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

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| issue date = 02/20/2012
| issue date = 02/20/2012
| title = Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate
| title = Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate
| author name = Borton K F
| author name = Borton K
| author affiliation = Exelon Generation Co, LLC
| author affiliation = Exelon Generation Co, LLC
| addressee name =  
| addressee name =  
Line 18: Line 18:


=Text=
=Text=
{{#Wiki_filter:Exeion Generation Company, L L Cwww.exeloncorp.com 4300 Winfield Road Warrenv ille, I L b0555 RS-12-033 February 20, 2012 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF
{{#Wiki_filter:Exeion Exeion Generation
-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos.
            .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.
NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
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:==
==Subject:==
Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate
Additional Information Additional    InformationSupporting SupportingRequest Request for for License License Amendment Regarding Regarding Measurement Uncertainty Uncertainty Recapture Power   Power Uprate Uprate


==References:==
==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


1.Letter from Craig Lambert (Exelon Generation Company, LLC) to U. S. NRC,"Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated June 23, 2011 2.Letter from B. Mozafari (U. S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Units 1 and 2 - Request for Additional Information RE: Measurement Uncertainty Recapture Power Uprate Request (TAC NOS. ME6587, ME6588, 6589, AND ME6590)," dated February 14, 2012 [ML 120270146]
20, 2012 February 20,2012 U.S.
3.Letter from B. Mozafari (U. S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2 and Braidwood Station, Units 1 and 2 - Request for Additional Information RE: Measurement Uncertainty Recapture Power Uprate Request (TAC NOS. ME6587, ME6588, 6589, AND ME6590)," dated February 14, 2012 [ML 120260936]
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 .
In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Facility Operating License Nos. NPF-72, NPF-77, NPF-37 and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively. Specifically, the proposed changes revisethe Operating License and Technical Specifications to implement an increase in rated thermal power of approximately 1.63% based on increased feedwater flow measurement accuracy. In References 2 and 3, the NRC requested additional information to support review of the proposed changes. In response to this request, EGC is providing the attached information for all of the requests with the exception of the Civil and Mechanical Branch [ECMB] Request 13 in Reference 2 and the Balance of Plant Branch [SBPB] Request 1 in Reference 3. EGC will be G en e-ration 10 CFR 50.90 Exeion .DnMoT'0n Generation 4300 Winfield Road Warrenville, IL 60555 10 CFR 50.90 RS-12-033 February 20,2012 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and STN 50-457 Byron Station, Units 1 and 2 Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455
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


==Subject:==
Braidwood and Braidwood  and Byron Stations Measurement Measurement Uncertainty Recapture License LicenseAmendment AmendmentRequest Request(MUR (MURLAR)
Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate
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-    -


==References:==
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,"
: 1. Letter from Craig Lambert (Exelon Generation Company, LLC) to u. S. NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated June 23,2011 2. Letter from B. Mozafari (U. S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Units 1 and 2 -Request for Additional Information RE: Measurement Uncertainty Recapture Power Uprate Request (TAG NOS. ME6587, ME6588, 6589, AND ME6590)," dated February 14, 2012 [ML 120270146]
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.
: 3. Letter from B. Mozafari (U. S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Byron Station, Unit Nos. 1 and 2 and Braidwood Station, Units 1 and 2 -Request for Additional Information RE: Measurement Uncertainty Recapture Power Uprate Request (TAG NOS. ME6587, ME6588, 6589, AND ME6590)," dated February 14, 2012 [ML 120260936]
margin.
In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Facility Operating License Nos. NPF-72, NPF-77, NPF-37 and NPF-66 for Braidwood Station, Units 1 and 2, and Byron Station, Units 1 and 2, respectively.
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,"
Specifically, the proposed changes revise the Operating License and Technical Specifications to implement an increase in rated thermal power of approximately 1.63% based on increased feedwater flow measurement accuracy.
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
In References 2 and 3, the NRC requested additional information to support review of the proposed changes. In response to this request, EGC is providing the attached information for all of the requests with the exception of the Civil and Mechanical Branch [ECMB] Request 13 in Reference 2 and the Balance of Plant Branch [SBPB] Request 1 in Reference
(+0.6°F) inin the peak peak containment containment air temperature for Unit 1. Unit              Unit22remains remainsbounded boundedby  bythe the analysis of record.
: 3. EGC will be February 20, 2012 U.S. Nuclear Regulatory Commission Page 2 providing the response to these two requests under separate transmittal as indicated in .
Exelon's Exelon's response response (Reference (Reference 2) to   to the NRC NRCRequestRequestfor  forAdditional Additional Information Information (Reference (Reference 3)    3)
EGC has reviewed the information supporting a finding of no significant hazards considerationand the environmental consideration provided to the NRC in Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be
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.


prepared in connection with the proposed amendment.
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
There are no regulatory commitments contained in this letter.
Should you have any questions concerning this letter, please contact Leslie E. Holden at (630) 657-3316.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 201h day of February 2012.
Respectfully, Kevin F. Borton Manager, Licensing - Power Uprate : Response to Request for Additional Information cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Braidwood Station NRC Senior Resident Inspector - Byron Station Illinois Emergency Management Agency - Division of Nuclear Safety February 20,2012 U.S. Nuclear Regulatory Commission Page 2 providing the response to these two requests under separate transmittal as indicated in Attachment
: 1. EGC has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in Reference
: 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration.
In addition, the additional information provided in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed amendment.
There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Leslie E. Holden at (630) 657-3316.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 20 th day of February 2012. Respectfully, Kevin F. Borton Manager, Licensing
-Power Uprate Attachment 1: Response to Request for Additional Information cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector
-Braidwood Station NRC Senior Resident Inspector
-Byron Station Illinois Emergency Management Agency -Division of Nuclear Safety Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR)
RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI)
February 20, 2012 ATTACHMENT I RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION (NON-PROPRIETARY)
Braidwood and Byron Stations Measurement Uncertainty Recapture License Amendment Request (MUR LAR) RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION (RAI) February 20, 2012 ATTACHMENT 1 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION (NON.PROPRIETARY)
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 1 NON-PROPRIETARY NRC/Mechanical and Civil Engineering Branch (EMCB)
NRC/EMCB Request I Section IV.1.A.ii.f of Attachment 7 to the license amendment request (LAR) discusses the structural evaluation of the lower and upper core support assemblies for the effects of increased heat generation rates.
Provide further information and confirm thata.the proposed MUR power uprate only affects the design loads associated with heat generation rates and all other design loads associated with the design of the reactor vessel internals are unaffected by the proposed MUR power uprate; b.all design loading conditions, as noted in Section 3.9.5.2 of the Byron and Braidwood updated final safety analysis report (UFSAR), were considered in the structural re-evaluation of the reactor vessel internal components to assess the impact of the proposed MUR power uprate; andc.the original design codes of record were utilized in the structural re-evaluation of the reactor vessel internal components.
Provide the maximum calculated stresses and cumulative fatigue usage factor for the most limiting component of the reactor vessel internals and their respective comparison with the Byron and Braidwood design basis acceptance criteria.
Response The Byron and Braidwood reactor vessel internal components analysis of record (AOR) was performed with conservative gamma heating rates.The Measurement Uncertainty Recapture (MUR) power uprate gamma heating rates were verified to remain bounded by the conservative heating rates used in the AOR.
All the design loading conditions noted in Section 3.9.5.2 of the Byron and Braidwood Updated Final Safety Analysis Report (UFSAR) were considered in the structural assessment of the reactor vessel internal components to assess the impact of the proposed MUR power uprate.
The design loads associated with the design of the reactor vessel internals remain bounded by the AOR.The Byron and Braidwood Units 1 and 2 reactor vessel internals components were designed prior to the introduction of Subsection NG of the American Society 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 a result, a plant-specific stress report of the reactor internals was not required. However, the design of the reactor internals is evaluated according to the Westinghouse Criteria which is similar to the criteria described in the Subsection NG of the ASME code. The Westinghouse acceptance criteria are the same as those used in the original design of the plant and its original licensing basis.
The maximum calculated stresses and cumulative fatigue usage factor for the most-limiting component of the reactor vessel internals are unaffected by the MUR power uprate and remain bounded by the AOR.Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 1 NON-PROPRIETARY NRCIMechanical and Civil Engineering Branch (EMCB) NRCIEMCB Request 1 Section IV.1.A.ii.f of Attachment 7 to the license amendment request (LAR) discusses the structural evaluation of the lower and upper core support assemblies for the effects of increased heat generation rates. Provide further information and confirm that a. the proposed MUR power uprate only affects the design loads associated with heat generation rates and all other design loads associated with the design of the reactor vessel internals are unaffected by the proposed MUR power uprate; b. all design loading conditions, as noted in Section 3.9.5.2 of the Byron and Braidwood updated final safety analysis report (UFSAR), were considered in the structural evaluation of the reactor vessel internal components to assess the impact of the proposed MUR power uprate; and c. the original design codes of record were utilized in the structural re-evaluation of the reactor vessel internal components.
Provide the maximum calculated stresses and cumUlative fatigue usage factor for the most limiting component of the reactor vessel internals and their respective comparison with the Byron and Braidwood design basis acceptance criteria.
Response The Byron and Braidwood reactor vessel internal components analysis of record (AOR) was performed with conservative gamma heating rates. The Measurement Uncertainty Recapture (MUR) power uprate gamma heating rates were verified to remain bounded by the conservative heating rates used in the AOR. All the design loading conditions noted in Section 3.9.5.2 of the Byron and Braidwood Updated Final Safety Analysis Report (UFSAR) were considered in the structural assessment of the reactor vessel internal components to assess the impact of the proposed MUR power uprate. The design loads associated with the design of the reacto'r vessel internals remain bounded by the AOR. The Byron and Braidwood Units 1 and 2 reactor vessel internals components were designed prior to the introduction of Subsection NG of the American Society 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 a result, a plant-specific stress report of the reactor internals was not required.
However, the design of the reactor internals is evaluated according to the Westinghouse Criteria which is similar to the criteria described in the Subsection NG of the ASME code. The Westinghouse acceptance criteria are the same as those used in the original design of the plant and its original licensing basis. The maximum calculated stresses and cumulative fatigue usage factor for the most-limiting component of the reactor vessel internals are unaffected by the MUR power uprate and remain bounded by the AOR.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 2 NON-PROPRIETARY NRCIEMCB Request 2 Section 3.9.5.1 of the Byron and Braidwood UFSAR describes the reactor vessel internals as three parts consisting of the lower core support structure, the upper core support structure, and the incore instrumentation support structure. Section IV of Attachment 7 to the LAR does not discuss the incore instrumentation support structures. Provide further information relative to the impact of the design conditions associated with the proposed MUR power uprate on the incore instrumentation support structures.
Response As stated in UFSAR Section 3.9.5.1, the in-core instrumentation support structures consist of an upper system to convey and support thermocouples penetrating the vessel through the head and a lower system to convey and support flux thimbles penetrating the vessel through the bottom.The proposed MUR power uprate impact on the incore instrumentation support structures, including both the upper support columns and the lower support columns was assessed. Since the current analyses loads (i.e. LOCA hydraulic forces and seismic loads) are not changing from the current analysis of record and remain bounded for the MUR power uprate, the stresses and the cumulative fatigue usage factors in these components remain unchanged from the current analysis of record.
NRCIEMCB Request 3 Provide further information and confirm that, for the proposed MUR power uprate conditions, the maximum deflection values allowed for the reactor vessel internal support structures, as noted in Table 3.9-4 of the Byron and Braidwood UFSAR, are maintained.
Response The design inputs, i.e. LOCA hydraulic and seismic forces and geometry, are not changing from the current analysis of record for the MUR power uprate; therefore, there is no impact on the allowable deflections provided in Byron and Braidwood UFSAR Table 3.9-4, "Maximum Deflections Allowed for Reactor Internal Support Structure." The values provided in UFSAR Table 3.9-4 remain valid for the MUR power uprate.
NRCIEMCB Request 4 Section IV.1.B.iv.1 of Attachment 7 to the LAR states that there is an approximate 1.2°F increase in temperature difference across the core (That increases approximately 0.6°F and TC,,d decreases approximately 0.6°F) from current operating conditions due to the MUR poweruprate.Section IV.1.A.i of Attachment 7 to the LAR discusses reactor vessel structural evaluation and states that due to operational restrictions, the MUR minimum vessel inlet and maximum vessel outlet temperatures are limited to 538.2°F and 618.4°F, respectively.
Provide further clarification on temperature effects relative to the values in Tables 3-1 and 3-2 of  to the LAR, the statements in Sections IV.1.B.iv. I and IV.1.A.i of the LAR, and the temperatures used in the analysis of record.
Furthermore, the lifting lug loads and evaluation are discussed in Section IV.1.A.i of Attachment 7 to the LAR. The terminology of "lifting lug" and its relation to and its inclusion in the proposed MUR power uprate license amendment is not clear. Provide further information to clarify which NRCIEMCB Request 2 Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 2 NON*PROPRIETARY Section 3.9.5.1 of the Byron and Braidwood UFSAR describes the reactor vessel internals as three parts consisting of the lower core support structure, the upper core support structure, and the incore instrumentation support structure.
Section IV of Attachment 7 to the LAR does not discuss the incore instrumentation support structures.
Provide further information relative to the impact of the design conditions associated with the proposed MUR power uprate on the incore instrumentation support structures.
Response As stated in UFSAR Section 3.9.5.1, the in-core instrumentation support structures consist of an upper system to convey and support thermocouples penetrating the vessel through the head and a lower system to convey and support flux thimbles penetrating the vessel through the bottom. The proposed MUR power uprate impact on the incore instrumentation support structures, including both the upper support columns and the lower support columns was assessed.
Since the current analyses loads (Le. LOCA hydraulic forces and seismic loads) are not changing from the current analysis of record and remain bounded for the MUR power uprate, the stresses and the cumulative fatigue usage factors in these components remain unchanged from the current analysis of record. NRCIEMCB Request 3 Provide further information and confirm that, for the proposed MUR power uprate conditions, the maximum deflection values allowed for the reactor vessel internal support structures, as noted in Table 3.9-4 of the Byron and Braidwood UFSAR, are maintained. . Response The design inputs, Le. LOCA hydraulic and seismic forces and geometry, are not changing from the current analysis of record for the MUR power uprate; therefore, there is no impact on the allowable deflections provided in Byron and Braidwood UFSAR Table 3.9-4, "Maximum Deflections Allowed for Reactor Internal Support Structure." The values provided in UFSAR Table 3.9-4 remain valid for the MUR power uprate. NRCIEMCB Request 4 Section IV. 1.B.iv. 1 of Attachment 7 to the LAR states that there is an approximate 1.2°F increase in temperature difference across the core (T hot increases approximately O.6°F and T cold decreases approximately O.6°F) from current operating conditions due to the MUR power uprate. Section IV.1.Ai of Attachment 7 to the LAR discusses reactor vessel structural evaluation and states that due to operational restrictions, the MUR minimum vessel inlet and maximum vessel outlet temperatures are limited to 538. 2°F and 618.4°F, respectively.
Provide further clarification on temperature effects relative to the values in Tables 3-1 and 3-2 of Attachment 1 to the LAR, the statements in Sections IV. 1.B.iv. 1 and IV.1.Ai of the LAR, and the temperatures used in the analysis of record. Furthermore, the lifting lug loads and evaluation are discussed in Section IV.1.Ai of Attachment 7 to the LAR. The terminology of "lifting lug" and its relation to and its inclusion in the proposed MUR power uprate license amendment is not clear. Provide further information to clarify which Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 3 NON-PROPRIETARY reactor vessel component is referred to as "lifting lug". Also, regarding the affected reactor vessel component, a.provide a table summarizing the comparison of design parameters for the current operation conditions, MUR power uprate conditions, and design basis conditions; and b.provide the maximum calculated stresses and cumulative fatigue usage factors at the most critical location of the affected component and their respective comparison with the Byron and Braidwood design basis acceptance criteria.
Response The MUR power uprate Reactor Coolant System (RCS) design conditions given in Tables 3-1 and 3-2 provide a Tag range in which the minimum Todd is 541.4°F and the maximum Thor is 620.9°F. The reactor vessel analysis of record (AOR) evaluated a minimum Tco,d of 538.2°F and a maximum Thot of 620.3°F. Therefore, the MUR power uprate maximum Thot of 620.9°F exceeds the maximum Thot evaluated in the reactor vessel AOR. Note that the MUR power uprate minimum Too,d is bounded by the minimum Too,d evaluated in the reactor vessel AOR.
Normally, a reconciliation would be necessary because the MUR power uprate maximum Thor is not bounded by the maximum Thor evaluated in the reactor vessel AOR. However, all Byron and Braidwood units have plant operational limits which restrict the minimum T"Id to 538.2°F and the maximum Thot to 618.4°F. The plant operational limits will remain in place for the MUR power uprate. Therefore, the minimum Tco,d and maximum Thot evaluated in the reactor vessel AOR bound those of the MUR power uprate when the plant operational limits are taken into consideration.
There are three lifting lugs oriented 120° apart around the external side of the reactor vessel closure head. The Integrated Head Package (IHP) lift rod assemblies attach to the lifting lugs through a lift rod clevis and clevis pin. Figures EMCB R4-1 and R4-2 depict how the lifting lugs are attached to the reactor vessel closure head.
The lifting lug mechanical loads identified for current operating conditions did not change due to the MUR power uprate.
Bottom Portion of IHP Figure EMCB R4 - 2:Detail of Figure EMCB R4 - 1:
Bottom Portion of IntegratedLifting Lug Attachment to Reactor Head Package to Reactor Vessel Closure HeadVessel Closure Head Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 3 NON-PROPRIETARY reactor vessel component is referred to as "lifting lug". Also, regarding the affected reactor vessel component
: a. provide a table summarizing the comparison of design parameters for the current operation conditions, MUR power uprate conditions, and design basis conditions; and b. provide the maximum calculated stresses and cumulative fatigue usage factors at the most critical location of the affected component and their respective comparison with the Byron and Braidwood design basis acceptance criteria.
Response The MUR power uprate Reactor Coolant System (RCS) design conditions given in Tables 3-1 and 3-2 provide a T avg range in which the minimum T cold is 541.4 OF and the maximum T hot is 620.9°F. The reactor vessel analysis of record (AOR) evaluated a minimum T cold of 538.2°F and a maximum T hot of 620.3°F. Therefore, the MUR power uprate maximum T hot of 620.9°F exceeds the maximum T hot evaluated in the reactor vessel AOR. Note that the MUR power uprate minimum T cold is bounded by the minimum T cold evaluated in the reactor vessel AOR. Normally, a reconciliation would be necessary because the MUR power uprate maximum T hot is not bounded by the maximum T hot evaluated in the reactor vessel AOR. However, all Byron and Braidwood units have plant operational limits which restrict the minimum T cold to 538.2°F and the maximum T hot to 618.4°F. The plant operational limits will remain in place for the MUR power uprate. Therefore, the minimum T cold and maximum T hot evaluated in the reactor vessel AOR bound those of the MUR power uprate when the plant operational limits are taken into consideration.
There are three lifting lugs oriented 120° apart around the external side of the reactor vessel closure head. The Integrated Head Package (IHP) lift rod assemblies attach to the lifting lugs through a lift rod clevis and clevis pin. Figures EMCB R4-1 and R4-2 depict how the lifting lugs are attached to the reactor vessel closure head. The lifting lug mechanical loads identified for current operating conditions did not change due to the MUR power uprate. Bottom Portion of IHP Lifting Lug 1.1 fT toO AMeMIL l Lift Rod Clevis and Clevis Pin Figure EMCB R4 -1: Bottom Portion of Integrated Head Package to Reactor Vessel Closure Head I't. I I , t I ! ! L I I i J , I J J I I I i Figure EMCB R4 -2: Detail of Lifting Lug Attachment to Reactor Vessel Closure Head Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 4 NON-PROPRIETARY NRC/EMCB Request 5 Section IV.1.A.iii of Attachment 7 to the LAR discusses the control rod drive mechanism (CRDM). In this section, it is stated that updated seismic and loss-of-coolant accident (LOCA) loads remain less than the allowable loads provided in the analysis of record. This statement implies that the seismic loads have been updated.Also, this statement is not consistent with Section IV.1.A.ii.e of Attachment 7 to the LAR where it is stated that the proposed MUR power uprate conditions do not affect the current design basis for seismic and LOCA loads. Provide further clarification.
Furthermore, Section IV.1.A.iii of Attachment 7 to the LAR states that CRDM is subjected to Tcold temperatures and reactor coolant system pressures and these are the only design parameters considered in the CRDM evaluation. Elaborate and confirm that:
a.the design basis loading conditions and operational requirements, as noted in Section 3.9.4 of the Byron and Braidwood UFSAR, have been considered in the structural evaluation of the control rod drive system for the proposed MUR power uprate conditions; and b.the control rod drive system will continue to be in compliance with the Byron and Braidwood design basis acceptance criteria under the proposed MUR power uprate conditions.
Response A seismic and loss of coolant accident (LOCA) loads assessment was completed as part of the MUR power uprate. The assessment concluded that MUR uprate conditions have no impact on the seismic/LOCA loads and the existing seismic/LOCA loads remain valid and unchanged for the MUR power uprate.
The CRDM assessment completed for the MUR uprate project considered all pressure and thermal design transients and load combinations noted in Section 3.9.4 of the Byron Braidwood UFSAR. The CRDM assessment concluded that the pressure and thermal design transients due to the MUR uprate have no impact on the CRDM qualification analyses of record. The CRDM qualification analyses of record demonstrated that Byron and Braidwood are in compliance with the ASME Code stress criteria.
NRC/EMCB Request 6 Provide further information and confirm that the design basis pressure and temperatures (normal operating and accident temperatures) used in the design of the containment structure, including the steel liner plate, and its internal structures remain bounding following the proposed MUR power uprate.
Response The design basis containment pressure and temperature for normal operation are delineated respectively in Byron/Braidwood Technical Specification 3.6.4 and 3.6.5. Assessments performed for the MUR power uprate concluded that these normal operation design parameters remain applicable.
Accident containment parameters were evaluated for the MUR power uprate. For primary system pipe breaks (i.e., LOCAs), as discussed in the MUR LAR submittal (Reference 1), Section 111.15.5, "LOCA Long Term Mass and Energy Release and Containment Response -
NRCIEMCB Request 5 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 4 NON-PROPRIETARY Section IV. 1.A.iii of Attachment 7 to the LAR discusses the control rod drive mechanism (GRDM). In this section, it is stated that updated seismic and loss-of-coolant accident (LOGA) loads remain less than the allowable loads provided in the analysis of record. This statement implies that the seismic loads have been updated. Also, this statement is not consistent with Section IV.1.A.ii.e of Attachment 7 to the LAR where it is stated that the proposed MUR power uprate conditions do not affect the current design basis for seismic and LOGA loads. Provide further clarification.
Furthermore, Section IV.1.A.iii of Attachment 7 to the LAR s.tates that GRDM is subjected to Tcold temperatures and reactor coolant system pressures and these are the only design parameters considered in the CRDM evaluation.
Elaborate and confirm that: a. the design basis loading conditions and operational requirements, as noted in Section 3.9.4 of the Byron and Braidwood UFSAR, have been considered in the structural evaluation of the control rod drive system for the proposed MUR power uprate conditions; and b. the control rod drive system will continue to be in compliance with the Byron and Braidwood design basis acceptance criteria under the proposed MUR power uprate conditions.
Response A seismic and loss of coolant accident (LOCA) loads assessment was completed as part of the MUR power uprate. The assessment concluded that MUR uprate conditions have no impact on the seismic/LOCA loads and the existing seismic/LOCA loads remain valid and unchanged for the MUR power uprate. The CRDM assessment completed for the MUR uprate project considered all pressure and thermal design transients and load combinations noted in Section 3.9.4 of the Byron Braidwood UFSAR. The CRDM assessment concluded that the pressure and thermal design transients due to the MUR uprate have no impact on the CRDM qualification analyses of record. The CRDM qualification analyses of record demonstrated that Byron and Braidwood are in compliance with the ASME Code stress criteria.
NRCIEMCB Request 6 Provide further information and confirm that the design basis pressure and temperatures (normal operating and accident temperatures) used in the design of the containment structure, including the steel liner plate, and its internal structures remain bounding following the proposed MUR power uprate. Response The design basis containment pressure and temperature for normal operation are delineated respectively in Byron/Braidwood Technical Specification 3.6.4 and 3.6.5. Assessments performed for the MUR power uprate concluded that these normal operation design parameters remain applicable.
Accident containment parameters were evaluated for the MUR power uprate. For primary system pipe breaks (i.e., LOCAs), as discussed in the MUR LAR submittal (Reference 1), Section 111.15.5, "LOCA Long Term Mass and Energy Release and Containment Response-Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 5 NON-PROPRIETARY UFSAR 6.2.1.3.1, Analysis Results," the containment peak pressure and temperature for the MUR remain bounded by the containment structure design pressure and temperature with margin.For secondary pipe breaks (Main Steam Line Breaks (MSLB)), as discussed in the MUR LAR submittal (Reference 1), Section 111.16.5, "Main Steam Line Break Mass and Energy Releases Inside Containment -
UFSAR 6.2.1.4, Analysis Results," the peak pressure remains bounded by the containment design pressure with margin and there is a very small calculated increase
(+0.6°F) in the peak containment air temperature for Unit 1. Unit 2 remains bounded by the analysis of record.
Exelon's response (Reference 2) to the NRC Request for Additional Information (Reference 3)
Request 10, summarized the temperatures and pressures from the LOCA and MSLB Mass and Energy Analyses for Byron
/Braidwood MUR.
As discussed in the UFSAR Section 6.2.1.1.3, "Containment Structure, Design Evaluation," the justification for the design temperatures selected for the liner and internal containment structures is that they are conservative when the duration of the peak temperature for the secondary side (i.e., steam line) break, the temperature lag between the containment atmosphere and the passive heat sinks such as the containment liner and internal structures, and the resistance to heat transfer provided by the materials used, are considered. This justification remains applicable for MUR power uprate because the duration remains short.
Figure 10-1, "Containment EQ Temperature and Pressure Profile," in Reference 2 shows that the MSLB temperature profile for the MUR power uprate falls below the containment design temperature of 280
°F less than 200 seconds after the onset of the MSLB.
The assessment performed for the MUR power uprate indicated that the structural effect of the MSLB temperature on the containment structure remains bounded by the LOCA case.
Therefore for both units the containment structure remains acceptable for both primary and secondary system pipe breaks.
For the containment internal structures, RCS initial pressure and temperature for MUR were reviewed and confirmed to be bounded by the inputs to the existing short-term LOCA mass and energy releases. Therefore the containment internal structures remain acceptable for the MUR power uprate.
NRC/EMCB Request 7 Section IV.1.A.iv "Reactor Coolant Piping and Supports" of Attachment 7 to the LAR discusses the effects of the proposed MUR power uprate mostly on a qualitative basis and the term "no significant changes" has been used in several areas to describe the impact of the proposed MUR power uprate. Discuss in more detail the information relative to the revised design conditions, before and after the proposed MUR power uprate, for those components evaluated under Section IV. 9.A.iv of Attachment 7 to the LAR.
Summarize the results of any additional evaluations performed for the affected components and indicate whether these components remain bounded by the current analysis of record. For those components that were not bounded by the analysis of record:
a.provide the maximum calculated stresses and cumulative fatigue usage factors at the most critical location; and b.provide further clarification that the re-evaluation was performed in accordance with the design basis code of record and-the affected components continue to remain in compliance with the Byron and Braidwood stations design basis acceptance criteria.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 5 NON-PROPRIETARY UFSAR 6.2.1.3.1, Analysis Results," the containment peak pressure and temperature for the MUR remain bounded by the containment structure design pressure and temperature with margin. For secondary pipe breaks (Main Steam Line Breaks (MSLB>>, as discussed in the MUR LAR submittal (Reference 1), Section 111.16.5, "Main Steam Line Break Mass and Energy Releases Inside Containment
-UFSAR 6.2.1.4, Analysis Results," the peak pressure remains bounded by the containment design pressure with margin and there is a very small calculated increase (+0.6°F) in the peak containment air temperature for Unit 1. Unit 2 remains bounded by the analysis of record. Exelon's response (Reference
: 2) to the NRC Request for Additional Information (Reference
: 3) Request 10, summarized the temperatures and pressures from the LOCA and MSLB Mass and Energy Analyses for Byron/Braidwood MUR. As discussed in the UFSAR Section 6.2.1.1.3, "Containment Structure, Design Evaluation," the justification for the design temperatures selected for the liner and internal containment structures is that they are conservative when the duration of the peak temperature for the secondary side (Le., steam line) break, the temperature lag between the containment atmosphere and the passive heat sinks such as the containment liner and internal structures, and the resistance to heat transfer provided by the materials used, are considered.
This justification remains applicable for MUR power uprate because the duration remains short. Figure 10-1, "Containment EQ Temperature and Pressure Profile," in Reference 2 shows that the MSLB temperature profile for the MUR power uprate falls below the containment design temperature of 280°F less than 200 seconds after the onset of the MSLB. The assessment performed for the MUR power uprate indicated that the structural effect of the MSLB temperature on the containment structure remains bounded by the LOCA case. Therefore for both units the containment structure remains acceptable for both primary and secondary system pipe breaks. For the containment internal structures, RCS initial pressure and temperature for MUR were reviewed and confirmed to be bounded by the inputs to the existing short-term LOCA mass and energy releases.
Therefore the containment internal structures remain acceptable for the MUR power uprate. NRCIEMCB Request 7 Section IV.1.A.iv "Reactor Coolant Piping and Supports" of Attachment 7 to the LAR discusses the effects of the proposed MUR power uprate mostly on a qualitative basis and the term "no significant changes" has been used in several areas to describe the impact of the proposed MUR power uprate. Discuss in more detail the information relative to the revised design conditions, before and after the proposed MUR power uprate, for those components evaluated under Section IV.1.A.iv of Attachment 7 to the LAR. Summarize the results of any additional evaluations petiormed for the affected components and indicate whether these components remain bounded by the current analYSis of record. For those components that were not bounded by the analysis of record: a. provide the maximum calculated stresses and cumulative fatigue usage factors at the most critical location; and b. provide further clarification that the re-evaluation was petiormed in accordance with the design basis code of record and*the affected components continue to remain in compliance with the Byron and Braidwood stations design basis acceptance criteria.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 6 NON-PROPRIETARY Response The conditions associated with the MUR power uprate were evaluated to determine the impact on the existing as-built design basis reactor coolant loop (RCL) analysis for the following:RCL piping stresses and displacements,Primary equipment nozzle loads (reactor pressure vessel (RPV) inlet and outlet nozzles, steam generator inlet and outlet nozzles, and reactor coolant pump (RCP) suction and discharge nozzles),Primary equipment support loads (RPV nozzle supports, steam generator columns and lateral bumpers, RCP columns and lateral supports, and pressurizer supports), andPressurizer surge line piping stresses and displacements including the effects of thermal stratification.
The following inputs were considered in the assessment:Nuclear Steam Supply System (NSSS) Design Parameters,NSSS design transients,Loss-of-coolant accident (LOCA) hydraulic forcing functions loads, andRPV motions due to LOCH.
The RCL piping assessment for the MUR power uprate was performed in accordance with the Byron/Braidwood design basis code of record (ASME, Section III, 1974 Edition, including Summer 1975 addenda).
The RCL thermal, deadweight, seismic, fatigue, LOCA and Main Steam / Feedwater line break analyses were reconciled to the design inputs as follows:
RCL Thermal Analysis The RCL piping in the existing design basis was evaluated for the conditions associated with a RCS hot leg upper bound temperature of 618.4°F, cross-over leg temperature of 555.4°F, and a cold leg temperature of 555.7°F. The reactor coolant upper bound temperatures for the MUR power uprate did not increase for the hot leg, they decreased by 0.6°F for the cross-over leg, and they decreased by 0.6°F for the cold leg as compared to the current design basis temperatures. The MUR power uprate upper bound thermal NSSS design parameters are bounded by the design basis analysis.
Considering the RCL MUR power uprate lower bound temperature case, there is a temperature operating window as follows: 9.8°F between the upper bound Thigh and lower bound Trout for the hot leg, 16.9°F between the upper bound Thigh and lower bound Teow for the cross-over leg, and 16.9°F between the upper bound Thigh and lower bound Tlow for the cold leg.The thermal piping stresses and displacements are dependent on the coefficient of thermal expansion and temperature difference between ambient to hot conditions. The coefficient of thermal expansion increases with an increase in temperature. The thermal piping loads and


thermal stresses for the lower bound temperatures are lower than the corresponding loads and stresses for the upper bound case. Therefore, the thermal stresses for the upper bound case are higher, and the upper bound case piping stresses, primary equipment nozzle loads, primary equipment support loads (including the reactor vessel, steam generator, reactor coolant pump and pressurizer), and the auxiliary line displacements at the connections to the RCL are limiting.
===Response===
Response Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 6 NON-PROPRIETARY The conditions associated with the MUR power uprate were evaluated to determine the impact on the existing as-built design basis reactor coolant loop (RCL) analysis for the following:
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 piping stresses and displacements,
(RCL) analysis analysisfor    forthe thefollowing:
* Primary equipment nozzle loads (reactor pressure vessel (RPV) inlet and outlet nozzles, steam generator inlet and outlet nozzles, and reactor coolant pump (RCP) suction and discharge nozzles),
following:
* Primary equipment support loads (RPV nozzle supports, steam generator columns and lateral bumpers, RCP columns and lateral supports, and pressurizer supports), and
  **       RCL piping RCL   piping stresses and displacements,
* Pressurizer surge line piping stresses and displacements including the effects of thermal stratification.
  **        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),
The following inputs were considered in the assessment:
  **       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),
* Nuclear Steam Supply System (NSSS) Design Parameters,
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,
* NSSS design transients,
* Loss-of-coolant accident (LOCA) hydraulic forcing functions loads, and
* Loss-of-coolant accident (LOCA) hydraulic         hydraulic forcing forcing functions functions loads,loads, and and
* RPV motions due to LOCA. The RCL piping assessment for the MUR power uprate was performed in accordance with the Byron/Braidwood design basis code of record (ASME, Section III, 1974 Edition, including Summer 1975 addenda).
* RPV motions due    due to to LOCH.
The RCL thermal, deadweight, seismic, fatigue, LOCA and Main Steam / Feedwater line break analyses were reconciled to the design inputs as follows: RCL Thermal Analysis The RCL piping in the existing design basis was evaluated for the conditions associated with a RCS hot leg upper bound temperature of 618.4°F, cross-over leg temperature of 555.4°F, and a cold leg temperature of 555.rF. The reactor coolant upper bound temperatures for the MUR power uprate did not increase for the hot leg, they decreased by 0.6°F for the cross-over leg, and they decreased by 0.6°F for the cold leg as compared to the current design basis temperatures.
LOCA.
The MUR power uprate upper bound thermal NSSS design parameters are bounded by the design basis analysis.
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).
Considering the RCL MUR power uprate lower bound temperature case, there is a temperature operating window as follows: 9.8°F between the upper bound Thigh and lower bound Ttowfor the hot leg, 16.9°F between the upper bound Thigh and lower bound Ttowfor the cross-over leg, and 16.9°F between the upper bound Thigh and lower bound Ttowforthe cold leg. The thermal piping stresses and displacements are dependent on the coefficient of thermal expansion and temperature difference between ambient to hot conditions.
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:
The coefficient of thermal expansion increases with an increase in temperature.
follows:
The thermal piping loads and thermal stresses for the lower bound temperatures are lower than the corresponding loads and stresses for the upper bound case. Therefore, the thermal stresses for the upper bound case are higher, and the upper bound case piping stresses, primary equipment nozzle loads, primary equipment support loads (including the reactor vessel, steam generator, reactor coolant pump and pressurizer), and the auxiliary line displacements at the connections to the RCL are limiting.
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.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 7 NON-PROPRIETARY Since there is no increase in upper-bound temperature in comparison to the hot leg, cross-over leg, and cold leg temperatures in the current RCL thermal analysis design basis, the current RCL thermal analysis design basis analysis remains bounding.
analysis.
RCL Deadweight and Seismic Analysis There is no change in deadweight because there is no change to the configuration of the RCL piping and supports due to the MUR power uprate. The seismic response spectrum does not change due to the MUR power uprate.. Therefore, it is concluded that there are no changes to RCL deadweight and seismic analyses for the MUR power uprate.
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:
RCL Fatigue and Surge Line Stratification There are no changes to the primary side NSSS design transients due to the MUR power uprate.Also, the pressurizer surge line transients do not change. Therefore, there is no impact on the piping for the MUR power uprate due to the NSSS design transients. There is no adverse effect on the fatigue evaluation of the RCL and pressurizer surge line, including the effects of thermal stratification. The pressurizer surge line stratification analysis continues to meet the code of record (ASME, Section III, 1986 Edition).
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.
LOCA Analysis The impact on the LOCA hydraulic forcing functions (HFFs) due to the MUR power uprate has been assessed for the accumulator and surge line breaks. Based on this assessment, the LOCA HFFs used in the existing RCL piping LOCA analyses remains bounding for the MUR power uprate.
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.
The impact on the RPV motions due to MUR power uprate has been assessed. Based on this assessment, the LOCA RPV motions used in the existing RCL piping LOCA analyses remains bounding for the MUR power uprate.
conditions. The Thecoefficient coefficientofof thermal expansion expansion increases increases with with an an increase increase in  in temperature.
Main Steam and Feedwater Line Break The design basis main steam and feedwater line break analyses remain valid for the MUR power uprate. Based on the NSSS design parameters, the main steam line and feedwater line break pressures decrease and the feedwater temperature decreases slightly for the MUR power uprate. A decrease in pressure will reduce the thrust and jet impingement forces; however a decrease in temperature may increase the forces due to fluid momentum.
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.
These small differences will offset each other such that the thrust and jet impingement forces used in the current analysis remain bounding.
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),
Based on the above, there are no changes due to the MUR power uprate to the piping or component qualification from the design basis, including: primary equipment nozzles and supports, Class 1 auxiliary piping analysis, and surge line stratification. The maximum primary and secondary stresses and the maximum fatigue usage factors associated with the existing design basis analysis are applicable to the MUR power uprate. The above components continue to remain in compliance with the Byron/Braidwood design basis acceptance criteria.
pressurizer), and    and the theauxiliary auxiliarylinelinedisplacements displacementsatatthe      the connections connections to  to the the RCL RCL are are limiting.
NRC/EMCB Request 8 Section I V. 1.A. v of Attachment 7 to the LAR discusses the evaluation of balance of plant (BOP) piping systems.Confirm that other BOP piping systems (e.g., chemical and volume control, auxiliary feedwater, fuel pool cooling, containment spray, essential service water, safety injection) that may be affected by the MUR uprate conditions have been evaluated and provide a complete list of BOP piping systems evaluated in support of MUR power uprate. Discuss the Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 7 NON-PROPRIETARY Since there is no increase in upper-bound temperature in comparison to the hot leg, over leg, and cold leg temperatures in the current RCL thermal analysiS design basis, the current RCL thermal analysis design basis analysis remains bounding.
limiting.
RCL Deadweight and Seismic Analysis There is no change in deadweight because there is no change to the configuration of the RCL piping and supports due to the MUR power uprate. The seismic response spectrum does not change due to the MUR power uprate. Therefore, it is concluded that there are no changes to RCL deadweight and seismic analyses for the MUR power uprate. RCL Fatigue and Surge Line Stratification There are no changes to the primary side NSSS design transients due to the MUR power uprate. Also, the pressurizer surge line transients do not change. Therefore, there is no impact on the piping for the MUR power uprate due to the NSSS design transients.
 
There is no adverse effect on the fatigue evaluation of the RCL and pressurizer surge line, including the effects of thermal stratification.
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.
The pressurizer surge line stratification analysis continues to meet the code of record (ASME, Section III, 1986 Edition).
bounding.
LOCA Analysis The impact on the LOCA hydraulic forcing functions (HFFs) due to the MUR power uprate has been assessed for the accumulator and surge line breaks. Based on this assessment, the LOCA HFFs used in the existing RCL piping LOCA analyses remains bounding for the MUR power uprate. The impact on the RPV motions due to MUR power uprate has been assessed.
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.
Based on this assessment, the LOCA RPV motions used in the existing RCL piping LOCA analyses remains bounding for the MUR power uprate. Main Steam and Feedwater Line Break The design basis main steam and feedwater line break analyses remain valid for the MUR power uprate. Based on the NSSS design parameters, the main steam line and feedwater line break pressures decrease and the feedwater temperature decreases slightly for the MUR power uprate. A decrease in pressure will reduce the thrust and jet impingement forces; however a decrease in temperature may increase the forces due to fluid momentum.
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.
These small differences will offset each other such that the thrust and jet impingement forces used in the current analysis remain bounding.
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).
Based on the above, there are no changes due to the MUR power uprate to the piping or component qualification from the design basis, including:
Edition).
primary equipment nozzles and supports, Class 1 auxiliary piping analysis, and surge line stratification.
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.
The maximum primary and secondary stresses and the maximum fatigue usage factors associated with the existing design basis analysis are applicable to the MUR power uprate. The above components continue to remain in compliance with the Byron/Braidwood design basis acceptance criteria.
uprate.
NRCIEMCB Request 8 Section IV. 1.A. v of Attachment 7 to the LAR discusses the evaluation of balance of plant (BOP) piping systems. Confirm that other BOP piping systems (e.g., chemical and volume control, auxiliary feedwater, fuel pool cooling, containment spray, essential service water, safety injection) that may be affected by the MUR uprate conditions have been evaluated and provide a complete list of BOP piping systems evaluated in support of MUR power uprate. Discuss the Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 8 NON-PROPRIETARY methodology used for evaluating BOP piping, including pipe supports, and provide further information relative to the design conditions in each BOP piping system, before and after the proposed MUR power uprate. Summarize the results of the additional evaluations performed for the affected piping systems and indicate whether these piping systems remain bounded by the current analysis of record. For those BOP piping systems not bounded by the current analysis of record:
The impact on the RPV motions due to MUR power                power uprate uprate has  has been been assessed.
a.provide the maximum calculated stresses and cumulative fatigue usage factors at the most critical location in each unbounded piping system; and b.provide further clarification that the re-evaluation of the piping system, including pipe supports, was performed in accordance with the design basis code of record and in compliance with the Byron and Braidwood stations design basis acceptance criteria.
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.
Furthermore, state whether any piping or pipe support modifications are required to support the proposed MUR power uprate.
uprate.
Response The following Byron and Braidwood Stations Balance of Plant/Nuclear Steam Supply System (BOP/NSSS) piping systems were assessed for MUR power uprate conditions:Main Steam System n Extraction Steam System n Condensate SystemCondensate Booster SystemHeater Drains System n Feedwater System n Steam Generator Blowdown System n Auxiliary Steam SystemAuxiliary Feedwater System n Chemical and Volume Control SystemFuel Pool Cooling System n Safety Injection System n Essential Service Water SystemComponent Cooling Water SystemContainment Spray SystemNon-Essential Service Water n Circulating Water It was determined that the following Byron and Braidwood Stations BOP/NSSS piping systems are not negatively impacted (i.e., an increase in temperature or pressure) by MUR power uprate: Steam Generator Blowdown System Auxiliary Feedwater System Chemical and Volume Control System Safety Injection System Containment Spray System Circulating Water Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 8 NON*PROPRIETARY methodology used for evaluating BOP piping, including pipe supports, and provide further information relative to the design conditions in each BOP piping system, before and after the proposed MUR power uprate. Summarize the results of the additional evaluations performed for the affected piping systems and indicate whether these piping systems remain bounded by the current analysis of record. For those BOP piping systems not bounded by the current analysis of record: a. provide the maximum calculated stresses and cumulative fatigue usage factors at the most critical location in each unbounded piping system; and b. provide further clarification that the re-evaluation of the piping system, including pipe supports, was performed in accordance.
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.
with the design basis code of record and in compliance with the Byron and Braidwood stations design basis acceptance criteria.
momentum.
Furthermore, state whether any piping or pipe support modifications are required to support the proposed MUR power uprate. Response The following Byron and Braidwood Stations Balance of Plant/Nuclear Steam Supply System (BOP/NSSS) piping systems were assessed for MUR power uprate conditions:
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
* Main Steam System
* Extraction Steam System
              *n Extraction Steam System
* Condensate System
            *n Condensate System
* Condensate Booster System
* Condensate Booster Booster System System
* Heater Drains System
* Heater Drains System
* F eedwater System
            *n F Feedwater eedwater System
* Steam Generator Blowdown System
            *n Steam Generator Blowdown Blowdown System System n* Auxiliary Steam SystemSystem
* Auxiliary Steam System
* Auxiliary Feedwater Feedwater System System n* Chemical and Volume Control      Control System System
* Auxiliary Feedwater System
* Fuel Pool Cooling System   System n* Safety Injection Injection System System n* Essential Service Water   Water System System
* Chemical and Volume Control System
* Component Cooling    Cooling Water Water System System
* Fuel Pool Cooling System
* Containment Containment SpraySpray System System
* Safety Injection 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.,
* Essential Service Water System
(i.e., an increase inin temperature or  or pressure) pressure) by by MUR MUR power power uprate:
* Component Cooling Water System
uprate:
* Containment Spray System
* Steam Generator Generator Blowdown Blowdown System System
* Non-Essential Service Water
* Auxiliary Feedwater Feedwater System System
* Circulating Water It was determined that the following Byron and Braidwood Stations BOP/NSSS piping systems are not negatively impacted (i.e., an increase in temperature or pressure) by MUR power uprate:
* Chemical   and and Volume Control Volume     Control System System
* Steam Generator Blowdown System
* Safety Injection  System Injection System
* Auxiliary Feedwater System
* Containment Spray Containment      Spray System System
* Chemical and Volume Control System
* Circulating Circulating Water Water
* Safety Injection System
 
* Containment Spray System
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.
* Circulating Water Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 9 NON-PROPRIETARY For these piping systems no further assessment was performed. These systems remain bounded.For the remaining systems (i.e., those that were assessed to have an increase in temperature and/or pressure) the methodology and acceptance criteria discussed in the paragraphs below were applied to assess the acceptability of the piping for the MUR power uprate.
performed. TheseThesesystems systems remain remain bounded.
Operating pressures and temperatures in each line under Current Licensed Thermal Power (CLTP)and MUR power uprate were reviewed against the design pressure and temperature of the line.For non-seismic piping, the increase in pressure was considered to be acceptable provided that the MUR power uprate operating pressure was bounded by the design pressure. As a result of the MUR power uprate, there were no non-seismic systems that exceeded the design pressure.
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.
For seismic piping, there were no pressure increases as a result of the MUR power uprate.The increase in temperature was considered to be acceptable provided that the MUR power uprate operating temperature did not increase by more than 1 % compared to CLTP operating temperature or the MUR operating temperature remained less than 150
uprate.
&deg;F.For lines that are currently qualified to be within Code thermal stress allowable, increasing the system temperature range by <1 % will not affect the acceptability of the piping
Operating pressures Operating    pressures and  and temperatures temperatures in    in each each line line under under Current Current Licensed Licensed Thermal Thermal Power (CLTP)
/support system.
(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.
Decreasing the system temperature will increase the allowable stress margin. For evaluating pipe thermal expansion stress, the temperature range is equal to the maximum operating temperature minus the normal ambient temperature, or 70&deg;F. This represents the largest change in temperature that the pipelines can experience.
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.
Typically, pipe thermal stress is not evaluated for operating temperatures less than 150&deg;F.
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.
For piping segments which do not pass the screening criterion (i.e.,<1 % change), a detailed review of pipe stress calculations is conducted to determine if margin exists to accommodate thermal expansion stresses at MUR power uprate.
pressure.
All of the systems, except for the heater drain piping and condensate booster piping are considered to remain bounded based on the above criteria.
piping, there For seismic piping,       there were no no pressure pressure increases increases as   as aaresult resultofofthe the MUR power uprate.
The heater drain system piping experiences a maximum temperature increase of 1.43%. The design basis analysis was found to bound the MUR condition because the design basis analysis used operating temperature of 187&deg;F while the CLTP operating temperature is 160.8&deg;F and the MUR operating temperature is 162.1&deg;F temperature. The condensate booster piping experiences a maximum temperature increase of 1.10%. The design basis analysis was found to bound MUR conditions because the design basis analysis used an operating temperature of 176&deg;F while the CLTP operating temperature is 161.0&deg;F and the MUR operating temperature is 162.0&deg;F. Therefore, the BOP/NSSS piping systems are considered to remain in compliance with their current design basis code of record and the Byron and Braidwood stations design basis acceptance criteria.
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&deg;&deg; 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.
Since there were no significant increases in piping temperatures, pipe support loads did not experience an appreciable increase.
Decreasing the system temperature will            will increase the allowable stress     stress margin.
Therefore, no pipe or pipe support modifications are required for MUR power uprate conditions.
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&deg;F.                          This represents 70&deg;F. This     represents the largest change in in temperature that the pipelinespipelines can experience.
NRC/EMCB Request 9 Section IV. I.A.viii of Attachment 7 to the LAR discusses the pressurizer structural evaluation. In this section of the LAR, it is stated that the revised design parameters have an insignificant impact on the fatigue analysis results. It is also stated that the proposed MUR power uprate has a negligible impact on the qualification of the pressurizer surge, spray, safety and relief nozzle structural weld overlay designs. Provide further information to support the above qualitative Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 9 NON-PROPRIETARY For these piping systems no further assessment was performed.
experience. Typically, pipe thermal stress        stress is is not not evaluated for operating temperatures less than 150&deg;F.              150&deg;F.
These systems remain bounded. For the remaining systems (i.e., those that were assessed to have an increase in temperature and/or pressure) the methodology and acceptance criteria discussed in the paragraphs below were applied to assess the acceptability of the piping for the MUR power uprate. Operating pressures and temperatures in each line under Current Licensed Thermal Power (CL TP) and MUR power uprate were reviewed against the design pressure and temperature of the line. For non-seismic piping, the increase in pressure was considered to be acceptable provided that the MUR power uprate operating pressure was bounded by the design pressure.
For piping segments which do not pass the screening criterion                criterion (i.e., <1<1 %% change),
As a result of the MUR power uprate, there were no non-seismic systems that exceeded the design pressure.
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.
For seismic piping, there were no pressure increases as a result of the MUR power uprate. The increase in temperature was considered to be acceptable provided that the MUR power uprate operating temperature did not increase by more than 1 % compared to CL TP operating temperature or the MUR operating temperature remained less than 150&deg; F. For lines that are currently qualified to be within Code thermal stress allowable, increasing the system temperature range by <1 % will not affect the acceptability of the piping/support system. Decreasing the system temperature will increase the allowable stress margin. For evaluating pipe thermal expansion stress, the temperature range is equal to the maximum operating temperature minus the normal ambient temperature, or 70&deg;F. This represents the largest change in temperature that the pipelines can experience.
uprate.
Typically, pipe thermal stress is not evaluated for operating temperatures less than 150&deg;F. For piping segments which do not pass the screening criterion (i.e., <1 % change), a detailed review of pipe stress calculations is conducted to determine if margin exists to accommodate thermal expansion stresses at MUR power uprate. All of the systems, except for the heater drain piping and condensate booster piping are considered to remain bounded based on the above criteria.
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.
The heater drain system piping experiences a maximum temperature increase of 1.43%. The design basis analYSis was found to bound the MUR condition because the design basis analysis used operating temperature of 187&deg;F while the CLTP operating temperature is 160.8&deg;F and the MUR operating temperature is 162.1&deg;F temperature.
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&deg;F while the CLTP operating temperature is                   is 160.8&deg;F 160.8&deg;F and and thethe MUR MURoperating operating temperature temperature is  is 162.1&deg;F temperature. The 162.1&deg;F                        Thecondensate condensateboosterboosterpiping piping experiences experiences aamaximum maximu mtemperature temperature increase of 1.10%.
The condensate booster piping experiences a maximu m temperature increase of 1.10%. The design basis analysis was found to bound MUR conditions because the design basis analysis used an operating temperature of 176&deg;F while the CL TP operating temperature is 161.0&deg;F and the MUR operating temperature is 162.0&deg;F. Therefore, the BOP/NSSS piping systems are considered to remain in compliance with their current design basis code of record and the Byron and Braidwood stations design basis acceptance criteria.
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&deg;F while the CLTP          CLTP operating operating temperature is 161.0&deg;F and the MUR          MUR operating operating temperature temperature is 162.0&deg;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.
Since there were no significant increases in piping temperatures, pipe support loads did not experience an appreciable increase.
criteria.
Therefore, no pipe or pipe support modifications are required for MUR power uprate conditions.
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.
NRCIEMCB Request 9 Section IV. 1.A. viii of Attachment 7 to the LAR discusses the pressurizer structural evaluation.
increase. Therefore, no pipe or pipe            pipe support modifications are required for MUR power  power uprate uprate conditions.
In this section of the LAR, it is stated that the revised design parameters have an insignificant impact on the fatigue analysis results. It is also stated that the proposed MUR power uprate has a negligible impact on the qualification of the pressurizer surge, spray, safety and relief nozzle structural weld overlay designs. Provide further information to support the above qualitative Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 10 NON-PROPRIETARY statements and to demonstrate compliance with the Byron and Braidwood design basis acceptance criteria. Also, provide a table summarizing the comparison of pressurizer design parameters for the current operation conditions, MUR power uprate conditions, and design basis conditions.
conditions.
Response Heat-up of the pressurizer from the cold condition to the hot standby condition is independent of plant power level and is unaffected by an uprate which may affect RCS temperatures and transients between hot standby and 100% power operation. The pressurizer maintains the RCS pressure and provides a cushion to accommodate changes in fluid volume and provides overpressure protection to the RCS. The temperature within the pressurizer is at the saturation temperature. Therefore, transients that will affect the fatigue analysis for pressurizer components are the result of changes to the fluid temperature entering the pressurizer, i.e., insurge/ outsurge through the surge line or spray through the spray line, or as a result in changes to the transients affecting the pressurizer pressure transients. Previous Westinghouse evaluations of design transients following an MUR power uprate show that the only transients that are affected are those that are the result of the feedwater changes and affect only the steam generator secondary side components. There are no transients affected that pertain to the pressurizer, temperature or pressure. Therefore, there is no impact on the pressurizer analysis as a result of MUR power uprate transient changes. Given that the transients are unchanged, the impact on the lower pressurizer components due to insurge/outsurge and the upper pressurizer components due to spray will change only if the temperature of the fluid changes, and then only if the temperature change increases. For this to happen, the RCS temperature for Thot, affecting insurge/outsurge, and Tcp,d, affecting the spray temperature, would have to decrease from the analyzed condition.
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
The Table EMCB R9-1 provides a comparison showing the temperature change across the pressurizer components evaluated for the design basis conditions, the current operating conditions, and at MUR power uprate conditions. It is seen from Table EMCB R9-1 that the temperature change for Th0t, affecting the lower pressurizer (AThot), is less at MUR power uprate
 
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)
(&deg;F)              (OF)
(&deg;F)                (OF)
(&deg;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 (&deg;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 (&deg;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 hours.
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 hours 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


conditions by 0.6 OF and is enveloped by the analysis of record (AOR). The temperature differential for the upper portion of the pressurizer is shown to exceed the current operating condition by 0.6 OF (ATWd). This is an increase of approximately 0.5% over the current operating condition AT.,d and is not considered to be significant.
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.
Also, since the baseline analysis, which is also the AOR, continues to envelope the MUR power uprate temperature differential, the AOR is not affected and remains applicable. Therefore, there is no change to the baseline analysis results due to the MUR power uprate resulting from changes to the RCS temperatures affecting the pressurizer.
The SG PORVs do not share mechanical      mechanical components.
An assessment of the pressurizer surge, spray, safety and relief nozzle for structural weld overlay (SWOL) was also performed as part of the MUR power uprate. The assessment concluded that the MUR power uprate would have no impact on the AOR for these components based on the findings previously noted. Therefore, the MUR power uprate is enveloped by the current SWOL analysis and is acceptable.
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.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 10 NON-PROPRIETARY statements and to demonstrate compliance with the Byron and Braidwood design basis acceptance criteria.
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.
Also, provide a table summarizing the comparison of pressurizer design parameters for the current operation conditions, MUR power uprate conditions, and design basis conditions.
1 PA34J.
Response Heat-up of the pressurizer from the cold condition to the hot standby condition is independent of plant power level and is unaffected by an uprate which may affect RCS temperatures and transients between hot standby and 100% power operation.
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).
The pressurizer maintains the RCS pressure and provides a cushion to accommodate changes in fluid volume and provides overpressure protection to the RCS. The temperature within the pressurizer is at the saturation temperature.
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.
Therefore, transients that will affect the fatigue analysis for pressurizer components are the result of changes to the fluid temperature entering the pressurizer, i.e., insurge/ outsurge through the surge line or spray through the spray line, or as a result in changes to the transients affecting the pressurizer pressure transients.
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.
Previous Westinghouse evaluations of design transients following an MUR power uprate show that the only transients that are affected are those that are the result of the feedwater changes and affect only the steam generator secondary side components.
Board. The Theoutput outputsignal signalfrom fromthethecontrollers controllersdrives drivesthe the hydraulic hydraulic pumppump motor to   to either open or close the valve.
There are no transients affected that pertain to the pressurizer, temperature or pressure.
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.
Therefore, there is no impact on the pressurizer analysis as a result of MUR power uprate transient changes. Given that the transients are unchanged, the impact on the lower pressurizer components due to insurge/outsurge and the upper pressurizer components due to spray will change only if the temperature of the fluid changes, and then only if the temperature change increases.
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.
For this to happen, the RCS temperature for T hot, affecting insurge/outsurge, and T cold, affecting the spray temperature, would have to decrease from the analyzed condition.
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.
The Table EMCB R9-1 provides a comparison showing the temperature change across the pressurizer components evaluated for the design basis conditions, the current operating conditions, and at MUR power uprate conditions.
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.
It is seen from Table EMCB R9-1 that the temperature change for T hot, affecting the lower pressurizer (L1 T hot), is less at MUR power uprate conditions by 0.6 OF and is enveloped by the analysis of record (AOR). The temperature differential for the upper portion of the pressurizer is shown to exceed the current operating condition by 0.6 OF (L1 T cold). This is an increase of approximately 0.5% over the current operating condition L1 T cold and is not considered to be significant.
associated control      cabinets.The   The   informationprovided information          providedreflects reflectsthe theconfiguration configurationfollowing followingthe  the implementation of the UPS modification.
Also, since the baseline analysis, which is also the AOR, continues to envelope the MUR power uprate temperature differential, the AOR is not affected and remains applicable.
modification. All      All power supplies      are Electrical supplies are     Electrical Class Class11E. E.
Therefore, there is no change to the baseline analysis results due to the MUR power uprate resulting from changes to the RCS temperatures affecting the pressurizer.
An assessment of the pressurizer surge, spray, safety and relief nozzle for structural weld overlay (SWOL) was also performed as part of the MUR power uprate. The assessment concluded that the MUR power uprate would have no impact on the AOR for these components based on the findings previously noted. Therefore, the MUR power uprate is enveloped by the current SWOL analysis and is acceptable.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 11 NON-PROPRIETARY Table EMCB R9-1:Comparison of Byron
/Braidwood Pressurizer Analysis Basis Baseline Current MUR Parameter Analysis Operating Operating (AOR)Conditions Conditions
(&deg;F)(&deg;F)(&deg;F)Tpressurizer 652.7 652.7 652.7 Thot 542.7 608 608.6 Tcad517.7542 541.4 AThot = Tpressurizer - Thot 110 44.7 44.1 ATcold = Tpressurizer -
135 110 7 111 3 Tcold..NRC/EMCB Request 10 discusses the evaluation of the reactor vessel internal components for flow induced vibration (FIV) impact under MUR power uprate conditions.
Also, Section IV.1.A.ii.e of Attachment 7 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 those design parameters, before and after MUR power uprate, which could potentially influence FIV response of the reactor internals. Also, discuss the comparison of alternating stress intensities to design basis allowable limits for the most critical components demonstrating compliance with the Byron and Braidwood design basis acceptance criteria.
Response Comparisons of flow induced vibration (FIV) design parameters before and after the MUR power uprate are provided in Table EMCB RIO-1.
Table EMCB R10-1: Comparison of FIV Evaluation Input Design Parameters Parameter Mechanical Design Flow (gpm/loop)Vessel Inlet Temperature (&deg;F) /
Fluid Density (ibm/ft3)Vessel Outlet Temperature (&deg;F) /
Fluid Density (Ibm/ft)
Current Analysis of Record 107,000542/47.369608/42.4535 MUR Power Uprate 107,000541.4/47.385608.6/42.411 Ratio 1.0-1.0-1.0 The MUR power uprate design conditions will slightly alter the Toad and Thot fluid densities, which will slightly change the forces, induced by flow.The corresponding Tcord and Thot fluid densities change by less than 0.1 % from the current analyzed condition. Therefore, the effect on the flow-induced vibration stresses (alternating stress intensities) due to MUR power uprate on the reactor internals remains unchanged from the current analysis of record.
Table EMCB R9-1: Parameter T pressurizer That T cold 11 That = T pressurizer
-That 11 T cold = T pressurizer
-Teold NRCIEMCB Request 10 Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 11 NON-PROPRIETARY Comparison of Byron/Braidwood Pressurizer Analysis Basis Baseline Current MUR Analysis Operating Operating (AOR) Conditions Conditions (OF) (OF) (OF) 652.7 652.7 652.7 542.7 608 608.6 517.7 542 541.4 110 44.7 44.1 135 110.7 111.3 Section IV.1.B.iii of Attachment 7 to the LAR discusses the evaluation of the reactor vessel internal components for flow induced vibration (FIV) impact under MUR power uprate conditions.
Also, Section IV.1.A.ii.e of Attachment 7 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 those design parameters, before and after MUR power uprate, which could potentially influence FIV response of the reactor internals.
Also, discuss the comparison of alternating stress intensities to design basis allowable limits for the most critical components demonstrating compliance with the Byron and Braidwood design basis acceptance criteria.
Response Comparisons of flow induced vibration (FIV) design parameters before and after the MUR power uprate are provided in Table EMCB R10-1. Table EMCB R10-1: Comparison of FIV Evaluation Input Design Parameters Parameter Current Analysis MUR Power Ratio of Record Uprate Mechanical Design Flow 107,000 107,000 1.0 (gpmlloop)
Vessel Inlet Temperature CF) 1 5421 541.41 -1.0 Fluid Density (Ibm/fe) 47.369 47.385 Vessel Outlet Temperature (OF) 1 6081 608.61 -1.0 Fluid Density (Ibm/fe) 42.4535 42.411 The MUR power uprate design conditions will slightly alter the T cold and T hot fluid densities, which will slightly change the forces, induced by flow. The corresponding T cold and That fluid densities change by less than 0.1 % from the current analyzed condition.
Therefore, the effect on the flow-induced vibration stresses (alternating stress intensities) due to MUR power uprate on the reactor internals remains unchanged from the current analysis of record.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 12 NON-PROPRIETARY NRC/EMCB Request 11 Discuss further information and confirm that the nuclear steam supply system component supports, as discussed in Section 3.9.3.4 of the Byron and Braidwood UFSAR, will continue to be in compliance with the Byron and Braidwood design basis acceptance criteria at the proposed MUR power uprate conditions.
Also, confirm that the operating temperatures for support elements, as defined in Table 3.9-17 of the Byron and Braidwood UFSAR, are not affected by the MUR power uprate.
Response The NSSS component supports, which include the reactor vessel, steam generator, reactor coolant pump, and pressurizer equipment supports, were assessed for the MUR power uprate as discussed in the response to EMCB R-7 and were shown to remain acceptable and bounded by the current design basis. Therefore, the NSSS component supports will remain in compliance with UFSAR Section 3.9.3.4.
The operating temperatures of the supports, as outlined in Table 3.9-17 of the UFSAR, are not affected by the MUR power uprate. The MUR power uprate does not require an increase in the ambient containment temperature design value. Further, the small changes to the NSSS design temperatures, as discussed in the response to EMCB R-7, do not require a change to the operating temperature of the supports attached to the steam generator, reactor coolant pump, reactor vessel, or pressurizer.
NRC/EMCB Request 12 Section IV.1.A.vi.1.b of Attachment 7 to the LAR discusses the structural evaluation of Byron and Braidwood Unit I replacement steam generators and states that a reconciliation analysis was performed to address the structural integrity of the entire steam generator pressure boundary for the MUR power uprate conditions. Discuss further information relative to, before and after uprate, the maximum stress intensity and the cumulative fatigue usage factors for the critical components of the primary and secondary sides, including nozzles, of the replacement steam generators and the respective service conditions. Also, confirm that the reconciliation analysis was performed in accordance with the original design code of record and in compliance with the Byron and Braidwood stations design basis acceptance criteria.
Response During the structural integrity analysis of the replacement steam generators (RSGs) on Unit 1 for MUR conditions it was concluded that the maximum primary and secondary side temperatures and pressures specified for MUR power uprate conditions were less than the primary and secondary side temperatures and pressures specified for the original analysis.
Therefore, there are no changes to the calculated stress values or limits for design conditions (i.e., name plate conditions).
However, a reconciliation analysis was performed for critical components of the replacement steam generators due to differences in the Level A & B (Normal and Upset), Level C (Emergency) and Level D (Faulted) condition loads. The stress intensities and cumulative usage factors for these service conditions for pre-MUR and post-MUR power uprate conditions are included in Tables EMCB R12-1 though R12-4.
The reconciliation analysis was performed in accordance with the original design code of record as required by the current Certified Design Specification. Specifically, the acceptance criteria NRCIEMCB Request 11 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 12 NON*PROPRIETARY Discuss further information and confirm that the nuclear steam supply system component supports, as discussed in Section 3.9.3.4 of the Byron and Braidwood UFSAR, will continue to be in compliance with the Byron and Braidwood design basis acceptance criteria at the proposed MUR power uprate conditions.
Also, confirm that the operating temperatures for support elements, as defined in Table 3.9-17 of the Byron and Braidwood UFSAR, are not affected by the MUR power uprate. Response The NSSS component supports, which include the reactor vessel, steam generator, reactor coolant pump, and pressurizer equipment supports, were assessed for the MUR power uprate as discussed in the response to EMCB R-7 and were shown to remain acceptable and bounded by the current design basis. Therefore, the NSSS component supports will remain in compliance with UFSAR Section 3.9.3.4. The operating temperatures of the supports, as outlined in Table 3.9-17 of the UFSAR, are not affected by the MUR power uprate. The MUR power uprate does not require an increase in the ambient containment temperature design value. Further, the small changes to the NSSS design temperatures, as discussed in the response to EMCB R-7, do not require a change to the operating temperature of the supports attached to the steam generator, reactor coolant pump, reactor vessel, or pressurizer.
NRCIEMCB Request 12 Section IV.1.A. vi.1.b of Attachment 7 to the LAR discusses the structural evaluation of Byron and Braidwood Unit 1 replacement steam generators and states that a reconciliation analYSis was performed to address the structural integrity of the entire steam generator pressure boundary for the MUR power uprate conditions.
Discuss further information relative to, before and after uprate, the maximum stress intenSity and the cumulative fatigue usage factors for the critical components of the primary and secondary sides, including nozzles, of the replacement steam generators and the respective service conditions.
Also, confirm that the reconciliation analysis was performed in accordance with the original design code of record and in compliance with the Byron and Braidwood stations design basis acceptance criteria.
Response During the structural integrity analysis of the replacement steam generators (RSGs) on Unit 1 for MUR conditions it was concluded that the maximum primary and secondary side temperatures and pressures specified for MUR power uprate conditions were less than the primary and secondary side temperatures and pressures specified for the original analysis.
Therefore, there are no changes to the calculated stress values or limits for design conditions (i.e., name plate conditions).
However, a reconciliation analysis was performed for critical components of the replacement steam generators due to differences in the Level A & B (Normal and Upset), Level C (Emergency) and Level D (Faulted) condition loads. The stress intensities and cumulative usage factors for these service conditions for pre-MUR and post-MUR power uprate conditions are included in Tables EMCB R12-1 though R12-4. The reconciliation analysis was performed in accordance with the original design code of record as required by the current Certified Design Specification.
Specifically, the acceptance criteria Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 13 NON-PROPRIETARY for the reconciliation of the pressure boundary components were those specified in the 1986 ASME B&PV Code with no Addenda, for Section III, Class 1 components. The Code acceptance criteria are unchanged from the original RSG analysis.
Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 13 NON-PROPRIETARY for the reconciliation of the pressure boundary components were those specified in the 1986 ASME B&PV Code with no Addenda, for Section III, Class 1 components.
The Code acceptance criteria are unchanged from the original RSG analysis.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 14 NON-PROPRIETARY Table EMCB R12-1: Stress Intensity (SI) and Fatigue Usage Factors (FUF) for Level A &
B Conditions MUR SI Range (ksi)Orig.SI Range (ksi)MUR SI Limit (ksi)Orig.SI Limit (ksi)Component I Location MUR FUF FUF Limit Orig.FUF Tubesheet Primary Head I Tubesheet Juncture 38.5*82.1 80.1 87.3 0.880 0.741 1.0 Secondary Shell / Tubesheet Juncture 86.4 85.4 95.087.30.160 0.223 1.0 Tubesheet Perforated Region 90.190.095.0 93.6 0.330 0.387 1.0 Primary Nozzle Primary nozzle67.8567.85 80.1 80.1 0.839 0.839 1.0 Primary nozzle safe end 57.37 57.3760.360.30.096 0.096 1.0 Primary Manway Cover 30.3 30.3 80.1 80.1 0.006 0.006 1.0 Shell/flange 46 46.0 80.1 80.1 0.121 0.121 1.0 See Table EMCB R12-4 for Average and Maximum Bolt Stresses Primary Head Support Pad 79.4 79.4 80 80 0.67 0.67 11.0 Primary Divider Plate 63.9 63.969.969.9 0.905 0.904 11.0 Small Nozzles
%" Nozzles 13.96 11.8326.726.7 0.81 0.679 1.0 Steam Drum
/Cone/Lower Shell Assembly74.2262.9 80.1 80.1 0.025 0.021 1.0 Stud 1.0 0.871 0.871 NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 14 Table EMCB R12*1: Stress Intensity (SI) and Fatigue Usage Factors (FUF) for Level A & B Conditions MUR Orlg. MUR Orlg. MUR Orlg. FUF Component f Location SIRange SI Range SI Limit SI Limit (ksi) (ksi) (ksi) (ksi) FUF FUF limit Tubesheet Primary Head I Tubesheet Juncture 38.5* 82.1 80.1 87.3 0.880 0.741 1.0 Secondary Shell I Tubesheet Juncture 86.4 85.4 95.0 87.3 0.160 0.223 1.0 Tubesheet Perforated Region 90.1 90.0 95.0 93.6 0.330 0.387 1.0 Primary Nozzle Primary nozzle 67.85 67.85 80.1 80.1 0.839 0.839 1.0 Primary nozzle safe end 57.37 57.37 60.3 60.3 0.096 0.096 1.0 Primary Manway Cover 30.3 30.3 80.1 80.1 0.006 0.006 1.0 Shelilflange 46 46.0 80.1 80.1 0.121 0.121 1.0 Stud See Table EMCB R12-4 for Average and 0.871 0.871 1.0 Maximum Bolt Stresses Primary Head Support Pad 79.4 79.4 80 80 0.67 0.67


===1.0 Primary===
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:
Divider Plate 63.9 63.9 69.9 69.9 0.905 0.904 1.0 Small Nozzles %" Nozzles 13.96 11.83 26.7 26.7 0.81 0.679 1.0 Steam Drum/ConefLower Shell Assembly 74.22 62.9 80.1 80.1 0.025 0.021 1.0 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 15 NON-PROPRIETARY Table EMCB R12-1: Stress Intensity (SI) and Fatigue Usage Factors (FUF) for Level A & B Conditions MUR SI Range (ksi)Orig.SI Range (ksi)MUR SI Limit (ksi)Orig.SI Limit (ksi)Component/Location Orig.FUF MUR FUF FUF Limit 8" Shell Cone Handhole Shell/cover/flange 80 67.3 57 80 1.0 0.074 0.256 Stud See Table EMCB R12-4 for Average and Maximum Bolt Stresses 1.0 0.987 0.975 6" Feedring Handhole Shell/cover/flange 80 78.0 80 76.5 1.0 0.823 0.374 Stud See Table EMCB R12-4 for Average and Maximum Bolt Stresses 0.823 2" Inspection Port Shell/cover/flange 80 77.6 80 65.8 0.214 Stud See Table EMCB R12-4 for Average and Maximum Bolt Stresses 0.864 1.0 0.84 1.0 0.205 1.0 0.807 Secondary Manway Flange/Steam Drum Head Diaphragm Cover 80 55.2 80 46.8 1.0 0.019 0.02 1.0 60.4 69.9 69.9 60.4 80 25.5 80 21.6 0.02 0.015 1.0 0.02 0.000 Stud See Table EMCB R12-4 for Average and Maximum Bolt Stresses 1.0 0.973 0.752 NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 15 Table EMCB R12-1: Stress Intensity (51) and Fatigue Usage Factors (FUF) for Level A & B Conditions MUR Orig. MUR Orig. MUR Orlg. FUF Component I Location SIRange SIRange SI Limit SI Limit (ksi) (ksi) (ksi) (ksi)
Table        R2 -1:   SG PORV SG  PORV Power Supplies Control Panel Control              Primary Power Primary SG PORV SG                                                      Backup Power Supply Backup (Controllers)
FUF FUF Limit 8" Shell Cone Handhole Shell/cover/flange 67.3 57 80 80 0.256 0.074 1.0 Stud See Table EMCB R12-4 for Average and 0.987 0.975
(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


===1.0 Maximum===
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.
Bolt Stresses 6" Feedring Handhole Shell/cover/flange 78.0 76.5 80 80 0.823 0.374 1.0 Stud See Table EMCB R12-4 for Average and 0.823 0.84 1.0 Maximum Bolt Stresses 2" Inspection Port Shell/cover/flange 77.6 65.8 80 80 0.214 0.205 1.0 Stud See Table EMCB R12-4 for Average and 0.864 0.807 1.0 I Maximum Bolt Stresses Secondary Manway Flange/Steam Drum Head 55.2 46.8 80 80 0.02 0.019 1.0 Diaphragm 60.4 60.4 69.9 69.9 0.02 0.015 1.0 Cover 25.5 21.6 80 80 0.02 0.000 1.0 Stud See Table EMCB R12-4 for Average and 0.973 0.752 1.0 Maximum Bolt Stresses Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 16 NON-PROPRIETARY Table EMCB R12-1: Stress Intensity (SI) and Fatigue Usage Factors (FUF) for Level A & B Conditions MUR SI Range (ksi)44.6*41.2*48.6 64.8*72 40.6 54.7 70.6 MUR FUF 0.538 0.74 0.401 0.688 0.608 0.652 0.652 0.021 Orig.FUF 0.476 0.559 0.209 0.239 0.546 0.545 0.545 0.010 Steam Drum Head/Steam Drum Juncture Primary Deck Lug/Steam Drum Juncture Shroud Lug Shroud Lug/ Shell Juncture Upper Vessel Support/ Steam Drum Juncture Orig.SI Range (ksi)44.6*41.2*48.6 80 61 34.4 46.4 59.8 MUR SI Limit (ksi)56.1 56.1 80.1 80.0 80 58.5 80 80.1 Orig.SI Limit (ksi)56.1 56.1 80.1 80.0 80 58.5 80.1 80.1 FUF Limit.1.01.01.01.01.01.0 1.0 1.0 Main Feedwater Nozzle Shell/nozzle juncture Nozzle Transition ring/Thermal sleeve 77.6 69.3 27.2*77.6 58.7 27.2*80 80 69.9 80 80 69.9 0.408 0.046 0.985 0.346 0.039 0.9451.01.0 1.01.01.01.0 1.0 0 26.8 70 22.7 70 80 69.5 80 58.9 0.035 0.048 80 71.3 80 60.4 0.033 0.049 80 76.7 65 80 0.080 0:059 NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 16 Table EMCB R12-1: Stress Intensity (51) and Fatigue Usage Factors (FUF) for Level A & B Conditions MUR Orig. MUR Orig. MUR Orig. FUF Component I Location SIRange SI Range SI Limit SI Limit (ksi) (ksi) (ksi) (ksi)
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:
FUF FUF Limit Pressure Boundary Attachments Seal Skirt Transition Juncture 44.6* 44.6* 56.1 56.1 0.538 0.476 1.0 Skirt Weld 41.2* 41.2* 56.1 56.1 0.74 0.559 1.0 Steam Drum Head/Steam Drum Juncture 48.6 48.6 80.1 80.1 0.401 0.209 1.0 Steam Drum / Trunion Juncture 64.8* 80 80.0 80.0 0.688 0.239 1.0 I Primary Deck Lug/Steam Drum Juncture 72 61 80 80 0.608 0.546 1.0 Shroud Lug 40.6 34.4 58.5 58.5 0.652 0.545
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.


===1.0 Shroud===
===Response===
Lug/ Shell Juncture 54.7 46.4 80 80.1 0.652 0.545 1.0 Upper Vessel Support! Steam Drum 70.6 59.8 80.1 80.1 0.021 0.010 1.0 I Juncture Main Feedwater Nozzle Shell/nozzle juncture 77.6 77.6 80 80 0.408 0.346
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,


===1.0 Nozzle===
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.
69.3 58.7 80 80 0.046 0.039
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.


===1.0 Transition===
===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


ringlThermal sleeve 27.2* 27.2* 69.9 69.9 0.985 0.945 1.0 Steam Outlet Nozzle Nozzle/Safe End Juncture 26.8 22.7 70 70 0 0 1.0 Nozzle 69.5 58.9 80 80 0.048 0.035 1.0 Steam Drum Head 71.3 60.4 80 80 0.049 0.033
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.


===1.0 Perforated===
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.


Zone 76.7 65 80 80 0.080 0:059 1.0 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 17 NON-PROPRIETARY Table EMCB R12-1: Stress Intensity (SI) and Fatigue Usage Factors (FUF) for Level A & B Conditions MUR SI Range (ksi)Orig.SI Range (ksi)MUR SI Limit (ksi)Orig.SI Limit (ksi)Component I Location Orig.FUF MUR FUF FUF Limit 10.19 26.7 15 26.7 11.83 26.7 46.3 56 73.8 79.8 Small Nozzles 3" Blowdown Nozzle 3" Recirculation Nozzle 3/" Nozzles Acoustic Sensor Pad Tubes 12.02 12.02 13.96 54.63 73.826.71.026.71.026.71.0560.811.0 0.85 0.5 0.81 0.928 0.938 0.679 0.777 79.80.190.191.0*BoidAtaiicized stress range values were determined using simplified elastic-plastic analysis in accordance with NB-3228.5.
===Response===
NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 17 Table EMCB R12-1: Stress Intensity (51) and Fatigue Usage Factors (FUF) for Level A & B Conditions MUR Orig. MUR Orig. MUR Orig. FUF Component I Location SIRange SIRange SI Limit SI Limit (ksi) (ksi) (ksi) (ksi) FUF FUF Limit Small Nozzles 3" Blowdown Nozzle 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 0.938 1.0 %" Nozzles 13.96 11.83 26.7 26.7 0.81 0.679 1.0 Acoustic Sensor Pad 54.63 46.3 56 56 0.81 0.777 1.0 Tubes 73.8 73.8 79.8 79.8 0.19 0.19 1.0 ------* Bold/Italicized stress range values were determined using simplified elastic-plastic analysis in accordance with NB-3228.5.
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)
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 18 NON-PROPRIETARY Table EMCB R12 Primary Membrane and Bending Stresses for Level C Conditions Component/Location MUR Pm/PL SI (ksi)Orig.Pm/PL SI (ksi)MURPm Sl Limit (ksi)Orig.Pm SI Limit (ksi)MUR PL/Pm+Pb SI (ksi)Orig.PL/Pm+Pb SI (ksi)MUR PL Pm+Pb SI Limit (ksi)Orig. PL Pm+Pb SI Limit (ksi)
(AFW) temperature to be the conservative assumptions for the steam generator tube rupture margin to overfill (MTO) analysis.
Secondary shell Primary Nozzle Primary Manway 29.6 29.2 38.79 38.79 49.9 49.2 Bounded by design conditions 64.65 64.65 Cover Shell/flange 24.39 21.31 38.8 13.33 13.33 38.8 38.8 21.31 21.31 38.8 58.2 24.39 58.2 58.2 21.31 58.2 Primary Head Support Pad Primary Divider Plate Small NozzlesSteam Drum
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.
/Cone/Lower Shell Assembly Bounded by design conditions Bounded by design conditions Bounded by design conditions Bounded by design conditions 8" Shell Cone Handhole 6" Feedring Handhole 2" Inspection Port 29.3 29.3 10.6 29.02 29.02 10.5 29.37 29.37 28 29.37 29.37 28 32.6 34.6 20.7 32.2 34.6 20.5 48.06 48.06 42 48.06 48.06 42 Secondary ManwayBounded by design conditions NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 18 Table EMCB R12-2 -Primary Membrane and Bending Stresses for Level C Conditions MUR Orig. MUR Orig. MUR Orig.PL/ MURPL Orig. PL PmSI PmSI PLlPm+ Pm+PbSI Pm+PbSI Component
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.
/ Location Pm/PL SI Pm/PL SI Limit Limit PbSI Pm+PbSI Limit Limit (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) Primary Head / Tubesheet
conservative.
/ 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 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 6" Feedring Handhole 29.3 29.02 29.37 29.37 34.6 34.6 48.06 48.06 2" Inspection Port 10.6 10.5 28 28 20.7 20.5 42 42 Secondary Manway Bounded by design conditions Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 19 NON-PROPRIETARY Table EMCB R12 Primary Membrane and Bending Stresses for Level C Conditions Component/Location MUR Pm/PL SI (ksi)Orig.Pm/PL SI (ksi)MUR Pm SI Limit (ksi)Orig.Pm SI Limit (ksi)MUR PL/Pm+Pb SI (ksi)Orig.PL/Pm+Pb SI (ksi)MUR PL Pm+Pb SI Limit (ksi)Orig. PL Pm+Pb SI Limit (ksi)Pressure Boundary Attachments Seal Skirt Transition Juncture 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 35.6 65.7 65.7 Primary Deck Lug/Steam Drum Juncture Shroud Lug 2.32 2.3 26.37 26.37 5.8 5.73 43.95 43.95 Shroud Lug/ Shell Juncture Upper Vessel Support/ Steam Drum Juncture Main Feedwater Nozzle Shell/nozzle juncture 29 28.7 43.8 43.8 46.6 46.1 65.7 65.7 Nozzle 28.6 28.3 43.8 43.8 28.6 28.3 65.7 65.7 Transition ring/Thermal sleeve 9.5 9.4 28 28 26.1 25.8 41.9 41.9 Steam Outlet Nozzle Bounded by design conditions Small Nozzles Bounded by design conditions Tubes 22.95 22.7 35.2 35.2 32.35 32 52.9 52.9 Tubes (external pressure)0.1680.166 1.424 1.424 65.2 29.8 29.8 65.7 43.8 43.8 65.2 65.7 24.9 24.63 36.9 36.9 26.9 26.61 65.7 65.7 Bounded by design conditions NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 19 Table EMCB R12-2 -Primary Membrane and Bending Stresses for Level C Conditions MUR Orig. MUR Orig. MUR Orig.PLI MURPL Orig. PL PmSI PmSI PLlPm+ Pm+PbSI Pm+Pb SI Component I Location Pm/PLSI Pm/PLSI Limit Limit PbSI Pm+PbSI Limit Limit (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi)
Pressure Boundary Attachments Seal Skirt Transition Juncture Bounded by design conditions Skirt Weld Bounded by design conditions Steam Drum Head/Steam Drum Bounded by design conditions Juncture Steam Drum / Trunion Juncture 28.8 28.5 39.4 39.4 36 35.6 65.7 65.7 Primary Deck Lug/Steam Drum 29.8 29.8 43.8 43.8 65.2 65.2 65.7 65.7 Juncture Shroud Lug 2.32 2.3 26.37 26.37 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! Steam Drum Bounded by design conditions Juncture Main Feedwater Nozzle Shell/nozzle juncture 29 28.7 43.8 43.8 46.6 46.1 65.7 65.7 Nozzle 28.6 28.3 43.8 43.8 28.6 28.3 65.7 65.7 Transition ringlThermal sleeve 9.5 9.4 28 28 26.1 25.8 41.9 41.9 Steam Outlet Nozzle Bounded by design conditions Small Nozzles Bounded by design conditions Tubes 22.95 22.7 35.2 35.2 32.35 32 52.9 52.9 , Tubes (external pressure) 0.168 0.166 1.424 1.424
--------'----....
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 20 NON-PROPRIETARY Table EMCB R12-3: Primary Membrane and Bending Stresses for Level D Conditions Component/Location MUR Pm/PL SI (ksi)Orig.Pm/PL SI (ksi)MUR Pm SI Limit (ksi)Orig.Pm SI Limit (ksi)MUR PUPm+Pb SI (ksi)Orig.PU Pm+Pb SI (ksi)MUR PL Pm+Pb SI Limit (ksi)Orig. PL Pm+Pb SI Limit (ksi)Primary Head/ Tubesheet /
Secondary Shell Primary Nozzle Primary nozzle 51.51 51.51 56 56 76.27 76.27 84 84 Primary nozzle safe end 27.7 27.7 48.9 48.9 39.93 39.93 72.36 72.36 Primary Manway Cover 13.33 13.33 56 56 24.39 24.39 84 84 Shell/flange 21.31 21.31 56 56 21.31 21.31 84 84 Primary Head Support Pad 15.9 15.9 56 56 53.9 53.9 84 84 Primary Divider Plate 35.9 35.452.552.5 61.2**60.4 67.5 67.5 Small Nozzles 3 " Nozzles 16.9 16.7 42.8 42.8 38 37.5 64 64Steam Drum/Cone
/Lower Shell Assembly 8" Shell Cone Handhole 40.9 40.4 56 56 40.9 40.4 84 84 6" Feedring Handhole 35.3 34.8 56 56 35.3 34.8 84 84 2" Inspection Port 55.1 54.4 56 56 60.9 60.1 84 84 Secondary Manway 47.7 47.1 56 56 47.7 47.1 84 84 56 29.6 29.1 56 84 67.8 68.7 84 56 46.2 56 35.6 84 61.5 60.7 84 NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 20 Table EMCB R12-3: Primary Membrane and Bending Stresses for Level D Conditions MUR Orig. MUR Orig. MUR Orig.PU MURPL Orig. PL PmSI PmSI PUPm+ Pm+Pb SI Pm+PbSI Component
/ Location PmlPLSI Pm/PL SI Limit Limit PbSI Pm+PbSI Limit Limit (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi)
Primary Head / Tubesheet
/ 29.6 29.1 56 56 68.7 67.8 84 84 Secondary Shell Primary Nozzle Primary nozzle 51.51 51.51 56 56 76.27 76.27 84 84 Primary nozzle safe end 27.7 27.7 48.9 48.9 39.93 39.93 72.36 72.36 Primary Manway Cover 13.33 13.33 56 56 24.39 24.39 84 84 Shell/flange 21.31 21.31 56 56 21.31 21.31 84 84 Primary Head Support Pad 15.9 15.9 56 56 53.9 53.9 84 84 Primary Divider Plate 35.9 35.4 52.5 52.5 61.2** 60.4 67.5 67.5 Small Nozzles %" Nozzles 16.9 16.7 42.8 42.8 38 37.5 64 64 Steam DrumlCone/Lower Shell 46.2 35.6 56 56 61.5 60.7 84 84 Assembly 8" Shell Cone Handhole 40.9 40.4 56 56 40.9 40.4 84 84 6" Feedring Handhole 35.3 34.8 56 56 35.3 34.8 84 84 2" Inspection Port 55.1 54.4 56 56 60.9 60.1 84 84 Secondary Manway 47.7 47.1 56 56 47.7 47.1 84 84 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 21 NON-PROPRIETARY Table EMCB R12-3: Primary Membrane and Bending Stresses for Level D Conditions Component/Location MUR Pm/PL SI (ksi)Orig.Pm/PL SI (ksi)MUR Pm SI Limit (ksi)Orig.Pm SI Limit (ksi)MUR PL/Pm+Pb SI (ksi)Orig.PLI Pm+Pb SI (ksi)MUR PL Pm+Pb SI Limit (ksi)Orig. PL Pm+Pb SI Limit (ksi)Pressure Boundary Attachments Seal Skirt Transition Juncture 8.81 8.7 49 49 26.826.573.5 73.5 Steam Drum Head/Steam Drum Juncture Steam Drum / Trunnion Juncture 35.8 35.3 56 56 42.7 42.2 84 84 Primary Deck Lug/Steam Drum Juncture Shroud Lug 39 38.5 49 49 43.4 42.973.573.5 Shroud Lug/ Shell Juncture 33.5 33.1 56 56 39.5 39.1 84 84 Upper Vessel Support/ Steam Drum Juncture Main Feedwater Nozzle Shell/nozzle juncture 33.9 33.5 56 56 83.8 83.3 84 84 Nozzle 7.9 7.8 56 56 29.5 29.1 84 84 Transition ring/Thermal sleeve 12.9 12.7 49 49 53 52.3 73.5 73.5 Steam Outlet Nozzle Pipe extension16.6816.47 42 42 43.38 42.82 63 63 Nozzle/Safe End Juncture 14.99 14.8 49 49 40.84 40.3273.573.5 Nozzle 25.43 25.1 56 56 54.82 54.12 84 84 Steam Drum Head 26.34 26 56 56 55.84 55.12 84 84 Perforated Zone 32.62 32.2 56 56 61 60.22 84 84 5635.3534.9 56 84 45.5 46.1 84 74 40.1 40.1 56 74 56 84 84 28 27.6 56 56 63.9 63.1 84 84 NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 21 Table EMCB R12-3: Primary Membrane and Bending Stresses for Level D Conditions MUR Orig. MUR Orig. MUR Orig.PU MURPL Orig. PL PmSI PmSI PUPm+ Pm+PbSI Pm+PbSI Component I Location PmlPL SI Pm/PL SI Limit Limit PbSI Pm+PbSI 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 Drum 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 Upper Vessel Support! Steam Drum 28 27.6 56 56 63.9 63.1 84 84 Juncture Main Feedwater Nozzle Shell/nozzle juncture 33.9 33.5 56 56 83.8 83.3 84 84 Nozzle 7.9 7.8 56 56 29.5 29.1 84 84 Transition ringlThermal sleeve 12.9 12.7 49 49 53 52.3 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 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 Head 26.34 26 56 56 55.84 55.12 84 84 Perforated Zone 32.62 32.2 56 56 61 60.22 84 84 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 22 NON-PROPRIETARY Table EMCB R12-3: Primary Membrane and Bending Stresses for Level D Conditions Orig.Pm/PL SI (ksi)MUR Pm SI Limit (ksi)Orig.Pm SI Limit (ksi)MUR PUPm+Pb SI (ksi)Orig.PL/Pm+Pb SI (ksi)68.68 67.8 30.6 30.6 42.8 42.8 42.8 42.8 42.1 42.1 61.2 61.2 Bounded by Steam Drum/Cone/Lower Shell Assembly Component/Location Small Nozzles 3" Blowdown Nozzle


3" Recirculation Nozzle Acoustic Sensor Pad Tubes MUR Pm/PL SI (ksi)21.1 21.1 31.4 MUR PL Pm+Pb SI Limit (ksi)64 64 84 Orig. PL Pm+Pb SI Limit (ksi)64 64 84 31 56 56 1.780 1.127 1.780 Tubes (external pressure) 1.142 A prorating factor corresponding to the SG secondary side level D loading has been applied to the Divider Plate MUR PUPm+Pb SI, making the reported value conservative. However, only primary stresses from divider plate level D loads need to be analyzed and since the primary side pressures are invariant between MUR and Original conditions, both the level D stresses and their ASME Code limits are unchanged.
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*
NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 22 Table EMCB R12-3: Primary Membrane and Bending Stresses for Level D Conditions , MUR Orig. MUR Orig. MUR Orig.PU MURPL Orig. PL I PmSI PmSI PUPm+ Pm+Pb SI Pm+PbSI Component I Location PmlPL SI Pm/PL SI Limit Limit PbSI Pm+PbSI Limit Limit (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi) (ksi)
Case                                    Description                                (ft3)
Small Nozzles 3" Blowdown Nozzle 21.1 30.6 42.8 42.8 42.1 61.2 64 64 3" Recirculation Nozzle 21.1 30.6 42.8 42.8 42.1 61.2 64 64 Acoustic Sensor Pad Bounded by Steam Drum/Cone/Lower Shell Assembly Tubes 31.4 31 56 56 68.68 67.8 84 84 Tubes (external pressure) 1.142 1.127 1.780 1.780 --------** A prorating factor corresponding to the SG secondary side level D loading has been applied to the Divider Plate MUR PUPm+Pb SI, making the reported value conservative.
(ft) 1          Low Tang, Low  T avg, 5% tube plugging, plugging, ANS 1971 + 20%,
However, only primary stresses from divider plate level D loads need to be analyzed and since the primary side pressures are invariant between MUR and Original conditions, both the level D stresses and their ASME Code limits are unchanged.
1                                                                                 +321 maximum AFW enthalpy 2          Low Tavg, Low  T avg, 0% tube plugging, plugging, ANS 1971 + 20%,
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 23 NON-PROPRIETARY Table EMCB R12-4: Average and Maximum Stresses for Studs/Bolts Component/Location MUR Average Stress (ksi)Orig.Average Stress (ksi)MUR Average Stress Limit (ksi)Orig.Average Stress Limit (ksi)MUR Maximum Stress (ksi)Orig.Maximum Stress (ksi)MUR PL Maximum Stress Limit (ksi)Orig.Maximum Stress Limit (ksi)Primary Manway Level A/B43.8 43.8 54.6 54.6 55.4 55.4 81.9 81.9 Level C34.7 34.752.652.6 76 76 78.9 78.9 Level D 34.7 34.7 87.5 87.5 76 76 125 125 8" Shell Cone Handhole Level A/B13.513.2557.7 57.7 49.7 48.777.977.9 Level C41 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 6" Feedring Handhole Level A/B41.4 40.6 57.7 57.7 69 67.677.977.9 Level C36.7 36.3 57.7 57.7 55.254.686.7 86.7 Level D 36.8 36.3 86.2 86.2 55.6 54.9 125 125 2" Inspection Ports Level A/B40.9 40.157.757.7 52.8 51.877.977.9 Level C41.1 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 Component I Location Manwa3l Level AlB Level C Level D 8" Shell Cone Handhole Level AlB LevelC Level D 6" Feedring Handhole Level AlB LevelC Level D 2" Ports Level AlB Level C Level D NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 23 Table EMCB R12-4: Average and Maximum Stresses for Studs/Bolts MUR Orig. MUR Orig. MUR Orig. MURPL Orig. Average Average Average Average Maximum Maximum Maximum Maximum Stress Stress Stress Stress Stress Stress Stress Stress (ksi) (ksi) Limit Limit (ksi) (ksi) Limit Limit (ksi) (ksi) (ksi) (ksi) 43.8 43.8 54.6 54.6 55.4 55.4 81.9 81.9 34.7 34.7 52.6 52.6 76 76 78.9 78.9 34.7 34.7 87.5 87.5 76 76 125 125 13.5 13.25 57.7 57.7 49.7 48.7 77.9 77.9 41 40.54 57.7 57.7 78.6 77.73 86.7 86.7 41.1 40.54 86.2 86.2 79.7 78.66 125 125 41.4 40.6 57.7 57.7 69 67.6 77.9 77.9 36.7 36.3 57.7 57.7 55.2 54.6 86.7 86.7 36.8 36.3 86.2 86.2 55.6 54.9 125 125 40.9 40.1 57.7 57.7 52.8 51.8 77.9 77.9 41.1 40.6 57.7 57.7 47.4 46.9 86.7 86.7 41 40.5 86.2 86.2 47.1 46.5 125 125 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 24 NON-PROPRIETARY Table EMCB R12-4: Average and Maximum Stresses for Studs/Bolts Component/Location Secondary ManwayLevel A/B Level C Level D MUR Average Stress (ksi)47.8 32.1 30.4 Orig.Average Stress (ksi)40.5 31.8 30.0 MUR Average Stress Limit (ksi)57.7 57.7 86.2 Orig.Average Stress Limit (ksi)57.7 57.7 86.2 MUR Maximum Stress (ksi)72.1 58.5 44.7 Orig.Maximum Stress (ksi)61.1 57.9 44.1 MUR PL Maximum Stress Limit (ksi)77.9 77.9 125 Orig.Maximum Stress Limit (ksi)77.9 77.9 125 Component I Location SecondarY Manwa)l Level AlB Levele Level D ,
2                                                                                 +314 maximum AFW enthalpy 3         High Tang, High  T avg, 5% tube plugging, plugging, ANS 1971 + 20%,
NON-PROPRIETARY Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 24 Table EMCB R12-4: Average and Maximum Stresses for Studs/Bolts MUR Orig. MUR Orig. MUR Orig. MURPL Orig. Average Average Average Average Maximum Maximum Maximum Maximum Stress Stress Stress Stress Stress Stress Stress Stress (ksi) (ksi) Limit Limit (ksi) (ksi) Limit Limit (ksi) (ksi) (ksi) (ksi) 47.8 40.5 57.7 57.7 72.1 61.1 77.9 77.9 32.1 31.8 57.7 57.7 58.5 57.9 77.9 77.9 30.4 30.0 86.2 86.2 44.7 44.1 125 125 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 25 NON-PROPRIETARY NRC/EMCB Request 13 Discuss further information to demonstrate that, for the expected post-uprate conditions, the spent fuel pool (SFP) structure, including SFP liner and the spent fuel racks, remain capable of performing their intended design functions and will continue to be in compliance with the Byron and Braidwood design basis code of record(s) and acceptance criteria.
3                                                                                  +458 maximum AFW enthalpy 4          High Tavg, High  T avg, 0% tube tube plugging, plugging, ANS 1971 +  + 20%,
Response During a February 1, 2012 clarification call between Exelon Generation Company (EGC) and the Nuclear Regulatory Commission (NRC) staff, EGC requested and the NRC staff agreed to allow EGC to provide a response to this request under a separate transmittal at a later date.
4                                                                                 +457 maximum AFW enthalpy 5         Low Tavg, Low  T avg, 5% tube plugging, plugging, ANS 1979 - 26,20',
NRCJEMCB Request 13 Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 25 NON-PROPRIETARY Discuss further information to demonstrate that, for the expected post-up rate conditions, the spent fuel pool (SFP) structure, including SFP liner and the spent fuel racks, remain capable of performing their intended design functions and will continue to be in compliance with the Byron and Braidwood design basis code of record(s) and acceptance criteria.
5                                                                                   +47 maximum AFW enthalpy 6         Low Tang, Low  T avg, 0% tube tube plugging, plugging, ANS 1979 - 2a,20',
Response During a February 1, 2012 clarification call between Exelon Generation Company (EGC) and the Nuclear Regulatory Commission (NRC) staff, EGC requested and the NRC staff agreed to allow EGC to provide a response to this request under a separate transmittal at a later date.
6                       AFW enthalpy
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 26 NON-PROPRIETARY NRC Balance of Plant (NRC/SBPB)
                                                                                    +52 maximum AFWenthalpy 7         High Tavg,    5% tube T avg , 5%    tube plugging, ANS 1979 1979 -- 2Q, 20',
NRC/SBPB Request 1 Technical Specification (TS) 3.7.4 for the steam generator (SG) power operated relief valves (PORVs)currently allows 24 hours completion time to restore all but one of the four PORVs when two or more PORVs are inoperable.
7                                                                                 +178 maximum AFW enthalpy 8         High TTa^,
Hence, the TS action statement would allow all four PORVs to be inoperable for up to 24 hours.The analysis for a steam generator tube rupture (SG TR)credits the use of two PORVs to cool down the reactor coolant system (RCS) rapidly to achieve a subcooling margin in order to start depressurizing the RCS to stop the break flow. The analysis identifies the most limiting single failure as a failure of a SG PORV on an intact SG.Thus, the licensee credits the SG PORVs with a high significance for successfully mitigating a SGTR.The current TS that allows 24 hours for all four PORVs to be inoperable (loss of function) may not be appropriate.
avg, 0% tube plugging, ANS 1979  1979 -- 2a, 20',
Justify the current TS action statement that allows all four SG PORVs to be inoperable based on the new SGTR analysis.Response Based on discussions during the February 1, 2012 clarification call between EGC and the NRC staff, the NRC staff revised this request in an e-mail dated February 8, 2012 (Reference 4).
8                                                                                  +176 maximum AFW enthalpy Low Tang, T avg, 5% 5%tube tubeplugging, plugging, ANS ANS 1979 - 2a, 20',
The NRC staff agreed to allow EGC to provide a response to this request under separate transmittal at a later date.
9                                                                              Limiting case minimum AFW enthalpy Low Tavg, T avg, 0%0% tube plugging, ANS 1979 1979 -- 2Q, 20',
NRC/SBPB Request 2 The licensee identifies the SG PORVs as being a key component in mitigating an SGTR from an overfill condition.
10                                                                                  +3 minimum AFW enthalpy High Tavg, High  T avg, 5% tube tube plugging, plugging, ANS 1979 1979 -- 2Q, 20',
The licensee identified an SG PORV failing to open on one of the intact SGs as the most limiting failure for the margin to overfill (MTO) analysis. The installation of an uninterruptible power supply was made to reduce the current vulnerability of a single failure making two SG PORVs inoperable.
11                                                                                +125 minimum AFW            enthalpy AFWenthalpy High Tang,    0%tube T avg , 0%    tube plugging, plugging, ANS 1979 1979 -- 2Q, 20',
In Table 1-2, "Steam Generator Tube Rupture Equipment List," the licensee states,"Table 1-2 identifies the systems, components, and instrumentation which are credited for accident mitigation." The Table 1-2 does not list the SG PORV controllers.
12                                                                                +123 minimum AFWenthalpy minimum      AFWenthalpy
Provide a description of the PORVs electrical systems to include power supplies to the controllers and circuitry, and include any other circuits that would affect the SG PORV's ability to perform its function; identify any shared components (i.e., electrical, mechanical, Instrumentation & Control, etc.); and justify not including the SG PORV controllers.
** ++ indicates increase in MTO from the Limiting Case. Case.
Response As described in Technical Specifications Bases 3.7.4, a Steam Generator (SG) Power Operated Relief Valve (PORV) is considered OPERABLE when it is capable of providing controlled relief of the main steam flow and capable of fully opening and closing on demand. The definition of


OPERABLE requires that all necessary attendant instrumentation, controls, normal or emergency electrical power required to perform its specified safety function are also capable of performing their related support functions. As such, the SG PORVs were listed as an assembly NRC/SBPB Request 1 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 26 NON-PROPRIETARY NRC Balance of Plant (NRC/SBPB)
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*
Technical Specification (TS) 3.7.4 for the steam generator (SG) power operated relief valves (PORVs) currently allows 24 hours completion time to restore all but one of the four PORVs when two or more PORVs are inoperable.
Case                                  Description Description (fe)
Hence, the TS action statement would allow all four PORVs to be inoperable for up to 24 hours. The analysis for a steam generator tube rupture (SGTR) credits the use of two PORVs to cool down the reactor coolant system (RCS) rapidly to achieve a sub cooling margin in order to start depressurizing the RCS to stop the break flow. The analysis identifies the most limiting single failure as a failure of a SG PORV on an intact SG. Thus, the licensee credits the SG PORVs with a high significance for successfully mitigating a SGTR. The current TS that allows 24 hours for all four PORVs to be inoperable (loss of function) may not be appropriate.
(ft) 1        Low Tavg, Low          10% tube plugging, T avg , 10%                    ANS 1971 plugging, ANS    1971 + 20%,
Justify the current TS action statement that allows all four SG PORVs to be inoperable based on the new SGTR analysis.
1 maximum AFW  AFW enthalpy
Response Based on discussions during the February 1, 2012 clarification call between EGC and the 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 to provide a response to this request under separate transmittal at a later date. NRC/SBPB Request 2 The licensee identifies the SG PORVs as being a key component in mitigating an SGTR from an overfill condition.
                                                                                    +337 maximum 2         Low Tavg, T avg, 0%0% tube plugging, ANS 1971 + 20%,
The licensee identified an SG PORV failing to open on one of the intact SGs as the most limiting failure for the margin to overfill (MTO) analysis.
2                                                                                 +353 maximum AFW enthalpy 3          High TTavg,    10%tube avg , 10%    tubeplugging,   ANS 1971 ++ 20%,
The installation of an uninterruptible power supply was made to reduce the current vulnerability of a single failure making two SG PORVs inoperable.
plugging, ANS 3                                                                                  +440 maximum AFW enthalpy 4          High Tavg, T avg, 0%0% tube tube plugging, ANS 1971      + 20%,
In Table 1-2, "Steam Generator Tube Rupture Equipment List," the licensee states, "Table 1-2 identifies the systems, components, and instrumentation which are credited for accident mitigation." The Table 1-2 does not list the SG PORV controllers.
1971 +
Provide a description of the PORVs electrical systems to include power supplies to the controllers and circuitry, and include any other circuits that would affect the SG PORV's ability to perform its function; identify any shared components (i.e., electrical, mechanical, Instrumentation
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
& Control, etc.); and justify not including the SG PORV controllers.
** ++ indicates increase increase in MTO MTD from the Limiting Limiting Case.
Response As described in Technical Specifications Bases 3.7.4, a Steam Generator (SG) Power Operated Relief Valve (PORV) is considered OPERABLE when it is capable of providing controlled relief of the main steam flow and capable of fully opening and closing on demand. The definition of OPERABLE requires that all necessary attendant instrumentation, controls, normal or emergency electrical power required to perform its specified safety function are also capable of performing their related support functions.
Case.
As such, the SG PORVs were listed as an assembly Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 27 NON-PROPRIETARY rather than listing individual components required to support the performance of their safety function.The SG PORVs do not share mechanical components.
The SG PORVs on a single electrical division share their normal source of 480 VAC from an Engineered Safety Feature (ESF) switchgear on that division.
On Unit 1, for example, Division 1 Motor Control Centers (MCCs) are supplied from ESF Switchgear 131X and Division 2 MCCs are supplied from ESF Switchgear 132X. The SG PORVs on a single division share a common process control cabinet.On Unit 1, for example, SG PORVs 1A and 1 D receive process control signals from cabinet 1 PA33J and SG PORVs 1 B and 1 C receive process control signals from cabinets 1 PA34J.The existing SG PORVs are fed from safety related 480V MCCs which feed a power transformer in the 1/2MS018JA, JB, JC, and JD SG PORV control panels (controllers).
The SG PORV controllers contain a 4KVA power transformer that reduces the 480VAC supply to a 125VAC control power source and a secondary AC power supply source that is subsequently rectified to a DC source and used to power the reversible hydraulic pump motor contained on the PORV operator.The SG PORV controllers receive a control signal from the pressure control loops associated with the steam line pressure controls from control cabinets 1/2PA33J and 1/2PA34J.
This signal is generated based on pressure control or demands from a Manual Auto (MA) Station on the Main Control Board. The output signal from the controllers drives the hydraulic pump motor to either open or close the valve.
Once installed the modified SG PORVs will incorporate a battery-backup Uninterruptible Power Supply (UPS) into the power feed to one of two SG PORV circuits per electrical division (SG PORV's 1/2D for Division I and 1
/2C for Division 2).The UPS will provide battery back-up power to the valve in the event of a loss of the UPS normal AC power supply.The SG PORV UPS modification also affects the power source to the Division 1 SG PORV process control cabinets. A loss of power to a safety related Division 1 would also result in the loss of power to the 1/2PA33J control cabinet since the cabinets are fed from two separate 120VAC distribution panels from Division I MCCs. The pressure modulating signals for SG PORVs 1/2A and 1/2D are processed in 1/2PA33J.
To resolve this issue, the 120VAC distribution panel feed to the 1/2PA33J cabinet is replaced with a feed from a Division I inverter backed instrument bus. The 1B and 1C SG PORV control circuits are processed in the 1/2PA34J panels which are currently fed from a Division 2 120VAC distribution panel and a Division 2 inverter backed instrument bus and thus would be unaffected by a Division 2 bus outage/failure.
Tables SBPB R2-1 and R2-2 below list the power supplies for the SG PORV controllers and associated control cabinets. The information provided reflects the configuration following the implementation of the UPS modification. All power supplies are Electrical Class 1 E.
Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 27 NON-PROPRIETARY rather than listing individual components required to support the performance of their safety function.
The SG PORVs do not share mechanical components.
The SG PORVs on a single electrical division share their normal source of 480 VAC from an Engineered Safety Feature (ESF) switchgear on that division.
On Unit 1, for example, Division 1 Motor Control Centers (MCCs) are supplied from ESF Switchgear 131X and Division 2 MCCs are supplied from ESF Switchgear 132X. The SG PORVs on a single division share a common process control cabinet. On Unit 1, for example, SG PORVs 1A and 10 receive process control signals from cabinet 1 PA33J and SG PORVs 1 Band 1 C receive process control signals from cabinets 1PA34J. The existing SG PORVs are fed from safety related 480V MCCs which feed a power transformer in the 1/2MS018JA, JB, JC, and JD SG PORV control panels (controllers).
The SG PORV controllers contain a 4KV A power transformer that reduces the 480VAC supply to a 125VAC control power source and a secondary AC power supply source that is subsequently rectified to a DC source and used to power the reversible hydraulic pump motor contained on the PORV operator.
The SG PORV controllers receive a control signal from the pressure control loops associated with the steam line pressure controls from control cabinets 1/2PA33J and 1/2PA34J.
This signal is generated based on pressure control or demands from a Manual Auto (MA) Station on the Main Control Board. The output signal from the controllers drives the hydraulic pump motor to either open or close the valve. Once installed the modified SG PORVs will incorporate a battery-backup Uninterruptible Power Supply (UPS) into the power feed to one of two SG PORV circuits per electrical division (SG PORV's 1/20 for Division 1 and 1/2C for Division 2). The UPS will provide battery back-up power to the valve in the event of a loss of the UPS normal AC power supply. The SG PORV UPS modification also affects the power source to the Division 1 SG PORV process control cabinets.
A loss of power to a safety related Division 1 would also result in the loss of power to the 1/2PA33J control cabinet since the cabinets are fed from two separate 120VAC distribution panels from Division 1 MCCs. The pressure modulating signals for SG PORVs 1/2A and 1/20 are processed in 1/2PA33J.
To resolve this issue, the 120VAC distribution panel feed to the 1/2PA33J cabinet is replaced with a feed from a Division 1 inverter backed instrument bus. The 1 Band 1 C SG PORV control circuits are processed in the 1/2PA34J panels which are currently fed from a Division 2 120VAC distribution panel and a Division 2 inverter backed instrument bus and thus would be unaffected by a Division 2 bus outage/failure.
Tables SBPB R2-1 and R2-2 below list the power supplies for the SG PORV controllers and associated control cabinets.
The information provided reflects the configuration following the implementation of the UPS modification.
All power supplies are Electrical Class 1 E.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 28 NON-PROPRIETARY Table SPBP R2
-1:SG PORV Power Supplies SG PORV Control Panel (Controllers)
Primary Power Supply Backup Power Supply Byron Station l MS018A 1 MS018JA 131 X28 1 MS018B 1 MS018JB 132X1 1 MS018C 1 MS018JC UPS from 132X5 UPS from Battery 1 MS018D 1 MS018JD UPS from 131X4 UPS from Battery 2MS018A 2MS018JA 231X28 2MS018B 2MS018JB 232X1 2MS018C 2MS018JC UPS from 232X5 UPS from Battery 2MS018D 2MS018JD UPS from 231X4 UPS from Battery Braidwood Station I MS018A l MS018JA 131 X28 1 MS018B 1 MS018JB 132X1 1 MS018C 1 MS018JC UPS from 132X5 UPS from Battery lMS018D lMS018JD UPS from 131X4 UPS from Battery 2MS018A 2MS018JA 231X28 2MS018B 2MS018JB 232X1 2MS018C 2MS018JC UPS from 232X5 UPS from Battery 2MS018D 2MS018JD UPS from 231X4 UPS from Battery Table SBPB R2-2: Control Cabinet Power SuppliesControl Cabinet 1 PA33J 1 PA34J 2PA33J 2PA34J Primary Power Supply Byron Station Instrument Bus 113 Instrument Bus 114 Instrument Bus 213 Instrument Bus 214 Backup Power Supply (120 VAC Distribution Panel) 131X1 132X1 231X1 232X1 Braidwood Station 1 PA33J 1 PA34J 2PA33J 2PA34J Instrument Bus 113


Instrument Bus 114 Instrument Bus 213 Instrument Bus 214 131X1 132X1 SG PORV 1MS018A 1MS018B 1MS018C 1MS018D 2MS018A 2MS018B 2MS018C 2MS018D 1MS018A 1MS018B 1MS018C 1MS018D 2MS018A 2MS018B 2MS018C 2MS018D Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 28 NON*PROPRIETARY Table SPBP R2*1: SG PORV Power Supplies Control Panel Primary Power Backup Power Supply (Controllers)
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
Supply Byron Station 1MS018JA 131X2B 1MS018JB 132X1 1MS018JC UPS from 132X5 UPS from Battery 1MS018JD UPS from 131X4 UPS from Battery 2MS018JA 231X2B 2MS018JB 232X1 2MS018JC UPS from 232X5 UPS from Battery 2MS018JD UPS from 231X4 UPS from Battery Braidwood Station 1MS018JA 131X2B 1MS018JB 132X1 1MS018JC UPS from 132X5 UPS from Battery 1MS018JD UPS from 131X4 UPS from Battery 2MS018JA 231X2B 2MS018JB 232X1 2MS018JC UPS from 232X5 UPS from Battery 2MS018JD UPS from 231X4 UPS from Battery Table SBPB R2*2: Control Cabinet Power Supplies Control Backup Power Cabinet Primary Power Supply Supply (120 VAC Distribution Panel) Byron Station 1PA33J Instrument Bus 113 131X1 1PA34J Instrument Bus 114 132X1 2PA33J Instrument Bus 213 231X1 2PA34J Instrument Bus 214 232X1 Braidwood Station 1PA33J Instrument Bus 113 131X1 1PA34J Instrument Bus 114 132X1 2PA33J Instrument Bus 213 231X1 2PA34J Instrument Bus 214 232X1 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 29 NON-PROPRIETARY NRC/SBPB Request 3 The licensee is making modifications to the auxiliary feedwater (AFW) flow control valves to include an air accumulator tank capable of supplying air for 30 minutes. In accordance with their analysis, AFW flow control is required longer than 30 minutes to mitigate the SGTR and for RCS cool down. In Attachment 5a, Section 11.2.E, Single Failure Considerations, the licensee states: In addition, since the failure of an intact SG PORV scenario assumes a loss of offsite power with an associated loss of Instrument Air (IA), the modification described in Section 11.2.F, Item 1, assures that AFW flow control is maintained throughout the event.
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,"
According to the licensee's evaluation, an SGTR event continues until break flow is terminated at 3458/3258 seconds (Units I and 2).
Uprate,"
Describe the basis for selecting 30 minutes, and explain how the amount of air that is required is determined and the amount of air available to support this function.
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.
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 seconds Unit 1 and 3,258 seconds for Unit 2).Attachment 5a to the MUR power uprate LAR (Reference 1), Section 11.2.F, "Modifications to Support MTO Single Failure Considerations," describes the plant modifications Byron and Braidwood Stations will be implementing to support the Steam Generator Margin to Overfill Reanalysis assumptions. Included in these modifications will be the installation of two instrument air accumulator tanks on each Unit (one per train) to provide a safety related air supply for the Auxiliary Feedwater (AFW) Flow Control Valves (FCVs)(AF005). The air accumulator tanks for the AFW FCVs (AF005) are only required for the first 30 minutes (1,800 seconds) post-SGTR event initiation for AFW flow control and isolation. After AFW flow is isolated to the ruptured SG, AFW flow control to the ruptured SG is no longer needed for the duration of the event. AFW flow to the non-ruptured SGs is controlled by throttling either the AFW FCVs (AF005) or the motor operated AFW valves (AF013); these valves are in series with each other.
S. NRC)
The Emergency Operating Procedures (EOPs) (1/2B(w)EP-0, Reactor Trip or Safety Injection Unit 1(2)) direct isolation of AFW to the ruptured SG with the motor operated AFW isolation valves (AF013). Following the installation of the air accumulator tanks, the EOPs will be revised to direct the closure of the AFW FCVs (AF005) via the controller in the Main Control Room at the same point in the procedure that they are directed to close the AFW (AF013) valve. If an AF0013 valve fails to close, then the EOPs will direct an operator to be dispatched to close the associated AF005 flow control valve locally. This action prevents the valve from failing open when the air supply from the accumulator tank is exhausted. It was determined that a 30 minute supply of air is sufficient to allow the operator to reach the AF005 valve and manually close it using the installed handwheel on the valve.
NRC) toto M.
The time assumed for the local closure of the AF005 valves is consistent with the current Byron and Braidwood design basis. Specifically, in UFSAR Sections 3.11.10, "High Energy Line Break (HELB)," 10.4.9.3, "Auxiliary Feedwater, Safety Evaluation," and 15.2.8.2, "Feedwater System Pipe Break, Analysis of Effects and Consequences," for feedline and main steamline breaks, NRC/SSPS Request 3 Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 29 NON-PROPRIETARY The licensee is making modifications to the auxiliary feedwater (AFW) flow control valves to include an air accumulator tank capable of supplying air for 30 minutes. In accordance with their analysis, AFW flow control is required longer than 30 minutes to mitigate the SGTR and for RCS cool down. In Attachment 5a, Section 1I.2.E, Single Failure Considerations, the licensee states: In addition, since the failure of an intact SG PORV scenario assumes a loss of off site power with an associated loss of Instrument Air (fA), the modification described in Section 1I.2.F, Item 1, assures that AFWflow control is maintained throughout the event. According to the licensee's evaluation, an SGTR event continues until break flow is terminated at 345813258 seconds (Units 1 and 2). Describe the basis for selecting 30 minutes, and explain how the amount of air that is required is determined and the amount of air available to support this function.
M. J.
Response As noted in the NRC's request above, the limiting Steam Generator Tube Rupture (SGTR) event continues until break flow is terminated (3,4S8 seconds Unit 1 and 3,2S8 seconds for Unit 2). Attachment Sa to the MUR power uprate LAR (Reference 1), Section II.2.F, "Modifications to Support MTO Single Failure Considerations," describes the plant modifications Byron and Braidwood Stations will be implementing to support the Steam Generator Margin to Overfill Reanalysis assumptions.
J. Pacilio Pacilio (Exelon (Exelon Generation GenerationCompany, Company, LLC),
Included in these modifications will be the installation of two instrument air accumulator tanks on each Unit (one per train) to provide a safety related air supply for the Auxiliary Feedwater (AFW) Flow Control Valves (FCVs)(AFOOS).
LLC), "Braidwood "Braidwood Station, Station, Units Units 11 and and 22 and and Byron Byron Station, Station, Unit Unit Nos.
The air accumulator tanks for the AFW FCVs (AFOOS) are only required for the first 30 minutes (1,800 seconds) post-SGTR event initiation for AFW flow control and isolation.
Nos. 1I and 2 - Request Request RE: Measurement for Additional Information RE: Measurement Uncertainty Uncertainty Power PowerUprate UprateRequest Request(TAC (TACNOS.
After AFW flow is isolated to the ruptured SG, AFW flow control to the ruptured SG is no longer needed for the duration of the event. AFW flow to the non-ruptured SGs is controlled by throttling either the AFW FCVs (AFOOS) or the motor operated AFW valves (AF013); these valves are in series with each other. The Emergency Operating Procedures (EOPs) (1/2B(w)EP-0, Reactor Trip or Safety Injection Unit 1 (2)) direct isolation of AFW to the ruptured SG with the motor operated AFW isolation valves (AF013). Following the installation of the air accumulator tanks, the EOPs will be revised to direct the closure of the AFW FCVs (AFOOS) via the controller in the Main Control Room at the same point in the procedure that they are directed to close the AFW (AF013) valve. If an AF0013 valve fails to close, then the EOPs will direct an operator to be dispatched to close the associated AFOOS flow control valve locally. This action prevents the valve from failing open when the air supply from the accumulator tank is exhausted.
NOS.
It was determined that a 30 minute supply of air is sufficient to allow the operator to reach the AFOOS valve and manually close it using the installed handwheel on the valve. The time assumed for the local closure of the AFOOS valves is consistent with the current Byron and Braidwood design basis. Specifically, in UFSAR Sections 3.11.10, "High Energy Line Break (HELB)," 1 0.4.9.3, "Auxiliary Feedwater, Safety Evaluation," and 1S.2.8.2, "Feedwater System Pipe Break, Analysis of Effects and Consequences," for feedline and main steamline breaks, Braidwood/Byron Stations MUR LAR Responseto RAI February 20, 2012 , page 30 NON-PROPRIETARYoperator action is credited to isolate auxiliary feedwater to the faulted steam generator within 20 minutes.The AF005 instrument air accumulators were sized to include 30 minutes of air supply as described above and additional capacity to account for:Stroking four valves (1 Train)from full open to full closed,Maximum air consumption rate for four electric to pneumatic signal converters (IY's),Maximum air consumption rate for four valve positioners, and10% allowance for leakage.The total volume required was determined to be 27.3 cubic feet (204 gallons). Additional conservatism exists since the tank size is 33.4 cubic feet (250 gallons). This ensures that adequate air is available to support the required function of AFW flow control and isolation.
ME6587, ME6587, ME6588, ME6588, 6589, 6589, AND AND ME659D),"
NRC/SBPB Request 4 Figure 11-5 of Attachment 5a shows the SG water volume on Unit I trending towards the maximum available quantity. At approximately 3200 seconds, the trend tapers off, resulting in a margin to overfill of approximately 94 cubic feet. At the same time other graphs show a sharp reduction in SG pressure, which logically corresponds to a second opening of the SG PORVs on the intact SGs. This action stops the upward trend and prevents the overfill condition. The 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 an overfill of the SG.In the updated final safety analysis report, Section 15.6.3.2, under the section describing major operator actions, the licensee's analysis credits operators for reopening pressurizer PORV, four minutes after establishing normal charging and letdown, in order to equalize the RCS and SG pressures.
ME6590)," dated dated November November 28,  28, 2011 2011 4 E-mail from Brenda Mozafari (U.  (U. S. NRC)
In Attachment 5a (page 11-10), the licensee states that the SG PORVs on the intact SGs automatically open, as necessary, to maintain RCS subcooling margin. The above mentioned graph trend shows a sharp pressure reduction at 3200 seconds, which is not indicative of SG PORV automatically controlling pressure at a prescribed setpoint a.Evaluate whether this operator action is credited to be performed within a specific time in order to prevent an overfill condition.
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,"
b.If operator action is required, identify the action as a critical operator action.
Overfill Analysis  Methodology," November November 2007.}}
c.Describe whether the new analysis changes the existing UFSAR analysis, and results in the major operator action opening a SG PORV rather than a pressurizer PORV after SI termination to stop an overfill condition from occurring.
Response The Steam Generator Tube Rupture/Margin to Overfill (SGTR/MTO) analysis methodology used in the new SGTR/MTO Analysis submitted in Attachment 5a to the MUR power uprate LAR (Reference 1) is different from the methodology in the current Analysis of Record (AOR) described in the UFSAR Section 15.6.3, "Steam Generator Tube Rupture." The methodology used in the current AOR SGTR/MTO analysis explicitly models operator actions after Safety Injection (SI) flow termination (i.e. securing Emergency Core Cooling (ECCS) flow), including the operator action to open the pressurizer PORV within a specific time in order to prevent an overfill condition. The SGTR/MTO analysis provided in Attachment 5a, "Steam Generator Tube Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 30 NON-PROPRIETARY operator action is credited to isolate auxiliary feedwater to the faulted steam generator within 20 minutes. The AF005 instrument air accumulators were sized to include 30 minutes of air supply as described above and additional capacity to account for:
* Stroking four valves (1 Train) from full open to full closed,
* Maximum air consumption rate for four electric to pneumatic signal converters (IY's),
* Maximum air consumption rate for four valve positioners, and
* 10% allowance for leakage. The total volume required was determined to be 27.3 cubic feet (204 gallons).
Additional conservatism exists since the tank size is 33.4 cubic feet (250 gallons).
This ensures that adequate air is available to support the required function of AFW flow control and isolation.
NRC/SBPB Request 4 Figure 1/-5 of Attachment 5a shows the SG water volume on Unit 1 trending towards the maximum available quantity.
At approximately 3200 seconds, the trend tapers off, resulting in a margin to overfill of approximately 94 cubic feet. At the same time other graphs show a sharp reduction in SG pressure, which logically corresponds to a second opening of the SG PORVs on the intact SGs. This action stops the upward trend and prevents the overfill condition.
The 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 an overfill of the SG. In the updated final safety analysis report, Section 15.6.3.2, under the section describing major operator actions, the licensee's analysis credits operators for reopening pressurizer PORV, four minutes after establishing normal charging and letdown, in order to equalize the RCS and SG pressures.
In Attachment 5a (page 11-10), the licensee states that the SG PORVs on the intact SGs automatically open, as necessary, to maintain RCS sub cooling margin. The above mentioned graph trend shows a sharp pressure reduction at 3200 seconds, which is not indicative of SG PORV automatically contrOlling pressure at a prescribed setpoint.
: a. Evaluate whether this operator action is credited to be performed within a specific time in order to prevent an overfill condition.
: b. If operator action is required, identify the action as a critical operator action. c. Describe whether the new analysis changes the existing UFSAR analysis, and results in the major operator action opening a SG PORV rather than a pressurizer PORV after SI termination to stop an overfill condition from occurring.
Response The Steam Generator Tube Rupture/Margin to Overfill (SGTR/MTO) analysis methodology used in the new SGTR/MTO Analysis submitted in Attachment 5a to the MUR power uprate LAR (Reference
: 1) is different from the methodology in the current AnalYSis of Record (AOR) described in the UFSAR Section 15.6.3, "Steam Generator Tube Rupture." The methodology used in the current AOR SGTR/MTO analysis explicitly models operator actions after Safety Injection (SI) flow termination (i.e. securing Emergency Core Cooling (ECCS) flow), including the operator action to open the pressurizer PORV within a specific time in order to prevent an overfill condition.
The SGTR/MTO analysis provided in Attachment 5a, "Steam Generator Tube Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 31 NON-PROPRIETARY Rupture Analysis report," of the MUR power uprate submittal (Reference 1) uses the NRC approved methodology described in WCAP-1 0698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill" (Reference 5).
Consistent with the WCAP-1 0698-P-A methodology, specific operator actions after SI termination are not used and the LOFTTR2 computer code is used to predict the transient responses that lead to pressure equalization (break flow termination) and to demonstrate the SG overfill condition is not reached. Therefore, actions taken after SI termination are not considered critical operator responses and as such are modeled to occur as conditions require as predicted by the LOFTTR2 computer code.
As discussed in Section 11.2.D, "Operator Action Times," of Attachment 5a of the MUR power uprate submittal (Reference 1), the critical operator responses are:
1.Isolate Auxiliary Feedwater (AFW) flow to the ruptured Steam Generator (SG), 2.Isolate the MSIV on the ruptured SG, 3.Initiate RCS cooldown, to initiate RCS depressurization, and 4.Terminate Safety Injection (SI) (secure Emergency Core Coolant (ECCS) flow).
These operator actions and the corresponding operator action times used for the analyses are summarized in Table 11-2, "Operator Action Times for Design Basis SGTR Analyses" of a of the MUR power uprate submittal (Reference 1). These actions are consistent with the actions in WCAP-1 0698-P-A (Reference 5) Table 2.3-2, "Operator Action Times for Design Basis SGTR Analysis." Also, consistent with the methodology in WCAP-1 0698-P-A (Reference 5) the times required for cooldown, depressurization, and pressure equalization are calculated using the LOFTTR2 program. The analyses do not model specific operator action times after SI termination.
In accordance with Emergency Operating Procedures (EOPs) (1/2B(w)EP-3), the same step that directs the operator to terminate RCS cooldown also directs the operators to maintain RCS temperature below the required temperature. This step occurs before SI termination and is a step that is monitored and acted on throughout the procedure. SI termination occurs at 2,311 seconds on Unit 1 and at 2,482 seconds on Unit 2. After SI termination, LOFTTR2 models the opening of two of the intact SG PORVs to maintain the required RCS temperature from theEOPs. This action is predicted by LOFTTR2 to occur at approximately 3,200 seconds (Unit I analysis). This modeling is consistent with the methodology in WCAP-10698-P-A.
NRC/SBPB Request 5 Calculation Westinghouse commercial atomic power (WCAP) -10698-P-A provides a general assessment of the MTO for Westinghouse type reactors. There were instances where the licensee deviated from the input parameters selected in WCAP-10698-P-A as the most conservative.
a.Decay heat is one of the input factors that influence MTO analyses and Thermal/Hydraulic analyses during a tube rupture. For the MTO analysis, the licensee states that plant specific sensitivities were performed for Bryon and Braidwood Units 1 and 2.
These studies concluded that the 1979-2a American Nuclear Society (ANS) decay heat factor was more conservative compared to the 1971 +20% ANS decay heat model specified in WCAP-10698-P-A.
Justify use of the 1979-2Q ANS decay heat factor was more conservative compared to the 1971 +20% ANS decay heat factor.
Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 31 NON-PROPRIETARY Rupture Analysis report," of the MUR power uprate submittal (Reference
: 1) uses the NRC approved methodology described in WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill" (Reference 5). Consistent with the WCAP-1 0698-P-A methodology, specific operator actions after SI termination are not used and the LOFTTR2 computer code is used to predict the transient responses that lead to pressure equalization (break flow termination) and to demonstrate the SG overfill condition is not reached. Therefore, actions taken after SI termination are not considered critical operator responses and as such are modeled to occur as conditions require as predicted by the LOFTTR2 computer code. As discussed in Section 11.2.0, "Operator Action Times," of Attachment 5a of the MUR power uprate submittal (Reference 1), the critical operator responses are: 1. Isolate Auxiliary Feedwater (AFW) flow to the ruptured Steam Generator (SG), 2. Isolate the MSIV on the ruptured SG, 3. Initiate RCS cooldown, to initiate RCS depressurization, and 4. Terminate Safety Injection (SI) (secure Emergency Core Coolant (ECCS) flow). These operator actions and the corresponding operator action times used for the analyses are summarized in Table 11-2, "Operator Action Times for Design Basis SGTR Analyses" of Attachment 5a of the MUR power uprate submittal (Reference 1). These actions are consistent with the actions in WCAP-10698-P-A (Reference
: 5) Table 2.3-2, "Operator Action Times for Design Basis SGTR Analysis." Also, consistent with the methodology in WCAP-10698-P-A (Reference
: 5) the times requJred for cooldown, depressurization, and pressure equalization are calculated using the LOFTTR2 program. The analyses do not model specific operator action times after SI termination.
In accordance with Emergency Operating Procedures (EOPs) (1/2B(w)EP-3), the same step that directs the operator to terminate RCS cooldown also directs the operators to maintain RCS temperature below the required temperature.
This step occurs before SI termination and is a step that is monitored and acted on throughout the procedure.
SI termination occurs at 2,311 seconds on Unit 1 and at 2,482 seconds on Unit 2. After SI termination, LOFTTR2 models the opening of two of the intact SG PORVs to maintain the required RCS temperature from the EOPs. This action is predicted by LOFTTR2 to occur at approximately 3,200 seconds (Unit 1 analysis).
This modeling is consistent with the methodology in WCAP-10698-P-A.
NRC/SBPB Request 5 Calculation Westinghouse commercial atomic power (WCAP) -10698-P-A provides a general assessment of the MTO for Westinghouse type reactors.
There were instances where the licensee deviated from the input parameters selected in WCAP-10698-P-A as the most conservative.
: a. Decay heat is one of the input factors that influence MTO analyses and Thermal/Hydraulic analyses during a tube rupture. For the MTO analysis, the licensee states that plant specific sensitivities were performed for Bryon and Braidwood Units 1 and 2. These studies concluded that the 1979-2a American Nuclear SOCiety (ANS) decay heat factor was more conservative compared to the 1971 +20% ANS decay heat model specified in WCAP-10698-P-A.
Justify use of the 1979-2a ANS decay heat factor was more conservative compared to the 1971 +20% ANS decay heat factor.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 32 NON-PROPRIETARY b.Similar to above, in determining the most conservative input values, the licensee chose to model the minimum AFWenthalpy of 0.03 Btu/lbm; whereas, WCAP-10698-P-A models the maximum temperature of AFW (maximum enthalpy) as the most conservative parameter in the analysis for MTO.
Justify how the use of the minimum AFW enthalpy is more conservative compared to using the maximum temperature (enthalpy) for AFW.
Response WCAP-1 0698-P-A (Reference 5) identified high decay heat and high Auxiliary Feedwater (AFW) temperature to be the conservative assumptions for the steam generator tube rupture margin to overfill (MTO) analysis. NSAL-07-1 1, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology" (Reference 6), identified a lower decay heat can be more limiting for some plants. To resolve the concerns of NSAL-07-1 1, plant-specific sensitivities were performed for Byron and Braidwood Units 1 and 2 to justify the decay heat model and AFW enthalpy assumed in the analysis. The Tables SBPB R5-1 and 2 show the impact on the Margin to Overfill (MTO) resulting from the sensitivity study. The study covered the Tav9 range and the steam generator tube plugging levels supported by the analysis provided in Attachment 5a. The impact on MTO provided is relative to the limiting case modeling the ANS 1979 - 2a decay heat model, low AFW enthalpy, low Tavg, and high steam generator tube plugging level.
The results show that use of the ANS 1971 + 20% decay heat model (cases 1 to 4) clearly provides more MTO margin than the ANS 1979 - 2a decay heat model (cases 5 to 8). The conservative direction for AFW enthalpy is studied using low decay heat. Comparing cases 5 to 8 with corresponding cases 9 to 12 show that minimum AFW enthalpy is conservative.
Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 32 NON-PROPRIETARY
: b. Similar to above, in determining the most conservative input values, the licensee chose to model the minimum AFW enthalpy of 0.03 Btu/Ibm; whereas, WCAP-10698-P-A models the maximum temperature of AFW (maximum enthalpy) as the most conservative parameter in the analysis for MTO. Justify how the use of the minimum AFW enthalpy is more conservative compared to using the maximum temperature (enthalpy) for AFW. Response WCAP-10698-P-A (Reference
: 5) identified high decay heat and high Auxiliary Feedwater (AFW) temperature to be the conservative assumptions for the steam generator tube rupture margin to overfill (MTO) analysis.
NSAL-07-11, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology" (Reference 6), identified a lower decay heat can be more limiting for some plants. To resolve the concerns of NSAL-07-11, plant-specific sensitivities were performed for Byron and Braidwood Units 1 and 2 to justify the decay heat model and AFWenthalpy assumed in the analysis.
The Tables SBPB R5-1 and 2 show the impact on the Margin to Overfill (MTO) resulting from the sensitivity study. The study covered the T avg range and the steam generator tube plugging levels supported by the analysis provided in Attachment 5a. The impact on MTO provided is relative to the limiting case modeling the ANS 1979 -20 decay heat model, low AFW enthalpy, low T avg, and high steam generator tube plugging level. The results show that use of the ANS 1971 + 20% decay heat model (cases 1 to 4) clearly provides more MTO margin than the ANS 1979 -20 decay heat model (cases 5 to 8). The conservative direction for AFW enthalpy is studied using low decay heat. Comparing cases 5 to 8 with corresponding cases 9 to 12 show that minimum AFW enthalpy is conservative.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 33 NON-PROPRIETARY Table SBPB R5-1:Byron/Braidwood Unit I Results of Sensitivity Study on MTO Case Description Impact on MTO*(ft)1 Low Tang, 5% tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+321 2 Low Tavg, 0% tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+314 3 High Tang, 5% tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+458 4 High Tavg, 0% tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+457 5 Low Tavg, 5%
tube plugging, ANS 1979 - 26, maximum AFW enthalpy
+47 6 Low Tang, 0% tube plugging, ANS 1979 - 2a, maximum AFW enthalpy
+52 7 High Tavg, 5% tube plugging, ANS 1979 - 2Q, maximum AFW enthalpy
+178 8 High Ta^, 0% tube plugging, ANS 1979 - 2a, maximum AFW enthalpy
+176 9 Low Tang, 5% tube plugging, ANS 1979 - 2a, minimum AFW enthalpy Limiting case 10 Low Tavg, 0% tube plugging, ANS 1979 - 2Q,+3 minimum AFW enthalpy 11 High Tavg, 5% tube plugging, ANS 1979 - 2Q, minimum AFW enthalpy
+125 12 High Tang, 0% tube plugging, ANS 1979 - 2Q,+123 minimum AFWenthalpy
* +indicates increase in MTO from the Limiting Case.
Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 33 NON-PROPRIETARY Table SBPB R5-1: Byron/Braidwood Unit 1 Results of Sensitivity Study on MTO Case Description Impact on MTO* (ft3) 1 Low T avg, 5% tube plugging, ANS 1971 + 20%, +321 maximum AFW enthalpy 2 Low T avg, 0% tube plugging, ANS 1971 + 20%, +314 maximum AFW enthalpy 3 High T avg, 5% tube plugging, ANS 1971 + 20%, +458 maximum AFW enthalpy 4 High T avg, 0% tube plugging, ANS 1971 + 20%, +457 maximum AFW enthalpy 5 Low T avg, 5% tube plugging, ANS 1979 -20', +47 maximum AFW enthalpy 6 Low T avg, 0% tube plugging, ANS 1979 -20', +52 maximum AFWenthalpy 7 High T avg , 5% tube plugging, ANS 1979 -20', +178 maximum AFW enthalpy 8 High T avg, 0% tube plugging, ANS 1979 -20', +176 maximum AFW enthalpy 9 Low T avg, 5% tube plugging, ANS 1979 -20', Limiting case minimum AFW enthalpy 10 Low T avg, 0% tube plugging, ANS 1979 -20', +3 minimum AFW enthalpy 11 High T avg, 5% tube plugging, ANS 1979 -20', +125 minimum AFWenthalpy 12 High T avg , 0% tube plugging, ANS 1979 -20', +123 minimum AFWenthalpy
* + indicates increase in MTO from the Limiting Case.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 34 NON-PROPRIETARY Table SBPB R5
-2:Byron/Braidwood Unit 2 Results of Sensitivity Study on IMO Case Description Impact on MTO*(ft)1 Low Tavg, 10% tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+337 2 Low Tavg, 0%
tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+353 3 High Tavg, 10% tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+440 4 High Tavg, 0% tube plugging, ANS 1971 + 20%, maximum AFW enthalpy
+472 5 Low Tang, 10% tube plugging, ANS 1979 - 2a, maximum AFW enthalpy
+67 6 Low Tavg, 0% tube plugging, ANS 1979 - 2a, maximum AFW enthalpy
+102 7 High Tang, 10% tube plugging, ANS 1979 - 2a, maximum AFW enthalpy
+176 8 High Tavg, 0% tube plugging, ANS 1979 - 2Q, maximum AFW enthalpy
+212 9 Low Tavg, 10% tube plugging, ANS 1979 - 2Q, minimum AFW enthalpy Limiting case 10 Low Tavg, 0%
tube plugging, ANS 1979 - 2a, minimum AFW enthalpy
+23 11 High Tavg, 10% tube plugging, ANS 1979 - 2a, minimum AFW enthalpy
+129 12 High Tavg, 0% tube plugging, ANS 1979 - 2a, minimum AFW enthalpy
+159* +indicates increase in MTO from the Limiting Case.
Braidwood/Byron Stations MUR LAR Response to RAI February 20,2012 Attachment 1, page 34 NON-PROPRIETARY Table SBPB R5-2: Byron/Braidwood Unit 2 Results of Sensitivity Study on MTO Case Description Impact on MTO* (fe) 1 Low T avg , 10% tube plugging, ANS 1971 + 20%, +337 maximum AFW enthalpy 2 Low T avg, 0% tube plugging, ANS 1971 + 20%, +353 maximum AFW enthalpy 3 High T avg , 10% tube plugging, ANS 1971 + 20%, +440 maximum AFW enthalpy 4 High T avg, 0% tube plugging, ANS 1971 + 20%, +472 maximum AFW enthalpy 5 Low T avg , 10% tube plugging, ANS 1979 -20, +67 maximum AFW enthalpy 6 Low T avg, 0% tube plugging, ANS 1979 -20, +102 maximum AFW enthalpy 7 High T avg, 10% tube plugging, ANS 1979 -20, +176 maximum AFW enthalpy 8 High T avg, 0% tube plugging, ANS 1979 -20, +212 maximum AFW enthalpy 9 Low T avg, 10% tube plugging, ANS 1979 -20, Limiting case minimum AFW enthalpy 10 Low T avg, 0% tube plugging, ANS 1979 -20, +23 minimum AFW enthalpy 11 High T avg , 10% tube plugging, ANS 1979 -20, +129 minimum AFW enthalpy 12 High T avg, 0% tube plugging, ANS 1979 -20, +159 minimum AFW enthalpy * + indicates increase in MTD from the Limiting Case.
Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 , page 35 NON-PROPRIETARY REFERENCES 1Letter from Craig Lambert (Exelon Generation Company, LLC) to U. S. NRC,"Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated June 23, 2011 2Letter from Kevin F. Borton (Exelon Generation Company, LLC) to U. S. NRC, "Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated December 9, 2011 3 Letter from N. J. DiFrancesco (U. S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. I and 2 - Request for Additional Information RE: Measurement Uncertainty Power Uprate Request (TAC NOS.
ME6587, ME6588, 6589, AND ME6590)," dated November 28, 2011 4E-mail from Brenda Mozafari (U. S. NRC) to Leslie Holden, et. al., Exelon Generation Company), "FW: Draft Balance of Plant RAls related to MUR dated June 23, 2011," dated February 8, 2012 5WCAP-1 0698-P-A,"SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," August 1987 6NSAL-07-1 1, "Decay Heat Assumption in Steam Generator Tube Rupture Margin-to-Overfill Analysis Methodology," November 2007.
REFERENCES Braidwood/Byron Stations MUR LAR Response to RAI February 20, 2012 Attachment 1, page 35 NON-PROPRIETARY 1 Letter from Craig Lambert (Exelon Generation Company, LLC) to u. S. NRC, "Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated June 23, 2011 2 Letter from Kevin F. Borton (Exelon Generation Company, LLC) to u. S. NRC, "Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate," dated December 9,2011 3 Letter from N. J. DiFrancesco (U. S. NRC) to M. J. Pacilio (Exelon Generation Company, LLC), "Braidwood Station, Units 1 and 2 and Byron Station, Unit Nos. 1 and 2 -Request for Additional Information RE: Measurement Uncertainty Power Uprate Request (TAC NOS. ME6587, ME6588, 6589, AND ME659D)," dated November 28, 2011 4 E-mail from Brenda Mozafari (U. S. NRC)toLeslieHolden,et.al., Exelon Generation Company), "FW: Draft Balance of Plant RAls related to MUR dated June 23, 2011," dated February 8,2012 5 WCAP-10698-P-A, "SGTR Analysis Methodology to Determine the Margin to Steam Generator Overfill," August 1987 6 NSAL-07-11, "Decay Heat Assumption in Steam Generator Tube Rupture Overfill AnalysiS Methodology," November 2007.}}

Latest revision as of 18:59, 6 February 2020

Additional Information Supporting Request for License Amendment Regarding Measurement Uncertainty Recapture Power Uprate
ML12052A113
Person / Time
Site: Byron, Braidwood  Constellation icon.png
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)
v • d • e


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,
  • 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:

  • n Condensate System
  • Condensate Booster Booster System System
  • Heater Drains System
  • n Steam Generator Blowdown Blowdown System System n* Auxiliary Steam SystemSystem
  • 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
  • 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:

  • Chemical and and Volume Control Volume Control System System
  • Safety Injection System Injection 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.

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