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05000254/FIN-2018010-01Minor Violation2018Q3On March 13, 2013, the licensee initiated AR 1487225 to document and evaluate an installed weld nonconformance that attached core spray keep fill line 114342LX to pipe support M983FH1. The licensees immediate evaluation of the nonconforming weld documented that the pipe support would perform its function of restraining the pipe for all loading conditions. Although the inspectors concluded the immediate evaluation provided a reasonable expectation the pipe support would perform its function of restraining the pipe, the inspectors noted the licensee did not provide a more detailed evaluation of the nonconformance using acceptance guidance per procedure OPAA108115, Operability Determinations (CM-1). As a result of the inspectors inquiry, the licensee initiated AR 417429, performed a more detailed operability evaluation in EC 625648, and concluded both the piping and pipe support would be able to meet design allowable stress limits with the nonconforming weld configuration. The inspectors reviewed the current design calculation for pipe support M983FH1, analysis 27.0200.1053.019.031 and performed a field walkdown of the installed piping and pipe support configuration for core spray keep fill line 114342LX to verify the adequacy of information provided used in EC 625648. In addition, licensee procedure OPAA108115 also required corrective action at the next opportunity, normally the next refueling outage, with a provision for deferral with proper documented justification. As a result of inspector inquiry regarding the timeliness of the corrective action, the licensee initiated AR 4177210 which documented the nonconformance repair was deferred from Q1R23 (March 2015) and Q1R24 (March 2017) without documented justification. The licensee plans to repair the nonconforming weld in the upcoming Q1R25 (March 2019) outage. The inspectors determined that this is a minor violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, related to the licensees evaluation and timeliness of corrective action for a safety-related pipe support nonconformance. Screening: The inspectors used Inspection Manual Chapter 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009 and determined that timeliness of corrective action and the lack of a detailed operability evaluation were minor issues. Specifically, the inspectors compared the weld nonconformance to a calculation error in Example 3a of Appendix E and concluded the issue was minor because licensee EC 625648 provided reasonable justification the nonconforming weld configuration will meet design allowable stress for all loading conditions without modification. Violation: This failure to comply with 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000254/FIN-2018003-02Failure to Follow Procedures for Forced Helium Dehydration of a Multipurpose Canister2018Q3The inspectors identified a Severity Level IV NCV of 10 CFR 72.150 when the licensee failed to follow procedures for the setup of the MPC FHD system. Specifically, during the setup for processing MPCs during the 2018 ISFSI loading campaign, the licensee failed to follow procedure OUMW671200, MPC Processing FHD for BWRs, Revision 1, Attachment 9, Step 1.2.1,which connected inlet and outlet hosing between the FHD skid and FHD manifold.
05000254/FIN-2018003-01Failure to Maintain the Design Basis for Residual Heat Removal Torus Suction Valve2018Q3The inspectors identified a Green finding and associated Non-Cited Violation(NCV) of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion III, Design Control, when the licensee performed an in-field adjustment to the torque switch settings on RHR torus suction valve 110017C and failed to ensure measures were established to assure the valve could continue to meet its design basis requirements.
05000254/FIN-2018010-02Minor Violation2018Q3In 2009, due to a control room envelope differential pressure test failure at another nuclear station, the licensee completed an engineering change to develop separate correction factors (uncertainty) for different test methods. The engineering change recommended procedure QCOS 575016, Control Room Envelope DP (Differential Pressure) Surveillance, be revised to perform additional testing using alternate test methods with reduced correction factors. However, the procedure was not revised. In January 2016, during the control room envelope differential pressure test, seven areas failed the acceptance criteria. The licensee utilized the engineering change and performed a temporary procedure change to use a different method with a reduced correction factor for testing. The test was successfully performed and acceptance criteria were met. After the test, the procedure reverted to the old method. A condition report was written regarding this issue and actions were assigned to perform repair of the control room boundary and determine if a more accurate instrument is needed. Although repairs were made to the control room boundary, the licensee has not yet determined if a more accurate instrument is needed. Also, the procedure was not revised to use the alternate methods. The inspectors determined this is a minor violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings for not having procedure appropriate to the circumstances. Screening: The inspectors determined this issue is not more than minor because the existing procedure is not incorrect but missing the steps the licensee could take when unsatisfactory results are obtained. Violation: This failure to comply with 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000254/FIN-2018002-01Failure to Have a Procedure Appropriate to Circumstances for Degraded Voltage Relays2018Q2A finding of very low safety significance (Green) and a Non-Cited Violation of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed on April 16, 2018, for the licensees failure to establish a preventive maintenance procedure for the safety-related degraded voltage relays that was appropriate to the circumstances. Specifically, the licensee failed to ensure that the first-time functional test and calibration for relay 1274B241 (Procedure MAQC773524, Quad Cities NOAD Unit 2 Tech Spec Undervoltage Relay and Degraded Voltage Relay Calibration) was at an appropriate frequency to ensure that the relay would perform its Technical Specification function.
05000254/FIN-2018001-04Enforcement Action: EA18021: EDG Non-conformance for Tornado Missiles (EGM 15002)2018Q1On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015, (ML15111A269) and revised on February 7, 2017 (ML16355A286). The EGM applies specifically to a structure, system, and component (SSC) that is determined to be inoperable for tornado-generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Quad Cities Nuclear Generating Station, the EGM provided for enforcement discretion of up to 3 years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the limiting conditions for operation (LCO), initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain in place until permanent repairs are completed or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. In 1967, the NRC issued general design criterion to which the Quad Cities Nuclear Generating Station was evaluated against. Quad Cities Updated Final Safety Analysis Report (UFSAR), Section 3.1, Conformance with NRC General Design Criteria, discusses this criterion and its applicability to the sites design. Specifically, UFSAR Section 3.1.1.2, Criterion 2Performance Standards, states, those systems and components essential to the prevention of accidents or to mitigation of their consequences shall be designed, fabricated, and erected to performance standards that will enable the facility to withstand, without loss of the capability to protect the public, the additional forces that might be imposed by natural phenomena such as earthquakes, tornadoes, flooding conditions, winds, ice, and other local site effects. Section 3.1.1.2 further states that plant equipment which is important to safety is designed to permit safe plant operation and to accommodate all design basis accidents for all appropriate environmental phenomena at the site without loss of their capability. On March 1, 2018, during an engineering review of the Quad Cities, Units 1 and 2 facility design, the licensee identified a nonconforming condition with the aforementioned general design criterion. Specifically, the licensee identified that the three EDG systems intake stacks, exhaust stacks, fuel oil storage tank vent lines, and diesel oil day tank vent lines were inadequately protected against tornado missiles. As a result of the nonconforming condition, the licensee declared the Units 1, 2, and 12 EDG systems inoperable and entered the Technical Specifications (TS) LCO required action statements. The condition was reported to the NRC in Event Notice 53235 as an unanalyzed condition and a condition that could have prevented fulfillment of a safety function. Corrective Actions: The licensee documented the inoperability and functionality of the affected SSCs and the applicable TS LCO action statements in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002 and documented the implementation of the compensatory measures to establish the SSCs as operable but nonconforming prior to expiration of the required LCO action statements. The licensees initial (and final) compensatory measures included: verification that procedures and training for a tornado watch or warning were in place to provide additional instructions for operators to respond in the event of tornados or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and emergency response procedures; establishment of a heightened level of station awareness and preparedness relative to identifying tornado missile vulnerabilities; and revision to procedure QCOA 001010, Tornado Watch-Warning, Severe Thunderstorm Warning, or Severe Winds, to include guidance for unobstructing and/or repairing crimped diesel fuel oil tank vent lines. Corrective Action References: IR 1281009: Tornado Missile Protection Unresolved Item and IR 4110003: EDG Non-Conformance for Tornado Missiles Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3: General Shutdown LCO (cascading or by reference from other LCOs); and TS 3.8.1: AC SourcesOperating. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies, was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS Accession No. ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee reviewed their initial compensatory measures to determine if more comprehensive compensatory measures were warranted. Upon their review, the licensee concluded that their initial compensatory measures were sufficient to satisfy both the short-term and long-term actions required by the EGM and therefore no additional actions were necessary for enforcement discretion. The disposition of this enforcement discretion closes URI05000254/201100904; 05000265/ 201100904: Tornado Missile Protection of the Emergency Diesel Generator Air Intake and Exhaust.
05000254/FIN-2018001-03Half Scram Due to Low Voltage on 24/48 Vdc System2018Q1A finding of very low safety significance (Green) and a Non-Cited Violation of Technical Specification 5.4.1, Procedures, was self-revealed on January 11, 2018, for the licensees failure to perform an equalizing charge on the Unit 1B 24/48 Vdc battery prior to returning the 24/48 Vdc battery to a normal configuration following a test discharge, which was required by station procedures. The failure to follow procedures led to a low voltage condition and caused a Unit 1B channel half scram in the reactor protection system.
05000265/FIN-2018001-02Failure to Establish Design Standard for Unit 2 Residual Heat Removal Service Water Pumps2018Q1The inspectors identified a finding of very low safety significance (Green) and a Non-Cited Violation of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the design bases standard and other requirements necessary to assure adequate quality were included in the design documents for the Unit 2 residual heat removal service water pumps. Consequently, the licensee failed to ensure the Unit 2 pumps were designed and constructed in accordance with the Standards of the Hydraulic Institute as identified in the Updated Final Safety Analysis Report.
05000254/FIN-2018001-01Repeat Use of Written Exams During Licensed Operator Requalification Examinations2018Q1The inspectors identified a Severity Level IV Non-Cited Violation of 10 CFR 55.49, Integrity of Examinations and Tests, due to the licensee engaging in an activity that compromised the integrity of an examination. Specifically, the Quad Cities 2015 Licensed Operator Requalification (LOR) written examinations were duplicated from the 2013 LOR written examinations, the 2017 LOR written examinations were duplicated from the 2015 LOR examinations, and four individuals were administered the same written examinations from the previous requalification examination cycle.
05000254/FIN-2017004-01Repeat Use of Written Exams during Licensed Operator Requalification Examinations2017Q4a. 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)
05000254/FIN-2017002-02Failure to Ensure Two Low Pressure ECCS Systems Operable in MODE 42017Q2The inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specification (TS) 3.0.1 on April 12, 2017,for the licensees failure to meet TS Limiting Condition for Operation (LCO) 3.5.2, Emergency Core Cooling Systems (ECCS)Shutdown. Specifically, on April 12, 2017, the licensee failed to ensure two low pressure ECCS subsystems were operable in Mode 4 in accordance with TS LCO 3.5.2 and failed to verify the LCO action conditions were met. Immediate corrective actions included restoring the 1A core spray pump to an operable status within 4 hours in order to comply with TS 3.5.2. This issue was entered into the licensees CAP as IR 3997127.The performance deficiency was determined to be more than minor, and a finding, because it impacted the Mitigating Systems Cornerstone attribute of Equipment Performance 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 finding was screened using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, against the questions in Exhibit 3, Mitigating Systems Screening Questions. The inspectors answered No to all of the questions and determined the finding could be screened as very low safety significance. The inspectors determined this finding affected the cross-cutting aspect of Human Performance, in the aspect of Work Management, because the licensee failed to ensure proper controls were in place while performing multiple activities which rendered multiple low pressure ECCS systems inoperable. In addition, the licensee failed to identify and manage the risk associated with performing multiple evolutions concurrentlyso that TS LCO 3.5.2 would be met and the required actions taken as necessary (H.5).
05000254/FIN-2017002-03Failure to Justify MSIV Maintenance Deferral2017Q2The inspectors identified a finding of very low safety significance for the licensees failure to provide an adequate technical justification for deferral of a preventative maintenance task to replace or refurbish the Unit 1 2D main steam isolation valve (MSIV) in accordance with WCAA120, Preventive Maintenance (PM) Database Revision Requirements. Specifically, overhaul or replacement of the 2D MSIV was deferred despite the historical performance of the valve, the as-found test results during Q1R24, and the amount of time that was available to plan for the overhaul to meet the maintenance strategy requirement of every seventh outage. Corrective actions for this issue included the licensee scheduling replacement of the Unit 1 2D MSIV during the next scheduled refueling outage (RFO). This issue was captured in the licensees corrective action program (CAP) as Issue Report (IR) 4017529.The inspectors determined the performance deficiency was more than minor because it was associated with the Initiating Events Cornerstone attribute of Equipment Performance and impacted the cornerstone objective because the MSIV preventive maintenance overhaul/replacement frequency was not effectively managed to ensure the reliability of the MSIV closure time performance to meet Technical Specification (TS) requirements on a consistent basis. The inspectors determined the finding could be evaluated using IMC 0609, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, Exhibit 1, for the Initiating Events screening questions and determined the finding was of very low safety significance. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, which states, Individuals use the decision making practices that emphasize prudent choices over those that are simply allowable (H.14).
05000254/FIN-2017002-04Failure to have Adequate Guidance in the Fire/Explosion Response Procedure2017Q2The inspectors identified a finding of very low safety significance and an associated non-cited violation of TS Section 5.4.1.c, Procedures, for the licensees failure to establish and maintain the fire response procedure. Specifically, Procedure QCOA 001012 Fire/Explosion, Revision 47, failed to provide adequate instructions to ensure that the reactor core isolation cooling (RCIC) system would not be potentially affected by a single spurious operation of any of its associated valves in the event of a fire in Fire Area TBII. The licensee entered the issue into their CAP as IR 2595878 and planned to revise the affected procedures.The performance deficiency was determined to be more-than-minor because it impacted the Mitigating Systems Cornerstone attribute of Protection Against External Events (Fire), and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the lack of adequate procedural guidance in the fire response procedure did not ensure a single spurious operation would not potentially impair the operation of RCIC system in the event of a fire in TBII. The finding was screened using IMC 0609, Appendix F, Attachment 1, Fire Protection Significance Determination Process Worksheet. The inspectors determined the finding required a detailed risk evaluation by a Senior Reactor Analyst. The finding screened as very low safety significance because the calculated total Delta Core Damage Frequency (CDF) was 9.5E7/yr per the detailed risk evaluation. 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.
05000254/FIN-2017002-01Failure to Establish aProcedure Appropriate for Calibration of RCIC Governor2017Q2A finding of very low safety significance and an associated non-cited violation of Technical Specification (TS) Section 5.4.1 was self-revealed for the licensees failure to establish a procedure as governed by Regulatory Guide 1.33, Revision 2, Appendix A that was appropriate for performing adjustments to the governing control system for the Unit 1 reactor core isolation cooling (RCIC) system. Specifically, on April 14, 2017, the licensee failed to ensure procedure QCIPM 130004, RCIC Woodward Governor EGM Control Box and Ramp Generator/Signal Convertor in Field Calibration, was appropriate for the accurate calibration of the RCIC system turbine governor actuator such that the system would be capable supplying its TS required flowrate of 400 gallons per minute (gpm). Immediate corrective actions included the licensee declaring the Unit 1 RCIC system inoperable and performing required calibrations at normal operating temperatures and pressures. Additional corrective actions included the licensee making procedural revisions to QCIPM 130004 to include specific guidance on performing turbine governor calibration adjustments and providing training to maintenance control system technicians on performing the procedure tasks and other related tasks that led to the inadequate adjustment. The issue was entered into the licensees CAP as IR 3998478.The performance deficiency was determined to be more than minor, and a finding,because it impacted the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective because the failure to properly calibrate the RCIC governor led to the system becoming inoperable. The inspectors determined the finding could be evaluated using IMC 0609, Appendix A, Significance Determination Process (SDP) for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, and determined that the finding required a detailed risk evaluation by a senior reactor analyst (SRA) because it resulted in the loss of the RCIC system function. A SRA performed a detailed risk evaluation of the performance deficiency using the Quad Cities SPAR Model and determined the total Delta Core Damage Frequency (CDF) was 7E9 (Green). The inspectors determined this finding affected the cross-cutting area of Human Performance, in the aspect of Training, because the licensee failed to ensure the technicians performing the calibration understood null voltage adjustments to the RCIC turbine governor could only be performed when the system was at a specified rated speed and pressure (H.9).
05000254/FIN-2017001-01Failure to Ensure Hardware Secure for Breaker MOC Switch Linkage2017Q1Green . A finding of very low safety significance and an associated NCV of 10 CFR 50, Appendix B, Criterion V was self -revealed on January 27, 2017, when the Unit 1C residual heat removal service water (RHRSW ) pump was started for a routine surveillance evolution and all expected annunciators and equipment failed to operate properly, which led to the licensee declaring the Unit 1C RHRSW pump inoperable. Specifically, t he licensee failed to establish a procedure for the mechanism operated contact (MOC) switch linkage arm that was appropriate to the circumstances to ensure the component would c ontinue to perform its function. Immediate corrective actions included reconnecting the MOC switch linkage arm assembly and test ing it by starting the 1C RHRSW pump prior to declaring the pump operable. In addition, the licensee planned procedure revisions to QCEPM 0200 11 that would specify a torque value to ensure the MOC switch linkage arm was adequately secured and could perform its function. Th is issue was entered into the licensees corrective action program as Issue Report 3967424 . The finding was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to ensure the MOC switch linkage arm was adequately fastened led to the failure of the component and its associated Unit 1C RHRSW pump d uring breaker operation on January 27, 2017. T he finding was determined to be of very low safety significance (Green), because the inspectors answered No to all of the questions in IMC 0609, Appendix A, The Significance Determination Process for Findings at Power , Exhibit 2, Mitigating Systems Screening Questions, Section A, Mitigating SSCs and Functionality. The inspectors determined this finding affected the cross- cutting area of human performance, in the aspect of avoid complacency, which state s, Individuals recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to recognize a potential risk and inherent latent issue for a condition identified in 2015 at Quad Cities, when a MOC switch failed to perform its function due to a missing nut in a different breakers linkage assembly. The licensee identified and corrected the 3 condition but failed to evaluate the cause of the missing nut because it did not impact the operability of the component . I n the 2015 instance, the MOC switch issue only affected indications for the component and had no adverse impact on the ability of the component to perform its function (H.12 ).
05000254/FIN-2016004-01Failure to Implement Foreign Materials Exclusion Controls2016Q4A finding of very low safety significance and an associated non-cited violation of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to implement foreign material exclusion (FME) controls during the implementation of modification Work Order 1649339, Modify the Target Rock to Increase the Volume per Engineering Change 394119, and was contrary to MAAA716008, Foreign Material Exclusion Program, Revision 9. The failure to implement FME controls during maintenance led to the failure of the Unit 2 Target Rock safety relief valve solenoid valve during surveillance testing on April 5, 2016. The licensees corrective actions included replacing the Target Rock safety relief valve solenoid valve. In addition, the licensee made procedure revisions to the standard template for welding activities to ensure that a FME plan is developed when performing butt welds or weld repairs. The licensee entered this issue into their corrective action program as Issue Report 2703233. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance. The inspectors determined the finding represented a potential loss of the valve function and, therefore, a detailed risk evaluation was required. A regional senior risk analyst performed a detailed risk evaluation and determined the finding was of very low safety significance. This finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety was an overriding priority. Specifically, during the implementation of Work Order 1649339 and subsequent revisions, the licensee failed to control and execute the work while following FME processes and procedures (H.5).
05000254/FIN-2016008-01Failure to Provide Appropriate Operating Instructions for Aligning a Battery Charger to the Station Black-Out Diesel Generator2016Q3A finding of very-low safety significance (Green) and an associated NCV of Technical Specification 5.4.1.a, Procedures, was self-revealed on December 2, 2014 when procedural guidance failed to be implemented as written. Specifically, Procedure QCOA 6100-17, Revision 12, Loss of SBO (Station Black-Out Normal 13.8kV Transformer T42R-6 Feed to 4kV Bus 61 and 71, included inappropriate guidance to cross-tie Bus 61 and Bus 71. The licensees procedural guidance as written were technically infeasible and could not be implemented due to breaker interlocks caused by the digital control system interface that precluded the 4kV buses 61 and 71 from being cross-tied. The licensee entered this finding into their Corrective Action Program as Issue Report 2487426 and Issue Report 2706435 and removed the guidance to cross-tie the 4KV buses from the procedure. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating System cornerstone attribute of design control and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of any structure, system, or component. Specifically, using other procedural guidance, operators were able to start both station black-out diesels within the hour. The inspectors did not assign a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance.
05000254/FIN-2016008-02Failure to Evaluate the Target Rock Relief Valve Accumulator per ASME Code2016Q3The inspectors identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, for licensees failure to assure that quality standards for the Target Rock Relief valve accumulator were specified and included in the design documents and that deviations were identified and controlled. Specifically, Engineering Change (EC 394119) fabricated the replacement Unit 2 Target Rock valve accumulator to American National Standard B31.1 Power Piping code requirements instead of the applicable American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section VIII requirements, without adequate justifications. The licensee captured this issue in their Corrective Action Program as IR 02708406 to evaluate the appropriate corrective actions and revise documentation as required. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating System cornerstone attribute of design control and adversely affected the cornerstone objective to ensure the availability, and reliability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of any affected structure, system, or components. This finding has a cross-cutting aspect in the area of Human Performance in the area of Design Margin because the licensee failed to maintain equipment within its design margins.
05000265/FIN-2016003-01Licensee-Identified Violation2016Q3Technical Specification 5.7.2, states, in part, that each high-radiation area, accessible to personnel with radiation levels > 1000 mrem/hr at 30 cm (12 in.) from the radiation source or from any surface which the radiation penetrates shall have doors that are locked to prevent unauthorized entry. Contrary to the above, on April 26, 2016, the licensee identified the locking mechanism for a door was non-functional and could not prevent unauthorized entry to the area. Specifically, a worker intentionally challenged the locking mechanism to the Unit 2 low pressure heater bay door when the latch opened. The individual left the area and later reported the issue to the radiation protection staff that promptly secured the area with an alternate locking mechanism and determined that the dose rates exceeded 1000 mrem/hr inside the area. The licensee documented this issue in Issue Report 2661096 and reported a PI occurrence under the Occupational Radiation Safety Cornerstone. The inspectors determined that this issue was of very low safety-significance (Green) after reviewing IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The inspectors determined that it was not an as-low-as-reasonably-achievable planning issue, there was neither overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. Therefore, the finding screened as Green (very-low safety significance).
05000254/FIN-2016002-01Failure to Maintain Parameters within Limits of TS 3.6.2.5 and 3.6.3.12016Q2The inspectors identified a finding of very low safety significance and an associated non-cited violation of Technical Specifications (TS) 3.6.2.5 and 3.6.3.1 for the licensees failure to take actions required by TS 3.6.2.5 and 3.6.3.1. Specifically, on May 25, 2016, the licensee failed to restore the Unit 2 drywell-to-suppression chamber differential pressure and primary containment oxygen concentration to within the TS specified limits, or reduce power below 15 percent rated thermal power as required by TS 3.6.2.5 and 3.6.3.1, Conditions A and B. The licensees corrective actions included restoring both parameters to within their specified limits on May 25, 2016. The violation was entered into the licensees corrective action program as Issue Report 2677621. The performance deficiency was determined to be more than minor and a finding because, if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to maintain drywell-to-suppression chamber differential pressure and primary containment oxygen concentration within their specified limits had the potential to lead to stresses that could challenge the structural integrity of the containment and/or lead to a combustible mixture inside the Unit 2 drywell following a loss of coolant accident, either of which could have challenged the assumptions of the safety analyses. The finding was screened against the Barrier Integrity Cornerstone and was determined to be of very low safety significance by the Region III senior reactor analyst using the insights from IMC 0609, Appendix H, Containment Integrity Significance Determination Process, Table 6.2, Phase 2 Risk SignificanceType B Findings at Full Power, because the duration of the condition was shorter than 3 days. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of Conservative Bias, which states, individuals use decision making practices that emphasize prudent choices over those that are simply allowable. A proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee failed to exercise prudent judgment when they raised power above 15 percent prior to meeting TS Limiting Condition for Operation 3.6.2.5 and 3.6.3.1 while still in the MODE of Applicability (MODE 1) (H.14).
05000254/FIN-2016002-02Failure to Identify Conditions Adverse to Quality2016Q2The inspectors identified a finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, for the failure to identify a condition adverse to quality. Specifically, the licensee failed to identify the installation of the low pressure coolant injection (LPCI) loop-select differential pressure indicating switches (DPISs) on both units beyond their performance-centered maintenance template recommended replacement frequency and beyond their environmental qualification (EQ) service life established in EQ binder EQ-83Q as conditions adverse to quality (CAQ) in their corrective action program (CAP). The licensees corrective actions included entering the non-conforming conditions into the CAP (Issue Report 2663100) and evaluating the CAQs for operability. The licensee determined the current DPISs were operable but non-conforming, and replaced all remaining LPCI loop-select DPISs. The failure to identify CAQs within the CAP was determined to be more than minor because if left uncorrected, it had the potential to lead to a more significant safety concern. Specifically, the failure to identify when safety-related structures, systems, or components (SSCs) are beyond their qualified life could lead to an SSC not being able to perform its specified safety function. The finding was screened against the Mitigating Systems Cornerstone and determined to be of very low safety significance because the SSC maintained its operability. The inspectors determined this finding affected the cross-cutting area of problem identification and resolution, in the aspect of Identification, which states, The organization implements a corrective action program with a low threshold for identifying issues. Individuals identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, the licensee failed to document a condition adverse to quality related to the LPCI loop select DPISs on both units in the CAP in a timely manner (P.1).
05000254/FIN-2016001-02Failure to Identify Structures, Systems, and Components as Safety-Related2016Q1A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion II, Quality Assurance, was identified by the inspectors for the licensees failure to identify the structures, systems, and components to be covered by the quality assurance program, in that they did not properly classify a component of the control room emergency ventilation system as safety-related. The licensee documented the issue in their corrective action program under Issue Report 2596725. Immediate corrective actions included replacing Differential Pressure Switch (DPS) 0579550 and revising the control room ventilation procedure to allow operators to disable the interlock between the A and B trains of the control room emergency ventilation system. The procedure change eliminated the need for the DPS to be classified as safety-related (and therefore corrected the violation) because in the event of a failure of the DPS, the system would still be able to perform its safety function. The performance deficiency was determined to be more than minor and a finding because it was associated with the Barrier Integrity Cornerstone attribute of Design Control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the B train of the control room emergency ventilation system is a habitability system that is provided to ensure control room operators are able to remain in the control room and operate the plant safely and to maintain the plant in a safe condition under accident conditions. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012. The inspectors determined the finding to be of very low safety significance (Green) in accordance with Exhibit 3, Barrier Integrity Screening Questions, because the finding only represented a degradation of the radiological barrier function provided for the control room and did not represent a degradation of the barrier function of the control room against smoke or toxic atmosphere. This finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current performance.
05000254/FIN-2016001-01Failure to Control Deviation from EQ Standard Results in Limit Switch Submergence2016Q1A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was self-revealed on February 2, 2016, when the operators received an alarm due to a steam leak in the Unit 1 main steam isolation valve room which resulted in the limit switch compartment for Unit 1 reactor core isolation cooling (RCIC) system motor-operated valve (MOV), MO 1130117 (outboard primary containment steam isolation valve), becoming submerged with water. Specifically, the licensee failed to ensure that deviations from design standard, Environmental Qualification Standard 74Q (EQ74Q), were controlled during original installation of MO 1130117 such that the valve would not be subjected to a spray or submergence environment. The licensee documented the issue in their corrective action program under Issue Report 2625523. Corrective actions included a temporary repair of the steam leak, removal of water from the limit switch compartment, and compensatory measures that included daily monitoring for steam leaks in the Unit 1 main steam isolation valve room. In addition, the licensee performed an extent of condition review of other valves in the main steam isolation valve room. Planned corrective actions included installing t-drains or weep holes in MOVs that the licensee deemed susceptible to spray or submergence. The performance deficiency was determined to be more than minor and a finding because it was associated with the Barrier Integrity Cornerstone attribute of Design Control and affected the cornerstone objective to provide reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to control any environmental qualification design deviations had the potential to impact the ability of MO 1130117 to close on an isolation signal and prevent radioactive releases to the environment. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process for Findings at Power, issued June 19, 2012. The inspectors determined the finding to be of very low safety significance (Green) in accordance with Exhibit 3, Barrier Integrity Screening Questions, because the inspectors answered No to all questions in Section B of Exhibit 3. This finding did not have a cross-cutting aspect because the performance deficiency was not indicative of current performance.
05000254/FIN-2015004-02Licensee-Identified Violation2015Q4Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established procedure QCOP 650007, Racking in a 4160 V Horizontal Type AMHG or G25 Circuit Breaker, as the implementing procedure for installing fuse blocks in safety-related breakers, an activity affecting quality. Contrary to the above, prior to August 21, 2015, the licensee failed to have a procedure for installing fuse blocks in safety-related breakers. Specifically, QCOP 350007 failed to ensure that fuse blocks for safety-related 4160 Volt breakers were properly installed to ensure the breakers would perform their function. The procedure did not provide the operators guidance to ensure the fuse blocks were fully seated, and on August 21, 2015, following post-maintenance testing for the Unit 1A RHR pump breaker and the system being declared operable, an equipment operator on rounds identified that the breaker closing springs were not charged. The licensee determined that the breaker fuse blocks were not fully seated following breaker maintenance. The licensee captured this issue in the CAP as IR 2550801. The inspectors evaluated the finding under Inspection Manual Chapter 0609, Appendix A, The SDP (Significance Determination Process) for Findings At-Power, issued June 19, 2012. The inspectors answered No to questions A1A4 in Exhibit 2, Mitigating Systems Screening Questions, and determined the finding was of very low safety significance (Green).
05000254/FIN-2015405-01Security2015Q4
05000254/FIN-2015004-01EAL Threshold Values Were Not Revised2015Q4An Unresolved Item (URI) was identified because additional information is required to determine whether a performance deficiency that is more than minor exists, and if a violation of 10 CFR 50.54(q)(2), which requires a licensee to develop and maintain an emergency plan that meets the requirements of 10 CFR 50.47(b), and 10 CFR Part 50, Appendix E, had occurred. The licensee identified an issue of concern when the Quad Cities General Abnormal procedures (QGAs) were revised with a new value for Minimum Steam Cooling Reactor Pressure Vessel Water Level (MSCRWL) but the associated EALs that use the MSCRWL value as an EAL threshold were not revised. On March 12, 2015, the QGAs were revised with a new value for MSCRWL. However, the site EALs that should use the revised QGA value as an EAL threshold value were not revised. The licensee scheduled the revisions of the QGAs to support implementation of changes that were associated with the diverse and flexible coping strategies (FLEX) implementation and the sites transition to new Optima2 fuel. Both of the changes were scheduled to be implemented in March 2015 during the Quad Cities Unit 1 Refueling Outage as part of a revision package. Because of the new fuel, the MSCRWL value changed from -166 inches to -190 inches. On April 28, 2015, the licensee identified the EALs were not changed to correspond with the new MSCRWL values incorporated in the QGAs. The specific EALs that are affected are MG2 and FG1, which are used to determine if a General Emergency should be declared based on the MSCRWL value. Since the value remained at -166 inches, the licensee concluded that the issue could have potentially caused, under certain conditions, the site to declare a General Emergency earlier than needed and issue an unnecessary Protective Action Recommendation (PAR) to the public. Following identification of the issue, the licensee implemented the appropriate changes to EALs MG2 and FG1 on April 30, 2015. Since there was a discrepancy between the QGAs and the EAL threshold values that could have affected the timely and accurate classification of a General Emergency, a potential performance deficiency exists. However, in order to determine if the performance deficiency is more than minor significance, additional information is needed. The URI was identified pending additional information and inspection follow-up. Specifically, additional information is required to: understand if the discrepancy in the MSCRWL values documented in the QGAs and the EALs would have led to an overclassification of a General Emergency and issuance of an unnecessary PAR; understand if there are events that could be postulated where the -166 inches could be exceeded without reaching the -190 inches; and understand the timeline from when the fuel was transitioned to Optima2 until the discovery of this issue. This information will assist the inspectors to determine if the performance deficiency is more than minor and if a violation of 10 CFR 50.54(q)(2) occurred.
05000254/FIN-2015003-02Failure to Establish Adequate Procedure to Preclude Unacceptable Preconditioning of the SBGT System2015Q3A finding of very low safety significance and an associated non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to establish a procedure appropriate to the circumstances that precluded unacceptable preconditioning of the standby gas treatment (SBGT) system during surveillance testing. The licensee performed an evaluation and concluded the SBGT system was operable and planned additional testing on the relay timing function. Other corrective actions included revising the applicable procedures such that unacceptable preconditioning would not occur. The licensee captured this issue in their CAP as IR 2524699. The finding was determined to be more than minor because it was associated with the Barrier Integrity Cornerstone attribute of Procedure Quality and affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the inadequate procedure had the potential to mask the ability of the SBGT system to initiate in time to prevent ex-filtration of radioactive gases during a design basis accident. The finding was determined to be of very low safety significance because it represented a degradation of the radiological barrier function for the SBGT system. This finding had a cross-cutting aspect of questioning attitude in the area of human performance because the licensee did not recognize the possibility of mistakes, latent problems, or inherent risk, even while expecting successful outcomes. Specifically, the licensee failed to recognize that performing the steps in the specified sequence could unacceptably precondition the time-delay relay for the SBGT system and mask the ability of the system to perform its function (H.12).
05000254/FIN-2015003-01Failure to Evaluate Degraded or Non-Conforming Conditions for Operability2015Q3A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to document degraded or non-conforming conditions in the corrective action program (CAP) and route or discuss the issue with Operations shift management so that operability of the affected components could be evaluated. Immediate corrective actions included entering the issues into the CAP and evaluating the issues for operability. The licensee captured the issue in the CAP as Issue Reports (IRs) 2537968 and 2537936. The finding was determined to be more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the failure to identify degraded, non-conforming, or unanalyzed conditions in the CAP and bring those conditions to the attention of Operations shift management so that the operability of safety-related systems, structures, and components (SSCs) may be evaluated could lead to those SSCs being in an inoperable condition without the appropriate Technical Specification (TS) actions taken. The inspectors concluded this finding was associated with the Mitigating Systems Cornerstone. The finding was determined to be of very low safety significance because the control room emergency ventilation (CREV) and high pressure coolant injection (HPCI) systems remained operable. This finding had a cross-cutting aspect of identification in the area of problem identification and resolution because the licensee did not identify issues completely, accurately, and in a timely manner in accordance with the program. Specifically, when degraded and non-conforming conditions were identified, licensee personnel failed to promptly capture the issues in the CAP (P.1).
05000254/FIN-2015003-03Failure to Adequately Inspect Relay Contacts for Oxidation Results in Relay Failure2015Q3A finding of very low safety significance and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to establish a preventive maintenance procedure for HFA relays that was appropriate to the circumstances. Immediate corrective actions included burnishing of the associated relay contacts and testing the associated relays. In addition, the licensee revised their relay inspection procedure and planned future relay replacements during the next refueling outage. The licensee entered the issue into their CAP as IR 2485051. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of Procedure Quality and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). Specifically, the failure to perform adequate preventive maintenance on the automatic depressurization system (ADS) logic HFA relay in 2013 resulted in the build-up of oxidation on the relay contacts. This build-up caused the relay to fail its next scheduled test in 2015. A senior reactor analyst performed a detailed risk evaluation and determined the finding was of very low safety significance. This finding had a cross-cutting aspect of operating experience in the area of problem identification and resolution, because the licensee did not systematically collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, the licensee identified several internal and external operating experience events related to relay contact oxidation and failed to implement changes to their relay inspection procedures to ensure that effective corrective actions were implemented (P.5).
05000254/FIN-2015002-01Failure to Conduct Post-Maintenance Testing Following Manual Operation of RCIC MOV2015Q2A finding of very low safety significance and associated NCV of Technical Specification 5.4, Procedures, was self-revealed on March 22, 2015, for the licensees failure to conduct procedurally required post-maintenance testing on reactor core isolation cooling (RCIC) motor operated valve (MOV) MO 1130161, following operation of the valve in the manual mode. Immediate corrective actions included manually engaging the motor clutch and functionally stroking the valve from the control room to verify operation. The licensee captured this condition in their CAP as Issue Report 2472416. The finding was determined to be more than minor because it was associated with the Mitigating Systems Cornerstone attribute of Equipment Performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the licensee was not able to ensure the operability of the RCIC system when they failed to conduct post-maintenance testing (PMT) on RCIC 1130161. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions. The inspectors answered No to all questions in Section A of Exhibit 2 and the finding screened as Green, or very low safety significance. This finding has a cross-cutting aspect in the area of Human Performance, Documentation, because the licensee did not maintain complete, accurate, and up-to-date documentation. Specifically, the licensee failed to document the status of the RCIC valve after placing it in the manual mode of operation to ensure that the required PMT was performed.
05000254/FIN-2015002-02Inadequate Zone of Protection for Electrical Bus Maintenance2015Q2A finding of very low safety significance and associated NCV of Technical Specification 5.4, Procedures, was self-revealed on March 14, 2015, for the licensees failure to implement a clearance order in accordance with procedure OPAA109101, Clearance and Tagging, for electrical maintenance on Bus 12, Cubicle 9. The clearance order failed to provide a safe zone of protection for all physical work to be performed under the clearance order or for required equipment protection. Immediate corrective actions included stopping all electrical work and verifying electrical work boundaries prior to re-commencing work. The licensee documented the issue in the corrective action program (CAP) under Issue Report 2468511. The finding was determined to be more than minor because, if left uncorrected, it could become a more significant safety concern. Specifically, the failure to properly control and de-energize equipment prior to performing maintenance could have an impact on safety-related equipment (including equipment damage and potential loss of off-site power). The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings. Because the finding impacted the Initiating Events Cornerstone and Unit 1 was shut down at the time of the event, the inspectors determined the finding could be further evaluated using IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings. The inspectors answered No to all questions in Exhibit 2 of IMC 0609, Appendix G, Attachment 1 and determined the finding was of very low safety significance (Green). This finding has a cross-cutting aspect in the area of Human Performance, Work Management because the licensee did not implement a process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, the licensee failed to plan, control, and execute a clearance order that provided a safe zone of protection for all physical work to be performed under the clearance order or for required equipment protection during maintenance on Bus 12, Cubicle 9 (H.5).
05000254/FIN-2015201-01Security2015Q2
05000254/FIN-2015001-01Failure to Establish and Maintain Service Life for Safety-Related Relay Results in Failure and Inoperability2015Q1A finding of very low safety significance (Green) and associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, was self-revealed on January 6, 2015, when an electrical maintenance worker found a tripped breaker in motor control center (MCC) 281, for the Unit 2 power feed to the common unit (Unit 0) fuel oil transfer pump (FOTP). The licensee determined that an HGA relay in the FOTP power transfer circuit had failed due to aging and not having any associated preventive maintenance task. The inspectors determined the licensee failed to establish and maintain the service life for the FOTP HGA relay, which was a performance deficiency. This also resulted in the inoperability of the Unit 0 emergency diesel generator (EDG) for longer than its technical specification allowed outage time, which was a violation of Technical Specification 3.8.1, AC SourcesOperating. The immediate corrective actions included replacing the failed relay and declaring the EDG operable following post-maintenance testing. The licensee captured the issue in their corrective action program (CAP) as Issue Report (IR) 2433389. The performance deficiency was determined to be more than minor and a finding because it was associated with the equipment performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency caused an unplanned inoperable condition for the Unit 0 EDG. The inspectors evaluated the finding using IMC 0609, Appendix A, The SDP for Findings At-Power, issued June 19, 2012. The issue resulted in the EDG being inoperable for longer than the Technical Specification (TS) allowed outage time. A detailed risk analysis was performed and determined the finding was of very low safety significance. This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because the licensee did not thoroughly evaluate issues to ensure that the resolution addressed causes and extent of conditions commensurate with their safety significance. Specifically, the licensee identified other EDG electrical component failures that occurred at the station where the causes were identified as failure to have associated preventive maintenance for the affected components and equipment. The extent of condition evaluations for those events failed to identify additional safety related components that did not have any associated preventive maintenance tasks or documented service life, including replacement schedules.
05000254/FIN-2015001-03Licensee-Identified Violation2015Q1Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, states 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. The licensee established procedure NFAA309, Attachment 4, Move Sheet as their implementing procedure for moving fuel during refuel outage Q1R23, an activity affecting quality. Procedure NFAA309, Attachment 4, Move Sheet for Q1R23 Shuffle 2 Northwest, step 7, states, in part, to move fuel bundle QAD224 from 5334 (NW) to 5140 (SE). Contrary to the above, on March 10, 2015, the licensee failed to accomplish the steps in licensee procedure NFAA309, Attachment 4, when fuel assembly QAD224 was inserted into core location 5140 in the incorrect orientation (NW). On March 13, 2015, the licensee implemented procedure NFAA309, Attachment 4, to retrieve the misoriented fuel assembly from its core location and reinsert it in the proper orientation. The issue was entered into the licensees CAP as IR 2467903. The failure to meet the requirements of NFAA309 was a performance deficiency. The performance deficiency was more than minor and a finding because it was associated with the Barrier Integrity Cornerstone attribute to provide reasonable assurance that physical design barriers (fuel cladding) protect the public from radionuclide releases caused by accidents or events. Specifically, the misoriented fuel bundle placed the core in a configuration for which shutdown margin (negative reactivity insertion needed to maintain the reactor shutdown under accident conditions) had not been previously analyzed. The inspectors evaluated the finding using the SDP in accordance with IMC 0609, Appendix G, Attachment 1, Shutdown Operations Significance Determination ProcessPhase 1 Initial Screening and Characterization of Findings, Exhibit 4 for the Barrier Integrity Cornerstone. Because the finding involved a fuel bundle misorientation in the reactor core, per the note in Exhibit 4, the finding screened as very low safety significance
05000254/FIN-2015001-02Failure to Ensure Standby Lineup Results in Steam Release in the HPCI Room2015Q1A finding and non-cited violation of very low safety significance (Green) was self-revealed for the licensees failure to ensure the Unit 1 high pressure coolant injection (HPCI) system was in a standby lineup configuration in accordance with station procedures. This represented a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The performance deficiency resulted in a steam release into the Unit 1 HPCI room. The licensee took immediate actions to terminate the steam release by closing the HPCI steam isolation valves. The licensee captured the issue in their corrective action program as IR 2450896. The performance deficiency was determined to be more than minor and a finding because it was associated with the configuration control attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors evaluated the finding using IMC 0609, Appendix A, The SDP for Findings At-Power, and answered No to all of the screening questions in Exhibit 2 Mitigating Systems, and concluded the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Work Management, because the organization failed to implement a process of planning, controlling, and executing work activities such that nuclear safety was the overriding priority; and the work process did not include the identification and management of risk commensurate to the work, and the need for coordination with different job activities. Specifically, the licensee failed to coordinate the simultaneous performance of two tests and ensure the HPCI system was in the proper lineup and configuration prior to test execution.
05000254/FIN-2015001-04Licensee-Identified Violation2015Q1Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented procedures of a type appropriate to the circumstances and be accomplished in accordance with these procedures. The licensee established procedure QCOA 020204, Reactor Recirc Pump Trip, as the implementing abnormal operating procedure for responding to a reactor recirculation pump trip transient condition, an activity affecting quality. Procedure 020204, step D.3 states, in part, to insert control rods in sequence to TARGET IN position using control rod move sheets. Contrary to the above, on January 27, 2015, the licensee failed to insert control rods in sequence according to the control rod move sheets. Specifically, the reactor operators inserted control rods in step 20, and two rods in step 19, before a qualified nuclear engineer identified that the operators had missed inserting the rod in step 21. The licensee immediately entered QCOA 030004, Mispositioned Control Rods, and inserted the missed rod per the qualified nuclear engineers guidance. The licensee documented this issue in their CAP as IR 2443241. The inspectors determined that the finding was more than minor because it was associated with the Mitigating Systems Cornerstone attribute of human performance and adversely affected the cornerstone objective ensuring the reliability, availability, and capability of systems that respond to initiating events. The inspectors screened the finding using Exhibit 2 for Mitigating Systems in IMC 0609, Appendix A, The Significance Determination Process for Findings At Power, and answered, Yes to question C.3 because the finding represented a mismanagement of reactivity by operators. The inspectors were directed to IMC 0609, Appendix M, SDP Using Qualitative Criteria, and the finding screened as having very low safety significance (Green) because the licensee determined that there was very little effect on their margin to thermal limits.
05000254/FIN-2014008-01Failure to Identify Aging Effects on Plant Equipment and Structures2014Q4The inspectors identified a finding of very low safety significance and associated NC V of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to effectively identify, evaluate, and document aging effects on plant equipment and structures as part of the licensees Aging Management Programs for a plant within its period of extended operation. The inspectors identified two corroded pipe supports and associated base plates in the Unit 1 high pressure coolant injection (HPCI) room as well as a severely corroded nut and stud on the 1/2 diesel generator cooling water pump outboard mechanical seal. These conditions had not been previously identified, evaluated, or documented. The licensee entered this finding into their Corrective Action Program. The performance deficiency was determined to be more than minor and a finding in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because if left uncorrected, the performance deficiency would have the potential to lead to a more significant safety concern. The finding screened as very low safety significance (Green) because the inspectors were able to answer No to each screening question, because the conditions had not yet affected structural integrity or operability of the systems. Specifically, the licensee confirmed the HPCI supports would be capable to perform their function and the remaining bolts on the mechanical seal were sufficient to prevent excessive leakage. The inspectors identified a cross-cutting aspect associated with this finding in the area of Human Performance, Resources component, because the licensee did not ensure that personnel, equipment, procedures, and other resources are adequate to assure nuclear safety by maintaining long term plant safety.
05000254/FIN-2014403-01Licensee-Identified Violation2014Q4
05000254/FIN-2014005-02Failure to Submit a Report Required by 10 CFR 50.72(b)(3)(xiii)2014Q4The inspectors identified a Severity Level IV non-citied violation of 10 CFR 50.72(b)(3)(xiii) when licensee personnel failed to submit a report required by 10 CFR 50.72 for a loss of emergency assessment capability when an unplanned loss of the station seismograph was identified. Specifically, the licensee declared the station seismograph non-functional on October 7, 2014, and failed to report the condition in accordance with 10 CFR 50.72. The licensee entered this issue into their CAP as Issue Report 2415243, A Potential Issue Related to the QDC Seismic Monitor. The inspectors determined that this issue had the potential to impact the regulatory process based, in part, on the generic communications input that 10 CFR 50.72 reports serve. Since the issue impacted the regulatory process, it was dispositioned through the traditional enforcement process. The inspectors determined that this issue was a Severity Level IV violation based upon Section 6.9, Inaccurate and Incomplete Information or Failure to Make a Required Report, example d.9 in the NRC Enforcement Policy. Example d.9 specifically stated, The licensee fails to make a report requirement by 10 CFR 50.72 or 10 CFR 50.73. Because a more-than-minor Reactor Oversight Process finding was not identified, there was no cross-cutting aspect associated with this violation.
05000254/FIN-2014008-02Non-Conservative DGCW Pump Break Horsepower Assumed in EDG Loading Analysis2014Q4The inspectors identified an unresolved item (URI) regarding the motor load measured by the licensee for the Unit 1 DGCW pump that was determined to be less than the vendor certified pump performance data and pump load break horsepower (BHP) requirement at maximum flow conditions. Field measured pump motor load data was evaluated by the licensee and utilized as a design input in the Electrical Transient Analysis Program (ETAP) analysis for the emergency bus loading on offsite power and for the bus loading when powered by the EDG. The licensee did not fully evaluate and reconcile the effects on electrical bus and EDG loading analyses for the required pump BHP for the expected pump flow conditions, when the Unit 1 DGCW pump impeller was replaced in 2011 under work order (WO) 01301062 and evaluated under engineering change (EC) 369825. This condition was entered into the corrective action program as AR2420101 and AR2420905. The inspectors noted the licensee did not use vendor certified pump performance data when they evaluated the effects of the pump impeller replacement on the motor load. Instead, the licensee used field measured electrical load data as a design input to the ETAP electrical calculations for bus load flow and EDG loading, which the inspectors determined resulted in the bus and EDG loading being non-conservative. Technical Specification surveillance requirement (SR) 3.8.1.15 required loading the EDG between 2340 and 2600 kW for 22 hours and between 2730 and 2860 kW for 2 hours. The inspectors determined for the 1/2 EDG (which the licensee identified had the worst 12 case loading), the margin between the actual EDG accident loading and the 2-hour minimum surveillance limit of 2730 kW was further reduced as a result of the increase in required pump brake horsepower from 90 to 101 BHP based on vendor certified pump performance data. The licensee concluded the pump brake horsepower increased only 5 HP, from 90 to 95 BHP, based on their field measurements of voltage and line current, but failed to reconcile the 5 HP load increase with vendor certified pump performance curve, which required approximately 101 BHP, or an 11 HP increase, at the established flow condition (1650 gpm). The licensee determined the additional increase in pump horsepower, when considering the vendor certified BHP for the pump at maximum flow conditions, remained within the minimum 2-hr load capability requirement established for TS SR 3.8.1.15. The inspectors determined the vendors pump load requirement derived from the certified pump performance curve of 101 BHP was approximately 6 HP more than the load determined by the licensee by field measurement of motor current and voltage. Field measurement of motor current and voltage was a licensee approved method in standard NES-EIC-11.01, Use of Analytical Software for AC Auxiliary Power System Analysis, to determine pump horsepower load for input to the ETAP analysis. The inspectors were concerned that if other load inputs to ETAP, derived from the measurement of voltage and current conditions, were also found to be less than actual maximum design basis load conditions, the EDG load could be adversely impacted. Resolution of this issue will be based on a licensee evaluation for the extent of condition for any additional impact on the subject bus loading conditions. Pending resolution, this item will be tracked as an unresolved item.
05000254/FIN-2014005-01HPCI Flood Barrier Degraded2014Q4A finding of very low safety significance (Green) and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to meet the requirements of QCTP 0130-11, Internal Flood Protection Program, and QCTS 0810-10,Reactor Building Internal Flood Barrier Surveillance, which require, in part, that internal flood protection requirements for emergency core cooling systems rooms are met. Specifically, the licensee failed to identify that a flood barrier for a fire protection pipe penetration into the Unit 2 high pressure coolant injection room was in a degraded condition. The licensee entered the condition into their CAP as Issue Report 2406984, IEMA U2 HPCI Flood Penetration Concern, and was able to immediately correct the degraded condition of the link-seal type barrier by tightening the bolts around the seal. The finding was determined to be more than minor because failing to identify degraded flood barriers could lead to safety-related equipment becoming susceptible to a flooding event. The finding was associated with the Mitigating Systems Cornerstone attribute of protection against external factors (flood hazard) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with Inspection Manual Chapter (IMC) 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power, issued June 19, 2012. The inspectors answered, No, to all of the Exhibit 2, Mitigating Systems Screening Questions, in section B for external events and determined the finding was of very low safety significance (Green). This finding had a cross-cutting aspect in the area of Human Performance, Consistent Process aspect because the licensee did not use a consistent and systematic approach to conducting flood barrier inspections (H.13).
05000254/FIN-2014008-03Testing of MSIVs with Instrument Air or Drywell Pneumatic System Aligned to Actuators2014Q4The inspectors identified an unresolved issue (URI) regarding the testing of the MSIVs. Specifically, the inspectors identified the MSIV closure timing surveillance tests were performed with non-safety related instrument air or the drywell pneumatic system aligned to the actuators. The inspectors were concerned that the surveillance test acceptance criteria could be non-conservative. The inspectors reviewed surveillance test Procedure QCOS 0250-04, MSIV Closure Timing, and noted the MSIV fast closure testing (required by TS surveillance requirement (SR) 3.6.1.3.6 and the IST Program) was performed with non-safety-related instrument air or the drywell pneumatic system aligned to the actuators. The MSIVs were designed with safety-related accumulators to provide pressure to assist in closing the valves; however these air accumulators would be expected to provide less pressure than the non-safety-related instrument air or drywell pneumatic systems. Technical Specification SR 3.6.1.3.6 required verification that the isolation time of each MSIV will be 3 seconds and 5 seconds. The inspectors observed test Procedure QCOS 0250-04 included separate closure time acceptance criteria for hot and cold conditions, but did not include any acceptance criteria adjustments for the use of non-safety-related instrument air or the drywell pneumatic system. The inspectors were concerned the surveillance test maximum closing time acceptance criteria ( 5 seconds for cold valves) could be non-conservative. This concern was previously addressed by NRC IN 85-84, Inadequate Inservice Testing of Main Steam Isolation Valves. At that time, the licensees review of IN 85-84 stated the MSIV air supply isolation valve was closed during testing. On February 22, 1989, 13 General Electric Nuclear Services Information Letter (SIL) No. 482 was issued to address the effect of non-safety related air on the closure time testing of some MSIVs. The SIL indicated that testing of two Boiling Water Reactor MSIVs equipped with hydraulic self-compensation mechanisms resulted in a time increase of about 0.1 seconds when the non-safety-related air supply was disconnected. The installed MSIVs were also equipped with hydraulic self-compensation mechanisms. The SIL No. 482 also recommended testing to verify the effect of removing non-safety related air supplies. The SIL stated that if the closure time with the air disconnected increased less than 0.3 seconds, it would be acceptable to leave the non-safety-related air connected during closure time testing. The 0.3 second criterion was based on a typical total allowable variation of 0.5 seconds. This 0.5 margin was based on an historical practice of setting MSIV closure times between 3.5 and 4.5 seconds (light to light). In addition, the historical accident analyses were based on the MSIVs closing in less than 10 seconds (plus an additional 0.5 second allowance for instrument/control response). The current alternate source term accident analysis was based on the MSIVs closing within 5 seconds (plus an additional 0.5 second allowance for instrument/control response) under accident conditions (UFSAR Section 15.6.4.5.1). Based on the current alternate source term analysis, there was no allowance for margin between the as-found surveillance test acceptance criterion, the TS limit, and the analytical limit. It was unclear to the inspectors whether the licensees basis to revise the test methodology to having non-safety air un-isolated was based on the SIL. The licensees evaluation of SIL No. 482, dated June 27, 1989, stated the MSIVs were tested with the non-safety-related instrument air or drywell pneumatic system aligned to the actuators. The evaluation also stated a special test was required to determine the effect of air pressure on MSIV closure time. The licensee determined the special test was performed with acceptable results to justify continued closure time testing with the non-safety related instrument air or the drywell pneumatic system aligned to the actuators. However, the actual test results were lost in the late 1980s due to a computer records failure. The licensee initiated AR02420923 on December 4, 2014, to address this issue. The licensee was able to obtain MSIV closure time special test results for eight similar MSIVs from Dresden Station, performed in May 1992. The results indicated an average closure time increase of less than 0.1 seconds with non-safety related air disconnected. However, there was considerable variation in the individual MSIV test results. Based on the Dresden special test results, the records indicating that a special test was performed successfully, the most recent as-left MSIV closure time test results, and documentation from General Electric, the licensee determined the MSIVs remained operable. Resolution of this issue will be based on additional analysis and/or testing by the licensee. This analysis/testing will determine if additional surveillance test acceptance criteria margin and/or a change in testing methodology will be required to ensure the MSIVs will close in the required time under the most limiting conditions. Specifically, the inspectors were concerned the current testing methodology (with the non-safety-related instrument air or drywell pneumatic system aligned to the actuators) could result in the MSIVs stroking faster than the most limiting accident conditions, with only safety-related accumulators available, which appeared to be a change from the testing methodology prior to IN 85-84. The inspectors were also concerned the recommendations of SIL No. 482 were not applicable to the current licensing/design basis because those recommendations were based on an assumed total allowable variation of 0.5 seconds. As discussed above, the alternate source term analysis 14 (performed after the SIL was issued) did not include any closure time margin beyond the as-found surveillance test acceptance criterion and the upper TS time limit. It appeared that any non-conservatism in the test methodology would be unacceptable unless the test acceptance criteria included explicit allowances for the difference between the test conditions and the most limiting accident conditions. Pending resolution, this item will be tracked as an unresolved item.
05000254/FIN-2014004-03Licensee-Identified Violation2014Q3The licensee identified a violation of TS 3.3.1.1, RPS Instrumentation, and TS LCO 3.0.4. Technical Specification 3.3.1.1 specifies that four channels of turbine condenser vacuum-low scram function are required to be operable in MODE 1. Technical Specification 3.3.1.1, Condition A, stated that if one channel is not operable, the channel of the associated trip system is to be placed in trip within 12 hours. Technical Specification 3.0.4 specifies the requirements that must be satisfied prior to making a MODE change if a limiting condition for operation (LCO) is not met. Limiting Condition for Operations 3.0.4 stated, in part, that when an LCO is not met, entry into a MODE or other specified condition in the Applicability shall only be made when the associated ACTIONS to be entered permit continued operation in the MODE or other specified condition for an unlimited period of time. Contrary to the above, from May 6, 2014 to May 16, 2014, the licensee failed to meet the provisions of TS 3.3.1.1 and LCO 3.0.4. Specifically, on May 16, 2014, RPS pressure switch 2-0503-B was declared inoperable. The licensee determined that the cause was the inadvertent closure of the C condenser pressure indicator root valve on May 6, 2014, at approximately 6:20 p.m. while Unit 2 was in MODE 2. Unit 2 entered MODE 1 at 10:52 p.m. on May 6, 2014. Therefore, the licensee transitioned to MODE 1 without the required number of channels and did not take the required action to place the channel or associated trip system in the trip condition within 12 hours. Section 4OA2.5 above provides additional background description for this licensee-identified violation. The licensee documented the conditions prohibited by TSs for pressure switch 2-0503-B in IR 1660714. Because the inspectors answered No to all questions in Section C of IMC 0609 Appendix A, The Significance Determination Process for Findings At Power, Exhibit 2 Mitigating Systems Screening Questions, the finding screened as very low safety significance (Green).
05000254/FIN-2014004-02Inadequate Evacuation Time Estimate Submittals2014Q3The inspectors identified a finding of very low safety significance (Green) with an associated non-cited violation of 10 CFR 50.54(q)(2) as required by 10 CFR 50.47(b)(10) and 10 CFR Part 50, Appendix E, Section IV.4, for failing to maintain the effectiveness of the Quad Cities Nuclear Power Station Emergency Plan, as a result of failing to provide the station evacuation time estimate (ETE) to the responsible offsite response organizations by the required date. Exelon submitted the Quad Cities Nuclear Power Station ETE to the NRC on December 12, 2012, prior to the required due date of December 22, 2012. The NRC completeness review found the ETEs to be incomplete due to Exelon fleet common and site-specific deficiencies, thereby preventing Exelon from providing the ETEs to responsible offsite response organizations and from updating site-specific protective action strategies as necessary. The NRC discussed its concerns regarding the completeness of the ETE, in a teleconference with Exelon on June 10, 2013, and on September 5, 2013, Exelon resubmitted the ETEs for its sites. The NRC again found the ETEs to be incomplete. The issue is a performance deficiency because it involves a failure to comply with a regulation that was under Exelons control to identify and prevent. The finding is more than minor because it is associated with the Emergency Preparedness Cornerstone attribute of procedure quality and because it adversely affected the cornerstone objective of ensuring that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The finding is of very low safety significance (Green) because it was a failure to comply with a non-risk significant portion of 10 CFR 50.47(b)(10). The licensee had entered this issue into their corrective action program (CAP) and re-submitted a new revision of the Quad Cities Nuclear Power Station ETE to the NRC on April 30, 2014. The cause of the finding is related to crosscutting element of Human Performance, Documentation (H.7).
05000254/FIN-2014007-02Inadequate Administrative Controls 05000265/2014007-022014Q3A finding of very low safety significance (Green) and an associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors when they determined that Technical Specification (TS) surveillance procedures contained inadequate acceptance criteria. The failure to have TS surveillance procedure acceptance criteria that ensured the Emergency Diesel Generator (EDG) loading would not exceed the maximum licensed limit was a performance deficiency. The issue was entered into the licensees CAP as IR 02389102, PIR Admin Controls For Allowed EDG Frequency Tolerance. The licensee had not had time to determine corrective actions before the end of the inspection. The performance deficiency was determined to be more than minor because it was associated with the design control attribute of the Mitigating Systems Cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences and is therefore a finding. Specifically, the licensee failed to ensure the acceptance criteria for EDG frequency and voltage would not affect the operability and reliability of the engine and safety related structures, systems or components. Using Manual Chapter 0609, Attachment 0609.04 Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, dated June 19, 2012, the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was a deficiency affecting the design or qualification of a mitigating structure, system or component. This finding has a cross-cutting aspect of resolution in the area of problem identification because the licensee did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance. Specifically, the licensee did not implement adequate administrative controls to their EDG testing procedures to ensure that the procedures adequately addressed the non-conservative TS.
05000254/FIN-2014007-01Inadequate Rounds Package Acceptance Criteria2014Q3A finding of very low safety significance (Green) was identified by the inspectors when they determined that non-licensed operator general area rounds and field checks were inadequate for the circumstances. The inspectors determined that the failure to have non-licensed operator rounds package acceptance criteria that met procedural requirements was a performance deficiency. The licensee entered this issue into the CAP as Issue Report (IR) 02385609, PIR Operator Rounds For HPCI Bearing Oil Lvl Differ between Units. The licensee had not had time to determine corrective actions before the end of the inspection. The performance deficiency was more than minor because it was associated with the procedure quality attribute of the mitigating systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability and capability to response to initiating events to prevent undesirable consequences and is therefore a finding. Using Manual Chapter 0609, Attachment 0609.04 Initial Characterization of Findings, and Appendix A The Significance Determination Process for Findings at Power, the finding was screened against the mitigating systems cornerstone and determined to be of very low safety significance (Green) because the finding was/did not: 1) a deficiency affecting the design or qualification of a mitigating structure, system or component, 2) represent a loss of system and/or function, 3) represent an actual loss of function of a single train for greater than its technical specification allowed outage time, 4) represent an actual loss of function of one or more non-technical specification trains of equipment designated as high safety-significant for greater than 24 hours and 5) did not involve the loss or degradation of equipment or function specifically designed to mitigate a seismic, flooding or severe weather event. The inspectors determined this finding affected the cross-cutting area of Human Performance in the aspect of Training. Specifically, the non-licensed operators should have been trained that an oil level not between the marked bands on the oil level indicator was an issue regardless of the rounds acceptance criteria for that parameter.
05000254/FIN-2014004-01Angle Iron Support Installed with Minimal Clearance to Unit 2 Torus Shell2014Q3A finding of very low safety significance (Green) and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to evaluate the impact of a conduit support installed in close proximity of the Unit 2 torus shell. Specifically, during installation of the conduit support, the licensee failed to provide instructions to ensure that sufficient clearance from the torus shell was provided to accommodate the torus wall movements predicted in the Updated Final Safety Analysis Report (UFSAR) torus design basis load cases. Immediate corrective actions included performing an operability evaluation under Issue Report (IR) 1672301 that determined the torus remained operable under all design basis events. The licensee has also corrected the condition by cutting the conduit support to ensure sufficient clearance to the torus wall is maintained. The performance deficiency was determined to be more than minor because the finding was associated with the design control attribute of both the Mitigating Systems and Barrier Integrity Cornerstones. The finding adversely affected the Mitigating Systems cornerstone attribute of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding also adversely affected the Barrier Integrity Cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. The inspectors determined the finding screened as very low safety significance (Green) because the licensees operability evaluation determined the torus remained operable under all design basis conditions. The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not representative of current performance because it was associated with a modification that occurred in the 1980s.
05000254/FIN-2014003-02Post Maintenance Test Fails to Ensure Battery Charger can Perform Function2014Q2A finding of very low safety significance and associated non-citied violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to meet the requirements of MA-AA-716-012, Post Maintenance Testing, which states, in part that post maintenance testing ensures that a component is able to perform its intended function and that the original deficiency is corrected. Specifically, licensee procedure QCEMS 0210-01 failed to include quantitative and qualitative acceptance criteria for determining that the Unit 1 250 VDC Battery Charger could perform its intended function. This issue was placed into the licensees CAP as IR 1631541. Immediate corrective actions included replacing the float potentiometer in the battery charger circuitry, replacing a thyristor in the voltage regulation circuitry, and correcting a loose solder connection identified in the battery charger circuitry. Planned corrective actions include revising procedure QCEMS 0210-01 to include acceptance criteria that ensure the battery chargers can satisfactorily perform their intended function. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors answered, No, to all of the Exhibit 2, Mitigating Systems Screening Questions, in Section A and determined the finding was of very low safety significance. This finding had a cross-cutting aspect of design margins in the area of Human Performance because the licensee did not operate and maintain the battery charger within design margins. Specifically, the licensees post maintenance testing acceptance criteria did not give them enough margin to prevent the battery from becoming inoperable (H.6).
05000265/FIN-2014003-01Seismic Scaffold in Contact with Safety-Related Equipment2014Q2A finding of very low safety significance and associated non-citied violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified by the inspectors for the licensees failure to meet the requirements of procedure MA-AA-796-024, Scaffold Installation, Inspection, and Removal, when scaffold Q0178 was built with one of its supports in rigid contact with the operable Unit 2 torus. Immediate corrective actions included modifying the scaffold such that it was no longer in contact with the Unit 2 torus. This issue was captured in the licensees CAP as IR 1639356. The finding was determined to be more than minor because the finding was associated with the Mitigating Systems Cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a scaffold built in contact with safety related equipment could damage the equipment and affect its availability and reliability. The inspectors determined the finding could be evaluated using the SDP in accordance with IMC 0609, Significance Determination Process, Appendix A, The Significance Determination Process (SDP) for Findings At-Power. The inspectors answered, No, to all of the Exhibit 2, Mitigating Systems Screening Questions, in section A and determined the finding was of very low safety significance. This finding has a cross-cutting aspect of documentation in the area of human performance because the licensee did not create and maintain complete, accurate and, up-to-date documentation. Specifically, the licensee did not completely and accurately evaluate the acceptability of a scaffold that was in contact with safety related equipment (H.7).
05000254/FIN-2014002-02Licensee-Identified Violation2014Q1On February 5, 2014, Mechanical Maintenance Department was performing a bi-annual Fire Protection Walkdown, and identified that the fire protection coating was missing on an I-beam that was a recently installed pipe support in the turbine building. This pipe support was installed by a licensee contractor as part of an engineering change to install above ground piping for safety-related service water. The licensees investigation discovered that the work instructions did not provide steps to replace the coating or to initiate the fire impairment in accordance with the licensees work planning process. Technical Specification 5.4.1.c. states that written procedures shall be established, implemented, and maintained covering the Fire Protection Program. The licensee established procedure QCAP 1500-01, Revision 32; Administrative Requirements for Fire Protection, as their implementing procedure for the Fire Protection Program. Step D.7.a.(1)(e) states, in part, that fire barriers (structural steel fire coating) protecting safety related or safe shutdown areas SHALL be intact when in Mode 1, 2, & 3. Additionally, Step D.7.a.(3) states, IF fire barrier inoperable, THEN compensatory and reporting requirements SHALL be followed per steps D.7.c and D.7.d. Contrary to the above, the licensee failed to have an intact fire barrier while in Mode 1 without implementing any compensatory measures or reporting requirements. This was a violation of Technical Specification 5.4.1.c. This finding is more than minor because it is associated with the Reactor Safety Mitigation Systems Cornerstone attribute of protection against external factors (fire) and adversely 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 finding was determined to be of very low safety significance (Green) because the reactor would be able to reach and maintain a safe shutdown condition. The licensee documented this issue in their corrective action program in IR 1617549, Fire Protection Coating Missing on Steel I-Beam.
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