ML21153A413
| 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 Nuclear Connecticut, inc.
5000 DominionBoulevard, GlenAllen, VA23060 D
DominionEnergy.com
@Q June2,2021 U.S.
Nuclear Regulatory Commission Serial No.21-166 Attention:Document ControlDesk NRA/SS RO Washington,DC 20555 Docket No.
50-423 License No.
NPF-49 MILLSTONEPOWER STATION UNIT3 RESPONSETO REQUEST FOR ADDITIONAL INFORMATIONREGARDING LICENSEAMENDMENTREQUEST FOR MEASUREMENT UNCERTAINTY RECAPTUREPOWERUPRATE Byletter dated November 19,2020(Agencywide Documents AccessandManagement System(ADAMS)
Accession No.ML20324A702),
Dominion Energy Nuclear Connecticut, Inc.
(DENC),
submitted alicense amendment request (LAR) totheNuclear Regulatory Commission (NRC) forMillstone Power Station,Unit No. 3(MPS3).Theproposed license amendment wouldincrease therated thermal power (RTP) level from3,650megawatts thermal (MWt) to3,709MWtintheMPS3operating license andinTechnicalSpecification (TS) 1.27, anincrease inRTPofapproximately 1.6%.
The proposed increase isreferred toasameasurement uncertainty recapture (MUR) power uprate andisbasedonutilizing aninstalled Cameron Technology USLLC(currently knownas
- Sensla, formerlyknown asCaldon)
Leading EdgeFlowMeterCheckPlus system
.as an ultrasonic flowmeter located ineachofthefourmainfeedwater lines supplying thesteam generators to improve plant calorimetric heatbalance measurement accuracy.
Theproposed changes wouldalsoinvolve aneditorial correction toTS2.1.1.1 andrevision toTS 3.7.1.1, Action Statement "a"
andTSTable 3.7-1, "Operable MSSVsVersus Maximum Allowable Power" toupdate themaximumallowable powerlevels corresponding tothenumberof operable i
mainsteamsafety valves persteamgenerator.
Inanemail dated April 8,2021,theNRCissued adraft request foradditional information (RAI) related totheproposed LAR.OnApril 20,2021,theNRCstaff conducteda conference call withDENCstaff toclarify therequest.
Inanemail dated April 22,2021, theNRCtransmitted thefinal version oftheRAl(ADAMS Accession No.ML21112A308).
DENCagreed torespond totheRAIwithin 45daysofissuance, ornolater thanJune7,
- 2021, I
I Attachment 1provides DENC'sresponse totheRAl.Attachment 2provides asupplement toclarify that themethodology selected forDENC'splanned LossofNormal Feedwater (LONF) reanalysis haschanged fromwhatispresented inMURLAR,Attachment 4,
Sections II.1.10 andIII.4-1.1
- Instead, theLONFeventrevision will beperformed using analternate, NRC-approved, methodology.
Serial No.21-166 Docket No.50-423 Page2of3 Ifyou have anyquestions orrequire additional information, please contact Shayan Sinha at(804) 273-4687.
Sincerely, Gerald T.Bischof Senior VicePresident Nuclear Operations
& Fleet Performance COMMONWEALTH OFVIRGINIA )
)
COUNTYOFHENRICO
)
Theforegoing document wasacknowledgedbefore me,inandfor theCounty andCommonwealth aforesaid, today byMr.Gerald T.Bischof, whois Senior VicePresident
- Nuclear Operations and Fleet Performance ofDominion Energy Nuclear Connecticut, Inc.
Hehasaffirmed before methat heisdulyauthorized toexecute andfile the foregoing document inbehalf ofthat
- company, and that thestatements inthedocument aretrue tothebest of his-knowledge andbelief.
Acknowledged before methis2 dayof
, 2021 MyCommission Expires:
CRAIG DSLY N taryu Notary Public Commonwealth ofVirginia Reg. #7518653 g
MyCommissionExpires December31,20-Attachments:
1, Response toRequest forAdditional Information Regarding License Amendment Request forMeasurement Uncertainty Recapture PowerUprate
- 2. Supplement toClarify Methodology forRevised LossofNormalFeedwater Reanalysis i
Commitments madeinthis letter:
None I
Serial No.21-166 Docket No.50-423 Page3of3 cc U.S. NuclearRegulatory 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, MailStop08-C2 11555 Rockville Pike Rockville, MD20852-2738 NRCSenior Resident Inspector Millstone PowerStation
- Director, Radiation Division Department ofEnergy andEnvironmental Protection 79ElmStreet
- Hartford, CT 06106-5127
Serial No.21-166 Docket No.50-423 ATTACHMENT 1
MILLSTONEPOWERSTATIONUNIT3 DOMINION ENERGYNUCLEARCONNECTICUT, INC.
Serial No.21-166 Docket No.50-423 Attachment 1,Page1of21 Byletter dated November19,2020(Agencywide Documents AccessandManagement System (ADAMS) AccessionNo.ML20324A702),
Dominion Energy Nuclear Connecticut, Inc.
(DENC),
submitted alicense amendment request (LAR) totheNuclear Regulatory Commission (NRC) forMillstonePowerStation, Unit No.3(MPS3).
Theproposed license amendment would increase therated thermal power(RTP) level from3,650megawatts thermal (MWt) to3,709 MWtintheMPS3operating license andinTechnical Specification (TS) 1.27, anincrease inRTPofapproximately 1.6%.
Theproposed increase isreferred toasameasurementuncertainty recapture(MUR) poweruprate andisbasedonutilizing aninstalled Cameron Technology USLLC(currently knownasSensia, formerly known asCaldon)
Leading EdgeFlow Meter CheckPlussystem asanultrasonic flowmeter located ineachofthefourmain feedwater lines supplying thesteamgenerators to improve plant calorimetric heat balance measurement accuracy.
Theproposed changes wouldalsoinvolve aneditorial correction toTS2.1.1.1 andrevision toTS3.7.1.1, Action Statement "a"
andTSTable 3.7-1, "Operable MSSVs Versus MaximumAllowable Power" toupdate themaximumallowable powerlevels corresponding tothenumberofoperable mainsteamsafety valves persteam generator.
Inanemail dated April 8,2021,theNRCissueda draft request foradditional information (RAI) related totheproposed LAR.OnApril 20,
- 2021, theNRCstaff conducted a
conference call withDENCstaff toclarify therequest.
In anemaildated April 22,2021, theNRCtransmitted thefinal version oftheRAI(ADAMS Accession No.ML21112A308).
DENCagreed torespond totheRAIwithin 45daysofissuance, ornolater thanJune7, 2021 Thisattachment provides DENC'sresponse totheRAI Theregulation at10CFR50.55a requires thatthereactor pressure vessel (RPV) be constructed, designed andanalyzed inaccordance withtheAmerican 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.
Fracture resistance ofRPVbeltline materials during PTSevents isevaluated bycalculating thenil-ductility temperature (RTNDT) forPTS(identified asRTPTS).
Section 50.61(b)(1) requires that PWRlicensees haveprojected values ofRTPTSaccepted bytheNRCforeachRPV beltline material.
Section 50.61(c)(2) requires that RTPTScalculations forRPVbeltline materials incorporate credible RPVsurveillance material testdatathat arereported as part oftheRPVmaterials sunteillance program required by10CFRPart50,Appendix H.
Serial No.21-166 Docket No.50-423 Attachment 1,Page2of21 NVIB-RAl-1 SectionIV.1.A.ii.e, "Mechanical Evaluation,"
inAttachment 4oftheLARstates that the MURpower uprate designconditions donotaffect thecurrent design basesforseismic andloss-of-coolant accident (LOCA) loads.
Thelicensee further stated that thestress levels caused by the flowinducedvibration onthecorebarrel assembly andupper internals arelowand remain well belowthematerial high-cycle fatigue endurance limit.
Summarize the mechanical evaluation that demonstrates thecorebarrel assembly and upperinternals arenotaffected bytheMURpoweruprate design conditions, including a
discussion onthe acceptance
- criteria, resultingstresses andfatigue endurance limits.
Thegoalofanalyzing flow-induced vibration (FIV)onthecorebarrel assembly andupper internals istoassess its impact onstructural integrity ofthecomponents contained within theMPS3reactor vessel.
Extensive evaluations were previouslycompleted insupport of theMPS3stretch poweruprate (SPU)
- LAR, which iscurrently underNRCreview, further increases theRTPlevel byapproximately 1.6%.
TheMURpowerincrease considers changing theoperating powerplus uncertainty butremainsbounded bythe SPUoperating powerplusuncertainty whichwaspreviously analyzed.
Basedonthis understanding, a comparison totheevaluations completed for theSPUincrease was relied uponforjustification oftheMURincrease.
Themethodology followed intheSPUevaluation inSPULARAttachment 5 (Reference 1-1),
Section 2.2.3involved 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.
Thechangeininput parameters isshowntobenegligible
(<1%), so the FIV evaluation completed fortheSPUprogram wasconsidered tobeapplicable tothe MUR
- program, asthese parameters aretheonlychange noted inthis evaluation between the twoprograms.
TheSPUevaluation included thereactor internals components that areconsidered tobe limiting withregards toFIV.Thesecomponents consist ofthecorebarrel inthelower internals assembly andtheguide tubes intheupper internals assembly.
Forother reactor internal components suchasthelower radial restraints, uppercoreplate alignment
- pins, lower support
- plate, andthelower support
- columns, thevibratory response isextremely small.
Thecalculated stresses wereobtained byscaling previously generated FIV stresses towhatwouldbeexpected after SPUimplementation.
Forconservatism, the values werescaled basedonthehotfunctional flowrates, whicharetypically about 4%
higher thantheMDFrate.
Furthermore, theseismic andLOCAanalyses fortheMPS3SPUprogram areconsidered tobeapplicable totheMURprogram.
Theseismic inputs andplant modelparameters fromtheSPUprogram arealsoapplicable totheMURpoweruprate.
Additionally, the
Serial No.21-166 Docket No.50-423 Attachment 1,Page3of21 conservative assumptions intheLOCAforce calculations boundthechanges found inthe MURprogram.
Therefore, existing SPUparameters weredetermined tobeapplicable to theMURconditions.
TheASMECode (1998 Editionwith2000Addenda, Section
- III, Division 1andSection II, PartD)combined with measured data,forms thebasis fortheacceptance criteria for mechanically induced stresses/strains produced by FIV.Although theparticular components analyzed do notconstitute a pressure
- boundary, theacceptability ofthe components impacted by the uprate relating toalternating stresses forhigh-cycle fatigue wasassessed.
Thealternating stresses werecalculated andcompared according tothe ASMECoderules onhigh-cycle fatigue ormeasuredexperimental dataonstrains limits intheMPS3SPULAR.
NVIBRAl-2 Section IV1.A.ii.f, "Structural Evaluation,"
in Attachment 4 oftheLAR states that evaluations wereperformed todemonstrate that the structural integrity ofreactor internal components isnotadversely affected bytheMURpower uprate design conditions.
The NRCstaff requests thelicensee to(a) summarize the structural evaluation that demonstrates thereactor internal components arenotadversely affectedbytheMUR poweruprate design conditions, including a discussion onthe acceptance criteria and resulting stresses and(b) discuss whether there areanycracksin any oftheRPVinternal components.
Ifthere areanycracks, discuss whether anevaluation has beenperformed andhowthis evaluation demonstrates sufficient structural integrity of thedegraded reactor internal component undertheMURpoweruprate conditions.
M M
Extensive evaluations werepreviously completed insupport oftheMPS3SPULAR, whichinvolved anRTPincrease of7%.TheMPS3MURpoweruprate LAR,which is currently under NRCreview, further increases theRTPlevel byapproximately 1.6%.
The MURpowerincrease considers changing theoperating powerplusuncertainty but remains boundedbytheSPUoperating powerplusuncertainty whichwaspreviously analyzed.
Basedonthis understanding, acomparison totheevaluations completed for theSPUincrease wasrelied uponforjustification oftheMURincrease.
Theinputs forallcomponents remained unchanged, thusnonewcalculations were performed, andtheresults fromtheSPUcalculations weredocumented intheSPULAR.
Fiveareas wereevaluated, including thefollowing:
Control RodInsertability Evaluation, Baffle BoltEvaluation, FIVEvaluation, Critical Reactor Internal Components Structural Evaluation, andanUpperandLowerCorePlate Evaluation.
TheFIVevaluation is addressed inNVIB-RAl-1, theUpperandLowerCorePlate Evaluation isaddressed in
Serial No.21-166 Docket No.50-423 Attachment 1,Page4of21 NVIB-RAl-3, andtheBaffle BoltEvaluation isaddressed inNVIB-RAl-4.
Further discussion ontheremainingtwoevaluations isincluded below.
M This evaluation determined themaximummechanical loadacting ontheguide tubes.
This loadwasthen compared totheallowable loadontheguidetubes.
Themethodology assessed the predicted lateral load/displacement against droptests that applied lateral loads/deflections tothe guide tubetoconfirm control rodinsertion.
Thesetests considered theeffects ofboth temporary (elastic) andpermanent (plastic) deformation.
Theapplicable loadings that were involved inthisevaluation include massflowand acoustic
- loads, system
- loads, and safe shutdown earthquake (SSE) loads.
Thesethree
- loads, aspreviously generated forthe SPU program, weredeemedtobeapplicable for useintheMURprogram.
This applicability wasconfirmedbasedoneither bounding or unchanging inputs tothespecified loading conditions.
MPS3usesa 17x17,96-inch style guide tube.
The allowable loads andacceptance criteria forthis style ofguide tubeandcontrol rod insertion wereshowntobemetforthe SPUanalysis documented inSPULARAttachment 5 (Reference 1-1),
Section 2.2.3 (which boundMURconditions).
Thepurpose ofthis evaluation wastoassesstheimpactof the MURpoweruprate program ontheMPS3reactor internal components including thelower coresupport plate atypical
- region, lower support
- columns, andcorebarrel nozzle weldments.
Thefollowing methodology was previously usedfortheSPUprogram evaluation inSPU LAR Attachment 5,Section 2.2.3:
1 Determine thechanges inloadconditions duetothepoweruprate.
Generally, the poweruprate wouldchange theloads ontheinternal components during normal andupset conditions duetoHeatGeneration Rates(HGR) fromgammaheating andthermal fluid transients.
2.ComparetheMPS3coresupport component design configurations toother design configurations thathavebeenusedfordetailed stress
- analyses, whichwere I
performed under similar loading conditions.
Anydifferences indesign dimensions needtobereconciled inthecomponent stress evaluation.
ifthere areanymajor design differences, component stresses needtobedetermined independently.
3.Compare thermal loadings (thermal transients andHGRs) usedincoresupport component
- analysis, which wereperformed under similar loading conditions tothe thermal loadings forMPS3uprate conditions.
4.Determine ifthethermal loadings ofsimilar components indetailed stress
- analyses, which wereperformed under similar loading conditions boundtheMPS3 thermal loadings.
Serial No.21-166 Docket No.50-423 Attachment 1,Page5of21 The primary inputs totheseevaluations include heatgeneration ratesanddesign transients.
Both oftheseinputs aredeemedtobeapplicable toboththeSPUandMUR programs, due toconservatisms included intheSPUanalysis.
No plant-specific ASME Codestress report waswritten forMPS3.Thereactor internal components were analyzed tomeettheintent oftheASMECode,Section III criteria.
Therefore, based on the previous evaluations andcurrent practices, theguidelines in Subsection NGofthe ASME Codewereusedforthisevaluation.
Inconclusion, the allowable loads andacceptance criteriafortheinternal components analyzed wereshown tobemetfortheSPUanalysis (which boundMURconditions).
Aspart ofthe10-year visual examinations, DENC hasnotidentified cracking inanyMPS3 RPVinternal components.
NVIB-RAl3 Section IV.1.A.ii.g, "Upper andLowerCorePlate Structural Analysis,"
inAttachment 4of theLARstates thatthermal design transients, heatgeneration
- rates, andoperating conditions affect thermal loads ontheupperandlower core plates.
Thelicensee stated that fortheMURpoweruprate, current analysis ofrecord (AOR) thermal design transients andheatgeneration rates remain applicable because theMURpower uprate operating conditions arebounded bytheoperating conditions inthe current AOR.The licensee further stated that themaximumprimary plussecondary stress intensity ofthe upperandlower coreplate andcumulative usagefactor remain acceptable.
Summarize howtheexisting structural analysis fortheupperandlower coreplates isstill applicable undertheMURpoweruprate conditions, including adiscussion onthe acceptance criteria andresulting stresses.
M Extensive evaluations werepreviously completed insupport oftheMPS3SPULAR, whichinvolved anRTPincrease of7%.TheMPS3MURpoweruprate LAR,whichis currently under NRCreview, further increases theRTPlevel byapproximately 1.6%.
The MURpowerincrease considers changing theoperating powerplusuncertainty but remains bounded bytheSPUoperating powerplusuncertainty which waspreviously analyzed.
Basedonthis understanding, acomparison totheevaluations completed for theSPUincrease wasrelied uponforjustification oftheMURincrease.
Thepurpose oftheUpperandLowerCorePlate Evaluation wastoassess theimpact on structural integrity ofMPS3reactor internals lower coreplate (LCP) andupper coreplate (UCP) withregard totheproposed MURpoweruprate program.
Themethodology for qualifying thestructural integrity oftheMPS3UCPandLCPwasinvestigated to determine thedriving inputs whichproduce thestresses ontheupperandlower core plates.
Itwasdetermined that theLCPandUCPgeometry andmaterial
, coreplate
Serial No.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 usedinthestress calculation fortheLCP andUCP. For theMURpoweruprate, itwaseither determined that theinputs listed have notchanged, that thechangeobserved isnegligible, orthattheobserved changeis bounded bythe AOR.
Theoriginal LCPand UCP evaluation documented effects ontheSPUprogram forMPS3 inSPULARAttachment 5 (Reference 1-1),
Section 2.2.3.
Thisevaluation madeuseof analyses completed for a similar plant during a replacement steamgenerator (RSG) program.
Therefore, thestructural qualification andfatigue evaluation istheoriginal design basis oftheMPS3reactor internals.
Basedonsimilarities indesign andthermal
- loading, theevaluation completed fortheLCPandUCPatthesimilar plant wasusedin thecurrent analysis fortheMPS3SPU program.
Theallowable stresses areobtainedfrom theASMECodeRequirements (1974
- edition, Division 1,Section
- Ill, Subsection NG)which isconsistent withtheoriginal evaluation.
Forthis evaluation theallowable
- stress, as prescribed bythe1974ASMEBoiler
- Code, wereusedastheacceptance criteria.
NVIB-RAl-4 Section IV.1.A.ii.h, "Baffle-Barrel Region Evaluations,"
inAttachment 4oftheLARstates that thebaffle bolts aresubjected toprimary loads consisting of deadweight, hydraulic pressure differentials, LOCAandseismic
- loads, andsecondary loads consisting of preload andthermalloads resulting fromreactor coolant system (RCS) temperatures and gammaheating rates.
Thelicensee stated thatitevaluated thebaffle former bolt maximumdisplacement attheMURpoweruprate design conditions.
The licensee concluded that theexisting thermal andstructural analysis ofthebaffle-barrel region results remain bounding fortheMURpoweruprate design conditions.
TheNRC staff requests thelicensee tosummarize(a) howtheexisting thermal andstructuralanalysis ofthebaffle barrel region remainbounding undertheMURpoweruprate design conditions, including adiscussion onthebaffle former bolt maximumdisplacement and stresses undertheMURpoweruprate design conditions and(b) theinspection history andresults offormer baffle bolts andplates.
M Extensive evaluations werepreviously completed insupport oftheMPS3SPULAR, whichinvolved anRTPincrease of7%.TheMPS3MURpoweruprate LAR,whichis currently underNRCreview, further increases theRTPlevel byapproximately 1.6%.
The MURpowerincrease considers changing theoperating powerplusuncertainty but remains bounded bytheSPUoperating powerplusuncertainty whichwaspreviously
Serial No.21-166 Docket No.50-423 Attachment 1,Page7of21 analyzed.
Based onthis understanding, acomparison totheevaluations completed for theSPU increase wasrelied uponforjustification oftheMURincrease.
Thebaffle bolt evaluation wasusedtoassess thestructural acceptability ofthebaffle-former boltsfor the MPS3MURpoweruprate 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 factorwasthencompared totheallowable factor.
Calculation ofthedamagefactor was completed using thefollowing methodology:
1.Thegeometry oftheMPS3 baffle-former bolts wasreviewed andcompared tothe bolt geometry usedtodetermine adesign fatigue curve.
Thisfatigue curvewas developed fromtest dataused for qualification ofthestandard four-loop,
- upflow, baffle-former bolts.
Thegeneration ofthis fatigue curveincluded themargins specified intheASMEBoilerand Pressure Vessel Code,Section
- Ill, Division 1,
1998Edition, forusing test results to qualify acomponent. Itwasassumed that the fatigue curve hadthesame"shape"as the design fatigue curve forstainless steel.
Thisisconsistent withtheprocedure provided intheASMECode,Appendix II, for performing fatigue tests using accelerated loadings.
2.Baffle-barrel temperatures weredetermined for normal andupsetservice conditions.
Thisanalysis addressed heating rates associated withthehistoric and projected fuel loading patterns.
3.Displacements ofthebaffle plates weredetermined atthe bolt locations, using the previous fatigue curve andthetemperatures determined in Steps 1and2.Using linear
- scaling, displacements werealsodetermined forother applicable load conditions.
4.Fatigue usagewasdetermined using Miner's Rule.
Theinputs tothis evaluation included RCStemperature andflow parameters for MUR andSPUconditions, thefour-loop,
- upflow, baffle-bolt design fatigue
- curve, design transients, HGRs,andbaffle-barrel temperatures.
Anevaluation oftheaforementioned inputs wascompleted todetermine applicability oftheSPUparameters fortheMUR program.
Allofthelisted inputs weredeemedtohavenotchanged andareapplicableto theMURprogram forMPS3.Therefore, theresults fromtheSPUprogram arealso applicable totheMURpoweruprate program.
Theconclusion reached isthat anacceptable damagefactor wascalculated fortheSPU
- program, andthis remains applicable totheMURpoweruprate program.
Therefore, the baffle-former-barrel configuration isacceptable fortheeffects oftheMUR,andthe calculated usageiswellbelowtheallowable fatigue usagelimit of1.0.Notethat this analysis considered afull baffle-former bolting pattern.
NoAcceptable Bolting Pattern Analysis (ABPA) methods wereapplied intheSPUortheMURevaluations.
Serial No.21-166 Docket No.50-423 Attachment 1,Page8of21 TheMPS3 reactor vessel coresupport barrel former baffle plates andbaffle bolting have been subjected tooneVT-3examination during each10-year ISIinterval.
Thefirst interval examwasperformed duringthe3R05outage (June 1995),
thesecond interval examwas performed during the 3R11outage(April 2007),
andthethird interval examwas performed during the 3R17 outage(April 2016).
Norelevant indications wereidentified onthecoresupport barrel baffle plates orbaffle bolting during anyofthese exams.
NVIBRAl-5 Section IV.1.A.iii.a, "Bottom Mounted Instrumentation (BMI),"
inAttachment 4oftheLAR discusses thestress analysis ofthe BMI guide tubes.Thelicensee stated that therange ofvessel coreinlet temperatures for theMUR power uprate is536.7 degrees Fahrenheit
(*F) to555.8*F, which islowerthantheRPV coreinlet temperature intheexisting analysis.
Thesetemperatures arebounded by the BMIguidetubedesign temperature of 560'F.
Clarify whether theexisting stress analysis of theBMIguide tubes isbasedonthe design temperature of560*F.
M TheBMIguide tubes stress analysis isbasedonthedesigntemperature of560*F.
This temperature isbounding ofboththepre-MURandpost-MURpower upratecoreinlet temperatures.
NVIB-RAl-6 Section IV.1.C.i, "Pressurized Thermal Shock(PTS)
Calculations,"
inAttachment 4 ofthe LARstates thatthelimiting reference temperature forPTS(RTPTS) value of130*F applies toLowerShell Plate B9820-2.
Thelicensee stated that this isachange fromthe AORthathada limiting RTPTSvalue of133*Fpertaining toIntermediate Shell Plate B9805-1.
TheNRCstaff requests thelicensee to(a) clarify thecause ofthechanges in theRTPTSvalue from133*F to130"Fandthelimiting beltline
- material, and(b) discuss whether theRTPTSvalue of130*Fwasderived basedontheMURpoweruprate conditions.
Serial No.21-166 Docket No.50-423 Attachment 1,Page9of21 W
h Per10CFR50.61,"Fracture toughness requirements forprotection against pressurized thermal shockevents
[Reference 1-3),"
theuseofresults fromtheplant-specific surveillance program may result inanRTPTS Valuethat ishigher orlower thanthevalue calculated without plant-specific surveillance data.
Asdescribed inMPS3FSARSection 5.3,themethodology described in Regulatory Guide(RG) 1.99, Revision 2[Reference 1-6)isperiodically updated toincorporate theeffects ofirradiation exposure using in reference legtemperature calculations.
RG 1.99includes positions thatdetermine Chemistry Factor (CF)xwithout theuse of surveillance data(Position 1.1) orwit.h theuse ofsurveillance data(Position 2.1).
TheAORforpressurized thermal shock(PTS) for MPS3 iscontained 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 intheSPULAR,Attachment 5:
Thelimiting material isIntermediate Shell Plate B9805-1, with the morelimiting RTers value occurring forcalculations usingtheRG 1.99, Rev.
2 Position 1.1 Chemistry
- Factor, asopposed tothePosition 2.1Chemistry Factorcalculated from credible surveillance data.
Themostlimiting RTers value at54EFPY[effective full poweryears) forPlate B9805-1 is133F.
Thus,theAORlimiting RTPTS Value iSConsidered tobe133F,which corresponds to Intermediate Shell Plate B9805-1 evaluated with aPosition 1.1CF(i.e.,
calculated without theuseofsurveillance data).
- However, using thecredible surveillance plate dataper Position 2.1,anRTPTS Value of1110F wascalculated forIntermediate Shell Plate B9805-1intheSPUanalysis (see Table 2.1.3-4 ofReference 1-1).
Since thesurveillance dataiscredible andthePosition 2.1calculation results inalower calculated RTPTS
- value, thePosition 2.1results could havebeenusedinlieu ofthe Position 1.1results fortheSPUanalysis per10CFR50.61.
- However, instead oftaking credit forthecredible surveillance data,theSPULARAttachment 5 conservatively reported thePosition 1.1CFresult of133FforIntermediate Shell Plate B9805-1, instead ofthe111Fvalue fromusing thePosition 2.1CF.
IftheSPUreport hadelected totakecredit forthecredible surveillance datafor Intermediate Shell Plate B9805-1, this material wouldnolonger havebeenthemost
Serial No.21-166 Docket No.50-423 Attachment 1,Page10of21 limiting material.
- Instead, thelimiting RTPTS Value wouldhavebeen130Fcorresponding toLower Shell Plate B9820-2(see Table 2.1.3-4 ofReference 1-1).
MPS3MUR RTPTS Evaluation TheRTPTS CalCulations supporting theMUR[Reference 1-4) areshowninTable1-1.
This analysis isbasedon fluence values derived fromMURpoweruprate conditions.
Consideration oftheMUR increased thepeak54EFPYfluence valuefromtheSPU analysis fromavalue of2.70 x 1019n/cm2 (SPU) to2.72x1019 n/cm2 (MUR).
Sincethe RTPTS Values arereportedas whole numbers,thisslight increase influence didnot change thehighest calculated RTeTS values.
Specifically,theRTPTS values corresponding toIntermediate Shell Plate B9805-1 using Position1.1(1330F),
Intermediate Shell Plate B9805-1usingPosition 2.1(111F),
and LowerShell PlateB9820-2(130F) are unchanged fromthefluence increase.
Itisalsonoted thatthePosition 2.1CFfor Intermediate Shell Plate B9805-1 for boththe SPU andMURutilize thesamesurveillance
- data, available inWCAP-16629-NP
[Reference 1-5). Thecredibility ofthedataisalso assessed inWCAP-16629-NP.
TheMURLARRTPTS analySiS takes credit forthe available credible surveillance datafor Intermediate Shell Plate B9805-1, andthus theRTPTS Value CorreSponding tothis material is111F.The133Fresult basedonthePosition 1.1CFis shown for information only.
As
- aresult, thelimiting RTPTS Value is130F,which correspondsto Lower Shell Plate B9820-2.
Thechangeinthelimiting RTPTS Value andmaterial istheresult ofdecisions madewith theimplementation of10CFR50.61 with respect tosurveillance data,and isnotaresult ofchanges influence projections, surveillance dataavailable since the SPU
- LAR, and/or material properties.
Thechange inthelimiting RTPTS Value from133Fto1300F andthe limiting beltline material complies with10CFR50.61andthePositions ofRegulatory Guide1.99, Revision 2.
i Asdescribed intheresponse toNVIB-RAl-6, part(a),
theRTPTS Value of130Fconsiders i
thefluence value derived fromMURpoweruprate conditions.
Thevalue of130Fwas unchanged fromtheanalysis ofrecord basedonthefluence increase resulting fromthe MURpower uprate conditions.
i
Serial No.21-166 Docket No.50-423 Attachment 1,Page11of21 Table1-1 MURRTPTS Calculations fortheMPS3Beltline andExtended Beltline Materials at54EFPY00 R.G.
Reactor Fluence(c)
Initial 1.99, CF(b)
ARTl@
GU GA Margi RTers Vessel Reactor Vessel Material (n/cm2, E> 1.0Me FF(c)RTmrr (b)
Rev.2 (F) o r(F) (F) (0F)n(0F)(0F)
Location V)
(F)
Position
.3 0
U W
m Intermediate Shell Plate B9805-1 2.1 26.7 2.72x1019 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 x1019 1.2671 0
39.3 0
17 34.0 73 LowerShell Plate B9820-1 1.1 51.0 2.72 x 1019 1.2671 10 64.6 0
17 34.0 109 LowerShell 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 2.72x 1019 1.2671 -50 40.3 0
20.1(e) 40.3 31 WeldSeams101-124 A,B,C 2.1 6.7 2.72 x 1019 1.2671 -50 8.5 0
4.2(e)8.5
-33 Intermediate toLowerShell Girth 1.1 31.8 2.72x 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 LowerShell Longitudinal Weld 1.1 31.8 2.72 x 1019 1.2671 -50 40.3 0
20.1(e) 40.3 31 Seams101-142 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 x1019 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 Inlet Nozzle B9806-4 1.1 58.0 0.0814 x 1019 0.3770 0
21.9 0
10.9(e) 21.9 44 Extended Beltline Inlet Nozzle R5-3 1.1 44.0 0.0814 x1019 0.3770
-10 16.6 0
8.3(e) 16.6 23 Inlet Nozzle R5-4 1.1 51.0 0.0814 x1019 0.3770 0
19.2 0
9.6(e) 19.2 38 Nozzle Shell Longitudinal Weld 1.1 39.8 0.0814 x1019 0.3770
-10 15.0 0
7.5(e) 15.0 20 101-122A Nozzle Shell Longitudinal Weld 11 39.8 0.0814 x1019 0.3770
-50 15.0 0
7.5(e) 15.0
-20 101-122B, 101-122C
.-...............-~....
Serial No.21-166 Docket No.50-423 Attachment 1,Page12of21 R.G.
Reactor Fluence(c)
Initial 1.99, CFW ARTr ou ca Margi RTPTS Vessel Reactor Vessel Material (n/cm2, E> 1.0Me FF(c)RTNDT Rev.2 (F) o T(F) (0F)(0F)n(F) (F)
Location V)
(F)
Position NozzleShell toIntennediate Shell 1.1 41.0 0.0814 x1019 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 x1019 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 x1019 0.3770 -50 28.4 0
14.2(e) 28.4 7
Inlet Nozzle Weld105-121D 1.1 75.4 0.0814 x1019 0.3770 -50 28.4 0
14.2(e) 28.4 7
Notes:
(a)The10CFR50.61
[Reference1-3) methodology wasutilized inthecalculation oftheRTers values.
(b)Values areconsistent with those utilized intheanalysis
- ofrecord, theSPUevaluation
[Reference 1-1).
(c)Maximumfluence values forthe beltline (2.72 x 1019 n/cm2) andextended beltline (8.14 x 1017 n/m2) considering MURconditions areconservatively used foreverymaterial intheregion.
FF= fluence factor ascalculated withthe10CFR50.61[Reference 1-3) methodology.
(d)A reduced catermisusedsince the surveillance dataisdeemed credible perWCAP-16629-NP
[Reference 1-5).
(e)Per10CFR50.61[Reference 1-3),
caneednot exceed 0.5*ARTuDT.
Therefore, thecahasbeenreduced.
(f)Thelimiting RTrrs value forMPS3is130F, which corresponds toLowerShell Plate B9820-2.
Notethat Intermediate Shell Plate B9805-1 resulted ina higher RTers value of133Fwhenthe surveillancedata was notused;
- however, theRTPTs value forIntermediate Shell Plate B9805-1 is111Fwhenthe credible surveillance data isused.Thus, taking credit for the credible IntermediateShell Plate B9805-1 surveillanc'e
- data, thelimiting material forMPS3is LowerShell Plate B9820-2.
Serial No.21-166 Docket No.50-423 Attachment 1,Page13of21 AppendixA to10CFR50establishes minimum criteria (General Design Criteria orGDC) forthesafeoperation oflight water reactors.
GDC10,"Reactor Design",
requires that the reactor core and associated
- coolant, control, andprotection systems bedesigned with appropriate margin to assure that specified acceptable fueldesign limits arenotexceeded during anycondition of normal operation,including theeffects ofanticipated operational occurrences.
Fuel design limits arechallenged 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 thatreactor coolant system and associated auxiliary,
- control, andprotection systems bedesigned withsufficient 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, and15.3.2 describes analyses of anticipated operational occurrences which could challenge thereactor coolant pressure boundary.
GDC28,"Reactivity limits",
requires reactivity control systems tobedesigned with appropriate limits onpotential reactivity increases sothe effects ofa postulated rod ejection accident canresult inneither damagetothe reactor coolant pressure boundary norresult insufficient disturbance toimpair thecorecooling capability.
Thetransient described inUFSARsection 15.4.8 helps todemonstrate that this criterion ismet.
GDC31,"Fracture prevention ofreactor coolant pressure boundary",
requires that the reactor pressure boundary be designed withsufficient margintoensure thatthe probability ofrapidly propagating failure isminimized under postulated accident conditions.
UFSARsections and15.2.8 and15.3.3 describes analyses ofpostulated accidents which could challenge thereactor coolant pressure boundary.
Regulatory Information Summary(RIS) 2002-03 (ADAMS Accession No.ML013530183) provides guidance toaddressees onthescopeanddetail ofinformation that should be provided toNRCforreviewing MURpoweruprate applications.
Theguidance states that inareasforwhichexisting AORboundplant operation attheproposed powerlevel, the staff will notconduct adetailed review.
Natural circulation cooldown isaportion ofthetransients described inUFSARsections 15.2.6, 15.2.7, 15.2.8, and15.3.2.
Thecurrentlicensing analysis forthis eventdocuments compliance withtheguidance inBranch Technical Position (BTP) 5-1.
Serial No.21-166 Docket No.50-423 Attachment 1,Page14of21 SNSB-RAl7 Thelicensee described thepowerlevel assumedwhenanalyzing accidents and transientsfor MPS3 inTable 11-1 ofAttachment 4 totheLAR.However, theNRCstaff notes that theinformation intheLAR,Table 11-1 is.not consistent withtheinformation in Table15.0-2 ofthe
- UFSAR, Revision 33.Specifically, there isa discrepancy forthe thermal powerlevel listed for accidents inUFSARsections 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, pleaseexplain thediscrepancy between thepower listed in Table 15.0-2oftheUFSARandTable 11-1 ofAttachment 4to theLARforaccidents 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 aportion oftheanalysiswasperformed assuming apower level thatdoesnotboundtheuprated power
- level, please provide ajustification or summarize theupdated analysis.
M SNSB-RAl-7 identifies discrepancies between the power level informationlisted inthe MPS3MURLARAttachment 4,Table ll-1 andMPS3FSAR Table 15.0-2.
DENC'sreview ofthetables identified twoprimary differences between the reported powers.First, FSAR Table15.0-2 inconsistently presents initial power information.
- Second, FSARTable 15.0-2includes powerlevels reflecting thetransient analyses that provide input totheMUR DNBdesign basis.
Thesetransient analyses arebasedona3666 MWt NSSSpower.
Regarding Item1,three general inconsistencies wereidentified inthe FSAR Table15.0-2reported powers.
Theinconsistencies include:
(a) lackofuseofanexplicit identifier forcorepower, (b) lack ofinclusion ofthepoweruncertainty, and(c) lackof inclusion of theconservative MURpowerlevel(3712 MWt) foreventswitha DNBacceptance criterion.
Theissue wasentered into DENC'scorrective action system andislimited to theFSARtable.
TheMPS3safety analyses described inFSARChapter 15use the appropriate powerlevels andarenotimpacted.
MURLARTable 11-1 hasbeenreviewed andisaffirmed todocument appropriate andconsistent inputs forpowerlevel.
Withrespect toItem2,select differences between FSARTable15.0-2 andLARTable Il-1reflect thetransient analyses that provide statepoint operating conditions into theMUR DNBdesign basis.
Select DNBcalculations wereperformed ataconservative, scaled powerlevel of101.7%of3650MWt(3712 MWt)alongwithstatepoint operating conditions generated byasystem transient analysis performed ataNSSSpowerlevel of 3666MWt.Thestatepoint conditions generated at3666MWtNSSSpowerhavebeen validated asapplicable fortheMURpoweruprate because theminorpowerlevel difference hasanegligible effect onthelocal coreconditions (temperature,
- pressure, and flow) relevant forthecalculation ofDNB.Thecombination ofthesystemtransient operating conditions (based onaNSSSpowerof3666MWt) andthescaled, conservative powerlevel of3712MWtresults inaconservative calculation ofDNBforcomparison to theDNBacceptance criteria.
Theapproach takenispresently summarized inFSAR
Serial No.21-166 Docket No.50-423 Attachment 1,Page15of21 Section 15.0.3.1.
Asdescribed intheMURLARandFSARSection 15.0.3.1, allDNB eventsshow acceptable DNBperformance.
Thispowerscaling isnotapplicable tothe FSARSection 15.1.3 and15.4.3 DNBevents identified inSNSB-RAl-7.
Table1-2outlines thecorrections required toconsistently report initial powerinFSAR Table 15.0-2.
- Further, Table1-2identifies thoseanalyses wheretheDNBcalculation wasperformed ata conservative, scaled corepowerof3712MWtalong with statepoint operating conditions generated at3666MWtNSSSpower.
EachFSARsection identified bytheNRCascontaining a power discrepancy isincluded.
TheFSARSection 15.1.2 is alsoincluded since its associated FSARTable15.0-2 entry isimpacted bythepower reporting inconsistencies, andthe transient analysis supporting theDNBcalculation was performed at3666NSSSpower.
Foralllisted
- events, thereferences for NRC approval and/or NRC-approved methods presented inLARTable II-1 remain accurate and valid.
i I
Serial No.21-166 Docket No.50-423 Attachment 1,Page16of21 Table12:MPS3FSARTable15.0-2 Corrections andClarifications FSARSection andEvent Current FSAR Corrected FSAR BasisforCorrection orDiscrepancy Table 15.0-2 Table15.0-2 initial Power Initial Power MWt MWt Note1 15.1.2
- FeedwaterSystem 0(NSSS) and 0NSSSand FSARTable 15.0-2 doesnotcurrently Malfunctions that Result in 3666(NSSS) 3666NSSS identify theconservative, scaled core anIncrease inFeedwater 0Coreand powerassumed inDNBcalculations.
Flow 3712Core(DNB) 15.1.3
- Excessive Increase 3666(NSSS) 3712Core(DNB)FSARTable15.0-2 doesnotcurrently inSecondary SteamFlow list theconservative corepower assumed inDNBcalculations.
No power scaling wasapplied as described inMPS3MURLAR Attachment 4,Section IL1.3.
15.3.1
- Partial Lossof 3666(NSSS) 3666 NSSS FSARTable15.0-2 doesnotcurrently Forced Reactor Coolant 3712(not 3712 Core (DNB)
Include the"core" identifier forpower Flow identified asacore level.
Aconservative, scaled core powerinTable powerisassumed inDNB 15.3.2
- Complete Lossof 15.0-2) calculations.
Forced Reactor Coolant Flow 15.3.3
- Locked Rotor 3666(NSSS) 3666NSSS FSAR Table15.0-2 doesnotcurrently RodsinDNB 3712(not 3712Core(DNB)include the"core" identifier forpower 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 powerscaling wasapplied as described inMPS3MUR LAR Attachment 4,Section II.1.17.
15.4.8
- RodEjection 0(NSSS) and 0Coreand FSARTable 15.0-2 doesnotcurrently 3650(not 3723Core include the"core" identifier for power identified asacore level noridentify that calorimetric powerinTable uncertainty wasapplied totheevent 15.0-2) analysis.
15.6.1
- Inadvertent 3666(NSSS) 3666NSSS FSARTable 15.0-2 doesnotcurrently Opening ofaPressurizer 3712Core(DNB)include theconservative, scaled core Safety orRelief Valve powerassumed inDNBcalculations.
Note1:Poweruncertainty isincluded intheinitial powerasappropriate foreachevent.
Serial No.21-166 Docket No.50-423 Attachment 1,Page17of21 SNSBRAl-8 Thelicensee provided anevaluation forthenatural circulation cooling event insupport of theMURpower uprate inSection 111.3-1 ofAttachment 4totheLAR.Intheir evaluation, thelicensee indicated that thecurrent AORwasperformed at3650MWt,which doesnot boundthepower level proposed intheMURpoweruprate.
Thelicensee hasrepeated thenatural circulation cooling analysisatconditions bounding theMURpoweruprate, andconfirmed thatRCS cooldown andadequate boronmixing canbe achieved.
- However, thelicensee did not describe RCSpressure control ordepressurization intheir evaluation.
Theinitial licensing evaluation, aswell asanevaluation performed insupport ofastretch poweruprate, confirmed that pressure control canbeachieved during a
natural circulation cooling event.
Please confirmthatRCSpressure control canbe maintained anddepressurization can be achieved inanatural circulation cooling event at conditions bounding theMURuprate.
M Thenatural circulation cooldown AORdescribed in theMPS3SPULAR(performed at 3650MWt)concluded thatRCSpressure control could bemaintained bytheuseof pressurizer auxiliary sprayandpressurizer power operated reliefvalves (PORVs),
and thattheRCSpressure could bereduced lowenoughto allow Residual HeatRemoval (RHR) systeminitiation inapproximately eleven hours (which iswellbefore therequired 24hours).
Theevaluation ofnatural circulation cooldown was reevaluated fora rated thermal powerof3723MWt,whichboundstheMURPowerUprate conditions.
This reevaluation determined that theconclusions fromthepreviously performed analyses for 3650MWtremained valid atMURconditions, including meeting therequired timeframe toinitiate RHR.Systemandcomponent evaluations attheMUR conditions have confirmed thatthepressurizer PORVscontinue tohavethecapability to control and reduce RCSpressure whenrequired.
Additional evaluations haveshown that the pressurizer auxiliary spray canalsocontrol andreduce RCSpressure whenneeded at MURconditions.
Therefore, RCSpressure control canbemaintained during a natural circulation cooldown atconditions bounding theMURuprate anddepressurizationto a pressure lowenough toinitiate RHRcanbeachieved.
GDC17,"Electric powersystems,"
ofAppendix A,"General Design Criteria forNuclear PowerPlants,"
to10CFR50,states inpad,"Anonsite electric powersystemandan offsite electric powersystemshall beprovided topermit functioning ofstructures,
- systems, andcomponents impodant tosafety.
Thesafety function foreachsystem (assuming theother system isnotfunctioning) shall betoprovide sufficient capacity and capability toassure that(1) specified acceptable fuel design limits anddesign conditions ofthereactor coolant pressure boundary arenotexceeded asa result ofanticipated operational occurrences and(2) thecoreiscooled andcontainment integrity andother vital functions aremaintained intheevent 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
discussionof the effect ofthepoweruprate onelectrical equipment.
. Forequipment that isnotbounded by existing analyses
- ofrecord, adetailed discussion should beincluded toidentify andevaluate thechanges related tothepoweruprate.
Specifically, this discussion should address thefollowingitems:
..D.grid stability."
Office ofNuclear Reactor Regulation office instruction LIC-105, "Managing Regulatory Commitments MadebyLicensees totheNRC",Revision 7(ADAMS Accession No.
Section 4.1, "Creation ofRegulatoryCommitments,"
states
- inpart, "Regulatory commitments...donot warrant either legally binding requirements or inclusion inupdated final safety analysis reports (UFSARs) orprograms subject toa formal regulatory change control mechanism."
EEEB-RAl-9 TheNRCstaff evaluated theLARforconsistencywith RIS2002-03, GDC17,andthe MPS3UFSAR.Specifically, asreferenced inAttachment 4oftheLAR,Section V,"NRC Regulatory Issue Summary2002-03 Topic:
Electrical Equipment Design,"RIS2002-03 specifies that adiscussion oftheeffect ofthepower uprate on electrical equipment be included intheLAR,specifically infour
- areas, oneofwhich isgrid stability.
Atypical grid stability study foranuclear powerplant assesses (1) theimpacts of the loss,through a
single
- event, ofthelargest capacity beingsupplied tothegrid, (2) the removal ofthe largest loadfromthegrid, (3) themostcritical transmission line if unavailable that results inthelossofoffsite
- power, (4) anyincreased maingenerator output adverse
- effects, and (5) confirms adequate reactive powersupport atthelowest post-contingency 345kV switchyard voltage.
TheNRCstaff hasdetermined that theLARfor MURimplementation doesnotprovide sufficient information aboutgrid stability andthe345kVswitchyard to complete its review.
Section 3.0,"Technical Analysis,"
inAttachment 1oftheLARstates, "Insupport of meeting theISO-NewEngland requirements, maingenerator upgrades will berequired totransmit additional megawatts electric (MWe) tothegrid atuprate conditions."
Section V.1.F.i, "Main Generator,"
inAttachment 4oftheLARstates, "Themaingenerator requires upgrades inorder toaccommodate theMURPowerUprate attherequired ISO-NewEngland powerfactor."
Please provide theMPS3combined maingenerator (MG) output (in megawatts (MW) and/or megavoltamperes (MVA) segregating thecontributions dueexclusively totheMG upgrade andMURimplementation (the NRConlyregulates theMURportion).
Serial No.21-166 Docket No.50-423 Attachment 1,Page19of21 Theproposed MURpoweruprate wouldincrease reactor thermal powerfrom3650MWt to3709MWt.
This increase inreactor thermal powerwouldyield anincrease inmain generator output ofapproximately 21MWe.Theexisting generator isnearing itsendof life expectancy,and the generator supplier (GE) hasissued manyTechnical Information Letters that would require significant repairs.
DENChaschosen toupgrade thegenerator rather thanonly conduct these repairs.Whilethecurrent generator hasa nameplate rating of1354.7 MVA,the upgraded generatorwill haveanameplate rating of1500MVA.
The1500MVArating would ensure that MPS3will meettherequirements forreactive
- loading, with apower factor of0.95 atthepoint ofinterconnection inaccordance withISO NewEngland "Schedule 22,Large Generator Interconnection Procedures
[Reference 1-10)."
- However, MPS3will notbeable toreach this full generator
- capacity, because the unit wouldbelimited bythelicensedreactor thermal powerthat isproposed intheMUR LAR.TheMPS3maingenerator will beoperating wellbelow this 1500MVAvalue when theunit isattheproposed MURpower level.
Insummary, therewouldbenoquantifiable generator outputincrease solely dueto generator
- upgrades, sotheexpected increase ofapproximately 21MWewouldbedueto theincrease inreactor thermal powerfromthe proposed MUR uprate.
EEEB-RAl10 Section V.1.D, "Grid Stability,"
inAttachment 4oftheLARstates that an Interconnection SystemImpact Studywill beperformed inaccordance withthe processes ofISO-NE Schedule 22,LargeGenerator Interconnection Procedures andwill evaluate theimpact oftheproposed interconnection onthereliability andoperation oftheNew England Transmission System.
Thelicensee alsoidentified thestudy would consist of a short circuit
- analysis, astability
- analysis, apowerflowanalysis (including thermal analysis and voltage analysis),
asystem protection analysis andanyother analyses that are deemed necessary bytheSystemOperator (ISO-NE) inconsultation withtheInterconnecting Transmission Owner(Eversource).
Section V.1.G, "Switchyard Interface,"
inAttachment 4oftheLARstates:
The345kVswitchyard isdiscussed inFSARSection 8.1.3.
AnInterconnection System Impact Study(refer toSection V1.D) will beperformed inaccordance with theprocesses ofReference V.1(see Regulatory Commitment inAttachment 5).
TheInterconnection System Impact Study will contain anyother analyses that aredeemednecessary bythe System Operator (ISO-NE) inconsultation withtheInterconnecting Transmission Owner (Eversource),
which includes evaluation ofthe345kVswitchyard anddistribution system.
Thisanalysis oftheswitchyard will beprepared byEversource andwill ensures the functionality oftheswitchyard anditsassociated components that would beaffected by theMURpoweruprate.
Serial No.21-166 Docket No.50-423 Attachment 1,Page20of21 InAttachment 5oftheLAR,thefollowing regulatory commitment wassubmitted bythe licensee:
(Commitment)
DENC will complete anInterconnection System Impact
- Study, including a
grid stability
- analysis, asdescribedinAttachment 4,Section V.I.D oftheMPS3MUR PowerUprate LAR submittal.
(ScheduledCompletion Date)
TheInterconnection System Impact Stability Study will becompleted prior toimplementation oftheMURPower Uprate forMPS3."
Please provide additionaldetails aboutwhatexplicit actions DENCwill takeincluding notifying theNRCifthegrid stability studyresults donotmeetDENC'sorthepreparer's designated standard considering effects on (1)MPS3gridresiliency, (2) 345kV switchyard functionality, and(3)GDC 17requirements foronsite andoffsite power systems.
M TheSystem Operator (ISO-NE) will beperforming a system impact study perReferences 1-7and1-8.
Thisstudy includes:
1.m:
Thisanalysis will beperformed todemonstrate compliance with voltage andthermal loading criteria andwill identify any systemupgrades to satisfy these criteria.
Theanalysis will verify thatthe power increase will nothave anysignificant adverse impact uponthereliability oroperating characteristics of thebulkpowersystem(grid).
Forsteady-state
- analysis, the maximum output for theMPS3generator will beusedduring summeroperation.
2.h:
Thisanalysis demonstrates that theproposed power increase will nothaveanysignificant adverse impactuponthestability,reliability, or operating characteristics ofthebulkpowersystems underthemost severe conditions.
Studies demonstrating dynamicperformance mustthensimulate conditions thatstress thesystembeyondtypical combinations ofload
- level, generation
- dispatch, andpowertransfers.
Forstability
- analysis, themaximum I
output fortheMPS3will beusedduring winter operation.
3.Short Circuit:
Thisanalysis will beconducted todemonstrate that theshort circuit duties will notexceed equipment capability andwill identify anysystemupgrades
=
required tosatisfy thecriterion.
Thebasecasewill include all generating facilities andelective transmission upgrades onthedatethestudy iscommenced that:
(i) aredirectly connected totheNewEngland Transmission System,(ii) are interconnected totheaffected systemsandmay havean impacton the interconnection request; and(iii) haveapending higher queuedposition andmay impact ontheMPS3interconnection request.
4.
Thesystem impact study will evaluate thecompliance ofthevoltage control capability with the requirements ofReference 1-9.
Serial No.21-166 Docket No.50-423 Attachment 1,Page21of21 Upon completion ofthestudy, DENCwill confirm theeffects ofgrid resiliency, switchyard functionality, andGDC-17compliance (which issummarized inMPS3FSARSection 3.1.2.17). If the studyproduces satisfactory results undertheMURuprated conditions, thenthe proposed changes fromtheMURLARwill beimplemented.
Ifthestudy produces unsatisfactoryresults undertheMURuprated conditions, theMURLARwill notbe implemented, and DENC will inform theNRCofthis outcome anddescribe theexpected actions for resolution.
References:
1-1.Dominion Energy Letter Serial No.07-0450, "Dominion Nuclear Connecticut, Inc.
Millstone PowerStation Unit 3 License AmendmentRequest Stretch Power Uprate,"
dated July 13,2007.
Attachment 5:MillstonePowerStation Unit 3Stretch PowerUprate Licensing Report
[ADAMS Accession No.ML072000400).
1-2.Millstone PowerStation Unit3 Safety Analysis
- Report, Revision 33.02,dated March31,2021.
1-3.CodeofFederal Regulations, 10CFR50.61, "Fracture ToughnessRequirements forProtection Against Pressurized Thermal Shock Events,"
dated January 4,2010.
1-4.Dominion EnergyLetter Serial No.20-043, "Dominion EnergyNuclear Connecticut, Inc.
Millstone PowerStation Unit 3 Proposed LicenseAmendment Request Measurement Uncertainty Recapture Power Uprate,"
datedNovember 19,2020[ADAMS Accession Number ML20324A703).
1-5.Westinghouse Report WCAP-16629-NP, Revision 0,"Analysis ofCapsule Wfrom theDominion Nuclear Connecticut Millstone Unit3 Reactor Vessel Radiation Surveillance Program,"
September 2006.
1-6.U.S.Nuclear Regulatory Commission, Office ofNuclear Regulatory
- Research, Regulatory Guide1.99, Revision 2,"Radiation Embrittlement ofReactor Vessel g
Materials,"
May1988[ADAMS Accession NumberML003740284).
1-7.ISONew England Planning Procedure 5-3,Guidelines forConducting and l
Evaluating Proposed PlanApplication Analysis.
1-8.ISONewEngland Planning Procedure 5-6,interconnection Planning Procedure forGeneration andElective Transmission Upgrades.
1-9.ISONewEngland 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 MILLSTONE POWERSTATIONUNIT3 DOMINIONENERGYNUCLEARCONNECTICUT, INC.
Serial No.21-166 Docket No.50-423 Attachment 2,Page1of1 Attachment 4,Section III.4-1.1 oftheMillstone PowerStation Unit3 (MPS3)
Measurement Uncertainty Recapture (MUR)
License AmendmentRequest(LAR) submittal, dated November 19,2020(ADAMS Accession No.ML20324A702),
discussed theplanned reanalysis ofselect Final Safety Analysis Report(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) eventinSection II.1.10 alsodescribed howthis event will bereanalyzed to implement WCAP-17642-P-A.
Thepurpose ofthis supplement istoclarify that the methodology selected fortheDominion Energy Nuclear Connecticut, Inc.(DENC) planned LONF reanalysis haschanged fromwhatispresented intheMUR LARAttachment 4,SectionsII.1.10 andIll.4-1.1.Instead, theLONFeventrevision will beperformed using analternate, NRC-approved, methodology.
Thecurrent FSARChapter 15.2.7 LONF event wasperformed using theWestinghouse RETRANtransient analysis method (WCAP-14882-P-A) [Reference 2-2).
TheLONF revision will beperformed usingthesame RETRAN methodasthecurrent LONF analysis.
Therevised analysis will beconfirmed each reloadusing theDominion Energy ReloadMethodology (VEP-FRD-42-A)
[Reference 2-3). Prior toPAD5implementation, DENCwill perform a review oftheLONFrevision and submitanyitemsdeemed necessary forNRCreview perthecriteria of10CFR50.59.
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 forWestinghouse Pressurized WaterReactor NON-LOCASafety Analyses,"
April 1999.
2-3.Topical Report VEP-FRD-42-A, Revision 2,MinorRevision 2,"ReloadNuclear Design Methodology,"
October 2017.