RS-12-033, Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate

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

ML12052A113
Person / Time
Site:	Byron, Braidwood
Issue date:	02/20/2012
From:	Borton K Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-12-033
Download: ML12052A113 (38)

Text

Exeion Exeion Generation

.DnMoT'0n Company, L LC www.exeloncorp.com 4300 Winfield Road 4300 Road Generation G en e-ration Warrenv ille, IILL60555 Warrenville, b0555 10 10 CFR CFR 50.90 RS-12-033 RS-12-033 February 20,2012 February 20, 2012 U. S. Nuclear U. Nuclear Regulatory Regulatory Commission Commission ATTN: Document Control Control Desk Washington, DC DC 20555-0001 20555-0001 Braidwood Station, Braidwood Station, Units Units 1 and 2 Operating License Facility Operating License Nos.

Nos. NPF-72 NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Byron Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. Nos. STN 50-454 and STN 50-455

Subject:

Additional Information Additional InformationSupporting SupportingRequest Request for for License License Amendment Regarding Regarding Measurement Uncertainty Uncertainty Recapture Power Power Uprate Uprate

References:

1. Letter from from Craig Craig Lambert Lambert (Exelon (Exelon Generation Generation Company, Company, LLC)LLC) to u. S.

to U. S. NRC, NRC, "Request for License Amendment Regarding Regarding Measurement MeasurementUncertainty Uncertainty Recapture Power Power Uprate,"

Uprate," dated dated June June23, 2011 23,2011

2. Letter fromfrom B.B. Mozafari Mozafari (U.(U.S. S. NRC)

NRC) to to M.

M.J.J. Pacilio Pacilio (Exelon (Exelon Generation Generation Company, LLC), "Byron Station, Unit LLC), "Byron Station, Unit Nos. Nos. 1 and 2, and and BraidwoodStation, Braidwood Station, Units 1 and and 22-- Request Requestfor forAdditional AdditionalInformation Information RE: Measurement Measurement Uncertainty Recapture Power Power Uprate Uprate Request Request(TAC (TAG NOS.NOS. ME6587, ME6587, ME6588, ME6588, 6589, AND ME6590),"dated AND ME6590)," dated February 14, 14, 2012 2012[ML [ML120270146]

120270146]

3. Letter from B. Mozafari from B. Mozafari (U.(U. S.

S. NRC) to M. M. J.

J. Pacilio (Exelon (Exelon Generation Generation Company, LLC), "Byron Station, LLC), "Byron Station, Unit Unit Nos.

Nos. 1 and 2 and andBraidwood BraidwoodStation, Station, Units 11 and and 22--Request Requestfor forAdditional AdditionalInformation Information RE: RE: Measurement Measurement Uncertainty Recapture Power Power Uprate Uprate Request Request(TAC (TAG NOS.

NOS. ME6587, ME6587, ME6588, ME6588, 6589, ANDAND ME6590),"

ME6590),"dated dated February February 14, 20122012 [ML[ML120260936]

120260936]

In Reference 1, 1, Exelon Exelon Generation Generation Company, Company, LLC LLC (EGC)

(EGC) requested requested an anamendment amendmenttotoFacility Facility Operating License License Nos. Nos. NPF-72, NPF-72,NPF-77, NPF-77,NPF-37 NPF-37and andNPF-66 NPF-66for forBraidwood BraidwoodStation, Station,Units Units11 and 2,2, and and ByronByron Station, Units respectively. Specifically, Units 1 and 2, respectively. Specifically, the the proposed changes changes revise revise the Operating Operating License License andand Technical Technical Specifications Specifications to implement an increase increase in in rated rated thermal thermal power of approximately1.63%

of approximately 1.63%based based onon increased feedwater flow flow measurement measurement accuracy. In In References References 22and and3,3,thetheNRC NRCrequested requestedadditional additionalinformation informationto tosupport supportreview reviewof ofthe the proposed proposed changes.

changes.InInresponse responsetotothis thisrequest, request,EGC EGCis isproviding providingthe theattached attachedinformation informationfor for all of the requests requestswith with the the exception exceptionof ofthe theCivil Civil and and Mechanical Mechanical Branch Branch[ECMB]

[ECMB] Request Request 13 13inin Reference Reference 22 and and the theBalance Balanceof ofPlant PlantBranch Branch[SBPB]

[SBPB] Request Request11in in Reference Reference3.3.EGC EGCwill willbe be

20, 2012 February 20,2012 U.S.

U.S. Nuclear Regulatory Commission Page 2 providing the providing the response to these two requests under under separate transmittal transmittal as as indicated indicated in in .

EGC has reviewed EGC reviewed the the information information supporting a finding finding of no no significant significant hazards hazards consideration and the environmental consideration provided to the NRC in in Reference Reference 1. The Theadditional additional information provided information providedininthis thissubmittal submittaldoes does notnot affect affect the the bases bases for for concluding that the proposed proposed license amendment does does notnot involve involve a significant hazards consideration.

significant hazards consideration. In In addition, addition, the additional information additional informationprovided providedininthisthissubmittal submittaldoes does not not affect affectthe the bases bases for concluding that neither an neither an environmental environmentalimpact impactstatement statementnor noran anenvironmental environmentalassessment assessment needs needs to be prepared in in connection with the with the proposed amendment.

amendment.

There are no regulatory commitments contained in in this letter.

letter.

Should you have any questions concerning this letter, letter, please please contact contact Leslie Leslie E.E. Holden Holden atat (630) 657-3316.

II declare declare under penalty of of perjury perjurythat that the the foregoing foregoingisistruetrueand and correct.

correct. Executed on on the the th 201h 20 dayof day ofFebruary February 2012.

Respectfully, Kevin F. Borton Manager, Licensing - Power Power Uprate Uprate : Response Response to to Request Request for for Additional Additional Information Information cc: NRC Regional Administrator, Region Region IIIIII NRC Senior Resident Inspector - Braidwood Braidwood Station Station NRC Senior Resident Inspector -- Byron Byron Station Illinois Emergency Management Agency - Division Division of ofNuclear NuclearSafety Safety

Braidwood and Braidwood and Byron Stations Measurement Measurement Uncertainty Recapture License LicenseAmendment AmendmentRequest Request(MUR (MURLAR)

LAR)

RESPONSE TO RESPONSE TO REQUEST REQUEST FOR FOR ADDITIONAL ADDITIONAL INFORMATION INFORMATION (RAI)

February 20, February 20, 2012 ATTACHMENT 1I ATTACHMENT RESPONSES TO RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION (NON-PROPRIETARY)

(NON.PROPRIETARY)

Braidwood/Byron StationsStations MUR MUR LAR LAR Response Response to to RAI RAI February 20, 2012 20,2012 Attachment Attachment 1, 1, page page 11 NON-PROPRIETARY NON-PROPRIETARY NRCIMechanical NRC/Mechanical and Civil Civil Engineering EngineeringBranch Branch(EMCB)

(EMCB)

NRCIEMCB NRC/EMCBRequest Request I1 Section IV.1.A.ii.f of Attachment 7 to the license amendment amendment request request (LAR)

(LAR) discusses the structural evaluation ofof the lower lower and upper support assemblies for the effects of upper core support of increased heat generation generation rates.

rates. Provide further information and confirm that a.

a. the proposed MUR power power uprate only only affects thethe design loads associated with with heat generation rates and all other design loads associated with with the design of of the the reactor vessel internals are unaffected by the proposed MUR power power uprate; uprate; b.

b. all design loading conditions, as noted in Section 3.9.5.2 of of the Byron and Braidwood updated final safety analysis report (UFSAR),

(UFSAR), were considered in the structural structural re-re-of the reactor vessel internal components to assess the impact of evaluation of of the proposed MUR power uprate; and c.

c. the original design codes of of record were utilized in the structural structural re-evaluation of ofthe the reactor vessel internal components.

Provide the maximum calculated stresses and cumulative cumUlative fatigue usage factor for the most limiting component of of the reactor vessel internals and their respective comparison with the Byron and Braidwood Braidwood design acceptance criteria.

design basis acceptance criteria.

Response

The Byron and Braidwood reactor vessel internal components analysis analysis of of record record (AOR)

(AOR) was performed with conservative gamma gamma heating heating rates. The Measurement Uncertainty Recapture (MUR) power uprate gamma heating rates were verified to remain remain bounded bounded by by the conservative conservative heating rates used in the AOR.

All the design loading conditions noted in in Section Section 3.9.5.2 3.9.5.2 ofof the the Byron Byron and and Braidwood Braidwood Updated Updated Final Safety Analysis Report (UFSAR) were considered in the structural assessment of of the reactor vessel internal components to assess the impact of the proposed MUR power power uprate.

uprate.

The design loads associated with the design of of the reactor reacto'r vessel internals internals remain remain bounded bounded by by the AOR.

The Byron and Braidwood Units 11 and 2 reactor reactor vessel internals internals components components were were designed designed introduction of Subsection NG of the American Society of prior to the introduction of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section III, and are not licensed to meet any specified edition or addenda of the ASME Code. As Asaaresult, result,aaplant-specific plant-specificstress stressreport reportofofthe the reactor internals was not required. However, the design of the reactor internals required. However, the design of the reactor internals is evaluated is evaluated according to the Westinghouse Criteria which is similar to the criteria described in the Subsection NG of the ASME code. The TheWestinghouse Westinghouseacceptance acceptancecriteria criteriaare arethe thesame sameas as those used in the original design of of the plant plant and and its its original original licensing licensing basis.

basis.

The maximum maximum calculated calculated stresses stressesand andcumulative cumulativefatigue fatigueusage usagefactor factorfor forthe themost-limiting most-limiting component of the reactor vessel internals internals are are unaffected unaffected by the MUR power power uprate uprate andand remain remain bounded by the AOR.

Braidwood/Byron Stations Braidwood/Byron Stations MUR MUR LAR LAR Response Response to to RAI February20,2012 February 20, 2012 Attachment 1, Attachment page 22 1, page NON-PROPRIETARY NON*PROPRIETARY NRCIEMCB Request NRCIEMCB Request 22 Section 3.9.5.1 Section 3.9.5.1 of of the the Byron Byron andand Braidwood Braidwood UFSAR UFSAR describes describes the the reactor reactor vessel internals as vessel internals as three parts consisting of the three parts consisting of the lower core lower core support support structure, structure, the the upper upper corecore support support structure, structure, and the incore the incore instrumentation instrumentation support structure. structure. Section Section IV IV of ofAttachment Attachment 77 to to the the LAR LAR does does notnot discuss the discuss the incore incore instrumentation instrumentation support structures. structures. Provide Provide further furtherinformation information relative relative toto the the impact of impact of the the design design conditions associated with with the the proposed MUR power uprate on the incore instrumentation support instrumentation support structures.

structures.

Response

Response As stated As stated in UFSAR Section 3.9.5.1, the in-core instrumentation support structures consist of an upper system upper system to to convey convey and and support thermocouples penetrating penetrating the vessel through the head and a lower system and system to to convey and support flux thimbles penetrating the vessel through the bottom.

The proposed MUR power uprate impact on the incore instrumentation support structures, both the including both the upper support columns and and the lower support columns was assessed. Since Since the current analyses loads (i.e. (Le. LOCA hydraulic forces and seismic loads) are not changing from the current analysis of record and remain remain bounded for the MUR power uprate, the stresses and the cumulative fatigue usage factors in these components remain unchanged unchanged from the current current analysis of record.

of record.

NRCIEMCB Request 3 Provide further further information and and confirm confirm that, that, for for the the proposed proposedMUR MURpower poweruprate uprateconditions, conditions,the the maximum deflection values allowed allowed for the reactor reactor vessel internal support support structures, structures, as noted in Table 3.9-4 of of the Byron and Braidwood UFSAR, are maintained. .

Response

The design inputs, i.e. Le. LOCA hydraulic and seismic forces and geometry, are not not changing changing fromfrom the current analysis of of record for the MUR power uprate; therefore, there is is nono impact on the allowable deflections provided in Byron and Braidwood Braidwood UFSAR UFSAR Table Table 3.9-4, 3.9-4, "Maximum "Maximum Deflections Allowed for Reactor Internal Support Structure." The Thevalues valuesprovided providedininUFSAR UFSAR Table 3.9-4 3.9-4 remain remain valid valid for for the the MUR MUR powerpoweruprate.

uprate.

NRCIEMCB Request Request 4 Section IV.1.B.iv.1 IV. 1.B.iv. 1 of ofAttachment Attachment 77to to the the LAR LAR states states that there is an approximate approximate 1.2°F 1.2°F increase in temperature difference increase in temperature difference across across the core core (That (Thot increases approximately approximately0.6°F O.6°FandandTC,,d Tcold decreases decreases approximately approximately 0.6°F) O.6°F) from from current current operating operating conditions conditions due due to to the the MUR MUR power power uprate.

uprate.Section IV.1.A.i IV.1.Ai of ofAttachment Attachment77to to the the LAR LAR discusses discusses reactor reactor vessel structural structural evaluation evaluation and and states states thatthat due due to to operational operational restrictions, restrictions, thetheMUR MURminimum minimum vessel vessel inlet inlet and and maximum maximum vessel vessel outlet outlet temperatures temperatures are are limited limited to to 538.2°F 538. 2°F andand 618.4°F, 618.4°F, respectively.

respectively. Provide Provide further further clarification clarification on on temperature temperature effects effects relative relative to to the the values values in in Tables Tables 3-1 3-1 and and 3-2 3-2 of of Attachment 1to to the the LAR, LAR, the thestatements statementsininSections SectionsIV.1.B.iv.

IV. 1.B.iv.I1andandIV.1.A.i IV.1.Aiofofthe theLAR, LAR, and and the the temperatures temperatures used used in in the the analysis analysisof ofrecord.

record.

Furthermore, Furthermore, the the lifting lifting lug loads loads and and evaluation evaluation are are discussed discussed in in Section Section IV.1.A.i IV.1.Ai of ofAttachment Attachment 77 to the the LAR. The Theterminology terminologyofof"lifting "liftinglug" lug"andandits itsrelation relation totoand anditsitsinclusion inclusionininthe theproposed proposed MUR MUR power power uprate uprate license license amendment amendment is is not not clear.

clear. Provide Providefurther furtherinformation informationtotoclarify clarifywhich which

Braidwood/Byron Stations Braidwood/Byron Stations MUR LAR Response to to RAI RAI February February 20,2012 20, 2012 Attachment Attachment 1, 1, page 3 NON-PROPRIETARY NON-PROPRIETARY reactor vessel reactor vessel component component is is referred referred toto as as "lifting "liftinglug".

lug". Also, Also, regarding regarding the the affected affected reactor reactor vessel component vessel component,

a. provide

a. provide a table table summarizing the the comparison of of design parameters parameters for the current current operation conditions, operation conditions, MUR power power uprate conditions, conditions, and design basis basis conditions; and conditions;

b. provide

b. provide thethe maximum calculated stresses stresses and and cumulative cumulative fatigue fatigue usage usage factorsfactors at at the the most most critical location of the affected of the affected component component and their respective comparison with with the Byron the Byron and Braidwood design basis basis acceptance acceptance criteria.

Response

Response The MUR The MUR power uprate Reactor Coolant System (RCS) design conditions given in in Tables 3-1 3-1 and 3-2 provide and provide aa Tag T avg range in which the the minimum minimumTodd T cold is is 541.4°F 541.4 OF andand thethemaximum maximumThor T hot is is 620.9°F. The 620.9°F. Thereactor reactorvessel vesselanalysis analysisofofrecordrecord(AOR)(AOR)evaluated evaluateda aminimum minimumTco,d T cold ofof 538.2°F 538.2°F and a maximum Thot and T hot ofof620.3°F.

620.3°F. Therefore, the MUR MUR power power uprate uprate maximum maximumThot T hot of 620.9°F exceeds exceeds the the maximum Thot evaluated in the reactor vessel T hot evaluated in the reactor vessel AOR. Note that AOR. Note that the MUR power uprate minimum minimum Too,d T cold isis bounded bounded by the minimum minimum Too,d evaluated in the reactor T cold evaluated reactor vesselvesselAOR. AOR.

Normally, a reconciliation would be Normally, be necessary necessary because because the the MUR MURpower poweruprate upratemaximummaximumThor T hot is is not bounded by the maximum Thor evaluated ininthe T hot evaluated the reactor reactor vessel vessel AOR.

AOR. However, However, all all Byron Byron and and Braidwood units have plant operational limits which restrict the minimum T"Id Braidwood T cold toto 538.2°F 538.2°F and and the maximum Thot T hot toto618.4°F.

618.4°F. The plant plant operational operational limitslimits will will remain remain inin place placefor forthetheMUR MUR power uprate. Therefore, Therefore,the theminimum minimumTco,d T cold and maximum maximum Thot evaluated in the reactor T hot evaluated reactor vessel vessel AOR bound those of the MUR power power uprate uprate when when thethe plant plant operational operational limits limits areare taken taken into into consideration.

There are three lifting lugs oriented 120° apart around the external side side of of the reactor reactor vesselvessel closure head. The The Integrated Head Package (IHP) lift rod assemblies attach to the liftinglugs Integrated Head Package (IHP) lift rod assemblies attach to the lifting lugs through a lift rod clevis and clevis pin. pin. Figures FiguresEMCB EMCBR4-1 R4-1 and andR4-2 R4-2depict depicthow howthe thelifting liftinglugs lugs are attached to the reactorreactor vessel vessel closure closure head. head.

The lifting lug mechanical mechanical loads loads identified identified for for current current operating operating conditions conditions did did notnot change change due due to to the MUR power uprate.

MUR power uprate.

Bottom Portion of of IHP IHP 1.1 fT toO AMeMILl I't. I I ,

t I  !  !

Lift Rod Clevis and Clevis Pin L I ,

Lifting Lug I i J I J J I I I i Figure EMCB Figure EMCBR4 R4--2: 2: Detail of Detail of Figure Figure EMCB EMCBR4 R4--1:1: Bottom Bottom Portion Portion ofofIntegrated Integrated Lifting Lug Lifting Lug Attachment Attachmentto to Reactor Reactor Head Head Package to Package to Reactor ReactorVessel VesselClosure ClosureHead Head Vessel Closure Vessel Closure Head Head

Braidwood/Byron Stations Braidwood/Byron Stations MUR LAR Response to to RAI RAI February February 20, 20, 2012 Attachment 1, page 4 Attachment NON-PROPRIETARY NRC/EMCB Request NRCIEMCB Request 55 Section IV.

Section IV.1.A.iii of Attachment 1.A.iii of Attachment 7 to to the the LAR LAR discusses discusses the the control control rod drive mechanism (CRDM). In this section, (GRDM). section, it is stated that that updated seismic and loss-of-coolant loss-of-coolant accident accident (LOCA)

(LOGA) loads remain less less than than the the allowable allowable loads loads provided provided in in the the analysis analysisof ofrecord.

record. This statement statement implies that the seismic loads have been updated. updated. Also, this this statement statement is not consistent with Section IV.1.A.ii.e Section IV.1.A.ii.e of Attachment 7 to to the the LAR where it is stated that the proposed MUR power power uprate conditions do uprate do not affect the current design basis for for seismic seismic andand LOGA LOCA loads.

loads. Provide further clarification.

Furthermore,Section IV.1.A.iii of Attachment 7 to Furthermore, to the the LAR s.tates states that GRDM CRDM is subjected to Tcold temperatures and reactor coolant system pressures and these are the only design Tcold parameters considered in in the the CRDM evaluation.

evaluation. Elaborate Elaborate and and confirm that:

a. the design basis loading conditions and

a. and operational operational requirements, requirements, as noted in Section 3.9.4 of thethe Byron and Braidwood UFSAR, have been been considered in the structural evaluation of of the control rod drive system for the the proposed proposed MUR power uprate conditions; and

b. the control rod drive system will continue to be in compliance with the Byron and b.

Braidwood design basis acceptance criteria under the proposed MUR power power uprate conditions.

Response

A seismic and loss of of coolant accident (LOCA) loads assessment assessment was completed completed as as part part of of the the MUR power uprate. The Theassessment assessmentconcluded concludedthat thatMURMURuprateuprateconditions conditionshave havenonoimpact impacton on the seismic/LOCA loads and the existing seismic/LOCA loads loads remain remain valid and unchanged unchanged for for the MUR power power uprate.

uprate.

The CRDM assessment completed for the MUR MUR uprate uprate project project considered considered allall pressure pressure and and thermal design transients and load combinations noted noted in in Section Section 3.9.4 3.9.4 of of the the Byron Byron Braidwood Braidwood UFSAR. The TheCRDM CRDM assessment assessmentconcluded concludedthat thatthe thepressure pressureand andthermal thermaldesign designtransients transients due to the MUR uprate uprate have no impact impact onon the CRDM CRDM qualification qualification analyses analyses of of record.

record. The The CRDM qualification analyses of of record demonstrated that Byron Byron and Braidwood Braidwood are in in compliance with the the ASME ASME Code Code stress stresscriteria.

criteria.

NRC/EMCB NRCIEMCB RequestRequest 66 Provide further further information information and andconfirm confirm that the design basis pressure and and temperatures (normal operating and and accident accident temperatures) temperatures) used in the design of of the the containment containmentstructure, structure, including the the steel liner liner plate, and and its internal structures remain bounding following following thethe proposed proposed MUR power poweruprate.

uprate.

Response

The design basis basis containment containment pressure pressure and and temperature temperature for for normal normal operation operation areare delineated delineated respectively in Byron/Braidwood Technical Specification Specification 3.6.4 3.6.4 and and 3.6.5.

3.6.5. Assessments Assessments performed for for the the MUR MUR power power uprate uprate concluded concludedthat thatthese thesenormalnormaloperation operationdesign designparameters parameters remain applicable.

applicable.

Accident Accident containment containment parameters parameters were were evaluated evaluated for for the the MURMUR power power uprate.

uprate. ForForprimary primary system system pipe pipe breaks breaks (i.e.,

(i.e., LOCAs),

LOCAs), as as discussed discussed in in the the MUR MUR LAR LAR submittal submittal (Reference (Reference 1), 1),

Section Section 111.15.5, "LOCALong 111.15.5, "LOCA LongTerm TermMassMassandandEnergy Energy Release Release and and Containment Response Response- -

Braidwood/Byron Braidwood/Byron Stations MUR LAR Response Response to RAI RAI February February 20,20, 2012 Attachment 1, page 5 Attachment NON-PROPRIETARY NON-PROPRIETARY UFSAR 6.2.1.3.1, Analysis UFSAR Analysis Results,"

Results," the the containment containment peak pressure and temperature temperature for for the the MUR bounded by MUR remain bounded by the the containment containment structure structure design pressure and temperature with with margin.

margin.

For For secondary secondary pipe pipe breaks breaks (Main (Main Steam Line Breaks (MSLB)), (MSLB>>, as as discussed discussed in in the MUR MURLAR LAR submittal (Reference submittal (Reference 1), Section 111.16.5, "MainSteam 111.16.5, "Main SteamLine LineBreak Break Mass Mass andand Energy Releases Releases Inside Containment Inside Containment-- UFSAR UFSAR 6.2.1.4, Analysis Analysis Results,"

Results," the peak peak pressure pressure remains remains bounded bounded by by the the containment containment design design pressure with with margin margin and there is is aa very very small small calculated calculated increase increase

(+0.6°F) inin the peak peak containment containment air temperature for Unit 1. Unit Unit22remains remainsbounded boundedby bythe the analysis of record.

Exelon's Exelon's response response (Reference (Reference 2) to to the NRC NRCRequestRequestfor forAdditional Additional Information Information (Reference (Reference 3) 3)

Request 10, summarized summarized the the temperatures temperatures and and pressures pressures from from thethe LOCA and and MSLB Mass and Energy Analyses for for Byron /Braidwood MUR.

Byron/Braidwood discussed in As discussed in the the UFSAR SectionSection 6.2.1.1.3, 6.2.1.1.3,"Containment "ContainmentStructure, Structure,Design DesignEvaluation,"

Evaluation,"the the justification justification for for the the design design temperatures temperatures selected for the liner and internal internal containment structures structures is is that that they they are are conservative when the duration duration of of the peak peak temperature temperaturefor forthe the secondary secondary side side (Le.,

(i.e., steam line) line) break, the temperature lag between the containment containment atmosphere and the the passive passive heatheatsinks sinks such such as asthe thecontainment containmentliner linerand andinternal internalstructures, structures, and the resistance resistance to to heat heat transfer transfer provided provided by bythe the materials materialsused,used,are are considered.

considered. This justification justification remains remains applicable applicable for for MUR power power uprate uprate because becausethe theduration durationremains remainsshort.

short.

Figure Figure 10-1, "Containment "Containment EQ Temperature Temperature and and Pressure Pressure Profile,"

Profile," in Reference 22 shows shows thatthat the MSLB MSLB temperature temperature profileprofile for for the MUR MUR powerpoweruprateupratefalls falls below belowthe thecontainment containmentdesign design temperature of 280°F 280° F less than 200 200 seconds seconds after after thethe onset onset of of the the MSLB.

MSLB.

The assessment assessment performed performed for for the MUR MURpowerpoweruprate uprateindicated indicatedthat thatthe thestructural structural effect effect ofof the the MSLB temperature on on the the containment containment structure structure remains bounded bounded by by the LOCA LOCAcase.case.

Therefore forfor both both units the containment structure remains acceptable acceptable for for both both primary primary and and secondary system pipe breaks.

For the containment containment internal internal structures, structures, RCS initial pressure RCS initial pressure and and temperature temperature for for MUR werewere reviewed and confirmed confirmed to be be bounded bounded by by thethe inputs inputs toto the the existing existing short-term short-term LOCA mass and energy releases. ThereforeThereforethe thecontainment containmentinternalinternal structures structures remain acceptable for for the MUR MUR power uprate.

uprate.

NRC/EMCB NRCIEMCB Request Request 7 Section IV.1.A.iv "Reactor Coolant Piping and Supports" of of Attachment 7 to the LAR discusses the effects of of the proposed MUR power power uprate mostly mostly on on aa qualitative basis and qualitative basis and the term term "no "no significant significant changes" changes" has has been been used in several several areasareas to describe the impact of of the proposed proposed MUR power power uprate. DiscussDiscussininmore moredetail detailthe theinformation informationrelative relativetotothe therevised reviseddesign design conditions, conditions, before and after after the proposed MUR power power uprate, uprate, for for those those components components evaluated evaluated under under Section Section IV. 9.A.iv of Attachment IV.1.A.iv Attachment 77 to to the the LAR.

LAR.

Summarize the the results results ofofany any additional additionalevaluations evaluationsperformedpetiormedfor forthe the affected affected components components and and indicate whether these these components remain bounded by the current analYSis analysis of record.

record. ForFor those components components that were not bounded that were bounded by by thethe analysis analysis of ofrecord:

record:

a. provide provide the the maximum maximum calculated stresses and calculated stresses and cumulative cumulative fatigue fatigue usage usage factors at at the the most most critical critical location; and and b.

b. provide provide further further clarification clarification that that the the re-evaluation re-evaluation was was performed petiormed in in accordance accordance with with the design basis code of design basis of record record and-the and*the affected affected components componentscontinuecontinuetotoremain remaininin compliance with with the the Byron Byron andand Braidwood Braidwood stations stations design basis acceptance design basis acceptance criteria.

criteria.

Braidwood/Byron Stations MUR LAR Response to Braidwood/Byron to RAI RAI February February 20,2012 20, 2012 Attachment Attachment 1, page 6 NON-PROPRIETARY NON-PROPRIETARY

Response

Response The conditions The conditions associated associated with the MUR power uprate were evaluated to determine the impact on the on the existing as-built design basis basis reactor reactor coolant coolant loop loop (RCL)

(RCL) analysis analysisfor forthe thefollowing:

following:

• • RCL piping RCL piping stresses and displacements,

• • Primary equipment Primary equipment nozzle loads (reactor (reactor pressure pressure vessel vessel (RPV)(RPV) inlet inlet and and outlet outletnozzles, nozzles, steam generator inlet and steam and outlet nozzles, nozzles, and reactor coolant pump (RCP) suction and discharge nozzles),

• • Primary equipment Primary equipment support loads (RPV (RPV nozzle nozzle supports, supports, steam steam generator generatorcolumns columnsand and lateral bumpers, RCP columns and lateral lateral lateral supports, supports, and and pressurizer pressurizersupports),

supports), and and

• Pressurizer surge line piping stresses and and displacements displacements including including the theeffects effectsof ofthermal thermal stratification.

following inputs were considered in the assessment:

The following

• Nuclear Steam Supply System (NSSS) Design Design Parameters, Parameters,

• NSSS design transients,

• Loss-of-coolant accident (LOCA) hydraulic hydraulic forcing forcing functions functions loads,loads, and and

• RPV motions due due to to LOCH.

LOCA.

The RCL piping assessment for the MUR power power uprate uprate was performed performed in in accordance accordance with the the Byron/Braidwood design basis basis code code of of record record (ASME, (ASME, Section SectionIII, III, 1974 1974Edition, Edition,including including Summer 1975 addenda).

The RCL thermal, deadweight, seismic, fatigue, LOCA LOCA and Main Main Steam Steam // Feedwater Feedwater line line break break analyses were reconciled reconciled to to the the design design inputs inputsas asfollows:

follows:

RCL Thermal Analysis The RCL piping in the existing design basis basis was evaluated evaluated for for the the conditions conditions associated associated with with a RCS hot leg upper upper bound bound temperature temperature of of618.4°F, 618.4°F, cross-over cross-overleg legtemperature temperatureofof555.4°F, 555.4°F, and a cold leg temperature of 555.7°F. 555.rF. The Thereactor reactorcoolant coolantupper upperboundboundtemperatures temperaturesfor for the MUR power power uprate uprate diddid not not increase increase for for the hot hot leg, leg, they they decreased decreased by by 0.6°F 0.6°F forfor the the cross-over leg, and they decreased by 0.6°F for the cold leg as as compared to the current current design basis temperatures. The MUR power uprate upper bound The MUR power uprate upper bound thermal NSSS design thermal NSSS design parameters are bounded by by the the design design basisbasisanalysis.

analysis.

Considering the RCL MUR power power uprate uprate lower lower bound bound temperature temperature case, case, there there isis aa temperature temperature operating operating windowwindowas asfollows:

follows:9.8°F 9.8°Fbetween betweenthe theupper upper bound bound Thigh Thigh and and lower lower bound bound Trout Ttowforforthe thehothot leg, leg, 16.9°F 16.9°F between betweenthe theupper upperbound bound Thigh Thighandandlower lowerbound boundTeow for the Ttowfor the cross-over cross-overleg,leg,andand16.9°F 16.9°Fbetween betweenthe theupper upper bound bound Thigh Thighandandlower lowerbound boundTlow for the cold Ttowforthe cold leg.

The thermal piping stresses stresses and displacements displacements are are dependent dependent on on the the coefficient coefficient of of thermal thermal expansion and temperature difference difference between between ambient ambient to to hot hot conditions.

conditions. The Thecoefficient coefficientofof thermal expansion expansion increases increases with with an an increase increase in in temperature.

temperature. The Thethermal thermalpiping pipingloadsloadsand and thermal stresses stresses for for the the lower lower bound bound temperatures temperatures are are lower lower thanthan the the corresponding corresponding loads loads and stresses for the upper upper bound bound case.

case. Therefore, Therefore,the thethermal thermalstresses stressesfor forthe theupper upperbound bound case are higher, higher, and the upper upper bound bound case case piping piping stresses, stresses, primary primary equipment equipment nozzle nozzle loads, primary primary equipment equipment support supportloads loads(including (includingthe thereactor reactorvessel, vessel,steam steamgenerator, generator, reactor coolant coolant pumppump and and pressurizer),

pressurizer), and and the theauxiliary auxiliarylinelinedisplacements displacementsatatthe the connections connections to to the the RCL RCL are are limiting.

limiting.

Braidwood/Byron Stations MUR LAR Response Braidwood/Byron Response to to RAI RAI February February 20, 20, 2012 Attachment 1, page 7 Attachment NON-PROPRIETARY NON-PROPRIETARY Since there Since there isis no no increase increase in in upper-bound upper-bound temperature temperature in in comparison comparison to the hot leg, cross-over leg, over leg, and and cold cold leg leg temperatures temperatures in in the the current current RCL RCL thermal thermal analysiS analysis design basis, the current RCL thermal analysis design basis current basis analysis analysis remains remains bounding.

bounding.

RCL Deadweight and Seismic Analysis RCL change in There is no change in deadweight because there is no change to the configuration of of the RCL piping RCL piping andand supports supports due due to to the the MUR MUR power uprate.uprate. The The seismic seismicresponse response spectrum spectrum change due to does not change to the the MUR MUR power uprate.uprate.. Therefore, it is concluded that there are no changes to changes to RCL RCL deadweight and seismic analyses for the MUR power uprate.

RCL Fatigue and Surge Line Stratification RCL changes to There are no changes to the the primary side NSSS design transients due to the MUR power Also, the pressurizer surge line transients do not change. Therefore, uprate. Also, Therefore, therethereisisnono impact on on the the piping piping for thethe MUR MUR power uprate due to the NSSS design transients. There Thereisis no adverse effect on no on the the fatigue fatigue evaluation of the RCL and pressurizer surge line, including thermal stratification.

the effects of thermal stratification. The The pressurizer pressurizersurgesurgeline linestratification stratification analysis analysis continues to meet the code of of record record (ASME, (ASME, Section Section III, III, 1986 1986Edition).

Edition).

LOCA Analysis The impact on the LOCA hydraulic forcing functions (HFFs) due to the MUR power power uprate uprate has been has been assessed assessed for the accumulator and surge line breaks. Based Based on on this thisassessment, assessment, the LOCA HFFs used in the existing RCL piping LOCA analyses remains bounding bounding for for the the MUR power power uprate.

uprate.

The impact on the RPV motions due to MUR power power uprate uprate has has been been assessed.

assessed. Based Basedon on this assessment, the LOCA RPV motions used in the existing RCL piping piping LOCA LOCA analyses analyses remains bounding for the MUR power power uprate.

uprate.

Main Steam and Feedwater Feedwater Line Line Break Break The design basis main steam and feedwater line line break break analyses remain remain valid valid for for the the MUR MUR power uprate. Based Based on on the the NSSS NSSS design designparameters, parameters, the themain mainsteam steamlinelineandandfeedwater feedwater line break pressures decrease and the feedwater temperature decreases slightly for the MUR power uprate. AA decrease decreaseininpressure pressurewillwillreduce reducethe thethrust thrustand andjetjetimpingement impingement forces; however a decrease in temperature may increase increase the forces due to fluid fluid momentum.

momentum.

These small differences will offset each other such that the thrust and and jet jet impingement impingement forces used in in the current current analysis analysis remain remain bounding.

bounding.

Based on the above, there are no changes due to the MUR power uprate uprate to the piping piping or or component qualification qualification from the design basis, including: including: primary primary equipment equipment nozzles nozzles and and supports, Class 11 auxiliary piping analysis, and surge line line stratification.

stratification. TheThemaximum maximumprimaryprimary and secondary secondary stresses stresses and the maximum maximum fatigue fatigue usage usage factors factors associated associated withwith the the existing existing design basis analysis analysis areare applicable to the MUR MUR power power uprate.

uprate. The Theabove abovecomponents components continue to to remain remain in in compliance compliance with with the the Byron/Braidwood Byron/Braidwooddesign designbasis basisacceptance acceptancecriteria.

criteria.

NRC/EMCB NRCIEMCB Request Request 88 Section Section I IV.

V. 1.A.

1.A.vvof ofAttachment Attachment 77 to to the LAR discusses discusses the evaluation evaluation of ofbalance balance of ofplant plant(BOP)

(BOP) piping systems.

piping systems. Confirm that other BOP piping other BOP piping systemssystems (e.g.,

(e.g., chemical chemical andand volume volume control, control, auxiliary auxiliary feedwater, feedwater, fuelfuel pool pool cooling, cooling, containment containment spray,spray, essential essential service service water, water, safety safety injection) injection) that that may may be be affected affected by by the the MUR MUR uprate uprate conditions conditions have have been been evaluated evaluated and and provide provide aa complete listlist of of BOP BOP piping piping systems systems evaluated evaluated in in support support of of MUR power power uprate. DiscussDiscussthe the

Braidwood/Byron Stations Braidwood/Byron Stations MUR LAR Response to to RAI RAI February February 20,2012 20, 2012 Attachment Attachment 1, page 8 NON-PROPRIETARY NON*PROPRIETARY methodology used methodology used for for evaluating evaluating BOP piping, piping, including including pipe pipe supports, supports, and provide further information relative information relative to to the the design design conditions in eacheach BOP BOP piping piping system, before before and and after after the the proposed MUR power power uprate.

uprate. Summarize Summarize the the results results of of the the additional evaluations performed for the for the affected piping piping systems and indicate whether these these piping systems remain remain bounded by the current analysis of the current analysis of record. record. For those BOP those BOP piping piping systems systems not bounded by the the current analysis of record:

analysis

a. provide the maximum calculated stresses and cumulative fatigue usage

a. usage factors factors at the the most critical location in each unbounded piping system; and

b. provide further clarification that the re-evaluation of

b. of the piping system, including pipe supports, was performed supports, performed in in accordance.

accordance with the design basis code of record and in compliance with compliance with the Byron and Braidwood stations design basis acceptance criteria.

Furthermore, state whether any piping or pipe support modifications are required to support the Furthermore, proposed MUR proposed MUR power power uprate.

uprate.

Response

The following Byron and Braidwood Stations Balance of of Plant/Nuclear Steam Supply Supply System System (BOP/NSSS) piping systems were assessed for MUR power uprate conditions:

• Main Steam System

• n Extraction Steam System

• n Condensate System

• Condensate Booster Booster System System

• Heater Drains System

• n F Feedwater eedwater System

• n Steam Generator Blowdown Blowdown System System n* Auxiliary Steam SystemSystem

• Auxiliary Feedwater Feedwater System System n* Chemical and Volume Control Control System System

• Fuel Pool Cooling System System n* Safety Injection Injection System System n* Essential Service Water Water System System

• Component Cooling Cooling Water Water System System

• Containment Containment SpraySpray System System

• Non-Essential Service Water Water n* Circulating Circulating Water It was determined that that the the following following Byron Byron and and Braidwood Braidwood Stations Stations BOP/NSSS BOP/NSSS piping piping systems systems are not not negatively negatively impacted (i.e.,

(i.e., an increase inin temperature or or pressure) pressure) by by MUR MUR power power uprate:

uprate:

• Steam Generator Generator Blowdown Blowdown System System

• Auxiliary Feedwater Feedwater System System

• Chemical and and Volume Control Volume Control System System

• Safety Injection System Injection System

• Containment Spray Containment Spray System System

• Circulating Circulating Water Water

Braidwood/Byron Braidwood/Byron Stations MUR LAR Response Response to to RAI RAI February 20,2012 20, 2012 Attachment Attachment 1, 1, page 9 NON-PROPRIETARY these piping For these piping systems systems no no further further assessment assessment was was performed.

performed. TheseThesesystems systems remain remain bounded.

the remaining For the remaining systems systems (i.e., those that were assessed to to have have an an increase increaseinintemperature temperature and/or pressure) the methodology and acceptance criteria criteria discussed in the paragraphs below below were applied to assess the acceptability acceptability of the piping piping forfor the the MUR MUR power power uprate.

uprate.

Operating pressures Operating pressures and and temperatures temperatures in in each each line line under under Current Current Licensed Licensed Thermal Thermal Power (CLTP)

(CL and MUR power TP) and power uprate uprate were were reviewed reviewed againstagainst the the design design pressure pressure and and temperature temperature of of the line.

For non -seismic piping, non-seismic piping, the the increase in pressure was considered to to be be acceptable acceptableprovided providedthatthat the MUR power uprate uprate operating operating pressure pressure was was bounded bounded by by the the design design pressure.

pressure. As a result result of the MUR MUR power power uprate, uprate, there there were were nono non-seismic non-seismic systems systems that exceeded the the design design pressure.

pressure.

piping, there For seismic piping, there were no no pressure pressure increases increases as as aaresult resultofofthe the MUR power uprate.

The increase increase in in temperature temperature was was considered considered to to be be acceptable acceptable provided that the MUR power uprate operating temperature did did not not increase increase by by more more than than 11%  % compared compared to CLTP CLTP operating temperature or the MUR MUR operating operating temperature remained remained less less than 150150°° F. For lines that are currently qualified to be within Code thermal stress stress allowable, increasing the system temperature range by <1 <1 %% will will not affect the acceptability of the piping/support system.

Decreasing the system temperature will will increase the allowable stress stress margin.

margin. For Forevaluating evaluating pipe thermal pipe thermal expansion expansion stress, stress, the the temperature range is equal to the maximum maximum operating operating temperature minusminus the normal ambient temperature, or 70°F. This represents 70°F. This represents the largest change in in temperature that the pipelinespipelines can experience.

experience. Typically, pipe thermal stress stress is is not not evaluated for operating temperatures less than 150°F. 150°F.

For piping segments which do not pass the screening criterion criterion (i.e., <1<1 %% change),

change), a detailed review of pipe stress calculations calculations is is conducted conducted to to determine determine if margin existsexists toto accommodate thermal expansion stresses at MUR MUR power poweruprate.

uprate.

All of of the systems, except for the heater heater drain drain piping piping and condensate booster booster piping piping are considered to remain bounded bounded basedbased on on the theabove abovecriteria.

criteria. The heater drain system piping piping experiences a maximum temperature increase of 1.43%. 1.43%. The The design design basis analysis analYSis was foundfound to bound the MUR condition because the design basis analysis used operating temperature of of 187°F while the CLTP operating temperature is is 160.8°F 160.8°F and and thethe MUR MURoperating operating temperature temperature is is 162.1°F temperature. The 162.1°F Thecondensate condensateboosterboosterpiping piping experiences experiences aamaximum maximu mtemperature temperature increase of 1.10%.

1.10%. The The design design basis analysis was found found to bound MUR MUR conditions conditions because the the design basis analysis used an an operating operating temperature temperature of 176°F while the CLTP CLTP operating operating temperature is 161.0°F and the MUR MUR operating operating temperature temperature is 162.0°F. Therefore, the BOP/NSSS piping piping systems are considered considered to remain in to remain in compliance compliance with with their their current current design design basis code of record and the Byron and Braidwood Braidwood stationsstations design basis acceptance acceptance criteria.

criteria.

there were Since there were no no significant significant increases increases in in piping piping temperatures, temperatures, pipe pipesupport support loads loads did did not not experience an an appreciable appreciable increase.

increase. Therefore, no pipe or pipe pipe support modifications are required for MUR power power uprate uprate conditions.

conditions.

NRCIEMCB Request NRC/EMCB Request 9 Section IV. I.A.viii IV. 1.A. viii ofofAttachment Attachment77totothe theLAR LARdiscusses discussesthe thepressurizer pressurizer structural evaluation. InIn this section of of the LAR, LAR, ititisis stated stated that the revised design design parameters parametershave havean aninsignificant insignificant impact on the fatigue fatigue analysis results. ItItis is also stated stated that that the the proposed proposedMUR MUR power uprate has has aa negligible negligible impact on the qualification qualification of of thethepressurizer pressurizer surge, spray, safety safety and andrelief reliefnozzle nozzle structural weld overlay designs. designs. Provide Providefurther furtherinformation information to support support the the above qualitative

Braidwood/Byron Stations MUR LAR Response Braidwood/Byron Response to to RAI RAI February February 20, 20, 2012 Attachment 1, page 10 Attachment NON-PROPRIETARY NON-PROPRIETARY statements and to demonstrate statements demonstrate compliance with the Byron and Braidwood design basis basis acceptance criteria.

acceptance criteria. Also, Also, provide provide aa table table summarizing summarizingthe thecomparison comparisonof ofpressurizer pressurizerdesigndesign parameters for the current operation conditions, conditions, MUR power power uprate conditions, conditions, and and design design basis conditions.

Response

Response Heat-up of Heat-up of the the pressurizer from from the the cold cold condition condition to the hot standby condition is independent of plant power level plant level and and is is unaffected unaffected by an uprate which may affect RCS temperatures and transients between hot standby and 100% power operation. The Thepressurizer pressurizermaintainsmaintainsthe theRCSRCS pressure and pressure and provides provides a cushion cushion to accommodate changes in fluid volume and provides protection to the RCS.

overpressure protection RCS. The The temperature temperaturewithin within thethepressurizer pressurizerisisat atthe thesaturation saturation temperature. Therefore, temperature. Therefore, transients transients that that will will affect affect the fatigue fatigue analysis for pressurizer components are the result of changes to the fluid temperature entering the pressurizer, pressurizer, i.e.,i.e.,

insurge/ outsurge through the surge line line or or spray spray through through the the spray sprayline, line, ororasasaaresult resultinin changes to the transients affecting the pressurizer pressure transients. Previous PreviousWestinghouse Westinghouse evaluations of design transients following an MUR power uprate show that the only only transients transients that are affected are those that are the result of of the feedwater feedwater changeschanges and and affect affect only only the the steam generator secondary side components. There Thereare areno notransients transientsaffected affectedthat thatpertain pertaintoto pressurizer, temperature the pressurizer, temperature or or pressure.

pressure. Therefore, there is no impact on the pressurizer result of analysis as a result of MUR MUR power power uprate uprate transient transient changes.

changes. Given Given thatthat the transients are unchanged, the impact on the lower pressurizer components due to insurge/outsurge and the unchanged, upper pressurizer components due to spray will change only ifif the temperature of of the fluid fluid changes, and then only if the temperature change increases.

changes, increases. For Forthis thistotohappen, happen,the theRCS RCS temperature for temperature for Thot, T hot, affecting insurge/outsurge, insurge/outsurge, and andTcp,d, T cold, affecting the spray spray temperature, temperature, would have to decrease from the analyzed condition.

The Table EMCB EMCB R9-1 R9-1provides providesaa comparison comparisonshowing showingthe thetemperature temperature change change across across the pressurizer components evaluated evaluated forfor the design basis conditions, the current design basis conditions, the current operating operating conditions, conditions, and at at MUR MUR power uprate conditions. ItIt isis seen seenfromfromTable TableEMCB EMCB R9-1 R9-1 that that the the temperature temperaturechange changefor forTh0t, T hot, affecting the lower lower pressurizer pressurizer(AThot),

(L1 T hot), isis less less at MUR MUR power uprate conditions by 0.6 OF OFand andisisenveloped enveloped by by the the analysis analysis of record (AOR). (AOR). The The temperature temperature differential forthe differential for theupper upper portion portionof ofthe thepressurizer pressurizer is is shown shown to to exceed the the current current operating operating condition by 0.6 OF (ATWd).

OF (L1 Thisisisan T cold). This anincrease increase of approximately approximately 0.5% over the the current current operating condition L1 AT.,d T cold and is not considered considered to to be besignificant.

significant.

Also, since the baseline analysis, which is is also also the AOR,AOR, continues continues to to envelope envelope the the MUR MUR powerpower uprate temperature differential, the AOR is not affected and remains applicable. applicable. Therefore, Therefore, there is is no no change to the baselinebaseline analysis analysis results results due due to to the the MUR MUR power power uprateuprate resulting resulting from from changes to the RCS temperatures affecting the to the RCS temperatures affecting the pressurizer. pressurizer.

An assessment of of the pressurizer pressurizer surge, surge, spray, safetysafety andand relief relief nozzle nozzle for for structural structural weld weld overlay (SWOL) was also performed as as part part of of the MUR MUR power power uprate.

uprate. The Theassessment assessment concluded that the MUR MUR power power uprate uprate would would have have no no impact impact on on the the AORAOR for for these these components components based on the findings previously noted. noted. Therefore, Therefore,the theMUR MURpower poweruprate uprateisisenveloped envelopedby bythethe current SWOL SWOL analysis analysis and and isis acceptable.

acceptable.

Braidwood/Byron Braidwood/Byron Stations MUR LAR Response to to RAI RAI February 20,2012 20, 2012 Attachment 1, 1, page 11 11 NON-PROPRIETARY Table EMCB EMCB R9-1: Comparison of Byron/Braidwood Pressurizer Analysis Basis Baseline Current MUR Analysis Operating Operating Parameter (AOR) Conditions Conditions (OF)

(°F) (OF)

(°F) (OF)

(°F)

T Tpressurizer pressurizer 652.7 652.7 652.7 That Thot 542.7 608 608.6 T cold Tcad 517.7 542 541.4 11 =

That = Tpressurizer AThot T pressurizer -- Thot That 110 44.7 44.1 11 ATcold =

T cold = Tpressurizer T pressurizer --

135 110.7 110 . 7 111.3 111 . 3 Teold Tcold NRC/EMCB Request 10 NRCIEMCB Section IV.1.B.iii Section of Attachment IV.1.B.iii of Attachment 7 to7the to theLARLAR discusses the the evaluation of the the reactor vessel internal components for flow induced vibration (FIV) (FIV) impact under MUR power uprate Also,Section IV.1.A.ii.e of conditions. Also, of Attachment 77 to the LAR states that the FIV stress levels on the core barrel assembly and upper internals are below the material high-cycle fatigue endurance limit and the proposed MUR uprated conditions do not affect the structural margin for FIV. Provide further information relative to to those those design design parameters, parameters, before and after MUR power power uprate, which could potentiallypotentially influence FIV FIV response of of the reactor internals.

internals. Also, Also, discuss the comparison of alternating stress intensities to design of alternating stress intensities to design basis allowable limits for for the the most critical components demonstrating compliance with the Byron and Braidwood design basis acceptance criteria.

acceptance criteria.

Response

Comparisons of of flow induced induced vibration vibration (FIV)(FIV) design design parameters parameters before before and and after afterthetheMUR MURpower power uprate are provided in Table EMCB EMCB RIO-1.R10-1.

Table EMCB EMCB R10-1: R10-1: Comparison Comparison of of FIV FIV Evaluation Input Design Parameters Input Design Parameters Current Analysis Analysis MUR Power MUR Power Parameter Ratio of Record Uprate Mechanical Design Flow (gpm/loop) 107,000 107,000 1.0 (gpmlloop)

Vessel Inlet Inlet Temperature Temperature (°F) CF) /1 542/

5421 541.4/

541.41

-1.0 Fluid Density Density (ibm/ft3)

(Ibm/fe) 47.369 47.385 Outlet Temperature Vessel Outlet Temperature (°F) (OF) /1 608/

6081 608.6/

608.61

-1.0 Fluid Density Density (Ibm/fe)

(Ibm/ft) 42.4535 42.411 The MUR MUR power power uprate uprate design designconditions conditionswill will slightly slightly alter alter the the Toad T cold and and Thot T hot fluid fluid densities, densities, which will slightly change the forces, induced slightly change induced by flow. The The corresponding corresponding Tcord T cold and and Thot That fluid fluid densities change by less than than 0.1 %  % from from thethe current current analyzed analyzed condition.

condition. Therefore, Therefore, thethe effect effect on the flow-induced vibration vibration stresses (alternating stress intensities) intensities) due to to MUR MUR power power uprate uprate on the reactor reactor internals internals remains remains unchanged unchanged from from the the current current analysis analysis ofof record.

record.

Braidwood/Byron Stations MUR LAR LAR Response Response to to RAI RAI February 20, 2012 Attachment 1, page 12 NON*PROPRIETARY NON-PROPRIETARY NRCIEMCB NRC/EMCB Request 11 Discuss further information Discuss information and confirm confirm that the the nuclear steam supply system component supports, as discussed in in Section 3.9.3.4 of the Byron and Braidwood UFSAR, UFSAR, will continue to be inin compliance compliance with with the Byron and Braidwood design basis acceptance criteria at the proposed MUR power power uprate uprate conditions.

conditions. Also, confirm that the operating temperatures for support elements, as defined in Table 3.9-17 of of the Byron and Braidwood UFSAR, are not affected by the MUR powerpower uprate.

uprate.

Response

The NSSS component supports, which include the reactor vessel, steam generator, reactor reactor pump, and pressurizer equipment supports, were assessed for the MUR power coolant pump, power uprate uprate as discussed in the response to EMCB R-7 and were shown to remain remain acceptable and and bounded bounded the current design by the design basis.

basis. Therefore, Therefore, thethe NSSS NSSS component componentsupports supportswill willremain remaininin compliance with UFSAR Section Section 3.9.3.4.

3.9.3.4.

The operating temperatures of the supports, as outlined in Table 3.9-17 of of the UFSAR, UFSAR, are are not not affected by the MUR power uprate. The MUR power uprate does not affected not require require an an increase increase in in the the ambient containment temperature design value. Further, Further, the thesmall smallchanges changesto tothe theNSSS NSSSdesigndesign temperatures, as discussed in the response to EMCB R-7, do not temperatures, not require require a change to the the operating temperature of the supports attached to the steam generator, reactor coolant pump, pump, reactor vessel, or pressurizer.

pressurizer.

NRC/EMCB Request 12 NRCIEMCB 12 IV.1.A.vi.1.b Section IV.1.A. vi.1.b of Attachment 7 to the LAR discusses the structural evaluation evaluation of ofByron Byron and Braidwood Unit I1replacement replacementsteam steamgenerators generatorsandandstates statesthat thataareconciliation reconciliationanalysis analYSis was performed to address the structural integrity of of the the entire entire steam steam generator generatorpressure pressure boundary for the MUR power power uprate conditions. DiscussDiscussfurther furtherinformation informationrelative relativeto,to, before before and after uprate, the maximum stress intenSity intensity and the cumulative fatigue usage factors for the critical components of the primary and secondary sides, including nozzles, of of the replacement steam generators and the respective service conditions. Also, Also, confirm confirmthat thatthe thereconciliation reconciliation analysis was performed in accordance with the original design code of of record and in compliance with the Byron andand Braidwood Braidwood stations stations design design basis acceptance acceptance criteria.

criteria.

Response

During the structural integrity analysis analysis of of the the replacement replacement steam steamgenerators generators(RSGs) (RSGs)on onUnit Unit11 for MUR conditions it was concluded that the maximum primary and and secondary secondary side side temperatures and pressures specified for MUR power power uprate uprate conditions conditions werewere lessless than than thethe primary and secondary side temperatures temperatures and and pressures pressures specified specifiedfor forthe theoriginal originalanalysis.

analysis.

Therefore, there are no no changes changes toto the the calculated calculated stress values or limits limits for for design design conditions conditions (i.e., name plate conditions).

However, a reconciliation analysis was performed for critical components of of the replacement steam generators due to differences in the Level A & B B (Normal and Upset),Upset), Level Level C C (Emergency) and Level D (Faulted) condition loads. The Thestress stress intensities intensities and cumulative cumulative usage factors for these service conditions for pre-MUR and post-MUR post-MUR power power uprate uprate conditions conditions are included inin Tables Tables EMCB EMCB R12-1 R12-1 though though R12-4.

R12-4.

The reconciliation analysis analysis was was performed performed in inaccordance accordancewithwiththe theoriginal originaldesign designcode codeofofrecord record as required by the current Certified Design Specification. Specifically, the acceptance Specifically, the acceptance criteria criteria

Braidwood/Byron Braidwood /ByronStations StationsMUR MUR LAR LAR Response to RAI February February 20,2012 20, 2012 Attachment Attachment 1,1, page page 13 NON-PROPRIETARY NON-PROPRIETARY for for the the reconciliation reconciliation of of the the pressure pressure boundary components were those specified specified in in the 1986 1986 ASME B&PV Code ASME Code with with no Addenda, Addenda, for for Section Section III, III, Class Class 11 components.

components. The Code acceptance criteria acceptance criteria are unchanged from the original RSG analysis.

analysis.

Braidwood/Byron Stations Stations MUR MUR LAR LAR Response Responseto toRAI RAI 2012 February 20, 2012 Attachment 1, 1, page page 14 14 NON-PROPRIETARY Table EMCB R12-1:

R12*1: Stress Intensity (SI) and Fatigue Usage Usage Factors Factors (FUF) for Level A & B Conditions (FUF) for MUR Orlg.

Orig. MUR Orlg.

Orig.

MUR Orlg.

Orig. FUF Component fI Location SIRange SI Range SI Range SI Limit SI SI Limit FUF FUF limit Limit (ksi) (ksi) (ksi) (ksi)

Tubesheet Primary Head I Tubesheet Juncture 38.5* 82.1 80.1 87.3 0.880 0.741 1.0 1.0 Secondary Shell I/ Tubesheet Juncture Juncture 86.4 85.4 95.0 87.3 0.160 0.223 1.0 1.0 Tubesheet Perforated Region Region 90.1 90.0 95.0 93.6 0.330 0.387 1.0 1.0 Primary Nozzle Primary nozzle 67.85 67.85 80.1 80.1 0.839 0.839 1.0 1.0 Primary nozzle safe end 57.37 57.37 60.3 60.3 0.096 0.096 1.0 1.0 Primary Manway Cover 30.3 30.3 80.1 80.1 0.006 0.006 1.0 1.0 Shell/flange Shelilflange 46 46.0 80.1 80.1 0.121 0.121 1.0 1.0 See Table EMCB EMCB R12-4 R12-4 for for Average Average and and Stud 0.871 0.871 1.0 1.0 Maximum Bolt Stresses Stresses Primary Head Support Support Pad Pad 79.4 79.4 80 80 0.67 0.67 0.67 1 1.0 1.0 Primary Divider Divider Plate Plate 63.9 63.9 69.9 69.9 0.905 0.904 0.904 1 1.0 1.0 Small Nozzles

%" Nozzles

%" 13.96 11.83 26.7 26.7 0.81 0.679 1.0 1.0 Steam Drum/Cone/Lower Drum/ConefLower Shell Assembly 74.22 62.9 80.1 80.1 0.025 0.021 1.0 1.0

Braidwood/Byron Braidwood /Byron Stations MUR LAR Response to to RAI RAI February February 20, 20, 2012 Attachment 1, 1, page 15 NON-PROPRIETARY NON-PROPRIETARY Table EMCB Table EMCB R12-1:

R12-1: Stress Stress Intensity Intensity (SI)

(51) and and Fatigue Fatigue Usage Usage Factors Factors (FUF)

(FUF) for for Level Level AA&&BB Conditions Conditions MUR Orig. MUR Orig.

Component I/ Location Location MUR MUR Orlg.

Orig. FUF SIRange SI Range SIRange SI Range SI SI Limit SI SI Limit Limit FUF FUF FUF Limit (ksi) (ksi) (ksi) (ksi) 8" Shell Cone Handhole 8"

Shell/cover/flange 67.3 57 80 80 80 0.256 0.074 1.0 1.0 See Table EMCB R12-4 for for Average Average and and Stud 0.987 0.975 1.0 1.0 Maximum Bolt Stresses 6" Feedring Handhole Shell/cover/flange 78.0 76.5 80 80 0.823 0.374 1.0 1.0 See Table EMCB R12-4R12-4 for for Average Average and and Stud Stud 0.823 0.84 1.0 1.0 Maximum Bolt Stresses Stresses 2" Inspection Port Shell/cover/flange 77.6 65.8 80 80 0.214 0.205 1.0 1.0 Stud See Table EMCB EMCB R12-4 R12-4 for for Average Average and and 0.864 0.807 1.0 1.0 Maximum Bolt Stresses Stresses I Secondary Manway Flange/Steam Drum Head 55.2 46.8 80 80 0.02 0.019 1.0 1.0 Diaphragm 60.4 60.4 69.9 69.9 0.02 0.015 1.0 1.0 Cover 25.5 21.6 80 80 80 80 0.02 0.000 0.000 1.0 1.0 Stud See Table EMCB R12-4R12-4 for for Average Average and and 0.973 0.752 0.752 1.0 1.0 Maximum BoltBolt Stresses Stresses

Braidwood/Byron Braidwood /Byron Stations Stations MUR LAR Response to to RAI RAI February February 20, 20, 2012 Attachment 1, Attachment 1, page page 16 NON-PROPRIETARY NON-PROPRIETARY Table EMCB R12-1:

Table R12-1: Stress Stress Intensity (51)

(SI) and and Fatigue Usage Factors (FUF)(FUF) for for Level AA &B B Conditions MUR MUR Orig. MUR Orig.

MUR MUR Orig. FUF Component I Location SI Range SIRange SI SI Range SI SI Limit SI SI Limit FUF FUF FUF Limit Limit.

(ksi)

(ksi) (ksi) (ksi) (ksi)

Pressure Boundary Attachments Seal Skirt Transition Juncture 44.6* 44.6* 56.1 56.1 0.538 0.476 1.0 1.0 Skirt Weld 41.2* 41.2* 56.1 56.1 0.74 0.559 1.0 1.0 Steam Drum Head/Steam Drum Juncture 48.6 48.6 80.1 80.1 0.401 0.209 1.0 1.0 I

Steam Drum / Trunion Juncture 64.8* 80 80.0 80.0 0.688 0.239 1.0 1.0 Primary Deck Lug/Steam Drum Juncture 72 61 80 80 80 0.608 0.546 1.0 1.0 Shroud Lug 40.6 34.4 58.5 58.5 0.652 0.545 1.0 1.0 Shroud Lug/ Shell Juncture 54.7 46.4 80 80 80.1 0.652 0.545 1.0 1.0 Upper Vessel Support!

Support/ Steam Drum I Juncture 70.6 59.8 80.1 80.1 0.021 0.010 1.0 1.0 Main Feedwater Nozzle Shell/nozzle juncture juncture 77.6 77.6 80 80 80 80 0.408 0.346 1.0 1.0 Nozzle 69.3 58.7 80 80 80 80 0.046 0.039 1.0 1.0 Transition ring/Thermal ringlThermal sleeve sleeve 27.2* 27.2* 69.9 69.9 69.9 0.985 0.945 1.0 1.0 Steam Outlet Nozzle Nozzle/Safe End Juncture 26.8 26.8 22.7 22.7 70 70 70 70 00 0 1.0 1.0 Nozzle 69.5 69.5 58.9 58.9 80 80 80 80 0.048 0.048 0.035 0.035 1.0 1.0 Steam Drum Head 71.3 71.3 60.4 60.4 80 80 80 80 0.049 0.049 0.033 0.033 1.0 1.0 Perforated Zone 76.7 76.7 65 65 80 80 80 80 0.080 0.080 0:059 0:059 1.0 1.0

Braidwood/Byron Braidwood/Byron Stations MUR LAR LAR Response Response to to RAI RAI February 20, 2012 February 20, 2012 Attachment 1, page 17 Attachment NON-PROPRIETARY NON-PROPRIETARY Table EMCB R12-1:

Table R12-1: Stress Stress Intensity (51)

(SI) and and Fatigue Usage Factors (FUF) (FUF) for A & B Conditions for Level A MUR MUR Orig. MUR Orig.

Component II Location MUR MUR Orig. FUF Component SIRange SI Range SIRange SI Range SI SI Limit Limit SI SI Limit FUF FUF FUF Limit (ksi) (ksi) (ksi) (ksi)

Small Nozzles 3" Blowdown Nozzle 3" 12.02 10.19 26.7 26.7 0.85 0.928 1.0 3" Recirculation Nozzle 12.02 15 26.7 26.7 0.5 0.5 0.938 1.0 3/" Nozzles

%" 13.96 11.83 26.7 26.7 0.81 0.679 1.0 Acoustic Sensor Pad Acoustic 54.63 46.3 56 56 56 56 0.81 0.81 0.777 1.0 1.0 Tubes Tubes 73.8 73.8 79.8 79.8 0.19 0.19 1.0 BoidAtaiicized stress range values were determined using

• Bold/Italicized using simplified simplified elastic-plastic elastic-plastic analysis analysis in in accordance accordance with with NB-3228.5.

NB-3228.5.

Braidwood/Byron Stations Stations MUR LAR Response to to RAI RAI February 20, 2012 1, page 18 Attachment 1, NON-PROPRIETARY Table EMCB R12-2 -- Primary Membrane and Bending Stresses for Level C C Conditions MUR Orig. MUR MURPL MUR PL Orig. PL MUR Orig. Orig.PL/

PmSI Pm Sl PmSI Pm SI PLlPm+

PL/Pm+ Pm+PbSI Pm+Pb SI Pm+PbSI Pm+Pb SI Component / Location Pm/PL SI Pm/PL SI Pm+PbSI Pm+Pb SI Limit Limit PbSI Pb SI Limit Limit (ksi) (ksi) (ksi)

(ksi) (ksi) (ksi) (ksi) (ksi)

Primary Head Primary/ Head Tubesheet

/ Tubesheet //

29.6 29.2 38.79 38.79 49.9 49.2 64.65 64.65 Secondary shell Primary Nozzle Bounded by design conditions Primary Manway Cover 13.33 13.33 38.8 38.8 24.39 24.39 58.2 58.2 Shell/flange 21.31 21.31 38.8 38.8 21.31 21.31 58.2 58.2 Primary Head Support Support Pad Pad Bounded by design conditions Primary Divider Plate Bounded by design conditions Small Nozzles Bounded by design conditions Steam Drum/Cone/Lower Shell Bounded by design conditions Assembly 8" Shell Cone Handhole 29.3 29.02 29.37 29.37 32.6 32.2 48.06 48.06 48.06 6" Feedring Handhole 29.3 29.02 29.37 29.37 34.6 34.6 48.06 48.06 48.06 48.06 2" Inspection Port 10.6 10.5 28 28 20.7 20.5 42 42 42 42 Secondary Manway Bounded by design conditions

Braidwood/Byron Braidwood/Byron Stations Stations MURMUR LAR LAR Response to RAI February February 20,2012 20, 2012 Attachment Attachment 1,1, page page 19 NON-PROPRIETARY NON-PROPRIETARY Table EMCB Table EMCB R12-2 R12-2 -- Primary Primary Membrane and Bending Stresses for Level C Conditions Conditions MUR Orig. MUR MURPL MUR PL Orig. PL MUR MUR Orig. Orig.PLI Orig.PL/

Component I/ Location Location PmSI Pm SI PmSI Pm SI PLlPm+

PL/Pm+ Pm+PbSI Pm+Pb SI Pm+Pb SI Component Pm/PLSI Pm/PL SI Pm/PLSI Pm/PL SI Pm+PbSI Pm+Pb SI Limit Limit PbSI Pb SI Limit Limit (ksi)

(ksi) (ksi)

(ksi) (ksi)

(ksi) (ksi) (ksi) (ksi) (ksi)

(ksi)

Pressure Boundary Attachments Pressure Seal Skirt Transition Juncture Seal Bounded by design conditions Skirt Weld Bounded by design conditions Steam Drum Head/Steam Drum Juncture Bounded by design conditions Steam Drum /I Trunion Juncture 28.8 28.5 39.4 39.4 36 36 35.6 65.7 65.7 Primary Deck Lug/Steam Drum Drum 29.8 29.8 43.8 43.8 65.2 65.2 65.7 65.7 Juncture Shroud Lug 2.32 2.3 2.3 26.37 26.37 5.8 5.8 5.73 43.95 43.95 Shroud Lug/ Shell Juncture 24.9 24.63 36.9 36.9 26.9 26.61 65.7 65.7 Upper Vessel Support/

Support! Steam Drum Drum Juncture Bounded by design conditions Main Feedwater Feedwater Nozzle Shell/nozzle juncture juncture 29 29 28.7 43.8 43.8 46.6 46.6 46.1 46.1 65.7 65.7 65.7 65.7 Nozzle 28.6 28.6 28.3 28.3 43.8 43.8 43.8 43.8 28.6 28.6 28.3 28.3 65.7 65.7 65.7 65.7 Transition ring/Thermal ringlThermal sleeve sleeve 9.5 9.5 9.4 9.4 28 28 28 28 26.1 26.1 25.8 25.8 41.9 41.9 41.9 41.9 Steam Outlet Outlet Nozzle Nozzle by design Bounded by design conditions conditions Small Nozzles Nozzles by design Bounded by design conditions conditions Tubes Tubes 22.95 22.95 22.7 22.7 35.2 35.2 35.2 35.2 32.35 32.35 32 32 52.9 52.9 52.9 52.9

, Tubes Tubes (external (external pressure) pressure) 0.168 0.168 0.166 0.166 1.424 1.424 1.424 1.424 -- -- -- --

Braidwood/Byron Braidwood/Byron Stations MURMUR LAR LAR Response to RAI February February 20, 20, 2012 Attachment Attachment 1, 1, page page 2020 NON-PROPRIETARY NON-PROPRIETARY Table EMCB Table EMCB R12-3:

R12-3: Primary Primary Membrane Membrane and and Bending Bending Stresses Stresses for for Level Level D D Conditions Conditions MUR Orig. MUR MURPL MUR PL Orig. PL Orig. PL MUR MUR Orig.

Orig. Orig.PU Component // Location Location PmSI Pm SI PmSI Pm SI PUPm+ Pm+Pb SI Pm+PbSI Pm+Pb SI Component PmlPLSI Pm/PL SI Pm/PL SI Pm/PL Pm+PbSI Pm+Pb SI (ksi) Limit Limit Limit PbSI Pb SI Limit Limit (ksi) (ksi)

(ksi) (ksi)

(ksi)

(ksi) (ksi) (ksi) (ksi) (ksi)

(ksi)

Primary Head Primary Head / Tubesheet Tubesheet /

29.6 29.1 56 56 68.7 67.8 84 84 84 84 Secondary Shell Primary Nozzle Primary nozzle Primary nozzle 51.51 51.51 56 56 76.27 76.27 84 84 84 84 Primary nozzle Primary safe end nozzle safe 27.7 27.7 48.9 48.9 39.93 39.93 72.36 72.36 Primary Manway Primary Manway Cover Cover 13.33 13.33 56 56 24.39 24.39 84 84 84 84 Shell/flange Shell/flange 21.31 21.31 56 56 21.31 21.31 84 84 84 84 Primary Head Support Support Pad Pad 15.9 15.9 56 56 53.9 53.9 84 84 84 84 Primary Divider Divider Plate Plate 35.9 35.4 52.5 52.5 61.2** 60.4 67.5 67.5 Small Nozzles 3

%"" Nozzles Nozzles 16.9 16.9 16.7 42.8 42.8 38 38 37.5 37.5 64 64 64 64 Steam Drum/Cone/Lower DrumlCone/Lower Shell Assembly 46.2 35.6 56 56 56 61.5 61.5 60.7 60.7 84 84 84 84 8" Shell Cone Handhole Handhole 40.9 40.9 40.4 40.4 56 56 56 56 40.9 40.9 40.4 40.4 84 84 84 84 6"

6" Feedring Feedring Handhole Handhole 35.3 35.3 34.8 34.8 56 56 56 56 35.3 35.3 34.8 34.8 84 84 84 84 2"

2" Inspection Inspection Port Port 55.1 55.1 54.4 54.4 56 56 56 56 60.9 60.9 60.1 60.1 84 84 84 84 Secondary Secondary Manway Manway 47.7 47.7 47.1 47.1 56 56 56 56 47.7 47.7 47.1 47.1 84 84 84 84

Braidwood/Byron Stations Stations MUR MUR LAR LAR Response Response totoRAI RAI February 20, 2012 Attachment 1, page Attachment 1, page 2121 NON-PROPRIETARY Table EMCB Table EMCB R12-3:

R12-3: Primary Primary Membrane Membrane and Bending Stresses for Level D D Conditions Conditions MUR Orig. MUR MURPL MUR PL Orig. PL MUR Orig. Orig.PU Orig.PLI PmSI Pm SI PmSI Pm SI PUPm+

PL/Pm+ Pm+PbSI Pm+Pb SI Pm+PbSI Pm+Pb SI Component I/ Location PmlPL Pm/PL SI Pm/PL SI Pm+PbSI Pm+Pb SI Limit Limit PbSI Pb SI Limit Limit (ksi) (ksi) (ksi)

(ksi) (ksi) (ksi) (ksi) (ksi)

Pressure Boundary Attachments Seal Skirt Transition Juncture 8.81 8.7 49 49 26.8 26.5 73.5 73.5 Steam Drum Head/Steam Drum 35.35 34.9 56 56 46.1 45.5 84 84 Juncture Steam Drum / Trunnion Juncture 35.8 35.3 56 56 42.7 42.2 84 84 Primary Deck Lug/Steam DrumDrum 40.1 40.1 56 56 74 74 84 84 Juncture Shroud Lug 39 38.5 49 49 43.4 42.9 73.5 73.5 Shroud Lug/ Shell Juncture 33.5 33.1 56 56 39.5 39.1 84 84 Support/ Steam Drum Upper Vessel Support!

28 27.6 56 56 63.9 63.1 84 84 Juncture Main Feedwater Nozzle Shell/nozzle juncture juncture 33.9 33.5 56 56 83.8 83.3 84 84 84 Nozzle 7.9 7.8 56 56 29.5 29.1 84 84 Transition ring/Thermal ringlThermal sleeve sleeve 12.9 12.7 49 49 53 52.3 73.5 73.5 73.5 Steam Outlet Nozzle Pipe extension 16.68 16.47 42 42 43.38 42.82 63 63 Nozzle/Safe End Juncture Juncture 14.99 14.8 49 49 40.84 40.32 73.5 73.5 Nozzle 25.43 25.1 56 56 54.82 54.12 84 84 Steam Drum Drum Head Head 26.34 26 56 56 56 55.84 55.84 55.12 55.12 84 84 84 84 Perforated Zone Zone 32.62 32.2 56 56 56 61 61 60.22 84 84 84 84

Braidwood/Byron Braidwood/Byron Stations MUR LAR LAR Response Response to to RAI RAI February 20,2012 20, 2012 Attachment 1, page 22 NON-PROPRIETARY NON-PROPRIETARY Table EMCB Table EMCBR12-3:

R12-3: Primary Membrane Membrane and Bending Stresses Stresses for for Level Level D D Conditions Conditions ,

MUR Orig. MUR MURPL MUR PL Orig. PL I MUR Orig. Orig.PU Orig.PL/

PmSI Pm SI PmSI Pm SI PUPm+ Pm+Pb SI Pm+PbSI Pm+Pb SI Component I/ Location Location Pm/PL SI PmlPL Pm/PL SI Pm+PbSI Pm+Pb SI Limit Limit PbSI Pb SI Limit Limit (ksi) (ksi) (ksi)

(ksi) (ksi) (ksi) (ksi) (ksi)

Small Nozzles 3" Blowdown Nozzle 3" 21.1 30.6 42.8 42.8 42.1 61.2 64 64 64 64 3" Recirculation Nozzle 3" 21.1 30.6 42.8 42.8 42.1 61.2 64 64 64 64 Acoustic Sensor Pad Bounded by Steam Drum/Cone/Lower Shell Assembly Assembly Tubes 31.4 31 56 56 56 68.68 67.8 84 84 84 84 Tubes (external pressure) 1.142 1.127 1.780 1.780 -- -- -- --

• • A prorating factor corresponding to to the SG SG secondary secondary side side level level DD loading loading has hasbeen beenapplied appliedtotothe theDivider DividerPlate PlateMURMURPUPm+Pb PUPm+PbSI, SI,making makingthethe reported value conservative. However, reported However,only onlyprimary primarystresses stressesfrom fromdivider dividerplate platelevel levelDDloads loadsneed needtotobe beanalyzed analyzedandandsince sincethe theprimary primaryside side pressures are invariant between MUR andand Original Original conditions, conditions, both both the the level level DDstresses stresses andand their theirASME ASMECodeCodelimits limitsare areunchanged.

unchanged.

Braidwood/Byron Braidwood/Byron Stations MUR LAR Response to to RAI RAI February 20, 2012 1, page 23 Attachment 1, NON-PROPRIETARY Table EMCB R12-4:

Table R12-4: Average AverageandandMaximum MaximumStresses Stresses for forStuds/Bolts Studs/Bolts MUR Orig. MURPL MUR PL Orig.

MUR Orig. MUR Orig.

Average Average Maximum Maximum Average Average Maximum Maximum Component I/ Location Stress Stress Stress Stress Stress Stress Stress Stress Limit Limit Limit Limit (ksi) (ksi) (ksi) (ksi)

(ksi) (ksi) (ksi) (ksi)

Primary Prima~ Manway Manwa3l A/B Level AlB 43.8 43.8 54.6 54.6 55.4 55.4 81.9 81.9 81.9 Level C 34.7 34.7 52.6 52.6 76 76 78.9 78.9 78.9 Level D 34.7 34.7 87.5 87.5 76 76 76 125 125 125 125 8" Shell Cone Handhole Level A/B AlB 13.5 13.25 57.7 57.7 49.7 48.7 77.9 77.9 Level C LevelC 41 40.54 57.7 57.7 78.6 77.73 86.7 86.7 Level D 41.1 40.54 86.2 86.2 79.7 78.66 125 125 125 6" Feedring Handhole Handhole A/B Level AlB 41.4 40.6 57.7 57.7 69 67.6 77.9 77.9 Level C LevelC 36.7 36.3 57.7 57.7 55.2 54.6 86.7 86.7 Level D 36.8 36.3 86.2 86.2 55.6 54.9 125 125 125 125 2" Inspection Ins~ection Ports Level A/B AlB 40.9 40.1 57.7 57.7 52.8 51.8 77.9 77.9 Level C 41.1 40.6 57.7 57.7 47.4 46.9 86.7 86.7 Level D 41 40.5 86.2 86.2 47.1 46.5 125 125 125 125

Braidwood/Byron Braidwood/Byron Stations Stations MUR LAR Response to to RAI RAI February February 20, 2012 Attachment 1, 1, page 24 NON-PROPRIETARY NON-PROPRIETARY Table EMCB Table EMCBR12-4:

R12-4: Average Average and and Maximum MaximumStresses Stresses for Studs/Bolts Studs/Bolts MUR Orig. MURPL MUR PL Orig.

MUR Orig. MUR Orig.

Orig.

Average Average Maximum Maximum Component I/ Location Average Average Maximum Maximum Component Stress Stress Stress Stress Stress Stress Stress Stress Limit Limit Limit Limit (ksi) (ksi) (ksi) (ksi)

(ksi) (ksi) (ksi) (ksi)

Secondary Manwa)l SecondarY Manway A/B Level AlB 47.8 40.5 57.7 57.7 72.1 61.1 77.9 77.9 Level C Levele 32.1 31.8 57.7 57.7 58.5 57.9 77.9 77.9

, Level D 30.4 30.0 86.2 86.2 44.7 44.1 125 125 125 125

---~

Braidwood/Byron Stations Braidwood/Byron Stations MUR MUR LARLAR Response to RAI RAI February February 20,2012 20, 2012 Attachment Attachment 1, 1, page 25 NON-PROPRIETARY NON-PROPRIETARY NRCJEMCB NRC/EMCB Request Request 13 Discuss further information Discuss information toto demonstrate demonstrate that, that, for the the expected post-up post-uprate rate conditions, conditions, the spent fuel pool spent pool (SFP)

(SFP) structure, structure, including SFP liner and the spent fuel racks, racks, remain capable capable of performing their intended design functions and will continue to be in compliance with with the the Byron Byron and Braidwood design basis code of record(s) and record(s) and acceptance criteria.

Response

During aa February During February 1, 1, 2012 2012 clarification clarification call call between between Exelon Exelon Generation Company (EGC) and the Nuclear Regulatory Commission (NRC) staff, EGC requested and the NRC staff staff agreed to allow EGC to provide a response to this request under a separate transmittal at a later date.

Braidwood/Byron Stations Braidwood/Byron Stations MUR LAR Response to RAI February February 20, 20, 2012 2012 Attachment Attachment 1, 1, page 26 26 NON-PROPRIETARY NON-PROPRIETARY NRC Balance NRC Balance of Plant (NRC/SBPB)

NRC/SBPB Request 1 NRC/SBPB Technical Specification Technical Specification (TS) (TS) 3.7.4 for the steam generatorgenerator (SG) (SG) power poweroperated operatedreliefreliefvalves valves (PORVs) currently (PORVs) currently allows 24 hours completion time to restore all but one restore all but one of of the the four four PORVs PORVs when two when two or more PORVs PORVs are inoperable.

inoperable. Hence, Hence, the the TSTS action action statement statement would allow all four PORVs to to be be inoperable inoperable for up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The analysis The analysis for a steam generator tube rupture (SGTR) (SG TR) credits credits the use of of two two PORVs PORVs to to cool cool down the down the reactor coolant system (RCS) (RCS) rapidly to achieve a sub subcooling cooling margin in order to start depressurizing the the RCS to stop the break flow. flow. TheThe analysis analysisidentifies identifiesthe the most mostlimiting limitingsingle single failure as a failure of of a SG PORV on an intact SG. Thus, Thus, thethe licensee licensee credits the SG SG PORVs with a high significance for successfully with successfully mitigating mitigating aa SGTR. The The current current TSTS that allows 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> for all four PORVs to be inoperable (loss of function) may not be appropriate.

Justify the current TS action statement that allows all four SG PORVs to to be inoperable based on the new SGTR analysis.

Response

Based on discussions during the the February February 1, 1, 2012 2012 clarification clarification callcall between betweenEGC EGCand andthetheNRC NRC staff, the NRC staff revised this request in an e-mail dated February 8, 2012 (Reference 4).

The NRC staff agreed to allow EGC EGC toto provide provide aa response response to to this this request under separate transmittal at a later date.

transmittal date.

NRC/SBPB Request Request 2 The licensee identifies the SG PORVs as being a key key component component in in mitigating mitigating anan SGTR from an overfill overfill condition.

condition. The licensee identifiedidentified anan SG PORV failing to to open open on one of of the intact SGs as the most limiting failure for the margin to overfill (MTO) analysis. analysis. TheThe installation installation ofof an uninterruptible power supply uninterruptible power supply was was made to reduce the currentcurrent vulnerability vulnerability of ofaa single single failure failure making twotwo SG PORVs inoperable.

In Table Table 1-2, "Steam Generator 1-2, "Steam Generator Tube Tube Rupture Equipment List," List," the licensee states, states, "Table "Table 1-2 1-2 identifies the systems, the systems, components, and instrumentation which are credited and instrumentation which are credited for accident for accident mitigation."

mitigation." The TheTable Table1-2 1-2 does does notnotlist listthe the SG SG PORV PORV controllers.

controllers.

Provide aa description description of ofthe the PORVs PORVs electrical electricalsystems systemstotoinclude includepower powersupplies suppliestotothe the controllers and circuitry, and include any other circuits that would controllers and circuitry, and include any other circuits that would affect the SG PORV's affect the PORV's ability ability to to perform perform itsits function; function; identify identifyany anyshared sharedcomponents components(i.e., (i.e.,electrical, electrical,mechanical, mechanical, Instrumentation Instrumentation && Control, Control, etc.);

etc.); and andjustify justifynot notincluding includingthe the SG PORV controllers.

SG PORV controllers.

Response

Response As As described described in in Technical Technical Specifications Specifications BasesBases 3.7.4, 3.7.4, aa Steam Steam Generator Generator(SG) (SG) Power PowerOperated Operated Relief Relief Valve (PORV)

(PORV) is is considered considered OPERABLE OPERABLE when when itit isis capable capable of of providing providing controlled controlled relief relief of the main steam flow and capable of of the main steam flow and capable of fully opening fully opening and closing closing on demand. The The definitionof definition of OPERABLE OPERABLE requires requires thatthat all all necessary necessaryattendant attendantinstrumentation, instrumentation,controls, controls,normal normaloror emergency emergencyelectrical electrical power powerrequired requiredto toperform performits itsspecified specified safety safetyfunction function are arealso alsocapable capableof of performing performing their related support functions. functions. As such, the SG SG PORVs PORVswere werelisted listedas asananassembly assembly

Braidwood/Byron Braidwood/Byron Stations Stations MUR Response to MUR LAR Response to RAI RAI February February 20,2012 20, 2012 Attachment 1, page 27 Attachment NON-PROPRIETARY rather than listing listing individual individual components required to support support the performance performanceof oftheir theirsafety safety function.

The SG PORVs do not share mechanical mechanical components.

components. The SG PORVs on on a single electrical division division share their normal source source of of 480 VAC from from anan Engineered Safety Feature (ESF) (ESF) switchgear on on that that division.

division. On Unit Unit 1, 1, for example, Division Division 11 Motor Motor Control Centers (MCCs) are supplied supplied fromfrom ESF Switchgear 131X 131X and and Division Division 22 MCCs MCCsare aresupplied supplied from ESF Switchgear 132X. 132X. TheThe SGSG PORVs PORVs on on a single division division share a common common process process control control cabinet. On Unit 1, 1, for example, SG PORVs PORVs 1A 1A and and 110 receive process D receive process control control signals signals from from cabinet 1PA33JPA33J andand SG PORVs 1Band B and 1CC receive receive process control signals from from cabinets cabinets 1PA34J.

1 PA34J.

The existing existing SG PORVs are fed fed from from safety safety related related 480V MCCs which which feed a powerpower transformer in the 1/2MS018JA, JB, JC, and and JDJD SG SG PORV PORV control (controllers). The SG control panels (controllers).

PORV controllers contain a 4KV 4KVA A power transformer that reduces the 480VAC supply to a 125VAC control control power source source and a secondary secondary AC power supply source that is subsequently subsequently rectified to a DC source and used used to to power power the the reversible reversible hydraulic hydraulic pump pump motor motorcontained contained on on the PORV operator. The SG PORV controllers controllers receive aa control control signal signal from from the the pressure pressure control control loops loops associated with the steam steam line line pressure pressurecontrols controls from from control control cabinets cabinets 1/2PA33J and 1/2PA34J.

1/2PA34J. This signal signal isis generated generated based on pressure pressure control or demands from a Manual Manual Auto (MA) Station Station onon the the Main Main Control Control Board.

Board. The Theoutput outputsignal signalfrom fromthethecontrollers controllersdrives drivesthe the hydraulic hydraulic pumppump motor to to either open or close the valve.

Once installed installed thethe modified modified SG PORVs will will incorporate a battery-backup Uninterruptible Uninterruptible PowerPower Supply (UPS) into the power feed to one of of two two SG PORV circuits per electrical division division (SG PORV's 1/20 Division 1 1/2D for Division and 11/2C I and /2C for Division 2). The UPS will provide provide battery back-up power to the valve in the event of a loss loss of the UPS UPS normal normal AC power supply.

The SG PORV UPS UPS modification modification also also affects the power power source source toto the Division 1 SG PORV PORV process process control control cabinets.

cabinets. A A loss loss of power to a safety safety related related Division Division 11 would would also also result result inin the loss loss of power to the 1/2PA33J 1/2PA33J control control cabinet cabinet since since the the cabinets cabinets are are fed fed from from two two separate 120VAC distribution distributionpanels panelsfrom fromDivision Division1I MCCs. MCCs. TheThepressure pressuremodulating modulating signals signals forfor SG SG PORVs 1/2A and 1/20 1/2D areare processed processed in 1/2PA33J. To resolve this issue, resolve this issue, the 120VAC 120VAC distribution distribution panel feed to to the 1/2PA33J 1/2PA33J cabinet is is replaced replaced with with aa feed feed from from aa Division Division 1I inverter inverter backed instrument instrument bus. The 1 bus. The Band 1B and 1C 1C SG PORV control circuits circuits are are processed in in the the 1/2PA34J panels which are currently currently fed from aa DivisionDivision 2 120VAC 120VAC distribution distribution panel and aa Division Division 2 inverter backed instrument bus and thus would be unaffected unaffected by by aa Division Division 22busbus outage/failure.

Tables SBPB SBPB R2-1 R2-1 and and R2-2 R2-2 below belowlist list the the power supplies for the SG PORV controllers and power supplies and associated control cabinets.

associated control cabinets.The The informationprovided information providedreflects reflectsthe theconfiguration configurationfollowing followingthe the implementation of the UPS modification.

modification. All All power supplies are Electrical supplies are Electrical Class Class11E. E.

Braidwood/Byron Stations MUR LAR Response Braidwood/Byron Response to to RAI RAI February February 20,2012 20, 2012 Attachment 1, page Attachment page 28 28 NON-PROPRIETARY NON*PROPRIETARY Table SPBP R2*1:

Table R2 -1: SG PORV SG PORV Power Supplies Control Panel Control Primary Power Primary SG PORV SG Backup Power Supply Backup (Controllers)

(Controllers) Supply Supply Byron Station Byron l MS018A 1MS018A 1 MS018JA 1MS018JA 131 X28 131X2B 1 MS018B 1MS018B 1 MS018JB 1MS018JB 132X1 1 MS018C 1MS018C 1 MS018JC 1MS018JC UPS from 132X5 UPS UPS from Battery UPS 1 MS018D 1MS018D 1 MS018JD 1MS018JD UPS from 131X4 UPS UPS from UPS from Battery 2MS018A 2MS018JA 231X28 231X2B 2MS018B 2MS018JB 232X1 2MS018C 2MS018JC UPS from 232X5 UPS UPS from Battery UPS 2MS018D 2MS018JD UPS from 231X4 UPS UPS from UPS from Battery Braidwood Station I MS018A 1MS018A l MS018JA 1MS018JA 131 X28 131X2B 1 MS018B 1MS018B 1 MS018JB 1MS018JB 132X1 1 MS018C 1MS018C 1 MS018JC 1MS018JC UPS from 132X5 UPS from Battery lMS018D 1MS018D lMS018JD 1MS018JD UPS from 131X4 UPS UPS from Battery 2MS018A 2MS018JA 231X28 231X2B 2MS018B 2MS018JB 232X1 2MS018C 2MS018JC UPS from 232X5 UPS from Battery 2MS018D 2MS018JD UPS from 231X4 UPS from Battery Table SBPB R2 -2:

R2*2: Control Cabinet Cabinet Power Supplies Backup Power Control Primary Power Supply Supply (120 VAC Supply (120 VAC Cabinet Distribution Panel)

Byron Station 11PA33J PA33J Instrument Bus 113113 131X1 131X1 11PA34J PA34J Instrument Bus Bus 114 114 132X1 132X1 2PA33J Instrument Instrument Bus Bus 213 213 231X1 231X1 2PA34J Instrument Instrument Bus Bus 214 214 232X1 Braidwood Station Station 11PA33J PA33J Instrument Instrument Bus Bus 113 113 131X1 131X1 11PA34J PA34J Instrument Instrument Bus Bus 114 114 132X1 132X1 2PA33J 2PA33J Instrument Instrument Bus Bus 213 213 231X1 2PA34J 2PA34J Instrument Bus 214 Instrument Bus 214 232X1

Braidwood/Byron Stations Braidwood/Byron Stations MUR MUR LAR LAR Response to RAI RAI February 20, 2012 February 20, 2012 Attachment 1, Attachment 1, page page 29 NON-PROPRIETARY NON-PROPRIETARY NRC/SBPB Request 3 NRC/SSPS The licensee is The is making making modifications modifications to to the the auxiliary feedwater (AFW) (AFW) flowflow control valves to include an air accumulator tank capable include capable of supplying supplying air for 30 air for 30 minutes.

minutes. In accordance accordance with with their analysis, their analysis, AFW flow control is required longer longer than than 30 30 minutes minutes to to mitigate mitigate the the SGTR SGTR and and for for RCS cool RCS cool down.down. In Attachment 5a, Section 1I.2.E, Section 11.2.E,Single SingleFailure Failure Considerations, Considerations, the licensee licensee states:

In addition, In addition, since since thethe failure failure of an an intact SG PORV scenario scenario assumes a loss of offsite power with offsite power with an an associated loss of Instrument Air (IA), (fA), the modification described in described in Section Section11.2.F, 1I.2.F, Item 1, assures that AFWflow AFW flow control is is maintained throughout the event.

According to the licensee's evaluation, an SGTR event continues until break flow is terminated 3458/3258 seconds (Units at 345813258 (Units 1 I and and 2).2).

Describe the basis for selecting 30 minutes, minutes, and explain how the amount of of air that is required is determined and the amount of of air available to support this function.

Response

As noted in the NRC's request above, the limiting Steam Generator Tube Rupture (SGTR) event continues until break flow is terminated (3,458 (3,4S8 seconds Unit 1 and 3,258 3,2S8 seconds for Unit Unit Attachment 5a 2). Attachment Sa toto the the MUR MUR power poweruprate uprateLAR LAR(Reference (Reference1), 1),Section Section11.2.F, II.2.F, "Modifications "Modificationstoto Support MTO Single Failure Failure Considerations,"

Considerations," describesdescribes the the plant plantmodifications modificationsByron Byronand and Braidwood Stations will be implementing to support the Steam Generator Margin to support the Steam Generator Margin to Overfill to Overfill assumptions. Included Reanalysis assumptions. Included in in these these modifications modifications will will be bethe the installation installationof oftwo two instrument air accumulator tanks on each each Unit Unit (one (one perper train) train) toto provide provide aasafety safetyrelated relatedair air supply for the Auxiliary Feedwater (AFW) Flow Control Valves (FCVs)(AF005). (FCVs)(AFOOS). The Theair air accumulator tanks for the AFW AFW FCVs FCVs (AF005)

(AFOOS) are are only only required required for forthethefirst first30 30minutes minutes(1,800(1,800 seconds) post-SGTR event initiation for AFW flow control and isolation. After AfterAFW AFWflow flowisis isolated to the ruptured SG, AFW AFW flow flow control control toto the the ruptured ruptured SG SG isisno nolonger longerneeded neededfor forthe the duration of the event. AFW AFWflow flowto tothe thenon-ruptured non-ruptured SGs SGsisiscontrolled controlled by bythrottling throttlingeither eitherthe the AFW FCVs (AF005)(AFOOS) or the motor operated AFW AFW valves (AF013); these valves are are in in series series with each other.

The Emergency Emergency Operating Operating Procedures Procedures(EOPs) (EOPs)(1/2B(w)EP-0, (1/2B(w)EP-0,Reactor ReactorTrip TripororSafety SafetyInjection Injection Unit 11(2)) direct isolation (2)) direct isolation of AFW AFWto to the the ruptured SG with with the the motor motor operated operatedAFW AFWisolation isolation valves (AF013).

(AF013). Following Following the the installation installation of of the air air accumulator accumulator tanks, tanks, thetheEOPs EOPswill will beberevised revised to direct the the closure closureof ofthe theAFW AFW FCVsFCVs (AF005)

(AFOOS) via via the the controller controller in in the the Main Main Control Control RoomRoom at at the same samepoint point inin the the procedure procedure that that they they are aredirected directedto toclose closethe theAFW AFW (AF013)

(AF013) valve. IfIf an an AF0013 valve fails fails toto close, close, then the EOPs EOPs will direct an operator to be will direct be dispatched dispatchedto toclose closethe the associated AF005 flow associated AFOOS flow control control valve locally. This action prevents the valve from locally. This action prevents the valve from failing open failing open when thethe air air supply supply fromfrom the accumulator tank is exhausted. ItItwas wasdetermined determinedthat thataa30 30minute minute supply ofof air is sufficient sufficient to to allow allowthe the operator operator to to reach the AF005AFOOS valvevalve andand manually manually close closeitit using the installed installed handwheel handwheel on on the thevalve.

valve.

The time assumed assumed for for the the local local closure closure of ofthe theAF005 AFOOSvalvesvalvesisisconsistent consistentwith withthe thecurrent currentByron Byron and Braidwood Braidwood design basis. basis. Specifically, Specifically,ininUFSAR UFSARSections Sections3.11.10, 3.11.10,"High"HighEnergy EnergyLine LineBreak Break (HELB)," 10.4.9.3, 10.4.9.3, "Auxiliary "Auxiliary Feedwater, Feedwater, Safety Safety Evaluation,"

Evaluation,"and and15.2.8.2, 1S.2.8.2,"Feedwater "FeedwaterSystem System Pipe Break, Analysis Analysis of of Effects Effects andand Consequences,"

Consequences,"for forfeedline feedlineand andmainmainsteamline steamlinebreaks,breaks,

Braidwood/Byron Stations Braidwood/Byron Stations MUR MUR LARLAR Response Response to to RAI RAI February 20, February 20, 2012 2012 Attachment 1, page 30 Attachment NON-PROPRIETARY NON-PROPRIETARY operator action operator action is is credited credited to to isolate auxiliary feedwater to the faulted steam generator within 20 minutes.

minutes.

instrument air The AF005 instrument air accumulators accumulators were were sized to include 30 minutes minutes of of air air supply supply asas described above and additional described additional capacity capacity to to account account for:

for:

• • Stroking four Stroking four valves (1 (1 Train) fromfrom full full open to full closed,

• • Maximum air consumption Maximum consumption rate for four electric electric toto pneumatic signal signal converters converters (IY's),

• Maximum air consumption consumption rate for four valve positioners, and

• 10% allowance for leakage.

The total required was determined to be 27.

total volume required 27.33 cubic cubic feet feet (204 (204 gallons).

gallons). Additional Additional conservatism exists conservatism existssince sincethe thetank tanksize sizeisis33.4 33.4cubic cubicfeet feet(250 (250 gallons).

gallons). This ensures that adequate air air is is available available to to support support the the required required function function of AFW flow control control and isolation.

NRC/SBPB Request 4 Figure 11-5 ofAttachment 1/-5 of Attachment 5a 5a shows the the SG water volumevolume on on Unit 1 I trending trending towards towards thethe maximum available maximum available quantity.

quantity. At At approximately approximately 3200 seconds, the trend tapers off, resulting in in aa margin to overfill of of approximately approximately 94 94 cubic cubic feet.

feet. At At the same time other graphs show a sharp sharp pressure, which logically corresponds to a second opening of reduction in SG pressure, of the SG PORVs on the the intact SGs.

SGs. This action stops the upward This action stops the upward trend and trend and prevents the the overfill condition.

condition. TheThe licensee does not identify a critical operator action to open the SG PORVs a second time within a certain time period as a condition to prevent prevent an overfill overfill of ofthe the SG.

In the updated final safety analysis report, report, Section 15.6.3.2, 15.6.3.2, under under the section describing describing major major operator operator actions, the licensee's analysis credits operators for reopening pressurizer pressurizer PORV, four minutes after establishing normal charging and letdown, in order order to equalize the RCS and SG pressures.

In Attachment Attachment 5a (page 11-10), 11-10),the the licensee licensee states that the SG PORVs on the intact SGs automatically automatically open, open, as necessary, to maintain RCS sub subcooling margin. The cooling margin. The above above mentioned mentioned graph trend shows aa sharp pressure reduction at at 3200 3200 seconds, which is not indicative of of SG PORV automatically controlling contrOlling pressure at a prescribed setpoint setpoint.

a. Evaluate whether whether this this operator operator action action isis credited creditedto tobe beperformed performedwithinwithinaaspecific specifictime timeinin order order to prevent prevent an an overfill overfill condition.

condition.

b. If If operator operator action is is required, required, identify identify thethe action action as a critical operator operator action.

action.

c. Describe whether whether the new analysis changes the the existing existing UFSAR UFSAR analysis, analysis, and and results results in in the major major operator action opening a SG SG PORV PORV rather rather than than aa pressurizer pressurizer PORVPORV afterafter SI termination to to stop stop anan overfill overfill condition condition from from occurring.

occurring.

Response

The Steam Steam Generator Generator Tube Tube Rupture/Margin Rupture/Margin to to Overfill Overfill (SGTR/MTO)

(SGTR/MTO) analysis analysis methodology methodology used used in the new SGTR/MTO Analysis submitted submitted in in Attachment Attachment 5a 5a toto the the MUR MUR powerpower uprate uprate LAR LAR (Reference 1) 1) isis different different from from the the methodology methodologyininthe thecurrent currentAnalysis AnalYSisofofRecord Record(AOR)

(AOR) described in in the UFSAR UFSAR SectionSection 15.6.3, 15.6.3, "Steam "Steam Generator Generator Tube Tube Rupture."

Rupture." The Themethodology methodology used in the current AOR SGTR/MTO analysis explicitly explicitly models models operator operator actions actions after after Safety Safety Injection (SI)

(SI) flow termination (i.e. (i.e. securing securing Emergency EmergencyCore CoreCooling Cooling(ECCS)

(ECCS)flow),

flow),including including the the operator operator action action to to open open thethe pressurizer pressurizer PORV PORV within within aa specific specific time time inin order order toto prevent prevent anan overfill overfill condition.

condition. The TheSGTR/MTO SGTR/MTOanalysis analysisprovided providedininAttachment Attachment5a, 5a,"Steam "SteamGenerator GeneratorTube Tube

Braidwood/Byron StationsStations MUR MUR LAR LAR Response ResponsetotoRAI RAI February 20, 2012 20,2012 Attachment Attachment 1, 1, page page31 31 NON-PROPRIETARY NON-PROPRIETARY Rupture Analysis report," of the MUR power uprate submittal (Reference 1) 1) uses uses the NRCNRC approved methodology methodology described described inin WCAP-1 0698-P-A, "SGTR Analysis WCAP-10698-P-A, Analysis Methodology Methodologyto to Determine the Margin to to Steam Steam Generator GeneratorOverfill" Overfill"(Reference (Reference5). 5).

Consistent with the WCAP-1 WCAP-1 0698-P-A methodology, specific operator operator actions actions after after SI SI termination are not used and the LOFTTR2 LOFTTR2 computer computer code code isis used used to to predict predict the the transient transient responses that lead to pressure equalization (break flow termination) and to demonstrate the SG overfill condition is not reached. Therefore, Therefore,actions actionstaken takenafter afterSISItermination terminationare arenot not considered critical operator responses and as such such are modeled modeled to occur occur asas conditions conditions require require as predicted byby the the LOFTTR2 LOFTTR2 computer computercode. code.

As discussed discussed in in Section Section11.2.D, "Operator Action Times," of 11.2.0, "Operator of Attachment Attachment 5a 5a of ofthe theMUR MURpowerpower uprate submittal (Reference (Reference 1), 1), the the critical critical operator operatorresponses responsesare: are:

1. Isolate Auxiliary Feedwater

1. Feedwater (AFW) (AFW) flowflow toto the the ruptured ruptured Steam SteamGenerator Generator(SG),(SG),

2. Isolate the MSIV on the ruptured ruptured SG, SG,

3. Initiate RCS cooldown, to to initiate initiate RCS RCS depressurization, depressurization, and and

4. Terminate Safety Injection Injection (SI)(SI) (secure (secure Emergency EmergencyCore CoreCoolant Coolant(ECCS)

(ECCS)flow).

flow).

These These operator operator actions actions and and the the corresponding corresponding operator operator action action times times used used for for the analyses analysesare are summarized summarized in in Table Table11-2, "Operator Action 11-2, "Operator ActionTimesTimes forforDesign Design Basis Basis SGTR Analyses" of Attachment a 5a of thethe MUR MURpower poweruprate upratesubmittal submittal(Reference (Reference1). 1).These These actions actions are consistent with with the actions actions in in WCAP-1 0698-P-A (Reference WCAP-10698-P-A (Reference 5) Table 2.3-2, 2.3-2, "Operator "OperatorAction Action Times Times for for Design Basis Basis SGTR SGTR Analysis."

Analysis." Also,Also,consistent consistentwith withthe themethodology methodologyininWCAP-1 0698-P-A WCAP-10698-P-A (Reference 5) the times required requJred for cooldown, cooldown, depressurization, depressurization, and pressurepressure equalization equalization are are calculated using the LOFTTR2LOFTTR2 program.program. The The analyses analysesdo donot notmodel modelspecific specific operator operatoraction action times after after SI SI termination.

termination.

In accordance with Emergency Operating Operating Procedures Procedures (EOPs)(EOPs) (1/2B(w)EP-3),

(1/2B(w)EP-3), the thesame samestep step that directs the operator to terminate RCS cooldown also directs the operators to maintain maintain RCS RCS temperature below the required temperature. This Thisstepstepoccurs occursbefore beforeSI SItermination terminationand andisisaa step that is monitored and acted on throughout the procedure. SI SI termination terminationoccurs occursat at2,311 2,311 seconds on Unit 1 and at 2,482 seconds on Unit Unit 2. After AfterSISItermination, termination,LOFTTR2 LOFTTR2models modelsthe the opening of two of the intact SG PORVs to maintain the required RCS temperature from the EOPs. This EOPs. This action action is is predicted predicted by LOFTTR2LOFTTR2 to occur at approximately 3,200 seconds seconds (Unit 1 I analysis). This analysis). This modeling modeling isis consistent consistent with with the the methodology methodologyinin WCAP-10698-P-A.

WCAP-10698-P-A.

NRC/SBPB Request Request 5 Calculation Westinghouse commercial commercial atomic atomic power (WCAP) -10698-P-A provides a general power(WCAP) assessment of the the MTOMTOfor forWestinghouse Westinghousetype typereactors.

reactors. There There were were instances instances where the licensee deviated deviated from from the input parameters parameters selected in WCAP-10698-P-A as the most conservative.

a. Decay heat is one of of the input input factors factors that that influence influence MTOMTO analyses and Thermal/Hydraulic during aa tube analyses during tube rupture.

rupture. For For the MTO analysis, the licensee states that plant the MTO plant specific sensitivities were performedperformed for for Bryon Bryon andand Braidwood BraidwoodUnits Units11and and2.2. These studies concluded that the 1979-2a American Nuclear Society SOCiety (ANS) decay heat factor was more conservative compared to the 1971 1971 +20%

+20% ANS ANS decay decay heat heatmodel modelspecified specifiedininWCAP-WCAP-10698-P-A.

Justify use of of the 1979-2a 1979-2Q ANS ANS decay heat factor was more conservative compared to the 1971 +20%

1971 +20% ANS decay heat heat factor.

factor.

Braidwood/Byron Stations MUR LAR LAR Response Response to to RAI RAI February 20, 2012 20,2012 Attachment 1, page 32 NON-PROPRIETARY

b. Similar to above, inin determining the most conservative input values, values, the licensee chose to the minimum model the minimum AFW AFWenthalpy enthalpy of of 0.03 Btu/lbm; WCAP-10698-P-A models the Btu/Ibm; whereas, WCAP-10698-P-A maximum temperature of AFW AFW (maximum enthalpy) as the most conservative parameter parameter in in the the analysis for MTO.

Justify how the use of the minimum AFW enthalpy is more conservative compared to using the maximum temperature (enthalpy) for the for AFW.

Response

WCAP-1 0698-P-A (Reference WCAP-10698-P-A (Reference 5)5) identified identified high high decay heat and high Auxiliary Feedwater Feedwater (AFW)

(AFW) temperature to be the conservative assumptions for the steam generator tube rupture margin to overfill (MTO) analysis.

overfill analysis. NSAL-07-1 NSAL-07-11, 1, "Decay Heat Assumption in Steam Generator Tube Margin-to-Overfill Analysis Rupture Margin-to-Overfill Analysis Methodology" (Reference 6), identifiedidentified a lower lower decay decay heat heat can be more limiting for some plants. To resolve the concerns of NSAL-07-1 To resolve the concerns of NSAL-07-11, 1, plant-specific Byron and Braidwood Units 1 and 2 to justify the decay heat sensitivities were performed for Byron heat model and model and AFWenthalpy AFW enthalpy assumed in the analysis. The TheTables TablesSBPB SBPBR5-1 R5-1 and and22show showthethe resulting from the sensitivity study. The impact on the Margin to Overfill (MTO) resulting The study studycovered covered Tav9 the T rangeand avg range andthethesteam steamgenerator generatortube tube plugging plugging levels levels supported supported by the analysis provided 5a. The in Attachment 5a. The impact impacton on MTO MTO provided provided isisrelative relativeto to the thelimiting limitingcase case modeling modelingthethe ANS 1979 - 2a decay heat model, low AFW enthalpy, 20 decay heat model, low AFW enthalpy, low T avg,low Tavg, and high steam generator generator tube plugging level.

The results show that use of the ANS 1971 + + 20% decay heatheat model model (cases (cases 11 to to 4) clearly clearly provides more MTO margin than the ANS ANS 1979 - 2a 20 decay heat heat model (cases 5 to 8). The model (cases 5 to 8). The conservative direction for AFW enthalpy is studied using low decay heat. heat. Comparing Comparingcases cases55toto 8 with corresponding cases 99 to to 12 12 show show that that minimum minimum AFW AFWenthalpy enthalpyisisconservative.

conservative.

Braidwood/Byron Stations MUR LAR LAR Response Response to to RAI RAI 20, 2012 February 20,2012 Attachment 1, page 33 NON-PROPRIETARY Table SBPB R5-1: Byron/Braidwood Unit 1I Byron/Braidwood Results of Sensitivity Study on MTO Results Impact on MTO*

Case Description (ft3)

(ft) 1 Low Tang, Low T avg, 5% tube plugging, plugging, ANS 1971 + 20%,

1 +321 maximum AFW enthalpy 2 Low Tavg, Low T avg, 0% tube plugging, plugging, ANS 1971 + 20%,

2 +314 maximum AFW enthalpy 3 High Tang, High T avg, 5% tube plugging, plugging, ANS 1971 + 20%,

3 +458 maximum AFW enthalpy 4 High Tavg, High T avg, 0% tube tube plugging, plugging, ANS 1971 + + 20%,

4 +457 maximum AFW enthalpy 5 Low Tavg, Low T avg, 5% tube plugging, plugging, ANS 1979 - 26,20',

5 +47 maximum AFW enthalpy 6 Low Tang, Low T avg, 0% tube tube plugging, plugging, ANS 1979 - 2a,20',

6 AFW enthalpy

+52 maximum AFWenthalpy 7 High Tavg, 5% tube T avg , 5% tube plugging, ANS 1979 1979 -- 2Q, 20',

7 +178 maximum AFW enthalpy 8 High TTa^,

avg, 0% tube plugging, ANS 1979 1979 -- 2a, 20',

8 +176 maximum AFW enthalpy Low Tang, T avg, 5% 5%tube tubeplugging, plugging, ANS ANS 1979 - 2a, 20',

9 Limiting case minimum AFW enthalpy Low Tavg, T avg, 0%0% tube plugging, ANS 1979 1979 -- 2Q, 20',

10 +3 minimum AFW enthalpy High Tavg, High T avg, 5% tube tube plugging, plugging, ANS 1979 1979 -- 2Q, 20',

11 +125 minimum AFW enthalpy AFWenthalpy High Tang, 0%tube T avg , 0% tube plugging, plugging, ANS 1979 1979 -- 2Q, 20',

12 +123 minimum AFWenthalpy minimum AFWenthalpy

• • ++ indicates increase in MTO from the Limiting Case. Case.

Braidwood/Byron Stations MUR LAR Response Braidwood/Byron Response to to RAI RAI February 20,2012 February 20, 2012 Attachment Attachment 1, page 34 NON-PROPRIETARY NON-PROPRIETARY Table SBPB R5 Table R5-2:-2: Byron/Braidwood Unit 2 Byron/Braidwood Results of Results of Sensitivity Sensitivity Study Study onon MTO IMO Case Impact on Impact on MTO*

Case Description Description (fe)

(ft) 1 Low Tavg, Low 10% tube plugging, T avg , 10% ANS 1971 plugging, ANS 1971 + 20%,

1 maximum AFW AFW enthalpy

+337 maximum 2 Low Tavg, T avg, 0%0% tube plugging, ANS 1971 + 20%,

2 +353 maximum AFW enthalpy 3 High TTavg, 10%tube avg , 10% tubeplugging, ANS 1971 ++ 20%,

plugging, ANS 3 +440 maximum AFW enthalpy 4 High Tavg, T avg, 0%0% tube tube plugging, ANS 1971 + 20%,

1971 +

4 +472 maximum AFW enthalpy 5 Low TTang, 10%tube avg , 10% tubeplugging, plugging, ANS ANS 1979 - 2a,20, 5 +67 maximum AFW enthalpy 6 Low TTavg, avg, 0% 0%tube tubeplugging, plugging, ANS ANS 1979 - 2a,20, 6 +102 maximum AFW enthalpy 7 High TTang, avg, 10%10%tube tubeplugging, plugging, ANS ANS 1979 - 2a, 20, 7 +176 maximum AFW enthalpy High Tavg, T avg, 0%0%tube tube plugging, plugging, ANS 1979 1979--2Q, 20, 8 +212 maximum AFW enthalpy Low Tavg, 10% tube T avg, 10% tube plugging, plugging, ANS ANS 1979 1979--2Q, 20, 9 Limiting case minimum AFW enthalpy Low Tavg, Low T avg, 0% tube plugging, plugging, ANS 1979 - 2a,20, 10 +23 minimum AFW enthalpy High Tavg, 10%tube T avg , 10% tube plugging, plugging, ANS 1979 1979 -- 2a, 20, 11 +129 minimum AFW enthalpy High Tavg, T avg, 0%0%tube tube plugging, plugging, ANS 1979 1979 -- 2a, 20, 12 +159 minimum AFW enthalpy

• • ++ indicates increase increase in MTO MTD from the Limiting Limiting Case.

Case.

Braidwood/Byron Stations MUR LAR LAR Response Response to to RAI RAI February 20, 2012 Attachment 1, 1, page page 3535 NON-PROPRIETARY REFERENCES 1

1 Letter from from Craig Generation Company, Craig Lambert (Exelon Generation Company, LLC) to U. u. S. NRC, NRC, "Request for for License License Amendment Regarding Measurement Uncertainty Uncertainty Recapture PowerPower Uprate,"

Uprate,"

dated June 23, 2011 2011 2 Letter from from Kevin Borton (Exelon Kevin F. Borton (Exelon Generation Generation Company, LLC) to U. u. S. NRC, NRC, "Additional "Additional Information Information Supporting Supporting Request Request for for License License Amendment Amendment Regarding Measurement Measurement Uncertainty Uncertainty Recapture Power Uprate," dated December December 9, 9,2011 2011 3 Letter from N. J. DiFrancesco (U. (U. S.

S. NRC)

NRC) toto M.

M. J.

J. Pacilio Pacilio (Exelon (Exelon Generation GenerationCompany, Company, LLC),

LLC), "Braidwood "Braidwood Station, Station, Units Units 11 and and 22 and and Byron Byron Station, Station, Unit Unit Nos.

Nos. 1I and 2 - Request Request RE: Measurement for Additional Information RE: Measurement Uncertainty Uncertainty Power PowerUprate UprateRequest Request(TAC (TACNOS.

NOS.

ME6587, ME6587, ME6588, ME6588, 6589, 6589, AND AND ME659D),"

ME6590)," dated dated November November 28, 28, 2011 2011 4 E-mail from Brenda Mozafari (U. (U. S. NRC)

NRC)toLeslieHolden,et.al., Exelon Generation to Leslie Holden, et. al., Exelon Company), "FW: Draft Balance of Plant RAls related to MUR dated June 23, 2011," dated February 8,8,2012 2012 5 WCAP-10698-P-A, WCAP-1 0698-P-A, "SGTR Analysis Methodology to to Determine the Margin to Steam Steam Generator Overfill," August August 1987 6 NSAL-07-11, Heat Assumption NSAL 1 1, "Decay Heat Assumption in in Steam Generator Tube Tube Rupture Rupture Margin-to-Margin-to-AnalysiS Methodology,"

Overfill Analysis Methodology," November November 2007.