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Watts Bar Nuclear Plant, Unit 2 - Transmittal of Initial Startup Report to the United States Nuclear Regulatory Commission
	ML16315A334
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Site:	Watts Bar
Issue date:	11/10/2016
From:	Simmons P R; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
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Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000November 10, 201610 cFR 50.4ATTN: Document Contro! DeskU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001Watts Bar Nuclear Plant, Unit 2Facility Operating License No. NPF-96NRC Docket No.50-391Subject Watts Bar Nuclear Plant, Unit 2, Transmittal of lnitial Startup Report tothe United States Nuclear Regulatory Gommissionln accordance with the requirements of the Watts Bar Nuclear (WBN) Plant, Dual UnitUpdated Final Safety Analysis Report (UFSAR), Chapter 14.2.6, 'Test Records," theTennessee Valley Authority (TVA) is submitting the lnitial Startup Report for WBN Unit 2.The enclosure to this letter provides the Final Startup Report for WBN Unit 2. lnitial fuelload, pre-critical testing, initial criticality, and low power physics testing, and powerascension testing are discussed in separate sections of the report. The report details thetest objectives, methodology, test results, and problems noted for each of the testsperformed.The test objectives and methodology were developed using the graded approach based oncriteria provided in Regulatory Guide (RG) 1.68, "lnitial Test Programs for Water-CooledNuclear Power Plants,' Revision 2. RG 1.68 was further utilized for the selection of plantstructures, systems, and components (SSCs), and design features to be included in the testprogram. During the power ascension testing program, power ascension tests, surveillanceinstructions, and other permanent plant tests and technical instructions were performed todemonstrate satisfactory operation of SSCs.The report addresses test activities and the results of tests performed during the periodNovember 2015 through October 2016. The following table provides a summary of the keyWBN Unit 2 milestones and associated dates.

U.S. Nuclear Regulatory CommissionPage 2November 10, 2016The WBN Plant Operations Review Committee has reviewed the report.There are no new regulatory commitments made in this letter. Please address anyquestions regarding this submittal to Gordon P. Arent d (d'23) 365-2004.Respectfully,@rM;Paul R. SimmonsSite Vice President, Watts Bar Nuclear Plant

Enclosure:

lnitial Startup Report to the Nuclear Regulatory Commission,Facility Operating License No. NPF-96, NRC Docket No. 50-391,Final Report November 2015 through October 2016cc (Enclosure):NRC RegionalAdministrator - Region llNRC Senior Resident lnspector - Watts Bar Nuclear PlantNRC Project Manager - Watts Bar Nuclear PlantWatts Bar Nuclear Plant, Unat2 - Milestone ActivitiesMilestoneDateWBN Unit 2 Facility Operating License, NPF-96October 22,2015lnitial Fuel Load CommencementDecember 4, 2015lnitial CriticalityMay 23,2016Test Plateau, 30o/o Reactor Thermal Power (RTP)June 16, 2016Test Plateau, 5Ao/o RTPJuly 16, 2016Test Plateau, 75o/o RTPJuly 29,2016Test Plateau, 90o/o RTPAugust 29,2016Test Plateau, lOOo/o RTPOctober 6 ,2016Commercial OperationOctober 19, 2016 WATTS BAR NUCLEAR PLAhITUNIT 2INTTIAL STARTUP REPORTTO THEUNITED STATESNUCLEAR REGULATORY COMMISSIONAPPROVAL SHEETPower Ascension Test Manager:TRG Meeting No./06TRG ChairmaniPlant Manager:! lo"a/:

TENNESSEE VALLEY AUTHORITYWATTS BARNUCLEARPLANTUNIT 2INITIAL STARTUP REPORTTO THEUNITED STATESNUCLEAR REGULATORY COMMISSIONFACILITY OPERATING LICENSE NO. NPF'.96NRC Docket No. 50-391Final ReportNovember 2015 through October 2016 TABLE OF CONTENTSSECTION1.02.03.04.05.0PAGE NUMBERLIST OF FIGURES ............... ivLIST OF ACRONYMS.............. ............ viiTNTRODUCTTON ...................1POWER ASCENSTON TEST PROGRAM (PATP) OVERVIEW ................32.1 Administration of the Program ............. .........32.2 lmplementation of the Pro9ram.............. .......62.3 Summary ....................8WATTS BAR UNIT 2 STARTUP CHRONOLOGY... ..............10INITIAL FUEL LOAD ..........254.1 Overview and Summary of lnitial Core Loading ...........254.2 lnitial Core Loading Sequence (2-PAT-2.0)........ .........264.3 Reactor System Sampling for Core Load (2-PAT-2.1)......... .........284.4 Response Check of Core Load Instrumentation After 8 Hour Delayin Fuel Movement (2-PAT-2.2)............... .....314.5 Pre-Power Escalation NIS Calibration Data (2-PET-102)..............................334.6 lnitial Core Loading (2-PET-105) ................35PRECRITTCAL TEST!NG.......... ...........415.1 Post Core Loading Precritical Test Sequence(2-PAT-3.0)............... .................415.2 Control Rod Drive Mechanism Timing and CERPI lnitial Calibration(2-PAT-3.1)............... .................4e5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-PAT-3.2)............... .................565.4 Rod Control and Rod Position lndication (CERPI) (2-PAT-3.4).....................615.5 Reactor Coolant Flow Coastdown (2-PAT-3.7)............... ..............695.6 Rod Drop Time Measurement and Stationary Gripper Release Timing(2-PAT-3.8)............... .................71al-TABLE OF CONTENTS (continued)SECTION5.0 PRECRITICALTESTING(continued)6.07.0PAGE NUMBER5.7 Reactor Trip System (2-PAT-3.10) .............755.8 Adjustment of Steam Flow Transmitters at Minimal Flow(2-PAT-3.11)............. .................7e5.9 Control Rod Drive Mechanism Timing (2-PET-106)............. .........81INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING ...........846.1 lnitial Criticality and Low Power Test Sequence (2-PAT-4.0).........................846.2 Reactivity Computer (ADRC) (2-PET-103) ............ ......866.3 lnitial Criticality and Low Power Physics Testing (2-PET-201).......................88POWER ASCENSTON TESTING ............. .............957.1 Test Sequence for 30% Plateau (2-PAT-5.0)......... ......957.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1).....................997.1.2 Automatic Steam Generator Level Control Transients atLow Power (2-PAT-5.3)............ ...1087.1.3 Calibration of Steam and Feedwater Flow lnstruments at30% Power (2-PAT-5.4) ............ ...................1167.2 Test Sequence for 50% Plateau (2-PAT-6.0)......... ....1207.2.1 Turbine Generator Trip with Coincident Loss of Offsite PowerTest (2-PAT-5.2)....... ...................1247.2.2 Automatic Reactor Control System (2-PAT-6.1)............... ..............1297.2.3 Automatic Steam Generator Level Control Transientsat 50% Power (2-PAT-6.2) ............... ............1417.2.4 Calibration of Steam and Feedwater Flow lnstrumentsat 50% Power (2-PAT-6.3) ............... ............1467.3 Test Sequence for 75o/o Plateau (2-PAT-7.0)......... ....1497.3.1 Calibration of Steam and Feedwater Flow lnstrumentsat75o/o Power (2-PAT-7.1) ............... ............152aal- l-TABLE OF CONTENTS (continued)SECTION7.0 POWER ASCENSION TESTING (continued)PAGE NUMBER7.4 Test Sequence for 100% Plateau (2-PAT-8.0)............... .............1557.4.1 Load Swing Test (2-PAT-1 .2)............ ...........1597.4.2 Large Load Reduction Test (2-PAT-1.3)............. ...........1627.4.3 Pipe Vibration Monitoring (2-PAT-1.4)......... ..................1777.4.4 LoosePartsMonitoringSystem(2-PAT-1.5)............ .....1827.4.5 Startup Adjustments of Reactor Control System (2-PAT-1.6).........1857.4.6 Operational Alignment of Process Temperaturelnstrumentation (2-PAT-1.7)......... ................1897.4.7 Thermal Expansion of Piping Systems (2-PAT-1.8).......................1967.4.8 Automatic Steam Generator Level Control (2-PAT-1.9) ............ .....2017.4.9 lntegrated Computer System (lCS) (2-PAT-1.10)...........................2057.4.10 RVLlSPerformanceTest(2-PAT-1.11).......... ...............2087.4.11 Common Q Post Accident Monitoring System (2-PAT-1.12)..........2127.4.12 RCS Flow Measurement (2-PAT-3.3)......... ...................2157.4.13 Calibration of Steam and Feedwater Flow lnstrumentsat 100 % Power (2-PAT-8.4) ............... .........2177.4.14 Shutdown From Outside the Main Control Room (2-PAT-8.5) .......2217.4.15 Plant Trip Evaluation for Equivalency of Test Performance of2-PAT-8.6, Plant Trip From 100o/o Power (Turbine Trip).................2267.4.16 Core Power Distribution Factors (2-PET-301)............. ...................2487.4.17 Operational Alignment of NIS (2-PET-304) ............ .......2537.4.18 Radiation Baseline Surveys (RCI-159) .........255aaal-rl-FIGURE2.0-14.6-14.6-25.3-16.3-17.1 .2-17.2.2-17.2.2-27.2.2-37.2.2-47.2.2-57.2.2-67.2.2-77.2.2-87.2.2-9LIST OF FIGURESPAGE NUMBERWBN PowerAscension Test Program Schedule Overview ..................9U2Cl Core Load Sequence ............39Unit 2 Core Load ICCR ...................40Pressurizer Spray Response ...........60ICRR vs Primary Water (N31, N32) ........ .........94SG #2 FW Flow Oscillation ...........115RCS Auctioneered T"* vs T6sSteady-State Operatioh.............. ...132RCS Auctioneered T"* vs T,"glncreasing T"* Transient .......... ....133RCS Auctioneered Trrs vs TpgDecreasing T"r, Transient.......... ...134Rod Speed and Direction Demandlncreasing T"rs Transient .......... ....135Rod Speed and Direction DemandDecreasing T"r, Transient.......... ...136Pressurizer Pressurelncreasing T"r, Transient .......... ....137Pressurizer PressureDecreasing T"r, Transient.......... ...138Pressurizer Level and Level Setpointlncreasing T"rs Transient .......... ....139Pressurizer Leve! and Level SetpointDecreasing T"", Transient.......... ...140al-v LIST OF FIGURES (continued)FIGURE PAGE NUMBER7.4.2-1 Generator Output vs Time(Large Load Reduction)........ .........1057.4.2-2 NIS Power vs Time(Large Load Reduction)........ .........1G07.4.2-3 Pressurizer Pressure vs Time(Large Load Reduction)........ .........1G77.4.24 Pressurizer Level vs Time(Large Load Reduction)........ .........1687.4.2-5 Control Bank D Position vs Time(Large Load Reduction)........ .........1097.4.2-O Tay6 vs Time(Large Load Reduction)........ .........1707.4.2-T Steam Dump Signal vs Time(Large Load Reduction)........ .........1717.4.2-8 Feedwater Pressure vs Time(Large Load Reduction)........ .........1721727.4.2-9 Steam Generator 1 Level vs Time(Large Load Reduction)........ .........1737.4.2-10 Steam Generator 2 Level vs Time(Large Load Reduction)........ .........1747.4.2-11 Steam Generator 3 Level vs Time(Large Load Reduction)........ .........1757.4.2-12 Steam Generator 4 Level vs Time(Large Load Reduction)........ .........1707.4.14-'l Main Control Room Abandonment30 Minute Stability.... ....2247.4.14-2 Main Control Room AbandonmentCooldown ...2257.4.15-1 Representative RCCA Position vs Time(100% Trip) ........ ..........2327.4.15-2 Main Turbine Speed vs Time(100Yo Trip)........ ..........233 FIGURE7 .4.15-37.4.15-47 .4.1 5-57 .4.15-67.4.15-77 .4.15-g7 .4.1 5-g7 .4.1 5-1 07.4.15-117.4.15-127 .4.1 5-1 37.4.15-147 .4.15-157 .4.15-16LIST OF FIGURES (continued)PAGE NUMBERSafety lnjection Alarm vs Time(100Yo Trip)........ ..........234Steam Generator Safety Valve Position Loops 1 &2 vs Time(100% Trip)........ ..........29sSteam Generator Safety Valve Position Loops 3 & 4 vs Time(100Yo Trip)........ ..........236Pressurizer Safety Valve Outlet Temperature vs Time(100% Trip)........ ..........237NIS Power vs Time(100% Trip)........ ..........238Hot Leg RTD Response vs Time(100o/o Trip)........ ..........299Steam Generator Levels vs Time(100% Trip)........ ..........240Pressurizer Pressure vs Time (1)( 1 00% Trip).. .. . . . ... ...... t .. .. .. ...... .. . .241Pressurizer Pressure vs Time (2)(100% Trip)......... .......242Pressurizer Level vs Time(100o/o Trip)........ .......243Pressurizer Level Controller Output vs Time(100Yo Trip)......... .......244RQQ Aryrage Temperature (Auctioneered) vs Time(100yo Trip)......... .......2459lCqT Pqqp Demand and T"o vs Time(100o/o Trip)......... .......246Main Feedwater Pump Flows and T"* vs Time(100Yo Trip)......... .......247avl-LIST OF ACRONYMSADRC Advanced Digital Reactivity ComputerAFW Auxiliary FeedwaterAOI Abnormal Operating lnstructionARO All Rods OutASME American Society of Mechanical EngineersAUX AuxiliaryBOL Beginning of LifeBYA Bypass A (Reactor Trip Breaker A)BYB Bypass B (Reactor Trip Breaker B)CBA Control Bank ACBB Control Bank BCBC Control Bank CCBD Control Bank DCERPI Computer Enhanced Rod Position lndicationCLA Cold Leg AccumulatorCOTS Channel Operational TestsCPS Counts Per SecondCR Condition ReportCRDM Control Rod Drive MechanismCTL ChronologicalTest LogCV Concurrent VerificationCVCS Chemical and Volume Control SystemDAS Data Acquisition SystemDCN Design Change NoticeDCS Distributed Contro! SystemDRWM Digital Rod Worth MeasurementeNuPOP Electronic Nuclear Parameters and Operations PackageFATF FuelAssembly Transfer FormFCV Flow ControlValveFHI Fuel Handling lnstructionF! Flow lndicatorFIC Flow lndicating ControllerFW FeedwaterGO General Operating InstructionHFP Hot Full PowerHS Hand SwitchHVAC Heating Ventilation and Air ConditioningHZP Hot Zero PowerICRR lnverse Count Rate RatioICS lntegrated Computer Systemaavt1 LIST OF ACRONYMS (continued)INPO lnstitute of Nuclear Power OperationsIR lntermediate RangeITC lsothermalTemperature CoefficientlV lndependent VerificationKBH Thousand Pounds Per HourLC Level ControllerLCP Loop Calculation ProcessorLCV Level ControlValveLPMS Loose Parts Monitoring SystemLPPT Low Power Physics TestLVDT Linear Variable Differential TransformerM&TE Measuring and Test EquipmentMCD Maximum Channel DeviationMCR Main Control RoomMED Maximum Expected DeviationMFP Main Feedwater PumpMFPT Main Feedwater Pump TurbineMFW Main FeedwaterMMI Man Machine lnterfaceMPPH Million Pounds Per HourMSIV Main Steam lsolation ValveMSV Main Steam ValveMTC Moderator Temperature CoefficientNl Nuclear lnstrumentationNIS Nuclear lnstrumentation SystemNOB Nuclear Operating BookNOTP Normal Operating Temperature & PressureNPG Nuclear Power GroupNRC Nuclear Regulatory CommissionNSSS Nuclear Steam Supply SystemNuPOP Nuclear Parameters and Operations PackageOPC Overspeed Protection ControllerOPDP NPG Standard Department Procedure (Operations)OPSP Over Power SetpointOTDT Overtemperature Delta TemperatureOTSP Over Temperature SetpointPAT Power Ascension TestPATP Power Ascension Test ProgramPDMS Power Distribution Monitoring SystemPET Power Escalation TestPIC Pressure lndicating Controlleraaavl- l- l-PIDPLSPMTPORCPRpsipsiapsidpsigPTIPZRQPTRRBSSRCCARCIRCPRCSRDTCRegRHRRPIRSARTRTARTBRTDRTPRVLISRWPRWSTSARSBASBBSBCSBDSESEQSFPSGSGBDSILIST OF ACRONYMS (continued)Point IdentificationPrecautions, Limitations and SetpointsPost Maintenance TestPlant Operations Review CommitteePower RangePounds per square inchPounds per square inch absolutePounds per square inch differentialPounds per square inch gaugePreoperational Test I nstructionPressurizerQuadrant Power Tilt RatioRod Bank Select SwitchRod Cluster Control AssemblyRadiological Control I nstructionReactor Coolant PumpReactor Coolant SystemRod Drop Test ComputerRegulatingResidual Heat RemovalRod Position lndicatorRedundant Sensor AlgorithmReactor TripReactor Trip Breaker AReactor Trip Breaker BResistance Tem perature DetectorRated Thermal PowerReactor Vessel Level lnstrumentation SystemRadiological Work PermitRefueling Water Storage TankSafety Analysis ReportShutdown Bank AShutdown Bank BShutdown Bank CShutdown Bank DSite EngineeringSequenceSpent Fuel PoolSteam GeneratorSteam Generator BlowdownSurveillance I nstructionaIX LIST OF ACRONYMS (continued)SOI System Operating lnstructionSR Source RangeSRF Statistical Reliability FactorSSP Site Standard PracticeSSPS Solid State Protection SystemSWIF Seal Water lnjection FilterTE Temperature ElementTl Technical lnstructionTR Temperature RecorderTRG Test Review GroupTRI Technical Requirements lnstructionTTD Time Trip DelayTVA Tennessee Valley AuthorityUC Urgent ChangeUFSAR Updated Final Safety Analysis ReportUSNRC United States Nuclear Regulatory CommissionVCT Volume Control TankWBN Watts Bar NuclearWO Work Order

1.0 INTRODUCTION

The lnitial Startup Report for the Watts Bar Unit 2 nuclear plant discusses theresults of testing performed from initial core load through full power operation.This report address each of the power ascension tests identified in Chapter 14 ofthe WBN Unit 2 UFSAR and other license commitments. The report includes adescription of the measured values of the operating conditions or characteristicsobtained during the testing program and a comparison of these values withdesign predictions and specifications. Any corrective actions that were requiredto obtain satisfactory operation are also described.WBN Unit 2 UFSAR Chapter 14.2.6, Test Records, requires the Startup Reportbe submitted within:(1) 90 days following completion of the Startup Test Program,(2) 90 days following resumption or commencement of commercial poweroperation, or(3) 9 months following initial criticality, whichever is earliest.lf the Startup Report does not cover all three events (i.e., initial criticality,completion of Startup Test Program, and resumption or commencement ofcommercial power operation), supplementary reports shall be submitted at leastevery 3 months until allthree events have been completed.Item (1) is being satisfied since the Power Ascension Test Program wascompleted on October 6, 2016.WBN Unit 2 Facility Operating License No. NPF-96 was issued on October 22,2015. lnitial Fuel load commenced with movement of the first fuel assembly at20:49 on December 4, 2015. Core loading was completed at 02:10 on December8, 2015. lnitial criticality was achieved at 02:16 on May 23,2016. Further testingwas successfully completed at the following plateaus:Test Plateou, % RTPDate Completed30June 16, 201650July 16, 201675July 29,201690August 29,2016100October 6 ,2016lnitial Fuel load, precritical testing, initial criticality and lowand power ascension testing are discussed in separateThe report details the test objectives, methodology, testnoted for each of the tests performed.power physics testing,sections of the report.results, and problems

1.0INTRODUCTION

(continued)Acceptance Criteria is defined as safety related performance parameters definedin the Design Output, vendor documents, TVA or vendor drawings, NRCcommitments, other licensing and design documents, and so forth, that must beexhibited during the performance of a PAT or PET. Failure to meet anacceptance criterion is considered to be a safety related issue.A 10CFR50.59 Evaluation per NPG-SPP-09.4, 10 CFR 50.59 Evaluationsof Changes, Tests, and Experiments or a Technical Evaluation perNPG-SPP-09.3, Plant Modifications and Engineering Change Control, may berequired and testing may be stopped. The subsequent course of action will bedetermined by the nature of the discrepancy and applicable TechnicalSpecifications. Failure to meet Acceptance Criteria will be documented in aCondition Report (CR).Review Criteria encompasses other performance parameters defined in theUFSAR, design criteria, vendor documents, drawings (TVA or vendor), otherlicensing, design, setpoint and operational documents that are expected to beexhibited during performance of a PAT or PET. These criteria should be viewedas a guide to possible measurement or design errors. Failure to meet thesecriteria do not by themselves constitute problems. While prudent measuresshould be taken to resolve any conflict between measurements and predictions,failure of these criteria do not require 10CFR50.59 Evaluations, per NPG-SPP-09.4, and do not require testing or power ascension to be stopped for resolution.Failure to meet these review criteria will be documented in a CR.

2.0 POWER ASCENSTON TEST PROGRAM (pATp) OVERVTEWThe PATP was developed from testing described in Chapter 14 of the WBNUnit 2 UFSAR and requirements specified in Regulatory Guide 1.68, Revision 2,"lnitial Test Programs for Water-Cooled Nuclear Power Plants". Testing of theNSSS followed Westinghouse test methodology.2.1 Administration of the ProgramThe Site Vice President had the overall responsibility for the PATP.Overall management of the PATP was directed by the Plant Manager who wasresponsible for:o Development and implementation of the PATP to ensure the PATP wasconducted in a safe and efficient manner while complying with licenseprovisions and other commitments.. Establishing the Power Ascension Testing Organization.o Advising senior management on PATP activities.o Establishing a Technical Review Group (TRG) as a subcommittee of thePORC to review PATP activities.o Providing final approval of Power Ascension Tests (PATs) and selectedother procedures.o Ensuring the PATP was conducted in accordance with applicable WBNAdministrative Procedures.o Providing approvalto proceed to the next PATP test plateau.o Providing final approval of all each test package and the Startup Report.The Power Ascension Test Manager was responsible for:o Notifoing the plant manager of major problems and of the completion ofeach major test phase (i.e., test sequence) of the program.. Ensuring the PATs and the PETs were available for NRC review aminimum of 60 days before the scheduled fuel load date.o Ensuring the technicaljustification and schedule, including power level forcompletion of delaying preoperational tests, were provided to the NRCstaff prior to fuel Ioad.

2.1 Administration of the Program (continued)o Ensuring the requirements of TVA-SPP-30.010, lnitial Synchronization ofTVA Generating Assets to TVA's Transmission System, were met.. Developing and implementing plans and schedules for the PATP.. Ensuring testing activities, including planning and scheduling, resulted insafe plant operations and that were not dependent on the performance ofuntested systems.. Coordinating and directing overall PATP testing and related activities andrequirements with appropriate support groups.. Supervising test personnel assigned to the power ascension testing group.o Assigning responsibilities to organizations for specific testingrequirements.o Participating in the review activities of the TRG, and acting as Chairman ofthe TRG.o Ensuring the test procedures were reviewed by the TRG.o Ensuring the Startup Report was reviewed by the TRG.o Ensuring additional startup Reports were prepared, reviewed, approvedand transmitted to the NRC as needed.o Ensuring the post-performance test results (i.e., test packages) werereviewed by TRG.o Ensuring test directors for the PATP were qualified, and met the minimumqualifications of ltem 1 and either ltem 2 or ltem 3 below and ensuringother required individuals (e.9. lndependent Verifiers (lV) and ConcurrentVerifiers (CV) were qualified to perform the tasks assigned:1. Knowledgeable of the test program administration, the systemdesign and operational requirements, and expected plantoperational characteristics during the test, andTrained as test coordinators in accordance with NPG-SPP-06.9.1,Conduct of Testing.

2.1Administration of the Program (continued)2. Possessed a bachelor degree in engineering or physical science,andHad two years experience in power plant testing or operation.Included in the two years was one year nuclear power plant testing,operating or training on a nuclear facility.3. Possessed a high school diploma or equivalent, andHad five years experience in power plant testing. lncluded in thefive years were two years of nuclear power plant experience.Credit for up to two years of related technical experience could besubstituted for experience on a one-for-one basis.Technical and administrative oversight of the PATP was performed by TRGwhich was composed of one representative, or their alternates, from each of thefollowing organizations:o Plant Operationso Reactor Engineeringo Site Engineering. Corporate Nuclear Fuelso Power Ascension Testingo WestinghouseTRG was charged with reviewing PATP testing activities for technical adequacyand affecUimpact on nuclear safety, and advising PORC and the plant manageron the disposition of those items reviewed. The responsibilities of TRG includedfinal review and recommendation of approval of al! PATP test procedures,revisions, and test results.Following completion of testing at each major test sequence of the PATP, testresults were reviewed by TRG to ensure required tests had been performed.TRG also ensured Acceptance Criteria were satisfied; test deficiencies hadproper dispositions, appropriate retesting had been completed, and test resultshad been reviewed by appropriate designated personnel prior to proceeding tothe next major test sequence. This review ensured that all required systemswere operating properly and that testing for the next major test sequence couldbe conducted in a safe and efficient manner.

2.2Implementation of the ProgramThe WBN PATP utilized information gained from operating and testing experience atother nuclear plants. This information was used in the development of the PATPtest procedures and schedules and to alert personnel to potential problem areas.Test procedures were developed utilizing information obtained from OperatingExperience (OE) database. The TVA Operating Experience Program identifies andevaluates experience gained from other TVA nuclear plants, INPO, NRC, equipmentsuppliers, and from other utilities. Significant operational experience and eventswere reviewed and integrated into appropriate PATP test procedures to ensurenuclear safety and reliability. To the extent practical, simulator-based training andtrial use of the PATP test procedures were performed on the WBN simulator tofamiliarize personnel with systems and plant operation and to assure technicaladequacy of the procedures under simulated plant conditions prior to field useduring power operation.The testing program was conducted by qualified personnel using approved plantadministrative, test, and operating procedures. The plant was taken from core loadto full power in a highly controlled, conservative, and documented manner whichdemonstrated, where practical:The plant is ready to operate in a manner which will not endanger thehealth and safety of the public.The plant has been properly constructed, and plant performance issatisfactory in terms of established design criteria.The plant meets licensing requirements and provides assurance of plantreliability for operation.o The plant is capable of withstanding anticipated transients and postulatedaccidents.The PATP was specified in seven PAT sequence procedures:. 2-PAT-2.0, lnitial Core Loading Sequence. 2-PAT-3.0, Post Core Loading Precritical Test Sequenceo 2-PAT-4.A, lnitial Criticality and Low Power Test Sequence. 2-PAT-5.0, Test Sequence for 30% Plateauo 2-PAT-6.0, Test Sequence for 50% Plateauo 2-PAT-7.0, Test Sequence for 75% Plateauo 2-PAT-8.0, Test Sequence for 1A0o/o Plateau 2.2lmplementation of the Program (continued)Each PAT sequence procedure called out the performance of other PATs, aswell as other designated plant procedures such as PETs, Sls, TRls, Tls, RClsand FHls. The sequence procedures specified the logica! performance ofrequired tests and procedures through each test plateau. The sequenceprocedures also specified general prerequisites, precautions and limitations, andadditional operational steps at each test plateau. The detailed test and normalplant procedures called out by the sequence procedures defined step-by-stepactions, specific prerequisites and limitations, signoffs, data taking requirements,and test acceptance and review criteria.The PATP commenced with the receipt of the Facility Operating License onOctober 22, 2015, and progressed with core loading, precritical testing, initialcriticality and low power physics testing, and power ascension testing. Core loadprocedures directed the initial core load in a prescribed manner which ensuredcore loading was accomplished in a safe and orderly fashion. Precritica! testingbrought the plant to hot standby conditions, made measurements, anddemonstrated that the plant was ready for critical operation. lnitial criticality onMay 23,2016, brought the Unit 2 reactor critica! for the first time. Low powerphysics testing performed measurements on the critical reactor to demonstrateconformance with design predictions prior to power operation. PAT brought theplant to full power, made minor plant instrumentation adjustments, anddemonstrated the plant's ability to withstand selected transients. Figure 2.0-1depicts the time line for the PATP.Plant events not directly associated with the PATP added to the duration of theprogram. These events are included in the chronology.

2.3 SUMMARYWatts Bar Nuclear Plant's Unit 2 Power Ascension Test Program began with the receiptof the operating license. The program encompasses; those preoperational type teststhat were deferred to the PATP, the prerequisites required to load the initial core, theinitia! core loading itself, post core loading tests, initial criticality, low power tests, and at-power tests.Completion of these tests verified that the unit was properly designed, constructed, andready to operate in a manner that will not endanger the health and safety of the public,meets contractual and licensing requirements, and provides assurance of plantreliability for operation. The PATP used Regulatory Guide 1.68, Revision 2 and theWBN UFSAR for development of the test requirements.lssues identified during the testing were resolved except for the following open itemsthat will be resolved by the Corrective Action Program.OPEN ITEMS:(1) CR 1208694 initiated WO 118122821 to design and install bracing on multipleMain Steam Traps as a result of visual inspection during 2-PAT-1.4, PipeVibration Monitoring.(2) UFSAR Table 14.2 2, Sheet 5 Test Method, refers to an evaluation of thermalexpansion at final ambient conditions. This final ambient condition evaluation willbe performed later and is tracked by Commitment 118008175.(3) CR 1208178 was initiated for 2-PT-1-81 being unavailable during testing2-PAT-1.6, Startup Adjustments of Reactor Control System, and will be repairedby WO 118121693.(4') CR 1171424 was written during the performance of 2-PAT-1.5, Loose PartsMonitoring System, for three channels taken out of service due to issues.These channels will be repaired under the following work orders.o WO 117845593 - Channel 101 Experiencing excessive noise and isalarming due to "llTA'rattling.. WO 117843208 - Channel 102 Accelerometerfound damagedo WO 117843209 - Channel 110 Suspect preamplifier FIGURE 2.0.1WBN POWER ASCENSION TEST PROGRAMSCHEDULE OVERVIEWi:=;sE:*-EEsb0.t roH.TSdfiernilolqi0PFr-o. I .!ron ls.I*l-Esl. :r-IIEEi*l-IIIt-II-TIgElll3oBi'Eg*q"PA.+a\l-qtnIo.aa\l-qYFLaa\tqt\,aFCIa\9rn-tOrr...H g EB,J;do3oCLyrrBIA+iEool-orrtt/c.gaaEl!tt,UtrIECLo'rEfIJ.g*octtooo 3.0 WATTS BAR UNIT 2 STARTUP CHRONOLOGYNote: Power Ascension Tests may be performed at multiple testing plateaus.The description of the individual PAT is documented in the section(plateau) in which it was completed.10122115 -Receipt of WBN Unit 2 Facility Operating License No. NPF-96 fromthe NRC.11119115 PAT-2.0, lnitial Core Load Sequence, was begun.-2-PAT-2.1, Reactor System Sampling for Core Load, was initiated.11119115 -Commenced movement of new fuel into the Spent Fuel Pool inpreparation for loading Unit 2 core.11122115 -All 193 fuel assemblies required for Unit 2 fuel load moved into theSpent Fue! Poo!.11123115 -Completed Spent Fuel Pool verification for Unit 2 Cycle 1, fuelassembly and component insert verification with no discrepanciesnoted.1213115 -RCl-159, Radiation Baseline Surveys, pre-fuel load surveys werecompleted. No Acceptance or Review Criteria is associated with thisprocedure.1214115 PAT-2.1, Reactor System Sampling for Core Load, completed withall criteria met.- Unit 2 entered Mode 6 at 20:04.-lnitiated transport of first fuel assembly to Unit 2 vessel at20:49.1215115 -Fuel movement was delayed due to issues with refueling machine at06:23.1215115 PAT-2.2, Response Check Of Core Load lnstrumentation After 8Hour Delay ln Fuel Movement, completed at 11:43 with all criteriamet and fuel movement resumed.1215115 -Debris was reported on the bottom of the fourth fue! assembly.Debris was removed and fuel movement resumed after appropriateapprovals. CR 1112204 was initiated.1216115 -Fuel movement was suspended again due to debris on a fuelassembly. Debris was cleared and fuel movement resumed.1218115 -lnitialfuel load for Unit 2 completed at 02:10.-2-PET-105, lnitial Core Loading completed.-2-Tl-28, Verification Of Core Load Prior To Vessel Closure, wascompleted at 12:29 with all criteria met.10 3.0 Watts Bar Unat 2 Startup Chronology (continued)1219115 PAT-2.0, lnitial Core Load Sequence, was completed and TRGapproved.-2-PAT-3.0, Post Core Loading and Precritical Test Sequence,pre-requisites were initiated.12110115 -Unit entered Mode 5, maintaining <105"F in RCS in preparation ofPAT at the Ambient Plateau.12112115 -RCl-159, Radiation Baseline Surveys - Post Fuel Load Survey wasfield work complete for applicable Ambient Plateau sections. NoAcceptance or Review Criteria were associated with this procedure.12116115 PAT-1.8, Thermal Expansion of Piping Systems, was field workcomplete for applicable Ambient Plateau sections with all criteria met.12123115 PNf-1.4, Pipe Vibration Monitoring, Section 6.5.6, Condensate -Short Cycle, field work complete with all Acceptance Criteria met.There was no Review Criteria for this test.11161'16 - PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete for applicable Ambient Plateau sections. There was noAcceptance or Review Criteria for these sections.1120116 PAT-3.10, Reactor Trip System, field work complete with all criteriamet.1124/,16 PAT-3.1, Control Rod Drive Mechanism and CERPI lnitia!Calibration, field work complete with allAcceptance Criteria met afterevaluation. CR 1128950 was written for high current amplitudes andclosed following Westinghouse evaluation that determined themeasurements to be acceptable. There was no Review Criteria forthis PAT.-2-PAT-3.8, Rod Drop Time Measurement and Stationary GripperRelease Timing, Mode 5 Performance, field work complete forapplicable Ambient Plateau sections with allAcceptance Criteria met.There was no Review Criteria for this test.1126116 -PAT testing on the plant primary side was suspended on 1126116until plant conditions allowed further testing.211116 -WO 112989715 Complete for WINCISE Site Acceptance Test (WNA-TP-02985-WBT). This WO satisfied UFSAR Table 14.2-2, Sheet 12,lncore lnstrumentation System Test Summary, Acceptance Criteria 1.-2-PAT-1.4, Pipe Vibration Monitoring, field work complete forSection 6.5.7 Condensate - Long Cycle with all criteria met.11 3.0 Watts Bar Un.t2 Startup Chronology (continued)3119116 -Unit entered Mode 4, RCS temperature >200"F and <350'F to allowPAT at the 250"F Plateau.3121116 PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for 250'F Plateau applicable sections with all criteria met.3124116 -RCS temperature increased to 300"F to facilitate PAT.3125116 PAT-1.11, RVLIS Performance Test, field work complete forapplicable 300'F Plateau sections with all criteria met.3/30/16 -Unit entered Mode 3, RCS temperature ) 350'F to allow furtherPower Ascension Testing at the 360"F Plateau.3131116 PAf-1.11, RVLIS Performance Test, field work complete forapplicable 360'F Plateau sections with all criteria met.-2-PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for applicable 360'F Plateau section with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable 360'F Plateau sections with all criteriaacceptable for continued heat-up.411116 -lnitiated plant heat-up to 400"F for PAT at01:14.-2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable 400"F Plateau sections with Review Criteria not met.CR 1156425 was written.-2-PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for applicable 400"F Plateau section with all criteria met.-lncreased RCS temperature to 450"F for testing.-2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable 450"F Plateau, with all criteria met.-2-PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for applicable 450'F Plateau sections with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems - field workcomplete for applicable 450'F Plateau sections. CR's were initiatedwithin the test for seven snubbers not performing as expected.Results indicated no issue with snubbers and approved to continue tonext plateau testing. (See Problem Report #1 of 2-PAT-1.8)L2 3.0 Watts Bar Unat 2 Startup Chronology (continued)412116 -Unit coo! down initiated to repair a check valve with excessiveleakage. Additionally, two RCS RTD's were replaced.-Unit entered Mode 4, RCS temperature >200'F and <350'F, at06:29.418116 -Unit re-entered Mode 3, RCS temperature 2 350"F.4110116 -RCS temperature increased to 500"F.-2-PAT-1.1 1, RVLIS Performance Test, field work complete forapplicable 500'F Plateau sections with all criteria met.4112116 PET-102, Pre-Power Escalation NIS Calibration Data, completedwith all criteria met.4113116 -RCS temperature increase to 557"F.-2-PAT-1.11, RVLIS Performance Test - field work complete forapplicable section of 557"F data taking only with al! criteria met forsteady state data collection.-2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for Steady State Data Collection, section 6.7, with allcriteria met.-2-PAT-1.4, Pipe Vibration Monitoring - Section 6.5.2, MainFeedwater Pump 2A Start and Steady State Operation on Recirc.,was field work complete on 4113116 with velocity and displacementAcceptance Criteria not met. CR 1161783 was initiated for anengineering evaluation which concluded equipment was acceptableas is. There was no Review Criteria for this test.4114116 -RCS at normal operating pressure.-2-PAT-1.4, Pipe Vibration Monitoring, completed for Section 6.5.1,Pressurizer Surge, Mode 3, with all criteria met for that section.-2-PAf-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable 557'F Plateau sections with an issueoutside containment on Protective Device PD07-2. WO 1 17755755was to resolve the issue with PD07-2 and investigate any possibleissues with PD07-1.Additionally, other components did not move as expected and wereevaluated and concluded to be within their working range. (SeeProblem Reports #2,#3 of 2-PAT-1.8).4116/16 PAT-3.2, Pressurizer Spray Capability and Continuous SprayFlow Setting, Section 6.1, Adjustment of the Pressurizer ManualSpray Bypass Valves, was completed. AIIAcceptance Criteria wasmet. Review Criteria for MCR alarms was not met with CRs1161382 and 1160969 written. A Westinghouse evaluationdetermined the PAT met the operability and design requirements ofthe pressurizer spray system.13 3.0 Watts Bar Unat 2 Startup Chronology (continued)4117116 -Unit was placed in Mode 4 for repairs to the Turbine DrivenAuxiliary Feedwater Pump and replacement of PD07-2 shimdetermined to require adjustment.511116 -Unit 2 in Mode 3 at 17:36.512116 -Unit 2 at NOTP at 23:00.513116 PAT-3.3, RCS Flow Measurement, field work complete with allcriteria met.514116 PNf-1.6, Startup Adjustments of Reactor Control System, fieldwork complete for Mode 3. This performance was data taking only.515116 PAT-1.4, Piping Vibration Monitoring, Section 6.5.3, MainFeedwater Pump 28 Start and Steady State Operation on Recirc.,was field work complete with steady state velocity and displacementexceeding the Acceptance Criteria. CR 1168287 was written for anengineering evaluation and resulted in adjustment of a loose hangerand a retest. The retest was completed on 5113116 with satisfactoryresults. There was no Review Criteria for this test.516116 PAT-1.T,OperationalAlignmentofProcessTemperaturelnstrumentation, field work complete with allAcceptance Criteria met.One Review Criteria was not met and CR 1168641 was initiated.517116 PAT-3.0, Attachment 1, field work complete with all AcceptanceCriteria met. CR 1168487 was written to document alternatecharging flow was not within anticipated range, however, it had noaffect on the test acceptance.-2-PAT-3.11, Adjustment of Steam Flow Transmitters at MinimalFlow, field work complete with all Review Criteria met. There was noAcceptance Criteria associated with this performance.518116 PAf -1.11 - RVLIS Performance Test, Section 6.1.3, field workcomplete and results indicated Acceptance Criteria would not bemet. The system was updated with the new RVLIS constantssupplied by Westinghouse to correct the abnormality. CR 1171130was initiated.-2-PAT-1.12, Common Q Post Accident Monitoring System, Section6.8, Pump Contact Data Collection at 557"F, was completed with allcriteria met.-2-PAT-3.7, Reactor Coolant Flow Coastdown, field work completewith all criteria met.L4 3.0Watts Bar Unat 2 Startup Chronology (continued)5112116-2-PAT-3.8 Rod Drop Time Measurement and Stationary GripperRelease Timing (Mode 3), field work complete. CR 1169659 writtenfor two rods failing a two sigma statistical evaluation. Additional roddrops were performed and the Acceptance Criteria was met. Therewas no Review Criteria.-2-PAT-1.4, Pipe Vibration Monitoring, Section 6.5.4, Turbine BypassValve 2-FCV-1-105 Transient, was completed with all criteria met.Section 6.5.5, Turbine Bypass Valve 2-FCV-1-111 Transient wascompleted onSl12h6 and re-tested on 5113116. Engineeringevaluation of the retest indicated satisfactory results. CR 1170319documented the engineering evaluation.-2-PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete with all criteria met after a volume booster adjustment withCR 1170159 and a retest on 2-FCV-1-108.-2-PAT-3.4, Rod Control and Rod Position lndication (CERPI) - fieldwork complete. The Acceptance Criteria was not met in Sections 6.4and 6.10. CRs 1168845, 1168881, and 1169602 were written todocument failure to meet criteria. The criteria was re-evaluated and itwas determined the Acceptance Criteria should be changed torequire each Rod Position lndication to indicate rod motion consistentwith the group demand indication for the full range of rod travel. Achange to the Westinghouse Acceptance Criteria and SAR ChangePackage No. U2-019 were approved and an urgent change to theprocedure incorporated the revised Acceptance Criteria. AllAcceptance Criteria were then met.-2-PAT-3.0, Post Core Loading and Precritical Test Sequence, TRGapproved.-2-PAT-4.0, lnitial Criticality and Low Power Test Sequence, inprogress.-A cool down to 360'F was initiated to replace a failed RTD on RCSLoop 3 Hot Leg. The unit was stabilized between 355-365'F at22:59.511311651151165t1611 651111165118/1 651201165t21t16-Unit was placed in Mode 4 at23:58 to facilitate repairs to the SolidState Protection System (SSPS).-Unit was returned to Mode 3 at 04:15.-Unit reached NOTP at 01:00. Response time testing of the replacedRCS RTD indicated it did not meet its Acceptance Criteria. A DCNwas initiated to revise the Acceptance Criteria to allow entry intoMode 2.15 3.0 Watts Bar Unat 2 Startup Chronology (continued)5123116 -Unit entered Mode 2 at0'l:04.-lnitial criticality at 02:16.-2-PET-201, lnitial Criticality and Low Power Physics Testing,completed with all criteria met.-2-PET-103, Reactivity Computer (ADRC), completed with all criteriamet.-2-PET-304, Operational Alignment of NlS, applicable sectionscompleted with all criteria met.5124116 PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 was written to document three channelsremoved from service.-2-PAT-1.10, lntegrated Computer System (lCS), applicable sectionscompleted with all criteria met. CR 1173586 was initiated for ICS PIDquality on several points but did not affect this plateau performance.CR 1174334 was initiated for exceeding the MED between T0457AMCR indicator 2-Tl-62-29, RCP 3 LWR RADIAL BRG Temp.-2-P4T4.0, lnitial Criticality and Low Power Test Sequence TRGapproved.-2-PAT-5.0 Test Sequence for 30% Plateau in progress.5125116 -Unit 2 entered Mode 1 at 03:33.5126116 PAT-5.3, Automatic Steam Generator Level Control Transients atLow Power, completed with all criteria met.5127116 PAT-5.1, Dynamic Automatic Steam Dump Control, completedSections 6.3, 6.4 and 6.5 with allAcceptance Criteria met after aprocedure and UFSAR revision per Westinghouse LetterLTR-SCS-16-23.-With reactor power between 13 and 14 percent the turbine was rolledfor testing in preparations for initial generator synchronization. Duringthe roll up an unanticipated noise was heard and the roll wasterminated. A second rollwas made later in the evening with similarresults. A decision was made to place the Unit in Mode 3 for turbinerepairs.5128116 -Unit 2 re-entered Mode 3 at 01:54 after a manual reactor trip forturbine repairs.5131116 -Unit 2 re-entered Mode 2 at 12:00.-Reactor taken critical at 13:39.-Unit 2 entered Mode 1 at 17:49.-RCI-159, Radiation Baseline Surveys, completed. No Acceptance orReview Criteria were associated with this procedure.L6 3.0 Watts Bar Unat 2 Startup Chronology (continued)618116619t1661111166113t16-Unit 2 was synchronized to the grid at 20:39 and holding at 15percent power to repair steam leaks.-Unit 2 turbine was manually tripped due to a non-isolable steam Ieak.-Unit 2 synchronized at 11:40.-Unit 2 received an automatic reactor trip with a safety injection due to#1 governor valve failing open causing a steam line pressuredecrease and subsequent Reactor Trip and Safety lnjection at 12:27.Unit was stabilized in Mode 3 following Reactor Trip.-Unit 2 in Mode 2 at01:39 after repairs to the governor valve.-Entry into Mode 1 was at 09:32.-Unit 2 synchronized to grid at 06:40.-Turbine manually tripped due to an non-isolable steam leak at 17:52.-Unit 2 synchronized to the grid at 13:23 and power increase initiatedto the 30% testing plateau.-2-PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-PAT-1.12, Common Q PostAccident Monitoring System,applicable sections were completed with all criteria met.-2-PAT-1.11, RVLIS Performance Test, applicable sections werecompleted with all criteria met.-Completed the initia! Flux Map in accordance with 2-T141, lncoreFlux Mapping, and 2-Sl-0-20, Hot Channel Factors Determination.-2-PAT-1.10, lntegrated Computer System (lCS), was completed withall criteria met. CR 1181784 was written to address a database errorbut did not affect this plateau performance.-2-PAT-1.4, Pipe Vibration Monitoring, completed with all criteria metfor observations at the 30% Plateau.-2-PAT-5.3, Automatic Steam Generator Level ControlTransients atLow Power, was completed with allAcceptance Criteria met. CR1181278 was initiated to document one Review Criteria not met. Anengineering evaluation determined this did not affect the performanceof the test nor invalidate any of the test results and testing shouldproceed to the next plateau.61141166115/1 66/3/1 66t4t166/5/1 6L1 3.0Watts Bar Unat 2 Startup Chronology (continued)6116/1 6-2-PAT-1 .7, Operational Alignment of Process Temperaturelnstrumentation, was completed with allAcceptance Criteria metTwo Review Criteria concerning parameters related to Delta T failed.The OTDT calculated by Eagle-21 and provided by the MMI cartsindicated approximately 158% and the MCR indicators maximumvalue is 150%. lt was expected the reading from Eagle-21 wasaccurate and the MCR meters were ranged such that they cannotread the higher value. Additional data was taken at higher powerranges and the meters came on scale with no issue. CR 118246 waswritten.-2-PAT-5.4 Calibration of Steam and Feedwater Flow lnstruments at30% Power was completed with all criteria met.-2-PAT-1.6 Startup Adjustments of Reactor Control System, wascompleted. This was data taking only with no Review or AcceptanceCriteria at this plateau.-2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete with all criteria met.-RCl-159, Radiation Baseline Surveys completed. No Acceptance orReview Criteria were associated with this procedure.-2-PAT-5.0, Test Sequence for 30% Plateau, was TRG approved.-2-PAT-6.0, Test sequence for 50% Plateau, performance sectionwas entered and power increase to 50% Plateau level initiated.-U-2 turbine tripped due to loss of 28 Main Feedwater Pump fromloss of MFP condenser vacuum with a subsequent automatic reactortrip as a result of the S/Gs reaching their low-low trip setpoint. Theplant was stabilized in Mode 3.-U-2 re-entered Mode 2.-U-2 re-entered Mode 1 and synchronized to the grid.-U-2 manually tripped the turbine due to a steam leak. Mode 2 wasentered and subsequently the reactor was tripped and the unitstabilized in Mode 3.-U-2 again entered Mode 2 and reactor critical at 03:20.-Unit entered Mode 1 at07:57.-U-2 synchronized to the grid at 13:36.6117 11661201166t23t16612411661261167121166115/1 618 3.0 Watts Bar Unat 2 Startup Ghronology (continued)717116 -U-2 reached 50% Plateau power level requirements for testing.-2-PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-PAT-1.1 1, RVLIS Performance Test, applicable sectionscompleted with all criteria met.-2-PAT-1.12, Common Q PostAccident Monitoring System,applicable sections were completed with all criteria met.718116 PAT-1.4, Pipe Vibration Monitoring, completed with al! criteria met.-2-PAT-1.6, Startup Adjustments of Reactor Control System,completed. This performance was data taking only with noAcceptance or Review Criteria at this plateau.-2-P AT -1. 7, Operational Al ig n ment of Process Tem peratu relnstrumentation, applicable sections completed with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems, completed withtwo issues being referred to Site Engineering for evaluation.Engineering review indicated it was acceptable to continue PowerAscension Testing. (See Problem Report 4 of 2-PAT-1.8).-2-PAT-1.10, lntegrated Computer System (lCS), completed with allcriteria met.-2-PAT-6.3, Calibration of Steam and Feedwater FIow lnstruments, at50% Power completed with all criteria met.719116 PAT-3.3, RCS Flow Measurement, completed with all criteria met.7113116 PAT-6.1, Automatic Reactor Control System, completed with allcriteria met.7114116 PAT-5.2. Turbine Generator Trip With Coincident Loss of OffsitePower Test, completed with all Acceptance Criteria met.CR 1192287 was written to document Tcold decreasing below the547'F Review Criteria.-Unit 2 entered Mode 3.-2-PAT-1.4, Pipe Vibration Monitoring, applicable section for transienttesting was completed with all criteria met.7116116 PAT-6.2. AutomaticSteam GeneratorLevel Control Transientscompleted with all criteria met.-2-PAT-6.0, Test Sequence for 50% Plateau, was approved by TRG.7117116 -Unit 2 re-entered Mode 2 after a planned trip with 2-PAT-5.2, TurbineGenerator Trip Coincident With Loss of Offsite Power Test.7118116 -Unit entered Mode 1 and synchronized to the grid.L9 3.0 Watts Bar Unat 2 Startup Chronology (continued)7119116 PAT-7.0, Test Sequence for 75o/o Plateau performance sectioninitiated.-2-PNf-1.2,Load Swing Test, was completed with allAcceptanceCriteria met. CR 1193637 was written for the Review Criteria notbeing met for an undershoot of steam header pressure. The ReviewCriteria required an undershoot of no more than 25 psi and the actualwas 28.5 psi. This test was originally scheduled for the previous 50%Plateau testing, however, issues with the turbine IMP lN controlsprevented performance during that plateau. The unit was held at45o/o power on the ascension to the 75% testing plateau to performthis test.-2-PAT-1.4 Pipe Vibration Monitoring, applicable sections for the loadswing were completed with all criteria met.7125116 -Unit 2 reached 75o/o Plateau testing power Ievel.-2-PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-P AT -1. 1 0, I ntegrated Computer System (lCS), completed.CR 1195476 written for failure of Acceptance Criteria. MCR indicator2-Tl-62-Tl comparison to ICS PlDT0127A (Regen Heat ExchLetdown Temp) was not within the MED. The CR was closed aftercalibration of the instrument.-2-PAT-1.1 1, RVLIS Performance Test, applicable sectionscompleted with all criteria met.-2-PAT-1.12, Common Q PostAccident Monitoring System,applicable sections were completed with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems, completed withall criteria met.7126116 PAT-1.4, Pipe Vibration Monitoring, completed with CR 1195665written on excessive vibration on the Main Steam Line Trap drain line.Temporary repairs to stabilize the line were initiated. All other criteriawere met.7127116 PAT-3.3, RCS Flow Measurement completed with a!! criteria met.7128116 PAT-1.6, Startup Adjustments of Reactor Control System, wascompleted with all criteria met.-2-PAT-7.1, Calibration of Steam and Feedwater Flow lnstruments at75o/o Power, was completed. Steam Flow and Feedwater Flow dataobtained in Section 6.1 of this PAT on7l26l16 was used to adjust thespan of the associated Steam Flow transmitters. Post calibrationdata was subsequently taken in accordance with Section 6.2 of thisPAT on 7128116. and all Review Criteria were met. There was noAcceptance Criteria for this PAT.20 3.0Watts Bar Unat 2 Startup Chronology (continued)-2-P AT -1. 7, O peration al Al i g n ment of Process Tem peraturelnstrumentation, was completed with allAcceptance Criteria met andall Review Criteria met upon the second performance. CR 1196243and CR 1196245 were generated for the initial failures. On thesecond data collection all Review Criteria were met and the CRsclosed.-2-PAT-1.9, Automatic Steam Generator Level Control, wascompleted with all criteria met.-2-PAf -7.0, Test Sequence for 75o/o Plateau, was TRG approved.-2-PAT-8.0, Test Sequence for 100% Plateau, performance sectionwas initiated. Due to increasing generator bushing temperatures andconcerns for further power increase, the original sequence of testingwas revised to perform 2-PAT-8.5, Shutdown From Outside the MainControl Room. An outage after the PAT performance was plannedfor repairs to the generator bushing.-Unit 2 power reduced to approximately 30% RTP.-2-PAT-8.5, Shutdown From Outside The Main Control Room wascompleted with all criteria met. The Unit was held in Mode 3 forequipment repairs.-Repairs were completed and Unit 2 startup initiated.-Unit 2 entered Mode 2 at 12:22 and the reactor was critical at 12:31 .-Unit entered Mode 1 at 16:14.8110/1 6-Unit 2 synchronized to the grid at 06:12. A delay in synchronizationoccurred due to particles in the thrust bearing wear trip fluid whichrequired flushing multiple times.-During power ascension an issue with increased temperatures on Cphase main generator bushing developed. This temperature issuewas noted prior to the Shutdown from Outside the Main Control Roomand resulted in a second planned outage.-Unit 2 was manually tripped at 03:06 and stabilized in Mode 3 for aplanned outage.-Unit 2 reactor critical at02:28.-Mode 1 entry at 08:32.-Unit 2 synchronized to the grid at 13:53.8113/1 68t22t167129t16811116813116817 116819t162L 3.0 Watts Bar Unat 2 Startup Chronology (continued)8123116 -Unit 2 reactor was manually tripped at 13:56 when the 2A MainTurbine Driven Feedwater Pump slowed and failed to providesufficient flow to maintain steam generator levels. The unit wasstabilized in Mode 3.8125116 -Unit 2 entered Mode 2 at 14:27 and the reactor was critical at 14:46.-Unit 2 entered Mode 1 at approximately 17:25.-Unit 2 synchronized to the grid at 23:19.8129116 -Unit 2 at93o/o rated thermal power allowing PAT testing tocommence at the 90% plateau.-2-PAT-1.6, Startup Adjustments of Reactor Control System,completed. For the g0% Plateau data collection was completedand satisfactory for this plateau. CR 1208178 was initiated beforeperformance because 2-PT-1-81 was unavailable for the test due toa steam leak. Results were acceptable for continuation to the100o/o plateau where measurements were repeated.-2-PAT-1 .7, Operational Alignment of Process Temperaturelnstrumentation, Review Criteria 5.2.8 and 5.2.C were not met buta CR was not written as the 2-PAT-1.7 performance was designedto correct the issue and the Acceptance Criteria were verified at the100Yo power plateau. All other Review and Acceptance Criteriawere met.-2-PAT-8.4, Calibration Of Steam And Feedwater Flow lnstrumentsat 100% Power, performance at 93% completed. All ReviewCriteria were met for Section 6.1. There was no AcceptanceCriteria for Section 6.1.8/30/16 -Unit 2 at > 98% rated thermal power allowing PAT testing tocommence at the 100o/o plateau.-2-PAT-1.5, Loose Parts Monitoring System was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-PAT-1.6, Startup Adjustments of Reactor Control System, datacollection was field work complete before the unit tripped.Subsequently, CR 1211020 was written for one failed AcceptanceCriteria. The failed criteria was due to full load steam pressurebeing below the expected value because T"rnwas at its maximumvalue. However, there is no safety or operational concern.Additionally, CR 1211015 was written for failed Review Criteria . 2-PT-1-81 was out of service, therefore, calibrations of the pressuretransmitter will be verified when 2-PT-1-81 is returned to serviceoutside the PAT program. CR 1208178 was previously written forthis issue and WO 118121693 will resolve the issue.22 3.0 Watts Bar Unit 2 Startup Chronology (continued)8/30/16 PAT-1.T,OperationalAlignmentofProcessTemperaturelnstrumentation, data collection was field work complete before unittripped. Data reduction was completed with failure to meetAcceptance Criteria for Loop 4 Taw. On917116 CR 121'1021 waswritten documenting this failed Acceptance Criteria. Although afailure, there was no safety concern or failure to meet the licensingbasis. All Review Criteria were met on this performance of thePAT.-2-PAT-1.8, Thermal Expansion of Piping Systems, completed withallcriteria met.-2-PAT-1.10, lntegrated Computer System (lCS), was completedfor the 100o/o Plateau. CR 1208754 was generated for failure ofmeeting the MED between indicator 2-Tl-062-0004 and ICS pointT0181A, RCP 1 No 1 Seal Outlet Temperature. A WO wasgenerated to calibrate and has subsequently closed.-2-PAT-1.1 1, RVLIS Performance Test, was completed with allcriteria met.-2-PAT-1.12, Common Q PostAccident Monitoring System, wascompleted with all criteria met.-2-PAT-3.3, RCS Flow Measurement, data collection was field workcomplete before unit tripped.-2-PAT-8.4, Calibration of Steam and Feedwater Flow lnstruments,at 100% Power, was field work complete. CR 1208875 was writtento document failure of Review Criteria on three steam flowtransmitters. WOs were initiated to respan the transmitters.-Unit 2 received an automatic Turbine Trip - Reactor Trip at21:09:13 due to a fault in the 28 Main Bank Transformer, resultingin a fire in the transformer. The unit was stabilized andsubsequently placed in Mode 4 for repairs.9115116 PAT-8.6, Plant Trip from 100o/o Power, was evaluated from datagathered during the actual plant trip on 8/30/16. AllAcceptanceCriteria was met. CR 1209770 was written to evaluate theequivalency of the data collected by the plant as well as oneReview Criteria which did not meet pressurizer level modulation tono load setpoint within 30 minutes. A Westinghouse evaluationconcluded the response was acceptable.-2-PAT-1.4, Pipe Vibration Monitoring, Section 6.6.19 was closedbased on an engineering walkdown evaluation CR 1211196.9125116 -Unit 2 entered Mode 1 at 01:53 after the Spare Main BankTransformer was placed in service for the failed 28 Main BankTransformer which was removed from site.9126116 -Unit 2 generator synchronized to the grid at 01:07.23 3.0Watts Bar Unat 2 Startup Chronology (continued)91281169129116-Unit 2 reached 100o/o power.-RCl-159, Radiation Baseline Surveys completed for 100o/o Plateau.No Acceptance or Review Criteria were associated with thisprocedure.2-PAT-1.9, Automatic Steam Generator Level Control, was fieldwork complete with all criteria met.-2-P Nf -1. 7, Operational Al ig n me nt of Process Tem peratu relnstrumentation, was completed with failure to meet AcceptanceCriteria for Loop 4Taw. However, on917116 CR 1211021had beenpreviously written documenting this failed Acceptance Criteria.-2-PAT-1.2,Load Swing Test, was field work complete with allAcceptance Criteria met. CR 1218746 was written for failure of oneReview Criteria for S/G Level response. Westinghouse evaluatedthe response to be adequate with no further testing required.-2-PAT-3.3, RCS Flow Measurement, was completed with allcriteria met.-2-PAT-8.4, Calibration of Steam and Feedwater Flow lnstruments,at 100o/o Power, was field work complete with all criteria met.2-PAT-1.3, Large Load Reduction Test, was field work completewith allAcceptance Criteria met. CR 1218917 was written forReview Criteria failure of S/G levels to remain within t15o/o of theprogram level. Westinghouse concluded the response wasacceptable.2-PAT-1.4, Pipe Vibration Monitoring, was completed with CR1208694 written for main steam traps excessive vibration as wasnoted at the 75o/o Plateau also. Civil Design generated WO118122821to design and install a restraint outside the PATP.2-PAT-1.6, Startup Adjustments of Reactor Control System, wasfield work complete on 10/3/16. Previously, CR 1211020 waswritten for one failed Acceptance Criteria. The failed criteria wasdue to full load steam pressure being below the expected valuebecause T"rnwas at its maximum value. However, there is nosafety or operational concern. Additionally, CR 1211015 waswritten for failed Review Criteria. Due to 2-PT-1-81 being out ofservice calibrations of the pressure transmitter will be verified when2-PT-1-81is returned to service outside the PAT program. CRs'1208178 and 1216904 were previously written for this issue.-2-PAT-8.0, Test Sequence for 100% Plateau was TRG approved.9t29t169/30/1 610131169127 11610t611624 4.04.1INITIAL FUEL LOADOverview and Summary of lnitial Core LoadingThe initial core loading at WBN Unit 2 was accomplished in approximately76 hours from December 4,2015, to December 8,2015, as directed by2-PAT-2.0, Initial Core Loading Sequence.Core loading was performed "wet" with the refueling cavity and the reactor vesselfilled with refueling concentration borated water at normal refueling levels. Thecore loading sequence was performed in accordance with an approved FuelAssembly Transfer Form (FATF). Actual movement of fuet was performed inaccordance with 2-FHl-7, Fuel Handling and Movement, as directed by 2-PET-105, lnitial Core Loading.The neutron monitoring station for lnverse Count Rate Ratio determinations wereestablished in the main control room to monitor source range detectors N-31 andN-32. ICRR plots were maintained for these detectors during all core loadingsequence steps and during delays in core loading to ensure that an adequatesubcritica! margin was maintained at all times.As a visual aid in tracking fuel movement evolutions and to ensure the core loadconfiguration was in accordance with the approved loading pattern prescribed onthe FATF, a core status display was maintained in the Main Control Room.RCS boron concentration was monitored during core loading to ensure that theboron concentration remained within prescribed limits.Some fuel assemblies were required by plan to be moved more than once,specifically those bearing primary neutron sources. As such, the core loadingsequence required two in-core fuel assembly movements to move the sourcebearing fuel assemblies from the reactor baffle wall to their final locations withinthe core. This was done to ensure that neutron counts could be monitored by theSource Range instrumentation at all times during core loading. After the corewas loaded, a video recording was made and verification of proper fuel assemblyposition and orientation was conducted. Fue! Related Components (FRCs) wereconfirmed to be inserted into the proper fuel assembly with the proper orientationin the Spent Fuel Pool prior to core load. The final core load configuration wasconsistent with the Westinghouse Core Loading Plan for Unit 2 Cycle 1.25 4.2 lnitial Gore Loading Sequence (2-PAT-2.0)This test started on 11119115 with prerequisites and completed on 1218115.1.0 Test ObiectivesThe objectives of this test were to:1.1 Sequence the procedures that established the prerequisitesrequired for the initial core loading of Unit 21.2 Define the sequence of operations and tests which were to beconducted during and following completion of the initial coreloading.The following PATs/PETs/RCl were sequenced for performance by2-PAT-2.0:o 2-PAT-2.1 Reactor System Sampling for Core Loado 2-PAT-2.2 Response Check of Core Load lnstrumentation After8 Hour Delay in Fuel Movemento 2-PET-102 Pre-Power Escalation NIS Calibration Datao 2-PET-105 lnitial Core Loadingo RCI-159

Radiation Baseline SurveysNote:

lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.2.0 Test MethodsPre-requisite actions started on 11119/15, prior to entry into Mode 6 toestablish prerequisite conditions in support of commencement of initialcore loading. The test continued through verification of core loading andwas field complete on 1218115, prior to the reassembly of the reactorvessel in preparation for Mode 5 entry.The major pre-requisites included the following:o Verification all Preoperational Test completed and test resultsapproved or technicaljustifications for delaying tests unti! after fuelload were approved by the Plant Managero Verification 2-PET-102,Pre-Power Escalation NIS Calibration Data,was successfully completed to the extent necessaryo RCI-159 Radiation Baseline Surveys commenced for the pre-fuelload survey26 4.2 lnitial Core Loading Sequence (2-PAT-2.0) (continued)o 2-PAT-2.1, Reactor System Sampling for Core Load, startedo Visual lnspection of the reactor vessel core support plate inaccordance with 2-PET-105, lnitial Core Loading completed Testingincluded the following:o 2-PAT-2.1, Reactor System Sampling for Core Load, completed on1214115 with all criteria met.o 2-PAT-2.2, Response Check of Core Load lnstrumentation After 8Hour Delay in Fuel Movement, completed on 1215115 with allcriteria meto 2-PET-102, Pre-Power Escalation NIS Calibration Data, applicablesections completed with all criteria meto 2-PET-105, lnitial Core Loading completed on 1218115 with allcriteria met.o RCI-159, Radiation Baseline Surveys, completed on 1213115.There was no Acceptance or Review Criteria associated with thisprocedure.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.0 ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAf-2.0.21 4.3 Reactor System Sampling for Gore Load (2-PAT-2.11This test was performed as part of test sequence 2-PAT-2.0, lnitial Core Loading.Testing was started on 11119115 and field work completed on 1214115.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the boron concentrations in the Reactor Coolant System(RCS), Residual Heat Removal (RHR) system and other directlyconnected portions of auxiliary systems are uniformly borated toprevent inadvertent dilution during core loading.1.2 Verify un-borated water sources are configured to preventinadvertent dilution during core loading.1.3 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 4, ReactorSystem Sampling For Core Loading Test Summary.2.0 Test MethodsThe preliminary actions of 2-PAT-2.1 researched logs and procedurallydriven Unit 2 activities that were completed and that were associated withthe preparations of the unit 2 water systems for entering Mode 6. Theresearch started with ensuring the RWST was borated to (3100 to 3300)ppm. Actual recorded samples of the RWST were 3326 ppm on0910712015 and 3224 ppm on 1'112112015. This confirms a correctlyborated RWST. This RWST water was subsequently used to fill the RCSand partially fill the Refueling Cana! and Cavity. Later boron sampleresults showed Refueling Canal at 3290 ppm and Refueling Cavity at3281ppm.Following the proper boration of the RWST and water transfer to the RCS,unit activities were verified that circulated water through:Both RHR A-A and B-B pump miniflowsBoth RHR pumpsRHR to CVCS LetdownBoth Charging pumpsBoth Containment Spray pumps (re-circulated borated RWST)Refueling Water Purification Pump BRefueling Water Purification Pump A was found to be tagged with aCaution Order 0-CO-2015-0048 stating that the pump has high vibration.WO 116318322 was previously written to address this issue. The volumeof potentially diluted water was conservatively calculated to be 13 gals.This small volume did not pose any risk to challenging any criteria listed inthis PAT.28 4.3 Reactor System Sampling for Core Load (2-PAT-2.1) (continued)The Boron lnjection Tank (BlT) was verified to have been borated andmixed via the performance of 2-5l-63-905, Boron lnjection Check ValveFlow During Refueling Outages.It was verified that both trains of Safety lnjection were circulated during theperformance of 2-Sl-63-906, Safety !njection Check Valve Full FlowTesting During Refueling Outages, on 1112512015.All4 Cold Leg Accumulators were verified by sample to be corectlyborated with the lowest reading 3175 ppm and the highest reading 3206ppm.The water in the Holdup Tank B (HUT B) was recirculated and sampled forboron concentration on 11104115 and found to be 3204 ppm boron; thiswater was used to fill the Fuel Transfer Canal. The Spent Fuel Pool (SFp)was sampled for boron concentration on 11123115 and found to be 3261ppm boron. Boric Acid Tanks B and C were sampled for boronconcentration on 11121115 and found to be 6919 and 6808 ppm boronrespectively.Problems encountered while running 2-Sl-63-905 and 2-5!-63-906resulted in a partial drain down moving water back to the RWST. Thissame water was again used to fill the Refueling Cavity and Chemistry re-performed 2-Sl-78-1, Reactor Coolant System and Refueling CanalRefueling Operations Boron Determination, to document compliance withthe refueling boron concentration requirements.Watts Bar Unit 2 systems connected to the Reactor Coolant System(RCS) were adequately borated and mixed to prevent a dilution event insupport of the initial core loading operations.2-PAT-2.1, Reactor System Sampling For Core Load, supported thisconclusion from research of the chemistry and operation logs.configuration contro! measures were in place to ensure that the RCS andconnected systems remain adequately borated for support of the initialcore loading operations. The Unit 2 Refueling Water Storage Tank(RWST) was at approximately 16.9% and based on calculations of recentmakeup to the RWST it was determined that the tank was adequatelyborated. Technical Specifications required Operations to validatecompliance with the RWST boron concentration and level prior to Mode 6.Technical Specifications required Watts Bar Unit 2 to maintain Mode 6surveillance instructions in frequency. Therefore, no additional sampling ormixing was required for 2-PAT-2.1.29 4.3 Reactor System Sampling for Gore Load (2-PAT-2.1) (continued)3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 Boron Concentration of samples meet requirements of theTechnical Specifications.3.1.1 The RCS Boron concentration is greater than or equal to3100 ppm and less than or equal to 3300 ppm.The RCS Boron as measured in the RHR TRAIN B systemwas 3284 ppm.3.1.2 The boron concentration final samples obtained from thedesignated sample points identified are uniformly boratedbetween 3100 ppm and 3300 ppm.The boron concentration for sample points met therequirements.3.1.3 The boron concentration of samples obtained from the BoricAcid Tanks (BAT B and BAT C) are within the limits ot 6120< Ce < 6990 ppm.Boron concentrations were 6919 ppm in BAT B and 6808ppm in BAT C.3.1.4 Un-borated water sources are configured to preventinadvertent dilution during core loading.2-Sl-62-1, Uncontrolled Boron Dilution Paths, wassatisfactorily completed for Mode 6.Review Criteria3.2 The boron concentrations for the Reactor Coolant System (RCS)and directly connected portions of the auxiliary systems are greaterthan or equal to 3100 ppm and less than or equal to 3300 ppm.Boron concentrations for the RCS and directly connected portionsof the auxiliary systems met the requirement.4.0 ProblemsThere were no significant problems encountered during the performanceof this test.30 4.4 Response Check of Core Load lnstrumentation After 8 Hour Delay inFuel Movement (2-P AT -2.21This test was performed as part of test sequence 2-PAT-2.0, lnitial Core LoadingSequence. Testing was started and completed on 1215115.1.0 Test ObiectivesThe objective of this test was to:1.1 Verify response of the Source Range Channels prior to resumptionof fuel loading following a delay of eight (8) hours or more.2.0 Test MethodsThree methods of testing were available for use:2.1 Statistical Evaluation Method using the Scaler Timer2.2 Statistical Evaluation Method using the Source Range Count Rateindications.2.3 Response Check of Core Load lnstrumentation Using PrimarySource Bearing Fuel Assembly Movement.The Statistical Evaluation Method using the Scaler Timer provided theverification of the Acceptance Criteria for resumption of fuel movement.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 The Source Range instrumentation for both channel N-31 and N-32were evaluated and determined to be acceptable for continuation offuel loading by meeting at least one Review Criteria.Review Criteria 3.2, below31 4.4 Response Check of Gore Load Instrumentation After 8 Hour Delay inFuel Movement (2-P AT -2.2) (continued)Review Criteria3.2 Statistical Evaluation Method using the Scaler Timer:Statistical Reliability Factor (SRF) for Source Range Channels shallbe > 0.5 and 3 1.4.Results indicated the SRF for Source Range Channel N-31 was1.2395 and SRF for Source Range Channel N-32 was 0.8609.3.3 Statistical Evaluation Method using the Source Range Count Rateindications:1. The Student F Distribution Test shall be satisfied by havingFexp s 3.179.2. The Student T Distribution Test shall be satisfied by havingTexp s 2.101.This method was originally chosen, however, problems wereencountered. See Problems below.3.4 Neutron instrumentation (Source Range Channels N-31 and N-32)are operational and indicates a positive (negative) change in countrate as the neutron level detected from a source is increased(decreased).This method was not used.4.0 Problemst1l No CR initiated:Section 6.2, Statistical Evaluation Using Source Range Count Ratelndications, method was attempted four (4) times with unsuccessfulresults. Based on only three assemblies loaded at the time ofperformance, low counts appeared to cause data scatter which wasobserved in monitored count rates. This failure of Section 6.2method resulted in the transition to Section 6.1, StatisticalEvaluation Method using the Scaler-Timer. Section 6.1 methodwas acceptable. No CR was initiated since this was a potentialscenario and the test provided alternative methods.32 4.5 Pre-Power Escalation NIS Calibration Data (2-PET-1021This test was performed as part of the test sequence 2-PAT-2.0, lnitial CoreLoading Sequence. The performance of 2-PET-102 was conducted viaWO 116884907. The WO started 0812512015 and was complete on 04112116withcalibration of all Power Range and lntermediate Range detectors.1.0 Test ObiectivesThe objectives of this test were to:1.1 Provide Nuclear lnstrumentation System (NlS) Power Range (PR)and lntermediate Range (lR) excore detector calibration data.1.2 lnitiate an adjustment of the NIS before startup for a new fuel cycle.Note: The calculation methodology in 2-PET-102 applied to the changesexpected to occur due to a refueling outage. This procedureaccommodated the calibrations to be performed for Cycle 1.2.0 Test MethodsThe normal method for determining calibration data after fuel reload andprior to startup is to ratio the sum of selected weighted assembly predictedpowers from the Beginning of Life (BOL) of the previous fuel cycle (Unit 1Cycle 1 was used as the reference condition) to the BOL of the upcomingcycle. This ensures a ratio based upon similar BOL core conditionsincluding the neutron energy spectrum and a nearly cosine axial fluxshape. This provides the most accurate excore Axial Offset indications forthe power range channels. This same methodology results in the mostaccurate power indications for the intermediate range channels.This same methodology was used to predict the lntermediate Range andPower Range calibration setpoints for the Unit 2 Cyclel startup, exceptthat Unit 1 Cycle 1 is used as the reference condition. ln this case,average composite values for the channels were used.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.33 4.5 Pre-Power Escalation NIS Calibration Data (2-PET-102) (continued)Acceptance Criteria3.1 Power Range Channel calibrations have been completed.The Power Range channel calibrations were completed via:o WO 115898162o WO 115898208o WO 115898252o WO 1158991873.2 lntermediate Range Channel predicted full power adjustments havebeen completed.lntermediate Range Channel predicted full power adjustments wereperformed. lR Gain Adjustment potentiometers were set to thevalues calculated in the PET via steps in Section 7.0 of the PET.3.3 lntermediate Range Channel OperationalTests (COTs) have beencompleted.The lntermediate Range channel COTs were completed via:o WO 117499181o WO 117499184Review CriteriaNone4.0 ProblemsThere were no significant problems encountered during the performanceof this test.34 4.6 lnitia! Core Loading (2-PET-1 05)This test was performed as part of test sequence 2-PAT-2.0, lnitial Core LoadingSequence. 2-PET-105 testing via WO 117370408 started on 1112312015 with theverification of Unit 2fuel assemblies and component inserts in the Spent FuelPool and completed on 1210812015 with the completion of fuel load and core loadverification.1.0 Test ObiectivesThe objectives of this test were to:1.1 ldentify the activities and requirements for fuel loading whichensure fuel loading is conducted in a cautious and controlledmanner:1.1.1 Specify the sequence for loading fuel assemblies into thereactor vessel such that the final core configuration isconsistent with that specified in the NuPOP for current fuelcycle. See Figure 4.6-1, U2C1 Core Load Sequence.1.1.2 Specify the fuel assembly identification number and type ofinsert for each core location.1.1.3 Establish the requirements for periodic and continuousneutron monitoring during each step of the core loadingprocess.1.1.4 Prescribe the steps necessary for obtaining and evaluatingneutron monitoring data during core loading.1.1.5 ldentify the neutron monitoring channels to be used duringeach step of the core loading sequence to ensure subcriticalconditions are maintained.1.2 Satisfied the requirements of UFSAR Table 14.2-2, Sheet 3, lnitialFuel Loading Test Summary.2.0 Test MethodsOnly data from "responding" detectors identified by the data package wasused in evaluating the safety of continued core loading. Prior tocompleting the loading of the initial nucleus of eight fuel assemblies,significant changes in the ICRR data were expected to occur due togeometry effects arising from changes in detector-to-fuel assemblycoupling. Therefore, the ICRR values were re-normalized followingmovement of source bearing fuel assemblies from the baffle wall to theirfinal location(s) in the core.35 4.6 lnitial Core Loading (2-PET-1 05) (continued)Changes in neutron flux level during and following fuel assembly insertionwas monitored for indications of abnormal and/or unstable reactivitybehavior.All fuel movement was performed in accordance with 2-FHl-7, FuelHandling and Movement.The core status display in the main control room was updated, asrequired, to reflect the actual physical location of all fuel assemblies andfuel related components at alltimes during the core loading evolution.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Note: Unit 2 Core Load ICRR plot is provided in Figure 4.6-2.Acceotance Criteria3.1 The core was successfully loaded in accordance with the Unit 2Cycle 1 Westinghouse Core Load Plan.Verification of successful core loading was provided via2-Tl-28, Physical Verification of Core Load Prior to Vessel Closure,(WO 117370592.) See Figure 4.6-1.3.2 At completion of each mini-core, the final count rate from anydetector shall not unexpectedly double from the initial count ratebefore the assembly was inserted.Neutron count rates observed during fuel movement did notunexpectedly double at any time.3.3 At completion of each mini-core the ICRR response from anydetector shall not be less than 0.5 as each fuel assembly isinserted.Neutron count rates observed during fuel movement did notunexpectedly double at any time and ICRR remained above 0.5,see Figure 4.6-2.3.4 Core loading operations are required to be immediately stoppedand the Containment Building evacuated if any of the followingconditions occur during core Ioading. Movements of an activesource bearing assembly, or detectorto-fuel assembly neutroniccoupling are anticipated type changes.36 4.6 Initial Core Loading (2-PET-1 05) (continued)3.4.1 An unanticipated simultaneous increase in the neutron countrate by a factor of > 2 on all "responding" neutron monitoringchannels.3.4.2 An unanticipated simultaneous increase in the neutron countrate on any individual "responding neutron monitoringchannel by a factor of > 5.Neutron count rates were acceptable and did not meet eithercriteria to warrant suspension of core loading operations orevacuation of the Containment Building.Review CriteriaAl! required Review criteria for this test were met as delineated below:3.5 Assessment of the ICRR response should be based on thepredicted ICRR response.ICRR plots maintained during core loading activities contained bothactual plant lcRR data as well as predicted ICRR data from the fuervendor, see Figure 4.6-2.3.6 Placement of initialfuel assemblies up to placement of primarysource assemblies in final core location should be detected by theICRR response.ICRR monitoring was maintained at all times during core loading,including loading of the first "mini-cores" and final movement ofprimary source bearing fuel assemblies.3.7 ICRR response should not be less than 0.8 for any fuel assembtyafter the primary source assemblies have been placed in their finallocations.ICRR data during core loading was determined to be less than 0.8following final placement of the two source bearing assemblies. CR1112886 was initiated to document violation of this ReviewCriterion. Violation of this criterion does not represent a failure ofthis test, as it only requires further evaluation by the fuel vendor.The fuel vendor was notified and agreed that the data wasacceptable.31 4.6 Initial Core Loading (2-PET-1 05) (continued)4.0 Problemst1] CR 1112049 was written to document a labeling issue. SourceRange labeling difference was noted between 2-PET-105 and theUnit 2 Main Control Room. Urgent Change 1 was processed for2-PET-105 to correct the labeling issue.l2l CR 1 112886 was initiated to document violation of the 0.8 ICRRlimit during core loading, as described in Section 3.7.t3l CR 11 12204 was initiated to document foreign material, laterdetermined to be glue, on the bottom nozzles on multiple fuelassemblies during initial core load. All assemblies noted to havedebris were cleaned prior to being loaded in the reactor.Efforts to remove the debris caused schedule delays.38 4.6lnitial Core Loading (2-PET-1 05) (continued)FIGURE 4.6.1U2C1 Core Load SequenceUZC1 Core Load SequencePERFORMED BY I39 4.6Initial Core Loading (2-PET-1 05) (continued)ilUHL,21.11.00.90.00.70.60.50.10.30.20.10.0FIGURE 4.6.2Unatz Core Load ICRRUnit 2Core LoadInverse Count Rate Ratio (ICRR)80 100 120 1a0ll&ber of nrel tage&lieso !{-31. t{-32rll-31 Fredictedrll-32 Fredicted40 5.0 PRECRITICAL TESTING5.1 Post Gore Loading Precritical Test Sequence (2-PAT-3.0)This test started on 121912015 and was completed on 05 115116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Serve as controlling document for establishing the required pre-requislte conditions to permit testing following the completion of2-PAT-2.0.1.2 Govern the sequence of tests performed in Mode 6 through Mode3.1.3 lmplement testing deferred from Pre-Operational Test lnstruction,2-PTl-062-03, HFT Charging and Letdown documented in CR1075347 and CR 1085430.The following PATs/PETs/RCl were sequenced for performance by2-PAT-3.0:o 2-PAT-1.4

Pipe Vibration Monitoringo 2-PAT-1.6

Startup Adjustments of Reactor Control Systemo 2-PAT-1.7

OperationalAlignment of Process Temperaturelnstrumentationo 2-PAT-1.8

Thermal Expansion of Piping Systemso 2-PAT-1.11* RVLIS Performance Testo 2-PAT-1.12* Common Q Post Accident Monitoring Systemo 2-PAT-3.1 Control Rod Drive Mechanism Timing and CERPI!nitial Calibrationo 2-PAT-3.2 Pressurizer Spray Capability and Continuous SprayFlow Settingo 2-PAT-3.3

RCS Flow Measuremento 2-PAT-3.4 Rod Control and Rod Position lndication (CERPI)o 2-PAT-3.7 Reactor Coolant Flow Coastdowno 2-PAT-3.8 Rod Drop Time Measurement and Stationary GripperRelease Timing. 2-PAT-3.10 Reactor Trip Systemc 2-PAT-3.11 Adjustment of Steam Flow Transmitters at MinimalFlowo 2-PAT-5.1

Dynamic Automatic Steam Dump Controlo 2-PET-106 Control Rod Drive Mechanism Timingo RCI-159

Radiation Baseline Surveys4L 5.1Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)2.0Note:

lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on121912015 and completed on 1211112015 and included verification of thefollowing major items:o 2-PAT-2.0 lnitial Core Loading Sequence completed.o 2-GO-7 Refueling Operations performed concurrently with2-PAT-3.0.o 2-GO-10 Reactor Coolant System Drain and Fill Operationperformed concurrently with 2-PAT-3.0.o RCI-159 Radiation Baseline Surveys, commenced for post-fuel loadactivities.Testing was performed at eight defined plateaus including, ambient(<105'F), 250"F, 300'F, 360'F,400'F,450"F, 500'F, and 557'F.This report is a summary therefore see individual test packages forspecific details at each plateau.Ambient Plateau testing included the following:o RCI-159, Radiation Baseline Surveys, Post Fuel Load Survey - fieldwork complete on 12112115. No Acceptance or Review Criteria wasassociated with this procedure.o 2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable sections on 12l16/15 with all criteria met.o 2-PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete for applicable sections on 1/16/16. There was noAcceptance or Review Criteria for this portion of testing.o 2-PAT-3.10, Reactor Trip System, field work complete on 1120116with all criteria met.o PAT-3.1, Control Rod Drive Mechanism and CERPI lnitialCalibration, field work complete on 1124116 with all AcceptanceCriteria met after evaluation of current amplitudes on twelve lift coilsdetermined the results to be acceptable for the designed operationof the rod control system. CR 1128950 was written for high currentamplitudes and closed following Westinghouse evaluation thatdetermined the measurements to be acceptable. There was noReview Criteria for this PAT.42 5.1 Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)o 2-PAT-3.8, Rod Drop Time Measurement and Stationary GripperRelease Timing, Mode 5 Performance, field work complete forapplicable sections on 1124116 with allAcceptance Criteria met. CR1128964 was written due to the RDTC plots for each rod wereinverted from the expected response. This did not impactperformance of the test and was resolved prior to the Mode 3performance. There was no Review Criteria for this test.. 2-PAT-1.4, Pipe Vibration Monitoring, field work complete forapplicable sections on211l16 with all criteria met.PAT testing on the plant primary side was suspended on 1126116 untilplant conditions and surveillance completions allowed further testing.Condensate was placed on modified long cycle which allowed thecompletion of the applicable portions of 2-PAT-1.4.On 3/15/16 preparations began for entering Mode 4 and PAT TestCoordinators began reviewing and completing pre-requisites for Mode 4testing. Mode 4, RCS temperature >200'F and <350'F, entry was madeon 3/19/16.The 250"F Plateau included the following:o 2-PAT-'1.12, Common Q Post Accident Monitoring System, fieldwork complete for applicable section on 3121116 with all criteria met.There is no Review Criteria associated with this PAT.The 300"F Plateau included the following:. 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 3125116 with al! criteria met.. The plant entered Mode 3, RCS temperature ) 350"F, on 3/30/16 at23:14 to allow further Power Ascension Testing.The 360"F Plateau included the following:o 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 3/31/16 with all criteria met.o 2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for applicable section on 3/31/16 with all criteria met.. 2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable sections on 3/31/16 with all criteriaacceptable for continued heat-up.Plant heat-up to 400 degrees was initiated on 411116 at01:14.43 5.1 Post Gore Loading Precritical Test Sequence (2-PAT-3.0) (continued)The 400'F Plateau included the following:o 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 411116. CR 1156311 documented some testdata was not taken with the RCP start but was collectedsatisfactorily from the plant computer. Review criteria was not meton the Reactor Coolant Pump Combination testing, The RVLISsystem was updated with the new constants supplied byWestinghouse to correct the abnormality and documented incR 1156425.o 2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for applicable sections on 411116 with all criteria met.After completion of the 400'F Plateau testing, plant heat-up to 450"F wasinitiated at 09:45 and completed at 12:16 on 411116.The 450"F Plateau inctuded the following:o 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 411116 with all criteria met.o 2-PAT-1.12, Common Q PostAccident Monitoring System, fieldwork complete for applicable section on 411116 with all criteria met.o 2-PAT-1.8, Thermal Expansion of Piping Systems - field workcomplete for applicable sections on 4/1/16. Problem Report #1 wasinitiated within the test for seven snubbers not performing asexpected. Results indicated no issue with the snubbers andapprovalwas received to continue to next plateau testing.Due to plant issues concerning check valve leakage, the decision wasmade on 412116 to cool down the RCS and make entry into Mode 4 toallow repairs. Mode 4 entry was made on 412116 at 06:29. Additionally,repairs on two RCS RTDs were made.Mode 3 re-entry was made on 418116 at 12:44.Plant condition of RCS temperature at 500'F was met on 4110116.The 500'F Plateau included the following:. 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 4110116 with all criteria met.44 5.1Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)Plant heatup to 557'F was completed at 15:30 on 4113116 with normaloperating pressure reached at 01:30 on 4114116. On 4117116 at 03:38 theunit was placed in Mode 4 for repairs to the Auxiliary Feedwater Pumpsand replacement ol PD07-2 shim determined to require adjustment. TheUnit was returned to Mode 3 on 5/1/16 at 17:36 and normal operatingtemperature and pressures on 512116 at 23:00.The 557'F Plateau included the following:c 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable section of 557'F data taking only on 4113116 with allcriteria met for steady state data collection. Additionally, Section6.1.3, Pump Combinations at 557'F, was field work complete on518116. Results (Section 6.1.4) indicated Acceptance Criteria wouldnot be met. The system was updated with the new RVLISconstants supplied by Westinghouse to correct the abnormality anddocumented in CR 1171130.o 2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for Data Collection Section 6.7 on 4113116 with allcriteria met. Section 6.8, Pump Contact Data Collection at 557'Fwas completed on 518116 with all criteria met.. 2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable sections on 4114116 with one issue outsidecontainment on PD07-2 which also required evaluation of PD07-1.Problem Report #2was initiated to resolve the issue with PD07-2and investigate any possible issues with PD07-1. AdditionallyProblem Report #3 was written to evaluate components not movingas expected. Both Problem Reports were closed and conditionswere acceptable to continue testing.o 2-PAT-1.4,Ptpe Vibration Monitoring, field work complete forSection 6.5.1 Pressurizer Surge - Mode 3, on 4114116 with allcriteria met for that section. Section 6.5.2, Main Feedwater Pump2A Start and Steady State Operation on Recirc., was field workcomplete on 4113116 with velocity and displacement AcceptanceCriteria not met. CR 1161783 was initiated for an engineeringevaluation which concluded acceptable as is. Section 6.5.3, MainFeedwater Pump 28 Start and Steady State Operation on Recirc.,was field work complete 5/5/16 with steady state velocity anddisplacement exceeding the Acceptance Criteria. CR 1168287 waswritten for an engineering evaluation and resulted in adjustment ofa loose hanger and a retest. The retest was completed on 5113116with satisfactory results. Section 6.5.4, Turbine Bypass Valve2-FCV-1-105 Transient was completed on 5112116 with all criteriamet. Section 6.5.5, Turbine Bypass Valve 2-FCV-1-111 Transientwas completed on 5112116 and re-tested on 5/13/16. Engineeringevaluation of the retest indicated satisfactory results.45 5.1Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)CR 1170319 documents engineering evaluation to accept-as-isfollowing the retest. Section 6.5.7 Condensate - Long Cycle wasfield work complete on2l1116 with all criteria met. There was noReview Criteria for 2-PAT-1.4.2-PAT-3.2, Pressurizer Spray Capability and Continuous SprayFlow Setting - Section 6.1, Adjustment of the Pressurizer ManualSpray Bypass Valves, was completed on 4116116. AllAcceptanceCriteria was met. Review criteria for MCR alarms was not met withCRs 1161382 and 1160969 written. A Westinghouse evaluationdetermined the PAT met the operability and design requirementsfor the pressurizer spray system. Additionally, CR 1161789 waswritten for proportional heater band not within the specifiedrequirement which was not an Acceptance Criteria.2-PAT-3.3, RCS Flow Measurement, field work complete on 5/3/16with all criteria met.2-P AT -1. 7, Operational AI i g n ment of Process Te m peratu relnstrumentation, field work complete on 5/6/16 with allAcceptanceCriteria met. One Review Criteria was not met and CR 1168641was initiated.2-PAT-1.6, Startup Adjustments of Reactor Control System, fieldwork complete for Mode 3 on 514116. This performance was datataking only.2-PAT-3.4, Rod Control and Rod Position lndication (CERPI), fieldwork complete on 5113116. The Acceptance Criteria was not met inSections 6.4 and 6.10. CRs 1 168845, 1168881, and 1169602 werewritten to document failure to meet criteria. The criteria was re-evaluated and it was determined the acceptance criteria should bechanged to require each Rod Position lndication to indicate rodmotion consistent with the group demand indication for the fullrange of rod travel. A change to the Westinghouse AcceptanceCriteria and SAR Change Package No. U2-019 were approved andan urgent change to the procedure incorporated the revisedAcceptance Criteria. All Acceptance Criteria for the final packagewere met.2-PAT-3.0, Attachment't,Testing Deferred from 2-PTl-062-03,- field work complete on 517116 with allAcceptance Criteria met.CR 1168487 was written to document alternate charging flow wasnot within anticipated range, however, it had no affect on the testacceptance.2-PAT-3.8, Rod Drop Time Measurement and Stationary GripperRelease Timing, field work complete on 5/11116. CR 1169659 waswritten for two rods failing a two sigma statistical evaluation. Threeadditional rod drops were performed and allAcceptance Criteriawas met. There was no Review Criteria for this test.46 5.1 Post Core Loading Precritical rest sequence (2-PAT-3.0) (continued)o 2-PAT-3.11, Adjustment of Steam Flow Transmitters at Minima!Flow, field work complete on 517116 with all Review Criteria met.There was no Acceptance Criteria associated with thisperformance.o 2-PAT-3.7, Reactor Coolant Flow Coastdown, field work completeon 5/8/16 with all criteria met. CR 1169224 was written todocument during removal of an instrument recorder a blown fusecaused alarms in the Main Control Room.o 2-PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete on 5112116 with all criteria met after a volume boosteradjustment with CR 1170159 and retest on 2-FCV-1-108.3.0 Test ResultsAcceptance/Review Criteria were contained within the test sequenced bythis test, except for Attachment 1, Testing Deferred from 2-PTl-062-03.Attachment 1 required Acceptance Criteria were met as delineated below.Acceotance Criteria3.1 Sum of RCP seal injection flow s 40 gpm, (6-13 gpm for each RCp).A. 2-Fl-62-1A = 9.0 gpm (6.6-12.4 gpm)B. 2-Ft-62-14A= 9.2 gpm (6.6-12.4 gpm)C. 2-Ft-62-2tA= 9.3 gpm (6.6-12.4 gpm)D. 2-Ft-62-40A = 9.2 gpm (6.6-12.4 gpm)Total Seal lnjection Flow Rate in gpm. = A + B + C + D =9.0+ 9.2 + 9.3 + 9.2 = 36.7 gpm3.2 The differential pressure across the following component at thegiven flowrate:DescriptionUNIDFIow RateAP(GIean)ActualAPSeallnjectionFilter B2-FLTR-62-9616-40 gpms7 psid6.0 psid3.3 lndication light 2-Xl-62-93 in MCR illuminated when2-HIC-62-93B was in manual.41 5.1 Post Gore Loading Precritical Test Sequence (2-PAT.3.0) (continued)3.4 2-FM-62-93E prevented 2-FCV-62-93 from going fully closed toensure Sea! Water flow rate of 33.5 11.5 gpm (32-35 gpm).Flow was 32.2 gpm.Review CriteriaNone4.0 Problemst1l CR 1168487 was written on 2-PAT-3.0, Attachment 1, Step 27.Although not Acceptance Criteria, alternate charging header flowwas anticipated to be approximately 89-103 gpm. Actual flow was82.7 gpm. This information was forwarded to engineering forevaluation, however, it had no affect on the acceptance of this test.Engineering evaluation calculated the minimum requirement at theregenerative heat exchanger temperature to be 74 gpm. The 82.7gpm exceeds this amount. At the current conditions the test wasacceptable and met the design specified criteria.Additional problems encountered are addressed in the followingdiscussions of each test sequenced by 2-PAT-3.0.48 5.2 Gontrol Rod Drive Mechanism Timing and CERPI Initia! Calibration(2-PAr-3.1)Performance of this test was directed by 2-PAT-3.0, Post Core LoadingPre-criticalTest Sequence, during the period from 1121116to 1124116. The testwas performed in Mode 5 at a RCS temperature of approximately 175'F.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the functionality of each CRDM for shutdown and controlrods in Mode 5 by:1.1.1 Verify each rod control system slave cycler provides itsassociated power cabinet with the appropriate commandsignal to obtain proper sequence timing of current suppliedto the CRDM coils.1.1.2 Verify CRDM coil current amplitudes are within acceptableranges.1.1.3 Verify the functionality of each shutdown and control roddrive mechanism.1.1.4 Verify manual mode stepping rate for shutdown and controlrods are within acceptable ranges.1.2 Verify the control bank overlap function in manual with minimaloverlap.1.3 Perform the initial calibration of the RPI in accordance with vendorproced u re WNA-TP-02576-WBT, C E RP I Calibration Proced ure.1.4 Partially satisfu the requirements of UFSAR Table 14.2-2, Sheet 7,Control Rod Drive Mechanism Timing Test Summary, and fullysatisfo it for Mode 5.2.0 Test MethodsThe CRDM functionality was verified by stepping ouUin each rod bank byapproximately 10 steps in individual bank select mode. CRDM currenttiming and amplitude measurements were taken during rod motion.Twelve of the CRDM amplitudes were outside the upper AcceptanceCriteria of the lift coil reduced current, however, the amplitudes wereevaluated as acceptable by Westinghouse.49 5.2 Gontrol Rod Drive Mechanism Timing and CERPI lnitial Galibration(2-PAT-3.1 ) (continued)The bank overlap circuitry was verified at minimal settings. Minimalsettings were set by adjusting the bank overlap thumbwheel switches suchthat control bank tip-to-tip distance was 15 steps and the all-rods-outposition was 25 steps withdrawn for each control bank. The bank overlapcircuitry functioned as designed and no issues were encountered. Notethat the bank overlap circuitry was also exercised during the performanceof the initial RPI calibration with the all-rods-out position set to 230 stepswithdrawn. The bank overlap circuitry functioned as designed with noissues.The initial RPI calibration was performed. First the bank zero adjustmentswere performed with all rods fully inserted. Next all shutdown and controlrods were withdrawn to the full out position of 230 steps withdrawn. Theshutdown banks were withdrawn first in individua! bank select and thecontrol rods were withdrawn in bank overlap. The bank position spancalibration and temperature null adjustments were performed with the rodsfully withdrawn. Next all control and shutdown rods were inserted tospecific demanded positions and data for each rod was obtained. Lastly,linearization adjustments were calculated based on the recorded data.The initia! RPI calibration was completed when the new linearizationadjustments were uploaded to train A and B of the RPI system. Note thatprior to the linearization adjustment and during the insertion of ControlBank B, the K14 and P6 rods had a 13 step rod-to-rod deviation whileinserted between CBB demanded positions of 170 steps withdrawn to 126steps withdrawn. The RPI system was not yet calibrated, therefore, theinitial calibration corrected the issue. Also note that 2-PAT-3.8 wasperformed following the initial RPI calibration and all rods were pulled toapproximately 50 steps withdrawn. All rods were within t2 steps of thedemanded position.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 Contro! Rod Drive Mechanism Timing3.1.1 Current Order TimingThe times at which the lift, movable, and stationary currentorders change, after the start of rod motion, are within10 msec. of the expected times during rod withdrawal andinsertion operations.50 5.2 Control Rod Drive Mechanism Timing and CERPI lnitial Calibration(2-PAT-3.1 ) (continued)Each CRDM current order timing was reviewed and allcurrent order timings were within 10 msec. of the expectedtimes.3.1.2 Coil Current AmplitudesStationary, movable, and lift currents are regulated bycircuitry internal to each power cabinet. The reduced andfull current nominal values are not critical, cannot beadjusted, but could be an indication of a regulation failure.Measured values outside the nominal ranges should beevaluated and documented by the system engineer.Lift Coil - FullNominal 4A amperes(35 to 47 .2 amperes)(equivalent to 438 to 590mVdc measured across a0.0125 ohm resistor)Lift Coil - ReducedNominal 16 amperes(13 to 19.7 amperes)(equivalent to 163 to 246mVdc measured across a0.0125 ohm resistor)Movable Gripper Coil -FullNominal 8 amperes(7 to 9.2 amperes)(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)Stationary Gripper CoilFullNominal 8 amperes(7 to 9.2 amperes)(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)Stationary Gripper Coil -ReducedNominal 4.4 amperes(3.8 to 4.8 amperes)(equivalent to 238 to 300mVdc measured across a0.0625 ohm resistor)51 5.2 Control Rod Drive Mechanism Timing and CERPI Initial Galibration(2-PAT-3. 1 ) (continued)Each current amplitude recorded in the test package werereviewed. All current amplitudes were within the AcceptanceCriteria with the exception of twelve CRDMs for the lift coilreduced current. The twelve lift coi! reduced currentamplitudes were evaluated and determined to be acceptablefor the designed operation of the rod control system.CR 1128950 was closed.3.1.3 Rod Withdrawal SpeedShutdown bank withdrawal speed nominal value is 64 (62 to66) steps per minute.Control bank withdrawal speed nominal value is 48(46 to 50) steps per minute.The Shutdown and Control banks withdrawal speeds mettheir Acceptance Criteria and were recorded as63.9 steps/min and 48.0 steps/min respectively.3.1.4 Rod Drive Mechanism OperabilityShutdown rod drive mechanisms operate with no indicationsof problems during the withdrawal and insertion stepping.Control rod drive mechanisms operate with no indications ofproblems during the withdrawal and insertion stepping.Each CRDM current trace was reviewed. Alltraces operatednormally and no abnormalities, such as movable/stationarygripper dragging or rod misstepping, were identified.3.2 Control Bank Overlap Demonstration3.2.1 The control rod bank overlap circuitry functions properlyduring the sequential withdrawal and insertion of ControlBanks in MANUAL mode.The control bank overlap circuitry functioned as designed.52 5.2 Contro! Rod Drive Mechanism Timing and GERPI lnitial Calibration(2-PAT-3. 1 ) (continued )3.2.2 The MCR rod speed demand display functions properly andindicates the rod stepping rate was within the range of 46 to50 steps/minute for Control Banks in Manual mode.The MCR rod speed demand display functioned as designedat 48.0 steps/min.3.2.3 The MCR group step counters function properly to indicategroup position and direction of rod motion during rodwithdrawal and insertion operations.The MCR group step counters functioned as designed.3.2.4 The MCR RPI functions properly to indicate individua! roddirection of motion during rod withdrawal and insertionoperations.The MCR RPls indicated the proper direction of motionduring rod withdrawal and insertion operations.3.2.5 The MCR rod direction indicator lights function properly toindicate the rod movement status and direction of rod motionduring rod withdrawal and insertion operations.The MCR rod direction indicator lights functioned asdesigned.Review CriteriaNone4.0 Problemst1] During the performance of the CRDM timing and amplitudemeasurements, the 1AC power cabinet- stationary group A coilamplitudes were lower than the expected value. WO 111522924was performed to inspect and reform backplane connector and cardedge pins for the 1AC power cabinet - stationary group A firing,regulation, and phase control cards. The issue was corrected andtesting continued.s3 5.2Control Rod Drive Mechanism Timing and CERPI lnitial Calibration(2-PAT-3. I ) (conti nued )l2lStep 6.3.3[7.1], WNA-TP-02576-WBT, Revision 2, Step 2.4.2.5,theall-rods-out position was 230 steps withdrawn, however, thecompensated position in the software was hardcoded to 231 stepswithdrawn. The performance of Step 2.4.2.5 was not impactedbecause it listed 230 steps t1 step. CR 1128373 was written andconcluded no changes to 2-PAT-3.1 were required.CR 1128918: Step 6.3.4[62], WNA-TP-02576-WBT, Revision 2,Appendix A.1 and A.2 forms were used for the linearizationadjustments. The "X Table C1" column values were not Watts BarU2 specific values. The Watts Bar U2 plant specific "X Table C1"values were used for the linearization adjustments. CR 1128918was written. Resolution was for Westinghouse to revise WNA-TP-02576-WBT.CR 1128950 Two rod amplitude measurements (F14 and D08)failed the procedure Acceptance Criteria and do not meet theWestinghouse expanded acceptance criteria in WBT-D-il20.Westinghouse has provided a letter (WBT-D-5604(3.8))documenting their evaluation and the acceptability for the 2 rodlocations that exceeded the Acceptance Criteria of 20 amps.Note that additional rod measurements were outside of theReduced Lift Current procedure Acceptance Criteria of 19.7 amps,however, WBT-D-5420 has been issued by Westinghouse thatstates reduce lift currents up to 20.0 amps are acceptable.2-PAT-3.1, Rev. 2, allows for evaluation of the currents outside ofthe Acceptance Criteria for successful completion of 2-PAT-3.1.Current orders outside of the Acceptance Criteria were evaluatedand deemed acceptable per Westinghouse. Also additionalmeasurements were obtained in Mode 3.During the performance of 2-TRl-85-1, Reactivity Control SystemsMovable ControlAssemblies (Modes 3, 4 and 5), rods common tothe 2BD power cabinet would not withdraw. Troubleshootingdetermined an issue with the 2BD movable gripper currentamplitudes. The firing and regulation cards for the 2BD movablegrippers were replaced with spares and the issue was corrected.No problems with this power cabinet occurred during 2-PAT-3.1testing.t3ll4It5I54 5.2Control Rod Drive Mechanism Timing and CERPI Initial Calibration(2-PAT-3.1 ) (continued)t6IStep 6.1l12lof Appendix E, rods common to the 1AC powercabinet - stationary group A did not have the expected reducedstationary gripper currents when the CRDM-DAQ was firstconnected. Under CR 1126661 and WO 117522924, the stationarygroup A firing, regulation, and phase cards were removed and bothbackplane connector and card edge pins were reformed. The issuewas corrected and testing was completed. No other problems withthis power cabinet occurred during 2-PAT-3.1 testing.During the performance of 2-TRl-85-1 with the 2-RBSS in the SBCand SBD positions, the CERPI monitor indicated 72 steps/min. Theactual speed of the SBC and SBD groups is approximately 64steps/min and is set at the SCD power cabinet. The indication didnot invalidate the performance of this test. CR 1126783 waswritten and closed to WO 115966328.Step 6.3.4[3], during SBD insertion, the SBC demand position onboth the ICS and RPI monitors followed the SBD demand position.The problem was due to communication issues between the rodcontrol system and the ICS. Testing continued because this issuedid not invalidate the performance of this test. CR 1128318 waswritten, and Post lssuance Change (PlC) 66181 was issued andimplemented by WO 117546244.Step 4.3[10]A, the 28 MG Set failed to sync in parallel with the 2AMG Set. The issue did not invalidate the performance of 2-PAT-3.1because only one MG set was required for the performance of thistest. CR 1126798 was written, and closed to WO 117531764.17It8l55 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-PAr-3.2)Performance of this test was directed by 2-PAT-3.0, Post Core LoadingPre-critical Test Sequence, during the period from 4115116to 514116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the pressure response to the opening of both normalPressurizer Spray Valves was within the allowable range specifiedby NSSS performance curyes.1.2 Verify the Pressurizer Bypass Spray Valves were throttled to anoptimum position such that during steady state operation:1.2.1 Spray line temperature was high enough to prevent the PZRSPRAY TEMP LO alarm from actuating.1.2.2 The equilibrium temperature for each spray line was greaterthan or equal to 540F.1.2.3 Pressurizer control bank heaters can maintain RCS pressureabove 2220 psig without backup heater operation.1.2.4 Surge line temperature was high enough to prevent the PZRSURGE LINE TEMP LO alarm from actuating.1.3 Verify the PZR SPRAY TEMP LO alarm would actuate ondecreasing spray line temperature of approximately 530F.1.4 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 13,Pressurizer Spray Capability And Continuous Spray Flow Setting TestSummary.55 5.3Pressurizer Spray Capability and Continuous Spray FIow Setring(2-P AT -3.2) (conti nued )2.0 Test Methods3.0This test established the optimal throttle positions for the PressurizerSpray Manual Bypass Valves, and also ensured the effectiveness of thenormal pressurizer spray by initiating full spray to reduce RCS pressure byapproximately 250 psi and compared the time to reduce pressure withWestinghouse performance curyes. During the performance of Section6.2, the validated RCS pressure DCS computer point being used tomonitor the depressurization of the RCS stopped updating at acceptablerate. Because of this the RCS narrow range indicators on the controlboard were used to determine when RCS pressure reached the triggervalue of 2000 psig. Subsequent review also determined that this pointdeviated further from the actual RCS pressure after it was no longer beingmonitored. This did not affect the ability to meet the Acceptance Criteriaof the test as ICS computer points were collected for use to analyzecompliance with Acceptance Criteria for the spray capability test. CR1168255 documents this issue.The spray line temperature low alarm was unable to be verified asintended during the performance of this test due to slight leakage pasteither the spray line FCV's or the spray bypass manual valves. Thisoccurred on both loops 1 and 2. This condition prevented meeting theReview Criteria associated with the spray line temperatures. Theoperation of the spray line temperature switches for each loop weresubsequently verified to be operating correctly by utilizing trend data fromthe plant computer. CRs 1 160969 and 1 161382 document this issue.Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceptance Criteria3.1Pressurizer pressure response to opening both Normal PressurizerSpray Valves is within the allowable range specified by NSSSperformance curves.The pressurizer spray response data was within the allottedresponse time as depicted on Figure 5.3-1.51 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-P AT -3.2) (conti nued )Review Criteria3.2 Pressurizer Manual Spray Bypass Valves 2-BW-68-552 and2-BW-68-555 are throttled to an optimum position during steady-state operation.All procedural criteria was met:1.Spray line equilibrium temperature is high enough to preventAnnunciator 2-><A-55-5A-89E, PZR SPRAY TEMP LO fromactuating.2-XA-55-5A-89E, PZR SPRAY TEMP LO, did not actuate.Equilibrium temperature for each spray line is greater than orequal to 550'F:Loop 1 Spray Line Temperature (lCS PID T0484A)Loop 2 Spray Line Temperature (lCS PID T0483A)Spray Iine equilibrium temperatures were > 550oF.Pressurazer control bank heaters can maintain RCS pressureabove 2220 psig without any Backup Heater operation.Backup Heater operation was not required to maintainpressure.Surge line equilibrium temperature is high enough to preventAnnunciator Alarm 2-XA-55-5A/89D, PZR SURGE LINETEMP LO, from actuating.2-KA-55-5A/89D, PzR SURGE LINE TEMP Lo, did notactuate.2.3.4.3.3Annunciator Alarm 2-XA-55-5A/89E, PZR SPRAY TEMP LOactuates on decreasing spray line temperature of approximately530'F (525"F to 535'F).Loop 1 spray line temperature would not decrease sufficiently toallow the low spray line temperature alarm to actuate. See CR1 160969.Loop 2 spray line temperature would not decrease sufficiently toallow the low spray line temperature alarm to actuate. See CR1161382.58 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-P AT -3.2) (conti nued )4.0 Problemst1] CR 1160969 was written because the loop 1 spray line temperaturewould not decrease sufficiently to allow the Iow spray linetemperature alarm to actuate. Westinghouse evaluation wasobtained which concluded that 2-PAT-3.2 met operability anddesign requirements for the pressurizer spray system.l2l CR 1161382 was written because the loop 2 spray line temperaturewould not decrease sufficiently to allow the low spray linetemperature alarm to actuate. Westinghouse evaluation wasobtained which concluded that 2-PAT-3.2 met operability anddesign requirements for the pressurizer spray system.13] CR 1161789 was written due to the master pressure controlleroutput being less than the desired range while setting the loop 2spray line bypass valve. The controller output was 37 percent whilethe desired range was a minimum of 46 percent. Performance ofSection 6.1.3 which performed the final setting of the spray bypassvalves, allowed the output of the master controller to be placed inthe desired range of the procedure.t4] CR 1168255 was written to document the issues experienced withthe ICS computer point for Validated Pressurizer Pressure(DCS0426) during the performance of Section 6.2.59 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-PAT*.2) (continued)FIGURE 5.3.1Pressurizer Spray Response2-PAT-3.2 Pressu rizer Spray Ca pa bility-50A.IlaeYoLa -10oobrLohlE -1soUl6oLc*-200-2502060 80Time (seconds!lower - - upper100- lcs DataL20140(-nQm (- -\\\\\L1\\\t:Il-ItrIFNIFII-'-1\\\\l-LT\\\\\IrLtrN\ri\-i\\\!L-vt:\:--- -; -\\rjj1ili r\- ---l. -- -{i--i---:il+\i\,,\\!L ----- i\ll\r*t60 5.4Rod Gontrol and Rod Position lndication (CERPI) (2-PAT-3.4)This test was performed in Mode 3 at NOTP as directed by 2-PAT-3.0, Post CoreLoading Precritical Test Sequence. lt performed the initial hot calibration of theComputer Enhanced Rod Position lndication (CERPI) system and functionaltesting of the Rod Control System. The performance of Section 6.0 of this testwas commenced on 05/05/16 and was completed on 05113116.1.0Test ObiectivesThe objectives of this test were to:1.2Perform the lnitial Hot Calibration of the Computer Enhanced RodPosition lndications (CERPI) system.Verify the Computer Enhanced Rod Position lndication system(CERPI) performs required indication function satisfactorily for eachshutdown and control rod over their entire range of travel and toverifo the rod position indication system alarm functions operateproperly. (UFSAR Table 14.2-2, Sheet 8, Rod Position lndicationSystem Test Summary).1.12.01.3 Demonstrate that the rod control system satisfactorily performs therequired control and indication functions, as required by UFSARTable 14.2-2, Sheet 10, Rod Control System Test Summary.Test MethodsThe rod position indication system completed the initial hot calibrationusing vendor instructions and 2-SI-85-3, Calibration of ComputerEnhanced Rod lndication Channels and Full Range Verification. TheCERPI system operated over the full length of travel and can operatewithout actuating rod-to-rod and rodto-bank deviation alarms by makingadjustments to the CERPI tunable parameters. This was consistent withvendor and Unit 1 operating experience. Therefore, the vendor CERPIAcceptance Criteria was revised and Urgent Change (UC) 2 wasprocessed for this test procedure to veriff that each rod indicates rodmotion consistent with the group demand over the full length of trave!.The Rod Control System and CERPI functionaltesting included controlsand indications. The functional testing included rod-to-bank and rod-to-rod deviation alarms, the C-11 annunciator, lntegrated Computer System(lCS) generated alarms, rod bottom bistables, rod bottom bypassbistables, rod control urgent and non-urgent alarms, main control roomdisplays, rod insertion limits, and control rod bank overlap circuitry.This test also documented 5 complete rod excursions (i.e., fullwithdrawaland insertions) of all shutdown and control rods per CR 234483. All 5excursions were successfully completed.6L 5.4 Rod Control and Rod Position lndication (CERPI) (2-PAT-3.4) (continued)3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceptance Criteria3.1 CERPI Calibration3.1.1 WNA-TP-02576-WBT, Watts Bar 2 ARPI SystemUpgrade CERPI Calibration Procedure, Section 2.6,lNlTlAL HOT CALIBRATION, was successfullycompleted and the linearity is within +12 steps at thesteps checked in the procedure.The initial hot calibration was successfully completed andthe linearity was demonstrated to be 112 steps at thesteps checked in the procedure.3.1.2 2-Sl-85-3, Calibration of Computer Enhanced Rodlndication Channels and Full Range Verification, wassuccessfully completed.2-S!-85-3 was successfully completed.3.2 Rod Control and lndication3.2.1 2.XA-55-4A.64F, C11 BANK D AUTO WITHDMWALBLOCKED, alarm window in control room was LIT whenControl Bank D was withdrawn above 220 steps.2-XA-55-4A-64F, C-11 alarm window annunciated at220steps.3.2.2 CERPI monitor alarm for:(1) Rod to Rod Deviation between two rods in a bankThe rod to rod deviation CERPI alarm was at12 steps.(2) Rod to Bank Deviation corresponding to > 12steps.The Rod to Bank Deviation alarm was at 12 steps.52 5.4Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.32-><A-55-4B-83D, PLANT COMPUTER GENERATEDALARM (SEE ICS), alarm window in control room wasLIT when ICS Computer detects the following conditions:(1) Deviation between rod position indicator for a rodand corresponding bank demand positi on >12steps.The deviation between ICS rod position and bankdemand was 12 steps.(2) Deviation between rod position indicator for a rodand average rod position >12 steps.The deviation between ICS rod position andaverage rod position was 13 steps.Rod bottom bistable indicators actuate at correct setpointsetting (below 20 steps withdrawn) as indicated by RPIindicators and rod bottom indicators on CERPI on 2-M-4.Each rod bottom bistable for all rods actuated below 20steps (20 to 19 steps) withdrawn.CERPI bypass indication for Control Banks B, C, and Dactuate at correct setpoint setting (below 35 stepswithdrawn) as indicated by CERPI Bank Demand digitaldisplay.CERPI bypass indication for Control Banks B, C, and Dactuated at 31 steps withdrawn.An Urgent Failure induced in a Power Cabinet and LogicCabinet caused local urgent failure alarm indicator lampat the respective cabinet and 2-Xl\-55-48-86A,CONTROL ROD URGENT FAILURE, ohnunciatorwindow to light.The Power Cabinet and Logic Cabinet local urgent failurealarm indicator lamp and 2-><A-55-48-86A annunciatorwindow functioned as designed and met all applicableAcceptance Criteria.3.2.43.2.53.2.663 5.4 Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.7A Non-Urgent Failure induced in each of the PowerCabinets and Logic Cabinet causes local non-urgentfailure alarm indicator lamp at respective cabinets and2-XI\-55.48-868, CONTROL ROD NON-URGENTFAILURE, ?rnunciator window to light.The Power Cabinet and Logic Cabinet local non-urgentfailure alarm indicator lamp and 2-)<A-55-48-868annunciator windows functioned as designed and met allapplicable Acceptance Criteria.A failure induced in the CERPI racks caused2-><A-55-4B-86C, CERPI TROUBLE, ?rnunciator windowto be LIT or REFLASH.The 2-><A-55-4B-86C annunciator window functioned asdesigned and met all applicable Acceptance Criteria.For shutdown and control rod banks having two groups,the group step counter for group 1 shall be 0 or 1 stepabove grou p 2 step counter over their full length of travel(i.e., 231 steps).The group step counters for group 1 were 0 or 1 stepabove grou p 2 step counters over their full length oftravel for shutdown and control rod banks having twogroups. All applicable Acceptance Criteria were met.2-><A-55-4B-87D, RODS AT BOTTOM, ornunciatorwindow was lit when one or more rods in CBA wereinserted in the normal sequence. Also, RODS ATBOTTOM, ornunciator window is not lit when controlrods were inserted or withdrawn in their normalsequence.2-xl\-55-4B -87 D annunciator window functioned asdesigned and met all applicable Acceptance Criteria.Each RPI indicated rod motion consistent with the groupdemand indication for the full range of rod travel.The RPI indicators for each rod indicated rod motionconsistent with the group demand indication for the fullrange of rod travel.3,2.93.2.93.2.103.2.1164 5.4 Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.12 The MCR rod speed demand display functions properlyand indicated the rod stepping rate (ROD SPEED) waswithin the range of 62 to 66 steps/minute for ShutdownBanks A and B in bank select mode.The MCR rod speed demand display indicated 64steps/min for both Shutdown Banks A and B.3.2.'13 The MCR rod speed demand display functions properlyand indicates the rod stepping rate (ROD SPEED) waswithin the range of 46 to 50 steps/minute for ControlBanks in bank select and in MANUAL mode.The MCR rod speed demand display indicated 48steps/min for all Control Banks in bank select and inManual mode.3.2.14 The MCR rod direction indicator lights functioned properlyto indicate the rod movement status and direction of rodmotion during rod withdrawal and insertion operations.The MCR rod direction indicator lights functioned asdesigned and met all applicable Acceptance Criteria.3.2.15 The MCR group step counters functioned properly toindicate group position and direction of rod motion duringrod withdrawal and insertion operations.The MCR group step counters as designed during bothwithdrawal and insertion.3.2.16 The MCR Computer Enhanced Rod Position lndicators(CERP!) function properly to indicate individual rodposition and direction of motion during rod withdrawaland insertion operations.The CERPI indicators functioned as designed to indicateindividual rod position and direction of motion duringwithdrawa! and insertion. All applicable AcceptanceCriteria were met.3.2.17 The rod insertion limits LO-LO upper limit was set to 211steps (Control Bank A)The rod insertion limits LO-LO upper limit was found tobe set at2ll steps (Control Bank A).65 5.4 Rod Gontro! and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.18 The rod insertion limits LO alarm actuated below 10steps above the insertion limit for any control bank.The rod insertion limits LO alarm actuated at 10 stepsabove the insertion limit for any control bank.3.2.19 The rod insertion limits LO-LO alarm actuated below 0steps above the insertion limit for any control bank.The rod insertion limits LO-LO alarm actuated at 0 stepsabove the insertion limit for any control bank.3.2.20 The control rod bank overlap circuitry functioned properlyduring the sequential withdrawal and insertion of Contro!Banks in MANUAL mode.The contro! bank overlap circuitry functioned as designedduring the sequential withdrawal and insertion of ControlBanks in MANUAL mode.3.2.21 Each RPI indicated rod motion consistent with the groupdemand indication for the ful! range of rod travel.The RPI indictors for each rod indicated rod motionconsistent with the group demand indication for the fullrange of rod travel.3.2.22 All rods were fully withdrawn and inserted five times.The test exercised all rods to fully withdrawn and fullyinserted five times.Review CriteriaNone4.0 Problernst1l CR 1168845 - Steps 6.4.3[20] and 6.4.4.1201, Acceptance Criteriafor RPI indication agreeing within t12 steps was not satisfied forShutdown Banks C and D for the full length of travel. ThisAcceptance Criteria was later changed by Urgent Change 2. SeeUC-2 description in this test report for further information.55 5.4 Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)CR 1168881 - Section 6.4.5, Rod position indications for rods inControl Banks A and B were at 12 steps from the demand positionat some positions over the full length of travel. Although not afailure in the Acceptance Criteria of Step 6.4.5[20] of within 12steps of the demand position, the rod-to-rod deviation alarms wereactuated at certain positions over the length of travel. This CRanticipated not meeting Steps 6.10[2.15]and 6.10[2.16]. Theacceptance of rod-to-rod deviations was later changed by UrgentChange 2. See UC-2 description in this test report for furtherinformation.CR 1169602 - Steps 6.10[2.15] and 6.10[2.16] Acceptance Criteriaof no rodto-rod and rodto-bank deviation alarms was not met forall 57 rods over the full length of travel. These steps were laterchanged by Urgent Change 2. See UC-2 description in this testreport for further information.CR 1 1692'17 - Step 6.4.6[1]C verified the C-1 1 Bank D AutoWithdrawal Block annunciator cleared during Control Bank Dinsertion between 219 and 214 steps on the step counters. Therecorded step counter position for Control Bank D was 212 stepswhen the C-11 annunciator cleared. Step 6.4.6[1]C is notAcceptance Criteria and the value at which the annunciator clearedis reasonable. Based on the CR evaluation, the vendor manualdescription of operation demonstrates (as well as the test engineerevaluation) there was not a problem encountered during 2-PAT-3.4performance with Control Rod Bank D permissive C-11. No FurtherAction Required.CR 1169282 - Step 6.7.2l7lcould not be performed as writtenbecause the performance of Step 6.7.2[6] cleared the urgent alarmupon seating the 4104 card. The card interlock is the only urgentalarm in the rod control system that clears upon restoring thesystem configuration. Step 6.7.2171was to verify rod motion wouldnot occur with a standing urgent alarm. These steps were notAcceptance Criteria and had no impact on successful completion ofthis test. Additionally, CR 1171247 was created to address andprovide justification associated with not performing the steps citedin CR 1169282.CR 1171254 - This CR was written to document certain steps andportions of sections which were repeated during testingperformance based on engineering judgment. These additionalperformances were used as a means to perform additionalcalibration of CERPI as specified in the Westinghouse CERPIcalibration procedure or as a means to restore from current testingl2l13Il4lt5It6I51 5.4 Rod Gontrol and Rod Position lndication (CERPI) (2-PAT-3.4) (continued)conditions and then later return. Section 6.1, Steps 6.1[1] through6.1111.19lwere repeated and Section 6.6, Steps 6.6[1]through6.6[9.5]were repeated one or more times. Section 6.6, Steps6.6[48] through 6.6[54] were repeated.This CR was created for documentation purposes only and has noimpact on the actual test results or verification of AcceptanceCriteria.l7l CR 1168538 - During performance of Section 6.1, Shutdown BankA was inserted to 118 steps demand position for CERPI hotcalibration. M14 indication drifted excessively for a couple of hoursuntil a stable rod position indication was reached.The PAT team and Westinghouse reviewed the M14 coil resistancevalues identified that the M14 drift was associated with a largetransient in coil resistance (i.e., change in coil stack temperature)and a larger value for the T_GAIN parameter for M14.Drift is a phenomena that occurs for ARPI/CERP! indicationsystems due to the analog coil stacks and their associatedtemperature dynamics and does not represent an actual change inrod position. This does not represent a deficiency in the design.Adjustments to the CERPltunable parameters were made tominimize drift. Therefore, this CR did not impact the successfulcompletion of this test procedure.l8l CR 1168899 - Step 6.4.5[8], the 2-)G-55-4B-87D, RODS ATBOTTOM, annunciator cleared as expected. However, the bell inthe MCR did not alarm. CR was closed to previously identifiedwork.tgl UC-1 was written to ensure that group step counters for ShutdownBank A, group 1, step counters displayed 56 steps prior to withdrawof Rod D-02 in Step 6.6 [40]. This allowed the group step counterto be updated to match the current rod position for rod that wascurrently capability of motion.t10] UC-2 was processed to revise Acceptance Criteria 5.2K and 5.2Uand related steps based on updated criteria provided byWestinghouse in WBT-D-5666, CERPI Acceptance Criteria,Revision 1, and a revision to UFSAR Table 14.2-2, Sheet 8.The update criteria stated "Each RPI indicates rod motionconsistent with the group demand indication for the full range of rodtravel."68 5.5Reactor Coolant Flow Goastdown (2-PAT-3.7)This test was performed as directed by 2-PAT-3.0, Post Core Loading PrecriticalTest Sequence, in Mode 3 at normaloperating temperature and pressure. Thetest was started and field work completed on 518116.1.0Test ObiectivesThe objectives of this test were to:1.1 Measure the rate at which reactor coolant flow changes subsequentto a simultaneous trip of all four reactor coolant pumps. Themeasured Flow Coastdown Time Constant is determined from theflow versus time data and compared to the Design Flow CoastdownTime Constant.1.2Measure the delay time associated with the low flow reactor tripand compare it to that value assumed in the accident analysis.Record the RCP Motor voltage decay during the transient forinformation only.1.4 This test satisfied the requirements of UFSAR Table 14.2-2,Sheet 15, Reactor Coolant FIow Coastdown Test Summary.2.0 Test Methods3.0All four reactor coolant pumps were simultaneously tripped, causing thereactor trip breakers to open on Low RCS Flow. Measurements weremade by recording reactor coolant loop elbow tap differentia! pressures(d/p), RCS low flow bistable state, reactor trip breaker position, reactorcoolant pump breaker position and reactor coolant pump motor voltagedecay data. Also recorded for information was the time of theundervoltage relay and associated time delay timer in the RCPsundervoltage circuit.Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 The Acceptance Criterion for core flow coastdown following thesimultaneous trip of the four reactor coolant pumps from full flowconditions, was that the measured flow coastdown time constant(TAUru) was greater than (>) design flow coastdown time constant(TAUD) of 11.72 seconds.Results indicated TAUu = 12.762 seconds1.369 5.5 Reactor Coolant Flow Coastdown (2-PAT-3.7) (continued)3.2 Acceptance criterion for the Total Low Flow Trip Delay Time is lessthan (<) 1.2 seconds.Results indicated Low Flow Trip Delay Time, Tr_r = 0.994 seconds3.3 Acceptance criterion for simultaneous trip of four reactor coolantpumps was that allfour pumps trip within (<) 100 msec. of eachother.Allfour pumps tripped within 20 msec.Review Criteria3.4 Review Criterion for coastdown flow data quality is that data from atleast 2 out of 3 flow transmitters in each RCS loop falls withinChauvenet's Criterion.All data points from all RCS flow transmitters fell within Chauvent'sCriterion.4.0 Problemst1] CR 1169224 - During performance of 2-PAT-3.7 post performanceactivities, Step 7.0[5], removing the test recorders from the Auxlnstrument Room, 8 out of 12 RCS low flow trip status lights lit inthe control room. Post event investigation revealed bistable fusesassociated with the affected flow loops (2-LPF-68-6A, 2-LPF-08-6B,2-L p F-69-29 A, 2-LpF-68-29 B, 2-Lp F-68-29D, 2-LPF-6848B, 2-LPF-68-71A, 2-LPF-68-71 B) blew during recorder disconnectionbecause of human performance issues.Since this was a post performance activity and all data wasrecorded during Section 6.0, there was no impact to the test resultsof 2-PAT-3.7.10 5.6 Rod Drop Time Measurement and stationary Gripper Release Timing(2-PAr-3.8)Portions of this test were performed in Mode 5 and again in Mode 3 as directedby 2-PAT-3.0, Post Core Loading Precritical Test Sequence.ln Mode 3 this test was performed in conjunction with the norma! Surveillancelnstruction 2-Sl-85-10, Rod Drop Time Measurement Using CERPI Rod DropTest Computer, to calculate the standard deviation of the rod drops and to directrequired additional rod drops for CR corrective actions and potential two-sigmadeviations.Prerequisites were started on 1122116 and the test was field work completed on1124116 for the Mode 5 performance. The Mode 3 performance was started on5111116 and completed on 5111116.1.0Test ObiectivesThe objectives of this test were to:1.3ln Mode 5, Section 6.2 of this Power Ascension Test (PAT) partiallywithdrew al! shutdown and control rods and demonstrated that allCRDMs unlatch and all rods fully insert into the core when thereactor trip breakers were opened.ln Mode 5, measured the Stationary Gripper Release times foreach control and shutdown rod.ln Mode 3, at Hot Standby conditions with full Reactor CoolantSystem (RCS) flow, measured the rod drop time and stationarygripper release time for each control and shutdown rod. WBN Unit2 Technical Specifications require rod drop time measurements,therefore, the normal Surveillance lnstruction 2-Sl-85-10, Rod DropTime Measurement Using CERPI Rod Drop Test Computer wasutilized.Meet the Mode 3 testing as required by UFSAR Table14.2-2, Sheet 9, Rod Drop Time Measurement And StationaryGripper Release Timing Test Summaryo Measure the stationary gripper release time for each controland shutdown rod. This measurement was performed inMode 5 and then repeated in Mode 3.1.11.21.41L 5.6 Rod Drop Time Measurement and Stationary Gripper Release Timing(2-PAT-3.8) (continued)1.5. Evaluate the data from rod drop time testing in the area ofthe dash pot entry looking for proper performance of thedecelerating devices (i.e. dashpots). This evaluation wasperformed in Mode 3 with the data collected during theperformance of 2-Sl-85-10, Rod Drop Time MeasurementUsing CERPI Rod Drop Test Computer.. Evaluate all 57 rod drop times in Mode 3 with the datacollected during the performance of 2-SI-85-10, Rod DropTime Measurement Using CERPI Rod Drop Test Computer.Ensure that four rod drops were performed in Mode 3 as requiredby corrective actions from Condition Repoft234483 action 003related to INPO SER 1-10.2.0Test MethodsThis PAT was written to supplement the norma! operating surveillance2-Sl-85-10, Rod Drop Time Measurement Using CERPI Rod Drop TestComputer and evaluated the rod drop time data.All rod drop times were used to calculate the standard deviation of the roddrop times. Two-sigma limits (i.e. plus or minus two times the standarddeviation) were used to evaluate drop times of the 57 rods. Those droptimes that were outside of the two-sigma limits were re-measured 3 (ormore) times and evaluated for consistency (i.e. within 50 milliseconds).Retesting the rods that fell outside of the two-sigma limits an additional 3(or more) times provided reasonable assurance of their properperformance during subsequent plant operations.This PAT measured the Stationary Gripper Release Time for each controland shutdown rod. The Stationary Gripper Release Time is a combinationof a Trip Signal Delay Time (i.e. Delay between Reactor Trip Breakeropening and the trip signal to the RDTC) and the delay between powerinterruption (i.e. trip signal to the RDTC) and the rod's initiation of its freefal!. The PAT evaluated the traces from the RDTC looking for a delay ofeach rod's initiation of free fall the RDTC's trip signal. The PAT alsomeasured the "Trip Signal Delay Time" while in Mode 5.This PAT also evaluated the traces from the rod drops in the area of thedecelerating devices (i.e. dash pots) entry looking for proper performanceof the dash pots.12 5.6 Rod Drop Time Measurement and Stationary Gripper Release Timing(2-PAT-3.8) (conti nued)3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 Each CRDM unlatches upon opening the Reactor Trip Breakers.Testing confirmed that each CRDM unlatched upon opening theReactor Trip Breakers in Mode 5 and Mode 3.3.2 The rod drop times for all shutdown and control rods, dropped fromthe fully withdrawn position, are within the limits specified in theTechn ical Specifi cations.This Acceptance Criteria was successfully met by performance of2-Sl-85-10, Rod Drop Time Measurement Using Rod Drop TestComputer.3.3 Rod drop time evaluations against the Two-Sigma statistical limitscalculated resulted in either criteria below being applicable:o Rod drop time was within the bounds of the lower and upperTwo-Sigma statistical limits.o Rods with rod drop times that fell outside of the bounds ofthe lower and upper Two-Sigma statistical limits have beendropped greater than or equal to 3 additional times. Theresults of the rod drop times were consistent (i.e. within aband of 50 milliseconds or less) and continue to meetTechnical Specification criteria specified in2-Sl-85-10, Rod Drop Time Measurement Using CERPI RodDrop Test Computer.During the Mode 3 performance, Rods D-2 and M-14 failed to meeta 2 sigma statistical limit for the first rod drop. Three additiona! roddrops were successfully performed. (See CR 1169659)3.4 The stationary gripper release time for all rods was <150 msec.This was the requirement in Unit 2 UFSAR Chapter 4.During the Mode 5 performance the stationary gripper release timewas conservatively determined to be 45 msec. Mode 3 releasetime was 50 msec.13 5.6 Rod Drop Time Measurement and stationary Gripper Release Timing(2-PAT-3.8) (conti nued)3.5 The Trip Signal Delay Time was <100 msec.; as accounted for in2-St-85-10.The Trip Signal Delay Time was determined to be 50 msec.Review CriteriaNone4.0 Problems11] CR 1128964: During the Mode 5 performance the RDTC plots foreach rod were inverted from the expected response. This issue didnot impact the performance of this test and resolution of the CRoccurred prior to the Mode 3 performance of this test.12) CR 1169659: Rods D-2 and M-14 did not meet a two-sigmastatistical limit for the first rod drop in Mode 3. Three additional roddrops were performed and allAcceptance Criteria met.74 5.7 Reactor Trip System (2-PAT-3.10)This test was performed in Mode 5 as directed by 2-PAT-3.0, Post Core LoadingPrecritical Test Sequence. Testing was started on 1113116 and field workcompleted on 1120116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Demonstrate proper functioning of the Reactor Trip System. Thisobjective was accomplished by demonstrating that:1.1.1 The reactor trip breakers can be opened manually from the MainControl Room (MCR)1.2.2 Interlocks permit momentary closure of both reactor tripbypass breakers and then cause a reactor trip.1.3.3 The reactor trip bypass breakers maintain the rod drivemechanisms energized when the associated reactor trip breaker isopened for test.1.4.4 With one reactor trip bypass breaker closed, placing theopposite SSPS train channel in test causes both reactor tripbreakers and the bypass breaker to open.1.2 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 19,Reactor Trip System Test Summary.2.0 Test Methods2.1 Performance of this test (2-PAT-3.10) was completed with the unitin Mode 5 and RCS pressure greater than 100 psig to satisff thereq u irements necessary for performance of 2-TRl 1,Reactivity Control Systems Movable Control Assemblies, asrequired by Technical Surveillance Requirement 3.1.7.During performance of 2-TRl-85-1 an bsue was encountered inwhich the rods controlled bythe 2BD power cabinet (SB Group 2,CB Group 2 and, CD Group 2) did not respond to outward roddemand. Further details on this bsue are documented in CR1126661 found in the write-up for 2-PAT-3.1.15 5.7Reactor Trip System (2-PAT-3.10) (continued)2.3Section 6.2 verified both Main Control Room (MCR) Reactor Triphandswitches (2-RT-1 and 2-RT-2) generated a reactor trip andthe associated indications appropriately.Section 6.3 verified the electric interlocks prevented both bypassbreakers from being closed simultaneously and resulted in areactor trip due to a general warning in both trains of SSPS.Sections 6.4 and 6.5 verified that when a reactor trip bypassbreaker was closed placing the opposite SSPS train in test resultedin a reactor trip due to the generation of simultaneous generalwarnings in both trains.section 6.6 verified that the bypass breakers maintained the controlrod drive mechanisms energized when the associated reactor tripbreaker was opened due to injection of a simulated ReactorProtection System trip signal on the associated SSPS train.2.53.02.23.2Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1Reactor trip breakers (RTA andwith hand switches 2-RT-1 andReactor trip breakers (RTA andboth hand switches 2-RT-1 andRTB) can be opened manually2-RT-2.RTB) were opened manually with2-RT-2.Electrical interlocks trip both reactor trip bypass breakers (BYA andBYB) when both bypass breakers are closed due to simultaneousgeneral warning reactor trip signals being sent to the Reactor TripBreakers (RTA and RTB)Electrical interlocks uocessfi.rlly tripped both reactor tri p bypassbreakers when both bypass breakers were closed.16 5.7 Reactor Trip System (2-PAT-3.10) (continued)3.3 With one reactor trip bypass breaker (BYA or ByB) closed, placingthe opposite SSPS train in test causes both reactor trip breakers(RTA and RTB) and the bypass breaker (BYA or ByB) to open dueto simultaneous general warning reactor trip signals being sent tothe reactor trip breakers (RTA and RTB)Section 6.4 tested bypass breaker B in conjunction with SSPSTrain A was completed successfully and all Acceptance Criteriawere met as stated.During performance of Section 6.5 a previously known andexpected indication issue related to placing SSPS Train BMultiplexer test switch in lNHlBlT, as originally documented in CR1126043, was encountered. After verification that indicationsreceived were the same as those previously documented in CR1126043 testing continued as the eroneous indications had noimpact upon performance of 2-PAT-3.10 and Section 6.5 wascompleted satisfactorily with the exception of a procedure errorwhich was identified in CR 1126802.3.4 The reactor trip bypass breakers (BYA or BYB) maintain the roddrive mechanisms energized when the associated reactor tripbreaker (RTA or RTB) is opened by injection of a simurated ReactorProtection System trip signal on the associated SSpS trainEach reactor trip bypass breaker maintained the rod drivemechanisms energized when the associated reactor trip breakerwas opened. Section 6.6 was completed satisfactorily with theexception of a procedure error which was identified in cR 1126802.Review CriteriaNone11 5.7 Reactor Trip System (2-PAT-3.10) (continued)4.0 Problemst1l CR 1126802 was generated because of procedure errors,which assumed breakers racked to the test position would stillget the GEN WARNING alarm and MCR light indication forbreaker position. lt was determined that the procedure errorswere minor and did not affect the test, including AcceptanceCriteria Steps 6.5[14]D, 6.6[13]D, and 6.6[27]C that verifiedbreaker position lights for bypass breakers not connected.After validation of the procedure error and its impacts, testingwithin 2-PAT-3.10 continued and Section 6.3 was completedsatisfactorily.Setup of Section 6.5 places the Reactor Trip Bypass Breaker B(BYB) in the test position; therefore, MCR indication lights forBYB are not illuminated. This section verifies that the reactortrips when Reactor Trip Bypass Breaker A (BYA) is closed andSSPS Train B is placed in Test; this was successfully performedwith associated lights illuminated for BYA.Section 6.6 verifies the ReactorTrip Bypass Breakers functionto prevent a Reactor Trip during testing of Reactor TripBreakers. During this section each of the Bypass Breakerswere installed in the connected position (one at a time) withappropriate !ight indication.78 5.8 Adjustment of Steam Flow Transmitters at Minimal FIow(2-PAT-3.11)This test was performed with the plant in Mode 3 at normal operating pressureand temperature, as specified in 2-PAT-3.0, Post Core Loading Pre-CriticalTestSequence. The test was started on 5/6/16 and field work complete on 517116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify/adjust the output of the eight steam flow transmitters for"zero" output with minimal steam flow.1.2 Satisfy the Mode 3 objective in the UFSAR Table 14.2-2, Sheet 21 ,Calibration Of Steam And Feedwater Flow lnstrumentation AtPower Test Summary.2.0 Test MethodsThe plant was in Mode 3 at normal operating temperature and normaloperating pressure. Steam flow was reduced to minimal by shutting aMSIV, one loop at a time. With the MSIV closed, each steam flowtransmitter on the associated main steam line was verified/adjusted for a"zeto" output. This was repeated for each main steam line.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaNoneReview Criteria3.1 At minimum steam flow, the output from each steam flowtransmitter and its associated loop reflects zero flow asdemonstrated by the following:A. D/P Test Point: 0.19829 Vdc (0.19641to 0.20016 Vdc)B. Flow Test Point: 0.2000 Vdc (0.1972to 0.2028 Vdc)C. Computer Test Point: 0.2000 Vdc (0.1972to 0.2028 Vdc)D. ComputerPoint: 0.0 KBH (-275to275 KBH)79 5.8 Adjustment of Steam Flow Transmitters at Minimal Flow(2-PAT-3.1 1) (continued)4.0The data below was collected and the output flow was verified/adjustedwithin the Review Criteria requirements.* adjustment madeProblemsThere were no significant problems encountered during the performanceof this test.TransmitterD/P TestPoint(Vdc)Flow TestPoint(Vdc)ComputerTest Point(Vdc)ComputerPoint(KBH)2-FT-1 -3A*0.1 981 00.1 99830.201 0342-FT-1 -380.1 96850.200550.200754z-FT-1-10A0.197820.200390.200271z-FT-1-10B*0.1 97800.200710.2a12162-FT-1-21A*0.1 98670.201 600.2001312-FT-1-21B0.1 99600.200380.2005562-FT-1-28A0.1981 10.200700.2005022-FT-1-2gB0.1 96900.200070.20051380 5.9Control Rod Drive Mechanism Timing (2-PET-106)This test was performed as part of test sequence 2-PAT-3.0, Post Core LoadingPrecritical Test Sequence. The test began via WO 117705850 on 04113/16 andwas field work completed on 516116.1.0Test ObiectivesThe objectives of this test were to:1.2Verify the acceptability of the Control Rod Drive Mechanism,(CRDM), current order timing, current order amplitudes, and rodwithdrawal speed.Partially satisfy the requirements of UFSAR Table 14.2-2,SheetT, Control Rod Drive Mechanism Timing Test Summary.2.0Test MethodsThe test was required to be performed following fuel toading. Since theCRDM latch assembly must be submerged in water for proper operation, aminimum RCS pressure of 100 psig was required. The test was run atnominal hot plant conditions. Reactor Engineering verification of currentboron concentration being adequate to perform this test by being equaltoor greater than the refueling boron concentration was required.With the reactor trip breakers closed and the lift coils verified to beconnected, a selected bank was withdrawn and then reinsertedapproximately 10 steps to obtain the CRDM readings.The test objectives were accomplished by monitoring the CRDM coilcurrent profiles to verify that the stationary gripper, movable gripper, andlift coil current order changes occur at the proper time during the 780msec. rod stepping cycle; that stationary, movable and lift coil currents areproperly regulated to full current values within acceptable ranges duringrod withdrawal and insertion operations; that shutdown bank rodwithdrawal speed is a nominal 64 steps/min and control bank rodwithdrawal speed is a nominal43 steps/min.Test Results3.0AII Acceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaCurrent Order TimingThe times at which the lift, movable, and stationary current orderschange, after the start of rod motion, are within 10 msec. of theexpected times during rod withdrawal and insertion operations.811.13.1 5.9 Control Rod Drive Mechanism Timing (2-PET-106) (continued)The lift, movable, and stationary current orders for al! CRDMs werewithin 10 msec. of the expected times during rod withdrawal andinsertion operations.3.2 Coil Current AmplitudesStationary, movable and lift currents are regulated by circuityinternal to each power cabinet. The reduced and full currentnominal values are not critical, cannot be adjusted, but could be anindication of a regulation failure. Measured Values outside thenominal ranges below should be evaluated and documented by thesystem engineer.3.2.1Lift Coil - fu!l35 to 47 .2 amperes(equivalent to 438 to 590mVdc measured across a0.0125 ohm resistor)3.2.2Lift coil -reduced13 to 19.7 amperes(equivalent to 163 to246mvdc measured acrossa 0.0125 ohm resistor)3.2.3Movable GripperCoil - full7 to 9.2 amperes(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)3.2.4StationaryGripper Coil - full7 to 9.2 amperes(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)3.2.5StationaryGripper coil -reduced3.8 to 4.8 amperes(equivalent to 238 to 300mVdc measured across a0.0625 ohm resistor)All stationary, movable, and lift currents amplitudes for all CRDMswere within the Acceptance Criteria with the exception of D08, 810,F14, F10, and D12lift coi! reduced currents which were greaterthan the 19.7 amperes criteria. This issue was previouslyevaluated in CR 1128950.These reduced lift currents were all Iessthan 21 amperes which was evaluated by Westinghouse asacceptable in Westinghouse Letter WBT-D-5604.82 5.9 Control Rod Drive Mechanism Timing (2-PET-106) (continued)3.3 Rod Withdrawal Speed3.3.1 Shutdown Bank withdrawal speed nominal 64 steps perminute.The measured Shutdown Bank withdrawal speed wasapproximately 64 steps per minute and did not exceed thenominalvalue.3.3.2 Control Bank withdrawal speed nominal4S steps per minute.The measured Control Bank withdrawal speed wasapproximately 48 steps per minute and did not exceed thenominalvalue.Review CriteriaNone4.0 ProblemsThere were no significant problems encountered during the performanceof this test.83 6.0 INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING6.1 lnitial Criticality and Low Power Test Sequence (2-PAT4.0)2-PAT-4.0 started with prerequisites on 5112116 and completed on 5124116.1.0 Test ObiectivesThe objective of this test was to:1.1 Provide governance of the sequence of the Power AscensionTesting in Mode 2.The following PATs/PETs were sequenced for performance by2-PAT-4.0:o 2-PET-201
lnitial Criticality and Low Power Physics Testingo 2-PET-103 Reactivity Computer (ADRC)o 2-PEf-304
OperationalAlignment of NISo 2-PAT-1.5
Loose Parts Monitoring Systemo 2-PAT-1.10* lntegrated Computer System (lCS). RCI-159
Radiation Baseline SurveysNote:
lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.2.0 Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on5112116 and completed on 5122116 and included verification of thefollowing major items:o Preoperationa! tests completed to allow entry into Mode 2o TVA-SPP-30.010, lnitial Synchronization of TVA Generating Assetsto TVA's Transmission System notificationo Reactivity Control Plans are developed to support testingo 2-PET-201, lnitial Criticality and Low Power Physics Testing, hasbeen initiatedo Section 4.0 of 2-PET-103, Reactivity Computer (ADRC), has beenperformedPrior to initiation of the performance section, a cool down was initiated on5116116 to 360oF to replace a failed hot Ieg RTD. RCS temperature wasstabilized between 355oF and 365oF at22:59 on 5/'t6/16. The unit wasplaced in Mode 4 on 5118116 at 23:58 to facilitate SSPS testing. Aftercompletion of testing Unit 2 re-entered Mode 3 at 04:15 on 5/20/16 andbegan a heat up to normaloperating temperature and pressure. NOTPwas reached on 5121116 at 01 :00. On 5123116 at 01:04 the unit enteredMode 2.84 6.1 lnitial Criticality and Low Power Test Sequence (2-PAT4.0) (continued)The reactor was taken critical on 5123116 at 02:16.2-PAT-4.0 governed initial criticality and the low power testing greater than3 percent and less than 5 percent reactor power. Applicable portions ofthe following procedures were initiated and completed as appropriate.o 2-PET-201, lnitial Criticality and Low Power Physics Testing -Completed 5123116 with all criteria met.o 2-PEf-103, Reactivity Computer (ADRC), completed 5/23116 withall criteria met.o 2-PET-304, Operational Alignment of NlS, completed 5/23116 withall criteria met.o 2-PAT-1.5, Loose Parts Monitoring System, was completed on5124116 with all criteria met. CR 1171424 documents threechannels removed from service.. 2-PAT-1.10, Integrated Computer System (lCS), completed 5124116CR 1174334 documents exceeding the MED between T0457A andMCR indicator 2-Tl-62-29, RCP 3 LWR RADIAL BRG Temp.o RCI-159, Radiation Baseline Surveys - completed 5/31/16. NoAcceptance or Review Criteria were associated with this procedure.Details of the performance of each PAT procedure is contained in theindividual summaries of the associated procedures.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.O ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAT-4.0.85 6.2Reactivity Computer (ADRC) (2-PET-1 03)This test was performed as part of test sequence 2-PAT-4.0, lnitial Criticality andLow Power Test Sequence. Field performance of 2-PET-103 was commencedon 05/15/16. The purpose of this procedure is to ensure that the AdvancedDigital Reactivity Computer (ADRC) is capable of reactivity measurements insupport of Low Power Physics Testing (LPPT) per 2-PET-201. This procedurewas completed on 05123116 following completion of LPPT.1.0Test ObiectivesThe objectives of this test were to:1.2Perform installation of the ADRCPerform the calibration and setup of the ADRC prior to reactivitymeasurements.Provide instructions for connecting/restoring the RCS Temperatureand Rods Move signals to/from the ADRC1.32.0 Test MethodsThis test provided instructions for setup and installation of the ADRC forLPPT. This test connected a RCS T"rn signal from the Unit 2 Auxiliarylnstrument Room to the ADRC, connected the Power Range detectorsTop and Bottom signals and the "Rods Move" signal to the ADRC, andprovided instructions on initial checkout of the reactivity computer.Proper installation was verified by performing the initial checkout and initialexponential test. The initial checkout ensured that the ADRC was loadedwith the correct constants and reactivity data consistent with WBN Unit 2Cycle 1 core design. After input data was confirmed, the initialexponential test was conducted using a simulated signal for reactor flux.This calculated reactivity was verified to be within 1.0o/o of the theoreticalvalue. This test ensured that the ADRC was correctly calculating reactivitywith appropriate input data.Once physics testing was complete, steps were given to remove allinstalled cables and return the plant to its original state.1.186 6.2 Reactivity Gomputer (ADRC) (2-PET-1 03) (continued)3.0 Test ResultsAII Acceptance/Review Criteria were met or resolved as delineated below.Acceptance Criteria3.1. The absolute value of the PREDICTED vs MEASURED enor, thepercent difference between the ADRC "predicted" reactivity and the"measured" reactivity is < 1.0% during the ADRC lnternalExponential Test.The difference between the "predicted" reactivity and "measured"reactivity was found to be -0.03%, within the 1.0% criteria specifiedby the procedure.Review CriteriaNone4.0 ProblemsThe following issues were encountered during Reactivity Computer setupper 2-PET-103:11] While verifying the inputs to the ADRC were correct, it was notedthat the value for the prompt neutron lifetime was inconsistentbetween the value stated in the eNuPOP compared to the valuebeing used by the ADRC. The eNuPOP listed a value of 19.718microseconds while the ADRC was found to have a value of 19.716seconds. Following consultation with the fuelvendor, it wasdetermined that both values were acceptable (per Westinghouseletter NF-TV-16-24) and showed the smalldifference due to beingcalculated by two separate versions of code. The value listed in theADRC was calculated using a later version of the ANC code. Aone-time-only change was generated for this procedure to allow forthis difference. The procedure originally stated that the values hadto be "identical." The one-time-only change allowed for the valuesto be "consistent."81 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-2011The test was performed as part of test sequence 2-PAT-4.0, lnitial Criticality andLow Power Test Sequence. Field performance of 2-PET-201was commencedon 5122116 and initial criticality was achieved at 02:16 on 512312016. The testwas completed on 05123116 with successful completion of initial criticality, rodworth measurements (using Dynamic Rod Worth Measurement (DRWM)method), boron endpoint measurements, and isothermaltemperature coefficienttesting.1.0 Test ObiectivesThe objectives of this test were to:1.1 Dilute the reactor to criticality in a cautious and controlled manner1.2 Perform Mode 2 Low Power Physics Testing in a cautious andcontrolled manner, including:1.2.1 Measuring the integral worth of the control and shutdown rodbanks.1.2.2 Measuring the ARO critical boron concentration.1.2.3 Measuring the ARO lTC.1.3 This test and associated Sls satisfied the requirements of UFSARTable 14.2-2:Sheet 22,lnitial Criticality Test Summary.Sheet 23, Determination Of Core Power Range For Physics TestingTest Summary.Sheet 24, Moderator Temperature Coefficient Test Summary.(2-S l-0-23, Moderator Tem perature CoefficientDetermination at BOL)Sheet 25, Rod And Boron Worth Measurements Test Summary.Sheet 26, Core Reactivity Balance, Acceptance Criteria 1.(2-SI-0-1 2, Core Reactivity)Note: Sheet 26, Core Reactivity Balance, Acceptance Criteria 2 isdocumented in 2-Sl-0-12, Core Reactivity, at full power.B8 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)2.0 Test Methodslnitial reactor startup was conducted via dilution to critical while allshutdown and control banks were fully withdrawn. The dilution began at65 gpm and the reactor was monitored by use of lcRR. see Figure 6.3-1,ICRR vs Primary Water (N31, N32)When the ICRR reached 0.3, the dilution was terminated. After criticalitywas achieved and power increased, contro! rods were inserted to zero thestartup rate with reactor power near 1x10-3 % power.With the reactor stable, a "bite check" was then performed to determine ifthe inserted worth of control Bank D was between 40 to 7s pcm. An RCSboration was performed to establish an inserted worth of 62 pcm. Areactor exponential test was then conducted while finding the point ofadding heat to set the physics testing range.With the Physics Testing Range met, the DRWM testing began bywithdrawing CBD in Manualto the full out position. Once flux reached theappropriate levelon the reactivity computer CBD was insertedcontinuously in individual bank select until 0-5 steps withdrawn. Whendata collection was complete, CBD was restored to the fullout position.This process, of measuring rod bank worth, was repeated for eachremaining control and shutdown bank, in individual bank select. Thereactor was then brought back to a stable condition in Manual with ControlBank D slightly inserted.The boron endpoint was then calculated using the measured bank worthdata by use of the ADRC. This information is used to determine the AROHZP No XE critical boron concentration. The ITC was then measured byinitiating a constant rate cooldown, at less than 30 deg F/hr. when datacollection was complete, a constant rate heatup, also at less than 30 degF/hr, was initiated. Both sets of data were analyzed to carcurate anaverage ITC and converted to a MTC, accounting for the Dopplerreactivity coefficie nt.3.0 Test ResultsAII Acceptance/Review Criteria were met or resolved as delineated below.B9 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)Acceptance Criteria3.1 Advanced Digital Reactivity Computer (ADRC) CheckoutThe indicated reactivity is within !4o/o or 1 1 pcm of the theoreticalreactivity for each reactor exponential measurement.lndicated reactivity during the reactor exponential test for ADRCcheckout was measured at 25.0 pcm with a predicted reactivity of24.8 pcm. This resulted in a difference of 0.94o/o or 0.2 pcmdifference.3.2 Control and Shutdown Bank Worths (DRWM criteria)The sum of the measured bank worths is greater than or equal to(100%-RWU) times the sum of the predicted bank worths.The RWU, Rod Worth Uncertainty, is given as 10o/o for Unit 2 Cycle1. The sum of the measured bank worths was measured to be1.2o/o gteater than the predicted bank worths. This value is greaterthan 90% (100%-RWU) of the predicted bank worths.3.3 Boron Endpoint MeasurementBoron endpoint Acceptance Criteria is verified in 2-Sl-0-12, CoreReactivity. (2-PET-201 verified that 2-Sl-0-12 was successfullycompleted). The Technical Specification Acceptance Criteria within2-Sl-0-12 is for measured Mode 2HZP ARO critical boronconcentration shall be within the reactivity equivalence of +1000pcm of the predicted HZP ARO critical boron concentrationThe Boron Endpoint Acceptance Criteria was met via performanceof 2-Sl-0-12 (WO 117827845) following data collection from2-PET-201. The Boron Endpoint was measured at 1089 ppm.The predicted value was 1034 ppm. This resulted in a difference of55 ppm, or -569.9 pcm.3.4 TemperatureCoefficientThe Moderator Temperature Coefficient (MTC) Acceptance Criteriais verified in 2-Sl-0-23, Moderator Temperature CoefficientDetermination at BOL (WO 115947713\.(2-PEl-201 verified that 2-Sl-0-23 was successfully completed.)The Technical Specification Acceptance Criteria within 2-SI-0-23are:3.4.1 The MTC is less than or equal to 0.0 pcm/'F atHZP.90 6.3 lnitial Griticality and Low Power Physics Testing (2-PET-201) (continued)3.4.2 The MTC is less than or equal to the Beginning of CycleMTC as-measured criterion specified in the COLR.Both MTC Acceptance Criteria were met by successfulperformance of 2-Sl-0-23. The MTC was measured by2-PET-201to be -3.515 pcm/'F, which is less than the 0.0 pcm/'Flimit and below the COLR limit of -3.33 pcm/'F.3.5 Zero Power Physics Testing RangeThe zero power physics testing range is determined such thatreactivity feedback from nuclear heating does not compromise themeasurements.The zero power physics testing range was determined to not haveany reactivity feedback affects prior to performing rod worth, BoronEndpoint or ITC testing.Review CriteriaThe Review Criteria are listed below with two noted failures.3.6 ADRC Checkout3.6.1 The indicated reactivity is within 12% or 11 pcm of thetheoretical reactivity for each measurement.Indicated reactivity during the reactor exponential test forADRC checkout was measured at 25.0 pcm with a predictedreactivity of 24.8 pcm. This resulted in a difference of 0.94o/oor 0.2 pcm difference.3.6.2 The reactivity traces do not exhibit excessive noise levelft2 pcm).During the determination of the physics testing range,reactivity traces were reviewed and confirmed to not exhibitexcessive noise outside of the specified Review Criteriatolerance.3.6.3 The reactivity indication is stable as a function of flux level(no obvious dependence on the flux input !evel).During the determination of the physics testing range,reactivity traces were reviewed and confirmed stableindication of reactivity as a function of flux level with noobvious dependence of the flux input level.9L 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)3.7 Rod Worth Measurement (DRWM Criteria)3.7.1 The measured worth of all banks are within +1Oo/o or +75pcm of the prediction, whichever is greater.All shutdown and control banks were within the +10% and175 pcm of the predicted values.3.7.2 The sum of the measured worths of all banks are within+(0.8.RWU)o/o of the prediction.BankMeasured(pcm)Predicted(pcm)M.P(pcm)100"(M/P-1)(%lCD1,304.21,339.5-34.3-2.60/oCC1,061 .31,052.78.6A.8o/oCB794.4743.251 .26.9o/oCA910.0951 .4-41 .4-4.4o/oSD437.7434.43.30.8o/oSC447.4434.413.03.0o/oSB1,055.71 ,017 .138.63.8o/oSA424.1389.434.78.9o/oTotal6,434.86,36 1 .173.71 .2o/oAll shutdown and control banks were within +10% and +Tspcm of the predicted values. The sum of the measuredworths of all banks were within +(0.8.RWU)% of theprediction.3.8 Boron Endpoint3.8.1 Measured ARO boron endpoint is within 150 ppm of thepredicted boron endpoint.The measured ARO boron endpoint was measured as 108gppm, which was 55 ppm higher than the predicted boronendpoint of 1034 ppm. CR 1173995 initiated to documentReview Criteria failure.3.8.2 Measured ARO boron endpoint is within +5gg pcmequivalent boron.The measured ARO boron endpoint was measured as-569.9 pcm different from predicted values.CR 1173995 captures this failed Review Criteria also.92 6.3 Initial Criticality and Low Power Physics Testing (2-PET-201) (continued)3.9 Temperature CoefficientThe Measured ITC is within +2 pcml"F of the predicted lTC.The ITC was measured as -5.305 pcm/"F with a predicted valueof -6.67 pcm/"F.4.0 Problemst1l CR 1173995: Both Review Criteria for Boron Endpoint results werenot met. The Boron Endpoint was measured to be -569.9 pcm or55 ppm from predicted value. The Acceptance Criteria were metfor reactivity balance.Reactor Engineering, Nuclear Fuel and Westinghouse concludedthat there were no safety concerns or issues resulting from thisdifference.93 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)FIGURE 6.3.1ICRR vs Primary Water (N31, N32)Inverse Count Rate Ratiot- F,-i,-', 6t5,k !,i:r. .1, ll-l :rt*F.r;Antual RiS C. -- ppm r.lEl stupsIt .i [: "tr ].. i r,, ,J {. i1lt.,Hi r-(. I' ), ' I" 19{LJj,iL'.1,r'.} 0L"U.1 'liIt1lriiiIlItItIiiiI1iltltiII,i.iiit1L,l1iI.*ii-.,tl'iitit-+ritll litl rliIIIi'-l1--i*-1It1I+il1,irJl'1l-I-+II1Il,11ii1iIiil{Iili i-'iti-1IiitiIi "'lli.lIr--i1i1l filT-III1ri+t--rI,llTIi1l1irr.lItiIt+!iiiil-il,1ij1iii1.lrl..i:Ii1I+iiii:jirtltiiitII1+Iifi1+liiffitt- f t.tit+4-++I1'l.--li.\+:i1iiIlIII+it.!i-t*.til_-TI,,1i" 'tIil1Il"Hi1i'lHilllr1i\llilruhrt-LT.=--ii-rrii'.ill1*J;,,1Iir'l,1\1.1ililit',rltl,,iLr--rtl,rltlr"tIri-i-i i-] .lil.It{",tla.ilf\;hTIt ,;i'*itl r'.t1in pp.p' ,-rn.lrii ":!;i,i!oi*ons+f rr#C-f ";*,*orld;*rlr-.lr..l;i, .r:i!Prinary lfater (9a1s)94 7.0 POWER ASCENSION TESTING7.1 Test Sequence for 30% Plateau (2-PAT-5.0)This test started on 5117 116 and was completed on 6/16116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Define the plant operational requirements in conjunction with2-GO-3, Unit Startup from Less than 4% Reactor Power to 30%Reactor Power.1.2 Ensure those requirements were met in order to permit powerescalation from Mode 2 conditions with reactor power < 5% RatedThermal Power (RTP) to 30%.1.3 Specifo the order of test performance at the 30% plateau.The following PATs/PETs were sequenced for performance by2-PAT-5.0:o 2-PAT-1.4
Pipe Vibration Monitoring. 2-PAT-1.5
Loose Parts Monitoring Systemo 2-PAT-1.6
Startup Adjustments of Reactor Control Systemo 2-PAT-1.7
OperationalAlignment of Process Temperaturelnstrumentationo 2-PAT-1.8
Thermal Expansion of Piping Systemso 2-PAT-1.10* Integrated Computer System (lCS)o 2-PAT-1.11* RVLIS Performance Testo 2-PAT-1.12* Common Q Post Accident Monitoring Systemo 2-PAT-5.1
Dynamic Automatic Steam Dump Controlo 2-PAT-5.3 Automatic Steam Generator Level Control,Transients at Low Powero 2-PAT-5.4 Calibration of Steam and Feedwater Flowlnstruments at 30% Powero 2-PET-301
Core Power Distribution Factorso 2-PET-304
Operational Alignment of NISo RCI-159
Radiation Baseline SurveyNote:
lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.This test and WO 116916855 for (WINCISE) Post-Critical SystemCalibration (WNA-TP-04724-WBT) satisfy UFSAR Table 14.2-2, Sheet 12,lncore lnstrumentation System Test Summary, Acceptance Criteria 2.95 7.1 Test Sequence for 30o/o Plateau (2-PAT-5.0) (continued)2.0 Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on5117116 and completed on 5125116 and included verification of thefollowing major items:c 2-P4T4.0, Initia! Criticality and Low Power Test Sequence, hasbeen completed.. NPG-SPP-10.4, Reactivity Management Program, ReactivityControl Plans were developed to support the planned testing forthis sequence.. WO 116916855 implemented vendor procedureWNA-TP-0 47 24-WBT, Westi ng house I ncore I nformationSurveillance & Engineering (WINCISE) Post Critical SystemCalibration.o WINCISE incore signa! quality verification was in progress byimplementation of applicable section of vendor procedure WNA-TP-04724-WBT.o Reactor power was S 5% RTPo RCS pressure was between2220 to 2250 psigo Section 6.3 of 2-PET-304, OperationalAlignment of NlS, to adjustthe Power Range High Flux Level Trip setpoints for testing at the30% Plateau was complete.On 5125116 the performance section of 2-PAT-5.0 was begun and a powerincrease to 6-9 percent was initiated. Mode 1,2 5o/o power, was reachedat 03:33 on 5125116.2-PAT-5.3, Automatic Steam Generator Level ControlTransients at LowPower, Section 6.1, was completed on 5126116 with all criteria met.RCI-159, Radiation Baseline Surveys, was completed on 5/31/16. NoAcceptance or Review Criteria were associated with this procedure.2-PAT-5.1, Dynamic Automatic Steam Dump Control, Sections 6.3, 6.4,and 6.5 were started on 5125116 and completed on 5127116. Section 6.3and 6.4 were completed with all criteria met. Section 6.5 was completedafter an Urgent Change to the procedure was approved by the TRG tochange the load rejection testing criteria per Westinghouse LetterLTR-SCS- 1 6-23 (LTR-PCSA-1 6-23). The revised Acceptance Criteriawas met for Section 6.5.On 5127116, with reactor power between 13 and 14 percent, the turbinewas rolled and subsequently stopped due to noise in the area of theturning gear. The turbine was shut down and subsequently re-rolled andthe noise repeated at approximately 400 rpm. The turbine was shutdownand the decision made to place the Unit in Mode 3 for turbine repairs.96 7.1Test Sequence for 30% Plateau (2-PAT-5.0) (continued)On 5128116 at 01:54 the Unit re-entered Mode 3 after a manua! reactortrip. The generator was purged and a clearance placed on the turbine forinspection. On 5/31/16 Unit 2 entered Mode 2 at 12:00 followed by takingthe reactor critical at 13:39. Mode 1 entry was made on 17:49 on 5/31/16.Unit 2 was synchronized to the grid on 613116 at 20:39 and powerincreased to 15 percent. As power increased a steam leak required amanualturbine trip on 614116 at 16:58. On 6/5/16 at11:40 the turbine wasagain tied to the grid and at 12:27 a Reactor Trip - Safety lnjectionoccurred due to the #1 governor valve failing to the open position.After repairs to the governor valve, as well as additional work on 28 MainFeed Pump, the unit was returned to Mode 2 on 618116 at 01:39. Mode 1was re-entered 6/8/16 at 09:32. On 6/9/16 at 06:40 the generator wassynchronized to the grid. An un-isolable steam leak required a turbine tripon 6/9/16 at 17:52. Repairs were made and Unit 2 was synchronized tothe grid at 13:23 on 6/11116. Power was increased to allow testingbetween 25 and 30 percent with the following Power Ascension Test beingcompleted as scheduled:o 2-PAT-1 .4, Pipe Vibration Monitoring, completed on 6/15/16 with allcriteria met for observations at the 30% Plateau.o 2-PAT-5.3, Automatic Steam Generator Level ControlTransients atLow Power was completed on 6/15/16 with allAcceptance Criteriamet. CR 1181278 was initiated to document one Review Criteria thatwas not met. An engineering evaluation determined this did not affectthe performance of the test nor invalidate any of the test results andtesting should proceed to the next plateau.o 2-PAT-1.5, Loose Parts Monitoring System, was completed on6113116 with all criteria met. CR 1171424 documents three channelsremoved from service.o 2-PAT-1.8, Thermal Expansion of Piping Systems, was field workcomplete on 6/15/16 with no issues noted.. 2-PAT-1.10, lntegrated Computer System (lCS), was completed on6114116 with all criteria met. CR 1181784 was written to address adatabase error but did not affect this plateau performance.o 2-PAT-1.11, RVLIS Performance Test, applicable sections werecompleted on 6/13/16 with all criteria met.o 2-PAT-1.12, Common Q Post Accident Monitoring System, applicablesections were completed on 6113116 with all criteria met.91 7.1Test Sequence for 30% Plateau (2-PAT-5.0) (continued). 2-PAT-1.7, Operational Alignment of Process Temperaturelnstrumentation, was completed on 6115116. All AcceptanceCriteria were met. Two Review Criteria concerning parametersrelated to Delta T failed. The OTDT calculated by Eagle-21 andprovided by the MMI carts indicated approximately 158% and theMCR indicators maximum value is 150%. lt was expected thereading from Eagle-21 was accurate and the MCR meters wereranged such that they cannot read the higher value. Additionaldata was taken at higher power ranges and the meters came onscale with no issue. CR 118246 was written.o 2-PAT-5.4 Calibration of Steam and Feedwater Flow lnstruments at30% Power was completed on 6/15/16 with all criteria met for the30% Plateau.o 2-PAT-1.6 Startup Adjustments of Reactor Control System wascompleted on 6/15/16. This was data taking only with no Review orAcceptance Criteria at this plateau.Additionally, Engineering completed the following procedures, with noissues, to support their testing at the 27-29 percent power level:o 2-T141- lncore Flux Mappingo 2-TRl-0 PDMS Operabilityo 2-Sl-0 Excore QPTR & Axial Flux Differenceo 2-Sl-92 lncore-Excore Cross Calibration Datao 2-T17.020 - PDMS Calibrationo 2-fl Calorimetric CalibrationAfter completion of all testing in this PAT it was noted that tempering flowisolations occured that did not meet the requirements of Westinghouselefter WAT-D-6432. CR 1182320 was written to document tempering flowisolations that occurred as part of testing at this plateau. Correctiveactions from this CR evaluated the length of the isolation and revised 2-SOI-2&3.01 adding a Precaution about WAT-D-6432.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.0 ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAT-5.0.98 7.1.'a Dynamic Automatic Steam Dump Control (2-PAT-5.1)This test was performed as part of test sequences 2-PAT-3.0, Post Core LoadingPrecritical Test Sequence, and 2-PAT-5.0, Test Sequence for 30o/o Plateau. Thetest began on 118116 and was field work completed on 5127116.The steam dump valves were tested without steam flow during sequence 2-PAT-3.0 in accordance with Sections 6.1 and 6.2.The steam dump valves were tested with steam flow during sequence 2-PAT-5.0in accordance with Sections 6.3, 6.4 and 6.5.The plant was less than 15% power, in Mode 1 with the main turbine notsynchronized to the grid.For Sections 6.6, the steam dump valves were tested for the deferral fromStartup with steam flow during Mode 3. This was done to confirm stroke times ina!! three of the following simulated scenarios: modulate open, trip close, and tripopen. Additionally, vibration testing on the valves which was deferred fromStartup Testing was performed.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the operation of the Steam Dump Control System. TheSteam Dump Control System has three modes of control; SteamPressure, Plant Trip, and Load Rejection=Each mode of controlwas tested to demonstrate stability following a small transient.1.2 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 28,Dynamic Automatic Steam Dump ControlTest Summary.1.3 Address additional scope of testing added to the Power AscensionTesting Program for the deferred Turbine Bypass System(Condenser Steam Dump Valves) testing. This test was to veriffthe (12) Steam Dump Valves stroke times are within acceptablelimits and to obtain vibration data on deferred Steam Dump valvesnot obtained during Hot Functional Testing in accordance with2-PAT-1.4.1.4 Satisfy the requirements of UFSAR Chapter 14,fable 14.2-1,Sheet 62, Main Steam System Test Summary by collectionvibration data on deferred Steam Dump valves not obtained duringHot Functional Testing in accordance with 2-PAf -1.4.99 7.1.1 Dynamic Automatic Steam Dump Gontrol (2-PAT-5.1) (continued)2.03.0Test MethodsThe steam dump control system is designed to maintain RCS averagetemperature by dumping steam to the condenser. This instructionfunctionally tested all three control modes (steam pressure control,plant trip control, and load reduction control) while reactor power was low(i.e., <15% power).The functional test included modulating the valves open and closed, andtripping open allsteam dump controlvalves using simulated signals whilesteam flow was isolated. The Steam Pressure controller was tested byvarying reactor power and observing the controller automaticallymaintained steam header pressure by changing steam flow to thecondenser. The Plant Trip controller was tested by simulating a reactortrip, varying reactor power, and observing controller parameters andoutput. The Load Rejection controller was tested by simulating the loss ofload permissive, and observing controller parameters and output.2-PAf-1.4, Pipe Vibration Monitoring, data was collected during theperformance of the PAT.Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaNote: There were no Acceptance Criteria for Sections 6.1 and 6.2.3.1 Section 6.3 (Steam Pressure Controller)3.1.1 After varying reactor power, the steam pressure controllermaintains steam header pressure stable, as demonstratedby neither the steam header pressure signal nor the steamdump demand signal showing divergent oscillations.During the transient neither the steam header pressuresignal nor the steam dump signal showed a divergentoscillation.3.1.2 After varying reactor power, steam pressure controllermaintains steam header pressure stable, as demonstratedby the steam dump control system remaining in automaticthroughout the transient.The steam pressure controller maintained steam headerpressure stable and the steam dump control systemremained in automatic throughout the transient.100 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)3.23.3Section 6.4 (Plant Trip Controller)3.2.1 After varying reactor power, the plant trip controllermaintains a stable Tavg as demonstrated by neither theRCS Tavg signal nor the steam dump demand signalshowing divergent oscillations.During the transient neither the RCS Tavg signal nor thesteam dump demand signal showed a divergent oscillation.3.2.2 After varying reactor power, the plant trip controllermaintains a stable Tavg as demonstrated by the steamdump control system remaining in automatic withoutdivergent oscil lations.The plant trip controller maintained Tavg stable andthe steam dump control system remained in Automaticthroughout the transient without divergent oscillations.Section 6.5 (Load Rejection Gontroller)3.3.1 The loss of load controller responds properly for the plantinput signals to the controller.The loss of load controller responded properly for the plantinput signals to the controller.3.4 Section 6.6 (Condenser Steam Dump valves stroke times)3.4.1 Condenser steam dump valves modulate open, trip closedand trip open stroke times are within acceptable limits.101 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)Modulation times and local/remote stroke times indicatedbelow are within Acceptance Criteria.Review Criteria3.5 Section 6.1 (Static Valve Timing - Modulation)3.5.1 The full open stroke length for each steam dump controlvalve is 2 314 inches to 3 inches.Review criteria were met as delineated below:Steam DumpValveExpectedStroke LenqthMeasuredStroke Lenqth2-FCV-1-1 0323t4-3n.2.8122-FCV-1-10423t4-3n.2.752-FCV-1-1 0523t4 - 3n.2.9752-FCV-1-1 062314 - 3n.2.8752-FCV-1-10723t4 - 3n.2.9222-FCV-1-1 0823t4 - 3n.2.812-FCV-1-1 0923t4 - 3n.2.8752-FCV-1-1 1023t4 - 3n.2,9122-FCV-1-11123t4 - 3n.2.8122-FCV-1-11223t4-3n.2.752-FCV-1-1 132314 - 3n.2.8122-FCV-1-11423t4-3n.2.812VALVEMODULATE OPEN<20 secTRIP CLOSES5secTRIP OPEN<3sec2-FCV-1-1 036.5 I 5.713.0 I 2.72.2 I 1.762-FCV-1-1046.2 I 4.612.25 12.902.32 12.122-FCV-1-1 0513.31 / 9.693.20 / 3.102.80 12.542-FCV-1-1 069.34 19.292.96 I 3.112.60 12.082-FCV-1-1075.40 I 6.202.74 12.502.90 12.332-FCV-1-1 089.11 / 5.892.90 l2.gg2.92 12.352-FCV-1-1 099.53 / 6.93.41 / 3.362.94 12.092-FCV-1-1 109.16 I 7.883.48 I 2.682.62 12.182-FCV-1-1114.02 13.463.08 I 3.182.56 I 1.622-FCV-1-1124.67 14.262.76 12.672.19 I 1.692-FCV-1-1 136.39 / 6.03.18 / 3.102.00 12.102-FCV-1-1149.96 I 9.413.00 / 3.502.95 12.30r02 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)3.5.2 The opening modulation time for each Steam Dump Contro!Valve is less than 20 seconds upon the receipt of a 5% to95% control signal step change.Review criteria were met as delineated below:Steam DumpValveOpen StrokeTimingRequirementActual OpenStroke Timing2-FCV-1-1 033 20 Seconds2.762-FCV-1-104s 20 Seconds4.812-FCV-1-1 05s 20 Seconds10.472-FCV-1-1 063 20 Seconds7.532-FCV-1-107s 20 Seconds7.102-FCV-1-1 083 20 Seconds10.912-FCV-1-1 09s 20 Seconds6.912-FCV-1-1 10= 20 Seconds5.912-FCV-1-111s 20 Seconds6.722-FCV-1-1123 20 Seconds3.962-FCV-1-1133 20 Seconds4.532-FCV-1-114= 20 Seconds9.083.5.3 The closing modulation time for each steam dump controlvalve is less than 20 seconds upon the receipt of a 95% to5% control signal step change.Review criteria were met as delineated below:Steam DumpValveClosed StrokeTimingRequirementActual ClosedStroke Timing2-FCV-1-1 033 20 Seconds5.322-FCV-1-1043 20 Seconds5.162-FCV-1-1 053 20 Seconds9.612-FCV-1-1 06s 20 Seconds5.222-FCV-1-1073 20 Seconds7.182-FCV-1-1 083 20 Seconds5.362-FCV-1-1 093 20 Seconds8.482-FCV-1-1 10s 20 Seconds7.962-FCV-1-1113 20 Seconds5.382-FCV-1-1123 20 Seconds5.482-FCV-1-113s 20 Seconds6.912-FCV-1-1143 20 Seconds11.92103 7.1.1 Dynamic Automatic Steam Dump Contro! (2-PAT-5.1) (continued)3.6 Section 6.2 (Static Valve Timing -Trip)3.6.1 All of the steam dump controlvalves trip open in 3 3 secondsfollowing a simulated Hl-Hl T"rn signal.Review criteria were met as delineated below:Steam DumpValveClosed StrokeTimingRequirementActual ClosedStroke Timing2-FCV-1-1 03s 3 Seconds1.232-FCV-1-104s 3 Seconds1.572-FCV-1-1 05s 3 Seconds2.412-FCV-1-1 06< 3 Seconds1.922-FCV-1-1A7s 3 Seconds2.462-FCV-1-1 08s 3 Seconds2.112-FCV-1-1 09< 3 Seconds1.772-FCV-1-1 10s 3 Seconds2.092-FCV-1-111s3 Seconds1 .852-FCV-1-1123 Seconds1.462-FCV-1-1133 Seconds1.612-FCV-1-1143 Seconds1.853.6.2 AII of the steam dump control valves trip closed in S 5seconds following a simulated block signal.Review criteria were met as delineated below:Steam DumpValveGlosed StrokeTimingRequirementActual ClosedStroke Timing2-FCV-1-1 03s 5 Seconds1.202-FCV-1-104s 5 Seconds1.422-FCV-1-1 05s 5 Seconds1.522-FCV-1-1 06s 5 Seconds1.702-FCV-1-107< 5 Seconds1.692-FCV-1-1 08s5 Seconds1.412-FCV-1-1 09s5 Seconds1.972-FCV-1-1 10s5 Seconds1.892-FCV-1-1115 Seconds1.912-FCV-1-112s 5 Seconds1.452-FCV-1-113s 5 Seconds1.702-FCV-1-114s 5 Seconds1.74104 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)4.03.7 Section 6.3 (Steam Pressure Controller)3.7.1 After varying reactor power, the steam pressure controllercontrols steam header pressure at setpoint (t 25 psi) withinnine minutes (three reset time constants).The steam pressure controller controlled steam headerpressure at setpoint (t 25 psi) within nine minutes aftervarying reactor power. Steam header pressure remainedwithin the setpoint throughout the transient.3.8 Section 6.4 (Plant Trip Controller)3.8.1 Before and after varying reactor power, the plant tripcontroller demand signal remained within 2.0o/o of thecalculated demand signal.The plant trip controller demand signal remained within 2.0o/oof the calculated demand signal before and after varyingreactor power.Note: There were no Review Criteria for Sections 6.5 and 6.6.Problems11] CR 1122945: During performance of WO 117441539, a jumperwas placed on the wrong termina! point in Step 4.3.2[9] and anunexpected alarm was received in the Main Control Room. Workwas stopped and the CR was initiated. Urgent Change, UC-1, wasmade to 2-PAT-5.1 Revision 2 to select a more accessible terminalpoint and labeled the terminal point for the jumper placement.CR 1123150: 2-FCV-1-105 failed to re-close following the trip opentest. The CR was written to troubleshoot. The needle valvebetween the positioner and the diaphragm on 2-FCV-1-105 wasfound to be closed. After verifying that the needle valve should bein the open position, the needle valve was opened and2-FCV-1-105 closed as expected. All other steam dump needlevalves were verified to be in the open position.l2l105 7.1.1 Dynamic Automatic Steam Dump Contro! (2-PAT-5.1) (continued)t3ICR 1124648: During performance of the steam dump sequencetest in Step 6.1 .21281, the Data Sheet 2 steam dump valve positionindications were not met at the three demand levels. The CRinitiated WO 117506695 which was implemented, calibrating thesteam dump controllers. ln addition, the upper limit switch for 2-FCV-1-103 was found to be sticking and it was replaced under thework order. The Step 6.1.21281, sequence test was then re-performed and the Data Sheet 2 steam dump valve positions werenot met again at similar demand positions as the first performance.The position indications that were not met in both sequence testswere associated with valve 20o/o open and 80% open positions.These positions are close to where the open and closed limitswitches actuate to turn on or off the red and green positionindication lights. lt was determined that the steam dump valvesequence was acceptable because in each sequence test eachbank of valves were full open before the next bank began to openand each valve modulated as required between fullopen and fullclosed demand positions. Data Sheet 2 was removed from theReview Criteria in Revision 3 to prevent additional unwarrantedconditional reports and repairs.CR 1124788 and CR 1127374: During performance of the steamdump valve modulation stroke timing test in Section 6.'1.3 through6.1.6, the greater than 12 second valve stroke time requirementwas not met.It was determined that the greater than 12 second stroke time is nota requirement in any design documentation. ln addition, there areno plant procedures that set up the valves to ensure a greater than12 second modulation stroke time. This criteria was removed from2-PAT-5.1 in Revision 3.CR 1174915: During transfer of steam dump control to the loss ofload controller, a diverging oscillation was observed in the loss ofload controller response. The loss of load controller response wasfound to be proper for the plant input signals to the controller andthe high gain settings of the controller. The evaluation of the loss ofload controller was documented in Westinghouse Letter LTR-SCS-16-23.l4lt5I106 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)t6IUrgent Change UC-1 to 2-PAT-5.1 Revision 0005 revised Step6.5.2[4] and Acceptance Criteria 5.1.3[A] to verifo the loss of loadcontroller responds properly for the plant input signals to thecontroller. Urgent Change UC-1 revised the remainder of the lossof load transient testing and evaluation in Sections 6.5.2 and 6.5.3.Acceptance Criteria 5.1.3[B] and Review Criteria 5.2.5 weredeleted. ln addition, the Westinghouse test scoping documentWATMBT-SU-2.8.5 Acceptance Criteria was revised to veriff theIoss of load controller responds properly. Also a SAR ChangePackage No. U2-021 was approved and issued that revisedChapter 14 Table 14.2-2 Sheet 28 to verify the load rejectioncontroller responds properly. These changes were based on theWestinghouse Letter LTR-SCS-16-23 which documented theproper response of the load rejection controller.CR 1'170159: 2-FCV-1-108 did not initially meet the trip openstroke time Acceptance Criteria of 3 seconds. A volume boosteradjustment was made to 2-FCV-1-108 under WO 117826339 andthe valve met the trip open stroke time when retested.CR 1170319: Piping vibration at2-FCY-1-111 did not meetAcceptance Criteria during the stroke test. Civil DesignEngineering evaluated the piping response and found it acceptable.17ltBlL01 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAr-5.3)This test was performed as part of test sequence 2-PAT-5.0, Test Sequence for30% Plateau. The test began on 5114116 and completed on 6l'15116.1.0Test ObiectivesThe objectives of this test were to:Demonstrate the proper operation and automatic response of theSteam Generator Level Control System for each Steam Generatorduring steady-state operation.2.01.2 Satisfy, in part, the requirements of UFSAR Table 14.2-2, Sheet 30,Automatic Steam Generator Level Control Test Summary.Test MethodsThe UFSAR requires tests be performed at various power levels from 5%through 100o/o reactor power. This PAT tested low power aspects of theUFSAR requirement. For Section 6.1 the plant was in Mode 1 at less than10% power with the main turbine not synchronized to the grid. For Section6.2 the plant was in Mode 1 at approximately 30% power after the MFWForward Flush/Back Flush Heatup had been completed. The test wasperformed in conjunction with maintenance work order activities to collectdata needed to calibrate and tune feedwater control system components.Actual testing in Section 6.1, Feedwater Bypass Control Valves, wasstarted on dayshift 5125116 and completed on nightshift 5/26/16. AllAcceptance and Review Criteria were met for Section 6.1, with noadditional tuning of the Feedwater Bypass Control Valve Controllers beingnecessary.Section 6.2, Transfer From Bypass To MFW Reg Valves, was initiated on6112116. During the performance of Section 6.2.1, Transfer From BypassTo MFW Reg Valve For SG No. 1, the Steam Generator leve! was notstable within the required + 2o/o during the 10 minute monitoring period.The decision was made to continue testing in accordance with Section6.2.4 for SG No. 4 and perform a re-test of Section 6.2.1 at a Iater time.During the performance of Section 6.2.4, Transfer From Bypass To MFWReg Valve For SG No. 4, the Main Feedwater Reg Valve, 2-FCV-003-0103, did not respond to a 30% demand. lt was determined that the airline to the MFW Reg Valve for SG No. 4 was Ieaking. WO 117904374was written and performed to repair the leak.1.1108 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued)On 6/13/16, Re-Test #1 was performed for Section 6.2.1, Transfer FromBypass To MFW Reg Valve For SG No. 1. All Acceptance Criteria weremet for SG No. 1 MFW Reg Valve with no adjustments being made.Repairs to the SG No. 4 air line were completed on 6113116 under WO117904374, and testing was resumed for the SG No. 4 MFW Reg Valve.All four MFW Reg Valves successfully met the Acceptance Criteria uponcompletion of testing in Section 6.2. At the conclusion of testing, CR1181278 was written to address areas of concern.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaSection 6.1 (Feedwater Bypass ControlValves)3.1.1 The indicated steam generator level undershoot was lessthan 4.0% below the final setpoint following automaticrecovery from high steam generator level.The steam generator level undershoot ranged from 1% to2o/o below the setpoint for al! four Steam Generators,following automatic recovery from high steam generatorlevel, which met the required Acceptance Criteria.3.1.2 The indicated steam generator level overshoot was less than4.0o/o above the setpoint following automatic recovery fromlow steam generator leve!.The steam generator level overshoot ranged from -1% to 0%above the setpoint for all four Steam Generators, followingautomatic recovery from high steam generator level, whichmet the required Acceptance Criteria.3.1109 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (continued)3,2Section 6.2 (Transfer from Bypass to MFW Control Valves)3.2.1 lndicated steam generator level returned to and remainedwithin t2o/o of the program levelwithin 10 minutes followingthe transfer of level control to the Main Feedwater Reg.Valves in automatic.All four steam generator level indications returned to andremained within t2o/o of the program level within 10 minutesfollowing the transfer of level control to the Main FeedwaterReg. Valves in automatic. CR 1181278 was written due toquestions regarding the wording of the Acceptance Criteriain 2-PAT-5.3 and is discussed under Problems.3.2.2 Demand signal oscillations for each of the Main FeedwaterReg. Valves were less than +6.00/o during steady stateoperation.Allfour Main Feedwater Reg Valves exhibited less than 16%oscillation of the Main Feedwater Reg Valves in Auto:o SG #1 - 2.17o/o. SG #2 - 1.24o/oo SG #3 - 2.81o/oo SG #4 -2.44YoCR 1181278 also addressed an issue with the procedure notbeing clear on the time data is recorded. This is discussedunder Problems.3.2.3 Feedwater flow oscillations to each Steam Generator wereless than t6.0% during steady state operation.The feedwater flow oscillations to each Steam Generator aredocumented below:o sG #1 - 3.41%o sG #2 - 6.39%o sG #3 - 5.59%o SG#4-4.75o/oThe feedwater flow oscillations to SGs # 1,3, and 4 met theAcceptance Criteria of less than +6 .0o/o during110 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (continued)steady state operation. CR 1181278 discussed in Problemsaddresses #2 SG data as acceptable.Review Criteria3.3 Section 6.1 (Feedwater Bypass ControlValves)3.3.1 lndicated Steam Generator level returned to and remainedwithin t2% of the level setpoint within 37.5 minutes followingautomatic recovery from high Steam Generator level.The indicated Steam Generator level returned to andremained within *2o/o of the level setpoint within 37.5 minutesfollowing automatic recovery from high Steam Generatorlevel as shown below:o SG #1 - 14 minuteso SG #2 - 13 minutes: 33 trl.1Zilitxf:3.3.2 lndicated Steam Generator level returned to and remainedwithin t2o/o of the level setpoint within 37.5 minutes followingautomatic recovery from low Steam Generator level.The indicated Steam Generator level returned to andremained within t2o/o of the level setpoint within 37.5 minutesfollowing automatic recovery from low Steam Generator levelas shown below:: :: #,1\ilIlil:i3.4 Section 6.2 (Transfer from Bypass to MFW Reg. Valves)3.4.1 The Main Feedwater Reg. Valve position was between theminimum and maximum positions given in Figure 1 of2-PAT-5.3 for the specific loop Main Steam Flow.111 7.1.2 Automatic Steam Generator Level ControlTransients at Low Power(2-PAT-5.3) (conti nued )Data indicated that the Main Feedwater Reg Valve positionsexceeded the maximum positions given in Figure 1 of2-PAT-S.3:. Reg Valve 2-FCV-3 41.0o/o. Reg Valve 2-FCV-348 - 45.40. Reg Valve 2-FCV-3 40.9Yo. Reg Valve 2-FCV-3-103 - 40.9%CR 1181278 was generated and Engineering was requestedto evaluate the data and provide recommendations.Engineering's recommendation was to proceed with powerAscension Testing to the 50% plateau.3.4.2 lndicated Steam Generator level was within t2o/o of theprogram Leve! within 10 minutes following Main Feed Reg.Valve being placed in AUTO and subsequent stableconditions steady state operations.Data verified that the indicated Steam Generator levels werewithin t2o/o of the program level within 10 minutes followingMain Feed Reg. Valve being placed in AUTO andsubsequent stable conditions for steady state operations.3.4.3 The Main Feedwater Header Pressure oscillations were lessthan 108 psi (peakto-peak) during steady state operations.(This limit was based on r3.0% of the instrument span of1800 psi).The Main Feedwater Header Pressure oscillations were lessthan 108 psi (peak-to-peak) during steady state operationsas shown below:o Main Feedwater Header Pressure OscillationFor SG #1 - 5 psio Main Feedwater Header Pressure OscillationFor SG #2 - 8 psio Main Feedwater Header Pressure OscillationFor SG #3 - 8 psio Main Feedwater Header Pressure OscillationFor SG #4 - 4 psi3.4.4 The Actual (measured) AP was within 25.0 psi of theProgram AP during steady state operation.The actual (measured) AP was 0.8 psi which met theAcceptance Criteria of being within 25.0 psi of the programAP during steady state operation.LL2 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued )4.0Problemst1] CR 1181278 was written due to questions regarding the wording ofthe Acceptance Criteria in 2-PAT-5.3. A comparison of 5.1.2Acceptance Criteria for Section 6.2 (Transfer from Bypass to MFWControlValves) to the Westinghouse document WBT-D4709 (LTR-PCSA-14-31) confirmed that the Acceptance Criteria in2-PAT-5.3 were written correctly. The steps in the body of theprocedure to perform the test and verify the Acceptance Criteriawere also reviewed with the originator of the CR. It was determinedthat the procedure for this section of the test was written correctlyand neither the test nor the results were invalidated by theconcerns in the CR.CR 1181278 documented another concern which stated "Theprocedure is not clear if the performer looks at the data before orafter a time. The procedure should say AFTER, because that is theapproximate time that the main feedwater is transferred into Auto.With clarification, allAcceptance Criteria are met." A review ofData Sheet 11 revealed that for each Main Feedwater Reg Valve,the column to record data contains notation which states "Data fromtime in Step 6.2.X[18]". Step 6.2.X[18] recorded the end time forthe 10 minute monitoring period. lt would have been better if DataSheet 11 would have stated "Data from monitoring period in Steps6.2.X1161 through 6.2.X[18]". Additionally, during the review, it wasdiscussed that the Acceptance Criteria was to monitor the DemandSignal oscillations for each of the Main Feedwater Reg Valves;however, the test kept the Reg valves in Manual instead of Autoduring this portion of testing. Fortunately, the Test Coordinatorscollected the appropriate data with the valves in Manual, thenswapped the controller position to Auto, as allowed by theprocedure, and collected the appropriate data in this condition. Thedata was analyzed with the valves in AUTO and it was determinedthat the Acceptance Criteria were met. Since the data was alsocollected with the valves in AUTO and the Acceptance Criteria weremet, there was no need to re-perform this section of the test.113 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued )l2lThe data recorded in the test for the feedwater flow oscillations toSG #2 was above the acceptable limit of 60/o for flow oscillations;however, CR 1181278 states "The feedwater was controlled anddid not oscillate. However, the maximum deviation was about6.39% of average flow." A review of the data indicated that thefeedwater flow to SG #2 started off with a deviation of >6o/oihowever, the controller brought the flow to within an acceptablerange in a steady manner and maintained an acceptable flow ratherthan oscillating for a period of time (see Figure 7.1.2-1) The factthat the feedwater flow stabilized within a range which was lessthan 6% without oscillating meets the intent of the AcceptanceCriteria.CR 1'181278 also requested Engineering to evaluate data on oneReview Criteria and provide recommendations. Engineering'srecommendation was to proceed with Power Ascension Testing tothe 50% plateau for the following Review Criteria:o Section 6.2, MFW Reg Valves were not between theminimum and maximum positions requiredWO 117904374 was initiated to repair a leak on the air Iine toli4Steam Generator MFW Reg. Valve. The valve was retested afterrepairs and passed Acceptance Criteria.lL4 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued )FIGURE 7.1 .2-1SG #2 FW Flow Oscillationflat^tiirc Hlclory L3-f un-zolrB 1I,!*6:()0 to ,.3-Jun-2or.6 1():56:00 cur)552oat-s tn700 -10 {olO:aa:OO EDT !O24,2.o gf ,.O:.lt:zo Ef ,O:3;t:@ EY 1O:52:rlO ET 1O:S.:ZO Ef lO:54:@ EDr1l-rm-20:ta l3-rn-aclj l'!-rrlt-aqt.3 1l-run-20L at-rr2q13 a3-rlli-zct3 af-rrr-20LEl'E##tCl$lF* 2 vAL:rD{rE n LEVEL t8t{ g+ Fr'(z) Fr@r@aa., (k:r,:l) z-FIC-ln3r arrE Lro{T -10 1rt(t) rso74 (El:t t) 30 a LEVEL scTFlolt{T 40 co r(a) Fo.rarA &:ui!) Jtr Gr z FEE irr rx a Fl..r, z@ a3o rtHS 2 VALIDATED115 II7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4)This test was performed with the plant stable at approximately 30% Power as part of2-PAT-5.0, Test Sequence for 30% Plateau. The test began on 6/13/16 and wascomplete on 6/15/16.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the output of the eight feedwater flow transmitters for "zero"output with minimal feedwater flow, collect data for determining thenew calibration spans for the steam flow transmitters1.2 Verify the calibration of the feedwater and steam flow transmitters,by comparing indicated flows between the Main Control Boardlndicators, the Protection System, and the Control System.1.3 Satisfy, in part, the 30% objective in the UFSAR Table 14.2-2,Sheet 21, Calibration Of Steam And Feedwater Flowlnstrumentation At Power Test Summary.2.0 Test MethodsAt approximately 30% power, each feedwater flow transmitter was placedin bypass and verified for "zero" output.At approximately 30% power, steam generator blowdown and temperingflow were isolated while data was collected. Steam generator blowdownand tempering flow were then reestablished and calculations/comparisonswere performed.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaNoneLL6 7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4) (continued)Review Criteria3.1 At zero DP, the output from each Feedwater Flow Transmitter andits associated loop reflect zero flow as demonstrated by thefollowing criteria:A. Computer Point: 0.000 KBH (-25.5 to 25.5 KBH)B. Flow Test point: 0.200 Vdc (0.1858to 0.2142Ydc1C. Computer Test Point: 0.200 Vdc (0.1858 to 0.2142 Vdc)D. DP Test point: 0.1983 Vdc (0.1848 to 0.2118 Vdc)The data below was collected and the output flow was verified withinthe Review Criteria requirements.3.2 The difference between the Feedwater Flow as measured in theProtection System and the Main Control Board lndicators is withint5.0% of the rated flow.Measured differences (% ERRORS) between -1.89Yo and +1.18%3.3 The difference between the feedwater flow as measured in theProtection System and the lndicated Computer Feedwater Flow iswithin t2.Oo/o of rated flow.Measured differences (% ERRORS) between -0.24o/o and +9.967oDESCRIPTIONSTEAM GENERATOR 1STEAM GENERATOR 22-FT-3-3542-FT-3-35B2'FT'3'48A2-FT-3-48BComputer Pont Flow (KBH) 5.2.A.15111Flow Test Pont 5.2.4.20.201480.200280.201540.20055Comp Test Point 5.2.A.30.200894.200320.200490.20029DP Test Point Voltaoe 5.2.A.40.202690.1 95890.1 97630.1 991 2DESCRIPTIONSTEAM GENERATOR 4STEAM GENERATOR 32-FT-3-90Az-FT-3-9082-FT-3-103A2-FT-3-103BComputer Point Flow (KBH) 5.2.A.142-31Flow Test Point 5.2.4.20.201040.200740.200560.20027Como Test Point 5.2.A.30.200670.200380.1 99560.20043DP Test Point Voltaoe 5.2.A.40.199450.197490.1 95890.1 9623LLl 7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4) (continued)3.4 The difference between the feedwater flow as measured in theProtection System and the Feedwater Flow Signal used for flowcontrol is within t2.0% of rated flow.Measured differences (% ERRORS) between -0.29Yo and -0.07o/o3.5 The difference between the steam flow as measured in theProtection System and the Main Gontrol Board lndicators is withint5.0% of the rated flow.Measured differences (% ERRORs) between -0.87o/o and +0.94%3.6 The difference between the steam flow as measured in theProtection System and the lndicated Computer Steam Flow iswithin t2.0o/o of rated flow.Measured differences (% ERRORS) between -0.10% and 0.00%3.7 The difference between the steam flow as measured in theProtection System and the Steam Flow Signal used for flow controlis within t2.0o/o of rated flow.Measured differences (% ERRORS) between -0.23o/o and 0.00%3.8 The difference between the feedwater flow as measured in theProtection System and the Steam Flow as measured in theProtection System is within t5.0o/o of rated flow.Measured differences (% ERRORs) between -2.620/o and +4.45o/o3.9 The difference between the feedwater flow as measured in theContro! System and the Steam Flow as measured in the ControlSystem is within t5.0% of rated flow.Measured differences (% ERRORS) between -2.57o/o and +4.560./0118 7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4) (conti nued )Additionally, a comparison of the corrected feedwater flows and steam flowsversus predicted design flow is provided:4.0 ProblemsThere were no significant problems encountered during the performanceof this test.119 7.2 Test Sequence for 50o/o Plateau (2-PAT-6.0)This test started on 5130116 and was completed on 7116/1 6.1.0 Test ObiectivesThe objectives of this test were to:1.1 ln conjunction with 2-GO4, Normal Power Operation, define theplant operational requirements and ensure those requirementswere met in order to permit power escalation from 30% RatedThermal Power (RTP) to 50%.1.2 Specifo the order of test performance at the 50% plateau.The following PATs/PETs were sequenced for performance by2-PAT-6.0:o 2-PAT-1.4
Pipe Vibration Monitoringo 2-PAT-1.5
Loose Parts Monitoring Systemo 2-PAT-1.6 " Startup Adjustments of Reactor Control Systemo 2-PAT-1.7
OperationalAlignment of Process Temperaturelnstrumentationo 2-PAT-1.8
Thermal Expansion of Piping Systemso 2-PAT-1.10* lntegrated Computer System (lCS)o 2-PAT-1.11* RVLIS Performance Testo 2-PAT-1.12" Common Q Past Accident Monitoring Systemo 2-PAT-3.3
RCS Flow Measuremento 2-PAT-5.2 Turbine Generator Trip With Coincident Loss ofOffsite Power Testo 2-PAT-6.1 Automatic Reactor Control Systemo 2-PAT-6.2 Automatic Steam Generator Level Control Transientsat 50o/o Powero 2-PAT-6.3 Calibration of Steam and Feedwater Flowlnstruments at 50 % Powero PET-301 . Core Power Distribution Factorso PET-304
OperationalAlignment of NIS. RCI-159
Radiation Baseline SurveysNote:
lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.L20 7.2Test Sequence for 50% Plateau (2-PAT-6.0) (continued)2.0 Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on5/30/16 and completed on 6117116 and included verification of thefollowing major items:. 2-PAT-5.0, Test Sequence for 30% Plateau, completeo NPG-SPP-10.4, Reactivity Management Program, ReactivityControl Plans were developed to support the planned testing forthis sequenceo Reactor power between 27o/o and 29% RTP with T"rn-T,"r mismatch+1.5 "F or less. RCS pressure is between 2220to 2250 psig. Section 6.5 of 2-PET-304, Operational Alignment of NlS, to adjustthe Power Range High Flux Level Trip setpoints for testing at the50% Plateau completePower increase to the 50% testing plateau was initiated on 6/17116 at11:40 and PAT testing in Section 6.1 of 2-PAT-6.0 was begun. On6120117 at 15:37, U-2 Turbine tripped due to the loss of 28 MainFeedwater Pump and subsequently an automatic Reactor Trip occurred at15:40 due to S/G levels reaching their low-low trip setpoint. The plant wasstabilized in Mode 3.Unit 2 re-entered Mode 2 on 6123116 at 17:37 and the reactor critical at17:53. Mode 1 entry was made at 03:00 on 6124116. The U-2 generatorwas synchronized to the grid at 13:58.A manual turbine trip was initiated on 6126116 at 09:45 due to a steamleak. Reactor power was reduced and the Unit entered Mode 2 at 11;44.At 15:26 the reactor was tripped manually and the unit stabilized in Mode3.Mode 2 was again entered on712116 at 03:00 with reactor criticality at03:20. U-2 entered Mode 1 at07:57 and was synchronized to the grid inthe afternoon at 13:36. On717116 the 50% Plateau power level testingwas reached and the 50% tests were commenced. Steady state testingincluded:o 2-PAT-1.4, Pipe Vibration Monitoring, completed on718116 with allcriteria met.o 2-PAT-1.5, Loose Parts Monitoring System, was completed on717116 with all criteria met. CR 1171424 documents three channelsremoved from service.o 2-PAT-1.6, Startup Adjustments of Reactor Control System, wascompleted on718116. This was data taking only with no Review orAcceptance Criteria at this plateau.L2L 7.2 Test Sequence for 50% Plateau (2-PAT-6.0) (continued). 2-PAT-1.7, Operational Alignment of Process Temperaturelnstrumentation, was completed on 718116 with allcriteria met.o 2-PAT-1.8, Thermal Expansion of Piping Systems, was field workcomplete on718116 with 2 issues referred to engineering forevaluation with Problem Report #4. Engineering review indicated itwas acceptable to continue Power Ascension Testing.o 2-PAT-1.10, lntegrated Computer System (lCS), was completed on718116 with all criteria met.o 2-PAT-1.11, RVLIS Performance Test, applicable sections werecompleted on117116 with all criteria met.o 2-PAT-1.12, Common Q Post Accident Monitoring System,applicable sections were completed on 7nh6 with al! criteria met.o 2-PAT-6.3, Calibration of Steam and Feedwater Flow lnstrumentsat 50% Power, was completed on 718116 with all criteria met.o 2-PAT-3.3, RCS Flow Measurement, was completed for the S0%Plateau on719116 with all criteria met.o RCI-159, Radiation Baseline Surveys, was completed for the 50%plateau onT110116. No Acceptance or Review Criteria wereassociated with this procedure.Transient tests were begun onTl'11116 and included the following:o 2-PAT-6.1, Automatic Reactor Control System, was completed on7113116 with all criteria met.o 2-PAT-6.2, Automatic Steam Generator Level ControlTransients,was completed on 7116116 with all criteria met.o 2-PAT-5.2, Turbine Generator Trip With Coincident Loss of OffsitePower Test, was completed on7114116 with all criteria met exceptone Review Criteria. CR 1192287 was written to document Tcoldgoing below the 547"F criteria.o 2-PAT-1.4, Pipe Vibration Monitoring, for transient testing wascompleted on7l14116 with all criteria met.2-PAT-1.2, Load Swing Test, originally scheduled for the 50% plateau,was revised to allow performance during 2-PAT-7.0 due to the inability ofthe turbine to be operated in IMP lN. Repairs to the circuitry wereevaluated during the outage and a procedure revision was made to allowperformance of the Load Swing Test in lMP OUT on the turbine controls.L22 7.2 Test Sequence for 50% Plateau (2-PAT-6.0) (continued)Additionally, Engineering completed the following procedures or applicablesections during the steady state period, with no issues, to support theirtesting at the 50% Plateau:o 2-T141- lncore Flux Mapping. 2-TRl-0 PDMS Operabilityo 2-Sl-0 Excore QPTR & Axial Flux Differenceo 2-PET-301 - Core Power Distribution Factorso 2-Sl-92 lncore-Excore Cross Calibration Data. 2-T17.020 - PDMS Calibration. 2-PET-304 - OperationalAlignment of NIS. 2-Tl Calorimetric Calibrationo 2-5!-0 Hot Channelfactors Determinationo 2-Sl-92 NIS Monthly Recalibration data. 2-Sl-0 lncore QPTRDetails of the performance of each PAT procedure is contained in theindividual summaries of the associated procedures as they are fullycompleted.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.0 ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAT-6.0.L23 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-PAr-5.2)This test was performed as part of 2-PAT-6.0, Test Sequence for 5Oo/o Plateau,and initiated in Mode 1. The test began pre-requisites on 7nn6 and was fieldwork completed on 7l'14116.1.0Test ObiectivesThe objectives of this test were to:1.1 Demonstrate Unit 2 response to a turbine generator trip with acoincident loss of offsite power (LOOP) is in accordance withdesign.1.2 Demonstrate that all four emergency diesel generators (EDG)automatically start, the Unit 2 EDGs connect to their respectiveshutdown board and provide power to the controls, indications, andequipment necessary to maintain Unit 2 in Hot Standby (Mode3)conditions for a minimum of 30 minutes.1.3 Demonstrate that operators can control plant parameters usingequipment available during a loss of offsite power.1.4 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 33, TurbineGenerator Trip With Coincident Loss Of Offsite Power TestSummary.1.5 Provide the steps necessary to protect Unit 1 operations.Test Methodslnitial conditions for Unit 2 include reactor power at approximately 30% ofrated thermal power, the main generator synchronized to the TVA grid,and electrical load greater than or equal to 120 MWe. All four dieselgenerators were in their normal standby condition.Unit 1 was in Mode 1 with alignment of the Unit 1 Reactor Coolant PumpBoards, Unit 1 Unit Boards, Common Boards and Shutdown Boards 1A-Aand 1B-B energized from normal power sources, the USST's associatedwith Mode 1 operation.The C-S CCS pump was aligned and in service to supply header 1Bl2B inaccordance with 0-SOl-70.01, "Component Cooling Water System". Theautomatic transfer of the 2A and 2C RCP boards to the A RCP Start buswas blocked and the automatic transfer of the 28 and 2D RCP boards tothe B RCP Start bus was blocked. The automatic transfer of the 2A-AShutdown board to the 2A-A Diesel Generator and the automatic transferof the 2B-B Shutdown board to the 2B-B Diesel Generator were notblocked. The automatic transfer of the 2A-A Shutdown board to the DCSST and the automatic transfer of the 2B-B Shutdown board to the CCSST was blocked. The maintenance supplies to the Unit 2 Shutdown2.0L24 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-PAT-5.2) (conti nued )Boards were verified in the "racked down/removed" position. The 6.9 KVB common board was in its normal alignment. The B common Board wasnot de-energized during the test to protect auxiliaries on both Units.The Unit 2 Main Turbine was manually tripped. Following the turbine trip,Operations concurrently and immediately performed the following:. Opened the normal power supply breaker to the 2A-A ShutdownBoard. The board did not transfer to its alternate power source,resulting in a dead board condition. Allfour emergency dieselgenerators (EDGs) started as expected. The 2A-A EDG connectedto and energized the 2A-A Shutdown Board.. Opened the normal power supply breaker to the 2B-B ShutdownBoard. The board did not transfer to its alternate power source,resulting in a dead board condition. The 2B-B EDG connected toand energized the 2B-B Shutdown Board as expected.o Operations ensured the U2 Main turbine and U2 Main Generatortripped.Following the Unit 2 generator trip, all four Unit 2 RCP Boards did nottransfer to their alternate power source and remained de-energized. Anautomatic reactor trip of Unit 2 occurred when voltage was lost to the Unit2 Reactor Coolant Pumps (RCPs). Operations then entered 2-E-0,Reactor Trip or Safety lnjection.With af l four Unit 2 reactor coolant pumps de-energized, the RCSdeveloped natural circulation conditions. Natural circulation parameterstook longer than Unit 1 to establish due to the very Iow decay heatgenerated in the new core. The Unit 2 Main Steam lsolation Valves(MSIV's) were manually closed to support testing with a simulated Loss ofOffsite Power configuration. Unit 2's Main Steam line pressure, steamgenerator pressure, and RCS temperature were maintained by the SGPORVs discharge to atmosphere. Auxiliary feedwater automaticallystarted, and steam generator level trended to post trip setpoint conditions.The test ran at least 30 minutes after the 2A-A and 2B-B 6.gkvShutdown Boards were energized from their respective emergency dieselgenerators without restoring offsite power. Unit 2 was restored to aplanned outage upon test completion at the direction of the shift Manager.125 7.2.1 Turbine Generator Trip with Goincident Loss of Offsite Power Test(2-PAT-5.2) (continued)3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 The 2A-A Diesel Generator automatically starts and connects to2A-A Shutdown Board following the Loss of Offsite Power transient.2A-A Diesel Generator automatically started and connected to itsrespective shutdown board.3.2 The 2B-B Diesel Generator automatically starts and connects to2B-B Shutdown Board following the Loss of Offsite Power transient.2B-B Diesel Generator automatically started and connected to itsrespective shutdown board.3.3 The Unit 2 Pressurizer Safety Valves do not open during thetest.Pressurizer Safeties did not open during the test.3.4 The Unit 2 Steam Generator Safety Valves do not open during thetest.Steam Generator Safety Valves did not open during the test.3.5 A Unit 2 Safety lnjection is not initiated during the test.No safety injection was initiated during testing.3.6 Hot standby (Mode 3) conditions on Unit 2 are maintained for atleast 30 minutes after the 2A-A and 2B-B 6.9kV Shutdown Boardsare energized from respective emergency diesel generators withoutrestoring offsite power.Hot Standby (Mode 3) conditions were maintained for at least 30minutes after 2A-A and 2B-B were energized lor their respectiveemergency diesel generators without restoring offsite power.L26 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-PAT-5.2) (conti nued )3.7 The 1A-A Diesel Generator automatically starts but does notconnect to the 1A-A Shutdown Board following the Loss of OffsitePower transient.1A-A Diesel Generator automatically started but did not connect tothe 1A-A Shutdown Board.3.8 The 1B-B Diesel Generator automatically starts but does notconnect to the 1BB Shutdown Board following the Loss of OffsitePower transient.1B-B Diesel Generator automatically started but did not connect tothe 1B-B Shutdown Board.Review Criteria3.9 The following Unit 2 parameters were maintained within theirrespective limits for at least 30 minutes immediately after de-energizing 2A-A and 2B-B Shutdown Boards, using equipmentavailable with offsite power removed from Unit 2:3.9.1 RCS Cold Leg Temperature (il7"F to 560oF and changingat a rate less than 50"F in one hour)This criteria was not met. RCS Cold Leg Temperaturesreduced below the minimum temperature of 547 degreesduring the 30 minute period. The rate of change was lessthan 50 degrees in one hour. CR 1192287 documented thisissue and was due to Turbine Driven AFW cooling sincereactor decay heat was minimal and no RCPs in service.3.9.2 Pressurizer Level (17Yo to 50%)Pressurizer level maintained between 25o/o and 34% duringthe test period.3.9.3 Pressurizer Pressure (2000 psig to 2335 psig)Pressurizer Pressure maintained between 2119 and 2244.8psig during the test period.3.9.4 Steam Generator Levels (17o/o to 60% narrow range andeither constant or trending toward 38% of narrow range)Steam Generator narrow range level maintained between30% and 39% during the test period.L21 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-P AT -5.2) (conti n ued )4.0Problemst1] CR 1192287 documented the Review Criteria was not met whenthe RCS cold leg temperature decreased below the minimumcriteria of 547oF. This deficiency is attributed to the TurbineDrive Auxiliary Feedwater Pump steam supply source coolingthe loop as it supplied AFW to the steam generators. Withminimal reactor decay heat and no RCPs running, the looptemperature was not maintained above the minimum criteria.121CR 1192023 was written to address the observation that the2A-A Diesel Generator appeared to be slower than expected intying on to its shutdown board. This was neither a Review orAcceptance Criteria for this test. Subsequent review by plantstaff did indicate the 2A-A Diesel Generator did not tie onto theboard within the Technical Specification limit and was declaredinoperable. The diesel generator was repaired by plantmaintenance and had no impact on meeting the PAT criteria asdelineated in UFSAR Chapter 14, Table 14.2.2, Sheet 33.L2B 7.2.2 Automatic Reactor Control System (2-PAT-6.1)This test was performed at the 50% test plateau as directed by2-PAT-6.0, Test Sequence for 50o/o Plateau. Testing was started on 6/18/16and field work completed on 7112116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Demonstrate the ability of the Automatic Rod Control System tomaintain the average RCS temperature (T",r) within acceptableIimits during both steady-state and transient conditions.1.2 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 31,Automatic Reactor Control System Test Summary.2.0 Test MethodsWith the Reactor Gontrol System (i.e. Rod Control) in manualand the Reactor Coolant System (RCS) at steady state conditions, RodControlwas placed in automatic to demonstrate that steady stateconditions could be maintained.Subsequently, with Rod Control in manual, T"w was varied from theReference Temperature (T,"fl by approximately +6 oF(+SoF to +7oF), bymanually changing the position of Control Bank D with no deliberateturbine load change. Rod Controlwas then placed in automatic todemonstrate the ability to restore and stabilizeTr- to within a t1.SoF deadband from T,"1via proper positioning of Control Bank D. The same testwas also performed for a T"* change of approximately -6oF(-SoF to -7oF) relative to Trer.The test was performed with reactor power approximately 45o/oto 47% ofRated Thermal Power (RTP) and RCS average temperature, pressurizerlevel and steam generator levels on program. The initialTaw - T,"smismatch was within tloF and RCS pressure was between 2200to 2250 psig.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.No control system settings were changed based on the performance ofthis test.Figures 7.2.2-1 through 7.2.2-g depict the performance results of theautomatic control systems.L29 7.2.2 Automatic Reactor Gontrol System (2-PAT-6.1) (continued)Acceotance Criteria3.1 No manual operator action or intervention is required to returnthe plant to stable conditions (i.e., auctioneered RCS Taw within11.soF of T,"r) for both steady-state and transient conditions.No manual operator action or intervention was required.3.2 For steady-state operation, and for both increasing anddecreasing T"*temperature transients, the Automatic RodControl System responds properly to automatically positioncontrol rods and return auctioneered RCS T"r, to within +1.soFof Tr*when the ARCS is placed in AUTO control mode.Rod Control properly responded to steady state and transientconditions to return T"r, to within 11.5"F of T'er.Review Criteria3.3 Pressurizer pressure tracks the response of auctioneered T"r,during the T"o transient tests and is controlled back toapproximately 2235 psig due to automatic pressurizer pressurecontrol.Pressurizer pressure tracked the response to Tavg and controlledback to approximately 2235 psig.3.4 Pressurizer level and level setpoint track the response ofauctioneere6 Tave during the T"o transient tests due to automaticpressurizer level control.Pressurizer level and level setpoint tracked the response to T"*.130 7.2.2 Automatic Reactor Gontrol System (2-PAT-6.1) (continued)4.0 Problemst1l CR 1190719 was written for two procedure deficiencies on2-PAT-6.1, Automatic Reactor Control System.Steps 6.2[8] and 6.3[12] said to ENSURE the passive summerindicated 72 steps/min. The passive summer does not indicate rodspeed. The passive summer indicates an enor signal in DegreesF. Steps 6.2[8] and 6.3[12]should have ENSURED the passivesummer indicated +5oF and - soF, respectively. The error wasidentified, discussed by PAT and Operations, CTL entry entered,and the test was continued. The +/- 5 degrees was verified duringperformance. Also Steps 6.2[10] and 6.3[14] verified that the rodspeed was72 steps/min at the time the rods were placed to auto."Step 6.3[3]1." was a typo and should have been deleted.131

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+{{#Wiki_filter:Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000November 10, 201610 cFR 50.4ATTN: Document Contro! DeskU.S. Nuclear Regulatory CommissionWashington, D.C. 20555-0001Watts Bar Nuclear Plant, Unit 2Facility Operating License No. NPF-96NRC Docket No.50-391Subject Watts Bar Nuclear Plant, Unit 2, Transmittal of lnitial Startup Report tothe United States Nuclear Regulatory Gommissionln accordance with the requirements of the Watts Bar Nuclear (WBN) Plant, Dual UnitUpdated Final Safety Analysis Report (UFSAR), Chapter 14.2.6, 'Test Records," theTennessee Valley Authority (TVA) is submitting the lnitial Startup Report for WBN Unit 2.The enclosure to this letter provides the Final Startup Report for WBN Unit 2. lnitial fuelload, pre-critical testing, initial criticality, and low power physics testing, and powerascension testing are discussed in separate sections of the report. The report details thetest objectives, methodology, test results, and problems noted for each of the testsperformed.The test objectives and methodology were developed using the graded approach based oncriteria provided in Regulatory Guide (RG) 1.68, "lnitial Test Programs for Water-CooledNuclear Power Plants,' Revision 2. RG 1.68 was further utilized for the selection of plantstructures, systems, and components (SSCs), and design features to be included in the testprogram. During the power ascension testing program, power ascension tests, surveillanceinstructions, and other permanent plant tests and technical instructions were performed todemonstrate satisfactory operation of SSCs.The report addresses test activities and the results of tests performed during the periodNovember 2015 through October 2016. The following table provides a summary of the keyWBN Unit 2 milestones and associated dates.
+U.S. Nuclear Regulatory CommissionPage 2November 10, 2016The WBN Plant Operations Review Committee has reviewed the report.There are no new regulatory commitments made in this letter. Please address anyquestions regarding this submittal to Gordon P. Arent d (d'23) 365-2004.Respectfully,@rM;Paul R. SimmonsSite Vice President, Watts Bar Nuclear Plant
+==Enclosure:==
+lnitial Startup Report to the Nuclear Regulatory Commission,Facility Operating License No. NPF-96, NRC Docket No. 50-391,Final Report November 2015 through October 2016cc (Enclosure):NRC RegionalAdministrator - Region llNRC Senior Resident lnspector - Watts Bar Nuclear PlantNRC Project Manager - Watts Bar Nuclear PlantWatts Bar Nuclear Plant, Unat2 - Milestone ActivitiesMilestoneDateWBN Unit 2 Facility Operating License, NPF-96October 22,2015lnitial Fuel Load CommencementDecember 4, 2015lnitial CriticalityMay 23,2016Test Plateau, 30o/o Reactor Thermal Power (RTP)June 16, 2016Test Plateau, 5Ao/o RTPJuly 16, 2016Test Plateau, 75o/o RTPJuly 29,2016Test Plateau, 90o/o RTPAugust 29,2016Test Plateau, lOOo/o RTPOctober 6 ,2016Commercial OperationOctober 19, 2016 WATTS BAR NUCLEAR PLAhITUNIT 2INTTIAL STARTUP REPORTTO THEUNITED STATESNUCLEAR REGULATORY COMMISSIONAPPROVAL SHEETPower Ascension Test Manager:TRG Meeting No./06TRG ChairmaniPlant Manager:! lo"a/:
+TENNESSEE VALLEY AUTHORITYWATTS BARNUCLEARPLANTUNIT 2INITIAL STARTUP REPORTTO THEUNITED STATESNUCLEAR REGULATORY COMMISSIONFACILITY OPERATING LICENSE NO. NPF'.96NRC Docket No. 50-391Final ReportNovember 2015 through October 2016 TABLE OF CONTENTSSECTION1.02.03.04.05.0PAGE NUMBERLIST OF FIGURES ............... ivLIST OF ACRONYMS.............. ............ viiTNTRODUCTTON ...................1POWER ASCENSTON TEST PROGRAM (PATP) OVERVIEW ................32.1 Administration of the Program ............. .........32.2 lmplementation of the Pro9ram.............. .......62.3 Summary ....................8WATTS BAR UNIT 2 STARTUP CHRONOLOGY... ..............10INITIAL FUEL LOAD ..........254.1 Overview and Summary of lnitial Core Loading ...........254.2 lnitial Core Loading Sequence (2-PAT-2.0)........ .........264.3 Reactor System Sampling for Core Load (2-PAT-2.1)......... .........284.4 Response Check of Core Load Instrumentation After 8 Hour Delayin Fuel Movement (2-PAT-2.2)............... .....314.5 Pre-Power Escalation NIS Calibration Data (2-PET-102)..............................334.6 lnitial Core Loading (2-PET-105) ................35PRECRITTCAL TEST!NG.......... ...........415.1 Post Core Loading Precritical Test Sequence(2-PAT-3.0)............... .................415.2 Control Rod Drive Mechanism Timing and CERPI lnitial Calibration(2-PAT-3.1)............... .................4e5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-PAT-3.2)............... .................565.4 Rod Control and Rod Position lndication (CERPI) (2-PAT-3.4).....................615.5 Reactor Coolant Flow Coastdown (2-PAT-3.7)............... ..............695.6 Rod Drop Time Measurement and Stationary Gripper Release Timing(2-PAT-3.8)............... .................71al-TABLE OF CONTENTS (continued)SECTION5.0 PRECRITICALTESTING(continued)6.07.0PAGE NUMBER5.7 Reactor Trip System (2-PAT-3.10) .............755.8 Adjustment of Steam Flow Transmitters at Minimal Flow(2-PAT-3.11)............. .................7e5.9 Control Rod Drive Mechanism Timing (2-PET-106)............. .........81INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING ...........846.1 lnitial Criticality and Low Power Test Sequence (2-PAT-4.0).........................846.2 Reactivity Computer (ADRC) (2-PET-103) ............ ......866.3 lnitial Criticality and Low Power Physics Testing (2-PET-201).......................88POWER ASCENSTON TESTING ............. .............957.1 Test Sequence for 30% Plateau (2-PAT-5.0)......... ......957.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1).....................997.1.2 Automatic Steam Generator Level Control Transients atLow Power (2-PAT-5.3)............ ...1087.1.3 Calibration of Steam and Feedwater Flow lnstruments at30% Power (2-PAT-5.4) ............ ...................1167.2 Test Sequence for 50% Plateau (2-PAT-6.0)......... ....1207.2.1 Turbine Generator Trip with Coincident Loss of Offsite PowerTest (2-PAT-5.2)....... ...................1247.2.2 Automatic Reactor Control System (2-PAT-6.1)............... ..............1297.2.3 Automatic Steam Generator Level Control Transientsat 50% Power (2-PAT-6.2) ............... ............1417.2.4 Calibration of Steam and Feedwater Flow lnstrumentsat 50% Power (2-PAT-6.3) ............... ............1467.3 Test Sequence for 75o/o Plateau (2-PAT-7.0)......... ....1497.3.1 Calibration of Steam and Feedwater Flow lnstrumentsat75o/o Power (2-PAT-7.1) ............... ............152aal- l-TABLE OF CONTENTS (continued)SECTION7.0 POWER ASCENSION TESTING (continued)PAGE NUMBER7.4 Test Sequence for 100% Plateau (2-PAT-8.0)............... .............1557.4.1 Load Swing Test (2-PAT-1 .2)............ ...........1597.4.2 Large Load Reduction Test (2-PAT-1.3)............. ...........1627.4.3 Pipe Vibration Monitoring (2-PAT-1.4)......... ..................1777.4.4 LoosePartsMonitoringSystem(2-PAT-1.5)............ .....1827.4.5 Startup Adjustments of Reactor Control System (2-PAT-1.6).........1857.4.6 Operational Alignment of Process Temperaturelnstrumentation (2-PAT-1.7)......... ................1897.4.7 Thermal Expansion of Piping Systems (2-PAT-1.8).......................1967.4.8 Automatic Steam Generator Level Control (2-PAT-1.9) ............ .....2017.4.9 lntegrated Computer System (lCS) (2-PAT-1.10)...........................2057.4.10 RVLlSPerformanceTest(2-PAT-1.11).......... ...............2087.4.11 Common Q Post Accident Monitoring System (2-PAT-1.12)..........2127.4.12 RCS Flow Measurement (2-PAT-3.3)......... ...................2157.4.13 Calibration of Steam and Feedwater Flow lnstrumentsat 100 % Power (2-PAT-8.4) ............... .........2177.4.14 Shutdown From Outside the Main Control Room (2-PAT-8.5) .......2217.4.15 Plant Trip Evaluation for Equivalency of Test Performance of2-PAT-8.6, Plant Trip From 100o/o Power (Turbine Trip).................2267.4.16 Core Power Distribution Factors (2-PET-301)............. ...................2487.4.17 Operational Alignment of NIS (2-PET-304) ............ .......2537.4.18 Radiation Baseline Surveys (RCI-159) .........255aaal-rl-FIGURE2.0-14.6-14.6-25.3-16.3-17.1 .2-17.2.2-17.2.2-27.2.2-37.2.2-47.2.2-57.2.2-67.2.2-77.2.2-87.2.2-9LIST OF FIGURESPAGE NUMBERWBN PowerAscension Test Program Schedule Overview ..................9U2Cl Core Load Sequence ............39Unit 2 Core Load ICCR ...................40Pressurizer Spray Response ...........60ICRR vs Primary Water (N31, N32) ........ .........94SG #2 FW Flow Oscillation ...........115RCS Auctioneered T"* vs T6sSteady-State Operatioh.............. ...132RCS Auctioneered T"* vs T,"glncreasing T"* Transient .......... ....133RCS Auctioneered Trrs vs TpgDecreasing T"r, Transient.......... ...134Rod Speed and Direction Demandlncreasing T"rs Transient .......... ....135Rod Speed and Direction DemandDecreasing T"r, Transient.......... ...136Pressurizer Pressurelncreasing T"r, Transient .......... ....137Pressurizer PressureDecreasing T"r, Transient.......... ...138Pressurizer Level and Level Setpointlncreasing T"rs Transient .......... ....139Pressurizer Leve! and Level SetpointDecreasing T"", Transient.......... ...140al-v LIST OF FIGURES (continued)FIGURE PAGE NUMBER7.4.2-1 Generator Output vs Time(Large Load Reduction)........ .........1057.4.2-2 NIS Power vs Time(Large Load Reduction)........ .........1G07.4.2-3 Pressurizer Pressure vs Time(Large Load Reduction)........ .........1G77.4.24 Pressurizer Level vs Time(Large Load Reduction)........ .........1687.4.2-5 Control Bank D Position vs Time(Large Load Reduction)........ .........1097.4.2-O Tay6 vs Time(Large Load Reduction)........ .........1707.4.2-T Steam Dump Signal vs Time(Large Load Reduction)........ .........1717.4.2-8 Feedwater Pressure vs Time(Large Load Reduction)........ .........1721727.4.2-9 Steam Generator 1 Level vs Time(Large Load Reduction)........ .........1737.4.2-10 Steam Generator 2 Level vs Time(Large Load Reduction)........ .........1747.4.2-11 Steam Generator 3 Level vs Time(Large Load Reduction)........ .........1757.4.2-12 Steam Generator 4 Level vs Time(Large Load Reduction)........ .........1707.4.14-'l Main Control Room Abandonment30 Minute Stability.... ....2247.4.14-2 Main Control Room AbandonmentCooldown ...2257.4.15-1 Representative RCCA Position vs Time(100% Trip) ........ ..........2327.4.15-2 Main Turbine Speed vs Time(100Yo Trip)........ ..........233 FIGURE7 .4.15-37.4.15-47 .4.1 5-57 .4.15-67.4.15-77 .4.15-g7 .4.1 5-g7 .4.1 5-1 07.4.15-117.4.15-127 .4.1 5-1 37.4.15-147 .4.15-157 .4.15-16LIST OF FIGURES (continued)PAGE NUMBERSafety lnjection Alarm vs Time(100Yo Trip)........ ..........234Steam Generator Safety Valve Position Loops 1 &2 vs Time(100% Trip)........ ..........29sSteam Generator Safety Valve Position Loops 3 & 4 vs Time(100Yo Trip)........ ..........236Pressurizer Safety Valve Outlet Temperature vs Time(100% Trip)........ ..........237NIS Power vs Time(100% Trip)........ ..........238Hot Leg RTD Response vs Time(100o/o Trip)........ ..........299Steam Generator Levels vs Time(100% Trip)........ ..........240Pressurizer Pressure vs Time (1)( 1 00% Trip).. .. . . . ... ...... t .. .. .. ...... .. . .241Pressurizer Pressure vs Time (2)(100% Trip)......... .......242Pressurizer Level vs Time(100o/o Trip)........ .......243Pressurizer Level Controller Output vs Time(100Yo Trip)......... .......244RQQ Aryrage Temperature (Auctioneered) vs Time(100yo Trip)......... .......2459lCqT Pqqp Demand and T"o vs Time(100o/o Trip)......... .......246Main Feedwater Pump Flows and T"* vs Time(100Yo Trip)......... .......247avl-LIST OF ACRONYMSADRC Advanced Digital Reactivity ComputerAFW Auxiliary FeedwaterAOI Abnormal Operating lnstructionARO All Rods OutASME American Society of Mechanical EngineersAUX AuxiliaryBOL Beginning of LifeBYA Bypass A (Reactor Trip Breaker A)BYB Bypass B (Reactor Trip Breaker B)CBA Control Bank ACBB Control Bank BCBC Control Bank CCBD Control Bank DCERPI Computer Enhanced Rod Position lndicationCLA Cold Leg AccumulatorCOTS Channel Operational TestsCPS Counts Per SecondCR Condition ReportCRDM Control Rod Drive MechanismCTL ChronologicalTest LogCV Concurrent VerificationCVCS Chemical and Volume Control SystemDAS Data Acquisition SystemDCN Design Change NoticeDCS Distributed Contro! SystemDRWM Digital Rod Worth MeasurementeNuPOP Electronic Nuclear Parameters and Operations PackageFATF FuelAssembly Transfer FormFCV Flow ControlValveFHI Fuel Handling lnstructionF! Flow lndicatorFIC Flow lndicating ControllerFW FeedwaterGO General Operating InstructionHFP Hot Full PowerHS Hand SwitchHVAC Heating Ventilation and Air ConditioningHZP Hot Zero PowerICRR lnverse Count Rate RatioICS lntegrated Computer Systemaavt1 LIST OF ACRONYMS (continued)INPO lnstitute of Nuclear Power OperationsIR lntermediate RangeITC lsothermalTemperature CoefficientlV lndependent VerificationKBH Thousand Pounds Per HourLC Level ControllerLCP Loop Calculation ProcessorLCV Level ControlValveLPMS Loose Parts Monitoring SystemLPPT Low Power Physics TestLVDT Linear Variable Differential TransformerM&TE Measuring and Test EquipmentMCD Maximum Channel DeviationMCR Main Control RoomMED Maximum Expected DeviationMFP Main Feedwater PumpMFPT Main Feedwater Pump TurbineMFW Main FeedwaterMMI Man Machine lnterfaceMPPH Million Pounds Per HourMSIV Main Steam lsolation ValveMSV Main Steam ValveMTC Moderator Temperature CoefficientNl Nuclear lnstrumentationNIS Nuclear lnstrumentation SystemNOB Nuclear Operating BookNOTP Normal Operating Temperature & PressureNPG Nuclear Power GroupNRC Nuclear Regulatory CommissionNSSS Nuclear Steam Supply SystemNuPOP Nuclear Parameters and Operations PackageOPC Overspeed Protection ControllerOPDP NPG Standard Department Procedure (Operations)OPSP Over Power SetpointOTDT Overtemperature Delta TemperatureOTSP Over Temperature SetpointPAT Power Ascension TestPATP Power Ascension Test ProgramPDMS Power Distribution Monitoring SystemPET Power Escalation TestPIC Pressure lndicating Controlleraaavl- l- l-PIDPLSPMTPORCPRpsipsiapsidpsigPTIPZRQPTRRBSSRCCARCIRCPRCSRDTCRegRHRRPIRSARTRTARTBRTDRTPRVLISRWPRWSTSARSBASBBSBCSBDSESEQSFPSGSGBDSILIST OF ACRONYMS (continued)Point IdentificationPrecautions, Limitations and SetpointsPost Maintenance TestPlant Operations Review CommitteePower RangePounds per square inchPounds per square inch absolutePounds per square inch differentialPounds per square inch gaugePreoperational Test I nstructionPressurizerQuadrant Power Tilt RatioRod Bank Select SwitchRod Cluster Control AssemblyRadiological Control I nstructionReactor Coolant PumpReactor Coolant SystemRod Drop Test ComputerRegulatingResidual Heat RemovalRod Position lndicatorRedundant Sensor AlgorithmReactor TripReactor Trip Breaker AReactor Trip Breaker BResistance Tem perature DetectorRated Thermal PowerReactor Vessel Level lnstrumentation SystemRadiological Work PermitRefueling Water Storage TankSafety Analysis ReportShutdown Bank AShutdown Bank BShutdown Bank CShutdown Bank DSite EngineeringSequenceSpent Fuel PoolSteam GeneratorSteam Generator BlowdownSurveillance I nstructionaIX LIST OF ACRONYMS (continued)SOI System Operating lnstructionSR Source RangeSRF Statistical Reliability FactorSSP Site Standard PracticeSSPS Solid State Protection SystemSWIF Seal Water lnjection FilterTE Temperature ElementTl Technical lnstructionTR Temperature RecorderTRG Test Review GroupTRI Technical Requirements lnstructionTTD Time Trip DelayTVA Tennessee Valley AuthorityUC Urgent ChangeUFSAR Updated Final Safety Analysis ReportUSNRC United States Nuclear Regulatory CommissionVCT Volume Control TankWBN Watts Bar NuclearWO Work Order
+==1.0 INTRODUCTION==
+The lnitial Startup Report for the Watts Bar Unit 2 nuclear plant discusses theresults of testing performed from initial core load through full power operation.This report address each of the power ascension tests identified in Chapter 14 ofthe WBN Unit 2 UFSAR and other license commitments. The report includes adescription of the measured values of the operating conditions or characteristicsobtained during the testing program and a comparison of these values withdesign predictions and specifications. Any corrective actions that were requiredto obtain satisfactory operation are also described.WBN Unit 2 UFSAR Chapter 14.2.6, Test Records, requires the Startup Reportbe submitted within:(1) 90 days following completion of the Startup Test Program,(2) 90 days following resumption or commencement of commercial poweroperation, or(3) 9 months following initial criticality, whichever is earliest.lf the Startup Report does not cover all three events (i.e., initial criticality,completion of Startup Test Program, and resumption or commencement ofcommercial power operation), supplementary reports shall be submitted at leastevery 3 months until allthree events have been completed.Item (1) is being satisfied since the Power Ascension Test Program wascompleted on October 6, 2016.WBN Unit 2 Facility Operating License No. NPF-96 was issued on October 22,2015. lnitial Fuel load commenced with movement of the first fuel assembly at20:49 on December 4, 2015. Core loading was completed at 02:10 on December8, 2015. lnitial criticality was achieved at 02:16 on May 23,2016. Further testingwas successfully completed at the following plateaus:Test Plateou, % RTPDate Completed30June 16, 201650July 16, 201675July 29,201690August 29,2016100October 6 ,2016lnitial Fuel load, precritical testing, initial criticality and lowand power ascension testing are discussed in separateThe report details the test objectives, methodology, testnoted for each of the tests performed.power physics testing,sections of the report.results, and problems
+==1.0INTRODUCTION==
+ (continued)Acceptance Criteria is defined as safety related performance parameters definedin the Design Output, vendor documents, TVA or vendor drawings, NRCcommitments, other licensing and design documents, and so forth, that must beexhibited during the performance of a PAT or PET. Failure to meet anacceptance criterion is considered to be a safety related issue.A 10CFR50.59 Evaluation per NPG-SPP-09.4, 10 CFR 50.59 Evaluationsof Changes, Tests, and Experiments or a Technical Evaluation perNPG-SPP-09.3, Plant Modifications and Engineering Change Control, may berequired and testing may be stopped. The subsequent course of action will bedetermined by the nature of the discrepancy and applicable TechnicalSpecifications. Failure to meet Acceptance Criteria will be documented in aCondition Report (CR).Review Criteria encompasses other performance parameters defined in theUFSAR, design criteria, vendor documents, drawings (TVA or vendor), otherlicensing, design, setpoint and operational documents that are expected to beexhibited during performance of a PAT or PET. These criteria should be viewedas a guide to possible measurement or design errors. Failure to meet thesecriteria do not by themselves constitute problems. While prudent measuresshould be taken to resolve any conflict between measurements and predictions,failure of these criteria do not require 10CFR50.59 Evaluations, per NPG-SPP-09.4, and do not require testing or power ascension to be stopped for resolution.Failure to meet these review criteria will be documented in a CR.
+.0 POWER ASCENSTON TEST PROGRAM (pATp) OVERVTEWThe PATP was developed from testing described in Chapter 14 of the WBNUnit 2 UFSAR and requirements specified in Regulatory Guide 1.68, Revision 2,"lnitial Test Programs for Water-Cooled Nuclear Power Plants". Testing of theNSSS followed Westinghouse test methodology.2.1 Administration of the ProgramThe Site Vice President had the overall responsibility for the PATP.Overall management of the PATP was directed by the Plant Manager who wasresponsible for:o Development and implementation of the PATP to ensure the PATP wasconducted in a safe and efficient manner while complying with licenseprovisions and other commitments.. Establishing the Power Ascension Testing Organization.o Advising senior management on PATP activities.o Establishing a Technical Review Group (TRG) as a subcommittee of thePORC to review PATP activities.o Providing final approval of Power Ascension Tests (PATs) and selectedother procedures.o Ensuring the PATP was conducted in accordance with applicable WBNAdministrative Procedures.o Providing approvalto proceed to the next PATP test plateau.o Providing final approval of all each test package and the Startup Report.The Power Ascension Test Manager was responsible for:o Notifoing the plant manager of major problems and of the completion ofeach major test phase (i.e., test sequence) of the program.. Ensuring the PATs and the PETs were available for NRC review aminimum of 60 days before the scheduled fuel load date.o Ensuring the technicaljustification and schedule, including power level forcompletion of delaying preoperational tests, were provided to the NRCstaff prior to fuel Ioad.
+.1 Administration of the Program (continued)o Ensuring the requirements of TVA-SPP-30.010, lnitial Synchronization ofTVA Generating Assets to TVA's Transmission System, were met.. Developing and implementing plans and schedules for the PATP.. Ensuring testing activities, including planning and scheduling, resulted insafe plant operations and that were not dependent on the performance ofuntested systems.. Coordinating and directing overall PATP testing and related activities andrequirements with appropriate support groups.. Supervising test personnel assigned to the power ascension testing group.o Assigning responsibilities to organizations for specific testingrequirements.o Participating in the review activities of the TRG, and acting as Chairman ofthe TRG.o Ensuring the test procedures were reviewed by the TRG.o Ensuring the Startup Report was reviewed by the TRG.o Ensuring additional startup Reports were prepared, reviewed, approvedand transmitted to the NRC as needed.o Ensuring the post-performance test results (i.e., test packages) werereviewed by TRG.o Ensuring test directors for the PATP were qualified, and met the minimumqualifications of ltem 1 and either ltem 2 or ltem 3 below and ensuringother required individuals (e.9. lndependent Verifiers (lV) and ConcurrentVerifiers (CV) were qualified to perform the tasks assigned:1. Knowledgeable of the test program administration, the systemdesign and operational requirements, and expected plantoperational characteristics during the test, andTrained as test coordinators in accordance with NPG-SPP-06.9.1,Conduct of Testing.
+.1Administration of the Program (continued)2. Possessed a bachelor degree in engineering or physical science,andHad two years experience in power plant testing or operation.Included in the two years was one year nuclear power plant testing,operating or training on a nuclear facility.3. Possessed a high school diploma or equivalent, andHad five years experience in power plant testing. lncluded in thefive years were two years of nuclear power plant experience.Credit for up to two years of related technical experience could besubstituted for experience on a one-for-one basis.Technical and administrative oversight of the PATP was performed by TRGwhich was composed of one representative, or their alternates, from each of thefollowing organizations:o Plant Operationso Reactor Engineeringo Site Engineering. Corporate Nuclear Fuelso Power Ascension Testingo WestinghouseTRG was charged with reviewing PATP testing activities for technical adequacyand affecUimpact on nuclear safety, and advising PORC and the plant manageron the disposition of those items reviewed. The responsibilities of TRG includedfinal review and recommendation of approval of al! PATP test procedures,revisions, and test results.Following completion of testing at each major test sequence of the PATP, testresults were reviewed by TRG to ensure required tests had been performed.TRG also ensured Acceptance Criteria were satisfied; test deficiencies hadproper dispositions, appropriate retesting had been completed, and test resultshad been reviewed by appropriate designated personnel prior to proceeding tothe next major test sequence. This review ensured that all required systemswere operating properly and that testing for the next major test sequence couldbe conducted in a safe and efficient manner.
+.2Implementation of the ProgramThe WBN PATP utilized information gained from operating and testing experience atother nuclear plants. This information was used in the development of the PATPtest procedures and schedules and to alert personnel to potential problem areas.Test procedures were developed utilizing information obtained from OperatingExperience (OE) database. The TVA Operating Experience Program identifies andevaluates experience gained from other TVA nuclear plants, INPO, NRC, equipmentsuppliers, and from other utilities. Significant operational experience and eventswere reviewed and integrated into appropriate PATP test procedures to ensurenuclear safety and reliability. To the extent practical, simulator-based training andtrial use of the PATP test procedures were performed on the WBN simulator tofamiliarize personnel with systems and plant operation and to assure technicaladequacy of the procedures under simulated plant conditions prior to field useduring power operation.The testing program was conducted by qualified personnel using approved plantadministrative, test, and operating procedures. The plant was taken from core loadto full power in a highly controlled, conservative, and documented manner whichdemonstrated, where practical:The plant is ready to operate in a manner which will not endanger thehealth and safety of the public.The plant has been properly constructed, and plant performance issatisfactory in terms of established design criteria.The plant meets licensing requirements and provides assurance of plantreliability for operation.o The plant is capable of withstanding anticipated transients and postulatedaccidents.The PATP was specified in seven PAT sequence procedures:. 2-PAT-2.0, lnitial Core Loading Sequence. 2-PAT-3.0, Post Core Loading Precritical Test Sequenceo 2-PAT-4.A, lnitial Criticality and Low Power Test Sequence. 2-PAT-5.0, Test Sequence for 30% Plateauo 2-PAT-6.0, Test Sequence for 50% Plateauo 2-PAT-7.0, Test Sequence for 75% Plateauo 2-PAT-8.0, Test Sequence for 1A0o/o Plateau 2.2lmplementation of the Program (continued)Each PAT sequence procedure called out the performance of other PATs, aswell as other designated plant procedures such as PETs, Sls, TRls, Tls, RClsand FHls. The sequence procedures specified the logica! performance ofrequired tests and procedures through each test plateau. The sequenceprocedures also specified general prerequisites, precautions and limitations, andadditional operational steps at each test plateau. The detailed test and normalplant procedures called out by the sequence procedures defined step-by-stepactions, specific prerequisites and limitations, signoffs, data taking requirements,and test acceptance and review criteria.The PATP commenced with the receipt of the Facility Operating License onOctober 22, 2015, and progressed with core loading, precritical testing, initialcriticality and low power physics testing, and power ascension testing. Core loadprocedures directed the initial core load in a prescribed manner which ensuredcore loading was accomplished in a safe and orderly fashion. Precritica! testingbrought the plant to hot standby conditions, made measurements, anddemonstrated that the plant was ready for critical operation. lnitial criticality onMay 23,2016, brought the Unit 2 reactor critica! for the first time. Low powerphysics testing performed measurements on the critical reactor to demonstrateconformance with design predictions prior to power operation. PAT brought theplant to full power, made minor plant instrumentation adjustments, anddemonstrated the plant's ability to withstand selected transients. Figure 2.0-1depicts the time line for the PATP.Plant events not directly associated with the PATP added to the duration of theprogram. These events are included in the chronology.
+.3 SUMMARYWatts Bar Nuclear Plant's Unit 2 Power Ascension Test Program began with the receiptof the operating license. The program encompasses; those preoperational type teststhat were deferred to the PATP, the prerequisites required to load the initial core, theinitia! core loading itself, post core loading tests, initial criticality, low power tests, and at-power tests.Completion of these tests verified that the unit was properly designed, constructed, andready to operate in a manner that will not endanger the health and safety of the public,meets contractual and licensing requirements, and provides assurance of plantreliability for operation. The PATP used Regulatory Guide 1.68, Revision 2 and theWBN UFSAR for development of the test requirements.lssues identified during the testing were resolved except for the following open itemsthat will be resolved by the Corrective Action Program.OPEN ITEMS:(1) CR 1208694 initiated WO 118122821 to design and install bracing on multipleMain Steam Traps as a result of visual inspection during 2-PAT-1.4, PipeVibration Monitoring.(2) UFSAR Table 14.2 2, Sheet 5 Test Method, refers to an evaluation of thermalexpansion at final ambient conditions. This final ambient condition evaluation willbe performed later and is tracked by Commitment 118008175.(3) CR 1208178 was initiated for 2-PT-1-81 being unavailable during testing2-PAT-1.6, Startup Adjustments of Reactor Control System, and will be repairedby WO 118121693.(4') CR 1171424 was written during the performance of 2-PAT-1.5, Loose PartsMonitoring System, for three channels taken out of service due to issues.These channels will be repaired under the following work orders.o WO 117845593 - Channel 101 Experiencing excessive noise and isalarming due to "llTA'rattling.. WO 117843208 - Channel 102 Accelerometerfound damagedo WO 117843209 - Channel 110 Suspect preamplifier FIGURE 2.0.1WBN POWER ASCENSION TEST PROGRAMSCHEDULE OVERVIEWi:=;sE:*-EEsb0.t roH.TSdfiernilolqi0PFr-o. I .!ron ls.I*l-Esl. :r-IIEEi*l-IIIt-II-TIgElll3oBi'Eg*q"PA.+a\l-qtnIo.aa\l-qYFLaa\tqt\,aFCIa\9rn-tOrr...H g EB,J;do3oCLyrrBIA+iEool-orrtt/c.gaaEl!tt,UtrIECLo'rEfIJ.g*octtooo 3.0 WATTS BAR UNIT 2 STARTUP CHRONOLOGYNote: Power Ascension Tests may be performed at multiple testing plateaus.The description of the individual PAT is documented in the section(plateau) in which it was completed.10122115 -Receipt of WBN Unit 2 Facility Operating License No. NPF-96 fromthe NRC.11119115 PAT-2.0, lnitial Core Load Sequence, was begun.-2-PAT-2.1, Reactor System Sampling for Core Load, was initiated.11119115 -Commenced movement of new fuel into the Spent Fuel Pool inpreparation for loading Unit 2 core.11122115 -All 193 fuel assemblies required for Unit 2 fuel load moved into theSpent Fue! Poo!.11123115 -Completed Spent Fuel Pool verification for Unit 2 Cycle 1, fuelassembly and component insert verification with no discrepanciesnoted.1213115 -RCl-159, Radiation Baseline Surveys, pre-fuel load surveys werecompleted. No Acceptance or Review Criteria is associated with thisprocedure.1214115 PAT-2.1, Reactor System Sampling for Core Load, completed withall criteria met.- Unit 2 entered Mode 6 at 20:04.-lnitiated transport of first fuel assembly to Unit 2 vessel at20:49.1215115 -Fuel movement was delayed due to issues with refueling machine at06:23.1215115 PAT-2.2, Response Check Of Core Load lnstrumentation After 8Hour Delay ln Fuel Movement, completed at 11:43 with all criteriamet and fuel movement resumed.1215115 -Debris was reported on the bottom of the fourth fue! assembly.Debris was removed and fuel movement resumed after appropriateapprovals. CR 1112204 was initiated.1216115 -Fuel movement was suspended again due to debris on a fuelassembly. Debris was cleared and fuel movement resumed.1218115 -lnitialfuel load for Unit 2 completed at 02:10.-2-PET-105, lnitial Core Loading completed.-2-Tl-28, Verification Of Core Load Prior To Vessel Closure, wascompleted at 12:29 with all criteria met.10 3.0 Watts Bar Unat 2 Startup Chronology (continued)1219115 PAT-2.0, lnitial Core Load Sequence, was completed and TRGapproved.-2-PAT-3.0, Post Core Loading and Precritical Test Sequence,pre-requisites were initiated.12110115 -Unit entered Mode 5, maintaining <105"F in RCS in preparation ofPAT at the Ambient Plateau.12112115 -RCl-159, Radiation Baseline Surveys - Post Fuel Load Survey wasfield work complete for applicable Ambient Plateau sections. NoAcceptance or Review Criteria were associated with this procedure.12116115 PAT-1.8, Thermal Expansion of Piping Systems, was field workcomplete for applicable Ambient Plateau sections with all criteria met.12123115 PNf-1.4, Pipe Vibration Monitoring, Section 6.5.6, Condensate -Short Cycle, field work complete with all Acceptance Criteria met.There was no Review Criteria for this test.11161'16 - PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete for applicable Ambient Plateau sections. There was noAcceptance or Review Criteria for these sections.1120116 PAT-3.10, Reactor Trip System, field work complete with all criteriamet.1124/,16 PAT-3.1, Control Rod Drive Mechanism and CERPI lnitia!Calibration, field work complete with allAcceptance Criteria met afterevaluation. CR 1128950 was written for high current amplitudes andclosed following Westinghouse evaluation that determined themeasurements to be acceptable. There was no Review Criteria forthis PAT.-2-PAT-3.8, Rod Drop Time Measurement and Stationary GripperRelease Timing, Mode 5 Performance, field work complete forapplicable Ambient Plateau sections with allAcceptance Criteria met.There was no Review Criteria for this test.1126116 -PAT testing on the plant primary side was suspended on 1126116until plant conditions allowed further testing.211116 -WO 112989715 Complete for WINCISE Site Acceptance Test (WNA-TP-02985-WBT). This WO satisfied UFSAR Table 14.2-2, Sheet 12,lncore lnstrumentation System Test Summary, Acceptance Criteria 1.-2-PAT-1.4, Pipe Vibration Monitoring, field work complete forSection 6.5.7 Condensate - Long Cycle with all criteria met.11 3.0 Watts Bar Un.t2 Startup Chronology (continued)3119116 -Unit entered Mode 4, RCS temperature >200"F and <350'F to allowPAT at the 250"F Plateau.3121116 PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for 250'F Plateau applicable sections with all criteria met.3124116 -RCS temperature increased to 300"F to facilitate PAT.3125116 PAT-1.11, RVLIS Performance Test, field work complete forapplicable 300'F Plateau sections with all criteria met.3/30/16 -Unit entered Mode 3, RCS temperature ) 350'F to allow furtherPower Ascension Testing at the 360"F Plateau.3131116 PAf-1.11, RVLIS Performance Test, field work complete forapplicable 360'F Plateau sections with all criteria met.-2-PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for applicable 360'F Plateau section with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable 360'F Plateau sections with all criteriaacceptable for continued heat-up.411116 -lnitiated plant heat-up to 400"F for PAT at01:14.-2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable 400"F Plateau sections with Review Criteria not met.CR 1156425 was written.-2-PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for applicable 400"F Plateau section with all criteria met.-lncreased RCS temperature to 450"F for testing.-2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable 450"F Plateau, with all criteria met.-2-PAT-1.12, Common Q Post Accident Monitoring System, field workcomplete for applicable 450'F Plateau sections with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems - field workcomplete for applicable 450'F Plateau sections. CR's were initiatedwithin the test for seven snubbers not performing as expected.Results indicated no issue with snubbers and approved to continue tonext plateau testing. (See Problem Report #1 of 2-PAT-1.8)L2 3.0 Watts Bar Unat 2 Startup Chronology (continued)412116 -Unit coo! down initiated to repair a check valve with excessiveleakage. Additionally, two RCS RTD's were replaced.-Unit entered Mode 4, RCS temperature >200'F and <350'F, at06:29.418116 -Unit re-entered Mode 3, RCS temperature 2 350"F.4110116 -RCS temperature increased to 500"F.-2-PAT-1.1 1, RVLIS Performance Test, field work complete forapplicable 500'F Plateau sections with all criteria met.4112116 PET-102, Pre-Power Escalation NIS Calibration Data, completedwith all criteria met.4113116 -RCS temperature increase to 557"F.-2-PAT-1.11, RVLIS Performance Test - field work complete forapplicable section of 557"F data taking only with al! criteria met forsteady state data collection.-2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for Steady State Data Collection, section 6.7, with allcriteria met.-2-PAT-1.4, Pipe Vibration Monitoring - Section 6.5.2, MainFeedwater Pump 2A Start and Steady State Operation on Recirc.,was field work complete on 4113116 with velocity and displacementAcceptance Criteria not met. CR 1161783 was initiated for anengineering evaluation which concluded equipment was acceptableas is. There was no Review Criteria for this test.4114116 -RCS at normal operating pressure.-2-PAT-1.4, Pipe Vibration Monitoring, completed for Section 6.5.1,Pressurizer Surge, Mode 3, with all criteria met for that section.-2-PAf-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable 557'F Plateau sections with an issueoutside containment on Protective Device PD07-2. WO 1 17755755was to resolve the issue with PD07-2 and investigate any possibleissues with PD07-1.Additionally, other components did not move as expected and wereevaluated and concluded to be within their working range. (SeeProblem Reports #2,#3 of 2-PAT-1.8).4116/16 PAT-3.2, Pressurizer Spray Capability and Continuous SprayFlow Setting, Section 6.1, Adjustment of the Pressurizer ManualSpray Bypass Valves, was completed. AIIAcceptance Criteria wasmet. Review Criteria for MCR alarms was not met with CRs1161382 and 1160969 written. A Westinghouse evaluationdetermined the PAT met the operability and design requirements ofthe pressurizer spray system.13 3.0 Watts Bar Unat 2 Startup Chronology (continued)4117116 -Unit was placed in Mode 4 for repairs to the Turbine DrivenAuxiliary Feedwater Pump and replacement of PD07-2 shimdetermined to require adjustment.511116 -Unit 2 in Mode 3 at 17:36.512116 -Unit 2 at NOTP at 23:00.513116 PAT-3.3, RCS Flow Measurement, field work complete with allcriteria met.514116 PNf-1.6, Startup Adjustments of Reactor Control System, fieldwork complete for Mode 3. This performance was data taking only.515116 PAT-1.4, Piping Vibration Monitoring, Section 6.5.3, MainFeedwater Pump 28 Start and Steady State Operation on Recirc.,was field work complete with steady state velocity and displacementexceeding the Acceptance Criteria. CR 1168287 was written for anengineering evaluation and resulted in adjustment of a loose hangerand a retest. The retest was completed on 5113116 with satisfactoryresults. There was no Review Criteria for this test.516116 PAT-1.T,OperationalAlignmentofProcessTemperaturelnstrumentation, field work complete with allAcceptance Criteria met.One Review Criteria was not met and CR 1168641 was initiated.517116 PAT-3.0, Attachment 1, field work complete with all AcceptanceCriteria met. CR 1168487 was written to document alternatecharging flow was not within anticipated range, however, it had noaffect on the test acceptance.-2-PAT-3.11, Adjustment of Steam Flow Transmitters at MinimalFlow, field work complete with all Review Criteria met. There was noAcceptance Criteria associated with this performance.518116 PAf -1.11 - RVLIS Performance Test, Section 6.1.3, field workcomplete and results indicated Acceptance Criteria would not bemet. The system was updated with the new RVLIS constantssupplied by Westinghouse to correct the abnormality. CR 1171130was initiated.-2-PAT-1.12, Common Q Post Accident Monitoring System, Section6.8, Pump Contact Data Collection at 557"F, was completed with allcriteria met.-2-PAT-3.7, Reactor Coolant Flow Coastdown, field work completewith all criteria met.L4 3.0Watts Bar Unat 2 Startup Chronology (continued)5112116-2-PAT-3.8 Rod Drop Time Measurement and Stationary GripperRelease Timing (Mode 3), field work complete. CR 1169659 writtenfor two rods failing a two sigma statistical evaluation. Additional roddrops were performed and the Acceptance Criteria was met. Therewas no Review Criteria.-2-PAT-1.4, Pipe Vibration Monitoring, Section 6.5.4, Turbine BypassValve 2-FCV-1-105 Transient, was completed with all criteria met.Section 6.5.5, Turbine Bypass Valve 2-FCV-1-111 Transient wascompleted onSl12h6 and re-tested on 5113116. Engineeringevaluation of the retest indicated satisfactory results. CR 1170319documented the engineering evaluation.-2-PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete with all criteria met after a volume booster adjustment withCR 1170159 and a retest on 2-FCV-1-108.-2-PAT-3.4, Rod Control and Rod Position lndication (CERPI) - fieldwork complete. The Acceptance Criteria was not met in Sections 6.4and 6.10. CRs 1168845, 1168881, and 1169602 were written todocument failure to meet criteria. The criteria was re-evaluated and itwas determined the Acceptance Criteria should be changed torequire each Rod Position lndication to indicate rod motion consistentwith the group demand indication for the full range of rod travel. Achange to the Westinghouse Acceptance Criteria and SAR ChangePackage No. U2-019 were approved and an urgent change to theprocedure incorporated the revised Acceptance Criteria. AllAcceptance Criteria were then met.-2-PAT-3.0, Post Core Loading and Precritical Test Sequence, TRGapproved.-2-PAT-4.0, lnitial Criticality and Low Power Test Sequence, inprogress.-A cool down to 360'F was initiated to replace a failed RTD on RCSLoop 3 Hot Leg. The unit was stabilized between 355-365'F at22:59.511311651151165t1611 651111165118/1 651201165t21t16-Unit was placed in Mode 4 at23:58 to facilitate repairs to the SolidState Protection System (SSPS).-Unit was returned to Mode 3 at 04:15.-Unit reached NOTP at 01:00. Response time testing of the replacedRCS RTD indicated it did not meet its Acceptance Criteria. A DCNwas initiated to revise the Acceptance Criteria to allow entry intoMode 2.15 3.0 Watts Bar Unat 2 Startup Chronology (continued)5123116 -Unit entered Mode 2 at0'l:04.-lnitial criticality at 02:16.-2-PET-201, lnitial Criticality and Low Power Physics Testing,completed with all criteria met.-2-PET-103, Reactivity Computer (ADRC), completed with all criteriamet.-2-PET-304, Operational Alignment of NlS, applicable sectionscompleted with all criteria met.5124116 PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 was written to document three channelsremoved from service.-2-PAT-1.10, lntegrated Computer System (lCS), applicable sectionscompleted with all criteria met. CR 1173586 was initiated for ICS PIDquality on several points but did not affect this plateau performance.CR 1174334 was initiated for exceeding the MED between T0457AMCR indicator 2-Tl-62-29, RCP 3 LWR RADIAL BRG Temp.-2-P4T4.0, lnitial Criticality and Low Power Test Sequence TRGapproved.-2-PAT-5.0 Test Sequence for 30% Plateau in progress.5125116 -Unit 2 entered Mode 1 at 03:33.5126116 PAT-5.3, Automatic Steam Generator Level Control Transients atLow Power, completed with all criteria met.5127116 PAT-5.1, Dynamic Automatic Steam Dump Control, completedSections 6.3, 6.4 and 6.5 with allAcceptance Criteria met after aprocedure and UFSAR revision per Westinghouse LetterLTR-SCS-16-23.-With reactor power between 13 and 14 percent the turbine was rolledfor testing in preparations for initial generator synchronization. Duringthe roll up an unanticipated noise was heard and the roll wasterminated. A second rollwas made later in the evening with similarresults. A decision was made to place the Unit in Mode 3 for turbinerepairs.5128116 -Unit 2 re-entered Mode 3 at 01:54 after a manual reactor trip forturbine repairs.5131116 -Unit 2 re-entered Mode 2 at 12:00.-Reactor taken critical at 13:39.-Unit 2 entered Mode 1 at 17:49.-RCI-159, Radiation Baseline Surveys, completed. No Acceptance orReview Criteria were associated with this procedure.L6 3.0 Watts Bar Unat 2 Startup Chronology (continued)618116619t1661111166113t16-Unit 2 was synchronized to the grid at 20:39 and holding at 15percent power to repair steam leaks.-Unit 2 turbine was manually tripped due to a non-isolable steam Ieak.-Unit 2 synchronized at 11:40.-Unit 2 received an automatic reactor trip with a safety injection due to#1 governor valve failing open causing a steam line pressuredecrease and subsequent Reactor Trip and Safety lnjection at 12:27.Unit was stabilized in Mode 3 following Reactor Trip.-Unit 2 in Mode 2 at01:39 after repairs to the governor valve.-Entry into Mode 1 was at 09:32.-Unit 2 synchronized to grid at 06:40.-Turbine manually tripped due to an non-isolable steam leak at 17:52.-Unit 2 synchronized to the grid at 13:23 and power increase initiatedto the 30% testing plateau.-2-PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-PAT-1.12, Common Q PostAccident Monitoring System,applicable sections were completed with all criteria met.-2-PAT-1.11, RVLIS Performance Test, applicable sections werecompleted with all criteria met.-Completed the initia! Flux Map in accordance with 2-T141, lncoreFlux Mapping, and 2-Sl-0-20, Hot Channel Factors Determination.-2-PAT-1.10, lntegrated Computer System (lCS), was completed withall criteria met. CR 1181784 was written to address a database errorbut did not affect this plateau performance.-2-PAT-1.4, Pipe Vibration Monitoring, completed with all criteria metfor observations at the 30% Plateau.-2-PAT-5.3, Automatic Steam Generator Level ControlTransients atLow Power, was completed with allAcceptance Criteria met. CR1181278 was initiated to document one Review Criteria not met. Anengineering evaluation determined this did not affect the performanceof the test nor invalidate any of the test results and testing shouldproceed to the next plateau.61141166115/1 66/3/1 66t4t166/5/1 6L1 3.0Watts Bar Unat 2 Startup Chronology (continued)6116/1 6-2-PAT-1 .7, Operational Alignment of Process Temperaturelnstrumentation, was completed with allAcceptance Criteria metTwo Review Criteria concerning parameters related to Delta T failed.The OTDT calculated by Eagle-21 and provided by the MMI cartsindicated approximately 158% and the MCR indicators maximumvalue is 150%. lt was expected the reading from Eagle-21 wasaccurate and the MCR meters were ranged such that they cannotread the higher value. Additional data was taken at higher powerranges and the meters came on scale with no issue. CR 118246 waswritten.-2-PAT-5.4 Calibration of Steam and Feedwater Flow lnstruments at30% Power was completed with all criteria met.-2-PAT-1.6 Startup Adjustments of Reactor Control System, wascompleted. This was data taking only with no Review or AcceptanceCriteria at this plateau.-2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete with all criteria met.-RCl-159, Radiation Baseline Surveys completed. No Acceptance orReview Criteria were associated with this procedure.-2-PAT-5.0, Test Sequence for 30% Plateau, was TRG approved.-2-PAT-6.0, Test sequence for 50% Plateau, performance sectionwas entered and power increase to 50% Plateau level initiated.-U-2 turbine tripped due to loss of 28 Main Feedwater Pump fromloss of MFP condenser vacuum with a subsequent automatic reactortrip as a result of the S/Gs reaching their low-low trip setpoint. Theplant was stabilized in Mode 3.-U-2 re-entered Mode 2.-U-2 re-entered Mode 1 and synchronized to the grid.-U-2 manually tripped the turbine due to a steam leak. Mode 2 wasentered and subsequently the reactor was tripped and the unitstabilized in Mode 3.-U-2 again entered Mode 2 and reactor critical at 03:20.-Unit entered Mode 1 at07:57.-U-2 synchronized to the grid at 13:36.6117 11661201166t23t16612411661261167121166115/1 618 3.0 Watts Bar Unat 2 Startup Ghronology (continued)717116 -U-2 reached 50% Plateau power level requirements for testing.-2-PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-PAT-1.1 1, RVLIS Performance Test, applicable sectionscompleted with all criteria met.-2-PAT-1.12, Common Q PostAccident Monitoring System,applicable sections were completed with all criteria met.718116 PAT-1.4, Pipe Vibration Monitoring, completed with al! criteria met.-2-PAT-1.6, Startup Adjustments of Reactor Control System,completed. This performance was data taking only with noAcceptance or Review Criteria at this plateau.-2-P AT -1. 7, Operational Al ig n ment of Process Tem peratu relnstrumentation, applicable sections completed with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems, completed withtwo issues being referred to Site Engineering for evaluation.Engineering review indicated it was acceptable to continue PowerAscension Testing. (See Problem Report 4 of 2-PAT-1.8).-2-PAT-1.10, lntegrated Computer System (lCS), completed with allcriteria met.-2-PAT-6.3, Calibration of Steam and Feedwater FIow lnstruments, at50% Power completed with all criteria met.719116 PAT-3.3, RCS Flow Measurement, completed with all criteria met.7113116 PAT-6.1, Automatic Reactor Control System, completed with allcriteria met.7114116 PAT-5.2. Turbine Generator Trip With Coincident Loss of OffsitePower Test, completed with all Acceptance Criteria met.CR 1192287 was written to document Tcold decreasing below the547'F Review Criteria.-Unit 2 entered Mode 3.-2-PAT-1.4, Pipe Vibration Monitoring, applicable section for transienttesting was completed with all criteria met.7116116 PAT-6.2. AutomaticSteam GeneratorLevel Control Transientscompleted with all criteria met.-2-PAT-6.0, Test Sequence for 50% Plateau, was approved by TRG.7117116 -Unit 2 re-entered Mode 2 after a planned trip with 2-PAT-5.2, TurbineGenerator Trip Coincident With Loss of Offsite Power Test.7118116 -Unit entered Mode 1 and synchronized to the grid.L9 3.0 Watts Bar Unat 2 Startup Chronology (continued)7119116 PAT-7.0, Test Sequence for 75o/o Plateau performance sectioninitiated.-2-PNf-1.2,Load Swing Test, was completed with allAcceptanceCriteria met. CR 1193637 was written for the Review Criteria notbeing met for an undershoot of steam header pressure. The ReviewCriteria required an undershoot of no more than 25 psi and the actualwas 28.5 psi. This test was originally scheduled for the previous 50%Plateau testing, however, issues with the turbine IMP lN controlsprevented performance during that plateau. The unit was held at45o/o power on the ascension to the 75% testing plateau to performthis test.-2-PAT-1.4 Pipe Vibration Monitoring, applicable sections for the loadswing were completed with all criteria met.7125116 -Unit 2 reached 75o/o Plateau testing power Ievel.-2-PAT-1.5, Loose Parts Monitoring System, was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-P AT -1. 1 0, I ntegrated Computer System (lCS), completed.CR 1195476 written for failure of Acceptance Criteria. MCR indicator2-Tl-62-Tl comparison to ICS PlDT0127A (Regen Heat ExchLetdown Temp) was not within the MED. The CR was closed aftercalibration of the instrument.-2-PAT-1.1 1, RVLIS Performance Test, applicable sectionscompleted with all criteria met.-2-PAT-1.12, Common Q PostAccident Monitoring System,applicable sections were completed with all criteria met.-2-PAT-1.8, Thermal Expansion of Piping Systems, completed withall criteria met.7126116 PAT-1.4, Pipe Vibration Monitoring, completed with CR 1195665written on excessive vibration on the Main Steam Line Trap drain line.Temporary repairs to stabilize the line were initiated. All other criteriawere met.7127116 PAT-3.3, RCS Flow Measurement completed with a!! criteria met.7128116 PAT-1.6, Startup Adjustments of Reactor Control System, wascompleted with all criteria met.-2-PAT-7.1, Calibration of Steam and Feedwater Flow lnstruments at75o/o Power, was completed. Steam Flow and Feedwater Flow dataobtained in Section 6.1 of this PAT on7l26l16 was used to adjust thespan of the associated Steam Flow transmitters. Post calibrationdata was subsequently taken in accordance with Section 6.2 of thisPAT on 7128116. and all Review Criteria were met. There was noAcceptance Criteria for this PAT.20 3.0Watts Bar Unat 2 Startup Chronology (continued)-2-P AT -1. 7, O peration al Al i g n ment of Process Tem peraturelnstrumentation, was completed with allAcceptance Criteria met andall Review Criteria met upon the second performance. CR 1196243and CR 1196245 were generated for the initial failures. On thesecond data collection all Review Criteria were met and the CRsclosed.-2-PAT-1.9, Automatic Steam Generator Level Control, wascompleted with all criteria met.-2-PAf -7.0, Test Sequence for 75o/o Plateau, was TRG approved.-2-PAT-8.0, Test Sequence for 100% Plateau, performance sectionwas initiated. Due to increasing generator bushing temperatures andconcerns for further power increase, the original sequence of testingwas revised to perform 2-PAT-8.5, Shutdown From Outside the MainControl Room. An outage after the PAT performance was plannedfor repairs to the generator bushing.-Unit 2 power reduced to approximately 30% RTP.-2-PAT-8.5, Shutdown From Outside The Main Control Room wascompleted with all criteria met. The Unit was held in Mode 3 forequipment repairs.-Repairs were completed and Unit 2 startup initiated.-Unit 2 entered Mode 2 at 12:22 and the reactor was critical at 12:31 .-Unit entered Mode 1 at 16:14.8110/1 6-Unit 2 synchronized to the grid at 06:12. A delay in synchronizationoccurred due to particles in the thrust bearing wear trip fluid whichrequired flushing multiple times.-During power ascension an issue with increased temperatures on Cphase main generator bushing developed. This temperature issuewas noted prior to the Shutdown from Outside the Main Control Roomand resulted in a second planned outage.-Unit 2 was manually tripped at 03:06 and stabilized in Mode 3 for aplanned outage.-Unit 2 reactor critical at02:28.-Mode 1 entry at 08:32.-Unit 2 synchronized to the grid at 13:53.8113/1 68t22t167129t16811116813116817 116819t162L 3.0 Watts Bar Unat 2 Startup Chronology (continued)8123116 -Unit 2 reactor was manually tripped at 13:56 when the 2A MainTurbine Driven Feedwater Pump slowed and failed to providesufficient flow to maintain steam generator levels. The unit wasstabilized in Mode 3.8125116 -Unit 2 entered Mode 2 at 14:27 and the reactor was critical at 14:46.-Unit 2 entered Mode 1 at approximately 17:25.-Unit 2 synchronized to the grid at 23:19.8129116 -Unit 2 at93o/o rated thermal power allowing PAT testing tocommence at the 90% plateau.-2-PAT-1.6, Startup Adjustments of Reactor Control System,completed. For the g0% Plateau data collection was completedand satisfactory for this plateau. CR 1208178 was initiated beforeperformance because 2-PT-1-81 was unavailable for the test due toa steam leak. Results were acceptable for continuation to the100o/o plateau where measurements were repeated.-2-PAT-1 .7, Operational Alignment of Process Temperaturelnstrumentation, Review Criteria 5.2.8 and 5.2.C were not met buta CR was not written as the 2-PAT-1.7 performance was designedto correct the issue and the Acceptance Criteria were verified at the100Yo power plateau. All other Review and Acceptance Criteriawere met.-2-PAT-8.4, Calibration Of Steam And Feedwater Flow lnstrumentsat 100% Power, performance at 93% completed. All ReviewCriteria were met for Section 6.1. There was no AcceptanceCriteria for Section 6.1.8/30/16 -Unit 2 at > 98% rated thermal power allowing PAT testing tocommence at the 100o/o plateau.-2-PAT-1.5, Loose Parts Monitoring System was completed with allcriteria met. CR 1171424 documents three channels removed fromservice.-2-PAT-1.6, Startup Adjustments of Reactor Control System, datacollection was field work complete before the unit tripped.Subsequently, CR 1211020 was written for one failed AcceptanceCriteria. The failed criteria was due to full load steam pressurebeing below the expected value because T"rnwas at its maximumvalue. However, there is no safety or operational concern.Additionally, CR 1211015 was written for failed Review Criteria . 2-PT-1-81 was out of service, therefore, calibrations of the pressuretransmitter will be verified when 2-PT-1-81 is returned to serviceoutside the PAT program. CR 1208178 was previously written forthis issue and WO 118121693 will resolve the issue.22 3.0 Watts Bar Unit 2 Startup Chronology (continued)8/30/16 PAT-1.T,OperationalAlignmentofProcessTemperaturelnstrumentation, data collection was field work complete before unittripped. Data reduction was completed with failure to meetAcceptance Criteria for Loop 4 Taw. On917116 CR 121'1021 waswritten documenting this failed Acceptance Criteria. Although afailure, there was no safety concern or failure to meet the licensingbasis. All Review Criteria were met on this performance of thePAT.-2-PAT-1.8, Thermal Expansion of Piping Systems, completed withallcriteria met.-2-PAT-1.10, lntegrated Computer System (lCS), was completedfor the 100o/o Plateau. CR 1208754 was generated for failure ofmeeting the MED between indicator 2-Tl-062-0004 and ICS pointT0181A, RCP 1 No 1 Seal Outlet Temperature. A WO wasgenerated to calibrate and has subsequently closed.-2-PAT-1.1 1, RVLIS Performance Test, was completed with allcriteria met.-2-PAT-1.12, Common Q PostAccident Monitoring System, wascompleted with all criteria met.-2-PAT-3.3, RCS Flow Measurement, data collection was field workcomplete before unit tripped.-2-PAT-8.4, Calibration of Steam and Feedwater Flow lnstruments,at 100% Power, was field work complete. CR 1208875 was writtento document failure of Review Criteria on three steam flowtransmitters. WOs were initiated to respan the transmitters.-Unit 2 received an automatic Turbine Trip - Reactor Trip at21:09:13 due to a fault in the 28 Main Bank Transformer, resultingin a fire in the transformer. The unit was stabilized andsubsequently placed in Mode 4 for repairs.9115116 PAT-8.6, Plant Trip from 100o/o Power, was evaluated from datagathered during the actual plant trip on 8/30/16. AllAcceptanceCriteria was met. CR 1209770 was written to evaluate theequivalency of the data collected by the plant as well as oneReview Criteria which did not meet pressurizer level modulation tono load setpoint within 30 minutes. A Westinghouse evaluationconcluded the response was acceptable.-2-PAT-1.4, Pipe Vibration Monitoring, Section 6.6.19 was closedbased on an engineering walkdown evaluation CR 1211196.9125116 -Unit 2 entered Mode 1 at 01:53 after the Spare Main BankTransformer was placed in service for the failed 28 Main BankTransformer which was removed from site.9126116 -Unit 2 generator synchronized to the grid at 01:07.23 3.0Watts Bar Unat 2 Startup Chronology (continued)91281169129116-Unit 2 reached 100o/o power.-RCl-159, Radiation Baseline Surveys completed for 100o/o Plateau.No Acceptance or Review Criteria were associated with thisprocedure.2-PAT-1.9, Automatic Steam Generator Level Control, was fieldwork complete with all criteria met.-2-P Nf -1. 7, Operational Al ig n me nt of Process Tem peratu relnstrumentation, was completed with failure to meet AcceptanceCriteria for Loop 4Taw. However, on917116 CR 1211021had beenpreviously written documenting this failed Acceptance Criteria.-2-PAT-1.2,Load Swing Test, was field work complete with allAcceptance Criteria met. CR 1218746 was written for failure of oneReview Criteria for S/G Level response. Westinghouse evaluatedthe response to be adequate with no further testing required.-2-PAT-3.3, RCS Flow Measurement, was completed with allcriteria met.-2-PAT-8.4, Calibration of Steam and Feedwater Flow lnstruments,at 100o/o Power, was field work complete with all criteria met.2-PAT-1.3, Large Load Reduction Test, was field work completewith allAcceptance Criteria met. CR 1218917 was written forReview Criteria failure of S/G levels to remain within t15o/o of theprogram level. Westinghouse concluded the response wasacceptable.2-PAT-1.4, Pipe Vibration Monitoring, was completed with CR1208694 written for main steam traps excessive vibration as wasnoted at the 75o/o Plateau also. Civil Design generated WO118122821to design and install a restraint outside the PATP.2-PAT-1.6, Startup Adjustments of Reactor Control System, wasfield work complete on 10/3/16. Previously, CR 1211020 waswritten for one failed Acceptance Criteria. The failed criteria wasdue to full load steam pressure being below the expected valuebecause T"rnwas at its maximum value. However, there is nosafety or operational concern. Additionally, CR 1211015 waswritten for failed Review Criteria. Due to 2-PT-1-81 being out ofservice calibrations of the pressure transmitter will be verified when2-PT-1-81is returned to service outside the PAT program. CRs'1208178 and 1216904 were previously written for this issue.-2-PAT-8.0, Test Sequence for 100% Plateau was TRG approved.9t29t169/30/1 610131169127 11610t611624 4.04.1INITIAL FUEL LOADOverview and Summary of lnitial Core LoadingThe initial core loading at WBN Unit 2 was accomplished in approximately76 hours from December 4,2015, to December 8,2015, as directed by2-PAT-2.0, Initial Core Loading Sequence.Core loading was performed "wet" with the refueling cavity and the reactor vesselfilled with refueling concentration borated water at normal refueling levels. Thecore loading sequence was performed in accordance with an approved FuelAssembly Transfer Form (FATF). Actual movement of fuet was performed inaccordance with 2-FHl-7, Fuel Handling and Movement, as directed by 2-PET-105, lnitial Core Loading.The neutron monitoring station for lnverse Count Rate Ratio determinations wereestablished in the main control room to monitor source range detectors N-31 andN-32. ICRR plots were maintained for these detectors during all core loadingsequence steps and during delays in core loading to ensure that an adequatesubcritica! margin was maintained at all times.As a visual aid in tracking fuel movement evolutions and to ensure the core loadconfiguration was in accordance with the approved loading pattern prescribed onthe FATF, a core status display was maintained in the Main Control Room.RCS boron concentration was monitored during core loading to ensure that theboron concentration remained within prescribed limits.Some fuel assemblies were required by plan to be moved more than once,specifically those bearing primary neutron sources. As such, the core loadingsequence required two in-core fuel assembly movements to move the sourcebearing fuel assemblies from the reactor baffle wall to their final locations withinthe core. This was done to ensure that neutron counts could be monitored by theSource Range instrumentation at all times during core loading. After the corewas loaded, a video recording was made and verification of proper fuel assemblyposition and orientation was conducted. Fue! Related Components (FRCs) wereconfirmed to be inserted into the proper fuel assembly with the proper orientationin the Spent Fuel Pool prior to core load. The final core load configuration wasconsistent with the Westinghouse Core Loading Plan for Unit 2 Cycle 1.25 4.2 lnitial Gore Loading Sequence (2-PAT-2.0)This test started on 11119115 with prerequisites and completed on 1218115.1.0 Test ObiectivesThe objectives of this test were to:1.1 Sequence the procedures that established the prerequisitesrequired for the initial core loading of Unit 21.2 Define the sequence of operations and tests which were to beconducted during and following completion of the initial coreloading.The following PATs/PETs/RCl were sequenced for performance by2-PAT-2.0:o 2-PAT-2.1 Reactor System Sampling for Core Loado 2-PAT-2.2 Response Check of Core Load lnstrumentation After8 Hour Delay in Fuel Movemento 2-PET-102 Pre-Power Escalation NIS Calibration Datao 2-PET-105 lnitial Core Loadingo RCI-159
+* Radiation Baseline SurveysNote:
+* lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.2.0 Test MethodsPre-requisite actions started on 11119/15, prior to entry into Mode 6 toestablish prerequisite conditions in support of commencement of initialcore loading. The test continued through verification of core loading andwas field complete on 1218115, prior to the reassembly of the reactorvessel in preparation for Mode 5 entry.The major pre-requisites included the following:o Verification all Preoperational Test completed and test resultsapproved or technicaljustifications for delaying tests unti! after fuelload were approved by the Plant Managero Verification 2-PET-102,Pre-Power Escalation NIS Calibration Data,was successfully completed to the extent necessaryo RCI-159 Radiation Baseline Surveys commenced for the pre-fuelload survey26 4.2 lnitial Core Loading Sequence (2-PAT-2.0) (continued)o 2-PAT-2.1, Reactor System Sampling for Core Load, startedo Visual lnspection of the reactor vessel core support plate inaccordance with 2-PET-105, lnitial Core Loading completed Testingincluded the following:o 2-PAT-2.1, Reactor System Sampling for Core Load, completed on1214115 with all criteria met.o 2-PAT-2.2, Response Check of Core Load lnstrumentation After 8Hour Delay in Fuel Movement, completed on 1215115 with allcriteria meto 2-PET-102, Pre-Power Escalation NIS Calibration Data, applicablesections completed with all criteria meto 2-PET-105, lnitial Core Loading completed on 1218115 with allcriteria met.o RCI-159, Radiation Baseline Surveys, completed on 1213115.There was no Acceptance or Review Criteria associated with thisprocedure.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.0 ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAf-2.0.21 4.3 Reactor System Sampling for Gore Load (2-PAT-2.11This test was performed as part of test sequence 2-PAT-2.0, lnitial Core Loading.Testing was started on 11119115 and field work completed on 1214115.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the boron concentrations in the Reactor Coolant System(RCS), Residual Heat Removal (RHR) system and other directlyconnected portions of auxiliary systems are uniformly borated toprevent inadvertent dilution during core loading.1.2 Verify un-borated water sources are configured to preventinadvertent dilution during core loading.1.3 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 4, ReactorSystem Sampling For Core Loading Test Summary.2.0 Test MethodsThe preliminary actions of 2-PAT-2.1 researched logs and procedurallydriven Unit 2 activities that were completed and that were associated withthe preparations of the unit 2 water systems for entering Mode 6. Theresearch started with ensuring the RWST was borated to (3100 to 3300)ppm. Actual recorded samples of the RWST were 3326 ppm on0910712015 and 3224 ppm on 1'112112015. This confirms a correctlyborated RWST. This RWST water was subsequently used to fill the RCSand partially fill the Refueling Cana! and Cavity. Later boron sampleresults showed Refueling Canal at 3290 ppm and Refueling Cavity at3281ppm.Following the proper boration of the RWST and water transfer to the RCS,unit activities were verified that circulated water through:Both RHR A-A and B-B pump miniflowsBoth RHR pumpsRHR to CVCS LetdownBoth Charging pumpsBoth Containment Spray pumps (re-circulated borated RWST)Refueling Water Purification Pump BRefueling Water Purification Pump A was found to be tagged with aCaution Order 0-CO-2015-0048 stating that the pump has high vibration.WO 116318322 was previously written to address this issue. The volumeof potentially diluted water was conservatively calculated to be 13 gals.This small volume did not pose any risk to challenging any criteria listed inthis PAT.28 4.3 Reactor System Sampling for Core Load (2-PAT-2.1) (continued)The Boron lnjection Tank (BlT) was verified to have been borated andmixed via the performance of 2-5l-63-905, Boron lnjection Check ValveFlow During Refueling Outages.It was verified that both trains of Safety lnjection were circulated during theperformance of 2-Sl-63-906, Safety !njection Check Valve Full FlowTesting During Refueling Outages, on 1112512015.All4 Cold Leg Accumulators were verified by sample to be corectlyborated with the lowest reading 3175 ppm and the highest reading 3206ppm.The water in the Holdup Tank B (HUT B) was recirculated and sampled forboron concentration on 11104115 and found to be 3204 ppm boron; thiswater was used to fill the Fuel Transfer Canal. The Spent Fuel Pool (SFp)was sampled for boron concentration on 11123115 and found to be 3261ppm boron. Boric Acid Tanks B and C were sampled for boronconcentration on 11121115 and found to be 6919 and 6808 ppm boronrespectively.Problems encountered while running 2-Sl-63-905 and 2-5!-63-906resulted in a partial drain down moving water back to the RWST. Thissame water was again used to fill the Refueling Cavity and Chemistry re-performed 2-Sl-78-1, Reactor Coolant System and Refueling CanalRefueling Operations Boron Determination, to document compliance withthe refueling boron concentration requirements.Watts Bar Unit 2 systems connected to the Reactor Coolant System(RCS) were adequately borated and mixed to prevent a dilution event insupport of the initial core loading operations.2-PAT-2.1, Reactor System Sampling For Core Load, supported thisconclusion from research of the chemistry and operation logs.configuration contro! measures were in place to ensure that the RCS andconnected systems remain adequately borated for support of the initialcore loading operations. The Unit 2 Refueling Water Storage Tank(RWST) was at approximately 16.9% and based on calculations of recentmakeup to the RWST it was determined that the tank was adequatelyborated. Technical Specifications required Operations to validatecompliance with the RWST boron concentration and level prior to Mode 6.Technical Specifications required Watts Bar Unit 2 to maintain Mode 6surveillance instructions in frequency. Therefore, no additional sampling ormixing was required for 2-PAT-2.1.29 4.3 Reactor System Sampling for Gore Load (2-PAT-2.1) (continued)3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 Boron Concentration of samples meet requirements of theTechnical Specifications.3.1.1 The RCS Boron concentration is greater than or equal to3100 ppm and less than or equal to 3300 ppm.The RCS Boron as measured in the RHR TRAIN B systemwas 3284 ppm.3.1.2 The boron concentration final samples obtained from thedesignated sample points identified are uniformly boratedbetween 3100 ppm and 3300 ppm.The boron concentration for sample points met therequirements.3.1.3 The boron concentration of samples obtained from the BoricAcid Tanks (BAT B and BAT C) are within the limits ot 6120< Ce < 6990 ppm.Boron concentrations were 6919 ppm in BAT B and 6808ppm in BAT C.3.1.4 Un-borated water sources are configured to preventinadvertent dilution during core loading.2-Sl-62-1, Uncontrolled Boron Dilution Paths, wassatisfactorily completed for Mode 6.Review Criteria3.2 The boron concentrations for the Reactor Coolant System (RCS)and directly connected portions of the auxiliary systems are greaterthan or equal to 3100 ppm and less than or equal to 3300 ppm.Boron concentrations for the RCS and directly connected portionsof the auxiliary systems met the requirement.4.0 ProblemsThere were no significant problems encountered during the performanceof this test.30 4.4 Response Check of Core Load lnstrumentation After 8 Hour Delay inFuel Movement (2-P AT -2.21This test was performed as part of test sequence 2-PAT-2.0, lnitial Core LoadingSequence. Testing was started and completed on 1215115.1.0 Test ObiectivesThe objective of this test was to:1.1 Verify response of the Source Range Channels prior to resumptionof fuel loading following a delay of eight (8) hours or more.2.0 Test MethodsThree methods of testing were available for use:2.1 Statistical Evaluation Method using the Scaler Timer2.2 Statistical Evaluation Method using the Source Range Count Rateindications.2.3 Response Check of Core Load lnstrumentation Using PrimarySource Bearing Fuel Assembly Movement.The Statistical Evaluation Method using the Scaler Timer provided theverification of the Acceptance Criteria for resumption of fuel movement.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 The Source Range instrumentation for both channel N-31 and N-32were evaluated and determined to be acceptable for continuation offuel loading by meeting at least one Review Criteria.Review Criteria 3.2, below31 4.4 Response Check of Gore Load Instrumentation After 8 Hour Delay inFuel Movement (2-P AT -2.2) (continued)Review Criteria3.2 Statistical Evaluation Method using the Scaler Timer:Statistical Reliability Factor (SRF) for Source Range Channels shallbe > 0.5 and 3 1.4.Results indicated the SRF for Source Range Channel N-31 was1.2395 and SRF for Source Range Channel N-32 was 0.8609.3.3 Statistical Evaluation Method using the Source Range Count Rateindications:1. The Student F Distribution Test shall be satisfied by havingFexp s 3.179.2. The Student T Distribution Test shall be satisfied by havingTexp s 2.101.This method was originally chosen, however, problems wereencountered. See Problems below.3.4 Neutron instrumentation (Source Range Channels N-31 and N-32)are operational and indicates a positive (negative) change in countrate as the neutron level detected from a source is increased(decreased).This method was not used.4.0 Problemst1l No CR initiated:Section 6.2, Statistical Evaluation Using Source Range Count Ratelndications, method was attempted four (4) times with unsuccessfulresults. Based on only three assemblies loaded at the time ofperformance, low counts appeared to cause data scatter which wasobserved in monitored count rates. This failure of Section 6.2method resulted in the transition to Section 6.1, StatisticalEvaluation Method using the Scaler-Timer. Section 6.1 methodwas acceptable. No CR was initiated since this was a potentialscenario and the test provided alternative methods.32 4.5 Pre-Power Escalation NIS Calibration Data (2-PET-1021This test was performed as part of the test sequence 2-PAT-2.0, lnitial CoreLoading Sequence. The performance of 2-PET-102 was conducted viaWO 116884907. The WO started 0812512015 and was complete on 04112116withcalibration of all Power Range and lntermediate Range detectors.1.0 Test ObiectivesThe objectives of this test were to:1.1 Provide Nuclear lnstrumentation System (NlS) Power Range (PR)and lntermediate Range (lR) excore detector calibration data.1.2 lnitiate an adjustment of the NIS before startup for a new fuel cycle.Note: The calculation methodology in 2-PET-102 applied to the changesexpected to occur due to a refueling outage. This procedureaccommodated the calibrations to be performed for Cycle 1.2.0 Test MethodsThe normal method for determining calibration data after fuel reload andprior to startup is to ratio the sum of selected weighted assembly predictedpowers from the Beginning of Life (BOL) of the previous fuel cycle (Unit 1Cycle 1 was used as the reference condition) to the BOL of the upcomingcycle. This ensures a ratio based upon similar BOL core conditionsincluding the neutron energy spectrum and a nearly cosine axial fluxshape. This provides the most accurate excore Axial Offset indications forthe power range channels. This same methodology results in the mostaccurate power indications for the intermediate range channels.This same methodology was used to predict the lntermediate Range andPower Range calibration setpoints for the Unit 2 Cyclel startup, exceptthat Unit 1 Cycle 1 is used as the reference condition. ln this case,average composite values for the channels were used.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.33 4.5 Pre-Power Escalation NIS Calibration Data (2-PET-102) (continued)Acceptance Criteria3.1 Power Range Channel calibrations have been completed.The Power Range channel calibrations were completed via:o WO 115898162o WO 115898208o WO 115898252o WO 1158991873.2 lntermediate Range Channel predicted full power adjustments havebeen completed.lntermediate Range Channel predicted full power adjustments wereperformed. lR Gain Adjustment potentiometers were set to thevalues calculated in the PET via steps in Section 7.0 of the PET.3.3 lntermediate Range Channel OperationalTests (COTs) have beencompleted.The lntermediate Range channel COTs were completed via:o WO 117499181o WO 117499184Review CriteriaNone4.0 ProblemsThere were no significant problems encountered during the performanceof this test.34 4.6 lnitia! Core Loading (2-PET-1 05)This test was performed as part of test sequence 2-PAT-2.0, lnitial Core LoadingSequence. 2-PET-105 testing via WO 117370408 started on 1112312015 with theverification of Unit 2fuel assemblies and component inserts in the Spent FuelPool and completed on 1210812015 with the completion of fuel load and core loadverification.1.0 Test ObiectivesThe objectives of this test were to:1.1 ldentify the activities and requirements for fuel loading whichensure fuel loading is conducted in a cautious and controlledmanner:1.1.1 Specify the sequence for loading fuel assemblies into thereactor vessel such that the final core configuration isconsistent with that specified in the NuPOP for current fuelcycle. See Figure 4.6-1, U2C1 Core Load Sequence.1.1.2 Specify the fuel assembly identification number and type ofinsert for each core location.1.1.3 Establish the requirements for periodic and continuousneutron monitoring during each step of the core loadingprocess.1.1.4 Prescribe the steps necessary for obtaining and evaluatingneutron monitoring data during core loading.1.1.5 ldentify the neutron monitoring channels to be used duringeach step of the core loading sequence to ensure subcriticalconditions are maintained.1.2 Satisfied the requirements of UFSAR Table 14.2-2, Sheet 3, lnitialFuel Loading Test Summary.2.0 Test MethodsOnly data from "responding" detectors identified by the data package wasused in evaluating the safety of continued core loading. Prior tocompleting the loading of the initial nucleus of eight fuel assemblies,significant changes in the ICRR data were expected to occur due togeometry effects arising from changes in detector-to-fuel assemblycoupling. Therefore, the ICRR values were re-normalized followingmovement of source bearing fuel assemblies from the baffle wall to theirfinal location(s) in the core.35 4.6 lnitial Core Loading (2-PET-1 05) (continued)Changes in neutron flux level during and following fuel assembly insertionwas monitored for indications of abnormal and/or unstable reactivitybehavior.All fuel movement was performed in accordance with 2-FHl-7, FuelHandling and Movement.The core status display in the main control room was updated, asrequired, to reflect the actual physical location of all fuel assemblies andfuel related components at alltimes during the core loading evolution.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Note: Unit 2 Core Load ICRR plot is provided in Figure 4.6-2.Acceotance Criteria3.1 The core was successfully loaded in accordance with the Unit 2Cycle 1 Westinghouse Core Load Plan.Verification of successful core loading was provided via2-Tl-28, Physical Verification of Core Load Prior to Vessel Closure,(WO 117370592.) See Figure 4.6-1.3.2 At completion of each mini-core, the final count rate from anydetector shall not unexpectedly double from the initial count ratebefore the assembly was inserted.Neutron count rates observed during fuel movement did notunexpectedly double at any time.3.3 At completion of each mini-core the ICRR response from anydetector shall not be less than 0.5 as each fuel assembly isinserted.Neutron count rates observed during fuel movement did notunexpectedly double at any time and ICRR remained above 0.5,see Figure 4.6-2.3.4 Core loading operations are required to be immediately stoppedand the Containment Building evacuated if any of the followingconditions occur during core Ioading. Movements of an activesource bearing assembly, or detectorto-fuel assembly neutroniccoupling are anticipated type changes.36 4.6 Initial Core Loading (2-PET-1 05) (continued)3.4.1 An unanticipated simultaneous increase in the neutron countrate by a factor of > 2 on all "responding" neutron monitoringchannels.3.4.2 An unanticipated simultaneous increase in the neutron countrate on any individual "responding neutron monitoringchannel by a factor of > 5.Neutron count rates were acceptable and did not meet eithercriteria to warrant suspension of core loading operations orevacuation of the Containment Building.Review CriteriaAl! required Review criteria for this test were met as delineated below:3.5 Assessment of the ICRR response should be based on thepredicted ICRR response.ICRR plots maintained during core loading activities contained bothactual plant lcRR data as well as predicted ICRR data from the fuervendor, see Figure 4.6-2.3.6 Placement of initialfuel assemblies up to placement of primarysource assemblies in final core location should be detected by theICRR response.ICRR monitoring was maintained at all times during core loading,including loading of the first "mini-cores" and final movement ofprimary source bearing fuel assemblies.3.7 ICRR response should not be less than 0.8 for any fuel assembtyafter the primary source assemblies have been placed in their finallocations.ICRR data during core loading was determined to be less than 0.8following final placement of the two source bearing assemblies. CR1112886 was initiated to document violation of this ReviewCriterion. Violation of this criterion does not represent a failure ofthis test, as it only requires further evaluation by the fuel vendor.The fuel vendor was notified and agreed that the data wasacceptable.31 4.6 Initial Core Loading (2-PET-1 05) (continued)4.0 Problemst1] CR 1112049 was written to document a labeling issue. SourceRange labeling difference was noted between 2-PET-105 and theUnit 2 Main Control Room. Urgent Change 1 was processed for2-PET-105 to correct the labeling issue.l2l CR 1 112886 was initiated to document violation of the 0.8 ICRRlimit during core loading, as described in Section 3.7.t3l CR 11 12204 was initiated to document foreign material, laterdetermined to be glue, on the bottom nozzles on multiple fuelassemblies during initial core load. All assemblies noted to havedebris were cleaned prior to being loaded in the reactor.Efforts to remove the debris caused schedule delays.38 4.6lnitial Core Loading (2-PET-1 05) (continued)FIGURE 4.6.1U2C1 Core Load SequenceUZC1 Core Load SequencePERFORMED BY I39 4.6Initial Core Loading (2-PET-1 05) (continued)ilUHL,21.11.00.90.00.70.60.50.10.30.20.10.0FIGURE 4.6.2Unatz Core Load ICRRUnit 2Core LoadInverse Count Rate Ratio (ICRR)80 100 120 1a0ll&ber of nrel tage&lieso !{-31. t{-32rll-31 Fredictedrll-32 Fredicted40 5.0 PRECRITICAL TESTING5.1 Post Gore Loading Precritical Test Sequence (2-PAT-3.0)This test started on 121912015 and was completed on 05 115116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Serve as controlling document for establishing the required pre-requislte conditions to permit testing following the completion of2-PAT-2.0.1.2 Govern the sequence of tests performed in Mode 6 through Mode3.1.3 lmplement testing deferred from Pre-Operational Test lnstruction,2-PTl-062-03, HFT Charging and Letdown documented in CR1075347 and CR 1085430.The following PATs/PETs/RCl were sequenced for performance by2-PAT-3.0:o 2-PAT-1.4
+* Pipe Vibration Monitoringo 2-PAT-1.6
+* Startup Adjustments of Reactor Control Systemo 2-PAT-1.7
+* OperationalAlignment of Process Temperaturelnstrumentationo 2-PAT-1.8
+* Thermal Expansion of Piping Systemso 2-PAT-1.11* RVLIS Performance Testo 2-PAT-1.12* Common Q Post Accident Monitoring Systemo 2-PAT-3.1 Control Rod Drive Mechanism Timing and CERPI!nitial Calibrationo 2-PAT-3.2 Pressurizer Spray Capability and Continuous SprayFlow Settingo 2-PAT-3.3
+* RCS Flow Measuremento 2-PAT-3.4 Rod Control and Rod Position lndication (CERPI)o 2-PAT-3.7 Reactor Coolant Flow Coastdowno 2-PAT-3.8 Rod Drop Time Measurement and Stationary GripperRelease Timing. 2-PAT-3.10 Reactor Trip Systemc 2-PAT-3.11 Adjustment of Steam Flow Transmitters at MinimalFlowo 2-PAT-5.1
+* Dynamic Automatic Steam Dump Controlo 2-PET-106 Control Rod Drive Mechanism Timingo RCI-159
+* Radiation Baseline Surveys4L 5.1Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)2.0Note:
+* lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on121912015 and completed on 1211112015 and included verification of thefollowing major items:o 2-PAT-2.0 lnitial Core Loading Sequence completed.o 2-GO-7 Refueling Operations performed concurrently with2-PAT-3.0.o 2-GO-10 Reactor Coolant System Drain and Fill Operationperformed concurrently with 2-PAT-3.0.o RCI-159 Radiation Baseline Surveys, commenced for post-fuel loadactivities.Testing was performed at eight defined plateaus including, ambient(<105'F), 250"F, 300'F, 360'F,400'F,450"F, 500'F, and 557'F.This report is a summary therefore see individual test packages forspecific details at each plateau.Ambient Plateau testing included the following:o RCI-159, Radiation Baseline Surveys, Post Fuel Load Survey - fieldwork complete on 12112115. No Acceptance or Review Criteria wasassociated with this procedure.o 2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable sections on 12l16/15 with all criteria met.o 2-PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete for applicable sections on 1/16/16. There was noAcceptance or Review Criteria for this portion of testing.o 2-PAT-3.10, Reactor Trip System, field work complete on 1120116with all criteria met.o PAT-3.1, Control Rod Drive Mechanism and CERPI lnitialCalibration, field work complete on 1124116 with all AcceptanceCriteria met after evaluation of current amplitudes on twelve lift coilsdetermined the results to be acceptable for the designed operationof the rod control system. CR 1128950 was written for high currentamplitudes and closed following Westinghouse evaluation thatdetermined the measurements to be acceptable. There was noReview Criteria for this PAT.42 5.1 Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)o 2-PAT-3.8, Rod Drop Time Measurement and Stationary GripperRelease Timing, Mode 5 Performance, field work complete forapplicable sections on 1124116 with allAcceptance Criteria met. CR1128964 was written due to the RDTC plots for each rod wereinverted from the expected response. This did not impactperformance of the test and was resolved prior to the Mode 3performance. There was no Review Criteria for this test.. 2-PAT-1.4, Pipe Vibration Monitoring, field work complete forapplicable sections on211l16 with all criteria met.PAT testing on the plant primary side was suspended on 1126116 untilplant conditions and surveillance completions allowed further testing.Condensate was placed on modified long cycle which allowed thecompletion of the applicable portions of 2-PAT-1.4.On 3/15/16 preparations began for entering Mode 4 and PAT TestCoordinators began reviewing and completing pre-requisites for Mode 4testing. Mode 4, RCS temperature >200'F and <350'F, entry was madeon 3/19/16.The 250"F Plateau included the following:o 2-PAT-'1.12, Common Q Post Accident Monitoring System, fieldwork complete for applicable section on 3121116 with all criteria met.There is no Review Criteria associated with this PAT.The 300"F Plateau included the following:. 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 3125116 with al! criteria met.. The plant entered Mode 3, RCS temperature ) 350"F, on 3/30/16 at23:14 to allow further Power Ascension Testing.The 360"F Plateau included the following:o 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 3/31/16 with all criteria met.o 2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for applicable section on 3/31/16 with all criteria met.. 2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable sections on 3/31/16 with all criteriaacceptable for continued heat-up.Plant heat-up to 400 degrees was initiated on 411116 at01:14.43 5.1 Post Gore Loading Precritical Test Sequence (2-PAT-3.0) (continued)The 400'F Plateau included the following:o 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 411116. CR 1156311 documented some testdata was not taken with the RCP start but was collectedsatisfactorily from the plant computer. Review criteria was not meton the Reactor Coolant Pump Combination testing, The RVLISsystem was updated with the new constants supplied byWestinghouse to correct the abnormality and documented incR 1156425.o 2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for applicable sections on 411116 with all criteria met.After completion of the 400'F Plateau testing, plant heat-up to 450"F wasinitiated at 09:45 and completed at 12:16 on 411116.The 450"F Plateau inctuded the following:o 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 411116 with all criteria met.o 2-PAT-1.12, Common Q PostAccident Monitoring System, fieldwork complete for applicable section on 411116 with all criteria met.o 2-PAT-1.8, Thermal Expansion of Piping Systems - field workcomplete for applicable sections on 4/1/16. Problem Report #1 wasinitiated within the test for seven snubbers not performing asexpected. Results indicated no issue with the snubbers andapprovalwas received to continue to next plateau testing.Due to plant issues concerning check valve leakage, the decision wasmade on 412116 to cool down the RCS and make entry into Mode 4 toallow repairs. Mode 4 entry was made on 412116 at 06:29. Additionally,repairs on two RCS RTDs were made.Mode 3 re-entry was made on 418116 at 12:44.Plant condition of RCS temperature at 500'F was met on 4110116.The 500'F Plateau included the following:. 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable sections on 4110116 with all criteria met.44 5.1Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)Plant heatup to 557'F was completed at 15:30 on 4113116 with normaloperating pressure reached at 01:30 on 4114116. On 4117116 at 03:38 theunit was placed in Mode 4 for repairs to the Auxiliary Feedwater Pumpsand replacement ol PD07-2 shim determined to require adjustment. TheUnit was returned to Mode 3 on 5/1/16 at 17:36 and normal operatingtemperature and pressures on 512116 at 23:00.The 557'F Plateau included the following:c 2-PAT-1.11, RVLIS Performance Test, field work complete forapplicable section of 557'F data taking only on 4113116 with allcriteria met for steady state data collection. Additionally, Section6.1.3, Pump Combinations at 557'F, was field work complete on518116. Results (Section 6.1.4) indicated Acceptance Criteria wouldnot be met. The system was updated with the new RVLISconstants supplied by Westinghouse to correct the abnormality anddocumented in CR 1171130.o 2-PAT-1.12, Common Q Post Accident Monitoring System, fieldwork complete for Data Collection Section 6.7 on 4113116 with allcriteria met. Section 6.8, Pump Contact Data Collection at 557'Fwas completed on 518116 with all criteria met.. 2-PAT-1.8, Thermal Expansion of Piping Systems, field workcomplete for applicable sections on 4114116 with one issue outsidecontainment on PD07-2 which also required evaluation of PD07-1.Problem Report #2was initiated to resolve the issue with PD07-2and investigate any possible issues with PD07-1. AdditionallyProblem Report #3 was written to evaluate components not movingas expected. Both Problem Reports were closed and conditionswere acceptable to continue testing.o 2-PAT-1.4,Ptpe Vibration Monitoring, field work complete forSection 6.5.1 Pressurizer Surge - Mode 3, on 4114116 with allcriteria met for that section. Section 6.5.2, Main Feedwater Pump2A Start and Steady State Operation on Recirc., was field workcomplete on 4113116 with velocity and displacement AcceptanceCriteria not met. CR 1161783 was initiated for an engineeringevaluation which concluded acceptable as is. Section 6.5.3, MainFeedwater Pump 28 Start and Steady State Operation on Recirc.,was field work complete 5/5/16 with steady state velocity anddisplacement exceeding the Acceptance Criteria. CR 1168287 waswritten for an engineering evaluation and resulted in adjustment ofa loose hanger and a retest. The retest was completed on 5113116with satisfactory results. Section 6.5.4, Turbine Bypass Valve2-FCV-1-105 Transient was completed on 5112116 with all criteriamet. Section 6.5.5, Turbine Bypass Valve 2-FCV-1-111 Transientwas completed on 5112116 and re-tested on 5/13/16. Engineeringevaluation of the retest indicated satisfactory results.45 5.1Post Core Loading Precritical Test Sequence (2-PAT-3.0) (continued)CR 1170319 documents engineering evaluation to accept-as-isfollowing the retest. Section 6.5.7 Condensate - Long Cycle wasfield work complete on2l1116 with all criteria met. There was noReview Criteria for 2-PAT-1.4.2-PAT-3.2, Pressurizer Spray Capability and Continuous SprayFlow Setting - Section 6.1, Adjustment of the Pressurizer ManualSpray Bypass Valves, was completed on 4116116. AllAcceptanceCriteria was met. Review criteria for MCR alarms was not met withCRs 1161382 and 1160969 written. A Westinghouse evaluationdetermined the PAT met the operability and design requirementsfor the pressurizer spray system. Additionally, CR 1161789 waswritten for proportional heater band not within the specifiedrequirement which was not an Acceptance Criteria.2-PAT-3.3, RCS Flow Measurement, field work complete on 5/3/16with all criteria met.2-P AT -1. 7, Operational AI i g n ment of Process Te m peratu relnstrumentation, field work complete on 5/6/16 with allAcceptanceCriteria met. One Review Criteria was not met and CR 1168641was initiated.2-PAT-1.6, Startup Adjustments of Reactor Control System, fieldwork complete for Mode 3 on 514116. This performance was datataking only.2-PAT-3.4, Rod Control and Rod Position lndication (CERPI), fieldwork complete on 5113116. The Acceptance Criteria was not met inSections 6.4 and 6.10. CRs 1 168845, 1168881, and 1169602 werewritten to document failure to meet criteria. The criteria was re-evaluated and it was determined the acceptance criteria should bechanged to require each Rod Position lndication to indicate rodmotion consistent with the group demand indication for the fullrange of rod travel. A change to the Westinghouse AcceptanceCriteria and SAR Change Package No. U2-019 were approved andan urgent change to the procedure incorporated the revisedAcceptance Criteria. All Acceptance Criteria for the final packagewere met.2-PAT-3.0, Attachment't,Testing Deferred from 2-PTl-062-03,- field work complete on 517116 with allAcceptance Criteria met.CR 1168487 was written to document alternate charging flow wasnot within anticipated range, however, it had no affect on the testacceptance.2-PAT-3.8, Rod Drop Time Measurement and Stationary GripperRelease Timing, field work complete on 5/11116. CR 1169659 waswritten for two rods failing a two sigma statistical evaluation. Threeadditional rod drops were performed and allAcceptance Criteriawas met. There was no Review Criteria for this test.46 5.1 Post Core Loading Precritical rest sequence (2-PAT-3.0) (continued)o 2-PAT-3.11, Adjustment of Steam Flow Transmitters at Minima!Flow, field work complete on 517116 with all Review Criteria met.There was no Acceptance Criteria associated with thisperformance.o 2-PAT-3.7, Reactor Coolant Flow Coastdown, field work completeon 5/8/16 with all criteria met. CR 1169224 was written todocument during removal of an instrument recorder a blown fusecaused alarms in the Main Control Room.o 2-PAT-5.1, Dynamic Automatic Steam Dump Control, field workcomplete on 5112116 with all criteria met after a volume boosteradjustment with CR 1170159 and retest on 2-FCV-1-108.3.0 Test ResultsAcceptance/Review Criteria were contained within the test sequenced bythis test, except for Attachment 1, Testing Deferred from 2-PTl-062-03.Attachment 1 required Acceptance Criteria were met as delineated below.Acceotance Criteria3.1 Sum of RCP seal injection flow s 40 gpm, (6-13 gpm for each RCp).A. 2-Fl-62-1A = 9.0 gpm (6.6-12.4 gpm)B. 2-Ft-62-14A= 9.2 gpm (6.6-12.4 gpm)C. 2-Ft-62-2tA= 9.3 gpm (6.6-12.4 gpm)D. 2-Ft-62-40A = 9.2 gpm (6.6-12.4 gpm)Total Seal lnjection Flow Rate in gpm. = A + B + C + D =9.0+ 9.2 + 9.3 + 9.2 = 36.7 gpm3.2 The differential pressure across the following component at thegiven flowrate:DescriptionUNIDFIow RateAP(GIean)ActualAPSeallnjectionFilter B2-FLTR-62-9616-40 gpms7 psid6.0 psid3.3 lndication light 2-Xl-62-93 in MCR illuminated when2-HIC-62-93B was in manual.41 5.1 Post Gore Loading Precritical Test Sequence (2-PAT.3.0) (continued)3.4 2-FM-62-93E prevented 2-FCV-62-93 from going fully closed toensure Sea! Water flow rate of 33.5 11.5 gpm (32-35 gpm).Flow was 32.2 gpm.Review CriteriaNone4.0 Problemst1l CR 1168487 was written on 2-PAT-3.0, Attachment 1, Step 27.Although not Acceptance Criteria, alternate charging header flowwas anticipated to be approximately 89-103 gpm. Actual flow was82.7 gpm. This information was forwarded to engineering forevaluation, however, it had no affect on the acceptance of this test.Engineering evaluation calculated the minimum requirement at theregenerative heat exchanger temperature to be 74 gpm. The 82.7gpm exceeds this amount. At the current conditions the test wasacceptable and met the design specified criteria.Additional problems encountered are addressed in the followingdiscussions of each test sequenced by 2-PAT-3.0.48 5.2 Gontrol Rod Drive Mechanism Timing and CERPI Initia! Calibration(2-PAr-3.1)Performance of this test was directed by 2-PAT-3.0, Post Core LoadingPre-criticalTest Sequence, during the period from 1121116to 1124116. The testwas performed in Mode 5 at a RCS temperature of approximately 175'F.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the functionality of each CRDM for shutdown and controlrods in Mode 5 by:1.1.1 Verify each rod control system slave cycler provides itsassociated power cabinet with the appropriate commandsignal to obtain proper sequence timing of current suppliedto the CRDM coils.1.1.2 Verify CRDM coil current amplitudes are within acceptableranges.1.1.3 Verify the functionality of each shutdown and control roddrive mechanism.1.1.4 Verify manual mode stepping rate for shutdown and controlrods are within acceptable ranges.1.2 Verify the control bank overlap function in manual with minimaloverlap.1.3 Perform the initial calibration of the RPI in accordance with vendorproced u re WNA-TP-02576-WBT, C E RP I Calibration Proced ure.1.4 Partially satisfu the requirements of UFSAR Table 14.2-2, Sheet 7,Control Rod Drive Mechanism Timing Test Summary, and fullysatisfo it for Mode 5.2.0 Test MethodsThe CRDM functionality was verified by stepping ouUin each rod bank byapproximately 10 steps in individual bank select mode. CRDM currenttiming and amplitude measurements were taken during rod motion.Twelve of the CRDM amplitudes were outside the upper AcceptanceCriteria of the lift coil reduced current, however, the amplitudes wereevaluated as acceptable by Westinghouse.49 5.2 Gontrol Rod Drive Mechanism Timing and CERPI lnitial Galibration(2-PAT-3.1 ) (continued)The bank overlap circuitry was verified at minimal settings. Minimalsettings were set by adjusting the bank overlap thumbwheel switches suchthat control bank tip-to-tip distance was 15 steps and the all-rods-outposition was 25 steps withdrawn for each control bank. The bank overlapcircuitry functioned as designed and no issues were encountered. Notethat the bank overlap circuitry was also exercised during the performanceof the initial RPI calibration with the all-rods-out position set to 230 stepswithdrawn. The bank overlap circuitry functioned as designed with noissues.The initial RPI calibration was performed. First the bank zero adjustmentswere performed with all rods fully inserted. Next all shutdown and controlrods were withdrawn to the full out position of 230 steps withdrawn. Theshutdown banks were withdrawn first in individua! bank select and thecontrol rods were withdrawn in bank overlap. The bank position spancalibration and temperature null adjustments were performed with the rodsfully withdrawn. Next all control and shutdown rods were inserted tospecific demanded positions and data for each rod was obtained. Lastly,linearization adjustments were calculated based on the recorded data.The initia! RPI calibration was completed when the new linearizationadjustments were uploaded to train A and B of the RPI system. Note thatprior to the linearization adjustment and during the insertion of ControlBank B, the K14 and P6 rods had a 13 step rod-to-rod deviation whileinserted between CBB demanded positions of 170 steps withdrawn to 126steps withdrawn. The RPI system was not yet calibrated, therefore, theinitial calibration corrected the issue. Also note that 2-PAT-3.8 wasperformed following the initial RPI calibration and all rods were pulled toapproximately 50 steps withdrawn. All rods were within t2 steps of thedemanded position.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 Contro! Rod Drive Mechanism Timing3.1.1 Current Order TimingThe times at which the lift, movable, and stationary currentorders change, after the start of rod motion, are within10 msec. of the expected times during rod withdrawal andinsertion operations.50 5.2 Control Rod Drive Mechanism Timing and CERPI lnitial Calibration(2-PAT-3.1 ) (continued)Each CRDM current order timing was reviewed and allcurrent order timings were within 10 msec. of the expectedtimes.3.1.2 Coil Current AmplitudesStationary, movable, and lift currents are regulated bycircuitry internal to each power cabinet. The reduced andfull current nominal values are not critical, cannot beadjusted, but could be an indication of a regulation failure.Measured values outside the nominal ranges should beevaluated and documented by the system engineer.Lift Coil - FullNominal 4A amperes(35 to 47 .2 amperes)(equivalent to 438 to 590mVdc measured across a0.0125 ohm resistor)Lift Coil - ReducedNominal 16 amperes(13 to 19.7 amperes)(equivalent to 163 to 246mVdc measured across a0.0125 ohm resistor)Movable Gripper Coil -FullNominal 8 amperes(7 to 9.2 amperes)(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)Stationary Gripper CoilFullNominal 8 amperes(7 to 9.2 amperes)(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)Stationary Gripper Coil -ReducedNominal 4.4 amperes(3.8 to 4.8 amperes)(equivalent to 238 to 300mVdc measured across a0.0625 ohm resistor)51 5.2 Control Rod Drive Mechanism Timing and CERPI Initial Galibration(2-PAT-3. 1 ) (continued)Each current amplitude recorded in the test package werereviewed. All current amplitudes were within the AcceptanceCriteria with the exception of twelve CRDMs for the lift coilreduced current. The twelve lift coi! reduced currentamplitudes were evaluated and determined to be acceptablefor the designed operation of the rod control system.CR 1128950 was closed.3.1.3 Rod Withdrawal SpeedShutdown bank withdrawal speed nominal value is 64 (62 to66) steps per minute.Control bank withdrawal speed nominal value is 48(46 to 50) steps per minute.The Shutdown and Control banks withdrawal speeds mettheir Acceptance Criteria and were recorded as63.9 steps/min and 48.0 steps/min respectively.3.1.4 Rod Drive Mechanism OperabilityShutdown rod drive mechanisms operate with no indicationsof problems during the withdrawal and insertion stepping.Control rod drive mechanisms operate with no indications ofproblems during the withdrawal and insertion stepping.Each CRDM current trace was reviewed. Alltraces operatednormally and no abnormalities, such as movable/stationarygripper dragging or rod misstepping, were identified.3.2 Control Bank Overlap Demonstration3.2.1 The control rod bank overlap circuitry functions properlyduring the sequential withdrawal and insertion of ControlBanks in MANUAL mode.The control bank overlap circuitry functioned as designed.52 5.2 Contro! Rod Drive Mechanism Timing and GERPI lnitial Calibration(2-PAT-3. 1 ) (continued )3.2.2 The MCR rod speed demand display functions properly andindicates the rod stepping rate was within the range of 46 to50 steps/minute for Control Banks in Manual mode.The MCR rod speed demand display functioned as designedat 48.0 steps/min.3.2.3 The MCR group step counters function properly to indicategroup position and direction of rod motion during rodwithdrawal and insertion operations.The MCR group step counters functioned as designed.3.2.4 The MCR RPI functions properly to indicate individua! roddirection of motion during rod withdrawal and insertionoperations.The MCR RPls indicated the proper direction of motionduring rod withdrawal and insertion operations.3.2.5 The MCR rod direction indicator lights function properly toindicate the rod movement status and direction of rod motionduring rod withdrawal and insertion operations.The MCR rod direction indicator lights functioned asdesigned.Review CriteriaNone4.0 Problemst1] During the performance of the CRDM timing and amplitudemeasurements, the 1AC power cabinet- stationary group A coilamplitudes were lower than the expected value. WO 111522924was performed to inspect and reform backplane connector and cardedge pins for the 1AC power cabinet - stationary group A firing,regulation, and phase control cards. The issue was corrected andtesting continued.s3 5.2Control Rod Drive Mechanism Timing and CERPI lnitial Calibration(2-PAT-3. I ) (conti nued )l2lStep 6.3.3[7.1], WNA-TP-02576-WBT, Revision 2, Step 2.4.2.5,theall-rods-out position was 230 steps withdrawn, however, thecompensated position in the software was hardcoded to 231 stepswithdrawn. The performance of Step 2.4.2.5 was not impactedbecause it listed 230 steps t1 step. CR 1128373 was written andconcluded no changes to 2-PAT-3.1 were required.CR 1128918: Step 6.3.4[62], WNA-TP-02576-WBT, Revision 2,Appendix A.1 and A.2 forms were used for the linearizationadjustments. The "X Table C1" column values were not Watts BarU2 specific values. The Watts Bar U2 plant specific "X Table C1"values were used for the linearization adjustments. CR 1128918was written. Resolution was for Westinghouse to revise WNA-TP-02576-WBT.CR 1128950 Two rod amplitude measurements (F14 and D08)failed the procedure Acceptance Criteria and do not meet theWestinghouse expanded acceptance criteria in WBT-D-il20.Westinghouse has provided a letter (WBT-D-5604(3.8))documenting their evaluation and the acceptability for the 2 rodlocations that exceeded the Acceptance Criteria of 20 amps.Note that additional rod measurements were outside of theReduced Lift Current procedure Acceptance Criteria of 19.7 amps,however, WBT-D-5420 has been issued by Westinghouse thatstates reduce lift currents up to 20.0 amps are acceptable.2-PAT-3.1, Rev. 2, allows for evaluation of the currents outside ofthe Acceptance Criteria for successful completion of 2-PAT-3.1.Current orders outside of the Acceptance Criteria were evaluatedand deemed acceptable per Westinghouse. Also additionalmeasurements were obtained in Mode 3.During the performance of 2-TRl-85-1, Reactivity Control SystemsMovable ControlAssemblies (Modes 3, 4 and 5), rods common tothe 2BD power cabinet would not withdraw. Troubleshootingdetermined an issue with the 2BD movable gripper currentamplitudes. The firing and regulation cards for the 2BD movablegrippers were replaced with spares and the issue was corrected.No problems with this power cabinet occurred during 2-PAT-3.1testing.t3ll4It5I54 5.2Control Rod Drive Mechanism Timing and CERPI Initial Calibration(2-PAT-3.1 ) (continued)t6IStep 6.1l12lof Appendix E, rods common to the 1AC powercabinet - stationary group A did not have the expected reducedstationary gripper currents when the CRDM-DAQ was firstconnected. Under CR 1126661 and WO 117522924, the stationarygroup A firing, regulation, and phase cards were removed and bothbackplane connector and card edge pins were reformed. The issuewas corrected and testing was completed. No other problems withthis power cabinet occurred during 2-PAT-3.1 testing.During the performance of 2-TRl-85-1 with the 2-RBSS in the SBCand SBD positions, the CERPI monitor indicated 72 steps/min. Theactual speed of the SBC and SBD groups is approximately 64steps/min and is set at the SCD power cabinet. The indication didnot invalidate the performance of this test. CR 1126783 waswritten and closed to WO 115966328.Step 6.3.4[3], during SBD insertion, the SBC demand position onboth the ICS and RPI monitors followed the SBD demand position.The problem was due to communication issues between the rodcontrol system and the ICS. Testing continued because this issuedid not invalidate the performance of this test. CR 1128318 waswritten, and Post lssuance Change (PlC) 66181 was issued andimplemented by WO 117546244.Step 4.3[10]A, the 28 MG Set failed to sync in parallel with the 2AMG Set. The issue did not invalidate the performance of 2-PAT-3.1because only one MG set was required for the performance of thistest. CR 1126798 was written, and closed to WO 117531764.17It8l55 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-PAr-3.2)Performance of this test was directed by 2-PAT-3.0, Post Core LoadingPre-critical Test Sequence, during the period from 4115116to 514116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the pressure response to the opening of both normalPressurizer Spray Valves was within the allowable range specifiedby NSSS performance curyes.1.2 Verify the Pressurizer Bypass Spray Valves were throttled to anoptimum position such that during steady state operation:1.2.1 Spray line temperature was high enough to prevent the PZRSPRAY TEMP LO alarm from actuating.1.2.2 The equilibrium temperature for each spray line was greaterthan or equal to 540F.1.2.3 Pressurizer control bank heaters can maintain RCS pressureabove 2220 psig without backup heater operation.1.2.4 Surge line temperature was high enough to prevent the PZRSURGE LINE TEMP LO alarm from actuating.1.3 Verify the PZR SPRAY TEMP LO alarm would actuate ondecreasing spray line temperature of approximately 530F.1.4 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 13,Pressurizer Spray Capability And Continuous Spray Flow Setting TestSummary.55 5.3Pressurizer Spray Capability and Continuous Spray FIow Setring(2-P AT -3.2) (conti nued )2.0 Test Methods3.0This test established the optimal throttle positions for the PressurizerSpray Manual Bypass Valves, and also ensured the effectiveness of thenormal pressurizer spray by initiating full spray to reduce RCS pressure byapproximately 250 psi and compared the time to reduce pressure withWestinghouse performance curyes. During the performance of Section6.2, the validated RCS pressure DCS computer point being used tomonitor the depressurization of the RCS stopped updating at acceptablerate. Because of this the RCS narrow range indicators on the controlboard were used to determine when RCS pressure reached the triggervalue of 2000 psig. Subsequent review also determined that this pointdeviated further from the actual RCS pressure after it was no longer beingmonitored. This did not affect the ability to meet the Acceptance Criteriaof the test as ICS computer points were collected for use to analyzecompliance with Acceptance Criteria for the spray capability test. CR1168255 documents this issue.The spray line temperature low alarm was unable to be verified asintended during the performance of this test due to slight leakage pasteither the spray line FCV's or the spray bypass manual valves. Thisoccurred on both loops 1 and 2. This condition prevented meeting theReview Criteria associated with the spray line temperatures. Theoperation of the spray line temperature switches for each loop weresubsequently verified to be operating correctly by utilizing trend data fromthe plant computer. CRs 1 160969 and 1 161382 document this issue.Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceptance Criteria3.1Pressurizer pressure response to opening both Normal PressurizerSpray Valves is within the allowable range specified by NSSSperformance curves.The pressurizer spray response data was within the allottedresponse time as depicted on Figure 5.3-1.51 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-P AT -3.2) (conti nued )Review Criteria3.2 Pressurizer Manual Spray Bypass Valves 2-BW-68-552 and2-BW-68-555 are throttled to an optimum position during steady-state operation.All procedural criteria was met:1.Spray line equilibrium temperature is high enough to preventAnnunciator 2-><A-55-5A-89E, PZR SPRAY TEMP LO fromactuating.2-XA-55-5A-89E, PZR SPRAY TEMP LO, did not actuate.Equilibrium temperature for each spray line is greater than orequal to 550'F:Loop 1 Spray Line Temperature (lCS PID T0484A)Loop 2 Spray Line Temperature (lCS PID T0483A)Spray Iine equilibrium temperatures were > 550oF.Pressurazer control bank heaters can maintain RCS pressureabove 2220 psig without any Backup Heater operation.Backup Heater operation was not required to maintainpressure.Surge line equilibrium temperature is high enough to preventAnnunciator Alarm 2-XA-55-5A/89D, PZR SURGE LINETEMP LO, from actuating.2-KA-55-5A/89D, PzR SURGE LINE TEMP Lo, did notactuate.2.3.4.3.3Annunciator Alarm 2-XA-55-5A/89E, PZR SPRAY TEMP LOactuates on decreasing spray line temperature of approximately530'F (525"F to 535'F).Loop 1 spray line temperature would not decrease sufficiently toallow the low spray line temperature alarm to actuate. See CR1 160969.Loop 2 spray line temperature would not decrease sufficiently toallow the low spray line temperature alarm to actuate. See CR1161382.58 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-P AT -3.2) (conti nued )4.0 Problemst1] CR 1160969 was written because the loop 1 spray line temperaturewould not decrease sufficiently to allow the Iow spray linetemperature alarm to actuate. Westinghouse evaluation wasobtained which concluded that 2-PAT-3.2 met operability anddesign requirements for the pressurizer spray system.l2l CR 1161382 was written because the loop 2 spray line temperaturewould not decrease sufficiently to allow the low spray linetemperature alarm to actuate. Westinghouse evaluation wasobtained which concluded that 2-PAT-3.2 met operability anddesign requirements for the pressurizer spray system.13] CR 1161789 was written due to the master pressure controlleroutput being less than the desired range while setting the loop 2spray line bypass valve. The controller output was 37 percent whilethe desired range was a minimum of 46 percent. Performance ofSection 6.1.3 which performed the final setting of the spray bypassvalves, allowed the output of the master controller to be placed inthe desired range of the procedure.t4] CR 1168255 was written to document the issues experienced withthe ICS computer point for Validated Pressurizer Pressure(DCS0426) during the performance of Section 6.2.59 5.3 Pressurizer Spray Capability and Continuous Spray Flow Setting(2-PAT*.2) (continued)FIGURE 5.3.1Pressurizer Spray Response2-PAT-3.2 Pressu rizer Spray Ca pa bility-50A.IlaeYoLa -10oobrLohlE -1soUl6oLc*-200-2502060 80Time (seconds!lower - - upper100- lcs DataL20140(-nQm (- -\\\\\L1\\\t:Il-ItrIFNIFII-'-1\\\\l-LT\\\\\IrLtrN\ri\-i\\\!L-vt:\:--- -; -\\rjj1il}}i r\- ---l. -- -{i--i---:il+\i\,,\\!L ----- i\ll\r*t60 5.4Rod Gontrol and Rod Position lndication (CERPI) (2-PAT-3.4)This test was performed in Mode 3 at NOTP as directed by 2-PAT-3.0, Post CoreLoading Precritical Test Sequence. lt performed the initial hot calibration of theComputer Enhanced Rod Position lndication (CERPI) system and functionaltesting of the Rod Control System. The performance of Section 6.0 of this testwas commenced on 05/05/16 and was completed on 05113116.1.0Test ObiectivesThe objectives of this test were to:1.2Perform the lnitial Hot Calibration of the Computer Enhanced RodPosition lndications (CERPI) system.Verify the Computer Enhanced Rod Position lndication system(CERPI) performs required indication function satisfactorily for eachshutdown and control rod over their entire range of travel and toverifo the rod position indication system alarm functions operateproperly. (UFSAR Table 14.2-2, Sheet 8, Rod Position lndicationSystem Test Summary).1.12.01.3 Demonstrate that the rod control system satisfactorily performs therequired control and indication functions, as required by UFSARTable 14.2-2, Sheet 10, Rod Control System Test Summary.Test MethodsThe rod position indication system completed the initial hot calibrationusing vendor instructions and 2-SI-85-3, Calibration of ComputerEnhanced Rod lndication Channels and Full Range Verification. TheCERPI system operated over the full length of travel and can operatewithout actuating rod-to-rod and rodto-bank deviation alarms by makingadjustments to the CERPI tunable parameters. This was consistent withvendor and Unit 1 operating experience. Therefore, the vendor CERPIAcceptance Criteria was revised and Urgent Change (UC) 2 wasprocessed for this test procedure to veriff that each rod indicates rodmotion consistent with the group demand over the full length of trave!.The Rod Control System and CERPI functionaltesting included controlsand indications. The functional testing included rod-to-bank and rod-to-rod deviation alarms, the C-11 annunciator, lntegrated Computer System(lCS) generated alarms, rod bottom bistables, rod bottom bypassbistables, rod control urgent and non-urgent alarms, main control roomdisplays, rod insertion limits, and control rod bank overlap circuitry.This test also documented 5 complete rod excursions (i.e., fullwithdrawaland insertions) of all shutdown and control rods per CR 234483. All 5excursions were successfully completed.6L 5.4 Rod Control and Rod Position lndication (CERPI) (2-PAT-3.4) (continued)3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceptance Criteria3.1 CERPI Calibration3.1.1 WNA-TP-02576-WBT, Watts Bar 2 ARPI SystemUpgrade CERPI Calibration Procedure, Section 2.6,lNlTlAL HOT CALIBRATION, was successfullycompleted and the linearity is within +12 steps at thesteps checked in the procedure.The initial hot calibration was successfully completed andthe linearity was demonstrated to be 112 steps at thesteps checked in the procedure.3.1.2 2-Sl-85-3, Calibration of Computer Enhanced Rodlndication Channels and Full Range Verification, wassuccessfully completed.2-S!-85-3 was successfully completed.3.2 Rod Control and lndication3.2.1 2.XA-55-4A.64F, C11 BANK D AUTO WITHDMWALBLOCKED, alarm window in control room was LIT whenControl Bank D was withdrawn above 220 steps.2-XA-55-4A-64F, C-11 alarm window annunciated at220steps.3.2.2 CERPI monitor alarm for:(1) Rod to Rod Deviation between two rods in a bankThe rod to rod deviation CERPI alarm was at12 steps.(2) Rod to Bank Deviation corresponding to > 12steps.The Rod to Bank Deviation alarm was at 12 steps.52 5.4Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.32-><A-55-4B-83D, PLANT COMPUTER GENERATEDALARM (SEE ICS), alarm window in control room wasLIT when ICS Computer detects the following conditions:(1) Deviation between rod position indicator for a rodand corresponding bank demand positi on >12steps.The deviation between ICS rod position and bankdemand was 12 steps.(2) Deviation between rod position indicator for a rodand average rod position >12 steps.The deviation between ICS rod position andaverage rod position was 13 steps.Rod bottom bistable indicators actuate at correct setpointsetting (below 20 steps withdrawn) as indicated by RPIindicators and rod bottom indicators on CERPI on 2-M-4.Each rod bottom bistable for all rods actuated below 20steps (20 to 19 steps) withdrawn.CERPI bypass indication for Control Banks B, C, and Dactuate at correct setpoint setting (below 35 stepswithdrawn) as indicated by CERPI Bank Demand digitaldisplay.CERPI bypass indication for Control Banks B, C, and Dactuated at 31 steps withdrawn.An Urgent Failure induced in a Power Cabinet and LogicCabinet caused local urgent failure alarm indicator lampat the respective cabinet and 2-Xl\-55-48-86A,CONTROL ROD URGENT FAILURE, ohnunciatorwindow to light.The Power Cabinet and Logic Cabinet local urgent failurealarm indicator lamp and 2-><A-55-48-86A annunciatorwindow functioned as designed and met all applicableAcceptance Criteria.3.2.43.2.53.2.663 5.4 Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.7A Non-Urgent Failure induced in each of the PowerCabinets and Logic Cabinet causes local non-urgentfailure alarm indicator lamp at respective cabinets and2-XI\-55.48-868, CONTROL ROD NON-URGENTFAILURE, ?rnunciator window to light.The Power Cabinet and Logic Cabinet local non-urgentfailure alarm indicator lamp and 2-)<A-55-48-868annunciator windows functioned as designed and met allapplicable Acceptance Criteria.A failure induced in the CERPI racks caused2-><A-55-4B-86C, CERPI TROUBLE, ?rnunciator windowto be LIT or REFLASH.The 2-><A-55-4B-86C annunciator window functioned asdesigned and met all applicable Acceptance Criteria.For shutdown and control rod banks having two groups,the group step counter for group 1 shall be 0 or 1 stepabove grou p 2 step counter over their full length of travel(i.e., 231 steps).The group step counters for group 1 were 0 or 1 stepabove grou p 2 step counters over their full length oftravel for shutdown and control rod banks having twogroups. All applicable Acceptance Criteria were met.2-><A-55-4B-87D, RODS AT BOTTOM, ornunciatorwindow was lit when one or more rods in CBA wereinserted in the normal sequence. Also, RODS ATBOTTOM, ornunciator window is not lit when controlrods were inserted or withdrawn in their normalsequence.2-xl\-55-4B -87 D annunciator window functioned asdesigned and met all applicable Acceptance Criteria.Each RPI indicated rod motion consistent with the groupdemand indication for the full range of rod travel.The RPI indicators for each rod indicated rod motionconsistent with the group demand indication for the fullrange of rod travel.3,2.93.2.93.2.103.2.1164 5.4 Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.12 The MCR rod speed demand display functions properlyand indicated the rod stepping rate (ROD SPEED) waswithin the range of 62 to 66 steps/minute for ShutdownBanks A and B in bank select mode.The MCR rod speed demand display indicated 64steps/min for both Shutdown Banks A and B.3.2.'13 The MCR rod speed demand display functions properlyand indicates the rod stepping rate (ROD SPEED) waswithin the range of 46 to 50 steps/minute for ControlBanks in bank select and in MANUAL mode.The MCR rod speed demand display indicated 48steps/min for all Control Banks in bank select and inManual mode.3.2.14 The MCR rod direction indicator lights functioned properlyto indicate the rod movement status and direction of rodmotion during rod withdrawal and insertion operations.The MCR rod direction indicator lights functioned asdesigned and met all applicable Acceptance Criteria.3.2.15 The MCR group step counters functioned properly toindicate group position and direction of rod motion duringrod withdrawal and insertion operations.The MCR group step counters as designed during bothwithdrawal and insertion.3.2.16 The MCR Computer Enhanced Rod Position lndicators(CERP!) function properly to indicate individual rodposition and direction of motion during rod withdrawaland insertion operations.The CERPI indicators functioned as designed to indicateindividual rod position and direction of motion duringwithdrawa! and insertion. All applicable AcceptanceCriteria were met.3.2.17 The rod insertion limits LO-LO upper limit was set to 211steps (Control Bank A)The rod insertion limits LO-LO upper limit was found tobe set at2ll steps (Control Bank A).65 5.4 Rod Gontro! and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)3.2.18 The rod insertion limits LO alarm actuated below 10steps above the insertion limit for any control bank.The rod insertion limits LO alarm actuated at 10 stepsabove the insertion limit for any control bank.3.2.19 The rod insertion limits LO-LO alarm actuated below 0steps above the insertion limit for any control bank.The rod insertion limits LO-LO alarm actuated at 0 stepsabove the insertion limit for any control bank.3.2.20 The control rod bank overlap circuitry functioned properlyduring the sequential withdrawal and insertion of Contro!Banks in MANUAL mode.The contro! bank overlap circuitry functioned as designedduring the sequential withdrawal and insertion of ControlBanks in MANUAL mode.3.2.21 Each RPI indicated rod motion consistent with the groupdemand indication for the ful! range of rod travel.The RPI indictors for each rod indicated rod motionconsistent with the group demand indication for the fullrange of rod travel.3.2.22 All rods were fully withdrawn and inserted five times.The test exercised all rods to fully withdrawn and fullyinserted five times.Review CriteriaNone4.0 Problernst1l CR 1168845 - Steps 6.4.3[20] and 6.4.4.1201, Acceptance Criteriafor RPI indication agreeing within t12 steps was not satisfied forShutdown Banks C and D for the full length of travel. ThisAcceptance Criteria was later changed by Urgent Change 2. SeeUC-2 description in this test report for further information.55 5.4 Rod Control and Rod Position lndication (GERPI) (2-PAT-3.4) (continued)CR 1168881 - Section 6.4.5, Rod position indications for rods inControl Banks A and B were at 12 steps from the demand positionat some positions over the full length of travel. Although not afailure in the Acceptance Criteria of Step 6.4.5[20] of within 12steps of the demand position, the rod-to-rod deviation alarms wereactuated at certain positions over the length of travel. This CRanticipated not meeting Steps 6.10[2.15]and 6.10[2.16]. Theacceptance of rod-to-rod deviations was later changed by UrgentChange 2. See UC-2 description in this test report for furtherinformation.CR 1169602 - Steps 6.10[2.15] and 6.10[2.16] Acceptance Criteriaof no rodto-rod and rodto-bank deviation alarms was not met forall 57 rods over the full length of travel. These steps were laterchanged by Urgent Change 2. See UC-2 description in this testreport for further information.CR 1 1692'17 - Step 6.4.6[1]C verified the C-1 1 Bank D AutoWithdrawal Block annunciator cleared during Control Bank Dinsertion between 219 and 214 steps on the step counters. Therecorded step counter position for Control Bank D was 212 stepswhen the C-11 annunciator cleared. Step 6.4.6[1]C is notAcceptance Criteria and the value at which the annunciator clearedis reasonable. Based on the CR evaluation, the vendor manualdescription of operation demonstrates (as well as the test engineerevaluation) there was not a problem encountered during 2-PAT-3.4performance with Control Rod Bank D permissive C-11. No FurtherAction Required.CR 1169282 - Step 6.7.2l7lcould not be performed as writtenbecause the performance of Step 6.7.2[6] cleared the urgent alarmupon seating the 4104 card. The card interlock is the only urgentalarm in the rod control system that clears upon restoring thesystem configuration. Step 6.7.2171was to verify rod motion wouldnot occur with a standing urgent alarm. These steps were notAcceptance Criteria and had no impact on successful completion ofthis test. Additionally, CR 1171247 was created to address andprovide justification associated with not performing the steps citedin CR 1169282.CR 1171254 - This CR was written to document certain steps andportions of sections which were repeated during testingperformance based on engineering judgment. These additionalperformances were used as a means to perform additionalcalibration of CERPI as specified in the Westinghouse CERPIcalibration procedure or as a means to restore from current testingl2l13Il4lt5It6I51 5.4 Rod Gontrol and Rod Position lndication (CERPI) (2-PAT-3.4) (continued)conditions and then later return. Section 6.1, Steps 6.1[1] through6.1111.19lwere repeated and Section 6.6, Steps 6.6[1]through6.6[9.5]were repeated one or more times. Section 6.6, Steps6.6[48] through 6.6[54] were repeated.This CR was created for documentation purposes only and has noimpact on the actual test results or verification of AcceptanceCriteria.l7l CR 1168538 - During performance of Section 6.1, Shutdown BankA was inserted to 118 steps demand position for CERPI hotcalibration. M14 indication drifted excessively for a couple of hoursuntil a stable rod position indication was reached.The PAT team and Westinghouse reviewed the M14 coil resistancevalues identified that the M14 drift was associated with a largetransient in coil resistance (i.e., change in coil stack temperature)and a larger value for the T_GAIN parameter for M14.Drift is a phenomena that occurs for ARPI/CERP! indicationsystems due to the analog coil stacks and their associatedtemperature dynamics and does not represent an actual change inrod position. This does not represent a deficiency in the design.Adjustments to the CERPltunable parameters were made tominimize drift. Therefore, this CR did not impact the successfulcompletion of this test procedure.l8l CR 1168899 - Step 6.4.5[8], the 2-)G-55-4B-87D, RODS ATBOTTOM, annunciator cleared as expected. However, the bell inthe MCR did not alarm. CR was closed to previously identifiedwork.tgl UC-1 was written to ensure that group step counters for ShutdownBank A, group 1, step counters displayed 56 steps prior to withdrawof Rod D-02 in Step 6.6 [40]. This allowed the group step counterto be updated to match the current rod position for rod that wascurrently capability of motion.t10] UC-2 was processed to revise Acceptance Criteria 5.2K and 5.2Uand related steps based on updated criteria provided byWestinghouse in WBT-D-5666, CERPI Acceptance Criteria,Revision 1, and a revision to UFSAR Table 14.2-2, Sheet 8.The update criteria stated "Each RPI indicates rod motionconsistent with the group demand indication for the full range of rodtravel."68 5.5Reactor Coolant Flow Goastdown (2-PAT-3.7)This test was performed as directed by 2-PAT-3.0, Post Core Loading PrecriticalTest Sequence, in Mode 3 at normaloperating temperature and pressure. Thetest was started and field work completed on 518116.1.0Test ObiectivesThe objectives of this test were to:1.1 Measure the rate at which reactor coolant flow changes subsequentto a simultaneous trip of all four reactor coolant pumps. Themeasured Flow Coastdown Time Constant is determined from theflow versus time data and compared to the Design Flow CoastdownTime Constant.1.2Measure the delay time associated with the low flow reactor tripand compare it to that value assumed in the accident analysis.Record the RCP Motor voltage decay during the transient forinformation only.1.4 This test satisfied the requirements of UFSAR Table 14.2-2,Sheet 15, Reactor Coolant FIow Coastdown Test Summary.2.0 Test Methods3.0All four reactor coolant pumps were simultaneously tripped, causing thereactor trip breakers to open on Low RCS Flow. Measurements weremade by recording reactor coolant loop elbow tap differentia! pressures(d/p), RCS low flow bistable state, reactor trip breaker position, reactorcoolant pump breaker position and reactor coolant pump motor voltagedecay data. Also recorded for information was the time of theundervoltage relay and associated time delay timer in the RCPsundervoltage circuit.Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 The Acceptance Criterion for core flow coastdown following thesimultaneous trip of the four reactor coolant pumps from full flowconditions, was that the measured flow coastdown time constant(TAUru) was greater than (>) design flow coastdown time constant(TAUD) of 11.72 seconds.Results indicated TAUu = 12.762 seconds1.369 5.5 Reactor Coolant Flow Coastdown (2-PAT-3.7) (continued)3.2 Acceptance criterion for the Total Low Flow Trip Delay Time is lessthan (<) 1.2 seconds.Results indicated Low Flow Trip Delay Time, Tr_r = 0.994 seconds3.3 Acceptance criterion for simultaneous trip of four reactor coolantpumps was that allfour pumps trip within (<) 100 msec. of eachother.Allfour pumps tripped within 20 msec.Review Criteria3.4 Review Criterion for coastdown flow data quality is that data from atleast 2 out of 3 flow transmitters in each RCS loop falls withinChauvenet's Criterion.All data points from all RCS flow transmitters fell within Chauvent'sCriterion.4.0 Problemst1] CR 1169224 - During performance of 2-PAT-3.7 post performanceactivities, Step 7.0[5], removing the test recorders from the Auxlnstrument Room, 8 out of 12 RCS low flow trip status lights lit inthe control room. Post event investigation revealed bistable fusesassociated with the affected flow loops (2-LPF-68-6A, 2-LPF-08-6B,2-L p F-69-29 A, 2-LpF-68-29 B, 2-Lp F-68-29D, 2-LPF-6848B, 2-LPF-68-71A, 2-LPF-68-71 B) blew during recorder disconnectionbecause of human performance issues.Since this was a post performance activity and all data wasrecorded during Section 6.0, there was no impact to the test resultsof 2-PAT-3.7.10 5.6 Rod Drop Time Measurement and stationary Gripper Release Timing(2-PAr-3.8)Portions of this test were performed in Mode 5 and again in Mode 3 as directedby 2-PAT-3.0, Post Core Loading Precritical Test Sequence.ln Mode 3 this test was performed in conjunction with the norma! Surveillancelnstruction 2-Sl-85-10, Rod Drop Time Measurement Using CERPI Rod DropTest Computer, to calculate the standard deviation of the rod drops and to directrequired additional rod drops for CR corrective actions and potential two-sigmadeviations.Prerequisites were started on 1122116 and the test was field work completed on1124116 for the Mode 5 performance. The Mode 3 performance was started on5111116 and completed on 5111116.1.0Test ObiectivesThe objectives of this test were to:1.3ln Mode 5, Section 6.2 of this Power Ascension Test (PAT) partiallywithdrew al! shutdown and control rods and demonstrated that allCRDMs unlatch and all rods fully insert into the core when thereactor trip breakers were opened.ln Mode 5, measured the Stationary Gripper Release times foreach control and shutdown rod.ln Mode 3, at Hot Standby conditions with full Reactor CoolantSystem (RCS) flow, measured the rod drop time and stationarygripper release time for each control and shutdown rod. WBN Unit2 Technical Specifications require rod drop time measurements,therefore, the normal Surveillance lnstruction 2-Sl-85-10, Rod DropTime Measurement Using CERPI Rod Drop Test Computer wasutilized.Meet the Mode 3 testing as required by UFSAR Table14.2-2, Sheet 9, Rod Drop Time Measurement And StationaryGripper Release Timing Test Summaryo Measure the stationary gripper release time for each controland shutdown rod. This measurement was performed inMode 5 and then repeated in Mode 3.1.11.21.41L 5.6 Rod Drop Time Measurement and Stationary Gripper Release Timing(2-PAT-3.8) (continued)1.5. Evaluate the data from rod drop time testing in the area ofthe dash pot entry looking for proper performance of thedecelerating devices (i.e. dashpots). This evaluation wasperformed in Mode 3 with the data collected during theperformance of 2-Sl-85-10, Rod Drop Time MeasurementUsing CERPI Rod Drop Test Computer.. Evaluate all 57 rod drop times in Mode 3 with the datacollected during the performance of 2-SI-85-10, Rod DropTime Measurement Using CERPI Rod Drop Test Computer.Ensure that four rod drops were performed in Mode 3 as requiredby corrective actions from Condition Repoft234483 action 003related to INPO SER 1-10.2.0Test MethodsThis PAT was written to supplement the norma! operating surveillance2-Sl-85-10, Rod Drop Time Measurement Using CERPI Rod Drop TestComputer and evaluated the rod drop time data.All rod drop times were used to calculate the standard deviation of the roddrop times. Two-sigma limits (i.e. plus or minus two times the standarddeviation) were used to evaluate drop times of the 57 rods. Those droptimes that were outside of the two-sigma limits were re-measured 3 (ormore) times and evaluated for consistency (i.e. within 50 milliseconds).Retesting the rods that fell outside of the two-sigma limits an additional 3(or more) times provided reasonable assurance of their properperformance during subsequent plant operations.This PAT measured the Stationary Gripper Release Time for each controland shutdown rod. The Stationary Gripper Release Time is a combinationof a Trip Signal Delay Time (i.e. Delay between Reactor Trip Breakeropening and the trip signal to the RDTC) and the delay between powerinterruption (i.e. trip signal to the RDTC) and the rod's initiation of its freefal!. The PAT evaluated the traces from the RDTC looking for a delay ofeach rod's initiation of free fall the RDTC's trip signal. The PAT alsomeasured the "Trip Signal Delay Time" while in Mode 5.This PAT also evaluated the traces from the rod drops in the area of thedecelerating devices (i.e. dash pots) entry looking for proper performanceof the dash pots.12 5.6 Rod Drop Time Measurement and Stationary Gripper Release Timing(2-PAT-3.8) (conti nued)3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 Each CRDM unlatches upon opening the Reactor Trip Breakers.Testing confirmed that each CRDM unlatched upon opening theReactor Trip Breakers in Mode 5 and Mode 3.3.2 The rod drop times for all shutdown and control rods, dropped fromthe fully withdrawn position, are within the limits specified in theTechn ical Specifi cations.This Acceptance Criteria was successfully met by performance of2-Sl-85-10, Rod Drop Time Measurement Using Rod Drop TestComputer.3.3 Rod drop time evaluations against the Two-Sigma statistical limitscalculated resulted in either criteria below being applicable:o Rod drop time was within the bounds of the lower and upperTwo-Sigma statistical limits.o Rods with rod drop times that fell outside of the bounds ofthe lower and upper Two-Sigma statistical limits have beendropped greater than or equal to 3 additional times. Theresults of the rod drop times were consistent (i.e. within aband of 50 milliseconds or less) and continue to meetTechnical Specification criteria specified in2-Sl-85-10, Rod Drop Time Measurement Using CERPI RodDrop Test Computer.During the Mode 3 performance, Rods D-2 and M-14 failed to meeta 2 sigma statistical limit for the first rod drop. Three additiona! roddrops were successfully performed. (See CR 1169659)3.4 The stationary gripper release time for all rods was <150 msec.This was the requirement in Unit 2 UFSAR Chapter 4.During the Mode 5 performance the stationary gripper release timewas conservatively determined to be 45 msec. Mode 3 releasetime was 50 msec.13 5.6 Rod Drop Time Measurement and stationary Gripper Release Timing(2-PAT-3.8) (conti nued)3.5 The Trip Signal Delay Time was <100 msec.; as accounted for in2-St-85-10.The Trip Signal Delay Time was determined to be 50 msec.Review CriteriaNone4.0 Problems11] CR 1128964: During the Mode 5 performance the RDTC plots foreach rod were inverted from the expected response. This issue didnot impact the performance of this test and resolution of the CRoccurred prior to the Mode 3 performance of this test.12) CR 1169659: Rods D-2 and M-14 did not meet a two-sigmastatistical limit for the first rod drop in Mode 3. Three additional roddrops were performed and allAcceptance Criteria met.74 5.7 Reactor Trip System (2-PAT-3.10)This test was performed in Mode 5 as directed by 2-PAT-3.0, Post Core LoadingPrecritical Test Sequence. Testing was started on 1113116 and field workcompleted on 1120116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Demonstrate proper functioning of the Reactor Trip System. Thisobjective was accomplished by demonstrating that:1.1.1 The reactor trip breakers can be opened manually from the MainControl Room (MCR)1.2.2 Interlocks permit momentary closure of both reactor tripbypass breakers and then cause a reactor trip.1.3.3 The reactor trip bypass breakers maintain the rod drivemechanisms energized when the associated reactor trip breaker isopened for test.1.4.4 With one reactor trip bypass breaker closed, placing theopposite SSPS train channel in test causes both reactor tripbreakers and the bypass breaker to open.1.2 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 19,Reactor Trip System Test Summary.2.0 Test Methods2.1 Performance of this test (2-PAT-3.10) was completed with the unitin Mode 5 and RCS pressure greater than 100 psig to satisff thereq u irements necessary for performance of 2-TRl 1,Reactivity Control Systems Movable Control Assemblies, asrequired by Technical Surveillance Requirement 3.1.7.During performance of 2-TRl-85-1 an bsue was encountered inwhich the rods controlled bythe 2BD power cabinet (SB Group 2,CB Group 2 and, CD Group 2) did not respond to outward roddemand. Further details on this bsue are documented in CR1126661 found in the write-up for 2-PAT-3.1.15 5.7Reactor Trip System (2-PAT-3.10) (continued)2.3Section 6.2 verified both Main Control Room (MCR) Reactor Triphandswitches (2-RT-1 and 2-RT-2) generated a reactor trip andthe associated indications appropriately.Section 6.3 verified the electric interlocks prevented both bypassbreakers from being closed simultaneously and resulted in areactor trip due to a general warning in both trains of SSPS.Sections 6.4 and 6.5 verified that when a reactor trip bypassbreaker was closed placing the opposite SSPS train in test resultedin a reactor trip due to the generation of simultaneous generalwarnings in both trains.section 6.6 verified that the bypass breakers maintained the controlrod drive mechanisms energized when the associated reactor tripbreaker was opened due to injection of a simulated ReactorProtection System trip signal on the associated SSPS train.2.53.02.23.2Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1Reactor trip breakers (RTA andwith hand switches 2-RT-1 andReactor trip breakers (RTA andboth hand switches 2-RT-1 andRTB) can be opened manually2-RT-2.RTB) were opened manually with2-RT-2.Electrical interlocks trip both reactor trip bypass breakers (BYA andBYB) when both bypass breakers are closed due to simultaneousgeneral warning reactor trip signals being sent to the Reactor TripBreakers (RTA and RTB)Electrical interlocks uocessfi.rlly tripped both reactor tri p bypassbreakers when both bypass breakers were closed.16 5.7 Reactor Trip System (2-PAT-3.10) (continued)3.3 With one reactor trip bypass breaker (BYA or ByB) closed, placingthe opposite SSPS train in test causes both reactor trip breakers(RTA and RTB) and the bypass breaker (BYA or ByB) to open dueto simultaneous general warning reactor trip signals being sent tothe reactor trip breakers (RTA and RTB)Section 6.4 tested bypass breaker B in conjunction with SSPSTrain A was completed successfully and all Acceptance Criteriawere met as stated.During performance of Section 6.5 a previously known andexpected indication issue related to placing SSPS Train BMultiplexer test switch in lNHlBlT, as originally documented in CR1126043, was encountered. After verification that indicationsreceived were the same as those previously documented in CR1126043 testing continued as the eroneous indications had noimpact upon performance of 2-PAT-3.10 and Section 6.5 wascompleted satisfactorily with the exception of a procedure errorwhich was identified in CR 1126802.3.4 The reactor trip bypass breakers (BYA or BYB) maintain the roddrive mechanisms energized when the associated reactor tripbreaker (RTA or RTB) is opened by injection of a simurated ReactorProtection System trip signal on the associated SSpS trainEach reactor trip bypass breaker maintained the rod drivemechanisms energized when the associated reactor trip breakerwas opened. Section 6.6 was completed satisfactorily with theexception of a procedure error which was identified in cR 1126802.Review CriteriaNone11 5.7 Reactor Trip System (2-PAT-3.10) (continued)4.0 Problemst1l CR 1126802 was generated because of procedure errors,which assumed breakers racked to the test position would stillget the GEN WARNING alarm and MCR light indication forbreaker position. lt was determined that the procedure errorswere minor and did not affect the test, including AcceptanceCriteria Steps 6.5[14]D, 6.6[13]D, and 6.6[27]C that verifiedbreaker position lights for bypass breakers not connected.After validation of the procedure error and its impacts, testingwithin 2-PAT-3.10 continued and Section 6.3 was completedsatisfactorily.Setup of Section 6.5 places the Reactor Trip Bypass Breaker B(BYB) in the test position; therefore, MCR indication lights forBYB are not illuminated. This section verifies that the reactortrips when Reactor Trip Bypass Breaker A (BYA) is closed andSSPS Train B is placed in Test; this was successfully performedwith associated lights illuminated for BYA.Section 6.6 verifies the ReactorTrip Bypass Breakers functionto prevent a Reactor Trip during testing of Reactor TripBreakers. During this section each of the Bypass Breakerswere installed in the connected position (one at a time) withappropriate !ight indication.78 5.8 Adjustment of Steam Flow Transmitters at Minimal FIow(2-PAT-3.11)This test was performed with the plant in Mode 3 at normal operating pressureand temperature, as specified in 2-PAT-3.0, Post Core Loading Pre-CriticalTestSequence. The test was started on 5/6/16 and field work complete on 517116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify/adjust the output of the eight steam flow transmitters for"zero" output with minimal steam flow.1.2 Satisfy the Mode 3 objective in the UFSAR Table 14.2-2, Sheet 21 ,Calibration Of Steam And Feedwater Flow lnstrumentation AtPower Test Summary.2.0 Test MethodsThe plant was in Mode 3 at normal operating temperature and normaloperating pressure. Steam flow was reduced to minimal by shutting aMSIV, one loop at a time. With the MSIV closed, each steam flowtransmitter on the associated main steam line was verified/adjusted for a"zeto" output. This was repeated for each main steam line.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaNoneReview Criteria3.1 At minimum steam flow, the output from each steam flowtransmitter and its associated loop reflects zero flow asdemonstrated by the following:A. D/P Test Point: 0.19829 Vdc (0.19641to 0.20016 Vdc)B. Flow Test Point: 0.2000 Vdc (0.1972to 0.2028 Vdc)C. Computer Test Point: 0.2000 Vdc (0.1972to 0.2028 Vdc)D. ComputerPoint: 0.0 KBH (-275to275 KBH)79 5.8 Adjustment of Steam Flow Transmitters at Minimal Flow(2-PAT-3.1 1) (continued)4.0The data below was collected and the output flow was verified/adjustedwithin the Review Criteria requirements.* adjustment madeProblemsThere were no significant problems encountered during the performanceof this test.TransmitterD/P TestPoint(Vdc)Flow TestPoint(Vdc)ComputerTest Point(Vdc)ComputerPoint(KBH)2-FT-1 -3A*0.1 981 00.1 99830.201 0342-FT-1 -380.1 96850.200550.200754z-FT-1-10A0.197820.200390.200271z-FT-1-10B*0.1 97800.200710.2a12162-FT-1-21A*0.1 98670.201 600.2001312-FT-1-21B0.1 99600.200380.2005562-FT-1-28A0.1981 10.200700.2005022-FT-1-2gB0.1 96900.200070.20051380 5.9Control Rod Drive Mechanism Timing (2-PET-106)This test was performed as part of test sequence 2-PAT-3.0, Post Core LoadingPrecritical Test Sequence. The test began via WO 117705850 on 04113/16 andwas field work completed on 516116.1.0Test ObiectivesThe objectives of this test were to:1.2Verify the acceptability of the Control Rod Drive Mechanism,(CRDM), current order timing, current order amplitudes, and rodwithdrawal speed.Partially satisfy the requirements of UFSAR Table 14.2-2,SheetT, Control Rod Drive Mechanism Timing Test Summary.2.0Test MethodsThe test was required to be performed following fuel toading. Since theCRDM latch assembly must be submerged in water for proper operation, aminimum RCS pressure of 100 psig was required. The test was run atnominal hot plant conditions. Reactor Engineering verification of currentboron concentration being adequate to perform this test by being equaltoor greater than the refueling boron concentration was required.With the reactor trip breakers closed and the lift coils verified to beconnected, a selected bank was withdrawn and then reinsertedapproximately 10 steps to obtain the CRDM readings.The test objectives were accomplished by monitoring the CRDM coilcurrent profiles to verify that the stationary gripper, movable gripper, andlift coil current order changes occur at the proper time during the 780msec. rod stepping cycle; that stationary, movable and lift coil currents areproperly regulated to full current values within acceptable ranges duringrod withdrawal and insertion operations; that shutdown bank rodwithdrawal speed is a nominal 64 steps/min and control bank rodwithdrawal speed is a nominal43 steps/min.Test Results3.0AII Acceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaCurrent Order TimingThe times at which the lift, movable, and stationary current orderschange, after the start of rod motion, are within 10 msec. of theexpected times during rod withdrawal and insertion operations.811.13.1 5.9 Control Rod Drive Mechanism Timing (2-PET-106) (continued)The lift, movable, and stationary current orders for al! CRDMs werewithin 10 msec. of the expected times during rod withdrawal andinsertion operations.3.2 Coil Current AmplitudesStationary, movable and lift currents are regulated by circuityinternal to each power cabinet. The reduced and full currentnominal values are not critical, cannot be adjusted, but could be anindication of a regulation failure. Measured Values outside thenominal ranges below should be evaluated and documented by thesystem engineer.3.2.1Lift Coil - fu!l35 to 47 .2 amperes(equivalent to 438 to 590mVdc measured across a0.0125 ohm resistor)3.2.2Lift coil -reduced13 to 19.7 amperes(equivalent to 163 to246mvdc measured acrossa 0.0125 ohm resistor)3.2.3Movable GripperCoil - full7 to 9.2 amperes(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)3.2.4StationaryGripper Coil - full7 to 9.2 amperes(equivalent to 438 to 575mVdc measured across a0.0625 ohm resistor)3.2.5StationaryGripper coil -reduced3.8 to 4.8 amperes(equivalent to 238 to 300mVdc measured across a0.0625 ohm resistor)All stationary, movable, and lift currents amplitudes for all CRDMswere within the Acceptance Criteria with the exception of D08, 810,F14, F10, and D12lift coi! reduced currents which were greaterthan the 19.7 amperes criteria. This issue was previouslyevaluated in CR 1128950.These reduced lift currents were all Iessthan 21 amperes which was evaluated by Westinghouse asacceptable in Westinghouse Letter WBT-D-5604.82 5.9 Control Rod Drive Mechanism Timing (2-PET-106) (continued)3.3 Rod Withdrawal Speed3.3.1 Shutdown Bank withdrawal speed nominal 64 steps perminute.The measured Shutdown Bank withdrawal speed wasapproximately 64 steps per minute and did not exceed thenominalvalue.3.3.2 Control Bank withdrawal speed nominal4S steps per minute.The measured Control Bank withdrawal speed wasapproximately 48 steps per minute and did not exceed thenominalvalue.Review CriteriaNone4.0 ProblemsThere were no significant problems encountered during the performanceof this test.83 6.0 INITIAL CRITICALITY AND LOW POWER PHYSICS TESTING6.1 lnitial Criticality and Low Power Test Sequence (2-PAT4.0)2-PAT-4.0 started with prerequisites on 5112116 and completed on 5124116.1.0 Test ObiectivesThe objective of this test was to:1.1 Provide governance of the sequence of the Power AscensionTesting in Mode 2.The following PATs/PETs were sequenced for performance by2-PAT-4.0:o 2-PET-201
+* lnitial Criticality and Low Power Physics Testingo 2-PET-103 Reactivity Computer (ADRC)o 2-PEf-304
+* OperationalAlignment of NISo 2-PAT-1.5
+* Loose Parts Monitoring Systemo 2-PAT-1.10* lntegrated Computer System (lCS). RCI-159
+* Radiation Baseline SurveysNote:
+* lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.2.0 Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on5112116 and completed on 5122116 and included verification of thefollowing major items:o Preoperationa! tests completed to allow entry into Mode 2o TVA-SPP-30.010, lnitial Synchronization of TVA Generating Assetsto TVA's Transmission System notificationo Reactivity Control Plans are developed to support testingo 2-PET-201, lnitial Criticality and Low Power Physics Testing, hasbeen initiatedo Section 4.0 of 2-PET-103, Reactivity Computer (ADRC), has beenperformedPrior to initiation of the performance section, a cool down was initiated on5116116 to 360oF to replace a failed hot Ieg RTD. RCS temperature wasstabilized between 355oF and 365oF at22:59 on 5/'t6/16. The unit wasplaced in Mode 4 on 5118116 at 23:58 to facilitate SSPS testing. Aftercompletion of testing Unit 2 re-entered Mode 3 at 04:15 on 5/20/16 andbegan a heat up to normaloperating temperature and pressure. NOTPwas reached on 5121116 at 01 :00. On 5123116 at 01:04 the unit enteredMode 2.84 6.1 lnitial Criticality and Low Power Test Sequence (2-PAT4.0) (continued)The reactor was taken critical on 5123116 at 02:16.2-PAT-4.0 governed initial criticality and the low power testing greater than3 percent and less than 5 percent reactor power. Applicable portions ofthe following procedures were initiated and completed as appropriate.o 2-PET-201, lnitial Criticality and Low Power Physics Testing -Completed 5123116 with all criteria met.o 2-PEf-103, Reactivity Computer (ADRC), completed 5/23116 withall criteria met.o 2-PET-304, Operational Alignment of NlS, completed 5/23116 withall criteria met.o 2-PAT-1.5, Loose Parts Monitoring System, was completed on5124116 with all criteria met. CR 1171424 documents threechannels removed from service.. 2-PAT-1.10, Integrated Computer System (lCS), completed 5124116CR 1174334 documents exceeding the MED between T0457A andMCR indicator 2-Tl-62-29, RCP 3 LWR RADIAL BRG Temp.o RCI-159, Radiation Baseline Surveys - completed 5/31/16. NoAcceptance or Review Criteria were associated with this procedure.Details of the performance of each PAT procedure is contained in theindividual summaries of the associated procedures.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.O ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAT-4.0.85 6.2Reactivity Computer (ADRC) (2-PET-1 03)This test was performed as part of test sequence 2-PAT-4.0, lnitial Criticality andLow Power Test Sequence. Field performance of 2-PET-103 was commencedon 05/15/16. The purpose of this procedure is to ensure that the AdvancedDigital Reactivity Computer (ADRC) is capable of reactivity measurements insupport of Low Power Physics Testing (LPPT) per 2-PET-201. This procedurewas completed on 05123116 following completion of LPPT.1.0Test ObiectivesThe objectives of this test were to:1.2Perform installation of the ADRCPerform the calibration and setup of the ADRC prior to reactivitymeasurements.Provide instructions for connecting/restoring the RCS Temperatureand Rods Move signals to/from the ADRC1.32.0 Test MethodsThis test provided instructions for setup and installation of the ADRC forLPPT. This test connected a RCS T"rn signal from the Unit 2 Auxiliarylnstrument Room to the ADRC, connected the Power Range detectorsTop and Bottom signals and the "Rods Move" signal to the ADRC, andprovided instructions on initial checkout of the reactivity computer.Proper installation was verified by performing the initial checkout and initialexponential test. The initial checkout ensured that the ADRC was loadedwith the correct constants and reactivity data consistent with WBN Unit 2Cycle 1 core design. After input data was confirmed, the initialexponential test was conducted using a simulated signal for reactor flux.This calculated reactivity was verified to be within 1.0o/o of the theoreticalvalue. This test ensured that the ADRC was correctly calculating reactivitywith appropriate input data.Once physics testing was complete, steps were given to remove allinstalled cables and return the plant to its original state.1.186 6.2 Reactivity Gomputer (ADRC) (2-PET-1 03) (continued)3.0 Test ResultsAII Acceptance/Review Criteria were met or resolved as delineated below.Acceptance Criteria3.1. The absolute value of the PREDICTED vs MEASURED enor, thepercent difference between the ADRC "predicted" reactivity and the"measured" reactivity is < 1.0% during the ADRC lnternalExponential Test.The difference between the "predicted" reactivity and "measured"reactivity was found to be -0.03%, within the 1.0% criteria specifiedby the procedure.Review CriteriaNone4.0 ProblemsThe following issues were encountered during Reactivity Computer setupper 2-PET-103:11] While verifying the inputs to the ADRC were correct, it was notedthat the value for the prompt neutron lifetime was inconsistentbetween the value stated in the eNuPOP compared to the valuebeing used by the ADRC. The eNuPOP listed a value of 19.718microseconds while the ADRC was found to have a value of 19.716seconds. Following consultation with the fuelvendor, it wasdetermined that both values were acceptable (per Westinghouseletter NF-TV-16-24) and showed the smalldifference due to beingcalculated by two separate versions of code. The value listed in theADRC was calculated using a later version of the ANC code. Aone-time-only change was generated for this procedure to allow forthis difference. The procedure originally stated that the values hadto be "identical." The one-time-only change allowed for the valuesto be "consistent."81 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-2011The test was performed as part of test sequence 2-PAT-4.0, lnitial Criticality andLow Power Test Sequence. Field performance of 2-PET-201was commencedon 5122116 and initial criticality was achieved at 02:16 on 512312016. The testwas completed on 05123116 with successful completion of initial criticality, rodworth measurements (using Dynamic Rod Worth Measurement (DRWM)method), boron endpoint measurements, and isothermaltemperature coefficienttesting.1.0 Test ObiectivesThe objectives of this test were to:1.1 Dilute the reactor to criticality in a cautious and controlled manner1.2 Perform Mode 2 Low Power Physics Testing in a cautious andcontrolled manner, including:1.2.1 Measuring the integral worth of the control and shutdown rodbanks.1.2.2 Measuring the ARO critical boron concentration.1.2.3 Measuring the ARO lTC.1.3 This test and associated Sls satisfied the requirements of UFSARTable 14.2-2:Sheet 22,lnitial Criticality Test Summary.Sheet 23, Determination Of Core Power Range For Physics TestingTest Summary.Sheet 24, Moderator Temperature Coefficient Test Summary.(2-S l-0-23, Moderator Tem perature CoefficientDetermination at BOL)Sheet 25, Rod And Boron Worth Measurements Test Summary.Sheet 26, Core Reactivity Balance, Acceptance Criteria 1.(2-SI-0-1 2, Core Reactivity)Note: Sheet 26, Core Reactivity Balance, Acceptance Criteria 2 isdocumented in 2-Sl-0-12, Core Reactivity, at full power.B8 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)2.0 Test Methodslnitial reactor startup was conducted via dilution to critical while allshutdown and control banks were fully withdrawn. The dilution began at65 gpm and the reactor was monitored by use of lcRR. see Figure 6.3-1,ICRR vs Primary Water (N31, N32)When the ICRR reached 0.3, the dilution was terminated. After criticalitywas achieved and power increased, contro! rods were inserted to zero thestartup rate with reactor power near 1x10-3 % power.With the reactor stable, a "bite check" was then performed to determine ifthe inserted worth of control Bank D was between 40 to 7s pcm. An RCSboration was performed to establish an inserted worth of 62 pcm. Areactor exponential test was then conducted while finding the point ofadding heat to set the physics testing range.With the Physics Testing Range met, the DRWM testing began bywithdrawing CBD in Manualto the full out position. Once flux reached theappropriate levelon the reactivity computer CBD was insertedcontinuously in individual bank select until 0-5 steps withdrawn. Whendata collection was complete, CBD was restored to the fullout position.This process, of measuring rod bank worth, was repeated for eachremaining control and shutdown bank, in individual bank select. Thereactor was then brought back to a stable condition in Manual with ControlBank D slightly inserted.The boron endpoint was then calculated using the measured bank worthdata by use of the ADRC. This information is used to determine the AROHZP No XE critical boron concentration. The ITC was then measured byinitiating a constant rate cooldown, at less than 30 deg F/hr. when datacollection was complete, a constant rate heatup, also at less than 30 degF/hr, was initiated. Both sets of data were analyzed to carcurate anaverage ITC and converted to a MTC, accounting for the Dopplerreactivity coefficie nt.3.0 Test ResultsAII Acceptance/Review Criteria were met or resolved as delineated below.B9 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)Acceptance Criteria3.1 Advanced Digital Reactivity Computer (ADRC) CheckoutThe indicated reactivity is within !4o/o or 1 1 pcm of the theoreticalreactivity for each reactor exponential measurement.lndicated reactivity during the reactor exponential test for ADRCcheckout was measured at 25.0 pcm with a predicted reactivity of24.8 pcm. This resulted in a difference of 0.94o/o or 0.2 pcmdifference.3.2 Control and Shutdown Bank Worths (DRWM criteria)The sum of the measured bank worths is greater than or equal to(100%-RWU) times the sum of the predicted bank worths.The RWU, Rod Worth Uncertainty, is given as 10o/o for Unit 2 Cycle1. The sum of the measured bank worths was measured to be1.2o/o gteater than the predicted bank worths. This value is greaterthan 90% (100%-RWU) of the predicted bank worths.3.3 Boron Endpoint MeasurementBoron endpoint Acceptance Criteria is verified in 2-Sl-0-12, CoreReactivity. (2-PET-201 verified that 2-Sl-0-12 was successfullycompleted). The Technical Specification Acceptance Criteria within2-Sl-0-12 is for measured Mode 2HZP ARO critical boronconcentration shall be within the reactivity equivalence of +1000pcm of the predicted HZP ARO critical boron concentrationThe Boron Endpoint Acceptance Criteria was met via performanceof 2-Sl-0-12 (WO 117827845) following data collection from2-PET-201. The Boron Endpoint was measured at 1089 ppm.The predicted value was 1034 ppm. This resulted in a difference of55 ppm, or -569.9 pcm.3.4 TemperatureCoefficientThe Moderator Temperature Coefficient (MTC) Acceptance Criteriais verified in 2-Sl-0-23, Moderator Temperature CoefficientDetermination at BOL (WO 115947713\.(2-PEl-201 verified that 2-Sl-0-23 was successfully completed.)The Technical Specification Acceptance Criteria within 2-SI-0-23are:3.4.1 The MTC is less than or equal to 0.0 pcm/'F atHZP.90 6.3 lnitial Griticality and Low Power Physics Testing (2-PET-201) (continued)3.4.2 The MTC is less than or equal to the Beginning of CycleMTC as-measured criterion specified in the COLR.Both MTC Acceptance Criteria were met by successfulperformance of 2-Sl-0-23. The MTC was measured by2-PET-201to be -3.515 pcm/'F, which is less than the 0.0 pcm/'Flimit and below the COLR limit of -3.33 pcm/'F.3.5 Zero Power Physics Testing RangeThe zero power physics testing range is determined such thatreactivity feedback from nuclear heating does not compromise themeasurements.The zero power physics testing range was determined to not haveany reactivity feedback affects prior to performing rod worth, BoronEndpoint or ITC testing.Review CriteriaThe Review Criteria are listed below with two noted failures.3.6 ADRC Checkout3.6.1 The indicated reactivity is within 12% or 11 pcm of thetheoretical reactivity for each measurement.Indicated reactivity during the reactor exponential test forADRC checkout was measured at 25.0 pcm with a predictedreactivity of 24.8 pcm. This resulted in a difference of 0.94o/oor 0.2 pcm difference.3.6.2 The reactivity traces do not exhibit excessive noise levelft2 pcm).During the determination of the physics testing range,reactivity traces were reviewed and confirmed to not exhibitexcessive noise outside of the specified Review Criteriatolerance.3.6.3 The reactivity indication is stable as a function of flux level(no obvious dependence on the flux input !evel).During the determination of the physics testing range,reactivity traces were reviewed and confirmed stableindication of reactivity as a function of flux level with noobvious dependence of the flux input level.9L 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)3.7 Rod Worth Measurement (DRWM Criteria)3.7.1 The measured worth of all banks are within +1Oo/o or +75pcm of the prediction, whichever is greater.All shutdown and control banks were within the +10% and175 pcm of the predicted values.3.7.2 The sum of the measured worths of all banks are within+(0.8.RWU)o/o of the prediction.BankMeasured(pcm)Predicted(pcm)M.P(pcm)100"(M/P-1)(%lCD1,304.21,339.5-34.3-2.60/oCC1,061 .31,052.78.6A.8o/oCB794.4743.251 .26.9o/oCA910.0951 .4-41 .4-4.4o/oSD437.7434.43.30.8o/oSC447.4434.413.03.0o/oSB1,055.71 ,017 .138.63.8o/oSA424.1389.434.78.9o/oTotal6,434.86,36 1 .173.71 .2o/oAll shutdown and control banks were within +10% and +Tspcm of the predicted values. The sum of the measuredworths of all banks were within +(0.8.RWU)% of theprediction.3.8 Boron Endpoint3.8.1 Measured ARO boron endpoint is within 150 ppm of thepredicted boron endpoint.The measured ARO boron endpoint was measured as 108gppm, which was 55 ppm higher than the predicted boronendpoint of 1034 ppm. CR 1173995 initiated to documentReview Criteria failure.3.8.2 Measured ARO boron endpoint is within +5gg pcmequivalent boron.The measured ARO boron endpoint was measured as-569.9 pcm different from predicted values.CR 1173995 captures this failed Review Criteria also.92 6.3 Initial Criticality and Low Power Physics Testing (2-PET-201) (continued)3.9 Temperature CoefficientThe Measured ITC is within +2 pcml"F of the predicted lTC.The ITC was measured as -5.305 pcm/"F with a predicted valueof -6.67 pcm/"F.4.0 Problemst1l CR 1173995: Both Review Criteria for Boron Endpoint results werenot met. The Boron Endpoint was measured to be -569.9 pcm or55 ppm from predicted value. The Acceptance Criteria were metfor reactivity balance.Reactor Engineering, Nuclear Fuel and Westinghouse concludedthat there were no safety concerns or issues resulting from thisdifference.93 6.3 lnitial Criticality and Low Power Physics Testing (2-PET-201) (continued)FIGURE 6.3.1ICRR vs Primary Water (N31, N32)Inverse Count Rate Ratiot- F,-i,-', 6t5,k !,i:r. .1, ll-l :rt*F.r;Antual RiS C. -- ppm r.lEl stupsIt .i [: "tr ].. i r,, ,J {. i1lt.,Hi r-(. I' ), ' I" 19{LJj,iL'.1,r'.} 0L"U.1 'liIt1lriiiIlItItIiiiI1iltltiII,i.iiit1L,l1iI.*ii-.,tl'iitit-+ritll litl rliIIIi'-l1--i*-1It1I+il1,irJl'1l-I-+II1Il,11ii1iIiil{Iili i-'iti-1IiitiIi "'lli.lIr--i1i1l filT-III1ri+t--rI,llTIi1l1irr.lItiIt+!iiiil-il,1ij1iii1.lrl..i:Ii1I+iiii:jirtltiiitII1+Iifi1+liiffitt- f t.tit+4-++I1'l.--li.\+:i1iiIlIII+it.!i-t*.til_-TI,,1i" 'tIil1Il"Hi1i'lHilllr1i\llilruhrt-LT.=--ii-rrii'.ill1*J;,,1Iir'l,1\1.1ililit',rltl,,iLr--rtl,rltlr"tIri-i-i i-] .lil.It{",tla.ilf\;hTIt ,;i'*itl r'.t1in pp.p' ,-rn.lrii ":!;i,i!oi*ons+f rr#C-f ";*,*orld;*rlr-.lr..l;i, .r:i!Prinary lfater (9a1s)94 7.0 POWER ASCENSION TESTING7.1 Test Sequence for 30% Plateau (2-PAT-5.0)This test started on 5117 116 and was completed on 6/16116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Define the plant operational requirements in conjunction with2-GO-3, Unit Startup from Less than 4% Reactor Power to 30%Reactor Power.1.2 Ensure those requirements were met in order to permit powerescalation from Mode 2 conditions with reactor power < 5% RatedThermal Power (RTP) to 30%.1.3 Specifo the order of test performance at the 30% plateau.The following PATs/PETs were sequenced for performance by2-PAT-5.0:o 2-PAT-1.4
+* Pipe Vibration Monitoring. 2-PAT-1.5
+* Loose Parts Monitoring Systemo 2-PAT-1.6
+* Startup Adjustments of Reactor Control Systemo 2-PAT-1.7
+* OperationalAlignment of Process Temperaturelnstrumentationo 2-PAT-1.8
+* Thermal Expansion of Piping Systemso 2-PAT-1.10* Integrated Computer System (lCS)o 2-PAT-1.11* RVLIS Performance Testo 2-PAT-1.12* Common Q Post Accident Monitoring Systemo 2-PAT-5.1
+* Dynamic Automatic Steam Dump Controlo 2-PAT-5.3 Automatic Steam Generator Level Control,Transients at Low Powero 2-PAT-5.4 Calibration of Steam and Feedwater Flowlnstruments at 30% Powero 2-PET-301
+* Core Power Distribution Factorso 2-PET-304
+* Operational Alignment of NISo RCI-159
+* Radiation Baseline SurveyNote:
+* lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.This test and WO 116916855 for (WINCISE) Post-Critical SystemCalibration (WNA-TP-04724-WBT) satisfy UFSAR Table 14.2-2, Sheet 12,lncore lnstrumentation System Test Summary, Acceptance Criteria 2.95 7.1 Test Sequence for 30o/o Plateau (2-PAT-5.0) (continued)2.0 Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on5117116 and completed on 5125116 and included verification of thefollowing major items:c 2-P4T4.0, Initia! Criticality and Low Power Test Sequence, hasbeen completed.. NPG-SPP-10.4, Reactivity Management Program, ReactivityControl Plans were developed to support the planned testing forthis sequence.. WO 116916855 implemented vendor procedureWNA-TP-0 47 24-WBT, Westi ng house I ncore I nformationSurveillance & Engineering (WINCISE) Post Critical SystemCalibration.o WINCISE incore signa! quality verification was in progress byimplementation of applicable section of vendor procedure WNA-TP-04724-WBT.o Reactor power was S 5% RTPo RCS pressure was between2220 to 2250 psigo Section 6.3 of 2-PET-304, OperationalAlignment of NlS, to adjustthe Power Range High Flux Level Trip setpoints for testing at the30% Plateau was complete.On 5125116 the performance section of 2-PAT-5.0 was begun and a powerincrease to 6-9 percent was initiated. Mode 1,2 5o/o power, was reachedat 03:33 on 5125116.2-PAT-5.3, Automatic Steam Generator Level ControlTransients at LowPower, Section 6.1, was completed on 5126116 with all criteria met.RCI-159, Radiation Baseline Surveys, was completed on 5/31/16. NoAcceptance or Review Criteria were associated with this procedure.2-PAT-5.1, Dynamic Automatic Steam Dump Control, Sections 6.3, 6.4,and 6.5 were started on 5125116 and completed on 5127116. Section 6.3and 6.4 were completed with all criteria met. Section 6.5 was completedafter an Urgent Change to the procedure was approved by the TRG tochange the load rejection testing criteria per Westinghouse LetterLTR-SCS- 1 6-23 (LTR-PCSA-1 6-23). The revised Acceptance Criteriawas met for Section 6.5.On 5127116, with reactor power between 13 and 14 percent, the turbinewas rolled and subsequently stopped due to noise in the area of theturning gear. The turbine was shut down and subsequently re-rolled andthe noise repeated at approximately 400 rpm. The turbine was shutdownand the decision made to place the Unit in Mode 3 for turbine repairs.96 7.1Test Sequence for 30% Plateau (2-PAT-5.0) (continued)On 5128116 at 01:54 the Unit re-entered Mode 3 after a manua! reactortrip. The generator was purged and a clearance placed on the turbine forinspection. On 5/31/16 Unit 2 entered Mode 2 at 12:00 followed by takingthe reactor critical at 13:39. Mode 1 entry was made on 17:49 on 5/31/16.Unit 2 was synchronized to the grid on 613116 at 20:39 and powerincreased to 15 percent. As power increased a steam leak required amanualturbine trip on 614116 at 16:58. On 6/5/16 at11:40 the turbine wasagain tied to the grid and at 12:27 a Reactor Trip - Safety lnjectionoccurred due to the #1 governor valve failing to the open position.After repairs to the governor valve, as well as additional work on 28 MainFeed Pump, the unit was returned to Mode 2 on 618116 at 01:39. Mode 1was re-entered 6/8/16 at 09:32. On 6/9/16 at 06:40 the generator wassynchronized to the grid. An un-isolable steam leak required a turbine tripon 6/9/16 at 17:52. Repairs were made and Unit 2 was synchronized tothe grid at 13:23 on 6/11116. Power was increased to allow testingbetween 25 and 30 percent with the following Power Ascension Test beingcompleted as scheduled:o 2-PAT-1 .4, Pipe Vibration Monitoring, completed on 6/15/16 with allcriteria met for observations at the 30% Plateau.o 2-PAT-5.3, Automatic Steam Generator Level ControlTransients atLow Power was completed on 6/15/16 with allAcceptance Criteriamet. CR 1181278 was initiated to document one Review Criteria thatwas not met. An engineering evaluation determined this did not affectthe performance of the test nor invalidate any of the test results andtesting should proceed to the next plateau.o 2-PAT-1.5, Loose Parts Monitoring System, was completed on6113116 with all criteria met. CR 1171424 documents three channelsremoved from service.o 2-PAT-1.8, Thermal Expansion of Piping Systems, was field workcomplete on 6/15/16 with no issues noted.. 2-PAT-1.10, lntegrated Computer System (lCS), was completed on6114116 with all criteria met. CR 1181784 was written to address adatabase error but did not affect this plateau performance.o 2-PAT-1.11, RVLIS Performance Test, applicable sections werecompleted on 6/13/16 with all criteria met.o 2-PAT-1.12, Common Q Post Accident Monitoring System, applicablesections were completed on 6113116 with all criteria met.91 7.1Test Sequence for 30% Plateau (2-PAT-5.0) (continued). 2-PAT-1.7, Operational Alignment of Process Temperaturelnstrumentation, was completed on 6115116. All AcceptanceCriteria were met. Two Review Criteria concerning parametersrelated to Delta T failed. The OTDT calculated by Eagle-21 andprovided by the MMI carts indicated approximately 158% and theMCR indicators maximum value is 150%. lt was expected thereading from Eagle-21 was accurate and the MCR meters wereranged such that they cannot read the higher value. Additionaldata was taken at higher power ranges and the meters came onscale with no issue. CR 118246 was written.o 2-PAT-5.4 Calibration of Steam and Feedwater Flow lnstruments at30% Power was completed on 6/15/16 with all criteria met for the30% Plateau.o 2-PAT-1.6 Startup Adjustments of Reactor Control System wascompleted on 6/15/16. This was data taking only with no Review orAcceptance Criteria at this plateau.Additionally, Engineering completed the following procedures, with noissues, to support their testing at the 27-29 percent power level:o 2-T141- lncore Flux Mappingo 2-TRl-0 PDMS Operabilityo 2-Sl-0 Excore QPTR & Axial Flux Differenceo 2-Sl-92 lncore-Excore Cross Calibration Datao 2-T17.020 - PDMS Calibrationo 2-fl Calorimetric CalibrationAfter completion of all testing in this PAT it was noted that tempering flowisolations occured that did not meet the requirements of Westinghouselefter WAT-D-6432. CR 1182320 was written to document tempering flowisolations that occurred as part of testing at this plateau. Correctiveactions from this CR evaluated the length of the isolation and revised 2-SOI-2&3.01 adding a Precaution about WAT-D-6432.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.0 ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAT-5.0.98 7.1.'a Dynamic Automatic Steam Dump Control (2-PAT-5.1)This test was performed as part of test sequences 2-PAT-3.0, Post Core LoadingPrecritical Test Sequence, and 2-PAT-5.0, Test Sequence for 30o/o Plateau. Thetest began on 118116 and was field work completed on 5127116.The steam dump valves were tested without steam flow during sequence 2-PAT-3.0 in accordance with Sections 6.1 and 6.2.The steam dump valves were tested with steam flow during sequence 2-PAT-5.0in accordance with Sections 6.3, 6.4 and 6.5.The plant was less than 15% power, in Mode 1 with the main turbine notsynchronized to the grid.For Sections 6.6, the steam dump valves were tested for the deferral fromStartup with steam flow during Mode 3. This was done to confirm stroke times ina!! three of the following simulated scenarios: modulate open, trip close, and tripopen. Additionally, vibration testing on the valves which was deferred fromStartup Testing was performed.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the operation of the Steam Dump Control System. TheSteam Dump Control System has three modes of control; SteamPressure, Plant Trip, and Load Rejection=Each mode of controlwas tested to demonstrate stability following a small transient.1.2 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 28,Dynamic Automatic Steam Dump ControlTest Summary.1.3 Address additional scope of testing added to the Power AscensionTesting Program for the deferred Turbine Bypass System(Condenser Steam Dump Valves) testing. This test was to veriffthe (12) Steam Dump Valves stroke times are within acceptablelimits and to obtain vibration data on deferred Steam Dump valvesnot obtained during Hot Functional Testing in accordance with2-PAT-1.4.1.4 Satisfy the requirements of UFSAR Chapter 14,fable 14.2-1,Sheet 62, Main Steam System Test Summary by collectionvibration data on deferred Steam Dump valves not obtained duringHot Functional Testing in accordance with 2-PAf -1.4.99 7.1.1 Dynamic Automatic Steam Dump Gontrol (2-PAT-5.1) (continued)2.03.0Test MethodsThe steam dump control system is designed to maintain RCS averagetemperature by dumping steam to the condenser. This instructionfunctionally tested all three control modes (steam pressure control,plant trip control, and load reduction control) while reactor power was low(i.e., <15% power).The functional test included modulating the valves open and closed, andtripping open allsteam dump controlvalves using simulated signals whilesteam flow was isolated. The Steam Pressure controller was tested byvarying reactor power and observing the controller automaticallymaintained steam header pressure by changing steam flow to thecondenser. The Plant Trip controller was tested by simulating a reactortrip, varying reactor power, and observing controller parameters andoutput. The Load Rejection controller was tested by simulating the loss ofload permissive, and observing controller parameters and output.2-PAf-1.4, Pipe Vibration Monitoring, data was collected during theperformance of the PAT.Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaNote: There were no Acceptance Criteria for Sections 6.1 and 6.2.3.1 Section 6.3 (Steam Pressure Controller)3.1.1 After varying reactor power, the steam pressure controllermaintains steam header pressure stable, as demonstratedby neither the steam header pressure signal nor the steamdump demand signal showing divergent oscillations.During the transient neither the steam header pressuresignal nor the steam dump signal showed a divergentoscillation.3.1.2 After varying reactor power, steam pressure controllermaintains steam header pressure stable, as demonstratedby the steam dump control system remaining in automaticthroughout the transient.The steam pressure controller maintained steam headerpressure stable and the steam dump control systemremained in automatic throughout the transient.100 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)3.23.3Section 6.4 (Plant Trip Controller)3.2.1 After varying reactor power, the plant trip controllermaintains a stable Tavg as demonstrated by neither theRCS Tavg signal nor the steam dump demand signalshowing divergent oscillations.During the transient neither the RCS Tavg signal nor thesteam dump demand signal showed a divergent oscillation.3.2.2 After varying reactor power, the plant trip controllermaintains a stable Tavg as demonstrated by the steamdump control system remaining in automatic withoutdivergent oscil lations.The plant trip controller maintained Tavg stable andthe steam dump control system remained in Automaticthroughout the transient without divergent oscillations.Section 6.5 (Load Rejection Gontroller)3.3.1 The loss of load controller responds properly for the plantinput signals to the controller.The loss of load controller responded properly for the plantinput signals to the controller.3.4 Section 6.6 (Condenser Steam Dump valves stroke times)3.4.1 Condenser steam dump valves modulate open, trip closedand trip open stroke times are within acceptable limits.101 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)Modulation times and local/remote stroke times indicatedbelow are within Acceptance Criteria.Review Criteria3.5 Section 6.1 (Static Valve Timing - Modulation)3.5.1 The full open stroke length for each steam dump controlvalve is 2 314 inches to 3 inches.Review criteria were met as delineated below:Steam DumpValveExpectedStroke LenqthMeasuredStroke Lenqth2-FCV-1-1 0323t4-3n.2.8122-FCV-1-10423t4-3n.2.752-FCV-1-1 0523t4 - 3n.2.9752-FCV-1-1 062314 - 3n.2.8752-FCV-1-10723t4 - 3n.2.9222-FCV-1-1 0823t4 - 3n.2.812-FCV-1-1 0923t4 - 3n.2.8752-FCV-1-1 1023t4 - 3n.2,9122-FCV-1-11123t4 - 3n.2.8122-FCV-1-11223t4-3n.2.752-FCV-1-1 132314 - 3n.2.8122-FCV-1-11423t4-3n.2.812VALVEMODULATE OPEN<20 secTRIP CLOSES5secTRIP OPEN<3sec2-FCV-1-1 036.5 I 5.713.0 I 2.72.2 I 1.762-FCV-1-1046.2 I 4.612.25 12.902.32 12.122-FCV-1-1 0513.31 / 9.693.20 / 3.102.80 12.542-FCV-1-1 069.34 19.292.96 I 3.112.60 12.082-FCV-1-1075.40 I 6.202.74 12.502.90 12.332-FCV-1-1 089.11 / 5.892.90 l2.gg2.92 12.352-FCV-1-1 099.53 / 6.93.41 / 3.362.94 12.092-FCV-1-1 109.16 I 7.883.48 I 2.682.62 12.182-FCV-1-1114.02 13.463.08 I 3.182.56 I 1.622-FCV-1-1124.67 14.262.76 12.672.19 I 1.692-FCV-1-1 136.39 / 6.03.18 / 3.102.00 12.102-FCV-1-1149.96 I 9.413.00 / 3.502.95 12.30r02 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)3.5.2 The opening modulation time for each Steam Dump Contro!Valve is less than 20 seconds upon the receipt of a 5% to95% control signal step change.Review criteria were met as delineated below:Steam DumpValveOpen StrokeTimingRequirementActual OpenStroke Timing2-FCV-1-1 033 20 Seconds2.762-FCV-1-104s 20 Seconds4.812-FCV-1-1 05s 20 Seconds10.472-FCV-1-1 063 20 Seconds7.532-FCV-1-107s 20 Seconds7.102-FCV-1-1 083 20 Seconds10.912-FCV-1-1 09s 20 Seconds6.912-FCV-1-1 10= 20 Seconds5.912-FCV-1-111s 20 Seconds6.722-FCV-1-1123 20 Seconds3.962-FCV-1-1133 20 Seconds4.532-FCV-1-114= 20 Seconds9.083.5.3 The closing modulation time for each steam dump controlvalve is less than 20 seconds upon the receipt of a 95% to5% control signal step change.Review criteria were met as delineated below:Steam DumpValveClosed StrokeTimingRequirementActual ClosedStroke Timing2-FCV-1-1 033 20 Seconds5.322-FCV-1-1043 20 Seconds5.162-FCV-1-1 053 20 Seconds9.612-FCV-1-1 06s 20 Seconds5.222-FCV-1-1073 20 Seconds7.182-FCV-1-1 083 20 Seconds5.362-FCV-1-1 093 20 Seconds8.482-FCV-1-1 10s 20 Seconds7.962-FCV-1-1113 20 Seconds5.382-FCV-1-1123 20 Seconds5.482-FCV-1-113s 20 Seconds6.912-FCV-1-1143 20 Seconds11.92103 7.1.1 Dynamic Automatic Steam Dump Contro! (2-PAT-5.1) (continued)3.6 Section 6.2 (Static Valve Timing -Trip)3.6.1 All of the steam dump controlvalves trip open in 3 3 secondsfollowing a simulated Hl-Hl T"rn signal.Review criteria were met as delineated below:Steam DumpValveClosed StrokeTimingRequirementActual ClosedStroke Timing2-FCV-1-1 03s 3 Seconds1.232-FCV-1-104s 3 Seconds1.572-FCV-1-1 05s 3 Seconds2.412-FCV-1-1 06< 3 Seconds1.922-FCV-1-1A7s 3 Seconds2.462-FCV-1-1 08s 3 Seconds2.112-FCV-1-1 09< 3 Seconds1.772-FCV-1-1 10s 3 Seconds2.092-FCV-1-111s3 Seconds1 .852-FCV-1-1123 Seconds1.462-FCV-1-1133 Seconds1.612-FCV-1-1143 Seconds1.853.6.2 AII of the steam dump control valves trip closed in S 5seconds following a simulated block signal.Review criteria were met as delineated below:Steam DumpValveGlosed StrokeTimingRequirementActual ClosedStroke Timing2-FCV-1-1 03s 5 Seconds1.202-FCV-1-104s 5 Seconds1.422-FCV-1-1 05s 5 Seconds1.522-FCV-1-1 06s 5 Seconds1.702-FCV-1-107< 5 Seconds1.692-FCV-1-1 08s5 Seconds1.412-FCV-1-1 09s5 Seconds1.972-FCV-1-1 10s5 Seconds1.892-FCV-1-1115 Seconds1.912-FCV-1-112s 5 Seconds1.452-FCV-1-113s 5 Seconds1.702-FCV-1-114s 5 Seconds1.74104 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)4.03.7 Section 6.3 (Steam Pressure Controller)3.7.1 After varying reactor power, the steam pressure controllercontrols steam header pressure at setpoint (t 25 psi) withinnine minutes (three reset time constants).The steam pressure controller controlled steam headerpressure at setpoint (t 25 psi) within nine minutes aftervarying reactor power. Steam header pressure remainedwithin the setpoint throughout the transient.3.8 Section 6.4 (Plant Trip Controller)3.8.1 Before and after varying reactor power, the plant tripcontroller demand signal remained within 2.0o/o of thecalculated demand signal.The plant trip controller demand signal remained within 2.0o/oof the calculated demand signal before and after varyingreactor power.Note: There were no Review Criteria for Sections 6.5 and 6.6.Problems11] CR 1122945: During performance of WO 117441539, a jumperwas placed on the wrong termina! point in Step 4.3.2[9] and anunexpected alarm was received in the Main Control Room. Workwas stopped and the CR was initiated. Urgent Change, UC-1, wasmade to 2-PAT-5.1 Revision 2 to select a more accessible terminalpoint and labeled the terminal point for the jumper placement.CR 1123150: 2-FCV-1-105 failed to re-close following the trip opentest. The CR was written to troubleshoot. The needle valvebetween the positioner and the diaphragm on 2-FCV-1-105 wasfound to be closed. After verifying that the needle valve should bein the open position, the needle valve was opened and2-FCV-1-105 closed as expected. All other steam dump needlevalves were verified to be in the open position.l2l105 7.1.1 Dynamic Automatic Steam Dump Contro! (2-PAT-5.1) (continued)t3ICR 1124648: During performance of the steam dump sequencetest in Step 6.1 .21281, the Data Sheet 2 steam dump valve positionindications were not met at the three demand levels. The CRinitiated WO 117506695 which was implemented, calibrating thesteam dump controllers. ln addition, the upper limit switch for 2-FCV-1-103 was found to be sticking and it was replaced under thework order. The Step 6.1.21281, sequence test was then re-performed and the Data Sheet 2 steam dump valve positions werenot met again at similar demand positions as the first performance.The position indications that were not met in both sequence testswere associated with valve 20o/o open and 80% open positions.These positions are close to where the open and closed limitswitches actuate to turn on or off the red and green positionindication lights. lt was determined that the steam dump valvesequence was acceptable because in each sequence test eachbank of valves were full open before the next bank began to openand each valve modulated as required between fullopen and fullclosed demand positions. Data Sheet 2 was removed from theReview Criteria in Revision 3 to prevent additional unwarrantedconditional reports and repairs.CR 1124788 and CR 1127374: During performance of the steamdump valve modulation stroke timing test in Section 6.'1.3 through6.1.6, the greater than 12 second valve stroke time requirementwas not met.It was determined that the greater than 12 second stroke time is nota requirement in any design documentation. ln addition, there areno plant procedures that set up the valves to ensure a greater than12 second modulation stroke time. This criteria was removed from2-PAT-5.1 in Revision 3.CR 1174915: During transfer of steam dump control to the loss ofload controller, a diverging oscillation was observed in the loss ofload controller response. The loss of load controller response wasfound to be proper for the plant input signals to the controller andthe high gain settings of the controller. The evaluation of the loss ofload controller was documented in Westinghouse Letter LTR-SCS-16-23.l4lt5I106 7.1.1 Dynamic Automatic Steam Dump Control (2-PAT-5.1) (continued)t6IUrgent Change UC-1 to 2-PAT-5.1 Revision 0005 revised Step6.5.2[4] and Acceptance Criteria 5.1.3[A] to verifo the loss of loadcontroller responds properly for the plant input signals to thecontroller. Urgent Change UC-1 revised the remainder of the lossof load transient testing and evaluation in Sections 6.5.2 and 6.5.3.Acceptance Criteria 5.1.3[B] and Review Criteria 5.2.5 weredeleted. ln addition, the Westinghouse test scoping documentWATMBT-SU-2.8.5 Acceptance Criteria was revised to veriff theIoss of load controller responds properly. Also a SAR ChangePackage No. U2-021 was approved and issued that revisedChapter 14 Table 14.2-2 Sheet 28 to verify the load rejectioncontroller responds properly. These changes were based on theWestinghouse Letter LTR-SCS-16-23 which documented theproper response of the load rejection controller.CR 1'170159: 2-FCV-1-108 did not initially meet the trip openstroke time Acceptance Criteria of 3 seconds. A volume boosteradjustment was made to 2-FCV-1-108 under WO 117826339 andthe valve met the trip open stroke time when retested.CR 1170319: Piping vibration at2-FCY-1-111 did not meetAcceptance Criteria during the stroke test. Civil DesignEngineering evaluated the piping response and found it acceptable.17ltBlL01 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAr-5.3)This test was performed as part of test sequence 2-PAT-5.0, Test Sequence for30% Plateau. The test began on 5114116 and completed on 6l'15116.1.0Test ObiectivesThe objectives of this test were to:Demonstrate the proper operation and automatic response of theSteam Generator Level Control System for each Steam Generatorduring steady-state operation.2.01.2 Satisfy, in part, the requirements of UFSAR Table 14.2-2, Sheet 30,Automatic Steam Generator Level Control Test Summary.Test MethodsThe UFSAR requires tests be performed at various power levels from 5%through 100o/o reactor power. This PAT tested low power aspects of theUFSAR requirement. For Section 6.1 the plant was in Mode 1 at less than10% power with the main turbine not synchronized to the grid. For Section6.2 the plant was in Mode 1 at approximately 30% power after the MFWForward Flush/Back Flush Heatup had been completed. The test wasperformed in conjunction with maintenance work order activities to collectdata needed to calibrate and tune feedwater control system components.Actual testing in Section 6.1, Feedwater Bypass Control Valves, wasstarted on dayshift 5125116 and completed on nightshift 5/26/16. AllAcceptance and Review Criteria were met for Section 6.1, with noadditional tuning of the Feedwater Bypass Control Valve Controllers beingnecessary.Section 6.2, Transfer From Bypass To MFW Reg Valves, was initiated on6112116. During the performance of Section 6.2.1, Transfer From BypassTo MFW Reg Valve For SG No. 1, the Steam Generator leve! was notstable within the required + 2o/o during the 10 minute monitoring period.The decision was made to continue testing in accordance with Section6.2.4 for SG No. 4 and perform a re-test of Section 6.2.1 at a Iater time.During the performance of Section 6.2.4, Transfer From Bypass To MFWReg Valve For SG No. 4, the Main Feedwater Reg Valve, 2-FCV-003-0103, did not respond to a 30% demand. lt was determined that the airline to the MFW Reg Valve for SG No. 4 was Ieaking. WO 117904374was written and performed to repair the leak.1.1108 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued)On 6/13/16, Re-Test #1 was performed for Section 6.2.1, Transfer FromBypass To MFW Reg Valve For SG No. 1. All Acceptance Criteria weremet for SG No. 1 MFW Reg Valve with no adjustments being made.Repairs to the SG No. 4 air line were completed on 6113116 under WO117904374, and testing was resumed for the SG No. 4 MFW Reg Valve.All four MFW Reg Valves successfully met the Acceptance Criteria uponcompletion of testing in Section 6.2. At the conclusion of testing, CR1181278 was written to address areas of concern.3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaSection 6.1 (Feedwater Bypass ControlValves)3.1.1 The indicated steam generator level undershoot was lessthan 4.0% below the final setpoint following automaticrecovery from high steam generator level.The steam generator level undershoot ranged from 1% to2o/o below the setpoint for al! four Steam Generators,following automatic recovery from high steam generatorlevel, which met the required Acceptance Criteria.3.1.2 The indicated steam generator level overshoot was less than4.0o/o above the setpoint following automatic recovery fromlow steam generator leve!.The steam generator level overshoot ranged from -1% to 0%above the setpoint for all four Steam Generators, followingautomatic recovery from high steam generator level, whichmet the required Acceptance Criteria.3.1109 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (continued)3,2Section 6.2 (Transfer from Bypass to MFW Control Valves)3.2.1 lndicated steam generator level returned to and remainedwithin t2o/o of the program levelwithin 10 minutes followingthe transfer of level control to the Main Feedwater Reg.Valves in automatic.All four steam generator level indications returned to andremained within t2o/o of the program level within 10 minutesfollowing the transfer of level control to the Main FeedwaterReg. Valves in automatic. CR 1181278 was written due toquestions regarding the wording of the Acceptance Criteriain 2-PAT-5.3 and is discussed under Problems.3.2.2 Demand signal oscillations for each of the Main FeedwaterReg. Valves were less than +6.00/o during steady stateoperation.Allfour Main Feedwater Reg Valves exhibited less than 16%oscillation of the Main Feedwater Reg Valves in Auto:o SG #1 - 2.17o/o. SG #2 - 1.24o/oo SG #3 - 2.81o/oo SG #4 -2.44YoCR 1181278 also addressed an issue with the procedure notbeing clear on the time data is recorded. This is discussedunder Problems.3.2.3 Feedwater flow oscillations to each Steam Generator wereless than t6.0% during steady state operation.The feedwater flow oscillations to each Steam Generator aredocumented below:o sG #1 - 3.41%o sG #2 - 6.39%o sG #3 - 5.59%o SG#4-4.75o/oThe feedwater flow oscillations to SGs # 1,3, and 4 met theAcceptance Criteria of less than +6 .0o/o during110 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (continued)steady state operation. CR 1181278 discussed in Problemsaddresses #2 SG data as acceptable.Review Criteria3.3 Section 6.1 (Feedwater Bypass ControlValves)3.3.1 lndicated Steam Generator level returned to and remainedwithin t2% of the level setpoint within 37.5 minutes followingautomatic recovery from high Steam Generator level.The indicated Steam Generator level returned to andremained within *2o/o of the level setpoint within 37.5 minutesfollowing automatic recovery from high Steam Generatorlevel as shown below:o SG #1 - 14 minuteso SG #2 - 13 minutes: 33 trl.1Zilitxf:3.3.2 lndicated Steam Generator level returned to and remainedwithin t2o/o of the level setpoint within 37.5 minutes followingautomatic recovery from low Steam Generator level.The indicated Steam Generator level returned to andremained within t2o/o of the level setpoint within 37.5 minutesfollowing automatic recovery from low Steam Generator levelas shown below:: :: #,1\ilIlil:i3.4 Section 6.2 (Transfer from Bypass to MFW Reg. Valves)3.4.1 The Main Feedwater Reg. Valve position was between theminimum and maximum positions given in Figure 1 of2-PAT-5.3 for the specific loop Main Steam Flow.111 7.1.2 Automatic Steam Generator Level ControlTransients at Low Power(2-PAT-5.3) (conti nued )Data indicated that the Main Feedwater Reg Valve positionsexceeded the maximum positions given in Figure 1 of2-PAT-S.3:. Reg Valve 2-FCV-3 41.0o/o. Reg Valve 2-FCV-348 - 45.40. Reg Valve 2-FCV-3 40.9Yo. Reg Valve 2-FCV-3-103 - 40.9%CR 1181278 was generated and Engineering was requestedto evaluate the data and provide recommendations.Engineering's recommendation was to proceed with powerAscension Testing to the 50% plateau.3.4.2 lndicated Steam Generator level was within t2o/o of theprogram Leve! within 10 minutes following Main Feed Reg.Valve being placed in AUTO and subsequent stableconditions steady state operations.Data verified that the indicated Steam Generator levels werewithin t2o/o of the program level within 10 minutes followingMain Feed Reg. Valve being placed in AUTO andsubsequent stable conditions for steady state operations.3.4.3 The Main Feedwater Header Pressure oscillations were lessthan 108 psi (peakto-peak) during steady state operations.(This limit was based on r3.0% of the instrument span of1800 psi).The Main Feedwater Header Pressure oscillations were lessthan 108 psi (peak-to-peak) during steady state operationsas shown below:o Main Feedwater Header Pressure OscillationFor SG #1 - 5 psio Main Feedwater Header Pressure OscillationFor SG #2 - 8 psio Main Feedwater Header Pressure OscillationFor SG #3 - 8 psio Main Feedwater Header Pressure OscillationFor SG #4 - 4 psi3.4.4 The Actual (measured) AP was within 25.0 psi of theProgram AP during steady state operation.The actual (measured) AP was 0.8 psi which met theAcceptance Criteria of being within 25.0 psi of the programAP during steady state operation.LL2 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued )4.0Problemst1] CR 1181278 was written due to questions regarding the wording ofthe Acceptance Criteria in 2-PAT-5.3. A comparison of 5.1.2Acceptance Criteria for Section 6.2 (Transfer from Bypass to MFWControlValves) to the Westinghouse document WBT-D4709 (LTR-PCSA-14-31) confirmed that the Acceptance Criteria in2-PAT-5.3 were written correctly. The steps in the body of theprocedure to perform the test and verify the Acceptance Criteriawere also reviewed with the originator of the CR. It was determinedthat the procedure for this section of the test was written correctlyand neither the test nor the results were invalidated by theconcerns in the CR.CR 1181278 documented another concern which stated "Theprocedure is not clear if the performer looks at the data before orafter a time. The procedure should say AFTER, because that is theapproximate time that the main feedwater is transferred into Auto.With clarification, allAcceptance Criteria are met." A review ofData Sheet 11 revealed that for each Main Feedwater Reg Valve,the column to record data contains notation which states "Data fromtime in Step 6.2.X[18]". Step 6.2.X[18] recorded the end time forthe 10 minute monitoring period. lt would have been better if DataSheet 11 would have stated "Data from monitoring period in Steps6.2.X1161 through 6.2.X[18]". Additionally, during the review, it wasdiscussed that the Acceptance Criteria was to monitor the DemandSignal oscillations for each of the Main Feedwater Reg Valves;however, the test kept the Reg valves in Manual instead of Autoduring this portion of testing. Fortunately, the Test Coordinatorscollected the appropriate data with the valves in Manual, thenswapped the controller position to Auto, as allowed by theprocedure, and collected the appropriate data in this condition. Thedata was analyzed with the valves in AUTO and it was determinedthat the Acceptance Criteria were met. Since the data was alsocollected with the valves in AUTO and the Acceptance Criteria weremet, there was no need to re-perform this section of the test.113 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued )l2lThe data recorded in the test for the feedwater flow oscillations toSG #2 was above the acceptable limit of 60/o for flow oscillations;however, CR 1181278 states "The feedwater was controlled anddid not oscillate. However, the maximum deviation was about6.39% of average flow." A review of the data indicated that thefeedwater flow to SG #2 started off with a deviation of >6o/oihowever, the controller brought the flow to within an acceptablerange in a steady manner and maintained an acceptable flow ratherthan oscillating for a period of time (see Figure 7.1.2-1) The factthat the feedwater flow stabilized within a range which was lessthan 6% without oscillating meets the intent of the AcceptanceCriteria.CR 1'181278 also requested Engineering to evaluate data on oneReview Criteria and provide recommendations. Engineering'srecommendation was to proceed with Power Ascension Testing tothe 50% plateau for the following Review Criteria:o Section 6.2, MFW Reg Valves were not between theminimum and maximum positions requiredWO 117904374 was initiated to repair a leak on the air Iine toli4Steam Generator MFW Reg. Valve. The valve was retested afterrepairs and passed Acceptance Criteria.lL4 7.1.2 Automatic Steam Generator Level Control Transients at Low Power(2-PAT-5.3) (conti nued )FIGURE 7.1 .2-1SG #2 FW Flow Oscillationflat^tiirc Hlclory L3-f un-zolrB 1I,!*6:()0 to ,.3-Jun-2or.6 1():56:00 cur)552oat-s tn700 -10 {olO:aa:OO EDT !O24,2.o gf ,.O:.lt:zo Ef ,O:3;t:@ EY 1O:52:rlO ET 1O:S.:ZO Ef lO:54:@ EDr1l-rm-20:ta l3-rn-aclj l'!-rrlt-aqt.3 1l-run-20L at-rr2q13 a3-rlli-zct3 af-rrr-20LEl'E##tCl$lF* 2 vAL:rD{rE n LEVEL t8t{ g+ Fr'(z) Fr@r@aa., (k:r,:l) z-FIC-ln3r arrE Lro{T -10 1rt(t) rso74 (El:t t) 30 a LEVEL scTFlolt{T 40 co r(a) Fo.rarA &:ui!) Jtr Gr z FEE irr rx a Fl..r, z@ a3o rtHS 2 VALIDATED115 II7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4)This test was performed with the plant stable at approximately 30% Power as part of2-PAT-5.0, Test Sequence for 30% Plateau. The test began on 6/13/16 and wascomplete on 6/15/16.1.0 Test ObiectivesThe objectives of this test were to:1.1 Verify the output of the eight feedwater flow transmitters for "zero"output with minimal feedwater flow, collect data for determining thenew calibration spans for the steam flow transmitters1.2 Verify the calibration of the feedwater and steam flow transmitters,by comparing indicated flows between the Main Control Boardlndicators, the Protection System, and the Control System.1.3 Satisfy, in part, the 30% objective in the UFSAR Table 14.2-2,Sheet 21, Calibration Of Steam And Feedwater Flowlnstrumentation At Power Test Summary.2.0 Test MethodsAt approximately 30% power, each feedwater flow transmitter was placedin bypass and verified for "zero" output.At approximately 30% power, steam generator blowdown and temperingflow were isolated while data was collected. Steam generator blowdownand tempering flow were then reestablished and calculations/comparisonswere performed.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.Acceotance CriteriaNoneLL6 7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4) (continued)Review Criteria3.1 At zero DP, the output from each Feedwater Flow Transmitter andits associated loop reflect zero flow as demonstrated by thefollowing criteria:A. Computer Point: 0.000 KBH (-25.5 to 25.5 KBH)B. Flow Test point: 0.200 Vdc (0.1858to 0.2142Ydc1C. Computer Test Point: 0.200 Vdc (0.1858 to 0.2142 Vdc)D. DP Test point: 0.1983 Vdc (0.1848 to 0.2118 Vdc)The data below was collected and the output flow was verified withinthe Review Criteria requirements.3.2 The difference between the Feedwater Flow as measured in theProtection System and the Main Control Board lndicators is withint5.0% of the rated flow.Measured differences (% ERRORS) between -1.89Yo and +1.18%3.3 The difference between the feedwater flow as measured in theProtection System and the lndicated Computer Feedwater Flow iswithin t2.Oo/o of rated flow.Measured differences (% ERRORS) between -0.24o/o and +9.967oDESCRIPTIONSTEAM GENERATOR 1STEAM GENERATOR 22-FT-3-3542-FT-3-35B2'FT'3'48A2-FT-3-48BComputer Pont Flow (KBH) 5.2.A.15111Flow Test Pont 5.2.4.20.201480.200280.201540.20055Comp Test Point 5.2.A.30.200894.200320.200490.20029DP Test Point Voltaoe 5.2.A.40.202690.1 95890.1 97630.1 991 2DESCRIPTIONSTEAM GENERATOR 4STEAM GENERATOR 32-FT-3-90Az-FT-3-9082-FT-3-103A2-FT-3-103BComputer Point Flow (KBH) 5.2.A.142-31Flow Test Point 5.2.4.20.201040.200740.200560.20027Como Test Point 5.2.A.30.200670.200380.1 99560.20043DP Test Point Voltaoe 5.2.A.40.199450.197490.1 95890.1 9623LLl 7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4) (continued)3.4 The difference between the feedwater flow as measured in theProtection System and the Feedwater Flow Signal used for flowcontrol is within t2.0% of rated flow.Measured differences (% ERRORS) between -0.29Yo and -0.07o/o3.5 The difference between the steam flow as measured in theProtection System and the Main Gontrol Board lndicators is withint5.0% of the rated flow.Measured differences (% ERRORs) between -0.87o/o and +0.94%3.6 The difference between the steam flow as measured in theProtection System and the lndicated Computer Steam Flow iswithin t2.0o/o of rated flow.Measured differences (% ERRORS) between -0.10% and 0.00%3.7 The difference between the steam flow as measured in theProtection System and the Steam Flow Signal used for flow controlis within t2.0o/o of rated flow.Measured differences (% ERRORS) between -0.23o/o and 0.00%3.8 The difference between the feedwater flow as measured in theProtection System and the Steam Flow as measured in theProtection System is within t5.0o/o of rated flow.Measured differences (% ERRORs) between -2.620/o and +4.45o/o3.9 The difference between the feedwater flow as measured in theContro! System and the Steam Flow as measured in the ControlSystem is within t5.0% of rated flow.Measured differences (% ERRORS) between -2.57o/o and +4.560./0118 7.1.3 Calibration of Steam and Feedwater Flow lnstruments at 30% Power(2-PAT-5.4) (conti nued )Additionally, a comparison of the corrected feedwater flows and steam flowsversus predicted design flow is provided:4.0 ProblemsThere were no significant problems encountered during the performanceof this test.119 7.2 Test Sequence for 50o/o Plateau (2-PAT-6.0)This test started on 5130116 and was completed on 7116/1 6.1.0 Test ObiectivesThe objectives of this test were to:1.1 ln conjunction with 2-GO4, Normal Power Operation, define theplant operational requirements and ensure those requirementswere met in order to permit power escalation from 30% RatedThermal Power (RTP) to 50%.1.2 Specifo the order of test performance at the 50% plateau.The following PATs/PETs were sequenced for performance by2-PAT-6.0:o 2-PAT-1.4
+* Pipe Vibration Monitoringo 2-PAT-1.5
+* Loose Parts Monitoring Systemo 2-PAT-1.6 " Startup Adjustments of Reactor Control Systemo 2-PAT-1.7
+* OperationalAlignment of Process Temperaturelnstrumentationo 2-PAT-1.8
+* Thermal Expansion of Piping Systemso 2-PAT-1.10* lntegrated Computer System (lCS)o 2-PAT-1.11* RVLIS Performance Testo 2-PAT-1.12" Common Q Past Accident Monitoring Systemo 2-PAT-3.3
+* RCS Flow Measuremento 2-PAT-5.2 Turbine Generator Trip With Coincident Loss ofOffsite Power Testo 2-PAT-6.1 Automatic Reactor Control Systemo 2-PAT-6.2 Automatic Steam Generator Level Control Transientsat 50o/o Powero 2-PAT-6.3 Calibration of Steam and Feedwater Flowlnstruments at 50 % Powero PET-301 . Core Power Distribution Factorso PET-304
+* OperationalAlignment of NIS. RCI-159
+* Radiation Baseline SurveysNote:
+* lndicates that the test is performed at multiple test plateaus.The description of the testing is documented in the section(plateau) in which it was completed.L20 7.2Test Sequence for 50% Plateau (2-PAT-6.0) (continued)2.0 Test MethodsPrerequisite actions for this Power Ascension Test (PAT) started on5/30/16 and completed on 6117116 and included verification of thefollowing major items:. 2-PAT-5.0, Test Sequence for 30% Plateau, completeo NPG-SPP-10.4, Reactivity Management Program, ReactivityControl Plans were developed to support the planned testing forthis sequenceo Reactor power between 27o/o and 29% RTP with T"rn-T,"r mismatch+1.5 "F or less. RCS pressure is between 2220to 2250 psig. Section 6.5 of 2-PET-304, Operational Alignment of NlS, to adjustthe Power Range High Flux Level Trip setpoints for testing at the50% Plateau completePower increase to the 50% testing plateau was initiated on 6/17116 at11:40 and PAT testing in Section 6.1 of 2-PAT-6.0 was begun. On6120117 at 15:37, U-2 Turbine tripped due to the loss of 28 MainFeedwater Pump and subsequently an automatic Reactor Trip occurred at15:40 due to S/G levels reaching their low-low trip setpoint. The plant wasstabilized in Mode 3.Unit 2 re-entered Mode 2 on 6123116 at 17:37 and the reactor critical at17:53. Mode 1 entry was made at 03:00 on 6124116. The U-2 generatorwas synchronized to the grid at 13:58.A manual turbine trip was initiated on 6126116 at 09:45 due to a steamleak. Reactor power was reduced and the Unit entered Mode 2 at 11;44.At 15:26 the reactor was tripped manually and the unit stabilized in Mode3.Mode 2 was again entered on712116 at 03:00 with reactor criticality at03:20. U-2 entered Mode 1 at07:57 and was synchronized to the grid inthe afternoon at 13:36. On717116 the 50% Plateau power level testingwas reached and the 50% tests were commenced. Steady state testingincluded:o 2-PAT-1.4, Pipe Vibration Monitoring, completed on718116 with allcriteria met.o 2-PAT-1.5, Loose Parts Monitoring System, was completed on717116 with all criteria met. CR 1171424 documents three channelsremoved from service.o 2-PAT-1.6, Startup Adjustments of Reactor Control System, wascompleted on718116. This was data taking only with no Review orAcceptance Criteria at this plateau.L2L 7.2 Test Sequence for 50% Plateau (2-PAT-6.0) (continued). 2-PAT-1.7, Operational Alignment of Process Temperaturelnstrumentation, was completed on 718116 with allcriteria met.o 2-PAT-1.8, Thermal Expansion of Piping Systems, was field workcomplete on718116 with 2 issues referred to engineering forevaluation with Problem Report #4. Engineering review indicated itwas acceptable to continue Power Ascension Testing.o 2-PAT-1.10, lntegrated Computer System (lCS), was completed on718116 with all criteria met.o 2-PAT-1.11, RVLIS Performance Test, applicable sections werecompleted on117116 with all criteria met.o 2-PAT-1.12, Common Q Post Accident Monitoring System,applicable sections were completed on 7nh6 with al! criteria met.o 2-PAT-6.3, Calibration of Steam and Feedwater Flow lnstrumentsat 50% Power, was completed on 718116 with all criteria met.o 2-PAT-3.3, RCS Flow Measurement, was completed for the S0%Plateau on719116 with all criteria met.o RCI-159, Radiation Baseline Surveys, was completed for the 50%plateau onT110116. No Acceptance or Review Criteria wereassociated with this procedure.Transient tests were begun onTl'11116 and included the following:o 2-PAT-6.1, Automatic Reactor Control System, was completed on7113116 with all criteria met.o 2-PAT-6.2, Automatic Steam Generator Level ControlTransients,was completed on 7116116 with all criteria met.o 2-PAT-5.2, Turbine Generator Trip With Coincident Loss of OffsitePower Test, was completed on7114116 with all criteria met exceptone Review Criteria. CR 1192287 was written to document Tcoldgoing below the 547"F criteria.o 2-PAT-1.4, Pipe Vibration Monitoring, for transient testing wascompleted on7l14116 with all criteria met.2-PAT-1.2, Load Swing Test, originally scheduled for the 50% plateau,was revised to allow performance during 2-PAT-7.0 due to the inability ofthe turbine to be operated in IMP lN. Repairs to the circuitry wereevaluated during the outage and a procedure revision was made to allowperformance of the Load Swing Test in lMP OUT on the turbine controls.L22 7.2 Test Sequence for 50% Plateau (2-PAT-6.0) (continued)Additionally, Engineering completed the following procedures or applicablesections during the steady state period, with no issues, to support theirtesting at the 50% Plateau:o 2-T141- lncore Flux Mapping. 2-TRl-0 PDMS Operabilityo 2-Sl-0 Excore QPTR & Axial Flux Differenceo 2-PET-301 - Core Power Distribution Factorso 2-Sl-92 lncore-Excore Cross Calibration Data. 2-T17.020 - PDMS Calibration. 2-PET-304 - OperationalAlignment of NIS. 2-Tl Calorimetric Calibrationo 2-5!-0 Hot Channelfactors Determinationo 2-Sl-92 NIS Monthly Recalibration data. 2-Sl-0 lncore QPTRDetails of the performance of each PAT procedure is contained in theindividual summaries of the associated procedures as they are fullycompleted.3.0 Test ResultsAllAcceptance/Review Criteria were contained within the tests sequencedby this test.4.0 ProblemsProblems encountered are addressed in the following discussions of eachtest sequenced by 2-PAT-6.0.L23 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-PAr-5.2)This test was performed as part of 2-PAT-6.0, Test Sequence for 5Oo/o Plateau,and initiated in Mode 1. The test began pre-requisites on 7nn6 and was fieldwork completed on 7l'14116.1.0Test ObiectivesThe objectives of this test were to:1.1 Demonstrate Unit 2 response to a turbine generator trip with acoincident loss of offsite power (LOOP) is in accordance withdesign.1.2 Demonstrate that all four emergency diesel generators (EDG)automatically start, the Unit 2 EDGs connect to their respectiveshutdown board and provide power to the controls, indications, andequipment necessary to maintain Unit 2 in Hot Standby (Mode3)conditions for a minimum of 30 minutes.1.3 Demonstrate that operators can control plant parameters usingequipment available during a loss of offsite power.1.4 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 33, TurbineGenerator Trip With Coincident Loss Of Offsite Power TestSummary.1.5 Provide the steps necessary to protect Unit 1 operations.Test Methodslnitial conditions for Unit 2 include reactor power at approximately 30% ofrated thermal power, the main generator synchronized to the TVA grid,and electrical load greater than or equal to 120 MWe. All four dieselgenerators were in their normal standby condition.Unit 1 was in Mode 1 with alignment of the Unit 1 Reactor Coolant PumpBoards, Unit 1 Unit Boards, Common Boards and Shutdown Boards 1A-Aand 1B-B energized from normal power sources, the USST's associatedwith Mode 1 operation.The C-S CCS pump was aligned and in service to supply header 1Bl2B inaccordance with 0-SOl-70.01, "Component Cooling Water System". Theautomatic transfer of the 2A and 2C RCP boards to the A RCP Start buswas blocked and the automatic transfer of the 28 and 2D RCP boards tothe B RCP Start bus was blocked. The automatic transfer of the 2A-AShutdown board to the 2A-A Diesel Generator and the automatic transferof the 2B-B Shutdown board to the 2B-B Diesel Generator were notblocked. The automatic transfer of the 2A-A Shutdown board to the DCSST and the automatic transfer of the 2B-B Shutdown board to the CCSST was blocked. The maintenance supplies to the Unit 2 Shutdown2.0L24 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-PAT-5.2) (conti nued )Boards were verified in the "racked down/removed" position. The 6.9 KVB common board was in its normal alignment. The B common Board wasnot de-energized during the test to protect auxiliaries on both Units.The Unit 2 Main Turbine was manually tripped. Following the turbine trip,Operations concurrently and immediately performed the following:. Opened the normal power supply breaker to the 2A-A ShutdownBoard. The board did not transfer to its alternate power source,resulting in a dead board condition. Allfour emergency dieselgenerators (EDGs) started as expected. The 2A-A EDG connectedto and energized the 2A-A Shutdown Board.. Opened the normal power supply breaker to the 2B-B ShutdownBoard. The board did not transfer to its alternate power source,resulting in a dead board condition. The 2B-B EDG connected toand energized the 2B-B Shutdown Board as expected.o Operations ensured the U2 Main turbine and U2 Main Generatortripped.Following the Unit 2 generator trip, all four Unit 2 RCP Boards did nottransfer to their alternate power source and remained de-energized. Anautomatic reactor trip of Unit 2 occurred when voltage was lost to the Unit2 Reactor Coolant Pumps (RCPs). Operations then entered 2-E-0,Reactor Trip or Safety lnjection.With af l four Unit 2 reactor coolant pumps de-energized, the RCSdeveloped natural circulation conditions. Natural circulation parameterstook longer than Unit 1 to establish due to the very Iow decay heatgenerated in the new core. The Unit 2 Main Steam lsolation Valves(MSIV's) were manually closed to support testing with a simulated Loss ofOffsite Power configuration. Unit 2's Main Steam line pressure, steamgenerator pressure, and RCS temperature were maintained by the SGPORVs discharge to atmosphere. Auxiliary feedwater automaticallystarted, and steam generator level trended to post trip setpoint conditions.The test ran at least 30 minutes after the 2A-A and 2B-B 6.gkvShutdown Boards were energized from their respective emergency dieselgenerators without restoring offsite power. Unit 2 was restored to aplanned outage upon test completion at the direction of the shift Manager.125 7.2.1 Turbine Generator Trip with Goincident Loss of Offsite Power Test(2-PAT-5.2) (continued)3.0 Test ResultsAll Acceptance/Review Criteria were met or resolved as delineated below.Acceotance Criteria3.1 The 2A-A Diesel Generator automatically starts and connects to2A-A Shutdown Board following the Loss of Offsite Power transient.2A-A Diesel Generator automatically started and connected to itsrespective shutdown board.3.2 The 2B-B Diesel Generator automatically starts and connects to2B-B Shutdown Board following the Loss of Offsite Power transient.2B-B Diesel Generator automatically started and connected to itsrespective shutdown board.3.3 The Unit 2 Pressurizer Safety Valves do not open during thetest.Pressurizer Safeties did not open during the test.3.4 The Unit 2 Steam Generator Safety Valves do not open during thetest.Steam Generator Safety Valves did not open during the test.3.5 A Unit 2 Safety lnjection is not initiated during the test.No safety injection was initiated during testing.3.6 Hot standby (Mode 3) conditions on Unit 2 are maintained for atleast 30 minutes after the 2A-A and 2B-B 6.9kV Shutdown Boardsare energized from respective emergency diesel generators withoutrestoring offsite power.Hot Standby (Mode 3) conditions were maintained for at least 30minutes after 2A-A and 2B-B were energized lor their respectiveemergency diesel generators without restoring offsite power.L26 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-PAT-5.2) (conti nued )3.7 The 1A-A Diesel Generator automatically starts but does notconnect to the 1A-A Shutdown Board following the Loss of OffsitePower transient.1A-A Diesel Generator automatically started but did not connect tothe 1A-A Shutdown Board.3.8 The 1B-B Diesel Generator automatically starts but does notconnect to the 1BB Shutdown Board following the Loss of OffsitePower transient.1B-B Diesel Generator automatically started but did not connect tothe 1B-B Shutdown Board.Review Criteria3.9 The following Unit 2 parameters were maintained within theirrespective limits for at least 30 minutes immediately after de-energizing 2A-A and 2B-B Shutdown Boards, using equipmentavailable with offsite power removed from Unit 2:3.9.1 RCS Cold Leg Temperature (il7"F to 560oF and changingat a rate less than 50"F in one hour)This criteria was not met. RCS Cold Leg Temperaturesreduced below the minimum temperature of 547 degreesduring the 30 minute period. The rate of change was lessthan 50 degrees in one hour. CR 1192287 documented thisissue and was due to Turbine Driven AFW cooling sincereactor decay heat was minimal and no RCPs in service.3.9.2 Pressurizer Level (17Yo to 50%)Pressurizer level maintained between 25o/o and 34% duringthe test period.3.9.3 Pressurizer Pressure (2000 psig to 2335 psig)Pressurizer Pressure maintained between 2119 and 2244.8psig during the test period.3.9.4 Steam Generator Levels (17o/o to 60% narrow range andeither constant or trending toward 38% of narrow range)Steam Generator narrow range level maintained between30% and 39% during the test period.L21 7.2.1 Turbine Generator Trip with Coincident Loss of Offsite Power Test(2-P AT -5.2) (conti n ued )4.0Problemst1] CR 1192287 documented the Review Criteria was not met whenthe RCS cold leg temperature decreased below the minimumcriteria of 547oF. This deficiency is attributed to the TurbineDrive Auxiliary Feedwater Pump steam supply source coolingthe loop as it supplied AFW to the steam generators. Withminimal reactor decay heat and no RCPs running, the looptemperature was not maintained above the minimum criteria.121CR 1192023 was written to address the observation that the2A-A Diesel Generator appeared to be slower than expected intying on to its shutdown board. This was neither a Review orAcceptance Criteria for this test. Subsequent review by plantstaff did indicate the 2A-A Diesel Generator did not tie onto theboard within the Technical Specification limit and was declaredinoperable. The diesel generator was repaired by plantmaintenance and had no impact on meeting the PAT criteria asdelineated in UFSAR Chapter 14, Table 14.2.2, Sheet 33.L2B 7.2.2 Automatic Reactor Control System (2-PAT-6.1)This test was performed at the 50% test plateau as directed by2-PAT-6.0, Test Sequence for 50o/o Plateau. Testing was started on 6/18/16and field work completed on 7112116.1.0 Test ObiectivesThe objectives of this test were to:1.1 Demonstrate the ability of the Automatic Rod Control System tomaintain the average RCS temperature (T",r) within acceptableIimits during both steady-state and transient conditions.1.2 Satisfy the requirements of UFSAR Table 14.2-2, Sheet 31,Automatic Reactor Control System Test Summary.2.0 Test MethodsWith the Reactor Gontrol System (i.e. Rod Control) in manualand the Reactor Coolant System (RCS) at steady state conditions, RodControlwas placed in automatic to demonstrate that steady stateconditions could be maintained.Subsequently, with Rod Control in manual, T"w was varied from theReference Temperature (T,"fl by approximately +6 oF(+SoF to +7oF), bymanually changing the position of Control Bank D with no deliberateturbine load change. Rod Controlwas then placed in automatic todemonstrate the ability to restore and stabilizeTr- to within a t1.SoF deadband from T,"1via proper positioning of Control Bank D. The same testwas also performed for a T"* change of approximately -6oF(-SoF to -7oF) relative to Trer.The test was performed with reactor power approximately 45o/oto 47% ofRated Thermal Power (RTP) and RCS average temperature, pressurizerlevel and steam generator levels on program. The initialTaw - T,"smismatch was within tloF and RCS pressure was between 2200to 2250 psig.3.0 Test ResultsAllAcceptance/Review Criteria were met or resolved as delineated below.No control system settings were changed based on the performance ofthis test.Figures 7.2.2-1 through 7.2.2-g depict the performance results of theautomatic control systems.L29 7.2.2 Automatic Reactor Gontrol System (2-PAT-6.1) (continued)Acceotance Criteria3.1 No manual operator action or intervention is required to returnthe plant to stable conditions (i.e., auctioneered RCS Taw within11.soF of T,"r) for both steady-state and transient conditions.No manual operator action or intervention was required.3.2 For steady-state operation, and for both increasing anddecreasing T"*temperature transients, the Automatic RodControl System responds properly to automatically positioncontrol rods and return auctioneered RCS T"r, to within +1.soFof Tr*when the ARCS is placed in AUTO control mode.Rod Control properly responded to steady state and transientconditions to return T"r, to within 11.5"F of T'er.Review Criteria3.3 Pressurizer pressure tracks the response of auctioneered T"r,during the T"o transient tests and is controlled back toapproximately 2235 psig due to automatic pressurizer pressurecontrol.Pressurizer pressure tracked the response to Tavg and controlledback to approximately 2235 psig.3.4 Pressurizer level and level setpoint track the response ofauctioneere6 Tave during the T"o transient tests due to automaticpressurizer level control.Pressurizer level and level setpoint tracked the response to T"*.130 7.2.2 Automatic Reactor Gontrol System (2-PAT-6.1) (continued)4.0 Problemst1l CR 1190719 was written for two procedure deficiencies on2-PAT-6.1, Automatic Reactor Control System.Steps 6.2[8] and 6.3[12] said to ENSURE the passive summerindicated 72 steps/min. The passive summer does not indicate rodspeed. The passive summer indicates an enor signal in DegreesF. Steps 6.2[8] and 6.3[12]should have ENSURED the passivesummer indicated +5oF and - soF, respectively. The error wasidentified, discussed by PAT and Operations, CTL entry entered,and the test was continued. The +/- 5 degrees was verified duringperformance. Also Steps 6.2[10] and 6.3[14] verified that the rodspeed was72 steps/min at the time the rods were placed to auto."Step 6.3[3]1." was a typo and should have been deleted.131
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