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05000373/FIN-2018003-082018Q3LaSalleFailure to Implement Engineering Change Results in Reactor Coolant Boundary LeakageThe inspectors documented a self-revealed finding of very low safety significance (Green) and associated NCVs of TS 5.4.1 Procedures, and TS 3.4.5 for the failure to implement EC 354539 to perform the final piping weld for the 1B33F067B bonnet vent line in the field, resulting in pressure boundary leakage when the weld failed at power.
05000374/FIN-2018003-072018Q3LaSallePotential Failure to Promptly Correct the Unit 2 Primary Containment Wall Cavity Leakage Condition and to Follow Corrective Action Program ProcessCondition description in AR 2420888 indicated that leakage through the Unit 2 primary containment wall has been a longstanding open issue. The leak was initially identified in 1998 when water leakage was noticed on the external side of the primary containment wall. The leakage was approximately 2025 drops per minute at the primary location and multiple areas near the 180 degree azimuth at construction joints on elevations 813 and 795. Another minor leak was noticed at a similar location near the 0 degrees azimuth. The condition was documented in AR 2269. The source of water leakage was determined to be a weld on a 2 fuel pool cooling drain line and work order 98109950 was initiated to repair the weld. The work was not scheduled and the work order was eventually cancelled. In 2010, the leakage was documented again in AR 1086083. A technical evaluation documented as ATI14709531847 in 2014 concluded that there was no adverse impact on structural adequacy of the containment. The technical evaluation stated that the leakage was to be repaired in the upcoming outage through work order 855785. Action request 2420888 was written in December 2014 to re-enter the condition in the CAP. It recommended corrective actions for liner ultrasound testing every other refueling outage, completion of weld repair, and performance of a technical evaluation for structural impact on the concrete, reinforcing steel, tendons, and liner. The technical evaluation assignment was closed to the evaluation documented under ATI14709531847 discussed above. The corrective action assignment for the weld repair was closed to a work order which has not been completed to-date. Based on the inspectors review, the licensee has deferred the actions to correct this condition identified in 1998. The inspectors question whether the continuous leakage could lead to deterioration of the concrete, corrosion of the reinforcement, or degradation of post tensioned tendons if it enters the tendon sheath or trumpet area; and therefore a condition adverse to quality. This unresolved item remains open pending additional inspector review of the issue with respect to regulatory requirements. Planned Closure Action: Inspectors will seek additional information from the licensee and the NRC will perform internal reviews to evaluate compliance with NRC regulations
05000373/FIN-2018003-062018Q3LaSallePotential Failure to Inspect Containment Post-Tensioned Tendons per Code Requirements and to Follow Corrective Action Program ProcessVertical and horizontal post tensioned tendons, along with reinforcing steel, are required to maintain structural integrity of the primary containment. There are a total of 120 vertical post tensioned tendons along the periphery of the primary containment wall, including 60 Group C tendons and 30 each of Groups A and B. Section 5.5.6 of the Technical Specifications describes the Inservice Inspection (ISI) program for post tensioning tendons and states that the Tendon Surveillance Program shall be in accordance with ASME Section XI, Subsection IWL as required by 10 CFR 50.55a. One Group B tendon (V213B) on Unit 1 was inspected in 1999 and according to the inspection records, water was identified on all components of the tendon. No presence of water is one of the acceptance criteria per Subsection IWL of the ASME Section XI. No condition report was found for this adverse condition. Subsequently, a condition involving degraded vertical tendons was identified during inspections in 2003 and documented in AR 157920. The degradation consisted of broken wires. The Root Cause Report (RCR) for this condition noted that 11 Group A tendons were found degraded and water induced corrosion was the root cause for tendon degradation. The evaluation concluded that five Group A tendons and all 30 Group B tendons in each unit were not susceptible to water intrusion because they were protected by welded covers. These tendons with welded covers were also determined to be inaccessible, and therefore exempt from future inspections requirements in accordance with provisions of ASME Section XI, IWL. The RCR did not address the condition of water found during the Group B tendon inspection in 1999. Additionally, to verify this assumption of welded covers providing protection from water intrusion, a corrective action was generated to inspect one Group A and one Group B inaccessible tendons during the next outage. Pertaining to inaccessible tendons, the inspectors noted the following requirements of 10 CFR 50.55a(b)(2)(viii)(E): Concrete containment examinations: Fifth provision. For Class CC applications, the applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or the result in degradation to such inaccessible areas. For each inaccessible area identified, the applicant or licensee must provide the following in the ISI Summary Report required by IWA6000: (1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation; and (3) A description of necessary corrective actions. After the licensee identified degraded group A tendon locations, to comply with the provision of 10 CFR 50.55a, the licensee documented in its 90 day post outage ISI reports information on the degraded A tendons in 2004 and 2005 for units 1 and 2, respectively. This information included an assumption that the extent of degradation did not apply to the Group B tendon locations because of a welded cover at locations that precluded entry of water. Additionally, a corrective action, CA 15792033, was generated to inspect one Group A and one Group B tendon during the next refueling outage to verify this assumption. The corrective action was closed without inspection of any Group B tendon based on a management decision following satisfactory inspection of a Group A tendon in 2006. The licensees decision failed to take into account the fact that the most recent inspection of a Group B tendon showed presence of water on tendon components and also that the welded closure details were different for tendons in the two groups. Subsequently, the licensee identified a concern regarding inadequate closure of this corrective action during its reviews for the license renewal application in 2014. Specifically, the licensee wrote AR 1658189 to document that due to the differences in the welded cover designs, the results of the Group A tendon inspection may not be applicable to Group B tendons. Therefore the critical assumption regarding the adequacy of Group B tendon covers remained unverified. In particular, the Group B tendon cover used dissimilar metal welds and water was found inside the cover during the most recent inspection. The licensee identified actions to perform inspections on two of the Group B tendons on each unit in addition to inspecting the tendon V213B where water was initially found. These actions were categorized as action tracking items (ACITs), items that do not represent conditions adverse to quality. Since water was found on all tendon components during the last inspection of a Group B tendon, and water induced corrosion was found to be the root cause of many tendon failures, the assumption in the RCR that the welded covers would prevent water intrusion needed to be validated through inspections. This unresolved item remains open pending additional inspector review of the issue with respect to regulatory requirements. Planned Closure Action: Inspectors will seek additional information from the licensee and the NRC will perform internal review to evaluate compliance with NRC regulations.
05000373/FIN-2018003-052018Q3LaSalleFailure to Translate Fuel Oil Relief Valve Setting into Design Drawing of Record.The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion III, Design Control, for the licensees failure to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to accurately translate the Division III EDG fuel oil relief valve set point from the design drawing of record, VPF341110, to the fuel oil pressure operator rounds alert value in the Division III EDG operating procedures.
05000373/FIN-2018003-042018Q3LaSalleFailure to Manage the Increase in Risk During a Battery Charger Capacity TesThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.65(a)(4) for the failure to manage risk when the licensee failed to adhere to procedure WCAA101, Revision 28, On-line Work Control Process. Specifically, procedural requirements regarding a dedicated operator for manual restoration actions and written instructions to credit the availability of the A RHRSW pump during the battery charger testing were not met.
05000373/FIN-2018003-032018Q3LaSalleFailure to Establish Goals to Monitor Steam Tunnel Check DampersIntroduction: The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50.65(a)(1) for the licensees failure to establish goals to monitor the performance of steam tunnel check dampers. Specifically, the licensees goals for functional failure and condition monitoring could always be satisfied given a two years monitoring period with only one testing opportunity.
05000373/FIN-2018003-022018Q3LaSalleFailure to Establish an Appropriate Inservice Testing ProcedureThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to prescribe procedures that were appropriate to the circumstances, for activities affecting quality, that included appropriate quantitative or qualitative acceptance criteria for determining that important activities had been satisfactorily accomplished. Specifically, the CSCS bypass line isolation valve IST procedure did not contain acceptance criteria to verify the necessary valve obturator movement.
05000373/FIN-2018003-012018Q3LaSalleFailure to Establish Heat Exchanger Inspection Procedures Appropriate for the CircumstancesThe inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions Procedures, and Drawings, for the licensees failure to ensure that activities affecting quality were prescribed by documented procedures of a type appropriate to the circumstances. Specifically, the licensee failed to ensure that procedure ERAA3401002 appropriately accounted for partially blocked HX tubes identified during HX inspections.
05000254/FIN-2018003-022018Q3Quad CitiesFailure to Follow Procedures for Forced Helium Dehydration of a Multipurpose CanisterThe 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-2018003-012018Q3Quad CitiesFailure to Maintain the Design Basis for Residual Heat Removal Torus Suction ValveThe 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.
05000461/FIN-2018002-052018Q2ClintonMinor ViolationThe inspectors reviewed AR 4116223, Blown Fuses during CPS 9080.23 8.4 for Fast Transfers. The inspectors selected this sample for review due to repetitive fuse failures within the safety-related Division 3 NUS Modules dating back to 2013. As appropriate, the inspectors verified the following attributes during their review: complete and accurate identification of the problem in a timely manner commensurate with its safety significance and ease of discovery; consideration of the extent of condition, generic implications, common cause, and previous occurrences; evaluation and disposition of operability/functionality/reportability issues; classification and prioritization of the resolution of the problem commensurate with safety significance; identification of corrective actions, which were appropriately focused to correct the problem; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. Description: While reviewing the historical ARs associated with the NUS fuse failures, the inspectors discovered licensee information indicating the NUS fuse failures were likely caused by voltage/current transients within the upstream, safety-related 480V to 120V regulating transformer. The purpose of the transformer was to regulate voltage and current to the downstream components including the NUS modules. However, degradation in the transformers ability to regulate voltage and current levels could create a condition where the voltage and current levels exceeded the NUS fuse rating causing fuse failure. The licensee documented the potential transformer degradation issue on September 20, 2013, in AR 1561455, Division 3, Group 1 Instruments Found De-energized during CPS 9080.23, Specifically, the licensee stated, The most probable cause of the failure of the NUS modules was the transient voltage overshoot of the regulating transformer causing the transient protection varistors on the five NUS modules to actuate, drawing a near fault current until the individual and line feed fuses blew. Station procedure PI-AA-125, Corrective Action Program, defined equipment failure as, damage to or degradation of a system, structure or component that may cause or contribute to the event. Based on the information documented in AR 1561455, the licensee identified transient voltage overshoots in the 480V to 120V regulating transformer, which was a degraded condition causing the NUS modules to fail. Per the licensee definition this would constitute an equipment failure. No further action was taken to identify and correct the regulating transformer degradation until the transformer failed on March 18, 2018, impacting multiple pieces of safety-related Division 3 equipment. Minor Violation: Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to this requirement, on September 20, 2013, the licensee identified a failure of the 480V to 120V regulating transformer, which manifested itself as a voltage overshoot causing the failure of the NUS modules, but failed to take actions to correct the condition. On March 18, 2018, the regulating transformer subsequently degraded further causing it to fail in a manner that tripped the upstream breaker and impacted additional pieces of safety-related Division 3 equipment. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. Specifically, the inspectors determined that although the transformer failure affected Division 3 equipment, the failure would not have impacted the Division 3 equipments ability to respond to a DBE or the capability to shut down the reactor and maintain it in a safe shutdown condition. Enforcement: The failure to comply with 10 CFR 50, Appendix B, Criterion XVI, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy.
05000461/FIN-2018002-042018Q2ClintonMinor ViolationThe inspectors reviewed AR 4082490, Reactor SCRAM from Trip of 1AP07EJ. The inspectors selected this sample for review due to the safety significance of the Division 1 and 2 safety-related transformers, which is the subject of the AR. This review focused on actions associated with newly installed Divisions 1 and 2 4160V to 480V transformers. As appropriate, the inspectors verified the following attributes during their review of the licensee's corrective actions for the above condition report and other related condition reports: classification and prioritization of the resolution of the problem commensurate with safety significance; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. The inspectors discussed the corrective actions and associated evaluations with licensee personnel. As a result of this review the inspectors identified the following minor violation: Minor Violation: The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow procedures associated with the CAP. Specifically, on May 10, 2018, the licensee identified discrepant results while testing safety-related transformers 0AP06E2 and 1AP12E2 but failed to enter this issue into the CAP in accordance with PIAA120, Issue Identification and Screening Process, Revision 8, Step 4.3.4, until prompted by the inspectors. Instead, the licensee evaluated the discrepant results within the work order and found them to be acceptable. The licensee generated AR 4137994, Insulation Power Factor Results For 0AP06E & 1AP12E, dated May 15, 2018, after being challenged by the inspectors regarding the need to enter the discrepant test results into the CAP. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The failure to document the discrepant values in the CAP did not adversely impact the safety-related transformers. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion II, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000461/FIN-2018002-032018Q2ClintonMinor ViolationDuring the inspection quarter, the inspectors reviewed a significant number of licensee CAP documents to assess the following performance attributes: complete, accurate, and timely documentation of the identified problem in the CAP; evaluation and timely disposition of operability and reportability issues; consideration of extent of condition and cause, generic implications, common cause, and previous occurrences; classification and prioritization of the problems resolution commensurate with the safety significance; and identification of negative trends associated with human or equipment performance that can potentially impact nuclear safety. Minor Performance Deficiency: The inspectors determined that issues which could impact the operability of TS-related equipment were generally entered into the CAP in a timely manner. However, operability determinations were not always performed within the timeframes established in Section 4.1 of Procedure OPAA108115, Operability Determinations (CM1), because some issue reports were not directly routed to the operating shift crew for review. The CAP software program used by the licensee included a standard set of questions which were normally answered by the individual entering the issue into the CAP. Depending on the answers to the questions, the CAP document routing could automatically bypass the operating shift crew for review. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The inspectors did not identify any instance where the failure to perform a timely operability determination had a significant consequence on licensed activities. However, the inspectors discussed the vulnerability between the CAP and the operability determination process with the licensee. The licensee implemented a standing order to require a shift review by the operating crew of condition reports not directly routed to the shift. In addition, the licensee is trending the number of condition reports which are returned by the Station Ownership Committee to the shift for review to determine whether further actions are warranted. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard.
05000461/FIN-2018002-022018Q2ClintonFailure to Establish Adequate Leak Rate Test Procedures for Shutdown Service Water Isolation Valve TestingThe inspectors identified a Green finding and a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to ensure testing of the shutdown service water (SX) isolation valves was performed with procedures which: (1) incorporated the requirements and acceptance limits contained in applicable design documents; and (2) included provisions for assuring that all prerequisites for the given test had been met. Specifically, the licensee failed to establish leak rate test procedures for SX boundary valves 1CC075A and 1CC076A that included provisions for ensuring the required differential test pressure was met during testing.
05000461/FIN-2018002-012018Q2ClintonFailure to Perform an Operability Determination for Suspected Leakage Past Shutdown Service Water Isolation ValvesThe inspectors identified a Green finding for the failure to perform an operability determination in accordance with Procedure OPAA108115, Operability Determinations (CM1). Specifically, the licensee failed to determine and document the operability status of the shutdown service water system and the ultimate heat sink after the discovery of leakage past the 1CC075A and 1CC076A isolation valves.
05000461/FIN-2018001-022018Q1ClintonFailure to Identify a Single Point Vulnerability Results in Manual Reactor ScramA self-revealed Green finding was identified for the licensees failure to identify a single point vulnerability in accordance with procedure ERAA2004, Revision 1. Specifically, during a site single point vulnerability review of the feedwater system, the licensee failed to identify a single point vulnerability that subsequently resulted in a loss of a feedwater heating string. The loss of the heater string caused a drop in temperature in the reactor of 100 degrees which prompted a manual scrambe initiated by the operators
05000461/FIN-2018001-012018Q1ClintonFailure to Follow Procedure Results in Unplanned Reactor Core Isolation Cooling UnavailabilityA self-revealed Green finding and 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 identified when the licensee failed to follow station procedure Clinton Power Station (CPS) 9030.01C034, RCIC (Reactor Core Isolation Cooling) Steam Line Flow E31N683A(B), E31N684A(B), Checklist. Specifically, the licensee failed to reset the isolation logic for the RCIC steam line outboard isolation valve prior to turning on the breaker for this valve. This resulted in the isolation of the steam supply to RCIC causing RCIC to become unavailable,and elevating the plant risk to Yellow.
05000255/FIN-2017004-012017Q4PalisadesImproperly Connected M&TE Leads to Unexpected AFU Fan TripA finding of very low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when the licensee failed to follow step 5.4.4.b of Technical Specification surveillance procedure RT85DA, Control Room Emergency Ventilation Filtration Testing A Train. Specifically, the licensee failed to properly connect maintenance and test equipment (M&TE) across flow transmitter test taps which caused V26A, the air filter unit (AFU) VF26A fan, to stop 17 seconds after operators started the fan from the control room. The licensee entered this issue into their Corrective Action Program (CAP) as condition report (CR) CRPLP201705234. Corrective actions included coaching the vendor on ensuring M&TE is properly connected to plant equipment and ensuring suitable field oversight of the vendor during re-performance of the surveillance.The issue was determined to be more than minor in accordance with IMC 0612, Appendix B, Issue Screening, because it was associated with the Barrier Integrity cornerstone attribute of Human Performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as having very low safety significance (Green) in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 3, because the inspectors answered "No" to all screening questions. The finding had across-cutting aspect in the area of Human Performance, in the Field Presence aspect, for the failure to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel (H.2).
05000237/FIN-2017004-012017Q4DresdenFailure to Follow Procedure,Results in Non-Functional Fire DoorThe inspectors identified a finding of very-low safety significance and associated NCV of Technical Specification 5.4.1.c for the licensees failure to implement the established Fire Protection Program procedures which ensure Fire Barrier Integrity. Specifically, the licensee ran an electrical cable through the doorway of an automatically closing fire door. This was contrary to Procedure DFPP 417501, which requires in part that fire doors must not be blocked open by props or any other material in its closing path. The licensee took immediate actions to restore the fire door, by removing the obstruction and entered the issue into their Corrective Action Program (CAP). The inspectors determined that the performance deficiency was more-than-minor because it affected the Mitigating Systems cornerstone objective since the electrical cable could have prevented the fire door from performing its function. The finding was of very-low safety significance per Task 1.4.3A of IMC 0609, Appendix F. Specifically, the total combustible loading on both sides of the affected fire door was representative of a fire duration less than 1.5 hours. The inspectors determined the finding had a cross-cutting aspect in the area of Human Performance, associated with the Training component, because the licensee failed to provide training and ensure knowledge transfer to maintain a knowledgeable, technically competent workforce and instill nuclear safety values. Specifically, the licensee believed the performance deficiency was caused by the one of the new temporary contractors brought onto the site to work in support of the D2R25 refueling outage. (H.9)
05000461/FIN-2017003-052017Q3ClintonFailure to Establish Secondary Containment Prior to Entering MODE 2The inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS LCO 3.0.4, for the failure to follow station procedure CCAA201, Plant Barrier Control Program, Revision 11. Specifically, the licensee entered MODE 2 from MODE 4 without meeting the requirements of LCO 3.0.4 for entering a mode when an applicable LCO is not met. The licensee had not met LCO 3.6.4.1 because the doors to the B reactor water cleanup room were both opened instead of being closed to make secondary containment operable as required in MODE 2. The licensee entered this issue into their CAP as AR 04017613. As corrective actions, the licensee planned to conduct training for site personnel.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely 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 failure to follow the station procedure by not identifying that the open doors required a plant barrier impairment (PBI) permit that would have identified the doors as a constraint to entering MODE 2 resulted in the unit transitioning to MODE 2 with the secondary containment inoperable. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone and determined 5 to be of very low safety significance because the finding only represented a degradation of a radiological barrier function provided for auxiliary building. The inspectors determined that this finding affected the cross-cutting are of human performance in the aspect of training, where the organization provides training and ensures knowledge transfer to maintain a knowledgeable, technically competent work force and instill nuclear safety values. Specifically, station personnel did not know the process for routing a PBI permit and did not know when a PBI permit was required. (H.9)
05000461/FIN-2017003-042017Q3ClintonFlow Control Valves Not Locked Out Results in Reactor Recirculation Pump RunbackThe inspectors documented a self-revealed finding of very low safety significance and an associated NCV of Technical Specification (TS) 5.4.1, Procedures, for the licensees failure to establish sufficient instructions in station procedure Clinton Power Station (CPS) 3103.01, Feedwater (FW), Revision 31e, for changing modes of operation for the nuclear steam supply system. Specifically, the station procedure did not provide instructions requiring the locking out the flow control valves (FCVs) to prevent a reactor recirculation FCV runback while changing the feedwater pump lineup resulting in an unexpected plant transient and 9.2 percent change in reactor power. The licensee entered this issue into their corrective action program (CAP) as Action Request (AR) 04007861. As corrective actions, the licensee revised their CPS 3103.01 procedure to require that the FCVs be locked out prior to shifting reactor feed water pumps. The performance deficiency was more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012,because the finding was associated with the procedure quality attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the failure to have adequate procedures for shifting feedwater pumps during a plant shutdown on May 7, 2017, resulted in an unexpected recirculation pump run back and a 9.2 percent change in reactor power. Using IMC 0609, Attachment 4, Initial Characterization of Findings, andAppendix A, The Significance Determination Process for Findings At-Power, issuedJune 19, 2012, the finding was screened against the Initiating Events cornerstone and determined to be of very low safety significance because the event did not cause a reactor scram. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of conservative bias, where individuals use decision making practices that emphasize prudent choices over those that are simply allowable and a proposed action is determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the procedure provided for the option to lockout the reactor 3 recirculation flow control valves if deemed necessary during a shift of the reactor feedwater pumps and the operations crew did not make the prudent choice of locking out the valves before determining that it was safe to proceed. (H.14)
05000461/FIN-2017003-032017Q3ClintonFailure to Perform Engineering Evaluation to Determine the Cause of Failure of SnubbersThe inspectors identified a finding of very low safety significance and an associated NCV of Title 10 Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to demonstrate compliance with the requirement as prescribed in procedure ERCL330, CPS Snubber Program, Revisions 1 and 2. Specifically, the licensee failed to perform engineering evaluations to determine the cause of failure of snubbers that did not satisfy their functional testing acceptance criteria. The licensee entered this issue into their CAP as ARs 04015242 and 04041302. As corrective actions, the licensee evaluated the components affected by the failed snubber and determined that no operability issues existed. The performance deficiency was determined to be more-than-minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it was associated with the Mitigating Systems cornerstone attribute of Protection against External Factors and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability for mitigating systems to respond to initiating events. Specifically, compliance with ERCL330 would ensure the failed snubber wasevaluated for the cause of failure, to ensure the licensee identified other snubbers that may have been vulnerable to the same type of deficiency. This would ensure that any potential undesired loading on the piping system could be avoided and the affected safety-related residual heat removal and reactor water cleanup piping systems could continue to perform their design function of maintaining the pressure boundary and structural integrity following a postulated design basis seismic event. The inspectors determined the finding could be evaluated using the Significance Determination Processin accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, dated June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, for the Mitigating Systems cornerstone and then Exhibit 4, External Events Screening Question. The finding screened as having very low safety significance because in each instance, the inspectors answered No to Questions 1 and 2 ofExhibit 4. The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of consistent process, where individuals use a consistent, systematic approach to make decisions. Specifically, the licensee failed to establish a systematic approach to evaluating snubbers that did not meet the acceptance criteria to ensure all required aspects were addressed. (H.13)
05000461/FIN-2017003-022017Q3ClintonFailure to Adequately Control Access in Locked High Radiation AreaA finding of very low safety significance and an associated NCV of TS 5.4.1 was self-revealed when individuals failed to adequately control access in locked high radiation areas (LHRAs). Specifically, the failure to meet all of the requirements of Procedure RPAA460, Attachment 5, represented a failure to comply with Radiation Work Permit CL1700518, C1R17 (Drywell) DW Bioshield Inservice Inspection Activities. This resulted in four individuals entering a LHRA that they had not been specifically authorized to enter. These individuals entered the incorrect location and were inside the area for approximately 2-3 minutes before they noticed that they were in the incorrect area. The individuals knew that they were in the incorrect location when they could not find the nozzles that they planned on inspecting. The individuals exited the area and were simultaneously told to exit the area by the radiation protection technician (RPT) providing remote coverage which demonstrated that the four workers were not in the authorized work area. Immediate corrective actions taken by the licensee included immediately suspending the work that was scheduled to take place within the bioshield associated with this job. Electronic dosimeters and dosimeters were immediately collected from the individuals that entered the area so the dose that was received could be known. The licensee also interviewed all the individuals that were involved in this bioshield entry, and the RPT that performed the brief. These interviews were conducted to understand which parts of the process associated with entry into LHRAs failed and led to this event transpiring. The licensee entered this event into their CAP as AR 04012075. As corrective actions the licensee planned to observe high radiation area and locked high radiation area briefs, for both in house and traveling RPTs. The licensee also planned to modify the bioshield as-low-as-reasonably-achievable (ALARA) plan template to label all accessible bioshield doors with elevation and azimuth.The inspectors determined that the performance deficiency was more-than-minor in accordance with IMC 0612, Appendix B, because the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone, and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation, in that, the workers entered an area that required the radiation dosimeter to be relocated to the workers knee, and the workers were wearing them on the head for the intended work location. The finding was determined to be of very-low safety significance (Green) in accordance with IMC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008, because: (1) it did not involve as-low-as-reasonably-achievable planning or work controls; (2) there was no overexposure; (3) there was no substantial potential for an overexposure; and (4) the ability to assess dose was not compromised.The inspectors determined this finding affected the cross-cutting area of human performance in the aspect of resources, where leaders ensure that personnel, 6 equipment, procedures and other resources are available and adequate to support nuclear safety. Specifically, radiation protection leadership failed to ensure that the RPT was capable of meeting the expectations for performing the LHRA briefing in accordance with station procedure RPAA460, Attachment 5. (H.1)
05000461/FIN-2017003-012017Q3ClintonMSIV TS Leakage Limits Exceeded Due to Condition Based Maintenance ApproachThe inspectors documented a self-revealed finding of very low safety significance and an associated NCV of TS limiting condition for operation (LCO) 3.6.1.3, for the failure to follow station procedure ERAA200, Preventative Maintenance Program, Revision 3. Specifically, the licensee utilized a condition-based maintenance approach on the main steam isolation valves (MSIVs) that failed to monitor and trend equipment performance so that planned maintenance could be performed prior to the MSIVs exceeding the TS leakage limits. The licensee entered this issue into their CAP as AR 04009845. As corrective actions, the licensee repaired and tested the valves prior to returning the unit to the modes of applicability.The performance deficiency was determined to be more than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, because it impacted the Barrier Integrity cornerstone attribute of configuration control and adversely 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 monitoring and trending of local leak rate tests on the MSIVs did not provide performance data that would allow planned maintenance to the valves prior to the valves failing resulting in exceeding TSleakage requirements for the MSIVs. Using IMC 0609, Attachment 4, Initial Characterization of Findings, and Appendix A, The Significance Determination Process for Findings at Power, Exhibit 2, October 7, 2016, the finding was screened against the Barrier Integrity cornerstone Reactor Containment and did represent an actual open pathway in the physical integrity of reactor containment. The inspectors proceeded to Appendix H, Containment Integrity Significance Determination Process, and determined that it was a Type B finding that was related to a degraded condition that has potentially important implications for the integrity of the containment, without affecting the likelihood of core damage. The inspectors used Figure 6.1, Road Map for LERF based Risk Significance for Evaluation of Type-B Findings at Full Power and determined this finding is of very low safety significance (Green). The inspectors determined that this finding affected the cross-cutting area of human performance in the aspect of design margins, where the organization operates and maintains equipment within design margins. Special attention is placed on maintaining fission product barriers, defense-in-depth and safety related equipment. Specifically, the procedure for testing the MSIVs utilized an administrative limit that provided no margin to correct performance prior the valves becoming inoperable. (H.6)
05000315/FIN-2017002-072017Q2CookLicensee-Identified ViolationTitle 10 CFR 50.71(e) required that the UFSAR be updated to assure that the latest information developed was in the UFSAR. In AR 2010 4194, Unit 1 and Unit 2 Small Break Loss of Cooling Accident (SBLOCA) Analyses, the licensee identified the March 2007 Unit 1 SBLOCA analysis had not incorporated into the UFSAR and was not included in the October 2008 UFSAR update provided to the NRC. The inspectors determined the failure to update the UFSAR by incorporating the newest SBLOCA analyses was contrary to 10 CFR 50.71e. The inspectors reviewed this issue in accordance with NRC IMC 0612 and the NRC Enforcement Policy. Violations of 10 CFR 50.71(e) are disposed using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the UFSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. The inspectors determined the performance deficiency was minor in that failure to update the UFSAR was not willful; did not impact a performance indicator; was not a material condition issue which could lead to a more significant safety issue, and did not impact the Mitigating Systems cornerstone objectives .
05000373/FIN-2017009-012017Q2LaSalleAnchor Darling Double Disc Gate Valve 1E22-F004 and 2E22-F004 Pressed-FitCollar Related 10 CFR Part 50, Appendix B, Criterion III ViolationIn 1990, the licensee had reviewed and accepted the vendors weak link analyses that provided the upper torque and thrust limits for all safety-related ADDDGV in service at the station. This analysis documentedthat the 1E22-F004 and 2E22-F004 valve stems were the weak link valve components in the closing direction (i.e.,provided enough closing thrust, thevalve stems would be the firstcomponent to becomenonfunctional).Therefore, theclosed thrust limit forthe 1E22-F004 and 2E22-F004 valves was approximately 260,000 lbf. The licensee had set up the valves ina manner that would ensure that the valveswould have enough torque and thrust tooperate under design basis conditions while staying below the maximum weak link limits. Maintenance and test records showed that thelicensee consistently verifiedthat these two valves were setup and maintained within this design window. Typical as-found and as-left closed thrust limits ranged from approximately between200,000240,000 lbf.As described in the licensees failure analysis report and as discussed above, the licensee identified that the pressed-fitcollar could relax its pre-load when operating the valve well within the established maximum closed thrust limitations. The licensees failure analysis report estimated that approximately 130,000 lbf was necessary to shift the collar up and relax the pre-load. Therefore,theteam concluded that the licensees weak link analysis was inadequate based upon the 2E22-F004 valve failure and associated failure analysiswhich determined that the pressed-fitcollar was a weaker component as compared to the valve stem. The team did not identify an associated performance deficiencyfor the inadequate weak link analysis. This determination was based upon the weak link analysis originating from the vendor in 1990, licensees review of that analysis, and latent design issue that had not been previously identified within the industry until recently identified by the licensee.Additionally, the team did not identify a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. This determination was based, in part, that correcting the unknown stem collar pre-torque issueafter receiving the 10 CFR Part 21 Flowserve notification would not necessarily have identified and corrected the non-conforming inadequate weak link design control issue. Enforcement: Title10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,inpart that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2, and as specified in the license application, for those structures, systems, and components to which this appendix apply are correctly translated into specifications, drawings, procedures, and instructions.Contrary to the above, since original plant construction, the licensee failed to ensure thatapplicable design basismaximumclosed thrust and torque valuesfor the safety-related Unit 1 and Unit 2 HPCS injection valves (1E22-F004, 2E22-F004)werecorrectly translatedinto specifications. Specifically, it was identified that the stem-to-wedgepre-torque credited within the design could relax by applying closed direction torque and thrust well within the specified design limitbecause that limit was based uponthe wrong weak link component. The loss of the stem-to-wedgepre-torque could subsequently break the wedge pin and result in stem-to-wedgethread degradation ultimately leading to valve failure.The NRC determined that issue was a Severity Level III Violation based upon Section6.1(c)(2) of the Enforcement Policy. Specifically, a system that is part of the primary success path and which functions or actuates to mitigate a design base accident or transient that either assumes the failure of or presents a challenge to the integrity of the fission product barrier not being able to perform its licensing basis safety function because it is not fully qualified.The NRC exercised enforcement discretion in accordance with Sections 3.10 of the Enforcement Policy and Section 3 of Part1 of the Enforcement Manual. Enforcement Policy Section 3.10 states that the NRC may exercise discretion for violations of NRC requirements by reactor licensees for which there are no associated performance deficiencies. This violation was entered into the Corrective Action Programas Issue Report3972901 and has been corrected by replacing the 1E22-F004 and 2E22-F004 valve stems with integral collars.
05000316/FIN-2017002-022017Q2CookUnit 2 CEQ Fan Failed SurveillanceGreen . A finding of very low safety significance was self -revealed on March 23, 2017, when one of the Unit 2 Containment Equalization (CEQ) Fans, 2 HV CEQ 1, failed its surveillance. Technical Specification (TS) 5.4.1, Procedures, requires that the applicable procedures covered in Regulatory Guide 1.33 are established, implemented, and maintained. Regulatory Guide 1.33 requires that maintenance that can affect the performance of safety -related equipment should be properly preplanned and performed in accordance with documented instructions appropriate to the circumstances. Contrary to these requirements, a preventative maintenance activity to grease the backdraft damper bearings of the CEQ fan resulted in the fan being left inoperable until the next scheduled surveillance approximately a month later. Due to inadequate work instructions, the damper was not cycled enough times following greasing, which resulted in a condition where more force than allowed by the Technical Specifications was required to open the damper. Due to an inadequate post -maintenance test, this was not detected until the next surveillance was performed. Upon failure of the surveillance, technicians re -greased the bearings, cycled the damper, and tested it satisfactorily. Although qualified, the technicians who first performed the maintenance were unaware of certain nuances associated with the CEQ fan dampers. This information was not described in the work instructions and the post -maintenance test did not validate the opening force. The issue was entered into the CAP and an apparent cause evaluation was performed by the licensee. The issue was greater than minor because it adversely affected the Procedure Quality attribute of the Mitigating Systems cornerstone. Specifically, the inadequate maintenance procedures adversely affected the availability, reliability, and capability of a system that responds to initiating events to prevent undesirable consequences (i.e., core damage). The finding screened to Green, or very low safety significance, based on IMC 0609 Appendix H, Containment Integrity Significance Determination Process, because CEQ fans are not important contributors to Large Early Release Frequency and Hydrogen Igniters remained available. The inspectors determined there was a cross- cutting aspect associated with the finding, namely, H.5., Work Management. Specifically, the licensee did not identify and manage risk nor coordinate between different work groups when it was recognized the normal maintenance group would not be working on the CEQ Fan. Further, the apparent cause evaluation identified a need to better coordinate the preventative maintenance activities with the surveillance tests.
05000315/FIN-2017002-052017Q2CookInadequate Design Control Measures to Ensure Leakage Remained Within AnalysisGreen . The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to have adequate design control measures verify that the Essential Service Water to Containment Spray (CTS) heat exchanger outlet valves were not leaking in excess of the limits of the Large Break Loss of Coolant Accident (LBLOCA) analysis. This finding was entered into the licensees CAP to evaluate adequate design control measures. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of the CTS system to respond to an initiating event 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 one of the trains of the CTS system. The inspectors did not identify a cross -cutting aspect associated with this finding because it was not reflective of current performance.
05000315/FIN-2017002-032017Q2CookFailure to Identify Parts Subject to a Part 21Green. A self -revealed finding and associated violation occurred on April 2, 2012, when the licensee failed to prevent installation of relays identified in a P art 21. Although the performance deficiency occurred in 2012, the consequence of the error did not manifest until March 2017, when a defective relay caused the Unit 2 control room indicating and display (CRID) 3 inverter to transfer and remain on the alternate power supply. Title 10 CFR 50 Appendix B, Criterion XV requires, in part, that Measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their installation. Contrary to this requirement, on April 22, 2012, the licensee failed to prevent installation of an AMETEK board, PC 201 with a defective relay. This led to a failure of the CRID 3 inverter on March 27, 2012. The licensee replaced the circuit board and restored CRID 3 to an operable status. The inspectors determined that the failure to prevent installation of defective parts into the safety related CRID system was a performance deficiency that warranted a significance determination. Using Attachment 0609.04, Initial Characterization of 4 Findings, dated October 7, 2016, Table 2, the inspectors determined that the finding affected the Mitigating Systems cornerstone. As a result, the inspectors evaluated the finding using IMC 0609, Attachment 1 Exhibit 2, dated June 19, 2012. The inspectors answered no to all the questions, therefore the finding screened as Green. Using Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016, Table 2, the inspectors determined that the finding affected the Mitigating Systems cornerstone. The inspectors did not identify a cross- cutting aspect associated with this finding because it was not reflective of current performance.
05000315/FIN-2017002-012017Q2CookFailure to Ensure the Unit 2 CCW Heat E xchanger Monitoring Program Could Demonstrate Its Continued Operability Between Maintenance IntervalsGreen . The inspectors identified a finding of very -low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulations , (CFR) Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to establish a heat exchanger monitoring program for the Unit 2 east component cooling water (CCW) heat exchanger that demonstrated it would perform satisfactorily in service and remain operable within its required range of physical conditions for the entire interval between heat exchanger maintenance inspections and c leanings. The licensee entered this finding into their Corrective Action Program (CAP) and, after a review of the Ultimate Heat Sink temperatures, determined the Unit 2 East CCW heat exchanger remained operable because the Ultimate Heat Sink temperatures had remained below the point where operability of the heat exchanger could be challenged. The performance deficiency was determined to be more- than- minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and it adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of the CCW system to respond to initiating events to prevent undesirable consequences. Specifically, the monitoring program established for the Unit 2 East CCW heat exchanger did not ensure its availability, reliability, and capability for the entire interval between heat exchanger maintenance inspections and cleanings. The finding screened as of very -low safety significance (Green) because although it affected the design or qualification of the Unit 2 East CCW heat exchanger, it did not result in the loss of operability or functionality of the heat exchanger . The inspectors determined this finding had an associated cross -cutting aspect, Design Margins, in the Human Performance cross -cutting area (H.6) because the licensee did not ensure the Unit 2 East CCW heat exchanger s heat transfer margin was carefully guarded after discovering excessive tube plugging above the acceptance criteria i n 2016. Specifically, special attention was not placed on maintaining the safety -related heat exchanger to ensure it would remain capable of performing its specified safety function within the required range of physical conditions during the entire interval between heat exchanger maintenance inspections and cleanings.
05000315/FIN-2017002-062017Q2CookSingle Point Failure Vulnerability in Annunciator SystemGreen . A self -revealed finding occurred on March 30, 2017, when operation of a work station for the control room annunciators caused a loss of all annunciators in the Unit 1 control room. Specifically, a software error coupled with an overflowing cache caused a single point fai lure of the Unit 1 annunciator. When in us e by a control room operator, Server 1 for the annunciator system failed and transferred functions to Server 2 . Server 2 also failed causing a loss of all annunciators for the Unit 1 control room. The licensee restored the system a few hours later and entered the condition into the corrective action program. The inspectors determined that the failure to design the system to preclude loss of a single active component from causing a loss of the annunciator system was a performance deficiency that warranted a significance determination. Using IMC 0612, the inspectors determined the finding was more than minor because it adversely impacted the mitigating system cornerstone objective to ensure the availability of systems that respond to initiating event. Using IMCC 0609, the inspectors determined that support of the Senior Risk Analyst (SRA) was needed because the condition resulted in the loss of a function, the annunciators. The S RA performed a simple detailed analysis and concluded the finding was of very low safety significance. The inspectors did not identify a cross -cutting aspect associated with this finding because it was not reflective of current performance.
05000315/FIN-2017002-042017Q2CookFailure to Report Deficiencies as Required by 10 CFR 50.46SL IV. The inspectors identified a Severity Level IV Violation of 10 CFR Part 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light -Water Nuclear Power Reactors. Specifically, the licensee failed to report the effects of the errors in the 5 LBLOCA Evaluation Model for the Unit 1 emergency core cooling systems. The inspectors determined that the failure to estimate and report the errors in the LB LOCA analyses were contrary to the requirements of 10 CFR 50.46 and was a performance deficiency. The performance deficiency was determined to be minor because the failure to report was not willful, did not impact a performance indicator, was not a material condition issue which could lead to a more significant safety issue, and did not impact the Mitigating Systems cornerstone objectives. The inspectors determined the failure to report was a Severity Level IV violation in accordance with Section 6.9 of the Enforcement Policy. A cross -cutting aspect was not assigned since the performance deficiency is minor.
05000461/FIN-2016009-032016Q4ClintonFailure to Amend the UFSAR Indicating Choice to Comply with 10 CFR 50.68(b)The team identified a Severity Level-IV NCV of 10 CFR 50.68, Criticality Accident Requirements, Paragraph (b)(8), for the licensee failure to amend the Updated Final Safety Analysis Report (UFSAR) to indicate they chose to comply with 10 CFR 50.68(b). Specifically, in 2005, the licensee chose to comply with 10 CFR 50.68(b) but did not amend the UFSAR following the issuance of the associated license amendment. The licensee captured this issue in their CAP as AR 02741851, reasonably confirmed compliance with 10 CFR 50.68(b) requirements (1) through (7) was maintained, and initiated plans to update the UFSAR to specifically indicate that Clinton Power Station chose to comply with 10 CFR 50.68(b). The Significance Determination Process does not specifically consider the impact to the regulatory process in its assessment of licensee performance. Therefore, it was necessary to address this violation, which potentially impacts the NRCs ability to regulate, using traditional enforcement to adequately deter non-compliance. Specifically, failure to update the UFSAR challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating License Amendment Request, and in preparation for and conduct of inspection activities. The team determined the traditional enforcement violation was a Severity Level-IV violation in accordance with Section 6.1.d.3 of the Enforcement Policy because the un-updated UFSAR had not been used to evaluate a facility or procedure change that resulted in a condition evaluated as having low-to-moderate or greater safety significance by the Significance Determination Process. However, it had a material impact on safety or licensed activities. Specifically, the un-updated UFSAR could be used to perform evaluations of facility or procedure changes, which would have the potential to result in unacceptable conditions and/or regulatory decisions. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000461/FIN-2016009-012016Q4ClintonNon Conservative Control Room Radiological Habitability AssessmentThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation and air-conditioning (HVAC) system outside air intake design resulting in a calculated control room dose following a loss of coolant accident that exceeded the applicable limit. The licensee captured this issue in their CAP as AR 02742442, completed an operability evaluation, and issued an NRC event notification. The performance deficiency was determined to be more-than-minor 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 performance deficiency resulted in the control room expected dose following a loss of coolant accident to exceed the applicable limits prompting an operability evaluation. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago.
05000461/FIN-2016009-022016Q4ClintonFailure to Scope SFP Temperature and Level Instruments into the Maintenance Rule ProgramThe team identified a finding of very-low safety significance (Green) and an associated NCV of Paragraph (b)(2)(i) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensee failure to scope non-safety related mitigating structure, systems, and components (SSCs) used within an emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an EOP used spent fuel pool (SFP) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their CAP as AR 02736193, performed an extent of condition to identify any other SSC addition to the EOPs requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected SSCs into the Maintenance Rule Program scope. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of SSC performance 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, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the SFP temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected EOP in 2015.
05000461/FIN-2016009-042016Q4ClintonFailure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator ActionsThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to verify the adequacy of design assumptions related to time critical operator actions made in calculations associated with the control room HVAC and RHR emergency SFP cooling functions. Subsequently, it was determined that operators did not fully understand the control room HVAC system operational demands and that the operational assumptions of the RHR emergency SFP cooling design were unrealistic. The licensee captured these issues into the CAP as AR 02739012, AR 03943566, and AR 02741909; reasonably demonstrated that SFP makeup sources would be available to cope with a prolonged loss of SFP cooling; conducted operator training; and provided refined procedural guidance to ensure the control room HVAC system would be operated consistent with the design assumptions. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely 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 pilot validations of the control room HVAC system operational assumptions demonstrated a significant reduction in margin due to, in part, a lack of operator understanding of the operational assumptions. Additionally, a preliminary review of procedures associated with SFP cooling and RHR determined the operational assumptions of the calculation related to RHR emergency SFP cooling were not bounding. The team determined that this finding was of very low safety significance (Green). Specifically, the control room HVAC system finding example only represented a degradation of the radiological barrier function provided for the control room in that it did not affect the control room barrier function against smoke or a toxic atmosphere. In addition, the finding example related to emergency SFP cooling did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the operation and engineering organizations did not effectively communicate and coordinate their respective roles in developing the control room HVAC system validation in a manner that supported nuclear safety.
05000461/FIN-2016009-052016Q4ClintonFailure to Promptly Identify that the Incapability of the RHR Design to Support TS Operability Requirements Was a CAQThe team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to promptly identify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit, RHR could not be realigned from shutdown cooling mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and AR 03948042, and planned to submit a License Amendment Request to align TS requirements with the design capabilities. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring TS functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding: (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out-of-service for greater than their TS allowed outage time; (3) involved non-TS trains of equipment; (4) involved a degradation of a functional RHR auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross-cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago.
05000461/FIN-2016009-062016Q4ClintonFailure to Follow the Operability Determination Process Following the Identification of a Control Room HVAC System Design IssueThe team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, for the licensee failure to follow the operability evaluation procedure after the identification of a significant design error associated with the control room HVAC system. Specifically, the licensee did not identify the affected safety function, and promptly restore or confirm system operability. The licensee captured these issues into the CAP as AR 03948266 and performed a preliminary engineering evaluation using another alternative analytical methodology that reasonably determined the control room HVAC system remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely 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 performance deficiency resulted in a condition where reasonable doubt on the operability of the control room HVAC system remained following the identification of a significant design error. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team identified that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not provide training to maintain a knowledgeable workforce that would facilitate an adequate implementation of the operability evaluation process following the identification of a non-conforming design-related issue.
05000255/FIN-2016009-012016Q3PalisadesFailure to Document 50.59 Evaluation for Removal of Eight Hour Operator Rounds from the FSARThe inspectors identified a Severity Level IV, Non-Cited Violation of Title 10 of the Code of Federal Regulations (CFR), Part 50.59, Changes, Tests, and Experiments, and an associated finding of very low safety significance (Green) for the licensees failure to maintain records of a change in the facility which included a written evaluation that provided the bases for the determination that the change did not require a license amendment. Specifically, the licensee failed to have a written evaluation that provided the bases for why removal of the 8-hour operator rounds credited to detect a Spent Fuel Pool (SFP) dilution event from the Final Safety Analysis Report did not require a license amendment. The licensee entered this issue into their Corrective Action Program (CAP) as CR-PLP-2016-03055 and issued a standing order to log SFP level every eight hours as an immediate corrective action. The licensees planned corrective actions include preparation of a 10 CFR 50.59 evaluation for the change. The inspectors determined that the failure to perform a 10 CFR 50.59 evaluation for the change to the Final Safety Analysis Report which removed the eight hour operator rounds credited to detect a SFP dilution event was contrary to 10 CFR 50.59(d)(1), and was a performance deficiency. The inspectors determined the performance deficiency was more than minor, and a finding, because it was associated with the barrier integrity cornerstone attribute of Configuration Control and adversely affected the associated Cornerstone Objective of ensuring that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the removal of the 8-hour operator rounds is associated with the boron concentration reactivity control in the SFP and could adversely affect the fuel claddings function to protect the public from radionuclide releases. In addition, the associated violation was determined to be more-than-minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The inspectors evaluated the finding using IMC 0609 Appendix A, The Significance Determination Process (SDP) for Findings at Power, dated June 19, 2012, Exhibit 3, for the Barrier Integrity cornerstone and were directed to further evaluate the significance of the finding using IMC 0609 Appendix M, Significance Determination Process Using Qualitative Criteria, dated April 12, 2012. The inspectors performed the qualitative evaluation described in IMC 0609, Appendix M, and determined the significance of the finding to be of very low safety significance (Green) by considering the availability of other measures the licensee had in place to detect a SFP dilution event. In accordance with Section 6.1.d of the NRC Enforcement Policy this violation is categorized as Severity Level IV because the resulting changes were evaluated by the SDP as having very-low safety significance (i.e., Green finding). The inspectors determined the associated finding had a cross-cutting aspect in the area of Human Performance because the licensee did not ensure their staff were adequately trained in the implementation of the 10 CFR 50.59 rule. Specifically, the licensee staff did not realize that a change which fundamentally alters the existing means of performing or controlling design functions (removal of the 8-hour operator rounds for detecting a SFP dilution event in lieu of an automatic alarm) is adverse and requires an evaluation.
05000266/FIN-2016002-022016Q2Point BeachSubmerged Safety-Related EDG Fuel Oil Transfer Pump CablesA finding of very low safety significance and associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors, for the failure to maintain emergency diesel generator (EDG) fuel oil transfer pump safety-related cables in an environment for which they were designed. Specifically, the licensee allowed the safety-related cables to be submerged in water, which was outside of their design, in manhole Z066B. The licensees corrective actions included pumping the water out of the manholes, repairing the failed sump pump, level switch, and alarm circuit; and performing an engineering evaluation to quantify the level of degradation as a result of the submergence. The performance deficiency 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, Attachment 0609.04, Initial Characterization of Findings, issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions to screen the finding. The finding screened as of very low safety significance (Green) because the inspectors answered "Yes" to the question does the SSC maintain its operability or functionality. Specifically, the submergence of the G01 and G02 EDG fuel oil transfer pump cables did not render the transfer pumps inoperable. This finding has a cross-cutting aspect Evaluation (P.2) in the area of problem identification and resolution, because the licensee did not thoroughly evaluate problems to ensure that resolutions address causes and extent of conditions, commensurate with their safety significance. Specifically the licensee failed to thoroughly investigate and prioritize the failure of the manhole alarm and pumping system according to the safety significance of the cables contained within the manholes which led to prolonged and unevaluated submergence of the cables.
05000266/FIN-2016002-032016Q2Point BeachSuitability of Reactor Protection System and Engineered Safeguards System ComponentsDuring the review of the Reactor Protection System (RPS), the inspectors identified an Unresolved Item (URI) associated with components in both units RPS and engineered safeguards (ESF) system which contained components known to degrade with age, including electrolytic capacitors. In some cases, these components may have been installed as original plant equipment. During the inspectors review of system health reports associated with both Units 1 and 2 RPS, and ESF system as an extent of condition review, the inspectors identified a URI associated with components in hundreds of safety-related RPS and ESF printed circuit boards, power supplies, amplifiers, transmitters, and other related components that potentially exceeded their design criteria for the time period that the components were installed for which no evaluations existed. The inspectors determined that this was an issue of concern in which more information was needed to determine if the issue constituted one or more violations of NRC requirements. Specifically, the inspectors determined that subcomponents, including but not limited to electrolytic capacitors, were installed in both safety trains of both units RPS and ESF components, in some cases for over 40 years without any documented evaluation of age-related degradation mechanisms. The inspectors needed to evaluate the licensees operability determinations that resulted from this inspection activity, any engineering evaluations to provide justification for suitability with respect to design control, recovery plans, a review of the proposed preventative maintenance activities, current failure rates and drift trending, and any other information provided by the licensee that may provide a technically defensible basis for the continued operation. The issue is unresolved pending further NRC review of the licensees evaluation.
05000266/FIN-2016002-062016Q2Point BeachIncorrect Wiring Causes Transformer LockoutA finding of very low safety significance and associated NCVs of TS 3.8.1, AC Sources-Operating and TS 3.8.2, AC Sources-Shutdown, were self-revealed for the licensees failure to follow procedure RMP 90569B, 1X03, Protective Relay Calibration and Testing. Specifically, a wiring error in the 1X03 connection box, which occurred in 2013, caused the 1X03 transformers differential protection circuity to lockout the transformer at current levels below the design protection values. The licensees corrective actions included correcting the improper wiring in the 1X03 connection box and evaluating other work performed by the same vendor during that timeframe. The inspectors determined that the finding was more than minor because it was associated with the Initiating Events cornerstone attribute of Equipment Performance and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically, the lockout of 1X03 caused a loss of one of the licensees offsite power lines and also caused a loss of power to multiple station battery chargers placing Unit 2 into limiting condition for operation (LCO) 3.0.3. 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, dated June 19, 2012, and Appendix A, The Significance Determination Process for Findings At-Power, Exhibit 1, Initiating Events Screening Questions, dated June 19, 2012. The inspectors answered Yes to the Support System Initiators question; therefore, a Detailed Risk Evaluation was required. Based on the conclusions in the Detailed Risk Evaluation, the SRA determined that the finding was of very low safety significance (Green). This finding has a cross-cutting aspect of Avoid Complacency (H.12), in the area of Human Performance, for failing to implement appropriate error reduction tools. Specifically, the incorrectly performed procedure step, in RMP 9056-9B, clearly specified which terminal point to land the wires on, the terminal points were clearly labeled, and the step required a concurrent verification; however, even with those barriers in place, the task performers still landed the wires on the wrong location.
05000266/FIN-2016002-052016Q2Point BeachFuel Assembly Move Sequence Planned IncorrectlyA finding of very low safety significance was identified by the inspectors, for the licensees failure to follow procedure REI 26.0, Fuel/Insert/Component Movement Planning. Specifically, the licensee failed to follow procedure REI 26.0, Step 5.5.7.b, which verified that the licensee would not place fuel assemblies with cooling times less than 295 days into spent fuel pool rack foot locations. The licensees corrective actions included completing additional spent fuel moves, which placed the spent fuel pool into an appropriate configuration. The inspectors determined that the finding was more than minor, because, if left uncorrected, it had the potential to become a more significant safety concern. Specifically, if the inspectors had not questioned the licensee about spent fuel pool rack foot locations, the spent fuel pool would have remained in an incorrect configuration. The inspectors concluded this finding was associated with the Barrier Integrity cornerstone. 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, dated June 19, 2012, and Appendix L, B.5.b Significance Determination Process, Table 2 Significance Characterization, The inspectors determined that the finding did not meet the criteria in Table 2 for a Greater-Than-Green significance; therefore, the finding was of very low safety significance (Green). This finding has a cross-cutting aspect of Avoid Complacency (H.12), in the area of Human Performance, for failing to implement appropriate error reduction tools. Specifically, the licensee became desensitized to overriding fuel placement constraints and failed to implement effective human performance tools to prevent the error.
05000266/FIN-2016002-042016Q2Point BeachViolation of Technical Specifications During Mode 4 Entry with LCO 3.6.6 Not MetA finding of very low safety significance and associated NCV of Technical Specification 3.0.4 was identified by the inspectors for the licensees failure to follow procedure OP 1A, Cold Shutdown to Hot Standby Unit 1 and checklist CL 2C, Mode 5 to Mode 4 Checklist. Specifically, the licensee entered Mode 4 from Mode 5 without meeting the requirements of LCO 3.0.4 for entering a Mode when an applicable LCO is not met. The licensee had not met LCO 3.6.6 because the control switches for two out of the required four containment accident recirculation fans were in their pullout position instead of the required automatic position. Corrective actions for this event included restoration of accident cooler fan control switches to automatic. Additional corrective actions included: performance of an apparent cause evaluation; changes to the licensees ORT 3 test procedures to restore accident fan cooler switches after completion of testing; updating OP 1A to include performance of a control room shift turnover checklist prior to changing modes; and planned enhancements to CL 2 series procedures to strengthen a note on the responsibility of the SRO when ensuring operability of LCOs. The inspectors determined that the finding was more than minor because it was associated with the Barrier Integrity cornerstone attribute of Human Performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, the failure to follow procedures OP 1A and CL 2C caused the licensee to unknowingly operate with multiple containment accident recirculation fans inoperable, which were required in Mode 4. 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, dated June 19, 2012, and Appendix G, Attachment 1, Shutdown Operations Significance Determination Process Phase 1 Initial Screening and Characterization of Findings, Exhibit 4, Barrier Integrity Screening Questions, dated May 9, 2014. The inspectors answered no to the Containment Barrier Screening Questions and determined the finding had very low safety significance (Green). This finding has a cross-cutting aspect of Challenge the Unknown (H.11), in the area of Human Performance, for failing to stop when faced with uncertain conditions. Specifically, when the licensee assessed the illuminated Safeguards Equipment Locked Off alarm, during their control board walk down, they confirmed that the safety injection pump control switch was in pullout and was a reason for the alarm to actuate; however, they failed to confirm that other inputs to the alarm were also not valid.
05000266/FIN-2016002-012016Q2Point BeachFailure to Perform Required Fire Watches in Areas Containing Transient CombustiblesA finding of very low safety significance and associated NCV of license condition 4.F was identified by the inspectors for the licensees failure to conduct required fire watch inspections in accordance with the licensees Fire Protection Program requirements. Specifically, while conducting fire protection walkdowns of both units residual heat removal (RHR) pipeway and heat exchanger rooms, the inspectors discovered numerous transient combustible items in areas that the licensee had controlled using tamper seals on the entrances in lieu of physical entry. The licensees corrective actions included documenting and quantifying the removal of the items from the zones and additional actions to perform additional evaluation of the fire zones. The finding was determined to be more than minor because the failure to conduct the required fire watch inspections was associated with the Initiating Events cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of preventing undesirable consequences (i.e., core damage). Specifically, the failure to conduct the required fire watch inspections or meet the alternate measures specified by the licensees engineers, allowed unanalyzed transient combustibles and ignition sources to be present in fire zones that contained both trains of both units RHR pumps, heat exchangers and associated equipment. The inspectors determined the finding could be evaluated in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 2, the inspectors determined the finding affected the Mitigating Systems cornerstone. The finding degraded fire protection defense-in-depth strategies, and the inspectors determined, using Table 3, that it could be evaluated using Appendix F, Fire Protection Significance Determination Process. The inspectors screened the issue under the Phase 1 Screening Question 1.3.1A, and determined that determined that the finding was of very low safety-significance (Green), because the inspectors determined that the impact of a fire would not prevent either reactor from reaching and maintaining safe shutdown (hot). This finding has a cross-cutting aspect of Bases for Decisions (H.10), in the area of human performance, because the licensees leadership did not ensure that the bases for operational and organizational decisions are communicated in a timely manner. Specifically, the licensee did not periodically verify the understanding of the individuals assigned to fire watches, in particular, that the relief from physical entry and application of a tamper seal required a thorough tour of the zones following any entry into those fire zones.
05000456/FIN-2016008-012016Q1BraidwoodFailure to Verify the Tripping Characteristic of Molded Case Circuit Breakers (MCCBs) Used as Isolation Devices for the 120 Vac Instrument Power SystemThe inspectors identified a finding of very-low safety significance and an associated NCV of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to test the 120 Vac molded case circuit breakers (MCCBs) used as isolation devices on the instrument power system. Specifically, although the licensee had committed to test circuit breakers used as isolation devices in response to Final Safety Analysis Report Question 40.73 in 1982, there was no evidence that these MCCBs had ever been tested. The licensee subsequently entered the issue into its Corrective Action Program. The finding was more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and affected the cornerstone objective of ensuring the availability of the safety-related instrument power system. Specifically, the licensee did not assure, by periodically verifying the time-current characteristic of the MCCBs, that the isolation devices would perform their safety function to isolate the nonsafety-related instrument bus from the safety-related instrument power bus before the safety bus could be affected by a fault on the nonsafety-related load. The inspectors determined that the finding was of very-low safety significance (Green) because it did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that there was no cross-cutting aspect associated with this finding because the finding was not indicative of the licensees current performance.
05000456/FIN-2016008-022016Q1BraidwoodFailure to Verify Air Intake for Diesel Driven Auxiliary Feedwater Pump was Adequately Protected from a High Energy Line BreakThe 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 the failure to verify the adequacy of the diesel driven Auxiliary Feedwater (AFW) pump design. Specifically, the licensee failed to verify the diesel driven AFW pump could perform its safe shutdown function following a high energy line break (HELB) in the Turbine Building. Since the diesels air intake was located in the Turbine Building, it would be impacted by a HELB. The licensee entered this issue into its Corrective Action Program and took immediate corrective actions by declaring the diesel driven AFW pump inoperable and then implementing a temporary plant modification to relocate the diesel air intake to the Auxiliary Building where it is not susceptible to a HELB to restore operability of the pump. The licensees planned corrective actions are to complete a permanent plant modification to relocate the air intake to a location that is not susceptible to a HELB. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to verify that the diesel driven AFW pump could perform its safety function following a HELB event in the Turbine Building did not ensure its availability, reliability, and capability to respond to the initiating event. Since the finding did represent an actual loss of function of at least a single Train for greater than its Technical Specification Allowed Outage Time, a Detailed Risk Evaluation was performed which concluded that the estimated change in core damage frequency was approximately 3.4E-7/yr., which represents a finding of very-low safety significance (Green). The inspectors did not identify a cross-cutting aspect associated with this finding because the finding was not indicative of the licensees current performance.
05000255/FIN-2015004-042015Q4PalisadesFailure to Perform a Required 50.59 Evaluation for Declassification of the CVCSThe inspectors identified a SL IV, NCV of 10 CFR, Part 50.59, Changes, Tests, and Experiments, and an associated finding of very-low safety significance (Green) for the licensees failure to maintain a record of the declassification of the Chemical Volume and Control System (CVCS) from safety-related to nonsafety-related, which includes a written evaluation that provides the bases for the determination that the change did not require a license amendment. The licensee entered this issue into their CAP, and after a review of the system, determined there was reasonable assurance that it could perform its function. The inspectors determined the underlying technical concern was a performance deficiency associated with the Mitigating Systems cornerstone that was more than minor because, if left uncorrected, would become a more significant safety concern. The underlying technical concern screened as a finding with very-low safety significance (Green) because, although it affected the design or qualification of the CVCS, it did not result in the loss of functionality of the CVCS. The violation was determined to be more than minor because the inspectors could not reasonably determine that the changes would not have ultimately required NRC prior approval. The violation was categorized as a SL IV in accordance with Section 6.1.d.2 of the NRC Enforcement Policy because the changes were evaluated by the SDP, described above, as having very-low safety significance (i.e., Green finding). The inspectors did not identify a cross-cutting aspect associated with the finding because the finding was not representative of current performance.
05000255/FIN-2015004-012015Q4PalisadesInadequate Dye Penetrant Examination of Pipe Lug WeldsThe inspectors identified a finding of very-low safety significance (Green), and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion IX, Control of Special Processes, for the licensees failure to perform a dye penetrant (PT) examination of the Safety Injection System (SIS) pipe lug welds in accordance with the American Society of Mechanical Engineers (ASME) Code Section XI requirements. The licensee entered this issue into the Corrective Action Program (CAP) as CR-PLP-2015-04191, repeated the PT examination of the affected SIS lug welds to meet the full extent of coverage required by the ASME Code, repeated examinations of other welds conducted by the PT examiner during the outage, and removed the PT examiner from further weld examination activities. This performance deficiency was determined to be more than minor because, if left uncorrected, the failure to perform a PT examination in accordance with the ASME Code requirements could result in acceptance and return to service of a component with an undetected crack that would increase the possibility of pipe leakage or failure. In addition, the failure to perform a PT examination in accordance with the ASME Code adversely affected the Mitigating System Cornerstone attribute of Equipment Performance, because it could result in failure to detect cracks in pipe welds, which would reduce the availability and reliability of the SIS mitigating system. The inspectors evaluated the finding in accordance with IMC 0609, Appendix A, The SDP for Findings At-Power, Exhibit 2, Mitigating Systems Screening Questions, and answered yes to screening question number 1. Although this finding adversely affected the design or qualification of the SIS pipe lugs, the finding screened as very-low safety significance (Green), because it did not result in the loss of operability or functionality of the affected SIS pipe segment. This finding had a cross-cutting aspect in the Field Presence component of the Human Performance cross-cutting area. Specifically, licensee leaders were not observed in the work areas of the plant to coach and reinforce standards or expectations for the licensees vendor staff to ensure deviation from standards and expectations were promptly corrected (H.2).
05000255/FIN-2015004-022015Q4PalisadesFailure to Identify Components Required to be Covered by the Quality Assurance ProgramThe inspectors identified a finding of very-low safety significance, and an associated NCV of 10 CFR, Part 50, Appendix B, Criterion II, Quality Assurance Program, for the licensees failure to identify all component cooling water (CCW) structures, systems, and components (SSC), which were required to be covered by the Quality Assurance Program (i.e., be safety-related). As a result, the licensee incorrectly credited nonsafety-related CCW components to remain functional during and following a design basis event (DBE). The licensee entered this finding into their CAP and, after performing operability determinations, concluded the system would still be capable of performing its function. The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance, and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as having very-low safety significance (Green) because, although it was a deficiency affecting the design or qualification of a mitigating SSC, the SSC maintained its operability. The inspectors did not identify a cross-cutting aspect associated with this finding because it was determined not to be representative of current performance.
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