ML21153A413

Response to Request for Additional Information Regarding License Amendment Request for Measurement Uncertainty Recapture Power Uprate

ML21153A413
Person / Time
Site:	Millstone
Issue date:	06/02/2021
From:	Gerald Bichof Dominion Energy Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
21-166
Download: ML21153A413 (27)

Text

Dominion Energy Connecticut, Nuclear inc.

5000 DominionBoulevard, Glen VA23060 Allen, D DominionEnergy.com

@Q June2,2021 U.S. Nuclear Regulatory Commission Serial No.21-166 Attention:Document Control Desk NRA/SS RO Washington,DC 20555 Docket No. 50-423 License No. NPF-49 MILLSTONE POWER STATION UNIT3 RESPONSETO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST FOR MEASUREMENT UNCERTAINTY RECAPTURE POWERUPRATE Byletter dated November 19,2020(Agencywide Documents Access andManagement System (ADAMS) Accession No.ML20324A702), Dominion Energy Nuclear Connecticut, ,

Inc. (DENC), submitted a license amendment request (LAR) totheNuclear Regulatory Commission (NRC) for M illstonePower Station,U nit No. 3 (MPS3). The proposed license amendment would increase therated thermal power (RTP) level from 3,650 megawatts thermal (MWt) to3,709 M Wt i n the M PS3 operating license and in Technical Specification (TS) 1.27, anincrease inRTPofapproximately 1.6%. The proposed increase isreferred  !

toasa measurement uncertainty recapture (MUR) power uprate andis based onutilizing aninstalled Cameron Technology USLLC(currently knownas Sensla, formerly known asCaldon) Leading EdgeFlowMeter CheckPlus system .as a n ultrasonic flow meter located ineachofthefour mainfeedwater lines supplying thesteam generators to improve plant calorimetric heat balance measurement accuracy. Theproposed changes would also involve aneditorial correction toTS2.1.1.1 andrevision toTS 3.7.1.1, Action Statement "a" andTSTable 3.7-1, "Operable MSSVs Versus Maximum Allowable Power" toupdate the maximum allowable power levelscorresponding tothe numberof operable i main steam safety valves persteam generator.

Inanemail dated April 8,2021, theNRCissued a draft request for additional information (RAI) related to the proposed LAR. On April 20, 2021, t he NRC staff conducted a l conference call with DENCstaff toclarifytherequest. Inanemail dated April 22,2021, the NRCtransmitted thefinal versionoftheRAl(ADAMS Accession No.ML21112A308). ,

DENCagreed torespond totheRAIwithin 45days ofissuance, ornolater than June7,  !

2021, I

I provides DENC's response tothe RAl. Attachment 2provides asupplement '

toclarify that themethodology selected forDENC's planned LossofNormal Feedwater (LONF) reanalysis has changed from w hat ispresented in MUR LAR, Attachment 4, Sections II.1.10 andIII.4-1.1 Instead,theLONFevent revision will beperformed using analternate, NRC-approved, methodology. l l

Serial No.21-166 Docket No.50-423 Page2 of3 Ifyou have anyquestions orrequire additional information, please contact Shayan Sinha at(804) 273-4687.

Sincerely, Gerald T.Bischof Senior Vice President -

Nuclear Operations & Fleet Performance COMMONWEALTH OFVIRGINIA )

)

COUNTY OFHENRICO )

Theforegoing document wasacknowledgedbefore me,inandfor County the andCommonwealth ,

I today aforesaid, byMr. Gerald T.Bischof, whois Senior Vice President-Nuclear Operations and FleetPerformance ofDominion Energy Nuclear Connecticut, Inc.Hehasaffirmed before methat heisduly authorized toexecute andfilethe foregoing document inbehalf ofthat company, and thatthestatements inthedocument aretrue tothebest of his-knowledge andbelief.

Acknowledged beforemethis 2 dayof , 2021 MyCommission Expires:

N taryu CRAIG D SLY Notary Public Commonwealth ofVirginia Reg. # 7518653 MyCommissionExpires December31, g

20-Attachments:

1, Response toRequest for Additional Information Regarding License Amendment Request

2. Supplement forMeasurement toClarify Uncertainty Methodology Recapture forRevised Power Uprate LossofNormal Feedwater l

Reanalysis i

Commitments madeinthis letter:None i

I

Serial No.21-166 Docket No.50-423 Page3of3 cc U.S. Nuclear Regulatory Commission Region I 2100 Renaissance Blvd, Suite 100 King of Prussia, PA19406-2713 R.V.Guzman Senior Project Manager U.S. NuclearRegulatory Commission OneWhite FlintNorth, Mail Stop08-C2 11555 Rockville Pike Rockville, MD20852-2738 NRCSenior Resident Inspector Millstone Power Station Director, Radiation Division Department ofEnergy andEnvironmental Protection 79ElmStreet Hartford, CT 06106-5127 j

Serial No.21-166 Docket No.50-423 ATTACHMENT1 POWER MILLSTONE STATION UNIT3 DOMINION ENERGYNUCLEAR CONNECTICUT, INC.

No.21-166 Serial DocketNo.50-423 1,Page1 of21 Attachment Byletter dated November19, 2020(Agencywide Documents Access andManagement System (ADAMS) AccessionNo.ML20324A702), Dominion Energy NuclearConnecticut, Inc.(DENC), submitted a license amendment request (LAR) totheNuclear Regulatory Commission (NRC) for MillstonePower Station, UnitNo.3(MPS3). Theproposed license amendment would increase therated thermal power (RTP) levelfrom3 ,650m egawatts thermal (MWt) to3,709 MWtin the MPS3operating license andinTechnical Specification (TS) 1.27, an increase in RTP of approximately 1.6%. The proposed increase isreferred toasa measurementuncertainty recapture(MUR) power uprate andisbased onutilizing aninstalled Cameron Technology USLLC(currently knownasSensia, formerly known asCaldon) Leading EdgeFlow Meter CheckPlussystem asan ultrasonic flow meter located ineachofthefourmain feedwater lines supplying thesteamgenerators to improve plant calorimetric heat balance measurement accuracy. Theproposed changes would also involve aneditorial correction toTS2.1.1.1 andrevision toTS3.7.1.1, Action Statement "a"andTSTable 3.7-1, "Operable MSSVs Versus Maximum Allowable Power" toupdate themaximum allowable powerlevels corresponding tothe number ofoperable main steam safety valves persteam generator.

Inanemail dated April 8,2021, the NRCissueda draft request for additionalinformation (RAI) related totheproposed LAR.On April 20, 2021, theNRCstaff conducted a conference call with DENCstaff toclarify therequest. In anemaildated April22,2021, theNRCtransmitted the finalversion oftheRAI(ADAMS Accession No.ML21112A308).  :

DENCagreed torespond totheRAIwithin 45days ofissuance, ornolater than June7, 2021 Thisattachment provides DENC's response tothe RAI .

Theregulation at10CFR50.55a requires thatthereactor pressure vessel (RPV) be constructed, designed andanalyzed inaccordance with theAmerican Society of Mechanical Engineers Boiler andPressure Vessel Code(ASME Code), Section Ill.

Theregulation at10CFR50.61 requires pressurized thermal shock (PTS) evaluations to ensure that adequate fracture toughness exists forRPVbeltline materials inpressurized water reactors (PWRs) toprotect against failure during aPTSevent. Fractureresistance ofRPVbeltline materials during PTSevents isevaluated bycalculating thenil-ductility temperature (RTNDT) for PTS(identified asRTPTS). Section 50.61(b)(1)requires that PWRlicensees haveprojected values ofRTPTSaccepted bytheNRCfor eachRPV beltline material. Section 50.61(c)(2) requires thatRTPTS calculations forR PV beltline materials incorporate credible RPVsurveillance material test data thatarereported as partofthe RPVmaterials sunteillance program required by10CFRPart 50,Appendix H.

Serial No.21-166 DocketNo.50-423 Attachment 1,Page2 of21 NVIB-RAl-1 SectionIV.1.A.ii.e, "Mechanical Evaluation," inAttachment 4 ofthe LARstates that the MURpower uprate designconditions donotaffect thecurrent design bases for seismic andloss-of-coolant accident (LOCA) loads. Thelicensee furtherstated thatthestress levels caused by the flowinduced vibration on thecore barrel assembly andupper internals arelowand remain well below thematerial high-cycle fatigue endurance limit.

Summarize themechanical evaluation that demonstrates thecore barrel assembly and upper internals arenotaffected bytheMURpower uprate design conditions, includinga discussion onthe acceptance criteria, resulting stresses andfatigue endurance limits.

Thegoal ofanalyzing flow-induced vibration (FIV)onthe corebarrel assembly andupper internals istoassess its impact onstructural integrity ofthecomponents contained within theMPS3reactor vessel. Extensive evaluations were previously completed insupport of theMPS3stretch power uprate (SPU) LAR, which involved a Rated Thermal Power (RTP) increase of7%.TheMPS3MURpower uprate LAR, which is currentlyunder NRCreview, further increases theRTP level byapproximately 1.6%. TheMURpowerincrease considers changing theoperating power plus uncertainty butremainsbounded bythe SPUoperating power plus uncertainty which waspreviously analyzed. Based onthis understanding, a comparison totheevaluations completed for the SPUincrease was relieduponfor justification oftheMURincrease.

Themethodology followed intheSPUevaluation inSPULARAttachment 5 (Reference 1-1), Section 2.2.3 involved scaling thestructural response ofthe FIV according to analytical andexperimental formulations. Specifically, byrelating parameters such asflow rate (Mechanical Design Flowor"MDF"), vessel inlet temperature, andvessel outlet temperature. Thechange ininputparameters isshown tobenegligible (<1%), so the FIV evaluation completed for theSPUprogram wasconsidered tobeapplicable tothe MUR program, asthese parameters arethe only change noted inthisevaluation between the twoprograms.

TheSPUevaluation included thereactor internals components that areconsidered tobe limitingwith regards toFIV. These components consist ofthecorebarrel inthelower internals assembly andthe guidetubes inthe upper internals assembly. Forother reactor internalcomponents such asthelower radial restraints, upper core plate alignment pins, lower support plate, andthelower support columns, thevibratory response isextremely small. Thecalculated stresses wereobtained byscaling previously generated FIV stresses towhatwould beexpected after SPUimplementation. Forconservatism, the values werescaled based onthehot functional flow rates, which aretypically about 4%

higherthan theMDFrate.

Furthermore, the seismic andLOCAanalyses forthe MPS3SPUprogram areconsidered tobeapplicable totheMURprogram. Theseismic inputs andplant model parameters from theSPUprogram arealso applicable totheMURpoweruprate. Additionally,the

Serial No.21-166 DocketNo.50-423 Attachment1,Page3of21 conservative assumptions inthe LOCAforce calculations bound thechanges foundinthe MURprogram. Therefore, existing SPUparameters were determined tobeapplicableto theMURconditions.

TheASMECode (1998 Edition with 2000Addenda, Section III, Division 1andSection II, Part D)combined with measured data, forms thebasis for theacceptance criteria for mechanically induced stresses/strains produced by FIV.Although theparticular components analyzed do not constitute a pressure boundary, theacceptability ofthe components impacted by the uprate relating toalternating stresses for high-cyclefatigue wasassessed. Thealternating stresses werecalculated andcompared accordingtothe ASMECode rules onhigh-cycle fatigue ormeasured experimental dataonstrains limits inthe MPS3SPULAR.

NVIBRAl-2 Section IV1.A.ii.f, "Structural Evaluation," in Attachment 4 oftheLAR states that evaluations were performed todemonstrate thatthe structural integrity ofreactorinternal components isnotadversely affected bytheMURpower uprate design conditions.The NRC staff demonstrates requests thereactor thelicensee internal (a) to summarize the structural evaluation components arenotadversely affected bytheMUR that power uprate design conditions, including a discussion onthe acceptance criteria and resulting stresses and(b) discuss whether there areanycracksin any ofthe RPVinternal components. If there areanycracks, discusswhether anevaluation has beenperformed andhow this evaluation demonstrates sufficient structural integrity of thedegraded reactor internal component under theMURpower uprate conditions.

M M  !

Extensive evaluations werepreviously completed insupport oftheMPS3SPULAR, which involved an RTPincrease of7%.TheMPS3MURpower uprate LAR,which is currently under NRCreview, further increases theRTPlevel byapproximately 1.6%.The MURpower increase considers changing theoperating powerplus uncertaintybut remains bounded bytheSPUoperating power plus uncertainty which waspreviously analyzed. Based onthis understanding, a comparison totheevaluations completed for theSPUincrease wasrelied upon for ofthe justification MURincrease.

Theinputs for all components remained unchanged, thus nonewcalculations were performed, andthe results from the SPUcalculations weredocumented inthe SPULAR.

Five areas wereevaluated, including thefollowing: Control RodInsertability Evaluation, Baffle Bolt Evaluation, FIVEvaluation, Critical Reactor Internal Components Structural Evaluation, andan Upper andLowerCorePlate Evaluation. TheFIVevaluation is addressed inNVIB-RAl-1, theUpper andLower CorePlate Evaluation isaddressed in

Serial No.21-166 DocketNo.50-423 Attachment 1,Page4of21 NVIB-RAl-3, andtheBaffle Bolt Evaluation isaddressed inNVIB-RAl-4. Further discussion ontheremainingtwoevaluations isincluded below.

M This evaluation determined themaximum mechanical load actingontheguide tubes. This load wasthen compared totheallowable load ontheguide tubes. Themethodology assessed thepredicted lateral load/displacement against drop tests that applied lateral loads/deflections tothe guide tube toconfirm control rodinsertion.These testsconsidered the effects ofboth temporary (elastic) andpermanent (plastic) deformation.

Theapplicable loadings that were involved inthis evaluation include massflow and acoustic loads,system loads, and safe shutdown earthquake (SSE) loads. These t hree loads, aspreviously generated forthe SPU program, were deemed tobeapplicable for useintheMURprogram. Thisapplicability wasconfirmed based oneither bounding or unchanging inputs tothe specified loading conditions.

MPS3usesa 17x17, 96-inch style guide tube. The allowable loads andacceptance criteria for thisstyle ofguide tube andcontrol rod insertion were shown tobemetfor the SPUanalysis documented inSPULARAttachment 5 (Reference 1-1), Section 2.2.3 (which bound M UR conditions).

Thepurpose ofthis evaluation was toassess theimpactof the MURpower uprate program onthe MPS3reactor internal components includingthelower coresupport plate atypical region,lower support columns, andcore barrel nozzle weldments. Thefollowing methodology was previously usedfortheSPU program evaluation inSPU LAR ,Section 2.2.3:

1 Determine thechanges inload conditions duetothepower uprate. Generally, the poweruprate would change theloads ontheinternal components during normal andupset conditions duetoHeat Generation Rates (HGR) from gammaheating andthermal fluidtransients.

2.Compare the MPS3core support component design configurations toother design configurations that havebeenusedfordetailed stressanalyses, which were I performed under similar loading conditions. Anydifferences indesign dimensions needtobereconciled inthecomponent stress evaluation.

if there areanymajor design differences, component stresses needtobedetermined independently.

3.Compare thermal loadings (thermal transients andHGRs) usedincore support component analysis, which wereperformed under loading similar conditions tothe thermal loadings for MPS3uprate conditions.

4. Determine ifthethermal loadings ofsimilar components indetailed stress analyses, which wereperformed under similar loadingconditions bound theMPS3 thermal loadings.

Serial No.21-166 Docket No.50-423 Attachment 1,Page5 of21 The primary inputs tothese evaluations include heatgeneration rates anddesign transients. Both ofthese inputs aredeemed tobeapplicable toboth theSPUandMUR programs, due toconservatisms included intheSPUanalysis.

No plant-specific ASME Codestress waswritten report forMPS3.Thereactor internal components were analyzed tomeetthe oftheASMECode,Section intent IIIcriteria.

Therefore, basedon the previous evaluations andcurrent practices, theguidelines in Subsection NG ofthe ASME Codewereusedfor this evaluation. Inconclusion, the allowable loads andacceptance criteria for the internalcomponents analyzed wereshown tobemetfor theSPUanalysis (which bound MURconditions).

Aspart ofthe10-year visual examinations, DENC hasnotidentified cracking inanyMPS3 RPVinternal components.

NVIB-RAl3 Section "Upper IV.1.A.ii.g, andLower CorePlateStructural Analysis," inAttachment 4 of theLAR states that thermal design transients, heatgeneration rates, andoperating conditions affect thermal loads ontheupper andlower core plates. Thelicensee stated that for the MURpower uprate, current ofrecord analysis (AOR) thermal design transients andheat generation rates remain applicable because the MURpower uprate operating conditionsarebounded bytheoperating conditions inthe current AOR.The licensee further stated that the maximum primaryplus secondary stress intensity ofthe upper andlower core plate andcumulative usage factor remain acceptable. Summarize howthe existingstructural analysis for theupperandlower core plates isstill applicable under the MURpower uprate conditions, adiscussion including onthe acceptance criteria andresulting stresses.

M Extensive evaluations werepreviously completed insupport oftheMPS3SPULAR, which involved an RTPincrease of7%.TheMPS3MURpower uprate LAR,which is currently under NRCreview, further increases theRTPlevel byapproximately 1.6%. The MURpowerincrease considers changing theoperating powerplus uncertainty but remains bounded bytheSPUoperating power plus uncertainty which was previously analyzed. Based onthis understanding, a comparison totheevaluations completed for theSPUincrease wasrelied upon for oftheMURincrease.

justification Thepurpose oftheUpper andLower Core Plate Evaluation wastoassess the impact on structural integrityofMPS3 reactor internals lowercore p (LCP) late and upper core p late (UCP) with regard totheproposed MURpower uprate program. Themethodology for qualifying thestructural integrity oftheMPS3UCPandLCP was investigated to determine thedriving inputs which produce thestresses ontheupper and lower core plates. Itwasdetermined thattheLCPandUCPgeometry andmaterial , coreplate

SerialNo.21-166 Docket No.50-423 Attachment 1,Page6of21 supports, HGRs andloads, fluid-thermal loads, thermal design transients (including number of cycles) andmechanical loads areall usedinthe stresscalculation fortheLCP andUCP. For theMURpower uprate, itwaseither determined that theinputs listed have notchanged, that the change observed isnegligible, orthat theobserved change is bounded bythe AOR.

Theoriginal LCPand UCP evaluation documented effects onthe SPUprogram forMPS3 inSPULARAttachment 5 (Reference 1-1), Section 2.2.3. This evaluation madeuseof analyses completed for a similar plant during a replacement steamgenerator (RSG) program. Therefore, thestructural qualification andfatigue evaluation istheoriginal design basis oftheMPS3reactor internals. Based onsimilarities indesign andthermal loading, the evaluation completed for theLCPandUCPatthesimilar plant wasused in the current analysis for theMPS3 SPU program.

Theallowable stresses areobtainedfrom theASMECodeRequirements (1974 edition, Division 1,Section Ill,Subsection NG) which isconsistent withthe original evaluation.

Forthis evaluation theallowable stress, as prescribed bythe 1974 ASMEBoiler Code, wereused astheacceptance criteria.

NVIB-RAl-4 Section IV.1.A.ii.h, "Baffle-Barrel Region Evaluations," inAttachment 4 oftheLARstates that thebaffle bolts aresubjected toprimary loads consisting of deadweight, hydraulic pressure differentials, LOCAandseismic loads, andsecondary loads consisting of preload andthermalloads resulting from reactor coolant system (RCS) temperatures and gammaheating rates. Thelicensee stated that itevaluated thebaffle former bolt maximumdisplacement attheMURpoweruprate design conditions. The licensee concluded that theexisting thermal andstructural analysis ofthebaffle-barrel region results remain bounding for theMURpower uprate design conditions. TheNRC staff requests ofthebaffle thelicensee barrel region remain (a) tosummarize howthe bounding existing under thermal andstructuralanalysis theMURpoweruprate design conditions, including a discussion onthe baffle former bolt maximumdisplacement and stresses under theMURpower uprate design conditions and(b) theinspection history andresults offormer baffle bolts andplates.

M Extensive evaluations werepreviously completed insupport oftheMPS3SPULAR, which involved anRTPincrease of7%.TheMPS3MURpower uprate LAR,which is currently under NRCreview, further increases theRTPlevel byapproximately 1.6%. The MURpower increase considers changing theoperating power plus uncertainty but remains bounded bytheSPUoperating powerplus uncertainty which waspreviously

SerialNo.21-166 Docket No.50-423 Attachment 1,Page7 of21 analyzed. Based onthis understanding, a comparison totheevaluations completed for the SPU increase wasrelied uponfor justification oftheMURincrease.

Thebaffle bolt evaluation wasusedtoassess thestructural acceptability ofthebaffle-former boltsfor the MPS3MURpower uprate program. Themethodology usedinthe SPUpower uprate evaluation inSPULARAttachment 5 (Reference 1-1),Section 2.2.3 determined thecumulative fatigue damageresulting fromthethermal loading. This cumulative fatigue damage factorwasthen compared tothe allowable factor. Calculation ofthedamage factor was completed using thefollowing methodology:

1.Thegeometry oftheMPS3 baffle-former bolts wasreviewed andcompared tothe bolt geometry usedtodetermine a design fatigue curve. This fatigue curve was developed from test data used for qualification ofthestandard four-loop, upflow, baffle-former bolts. Thegeneration ofthis fatigue curve included themargins specified intheASMEBoilerand Pressure Vessel Code, Section Ill,Division 1, 1998Edition, forusing testresults to qualify a component. It wasassumed that the fatigue curve hadthesame"shape"as the design fatigue curve forstainless steel.

This isconsistent with theprocedure provided intheASMECode, Appendix for II, performing fatigue tests using accelerated loadings.

2. Baffle-barrel temperatures weredetermined for normal and upsetservice conditions. Thisanalysis addressed heating rates associated withthehistoric and projected fuelloading patterns.

3.Displacements ofthebaffle plates weredetermined atthe bolt locations, using the previous fatigue curve andthetemperatures determined in Steps 1 and2.Using linear scaling, displacements werealso determined forother applicable load conditions.

4. Fatigue usage wasdetermined using Miner's Rule.

Theinputs tothis evaluation included RCStemperature andflow parameters for MUR andSPU conditions, thefour-loop, upflow, baffle-bolt design fatigue curve, design transients, HGRs,andbaffle-barrel temperatures. An evaluation oftheaforementioned inputs wascompleted todetermine applicability oftheSPUparameters for theMUR program. Allofthelisted inputs weredeemed tohave not changed andareapplicableto theMURprogram for MPS3.Therefore, theresults fromtheSPUprogram arealso applicable totheMURpower uprate program.

Theconclusion reached isthat anacceptable damage factorwascalculated fortheSPU program, andthis remains applicable totheMURpower uprate program. Therefore, the baffle-former-barrel configuration isacceptable for theeffects oftheMUR, andthe calculated usage iswell below theallowable fatigue usage limit of1.0. Notethat this analysis considered a full baffle-former bolting pattern. No Acceptable Bolting Pattern Analysis (ABPA) methods wereapplied intheSPUortheMURevaluations.

Serial No.21-166 DocketNo.50-423 Attachment1,Page8 of21 TheMPS3 reactor vessel core supportbarrel former baffle platesandbaffle boltinghave beensubjected tooneVT-3 examination during each 10-yearISIinterval.

Thefirst interval examwasperformed duringthe 3R05outage (June 1995),thes econd interval exam was performed during the 3R11outage (April 2007), andthethird interval examwas performed during the 3R17 outage (April 2016). No relevant indicationswereidentified onthecore support barrel baffle platesorbaffle bolting duringanyofthese exams.

NVIBRAl-5 SectionIV.1.A.iii.a, "Bottom Mounted Instrumentation inAttachment (BMI)," 4 ofthe LAR discusses the stress analysis oftheBMI guide tubes. Thelicensee stated thatthe range ofvessel core inlet temperatures theMUR power uprate for is536.7 degrees Fahrenheit

(*F)to 555.8*F, which islower than the RPV coreinlet temperature intheexisting analysis.These temperatures arebounded by the BMI guide tube design temperature of 560'F.Clarify whether the existing stress analysis of theBMIguide tubes isbased onthe designtemperature of560*F.

M TheBMIguide tubes stress analysis isbased onthedesigntemperature of560*F. This temperature isbounding ofboth thepre-MUR andpost-MURpower uprate core inlet temperatures.

NVIB-RAl-6 SectionIV.1.C.i, "Pressurized ThermalShock (PTS) inAttachment 4 ofthe Calculations,"

LARstates that thelimiting reference temperature PTS(RTPTS) for value of130*F toLower applies Shell Plate B9820-2. Thelicensee statedthatthisisa change fromthe AORthat hada limiting RTPTSvalue of133*F toIntermediate pertaining Shell Plate B9805-1. TheNRCstaff theRTPTSvalue from requests 133*F thelicensee to130"F (a) to clarify andthelimiting thecause ofthechanges material, beltline and(b) in discuss whether theRTPTSvalue of130*F wasderived basedon theMURpoweruprate conditions.

l

SerialNo.21-166 Docket No.50-423 Attachment 1,Page9 of21 W

h Per10CFR50.61,"Fracture toughness requirements for protection against pressurized thermal shock events [Reference 1-3)," theuseofresults fromtheplant-specific program surveillance may result in anRTPTS Value that ishigher orlower than the value calculatedwithout plant-specific surveillance data.Asdescribed inMPS3FSARSection 5.3,themethodology described in Regulatory Guide (RG) 1.99, Revision 2 [Reference 1-6)isperiodicallyupdated toincorporate theeffects ofirradiation exposure using in reference legtemperature calculations. RG 1.99includes positions that determine Chemistry Factor(CF)xwithout theuse of surveillance data (Position 1.1) orwit.h theuse ofsurveillance data (Position 2.1).

TheAORfor pressurized thermal shock (PTS) forMPS3 is contained inTable 2.1.3-4 of theSPULAR,Attachment 5 [Reference 1-1).Thelimiting RTPTs value (consistent with theSPULAR) isshowninTable 5.2-7 oftheMPS3FSAR [Reference 1-2).

Asdescribed Section 2.1.3.2.4 inthe SPULAR,Attachment 5:

Thelimiting material isIntermediate ShellPlate B9805-1, with the morelimiting RTers valueoccurring for calculations using theRG 1.99, Rev. 2 Position 1.1 ChemistryFactor, asopposed tothe Position2.1Chemistry Factorcalculated from credible surveillance data. Themostlimiting RTers value at54EFPY[effective full power years)for Plate B9805-1 is133F.

Thus, theAORlimiting RTPTS Value iSConsidered tobe133F,which corresponds to Shell Intermediate Plate B9805-1 evaluated withaPosition 1.1 CF(i.e., calculated without theuseofsurveillance data). However, using thecredible surveillance plate data per Position 2.1,anRTPTS Value of1110F wascalculated for Intermediate ShellPlate B9805-(see 1 intheSPUanalysisTable 2.1.3-4 ofReference 1-1).

Since thesurveillancedata iscredible andthePosition 2.1calculation resultsina lower calculatedRTPTSvalue, thePosition 2.1results could havebeenusedinlieu ofthe Position 1.1 results for theSPUanalysis per10CFR50.61. However, instead oftaking creditforthecredible surveillance data, theSPULARAttachment 5 conservatively the reported Position1.1 CFresult of133Ffor Intermediate Shell Plate B9805-1, instead ofthe111Fvalue from using thePosition 2.1CF.

IftheSPU report hadelected totakecredit forthecredible surveillance datafor IntermediateShellPlate B9805-1, this material would no longer havebeenthemost

Serial No.21-166 DocketNo.50-423 Attachment 1,Page 10of21 limiting material. Instead, thelimiting RTPTS Value would havebeen130Fcorresponding toLower Shell Plate (see B9820-2 Table 2.1.3-4ofReference 1-1).

MPS3MUR RTPTS Evaluation TheRTPTS CalCulations supporting theMUR[Reference 1-4) areshowninTable 1-1. This analysis isbasedon fluence values derived fromMUR poweruprate conditions.

Consideration oftheMUR increased thepeak54 EFPYfluence value from theSPU analysis from a value of2.70 x 1019n/cm2 (SPU) to 2.72 x 1019 n/cm2 (MUR). the Since RTPTS Values arereportedas whole numbers, this slight increase influence didnot change thehighest calculated RTeTS values. Specifically,the RTPTS values corresponding toIntermediate Shell Plate B9805-1 using Position 1.1 (1330F), Intermediate Shell Plate B9805-1 using Position 2.1(111F), and LowerShell Plate B9820-2 (130F) are unchanged fromthefluence increase. It isalso noted that thePosition 2.1CF for Intermediate Shell Plate B9805-1 forboththe SPU andMURutilize thesamesurveillance data, available inWCAP-16629-NP [Reference 1-5). Thecredibility ofthedata isalso assessed inWCAP-16629-NP.

TheMURLARRTPTS analySiS takes creditforthe available credible surveillance data for Intermediate Shell Plate B9805-1, and thus the RTPTS Value CorreSponding tothismaterial is111F. The133Fresult based onthe Position1.1 CFis shown for information only. As a result, the limiting RTPTS Value is130F, which correspondsto Lower Shell PlateB9820-2.

Thechange inthelimiting RTPTS Value andmaterial isthe result ofdecisions madewith theimplementation of10CFR50.61 with respect tosurveillance data,and is not a result ofchanges influence projections, surveillance data available since the SPU LAR, and/or material properties. Thechange inthelimiting RTPTS Value from 133Fto1300F andthe limiting beltline material complies with 10CFR50.61 andthePositions ofRegulatory Guide 1.99, Revision 2.

i l

Asdescribed intheresponse toNVIB-RAl-6, part(a), theRTPTS Value of130Fconsiders i thefluence value derived from MURpower uprate conditions. Thevalue of130Fwas l unchanged from theanalysis ofrecord based onthefluence increase resultingfrom the j MURpower uprate conditions.  !

i

SerialNo.21-166 Docket No.50-423 Attachment 1,Page11of21 Table 1-1 MUR RTPTS Calculations for theMPS3Beltline andExtended Beltline Materials at54EFPY00 R.G.

Reactor Fluence(c) Initial 1.99, CF(b) (b) ARTl@ GU GA Margi RTers Vessel Reactor Vessel Material (n/cm2, E > 1.0 Me FF(c)RTmrr Rev. 2 (F) r(F) (F) (0F)n (0F)(0F)

Location Position V) (F) o Intermediate Shell Plate B9805-1

.3 0 U W m 2.1 26.7 2.72 x 1019 1.2671 60 33.8 0 8.5(d) [7g (([

Intermediate Shell Plate B9805-2 1.1 31.0 2.72 x 1019 1.2671 10 39.3 0 17 34.0 83 Intermediate Shell Plate B9805-3 1.1 31.0 2.72 x 1019 1.2671 0 39.3 0 17 34.0 73 Lower Shell Plate B9820-1 1.1 51.0 2.72 x 1019 1.2671 10 64.6 0 17 34.0 109 Lower Shell Plate B9820-2 1.1 44.0 2.72 x 1019 1.2671 40 55.8 0 17 34.0 130(0 Beltline LowerShell Plate B9820-3 1.1 37.0 2.72 x 1019 1.2671 20 46.9 0 17 34.0 101 Intermediate Shell Longitudinal1.1 31.8 2.72 x 1019 1.2671 -50 40.3 0 20.1(e) 40.3 31 Weld Seams 101-124 A,B,C 2.1 6.7 2.72 x 1019 1.2671 -50 8.5 0 4.2(e)8.5 -33 Intermediate toLower ShellGirth 1.1 31.8 2.72 x 1019 1.2671 -50 40.3 0 20.1(e) 40.3 31 WeldSeam101-171 2.1 6.7 2.72 x 1019 1.2671 -50 8.5 0 4.2(e) 8.5 -33 Lower Shell Longitudinal Weld 1.1 31.8 2.72 x 1019 1.2671 -50 40.3 0 20.1(e) 40.3 31 Seams 101-142 A,B,C 2.1 6.7 2.72x 1019 1.2671 -50 8.5 0 4.2(e)8.5 -33 Nozzle Shell Plate B9804-1 1.1 31.0 0.0814 x 1019 0.3770 40 11.7 0 5.8(e) 11.7 63 Nozzle Shell Plate B9804-2 1.1 51.0 0.0814 x 1019 0.3770 20 19.2 0 9.6(e) 19.2 58 Nozzle Shell Plate B9804-3 1.1 31.0 0.0814 x 1019 0.3770 0 11.7 0 5.8(e) 11.7 23 Inlet Nozzle B9806-3 1.1 58.0 0.0814 x 1019 0.3770 10 21.9 0 10.9(e) 21.9 54 Extended Inlet Nozzle B9806-4 1.1 58.0 0.0814 x 1019 0.3770 0 21.9 0 10.9(e) 21.9 44 Beltline Inlet Nozzle R5-3 1.1 44.0 0.0814 x 1019 0.3770 -10 16.6 0 8.3(e) 16.6 23 Inlet Nozzle R5-4 1.1 51.0 0.0814 x 1019 0.3770 0 19.2 0 9.6(e) 19.2 38 Nozzle Shell Longitudinal Weld 1.1 39.8 0.0814 x 1019 0.3770 -10 15.0 0 7.5(e) 15.0 20 101-122A Nozzle Shell Longitudinal Weld 11 39.8 0.0814 x 1019 0.3770 -50 15.0 0 7.5(e) 15.0 -20 101-122B, 101-122C

.-...............-~.... ----..------ ---..... . --.----.--.------.--..---..- -.-..--------........ ...........- ............... . .. .... . . .. .. . ... ..-.--. ..---... . . . . .

SerialNo.21-166 Docket No.50-423 Attachment 1,Page12of21 R.G.

Reactor 1.99, CFW Fluence(c) Initial Margi RTPTS Vessel Reactor Vessel Material (n/cm2, E > 1.0 Me FF(c)RTNDT

• ARTr ou ca Location Rev. 2 (F) V) (F) o T(F) (0F)(0F)n (F) (F)

Position NozzleShell toIntennediate Shell 1.1 41.0 0.0814 x 1019 0.3770 -40 15.5 0 7.7(e) 15.5 -9 Girth Weld 103-121 Inlet Nozzle Weld 105-121A 1.1 45.3 0.0814 x 1019 0.3770 -60 17.1 0 8.5(e) 17.1 -26 Inlet Nozzle Weld 105-121B 1.1 75.4 0.0814 x 1019 0.3770 -50 28.4 0 14.2(e) 28.4 7 Inlet Nozzle Weld105-121C 1.1 75.4 0.0814 x 1019 0.3770 -50 28.4 0 14.2(e) 28.4 7 Inlet Nozzle Weld105-121D 1.1 75.4 0.0814 x 1019 0.3770 -50 28.4 0 14.2(e) 28.4 7 Notes:

(a) The10CFR50.61 [Reference1-3) methodology wasutilized inthe calculation ofthe RTers values.

(b)Values areconsistent with those utilized inthe analysis ofrecord, the SPUevaluation [Reference 1-1).

(c) Maximumfluence values for the beltline (2.72 x 1019 n/cm2) andextended beltline (8.14 x 1017 n/m2) considering MURconditions areconservatively used for every material inthe region. FF= fluence factor ascalculated withthe 10CFR50.61 [Reference 1-3) methodology.

(d)A reduced catermisused since the surveillance datais deemed credible WCAP-16629-NP per [Reference 1-5).

(e) Per10CFR50.61 [Reference 1-3), caneed not exceed 0.5

• ARTuDT. Therefore, the cahasbeen reduced.

(f)Thelimiting RTrrs value for MPS3is130F, which corresponds toLower Shell Plate B9820-2. Note that Intermediate Shell Plate B9805-1resulted ina higher RTers value of133Fwhenthe surveillancedata was not used; however, the RTPTs value for Intermediate Shell Plate B9805-1 is111Fwhenthe credible surveillance data isused. Thus, taking credit for the credible IntermediateShell Plate B9805-1 surveillanc'e data, the limiting material MPS3is for Lower Shell Plate B9820-2.

No.21-166 Serial DocketNo.50-423 Attachment 1,Page13of21 Appendix A to 10CFR50establishes minimum criteria (General Design CriteriaorGDC) forthesafeoperation oflight water reactors. GDC10, "Reactor Design", requires that the reactorcore and associated coolant, control, andprotection systems bedesigned with appropriatemargin to assure that specified acceptable fueldesign limits arenot exceeded duringanycondition of normal operation,including the effects ofanticipated operational occurrences. Fuel design limits are challenged bytransients described intheMPS3 Updated Final Safety Analysis Report (UFSAR) sections 15.1.3, 15.3.1, 15.3.2, 15.4.3, and15.6.1.

GDC15,"Reactor coolant system design", requires that reactor coolant system and associated auxiliary, control, andprotection systems bedesigned with sufficient margin toassure thatthedesign conditionsof the reactor coolant pressure boundary arenot exceeded during anycondition ofnormal operation, including anticipated operational occurrences. UFSARsections 15.2.6, 15.2.7, and 15.3.2 describes analyses of anticipatedoperational occurrences which could challenge the reactor coolant pressure boundary.

GDC28,"Reactivity limits", requires reactivity control systems tobe designed with appropriatelimits onpotential reactivityincreases sothe effects ofa postulated rod ejection accident canresult inneither damage tothe reactor coolant pressure boundary norresult insufficient disturbance toimpair thecore cooling capability. Thetransient describedinUFSARsection 15.4.8 helps todemonstrate that this criterion is met.

GDC31,"Fracture prevention ofreactor coolant pressure boundary", requires that the reactorpressure boundary be designed with sufficient margin toensure that the probabilityofrapidly propagating failure isminimized under postulated accident conditions.UFSARsections and15.2.8 and15.3.3 describes analyses ofpostulated accidentswhich could challenge thereactor coolant pressure boundary.

RegulatoryInformation Summary (RIS) 2002-03 (ADAMS Accession No.ML013530183) providesguidance toaddressees onthescope anddetail ofinformation thatshould be providedtoNRCfor reviewing MURpower uprate applications. Theguidance states that inareasfor which existing AORbound plant operation atthe proposed power level, the will staff notconduct a detailed review.

Naturalcirculation cooldown isa portion ofthetransients described inUFSARsections 15.2.7, 15.2.6, 15.2.8, and15.3.2. Thecurrentlicensing analysis forthis eventdocuments compliance with the guidance inBranch Technical Position (BTP) 5-1.

Serial No.21-166 DocketNo.50-423 Attachment 1,Page14of21 SNSB-RAl7 Thelicensee described thepowerlevel assumed whenanalyzing accidents and transientsfor MPS3 inTable 11-1 ofAttachment4 tothe LAR.However, theNRCstaff notes that theinformation intheLAR,Table 11-1is.not consistent with the informationin Table 15.0-2 ofthe UFSAR, Revision 33.Specifically, there isa discrepancy for the thermal power level listed for accidents inUFSAR sections 15.1.3, 15.3.1,15.3.2,15.4.3, 15.4.8, 15.6.1, andthe DNB analysis inUFSARsection 15.3.3. Toensure thatvarious analyses canbe accepted without detailed review, please explain thediscrepancy between thepower listed in Table 15.0-2 ofthe UFSARandTable 11-1ofAttachment 4 to theLARfor accidents described in UFSAR sections 15.1.3, 15.3.1, 15.3.2,15.4.3,15.4.8, 15.6.1,andtheDNBanalysis in UFSAR section15.3.3. Foranyofthese cases, ifthe discrepancy exists because a portion oftheanalysiswasperformed assuming a power levelthat doesnotbound theuprated power level, please provide a justificationor summarize theupdated analysis.

M SNSB-RAl-7 identifies discrepancies betweenthe power level informationlisted inthe MPS3MURLARAttachment 4,Table andMPS3FSAR Table 15.0-2.

ll-1 DENC's review ofthe tables identified twoprimary between differences the reported powers. FSAR First,  !

Table 15.0-2 inconsistently presents initial power information. Second, FSARTable 15.0-2 includes power levels reflecting thetransient analyses that provide input totheMUR DNBdesign basis. These transient analyses arebased ona 3666 MWt NSSSpower.

l Regarding Item 1,three general inconsistencies wereidentified inthe FSAR Table15.0-2 reported powers. Theinconsistencies (a) include:lack ofuseofanexplicit identifier forcore power, (b)lack ofinclusion MURpowerlevel ofthepower uncertainty, with (c) and lackof inclusion of a DNB acceptance theconservative (3712MWt) forevents criterion. Theissue wasentered intoDENC's corrective action system andislimited to theFSARtable. TheMPS3safety analysesdescribed inFSARChapter 15use the appropriate power levels andarenotimpacted. MURLARTable 11-1hasbeenreviewed andisaffirmed todocument appropriate andconsistent inputs for power level. ,

l Withrespect toItem 2,select differences between FSARTable 15.0-2 andLARTable Il- !

1reflectthe transient analyses that provide statepointoperating conditions intotheMUR DNBdesign basis. Select DNBcalculations wereperformed ata conservative, scaled powerlevel of101.7% of3650MWt (3712 MWt) along with statepoint operating conditions generated bya system transient analysis performed ata NSSSpower level of 3666MWt.The statepoint conditions generatedat 3666 MWt NSSS power h ave been validated as applicable for theMURpoweruprate because theminorpowerlevel difference hasa negligible effect onthelocalcore conditions (temperature, pressure, and flow)relevantthecalculation for ofDNB.Thecombination ofthe system transient operating conditions (based ona NSSSpower of3666 MWt) andthe scaled, conservative power level of3712MWtresults ina conservative calculation ofDNBfor comparison to theDNBacceptance criteria. Theapproach taken ispresently summarized in FSAR

Serial No.21-166 DocketNo.50-423 Attachment 1,Page15of21 Section 15.0.3.1. Asdescribed intheMURLARandFSARSection 15.0.3.1,allDNB events show acceptable DNBperformance. This power isnotapplicable scaling tothe FSARSection 15.1.3 and15.4.3 DNBevents inSNSB-RAl-7.

identified Table 1-2outlines thecorrections required toconsistently initial report power inFSAR Table 15.0-2. Further, Table 1-2 identifiesthose analyses wheretheDNBcalculation wasperformed ata conservative, scaled core power of3712MWtalong withstatepoint operating conditions generated at3666 MWtNSSSpower. EachFSARsection identified bytheNRCascontaining a power discrepancy isincluded. TheFSARSection 15.1.2 is also included since itsassociated FSAR Table15.0-2 entryimpacted is by thepower reporting inconsistencies, andthe transient analysis supportingtheDNBcalculationwas performed at3666NSSSpower.

Forall listed events, thereferencesfor NRC approval and/orNRC-approved methods presented inLARTable II-1 remain accurate and valid.

I i

I

Serial No.21-166 Docket No.50-423 Attachment 1,Page16of21 Table 12:MPS3FSARTable 15.0-2 Corrections andClarifications FSARSection andEvent Current FSAR Corrected FSAR Basis for Correction orDiscrepancy Table15.0-2 Table 15.0-2 initial Power Initial Power MWt MWt Note1 15.1.2-FeedwaterSystem 0 (NSSS)and 0 NSSSand FSARTable 15.0-2 does notcurrently Malfunctions that Result in 3666(NSSS) 3666NSSS identify the conservative, scaled core anIncrease inFeedwater 0 Coreand power assumed inDNBcalculations.

Flow 3712Core (DNB) 15.1.3 Excessive Increase 3666(NSSS) 3712Core(DNB)FSARTable 15.0-2 does notcurrently inSecondary Steam Flow list the conservative core power assumed inDNBcalculations. No power scaling wasapplied as described inMPS3MURLAR Attachment 4,Section IL1.3.

15.3.1-Partial Lossof 3666(NSSS) 3666 NSSS FSARTable 15.0-2 doesnotcurrently Forced ReactorCoolant 3712(not 3712 Core (DNB) Include the "core" identifierforpower Flow identified asacore level. A conservative, scaled core powerinTable power isassumed inDNB 15.3.2-Complete Loss of 15.0-2) calculations.

Forced ReactorCoolant Flow 15.3.3-Locked Rotor-3666(NSSS) 3666NSSS FSAR Table 15.0-2 doesnot currently RodsinDNB 3712(not 3712Core(DNB)include the "core" identifierforpower identified asacore level. A conservative, scaled core powerinTable poweris assumed inDNB 15.0-2) calculations.

15.4.3 RCCA 3666(NSSS) 3712Core(DNB)Although FSAR Table 15.0-2 presently Misalignment 3712(Core) lists the conservative corepower assumed inDNBcalculations, no power scaling wasapplied as described inMPS3MUR LAR Attachment 4,Section II.1.17.

15.4.8 RodEjection 0 (NSSS) and 0 Core and FSARTable 15.0-2 does notcurrently 3650(not 3723Core include the "core" identifier forpower identified asa core level noridentify thatcalorimetric powerinTable uncertainty wasapplied tothe event 15.0-2) analysis.

15.6.1 Inadvertent 3666 (NSSS) 3666NSSS FSARTable 15.0-2 does not currently Opening ofa Pressurizer 3712Core include (DNB)power the c onservative, scaled core orRelief Safety Valve assumed inDNBcalculations.

Note1:Power uncertaintyisincluded intheinitial power for asappropriate each event.

Serial No.21-166 Docket No.50-423 Attachment 1,Page17of21 SNSBRAl-8 Thelicensee provided anevaluation for the naturalcirculation cooling event insupport of theMURpower uprate inSection 111.3-1 ofAttachment 4tothe LAR.Intheir evaluation, thelicensee indicated that the current AORwasperformed at3650MWt,which does not boundthe power level proposed intheMURpower uprate. Thelicensee hasrepeated thenatural circulation cooling analysis atconditions bounding theMURpoweruprate, andconfirmed thatRCS cooldown andadequate boronmixing canbe achieved.

However, the licensee did not describe RCSpressure controlordepressurization intheir evaluation. Theinitial licensing evaluation, aswell asanevaluation performed insupport ofa stretch poweruprate, confirmed that pressure control canbeachieved during a natural circulation cooling event. Please confirm that RCSpressure control canbe maintained anddepressurization can be achieved ina natural circulation cooling event at conditions bounding the MURuprate.

M Thenatural circulation cooldown AORdescribed in the MPS3SPULAR(performed at 3650MWt) concluded that RCSpressure control could bemaintained bytheuseof pressurizer auxiliary spray andpressurizer power operated reliefvalves (PORVs), and thattheRCSpressure could bereduced lowenoughto allow Residual Heat Removal (RHR) system initiation inapproximately eleven hours (which iswellbefore therequired 24hours). Theevaluation ofnatural circulationcooldown was reevaluated for a rated thermal power of3723 MWt, which bounds theMURPower Uprate conditions. This reevaluation determined that theconclusions from thepreviously performed analyses for 3650MWtremained valid atMURconditions, including meeting therequired timeframe toinitiate RHR.System andcomponent evaluations attheMUR conditions have confirmed that thepressurizer PORVscontinue tohavethecapabilityto control and reduce RCS pressure whenrequired. Additional evaluations haveshown that the pressurizer spray auxiliary can also c ontroland reduce RCS pressure whenneeded at MURconditions. Therefore, RCSpressure control canbemaintained during a natural circulationcooldown atconditions bounding theMURuprate anddepressurizationto a pressure lowenough toinitiate RHRcanbeachieved.

GDC17,"Electric power systems," ofAppendix A,"General Design Criteria for Nuclear Power P lants," to10 CFR 50, states in pad, "An onsite electric power s ystem andan offsiteelectric powersystem shall beprovided topermit functioning ofstructures, systems, andcomponents impodant tosafety. Thesafety function foreachsystem (assuming theother system isnotfunctioning) shall betoprovide sufficient capacity and capabilitytoassure that (1) specified acceptable fuel design anddesign limits conditions ofthereactor coolant pressure boundary arenotexceeded asa result ofanticipated operational occurrences and(2) thecore iscooled andcontainment integrity andother vital functions aremaintained inthe event ofpostulated accidents."

Serial No.21-166 Docket No.50-423 Attachment 1,Page18of21 RIS2002-03, "Guidance ontheContent ofMeasurement Uncertainty Recapture Power Uprate Applications," Section V, "Electrical Equipment Design," states inpart, "A discussion of the effect ofthe power uprate onelectrical equipment. . Forequipment that isnotbounded by existing analyses ofrecord, a detailed discussion should beincluded toidentify andevaluate thechanges related tothepoweruprate. Specifically, this discussion should address the following items: ..D. grid stability."

OfficeofNuclear Reactor Regulation officeinstruction LIC-105, "Managing Regulatory Commitments MadebyLicensees totheNRC",Revision 7 (ADAMS Accession No.

ML16190A013), Section 4.1, "Creation ofRegulatory Commitments," states inpart, "Regulatory commitments...donot warrant either legally binding requirements or inclusioninupdated final safety analysis reports (UFSARs) orprograms subject toa formalregulatory change controlmechanism."

EEEB-RAl-9 TheNRCstaff evaluated theLARfor consistencywith RIS2002-03, GDC17, andthe MPS3UFSAR. Specifically, asreferenced inAttachment 4 oftheLAR,Section V,"NRC Regulatory Issue Summary 2002-03 Topic: Electrical Equipment Design," RIS2002-03 specifiesthat a discussion oftheeffect ofthe power uprate on electrical equipment be includedinthe LAR,specifically infour areas, oneofwhich isgrid stability. A typical grid stability study fora nuclear power plant assesses (1)the impacts of the loss, through a single event, ofthelargest capacity being supplied tothegrid, (2) the removal ofthe load largest from thegrid, (3)the m ost critical transmission lineifu navailable that results intheloss ofoffsite power, (4)any increased main generator output adverse effects, and (5)confirms adequate reactive powersupport atthelowest post-contingency 345kV switchyard voltage. TheNRCstaff hasdetermined thatthe LARfor MURimplementation doesnotprovide sufficient information about grid stabilityandthe345kVswitchyard to complete its review.

Section3.0, "Technical Analysis," inAttachment 1 oftheLARstates, "In support of meeting theISO-New England requirements, maingenerator upgrades will berequired totransmit additional megawatts electric (MWe) to theg ridat uprate conditions."

SectionV.1.F.i, "Main Generator," inAttachment 4ofthe LARstates, "The main generator requiresupgrades inorder toaccommodate the MURPower Uprate atthe required ISO-NewEngland power factor."

Pleaseprovide theMPS3combined maingenerator (MG) output (in megawatts (MW) mega and/or voltamperes (MVA) segregating the c ontributionsdue exclusively totheMG upgrade andMURimplementation (the NRConly regulates the MURportion).

Serial No.21-166 Docket No.50-423 Attachment 1,Page19of21 Theproposed MUR power uprate would increase reactor thermal power from 3650MWt to3709MWt. This increase inreactor thermal power would yield an increase inmain generator output ofapproximately 21MWe.Theexisting generator isnearing its endof expectancy,and the generator supplier life (GE) hasissued manyTechnical Information Letters that would require significant repairs. DENChaschosen toupgrade the generator ratherthan only conduct these repairs. While thecurrent generator hasa nameplate rating of1354.7 MVA,the upgraded generatorwill have a nameplate rating of1500 MVA.

The1500MVArating would ensure that MPS3will meettherequirements for reactive

loading, with a power factor of0.95 atthe point ofinterconnection inaccordance with ISO NewEngland "ScheduleLarge Generator Interconnection Procedures 22, [Reference 1-10)."

However, MPS3will notbeable toreach this full generator capacity, because the would unit belimited bythelicensedreactor thermal power thatisproposed intheMUR LAR.TheMPS3main generator willbeoperating well below this1500 MVA value when theunitisatthe proposed MURpower level.

Insummary, there wouldbe no quantifiable generator outputincrease solely dueto generatorupgrades, sothe expected increase ofapproximately 21MWewould bedueto theincrease inreactor thermal power from theproposed MUR uprate.

EEEB-RAl10 SectionV.1.D, "Grid Stability," inAttachment 4 ofthe LARstates that an Interconnection System Impact Study will beperformed inaccordance with theprocesses ofISO-NE Schedule 22,Large Generator Interconnection Procedures andwill evaluate theimpact oftheproposed interconnection on thereliability andoperation oftheNew England Transmission System. Thelicensee also identified thestudy would consist of a short circuit analysis, astability analysis, a power flow analysis (includingthermal analysis and voltageanalysis), a system protection analysis andanyother analyses that aredeemed necessary bytheSystem Operator (ISO-NE) in consultation with t he Interconnecting Transmission Owner (Eversource).

SectionV.1.G, "Switchyard Interface," inAttachment 4 ofthe LARstates:

The345kVswitchyard isdiscussed inFSARSection 8.1.3.

An Interconnection System ImpactStudy (refer 1.D) to Section V willbe performed inaccordance with the p rocesses ofReference System Impact (see V.1 Regulatory Study will contain Commitment anyother inAttachment analyses that 5).

aredeemed TheInterconnection necessary bythe System Operator (ISO-NE) in consultation with the I nterconnecting Transmission Owner (Eversource), which includes evaluation ofthe 345kVswitchyard anddistribution system.

Thisanalysis oftheswitchyard will beprepared byEversource andwill ensures the ofthe functionality switchyard andits associated components thatwould beaffected by theMURpower uprate.

SerialNo.21-166 Docket No.50-423 Attachment1,Page20of21 InAttachment 5oftheLAR,thefollowing regulatory commitment wassubmitted bythe licensee:

(Commitment) DENC willcomplete anInterconnection System Impact Study, including a grid analysis, stability asdescribedinAttachment 4,Section V.I.DoftheMPS3MUR PowerUprate LAR submittal. (ScheduledCompletion Date) TheInterconnection System Impact StabilityStudy will becompleted prior toimplementation ofthe MURPower Uprate forMPS3."

Please provide additional details about whatexplicit actions DENCwill take including notifying theNRCif thegrid stability studyresults donotmeetDENC's orthe reparer's p

designated standard considering effects on (1) MPS3grid resiliency,(2) 345 kV switchyard functionality, and(3)GDC 17requirements foronsite andoffsite power systems.

M TheSystem Operator (ISO-NE) willbeperforming a system impact study perReferences 1-7and1-8. This study includes:

1. m:

with voltage andthermal This analysis loading beperformed todemonstrate compliance will criteria andwill identify any systemupgrades to satisfythese criteria.Theanalysis will verifythat thepower increase will nothave anysignificant adverse impact uponthereliability oroperating characteristics of thebulk power system (grid). Forsteady-state analysis, the maximum output for theMPS3generator will beused during summer operation.

2.h: This analysis demonstrates that theproposedpower increase will nothaveanysignificant adverse impact uponthestability,reliability, or operating conditions.

characteristics Studies ofthebulk demonstrating powersystems dynamic performance under themost severe mustthensimulate l

l conditions that stress thesystem beyond typical combinations ofloadlevel, generation dispatch, andpower transfers.Forstability analysis,themaximum I output 3.Short for theMPS3will Circuit: This analysis beused will during winter beconducted operation.

todemonstrate that theshort circuit l

l dutieswill notexceed equipment capabilityandwill identify anysystem upgrades =

required tosatisfy thecriterion. Thebasecasewill include allgenerating facilities andelective transmission upgrades onthedate thestudy iscommenced that:(i) aredirectly connected totheNew England Transmission System, (ii)are interconnected to theaffected systems andmay havean impact on the interconnection request; and(iii) have a pending higher queued position andmay impact ontheMPS3interconnection request.

4. Thesystem impact study will evaluate the compliance ofthe voltage controlcapability with the requirements ofReference 1-9.

Serial No.21-166 Docket No.50-423 Attachment 1,Page21of21 Upon completion ofthestudy, DENCwill confirmtheeffects ofgrid resiliency, switchyard functionality, andGDC-17compliance (which issummarized inMPS3FSARSection 3.1.2.17). If the studyproduces satisfactory resultsunder theMURuprated conditions, then the proposed changes from theMURLARwill beimplemented. Ifthe study produces unsatisfactory results under MURuprated the conditions, theMURLARwill notbe implemented, and DENC will inform theNRCofthis outcome anddescribe theexpected actions for resolution.

References:

1-1.Dominion Energy LetterSerial No. 07-0450, "Dominion Nuclear Connecticut, Inc.

Millstone PowerStation Unit 3 License Amendment Request Stretch Power Uprate," dated July13, 2007. Attachment 5:Millstone Power Station Unit 3 Stretch Power Uprate LicensingReport [ADAMS Accession No.ML072000400).

1-2.Millstone PowerStation Unit 3 Safety Analysis Report, Revision 33.02, dated March 31,2021.

1-3.CodeofFederal Regulations, 10CFR50.61, "Fracture ToughnessRequirements for Protection AgainstPressurized Thermal Shock Events," dated January 4,2010.

1-4.Dominion Energy Letter Serial No.20-043, "Dominion EnergyNuclear Connecticut, Inc.Millstone Power StationUnit 3 Proposed LicenseAmendment Request Measurement Uncertainty Recapture Power Uprate," dated November 19,2020[ADAMS Accession Number ML20324A703).

1-5.Westinghouse Report WCAP-16629-NP, Revision 0,"Analysis ofCapsule Wfrom theDominion Nuclear Connecticut Millstone Unit 3 Reactor Vessel Radiation Surveillance Program," September 2006.

1-6.U.S.Nuclear Regulatory Commission, Office ofNuclear Regulatory Research, Regulatory Guide Revision 1 .99, 2, "Radiation Embrittlement ofReactor Vessel g Materials," May1988 [ADAMS Accession Number ML003740284). l 1-7.ISONew England Planning Procedure 5-3,Guidelines forConducting and Evaluating Proposed Application Plan Analysis.

1-8.ISONewEngland Planning Procedure 5-6, interconnection Planning Procedure for Generation andElective Transmission Upgrades.

1-9.ISONew England Operating Procedure No.14 Technical Requirements for Generators, DemandResponse Resources, AssetRelated Demandsand Alternative Technology Regulation Resources.

1-10. ISO-New England Schedule 22 Large Generator Interconnection Procedures.

[

Serial No.21-166 Docket No.50-423 ATTACHMENT 2 Reanay.sis POWER MILLSTONE STATION UNIT3 DOMINION ENERGY NUCLEAR CONNECTICUT, INC.

SerialNo.21-166 Docket No.50-423 Attachment 2,Page1of1 , Section III.4-1.1 of theMillstone PowerStation Unit3 (MPS3)

Measurement Uncertainty Recapture (MUR) License Amendment Request (LAR) submittal, dated November 19,2020(ADAMS Accession No.ML20324A702), discussed theplanned reanalysis ofselect Final Safety AnalysisReport (FSAR) Chapter 15events to supportimplementation of NRC-approved WCAP-17642-P-A, "Westinghouse Performance Analysis andDesign Model (PAD5)," [Reference 2-1). Thediscussion of theLossofNormal Feedwater (LONF) event inSection also II.1.10 described howthis event willbereanalyzed to implement WCAP-17642-P-A. Thepurpose ofthis supplement istoclarify thatthemethodology selected for theDominion Energy Nuclear Connecticut, Inc. (DENC) planned LONF reanalysis has changed from w hati s presented intheMUR LARAttachment 4,SectionsII.1.10 andIll.4-1.1.Instead, theLONFevent revision will beperformed using analternate, NRC-approved, methodology.

Thecurrent FSARChapter 15.2.7 LONF event wasperformed using theWestinghouse RETRAN transient analysis method (WCAP-14882-P-A) [Reference 2-2). TheLONF revision willbe performed using thesame RETRAN method as thecurrent LONF analysis. Therevised analysis will beconfirmed each reloadusing the Dominion Energy Reload Methodology (VEP-FRD-42-A) [Reference 2-3). PriortoPAD5implementation, DENCwill perform a review oftheLONFrevision and submitanyitems deemed necessary forNRCreview perthe criteriaof10CFR50.59. TheLONFrevision will reflect theMUR-power level.

References:

2-1 WCAP-17642-P-A, Revision 1,"Westinghouse Performance Analysis andDesign Model (PAD5)," November 2017.

2-2.WCAP-14882-P-A andWCAP-15234-A, "RETRAN-02 Modelingand Qualification for Westinghouse Pressurized Water ReactorNON-LOCA Safety Analyses," April 1999.

2-3.Topical Report VEP-FRD-42-A, Revision 2,Minor Revision 2,"Reload Nuclear Design Methodology," October 2017.