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| Q | Significance | CCA | Identified by | Title | Description | |
|---|---|---|---|---|---|---|
| 05000316/FIN-2018003-01 | 2018Q3 | Green | H.5 | Self-revealing | Misaligned Heater Level Column Valves Leads to Manual Reactor Trip | A self-revealed, Green finding was identified when the operators manually tripped the Unit 2 reactor in response to a hi-hi level in the Left Moisture Separator Drain Tank. On May 6, 2018, the Unit 2 reactor was at approximately 12 percent power following a startup at the conclusion of the spring 2018 refueling outage. While the station continued to make preparations to start the main turbine and synchronize with the grid, the moisture separator drain tank hi level alarm was received and remained standing for the better part of the shift. The drain tank collects condensed steam and water from the moisture separator reheater and associated high pressure turbine exhaust lines and routes it either to the condenser or #4 feedwater heaters. The day shift operators were hesitant to continue on with starting the main turbine until the cause of the alarm could be determined. Due to a series of miscommunications between day shift, night shift, the outage control center, and personnel performing troubleshooting, the night shift crew believed it was acceptable to continue with the turbine startup with the alarm still standing. The turbine was synchronized to the grid and power was stabilized at approximately 29 percent power with the alarm in for most of the turbine startup and synchronization. The alarm cleared for a period of time at 29 percent power, but then returned along with the hi-hi drain tank level alarm. Per the alarm response procedures, the operators tripped the reactor and main turbine to protect the turbine from excessive water in the system. Later investigation by the site revealed that the level columns for the #4 feedwater heaters had been left isolated following work and testing associated with the replacement of the #5 feedwater heaters. While the Operations Department had completed a valve lineup on the system per their startup procedures, which put the level columns in service, the Projects Department had not finished all of the work on the heaters at the time the lineup was performed. As a result, workers subsequently isolated the columns to complete testing after the Operations lineup was complete. A step in the Projects test procedure EC51366TP001 directed workers to specifically inform the operators that the level columns were isolated following testing and that the system was ready to be lined up per operations procedures. However, the workers did not provide that detail, and simply stated that the test was complete. As a result, operations did not know the valves had been taken out of alignment. Contributing to the issue, the outage schedule did not provide any logic ties to ensure all work was complete on the heaters before allowing operations to do their valve lineups. With the level columns isolated during startup, the #4 heaters indicated an erroneous level. This resulted in the operators believing that the heaters were at a normal operating level when in fact, they were full. Therefore, when the operators (per procedure) opened a high pressure turbine exhaust valve to the 4A heater, this created a pathway for water to flow from the #4 heaters, through the high pressure turbine exhaust lines, and into the moisture separator drain tank. The excessive flow of water caused the hi and hi-hi alarms in the drain tank which then led to the reactor/turbine trip. |
| 05000315/FIN-2018010-01 | 2018Q3 | NRC identified | Record Retention Requirements of the Boron Injection Tank and its Associated Support Structure | The inspectors identified an Unresolved Item concerning the Title 10 of the Code of Federal Regulations, Part 50, Appendix B, and ASME Code requirements for the BIT and its associated support structure calculation of record. Updated Final Safety Analysis Report (UFSAR) Section 2.9.2 delineated the BIT Seismic Classification as Class 1. The BIT was part of the Emergency Core Cooling System piping system, and is Seismic Class I. In addition, UFSAR Table 6.2-1 and UFSAR Table 6.2-3 delineated the BIT was designed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Class C. Additionally, Subsection C under Section IIII Article N-2111, stated, in part, The requirements of Section VIII of the Code shall apply to the materials, design, fabrication, inspection and testing, and certification of Class C vessels.... The inspectors reviewed Drawing No. 113E275; 900 Gallon BIT; Revision 5 which contained the design specification for the BIT. Also the inspectors reviewed Struthers Wells Calculation No. 2-70-07-30717; Seismic Stress Calculations for BITs; 07/02/1970 which contained the BIT support structure qualification. The inspectors reviewed Calculation No. DC-D-12-MSC-8 Attachment A, page A.10-10 and page A.9-28; Revision 2 which contained the applied nozzle loads at the BIT inlet and outlet nozzles. Lastly, the inspectors reviewed Document No. 546 CRI 109890; Westinghouse Purchase Order for BIT; 06/22/1970 which contained design requirements for the BIT. During the review of aforementioned design basis documents the inspectors identified the following examples in which the licensee did not have a calculation of record to address the following ASME code requirements: ASME Section VIII, Division 1, Subsection A, General Requirements, Part UG-22 titled Loading states, in part, the loadings to be considered in designing a vessel shall include: Internal or external design pressure (as defined in Par. UG-21), Impact loads, including rapidly fluctuating pressures: Weight of the vessel and normal contents under operating or test conditions. (This includes additional pressure due to static head of liquids), Superimposed loads such as other vessels, operating equipment, insulation, corrosion-resistant or erosion-resistant linings and piping, Wind loads, and earthquake loads where required, Reactions of supporting lugs, rings, saddles or other types of supports (see Appendices D and G) and the effects of temperature gradients on maximum stress. The inspectors identified that the licensee did not have a calculation of record to address the applied loadings due to dead weight of the vessel, fluid weight inside of the vessel, design temperature of 300 degrees Fahrenheit and earthquakes (Operating Basis Earthquake and Safe Shutdown Earthquake) on the BIT vessel shell and head ASME Section VIII, Division 1, Subsection A, General Requirements, Part UG-54 titled Supports states, in part, All Vessels shall be supported and the supporting members shall be arranged and/or attached to the vessel in such a way as to provide for the maximum imposed loadings (see Par. UG-22).. The inspectors identified that the licensee did not have a calculation of record to address the applied loadings due to the superimposed piping loads at the BIT inlet and outlet nozzle to the BIT support structure as well as the applied loading due to the design temperature of 300 degrees Fahrenheit. Secondly, the inspectors identified that no calculation of record existed for the welded connection between the support legs and the baseplate. Thirdly, no calculation of record existed for the welded connection between the support legs and the BIT. Lastly, the self-weight and self-weight seismic excitation of the support structure was not considered in the applied stresses of the support structure calculation of record. In response to the inspectors concern, the licensee initiated AR 2018-7104, Lack in Documentation for BIT 1-TK-11, 07/12/2018. In addition, the licensee performed an operability review and reasonably determined the BIT remained operable. Near the end of the inspection period, the licensee provided the inspectors additional information relevant to the calculation record retention requirements as defined by the ASME Code and the DC COOK Quality Assurance Program Document which will require additional review to determine whether a violation exists. Therefore, this issue is considered an unresolved item pending completion of inspector review and evaluation and discussion with the Office of Nuclear Reactor Regulation and Office of the General Counsel. | ||
| 05000315/FIN-2018003-02 | 2018Q3 | Licensee-identified | Site Specific Shielding and Barriers for HI-TRAC Transfer Cask Require NRC Approval Prior to Use | Certificate of Compliance (CoC) 1014, Amendment 9, Design Feature, Section 3.9, Environmental Temperature Requirements, requires building ambient temperatures be less than 110 degrees Fahrenheit during canister processing based upon the thermal analysis in the Holtec HI-STORM Final Safety Analysis Report, Revision 13. The thermal model documented in the Final Safety Analysis Report, Revision 13, Section 4.5.1, HI-TRAC Thermal Model, states that heat is passively rejected to the ambient from the outer surface of the HI-TRAC transfer cask by natural convection and thermal radiation. However, at D.C. Cook, the licensee uses additional shielding materials for as low as reasonably achievable purposes that are in contact with and in the general area of the HI-TRAC. The licensee requested Holtec to perform a site-specific thermal analysis, HI2177676, Thermal Evaluation of Shielding Package around the HI-TRAC at DC Cook, to include the shielding material in the thermal model. The analysis contained inputs that were different than the design basis calculation inputs, which were previously incorporated into Design Feature Section 3.9 and Approved Contents Section 2.4. The licensee performed a 10 CFR 72.48 Screening and Evaluation 2018013902, which concluded that shielding could be used without prior NRC approval and subsequently issued 212CR0017, which revised the 72.212 Report. The licensee implemented administrative controls on building temperature and fuel assembly heat load limits based upon the site specific thermal analysis. This unresolved item is being opened to determine if: A) the licensee is in compliance with Design Feature, Section 3.9, Environmental Temperature Requirements; B) the Design Feature Section 3.9 and Approved Contents Section 2.4 are non-conservative at D.C. Cook; and C) the licensee is in compliance with 10 CFR 72.48. Planned Closure Actions: Region III will coordinate with the Division of Spent Fuel Management in the NRC Office of Nuclear Materials Safety and Safeguards. Corrective Action References: AR 20184056; AR 20186342; AR 20186642 | ||
| 05000315/FIN-2018002-06 | 2018Q2 | 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 | 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 | 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 | 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 | |
| 05000315/FIN-2018002-02 | 2018Q2 | 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 | 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. |
| 05000315/FIN-2018001-02 | 2018Q1 | Green | NRC identified | Operation of Letdown System Leads to Voiding and Subsequent Relief Valve Lift | The inspectors identified a finding of very low safety significanceand associated Non-Cited Violation of Technical Specification 5.4, Procedures, when the licensee failed to maintain a procedure for operating the letdown system. As a result, a water-hammer occurred which caused a safety-related relief valve to lift, which discharged reactor coolant to the Pressurizer Relief Tank until letdown was isolated | |
| 05000315/FIN-2018001-01 | 2018Q1 | Green | Self-revealing | Failure of Unit 1 Turbine Driven Auxiliary Feedwater Pump to Reach Rated Speed | A self-revealed finding of very low safety significance with an associated Non-Cited Violation of Technical Specification 5.4 Procedures, occurred on December 21, 2017, when the Unit 1 Turbine-Driven Auxiliary Feedwater Pump failed to reach rated speed during a surveillance. Procedure 12MHP5021056008, Turbine-Driven Auxiliary Feedwater Pump Governor Valve Maintenance, was not appropriate for the circumstances in that direction was not given to check that the governor valve could fully open following maintenance on the governor valve. | |
| 05000315/FIN-2017004-04 | 2017Q4 | Green | H.14 | NRC identified | Failure to Verify the Adequacy of the Design for a Temporary Modification | A finding and associated violation of 10 CFR 50 Appendix B Criterion III self-revealed when licensee personnel could not obtain a water sample from a location designated as a connection point for a safety related temporary modification. Specifically, the licensee developed a temporary modification to add water to CCW but failed to verify the adequacy of the design in that the licensee did not validate the connection point could supply sufficient water as a source for CCW make-up. As an immediate action the licensee reestablished flow through the valves. The inspectors determined that the licensees failure to verify the adequacy of the design for the temporary modification was more than minor because it was associated with equipment performance attribute of Mitigating Systems cornerstone and adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The finding was of very low safety significance (Green) because the finding affected the qualification of CCW but did not render it inoperable. In this case, CCW remained operable based on credit taken for isolation valve capability. The finding includes a cross-cutting aspect in the human performance area of H.14, Conservative Bias. |
| 05000315/FIN-2017004-01 | 2017Q4 | Green | NRC identified | Failure to Correct Numerous Anchor Darling Double Disc Gate Valve Non-Conformances | The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action for the licensees failure to correct a design non-conformance reported to the licensee through two related 10 CFR Part 21 reports. In March 2013, the licensee identified that 28 safety-related Anchor Darling double disc gate valves (ADDDGVs) may not have been assembled with an assumed amount of valve stem to wedge pre-torque before the stem was pinned into the wedge. The licensee had restored compliance to only one of these valves and had no plans to restore quality to the remaining 27 valves prior to the inspection. The licensee entered the inspectors conclusions into their corrective action program (CAP) as AR 201710399. At the end of this inspection the licensees plan was to restore compliance by either correcting the Part 21 issue or changing the design to accept the stem not having any pre-torque into the wedge.The performance deficiency was determined to be more than minor because if left uncorrected could become a more significant safety concern. Specifically, the failure to correct the design deficiencies could result in the valve pin breaking and consequential valve damage if the valves were operated at a high enough torque and/or thrust value(s). 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 determination which concluded that all 28 valve wedge pins had not sheared based upon the known historic operational history, pin material properties, and for using stem to wedge thread friction in some cases. The inspectors determined that this finding was not indicative of recent performance and therefore did not have a cross-cutting aspect assigned. | |
| 05000315/FIN-2017004-02 | 2017Q4 | NRC identified | Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown | Refueling Outage Activities a. Inspection Scope The inspectors reviewed the Outage Safety Plan (OSP) and contingency plans for the Unit 1 refueling outage (RFO), conducted September 13 through November 26, to confirm that the licensee had appropriately considered risk, industry experience, and previous site-specific problems in developing and implementing a plan that assured maintenance of defense-in-depth. During the RFO, the inspectors observed portions of the shutdown and cooldown processes and monitored licensee controls over the outage activities listed below: licensee configuration management, including maintenance of defense-in-depth commensurate with the OSP for key safety functions and compliance with the applicable TS when taking equipment out of service; implementation of clearance activities and confirmation that tags were properly hung and equipment appropriately configured to safely support the work or testing; installation and configuration of reactor coolant pressure, level, and temperature instruments to provide accurate indication, accounting for instrument error; controls over the status and configuration of electrical systems to ensure that TS and OSP requirements were met, and controls over switchyard activities; monitoring of decay heat removal processes, systems, and components; controls to ensure that outage work was not impacting the ability of the operators to operate the spent fuel pool cooling system; reactor water inventory controls including flow paths, configurations, and alternative means for inventory addition, and controls to prevent inventory loss; controls over activities that could affect reactivity; maintenance of secondary containment as required by TS; licensee fatigue management, as required by 10 CFR 26, Subpart I; refueling activities, including fuel handling and reactor assembly/disassembly; startup and ascension to full power operation, tracking of startup prerequisites, walkdown of the containment to verify that debris had not been left which could block emergency core cooling system suction strainers, and reactor physics testing; and licensee identification and resolution of problems related to RFO activities. Documents reviewed are listed in the Attachment to this report. Inspections activities performed in the third quarter coupled with those in the fourth quarter constituted one RFO sample as defined in IP 71111.2005. b. Findings (Opened) Unresolved Item 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown Introduction: Shortly after the shutdown for the Unit 1 refueling outage in September 2017, the licensee was establishing conditions in the charging and letdown system for the upcoming cooldown. After lowering letdown flow and attempting to adjust pressure, a letdown safety valve lifted and failed to completely reseat. Review of plant parameters following the event revealed that the evolution created saturation conditions in the letdown system. Subsequently, the steam bubbles collapsed causing a water hammer that lifted and damaged a relief in the system. The event was discussed in Section 4OA3 of Inspection Report 05000315/05000316/2017003. Description: The inspectors reviewed the licensees follow up of the issue in the CAP and spoke to personnel in the operations and maintenance departments. The licensee identified potential issues in the areas of procedure adequacy, operator performance, and equipment performance. However, the inspectors could not reconcile information on plant conditions with licensees statements regarding the cause. Because of ambiguity regarding the cause, the inspectors could not determine whether the corrective actions taken by the licensee were adequate. The licensee determined that an apparent cause evaluation need not be done therefore the inspectors reviewed available data, including plant computer data and a prior event from 2004. Since it is unclear what, if any, performance deficiency exists associated with this issue, the inspectors determined an unresolved item (URI) was necessary pending further follow up of the issue.Following the lifting of the safety valve, the licensee isolated letdown to stop the remaining leakage through the valve. The licensee then cycled the valve sufficiently enough for it to reseat so letdown could be restored and the cooldown continued. The safety valve was later discovered to be damaged from the event, so it was also repaired. Walkdowns were also conducted of the letdown piping to ensure no damage had occurred during the pressure transient. As part of their corrective actions, the licensee made some changes to the letdown procedure, recalibrated a letdown flow control valve, and developed actions to cover the event and lessons-learned in training. However, as stated above, the inspectors were unable to determine if these were sufficient to address the prevailing cause of the issue. The inspectors developed a series of questions for the licensee to explore more of the details behind the various potential issues. In order close the URI, the inspectors need to review the licensees response to questions provided and review available documentation of the event. (URI 05000315/201700402, Unit 1 Letdown System Safety Valve Lift During Preparations for Cooldown) | ||
| 05000315/FIN-2017004-03 | 2017Q4 | Green | H.1 | NRC identified | Failure to Promptly Correct The CAQ by Not Testing the CCW Leak Isolation Valves | The inspectors identified a finding and associated NCV of Title 10 of the Code of Federal Regulations (CFR) Title 50, Appendix B Criterion XVI for failing to promptly correct a condition adverse to quality (CAQ). Specifically, in Inspection Report (IR) 05000315/3162015008 the NRC issued an NCV of 10 CFR 50 Appendix B Criterion III for the licensees failure to leak test isolation valves between redundant trains of the component cooling water (CCW) systems for Units 1 and 2. Despite opportunities to restore compliance, for Unit 1, the licensee suffered the violation from November 17, 2015, through November 4, 2017. As of December 31, 2017, the licensee continues to be in violation on Unit 2. The licensee tested the Unit 1 isolation valves during the fall 2017 outage and has scheduled testing of the Unit 2 valves in the spring 2018 outage. The inspectors determined that the licensees failure to promptly correct the CAQ by not testing the CCW leak isolation valves or otherwise restoring compliance was more than minor. The inspectors determined the issue was more than minor because it adversely affected the Mitigating Systems cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The issue was not greater than green because it did not render CCW inoperable. The inspectors determined the finding included a cross-cutting aspect of H.1, Resources. |
| 05000315/FIN-2017004-05 | 2017Q4 | Green | Licensee-identified | Licensee-Identified Violation | The following violation of very low significance (Green) was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. Title 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings states, in part, that activities affecting quality shall be prescribed by documented instructions, procedures, or drawings appropriate to the circumstances and shall be accomplished in accordance with those instructions, procedures, and drawings. Equipment tagging is a safety related process implemented by procedure 12OHP2110CPS001, Clearance Permit System. Contrary to 12OHP2110CPS001 step 4.4.3, which directs operators to comply with the tagout on the Unit 2 East Motor Driven AFW pump room cooler, the operators mistakenly secured and tagged the Unit 1 East Motor Driven AFW pump room cooler instead. This rendered the Unit 1 East Motor Driven AFW pump inoperable. The violation occurred at 0219 on September 6, 2017, and concluded at 0623 the same day after the error was realized and corrected. The licensee entered the issue into their CAP as AR20178509. The finding screened to Green because there was no loss of system function, nor loss of a train for greater than the Technical Specification allowed outage time. | |
| 05000315/FIN-2017007-01 | 2017Q3 | Green | P.3 | NRC identified | Failure to Correct Operable, but Non - Conforming Conditions | The inspectors identified a finding of very low safety significance and an associated non -cited violation of 10 Code of Federal Regulations ( CFR ) Part 50, Criterion V for three examples where the licensee failed to follow procedures associated with the licensees quality assurance program. This issue resulted in the licensee not properly classifying some structures, systems and components (SSCs) as operable, but non- conforming, consistent with station procedures . The inspectors determined that the failure to properly classify the above SSCs as operable, but non -conforming, was within the licensees ability to foresee and correct and was, therefore, a performance deficiency. This performance deficiency was considered more than minor, because it adversely affected the Design Control attribute 3 of Reactor Safety Barrier Integrity, ensuring that SSCs would remain functional during a design basis event. Specifically, station procedures required that prompt action be taken to address operable, but non -conforming conditions. The inspectors evaluated the finding using the Significance Determination Process in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At -Power, Exhibit 3, dated June 19, 2012. The finding was of very low safety significance (Green), because there was no actual loss of safety function for the affected SSCs. The inspectors determined this finding affected the cross -cutting area of problem identification and resolution in the aspect of resolution, specifically to ensure that the organization takes effective corrective actions to address issues in a timely manner commensurate with their safety significance. (P.3) |
| 05000315/FIN-2017002-05 | 2017Q2 | Green | NRC identified | Inadequate Design Control Measures to Ensure Leakage Remained Within Analysis | Green . The inspectors identified a finding of very low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to have adequate design control measures verify that the Essential Service Water to Containment Spray (CTS) heat exchanger outlet valves were not leaking in excess of the limits of the Large Break Loss of Coolant Accident (LBLOCA) analysis. This finding was entered into the licensees CAP to evaluate adequate design control measures. The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the capability of the CTS system to respond to an initiating event to prevent undesirable consequences. The finding screened as of very low safety significance (Green) because it did not result in the loss of operability or functionality of one of the trains of the CTS system. The inspectors did not identify a cross -cutting aspect associated with this finding because it was not reflective of current performance. | |
| 05000315/FIN-2017405-01 | 2017Q2 | Green | Licensee-identified | Licensee-Identified Violation | ||
| 05000316/FIN-2017002-02 | 2017Q2 | Green | H.5 | Self-revealing | Unit 2 CEQ Fan Failed Surveillance | Green . A finding of very low safety significance was self -revealed on March 23, 2017, when one of the Unit 2 Containment Equalization (CEQ) Fans, 2 HV CEQ 1, failed its surveillance. Technical Specification (TS) 5.4.1, Procedures, requires that the applicable procedures covered in Regulatory Guide 1.33 are established, implemented, and maintained. Regulatory Guide 1.33 requires that maintenance that can affect the performance of safety -related equipment should be properly preplanned and performed in accordance with documented instructions appropriate to the circumstances. Contrary to these requirements, a preventative maintenance activity to grease the backdraft damper bearings of the CEQ fan resulted in the fan being left inoperable until the next scheduled surveillance approximately a month later. Due to inadequate work instructions, the damper was not cycled enough times following greasing, which resulted in a condition where more force than allowed by the Technical Specifications was required to open the damper. Due to an inadequate post -maintenance test, this was not detected until the next surveillance was performed. Upon failure of the surveillance, technicians re -greased the bearings, cycled the damper, and tested it satisfactorily. Although qualified, the technicians who first performed the maintenance were unaware of certain nuances associated with the CEQ fan dampers. This information was not described in the work instructions and the post -maintenance test did not validate the opening force. The issue was entered into the CAP and an apparent cause evaluation was performed by the licensee. The issue was greater than minor because it adversely affected the Procedure Quality attribute of the Mitigating Systems cornerstone. Specifically, the inadequate maintenance procedures adversely affected the availability, reliability, and capability of a system that responds to initiating events to prevent undesirable consequences (i.e., core damage). The finding screened to Green, or very low safety significance, based on IMC 0609 Appendix H, Containment Integrity Significance Determination Process, because CEQ fans are not important contributors to Large Early Release Frequency and Hydrogen Igniters remained available. The inspectors determined there was a cross- cutting aspect associated with the finding, namely, H.5., Work Management. Specifically, the licensee did not identify and manage risk nor coordinate between different work groups when it was recognized the normal maintenance group would not be working on the CEQ Fan. Further, the apparent cause evaluation identified a need to better coordinate the preventative maintenance activities with the surveillance tests. |
| 05000315/FIN-2017002-06 | 2017Q2 | Green | Self-revealing | Single Point Failure Vulnerability in Annunciator System | Green . A self -revealed finding occurred on March 30, 2017, when operation of a work station for the control room annunciators caused a loss of all annunciators in the Unit 1 control room. Specifically, a software error coupled with an overflowing cache caused a single point fai lure of the Unit 1 annunciator. When in us e by a control room operator, Server 1 for the annunciator system failed and transferred functions to Server 2 . Server 2 also failed causing a loss of all annunciators for the Unit 1 control room. The licensee restored the system a few hours later and entered the condition into the corrective action program. The inspectors determined that the failure to design the system to preclude loss of a single active component from causing a loss of the annunciator system was a performance deficiency that warranted a significance determination. Using IMC 0612, the inspectors determined the finding was more than minor because it adversely impacted the mitigating system cornerstone objective to ensure the availability of systems that respond to initiating event. Using IMCC 0609, the inspectors determined that support of the Senior Risk Analyst (SRA) was needed because the condition resulted in the loss of a function, the annunciators. The S RA performed a simple detailed analysis and concluded the finding was of very low safety significance. The inspectors did not identify a cross -cutting aspect associated with this finding because it was not reflective of current performance. | |
| 05000315/FIN-2017002-04 | 2017Q2 | Severity level IV | NRC identified | Failure to Report Deficiencies as Required by 10 CFR 50.46 | SL IV. The inspectors identified a Severity Level IV Violation of 10 CFR Part 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light -Water Nuclear Power Reactors. Specifically, the licensee failed to report the effects of the errors in the 5 LBLOCA Evaluation Model for the Unit 1 emergency core cooling systems. The inspectors determined that the failure to estimate and report the errors in the LB LOCA analyses were contrary to the requirements of 10 CFR 50.46 and was a performance deficiency. The performance deficiency was determined to be minor because the failure to report was not willful, did not impact a performance indicator, was not a material condition issue which could lead to a more significant safety issue, and did not impact the Mitigating Systems cornerstone objectives. The inspectors determined the failure to report was a Severity Level IV violation in accordance with Section 6.9 of the Enforcement Policy. A cross -cutting aspect was not assigned since the performance deficiency is minor. | |
| 05000315/FIN-2017002-01 | 2017Q2 | Green | H.6 | NRC identified | Failure to Ensure the Unit 2 CCW Heat E xchanger Monitoring Program Could Demonstrate Its Continued Operability Between Maintenance Intervals | Green . The inspectors identified a finding of very -low safety significance (Green) and associated NCV of Title 10 of the Code of Federal Regulations , (CFR) Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to establish a heat exchanger monitoring program for the Unit 2 east component cooling water (CCW) heat exchanger that demonstrated it would perform satisfactorily in service and remain operable within its required range of physical conditions for the entire interval between heat exchanger maintenance inspections and c leanings. The licensee entered this finding into their Corrective Action Program (CAP) and, after a review of the Ultimate Heat Sink temperatures, determined the Unit 2 East CCW heat exchanger remained operable because the Ultimate Heat Sink temperatures had remained below the point where operability of the heat exchanger could be challenged. The performance deficiency was determined to be more- than- minor because it was associated with the Mitigating Systems cornerstone attribute of Equipment Performance and it adversely affected the associated cornerstone objective to ensure the availability, reliability, and capability of the CCW system to respond to initiating events to prevent undesirable consequences. Specifically, the monitoring program established for the Unit 2 East CCW heat exchanger did not ensure its availability, reliability, and capability for the entire interval between heat exchanger maintenance inspections and cleanings. The finding screened as of very -low safety significance (Green) because although it affected the design or qualification of the Unit 2 East CCW heat exchanger, it did not result in the loss of operability or functionality of the heat exchanger . The inspectors determined this finding had an associated cross -cutting aspect, Design Margins, in the Human Performance cross -cutting area (H.6) because the licensee did not ensure the Unit 2 East CCW heat exchanger s heat transfer margin was carefully guarded after discovering excessive tube plugging above the acceptance criteria i n 2016. Specifically, special attention was not placed on maintaining the safety -related heat exchanger to ensure it would remain capable of performing its specified safety function within the required range of physical conditions during the entire interval between heat exchanger maintenance inspections and cleanings. |
| 05000315/FIN-2017002-03 | 2017Q2 | Green | Self-revealing | Failure to Identify Parts Subject to a Part 21 | Green. A self -revealed finding and associated violation occurred on April 2, 2012, when the licensee failed to prevent installation of relays identified in a P art 21. Although the performance deficiency occurred in 2012, the consequence of the error did not manifest until March 2017, when a defective relay caused the Unit 2 control room indicating and display (CRID) 3 inverter to transfer and remain on the alternate power supply. Title 10 CFR 50 Appendix B, Criterion XV requires, in part, that Measures shall be established to control materials, parts, or components which do not conform to requirements in order to prevent their installation. Contrary to this requirement, on April 22, 2012, the licensee failed to prevent installation of an AMETEK board, PC 201 with a defective relay. This led to a failure of the CRID 3 inverter on March 27, 2012. The licensee replaced the circuit board and restored CRID 3 to an operable status. The inspectors determined that the failure to prevent installation of defective parts into the safety related CRID system was a performance deficiency that warranted a significance determination. Using Attachment 0609.04, Initial Characterization of 4 Findings, dated October 7, 2016, Table 2, the inspectors determined that the finding affected the Mitigating Systems cornerstone. As a result, the inspectors evaluated the finding using IMC 0609, Attachment 1 Exhibit 2, dated June 19, 2012. The inspectors answered no to all the questions, therefore the finding screened as Green. Using Attachment 0609.04, Initial Characterization of Findings, dated October 7, 2016, Table 2, the inspectors determined that the finding affected the Mitigating Systems cornerstone. The inspectors did not identify a cross- cutting aspect associated with this finding because it was not reflective of current performance. | |
| 05000315/FIN-2017404-01 | 2017Q2 | Green | P.3 | NRC identified | Security | |
| 05000315/FIN-2017002-07 | 2017Q2 | Severity level IV | Licensee-identified | Licensee-Identified Violation | Title 10 CFR 50.71(e) required that the UFSAR be updated to assure that the latest information developed was in the UFSAR. In AR 2010 4194, Unit 1 and Unit 2 Small Break Loss of Cooling Accident (SBLOCA) Analyses, the licensee identified the March 2007 Unit 1 SBLOCA analysis had not incorporated into the UFSAR and was not included in the October 2008 UFSAR update provided to the NRC. The inspectors determined the failure to update the UFSAR by incorporating the newest SBLOCA analyses was contrary to 10 CFR 50.71e. The inspectors reviewed this issue in accordance with NRC IMC 0612 and the NRC Enforcement Policy. Violations of 10 CFR 50.71(e) are disposed using the traditional enforcement process because they are considered to be violations that potentially impede or impact the regulatory process. The inspectors reviewed Section 6.1.d.3 of the NRC Enforcement Policy and determined this violation was Severity Level IV because the licensees failure to update the UFSAR as required by 10 CFR 50.71(e) had not yet resulted in any unacceptable change to the facility or procedures. The inspectors determined the performance deficiency was minor in that failure to update the UFSAR was not willful; did not impact a performance indicator; was not a material condition issue which could lead to a more significant safety issue, and did not impact the Mitigating Systems cornerstone objectives . | |
| 05000316/FIN-2017001-02 | 2017Q1 | Green | Self-revealing | Improper Disconnect Operation | Green . A self -revealed finding and associated NCV occurred on January 10, 2017, when the licensee caused a loss of a qualified off-site circuit while opening a disconnect on the Unit 2 reserve feed transformer. Regulatory Guide 1.33 requires procedures for operating the onsite and offsite electrical distribution system; however the licensee did not develop a procedure or instruction for operating the electrical distribution system. Licensee personnel opened a disconnect to the Unit 2 reserve auxiliary transformer with the transformer energized but unloaded. This action resulted in trip of an upstream breaker and unplanned Technical Specification entry for the opposite unit. The licensee recovered the offsite circuit for Unit 1. The licensee entered the issue into the corrective action program (CAP) as Action Request ( AR ) 2017 0346. The inspectors determined that the failure to develop, implement , and maintain procedures or work instructions for the electrical distribution system was a performance deficiency. The performance deficiency impacted the mitigating system performance objective of ensuring the availability of systems that respond to initiating events. The finding was not greater than green in accordance with IMC 0609, Appendix A, Exhibit 2, dated June 19, 2012, because the answer to all four questions was no. The finding does not include a cross -cutting aspect because the licensee followed guidance for operating the disconnect that existed for the life of the plant and is therefore not reflective of current performance. | |
| 05000315/FIN-2017001-03 | 2017Q1 | Green | P.2 | Self-revealing | Failure to Control Nonconforming Delivery Valve Holders on Emergency Diesel Generators | Green . A self -revealed finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion XV, Nonconforming Materials, Parts, or Components, occurred when the delivery valve holder (DVH) on a fuel injection pump failed during a run of the 1AB emergency diesel generator ( EDG ). Each cylinder on an EDG has a fuel injection pump. The DVH is the part of the fuel injection pump where the high pressure fuel line meets the pump discharge. A thru- wall crack developed from a machined portion inside the DVH that had too sharp of a corner. This same phenomenon occurred onsite and caused a leak in 2013 as well. In 2013, the licensee identified the tight radius as an issue and also identified a particular manufacturing lot of DVHs that might have the tight radius. Contrary to their commercial grade dedication (CGD) procedures, the licensee did not 3 update their CGD plan for these parts to include the radius as a critical characteristic. Further, the licensee relied on informal communications from the commercial grade supplier of the parts to conclude only a certain subset of the suspected lot of DVHs were susceptible to cracking. Finally, several management -approved actions to remove all affected DVHs of the lot were not performed, as there was the belief by some that only certain DVHs were affected. As a result, the licensee installed many DVHs from the suspect lot they thought were acceptable. However, in December 2016, one of the DVHs thought to be acceptable developed a leak during an EDG run. The radius was discovered to be out of tolerance, as were numerous other radii in DVHs across all of the EDGs which were from the suspect manufacturing lot. The licensee declared three of the four onsite EDGs inoperable, replaced DVHs, and commenced a root cause evaluation to address the issue. The issue was more- than -minor because it adversely affected the Design Control attribute of the Mitigating Systems cornerstone. Specifically, allowing nonconforming parts to be installed on safety -related equipment without proper controls or evaluation adversely affects 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 Green because performance testing of representative DVHs and engine analysis demonstrated that the EDGs in the as -found condition would have been able to perform their safety functions for the required lengths of time. The inspectors determined the issue had a cross -cutting aspect in the Problem Identification and Resolution area, specifically, P.2, Evaluation. Despite identifying a defect on a safety related part due to a failure in 2013, the licensee failed to properly evaluate the condition and ensure all susceptible parts were accounted for. Specifically, the failure to follow station processes for corrective action and CGD resulted in a defective part causing a leak on an EDG |
| 05000315/FIN-2017001-04 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | In LER 05000315 2015 004, the licensee identified multiple violations of TS 3.4.11, which requires each PORV and associated block valve to be operable. The licensee identified that on 3 occasions for Unit 1 and 4 for Unit 2 that one PORV was not operable when the supporting control air compressor was out of service. In these 7 cases, the licensee failed to close the associated block valve within 1 hour, as required by Condition B. Further, the licensee also failed to be in mode 3 within 6 hours and in some cases mode 4 as required by Condition H. The licensee failed to meet these requirements as follows: Unit 1: February 27, 2013 November 8, 2014 May 6, 2015 Unit 2: May 14, 2013 January 16, 2014 31 March 24, 2015 May 4, 2015 Th e inspectors evaluated the condition in accordance with IMC 0612, Appendix B and determined the issue was more than minor because it adversely affected the mitigating system cornerstone objective of ensuring the availability of systems that respond to initiating events. The inoperability impacted the equipment performance attribute of availability. Using IMC 0609, Appendix A , Exhibit 2, the inspectors answered all the questions no; therefore, the inspectors determined the finding was of very low safety significance. The licensee documented the issue in AR 2015 11204. | |
| 05000315/FIN-2017001-01 | 2017Q1 | Green | H.4 | Self-revealing | Failure to Brief Worker Entry to High Radiation Area Resulting in the Unplanned Dose Rate Alarm | Green . A finding of very -low safety significance and an associated NCV of Technical Specification 5.7.1.b was self -revealed for the failure to a make radiation worker aware of the radiation dose rate before entering a high radiation area. The failure to brief the worker resulted in an unplanned electronic dosimeter dose rate alarm. The worker immediately exited the area and reported the event to the radiation protection staff. The license e entered the event into their CAP as AR 2016 13827. The inspectors determined that the performance deficiency was more than minor i n accordance with IMC 0612, Appendix B, because the finding impacted the program and process attribute of the Occupational Radiation Safety Cornerstone and adversely affected the cornerstone objective of ensuring adequate protection of worker health and safety from exposure to radiation. Specifically, worker entry into a high radiation area without an adequate briefing could lead to unintended dose. The inspectors also identified an example in IMC 0612, Appendix E, which is similar to the performance issue. Therefore, t he finding was determined to be of very -low safety significance in accordance with I MC 0609, Appendix C, Occupational Radiation Safety Significance Determination Process, dated August 19, 2008. The violation was of very- low safety significance (Green) because: (1) it did not involve as -low -as-reasonably -achievable planning or work controls, (2) there was no overexposure, (3) there was no substantial potential for an overexposure, and (4) the ability to assess dose was not compromised. The inspectors concluded that the cause of the finding involved a cross -cutting 4 component in the human performance area, H.4, in the area of teamwork and communication and coordination across organizational boundaries, specifically between radiation protection staff and the individual . This resulted i n the worker proceeding into areas that they were not briefed to enter which contained unknown dose rates . |
| 05000315/FIN-2017001-05 | 2017Q1 | Green | Licensee-identified | Licensee-Identified Violation | The inspectors reviewed AR 2017 0503, VT- 2 examination not completed. The AR documented several cases where the licensee identified that safety related equipment had been returned to service without the necessary American Society of Mechanical Engineers Code required exams or evaluations being done to satisfy post -maintenance test (PMT) requirements. TS 5.4, Procedures, requires, in part, that written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in RG 1.33, Revision 2, Appendix A, February 1978. RG 1.33 Section 9 states, in part, that maintenance that can affect the performance of safety -related equipment should be properly preplanned and performed in accordance with written procedures or documented instructions appropriate to the circumstances. Contrary to this requirement, safety related valves 2 NCR 252 and 2 CMO 410 were returned to service approximately December 9, 2016 and November 13, 2016, respectfully, following maintenance without the required American Society of Mechanical Engineers visual inspections having been planned into the work orders. Once identified, the PMTs were verified complete or acceptable on January 24, 2017 for 2 NCR 252 and January 30, 2017 for 2CMO 410. The issue was more than minor because it adversely affected the Procedure Quality attribute of the Mitigating Systems cornerstone and was a programmatic issue. The finding screened as a Green NCV because operability was maintained as verified later via the appropriate PMTs . The licensee documented the issue in the aforementioned AR | |
| 05000316/FIN-2016004-01 | 2016Q4 | Green | H.9 | NRC identified | Designated Individual not at Airlock | Green. The inspectors identified a finding and associated NCV of Technical Specification (TS) 5.4.1 for failing to station a designated individual at the airlocks. Licensee procedure 2OHP4030227041, Revision 34 required that a designated person be available at the airlock at all times during fuel handling if both air lock doors are open. TS 5.4.1, Procedures, requires, in part, that the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, be established, implemented, and maintained. Regulatory Guide 1.33 states, in part, that general plant operating procedures for refueling and core alterations should be covered by written procedures. Contrary to this requirement, on October 18, 2016, the licensee failed to implement procedure 2OHP4030227041, Refueling Integrity. In response to the inspectors concern, the licensee stationed the designated individual. The licensee entered the issue into their CAP as AR210611898. The issue screened as more than minor because it adversely affected the Human performance attribute of the barrier integrity cornerstone. The inspectors concluded the issue was of very low safety significance using IMC 0609 Appendix G, Attachment 1 dated May 9, 2014 because the issue did not increase Core Damage Frequency or Large Early Release Frequency. The finding included a cross-cutting aspect of H.9, training, because operations staff had an incorrect understanding of the procedural requirements. |
| 05000316/FIN-2016404-01 | 2016Q4 | Severity level Enforcement Discretion | Licensee-identified | Licensee-Identified Violation | ||
| 05000315/FIN-2016201-01 | 2016Q4 | Green | H.8 | NRC identified | Security | |
| 05000315/FIN-2016201-02 | 2016Q4 | Green | NRC identified | Security | ||
| 05000316/FIN-2016004-02 | 2016Q4 | Green | H.12 | Self-revealing | Moisture Separator Reheater Rupture | Green. A self-revealed finding of very low safety significance (Green), occurred on July 6, 2016, when a portion of the Unit 2 Right Moisture Separator Reheater (MSR) B bellows assembly ruptured, causing a steam leak which damaged the adjacent turbine building wall. There were no associated violations of regulatory requirements since the piping was non-safety-related. Reacting to the rupture, operators tripped the reactor and isolated the leak by shutting the Main Steam Isolation Valves. While addressing a number of issues with the MSRs that occurred following a re-design of the internals in 2010, the licensee changed the design of the rods that hold the bellows assembly on each MSR pipe together. The design change called for tack welds to only be used on the end nuts of the rod. Contrary to the design change (EC51875), tack welds were placed on other nuts as well. The tack welds were determined to have changed the material properties of the rod in the vicinity of the welds, which caused cracking to initiate during operation. Eventually, the cracks grew to a point where two rods completely severed, causing the bellows to tear and rupture. Following the safe shutdown, the licensee repaired the bellows, inspected other rods, and restarted the plant. The issue was entered into their Corrective Action Program (CAP) as Action Request (AR)20167865. The issue was more than minor because it adversely affected the Design Control Attribute of the Initiating Events cornerstone because it resulted in a reactor trip and Unusual Event. Per the Significance Determination Process, a detailed risk evaluation was required because during the rupture operators had to close the Main Steam Isolation Valves, which isolated the main condenser (the preferred post-trip decay heat removal path). An NRC Regional Senior Reactor Analyst performed the evaluation and concluded the finding was of very low risk significance (Green). The inspectors determined the finding had an associated cross-cutting aspect in the Human Performance Area, specifically, H.12, Avoid Complacency. Specifically, site personnel did not plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. |
| 05000315/FIN-2016406-01 | 2016Q4 | Green | NRC identified | Security | ||
| 05000315/FIN-2016003-04 | 2016Q3 | Severity level IV | Licensee-identified | Licensee-Identified Violation | The following violation of very low significance (Green) or Severity Level IV was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV. The licensee identified a finding and NCV of 10 CFR 50.59 for the failure to properly evaluate modifications to the facility. Specifically, Evaluation 2010001600 failed to addresses each of the changes associated with proposed changes to the UFSAR to incorporate revised analysis to control room and offsite radiological dose consequences for the design bases accident. While preparing a revision to evaluation 2010001600, the preparer noted numerous discrepancies which were documented in AR 201214068. Because of the complex nature of the revisions and licensing history of Cooks dose consequence analysis, the licensee determined use of the 50.59 process would not resolve the discrepancies in the current license bases and concluded the appropriate corrective action would be preparation and submittal of a license amendment to fully implement alternate source term as part of the D. C. Cook license. The licensee has submitted the license amendment request and it is now approved. The inspectors determined that the failure of the licensee to demonstrate via the 50.59 process that the change to the licensee could be made without prior NRC approval was a performance deficiency. In accordance with the NRC Enforcement Policy, August 2016, changes made contrary to 10 CFR 50.59 impact the agencys ability to regulate and warrant enforcement. Examples provided in this Enforcement Policy stipulate that an associated SDP finding of very low safety significance would result in a Severity Level IV finding. The inspectors did not identify any physical changes to the facility based on the changes made to the UFSAR nor any other changes suitable for review in the SDP. Therefore, the inspectors determined there was no associated ROP finding and the issue was a Severity Level IV violation of 50.59. 10 CFR 50.59 requires in part, that a licensee must obtain a license amendment prior to implementing a change if it would result in a departure from a method described in the UFSAR used in establishing the design bases or in the safety analysis. Contrary to this requirement on February 25, 2010, evaluation 20100016 accepted changes to the UFSAR that affected methodologies for control room and offsite dose that departed from methods described in the UFSAR. Because the licensee identified this issue in AR 201214068 and took corrective actions to submit a license amendment and is Severity Level IV, the inspectors conclude that this issue may be disposed as a licensee identified NCV. | |
| 05000315/FIN-2016009-02 | 2016Q3 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a finding of very-low safety significance (Green) and associated NCV of License Conditions 2.C.4 and 2.C.3.o for Units 1 and 2 respectively for the licensees failure to establish an appropriate Monitoring Program in accordance with NFPA 805, Section 2.6. Section 2.6 of NFPA 805 required, in part, that monitoring shall ensure that the assumptions in the engineering analysis remain valid. Contrary to the above, the licensee failed to ensure that the assumptions in the engineering analysis remained valid for the availability and reliability of the auxiliary feedwater pumps in the Monitoring Program. The licensee used Maintenance Rule availability criteria to monitor the auxiliary feedwater pumps which did not bound the Fire PRA assumptions for the unavailability of the components. The performance deficiency was determined to be more-than-minor because the issue adversely impacted the Mitigating Systems cornerstone objective to ensure the capability of systems that respond to initiating events and prevent undesirable consequences due to external events such as fire. Specifically, the failure to adequately monitor plant equipment credited for post-fire SSD could result in that equipment being unavailable for longer periods of time than had been analyzed. The inspectors screened the finding using Inspection Manual Chapter 0609, Significance Determination Process, Appendix F, Fire Protection Significance Determination Process. Since the reactor was still able to reach and maintain a SSD condition, the finding screened as very-low safety significance (Green). The licensee entered the issue into the CAP as AR 2016-7239, NFPA 805 Monitoring Program FPRA\Maintenance Performance, and revised Maintenance Rule administrative procedures to consider the unavailability criteria of components and the impact on the fire PRA. | |
| 05000316/FIN-2016003-03 | 2016Q3 | Green | H.7 | Self-revealing | Wetting of Safety-Related Battery Charger | A finding of very low safety significance with an associated NCV of TS 5.4, Procedures, was self-revealed on June 21, 2016, when safety-related N-Train Battery Charger 2BCB was found soaked with water from a roof leak above. The licensee failed to follow administrative procedures for control of temporary catch basins. TS 5.4 states, in part, that the applicable procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978, be established, implemented, and maintained. Regulatory Guide 1.33 states, in part, that maintenance that can affect the performance of safety-related equipment should be properly preplanned and performed in accordance with written procedures, documented instructions, or drawings appropriate to the circumstances. Contrary to this requirement, the licensee installed and subsequently removed a drip-catch above battery charger 2BCB that was being used to protect the charger from a water leak in the area pending roof repairs. On June 3, 2016, the Performance Assurance department noted the catch had been installed outside of any formal process. In response, the licensee removed the catch but did not put anything in its place to protect the charger. On June 21, 2016, a severe rainstorm occurred, resulting in the wetting of the charger. The other charger was in-service at the time, so there was no impact to the affected N-Train distribution system. In response, the licensee added another protective device, dried out, inspected, and tested the charger. It was restored to operable status on June 23. The issue 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 finding screened as Green, or very low safety significance, because there was no loss of system operability. The finding had an associated cross-cutting aspect in the Human Performance area, specifically, H.7., Documentation. Had the licensee kept their leak detection log up-to-date with the addition of the catch over the charger initially, it would have prompted the licensee to ensure the repairs to the roof were complete before removing the barrier. Further, it would not have been identified as an issue by Performance Assurance. |
| 05000315/FIN-2016009-01 | 2016Q3 | Green | NRC identified | Inadequate Resolution for Double-Break Circuits Design for Several Valves | The inspectors identified a finding of very-low safety significance (Green) and an associated Non-Cited Violation of license conditions 2.C(4) and 2.C(3)(o) for the licensees failure to implement the approved. Specifically, the licensee failed to analyze the double-break circuits design for valves using risk-informed, performance-based techniques for several fire areas. In the event of a fire in several fire areas, fire induced-circuit failures (i.e., inter-cable shorting) for a double-break design for several valves (i.e., Power Operated Relief Valves) could potentially result in spurious operation of the valves. The circuit analysis for these valves in these areas was analyzed using the deterministic approach instead risk-informed, performance-based techniques. The licensee entered the issue into their Corrective Action Program and took credit for existing fire protection features and controls as compensatory measures and planned to review the multiple spurious operations Expert Panel Report and properly disposition the scenario. The performance deficiency was determined to be more-than-minor because if left uncorrected, it would potentially lead to a more significant safety concern. Specifically, the failure to properly evaluate and disposition all potential fire-induced circuit failures for all cables in a fire area could impair the plants ability to safely shutdown in the event of a fire. The performance deficiency was also associated with the Mitigating Systems cornerstone. The finding was of very-low safety significance because it did not impact the reactors ability to reach and maintain a safe shutdown condition. This finding did not have a cross-cutting aspect because it was not representative of current licensee performance. | |
| 05000315/FIN-2016003-02 | 2016Q3 | Green | Self-revealing | Charging System Thru-Wall Leak | A finding of very low safety significance with an associated NCV of Title 10 of the Code of Federal Regulations (CFR) 50 Appendix B, Criterion III, Design Control, was self-revealed when a thru-wall leak was identified on a branch connection off of the Unit 1 west coolant charging pump (CCP) discharge piping while it was in-service. The licensee failed to ensure the branch line design would remain intact when subjected to the vibratory conditions in the line. As a result, a vibration induced fatigue crack developed. This design issue caused a thru-wall leak on a similar line associated with the opposite-train charging pump in 2011. When the licensee addressed the prior leak, assumptions were made regarding the Unit 1 west CCP line. Since the length was slightly different, the belief was it would not be subject to the same increase in vibrations. However, when measuring the vibrations after the recent leak was identified, the results indicated the same elevated vibrations existed. The licensee secured the pump to stop the leak, declared the B train of the emergency core cooling system (ECCS) inoperable, and repaired the leaking weld. The issue was more than minor because it adversely affected the Design Control 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 finding screened as Green, or very low safety significance, because there was no loss of system function and the repairs were completed within the 72 hour timeframe allowed by Technical Specifications (TS). No cross-cutting aspect was assigned because the issue occurred in 2011 and was not reflective of current licensee performance. | |
| 05000315/FIN-2016407-01 | 2016Q3 | Green | P.2 | NRC identified | Security | |
| 05000315/FIN-2016003-01 | 2016Q3 | Green | H.11 | Self-revealing | Improper Backfill Severs Fire Main | A self-revealed finding of very low safety significance (Green) and associated NCV of the license condition for a fire protection program occurred when the licensee failed to ensure excavation activities preserved the functionality of the fire main. Specifically, the licensee improperly backfilled an excavation performed to inspect buried piping. The improper backfill led to a catastrophic failure of the fire main. The performance deficiency was a violation. License conditions 2.C(4) and 2.C(3)(o) of the Donald C. Cook Nuclear Power Plant, Unit 1 and Unit 2 Operating Licenses, respectively, require, in part, that the licensee implement and maintain in effect all provisions of the approved fire protection program that comply with 10 CFR 50.48(a) and 10 CFR 50.48(c), National Fire Protection Association (NFPA) Standard NFPA 805, as specified in the licensees amendment request dated July 1, 2011, as supplemented, and as approved in the Safety Evaluation dated October 24, 2013. Section 3.3.1.1(3) of NFPA 805 requires that, Administrative controls addressing the review of plant modifications and maintenance to ensure that both fire hazards and the impact on plant fire protection systems and features are minimized. Immediate actions included isolation of the faulted section of the fire main and repair of the break. The issue has been entered into the CAP as AR20167626. The inspectors determined that the finding was more than minor because the performance deficiency was more than minor, because it impacted the mitigating system cornerstone attribute of protection against external factors and adversely impacted the cornerstone objective of ensuring the availability of systems to respond to initiating events to prevent undesirable consequences. Using Appendix F, Attachment 1 dated September 20, 2013, the inspector determined that the licensee probable risk assessment should be reviewed to determined significance. With the short duration, the licensee determined the delta cdf to be less than (1e6). These results were reviewed and accepted by the Senior Reactor Analyst. The inspectors determined the finding included a cross cutting aspect of Challenge the Unknown, H.11, in the human performance area. |
| 05000315/FIN-2016002-03 | 2016Q2 | Green | H.3 | NRC identified | Procedure Failed to Establish Two Valve Isolation Between the Reactor Coolants System And Containment Atmosphere | The inspectors identified a finding of very low safety significance (Green) and associated NCV of TS 5.4.1, Procedures, for the licensee failing to maintain procedure 1OHP4021002001, reactor coolant system (RCS) fill and vent. When the licensee updated the procedure to Revision 42 they inadvertently omitted a step to shut valve 1RC144L3, RCS half-loop gauge glass isolation valve. Closing this valve establishes two valve isolation between American Society of Mechanical Engineers (ASME) code class 1 piping and non-code class piping. The licensee closed the valve and updated the affected procedure in response to the inspectors inquiries. The inspectors determined the issue was more than minor in accordance with IMC 0612 because, if left uncorrected the issue would have become more significant concern. Specifically, the licensee would have entered a higher operating mode without establishing the reactor coolant pressure boundary. The issue was not greater than green because actions taken by the licensee established the boundary prior to mode ascension. Therefore, the inspectors concluded the finding was of very low safety significance. The inspectors determined the issue included a cross cutting aspect of H.3, configuration management in the human performance area. For this issue, the licensee failed to update all applicable procedures following a change in methodology for filling coolant loops. |
| 05000315/FIN-2016002-04 | 2016Q2 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified an NCV of TS 5.4.1, procedures, for failing to establish a procedure for testing of containment equalization fans. Specifically, contrary to the requirements of TS 5.4.1 to establish procedures identified in regulatory guide 1.33, the licensee did not establish a procedure for surveillance testing of a section of the circuity for the containment equalization fan. On May 15, 2016, while performing a walkdown for a planned project on the solid state protection system, the licensee identified that a portion of the circuit was not verified to ensure the integrity of the circuit path to actuate a containment isolation fan. After discovery, the licensee performed a test to verify circuit integrity for containment equalization fans on both units. The condition of an inadequate surveillance existed for many years. Since all fans passed the surveillance test, the licensee confirmed the failure to test had not masked an inoperable condition. The issue was more than minor because it impacted the mitigating system cornerstone objective of ensuring the reliability of systems that respond to plan events. Because the licensee demonstrated that the circuit was operable, the inspectors concluded it was not inoperable longer than permitted by technical specifications and determined that issue was of very low safety significance. The licensee entered this condition into the corrective action program as AR 20166136. | |
| 05000315/FIN-2016002-02 | 2016Q2 | Green | H.1 | NRC identified | Containment Closure Requirements during Unit 1 2016 Refueling Outage | The inspectors identified a finding of very low safety significance with an associated NCV of TS 5.4, Procedures, for the failure to implement all of the requirements of PMP4100SDR001, Plant Shutdown Safety and Risk Management, pertaining to the closure of containment airlocks in the event shutdown cooling is lost. Contrary to TS 5.4, the licensee failed to implement the procedure as demonstrated by lack of closure requirement knowledge by containment closure attendants, failure to include isolation valves on ice lines, missing a shiftly check, and lack of required anti-contamination clothing. The licensee corrected the issues and entered them into the CAP. The issue was greater than minor because it adversely affected the Human Performance attribute of the Barrier Integrity cornerstone, whose objective is to provide assurance that principal design barriers (e.g., containment) can protect the public from radionuclide releases. Additionally, the inspectors were informed by IMC 0612, Power Reactor Inspection Reports, dated May 6, 2016, because the issue was programmatic in nature and could lead to more significant issues if left uncorrected. The finding screened as Green per IMC 0609 Appendix H, Containment Integrity Significance Determination Process, because the inspectors determined despite the issues identified, containment closure could be achieved within the time-to-boil. The inspectors determined the finding had an associated cross-cutting aspect of H.1, Resources, because leaders did not ensure personnel, equipment, procedures, and other resources were available and adequate to support nuclear safety. |
| 05000315/FIN-2016002-01 | 2016Q2 | Green | H.5 | NRC identified | Unanalyzed Scaffold Near Safety-Related Equipment | A finding of very low safety significance and associated NCV of Technical Specifications (TS) 5.4, Procedures, was identified by the inspectors. The licensee constructed a scaffold storage rack adjacent to the Unit 1 Component Cooling Water (CCW) Surge Tank without a seismic evaluation. Specifically, contrary to 12MHP5021SCF00, Scaffolding Procedure, the seismic adequacy of the rack was not evaluated prior to construction. The rack was built several weeks before the Unit 1 Spring 2016 refueling outage (RFO) and was assessed by inspectors during the outage as part of walkdown of the area. The licensee removed the scaffold storage rack and entered the issue into their Corrective Action Program (CAP). The issue was greater than minor because it adversely affected the Protection Against External Factors attribute of the Mitigating Systems cornerstone, whose objective is to ensure the capability, reliability, and availability of systems that respond to initiating events to prevent undesirable effects such as core damage. Further, the inspectors determined examples 3.j and 3.k of IMC 0612 Appendix E, Examples of Minor Issues, applied as there was reasonable doubt concerning the operability of the CCW system with the as-found condition of the scaffold. The finding screened as Green, or very low safety significance, utilizing IMC 0609, Significance Determination Process. Specifically, a seismic evaluation later demonstrated that safety functions were maintained. The inspectors determined the finding had a cross cutting aspect of Work Management (H.5.). Specifically, the work management process for scaffold construction was not implemented with nuclear safety as an overriding priority. The process did not identify and manage risks associated with work in the field nor did it ensure coordination between different work groups and activities. |
| 05000315/FIN-2016001-01 | 2016Q1 | Green | H.14 | NRC identified | Incorrect Auxiliary Feedwater Mission Time | The inspectors identified a finding of very low safety significance and associated NCV of with Title 10 of the Code of Federal Regulations (CFR) Part 50, Appendix B, Criterion III, Design Control. Specifically, the licensee failed to ensure that regulatory requirements and design bases were correctly translated into specifications and procedures, in that the licensee used an incorrect mission time for the turbine driven auxiliary feedwater (TDAFW) pump to determine operability. The licensee developed a procedure that permitted continued operability of the TDAFW pump without room ventilation provided room temperature remained below 104 F. The underlying engineering document assumed TDAFW pump mission time was 4 hours; however, this assumption was not supported by current license bases documents. This condition violates 10 CFR 50 Appendix B Criterion III, which requires licensees to establish measures to assure that applicable regulatory requirements and the design bases, as defined in 10 CFR 50.2 and as specified in the license application, for those systems structures and components to which the Appendix applies, are correctly translated onto specifications, drawings, procedures and instructions. The licensee has since restored the room coolers to an operable status, thus, no current safety concern exists. The licensee has entered the condition into the corrective action program (CAP). The licensees use of an incorrect mission time was a performance deficiency that warranted a significance review. Using IMC 0612 appendix B dated September 7, 2012, the inspectors determined that the finding was more than minor because it was associated with the Mitigating System cornerstone objective to ensure the availability, reliability and capability of systems that respond to initiating events and adversely affected the attribute of design control. Specifically, the licensee applied an incorrect mission time when determining room temperatures to ensure TDAFW pump operability. Using IMC 0609 Appendix A, Exhibit 21, dated June 19, 2012, the inspectors answered no to Questions A. 1 thru 4. In particular, control room logs document about 6 hours with the TDAFW room ventilation not functioning; therefore the inspectors determined that the pump would not have been inoperable for longer than the 72 hour completion time in technical specifications. The inspectors also identified a cross cutting aspect of H.14, conservative bias, in the human performance area. |
| 05000315/FIN-2016001-02 | 2016Q1 | Green | Licensee-identified | Licensee-Identified Violation | The licensee identified a finding of very low safety significance (Green) with an associated NCV of 10 CFR 50 Appendix B, Criterion III, Design Control, for the failure to ensure appropriate quality standards were specified and included in design documents associated with the Unit 1 and Unit 2 ESW strainer backwash valves. Specifically, this resulted in the use of non-dedicated parts in the backwash valves. The backwash function of the ESW strainers was originally classified as non-safety-related. However, in 2007, the backwash function became safety-related. When this change occurred, the Safety Classification Determination (SCD), which documented the safety classification of the various parts of the valves, was not updated accordingly. During a maintenance period on the ESW system in 2015, some licensee personnel questioned the adequacy of the SCD. The licensee later determined that non-dedicated replacement parts had been used in some of the strainer backwash valves since 2007. The issue was more than minor because per IMC 0612 Appendix B, it adversely affected the Mitigating Systems cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences. The issue screened as Green based on the guidance in IMC 0609 Appendix A, Exhibit 2. Specifically, the finding was associated with the design or qualification of a mitigating SSC where the operability was maintained. |