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05000263/FIN-2018012-022018Q3MonticelloFailure to Implement Adequate Freeze Protection Monitoring for Condensate Storage Tank Instrumentation Piping in Response to Industry Operating ExperienceThe inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to establish measures to ensure conditions adverse to quality are promptly identified and corrected. Specifically, the licensee failed to identify that monitoring of the CST instrument line heat tracing performed every 30 days was inadequate to assure the safety-related CST level instrumentation remained operable during extreme cold weather conditions
05000263/FIN-2018012-012018Q3MonticelloInboard Main Steam Isolation Valve Closure Time Test Acceptance Criteria Did Not Account for the Design Basis Accident Containment Back Pressure and Pneumatic Supply Operating PressureThe inspectors identified a Green finding and an associated NCV of Title 10 of the Code of Federal Regulations(CFR), Part 50, Appendix B, Criterion XI, Test Control, for the failure to assure that applicable requirements and acceptance limits contained in the inboard main steam isolation valve (MSIV) design documents were incorporated into their test procedure. Specifically, the inboard MSIV closure time acceptance criteria contained in Functional Test Procedure 0255-07-IA-2, Main Steam Isolation Valve Functional Checks Test, did not account for the elevated containment pressure and the expected lower pneumatic supply pressure expected during design basis accidents.
05000315/FIN-2018010-012018Q3CookRecord Retention Requirements of the Boron Injection Tank and its Associated Support StructureThe 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.
05000263/FIN-2018012-032018Q3MonticelloLicensee-Identified ViolationThis violation of very-low safety significance was identified by the licensee and has been entered into the licensee CAP. Therefore, this finding being treated as an NCV, consistent with Section 2.3.2 of the Enforcement Policy.Enforcement:Violation: Title 10 CFR Part 50, Appendix B, Criterion III, Design Control, requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions. Updated Final Safety Analysis Report, Appendix I,Evaluation of High Energy Line Breaks Outside Containment,Table I.5-2, Table of System Effects,Revision 36P, listed the Division II emergency power system as available during HELBs outside containment. Contrary to the above, on July 29, 1974, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically,the Division II emergency power system would not be available during a HELB outside containment.Procedure B.09.07-05, Operations Manual Section 4.16 kV Station Auxiliary, Revision 53,had actions that required entry into the lower 4kV area to permit repowering Division II emergency power systems but this area would be inaccessible during the event. Significance: The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of Design Control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences.Specifically, the performance deficiency resulted in a condition were the Division II emergency power system would not be available during HELBs outside containment. The inspectors assessed the significance of the finding using the SDP in accordance with IMC 0609, 11 Appendix A, The Significance Determination Process for Findings At-Power, using Exhibit 2, Mitigating System Screening Questions,and concluded the violation was of very-low safety or security significance (Green)because the licensee reasonably demonstrated an alternate strategy was available to timely reach and maintain cold shutdown conditions. Corrective Action References: CAP501000011837, CAP 50100001593
05000282/FIN-2018011-022018Q2Prairie IslandPotential 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-2018011-012018Q2Prairie IslandFailure to Justify Load Combinations Used in Main Steam Piping Stress AnalysisInspectors 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.
05000454/FIN-2018010-042018Q1ByronUse of 10 CFR 50.54(x) for Unit AFW Cross-TieIn 2008, the licensee added steps to Emergency Operating Procedure (EOP) 1/2BFR-H.1, Response to Loss of Secondary Heat Sink, to use the MDAFW train of a non-accident unit to combat a loss of all feedwater event in the opposite unit by using a recently installed unit cross-tie. The EOPs also directed operators to enter the technical specification LCO action statement for the unit donating the MDAFW train because the MDAFW trains were not designed and licensed to be shared between the reactor units.In 2011, the resident inspectors noted that the EOP change resulted in more than a minimal increase in the likelihood of occurrence of a malfunction of a SSC important to safety previously evaluated in the Updated Final Safety Analysis Report because the Updated Final Safety Analysis Report described the MDAFW trains as non-shared systems. However, the licensee implemented this change without prior NRC approval. As a result, the inspectors documented a Severity Level IV NCV of 10 CFR 50.59 in Inspection Report 05000454/2011004; 05000455/2011004 as NCV 05000454/2011004-02; 05000455/2011004-02, Modification of the Auxiliary Feedwater System Without Prior NRC Approval (REF: Accession No. ML 113070678).As corrective actions to this NCV, the licensee removed the steps in the EOPs that directed the unit cross-tie to be used and removed credit for the cross-tie in the stations Probabilistic Risk Assessment model. However, on August 8, 2017, the licensee added direction in EOP1/2BFR-H.1 to use the Unit Auxillary Feedwater cross-tie by invoking 10 CFR 50.54(x). Specifically, the change added a note and a caution that provided direction to initiate the MDAFW unit cross-tie before bleed and feed. The note stated:If at any time it has been determined that restoration of feed flow to any SG is untimely or may be ineffective in heat sink restoration, then the AF crosstie should be implemented per Step 5 (Page 8). The caution stated: The AF crosstie should be implemented per Step 5 if other attempts to restore feed flow to the SG(s) will not prevent the initiation of feed and bleed. Step 5 provided instructions on how to perform the cross-tie and did not include instructions on when to initiate it. The caution also stated Use of the AF crosstie requires invoking 50.54(x).During this inspection period, the inspectors challenged the use of 10 CFR 50.54(x) to implement this permanent change. In addition, the inspectors noted that the licensees 10 CFR 50.59 screening for the procedure change did not include in its review the added note and caution statements. Because the added note and caution were the only procedure provisions that provided direction on when to use the MDAFW cross-tie, the 10 CFR 50.59 screening did not review the instructions about when to use the MDAFW cross-tie. As a result, the screening failed to determine that the change may have required a technical specification change and, thus, a license amendment as originally planned.At the end of the inspection, the NRC continued to evaluateif a performance deficiency and or violation occurred. This Unresolved Item will remain open pending the outcome of this ongoing review.
05000454/FIN-2018010-032018Q1ByronFailure to Verify the Adequacy of the Air Pressure Regulator SetpointValue for Containment Isolation Valves 1(2)RF026The inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, AppendixB, Criterion III, Design Control, for the failure to verify the adequacy of the air pressure regulator setpoint value for air-operated containment isolation valves 1(2)RF026. Specifically, these safety-related valves were located inside containment but the licensee did not verify that their air pressure regulator setpoint value was adequate to provide the motive force necessary to close them against containment accident pressure and within their allowable stroketimes.
05000454/FIN-2018010-022018Q1ByronFailure to Periodically Test Instrument Air Check Valves Associated with Air-Operated Containment Isolation ValvesThe inspectors identified a Green finding and an associated NCV of 10 CFR Part 50, AppendixB, Criterion XI, Test Control, for the failure to periodically test instrument air check valves associated with air-operated containment isolation valves. Specifically, the licensee was not periodically testing the check valves designed to close and maintain sufficient pneumatic pressure in the accumulator tanks installed to closed air-operated containment isolation valves 1(2)RF026 and 1(2)RF027 in response to a containment isolation signal.
05000454/FIN-2018010-012018Q1ByronFailure to Prescribe Motor Driven Auxiliary Feedwater Pump Test Procedures that Accounted for the Allowed Emergency Diesel Generator Frequency VariationThe inspectors identified a Green finding and an associated Non-Cited Violation (NCV)of Title 10 of the Code of Federal Regulations (CFR),Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the failure to prescribe motor driven auxiliary feedwater pump test procedures that accounted for the allowed emergency diesel generator frequency variation. Specifically, the motor driven auxiliary feedwater pump surveillance procedures would allow a pump with degraded and unacceptable performance to meet the test acceptance criteria based upon the test being performed at nominal frequency and not accounting for potentially lower, allowable, emergency diesel generator frequency.
05000249/FIN-2018012-012018Q1DresdenFailure to Ensure that Thermal Overload Relays are Sized Properly for Throttling Motor Operated ValvesThe team identified a finding having very-low significance and an associated Non-Cited Violation of Title 10 of the Code of Federal Regulations,Part 50, Appendix B, Criterion III, Design Control.Specifically, Dresden had not verified that thermal overload relays on Unit 3 safety-related motor operated valves 3-1301-3, 3-1501-21A & 21B, 3-1501-18A & 18B, 3-1501-38A & 38B, 3-3-2301-10, 3-1501-3A & 3B, were properly sized to support the design function of repetitive jogging and throttling the valves in response to design basis transients or accidents.
05000315/FIN-2017004-052017Q4CookLicensee-Identified ViolationThe 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-2017004-042017Q4CookFailure to Verify the Adequacy of the Design for a Temporary ModificationA 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-2017004-032017Q4CookFailure to Promptly Correct The CAQ by Not Testing the CCW Leak Isolation ValvesThe 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-2017004-022017Q4CookUnit 1 Letdown System Safety Valve Lift During Preparations for CooldownRefueling 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-2017004-012017Q4CookFailure to Correct Numerous Anchor Darling Double Disc Gate Valve Non-ConformancesThe 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.
05000440/FIN-2017008-032017Q4PerryFailure to Verify the Capability to Manually Backwash the Emergency Service WaterStrainer during Loss of Offsite PowerThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control,for the failure to verify the capability to manually backwash the emergency service water (ESW) strainer during a LOOP. Specifically, the licensee credited the capability to manually backwash the ESW strainers during a LOOP. However, the associated differential pressure alarm setpoint did not ensure sufficient time to complete this activity because the alarms were set at the same value as the design differential pressure value assumed by the hydraulic calculations. The licensee captured the issue in their CAP as CR-2017-09033, reasonably determined ESW remained operable, and planned to revise the associated calculation and the alarm setpoint to ensure sufficient time to perform the required manual actions during a LOOP.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency did not assure the ESW capability to supply the required minimum flow to its supported components. 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. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the alarm set point was established more than 3 years ago.
05000440/FIN-2017008-022017Q4PerryInadequate Evaluation of Emergency Closed Cooling System Pipe SupportThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to consider all stresses resulting from the emergency closed cooling system as built pipe support 1P42-H1080 connection details. Specifically, the evaluation for the pipe support did not address the impact of rigid connections at both ends of the W8 steel post and of the lateral load on W21 auxiliary steel beam. The licensee captured the issues in their CAP as CR-2017-08986 and CR-2017-09043, reasonably determined the support remained operable, and planned to revise the affected structural analyses.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability,reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.Specifically, the failure to analyze actual pipe configuration and to evaluate the W21 beam did not ensure the emergency closed cooling system and its safety-related supported loads would remain available and capable of providing their accident mitigating function. 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. The team determined that this finding had a cross-cutting aspect in the area of human performance because the licensee did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the licensee did not recognize this latent issue when revising the structural evaluation in 2015.
05000440/FIN-2017008-012017Q4PerryFailure to Address the Susceptibility of the Condensate Storage TankLow Level Instrument Lines to FreezeThe team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations(CFR),Part 50, Appendix B, Criterion III, Design Control, and 10 CFR 50.63, Loss of All Alternating Current Power, for the licensees failure to evaluate the capability to transfer the high pressure core spray (HPCS)and the reactor core isolation cooling (RCIC) pumps suction source from the condensate storage tank (CST)to the suppression pool during cold weather conditions. Specifically, (1) monitoring of the CST level instrument lines heat tracing was inadequate to detect a credible common mode failure before the instrument lines would freeze and be rendered inoperable during normal operation, (2)the licensee did not address the condensate (CST) level instrument lines susceptibility to freeze during a cold weather loss of off-site power (LOOP) event with or without a design basis transient or accident, and (3)the licensee incorrectly evaluated the capability to transfer the HPCS pump suction source from the CST to the suppression pool during a cold weather station blackout (SBO) event. The licensee captured the issues within their Corrective Action Program (CAP) as Condition Report(CR) CR-2017-08685, CR-2017-08930, and CR-2017-09006. Corrective actions implemented included: increased the CST level instrument line heat tracing circuit monitoring frequency, revised the affected procedures ensured HPCS and RCIC are adequately aligned to the suppression pool during LOOP design basis events, and ensured a timely transfer of the HPCS and RCIC pump suctions to the suppression pool during a SBO. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability,reliability, and capability of systems that respond to initiating events to prevent undesirable consequences.Specifically, the failure of the HPCS and or RCIC pumps to automatically transfer their suction source from the CST to the suppression pool upon reaching a low CST water level condition could damage the pump(s) thus preventing them to be used to mitigate a transient or accident. A detailed risk evaluation was performed and determined that the finding was of very-low safety-significance (Green). The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the CST instrument lines were designed and the SBO coping strategy during cold weather was established more than 3 years ago.
05000255/FIN-2017007-032017Q4PalisadesContainment Dome Truss AnalysisThedome truss system was originally designed to support the containment liner plate and wet concrete during the construction of the containment dome (i.e., the liner plate initially acted as a form and the truss supported the form). After the concrete cured, the dome truss system was lowered away from the liner and was used to support the safety injection tanks(SITs) and CS system piping and their associated supports. The CS and SIT systemsare both safety-related which were required to be evaluated for seismic loads (self-weight and externally applied loads). The dome truss system would have alsobeen required to be evaluated for seismic loads. The UFSAR,Section 6.1,described the safety-related design function of the SITsystemwas to prevent fuel and cladding damage that could interfere with adequate emergency core cooling, and to limit the cladding-water reaction to less than approximately 1percentfor all break sizes in the primary system piping up to and including the double-ended rupture of the largest primary coolant pipe, for any break location, and for the applicable break time. Also,the SITsystem also functions to provide rapid injection of large quantities of borated water for added shutdown capability during rapid cooldown of the primary system caused by a rupture of a main steam line. UFSAR Section 6.2.1 described the safety-related design function of the CSsystem was to limit the containment building pressure rise and reduce the airborne radioactivity in containment by providing a means for spraying the containment atmosphere after occurrence of a LOCAor a main steam line break.The inspectors requested the design basis analysis of the dome truss system that considers the LOCA loading on the dome truss system as well as the seismic loading due to the applied design loads from the CS and SITsystem. During the time of the inspection, the licensee was unable to locate the dome truss analysis. In response to the inspectors concern, the licensee entered the issue into their CAP asCR 2017-05016, Dome Trusses, dated November 1, 2017. The licensee is investigating the containment dome truss analysis further with the vendor of the dome truss system.This issue is a URI pending additional inspector review of the design basis analysis for the containment dome truss system. (URI 05000255/2017007-03; Containment Dome Truss Analysis)
05000255/FIN-2017007-022017Q4PalisadesContainment Spray Pipe Support Strap DeficienciesThe inspectors identified a finding of low safety significance (Green) and an associated potential NCV of Title 10of theCode of Federal Regulations,Part 50, Appendix B, Criterion III, Design Control, for failure to meet Updated Final Safety Analysis Reportrequirements for containment spraypiping supports, specifically straps. Specifically, the inspectors identified that Calculation No. EA-SP-03369-02, Revision 0, used inelastic acceptance limits for the pipe straps which connect the pipe to the pipe support, in order to demonstrate Class I compliance which was not in accordance with the design and licensing basis specification. The license entered the issue into their Corrective Action Programas CR-PLP-2017-05246, Spray Pipe Support,dated November 14, 2017. The licensee performed an analysis to establish reasonable assurance of operability and the inspectors with support from the Office from the Nuclear Reactor Regulation reviewed this operability and no performance deficiencies were identified.The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity Cornerstone attribute of design control and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public fromradionuclide releases caused by accidents or events. This finding is of very-low safety significance (Green) because there was no actual reactor containment barrier degradation. The inspectors did not identify a cross-cutting aspect associated with thisfinding because this was a legacy design issue; and therefore, was not reflective of current performance.
05000255/FIN-2017007-012017Q4PalisadesFailure to Periodically Test the Emergency Diesel Generators Capacity to Start and Accelerate Design Basis Sequenced LoadsThe team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations, Part50, Appendix B, Criterion XI, Test Control, for the failureto periodically test the emergency diesel generators(EDGs) capability to start and accelerate all of the sequenced loads within the applicable design voltage and frequency transient and recovery limits.Specifically, EDG testingactivities did not demonstrate that all of the EDG auto-sequenced loads started and accelerated within the applicable voltage and frequency limits during start-up and recovery. In addition, the licensee did not perform adequate post-modification testing after replacing the EDG governor controller system or voltage regulators. Thelicensee captured theseissuesin their Corrective Action Programas Condition Report (CR)2017-05265 and CR 2017-05283, and performed an operability evaluation which reasonably determined the affected structures, systems, and componentswere operable.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems thatrespond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) becauseit did not result in the loss ofoperability or functionality of mitigating systems. Specifically, the licensee evaluated the most recent voltage and frequency data from the last EDG output breaker testsin which the data recorder was left running after the output breaker shut and reasonably determined that the EDGs and the affected loads were operable. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the associated testingprocedures were established more than 3years ago.
05000266/FIN-2017002-022017Q3Point BeachFailure to Identify Non-Conforming Conditionsafter Receipt of Anchor Darling Double Disc Gate Valve Related Part 21 ReportThe inspectors identified a finding of very-low safety significance (Green), and an associated (NCV) of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action,for the licensees failure to identify a condition adverse to quality. Specifically, after receiving and reviewing the Flowserve 10 CFR Part 21 report, the licensee misunderstood the information provided and failed to identify 36 safety-related valves that were nonconforming. Of these 36 valves, 14 were identified as being susceptible to pin failure based on their torque setting, 6 of which had open or close safety functions. The licensee captured the inspectors concern in the CAP as AR 02212531, and AR 02212915. In addition, the licensee performed operability evaluations that concluded the affected valves remained operable.The performance deficiency was more-than-minor because it was associated with the equipment performance attribute of the Mitigating System and Initiating Event cornerstones, and adversely affected the cornerstone individual objectives. Using IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the finding screened as of very-low safety significance (Green) by answering No to the questions contained in Exhibit 1, and in accordance with Exhibit 2, it did not result in the loss of operability or functionality of mitigating systems. The team did not identify a cross-cutting aspect associated with this finding because the most significant cause for the error was not reflective of current performance. Specifically, the Part 21 report and associated review by the licensee occurred in February 2013.
05000440/FIN-2017003-012017Q3PerryLicensee-Identified ViolationTechnical Specification 5.5.1, states in part, that the ODCM shall contain the conduct of the Radiological Environmental Monitoring Program (REMP). The ODCM, Revision 20, includes Table 5.1 1 ODCM REMP Locations and Section 3.12.1.c, which states in part, With milk or broadleaf vegetation samples unavailable from one or more of the sample locations...identify specific locations for obtaining replacement samples and add them within 30 days to the Radiological Environmental Monitoring Program given in the ODCM. Contrary to the above, as of August 11, 2017, substantive changes to the REMP identified by the licensee in 2015 were not incorporated into the ODCM. Specifically, the licensee identified that a milk sampling location was no longer available and that the expansion of broadleaf vegetation sampling was required. Additionally, the licensee relocated collection sites for water and sediment samples that were not reflected in the ODCM. The licensee documented this issue in CR 2017 08353. The inspectors determined that this REMP issue was of very low safety significance (Green) after reviewing IMC 0609, Appendix D, Public Radiation Safety SDP, dated 23 February 12, 2008. The inspectors determined that this finding was associated with the Environmental Monitoring Program, therefore, the finding screened as Green (verylow safety significance).
05000266/FIN-2017007-012017Q3Point BeachFailure to Correct a Condition Adverse to Quality Associated with a Seismic Interaction of the Motor-Driven Auxiliary Feedwater PipingThe NRC identified a finding of very-low safety significance (Green) and an associated NCV of Title 10, Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to correct a Condition Adverse to Quality (CAQ) associated with a seismic piping interaction affecting the Motor Driven Auxiliary Feedwater (MDAFW) system. Specifically, the licensee identified a flange clearance to the Unit 1 MDAFW suction piping was nonconforming and captured it in the Corrective Action Program (CAP) as Action Request (AR) 01684524. However, the licensee closed the AR without correcting the CAQ. The licensee captured the inspectors concern in the CAP as AR 02212810 and performed an evaluation that reasonably concluded the MDAFW remained operable.The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external factors and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. 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 the stresses resulting from the seismic interaction would reasonably be bounded by the applicable stress operability limits. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance because the performance deficiency occurred more than 3 years ago. Specifically, the licensee closed AR 01684524 without correcting this CAQ on September 20, 2011.
05000373/FIN-2017009-012017Q2LaSalleAnchor Darling Double Disc Gate Valve 1E22-F004 and 2E22-F004 Pressed-FitCollar Related 10 CFR Part 50, Appendix B, Criterion III ViolationIn 1990, the licensee had reviewed and accepted the vendors weak link analyses that provided the upper torque and thrust limits for all safety-related ADDDGV in service at the station. This analysis documentedthat the 1E22-F004 and 2E22-F004 valve stems were the weak link valve components in the closing direction (i.e.,provided enough closing thrust, thevalve stems would be the firstcomponent to becomenonfunctional).Therefore, theclosed thrust limit forthe 1E22-F004 and 2E22-F004 valves was approximately 260,000 lbf. The licensee had set up the valves ina manner that would ensure that the valveswould have enough torque and thrust tooperate under design basis conditions while staying below the maximum weak link limits. Maintenance and test records showed that thelicensee consistently verifiedthat these two valves were setup and maintained within this design window. Typical as-found and as-left closed thrust limits ranged from approximately between200,000240,000 lbf.As described in the licensees failure analysis report and as discussed above, the licensee identified that the pressed-fitcollar could relax its pre-load when operating the valve well within the established maximum closed thrust limitations. The licensees failure analysis report estimated that approximately 130,000 lbf was necessary to shift the collar up and relax the pre-load. Therefore,theteam concluded that the licensees weak link analysis was inadequate based upon the 2E22-F004 valve failure and associated failure analysiswhich determined that the pressed-fitcollar was a weaker component as compared to the valve stem. The team did not identify an associated performance deficiencyfor the inadequate weak link analysis. This determination was based upon the weak link analysis originating from the vendor in 1990, licensees review of that analysis, and latent design issue that had not been previously identified within the industry until recently identified by the licensee.Additionally, the team did not identify a violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action. This determination was based, in part, that correcting the unknown stem collar pre-torque issueafter receiving the 10 CFR Part 21 Flowserve notification would not necessarily have identified and corrected the non-conforming inadequate weak link design control issue. Enforcement: Title10 CFR Part 50, Appendix B, Criterion III, Design Control, requires,inpart that measures shall be established to assure that applicable regulatory requirements and the design basis, as defined in 50.2, and as specified in the license application, for those structures, systems, and components to which this appendix apply are correctly translated into specifications, drawings, procedures, and instructions.Contrary to the above, since original plant construction, the licensee failed to ensure thatapplicable design basismaximumclosed thrust and torque valuesfor the safety-related Unit 1 and Unit 2 HPCS injection valves (1E22-F004, 2E22-F004)werecorrectly translatedinto specifications. Specifically, it was identified that the stem-to-wedgepre-torque credited within the design could relax by applying closed direction torque and thrust well within the specified design limitbecause that limit was based uponthe wrong weak link component. The loss of the stem-to-wedgepre-torque could subsequently break the wedge pin and result in stem-to-wedgethread degradation ultimately leading to valve failure.The NRC determined that issue was a Severity Level III Violation based upon Section6.1(c)(2) of the Enforcement Policy. Specifically, a system that is part of the primary success path and which functions or actuates to mitigate a design base accident or transient that either assumes the failure of or presents a challenge to the integrity of the fission product barrier not being able to perform its licensing basis safety function because it is not fully qualified.The NRC exercised enforcement discretion in accordance with Sections 3.10 of the Enforcement Policy and Section 3 of Part1 of the Enforcement Manual. Enforcement Policy Section 3.10 states that the NRC may exercise discretion for violations of NRC requirements by reactor licensees for which there are no associated performance deficiencies. This violation was entered into the Corrective Action Programas Issue Report3972901 and has been corrected by replacing the 1E22-F004 and 2E22-F004 valve stems with integral collars.
05000454/FIN-2017001-012017Q1ByronLicensee-Identified Violation

On March 11, 2017 , with Unit 1 shutdown and in a refueling outage, pipefitters as signed to cut out and replace service water valve 1WS413 discovered that piping was blocked upstream of the valve and the work scope was appropriately changed to remove the blocked piping. Taking action they believed was allowed by the work instructions, the pipefitters opened a pipe union and removed the pipe. They then set the removed section containing valve 1WS023C on a nearby tripod to continue work. A system engineer performing a walkdown in the area identified that the removed valve had a clearance (danger) tag on it and immediately stopped work and contacted the operations department. Technical Specification 5.4.1 requires , in part , that written procedures be established, implemented and maintained covering the procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. One administrative procedure recommended in Appendix A is , Equipment Control ( e.g. locking and tagging). OP AA 109 101, Clearance and Tagging, accomplished the locking and tagging requirement for Byron Station. Section 5.2, Danger Tags, established standards for implementation of the tagging process. Step 5.2.2 stated , A component with a Danger Tag attached to it shall not be physically removed from the system. Contrary to the requirements stated above, a component with a danger tag attached was physically removed from the system on March 11, 2017. Specifically, pipefitters disconnected a pipe union and removed associated service water piping from the system that contained valve 1WS023C which had a clearance (danger) tag attached.

The licensee immediately verified that the cooler the piping served was out -of-service on both the supply and return sides with a clearance boundary in place and drained so that the workers were not exposed to a pressurized sourc e. The workers immediately acknowledged their error stating they did not see the tag because they were focused on the demolition activities. The issue was entered into the licensees CAP as IR 03984215 , and the maintenance organization conducted a stand down to reinforce the station standards for compliance with the clearance procedure. The inspectors determined that this issue was more than minor because the performance deficiency adversely impacted the Configuration Control attribute of the Initiating Events Cornerstone objective to limit the likelihood of events that upset plant stability and challenge safety functions during shutdown operations. The inspectors determined the issue was of very low safety significance , or Green by answering No to all screening questions in IMC 0609, Appendix G, Shutdown Operations Significant Determination Process, Exhibit 2, Initiating Events Screening Questions.

05000461/FIN-2016009-012016Q4ClintonNon Conservative Control Room Radiological Habitability AssessmentThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to use a technically appropriate analytical methodology in the control room radiological habitability calculation. Specifically, the licensee used a methodology that inappropriately characterized the control room heating, ventilation and air-conditioning (HVAC) system outside air intake design resulting in a calculated control room dose following a loss of coolant accident that exceeded the applicable limit. The licensee captured this issue in their CAP as AR 02742442, completed an operability evaluation, and issued an NRC event notification. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in the control room expected dose following a loss of coolant accident to exceed the applicable limits prompting an operability evaluation. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected calculations were performed more than 3 years ago.
05000461/FIN-2016009-022016Q4ClintonFailure to Scope SFP Temperature and Level Instruments into the Maintenance Rule ProgramThe team identified a finding of very-low safety significance (Green) and an associated NCV of Paragraph (b)(2)(i) of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for the licensee failure to scope non-safety related mitigating structure, systems, and components (SSCs) used within an emergency operating procedure (EOP) into Maintenance Rule Program. Specifically, an EOP used spent fuel pool (SFP) low-level and high-temperature parameters as distinct entry criteria but the associated components were not included in the scope of the Maintenance Rule Program. The licensee captured the team concerns in their CAP as AR 02736193, performed an extent of condition to identify any other SSC addition to the EOPs requiring them to be added to the Maintenance Rule Program scope, and initiated plans to incorporate the affected SSCs into the Maintenance Rule Program scope. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of SSC performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, a key aspect of the Maintenance Rule is to ensure that maintenance activities are performed in a manner that provide reasonable assurance that SSCs within its scope perform reliably and are capable of providing their intended Maintenance Rule function(s). In the case of the SFP temperature instruments, the licensee was not performing preventive maintenance to ensure that degradation, such as instrument drift, did not adversely affect their ability to detect and alarm EOP entry conditions such that mitigating actions could be implemented to preserve secondary containment. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of human performance because the licensee did not use a systematic process for evaluating and implementing changes when updating the affected EOP in 2015.
05000461/FIN-2016009-042016Q4ClintonFailure to Verify the Adequacy of Design Assumptions Related to Time Critical Operator ActionsThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensee failure to verify the adequacy of design assumptions related to time critical operator actions made in calculations associated with the control room HVAC and RHR emergency SFP cooling functions. Subsequently, it was determined that operators did not fully understand the control room HVAC system operational demands and that the operational assumptions of the RHR emergency SFP cooling design were unrealistic. The licensee captured these issues into the CAP as AR 02739012, AR 03943566, and AR 02741909; reasonably demonstrated that SFP makeup sources would be available to cope with a prolonged loss of SFP cooling; conducted operator training; and provided refined procedural guidance to ensure the control room HVAC system would be operated consistent with the design assumptions. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the pilot validations of the control room HVAC system operational assumptions demonstrated a significant reduction in margin due to, in part, a lack of operator understanding of the operational assumptions. Additionally, a preliminary review of procedures associated with SFP cooling and RHR determined the operational assumptions of the calculation related to RHR emergency SFP cooling were not bounding. The team determined that this finding was of very low safety significance (Green). Specifically, the control room HVAC system finding example only represented a degradation of the radiological barrier function provided for the control room in that it did not affect the control room barrier function against smoke or a toxic atmosphere. In addition, the finding example related to emergency SFP cooling did not cause SFP temperature to exceed the maximum analyzed limit, a detectible release of radionuclides, water inventory to decrease below the analyzed limit, or an adverse effect to the SFP neutron absorber or fuel loading pattern. The team determined that the finding had a cross-cutting aspect in the area of Human Performance because the operation and engineering organizations did not effectively communicate and coordinate their respective roles in developing the control room HVAC system validation in a manner that supported nuclear safety.
05000461/FIN-2016009-052016Q4ClintonFailure to Promptly Identify that the Incapability of the RHR Design to Support TS Operability Requirements Was a CAQThe team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensee failure to promptly identify that the incapability of the residual heat removal (RHR) design to support Technical Specifications (TS) operability requirements was a condition adverse to quality. Specifically, when reactor water temperature was greater than 150 degrees Fahrenheit, RHR could not be realigned from shutdown cooling mode of operations to provide the TS required functions of the emergency core cooling system, suppression pool cooling, containment spray, and feedwater leakage control system. The licensee captured this issue in their Corrective Action Program (CAP) as Action Request (AR) 02742439 and AR 03948042, and planned to submit a License Amendment Request to align TS requirements with the design capabilities. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and adversely affected the associated cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the performance deficiency resulted in voluntarily declaring TS functions inoperable when performing shutdown cooling operations, which did not ensure the associated mitigating systems availability or capability to respond to an initiating event. The team determined that this finding was of very low safety significance (Green). Specifically, there were no known instances where the finding: (1) represented a loss of system safety function; (2) represented an actual loss of safety function of at least a single train or two separate safety systems out-of-service for greater than their TS allowed outage time; (3) involved non-TS trains of equipment; (4) involved a degradation of a functional RHR auto-isolation on low reactor vessel level; (5) impacted external event protection; or (6) involved fire brigade issues. The team did not identify a cross-cutting aspect associated with this finding because it did not reflect current licensee performance since the performance deficiency occurred more than 3 years ago.
05000461/FIN-2016009-062016Q4ClintonFailure to Follow the Operability Determination Process Following the Identification of a Control Room HVAC System Design IssueThe team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instruction, Procedures, and Drawings, for the licensee failure to follow the operability evaluation procedure after the identification of a significant design error associated with the control room HVAC system. Specifically, the licensee did not identify the affected safety function, and promptly restore or confirm system operability. The licensee captured these issues into the CAP as AR 03948266 and performed a preliminary engineering evaluation using another alternative analytical methodology that reasonably determined the control room HVAC system remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of human performance and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, the performance deficiency resulted in a condition where reasonable doubt on the operability of the control room HVAC system remained following the identification of a significant design error. The finding screened as of very-low safety significance (Green) because it only represented a degradation of the radiological barrier function provided for the control room. Specifically, the finding did not affect the control room barrier function against smoke or a toxic atmosphere. The team identified that the finding had a cross-cutting aspect in the area of Human Performance because the licensee did not provide training to maintain a knowledgeable workforce that would facilitate an adequate implementation of the operability evaluation process following the identification of a non-conforming design-related issue.
05000461/FIN-2016009-032016Q4ClintonFailure to Amend the UFSAR Indicating Choice to Comply with 10 CFR 50.68(b)The team identified a Severity Level-IV NCV of 10 CFR 50.68, Criticality Accident Requirements, Paragraph (b)(8), for the licensee failure to amend the Updated Final Safety Analysis Report (UFSAR) to indicate they chose to comply with 10 CFR 50.68(b). Specifically, in 2005, the licensee chose to comply with 10 CFR 50.68(b) but did not amend the UFSAR following the issuance of the associated license amendment. The licensee captured this issue in their CAP as AR 02741851, reasonably confirmed compliance with 10 CFR 50.68(b) requirements (1) through (7) was maintained, and initiated plans to update the UFSAR to specifically indicate that Clinton Power Station chose to comply with 10 CFR 50.68(b). The Significance Determination Process does not specifically consider the impact to the regulatory process in its assessment of licensee performance. Therefore, it was necessary to address this violation, which potentially impacts the NRCs ability to regulate, using traditional enforcement to adequately deter non-compliance. Specifically, failure to update the UFSAR challenges the regulatory process because it serves as a reference document used, in part, for recurring safety analyses, evaluating License Amendment Request, and in preparation for and conduct of inspection activities. The team determined the traditional enforcement violation was a Severity Level-IV violation in accordance with Section 6.1.d.3 of the Enforcement Policy because the un-updated UFSAR had not been used to evaluate a facility or procedure change that resulted in a condition evaluated as having low-to-moderate or greater safety significance by the Significance Determination Process. However, it had a material impact on safety or licensed activities. Specifically, the un-updated UFSAR could be used to perform evaluations of facility or procedure changes, which would have the potential to result in unacceptable conditions and/or regulatory decisions. Traditional enforcement violations are not assessed for cross-cutting aspects.
05000341/FIN-2016007-012016Q3FermiInadequate Procedure for Addressing Non-Functional MDCT Fan Motor Brake SystemThe team identified a finding of very-low safety significance (Green) and an associated NCV of Title 10 of the Code of Federal Regulations (CFR), Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensee failure to establish procedures that were appropriate for addressing non-functional mechanical draft cooling tower (MDCT) fan motor brakes. Specifically, a license procedure contained instructions for addressing the impact of non-functional MDCT fan motor brakes to the ultimate heat sink operability that were inconsistent with the applicable Technical Specification requirements. The licensee captured this issue in their Corrective Action Program (CAP) as CARD 16-26762, verified that all MDCT fan brake systems were functional, revised the affected procedure to restore compliance, and issued a night order to notify control room licensed nuclear operators of the revised procedure. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a tornado event. Specifically, a historic review for the last 12 months revealed that the minimum required number of MDCT fans remained operable to mitigate the consequences of a tornado. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the procedure instructions for addressing MDCT fan motor brake non-functionality were developed more than 3 years ago.
05000341/FIN-2016007-032016Q3FermiFailure to Verify the Ability to Manually Throttle Safety-Related MOVs during a DBAThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the capability to manually throttle safety-related motor-operated valves (MOVs) during a DBA. Specifically, the licensee did not verify that the protective devices would allow manually throttling safety-related MOVs during a DBA without tripping. The licensee captured this issue in their CAP as CARD 16-26763, performed a preliminary protective device evaluation to reasonably determine the maximum number of throttling cycles each MOV can incur without tripping the associated thermal overload, and incorporated these limits into an operations night order. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed informal analyses to evaluate the installed protective devices for the throttling MOVs and reasonably determined that tripping would not occur. The team determined that the associated finding had a cross-cutting aspect in the area of Human Performance because work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the engineers that performed the affected calculation, which was approved on December 2013, did not communicate and coordinate with operations or the MOV engineer to determine if the plant had throttling MOVs that required additional analysis.
05000341/FIN-2016007-042016Q3FermiFailure to Periodically Test the EDG Capability to Start and Accelerate All of the Sequenced Loads Within the Applicable LimitsThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the failure to periodically test the emergency diesel generator (EDG) capacity to start and accelerate all of the sequenced loads within the applicable limits. Specifically, surveillance requirement (SR) activities did not demonstrate that all of the EDG auto-sequenced loads started and accelerated within the applicable voltage and frequency limits during start-up and recovery. In addition, the licensee did not timely evaluate the surveillance data collected for the residual heat removal pump motors. The licensee captured this issue in their CAP as CARD 16-26535 and CARD 16-26536, and performed an operability evaluation which reasonably determined the affected systems, structures, and components (SSCs) were operable but nonconforming. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee evaluated the most recent data and reasonably determined that the EDGs and the affected loads were operable. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the associated SR procedures were established more than 3 years ago.
05000341/FIN-2016007-052016Q3FermiFailure to Leak Test All Division 2 NIAS Boundary Isolation ValvesThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, for the licensees failure to leak test all Division 2 non-interruptible control air system (NIAS) boundary isolation valves. Specifically, the periodic NIAS leak testing did not account for the potential leakage of two valves used to isolate the NIAS safety-related system from the nonsafety-related interruptible control air system. The licensee captured this issue in their CAP as CARD 16-26389 and performed an operability evaluation which reasonably determined that Division 2 of NIAS remained functional. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, it was associated with the Barrier Integrity cornerstone attribute of SSC and barrier performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee used available data from a recent event and reasonably determined that system out-leakage was within the design limit. In addition, with respect to the Barrier Integrity cornerstone, the finding only represented a potential degradation of the control room and standby gas ventilation systems. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the test procedures were established more than 3 years ago.
05000341/FIN-2016007-062016Q3FermiFailure to Ensure that the MSIVs Would Close Within the TS Required Timeframe and as Described in the UFSARThe team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the main steam isolation valves (MSIVs) would close within the Technical Specification time requirements and with the motive forces described in the Updated Final Safety Analysis Report. Specifically, the SR procedures did not account for the steam flow closing force, accumulator pressure variances, and containment pressure when verifying that the MSIVs will close within the SR time acceptance criteria. In addition, the licensee had not demonstrated that the MSIVs would close with air pressure and/or spring force against peak containment pressure as described in the Updated Final Safety Analysis Report. The licensee captured this issue in their CAP as CARD 16-27189 and CARD 16-26697, and performed evaluations that reasonably determined the affected MSIVs remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. In addition, it was determined to be more-than-minor because it was associated with the Initiating Event cornerstone attribute of design control and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. The finding screened as of very-low safety significance (Green) because it did not result in exceeding the reactor coolant system leak rate for a small LOCA or affected other systems used to mitigate a LOCA. In addition, it did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components, and it did not involve an actual reduction in the function of hydrogen igniters in the reactor containment. The team did not identify a cross-cutting aspect associated with this finding because it was not reflective of current performance. Specifically, the most significant cause for the performance issues discussed had existed for at least 3 years.
05000341/FIN-2016007-082016Q3FermiInadequate Containment Debris Inspections Acceptance CriteriaThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to establish procedures that were appropriate to inspect containment debris. Specifically, the emergency core cooling system (ECCS) suction strainer and containment coating inspection procedures contained acceptance criteria that was inconsistent with the applicable design documents. The licensee captured this issue in their CAP as CARD 16-26128 and CARD 16-26585, and reasonably determined that the concern did not impact the affected SSCs functionality based on recent inspection results. The performance deficiency was determined to be more-than-minor because it was associated with the procedure quality attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, a review of recent inspection did not find a condition that reasonably challenged the applicable design analysis and all loose material identified during the inspections was removed. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the affected procedures were established more than 3 years ago.
05000341/FIN-2016007-092016Q3FermiFailure to Evaluate the Acceptability of Drywell Coatings with Respect to Potential ECCS Suction Strainer BlockageThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to evaluate the acceptability of drywell coatings with respect to potential ECCS suction strainer blockage. Specifically, the licensee had not ensured that coating Carbo Zinc 11 would remain attached to the base metal during a DBA and the ECCS suction strainer calculations did not account for this material as a potential source of debris blockage. The licensee captured this issue in their CAP as CARD 16-26581 and reasonably determined that the affected coating system would remain adhered during a LOCA by comparing Carbo Zinc 11 installation documents against DBA test reports for this coating. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee compared Carbo Zinc 11 installation documents against DBA test reports for this coating and reasonably concluded that this coating system would remain adhered in the event of a LOCA. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the associated evaluations were performed more than 3 years ago.
05000341/FIN-2016007-102016Q3FermiNon Conservative ECCS Suction Strainer Min-K Combined Generation and Transport FactorsThe team identified a finding of very-low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to use the min-K insulation debris generation and transport factors contained in the ECCS suction strainer licensing basis. Specifically, the licensee used non-conservative min-K insulation debris generation and transport factor values. The licensee captured this issue in their CAP as CARD 16-26800 and performed an operability evaluation that reasonably determined, based on industry test data, the existing calculation had sufficient conservatism to accommodate the effects of the additional debris volume. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation and reasonably determined that the existing calculation had sufficient conservatism to accommodate the effects of the additional debris volume. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the associated evaluations were performed more than 3 years ago.
05000341/FIN-2016007-112016Q3FermiFailure to Apply Design Control Measures to a Design Change Associated with NIAS Accumulator CapabilityThe team identified a finding of very low safety significance (Green), and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to apply design control measures to a design change associated with NIAS accumulator capacity. Specifically, the licensee did not verify that the reduced accumulator capacity was adequate during the entire time period that the compressor is expected to not be running, and ensure that operability limits and calibration tolerances contained in procedures were consistent with the new design. The licensee captured this issue in their CAP as CARD 16-26208, CARD 16-26561, and CARD 16-26607, and reasonably concluded that NIAS remained functional. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. In addition, it was associated with the Barrier Integrity cornerstone attribute of SSC and barrier performance and affected the cornerstone objective of providing reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed a bounding assessment that reasonably determined that the accumulator would maintain adequate pneumatic supply. In addition, with respect to the Barrier Integrity cornerstone, the finding only represented a potential degradation of the control room and standby gas ventilation systems. The team determined that the associated finding had a cross-cutting aspect in the area of Human Performance because the licensee did not carefully guarded margins and changed them only through a systematic and rigorous process. Specifically, the licensee failed to review and identify all of the design attributes associated with NIAS system before significantly reducing the accumulator capacity design margin in February 2016.
05000341/FIN-2016007-122016Q3FermiFailure to Identify an Out-of-Specification Pressure Reading on the Nitrogen Supply to the A MDCT Fan Motor Brake SystemThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify that the A MDCT fan motor brake system 100 psi nitrogen supply cylinder pressure did not meet the low-pressure acceptance criterion. Specifically, although the license had discovered this condition adverse to quality (CAQ), it was not captured into the CAP and was not corrected for a period of 7 consecutive days following its discovery. The licensee captured this issue in their CAP as CARD 16-26214, verified that the pressure of all MCDT fan motor brake cylinders were within limits, evaluated past operability, and performed a causal investigation. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of protection against external events and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not involve the loss or degradation of equipment or function specifically designed to mitigate a tornado event. Specifically, the licensee reviewed the pressure readings of the other nitrogen system supply cylinders and reasonably determined that their available pressure at the time would have compensated for the 100 psi cylinder low-pressure. The team determined that the associated finding had a cross-cutting aspect in the area of Human Performance because work groups did not communicate and coordinate their activities within and across organizational boundaries to ensure nuclear safety is maintained. Specifically, the nuclear operators and the control room licensed nuclear operators did not communicate and coordinate their activities to ensure the degraded condition was captured in the CAP.
05000341/FIN-2016007-142016Q3FermiFailure to Identify that an Inadequate Minimum MSIV Accumulator Air Pressure Setpoint Was CAQThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify that an inadequate minimum MSIV accumulator air pressure setpoint was CAQ. Specifically, a licensee engineering evaluation concluded that the minimum MSIV accumulator air pressure setpoint was inadequate but the condition was not captured in the CAP and, as a result, corrective actions were not implemented. The licensee captured this issue in their CAP as CARD 16-26697 and reasonably determined the MSIVs remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Barrier Integrity cornerstone attribute of design control and affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. The finding screened as of very-low safety significance (Green) because the finding did not represent an actual open pathway in the physical integrity of reactor containment, containment isolation system, or heat removal components, and it did not involve an actual reduction in the function of hydrogen igniters in the reactor containment. Specifically, the finding did not result in an actual open pathway and the MSIVs do not affect the function of heat removal components and hydrogen igniters. The team did not identify a cross-cutting aspect associated with this finding because it was not confirmed to reflect current performance due to the age of the performance deficiency. Specifically, the calculation that concluded that the minimum air pressure setpoint was inadequate was performed in 1997.
05000341/FIN-2016007-152016Q3FermiFailure to Identify that a Non-Conservative Min-K Insulation Volume Calculation Error Was Nonconforming to the ECCS Suction Strainer Licensing BasisThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify that a non-conservative min-K insulation volume calculation error was nonconforming to the ECCS suction strainer licensing basis. Specifically, the licensee identified the non-conservative calculation error and captured it in the CAP as CARD 11-21153. However, the licensee did not identify any regulatory basis requiring this condition to be addressed and, as a result, closed the CARD without correcting the CAQ. The licensee captured this issue in their CAP as CARD 16-26292 and CARD 16-26800, and performed an engineering functional assessment that reasonably determined the affected SSCs remained operable. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of equipment performance and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation and reasonably determined the affected SSCs remained operable. The team determined that this finding had a cross cutting aspect in the area of Human Performance because the licensee did not propose an action that was determined to be safe in order to proceed, rather than unsafe in order to stop. Specifically, the licensee determined that no regulatory basis was associated with the non-conservative error because they could not find any requirement that specifically described the physical configuration and condition addressed in CARD 11-21153 when evaluating the problem in 2015.
05000341/FIN-2016007-162016Q3FermiFailure to Timely Identify, Document, and Evaluate Conditions that Challenge OperabilityThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, for the licensees failure to timely identify, document, and evaluate nonconforming conditions that called the operability of one or more SSCs into question. Specifically, the licensee was not timely in capturing and evaluating ten CAQs identified during this inspection in their CAP and in accordance with their procedures, which resulted in untimely operability determinations. The licensee captured this issue in their CAP as CARD 16-26633, CARD 16-26776, CARD 16-26534, and CARD 16-26678, and completed the associated operability determinations, which reasonably determined the affected SSCs remained operable. The performance deficiency was determined to be more-than-minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed operability evaluations that reasonably determined that all of the affected SSCs remained operable. The team determined that this finding had a cross cutting aspect in the area of Human Performance because the licensee did not use a consistent, systematic approach to make decisions. Specifically, the licensee did not use the CAPs systematic process to identify CAQs and make timely and adequate prompt operability decisions.
05000341/FIN-2016007-022016Q3FermiFailure to Verify the Adequacy of the Voltage Supplied to Transformer #64 Load Tap ChangerThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the failure to verify the adequacy of the voltage supplied to the transformer #64 load tap changer. Specifically, the licensee did not perform calculations to verify that the load tap changer controls and actuator would have adequate voltage to be able to reset the degraded voltage relays following a design basis accident (DBA). The licensee captured this issue in their CAP as CARD 16-26702 and performed an operability evaluation that reasonably determined the voltage would marginally be acceptable to operate the load tap changer controls and actuator. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation that reasonably showed voltage would be marginally acceptable to operate the load tap changer controls and actuator when required during a DBA. The team determined that the associated finding had a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not conduct a self-critical and objective assessment of its programs and practices. Specifically, the licensee reviewed the applicability of a similar violation issued to a different licensee during the 2015 Component Design Bases Inspection Self-Assessment and concluded that it did not apply to Fermi.
05000341/FIN-2016007-072016Q3FermiFailure to Ensure that Protective Devices for the Loads Required at the Beginning of a LOCA Would Not Trip Under Degraded Voltage ConditionsThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion III, Design Control, for the licensees failure to ensure that the protective devices for the loads required at the beginning of a loss of coolant accident (LOCA) would not trip under degraded voltage conditions. Specifically, the licensee did not verify that the connected Class 1E loads would not be damaged or become unavailable during a LOCA concurrent with a degraded voltage condition between the degraded voltage dropout setting and the loss of voltage setting for the degraded voltage time delay of 7.3 seconds and subsequent reconnection to the EDG. The licensee captured this issue in their CAP as CARD 16-26533 and performed a preliminary evaluation that reasonably determined the protective devices would not actuate during this condition. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation and reasonably determined that protective devices would not actuate during a degraded voltage concurrent with a LOCA. The team determined that the associated finding had a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not conduct a self-critical and objective assessment of its programs and practices. Specifically, the licensee evaluated a similar violation issued at a different licensee during the 2016 Component Design Bases Inspection Self-Assessment and concluded that no corrective actions were required.
05000341/FIN-2016007-132016Q3FermiFailure to Identify a CAQ Associated with Over-Dutied 480V Safety-Related Switchgear BreakersThe team identified a finding of very-low safety significance (Green) and an associated NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the licensees failure to identify that over-dutied 480V safety-related switchgear breakers were nonconforming to the licensing basis. Specifically, the licensee did not identify that this condition was nonconforming to the licensing basis and, as a result, did not promptly correct the CAQ. The licensee captured this issue in their CAP as CARD 16-26209 and CARD 16-26210, and performed an operability evaluation that reasonably determined the affected buses were operable but nonconforming. The performance deficiency was determined to be more-than-minor because it was associated with the Mitigating Systems cornerstone attribute of design control and affected the cornerstone objective of ensuring the availability, reliability, and capability of mitigating systems to respond to initiating events to prevent undesirable consequences. The finding screened as of very-low safety significance (Green) because it did not result in the loss of operability or functionality of mitigating systems. Specifically, the licensee performed an operability evaluation and reasonably concluded that the associated buses remained operable. The team determined that the associated finding had a cross-cutting aspect in the area of Problem Identification and Resolution because the licensee did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the licensee failed to recognize that the condition was nonconforming to the licensing basis because they did not thoroughly evaluate the discovery of the over-dutied breakers and extent of condition.
05000456/FIN-2016007-012016Q3BraidwoodIdentification of SCAQs in Accordance with the QATRThe team identified an Unresolved Item (URI) regarding the identification of significant conditions adverse to quality (SCAQs) in the CAP. Specifically, the team determined that the CAP, as implemented by PI-AA-125,Corrective Action Program, and PI-AA-120, Issue Identification and Resolution, appeared to not ensure that SCAQs were appropriately identified and corrected to prevent recurrence. Chapter 16 of the Braidwood Quality Assurance Topical Report (QATR) describes the licensees program to identify and correct conditions adverse to quality. Procedure PI-AA-125 implemented the requirements established in the QATR. During this inspection, the team reviewed the CAP procedure to determine how it ensured that SCAQs were identified and resolved. As part of this review, the team requested a copy of identified SCAQs over the last two years and were subsequently informed that none had been identified. Issue #1 - The team reviewed the QATR and noted that the following requirements applied: Section 2.1 stated that measures are required to assure that the cause of any significant condition adverse to quality is determined and that corrective actions to prevent recurrence (CAPRs) are implemented. Section 2.2.1, Significant Conditions Adverse to Quality, stated that in cases of significant conditions adverse to quality the cause of the condition must be determined and documented, the resolution determined and documented, and the corrective actions taken and documented to prevent recurrence. Step 2.116 of Appendix D of the QATR defined a significant condition adverse to quality as, a condition, which if left uncorrected, could have a serious effect on safety or operability. The team reviewed procedure PI-AA-125 and PI-AA-120, which delineated the process for the identification and screening of issues, and identified that these procedures did not include a provision to classify an identified issue as a SCAQ. The team also noted that the definition of a SCAQ was not being used to determine whether a RCE was needed; therefore, a CAPR did not appear to be directly associated with a SCAQ. Based on the above, the team questioned whether CAP procedure PI-AA-125 prescribed a process through which SCAQs were identified and documented, and corrective actions taken and documented to prevent recurrence as required by the QATR. The team discussed this issue with the licensee. The licensee stated that since the terms SCAQ and condition adverse to quality (CAQ) were not explicitly defined in NRC regulations, that they had created a graded approach of significance level and likelihood (which included risk and uncertainty) to ensure that items were properly dispositioned and the level of resources and rigor applied appropriately followed the CAP governance. The licensee further stated that the graded approach, along with a well-trained management team that has nuclear safety and conservative decision-making as their primary focus, provided for an effective CAP. Finally, the licensee stated that even if a CAPR was not issued, that CAs would prevent recurrence of the events entered into the CAP. The team questioned whether a CAPR and a CA would be equally effective as corrective actions to prevent the recurrence of issues dispositioned in the CAP. The licensee agreed that the two types of CAs were treated differently. For example, 1) the MRC was required to assess changes to the intent of a CAPR, which was not required for a CA, 2) an effectiveness review may not necessarily be assigned if an issue was corrected using only a CA, and 3) if there was a desire to suspend or modify a previously implemented CAPR, then a risk analysis and MRC concurrence was necessary; which was not the case for a CA. At the end of the inspection it was not clear how procedures PI-AA-120 and PI-AA-125 ensured that SCAQs were identified and documented, and corrective actions taken and documented to prevent recurrence. Additionally, it was not clear if the licensees process implemented the requirements in the QATR. Resolution of this issue will be based on additional NRC review to determine if a violation of NRC requirements occurred. Issue #2 - The team identified an example of a potential SCAQ for which the licensee implemented CAs that failed to prevent the issue from recurring. Specifically, for a December 30, 2013 oil leak on the inboard bearing housing of the Unit 1 Train B (1B) SX pump, the licensees CAs restored operability, but were not adequate to prevent recurrence and consequently an oil leak recurred on November 18, 2014. Both of these oil leaks resulted in the licensee declaring the 1B SX pump inoperable and required entry into Technical Specification (TS) Limited Condition for Operation (LCO) 3.7.8 (reference Non-Cited Violation (NCV) 05000456/201400502; Failure to Correct Undersized Essential Service Water Pump Bearing Casing Drain Line Resulted in System Inoperability). The team questioned whether the oil leaks on the inboard pump bearing housing of the 1B SX pump should have been categorized as a SCAQ as defined in the licensees QATR. Specifically, QATR Section 2.116, Definitions, defined a SCAQ as, A condition, which if left uncorrected, could have a serious effect on safety or operability. In this case, although the oil leakage at the inboard pump bearing housing first identified in 2013 was specifically addressed through repairs, the CAs were not adequate to prevent recurrence and a second oil leak occurred in 2014 that caused a serious effect on the operability of the 1B SX pump (i.e. rendered the 1B SX pump inoperable). Additionally, the team considered this issue to have a potentially serious effect on operability, because if left uncorrected the oil leakage would have depleted the oil supply reservoir resulting in a loss of lubrication to the pump shaft bearings that could damage the pump shaft and require substantial repairs to return the pump to operation. The team discussed this issue with the licensee. The licensees response was that because there was no potential for common cause failure, and there was no significant change to plant risk after removing the 1B SX pump from service, the events discussed above were appropriately screened as Significance Level 3 issues. The licensee also stated that a SCAQ would typically be assigned for a Significance Level 1 or 2 issue, but even if an issue was assigned this level of significance, it would not necessarily be categorized as a SCAQ. At the end of the inspection it was not clear how the definition of SCAQ in the QATR was utilized in the CAP. Resolution of this issue will be based upon additional NRC review and a determination of whether the failure of the 1B SX pump constituted a SCAQ as defined in the QATR.
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