 at 97 percent power, two individuals simultaneously traversing through opposite doors of a secondary containment airlock each failed to hold the airlock interlock push button for two seconds prior to opening their respective doors resulting in a momentarily inoperability of secondary containment. Operability was restored upon the immediate closure of one of the two doors. Subsequently, maintenance was unable to recreate the condition and satisfactorily performed Surveillance Test Procedure (STP) 3.6.4.102, Secondary Containment Airlock Verification, and GMPELEC44,Section A5.1,Airlock Door Interlock Checks.The licensee entered this
05000331/FIN-2018002-022018Q2Duane ArnoldMinor ViolationMinor Violation: On June 19, 2016, while operating at 82 percent power, two secondary containment access airlock doors were opened simultaneously during surveillance testing as part of STP 3.6.4.102, Secondary Containment Airlock Verification. The inspectors determined this event was caused by inadequate procedural guidance which directed the user to attempt to open one airlock door while the other door was already open. During this test, the interlock failed because the permanent magnets had rotated and were misaligned. This failure could have been identified without challenging airlock interlock integrity if the second airlock door wasnt held open. The failure to have adequate procedural guidance for testing the secondary containment airlock doors was a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, which requires licensees to have procedures appropriate to the circumstance when performing safety-related activities. In response to this issue, the licensee immediately closed the airlock doors. In addition, the licensee submitted a TS change request to address the concurrent opening of two secondary containment airlock doors. The licensees corrective action program is tracking the TS change as CR 02034076, Secondary Containment Airlock Doors #225 and 228 Both Opened. Screening: The issue screened as minor because all of the questions associated with a minor issue found in IMC 0612, Appendix B were answered No due to the licensee reestablishing secondary containment operability immediately after the second airlock door opened. In addition, the inspectors considered the failure to have an appropriate procedure was less than a Severity Level IV violation in accordance with the NRCs Enforcement Policy. Violation: The failure to comply with 10 CFR Part 50, Appendix B, Criterion V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes LER 05000331/2016001.
05000331/FIN-2018002-012018Q2Duane ArnoldInappropriate Procedural Guidance Resulted in Loss of Scram Function and Failure to Enter Technical Specification Limiting Condition for OperationThe inspectors identified a finding of very low safety significance (Green) and a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have procedures appropriate to the circumstance for testing the main steam isolation valve (MSIV) and turbine stop valve (TSV) closure functions. Specifically, STP 3.3.1.117, MSIV Functional Test, and STP 3.3.1.119, Main Turbine Stop and Combined Intermediate Valves Test, directed the use of a reactor protection system test box which disabled the MSIV and the TSV closure automatic reactor scram functions while testing specific combinations of MSIVs and TSVs and failed to require entry into appropriate Technical Specification Limiting Condition for Operation action statements.
05000346/FIN-2018002-042018Q2Davis BesseMisapplication of the Operability Determination ProcessThe NRC identified a finding of Green significance due to the licensees misapplication of NOPOP1009, Operability Determinations and Functionality Assessments. Specifically, the licensee failed to apply the Operability Determination process in accordance with procedures.
05000346/FIN-2018002-032018Q2Davis BesseFailure to Perform a Procedure Affecting QualityThe NRC identified a finding of Green significance and an associated non-cited violation of 10 Code of Federal Regulation(CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, due to the licensees failure to implement DBOP03006, Miscellaneous Instrument Shift Checks, Specifically, the licensee declared SFAS Channel 1 operable without performing the required channel check.
05000346/FIN-2018002-022018Q2Davis BesseFailure to Apply Technical Specification for Safety Features Actuation SystemInstrumentationThe NRC identified a finding of Green significance and an associated Non-Cited Violation of Technical Specification 3.3.5.b, Safety Features Actuation System (SFAS) Instrumentation, for the licensees failure to place the reactor in Mode 3 within six hours of identifying that two channels of Safety Features Actuation System Borated Water Storage Tank level instrumentation were inoperable. Specifically, the licensee inappropriately exited Technical Specification 3.3.5.b, and failed to place the reactor in Mode 3 while two Borated Water Storage Tank level instruments were inoperable for more than six hours.
05000346/FIN-2018002-012018Q2Davis BesseFailure to Follow the Makeup and Purification ProcedureA self-revealed Green finding and associated Non-Cited Violation of Technical Specification 5.4.1.a, Procedures, was identified when the licensee failed to follow station procedure DBOP06006, Makeup and Purification System. Specifically, the licensee failed to open MU177, the Make-Up Filter 1 Outlet Isolation valve, which resulted in the isolation of letdown while swapping make-up filters.
05000346/FIN-2017004-012017Q4Davis BesseFailure to Maintain Procedures Associated with Ventilation Air Monitoring Assessment ProgramThe inspectors identified a finding of very-low safety significance and an associated NCV of Technical Specification 5.4.1 for the failure to maintain procedures for station vent releases during planned scenarios. Specifically, the inspectors identified multiple procedures that were not updated when the station vent monitors were replaced in 2014. This issue has been entered into the licensees Corrective Action Program as CR201710817. Corrective actions taken included the issuance of a Standing Order for collecting samples during accident conditions, provided Just-In-Time training for chemistry technicians, and revision of the outdated procedures. The performance deficiency was determined to be more-than-minor in accordance with Inspection Manual Chapter 0612, Appendix B, Issue Screening. Specifically, if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern in that the failure to maintain procedures to collect station vent samples under all predicted conditions could result in the inability to measure the amount of gaseous radioactivity leaving the plant and to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The finding was assessed using Inspection Manual Chapter 0609 Appendix D, Public Radiation Safety Significance Determination Process, and was determined to be of very-low safety significance because the issue involved radioactive effluent releases, but did not: (1) represent a substantial failure to implement the Radioactive Effluent Release Program; or (2) result in public exposure that exceeded the dose values in Appendix I to 10 CFR, Part 50, and/or 10 CFR, Part 20.1301(e) limits. The inspectors determined that the finding had a cross-cutting component in the area of Human Performance, in the aspect of Work Management: specifically, the organization did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. (H.5)
05000346/FIN-2017004-022017Q4Davis BesseInterface Between New Accident Range Ventilation Monitors and the Emergency Preparedness Dose Assessment ProgramDuring inspection activities associated with the accident range station vent monitor, the inspectors identified an unresolved item (URI) associated with the interface between the monitor and the Dose Assessment Program used to project dose to members of the public during potential accident conditions. Description: The licensee replaced the accident range station vent monitors in 2014 using ECP 040006, Replace Kaman Radiation Monitors. The replacement monitors were manufactured by a different company than the original monitors, had different detection capabilities, different system calibration, and different computer hardware to convert detector output into usable information. The licensee could not immediately provide specifics regarding the interface between the new accident range monitors and the program used during accident conditions for providing dose projections and the resulting protective action recommendations. The inspectors focus of concern was how the new accident range monitors accounted for the potentially rapidly changing mixture of radioactive gases during the early phase of a postulated accident. Consequently, this issue remains under review by the NRC awaiting for additional information from the licensee to verify the new monitor interface to determine if it represents a performance deficiency and is categorized as a URI. (URI 05000346/201700403, Interface Between New Accident Range Ventilation Monitors and the Emergency Preparedness Dose Assessment Program)
05000346/FIN-2017004-032017Q4Davis BesseFailure to Prescribe Appropriate Work Instructions for an Activity Affecting QualityA self-revealed finding with an Apparent Violation (AV) of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and an associated violation of technical specification (TS) 3.7.5, Emergency Feedwater (EFW), was identified on September 13, 2017, due to the licensees apparent failure to prescribe appropriate work instructions for an activity affecting quality of the safety-related auxiliary feedwater (AFW) system. Specifically, the licensee apparently did not provide appropriate instructions to maintain an adequate amount of oil in the AFW turbine bearing oil sumps, resulting in the failure of AFW 1 on September 13, 2017. The licensee entered this issue into the CAP as CR201709443 and CR201709857, immediately replaced the damaged bearing, and updated the lubrication manual data sheets to include sight glass marking dimensions per vendor guidance. The apparent performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and potentially adversely affected the cornerstone objective of ensuring the availability, capability and reliability of equipment that respond to initiating events. Specifically, the apparent performance deficiency resulted in the failure of the AFW 1. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and IMC 0609 Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012, the finding was screened against the mitigating systems cornerstone. The inspectors determined the finding represented an apparent actual loss of function of at least a single train for greater than its technical specification allowed outage time. Therefore, a detailed risk evaluation will be performed by a regional senior reactor analyst. Because the safety characterization of this finding is not yet finalized, it is being documented with a significance of to be determined (TBD). The inspectors determined this finding affected the cross-cutting aspect of challenge the unknown in the area of Human Performance, where individuals stop when faced with uncertain conditions. Risks are evaluated and managed before proceeding. Specifically, licensee personnel apparently did not stop when faced with uncertain conditions in the preventive maintenance procedure for replacing the AFPT sight glasses. Although the replacement of the AFPT 1 inboard bearing sight glass occurred in 1997, the licensee had the opportunity to challenge the lack of detail in the work instructions in late 2014 when the AFPT 2 outboard bearing sight glass was replaced. (H.11)
05000346/FIN-2017004-042017Q4Davis BesseFailure to Document a Degraded Condition on the AFPT 1 Outboard BearingThe inspectors identified a finding of very low safety significance for the licensees failure to document a degraded condition of a safety-related system in the corrective action program (CAP), as required by licensee procedure, NOPLP2001. Specifically, during planned maintenance on auxiliary feedwater pump turbine (AFPT) 1, the licensee identified scoring on the outboard turbine bearing and failed to generate a condition report detailing the issue. The licensee entered this issue into the CAP as condition report (CR) 201712487 for evaluation. The inspectors determined the performance deficiency was more than minor because if left uncorrected it had the potential to lead to a more significant safety concern. Specifically, the failure to document a degraded condition in the CAP did not allow the organization to properly assess the issue. Therefore, the underlying cause may not have been appropriately addressed. Using IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 7, 2016, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, issued May 9, 2014, the inspectors determined the finding to be of very low safety significance (Green) because the inspectors answered no to all questions in Exhibit 3 of Appendix G, Attachment 1. The inspectors determined this finding affected the cross-cutting aspect of identification in the area of Problem Identification and Resolution, where the organization implements a corrective action program with a low threshold for identifying issues and individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to completely identify the degraded condition, resulting in the failure to document the issue. (P.2)
05000254/FIN-2017004-012017Q4Quad CitiesRepeat Use of Written Exams during Licensed Operator Requalification Examinationsa. Inspection ScopeThe following inspection activities were conducted during the weeks of October 9 and October 16, 2017, to assess: (1) the effectiveness and adequacy of the facility licensees implementation and maintenance of its systems approach to training (SAT) based LORT Program put into effect to satisfy the requirements of 10 CFR 55.59; (2) conformance with the requirements of 10 CFR 55.46 for use of a plant referenced simulator to conduct operator licensing examinations and for satisfying experience requirements; and (3) conformance with the operator license conditions specified in 10 CFR 55.53. The documents reviewed are listed in the Attachment to this report.Licensee Requalification Examinations (10 CFR 55.59(c); SAT Element 4 as Defined in 10 CFR 55.4): The inspectors reviewed the licensees program for development and administration of the LORT biennial written examination and annual operating tests to assess the licensees ability to develop and administer examinations that are acceptable for meeting the requirements of 10 CFR 55.59(a).- The inspectors conducted a detailed review of one biennial requalification written examination versions to assess content, level of difficulty, and quality of the written examination materials. (02.03)- The inspectors conducted a detailed review of ten job performance measures and four simulator scenarios to assess content, level of difficulty, and quality of the operating test materials.(02.04)- The inspectors observed the administration of the annual operating test to assess the licensees effectiveness in conducting the examination(s), including the conduct of pre-examination briefings, evaluations of individual operator and crew performance, and post-examination analysis. The inspectors evaluated the performance of one crew in parallel with the facility evaluators during two dynamic simulator scenarios, and evaluated various licensed crew members concurrently with facility evaluators during the administration of several job performance measures. (02.05)- The inspectors assessed the adequacy and effectiveness of the remedial training conducted since the last requalification examinations and the training planned for the current examination cycle to ensure that they addressed weaknesses in licensed operator or crew performance identified during training and plant operations. The inspectors reviewed remedial training procedures and individual remedial training plans. (02.07) Conformance with Examination Security Requirements (10 CFR 55.49): The inspectors conducted an assessment of the licensees processes related to examination physical security and integrity (e.g., predictability and bias) to verify compliance with 10 CFR 55.49, Integrity of Examinations and Tests. The inspectors observed the implementation of physical security controls (e.g., access restrictions and simulator I/O controls) and integrity measures (e.g., security agreements, sampling criteria, bank use, and test item repetition) throughout the inspection period. (02.06)Conformance with Operator License Conditions (10 CFR 55.53): The inspectors reviewed the facility licensee's program for maintaining active operator licenses and to assess compliance with 10 CFR 55.53(e) and (f). The inspectors reviewed the procedural guidance and the process for tracking on-shift hours for licensed operators, and which control room positions were granted watch-standing credit for maintaining active operator licenses. Additionally, medical records for seven licensed operators were reviewed for compliance with 10 CFR 55.53(I). (02.08)Conformance with Simulator Requirements Specified in 10 CFR 55.46: The inspectors assessed the adequacy of the licensees simulation facility (simulator) for use in operator licensing examinations and for satisfying experience requirements. The inspectors reviewed a sample of simulator performance test records (e.g., transient tests, malfunction tests, scenario based tests, post-event tests, steady state tests, and core performance tests), simulator discrepancies, and the process for ensuring continued assurance of simulator fidelity in accordance with 10 CFR 55.46. The inspectors reviewed and evaluated the discrepancy corrective action process to ensure that simulator fidelity was being maintained. Open simulator discrepancies were reviewed for importance relative to the impact on 10 CFR 55.45 and 55.59 operator actions as well as on nuclear and thermal hydraulic operating characteristics. (02.09)Problem Identification and Resolution (10 CFR 55.59(c); SAT Element 5 as Defined in 10 CFR 55.4): The inspectors assessed the licensees ability to identify, evaluate, and resolve problems associated with licensed operator performance (a measure of the effectiveness of its LORT Program and their ability to implement appropriate corrective actions to maintain its LORT Program up to date). The inspectors reviewed documents related to licensed operator performance issues (e.g., licensee condition/problem identification reports including documentation of plant events and review of industry operating experience from previous 2 years). The inspectors also sampled the licensees quality assurance oversight activities, including licensee training department self-assessment reports. (02.10)This inspection constituted one Biennial LOR Program inspection sample as defined in IP 71111.1105.b. FindingsIntroduction: While performing an assessment of the licensees processes related to examination physical security and integrity (e.g. predictability and bias) to verify compliance with 10 CFR 55.49, Integrity of Examinations and Tests, the inspectors 10 identified that Quad Cities 2015 LOR written examinations were duplicated from the 2013 LOR examinations, that 2017 LOR written examinations were duplicated from the 2015 LOR examinations, and that four individuals were administered the same written examinations from the previous exam cycle.Description: The inspectors identified that, with few exceptions, the licensee had duplicated or reused questions from the 2015 written exam when they created the 2017 written exam. The licensee created six LOR written exam versions (i.e., AF), one for each crew. For the 2017 biennial exam, the licensee essentially swapped exam versions from 2015 that were given to each crew (i.e., the 2015 Version A was given to crew B in 2017 and Version B was given to crew A, etc.). The inspectors noted that no crew received the same exam version in 2017 as they did in 2015. However, due to crew personnel adjustments/realignments, the inspectors requested the licensee to investigate if, and how many, operators were going to receive the same exam in 2017 as in 2015. The licensee identified that one reactor operator had already taken the same exam in 2017 that they were given in 2015. In addition, the licensee also identified that two additional licensed operators were scheduled to take the same exam they had taken in 2015, but they had not yet been given the exam due to the exam schedule. After discussing the issue and concern with the inspectors, the licensee decided to administer those two individuals different exam versions to which they had not been previously exposed. In addition, the inspectors inquired how long the particular set of exam versions had been reused and swapped among the crews (i.e., before 2015). The licensee reviewed biennial written exams in 2013 and 2011 and determined the exam content was different and stated, there was no predictable pattern in exam versions. After reviewing all of the 2013 exam versions, the inspectors identified that three versions were a mixture of questions between reused and new questions. For example, 2013 Version A was a mixture of questions of 2015 exam Versions C and D and twounique questions. The 2013 Version B was a mixture of 2015 Version C and D and seven unique questions. The 2013 Version F was a mixture of 2015 D and F and fiveunique questions. The three remaining versions from 2013 were replicated in 2015, but given to different crews. The inspectors requested the licensee determine the number of personnel that took the same exam in 2015 as in 2013, and the licensee identified three individuals who were given the same exam in 2013 and 2015 (two senior reactor operators and one reactor operator). The inspectors are considering this issue to be an unresolved item (URI) concerning whether the repeated use of a biennial written examination for sequential requalification programs (consecutive 24 month periods), and the resulting predictability induced to the examination process, constitutes a violation of 10 CFR 55.49, Integrity of Examinations and Tests. The inspectors have requested the licensee provide the written examinations in question to the inspectors for further review. The inspectors will review individual questions of the written examinations in order to determine if there were sufficient differences between the examinations to characterize the examinations as either different or similar. The results of the review will be used to determine if a violation of 10 CFR 55.49 requirements exists. (URI 05000254/201700401; 05000265/201700401: Repeat Use of Written Exams during Licensed Operator Requalification Examinations)
05000373/FIN-2017003-012017Q3LaSalleInadequate Maintenance Rule Monitoring of the Low Pressure Core Spray SystemThe inspectors identified a Green NCV of Title 10 of the Code of Federal Regulation(CFR) 50.65(a)(1) for the failure to monitor the performance of the Unit 1 low pressure core spray (LPCS) system against licensee-established goals. Specifically, the licensee did not identify and properly account for a maintenance rule functional failure (MRFF) of the Unit 1 LPCS min-flow valve differential pressure switch, which demonstrated that performance of the Unit 1 LPCS system was not being controlled in accordance with the maintenance rule. The Licensees immediate corrective actions included entering this issue into their corrective action program (CAP), re-evaluating and classifying the LPCS min-flow valve differential pressure switch failure as a MRFF, and entering the system into (a)(1) status. This finding was entered into the licensees CAP as action request (AR) 4029999.The performance deficiency was determined to be more-than-minor in accordance with IMC 0612 Appendix B, Issue Screening, dated September 7, 2012, because it was associated with the Equipment Performance attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee did not properly classify the May 17, 2017, failure of the LPCS min-flow valve differential pressure switch as a MRFF. When properly classified, this failure caused the maintenance rule performance criteria for the LPCS system to be exceededcausing the system to receive additional remedial station attention. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued October 16, 2016, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, dated June 19, 2012, the inspectors determined that this maintenance rule program-based finding is of very low safety significance (Green) since it was not a deficiency affecting the design or qualification of a mitigating structure, system, or component, it did not represent the loss of a system and/or function, it did not represent an actual loss of function of at least a single train or two separate safety systems out-of-service for greater than its technical specifications allowed outage time, and it did not represent an actual loss of a non-technical specification equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. The inspectors determined this finding affected the cross-cutting area of Problem Identification and Resolution in the aspect of Evaluation, where the organization thoroughly evaluates issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the Licensee failed to thoroughly evaluate the failure of the Unit 1 LPCS min-flow valve differential pressure switch on May 17, 2017 (P.2).
05000440/FIN-2015003-032015Q3PerryLicensee-Identified ViolationDuring a review of items entered in the licensees CAP, the inspectors identified that corrective action item (CR 201416769) documented direct and root causes for the reactor scram that occurred on November 7, 2014. The inspectors reviewed the root cause analysis and the corrective actions taken to prevent recurrence. The direct cause of the event was injection of a false feed flow runback signal, caused by the redundant reactivity control system (RRCS) self-test feature, into the digital feedwater control system DFWCS) which caused both contacts in the A and B divisions to close simultaneously, thus actuating a real feedwater runback. The licensee determined that the design was not adequate to prevent this event from occurring and that the root cause of the event was a latent design flaw from the original digital upgrade design package. The latent design flaw was identified by the licensee as a violation of 10 CFR 50, Appendix B, Criterion III, Design Control, which requires in part, design control measures for verifying or checking the adequacy of the design. The corrective actions, which consisted of physical modifications to plant equipment, were previously reviewed by inspections conducted during the refueling outage, March and April of 2015, and documented in Perry Integrated IR 2015002. The inspectors evaluated the licensee-identified violation using IMC 0612, Appendix B, Issue Screening, and determined that the deficiency was more than minor because it was associated with the Initiating Events Cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors further evaluated the issue in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, and determined that the safety significance of this event was very low because, in accordance with the initiating events screening questions, all safety systems functioned as designed and the scram was not complicated. This issue is also discussed in Sections 4OA3.1 and 4OA7.
05000440/FIN-2015003-012015Q3PerryInadequate Operating Procedure for Diesel Generator Building Ventilation SystemThe inspectors identified a finding of very low safety significance and associated non-cited violation (NCV) of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure as of July 8, 2015, to establish and maintain an adequate procedure for operation of the Diesel Generator Building Ventilation System (DGBVS). Specifically, the DGBVS operating procedure did not ensure that diesel room temperature would remain below limits during testing. The failure to establish and maintain an adequate procedure was a performance deficiency and resulted in the Division 2 Diesel Generator room temperatures exceeding specified limits. The performance deficiency was more than minor, and thus 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. The inspectors determined that the finding was of very low safety significance because the finding is a deficiency affecting the design or qualification of a mitigating structure, system, and component (SSC) that maintained its operability. This finding has a cross-cutting aspect in the area of human performance, design margins, because the licensee did not incorporate the degree of redundancy specified in the Updated Safety Analysis Report for DGBVS into the applicable operating procedures.
05000440/FIN-2015003-022015Q3PerryFailure to Properly Implement Steps Outlined in a Technical Specification Surveillance ProceA finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1., Procedures, was self-revealed on August 5, 2015, when an unexpected isolation of the reactor core isolation cooling (RCIC) system occurred as a result of the licensees failure to properly implement the steps outlined in TS Surveillance Procedure, SVIE31T5395B, RCIC Steam Line Flow High Channel Functional for 1E31N684B. Specifically, during performance of the surveillance, several steps were marked as not applicable that were applicable to prevent the isolation of the RCIC system. As a result, the licensee failed to lift leads as required by the procedure and the RCIC steam supply inboard isolation valve then closed when the isolation trip signal was applied during the test. The licensee took immediate actions to restore system operability and availability and conducted a human performance event response investigation. A standing order for both Operations and Instrumentation and Controls personnel was initiated addressing interim actions for control room surveillance performance and to reinforce maintenance fundamentals and human performance behaviors. The licensees failure to properly implement the steps in the procedure was a performance deficiency that was determined to be more than minor, and thus a finding, because it was associated with the Mitigating Systems Cornerstone attribute of equipment performance and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of one or more non-Technical Specification trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of human performance, avoid complacency, for failing to recognize and plan for the possibility of mistakes, and for failure to implement appropriate error reduction tools, such as proper self-checks and peer checks, which resulted in an isolation of the RCIC system.
05000346/FIN-2015502-012015Q2Davis BesseLicensee-Identified ViolationThe licensee-identified a finding of very low safety significance (Green) and an associated violation of 10 CFR 50.54 (q)(2) and 10 CFR Part 50.47(b)(14). Title 10 CFR 50.54(q)(2), requires, in part, that a holder of a license under this part, shall follow and maintain the effectiveness of an emergency plan that meets the requirements of Appendix E, of Part 50, and for nuclear power reactor licensees, the planning standards of 10 CFR 50.47(b). Title 10 CFR 50.47(b)(14) requires, in part, that Periodic Exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. Section 8.1.2.c.6 of the Davis-Besse Nuclear Power Station Emergency Plan, Revision 30, states, Semiannual Health Physics drills will be conducted which involve response to, and analysis of, simulated elevated airborne and liquid samples and direct radiation measurements in the environment. Contrary to the above, from mid-2013 to the end of 2014, the licensee failed to comply with the established drill and exercise program. Specifically, the health physics drill objectives were only being partially met during this time period. The drill scenarios were limited and did not provide an opportunity for the participants to complete sampling/analysis of liquid samples. As part of the corrective actions after the discovery of this issue, the licensees Emergency Response Staff conducted a drill on December 11, 2014, to ensure that all aspects of the Health Physics drill objectives were met. The performance deficiency was more than minor because the issue was associated with the Emergency Preparedness cornerstone and adversely affected the associated cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the drill scenarios were limited in scope and health physics objectives specified in the emergency plan, were not carried out fully during exercises and drills, that were conducted in the middle of 2013 through the end of 2014. The NRC determined that this was a failure to comply with the licensees emergency plan and a degradation of a planning standard function in accordance with 10 CFR, Part 50.47(b)(14), and was a very low safety significance issue (Green) as indicated in Inspection Manuel Chapter 0609, Emergency Preparedness SDP, Appendix B, Attachment 2, Failure to Comply Significance Logic. Because the finding is of very low safety significance (Green) and it was entered into the licensees Corrective Action Program as Condition Report, CR-2014-16715, this violation is being treated as an non-cited violation, consistent with Section 2.3.2 of the NRC Enforcement Policy.
05000255/FIN-2015002-012015Q2PalisadesFailure to Take Appropriate Corrective Action for the Charging System While in Maintenance Rule (a)(1) StatusGreen. An NRC-identified finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) 50.65(a)(1) was identified for the failure to take appropriate corrective actions for the charging system, while in Maintenance Rule (a)(1) status, when performance or condition goals were not met. Specifically, on April 2, 2015, the front cap of the B charging pump cracked, causing volume control tank (VCT) level and pressure to lower, most likely due to excessive local cavity pressures in the pump caused by the suction accumulator pressure being out of specification. Accumulator pressures being out of specification, which causes pressure oscillations and vibrations in the charging pumps and their associated suction and discharge piping, was a similar cause to previous maintenance rule system functional failures that occurred in 2013 and 2014, which transitioned the system to (a)(1) status in July 2014. The licensee documented the issue in their corrective action program (CAP), conducted an equipment apparent cause evaluation (EACE) for the most recent failure, and revised the Maintenance Rule (a)(1) Action Plan to address the on-going issues with the suction and discharge accumulators. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612 because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The charging system provides the critical safety functions of pressure and inventory control in the emergency operating procedures. The finding screened as having very low safety significance (i.e., Green) based on answering No to all the screening questions under the Mitigating Structures, Systems, and Components (SSCs) and Functionality section of the significance determination process (SDP). The finding had a cross-cutting aspect of Evaluation in the Problem Identification and Resolution area. Specifically, the organization did not thoroughly evaluate previous data on the suction and discharge accumulators pressures being out of specification and what affect that may have on the system. Also, when the accumulator pressures were found out of specification, sometimes that information was not documented in condition reports (CRs), nor were the preventive maintenance (PM) frequencies re-evaluated in a technical and rigorous manner to ensure the correct PM activities were being conducted on these components in a timely manner to assure system reliability.
05000255/FIN-2015002-022015Q2PalisadesFailure to Wear Prescribed Respiratory ProtectionGreen. A self-revealed finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1 was identified for insulation work activities during the refueling outage associated with pressurizer spray valve CV1057. Specifically, prior to the work beginning, the licensee determined that the use of powered air purifying respirators would be required to minimize worker dose and maintain exposures as-low-as-reasonably-achievable (ALARA), but the work was performed using only face shields, and as a result a worker was contaminated externally and internally. Corrective actions included creation of an administrative requirement to revise any radiation work permit (RWP) task that required respiratory protection to more clearly state the requirements. The inspectors determined that the performance deficiency was more than minor in accordance with IMC 0612 because it was associated with the Program and Process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to wear required respiratory protection during the reinsulating of CV1057 resulted in personal contamination and the intake of radioactive material. The inspectors concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The finding was determined to be of very low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, because it was not an ALARA planning issue, there was neither an overexposure nor a substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors concluded that the cause of the issue involved a cross-cutting aspect in the area of Human Performance, Basis for Decisions. Specifically, the bases for operational decisions were communicated in a timely manner.
05000255/FIN-2015002-032015Q2PalisadesLicensee-Identified ViolationTitle 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements in 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(4) states, A standard emergency classification and action level scheme, the bases of which includes facility system and effluent parameters, is in use by the nuclear facility licensee, and State and local response plans call for reliance on information provided by facility licensees for determinations of minimum initial offsite response measures. Section 4.1 of the Palisades Site Emergency Plan states, in part, that a conservative philosophy for classification shall be used to declare the highest classification for which an EAL has been exceeded and that Palisades EALs can be found in the Site Emergency Plan, Supplement 1 EAL Wall Charts. Site Emergency Plan, Supplement 1 EAL Wall Charts requires, in part, the declaration of an Unusual Event for EAL HU 1.1 for a seismic event if identified by both: (1) earthquake felt in plant, and (2) National Earthquake Information Center. The Palisades EAL Technical Basis defines a felt earthquake as, An earthquake of sufficient intensity such that: (a) the vibratory ground motion is felt at the nuclear plant site and recognized as an earthquake based on a consensus of control room operators on duty at the time, and (b) for plants with operable seismic instrumentation, the seismic switches of the plant are activated. Contrary to the above, on May 2, 2015, the licensee control room staff declared EAL HU 1.1 without meeting the threshold criteria stated in the EAL. Specifically, the control room staff based the Notification of Unusual Event declaration on outside personnel reports and information from the National Earthquake Center, but without the consensus of control room operators on duty at the time or a valid recorded indication on the operable seismic instrumentation. The NRC determined that the EAL Overclassification, which resulted in no unnecessary protective action recommendations, was of very low safety significance (i.e., Green) as specified in IMC 0609, EP Significance Determination Process, Appendix B, Figure 5.41, Significance Determination for Ineffective EALs and Overclassification. As such the NRC determined this to be an NCV in accordance with Section 2.3.2 of the Enforcement Policy. The licensee entered this issue into their CAP as CR 201508137.
05000255/FIN-2015002-042015Q2PalisadesLicensee-Identified ViolationThe licensee identified an NCV of TS 5.4.1, Procedures for an inadequate procedure that failed to ensure all of the regulatory requirements for fit testing of respiratory protection were satisfied before use. Technical Specification 5.4.1 required, in part, that written procedures shall be established, written, and maintained for respiratory protection. The licensees procedure governing respirator fit testing was ENRP505, Portacount Respirator Fit Testing, and provided the fleet standard for performing personnel respirator fit testing for tight-fitting respirators. Title 29 CFR 1910.134, Respiratory Protection. Specifically, 29 CFR 1910.134(f)(8) states that, Fit testing of tight-fitting atmosphere-supplying respirators and tight-fitting PAPRs shall be accomplished by performing quantitative or qualitative fit testing in the negative pressure mode, regardless of the mode of operation (negative or positive pressure) that is used for respiratory protection. Contrary to the above, on June 2, 2013, multiple respirator fit tests of tight fitting PAPRs were performed in only the positive pressure mode. The inspectors reviewed IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The inspectors determined that it was not an ALARA planning issue, there was neither an overexposure nor a substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. Therefore the finding screened as having very low safety significance (i.e., Green). The licensee entered this issue into their CAP as CRPLP20132469.
05000331/FIN-2015001-032015Q1Duane ArnoldLicensee-Identified ViolationDuane Arnold TS 5.4, Procedures, Section 5.4.1.a requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Regulatory Guide 1.33, Revision 2, Appendix A, contains, in part under Section 8.b(2)(t), surveillance test procedures for inspection of the reactor coolant system pressure boundary. Contrary to the above, on November 8, 2014, the licensee failed to properly implement surveillance test procedure (STP) 3.10.102, Non Nuclear Heat Class 1 Ten Year System Leakage Pressure Test, Revision 32. Specifically, during the Fall 2014 refueling outage, licensee personnel identified leakage during visual undervessel inspections per STP 3.10.102. Although several CRs were generated to capture the identified leakage locations and approximate leakage rates from control rod drive mechanism (CRDM) flanges, the personnel failed to fully implement STP 3.10.102, Attachment 3 requirements to perform a detailed inspection of the associated CRDM flanges to identify the leakage source and to verify pressure boundary integrity. Had this identification/verification been performed, STP 3.10.1-02, Attachment 3, further required implementation of GMPTEST66, CRD (**-**) Troubleshooting Procedure, Revision 8, for CRDM flange leakage. Because CRs were written, the licensee personnel considered the under-vessel inspection results satisfactory and moved forward in the STP. Upon further review of the completed STP, the licensee identified that required detailed inspections were not performed for the CRDM flange leaks. The licensee entered the issue into the CAP and successfully re-performed STP 3.10.102 after resolving the leakage issues. Because the inspectors answered No to all questions under Exhibit 4 of IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02006364. Corrective actions included a revision to STP 3.10.102 to more clearly define under-vessel visual inspection requirements.
05000331/FIN-2015001-042015Q1Duane ArnoldLicensee-Identified ViolationTitle 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances. Contrary to the above, on September 27 and 29, 2014, the licensee failed to prescribe an instruction appropriate to the circumstances associated with the replacement of shielded cables between the A and C RHRSW pump motors and the associated 4kV supply breakers. Specifically, SPECE512, Cable and Wire Installation, Revision 14, did not ensure that shielded cables be grounded only at the switchgear end, and that the cables be routed back through ground fault (ring) current transformers in the cabinet before being grounded. This resulted in the improper development of work instructions used in the installation of replacement cables for the A and C RHRSW pumps and a resultant non-conforming condition which was discovered by the licensee during an extent of condition review in March of 2015. Because the SSCs maintained operability based on the deficiency affecting the design of the SSCs, the finding screened as very low safety significance (Green). The above issue was documented in the licensees CAP as CR 02023605. Immediate corrective actions included a determination of operability (the ground fault protection had no required safety supporting function for the RHRSW pumps and switchgear), equipment configuration control until resolution was taken, re-routing of the affected cables to restore full design, and a revision to SPEC-E512 to clearly describe shielded cable installation requirements.
05000263/FIN-2015001-012015Q1MonticelloFailure to Identify High Pressure Coolant Injection (HPCI) Seismic Support NonconformanceThe inspectors identified a finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to promptly identify conditions adverse to quality, such as deficiencies, deviations, and nonconformances. Specifically, on February 11, 2015, the inspectors identified a safety related seismic support for high pressure coolant injection (HPCI) turbine trip instrumentation that was not rigidly attached, supported, and restrained in accordance with plant construction code and installation specifications, a nonconformance which the licensee had failed to identify since initial plant construction. Corrective actions for this issue included repairs to the seismic support to rigidly connect the instrument line restraint and installation of a standalone support for the instrument tray. This issue was entered into the licensees corrective action program (CAP 1465906). The inspectors determined that the failure to promptly identify an HPCI instrument line support nonconformance was a performance deficiency requiring evaluation. The inspectors determined that the issue was more than minor because it adversely impacted the Mitigating Systems Cornerstone attribute of Protection Against External Factors, and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined that it was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, and the aspect of Identification because the licensee failed to implement a CAP with a low threshold for identifying issues (P.1).
05000263/FIN-2015001-052015Q1MonticelloTwo Emergency Diesels Inoperable Due to Human ErrorA self-revealing finding of very low safety significance and an NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was identified on December 28, 2014, due to the failure to properly implement Procedure 0187-02B, 12 Emergency Diesel Generator /12 ESW (Emergency Service Water) Monthly Pump and Valve Tests. Specifically, operations personnel failed to comply with Step 42 which directed the 12 EDG local governor control switch to be lowered to idle setting. The failure to implement the actions directed by Step 42 resulted in the 11 EDG being inoperable. Corrective actions for this issue included procedure revisions to require: protection/flagging of redundant equipment when technical specification equipment is declared inoperable for any reason, including planned maintenance and surveillance; peer checking or concurrent verification for manipulation of operable technical specification related equipment; and all equipment manipulations require a hard match (between procedure and equipment labeling). This issue was entered into the licensees corrective action program (CAP 1460675). The issue was more than minor because if left uncorrected, the failure to properly implement procedures associated with safety-related equipment would have the potential to lead to a more significant safety concern. Specifically, the failure to follow procedure resulted in the 11 EDG being made inoperable coincident with the 12 EDG being inoperable. The inspectors utilized IMC 0609 and determined that the issue was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Human Performance, Avoid Complacency aspect because of a failure of individuals to implement error reduction tools (H.12).
05000263/FIN-2015001-062015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.1, Offsite Dose Calculation Manual, (ODCM) which requires in part, that licensee initiated changes to the ODCM shall be effective after approval of the plant manager. Contrary to the above, ODCM01.01 Revision 6 and ODCM02.01 Revision 10, were not approved by the plant manager prior to implementation. This was identified by the licensee as part of the self-assessment process. The licensee documented this issue in the corrective action program (CAPs 1455999 and 1462092). This finding was determined to be of very-low safety significance (Green) because it was not a failure to implement an effluent program and public dose did not exceed Appendix I of 10 CFR 20.1301(e) criteria.
05000263/FIN-2015001-072015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Technical Specification 5.5.11 which requires in part, that the Primary Containment Leakage Rate Testing (LRT) Program shall be in accordance with the guidelines contained in RG 1.163, Performance-based Containment Leak-Test Program, dated September, 1995. RG 1.163 directs use of ANSI/ANS56.81994, Containment System Leakage Testing Requirements as an acceptable testing standard. ANSI/ANS56.81994 states, in part that for pressure decay testing, temperature shall be recorded at the start and end of each test, and the leakage rate shall be calculated using a specific formula which incorporates this temperature data to temperature-compensate the volume lost. Contrary to these requirements, the licensees Containment Leakage Rate Testing Program failed to include direction to take temperature data and perform temperature compensation, which resulted in a failure to perform testing in accordance with the ANSI standard and RG 1.163. Specifically, during this time, the licensee failed to correctly perform pressure decay testing for approximately 44 containment penetrations, including the Personnel Airlock. Upon discovery, engineers performed a bounding engineering analysis which verified the containment barrier remained operable but nonconforming and entered the issue into the corrective action program (CAPs 1463917 and 1465869). The performance deficiency was more than minor because the issue is associated with the barrier performance reliability attribute of the Barrier Integrity cornerstone and adversely affected the associated cornerstone objective to provide reasonable assurance that the physical containment barrier protects the public from radionuclide releases. Specifically, the repeated failure to ensure containment leakage testing met technical specification and regulatory requirements was programmatic, affected multiple components, adversely affected LRT test accuracy, and consequently impacted the licensees ability to verify the containment barrier remained operable. The finding was of very low safety significance because the finding did not represent an actual open pathway in the physical integrity of the containment barrier and did not result in a loss of containment barrier operability. (Green)
05000263/FIN-2015001-082015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. Criterion V which requires in part, that activities affecting quality be prescribed by procedures appropriate to the circumstances. Contrary to this requirement, between May 22, 2011 and February 5, 2014, MNGP startup instructions and procedures, C.1 Startup Procedure, 2167 Plant Startup, and 0118 Reactor Vessel Temperature Monitoring, were not appropriate to the circumstances. Specifically, during this time these procedures allowed reactor coolant system pressure to be decreased below 0 psig seven times during reactor startup activities, which was outside of the pressure parameter inputs to the analysis that is the basis for the pressure/temperature limit curves of TS 3.4.9. The licensees analysis showed that there was no impact on RPV integrity due to the existence of the partial vacuum conditions. This issue was identified by the licensee as a result of an operating experience review. The licensee entered this issue into the corrective action program (CAPs 1425020 and 1427529) and initiated action to revise the PTLR limits and submit them for NRC review. The performance deficiency was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Procedure QualityRoutine Operations Performance, and had the potential to adversely affect the associated cornerstone objective of providing reasonable assurance that a physical design barrier, the reactor coolant system, protects the public from radionuclide releases caused by accidents or events. The finding screened as very low safety significance because analysis determined that there was no change in risk to the RCS boundary due to the performance deficiency. (Green)
05000263/FIN-2015001-092015Q1MonticelloLicensee-Identified ViolationThe licensee identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.47(b)(14) and 10 CFR Part 50, Appendix E, Section IV.F.1. In part, Title 10 CFR 50.47(b)(14) states, Periodic exercises are (will be) conducted to evaluate major portions of emergency response capabilities, periodic drills are (will be) conducted to develop and maintain key skills, and deficiencies identified as a result of exercises or drills are (will be) corrected. Additionally, Title 10 CFR Part 50, Appendix E, Section IV.F.1 states, The program to provide for: (a) The training of employees and exercising, by periodic drills, of emergency plans to ensure that employees of the licensee are familiar with their specific emergency response duties, and (b) The participation in the training and drills by other persons whose assistance may be needed in the event of a radiological emergency shall be described. The Monticello Emergency Plan, Section 8.1.2.4, describes the required demonstration periodicity for drill and exercises. Contrary to the above, on January 1, 2015, the licensee failed to perform four emergency preparedness drill objectives at the required frequency listed in the Monticello Emergency Plan, Section 8.1.2.4. Specifically, Objectives 11.01, 11.03, and 11.04 were required to be performed annually and were not performed in 2014. Additionally, Objective 11.04 was required to be performed semi-annually and was only performed once in 2014. All missed objectives were associated with radiological exposure controls. The NRC determined that the failure to comply with the established drill and exercise program was a degradation of a planning standard function in accordance with 10 CFR 50.47(b)(14) and was a very low safety significance issue (Green) as indicated in IMC 0609, Emergency Preparedness SDP, Appendix B, Attachment 2, Failure to Comply Significance Logic. The licensee entered this issue in the corrective action program (CAP 1463920). As such, the NRC determined this to be an NCV in accordance with Section 2.3.2 of the Enforcement Policy.
05000440/FIN-2015001-012015Q1PerryFailure to Initiate A Transient Combustible PermitThe inspectors identified a finding of very low safety significance and associated NCV of Perry Operating License Condition 2.C(6) for failure to follow the site Fire Protection Program. Specifically, a large quantity of material from the previous space utilized as the Diesel Maintenance Shop had been placed in the Diesel Generator (DG) Hallway to allow reconstruction of the space as a storage area for post-Fukushima equipment and awaiting completion of a new maintenance shop location. However, as of the inspectors observations on February 3, 2015, the licensee failed to evaluate the impact of this large quantity of combustibles or to issue a transient combustible permit as required by Perry Administrative Procedure (PAP) 1910, Fire Control Program. This finding was entered into the licensees corrective action program for resolution as Condition Report 201501280 and immediate corrective action was taken to evaluate and issue a transient combustible permit for the DG Hallway. The failure to comply with the site Fire Protection Program was determined to be more than minor performance deficiency because it was associated with the protection against external factors (i.e., fire) attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to evaluate the fire impact of the stored material and process a permit for the excess combustible material stored in the DG Hallway fire area prevented the licensee from initiating compensatory fire watch actions, and additionally did not address the potential issue of restricting the availability of fire protection equipment in the area. The inspectors determined that the finding was of very low safety significance because the impact of a fire would have been limited to no more than one train of equipment important to safety. The finding has a cross-cutting aspect in the area of human performance, work management, in that the licensee work process did not provide for management of the risk commensurate to the work and the need for coordination with different groups or job activities, specifically fire safety personnel.
05000440/FIN-2015001-022015Q1PerryLiquid Penetrant Testing Procedure Was Not Qualified for Its Full Applicability RangeThe inspectors identified a finding of very low safety significance and associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to properly qualify a non-destructive testing procedure in accordance with applicable codes. Specifically, a liquid penetrant testing procedure was not qualified for its full applicability temperature range in accordance with American Society for Mechanical Engineers (ASME) Code, Section V, Non-Destructive Examination. This finding was entered into the licensees corrective action program as Condition Report 201503175. The failure to qualify a liquid penetrant testing procedure in accordance with ASME Section V was determined to be a more than minor performance deficiency because if left uncorrected, it has the potential to lead to a more significant safety concern. Specifically, since the liquid penetrant testing procedure was not qualified for its full applicability temperature range, liquid penetrant examinations would not be assured to detect flaws in the unqualified temperature range and as a consequence, the potential would exist for a rejectable flaw to go undetected, unknowingly impacting the operability of the inspected system. The inspectors determined the finding was of very low safety significance because it did not result in the loss of operability or functionality for any mitigating systems; thus, the inspectors answered No to the screening questions. The licensee completed a review of liquid penetrant examination records, and did not find an example where the procedure was implemented in the unqualified temperature range. The inspectors did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the inadequate qualifications were performed more than 3 years ago.
05000331/FIN-2015001-012015Q1Duane ArnoldFailure to Classify and Declare a Notification of Unusual EventThe inspectors identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) 50.54(q)(2), and 10 CFR 50.47(b)(4) for the failure of the licensee to classify and declare a Notification of Unusual Event. Specifically, on June 30, 2014, the licensee failed to classify and declare a Notification of Unusual Event after a control room instrument peaked at a wind speed that exceeded the Unusual Event Emergency Classification Level threshold for 4 seconds. The licensee entered the issue into the corrective action program (CAP) as condition report (CR) 01975495. Corrective actions included procedure changes to ensure available indications for wind speed are monitored during high wind events. The failure to classify and declare a Notice of Unusual Event when conditions warranted was a performance deficiency. The finding was more than minor because it adversely affected the emergency response organization (ERO) performance attribute of the Emergency Preparedness (EP) cornerstone objective to ensure that licensees are capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Because the finding only involved a failure to declare a Notification of Unusual Event, the finding screened as being of very low safety significance (Green). This finding was associated with the cross-cutting aspect of avoid complacency in the area of Human Performance, because control room operators did not walk-down instrumentation that was available to them in the control room. (H.12)
05000263/FIN-2015001-022015Q1MonticelloFailure to Maintain Fire Protection Program Procedures for Control of Portable Heater/Extension Cord Fire HazardsA finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1.d was self-revealed when the licensee failed to maintain procedures for Fire Protection Program Implementation to ensure that ignition sources (space heaters) were properly controlled to prevent plant fires. Specifically, on January 26, 2015, the licensee failed to maintain Fire Protection Program implementation procedures to include controls to ensure space heaters used in the plant stayed within allowable load ratings and were plugged directly into outlets without the use of extension cords. This resulted in a fire in the plant recombiner building which was extinguished within 13 minutes, nearing the 15 minute time limit at which a Notification of Unusual Event (NOUE) would have needed to be declared. It also resulted in a space heater causing an overloaded outlet at a location in the reactor building, near A residual heat removal (RHR) equipment. Upon discovery of the recombiner area fire, the licensee dispatched the fire brigade to ensure the fire was extinguished, performed extent of condition walkdowns in the plant, and took action to improve controls on extension cord and portable heater use in the power block. This issue was entered into the licensees corrective action program (CAP 1463506). The inspectors determined that the failure to maintain fire program procedures to ensure ignition sources (space heaters) were appropriately controlled was a performance deficiency requiring evaluation. The inspectors determined the issue was more than minor because, if left uncorrected, the failure to adequately control portable heater related fire hazards in the plant could lead to more significant safety concerns. In addition, the finding was more than minor because it was associated with the Initiating Events Cornerstone attribute of Protection Against External Factorsincluding fire, and affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors assessed the significance of this finding in accordance with IMC 0609 and determined that it was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Evaluation aspect because of the failure to thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000263/FIN-2015001-032015Q1MonticelloFailure to Maintain a Standard Emergency Action Level Scheme for FloodingThe inspectors identified a finding of very low safety significance and an NCV of Title 10 CFR 50.54(q)(2) and 10 CFR 50.47(b)(4) for the licensees failure to maintain the effectiveness of the emergency plan. Specifically, from May 28, 2014, until February 26, 2015, the HA1.6 Emergency Action Level (EAL) threshold was in conflict with the EAL basis for the alert classification. Additionally, both the revised EAL threshold and original NRC-approved safety evaluation report EAL threshold were later found to be greater than the actual river level that could lead to damage of safe shutdown equipment. The licensees corrective actions documented that the current river level was 906 and if flooding were to occur the licensee would have relied on Procedure A.6, "Acts of Nature," and that an event response team would have been formed to monitor river level during the duration of a flood event. The licensee concluded that the shift manager, Event Response team, and plant management would have monitored for indication of degraded performance of equipment or structures necessary for safe shutdown for event classification escalation to the Alert level. The licensee entered this issue into the Corrective Action Program (CAP 1454593). The inspectors determined that establishing a flooding EAL threshold that was in conflict with approved EAL basis as required by 10 CFR 50.47(b)(4), and subsequent failure to determine the actual level that could lead to damage of safe shutdown equipment for the alert classification High River Level EAL HA1.6 was a performance deficiency. The inspectors determined that the issue was more than minor because it is associated with the Procedure Quality attribute of the Emergency Preparedness (EP) cornerstone and adversely affected the cornerstone objective to ensure 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 assessed the significance of this finding in accordance with IMC 0609 and determined that it was of very low safety significance. The inspectors determined that the contributing cause that provided the most insight into the performance deficiency was associated with the cross-cutting area of Problem Identification and Resolution, Evaluation aspect because the licensee did not thoroughly evaluate the identified engineering error issue to ensure that resolutions address causes and extent of conditions commensurate with their safety significance (P.2).
05000331/FIN-2015001-022015Q1Duane ArnoldFailure to Report Required Monitoring Results to the NRCThe inspectors identified a Severity Level (SL) IV NCV of 10 CFR 20.2206 for the licensees failure to report results of individual radiation exposure monitoring for individuals required to be monitored by 10 CFR 20.1502. Specifically, on or before April 30, 2014, the licensee failed to report results for all individuals requiring monitoring for the calendar year 2013 to the NRCs Radiation Exposure Information and Reporting System (REIRS) database. The issue was entered into the licensees CAP as CR 02028468. Immediate corrective actions included the resubmittal of radiation exposure data to the REIRS database, which included radiation exposure for all individuals that were required to be monitored. The violation of 10 CFR 20.2206 was assessed in accordance with the traditional enforcement path in IMC 0612, Appendix B, Issue Screening. The inspectors determined that traditional enforcement did apply because reporting failures impact the regulatory process. In accordance with the NRC Enforcement Policy, Section 6.9(d)(2), failures to make a timely written report as required by 10 CFR 20.2206 are categorized as SL IV violations. Cross-cutting aspects are not assigned to traditional enforcement violations.
05000263/FIN-2015001-042015Q1MonticelloInadequate Evaluation of Operating Crew During Simulator AssessmentThe inspectors identified an URI on March 16, 2015, due to the licensees potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. In accordance with IMC 0612, Power Reactor Inspection Reports, the inspectors determined that this issue represented an URI because more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. On March 16, 2015, the NRC inspectors observed a potential failure to properly assess and critique a senior reactor operators performance during a simulator self-assessment in accordance with Procedure MTCP03.49, Conduct of Training Cycle Self-Assessments. Specifically, during an NRC observation of a Licensed Operator Training self-assessment and emergency preparedness objective demonstration, the inspector observed that the evaluators may not have adequately critiqued a knowledge deficiency in the Interpreting and Diagnosing Events competency area when evaluating a Shift Managers (SM) performance. The Shift Managers performance could have adversely impacted EAL classification during a graded self-assessment. This assessment included an evaluated Drill/Exercise Performance (DEP) opportunity for the EAL classification in question. During the inspectors observation, they noted that the critique session did not appear to adequately probe why the classification-related performance weaknesses occurred, and did not appear to determine a course of specific actions for the crew to take to improve individual performance relative to the SMs role in the EAL classification. Specifically, the inspectors noted that at the end of the critique, this item was not discussed as an item needing resolution, nor was it discussed that the SM had a challenge to his qualifications and needed potential remediation, which appeared to be contrary to the sites MTCP0349 procedure. These discussions and follow-up actions did not take place until after the critique had concluded and the NRC inspectors raised questions about the SMs misinterpretation of Safety Parameters Display System (SPDS) and his overall performance. This item represents an issue of concern about which more information is required to determine if a violation exists and if the performance deficiency is More-than-Minor. The NRC inspectors will work to obtain additional guidance and clarification/interpretation of the existing guidance in order to resolve this issue. Corrective actions for this issue included disqualifying the individual, developing a remediation plan, and initiating procedure changes to improve the critique process. This issue was entered into the corrective action program as CAP 1470975. (URI 05000263/201500104, Inadequate Evaluation of Operating Crew During Simulator Assessment)
05000282/FIN-2014005-032014Q4Prairie IslandFailure to Follow Procedures during EDG 24 Hour Load TestThe inspectors identified a finding of very low safety significance and a NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, on September 29, 2014, due to the licensees failure to follow procedure during the performance of SP 1335, D2 Diesel Generator 18 Month 24 Hour Load Test. Specifically, operations personnel failed to comply with steps within SP 1335 which directed that the emergency diesel generators (EDGs) kVAR loading be adjusted until a power factor of less than or equal to 0.85 was achieved or Bus 16 voltage was between 4350 and 4375 volts. An extent of condition review determined that operations personnel failed to comply with a similar procedure step during the 24 hour load test of the D1 EDG performed in May 2013. As a result, the licensee had to re-perform the tests, which resulted in additional EDG inoperability and unavailability. Corrective actions for this issue included training the operators on the need to maintain the power factor or bus voltage within limits during testing, requiring all data collected by the operations department during Technical Specification (TS) surveillance testing to be independently verified, and requiring all TS surveillance requirement results to be reviewed and approved by two senior reactor operators. The inspectors determined that this finding was more than minor because it was associated with the human performance attribute of the Mitigating Systems cornerstone and impacted the cornerstones objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operations personnel were required to declare the D1 and D2 EDGs inoperable and unavailable to perform their safety functions while the 24 hour load testing was re-performed. The inspectors concluded that this issue was of very low safety significance because each question provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, was answered No. This finding was cross-cutting in the Human Performance, Avoid Complacency area because operations personnel failed to implement appropriate error reduction tools to ensure that the power factor or bus voltage requirements were met during the surveillance test (H.12).
05000255/FIN-2014005-012014Q4PalisadesFailure to Follow Procedure for Storage of Equipment in the Vicinity of Safety-Related EquipmentThe inspectors identified a finding of very low safety significance (Green) with an associated non-citied violation of Technical Specification (TS) 5.4.1, Procedures and Programs, for the failure to follow site procedures covering the storage of material in the vicinity of safety-related equipment. Specifically, on three occasions the inspectors identified ladders at ladder station 42 in the 590 elevation of the component cooling water room that were either in contact with safety-related equipment or were capable of toppling into safety-related equipment. For immediate corrective actions, licensee personnel properly stored the ladder after each issue was identified by the inspectors. This issue is documented in the licensees corrective action program (CAP) as Condition Report CR-PLP-2015-00126. The performance deficiency was determined to be more than minor based on Inspection Manual Chapter (IMC) 0612, Appendix E, Example 4.a, which determined that low-level procedural errors without a safety consequence are more than minor when they become a repetitive/routine occurrence. Specifically, unrestrained ladders could impact safetyrelated equipment during a design basis seismic event. The inspectors evaluated the significance of the finding in accordance with IMC 0609, Attachment 4, Initial Characterization of Findings. In accordance with Table 2, the finding was determined to affect the Mitigating Systems Cornerstone. The inspectors answered No to the questions in Table 3 and continued the significance evaluation in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. The inspectors answered No to the Mitigating Systems Screening Questions contained in Exhibit 2 and determined the finding was of very low safety significance (Green). This finding was associated with a cross-cutting aspect of Identification in the Problem Identification and Resolution cross-cutting area (P1).
05000440/FIN-2014005-032014Q4PerryFailure to Follow Licensee Procedure to Properly Screen and Evaluate Temporary Changes to Plant Facilities / Structures, Systems, or ComponentsThe inspectors identified a finding of very low safety significance and associated NCV of Technical Specification 5.4.1.a, Procedures, for the licensees failure to implement the requirements of Nuclear Operating Business Practice (NOBP)LP4003A, FENOC 10 CFR 50.59 User Guidelines. This finding was entered into the licensees corrective action program for resolution as Condition Report 201500284. The inspectors determined that the failure to complete a Regulatory Applicability Determination (RAD) specified in NOBPLP4003A was a performance deficiency. The performance deficiency was more than minor, and thus a finding, because it was associated with the procedure quality attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was determined to have very low safety significance because the finding: (1) was not a design or qualification issue confirmed not to result in a loss of operability or functionality; (2) did not represent an actual loss of safety function and/or system; (3) did not result in the loss of one or more trains of TS equipment; and (4) does not represent the loss of a non-TS train of equipment. The finding has a cross-cutting aspect in the area of human performance associated with the change management component, in that leaders use a systematic process for evaluating and implementing change so that nuclear safety remains the overriding priority.
05000346/FIN-2014005-012014Q4Davis BesseAdditional Review of Medical Records NeededThe inspectors identified an URI concerning the auditable condition of the medical records and completeness of information that may affect the conditions required as part of a license operators license. The inspectors reviewed a sample of the licensed operator medical records during the inspection. The inspectors noted that the medical records were difficult to review because of the lack of a succinct filing within the records themselves. The inspectors found that two of the license operator medical files contained a change in information that was relevant to the license condition. One of the changes was not known by the licensee or the NRC during a period of time that the licensed operator was standing watch in a licensed position. The involved operators licenses are currently on administrative hold and the operators are not allowed to stan watch in license positions. As part of the biennial review inspection conducted during the week of December 8 through December 12, 2014, the inspectors reviewed a sample of the licensed operato medical records for compliance with 10 CFR Part 55. The inspectors reviewed seven medical records for accuracy and compliance with the license conditions. During the review, the inspectors noted that one licensed operator had failed to meet the conditions required on the license, had delayed in notifying the facility licensee, and had continued to stand watch. The inspectors also noted that another operator had failed to meet a condition of his license; but the license had previously been placed on administrative hold, and the operator had not stood watch during the failure to meet the license condition. The inspectors expanded the inspection sample and selected five additional records for review. None of the additional records were identified as containing a potential noncompliance. In total, 12 of the 58 operator license medical records at the site wer reviewed. The involved operators licenses are currently on administrative hold, and the operators are not allowed to stand watch in license positions. In reviewing the requirements of 10 CFR Part 55, the inspectors focused on the following: A person must be authorized by a license issued by the Commission to perform the function of an operator or a senior operator as defined in this part (10 CFR 55.3) The licensee shall comply with any other conditions that the Commission ma impose to protect health or to minimize danger to life or property. (10 CFR 55.53(l)) The inspectors determined that 2 of the 12 records reviewed (approximately 17 percent) had discrepancies which involved potential violations of 10 CFR 55. The records reviewed were found to have issues involving the auditable condition of the information contained in the records, and further inspection of the medical records should be performed to determine if additional discrepancies exist. (URI 05000346/201400501)
05000301/FIN-2014005-012014Q4Point BeachFailure to Promptly Correct a Failed Emergency Diesel Generator Day Tank Room HeaterA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified by the inspectors for the failure to promptly repair the non-functional HX272A, a safety-related room heater for the G04 Emergency Diesel Generator (EDG) day tank room. Specifically, HX272A was identified failed in June 2012 and was not corrected until November 2014 but not before inspectors identified that the redundant room heater, HX272B, had also failed and the room temperature had dropped below the design basis temperature of 50 degrees Fahrenheit. The licensee repaired HX272A on November 25, 2014 and also installed a thermometer in the fuel oil day tank room for operators to monitor room temperature. The licensee entered the issue into their CAP as action request (AR) 02018260 and AR 02008296. The inspectors determined that failing to promptly repair safety-related room heater, HX272A, G04 EDG day tank room heater was contrary to 10 CFR 50 Appendix B, Criterion XVI and was a performance deficiency. The inspectors determined that the finding was more than minor, because, if left uncorrected, it could have the potential to become a more significant safety concern. Specifically, the inspectors found both safety-related heaters non-functional in the fuel oil day tank room with outside air blowing into the room through a ventilation damper. The outside temperature was approximately 17 degrees Fahrenheit, and while the licensee determined that at the time their fuel oil cloud point was approximately zero degrees Fahrenheit, the licensees specification for fuel oil cloud point allowed for a fuel oil cloud point of up to 25 degrees Fahrenheit. Additionally, if the fuel oil day tank room temperatures dropped below freezing, the fire sprinkler piping within the room could have actuated and/or ruptured and adversely affected the safety-related fuel oil transfer pumps within the room. The inspectors determined the finding could be evaluated using the SDP in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2, Mitigating Systems Screening Questions, dated June 19, 2012. The inspectors concluded that the finding was of very low safety significance because the inspectors answered "No" to the Mitigating Systems screening questions. This finding has a cross-cutting aspect of Work Management (H.5), in the area of Human Performance, for failing to implement a process of planning, controlling, and executing work activities such that nuclear safety is an overriding priority.
05000346/FIN-2014005-032014Q4Davis BesseLicensee-Identified ViolationThe following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of Section 2.3.2 of the NRC Enforcement Policy for being dispositioned as an NCV. Wrong Channel Error Renders Safety Features Actuation System Channel 2 Inoperable During Testing on Channel 4 Appendix B of 10 CFR Part 50, Criterion V, Instructions, Procedures, Drawings requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Contrary to this requirement, on November 3, 2014, two reactor operators failed t adequately perform procedure DBSC03113, "SFAS (Safety Features Actuation System) Channel 4 Functional Test." Specifically, the operators errantly began performing the test procedure on SFAS Channel 2 instead of SFAS Channel 4, as the procedure required. During the performance of the initial procedures steps, the operators placed the SFAS Channel 2 test trip bypass switch in the "reactor coolant pressure" position. This resulted in the "reactor coolant system pressurelow" and the "reactor coolant system pressurelow low" functions for SFAS Channel 2 being rendere inoperable and an unplanned entry into TS 3.3.5, Condition A. Shortly thereafter, the operators recognized that they were performing their actions on the wrong SFAS channel, stopped all associated activities, and reported the error to the on-shift unit supervisor in the control room. The operators were relieved from all licensed duties, and all SFAS test activities were halted. Within the hour, control room personnel had returned the SFAS Channel 2 test trip bypass switch to its normal position and restored all functions for SFAS Channel 2 to an operable status. The DBSC03113 procedure was successfully performed on SFAS Channel 4 on November 8, 2014. The objective of the Mitigating Systems Cornerstone of Reactor Safety is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). A key attribute of this objective is human performance, and specifically, configuration control. In accordance with NRC IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, the inspectors determined that the violation was of more than minor significance in that it had a direct impact on this cornerstone objective. The licensees failure to complete DBSC03113 as written resulted in the unplanned inoperability of SFAS Channel 2 for the "reactor coolant system pressurelow" and the "reactor coolant syste pressurelow low" functions. The licensee had entered this issue into their CAP as CRs 201416542, 201416919, 201417011, and 201417037. A full apparent cause evaluation was performed and corrective actions included, but were not limited to: A lessons learned communication was provided to each operating crew prior to their next shift, and a site-wide human performance communication on the issue was developed and promulgated; The cabinet door key for SFAS Channel 4, which unintentionally was able to open the SFAS Channel 2 door, was replaced and this physical vulnerability removed; and As part of the extent of condition, it was validated that the keys for other similar safety, control, and instrumentation cabinets in the control room functioned onl in their proper and respective door locks.
05000266/FIN-2014005-022014Q4Point BeachLicensee-Identified ViolationThe licensee identified a NCV of TS 5.4.1, Procedures for the failure to follow the defined heavy load shipping path inside containment as specified in procedure, SLP1, Safe Load Path and Rigging Manual, which resulted in the movement of the polar crane main block over exposed reactor fuel. The licensees TS 5.4.1 required, in part, that written procedures shall be implemented covering refueling activities. The licensees refueling procedure governing the movement of the unit 1 containment crane was SLP1, which described the predefined safe load travel paths and laydown areas in containment during refueling operations that have been pre-analyzed per the FSAR and NUREG0612. Procedure SLP1 stated that the main load block of the polar crane was considered a heavy load because it is not single failure proof and weighed approximately 8,550 pounds; and therefore, shall not be moved over the reactor vessel when the head is removed and fuel is in the vessel, with the exception to lift the vessel internals. Contrary to the above, on October 11, 2014, while unit 1 was in mode 6 with the reactor vessel head removed, the cavity flooded in excess of 23 feet, and irradiated fuel in the reactor vessel during defueling; the licensee moved the main load block of the polar crane over the reactor vessel during the performance of daily crane checks. The licensee entered this issue into the CAP as AR 01998150 and AR 02020076. The inspectors determined that this issue was of very low safety significance (Green) after reviewing IMC 0609, Appendix G, Shutdown Operations Significance Determination Process, Attachment 1, dated May 9, 2014. The inspectors answered "No" to all questions in Exhibit 2 for Initiating Events. Therefore the finding screened as Green (very low safety significance).
05000331/FIN-2014005-072014Q4Duane ArnoldIneffective Radiological Engineering Controls Resulted in Unplanned and Unintended Radiological Intakes to WorkersA finding of very-low-safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Section 20.1701 was self-revealed during work activities associated with the failure to implement effective radiological engineering controls during reactor pressure vessel (RPV) disassembly that resulted in personal contaminations and unplanned and unintended radiological intakes to workers. Specifically, on October 5, 2014, several individuals working on the refuel floor were contaminated and several received small intakes of radioactive material while venting the RPV head. The licensee entered the issue into the Corrective Action Program as condition report 01996216. Corrective actions included revising applicable procedures for RPV flood-up with the RPV vented to atmosphere on the refuel floor. The finding was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected th cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, the failure to implement effective radiological engineering controls during RPV disassembly resulted in personal contaminations and low dose intakes to several workers. The inspectors also concluded that the radiological hazards had the potential to result in higher exposures to the individuals had the circumstances been slightly altered. The finding was determined to be of very-lowsafety significance because it was not an ALARA planning issue; there was neither overexposure nor a substantial potential for an overexposure; and the licensees ability to assess dose was not compromised. This finding was associated with the crosscutting aspect of operating experience in the area of Problem Identification and Resolution because the licensee did not systematically implement relevant external operating experience in a timely manner. (P.5)
05000346/FIN-2014005-022014Q4Davis BesseInadequate Procedural Guidance During Restoration From Valve Maintenance Results in Feedwater Heater System and Plant Power TransientA self-revealed finding of very low safety significance and associated NCV of Technical Specification (TS) 5.4.1(a) were identified when the licensee failed to provide proper procedural guidance for the restoration from valve maintenance on HD291G, a manual isolation valve for the level controller for HD291A, the emergency drain valve for High Pressure (HP) Feedwater Heater No. 14, on November 13, 2014. Specifically, the licensee's restoration instructions did not isolate HD291A prior to restoring its associated level controller. As a result, when a perturbation in the level controller during restoration caused HD291A to rapidly reposition to the fully open position, the resulting HP Feedwater Train 1 transient caused HP Feedwater Heaters 14, 15, and 16 to trip The change in plant efficiency that resulted momentarily drove plant power slightly above 100 percent This finding was associated with the Initiating Events Cornerstone of reactor safety and was of more than minor significance because it directly impacted the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The inspectors evaluated the finding using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power. Using Exhibit 1, the inspectors determined that the finding screened as very low safety significance because all screening questions for the Initiating Events Cornerstone of reactor safety were answered No. This finding als was determined to have a cross-cutting component in the area of human performance, work management aspect, because during the work planning process for this maintenance activity the licensee failed to identify the risk associated with not isolating the HP Feedwater Heater No. 14 Emergency Drain Valve, HD291A, prior to restoring its associated level controller to service.
05000331/FIN-2014005-012014Q4Duane ArnoldConstruction Code Used during a Replacement Activity Not Reconciled with the Owner's RequirementsA finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Section 50.55a, Codes and Standards, was identified by the inspectors for the failure to reconcile the construction code and owners requirements when replacing rod hangers associated with the high pressure coolant injection (HPCI) system. The licensee subsequently performed a code reconciliation and concluded the applicable construction code requirements were met. The licensee captured this issue in its Corrective Action Program as condition report 01999594. The performance deficiency 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 HPCI to respond to initiating events to prevent undesirable consequences. Specifically, the failure to reconcile the construction code and owners requirements when replacing HPCI support rod hangers reduced confidence in the systems capability to meet its mitigating function consistent with its design basis. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. This finding had a cross-cutting aspect of procedure adherence in the area of Human Performance because the licensee failed to follow American Society for Mechanical Engineers Section XI, Administrative Manual for Repair, Replacement, and Modification. (H.8)
05000331/FIN-2014005-022014Q4Duane ArnoldLiquid Penetrant Testing Procedures Not Qualified for their Full Applicability RangeA finding of very low safety significance (Green) and an associated non-cited violation of Title 10 of the Code of Federal Regulation, Part 50, Appendix B, Criterion IX, Control of Special Processes, was identified by the inspectors for the failure to properly qualify nondestructive testing procedures in accordance with applicable codes. Specifically, liquid penetrant testing procedures were not qualified for their full applicability temperature ranges in accordance with American Society for Mechanical Engineers (ASME) Code, Section V, Nondestructive Examination. The licensee entered this issue into the Corrective Action Program as condition report 01950601 and 01999596. As an immediate corrective action, the licensee reviewed completed liquid penetrant examination records and did not find an example where the procedures were implemented at the unqualified temperature range. The performance deficiency was determined to be more than minor because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, since the liquid penetrant testing procedures were not qualified for their full applicability temperature ranges, liquid penetrant examinations were not assured to detect flaws in the unqualified temperature ranges. As a consequence, the potential would exist for a rejectable flaw to go undetected affecting the operability of the affected system. This finding affected the Initiating Event, Mitigating System, and Barrier Integrity cornerstones. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality. The inspectors did not identify a cross-cutting aspect associated with this finding because the inadequate qualifications were performed more than three years ago and was not confirmed to reflect current performance.
05000331/FIN-2014005-042014Q4Duane ArnoldFailure to Accomplish Procedure for Leaking Pipe SnubberA finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to accomplish procedure EN-AA-203-1001, Operability Determinations/Functionality Assessments. Specifically, on May 8, 2014, the licensee failed to properly evaluate functionality of a leaking pipe snubber associated with the A core spray subsystem, the resultant operability impact on the Technical Specification affected systems, and the extent of condition. The licensee entered the inspectors concerns into the Corrective Action Program as condition report 02003867 and 02010686. Corrective actions included coaching/training of licensed operators during requalification training and management review committee members, and changes to applicable snubber program procedures. The performance deficiency was determined to be more than minor because it impacted the Mitigating System cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Because the finding did not involve the total loss of any safety function, the finding screened as very low safety significance (Green). This finding was associated with the cross-cutting aspect of consistent process in the area of Human Performance, because the licensees inconsistent application of the systematic operability/functionality determination process to evaluate the leaking snubber led to prolonged exposure of the extent of cause that affected several safety-related systems. (H.13)
05000440/FIN-2014005-042014Q4PerryFailure to Follow Procedures During Dry Cask OperationsThe inspectors identified a Severity Level IV NCV of very low safety significance of 10 CFR Part 72.150, Instructions, Procedures, and Drawings, for the licensees failure to follow procedures important to safety during dry cask operations. The licensee entered each example identified into its corrective action program as Condition Reports 201411637 and 201414279. The violation was determined to be more than minor in that both examples identified deficiencies in the performance of dry cask operations important to safety. In this determination, the inspectors considered example 4.a in IMC 0612, Appendix E, Examples of Minor Issues, dated August 11, 2009, and concluded that, while the errors did not result in any actual safety concern, there were multiple examples of procedural non-compliance. Additionally, if left uncorrected, a failed weld could lead to a release of radioactive materials to the environment and a malfunction of the Fuel Handling Building crane could lead to a more significant safety concern such as a load drop. The significance of the violation was found to be similar to SLIV example 6.5.d.3, of the NRCs Enforcement Policy, in that the licensee failed to adequately implement Quality Assurance processes or procedures. The issue was not found to be similar to any examples of higher significance; as such, the violation screened as a SLIV violation. Since traditional enforcement was used to disposition the violation, a cross-cutting aspect is not applicable.
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