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| Q | Site | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|---|
| 05000341/FIN-2018002-01 | 2018Q2 | Fermi | Green | H.14 | Self-revealing | Failure to Document a Condition Assessment Resolution Document for Reactor Recirculation Motor-Generator Set A Brush Gear Sparking | A 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-2018002-04 | 2018Q2 | Fermi | TBD | H.11 | Self-revealing | Failure to Identify a Condition Adverse to Quality on Division 2 Residual Heat Removal Service Water Outlet Flow Control Valve | A 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-2018002-03 | 2018Q2 | Fermi | Green | P.2 | Self-revealing | Failure to Adequately Evaluate the Operability of Emergency Diesel Generator11 | A 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-2018002-02 | 2018Q2 | Fermi | Green | P.2 | Self-revealing | Inadequate Preventative Maintenance in Residual Heat Removal Service Water System Outlet Flow Control Valves Results in Lower Bonnet (Backseat) Bushing Failure | A 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 |
| 05000346/FIN-2018002-01 | 2018Q2 | Davis Besse | Green | H.12 | Self-revealing | Failure to Follow the Makeup and Purification Procedure | A self-revealed Green finding and associated Non-Cited Violation of Technical Specification 5.4.1.a, Procedures, was identified when the licensee failed to follow station procedure DBOP06006, Makeup and Purification System. Specifically, the licensee failed to open MU177, the Make-Up Filter 1 Outlet Isolation valve, which resulted in the isolation of letdown while swapping make-up filters. |
| 05000461/FIN-2018002-05 | 2018Q2 | Clinton | Severity level Minor | NRC identified | Minor Violation | The inspectors reviewed AR 4116223, Blown Fuses during CPS 9080.23 8.4 for Fast Transfers. The inspectors selected this sample for review due to repetitive fuse failures within the safety-related Division 3 NUS Modules dating back to 2013. As appropriate, the inspectors verified the following attributes during their review: complete and accurate identification of the problem in a timely manner commensurate with its safety significance and ease of discovery; consideration of the extent of condition, generic implications, common cause, and previous occurrences; evaluation and disposition of operability/functionality/reportability issues; classification and prioritization of the resolution of the problem commensurate with safety significance; identification of corrective actions, which were appropriately focused to correct the problem; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. Description: While reviewing the historical ARs associated with the NUS fuse failures, the inspectors discovered licensee information indicating the NUS fuse failures were likely caused by voltage/current transients within the upstream, safety-related 480V to 120V regulating transformer. The purpose of the transformer was to regulate voltage and current to the downstream components including the NUS modules. However, degradation in the transformers ability to regulate voltage and current levels could create a condition where the voltage and current levels exceeded the NUS fuse rating causing fuse failure. The licensee documented the potential transformer degradation issue on September 20, 2013, in AR 1561455, Division 3, Group 1 Instruments Found De-energized during CPS 9080.23, Specifically, the licensee stated, The most probable cause of the failure of the NUS modules was the transient voltage overshoot of the regulating transformer causing the transient protection varistors on the five NUS modules to actuate, drawing a near fault current until the individual and line feed fuses blew. Station procedure PI-AA-125, Corrective Action Program, defined equipment failure as, damage to or degradation of a system, structure or component that may cause or contribute to the event. Based on the information documented in AR 1561455, the licensee identified transient voltage overshoots in the 480V to 120V regulating transformer, which was a degraded condition causing the NUS modules to fail. Per the licensee definition this would constitute an equipment failure. No further action was taken to identify and correct the regulating transformer degradation until the transformer failed on March 18, 2018, impacting multiple pieces of safety-related Division 3 equipment. Minor Violation: Title 10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, requires conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and corrected. Contrary to this requirement, on September 20, 2013, the licensee identified a failure of the 480V to 120V regulating transformer, which manifested itself as a voltage overshoot causing the failure of the NUS modules, but failed to take actions to correct the condition. On March 18, 2018, the regulating transformer subsequently degraded further causing it to fail in a manner that tripped the upstream breaker and impacted additional pieces of safety-related Division 3 equipment. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. Specifically, the inspectors determined that although the transformer failure affected Division 3 equipment, the failure would not have impacted the Division 3 equipments ability to respond to a DBE or the capability to shut down the reactor and maintain it in a safe shutdown condition. Enforcement: The failure to comply with 10 CFR 50, Appendix B, Criterion XVI, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. | |
| 05000373/FIN-2018010-01 | 2018Q2 | LaSalle | Green | H.12 | NRC identified | Failure to Translate Reactor Building Superstructure Design Basis | Inspectors identified a Green finding and associated Non-Cited Violation of Title 10of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, for licensees failure to assure that applicable Updated Final Safety Analysis Report described design basis for the Reactor Building (RB) superstructure were correctly translated to field documents was a performance deficiency. Specifically, Updated Final Safety Analysis Report Tables 3.8-9 and 3.8-11 define the design basis load combinations and the corresponding design stress limits applicable to the RB superstructure. Design calculation L-003415 evaluates these load combinations and applies RB overhead crane lifting limitations which ensures these design basis are met. The licensee failed to translate these limitations into specifications, drawings, procedures, or instructions which would ensure the specified stress limits for RB design basis load combinations would not be exceeded while operating the RB overhead crane. |
| 05000315/FIN-2018002-06 | 2018Q2 | Cook | Severity level Minor | NRC identified | Minor Violation | Technical Specification (TS) 5.4, Procedures, requires that the applicable procedures recommended in Regulatory Guide 1.33 be established, implemented, and maintained. Regulatory Guide 1.33 states that maintenance that could affect the performance of safety-related equipment should be properly pre-planned and performed in accordance with procedures appropriate to the circumstances. Contrary to this requirement, procedure 12EHP4030056218, Automatic Operation of Auxiliary Feedwater Pumps, was not performed as written in the procedure. Specifically, pages were skipped which resulted in the 2CD EDG inadvertently starting during the surveillance. Screening: The issue resulted in momentary loss of the T21C and T21D vital busses until the 2CD EDG reached rated speed and connected to the busses. The reactor was defueled at the time. One train of spent fuel pool cooling was lost for several minutes, but the other train stayed in service and there was no apparent change in spent fuel pool temperature. The issue screened as minor based on the guidance in IMC 0612 Appendix E because there were no safety consequences and there was no transient of any significance. Violation: This failure to comply with TS 5.4 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. | |
| 05000315/FIN-2018002-05 | 2018Q2 | Cook | Severity level Minor | NRC identified | Minor Violation | While there did appear to be a reduction in operational errors being made in the field while manipulating equipment (such as during clearance activities and in performing certain evolutions) the inspectors noted a trend in configuration control issues. Most of these dealt with some kind of operation department interface or coordination with another department. In one case, valves associated with feedwater heater level control were left closed following a project to replace some of the heaters, which contributed to a manual reactor trip due to high moisture-separator drain tank level when starting the plant following the Unit 2 refueling outage. Other examples were Chemistry and Operations department coordination on an non-essential service water (NESW) valve alignment which led to NESW being isolated to generator seal oil cooling during plant startup, poor coordination between Maintenance and Operations which resulted in a containment penetration being left open, a pressure gauge remaining isolated after the Projects department completed the heater drain pump replacements, and the failure to ensure that valve-closure tests were done following the feedwater heater replacements. Another identified trend was in the area of post-maintenance testing (PMT). During the refueling outage on Unit 2, both the NRC and the licensee identified instances of improper PMTs being scheduled for safety-related equipment. Inspectors identified work on an EDG fuel oil transfer pump that did not have an in-service test (IST) scheduled. The licensee identified the lack of a time response test following a motor-driven AFW pump motor replacement, was a repeat issue from the previous outage. The licensee also identified the lack of an IST following a seal replacement on a CCW pump. In each case, the issues were discovered and corrected before equipment was restored to fully operable status. In response to the trend, the licensee reviewed other work on safety-related equipment for the outage to confirm the proper PMTs would be done. No other issues were identified. Finally, early in the observation period, the inspectors noted a trend in procedure quality for maintenance activities on safety-related equipment. There were instances regarding Turbine-Driven Auxiliary Feedwater (TDAFW) pump linkages where better procedure direction could have precluded binding and governor-valve travel issues. Additionally, while replacing a TDAFW governor, a snap ring was inadvertently left out of a coupling due to insufficient procedure detail. Regarding the EDGs, the licensee discovered instructions for assembly of air start check valves did not contain the torque guidance that the vendor drawings stipulated. In response to this trend, the licensee started to perform deliberate reviews of OE before maintenance on some safety-related equipment, to verify maintenance instructions had up-to-date guidance before starting work. No violations or findings were identified by the inspectors. 12 Licensee management acknowledged the issues discussed by the inspectors. | |
| 05000315/FIN-2018002-04 | 2018Q2 | Cook | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: Title 10 Code of Federal Regulations; Part 20.1501(c) requires that the licensee shall ensure that instruments and equipment used for quantitative radiation measurements are calibrated periodically for the radiation measured. Contrary to the above, between November 2012 and May 2017 the licensee used the liquid scintillation counter for quantitative radiation measurements outside the range of equipment capability and the system calibration. The licensee analyzed the impact on the annual effluent reports and UFSAR limits between 1/8/2013 and 5/3/2017. The licensee entered the violation on the corrective action program. Licensee Identified Non-Cited Violation Significance/Severity Level: Green. The inspectors determined the performance deficiency was more than minor because it adversely affected the Plant Facilities/Equipment and Instrumentation attribute of the Public Radiation Safety Cornerstone objective to ensure adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors assessed the significance of the finding usingSDP Appendix D and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20174835 | |
| 05000315/FIN-2018002-03 | 2018Q2 | Cook | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy Violation: Title 10 of the Code of Federal Regulations (10 CFR) 50.47 b(8) requires that licensee emergency plans meet the standard of having adequate emergency facilities. The Cook Plant Emergency Plan states that the Technical Support Center (TSC) (an emergency facility) will be constructed to provide the same degree of radiological habitability as the Control Room under accident conditions. Contrary to the above, from January 24 to 30, 2018, the licensee failed to maintain the TSC as an adequate emergency facility, by installing a portable air conditioning unit in the Shift Managers office which compromised the ability of the TSC ventilation system to fulfill its function of providing the necessary radiological protection for the TSC. Specifically, the exhaust from the portable unit was routed to an existing ventilation duct of the TSC ventilation system, and a panel on one of the ventilation units was opened, exposing the TSC to the turbine building environment. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Facilities and Equipment attribute of the Emergency Preparedness cornerstone, whose objective is 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. The inspectors assessed the significance of the finding usingSDP Appendix B and concluded the violation was of very low safety or security significance (Green). Corrective Action Reference: AR20180952 | |
| 05000373/FIN-2018002-01 | 2018Q2 | LaSalle | Green | Self-revealing | Failure to Implement a Preventative Maintenance Strategy for Residual Heat Removal Service Water Pump Shorting Relays | A self-revealed Green finding of very low safety significance was identified for the licensees failure to implement a preventative maintenance (PM) strategy for the residual heat removal service water (RHRSW) pump shorting relays in accordance with procedure MAAA716210, Performance Centered Maintenance (PCM) Process, Revision 11. Specifically, a PCM template was issued in 2002 that required periodic as-found testing and calibration for control and timing relays, but a maintenance strategy was never implemented. As a result, one of the normally closed contacts on the Unit 1 D RHRSW pump shorting relay developed a high contact resistance and prevented the Unit 1 D RHRSW pump from starting. | |
| 05000373/FIN-2018002-02 | 2018Q2 | LaSalle | Green | NRC identified | Failure to Follow Procedure and Perform Database Revision Review Requirements | The inspectors identified a Green finding of very low safety significance for the licensees failure to follow procedure NSWPWM03, Predefine Database Revisions, Revision 0, for retiring procedure LESGM108, Inspection of 480V Motor Control Center Equipment, that performed bus bar inspection on Division 3 motor control centers. Specifically, instead of completing NSWPMW03, step 6.5, Database Revision Review Requirements, to retire the bus bar inspections for Division 3 motor control centers, the licensee retired the procedure based solely on having previously retiring the bus bar inspections for Division 1 and Division 2 in 2002,and did not performthe required review. | |
| 05000373/FIN-2018002-03 | 2018Q2 | LaSalle | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significant was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section 2.3.2 of the Enforcement Policy. Violation: Technical Specification LCO 3.4.4 (applicable for Modes 1, 2 and 3) states: The safety function of 12 safety relief valves (S/RVs) shall be OPERABLE, and Action Statement A states that One or more required S/RVs inoperableA.1 be in mode 3 in 12 hours and A.2 be in Mode 4 in 36 hours. Technical Specification SR 3.4.4.1 states that Verify the safety function lift setpoints of the required S/RVs are as follows Number of S/RVs Setpoint (psig 2 1205 36. 3 1195 35. 2 1185 35. 4 1175 35. 2 1150 34. Contrary to the above, during portions of previous Unit 1 and 2 operating cycles from 2012 through January of 2017, two main steam S/RVs did not meet these lift pressure setpoint requirements. Specifically S/RV 2B21F013C lifted at 1131 psig instead of from 1139.8 to 1210.2 psig and S/RV 2B21F013L lifted at 1130 psig instead of from 1159.2 to 1230.8 psig (reference: Licensee Event Report 05000374/201700400; 01, Two Main Safety Relief Valves Failed Inservice Lift Inspection Pressure Test. Significance/Severity: This licensee identified finding affected the Initiating Events Cornerstone and was screened in accordance with Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At Power. The two affected SRVs lifted low outside of their setpoint band, which was conservative with respect to maintaining the reactor coolant system overpressure protection safety function of these valves. Therefore, the inspectors determined that this finding is of very low safety significance (Green) because after a reasonable assessment of degradation, the finding would not have resulted in exceeding the reactor coolant system leak rate for a small LOCA and did not affect other systems used to mitigate a loss-of-coolant accident. Corrective Action Reference: AR 3974669 | |
| 05000373/FIN-2018002-04 | 2018Q2 | LaSalle | Severity level Minor | NRC identified | Minor Violation - Follow-up of Events and Notices of Enforcement Discretion | Minor Violation: For S/RV 2B21F013L, serial number N63790050012 (hereafter referred to as S/RV 12), the licensee completed a work group evaluation as documented in AR 03975216ACIT No. 3 to investigate the cause for two S/RVs that failed a set pressure lift test out of specification low. For ACIT No. 3, the licensee staff incorporated a vendor letter that documented the results of the S/RV vendors review of the S/RV 12 condition and which recorded an out of tolerance spring condition. It stated that The spring was measured and rate tested. The free height was found to be below the minimum original equipment manufacturer specified tolerance. The licensees vendor subsequently replaced the nonconforming spring with a new spring. In prior vendor correspondence with the licensee (reference E-mail dated June 24, 2015), the vendor stated that Typically we contribute a low as-found lift to an out-of-tolerance spring rate or free height dimension. Therefore, the nonconforming spring free height dimension may have caused the low as-found lift setpoint failure for this valve and as such was relevant (e.g. material) to the determination of a failure cause that was reported in LER 05000374/201700400 and 01. However, the licensee failed to identify this during their cause investigation and erroneously reported in LER 05000374/201700400 and 01 that The vendor reported for both valves that all the spring tolerances were within the acceptance limits. The licensee documented this violation in AR 04134591, Potential Minor Violation for Unit 2 LER 20170401. The licensee also submitted a revision to the LER as LER 05000374/201700402 Screening: The significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which could impede the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The inspectors determined that this issue was a Severity Level IV violation based on Example 6.9.d.10 in the NRC Enforcement Policy which states, A failure to identify all applicable reporting codes on a Licensee Event Report that may impact the completeness or accuracy of other information (e.g. performance indicator data) submitted to the NRC. In accordance with the Section 2.2.1.c of the NRC enforcement policy, the severity level of a violation involving the failure to make a required report to the NRC will depend on the significance of and the circumstances surrounding the matter that should have been reported. The NRC had not relied on information in this LER report to make a regulatory decision, and the inspector answered no to each of the more than minor screening questions in Appendix B of IMC 0612 for the issue of concern. Therefore, the NRC determined this was a minor violation because it was associated with a minor performance deficiency. Violation: Failure to comply with 10 CFR 50.9 Completeness and accuracy of information and accurately report the nonconforming S/RV 12 spring tolerance in LER 05000374/201700400 and 01 to the NRC constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. | |
| 05000315/FIN-2018002-02 | 2018Q2 | Cook | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy. Enforcement: Violation: License conditions 2.C.(4) (Unit 1) and 2.C.(3)(o) (Unit 2) require implementation of the approved fire protection program. Per the Cook NFPA 805 Fire Protection Program Manual Sections 3.11.2 and 3.11.4, fire seals shall have at least a three hour fire rating. Contrary to the above, on February 6, 2018, the licensee identified multiple fire seals (many of which were between the control rooms and the cable spreading area underneath) that were degraded to the point that they could no longer meet the three hour rating requirement of Sections 3.11.2 and 3.11.4 of the Cook NFPA 805 Fire Protection Program Manual. Specifically, inadequate controls in the fire seal maintenance procedure and unclear guidance for Performance Verification department inspections led to a deterioration in seal quality. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it adversely affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone, whose objective is to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e. core damage). The inspectors assessed the significance of the finding usingSignificance Determination Process Appendix F and concluded the violation was of very low safety significance (Green).Corrective Action Reference: AR20181208 | |
| 05000316/FIN-2018002-01 | 2018Q2 | Cook | Green | H.11 | Self-revealing | Steam Dump Closure Caused by Human Error | On May 10, 2018, a Green self-revealed finding and associated Non-Cited Violation occurred when licensee personnel caused the Unit 2 steam dump valves to the condenser to close. Specifically, when tuning the controller for the steam dump valves, licensee personnel left the controller in automatic, resulting in the closure of all the steam dump valves. This caused both the steam generator power operated relief valves and a steam generator safety valve to lift. |
| 05000440/FIN-2018410-01 | 2018Q2 | Perry | Green | H.3 | NRC identified | Security | |
| 05000440/FIN-2018002-01 | 2018Q2 | Perry | Green | H.12 | NRC identified | Failure to Control Transient Combustible Materials in a Designated Combustible Control Zone | The inspectors identified a finding of very low safety significance (Green) and an associated non-cited violation (NCV) of Perry Operating License Condition 2.C(6), Fire Protection, for the licensees failure to control transient combustible materials in a designated combustible control zone within fire area 1AB1g on Auxiliary Building elevation 574 10. Specifically, on May 16, 2018, the inspectors identified transient combustible materials left unattended in the designated combustible control zone in the corridor outside the emergency core cooling system (ECCS) pump rooms, which exceeded the ten pound limit established in the Fire Protection Program document, PAP1910, for ordinary combustibles (loose) in designated combustible control zones without a transient combustible permit. |
| 05000454/FIN-2018002-03 | 2018Q2 | Byron | Severity level Minor | NRC identified | Minor Violation | Minor Violation: The inspectors identified multiple instances of a failure to perform inservice testing in accordance with written procedures appropriate for the circumstances during this inspection period: 1. On March 30, 2018, the licensee performed 1BOSR 5.5.8.DO2, Test of the Diesel Oil Transfer System, and declared the 1B diesel oil transfer pump inoperable due to flow results being low out of specification. Subsequently, the licensee determined that the instrument setup was incorrect in that an incorrect value was entered into the flow meter for pipe diameter. The licensee declared the surveillance invalid and scheduled a time to re-perform the activity. Acceptable system flow rates were achieved a week later when the correct pipe diameter was used for the instrument setup. 2. On April 26, 2018, while observing the licensee perform 2BOSR 5.5.8.CS.52C, Comprehensive Inservice Testing (IST) Requirements for Containment Spray Pump 1CS01PB, the inspectors noted that the pump suction pressure and discharge pressure test gauges were not installed as described in the Precautions and Limitations section of the procedure. After the inspectors asked how the installed configuration satisfied the procedure requirement, the licensee suspended the test to obtain clarification. After some deliberation between engineers and operators attempting to identify the correct instrument location, the test data was recorded with the instruments at different locations for data gathering and comparison. The licensee verified that pump performance had sufficient margin, including the introduced error, to remain operable and available to perform its safety-related function as expected.3. On May 1, 2018, while observing the licensee perform 2BOSR 5.5.8.SX.51C, Comprehensive Inservice Testing (IST) Requirements for the Essential Service Water (SX) Pump 2SX01PA and Unit 2 SX Pumps Discharge Check Valves, the inspectors noted that operators were not taking data from the ultrasonic flow meter in accordance with the procedure. Specifically, the instrument was not set up to indicate time and flow so that an average flow could be determined as required by a Note in the procedure. Instead the operators were recording instantaneous flowrate. When the inspector asked for clarification and the operators and technicians deferred to their supervisors, the licensee suspended the test to obtain clarification. The test was performed again after the instrument was set up correctly and operators were briefed on how to obtain the correct data.Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions. Contrary to the above, for the diesel fuel oil transfer pump surveillance, 1BOSR 5.5.8.DO2, the procedure listed an incorrect pipe diameter value that was subsequently entered into the flow meter resulting in unacceptable test results; for the containment spray pump surveillance, 2BOSR 5.5.8.CS.52C, the licensee potentially introduced an unaccounted for error in the surveillance test method by not setting up test equipment in accordance with the procedure; and for the SX surveillance, 2BOSR 5.5.8.SX.51C, the licensee introduced a potential error in the surveillance test by not determining an average flow rate as discussed in the procedure Note.Screening: The failure to perform inservice testing in accordance with written procedures appropriate for the circumstances was a performance deficiencyin each of the listed 11 examples. The performance deficiency was determined to be minor in each case because the inspectors answered No to all of the more-than-minor screening questions in IMC 0612, Appendix B. The licensee generated the following issue reports (IRs) to document these issues:AR 04121539, Ultrasonic Flow Measurement Installation IssueAR 04122295, PCR (procedure change request) 1/2BOSR 5.5.8.DO1 AR 04131201, Engineering Clarification Needed on ASME Precaution AR 04133585, NRC ID: Potential Concerns With Execution of 2A SX Pump Surveillance Violation: These failures to comply with 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, constituted minor violations that are not subject to enforcement action in accordance with the NRCs Enforcement Policy | |
| 05000454/FIN-2018410-01 | 2018Q2 | Byron | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000263/FIN-2018002-01 | 2018Q2 | Monticello | Green | Licensee-identified | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensee corrective action program and is being treated as a Non-Cited Violation, consistent with Section2.3.2 of the Enforcement Policy.Enforcement: Violation: Title 10 CFR 50.54(q)(2) requires that a holder of a nuclear power reactor operating license follow and maintain the effectiveness of an emergency plan that meets the requirements of 10 CFR Part 50, Appendix E and the planning standards of 10 CFR 50.47(b). Title 10 CFR Part 50.47(b)(8) requires, in part, that a licensee must provide and maintain adequate emergency facilities and equipment to support the emergency response plan.Contrary to the above requirements, on March 23, 2018, the licensee identified the site failed to maintain the effectiveness of the emergency plan by not providing and/or maintaining equipment capable of measuring the Immediately Dangerous to Life and Health (IDLH) concentrations for several toxic chemicals as required to properly classify an Alert Emergency Action Level (EAL). Specifically, while performing an emergency equipment inventory, the licensee identified that detector tubes (Draeger tubes) available to measure chlorine gas concentrations were not capable of measuring the IDLH concentration of 10 ppm required to identify the threshold level for classifying an Alert EAL (HA 3.1) since the measurement range of the available sample tubes was 50500 ppm.The inability to properly classify the Alert EAL represented a Loss of Emergency Assessment Capability and resulted in the licensees submission of Event Notification Report # 53298 in accordance with the requirements of 10 CFR 50.72(b)(3)(xiii). An immediate extent of condition review performed by the licensee identified additional deficiencies in adequate sampling methods for determining IDLH concentrations for Butadiene, Ethylene Dichloride, and Gasoline. Additionally, the licensee identified that in April 2015 there was missed opportunity to correct this deficiency when an Emergency Preparedness (EP) Coordinator, performing a Control Room Emergency Equipment Inventory, identified the need to order and replace the existing chlorine detector tubes. The EP Coordinator added the incorrect detector tubes to the existing inventory form without validating the tubes detection range and accuracy to ensure it was capable of detecting the IDLH threshold concentration level of 10 ppm.Upon identification of the issue, the licensee implemented compensatory measures for determining the EAL classification and entered the issue into the corrective action program (CR 501000009876). On May 08, 2018, the licensee implemented the sites new EAL classification procedure that was developed using NEI 9901, Revision 6, which does not require atmospheric sampling (use of detection tubes) for classification of EAL HA 3.1.Significance/Severity Level: Using IMC 0609, Appendix B, Emergency Preparedness Significance Determination Process, Table 5.81, the inspectors determined this finding was 10 of very low safety significance (Green) because a significant amount of equipment necessary to implement the E-plan was not available or functional to the extent that any key ERO member could not perform his/her assigned functions, in the absence of compensatory measures (Degraded Planning Standard), specifically the ability to accurately classify the Alert EAL. Determining the finding significance using IMC 0609, Appendix B, Table 5.41, results in the same finding significance (very low significance) since the performance deficiency would have rendered an EAL initiating condition ineffective such that the Alert would have been declared in a degraded manner.Corrective Action Reference: 501000009876, CR Toxic Gas Detector Tube. | |
| 05000266/FIN-2018002-01 | 2018Q2 | Point Beach | Green | H.7 | NRC identified | Primary Auxiliary Building Floor Plug Removal Creates Unanalyzed Flood Path | A Green finding and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations Part 50, Appendix B, Criterion III, Design Control, was identified by the inspectors for the licensees failure to ensure that applicable regulatory requirements and design basis, for structures, systems, and components, were translated into procedures. Specifically, the licensee failed to include the floor plugs on the 26 level of the primary auxiliary building as credited flood barriers in procedure NP 8.4.7, PBNP Flooding Program. |
| 05000266/FIN-2018002-02 | 2018Q2 | Point Beach | Green | NRC identified | Unanalyzed Condition for Tornado Generated Missiles | On June 10, 2015, the NRC issued Regulatory Issue Summary (RIS) 201506, Tornado Missile Protection (ML15020A419), focusing on the requirements regarding tornado-generated missile protection and required compliance with the facility-specific licensing basis. The RIS also provided examples of noncompliance that had been identified through different mechanisms and referenced Enforcement Guidance Memorandum (EGM) 15002, Enforcement Discretion For Tornado Generated Missile Protection Non-Compliance, which was also issued on June 10, 2015, (ML15111A269) and revised on February 7, 2017, (ML16355A286). The EGM applies specifically to a SSC that is determined to be inoperable for tornado generated missile protection. The EGM stated that a bounding risk analysis performed for this issue concluded that tornado missile scenarios do not represent an immediate safety concern because their risk is within the LIC504, Integrated Risk-Informed Decision-Making Process for Emergent Issues, risk acceptance guidelines. In the case of Point Beach, the EGM provided for enforcement discretion of up to three years from the original date of issuance of the EGM. The EGM allowed NRC staff to exercise this enforcement discretion only when a licensee implements, prior to the expiration of the time mandated by the LCO, initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. In addition, licensees were expected to follow these initial compensatory measures with more comprehensive compensatory measures within approximately 60 days of issue discovery. The comprehensive measures should remain in place until permanent repairs are completed, or until the NRC dispositions the non-compliance in accordance with a method acceptable to the NRC such that discretion is no longer needed. Table 1.31 of the Point Beach Final Safety Analysis Report (FSAR) states, in part, that SSCs, which are essential to the prevention and mitigation of nuclear accidents, shall be designed, fabricated, and erected to withstand the forces that might reasonably be imposed by the occurrence of an extraordinary natural phenomenon, such as a tornado. On March 1, 2018, the licensee initiated AR 02252240, identifying a nonconforming condition of Table 1.31. Specifically, on both units 1 and 2, the steam supply lines and exhaust stacks for the turbine-driven auxiliary feedwater pumps, the main steam isolation valves, the atmospheric steam dumps, the main steam safety valves, and the vents for T175B bulk fuel oil storage tank were not adequately protected from tornado-generated missiles. The licensee declared the affected SSCs inoperable and promptly implemented compensatory measures designed to reduce the likelihood of tornado-generated missile effects. The condition was reported to the NRC as Event Notice 53239 as an unanalyzed condition and potential loss of safety function. Enforcement discretion was previously authorized and documented in Inspection Report 05000266/2018001 (ADAMS Accession Number ML18128A229). Corrective Actions: The licensee documented the inoperability of the SSCs and the affected TS LCO conditions in the CAP and in the control room operating log. The shift manager notified the NRC resident inspector of implementation of EGM 15002, and documented the implementation of the compensatory measures to establish the SSCs operable but nonconforming prior to expiration of the LCO required action. The licensees immediate compensatory measures included: review and revision of procedures for a tornado watch and a tornado warning to provide additional instructions for operators preparing for tornados and/or high winds, and a potential loss of SSCs vulnerable to the tornado missiles; confirmation of readiness of equipment and procedures dedicated to the Diverse and Flexible Coping Strategy (FLEX); verification that training was up to date for individuals responsible for implementing preparation and response procedures; and establishment of a heightened station awareness and preparedness relative to identified tornado missile vulnerabilities. The licensees longer term compensatory measure was to modify AOP13C, Severe Weather Conditions procedure, to include actions for removing potential airborne hazards and damage assessments for systems with a vulnerability to damage from tornado missiles. Corrective Action Reference: AR 2252240 Enforcement: Violation: The enforcement discretion was applied to the required shutdown actions of the following TS LCOs for both units: TS 3.0.3, General Shutdown LCO (cascading or by reference from other LCOs); TS 3.7.1, Main Steam Safety Valves (MSSVs); TS 3.7.2, Main Steam Isolation Valves (MSIVs) and Non-Return Check Valves; TS 3.7.4, Atmospheric Dump Valve (ADV) Flowpaths; TS 3.7.5, Auxiliary Feedwater (AFW); TS 3.8.1; AC Sources Operating; and TS 3.8.3, Diesel Fuel Oil and Starting Air. Severity/Significance: The subject of this enforcement discretion, associated with tornado missile protection deficiencies, was determined to be less than red (i.e., high safety significance) based on a generic and bounding risk evaluation performed by the NRC in support of the resolution of tornado-generated missile non-compliances. The bounding risk evaluation is discussed in Enforcement Guidance Memorandum 15002, Revision 1, Enforcement Discretion for Tornado-Generated Missile Protection Non-Compliance, and can be found in ADAMS Accession Number ML16355A286. Basis for Discretion: The NRC exercised enforcement discretion in accordance with Section 2.3.9 of the Enforcement Policy and EGM 15002 because the licensee initiated initial compensatory measures that provided additional protection such that the likelihood of tornado missile effects were lessened. The licensee implemented more comprehensive compensatory measures to address the nonconforming conditions within the required 60 days. These comprehensive actions are to remain in place until permanent repairs are completed, which, for Point Beach, were required to be completed by June 10, 2018, or until the NRC dispositioned the non-compliance in accordance with a method acceptable to the NRC, such that discretion was no longer needed.On April 26, 2018, the licensee submitted a request to extend the enforcement discretion in letter titled Request to Extend Enforcement Discretion Provided in Enforcement Guidance Memorandum 15002 for Tornado-Generated Missile Protection Non-conformances Identified in Response to Regulatory Issues Summary 201506, Tornado Missile Protection. On May 21, 2018, the NRC approved this request and extended the enforcement discretion until June 10, 2020. The disposition of this enforcement discretion closes LER 201800100. | |
| 05000255/FIN-2018415-01 | 2018Q2 | Palisades | Green | NRC identified | Security | ||
| 05000456/FIN-2018002-03 | 2018Q2 | Braidwood | Green | H.4 | Self-revealing | Inadequate Test Activity Coordination Results in Unintended Valve Actuation and Reactor Coolant System Pressure Drop | A self-revealed finding of very low safety significance (i.e., Green) and an associated NCV of Technical Specification 5.4, Procedures, was identified for the licensees failure to have properly coordinated testing activities associated with redundant Unit 1 pressurizer pressure instruments in accordance with the stations procedural requirements governing such testing. Specifically, during the licensees 20th Unit 1 refueling outage, on April 23, 2018, redundant pressurizer pressure instrumentation channels were inadvertently subjected to simultaneous testing activities. This resulted in the coincidence logic for both of the units pressurizer power-operated relief valves (PORVs) being satisfied and the PORVs opening to depressurize the RCS from approximately 345 pounds per square inch gauge (psig) to approximately 320 psig |
| 05000255/FIN-2018011-03 | 2018Q2 | Palisades | Green | Licensee-identified | Licensee-Identified Violation | License condition 2.C(3)requires the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program that complies with Title 10of the Code of Federal Regulations(CFR), Part50.48(a) and 10 CFR 50.48(c), NFPA Standard NFPA 805, as approved in the Safety Evaluation Report (SER)dated February 27, 2015. Section 2.4.3.3 of NFPA 805 states, in part, that the Probabilistic Safety Assessment (PSA)(Probabilistic RiskAssessment (PRA))approach, methods, and data shall be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant.Contrary to the above, from February 27, 2015, until May 14, 2018, the licensee failed to base the PSA (PRA) approach, methods, and data on the as-built and as-operated and maintained plant.Specifically, the licensees PSA (PRA) model/analysis credited the suppression system located in the cable spreading room to suppress a type 2 fire scenarios, whereas the actual room contained numerous obstructions by the stacked cable trays located near the ceiling that interfered with the water spray pattern discharged from the sprinklers from providing adequate water density pattern to suppress a fire in areas below the cable trays which contained electrical panels.Significance/Severity Level: The performance deficiency was determined to be more-than-minor, and therefore, a finding because the performance deficiency, if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, the licensees failure to correctly model/analyze the as-built condition of the suppression system located in the cable spreading room in the PRA could potentially affect the risk associated with a fire in the room and could result in inappropriately screening out the effects of otherchanges associated with the fire area.The finding was of very-low safety significance (Green). While there may be a change to the plants baseline risk as a result of this issue, this is a fire modeling issue only; no physical plant fire protection feature was altered by the fire PRA model. Therefore, there was no increase in actual core damage risk to the physical plant. | |
| 05000454/FIN-2018002-02 | 2018Q2 | Byron | Green | Licensee-identified | Licensee-Identified Violation | A violation of very low safety significance was identified by the licensee, has been entered into the licensees corrective action program, and is being treated as a non-cited violation, consistent with Section 2.3.2 of the Enforcement Policy Violation: Title 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished.Licensee procedure ERAA321, Administrative Requirements for Inservice Testing, stated in Step 4.10.5, that acceptance criteria are established using the reference values and the applicable ASME (American Society of Mechanical Engineers) Code. Paragraph ISTA3160, Test and Examination Procedures, of the ASME Operation and Maintenance of Nuclear Power Plants (OM) Code required in part that, Tests and examinations shall be performed in accordance with written procedures. The procedures shall contain the Owner-specified reference values and acceptance criteria. Paragraph ISTA9230, Inservice Test and Examination Results, of the ASME OM Code required, in part, that The results of tests and examinations shall be documented and shall include the following: comparison with allowable ranges of test and examination values, and analysis deviations and requirements for corrective action.Contrary to the above, from July 1, 2016, to May 30, 2018, the licensees procedures did not clearly document acceptance range, alert range, and required action values for the diesel oil (DO) transfer pump IST surveillance tests in accordance with the ASME OM Code. This resulted in several instances where the pump being tested did not meet IST criteria, but no action was taken. Significance/Severity Level: The inspectors determined the performance deficiency was more than minor because it was associated with the Procedural Quality attribute of the Mitigating Systems Cornerstone and adversely impacted the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the failure to clearly identify the acceptance criteria, alert range and required action ranges resulted an in organizational failure to declare the pumps inoperable and to perform required analysis of the equipments condition. The inspectors assessed the significance of the finding using SDP Appendix A and concluded the issue was of very low safety significance (i.e., Green).Corrective Action References: (1) AR 04142617, Acceptance Criteria Not Clearly Listed in DO Pump Procedures, and (2) AR 04142370, DO Pump Test Packages are Not Routed to the IST Coordinator. | |
| 05000455/FIN-2018002-01 | 2018Q2 | Byron | Green | H.12 | Self-revealing | Overspeed Trip of 2B Auxiliary Feedwater Pump During Surveillance | A finding of very low safety significance and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, was self-revealed when the 2B diesel-driven auxiliary feedwater (AF) pump tripped on overspeed during a quarterly inservice test (IST). Specifically, operators with portable instrumentation used an erroneous speed value to adjust pump speed beyond the range specified in the procedure resulting in a pump overspeed trip, entry into a 72-hour technical specification (TS) required action statement, and unplanned pump unavailability with an associated change in Unit 2 risk from green to yellow. |
| 05000461/FIN-2018050-01 | 2018Q2 | Clinton | TBD | H.2 | Self-revealing | Failure to Follow Multiple Procedure | On May 17, 2018, a To-Be-Determined (TBD) finding and an associated Apparent Violation of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and Technical Specification 3.8.2, Condition B.3,were self-revealed for the licensees failure to follow multiple procedures that affected quality.This resulted in the unavailability and inoperability of the Division 2 Emergency Diesel Generator when it was relied upon for plant safety |
| 05000456/FIN-2018002-02 | 2018Q2 | Braidwood | Green | H.3 | Self-revealing | Work Instruction Error Results in Reactor Coolant System Pressure Transient | A self-revealed finding of very low safety significance (i.e., Green) was identified due to the licensees failure to follow work instructions while performing a digital upgrade to plant control systems. Specifically, while performing maintenance on the volume control tank (VCT) level transmitter on April 10, 2018, maintenance personnel failed to properly track the steps being performed while simultaneously working on multiple packages. This resulted in the Unit 1 reactor coolant system (RCS) experiencing a pressure transient and the actuation of a VCT relief valve. |
| 05000255/FIN-2018011-02 | 2018Q2 | Palisades | NRC identified | Failure to Set Action Levels to Ensure that the Assumptions in the Engineering Analysis Remain Valid | The inspectors reviewed a sample of equipment located in the fire areas selectedfor inspection to determine if the licensee had established a proper method of monitoring that equipment as required by NFPA 805, Section 2.6. Section 2.6 of NFPA 805 required that, A monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee utilized Procedure EN-DC-357, NFPA 805 Monitoring Program, Revision 2,to ensure that, the assumptions in the NFPA 805 engineering analyses remain valid by executing an effective and ongoing monitoring program.The inspectors selected the high pressure air compressor (C-6B) and high pressure safety injection pump (P-66B), both of which were located in the West Safeguards Room. The licensee considered these components to be high-safety significant (HSS) structures, systems, or components (SSCs). The licensee chose to monitor the unavailability of these components utilizing the Maintenance Rule (10 CFR 50.65).The licensee set the Maintenance Rule allowable unavailability action level threshold for the high pressure air compressorat 5E-2 (5percent)whereas they assumed in their fire PRA an unavailability of 9.86E-3 (approximately 1percent). For the high pressure safety injection pump the licensee set the Maintenance Rule allowable unavailability at 1.5E-2 (1.5percent) whereas they assumed in their fire PRA an unavailability of 6.32E-3 (approximately 0.6percent). The inspectors believed that by relying on the less conservative action level thresholds in the Maintenance Rule the licensee failed to ensure that the assumptions in the engineering analysis (fire PRA) remained valid.The licensee stated in Procedure EN-DC-357, Section 1.0, Purpose, that, The NFPA 805 Monitoring Program ensures that the assumptions in the NFPA 805 engineering analyses remain valid by executing aneffective and ongoing monitoring program. Under Section 3.0, Definitions, the licensee defined, Action Level Threshold, as, When establishing the action level threshold for reliability and availability, the action level should be no lower than the Fire Probabilistic Safety Analysis (also called fire PRA) assumptions. The licensee stated in Section 5.3.3(c) that, If HSS SSCs have been identified in using the Maintenance Rule guidelines, the associated SSC specific performance criteria may be established as in the Maintenance Rule, provided the criteria are consistent with the Fire Probabilistic safety Analysisassumptions... The inspectors believed that Procedure EN-DC-357 required the licensee set the action level thresholds no lower than the fire PRA assumptions. Procedure section 5.3.4(b)(1) required that HSS equipment that is not sufficiently tracked in the Maintenance Rule be added to the NFPA 805 Monitoring Database. The licensee did not add the high pressure air compressor and the high pressure safety injection pump into the NFPA 805 Monitoring Database. In the SER 2015-2-27 dated February 27, 2015, in which the staff approved the licensee NFPA805 License Amendment Request, the staff noted that the licensee will develop an NFPA 805 Monitoring Program consistent with Frequently Asked Question (FAQ)10-0059. The staff also noted that the stated development of the Monitoring Program would include a review of existing surveillance, inspection, testing, compensatory measures, and oversight 8processes for adequacy. The staff concluded in SER 2015-2-27 that since the final values for availability and reliability, as well as the performance criteria for the SSCs being monitored, have not been established for the Monitoring Program as of the date of this SER, completion of the licensee's NFPA 805 Monitoring Program is an implementation item. Furthermore, the staff concluded that there is reasonable assurance that the licensee will develop a Monitoring Program that meets the requirements specified in Sections 2.6.1, 2.6.2, and 2.6.3 of NFPA 805Section 2.6 of NFPA 805 stated in part that, Monitoring shall ensure that the assumptions in the engineering analysis remain valid. The licensee interpreted this statement to mean that utilizing the existing Maintenance Rule unavailability values is consistent with its commitment in SER 2015-2-27 and would allow the site to appropriately monitor the availability and reliability of fire protection systems and features. The licensee also performed sensitivity studies on the differences in the unavailability values of fire protection systems and features between the Maintenance Rule criteria and the fire PRA values and determined that they were not risk-significant. The inspectors questioned the appropriateness of the licensees interpretation of assumptions as described in Section 2.6 of NFPA 805 above. The inspectors believed that the licensee should monitor the unavailability of fire protection systems and features utilizing the same values as thosedocumented in the fire PRA associated with the NFPA 805 License Amendment Request. The licensee further stated that they were waiting for guidance from the NRCs Office of Nuclear Reactor Regulation and the industry who were working on revising guidance in FAQ10-0059, NFPA 805 Monitoring, to determine if they needed to change their approach. That guidance document was in the process of being revised during the inspection. The inspectors needed to determine if the licensees approach to monitoring the availability and reliability of the fire protection systems and features using the Maintenance Rule monitoring values in order to ensure that the assumptions in the engineering analysis remained valid was an acceptable approach.Planned Closure Action(s): The inspectors will await clarification from the Office of Nuclear Reactor Regulation in order to determine if a performance deficiency exists.Licensee Action(s): The licensee plans to follow the resolution of FAQ 10-0059, Revision 6, and take the appropriate corrective actions based on the guidance provided in that FAQ. | ||
| 05000255/FIN-2018011-01 | 2018Q2 | Palisades | Green | NRC identified | Failure to Maintain Adequate Fire Protection System Functional Test Procedure | The inspectors identified a finding of very-low safety significance and associated violation of Technical Specification 5.4.1, Procedures,for the licensees failure to maintain fire protection system functional test procedure. Specifically, the licensee failed to maintain Procedure RO-52, Fire Suppression Water System Functional Test and Fire Pump Capacity Test, by failing to include appropriate acceptance criteria in the procedure to demonstrate fire protection system functionality. | |
| 05000457/FIN-2018002-01 | 2018Q2 | Braidwood | Green | H.9 | Self-revealing | Inadequate Detail in Maintenance Work Instructions Resulted in Failed Gearbox Oil Cooler Head Gasket and Inoperable 2B Auxiliary Feedwater Pump | A self-revealed finding of very low safety significance (i.e., Green) and an associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, was identified for the licensees failure to have adequate detail within their maintenance work instructions to enable proper reassembly of the 2B auxiliary feedwater (AF)pump gearbox oil cooler. Specifically, during the licensees 19th Unit 2 refueling outage in April 2017, the gearbox oil cooler closure head was reassembled following scheduled maintenance using an excessive amount of room temperature vulcanizing silicone (RTV) on the joint and an insufficient amount of torque on the closure head bolting. As a result, on March 16, 2018, the closure head joint failed causing several hours of unplanned inoperability and unavailability for the 2B AFPump. |
| 05000461/FIN-2018050-02 | 2018Q2 | Clinton | Green | H.12 | Self-revealing | Failure to Promptly Identifya Condition Adverse to Quality | On May 17, 2018,a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, were self-revealed for the licensees failure to promptly identify that the safety-related Division 2 EDG had its starting air receivers isolated, which was a condition adverse to quality that rendered the EDG inoperable and unavailable |
| 05000461/FIN-2018050-03 | 2018Q2 | Clinton | Green | H.6 | Self-revealing | Equipment Operator Rounds Points Inadequate Acceptance Criteria | On May 17, 2018,a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, were self-revealed for the licensees failure to include appropriate quantitative acceptance criteria for the Division 2 EDG parameters to ensure the Division 2 EDG could perform its safety function |
| 05000461/FIN-2018002-03 | 2018Q2 | Clinton | Severity level Minor | NRC identified | Minor Violation | During the inspection quarter, the inspectors reviewed a significant number of licensee CAP documents to assess the following performance attributes: complete, accurate, and timely documentation of the identified problem in the CAP; evaluation and timely disposition of operability and reportability issues; consideration of extent of condition and cause, generic implications, common cause, and previous occurrences; classification and prioritization of the problems resolution commensurate with the safety significance; and identification of negative trends associated with human or equipment performance that can potentially impact nuclear safety. Minor Performance Deficiency: The inspectors determined that issues which could impact the operability of TS-related equipment were generally entered into the CAP in a timely manner. However, operability determinations were not always performed within the timeframes established in Section 4.1 of Procedure OPAA108115, Operability Determinations (CM1), because some issue reports were not directly routed to the operating shift crew for review. The CAP software program used by the licensee included a standard set of questions which were normally answered by the individual entering the issue into the CAP. Depending on the answers to the questions, the CAP document routing could automatically bypass the operating shift crew for review. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The inspectors did not identify any instance where the failure to perform a timely operability determination had a significant consequence on licensed activities. However, the inspectors discussed the vulnerability between the CAP and the operability determination process with the licensee. The licensee implemented a standing order to require a shift review by the operating crew of condition reports not directly routed to the shift. In addition, the licensee is trending the number of condition reports which are returned by the Station Ownership Committee to the shift for review to determine whether further actions are warranted. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard. | |
| 05000461/FIN-2018002-04 | 2018Q2 | Clinton | Severity level Minor | NRC identified | Minor Violation | The inspectors reviewed AR 4082490, Reactor SCRAM from Trip of 1AP07EJ. The inspectors selected this sample for review due to the safety significance of the Division 1 and 2 safety-related transformers, which is the subject of the AR. This review focused on actions associated with newly installed Divisions 1 and 2 4160V to 480V transformers. As appropriate, the inspectors verified the following attributes during their review of the licensee's corrective actions for the above condition report and other related condition reports: classification and prioritization of the resolution of the problem commensurate with safety significance; and completion of corrective actions in a timely manner commensurate with the safety significance of the issue. The inspectors discussed the corrective actions and associated evaluations with licensee personnel. As a result of this review the inspectors identified the following minor violation: Minor Violation: The inspectors identified a violation of 10 CFR 50, Appendix B, Criterion II, Quality Assurance Program, for the failure to follow procedures associated with the CAP. Specifically, on May 10, 2018, the licensee identified discrepant results while testing safety-related transformers 0AP06E2 and 1AP12E2 but failed to enter this issue into the CAP in accordance with PIAA120, Issue Identification and Screening Process, Revision 8, Step 4.3.4, until prompted by the inspectors. Instead, the licensee evaluated the discrepant results within the work order and found them to be acceptable. The licensee generated AR 4137994, Insulation Power Factor Results For 0AP06E & 1AP12E, dated May 15, 2018, after being challenged by the inspectors regarding the need to enter the discrepant test results into the CAP. Screening: This issue screened as minor because all the questions associated with a minor issue found in IMC 0612, Appendix B, were answered No. The failure to document the discrepant values in the CAP did not adversely impact the safety-related transformers. Enforcement: This failure to comply with 10 CFR 50, Appendix B, Criterion II, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. Enforcement: The inspectors did not identify a violation of regulatory requirements associated with this minor finding because the procedure the licensee failed to follow was a self-imposed standard. | |
| 05000461/FIN-2018002-02 | 2018Q2 | Clinton | Green | NRC identified | Failure to Establish Adequate Leak Rate Test Procedures for Shutdown Service Water Isolation Valve Testing | The inspectors identified a Green finding and a Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to ensure testing of the shutdown service water (SX) isolation valves was performed with procedures which: (1) incorporated the requirements and acceptance limits contained in applicable design documents; and (2) included provisions for assuring that all prerequisites for the given test had been met. Specifically, the licensee failed to establish leak rate test procedures for SX boundary valves 1CC075A and 1CC076A that included provisions for ensuring the required differential test pressure was met during testing. | |
| 05000461/FIN-2018002-01 | 2018Q2 | Clinton | Green | H.6 | NRC identified | Failure to Perform an Operability Determination for Suspected Leakage Past Shutdown Service Water Isolation Valves | The inspectors identified a Green finding for the failure to perform an operability determination in accordance with Procedure OPAA108115, Operability Determinations (CM1). Specifically, the licensee failed to determine and document the operability status of the shutdown service water system and the ultimate heat sink after the discovery of leakage past the 1CC075A and 1CC076A isolation valves. |
| 05000331/FIN-2018002-01 | 2018Q2 | Duane Arnold | Green | H.1 | NRC identified | Inappropriate Procedural Guidance Resulted in Loss of Scram Function and Failure to Enter Technical Specification Limiting Condition for Operation | The inspectors identified a finding of very low safety significance (Green) and a non-cited violation of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to have procedures appropriate to the circumstance for testing the main steam isolation valve (MSIV) and turbine stop valve (TSV) closure functions. Specifically, STP 3.3.1.117, MSIV Functional Test, and STP 3.3.1.119, Main Turbine Stop and Combined Intermediate Valves Test, directed the use of a reactor protection system test box which disabled the MSIV and the TSV closure automatic reactor scram functions while testing specific combinations of MSIVs and TSVs and failed to require entry into appropriate Technical Specification Limiting Condition for Operation action statements. |
| 05000331/FIN-2018011-01 | 2018Q2 | Duane Arnold | Green | NRC identified | Failure to Translate Environmental Qualification Requirements into Maintenance Procedures/Instructions | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited Violation (NCV) of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to establish measures to assure that Environmental Qualification (EQ) requirements for qualified components correctly translated into procedures and instructions. Specifically, the inspectors identified two examples of the licensees failure to ensure that the EQ requirements for O-ring installed in EC290 connector/plug-in cable assemblies were translated into the associated maintenance procedures and instructions(i.e.,EQ Files, warehouses storage requirements). The licensee failed to correctly establish an end-of-life replacement schedule for the O-ring used in the cable assemblies installed in the dry well and failed to establish a 2-year shelf-life for the O-ring stored in the warehouse. | |
| 05000331/FIN-2018002-02 | 2018Q2 | Duane Arnold | Severity level Minor | NRC identified | Minor Violation | Minor Violation: On June 19, 2016, while operating at 82 percent power, two secondary containment access airlock doors were opened simultaneously during surveillance testing as part of STP 3.6.4.102, Secondary Containment Airlock Verification. The inspectors determined this event was caused by inadequate procedural guidance which directed the user to attempt to open one airlock door while the other door was already open. During this test, the interlock failed because the permanent magnets had rotated and were misaligned. This failure could have been identified without challenging airlock interlock integrity if the second airlock door wasnt held open. The failure to have adequate procedural guidance for testing the secondary containment airlock doors was a violation of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures and Drawings, which requires licensees to have procedures appropriate to the circumstance when performing safety-related activities. In response to this issue, the licensee immediately closed the airlock doors. In addition, the licensee submitted a TS change request to address the concurrent opening of two secondary containment airlock doors. The licensees corrective action program is tracking the TS change as CR 02034076, Secondary Containment Airlock Doors #225 and 228 Both Opened. Screening: The issue screened as minor because all of the questions associated with a minor issue found in IMC 0612, Appendix B were answered No due to the licensee reestablishing secondary containment operability immediately after the second airlock door opened. In addition, the inspectors considered the failure to have an appropriate procedure was less than a Severity Level IV violation in accordance with the NRCs Enforcement Policy. Violation: The failure to comply with 10 CFR Part 50, Appendix B, Criterion V, constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes LER 05000331/2016001. | |
| 05000282/FIN-2018011-01 | 2018Q2 | Prairie Island | Green | H.6 | NRC identified | Failure to Justify Load Combinations Used in Main Steam Piping Stress Analysis | Inspectors identified a Green finding and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations, Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to correctly translate provisions from specified quality standards for load combinations into piping analysis. Specifically, in the analysis for the Class I main steam piping, the licensee combined the seismic Operating Basis Earthquake and safety relief valve operating loads by Square Root of Sum of Squares. Prairie Island Updated Safety Analysis Report and the Engineering Manual for piping system stress analysis do not permit the Square Root of Sum of Squares method for combining these loads. |
| 05000282/FIN-2018011-02 | 2018Q2 | Prairie Island | NRC identified | Potential Failure to Protect Class I Structures, Systems,and Components from Tornado Generated Missiles | Inspectors identified a number of structure, systems,and components (SSCs) that lacked protection from tornado generated missiles. The following SSCs were identified: Division 1 and Division 2 Emergency Diesel Generators (D1/D2 EDGs)engine exhaust, fuel oil day tank vents, and main fuel oil storage tanks vents; and Diesel Driven Cooling Water Pumps (DDCWPs) main fuel storage tank vents, day tank vents, engine exhausts, and rooms ventilation intake and exhaust equipment. In various cases susceptible SSCs for redundant equipment (e.g. fuel tank vents) were right next to or within a few feet of each other such that a single missle could affect both trains of the system A review of the sites licensing bases, including the original FSAR, identified the D1/D2 EDGs and the DDCWPs as Class I, safety-related SSCs, which are required to be designed to withstand, without loss of capability, environmental phenomena including tornadoes and tornado generated missiles. Specifically, the current USAR Table 12.2-1, Classification Of Structures, Systems and Components, list both systems as Class I and has two notes of interest. Note 1 applies to the Diesel Generators and their associated (Main) Fuel Oil Storage Tank, which states, in part, The indicated Design Class I is applicable to D1/D2 Diesel Generators and associated(emphasis added) safety related components and systems. The second note is listed at the beginning of the Table, which states,in part,To determine detail design classifications and boundaries separating different design classes within the overall classification scheme listed here, refer to controlled drawings. A review of controlled drawings, including NF-39255-1, Flow Diagram Diesel Generators D1 & D2 Unit 1 & 2,Revision 85, and NF-39232, Flow Diagram Fuel & Diesel System Unit 1 & 2, Revision 86,showed the fuel oil vents for the main storage tanks, fuel oil vents for the day tanks,engine exhaust piping,mufflers, and silencers for the D1/D2 EDGs and DDCWPs were classified as safety-related Class I SSCs. A review of the current UFSAR identified the following sections of interest:The USAR Section 1.5.I, Overall Plant Requirements, Criterion 2 -Performance Standards, Answer, established in part The system and components designated Class I in Section 12, in conjunction with administrative controls and analysis, as applicable, are designed to withstand, without loss of capability to protect the public, the most severe environmental phenomena ever experienced at the site with appropriate margins included in the design for uncertainties in historical dataThe USAR Section 12.2.1.1.a, Classification of Structures and Components, defines Design Class I as Those structures and components including instruments and controls whose failure might cause or increase the severity of a loss-of-coolant accident or result in an uncontrolled release of substantial amounts of radioactivity, and those structures and components vital to safe shutdown and isolation of the reactor.The USAR Section 12.2.5.1.g.1, Protection for Class I Items, establishes, in part, that Class I items are protected against damage from: Missiles from different sources.These sources comprise: Tornado created missiles.The USAR Section 12.2.1.3.2.c., Tornado Loads, defines the design tornado driven missile as assumed equivalent to an airborne 4 x 12 x 120 plank travelling end-on at 300 mph, or a 4000 lbs automobile flying through the air at 50 mph and at not more than 25 feet above ground level.Based on the above, the inspectors were concerned the susceptible SSCs could lose the capability to perform their safety-related function if they were impacted by tornado generated missiles. For example, an impact to the fuel oil vents could crimp the vent path resulting in a vacuum inside the tanks that could collapse the tank and/or cause the associated fuel transfer pump to lose net positive suction head The licensee provided a position paper proposing the susceptible SSCs identified by the inspectors were meeting their current licensing bases and no further actions were required. The inspectors disagreed, but decided to request support from the Office of Nuclear Reactor Regulation (NRR) to obtain clarification on the sites licensing bases related to tornado generated missiles. Planned Closure Action: The inspectors have requested NRR to provide clarification on the sites current licensing bases regarding tornado generated missiles required protection.Licensee Action: Licensee is considering doing a self-review of design and licensing basis of the fuel oil storage tank vent lines to understand and clarify design class of the lines Corrective Action Reference:501000012997 | ||
| 05000282/FIN-2018002-01 | 2018Q2 | Prairie Island | NRC identified | Results of ISFSICask Array Dose Calculation Not Incorporated into FSAR | Prairie Island ISFSI FSAR, as updated, Revision 18, Section A7A.7 evaluates off-site dose rates for an array of ISFSI casks. In this dose rate calculation, explicit modeling credit is given to the earthen berm that surrounds the Prairie Island ISFSI as discussed in Section A7A.7.1. The earthen berm provides radiation shielding for the ISFSI. This calculation allows the licensee to demonstrate, in part, compliance with Title 10 of the Code of Federal Regulations (CFR) 72.104(a) which requires, in part, that, During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ. Calculation TN40HT0502, TN40HT Far Field Shielding Calculations, Revision 0, was performed by the licensee in support of a License Amendment Request (LAR) to modify the Prairie Island ISFSI TN40 cask design (designated as TN40HT casks). The TN40HT LAR was submitted to the NRC by the licensee on March 28, 2008. This dose rate calculation does not credit the earthen berm and, in part, also allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a). The licensee also provided this calculation directly to the NRC in a February 29, 2012, letter in response to a Request for Supplemental Information (RSI) from the NRC associated with the license renewal application for the Prairie Island ISFSI. Although the results from calculation TN40HT0502 for a single cask was incorporated into the Prairie Island ISFSI FSAR, Revision 18, in Tables A7A.22 and A7A.61, the results from TN40HT0502 for an array of casks which, in part, allows the licensee to demonstrate, in part, compliance with 10 CFR 72.104(a), has not been incorporated into the ISFSI FSAR, Revision 18.Title 10 CFR 72.70, Safety analysis report updating requires, in part, that (a) Each specific licensee for an ISFSI shall update periodically, as provided in paragraphs (b) and (c) of this section, the FSAR to assure that the information included in the report contains the latest information developed (b) Each update shall contain all the changes necessary to reflect information and analyses submitted to the Commission by the licensee or prepared by the licensee pursuant to Commission requirement since the submission of the original FSAR or, as appropriate, the last update to the FSAR under this section. The update shall include the effects of: (2) All safety analyses and evaluations performed by the licensee in support of approved license amendments.This Unresolved Item is being opened to determine whether or not the licensee is required to update the ISFSI FSAR with the results of calculation TN40HT0502 for an array of casks in accordance with 10 CFR 72.70.Planned Closure Action: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Material Safety and Safeguards to determine whether or not calculation TN40HT0502 is subject to the FSAR updating requirements of 10 CFR 72.70 for the Prairie Island ISFSI. | ||
| 05000331/FIN-2018001-01 | 2018Q1 | Duane Arnold | Self-revealing | Failure to Perform Nondestructive Examination of Main Steam Isolation Valve 4415 Following Machining of Valve Bore | On February 28, 2018, while the licensee was planning contingency work order packages for possible MSIV repairs during the upcoming refuel outage (RFO) 26, a self-revealing failure to perform an NDE of MSIV CV 4415 during RFO 25 was identified. During RFO 25, inboard MSIV CV 4415 failed local leak rate testing. The valve was disassembled; the valve bore was machined, and the valve was reassembled. Subsequently, the valve passed local leak rate testing. However, the licensee failed to perform post-machining NDE of the valve bore as required by plant design. The MSIV was purchased and installed in accordance with USAS B 31.1.0. The General Requirements of USAS B31.1.0, Part 107, Valves, states that (a) Valves complying with the standards and specifications listed in Table 126.1 may be used within the specified pressure-temperature ratings, (b) Valves not complying with Paragraph (a) above shall be of a design, or equal to the design, which the manufacturer recommends for the service as stipulated in Paragraph 102.2.2. Purchase Specification General Electric Spec. No. 21A9230, Revision 2, requires in Part 6.2.4, Castings for pressure-containing parts shall be 100 percent examined by radiography and all accessible surfaces, including machined surfaces, shall be examined by either liquid penetrant or magnetic particle methods following heat treatment. Re-examination of repaired areas shall be by the above techniques following heat treatment. Contrary to the above, the licensee failed to perform either liquid penetrant or magnetic particle testing following machining of the valve bore. The inspectors determined that the failure to perform design required NDE was a performance deficiency. However, the inspectors cannot assess whether the performance deficiency is more than minor until they review the NDE results. Planned Closure Actions: The inspectors will review the results of the NDE after it is performed during RFO 26 and determine the significance of the violation. Licensee Action: The licensee will perform an NDE of the MSIV bore during RFO 26. Corrective Action Reference: CR 02251009; Missed NDE During Rebuild Of MSIV CV4415. On January 10, 2018, the inspectors evaluated the licensees response to a B Channel half SCRAM condition. The half SCRAM was caused by the failure of Average Power Range Monitor (APRM) B. The inspectors determined the operators appropriately implemented pertinent procedures and Technical Specification requirements which allowed the operators to bypass APRM B and reset the half SCRAM. The licensee performed simple troubleshooting and determined the power supply for APRM B had failed. The power supply had recently been replaced. The inspectors determined the licensee had replaced the power supply in accordance with NextEra procedures and that the APRM power supply failure was not within the licensees ability to foresee and prevent. The licensee entered this minor issue into their corrective action program as CR 02243984; E/S 265 No Voltage Output. The inspectors discussed the corrective actions and associated evaluations with licensee personnel. | ||
| 05000331/FIN-2018010-01 | 2018Q1 | Duane Arnold | Green | NRC identified | Failure to have Adequate Pre-Fire Plans | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited Violation of Title 10 of the Code of Federal Regulations, Part50.48(c), and National Fire Protection Association (NFPA)805, Section 3.4.2,Pre-Fire Plans. Specifically, the inspectors identified two examples for the licensees failure to have current and detailed pre-fire plans. The first example for the failure to provide adequate guidance in the pre-fire plans for smoke and heat removal in the event of a fire in switchgear rooms. The second example was for the failure to show the addition of the Flexible Coping Strategiesbattery packsas a potential hazard in the pre-fire plan for the battery roomcorridor. | |
| 05000331/FIN-2018010-02 | 2018Q1 | Duane Arnold | Green | NRC identified | Failure to Include Operator Action in the Plant Operating Procedure | The inspectors identified a finding of very-low safety significance (Green), and associated Non-Cited ViolationofTitle10 of the Code of Federal Regulations, Part 50.48(c), and NFPA805, Section 4.2.4.1.6,Operations Guidance. Specifically, during the transition process to NFPA 805, performance-based standard fortheFire Protection Program, the licensee inadvertently removed a required operator action in the control room from plant operating procedure AOP 913, Fire. | |
| 05000331/FIN-2018411-01 | 2018Q1 | Duane Arnold | Green | Licensee-identified | Licensee-Identified Violation |