Semantic search

Jump to navigation Jump to search
 QSignificanceCCAIdentified byTitleDescription
05000341/FIN-2018003-022018Q3GreenH.7Self-revealingFailure to Ensure Electrolytic Capacitors Installed in the Plant Did Not Have Expired Shelf LivesA finding of very low safety significance with an associated non-cited violation of 10 CFR 50, Appendix B, Criterion VIII, Identification and Control of Materials, Parts, and Components was self-revealed when the reactor water cleanup system inlet flow square root converter failed, resulting in a failure of the reactor water cleanup (RWCU) differential flow instrument and loss of automatic isolation function of the RWCU isolation valves. Specifically, electrolytic capacitors were installed in the RWCU system logic that had expired shelf lives, resulting in failures of the automatic isolation function of the RWCU system.
05000341/FIN-2018003-012018Q3GreenP.3Self-revealingFailure to Apply Torque Values Described in Maintenance Procedure for Flexible Couplings on Emergency Diesel Generator 12A finding of very low safety significance with an associated non-cited violation of Technical Specification 5.4.1.a was self-revealed when plant operators discovered a pencil-thick lube oil leak coming from a flexible coupling on emergency diesel generator 12 during planned surveillance testing. Specifically, a lube oil leak developed when the flexible coupling located between the engine driven lube oil pump and the lube oil filter failed due to improper torque applied to the coupling On April 20, 2018, the licensee was performing a routine slow start surveillance of emergency diesel generator 12 (EDG12), when plant operators noted a pencil-thick lube oil leak from the flexible coupling fastener located between the engine driven lube oil pump and the lube oil filter with the engine running in idle. Plant operators subsequently shut down the engine, discontinued the surveillance, and EDG12 was declared inoperable. The licensee performed an investigation and found the flexible coupling fastener was torqued to 120 in/lbs. Maintenance procedure 35.307.008, Emergency Diesel Generator Engine General Maintenance, Enclosure X, Revision 44 required a torque value of 240260 in/lbs for the size of piping the fastener was on. The coupling was last disturbed in 2011, and the maintenance procedure at that time did not contain information regarding torque values for flexible couplings. A similar flexible coupling fastener failed in 2016 due to inadequate work instructions for torqueing flexible couplings (NCV 05000341/201600401, ADAMS Accession Number ML17030A328), and corrective actions were developed to use the vendor recommended values that had already been added to the maintenance procedure as Enclosure X in 2014. However, the corrective actions did not require all flexible couplings to be checked to ensure they were appropriately torqued. Opportunities existed for the licensee to ensure these flexible couplings were properly torqued according to vendor recommendations, either through scheduled maintenance online or during refueling and forced outages. Therefore, on April 20, 2018, another flexible coupling that was not checked as an extent of condition failed due to an under torqued condition.
05000341/FIN-2018003-032018Q3GreenH.11Self-revealingFailure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal Service Water Outlet Flow Control ValveA finding of very low safety significance and associated non-cited violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and TS 3.7.1 Residual Heat Removal Service Water (RHRSW) System, were self-revealed for the licensees failure to identify a condition adverse to quality on the Division 2 RHRSW outlet flow control valve E1150F068B. Specifically, troubleshooting and the associated post maintenance testing failed to identify and correct a failed anti-rotation key which resulted in an inoperable Division 2 RHRSW system for longer than its TS 3.7.1 allowed outage time.
05000341/FIN-2018002-012018Q2GreenH.14Self-revealingFailure to Document a Condition Assessment Resolution Document for Reactor Recirculation Motor-Generator Set A Brush Gear SparkingA self-revealed Green finding was identified for failure to document a Condition Assessment Resolution Document (CARD) for 5-inch rooster tail sparking on reactor recirculation motor-generator set A brush gear, which ultimately resulted in a manual recirculation pump A trip and plant transient.
05000341/FIN-2018002-042018Q2TBDH.11Self-revealingFailure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal Service Water Outlet Flow Control ValveA self-revealed TBD finding and an associated apparent violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, and Technical Specification 3.7.1 Residual Heat Removal Service Water (RHRSW) System, were identified for failure to identify a condition adverse to quality while performing corrective maintenance on Division 2 RHRSW outlet flow control valve E1150F068B prior to returning the Division 2 RHRSW system to service. Specifically, troubleshooting and associated post maintenance testing failed to identify and correct a failed anti-rotation key which resulted in an inoperable Division 2 RHRSW system for longer than its Technical Specification 3.7.1 allowed outage time.
05000341/FIN-2018002-032018Q2GreenP.2Self-revealingFailure to Adequately Evaluate the Operability of Emergency Diesel Generator11A finding of very low safety significance was self-revealed for the licensees failure to adequately evaluate the operability of a condition adverse to quality identified on Emergency Diesel Generator (EDG) 11. Specifically, a lube oil leak was evaluated as having no impact to the operation of the emergency diesel generator. However, during the next surveillance run of EDG 11, the engine had to be shut down and declared inoperable due to the lube oil leak degrading during operation.
05000341/FIN-2018002-022018Q2GreenP.2Self-revealingInadequate Preventative Maintenance in Residual Heat Removal Service Water System Outlet Flow Control Valves Results in Lower Bonnet (Backseat) Bushing FailureA self-revealed Green finding and associated non-cited violation (NCV) of 10 Code of Federal Regulations (CFR) Part 50, Appendix Criterion V, Instructions, Procedures, and Drawings were identified for failure to ensure activities affecting quality were prescribed in a manner consistent with the circumstances to the residual heat removal service water system(RHRSW). Specifically, preventative maintenance procedure M681 failed to establish an appropriate interval and guidance for periodic valve internals inspections on the Division 2 RHRSW system outlet flow control valve to prevent significant degradation from galvanic corrosion given known internal and external operating experience
05000341/FIN-2018001-012018Q1GreenSelf-revealingFailure to Incorporate Vendor Recommendations into Maintenance Instructions on the Division 1 Control Complex Heating Ventilation and Air Conditioning Supply Fan MotorA finding of very low safety significance and associated non-cited violation of 10CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed for the licensees failure to include instructions for maintenance on the Division 1 Control Complex heating ventilation and air conditioning (CCHVAC) supply fan motor requiring a matching belt set per the recommendation in the vendor manual. Specifically, by installing an unmatched belt set, vibrations degraded past the shutdown limit, rendering Division 1 CCHVAC inoperable.
05000341/FIN-2017004-032017Q4GreenH.14Self-revealingFailure to Perform Fit Testing on Self-Contained Breathing Apparatus RespiratorsA finding of very-low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 20.1703(c)(6) was identified by the inspectors for the licensees failure to perform fit testing for the self-contained breathing apparatus (SCBA) style of respirators utilized. The licensee entered this issue into the Corrective Action Program as Condition Assessment Resolution Document (CARD) 1725155. Corrective actions included working with MSA to get both the air purifying and SCBA respirators National Institute for Occupational Safety and Health approved for both the rubber and Kevlar harnesses, and replacing all face pieces with the same style harness. The performance deficiency was determined to be more-than-minor because it was associated with the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure adequate protection of the worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, the respirator fit testing was being used to verify respirator protection factors for workers and the failure to verify the appropriate protection factor for the SCBA style of respirator affected the licensees ability to control and limit the intake of airborne radioactivity and other hazards. The finding was determined to be of very-low safety significance (Green) because the finding did not involve: (1) as-low-as-reasonably-achievable planning and controls, (2) a radiological overexposure, (3) a substantial potential for an overexposure, or (4) a compromised ability to assess dose. The inspectors determined that the finding had a cross-cutting aspect of human performance, conservative bias. Specifically, although the licensee questioned the practice on multiple occasions, the licensee was not able to determine that the practice was unsafe, and therefore continued the practice (IMC 0310, H.14).
05000341/FIN-2017004-022017Q4NRC identifiedDivision 2 Residual Heat Removal Service Water System Outlet Flow Control Valve Lower Bonnet (Backseat) Bushing FailureThe inspectors evaluated the licensee's handling of selected degraded performance issues involving the following risk-significant structures, systems, and components (SSCs):Residual heat removal service water system. 13 The inspectors assessed performance issues with respect to the reliability, availability, and condition monitoring of the SSCs. Specifically, the inspectors independently verified the licensee's handling of SSC performance or condition problems in terms of:appropriate work practices;identifying and addressing common cause failures;scoping of SSCs in accordance with 10 CFR 50.65(b);characterizing SSC reliability issues;tracking SSC unavailability;trending key parameters (condition monitoring);10 CFR 50.65(a)(1) or (a)(2) classification and reclassification; andappropriateness of performance criteria for SSC functions classified (a)(2) and/or appropriateness and adequacy of goals and corrective actions for SSC functions classified (a)(1).In addition, the inspectors verified problems associated with the effectiveness of plant maintenance for risk-significant SSCs were entered into the licensee's corrective action program with the appropriate characterization and significance. Selected CARDs were reviewed to verify corrective actions were appropriate and implemented as scheduled.This inspection constituted one quarterly maintenance effectiveness inspection samples as defined in IP 71111.12.FindingsDivision 2 Residual Heat Removal Service Water System Outlet Flow Control Valve Lower Bonnet (Backseat) Bushing FailureIntroduction. The inspectors identified an unresolved item (URI) to further evaluate the events and causes of a failure of the Division 2 RHRSW system outlet Flow Control Valve (FSC) lower bonnet (backseat) bushing. Specifically, additional information was needed to determine if one or more performance deficiencies exist.Description. On October 23, 2017, the Division 2 RHRSW system was started to support weekly addition of biocide to the Division 2 RHR reservoir (ultimate heat sink) as a preventative measure to minimize raw water system fouling, which typically entailed running both Division 2 RHRSW pumps for approximately 12 hours. Approximately 20 minutes after system startup, the control room received an overhead annunciator alarm for reactor building south west quad leakage to floor drain sump high along with indication that the reactor building south west quad sump pumps were running. A non-licensed operator was dispatched to the field to investigate the alarms and identified the Division 2 RHRSW outlet Flow Control Valve (FCV) (E1150F068B), located in the Division 2 RHR heat exchanger room in the reactor building, had a significant packing leak calculated to be approximately 16 gallons per minute. The leakage did not impact any other plant equipment in the local area and was captured by the Division 2 RHR heat exchanger room floor drains, which discharge into the reactor building south west quad room sump. Control room operators subsequently shutdown the Division 2 RHRSW pumps to stop the packing leakage and declared the Division 2 RHRSW system inoperable 14 The licensee formed an emergent issues team to further investigate the issue.Following valve disassembly and inspection, the licensee identified the valve lower bonnet (backseat) bushing no longer had sufficient thread engagement to remain in place and that the valve packing had been ejected from the valve stuffing box. A temporary modification was implemented to install a new backseat bushing welded directly to the valve bonnet. The system was subsequently returned to service on October 27, 2017.The Division 2 RHRSW outlet FCV is a safety-related, 24inch Powell globe valve with a motor operator. The primary safety function of the outlet FCV is to fully open to support heat transfer from the Division 2 RHR heat exchanger to the ultimate heat sink. The valve remains fully open during RHRSW pump operation (combined pump flow on the order of 10,000 gallons per minute) and generally is not throttled other than during initial startup of the pumps for a short period of time to help mitigate any potential water hammer events.The licensee completed a root cause analysis documented in CARD 1728611 at the end of the inspection period. The direct cause of the Division 2 RHRSW outlet FCVpacking leakage was determined to be the valve bonnet carbon steel threads corroded to the point of no longer functioning as an adequate mechanical connection. This resulted in the backseat bushing detaching from the valve bonnet allowing the packing to be ejected. The root cause was determined to be previous operating experience resolution for galvanic corrosion for valves in the safety-related service water systems was less than adequate resulting in a failure to recognize the vulnerability of galvanic corrosion on passive valve components. Contributing causes consisted of (1) RHRSW system operation produces significant valve vibration levels and periodic wetting and then drying conditions promoting a corrosive environment and (2) high levels of ionic impurities, as measured by chloride concentration, in RHRSW accelerate galvanic corrosion.The inspectors reviewed the root cause analysis report and several previous issues associated with the Division 2 RHRSW outlet FCV. Those events included, but were not limited to:On May 22, 2017, while placing Division 2 RHRSW in service for biocide treatment of the Division 2 RHR reservoir, the Division 2 RHRSW outlet FCVfailed to fully open. Troubleshooting discovered the direct cause was failure of the anti-rotation bushing stem key due to broken tack welds caused by high vibration during system operation. Previous troubleshooting on what was believed to be an indication issue on May 5, 2017 for the Division 2 RHRSW outlet FCV was inadequate and did not identify the failure of the anti-rotation key. As a result, the RHRSW FCV was returned to service on May 7, 2017, and subsequently failed on the next on-demand stroke on May 22, 2017. The licensee submitted Licensee Event Report 05000341/201700300 to report this event in accordance with 10 CFR 50.73(a)(2)(i)(B) as a condition prohibited by Technical Specification 3.7.1 and 10 CFR 50.73(a)(2)(v)(B) as a condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to remove residual heat. The system was returned to service on May 24, 2017. On September 28, 2017, while Division 2 RHRSW was out of service for planned valve performance monitoring, a partial stem-to-disc separation was detected.This additional monitoring was put in place based on previous industry operating experience of potential stem-to-disc separation following anti-rotation key failures. Upon further investigation and valve-disassembly, the stem-to-disc jam nut tack welds were found broken and the stem had unthreaded approximately 0.225 inches from the disc. Repairs were performed to replace the broken tack welds on the disc jam nut. The disc guide pin was also identified to be damaged and the licensee performed an engineering evaluation to permanently remove the disc guide pin. A broken tack weld was also noted on the backseat bushing which was repaired. The system was returned to service on October 3, 2017.The inspectors questioned the potential relationships between the aforementioned events given the potential for each event to have influenced the eventual failure of the backseat bushing. The inspectors needed additional information to determine whether or not the valve, including the backseat bushing, was subject to an over thrust condition as a result of one or a combination of irregular limit switch settings, anti-rotation key failure, broken and subsequent removal of the disc guide pin, stem-to-disc unthreading, and various broken tack welds. Other additional information was needed in order to determine:if the Division 2 RHRSW outlet FCV was of appropriate design for the known conditions of high vibrations, periods of cavitation on startup and shutdown, and a highly susceptible corrosive environment due to periods of wet and dry conditions with known dissimilar metals highly susceptible to galvanic corrosion;the technical basis behind not including globe valves in the corrosion monitoring program following previously noted and evaluated concerns of RHRSW system susceptibility from years past; andthe technical basis and management of chemistry controls on the RHR reservoirs.Because the licensee completed their root cause evaluation at the end of the inspection period and additional information was required to determine if one or more performance deficiencies exists associated with the various Division 2 RHRSW outlet FCV problems, this issue is being treated as an unresolved item pending receipt of additional information and subsequent inspector review. (URI 05000341/201700402, Division 2 Residual Heat Removal Service Water System Outlet Flow Control Valve Lower Bonnet (Backseat) Bushing Failure)
05000341/FIN-2017004-012017Q4NRC identifiedReactor Recirculation Motor-Generator Set ABrush Gear FailureInspection ScopeDuring the course of the inspection period, the inspectors performed several observations of licensed operator performance in the plants control room to verify that operator performance was adequate and that plant evolutions were being conducted in accordance with approved plant procedures. Specific activities observed that involved a heighted tempo of activities or period of elevated risk included, but were not limited to:Planned downpower, performance of turbine stop and control valve testing, and main steam isolation valve testing on November 4 and 5, 2017;Subsequent response to a manual trip of reactor recirculation motor-generatorset A and single loop operation on November 26, 2017; andPower suppression testing on December 30, 2017. The inspectors evaluated the following areas during the course of the control room observations: licensed operator performance;crews clarity and formality of communications;ability to take timely actions in the conservative direction;prioritization, interpretation, and verification of annunciator alarms (if applicable); correct use and implementation of procedures;control board (or equipment) manipulations;oversight and direction from supervisors; andability to identify and implement appropriate TS actions and Emergency Plan actions and notifications (if applicable). The performance in these areas was compared to pre-established operator action expectations, procedural compliance and task completion requirements. Documents reviewed are listed in the Attachment to this report.These observation activities by the inspectors of operator performance in the plants control room constituted one quarterly licensed operator heightened activity/risk sample as defined in IP 71111.11.b. Findings Reactor Recirculation Motor-Generator Set A Brush Gear Failure Introduction. The inspectors identified an unresolved item (URI) to further evaluate the events and causes of a failure of the reactor recirculation motor-generator set A brush gear assembly. Specifically, the licensee had not yet completed their root cause evaluation of the event at the end of the inspection period.Description. On November 26, 2017, at approximately 2:38 p.m. while the plant was operating at 100 percent reactor thermal power, the control room unexpectedly received a Recirculation System A Generator Field Ground alarm with no other abnormal indications noted in the control room. A non-licensed operator was dispatched to the reactor recirculation motor-generator room in the reactor building and immediately noted an acrid odor along with significant arcing and sparking on the generator end of reactor recirculation motor-generator set A. Subsequently, at 2:56 p.m., control room operators manually tripped reactor recirculation motor-generator set A and entered abnormal operating procedure 20.138.01, reactor recirculation pump trip. In accordance with the abnormal operating procedure the control room operators promptly inserted control rods and stabilized the plant at approximately 35 percent reactor thermal power in single loop operation. The abnormal operating procedure was exited at 4:37 p.m. following completion of associated operator actions and plant stabilization.The licensee formed an emergent issues team to further investigate the issue. Significant damage was identified on the brush gear assembly mounted on the generator end of the reactor recirculation motor-generator set A including, but not limited to, severely worn slip rings, damaged carbon brushes and holders, and melted insulating materials. Initial evaluation by the licensee indicated the carbon brushes had inadequate spring tension to remain in contact with the slip rings, potentially as a result of a brush gear inspection procedure change that allowed for increased wear on the carbon brushes prior to replacement. Additionally, it was determined that sparks were observed on the brush gear assembly the day before the event, however, no discernable action was taken to address the issue.The brush gear assembly was repaired, however, multiple attempts were made to restart reactor recirculation motor-generator set A unsuccessfully. Further troubleshooting identified that the reactor recirculation motor-generator set A starting circuitry voltage was too low to support a restart during single loop operations. Additionally, a rectifier diode pigtail connection in the automatic voltage regulator was found damaged. The rectifier diode pigtail connection was subsequently repaired. A temporary modification was developed and implemented to increase the reactor recirculation motor-generator set A starting circuitry voltage. The reactor recirculation motor-generator set A was successfully restarted on December 5, 2017. Single loop operations were subsequently exited and the unit returned to 100 percent rated thermal power on December 6, 2017.The licensee entered this issue in the corrective action program as CARD 1729439. Because the licensee had yet to complete their root cause investigation and analysis of the event, the issue is being treated as an unresolved item pending the inspectors review of the licensees completed cause evaluation and proposed corrective actions. (URI 05000341/201700401, Reactor Recirculation Motor-Generator Set A Brush Gear Failure)
05000341/FIN-2017003-022017Q3GreenH.1NRC identifiedTechnical Specification Allowed Outage Time Exceeded for Electrical Power Distribution Systems Due to Auxiliary Equipment Out of ServiceThe inspectors identified a Non-Cited Violation (NCV) of Technical Specification (TS) 3.8.7 Distribution Systems Operating, for the licensees failure to either restore inoperable Division 1 and Division 2 AC electrical power distribution subsystems to operable status within 8 hours or be in Mode 3 in 12 hours. Specifically, electrical power distribution subsystems required by the above limiting condition for operation were inoperable due to their respective subdivisions of Residual Heat Removal (RHR) switchgear room ventilation systems being out of service and therefore unavailable to provide the technical specification support function of attendant cooling that was needed for the associated electrical systems to perform their specified safety functions. The licensee entered the issue into its corrective action program as CARD 1726749.The failure to comply with TS 3.8.7 by either restoring inoperable electrical power subsystems to operable status within 8 hours, or be in Mode 3 in 12 hours was a performance deficiency. The performance deficiency was more than minor, and therefore a finding, because it was associated with the Configuration Control attribute of the Mitigating Systems Cornerstone, and adversely affected the associated cornerstone objective to ensure availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The inspectors determined that the finding was of very low safety significance (Green) because it did not represent an actual loss of function of at least a single train for greater than its technical specification allowed outage time, or two separate safety systems out-of-service for greater than its technical specification allowed outage time. The inspectors determined that the violation had a cross-cutting aspect in the area of human performance, resources, because the licensee failed to ensure that the RHR Complex Heating and Ventilation procedure was adequate to support nuclear safety (H.1).
05000341/FIN-2017003-012017Q3Severity level IVNRC identifiedFailure to Satisfy 10 CFR 50.73 Reporting Requirements for Primary Containment Isolation Valve ActuationsThe inspectors identified a Severity Level IV NCV of the NRCs reporting requirements Title 10 of the Code of Federal Regulations (CFR), Part50.73(a)(1), Licensee Event Report (LER) System. The licensee failed to submit a required LER or provide a telephone notification within 60 days after discovery on March 24, 2017, of a condition that resulted in the invalid actuation of containment isolation signals affecting containment isolation valves in more than one system. The licensee entered this issue into its corrective action program to evaluate the cause for its failure to satisfy the reporting requirements and to identify appropriate corrective actions.Subsequently, the licensee made a telephone notification on July 14, 2017 to the NRC Operations Center via the Emergency Notification System to report the event (Event Notice 52859).Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined the performance deficiency was of minor significance based on No answers to the more-than-minor screening questions. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding.
05000341/FIN-2017405-012017Q3GreenP.6NRC identifiedSecurity
05000341/FIN-2017407-012017Q3GreenNRC identifiedSecurity
05000341/FIN-2017002-032017Q2GreenLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50.59(d)(1) requires, in part, that the licensee maintain records of changes to the facility, of changes in procedures, and of tests and experiments made pursuant 10 CFR 50.59(c). These records must include a written evaluation which provides the bases for the determination that the change, test, or experiment does 38 not require a license amendment pursuant to Paragraph (c)(2) of this section. 10 CFR 50.59(c)(2)(ii) requires that a licensee shall obtain a license amendment pursuant to 10 CFR 50.90 prior to implementing a proposed change, test, or experiment if the change, test, or experiment would result in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR (as updated). Technical Specification (TS) 3.3.1.1, Reactor Protection System Instrumentation, states the RPS instrumentation for each function in Table 3.3.1.1 1 shall be operable. As specified in Table 3.3.1.1 1, Function 5, Main Steam Isolation Valve - Closure (8 channels) and Function 9, Turbine Stop Valve - Closure (4 channels) are required to be operable in Mode 1. TS 3.3.1.1, Required Action C.1 states with one or more functions with RPS trip capability not maintained to restore RPS trip capability in 1 hour. Condition C was applicable to both the main steam isolation valve and turbine stop valve RPS logic functional testing. Contrary to the above, on or about August 19, 2016, the licensee failed to perform and maintain a written evaluation as required by 10 CFR 50.59(d)(1) to demonstrate a change to its facility did not require a license amendment. Specifically, the licensee incorrectly concluded no license amendment was required in its 10 CFR 50.59 evaluation prior to implementing surveillance test procedures 24.110.05, RPS Turbine Control and Stop Valve Functional Test, Revision 44 and 24.137.01, Main Steam Line Isolation Channel Functional Test, Revision 40. The revised procedures incorporated a change that resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a structure, system, or component important to safety previously evaluated in the FSAR (as updated) as specified by Section (c)(2)(ii). Specifically, the use of the test box resulted in the loss of two RPS trip functions by bypassing m ore than the TS minimum allowed inputs per channel to maintain functionality, violating the requirements of TS 3.3.1.1 during testing on September 22 and 23, 2016. In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, Table 3, SDP Appendix Router, the inspectors determined this finding affected the Mitigation Systems Cornerstone, specifically the Reactivity Controls Systems contributor , and would require review using IMC 0609, Append ix A, The Significance Determination Process (SDP) for Findings At -Power, June 19, 2012. The inspectors performed a Phase 1 SDP review of this finding using the guidance provided in IMC 0609, Appendix A, Exhibit 2, Mitigating Systems Screening Questions, and determined this finding was a licensee performance deficiency of very low safety significance ( Green ) because it did not affect a single RPS trip signal to initiate a reactor scram AND the function of other redundant trips or diverse methods of reactor shutdown. Violations of 10 CFR 50.59 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. This violation was also associated with a finding t hat has been evaluated by the SDP and communicated with a SDP color reflective of the safety impact of the deficient licensee performance. The SDP, however, does not specifically consider regulatory process impact. Thus, although related to a common regulatory concern, it is necessary to address the violation and finding using different processes to correctly reflect both the regulatory importance of the violation and the 39 safety significance of the associated finding. In accordance with Section 6.1.d.2 o f the NRC Enforcement Policy, this violation was categorized as Severity Level IV. This violation was entered into the licensees corrective action program as CARD 17 20163.
05000341/FIN-2017002-012017Q2GreenH.7Self-revealingInadequate Work Instructions for Maintenance on EDG 14Green . A finding of very low safety significance with an associated Non- Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self -revealed when plant operators were not able to shut down emergency diesel generator (EDG) 14 using the manual emergency stop button during surveillance testing. Consequently, operators shut down the engine and removed it from service. The licensee failed to have work instructions for maintenance on the safety -related EDG appropriate to ensure the emergency overspeed switch (EOS) oil seal was properly installed to prevent oil intrusion into the switch housing. The licensee entered this violation into its corrective action program for evaluation and identification of appropriate corrective actions. The licensee replaced the EOS and revised the maintenance procedure and work order guidance for proper oil seal installation on the EOS. The finding was of more than minor safety significance because it was associated with the Equipment Performance at tribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the EO S failure during surveillance testing due to oil intrusion resulted in unplanned inoperability and unavailability of an onsite emergency power source. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time nor did it represent a loss of function of a non- TS train designated as high safety significant in accordance with the licensees Maintenance Rule Program . The inspectors concluded this finding affected the cross - cutting area of human performance and the cross -cutting aspect of documentation. Plant activities are governed by comprehensive, high- quality, programs, processes and procedures. In this case , the licensee determined its maintenance procedure and work order guidance were not adequate to ensure the EOS oil seal and upper air start distributor gasket were properly installed to prevent oil leakage from the air start distributor from getting into the EOS housing. (IMC 0310, H.7)
05000341/FIN-2017002-022017Q2GreenP.3NRC identifiedUnacceptable Preconditioning of High Pressure Coolant Injection System Air Operated Valve Prior to Stroke Time Test MeasurementGreen . The inspectors identified a finding of very low safety significance with an associated Non -Cited Violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings. The licensee failed to establish an adequate procedure to perform required stroke time testing for high pressure coolant injection (HPCI) turbine barometric condenser condensate drain line inboard isolation valve E4100 F026. The surveillance test procedure resulted in unacceptable preconditioning of the valve prior to the stroke time test measurement. The licensee entered this issue into its corrective action program for evaluation and initiated a corrective action to revise the test procedure. The finding was of more than minor significance because it was associated with the Procedure Quality attribute of the Barrier Integrity Cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, cycling the air -operated valve prior to performing the stroke time measurement masked the actual as -found condition of the valve, invalidating the test results. Because the preconditioning altered the as-found condition of the valve, the data collected through the performance of the surveillance test was not fully indicative of the true valve performance trend. Therefore, this performance deficiency had a direct effect on the licensees ability to trend as -found data for the purpose of assessing the reliability of the air -operated valve. The finding was a licensee performance deficiency of very low safety significance because it represented only a degradation of the radiological barrier function provided for the auxiliary building and was not an actual loss of the barrier function provided by the HPCI system pressure boundary as a closed system outside containment . The inspectors concluded this finding affected the cross - cutting area of problem identification and resolution, in particular the cross -cutting aspect of resolution. The organization takes effective corrective actions to address issues in a timely manner, commensurate with their safety significance. Corrective actions resolve and correct the identified issues, including causes and extent of condition. In this case, corrective actions for the previous inspector -identified preconditioning issue did not effectively address the extent of condition involving potential preconditioning of other HPCI system air -operated valve s in other surveillance testing procedures. (IMC 0310, P.3)
05000341/FIN-2017007-012017Q2GreenNRC identifiedFailure to Correct a Design Deficiency that Mis-Quantified Unidentified LeakageGreen . The inspectors identified a finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR) , Part 50, Appendix B, Criterion XVI, Corrective Actions , for the licensees failure to correct a design deficiency that mis -quantified unidentified leakage from reactor coolant system (RCS ) pressure boundary . Specifically, in April 2007, the licensee identified that the driver mount drain for the reactor recirculation pump could potentially drain leakage from nearby pipe cracks to the identified leakage collection point. However, the licensee had 3 not correct ed this design deficiency as of the start of this inspection. The licensee documented this issue into the CAP as Condition Assessment Resolution Document (CARD) 17 25489 and developed a night order to direct the operators how to calculate unidentified leakage. The licensee also planned to revise procedure 24.000.02 as an interim measure until the modification was implemented. The inspectors determined that the licensees failure to correct the design deficiency that mis-quantified unidentified leakage is a performance deficiency that is reasonably within the licensees ability to foresee and correct. The inspectors determined that this issue is more than minor because if left uncorrected, the performance deficiency has the potential to lead to a more significant safety concern. Specifically, leakage that would normally be collected and measured as unidentified leakage could be collected and measured as identified leakage, leading to a potential violation of the TS unidentified leakage rate . Because the finding did not represent a loss of system or function, or represent an actual loss of function of at least a single train for greater than its Technical Specification (TS) Allowed Outage Time, or represent an actual loss of function of one or more non TS trains of equipment designated as high safety -significant in the licensees Maintenance Rule Program, it was screened as very low safety significance. The inspectors did not identify a cross -cutting aspect since the issue originated more than three years ago.
05000341/FIN-2017010-012017Q2GreenNRC identifiedFailure to Account for Internal Heat Rise for Protective Devices SettingsGreen . The inspectors 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 III, Design Control, for the licensees failure to ensure that the protective devices installed in Motor Control Centers ( MCCs ) would not spuriously trip during design basis events. Specifically, the licensee did not account for 18 degrees Fahrenheit (F) heat rise inside the MCCs. Protective devices inside MCCs located in harsh environment were evaluated and sized for a maximum elevated temperature up to 156 degrees F instead of 174 degrees F . The licensee captured the inspectors concern into their Corrective Action Program ( CAP ) as CARD 17 -24412. The performance deficiency was determined to be more- than- minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very -low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, as an immediate corrective action, the licensee performed a preliminary evaluation that reasonably concluded the overcurrent protection devices within the scope of DC -6475 would not spuriously trip. The finding did not have a cross -cutting aspect associated with it because it was not representative of current performance.
05000341/FIN-2017010-022017Q2GreenNRC identifiedFailure to Correctly Calculate the Post Accident Operating TimeGreen . The inspectors identified a finding of very -low safety significance (Green) and an associated NCV of 10 CFR 50.49, Paragraph (e) (1), Environmental Qualification of Electrical Equipment Important to Safety for Nuclear Power Plants, for the licensees failure to include the correct time -dependent temperature for EQ components in their EQ Program. Specifically, the inspectors identified two examples where the licensees EQ files failed or incorrectly calculated the Loss of Coolant Accident Post Accident Operating Time for EQ components . The licensee captured the inspectors concern into their CAP as CARD 17 -24760 and 17 -24619. 3 The performance deficiency was determined to be more- than- minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very -low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, as an immediate corrective action, the licensee per formed a preliminary assessment and calculated the Loss of Coolant Accident Post Accident Operating Time for these two EQ components and determined that the equipment remained qualified for the environmental conditions. The finding did not have a cross -cutting aspect associated with it because it was not representative of current performance
05000341/FIN-2017010-032017Q2GreenNRC identifiedFailure to Translate Environmental Qualification Requirements into Maintenance ProceduresGreen . The inspectors identified a finding of very -low safety significance (Green) and asso ciated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for failure to translate Environmental Qualification Requirements into Maintenance Procedures. Specifically, the licensee failed to ensure that the Environmental Qualification requirement to replace the top cover gasket on NAMCO EA740 Series Limit Switches was translated to the associated maintenance procedure. In addition, the licensee also faile d to ensure that the Environmental Qualification requirement to inspect MCC gaskets was translated to the associated maintenance procedure. The licensee captured the inspectors concern into their CAP as CARD 17 -24629 and 17 -24444. The performance deficiency was determined to be more- than- minor because it was associated with the Mitigating Systems cornerstone attribute of procedure control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very -low safety significance (Green) because it did not result in the loss of operability or functionality of any structure, system, or component . The finding did not have a cross -cutting aspect associated with it because it was not representative of current performance.
05000341/FIN-2017010-042017Q2GreenLicensee-identifiedLicensee-Identified ViolationTitle 10 CFR 50.49 , Paragraph (k) , stated, in part, that licenses are not required to requalify electric equipment important to safety in accordance with the provisions of this section if the Commission has previously required qualification of that equipment in accordance with, Guidelines for Evaluating Environmental Qualification of Class 1E Electrical Equipment in Operating Reactors, November 1979 (DOR Guidelines), or NUREG -0588, Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment. Paragraph (k)(1), stated, in part, that replacement equipment must be qualified in accordance with the provisions of this section unless there are sound reasons to the contrary. Contrary to these requirements, as of May 22, 2017, the licensee failed to replace Category II equipment with Category I qualified equipment. Specifically, General Electric vertical pump Residual Heat Removal motors were qualified to NUREG-0588 Category II originally. When they were refurbished by Westinghouse in 1990s, a mixed qualification was developed via analysis only instead of testing of specimens as required for Category I qualified equipment. This violation was identified by the licensee during self -assessment and was entered into t he licensees CAP as CARD 16 -25611. The licensee plan to replace the pump motors with new motors that meet NUREG-0588 Category I requirement. The inspectors determined that the performance deficiency was of more -than -minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. Specifically, the licensees failure to replace the Residual Heat Removal pump motor with qualified Category I motors did not ensure the reliability of the replacement motors . 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. Because the finding impacted the Mitigating Systems cornerstone, the team screened the finding through IMC 0609 Appendix A, The Significance Determination Process for Fin dings At -Power, issued on June 19, 2012, using Exhibit 2, Mitigating Systems Screening Questions. The finding screened as of very -low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems
05000341/FIN-2017404-012017Q2GreenH.1Self-revealingSecurity
05000341/FIN-2017009-012017Q1WhiteNRC identifiedFailure to Maintain the Effectiveness of the Sites Emergency PlanPreliminary White. An NRC identified finding preliminarily determined to be of low to moderate safety significance (White), and an associated apparent violation of Title 10 of the Code of Federal Regulations (10 CFR) 50.54(q)(2) and 10 CFR 50.47(b)(9) was identified for the licensees failure to maintain the effectiveness of its emergency plan and use adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency. Specifically, the licensee failed to maintain the ability to accurately declare an Emergency Action Level (EAL) classification, RG-1.1, and develop and issue accurate protective action recommendations (PARs) during the implementation of the sites Emergency Plan in response to a rapidly progressing accident. The licensee inaccurately analyzed the effect of increasing background radiation on the sites Standby Gas Treatment System accident range radiation monitor (AXM) indications based on the installed configuration of the AXM. As configured, the AXM could provide inaccurate indications of radioactive releases that are used as the licensees basis for determining EAL classification and development of PARs. The licensee documented the issue in the corrective action program as CR-16-29230, and actions were completed to restore the accuracy of the indications provided by the AXM. The inspectors determined that the licensees failure to maintain the effectiveness of its emergency plan and use adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency was a performance deficiency; the cause was reasonably within the licensees ability to foresee and correct; and should have been prevented. The inspectors determined the issue was more than minor because it adversely affected the emergency preparedness cornerstone objective to ensure that the licensee is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. Specifically, the finding would result in the potential over classification of an emergency event and the potential issuance of unnecessary or early PARs. 3 The inspectors applied Inspection Manual Chapter (IMC) 0609, Appendix B, Section 5.9. to screen this finding, and determined the licensee failed to maintain the risk significant planning standard (RSPS) identified in 10 CFR 50.47(b)(9) by ensuring adequate methods, systems, and equipment for assessing and monitoring actual or potential offsite consequences of a radiological emergency condition are in use. Using Table 5.9-1, the inspectors determined the sites dose assessment process was incapable of providing technically adequate estimates of radioactive material releases to the environment or projected offsite doses in some cases (specifically a rapidly progressing accident scenario). This significance example corresponds to a Degraded RSPS Function, which is a finding of low to moderate safety significance (White). The inspectors determined no cross-cutting aspects were associated with the performance deficiency.
05000341/FIN-2017001-012017Q1GreenH.7Self-revealingInadequate Work Instructions for Maintenance on EDG 14Green . A finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations ( 10 CFR) 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self -revealed when plant operators discovered a thick white smoke plume coming from the emergency diesel generator (EDG) 14 engine exhaust manifold during surveillance testing. Consequently, operators shut down the engine and removed it from service. The licensee failed to have work instructions for maintenance on the safety -related EDG appropriate to ensure insulation blankets on the engines exhaust manifold were replaced with insulation blankets conforming to the approved engineering design. The licensee entered this violation into its corrective action program for evaluation and identification of appropriate corrective actions. The licensee replaced the insulation blankets with insulation blank ets conforming to the approved engineering design. The finding was of more than minor safety significance because it was related to the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective o f ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, operators shutdown the engine after discovering a thick white smoke plume coming from the engines exhaust manifold , which resulted in unplanned inoperability and unavailability of this onsite emergency power source. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time nor did it represent a loss of function of a non -TS train designated as high safety significant in accordance with the licensees Maintenance Rule Program . The inspectors concluded this finding affected the cross -cutting area of human performance and the cross- cutting aspect of documentation. Plant activities are governed by comprehensive, high -quality, programs, processes and procedures. Design documentation, procedures, and work pack ages are complete, thorough, accurate, and current. In this case, the licensees process for implementing and maintaining engineering configuration control of the newly designed EDG exhaust manifold insulation blankets was inadequate because 3 it did not follow the licensees formal engineering configuration management process. (IMC 0310, H.7)
05000341/FIN-2017001-022017Q1GreenH.1Self-revealingFailure to Maintain Adequate SLC Storage Tank Boron ConcentrationGreen . A finding of very low safety significance with an associated Non- Cited Violation of TS 3.1.7, Standby Liquid Control (SLC) System, was self -revealed when the licensee measured the boron concentration in the SLC storage tank and discovered the concentration was below the minimum requirement of 8.5 percent. Specifically, the licensee failed to adequately monitor and identify a decreasing trend in SLC storage tank sodium pentaborate concentration concurrent with known dilution of the SLC storage tank during pump and valve testing. The licensee entered this violation into its corrective action program for evaluation and identifi cation of appropriate corrective actions and restored the SLC sodium pentaborate concentration to within TS limits. The finding was of more than minor safety significance because it was associated with the Equipment Performance attribute of the Mitigating Systems Cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, a lower than allowable sodium pentaborate concentration affected the SLC systems ability to shut down the reactor during a design basis event. The finding was determined to be a licensee performance deficiency of very low safety significance during a detailed Significance Determination Process review since the delta core damage frequency ( CDF ) was determined to be less than 1.0E 6/year. The inspectors concluded this finding affected the cross -cutting area of human performance and the cross -cutting aspect of resources. Specifically, the licensee failed to ensure equipment and procedures were adequate to support nuclear safety . Th is issue would have been avoided if the system monitoring plan was trending tank level via a pressure indicator . Also, chemistry had no administrative limits in their procedure to add boron prior to the minimum TS limit was reached and the system engineer was not a reviewer on the routine surveillance procedure and was not trending the concentration as a backup. (IMC 0310, H.1 )
05000341/FIN-2017001-032017Q1GreenLicensee-identifiedLicensee-Identified ViolationTS 3.3.5.1, ECCS Instrumentation, states the ECCS instrumentation for each function in Table 3.3.5.1 1 shall be operable. As specified in Table 3.3.5.1 1, Function 3b, HPCI System High Drywell Pressure (4 channels) and Function 3f, HPCI System Manual Initiation (1 channel) are required to be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. TS 3.3.5.1, Required Action A.1 states with one or more channel(s) inoperable, immediately enter the condition referenced in Table 3.3.5.1 1 for the channel. Table 3.3. 5.1 1, Function 3b, references Condition B for inoperable HPCI System High Drywell Pressure channels. Required Action B.2 states declare the HPCI system inoperable within 1 hour from discovery of loss of HPCI initiation capability and Required Action B.3 states place the affected channel(s) in trip within 24 hours. Table 3.3.5.1 1, Function 3f, references Condition C for an inoperable HPCI System Manual Initiation channel. Required Action C.2 states restore the channel to operable status within 24 hours . If the required actions and associated completion 42 times of Condition B or C are not met, Required Action G.1 states immediately declare the associated supported feature (i.e., HPCI system) inoperable. TS 3.5.1, ECCS Operating, states, in part, each ECCS injection subsystem shall be operable in Modes 1, 2, and 3, except HPCI is not required to be operable with reactor steam dome pressure less than or equal to 150 psig. With the HPCI system inoperable, Required Action E.1 states immediately verify by administrative means RCIC system is operable and Required Action E.2 states restore HPCI system to operable status in 14 days. If the required actions and associated completion times of Condition E are not met, Required Action I.1 states be in Mode 3 in 12 hours. LCO 3.0.4.b is not applicable to HPCI. TS 3.3.5.2, RCIC System Instrumentation, states the RCIC instrumentation for each function in Table 3.3.5.2 1 shall be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. As specified in Table 3.3.5.2 1, Function 4, RCIC System Manual Initiation (one channel per valve) is required to be operable. TS 3.3.5.2, Condition A states with one or more channels inoperable, immediately enter the condit ion referenced in Table 3.3.5. 21 for the channel. Table 3.3.5.2 1, Function 4, references Condition C for an inoperable RCIC System Manual Initiation channel. Required Action C.1 states restore the channel to operable status within 24 hours. If the required actions and associated c ompletion times of Condition C are not met, Required Action E.1 states immediately declare the RCIC system inoperable. TS 3.5.3, RCIC System, states the RCIC system shall be operable in Modes 1, 2, and 3 with reactor steam dome pressure greater than 150 psig. With the RCIC system inoperable, Required Action A.1 states immediately verify by administrative means HPCI system is operable and Required Action A.2 states restore RCIC system to operable status in 14 days. If the required actions and associated completion times of Condition A are not met, Required Action B.1 states be in Mode 3 in 12 hours. LCO 3.0.4.b is not applicable to RCIC. TS 3.0.4, Limiting Condition for Operation (LCO) Applicability, Paragraph (a) states, in part, when a LCO is not met , entry into an operational mode or other specified condition in the applicability shall only be made when the associated actions to be entered permit continued operation in the operational mode or other specified condition in the applicability for an unli mited period of time. This specification shall not prevent changes in modes or other specified conditions in the applicability that are part of a shutdown of the unit. Contrary to the above: 1. On six occasions (February 10, 2014, April 16, 2014, March 19, 2015, September 13, 2015, May 3, 2016, and November 7, 2016 ), the licensee entered Mode 3 following plant shutdowns without declaring the HPCI system instrumentation functions of high drywell pressur e and manual initiation inoperable and entering LCO 3.3.5.1. During the shutdowns, Fermi 2 was in Mode 3 for up to fifteen hours with reactor steam dome pressure greater than 43 150 psig without the licensee satisfying TS 3.3.5.1, Required Actions A.1, B.2, and G.1. This is a violation of TS 3.3.5.1. With HPCI inoperable as specified by TS 3.3.5.1, Required Actions B.2 and G.1, the licensee did not satisfy TS 3.5.1, Required Action E.1. This is a violation of TS 3.5.1. 2. On six occasions (February 10, 2014, A pril 16, 2014, March 19, 2015, September 13, 2015, May 3, 2016, and November 7, 2016 ), the licensee entered Mode 3 following plant shutdowns without declaring the RCIC system instrumentation function of manual initiation inoperable and entering LCO 3.3.5.2 . During the shutdowns, Fermi 2 was in Mode 3 for up to fifteen hours with reactor steam dome pressure greater than 150 psig without the licensee satisfying TS 3.3.5.2, Required Action A.1. This is a violation of TS 3.3.5.2. 3. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without declaring the HPCI system instrumentation functions of high drywell pressure and manual initiation inoperable and entering LCO 3.3.5.1. For up to nineteen hours during this time, the licensee did not satisfy TS 3.3.5.1, Required Actions A.1, B.2, and G.1. This is a violation of TS 3.3.5.1. With HPCI in operable as specified by TS 3.3.5.1, Required Actions B.2 and G.1, the licensee did not satisfy TS 3.5.1, Required Action E.1. This is a violation of TS 3.5.1. 4. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without declaring the RCIC system instrumentation function of manual initiation inoperable and entering LCO 3.3.5.2. For up to nineteen hours during this time, the licensee did not satisfy TS 3.3.5.2, Required Action A.1. This is a violation of TS 3.3.5.2. 5. On six occasions (March 28, 2014, April 21, 2014, April 3, 2015, November 25, 2015, May 12, 2016, and November 11, 2016 ), the licensee entered Mode 2 with reactor steam dome pressure greater than 150 psig during plant startups without meeting the LCOs of TS 3.3.5.1 and TS 3.3.5.2 for HPCI and RCIC systems instrumentation functions of high drywell pressure (HPCI only) and m anual initiation ( both HPCI and RCIC) . This is a violation of TS 3.0.4. This violation was entered into the licensees corrective action program as CARD 16 26153. The violation was determined to be of very low safety significance (Green) during a detailed Significance Determination Process review since the CDF was determined to be less than 1.0E -7/year.
05000341/FIN-2016004-012016Q4GreenSelf-revealingInadequate Work Instructions for Maintenance on Flexible Couplings for EDGsA finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was self-revealed when plant operators discovered an oil leak coming from a flexible coupling upstream of the emergency diesel generator (EDG) 12 lube oil heater during surveillance testing. The licensee failed to have work instructions for maintenance on safety-related EDGs appropriate to the circumstances to ensure flexible coupling fasteners were correctly torqued as specified by the manufacturer to prevent leakage. The licensee entered this violation into its corrective action program (CAP) as Condition Assessment Resolution Document (CARD) 16-25666 and replaced the leaking flexible coupling. This performance deficiency was of more than minor safety significance because it was related to the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the EDG 12 flexible coupling oil leak resulted in unplanned inoperability and unavailability of this onsite emergency power source. The finding was determined to be of very low safety significance because it did not represent an actual loss of function of a single train for greater than its Technical Specification (TS) allowed outage time nor did it represent a loss of function of a non-TS train designated as high safety significant in accordance with the licensees Maintenance Rule Program. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
05000341/FIN-2016004-022016Q4GreenH.9NRC identifiedFailure to Correctly Interpret and Implement TS Requirements for LOP Instrumentation and AC Electrical Power FunctionsThe inspectors identified a finding of very low safety significance with an associated NCV of TS 3.3.8.1, Loss of Power (LOP) Instrumentation, and TS 3.8.1, AC (Alternating Current) Sources Operating. The licensee failed to satisfy applicable action requirements for inoperable loss of voltage and degraded voltage instrument channels, inoperable EDGs, and an inoperable offsite power circuit when power was lost to the station transformer 64 auto voltage tap changer and one-half of the instrument channels for engineered safety features bus 64C due to failure of line side potential transformer fuses on April 24, 2016. The licensee entered this performance deficiency into its CAP as CARDs 16-23392, 16-25194 and 16-28120. As an immediate corrective actions the licensee established an expectation to enter Limiting Condition for Operation (LCO) 3.3.8.1 when any of the LOP instrumentation channels are tripped. Other corrective actions included additional training for licensed operators. This performance deficiency was of more than minor safety significance because a failure to correctly implement TS LCO requirements has the potential to lead to a more significant safety concern if left uncorrected. Specifically, a failure to declare an LCO not met, enter the applicable condition(s), and follow the applicable actions could reasonably result in operations outside of established safety margins or analyses. The finding was determined to be of very low safety significance during a detailed Significance Determination Process review since the delta core damage frequency (CDF) was determined to be less than 1.0E-6/year. The inspectors concluded this finding affected the cross-cutting area of human performance and the cross-cutting aspect of training. Specifically, licensed operators failed to correctly apply the TS LCO requirements for inoperable LOP instrument channels and inoperable AC power sources due to lack of knowledge and unfamiliarity with the equipment conditions they faced during the event (IMC 0310, H.9).
05000341/FIN-2016004-042016Q4GreenNRC identifiedInadequate Testing of SGTS FiltersThe inspectors identified a finding of very low safety significance with an associated NCV of TS 5.5.7, Ventilation Filter Testing Program. The licensee failed to perform testing of the standby gas treatment system (SGTS) high-efficiency particulate air (HEPA) filters that demonstrated a penetration and system bypass of less than 0.05 percent. The licensee entered this violation into its CAP as CARD 1628812. The licensee declared the Division 1 SGTS subsystem inoperable until testing was performed satisfactorily and evaluated the extent of condition on the control room filtration system. This performance deficiency was of more than minor safety significance because it was associated with the procedure quality attribute for the control room and auxiliary building and adversely affected the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, by not adequately testing the SGTS HEPA filters, the ability of the SGTS to collect and treat the design leakage of radionuclides from the primary containment to the secondary containment during an accident could not be assured. The finding was determined to be of very low safety significance because it involved only a degradation of the radiological barrier function provided by the SGTS. The inspectors concluded that because this condition has existed for greater than three years, this issue would not be reflective of current licensee performance and no cross-cutting aspect was identified.
05000341/FIN-2016004-052016Q4GreenH.12Self-revealingFailure to Lock an Area Meeting Locked High Radiation Area ConditionsA finding of very low safety significance with an associated NCV of TS 5.7.2, High Radiation Area, was self-revealed when a locked high radiation area (LHRA) was found to be unlocked. The licensee immediately locked the LHRA and performed follow-up surveys. Subsequent actions included providing additional training for radiation protection technicians. This issue was entered into the licensees CAP as CARD 16-28186. The inspectors determined the performance deficiency was more than minor because it impacted the program and process attribute of the Occupational Radiation Safety cornerstone and adversely affected the cornerstone objective to ensure adequate protection of worker health and safety from exposure to radiation from radioactive material during routine civilian nuclear reactor operation. Specifically, not locking LHRAs could lead to inadvertent worker entry into high dose rate areas without knowledge of the radiological conditions. The finding was determined to be of very low safety significance because it did not involve as-low-as-reasonably-achievable planning for work controls, there was no overexposure nor substantial potential for an overexposure, and the licensees ability to assess dose was not compromised. The inspectors determined the finding affected the cross-cutting area of human performance and the cross-cutting aspect of avoid complacency because individuals did not plan for the possibility of mistakes and implement appropriate error reduction tools. Specifically, the radiation protection technician did not ensure a lock verification was performed on the padlock as required by station procedures (IMC 0310, H.12).
05000341/FIN-2016004-062016Q4GreenLicensee-identifiedLicensee-Identified ViolationTitle 10 of the CFR, Section 50.72(a)(1)(ii), requires, in part, that the licensee shall notify the NRC Operations Center via the Emergency Notification System of those non-emergency events specified in Paragraph (b) that occurred within three years of the date of discovery. 10 CFR 50.72(b)(3) requires, in part, that the licensee shall notify the NRC as soon as practical and in all cases within eight hours of the occurrence of any of the applicable conditions. 10 CFR 50.72(b)(3)(v)(C) requires, in part, that the licensee report any event or condition, that at the time of discovery, could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. 10 CFR 50.73(a)(1) requires, in part, that the licensee submit an LER for any event of the type described in this paragraph within 60 days after the discovery of the event. 10 CFR 50.73(a)(2)(v)(C) requires, in part, that the licensee report any event or condition that could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above: 1. Between September 1, 2013 and September 30, 2016, the licensee failed to notify the NRC Operations Center via the Emergency Notification System of numerous non-emergency events specified in Paragraph (b) within eight hours of the events. These events involved the loss of safety function of the secondary containment when secondary containment pressure exceeded the TS limit due to known effects of high winds. 2. The licensee failed to submit required LERs within 60 days after the discovery of numerous events between September 1, 2013 and September 30, 2016. These events involved the loss of safety function of the secondary containment when secondary containment pressure exceeded the TS limit due to known effects of high winds. Violations of 10 CFR 50.72 and 10 CFR 50.73 are dispositioned using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to make reports to the NRC as required by 10 CFR 50.72(a)(1)(ii) and 10 CFR 50.73(a)(1). The licensee entered this violation into its CAP as CARD 16-27023. Title 10 of the CFR, Section 20.1501, requires, in part, that each licensee shall make, or cause to be made, surveys of areas that may be necessary for the licensee to comply with the regulations in this part and are reasonable for the circumstances to evaluate the magnitude and extend of radiation levels and the potential radiological hazards of the radiation levels. 10 CFR 20.1902(b) states that the licensee shall post each high radiation area with a conspicuous sign or signs bearing the radiation symbol and the words CAUTION, HIGH RADIATION AREA or DANGER, HIGH RADIATION AREA. Contrary to the above, on August 17, 2016, the licensee failed to conduct reasonable surveys to evaluate radiation levels to ensure compliance with the posting requirements of 10 CFR 20.1902(b) during activities known to cause changes in radiation levels. Specifically, the licensee failed to ensure surveys were performed while draining the annulus of the multi-purpose canister, which is an evolution known to change radiological conditions. An unposted high radiation area was identified several hours later when radiation protection personnel entered the area to perform surveys to ensure compliance with the containers Certificate of Compliance. This violation was entered into the licensees CAP as CARD 16-26586. The finding was assessed in accordance with IMC 0609, Appendix C, Occupational Radiation Safety SDP and determined to be of very-low safety significance because it did not involve as-low-as-reasonably-achievable planning or work controls, there was no overexposure nor substantial potential for an overexposure, and the ability to assess dose was not compromised
05000341/FIN-2016004-032016Q4Severity level IVNRC identifiedFailure to Satisfy 10 CFR 50.73 Reporting Requirements for Loss of LOP Instrumentation and AC Electrical Power Safety FunctionsThe inspectors identified a Severity Level IV NCV of the NRCs reporting requirements in 10 CFR 50.73(a)(1), Licensee Event Report System. The licensee failed to submit a required Licensee Event Report (LER) within 60 days after discovery on September 16, 2016, of an operation or condition which was prohibited by the plants TSs and an event or condition that could have prevented the fulfillment of the safety function to remove residual heat and mitigate the consequences of an accident. The inspectors concluded the licensee failed to satisfy the applicable regulatory reporting requirements due to unwarranted delay in evaluating conditions from the event with respect to compliance with the TSs and reporting requirements. The licensee subsequently submitted LER 05000341/2016-009-00, Emergency Diesel Generator Inoperable Due to Open Circuit on Loss of Power Instrumentation, on December 20, 2016, to report the event. The licensee entered this issue into its CAP as CARD 16-30164. Consistent with the guidance in IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, dated September 7, 2012, the inspectors determined the performance deficiency was of minor significance based on No answers to the more-than-minor screening questions. In accordance with Section 6.9.d.9 of the NRC Enforcement Policy, this violation was categorized as Severity Level IV because the licensee failed to report as required by 10 CFR 50.73(a)(1). No cross-cutting aspect is associated with this traditional enforcement violation because the associated performance deficiency was determined to be of minor significance and therefore not a finding.
05000341/FIN-2016007-012016Q3GreenNRC identifiedInadequate Procedure for Addressing Non-Functional MDCT Fan Motor Brake SystemThe 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 V, Instructions, Procedures, and Drawings, for the licensee failure to establish procedures that were appropriate for addressing non-functional mechanical draft cooling tower (MDCT) fan motor brakes. Specifically, a license procedure contained instructions for addressing the impact of non-functional MDCT fan motor brakes to the ultimate heat sink operability that were inconsistent with the applicable Technical Specification requirements. The licensee captured this issue in their Corrective Action Program (CAP) as CARD 16-26762, verified that all MDCT fan brake systems were functional, revised the affected procedure to restore compliance, and issued a night order to notify control room licensed nuclear operators of the revised procedure. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a tornado event. Specifically, a historic review for the last 12 months revealed that the minimum required number of MDCT fans remained operable to mitigate the consequences of a tornado. 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 procedure instructions for addressing MDCT fan motor brake non-functionality were developed more than 3 years ago.
05000341/FIN-2016007-032016Q3GreenH.4NRC identifiedFailure to Verify the Ability to Manually Throttle Safety-Related MOVs during a DBAThe 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 failure to verify the capability to manually throttle safety-related motor-operated valves (MOVs) during a DBA. Specifically, the licensee did not verify that the protective devices would allow manually throttling safety-related MOVs during a DBA without tripping. The licensee captured this issue in their CAP as CARD 16-26763, performed a preliminary protective device evaluation to reasonably determine the maximum number of throttling cycles each MOV can incur without tripping the associated thermal overload, and incorporated these limits into an operations night order. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed informal analyses to evaluate the installed protective devices for the throttling MOVs and reasonably determined that tripping would not occur. The team determined that the associated finding had a cross-cutting aspect in the area of Human Performance because work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the engineers that performed the affected calculation, which was approved on December 2013, did not communicate and coordinate with operations or the MOV engineer to determine if the plant had throttling MOVs that required additional analysis.
05000341/FIN-2016007-042016Q3GreenNRC identifiedFailure to Periodically Test the EDG Capability to Start and Accelerate All of the Sequenced Loads Within the Applicable LimitsThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to periodically test the emergency diesel generator (EDG) capacity to start and accelerate all of the sequenced loads within the applicable limits. Specifically, surveillance requirement (SR) activities did not demonstrate that all of the EDG auto-sequenced loads started and accelerated within the applicable voltage and frequency limits during start-up and recovery. In addition, the licensee did not timely evaluate the surveillance data collected for the residual heat removal pump motors. The licensee captured this issue in their CAP as CARD 16-26535 and CARD 16-26536, and performed an operability evaluation which reasonably determined the affected systems, structures, and components (SSCs) were operable but nonconforming. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee evaluated the most recent data and reasonably determined that the EDGs and the affected loads were operable. 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 associated SR procedures were established more than 3 years ago.
05000341/FIN-2016007-052016Q3GreenNRC identifiedFailure to Leak Test All Division 2 NIAS Boundary Isolation ValvesThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to leak test all Division 2 non-interruptible control air system (NIAS) boundary isolation valves. Specifically, the periodic NIAS leak testing did not account for the potential leakage of two valves used to isolate the NIAS safety-related system from the nonsafety-related interruptible control air system. The licensee captured this issue in their CAP as CARD 16-26389 and performed an operability evaluation which reasonably determined that Division 2 of NIAS remained functional. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, it was associated with the Barrier Integrity cornerstone attribute of SSC and barrier 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. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee used available data from a recent event and reasonably determined that system out-leakage was within the design limit. In addition, with respect to the Barrier Integrity cornerstone, the finding only represented a potential degradation of the control room and standby gas ventilation systems. 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 test procedures were established more than 3 years ago.
05000341/FIN-2016007-062016Q3GreenNRC identifiedFailure to Ensure that the MSIVs Would Close Within the TS Required Timeframe and as Described in the UFSARThe 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 licensees failure to ensure that the main steam isolation valves (MSIVs) would close within the Technical Specification time requirements and with the motive forces described in the Updated Final Safety Analysis Report. Specifically, the SR procedures did not account for the steam flow closing force, accumulator pressure variances, and containment pressure when verifying that the MSIVs will close within the SR time acceptance criteria. In addition, the licensee had not demonstrated that the MSIVs would close with air pressure and/or spring force against peak containment pressure as described in the Updated Final Safety Analysis Report. The licensee captured this issue in their CAP as CARD 16-27189 and CARD 16-26697, and performed evaluations that reasonably determined the affected MSIVs remained operable. 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 (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. In addition, it was determined to be more-than-minor because it was associated with the Initiating Event cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as of very-low safety significance (Green) because it did not result in exceeding the reactor coolant system leak rate for a small LOCA or affected other systems used to mitigate a LOCA. In addition, it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components, and it did not involve an actual reduction in the function of hydrogen igniters in the reactor containment. The team did not identify a cross-cutting aspect associated with this finding because it was not reflective of current performance. Specifically, the most significant cause for the performance issues discussed had existed for at least 3 years.
05000341/FIN-2016007-082016Q3GreenNRC identifiedInadequate Containment Debris Inspections Acceptance CriteriaThe team 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 establish procedures that were appropriate to inspect containment debris. Specifically, the emergency core cooling system (ECCS) suction strainer and containment coating inspection procedures contained acceptance criteria that was inconsistent with the applicable design documents. The licensee captured this issue in their CAP as CARD 16-26128 and CARD 16-26585, and reasonably determined that the concern did not impact the affected SSCs functionality based on recent inspection results. The performance deficiency was determined to be more-than-minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, a review of recent inspection did not find a condition that reasonably challenged the applicable design analysis and all loose material identified during the inspections was removed. 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 procedures were established more than 3 years ago.
05000341/FIN-2016007-092016Q3GreenNRC identifiedFailure to Evaluate the Acceptability of Drywell Coatings with Respect to Potential ECCS Suction Strainer BlockageThe 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 licensees failure to evaluate the acceptability of drywell coatings with respect to potential ECCS suction strainer blockage. Specifically, the licensee had not ensured that coating Carbo Zinc 11 would remain attached to the base metal during a DBA and the ECCS suction strainer calculations did not account for this material as a potential source of debris blockage. The licensee captured this issue in their CAP as CARD 16-26581 and reasonably determined that the affected coating system would remain adhered during a LOCA by comparing Carbo Zinc 11 installation documents against DBA test reports for this coating. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and 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 of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee compared Carbo Zinc 11 installation documents against DBA test reports for this coating and reasonably concluded that this coating system would remain adhered in the event of a LOCA. 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 associated evaluations were performed more than 3 years ago.
05000341/FIN-2016007-102016Q3GreenNRC identifiedNon Conservative ECCS Suction Strainer Min-K Combined Generation and Transport FactorsThe 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 licensees failure to use the min-K insulation debris generation and transport factors contained in the ECCS suction strainer licensing basis. Specifically, the licensee used non-conservative min-K insulation debris generation and transport factor values. The licensee captured this issue in their CAP as CARD 16-26800 and performed an operability evaluation that reasonably determined, based on industry test data, the existing calculation had sufficient conservatism to accommodate the effects of the additional debris volume. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and 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 of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation and reasonably determined that the existing calculation had sufficient conservatism to accommodate the effects of the additional debris volume. 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 associated evaluations were performed more than 3 years ago.
05000341/FIN-2016007-112016Q3GreenH.6NRC identifiedFailure to Apply Design Control Measures to a Design Change Associated with NIAS Accumulator CapabilityThe 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 licensees failure to apply design control measures to a design change associated with NIAS accumulator capacity. Specifically, the licensee did not verify that the reduced accumulator capacity was adequate during the entire time period that the compressor is expected to not be running, and ensure that operability limits and calibration tolerances contained in procedures were consistent with the new design. The licensee captured this issue in their CAP as CARD 16-26208, CARD 16-26561, and CARD 16-26607, and reasonably concluded that NIAS remained functional. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, it was associated with the Barrier Integrity cornerstone attribute of SSC and barrier 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. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed a bounding assessment that reasonably determined that the accumulator would maintain adequate pneumatic supply. In addition, with respect to the Barrier Integrity cornerstone, the finding only represented a potential degradation of the control room and standby gas ventilation systems. The team determined that the associated finding had a cross-cutting aspect in the area of Human Performance because the licensee did not carefully guarded margins and changed them only through a systematic and rigorous process. Specifically, the licensee failed to review and identify all of the design attributes associated with NIAS system before significantly reducing the accumulator capacity design margin in February 2016.
05000341/FIN-2016007-122016Q3GreenH.4NRC identifiedFailure to Identify an Out-of-Specification Pressure Reading on the Nitrogen Supply to the A MDCT Fan Motor Brake SystemThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify that the A MDCT fan motor brake system 100 psi nitrogen supply cylinder pressure did not meet the low-pressure acceptance criterion. Specifically, although the license had discovered this condition adverse to quality (CAQ), it was not captured into the CAP and was not corrected for a period of 7 consecutive days following its discovery. The licensee captured this issue in their CAP as CARD 16-26214, verified that the pressure of all MCDT fan motor brake cylinders were within limits, evaluated past operability, and performed a causal investigation. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a tornado event. Specifically, the licensee reviewed the pressure readings of the other nitrogen system supply cylinders and reasonably determined that their available pressure at the time would have compensated for the 100 psi cylinder low-pressure. The team determined that the associated finding had a cross-cutting aspect in the area of Human Performance because work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the nuclear operators and the control room licensed nuclear operators did not communicate and coordinate their activities to ensure the degraded condition was captured in the CAP.
05000341/FIN-2016007-142016Q3GreenNRC identifiedFailure to Identify that an Inadequate Minimum MSIV Accumulator Air Pressure Setpoint Was CAQThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify that an inadequate minimum MSIV accumulator air pressure setpoint was CAQ. Specifically, a licensee engineering evaluation concluded that the minimum MSIV accumulator air pressure setpoint was inadequate but the condition was not captured in the CAP and, as a result, corrective actions were not implemented. The licensee captured this issue in their CAP as CARD 16-26697 and reasonably determined the MSIVs remained operable. 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 (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding screened as of very-low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components, and it did not involve an actual reduction in the function of hydrogen igniters in the reactor containment. Specifically, the finding did not result in an actual open pathway and the MSIVs do not affect the function of heat removal components and hydrogen igniters. 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 calculation that concluded that the minimum air pressure setpoint was inadequate was performed in 1997.
05000341/FIN-2016007-152016Q3GreenH.14NRC identifiedFailure to Identify that a Non-Conservative Min-K Insulation Volume Calculation Error Was Nonconforming to the ECCS Suction Strainer Licensing BasisThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify that a non-conservative min-K insulation volume calculation error was nonconforming to the ECCS suction strainer licensing basis. Specifically, the licensee identified the non-conservative calculation error and captured it in the CAP as CARD 11-21153. However, the licensee did not identify any regulatory basis requiring this condition to be addressed and, as a result, closed the CARD without correcting the CAQ. The licensee captured this issue in their CAP as CARD 16-26292 and CARD 16-26800, and performed an engineering functional assessment that reasonably determined the affected SSCs remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation and reasonably determined the affected SSCs remained operable. The team determined that this finding had a cross cutting aspect in the area of Human Performance because the licensee did not propose an action that was determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee determined that no regulatory basis was associated with the non-conservative error because they could not find any requirement that specifically described the physical configuration and condition addressed in CARD 11-21153 when evaluating the problem in 2015.
05000341/FIN-2016007-162016Q3GreenH.13NRC identifiedFailure to Timely Identify, Document, and Evaluate Conditions that Challenge OperabilityThe team 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 timely identify, document, and evaluate nonconforming conditions that called the operability of one or more SSCs into question. Specifically, the licensee was not timely in capturing and evaluating ten CAQs identified during this inspection in their CAP and in accordance with their procedures, which resulted in untimely operability determinations. The licensee captured this issue in their CAP as CARD 16-26633, CARD 16-26776, CARD 16-26534, and CARD 16-26678, and completed the associated operability determinations, which reasonably determined the affected SSCs remained operable. The performance deficiency was determined to be more-than-minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed operability evaluations that reasonably determined that all of the affected SSCs remained operable. The team determined that this finding had a cross cutting aspect in the area of Human Performance because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensee did not use the CAPs systematic process to identify CAQs and make timely and adequate prompt operability decisions.
05000341/FIN-2016003-032016Q3GreenH.8NRC identifiedFailure to Follow Procedures During Concrete Placement of Flexible Storage Facility BuildingsThe inspectors identified a finding of very low safety significance when licensee personnel failed to follow the applicable procedure and design specification during concrete placement for installation of Diverse and Flexible Coping Strategies (FLEX) Buildings 1 and 2, identified as Flexible Storage Facility Buildings (FSF1 and FSF2). Specifically, the licensee failed to meet the requirements for limiting concrete pour heights and for treatment at cold joints. No violation of regulatory requirements was identified because construction of the FSF Buildings was not covered under 10 CFR 50, Appendix B. The finding was of more than minor safety significance because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the licensee failed to follow the instructions provided in the design specification and the plant procedure for concrete placement leading to potential degradation of the FSF building walls required for protection of the components needed for implementation of the FLEX in response to NRC Order EA12049, Order Modifying Licenses with Regard to Requirements for Mitigation Strategies for Beyond-Design-Basis External Events. In accordance with the NRC staff guidance for disposition of findings associated with NRC Order EA12049, the finding was presented to a cross-regional panel, which determined the finding to be a licensee performance deficiency of very low safety significance based on a qualitative evaluation of the potential consequences of the issue. The inspectors concluded this finding affected the cross-cutting area of human performance and the cross-cutting aspect of procedure adherence because licensee personnel failed to review and follow the applicable procedures and instructions while performing concrete placement work (IMC 0310, H.8).
05000341/FIN-2016007-022016Q3GreenP.6NRC identifiedFailure to Verify the Adequacy of the Voltage Supplied to Transformer #64 Load Tap ChangerThe 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 failure to verify the adequacy of the voltage supplied to the transformer #64 load tap changer. Specifically, the licensee did not perform calculations to verify that the load tap changer controls and actuator would have adequate voltage to be able to reset the degraded voltage relays following a design basis accident (DBA). The licensee captured this issue in their CAP as CARD 16-26702 and performed an operability evaluation that reasonably determined the voltage would marginally be acceptable to operate the load tap changer controls and actuator. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation that reasonably showed voltage would be marginally acceptable to operate the load tap changer controls and actuator when required during a DBA. The team determined that the associated finding had a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not conduct a self-critical and objective assessment of its programs and practices. Specifically, the licensee reviewed the applicability of a similar violation issued to a different licensee during the 2015 Component Design Bases Inspection Self-Assessment and concluded that it did not apply to Fermi.
05000341/FIN-2016007-072016Q3GreenP.6NRC identifiedFailure to Ensure that Protective Devices for the Loads Required at the Beginning of a LOCA Would Not Trip Under Degraded Voltage ConditionsThe 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 licensees failure to ensure that the protective devices for the loads required at the beginning of a loss of coolant accident (LOCA) would not trip under degraded voltage conditions. Specifically, the licensee did not verify that the connected Class 1E loads would not be damaged or become unavailable during a LOCA concurrent with a degraded voltage condition between the degraded voltage dropout setting and the loss of voltage setting for the degraded voltage time delay of 7.3 seconds and subsequent reconnection to the EDG. The licensee captured this issue in their CAP as CARD 16-26533 and performed a preliminary evaluation that reasonably determined the protective devices would not actuate during this condition. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation and reasonably determined that protective devices would not actuate during a degraded voltage concurrent with a LOCA. The team determined that the associated finding had a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not conduct a self-critical and objective assessment of its programs and practices. Specifically, the licensee evaluated a similar violation issued at a different licensee during the 2016 Component Design Bases Inspection Self-Assessment and concluded that no corrective actions were required.
Retrieved from "https://kanterella.com/Special:Ask/-5B-5BCategory:Finding-5D-5D-20-5B-5BSite::Fermi-5D-5D/-3FQuarter%3DQ/-3FSignificance/-3FCross-20cutting-20aspect%3DCCA/-3FIdentified-20by/-3FTitle/-3FDescription/mainlabel%3D/order%3Ddescending/sort%3DStart-20date/offset%3D0/format%3Dtable/searchlabel%3D-5BFindings-20Table-20View-20All-5D"