Text

Braidwood Station, Units 1 and 2 - 10 CFR 50.59 Summary Report for June 19, 2010 Through June 18, 2012
	ML12349A363
Person / Time
Site:	Braidwood
Issue date:	12/14/2012
From:	Enright D J; Exelon Generation Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	BW120121 LS-AA-104-1001, Rev 3
	Download: ML12349A363 (46)
	v • d • e

December 14, 2012 BW120121 U.S.Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.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 10 CFR 50.59(d)(2)

Subject:

10 CFR 50.59 Summary Report Pursuant to the requirements of 10 CFR 50.59,"Changes, tests, and experiments," paragraph (d)(2), Braidwood Station is providing the required report for Facility Operating License Numbers NPF-72 and NPF-77.This report is being provided for the time period of June 19,2010 through June 18,2012, and consists of the 10 CFR 50.59 Coversheets for changes to the facility or procedures as described in the Updated Final Safety Analysis Report (UFSAR), and tests or experiments not described in the UFSAR.Please direct any questions regarding this submittal to Mr.Chris VanDenburgh, Regulatory Assurance Manager, at (815)417-2800.Respectfully, Daniel J.Enright Site Vice President Braidwood Station

Attachment:

Braidwood Station 10 CFR 50.59 Summary Report cc: NRR Project Manager-Braidwood Station Illinois Emergency Management Agency-Division of Nuclear Safety US NRC Regional Administrator, Region III US NRC Senior Resident Inspector (Braidwood Station)

ATTACHMENT Braidwood Station 10 CFR 50.59 Summary Report Evaluation No.Title BRW-E-201 0-178, Revision 0 1A Delta Temperature

/Temperature Average Loop-Two Thot Resistance Temperature Detector (RTD)Operation for BR1 C16 BRW-E-2011-3, Revision 0 Dry Cask Storage Project-Fuel Handling BUilding (FHB)Crane Upgrade to Single Failure Proof BRW-E-2011-040, Revision 0 Installation of Bypass Test Capability for 7300 Process Protection System and Nuclear Instrumentation System BRW-E-2011-099, Revision 0 Isolation of the1D Loop Pressurizer Spray Path to Terminate 1 RY8050 Leakage BRW-E-2011-100, Revision 0 Revision 4 of BwOP CV-16 Pressurizer Auxiliary Spray Operation BRW-E-2011-122, Revision 0 Installation of Backup Power Supplies for the Steam Generator Power Operated relief Valves (SG PORV's)BRW-E-2011-216, Revision 1 Steam Generator Margin to Overfill (SG MTO)Power Operated Relief Valve (PORV)Trim Replacement and Valve Block Installation BRW-E-2012-085, Revision 0 Change In-Core Decay Time forA1 R16 BRW-E-2012-120, Revision 0 Eliminating Action 3.3.y.D from Technical Requirements Manual (TRM)Section RM 3.3.y BRW-E-2012-155, Revision 0 Lake ScreenHouseTravelina Screen Level control BRW-E-2012-162, Revision 0 Steam Generator Margin to Overfill (SG MTO)Power Operated Relief Valve (PORV's)Trim Replacement and Valve Block Installation BRW-E-2012-186, Revision 0 Close the Isolation Valves for the Relief Valves on each of the Main Steam Isolation Valve (MSIV)Hvdraulic Accumulator Manifolds BRW-E-2012-201, Revision 0 Change In-Core Decay Time for A2R16 Revision 1 Unit 2 Main Turbine Roll for Ventilation Testing W-E-2012-245, Revision 0 Increase Resin Volume in Mixed Bed Demineralizers to 35 Cubic Feet.

50.59 REVIEW COVERSHEET FORM LS-AA-I04-100l Revision 3 SO.59 Evaluation No.: BRW-E-2016-178 Rev.No.: 0 Page I of 12 StatiolllUDits:

Braidwood Unit I ActivttyJDoeument Number: Iemaorary Change EC 382328/lBwISR 3.3.1.16-2 Revision Number: 0.1 Title: IA DItrA Loop-Iwo'[hot RID Operatiog tor CYde BRICI6 NOm: For 50.59 Evaluations.

information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CPR 50.59(d)(2).

Descrlption of Activity: The temporary modification will electronically defeat one of three Reactor Coolant System (RCS)Narrow Range hot leg temperature (Thot)Resistance Temperature Detectors (RIDs)in the lA Delta Temperatureffemperature Average (DIlrA)instrument loops.specificaDy.

the lA DIlfA Al Thot RID.Each DTlrA Thot instrument loop has three RIDs that are averaged together to provide representative hot leg temperature indication.

The instrument loop will be reconfigured to provide Thot.indication for the lA loop based on the average of the remaining (A2 and A3)RIDs**To account for the removal of theA I RIDinput.the corresponding ReS Temperature lnstl'UIDent Alignment procedure (which nonnalizCIJ dT power indication to secondary calorimetric).will also be revised.Reason for Activity: The active and spare RID elements for the lA Delta TemperaturdI'emperature Average (DTIr A)loop Narrow Range{NR)Thot U Al"RID (1m-0411AlIl'fE..04lOAl)have degraded and.are providing low temperature readings*.Each RID is provided with two elements.active and spare.to provide backup capability in the event of the failure of the RID active element.*.The ac:tive ete D1e nt for the.IA.At RID has failed and troubleshooting determined that the spare element has low resistanCe to shield**The low input for the Al RID is.resulting in inaccurate.overall Thot indication for.the lA loop***Since the spare elementoot reliable, there is no longer a usable Al RID element.Therefore.

the temporary modification will remove the IA-thot At electronic signal from the averaging logic thus allowing for twoThotRTD operation for the remaining cycle (BRlCI6)until the RID can be replaced.The issue with the performance of the AIRID was entered into CAP (1R*1145482)and the RID will be required to be replaced during the next.refueling outage (A lRI6).Effect of Activity: The Thot Avg signal along with a Tcoldsignal for each loop are used to develop.a.Delta Temperature (dT)signal and Temperature Average (Tavg)signal for use in Reactor Trip SysteM (RTS)and Engineered Safeguards Feature Actuation System (ESFAS)instrumentation.

.The temporary modification wiD restore operation of the lA RCS Thot instrument loop with the remaining 2 RTDs to provide a representative indication of the hot leg tentperature

..This will in turn restore the dT andIavgsignalsfor the IA DTlfAloop to a representative value.Though only 2 out of the original 3 RIDs are being used, the configuration will still provide a reasonably accurate hot leg temperature indication with regard to temperature streaming within the bot leg primary coolant flow.Since the instrument uncertainty calculation accounts for operation with a failed Thot RID.there are 00 changes required with respect to RTS or ESFAS setpoints or AllOwable Values as stated in the Technical Resource Manual (TRM)or Tecbnical Specifications (LeO and Bases).*Consequently.

the overall function and operation of the associated RTS and ESFAS functiOns and associated control functions are unaffected by this change.swmnary of Conclusion for the Activity's SO.59 Review: The temporary modification will restore the Thot Avg signal for the.1A DTlfA loop to its expected value.The proposed activity is classified as a temporary change and will only be installed until the next refueling outage (AI R 16.spring of 2012).DuringAIRI6,this RID will be replaced with a new RID that will be calibrated and Response Time Tested (RTI')in accordance with Station procedures.

The remaining 2 hot leg RIDs will still sample a broad area of flow and averaging the 2 RIDs still provides a representative indication of hot leg temperature

..Additionally.

as part of the temporary modification.

the hot leg temperatures for the IA DIlrA loop will be monitored monthly, and.adjustments made to the correction factor (as needed)for the Thot Avg summator modUle in accordance with the methodology/guidance provided in 12523.Deviation alarms will also ensure that the temperature indication from the lA loop does not vary significantly from the other loops.The instrument uncertainties used in the development of the setpoints and allOwable values are based on 2 of 3 RIDs used to develop hot leg temperature.

Therefore, there is no increase in the consequence of an accident or malfunction.

The RTS and ESFAS instrumentation are used to detect and initiate a trip in response to an accident or transient.

but have no impact on the frequency of occurrence of the accidents or transients.

50.59 REVIEW COVERSHEET FORM LS-AA*104-100 I Revision I SO.59 Evaluation No.: BRW-E-2810:118 Rev.No.: 8 Page 2 of U AetivitylDocument Number: Temporaa Change Ee 382328/18"I8R 3.3.1.10:2 Revision Number: 8.2 Since the control input is based upon the most limiting temperature measurement based on auctioneered high Tavg, control is still representative of the bulk ReS temperature if the I B,IC or I D loop are controlling.

If the indication forIA Thot based on 2 RIDs results in a higher Tavg signal than the other loops.then the control input is based on a more conservative inpuL Therefore.

the control and interlock functions associated with That operation will not result in conditions outside of those established in any analysis for normal operation.

Therefore.

there is no increase in the consequence of an accident or malfunction.

The control loops will continue to operatenorma1ly and in a conservative direction with regard to variations in temperature indication from the IA hot leg.Therefore, the plant will not be operated in a new manner or with different limits.Therefore.

the temporary change to the number of RIDs used for IA hot leg temperature does not create an accident of a different type previously evaluated in the UFSAR.This temporary change does not create a different faUure mode for the SSC used in the I A hot leg temperature instrumentation

.*The resultant Thot average indication generated will be based on 2 RIDs instead of 3.but the.nature of the output does not change**Therefore.

the change does.not affect the operation of the components in the instrument loops or input signal to the protection or control loops.The change does not introduce any new connections or interactions between the channels of other instrumentation loops.will continue to perform the same amtdoes not create the possibility ola diffetentmalfuncti0n important to safety as previously evaluated in the.lJfSAR.Uncertainties.

setpoints and allowable values for the.instrument loops were developed based on 2.out oHRID operation.

Therefore.

the response of the Overtemperature Delta Temperature (OTDT)and Overpower Delta Temperature (OPDT)trip functi0lJS.

is.bounded by the existing.analYSis.for the.initiating accidents or.transients,.

Rod Control is.based on the average temperature of the loop with thebighest temperature.

A low average temPCfllture measurement from any loop temperature control channel will cause no control action.Ahigh average temperature measurement will cause rod insertion (safe direction).

Therefore, variation in the IA hot leg instrument loops will not affect rod motion or will insert control rods conservatively.with respect to margins to DeParture from Nucleate Boiling (DNB)*.Since the auctioneered Tavg will be representative

.0rcoDServative.

the.dumpsYStemwiU still be.able to control RCS temperatures and pressures.

within limits considered in the Therefore, fuel clad or RCS design basis limits wiD not be exceeded.The Westinghouse methodology (WCAP.I2S83)for RTS and ESFAS trip functions was previously.reviewed and approved by the NRC as documented in BraidWOOdTS Antendments42.and 100 (via NRC SERs)for the intended purpose of establishing the methodology forproteetion system setpaints, The SER associated with RID bypass piping elimination only describes normal operation with averaging of 3 hot leg RIDs and does not discuss the use of two Thot RIDs,.However.the approved setpaint methodology addresses operation in this configuration and is consistent with the setpointmethodology discussed in TS B 3.3.landB 3.3.2.The use of two Thot RIDS for operation has been previously approved by the NRC at several other utilities during RID bypass elimination licensing.

Since the WCAP methodology based on 2 of 3 RIDs used in the hot leg was approved for Braidwood as well as other.utilities.

the use of 2 of 3 RIDs.as a result of the temporary modification is acceptable without prior NRC approval.Based on the above, prior NRC approval is not required prior to the implementation of the temporary modification.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)o AppUc:abiUty Review o 50.5'Screening 1'81 58.5'Evaluation 50.59 Screening No.58.59 Evaluation No.BRW-E*20Io-178 Rev.Rev.o o 50.59 REVIEW COVERSHEET

<forEVAL)): 'uM""" slttlmnJl1'*'

AdIYtty'IDoc-." N......r..IC 3ft351*UDAl n_I1H73n.Dry r.SCm_Pnlcd-fBI CAM U.......,.gea Proof LS-M*I04-IOOI Revision 3 Pap I of3 ReYliMII NIu.....r..OM 101 0111712011 NOlli: For 50.59 Eval'informal:ioa on Ibis ferm will provide the.for pref*'iD.the bieDDial summary report IUbmitrI:d to abe NRC iD Kcordaace wiIb the I1!qUimDeDu of 10 CfR 5O.59(d)(2).01 AdtriIJ: (Provide I brief.conciJe dclcriptiOll of bat the propoIed IlCtivity inYOlves.)

This acDvhy iDvoIves lbc qualificaboa of the recently UParaded Fuel HudliD, Build.iq (fHB)aue (OHC030)to Ii aJe failure proof.Previous revWOOl of this EC performed the physical wan to replace abee.ill.moUey system (troOey.hoistl and boots)aDd the aane NIlWlY rail clips to iDcreae&be capKiry IIId to satisfy abe ASME 00-1 require 1bia reviJioa (004)iDvolvcs no physical wolt but raIher provides necelM"Y.,wneD" aDd supportiD, documenIatioa 10 demo"" lba&the FHB crane may be classified

..siap failan proof in KcordaDce wi repJatcry Juidelina aod iDduIary doc:umeDta:

ASMB OG-I (for the maiD bo'syIte IIId the trolley deck).NUREO-0554 (for the 1Ui!iary hoist)ud NtJIlE(M)6J2.

Appcudb C (for lbe bridJC ry tall).This JUidaacc wu applied IIId complied with to upp'Ide abe CT'IDe 10 siDaie failure proof.Sublequently.

abe uppwIed c:nae IDly be UIIld to perform itl oormaI dcsip fwK:tiOIl includiD,lbe fubn movemeal of fueJ casb peDdiD.repJalory review aDd approval via IOCFR72.Wilh ODe eucpcioa, Ibis IlCtivity dora DOl addreu IiftiD'devica and NRC JUidance reprdin, such.This Klivity also replaca the c:raoe nmwlY rail clips to increase rhe CllpllCity and to fy the ASME NOO-I requiRmeata.

A trolley beam IUJlPOI1ed from the cnae Jirder'nI also ualyzed for lifted 10*1 ud., 1oIIdiD...pM of the desip cbanp.__lor AdMtJ: wh)'die propelled activity II bei.., perfDmlllCd.)

A siJlp failure fWOOf c:rue II DeCaIIrY to tupplI1the 1DdepcDdc..Spe Fuel St<<qe lnstaIJaion (lSF'S1)project where sprat ruel will evanully be IIored In dry c contaUlm outdoors (pcDdiD, repIaIory revie aDd lppfOVai via IOCFR72).The uppded siap-1'aiJI.ft prooIFHB crane is iDleDded 10 millimizo the iDCreallld ri k assoclllCd

'dllbc iJIoaIcd frequcacy of handlin, heavy IoIdI iaside the FHB II wen*provide*meuurc: 01 defe-ilHlepdL The potcaliaJ cOIlIIlqUeIJCCl from aD ISPS)calk drop evall Cd.amqed spent fuel'ide the c)WImIIIed upJI"Ide 01 abe FHB c:nae 10 sinp faiJlft proof.Ju I result of Ibis IC1iviry.the C drop acdde:D1 is DO Ion,er cOftlidered I credible evCllL The Fuel HaadliDJ Bwldio, Crue is briJl'modifted to meet abc siaaJe faillft proof c:ritaia of ASME NOO-I.As part of the smale faillft proof cnoe crileria, the fuIJ lnDIVene horizontal seismic 10*1 (pe:rpeodjcu1w to the c:nae NIlway rail).aDd the c:nae supportiq strueblre (abe Fuel HIIIdIiDa BuiJdinJ)should be COIIIiItcaI with the desip requiremeDI.I for 2004.A coupled model for me crane aDd the Fuel HaDdIin.BuildiDJstt\CRD'e wu aaaJyzed (sec cakulalioa 4.L4-BRW-IG-1-.5)IIXI the resultiq fon:es aDd IDOIDCIlII evalualed for aiDeal COqJODed The new JOIIds lime crane runwa), rail are applied to 0111)'ODe c:nae runwlY rail The oriPoal c:rue run.y rail clips do DOl hive eaoup CIplCiay to resi abe load from lbe borizoDtai rorer, Therefore, DeW clipl have beeD desiped to incn:aae the clpKity and sad fy the ASMB OCJ.I requimDcDtl.(tee CakulaDoo 8.I..9-BRW-ID-OI05-S).

The aew rail dipl Mft iDItaIJed pel'revisioa 2 of Ibis£C.AddidoaaJly.

the IIrUCbI"e WII analyzed 1*1 orrev"'002.The'0.59 Bvaluatiall is rc:vUed u 1*1 of revision 04 oftbe dcsip cbanFIdlftsIIbose chID The trolley beam (TB-I)supported fro abe aane JinIcr w analyzed such that electric hoist could be mouJlted 011 the beam 10 be UICd for cquipmeat IDd milc:cUaneous rools.This" ualyzed io veodor caJcuIMioa 136272-'2.

m.t ftI AdtritJ: (Discua bow abe activity irnpKtl p....openbo desip or aalety aoaIyses delcribed io the UFSAR.)The proposed IC1iviry wiD be tr'IDSI*aR to pJ.t ope:Btio 1beIe c do DOl advmely alter FHB c:raoe openDoa.Tbo c:h does DOl advenely imper:t the UFSAR delcribed desip fuDctiOll 01 me FHB cnae conlrOl Tbe crue raMi ismic:ally qualified for desip bail eartbqu.kr:s

'dI*12.5-to11 ad.The loads OQ the fHB struel1lrc:

remaiD.cccptabIe.

A c:oupled model for abc c:nne and the Fuel RaadliD, BuildiD,5IrUCtUI'e analyzed (see calculaboa 4.I....BRW-lo-oo:5I-S) and o 50.59 REVIEW COVERSHEET (for EVAL)StadoalU.):.""""""""""*, Adfrity'lDeam'" NIlI_!I':.

Ie Jft35J*UfSAI Qepn'1J.f13 LS-AA-I04-IOOI Revision 3 Pqe2of3 R........NDlIIIIN!I':

_OH..-r-II.:lOlm..._

the rauJtin, forca ad RIOmelUl to'cridcaJ compoaau.The new 10Ids al the c:rane NnWIY rail are applied to oaJy ooe crue nmwlY rail The DCW c:ntlC nmway rail dips (illSlalJcd vii In.2 of dliI EC)have IUfficieaI cap.:ity 10 raid the load from!he dl:coupJcd hDrizollll1 fon:e aad urisly!he ASMB NOQ..I require (sec CaJculaboa The troUey beam ('rD-l)from the crane Jirdcr baa been demoftlldled 10 be qualified for the assoc:ia&led Io.da.The EvaJUIIion ilrevised pu1 ofrevWoa 04 oftfR ap cbaQ toIbotc There it DO cblDae lO p hues of any sse described in die UFSAR;however.1 new melbod of evaluatioa it uaed in the saI'dy aaalysi tbal is DO(described in the UFSAR.Movemenl of heavy loads over the spcm fUel pool will still be p-ohibilcd wilbout evaluatiOG to'the specific purpose.Also.inlCrlocb (elec1rical) c:oatinue lO prcvcnllDln'emCDl of heavy bdI 0 er abe ape fuel poot.ac1udina the.I area, durio, cask 10Idi1Jl Ktivirics.

The drop aaaIysiJ dcsaibed in the UFSAR.Section wiD DO 10 be credible wilb the JinaJe faille poofAIB crane ill c:oajuoc:tion wilb the use of sinJlc failure proof riuiDJ.0tbc:nriJe.

lbcR lie DO other IC'CideIIII iaIed'dllhe FHB crane clelcribed ill the UFSAJl No chili weR found to be advenc 10 the dcsip it or safely anaJysis of the FIlII cnoc.SIu_uyoi Ccaai.ui*for AdMtJ'l5I.5J Rmew: (Provide justification for abe cODClusioa, iacludina lUf6cienl delaillo recopble and unde:rsllDd the DliaIlfJumeau Iadiq lO IJlc CODCJusjoa.

Provide more Iba I simple meDlUI Scrc:cniDa

.50.59 EvaJualioa., or I Lice Ameadmc1ll Requesa.II applicable.

DOl rc:quircd.)

ThiJ activity been COGIcrVllivdy scrceaed lince I DCW methodoIOI)'

used.Spec:ifially, the proposed activhy an aJlCrUlive cvaJuadoa mclbodolOlY I'CCOIDIDCIIdcd by the NRC (ASMB NOG-I)in Cltllblishio.

lhc fHB crane dcsip bMa 8Dalyscs.The Braidwood SwioIlllpll'llded FHB CJ'IIIe is iJi c:onformaDcc wilb lhc rep1alory ad iDdusuy JUidacc documeau refemad in NRC JUS 200S-251Dd NRC RlS 2005-25, Supplemau I;and,lbcrcfore.

qualifia'DJIc failure JROf.1be ICtivity provides dde.iJt.dcpdJ meaIInI uch.electrical interlocb th.a prevCDI heavy Jo.d IIIOVeIIIeDl over irnldiaIed fuel weU use of I liD....failure proof bandJln, lystI:m 10 pre'VCDl1oad drvpa.Scctioa'.1.6 of NUREO-0612, definel dlc criteria for liccnaecs to implemc 1 sin&Jc faUun: proof baDdU11I syllCm and rcfcmIca Appendix C to NtJREG..06 J2 for JUiddina to uPJ1adc exiltin, cruea.The ICdvity wiJl reduce dlc frequcacy of occurrence of I c drop accidan (i.e., uw.ka ir no loaF c:nd1b1e);

willno advcne affecl on the Fuel Haadllq Buildm,.IU\IClln or speDr fuel pool;and, lbercfcn.miaimizc the comequences of accidents or malftmctioDs.

1be cruc wiD colltiauc ro ftmc:tioa IDd openIC dclcribed in the UFSAR wilb procedwaI levisi 10 1 small exlCDl (such..rcmovll of the requirclDCllllO use II'CSO'IiDt 00 the main book bC1I over abe speal fuel pool)to reflecl the liJIaJe failure proof MbIre ol the auc.Tbeaefore.

proc:ed Ire DOl adversely alfecred.The c:ranc baa been lCSICd per the l'CqunmcDtl of ASME NOG-I.which mcclI the rcqt;IiRJDCIIII ofNlJREG.()5j.41Dd ANSJ B30.21O demoasD:'lle ill compliance with the upar_crircria.While DDt Idve:ne, dlc uppaIe ollhe craDC 10 liq1e fail," proof baa beea performed to ne er metbodolopa thuI described lJI the UFSAJl However, these JDdbodoIoIies have bealrevicwecl ad appovcd by the NRC fot io ulJll'llU' existiat c:naa.These newer mcthodoloJia will be added 10 tfR UFSAR to ref1ec:11heir use aDd subleqQt:lll cornpliuc:e wilb Lbc reqliftmeolL Baaed 011 Ibis.lberc iI DO impK1 OG tfR Tec:Jmic:aI Spedficllion or Opcntin.Lice It II concluded rbaI dlc propoted lCuvily may be implcmcated wilboul: prior approval from abe NRC.A.IiM......ADaclI all'0.'9 Revic forma completEd.

1IIppI'OpUIe.(NOTE: if both 1 ScRemna&ad EvaJuation lie complct.cd, DO ScrecDiIll No.is rcquired.)

50.5'Sc:reeninl No.50.5'Evaluadon No.BRW-E-2011-3 50.59 REVIEW COVERSHEET (for EVAL)

AcdvitylDoeumat Number: BC 36'351" UFSAR Chagge'13-073 Tide: Da Cask Storace ProW-FBB Crane Upgrade to SigJe-Failure Proof Forms Aamched: (Check all that apply.)H AppBcabWty Review tj.

LS-AA-I04-IOOI Revision 3 Page 3 of3 RevisioD Number: 004/00 Rev._Rev._00_

StationlUnit(s):

50.59 REVIE"V COVERSHEET FORlVI Braidwood Units 1&2 LS-AA-I 04-1 00 I Revision 3 Page I of 5 AclivitylDocument Number: EC 379516 (Unit 1)EC 379517 (Unit 2)Revision Number: 0, 0 Title: Installation of Bypass Test Capability for 7300 Process Protection System and Nuclear Instrumentation System NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity: (Provide a brief.concise description of what the proposed activity involves.)

The proposed activity (Bypass Test Instrumentation (BTl)Modification) will make hardware moditications to both the Nuclear Instrumentation System (NIS)and 7300 Systems to allow bypassing an instrument channel without lifting leads or installing temporary jumpers.A bypass panel will be installed in the NIS cabient and will provide a second source of 118 V AC power in place of the output of a bistable function.Replacement test cards for the 7300 System will provide the capability to bypass a channel.In addition, an alarm window will he changed in the Main Control Room (MCR)to annunciate a channel in bypass and the Sequence of Events Recorder will provide information of the specitic channel which has heen bypassed.The current Technical Specification allows an inoperable channel to he placed in bypass to allow testing of another channel.A license amendment is required to implement routine surveillance testing with a channel in bypass.Status of the bypassed condition will be provided hoth in the Main Control Room and locally via lights on the cards in the cabinets.Both systems will retain the ability to test with a channel in the trip condition.

This 50.59 evaluation reviews the installation and design of the BTl system.In order to test in bypass a License Amendment Request is required to revise the current Technical Specifications.

The following 7300 channels will be affectedI.Loss of Reactor Coolant Flow-Reactor Trip)OverTemperature-Reactor Trip.1 OverPower-Reactor Trip 4.OverTemperature-Turbine Runback**5.Overpower 6T-Turbine Runback**6.Low-Low T avg (P-12)-ESFAS Interlock 7.Low T"e-Feedwater Isolation (coincident with RX Trip)8.Pressurizer Pressure-Low

-Reactor Trip 9.Pressurizer Pressure-High

-Reactor Trip 10.Pressurizer Pressure-Low

-Safety injectionII.Pressurizer Pressure-P-I I ESFAS Interlock 12.Pressurizer Level-High

-Reactor Trip 13.Steam Generator Level-Low-Low

-Reactor Trip and Aux Feedwater Initiation 14.Steam Generator Level-High-High

-Turbine Trip and Feedwater Isolation 15.Steam Generator Level-Low Alarm**16.Steam Flow Feed Flow Mismatch**

17.Steam Flow Loop**18.Steamline Pressure-Low

-Safety Injection and SteamIine Isolation 19.Steamline Pressure Negative Rate-High-Steamline Isolation 20.Steamline Pressure Low Alarm**21.Turbine Impulse Pressure (P-13)22.Containment Pressure-High-I

-Safety Injection 23.Containment Pressure-High-2

-Steamline Isolation 24.Containment Pressure-High-3

-Containment Spray Actuation&Phase B Isolation 25.Reactor Coolant System Temperature Wide Range Loop**26.Reactor Coolant System Pressure Interlock (wide range)**27.RWST Level Low-Low Switchover to Containment Sump Interlock and Alarm StationlUnit(s):

50.59 REVIE"V COVERSHEET FORlVI Braidwood Units I&2 lS-AA-!04-1 00 I Revision 3 Page 2 of 5 ActivitylDocument Number: EC 379516 (Unit 1)EC 379517 (Unit 2)Revision Number: 0,0 Title: lnstallation of Bypass Test Capability for 7300 Process Protection System and Nuclear Instrumentation System The following NIS Comparators to be bypassed include:*Power Range-High Flux Reactor Trip (Low setpoint)*Power Range-High Flux Reactor Trip (High setpoint)*Power Range-Overpower Rod Stop C-2***Power Range-P-I 0 Permissive

Power Range-P-8 Permissive

Power Range-High Flux Positive Rate Reactor Trip.**The above loops will have the new bypass cards/pane!

installed as part of this change but are not addressed by Technical Specifications.

Reason for Activity: (Discuss why the proposed activity is being performed.)

Installation of the new hardware will allow a channel to be bypassed without lifting leads or installing jumpers.The current plant Technical Specifications allow an inoperable channel to be bypassed to allow surveillance testing of other channels.The ability to place the channel in trip will still exist with the new hardware.Therefore this function is not affected.With a Technical Specification change, the plant would be able to perform routine testing with a channel in bypass instead of a tripped condition.

The Reactor Trip System (RTS)and Engineered Safety Features Actuation System (ESFAS)utilize 2-out-of-3 andout-of4 coincidence logic from redundant channels to initiate protective actions.Within these systems, analog channel comparators are currently placed in the tripped state for channel testing.With a channel in the tripped condition.

a second comparator trip in a redundant channel caused by human error, a spurious transient.

or channel failure would initiate a reactor trip or safeguards actuation.

With a channel in the tripped condition.

the logic now becomesaI of 2 oraI of 3.With implementation of this technical specification change, the spurious reactor trip or safeguards actuation will be avoided since the partial trip conditions that would have been present are eliminated by placing the channel in bypass, and the coincident logic is maintained as requiring signals from two additional channels to actuate the protective function.The logic with a channel in bypass becomes 2 of 2 or 2 of 3.This would reduce the likelihood of an inadvertent plant trip due to a spurious signal on another channel or human error.Installation of the bypass circuitry in the permissive logic circuits (lout of 2 and 2 out of 4)allow testing of the circuits without installation of jumpers and will ensure the entire circuit can be tested from the control cabinets to the solid state protection cabinets.The change to Technical Specifications 3.J.l for RTS instrumentation and 3.3.2 for Engineered Safety Feature Actuation System (ESFAS)Instrumentation to allow surveillance testing in bypass will be submitted to the NRC as part of a license amendment request (LAR).Effect of Activity: (Discuss how the activity impacts plant operations.

design bases, or safety analyses described in the UFSAR.)The RTS and ESFAS currently utilize 2-out-of-3 and 2-out-of-4 coincidence logic from redundant channels to initiate protective actions.Within these systems.analog channel comparators.

with the exceptions of the Nuclear Instrumentation System (NIS)out-of-2 functions and the ESFAS containment spray function are currently placed in the tripped state for channel testing or in response to a channel being out of service.With a channel in the tripped condition.

a second comparator trip in a redundant channel caused by human error, a spurious transient.

or channel failure would initiate a reactor trip or sateguards actuation.

StationlU nit(s): 50.59 REVIEW COVERSHEET FORiVI Braidwood Units 1&2 LS-AA-I 04-1 00 I Revision]Page]of 5 ActivitylDocument Number: EC 379516 (Unit 1)EC 379517 (Unit 2)Revision Number: 0,0 Title: Installation of Bypass Test Capability for 7300 Process Protection System and Nuclear Instrumentation System The current Plant Technical Specitications allow an inoperable channel to be bypassed to allow surveillance testing of other channels.To accomplish this.the plant must install jumpers or lift leads in the control cabinets.This moditication will install the hardware necessary to allow an inoperable channel to be bypassed without installation of jumpers.The permissive circuits will be modified to include channel bypass test card.This will allow a channel to be placed in test which will allow the permissive logic to be met while the loop is being tested.The bypass test card bypass function will allow the circuit 10 be tested from the comparator output in the control cabinets (7]00 or NIS)to the Solid State Protection System (SSPS)cabinets to ensure the SSPS relays function properly.Administrative Controls (Procedures, Indication and Alarm)will ensure the circuits are properly contigured during and after testing.A technical specification revision (LAR)is required to implement the change to allow a channel to be bypassed for routine testing.With the Technical Specification amendment.

this modification includes bypass circuitry for NIS power range reactor trip functions and the 7300 Process Protection System reactor trip functions and ESF functions identitied in Tables I&2 of WCAP-17349 (Reference 6)and listed above.With implementation of the BTl and the Technical Specification change.spurious reactor trips or sateguards actuation will be reduced.This is because the partial trip conditions that would have been present are eliminated.

and the coincident logic is maintained as requiring signals from two additional channels to actuate the protective function.This provides the benefits of reducing the challenges to the plant safety systems which may result from spurious actuations and thus potentially increasing plant availability.

Administrative controls are provided to prevent the simultaneous bypassing of more than one redundant protection set at anyone time.and to restore the system to normal operation.

Summary of Condusion for the Activity's 50.59 Review: (Provide justification for the conclusion, including sufticient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation.

or a License Amendment Request, as applicable, is not required.)

I.The new hardware of the bypass trip instrumentation (BTl)was designed in accordance with the existing licensing basis.The new BTl components are highly reliable, have similar failure modes and qualitied consistent with the original UFSAR design requirements and will normally function in the same manner (2 out of 4 logic or 2 out of 3 logic).A number of Chapter 15 accidents require these channels to operate or could be the initiator of the Chapter 15 transient.

such as inadvertent satety injection.

Given that the hardware is qualified consistent with original design basis requirements and is expected to have similar reliability as the existing components, the installation of the BTl modification does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.2.The BTl system will not affect the function or operability of the protection grade channels that will be tested using the BTl system.Measures (e.g., reliability evaluation, isolation capability evaluation and qualification testing)have been taken to ensure that the BTl will not be subject to common mode failure or affect the operability of the protection systems it will be installed in or the protection functions it will interface with.Additionally, the BTl System has been designed to fail safe for credible failures.A bypass relay failure results in the channel that is being bypassed returning to its normal operation conditions (i.e., the channel will be placed on-line in the non-tripped state).Potential relay contact failures that would inadvertently place a channel in bypass are detectable by observation of the bypass status lights and redundant channels are capable of pertorming the required actuation feature.Therefore, the installation of the BTl modification does not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.3.Implementation of the bypass testing capability does not atfect the integrity of the fission product barriers utilized for mitigation of radiological dose consequences as a result of an accident.Plant response as modeled in the safety analyses is unaffected.

Hence, the offsite mass releases used as input to the dose calculations are unchanged from those previously assumed.Therefore, it is concluded that the installation of the BTl modification does not resuh in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.

StationlUnit(s):

50.59 REVIEW COVERSHEET FOAAI Braidwood Units I&2 LS-AA-I 04-1 00 I Revision 3 Page 4 of 5 ActivitylDocument Number: EC 379516 (Unit 1)EC 379517 (Unit 2)Revision Number: 0.0 Title: Installation of Bypass Test Capability for 7300 Process Protection System and Nuclear Instrumentation System 4.The Bll system is designed to perform its intended protective functions (i.e..testing/troubleshooting in the bypassed condition) without increasing the consequences of a malfunction of equipment important to satety previously evaluated in Ihe UFSAR.Credible relay failures (open coil)will prevent the channel trom being bypassed or will return the channel to its normal condition if it was in a bypassed state.This would ensure the protective teatures of mitigating systems would still perform as required.Potential relay contact failures that would inadvertently place a channel in bypass are detectable hy observation of the bypass status lights.Redundant channels would still be capable of actuating the protective feature.similar to when a channel is currently placed in bypass when the channel is inoperable.

Therefore.

it is concluded that the installation of the BTl moditication does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.5.The new Bll hardware does not aftect accident initiation sequences or response scenarios as modeled in the satety analyses.No new operating contiguration is being imposed by the proposed change.The bypass test instrumentation has heen designed to applicable industry standards.

Fault conditions.

failure detection.

reliability and equipment qualitication have been considered.

Therefore.

it is concluded that the installation of the Bll moditication does not create a possibility for an accident of a ditferent type than any previously evaluated in the UFSAR.6.The BTl system will not aUect the function or operability of the protection grade channels that will be tested using the Bll system.Measures have been taken to ensure that the Bll will not be subject to common mode tailure or affect the operability of the protection systems it will be installed in or the protection tunctions il will interlace with.Additionally.

the Bll System has been designed to fail sate for credible tailures.A bypass relay failure results in the channel that is being bypassed returning to its normal operation conditions (i.e.*the channel will be placed on-line in the non-tripped state).Potential relay contact tailures are detectable by observation of the bypass status lights and redundant channels would perform the required protective teature.Based on this assessment.

it is determined that the installation of the BTl moditication does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR.7.Implementation of the bypass testing capability does not affect the integrity of the tission product barriers utilized tor mitigation of radiological dose consequences as a result of an accident.Plant response as modeled in the satety analyses is unaffected.

Hence.the otfsite mass releases used as input to the dose calculations are unchanged from those previously assumed.Therefore.

it is concluded that the installation of the Bll moditication does not result in a design basis limit for a tission product barrier as described in the UFSAR being exceeded or altered.8.The assumptions and methods used in the plant accident analyses are not affected by the installation of the Bll.The RTS and ESFAS design and licensing basis continue to be met as discussed in Reterence 6.The equipment is designed and qualitied consistent with the orignal licensing basis requirements discussed in UFSAR Sections 7.1.7.2 and 7.3.Therefore.

it is concluded that the installation of Ihe Bll modification does not result in a departure trom a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.Based on the above.the proposed hardware modification doesnotresult in an unreviewed safety question and may be implemented without NRC approval.However.routine surveillance testing with a channel in bypass would require a Technical Specification Change and NRC approval.Attachments:

Attach all 50.59 Review forms completed.

as appropriate.

Forms Attached: (Check all that apply.)

StationlUnit(s):

50.59 REVIEW COVERSHEET FORlVI Braidwood Units 1&2 LS-AA-I 04-1 00 I Revision.3 Page 5 of 5 ActivitylDocument Number: EC 379516 (Unit 1)EC 379517 (Unit 2)Revision Number: 0,0 Title: Installation of Bypass Test Capability for 7300 Process Protection System and Nuclear Instrumentation System o Applicability Review o 50.59 Screening[gI 50.59 Evaluation 50.59 Screening No.50.59 Evaluation No.BRW*E*2011*40 Rev.Rev.0------

50.59 REVIEW COVERSHEET FORl\tI LS-AA-I04-IOO1 Revision 3 Page lof2 StationlUnit(s):

Braidwood I 01 ActivitylDocument Number: 50.59 Evaluation

BRW*E*2011*99 Revision Number: 2.Title: Isolation of the lD Loop Pressurizer Spray Flow Path to Terminate lRY8050 Leakage NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the of 10 CPR 50.59(d)(2).

Description of Activity: (Provide a brief, concise description of what the proposed activity involves.)

This activity is a configuration change to the normal line up for the pressurizer spray line from the ID reactor coolant loop through IRY455B Pressurizer Spray Valve.The proposed activity will isolate the pressurizer spray valve IRY455B by closing the pressurizer isolation valves IRY022.Upstream Isolation for Loop D Pressurizer Spray Valve, and IRY023, Downstream Isolation for Loop D Pressurizer Spray Valve, Procedure OP-AA-108-1 0 I, Control Of Equipment And System Status, has an attachment for review of abnormal component positions.

The abnormal component position review (ACPS#11-081)per OP-AA-108-1O I Attachment 2 has concluded a 10CFR50.59 review is required for the isolation of the prsessurizer spray flow path.This IOCFR50.59 review is in support of the ACPS#11-081.Reason for Activity: (Discuss why the proposed activity is being performed.)

During Cycle 16 on Unit I.Reactor Coolant System Unidentified Leakage increased above the base line value.A leak inspection performed inside Containment identified a mechanical joint leak originating from valve I RY8050, Pressurizer Spray Valve lRY455B Bypass Flow Valve.The proposed activity will isolate the mechanical joint leak by closing the upstream isolatjon valve lRY022 and down stream isolation valve lRY023 in order to isolate valve IRY8050 from the reactor coolant system.Effect of Activity: (Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)Two separate, automatically controlled spray valves ORY455B and IRY455C)with remote manual overrides are used to initiate pressurizer spray.In parallel with each spray valve is a manual throttle valve ORY8050 and IRY80SD which permits a small continuous flow through both spray lines to reduce thermal stresses and thermal shock when the spray valves open, and to help maintain uniform water chemistry and temperature in the pressurizer.

A manual valve is located upstream (IRY022 for valve IRY455B;IRY024 for valve IRY455C)and downstream ORY023 for valve IRY455B;IRY025 for valve I RY455C)of each spray valves for isolation.

The Pressurizer Spray valves are normally maintained in the closed position and open automatically in response to signals from the pressurizer pressure control system.Closing the IRY022 and I RY023 valves to isolate the leaking bypass valve IRY8050 will result in isolation of the pressurizer spray flow path from the ID reactor coolant system loop to the pressurizer.

Pressurizer spray valve IRY455C, from the IC reactor coolant system loop, will continue to be available to operate as part of the pressurizer pressure control system and the bypass for pressurizer spray valve I RY 455C wjJI remain in service.The design basis of the pressurizer spray system ensures that at least 50%of the nominal capacity of the system will be retained with one reactor coolant pump out of service.TQerefore the function of the pressurizer spray system will be maintained by theIC pressurizer spray loop, The Westinghouse Owners Groul)documents.

including Emergency Procedures and Abnormal Operating Procedures.

were reviewed to determine how the spray valves are used in abnormal/emergency situations.

Both types of I)fOcedures use the spray valves when available, however.the Pressurizer PORVs are relied upon as the safety related components.

The spray valves and control systems are non safety related equipment The pressurizer spray system is modeled in some of the UFSAR Chapter 15 non-LOCA analyses.However.since the pressurizer spray system is a non-safety control system.it is only modeled if doing so results in more severe results in the analysis.For transients analyzed to address peak RCS pressure.pressurizer sprays are not modeled.For transients analyzed to address Departure from Nucleate Boiling (DNBl where jt is conservative to minimize pressure.the sprays are modeled.Therefore.

if the sprays are unavailable or if the performance is 50.59 REVIEW COVERSHEET FORlVI LS-AA-I04-IOOI Revision 3 Page 2 of2 StationlUnit(s):

Braidwood 101 ActivitylDocument Number: 50.59 Evaluation

BRW*E*2011*99 Revision Number:!!Title: Isolation of the ID Loop Pressurizer Spray Flow Path to Terminate lRY8050 Leakage degraded, the analysis results for those events would improve and be bounded by those currently reported in the UFSAR The pressurizer spray system is not modeled in the UFSAR Chapter 15 LOCA or steam generator tube rupture (SGTR)analyses.Summary of Conclusion for tbe Activity's 50.59 Review: (Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The isolation of the pressurizer spray flow path will isolate the lRY8050 mechanical joint leak from the reactor coolant system, The spray function of spray valve IRY 455B is not credited in the safety analyses when the operation of the valve benefits the results of the analysis.The spray valves are assumed to operate only in cases where a lower pressure in the pressurizerlRCS would result in more limiting results (i.e" DNB cases).The nominal Pressurizer conditions (Pressure, Level.Temperature) are inputs to the safety analyses and are maintained because the setpoints for the Pressurizer Pressure and Level Control systems are not affected by this activity, In the absence of the ID pressurizer spray loop, the function of the pressurizer spray system will be maintained by the IC pressurizer spray loop.Westinghouse has performed an evaluation that concluded the reduction in spray flow.due to one pressurizer spray flow path isolated.does not result in increased challenge to the PORVs.There is sufficient margin for one pressurizer spray to maintain Pressurizer pressure below the pressure that would initiate automatic action of a PORV, Based on the aforementioned.

this activity can be implemented without NRC ApprovallLicense Amendment.

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)0 Applicability Review 0 50.59 Screening 50.59 Screening No.Rev.[8J 50.59 Evaluation 50.59 Evaluation No.BRW*E*2011*99 Rev.0 StationlUnit(s):

Braidwood 50.59 REVIEW COVERSHEET FORM LS-AA-I04-1001 Revision 3 Page 1 of2 ActivitylDocument Number: BRW-E-2011-100 Title: REVISION 4 OF BwOP CV-16.PZR AUXILIARY SPRAY OPERATION Revision Number: J!NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CPR 50.59(d)(2).

Description of Activity: (Provide a brief, concise description of what the proposed activity involves.)

The procedure change in this activity provides the instructions for the use of auxiliary spray in the event that normal spray is not available, during Cycle 16 of Unit I, while the 10 normal pressurizer spray loop is isolated to contain a mechanical joint leak at the lRY8050 valve.This change is only applicable to Cycle 16 of Unit 1 due to the scope of review performed by Westinghouse as relates to this issue.Although the UFSAR states that auxiliary spray is normally used when the reactor coolant pumps (RCPs)are not in operation (RCPs are the driving head for normal spray), the procedure change in this activity will allow the use of auxiliary spray when RCPs are in operation.

Use of the system will be limited to the time it takes to place the unit in a condition that pressurizer spray is not required.RCP operation has no effect on the use of auxiliary spray or the ability of auxiliary spray to lower pressure.Westinghouse has determined that the use of auxiliary spray in support of a unit shutdown is acceptable provided the established thermal fatigue guidelines are observed.The procedure change in this activity provides the required information for monitoring the temperature limits associated with use of auxiliary spray during Cycle 16 of Unit 1.Reason for Activity: (Discuss why the proposed activity is being performed.)

The procedure change is required to provide the instructions required for operation of the auxiliary spray valve for RCS pressure control.The procedure will only be used in the event that normal pressurizer spray is not available and will only be used to shutdown the unit.Effect of Activity: (Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)Inadvertent depressurization of the RCS is listed as a design transient in UFSAR section 3.9.1.1 inadvertent auxiliary spray operation is listed as one of the initiators of the event, however pressurizer safety valve actuation causes the most severe transient and is used as an umbrella case to conservatively represent the reactor coolant pressure and temperature variations arising from any of the events.The operation of the auxiliary spray valve and the potential for an open failure of the auxiliary spray valve is bounded by the inadvertent opening of a safety valve.For design purposes this transient is assumed to occur 20.times during the 4O-year design life of the plant.."..Westinghouse hasdeterminedthat the use of auxiliary spray in support of a unit shutdown is acceptable provided the established thermal fatigue guidelines are observed.The procedure change in this activity provides the required information for monitoring the temperature limits associated with use of auxiliary spray.

StadonlUnit(s):

Braidwood 50.59 REVIEW COVERSHEET FORM LS-AA-I04-1001 Revision 3 Page 2 of2 , AcdvitylDocument Number: BRW*E*2011*I00 Title: REVISION 4 OF BwOP CY*I'.PZR AUXILIARY SPRAY OPERATION Revision Number:!l Summary of Conclusion for the Acdvity's SO.59 Review: (Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

The procedure change in this activity provides the instructions for the use of auxiliary spray for RCS pressure control in the event that normal spray is unavailable during Cycle 16 of Unit 1.The procedure will only be used to support a unit shutdown for RCS pressure control.Westinghouse has determined that the use of auxiliary spray in support of a unit shutdown is acceptable provided the established thermal fatigue guidelines are observed.The procedure monitors the thermal fatigue guidelines.

Since the open failure of the auxiliary spray valve is bounded by the failure of a pressurizer safety valve open failure, there is not more than a minimal increase in the frequency of occurrence of an accident.Use of the auxiliary spray valve will not cause more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety because the auxiliary spray valve (lCV8145)is a fail closed valve.In the event the valve fails open, the subsequent RCS depressurization is bounded by the open failure of a pressurizer safety valve.The revision to the procedure in this activity does not result in more than a minimal increase in the consequences of an accident or malfunction of an SSC important to safety previously evaluated in the UFSAR.The pressurizer spray valves are a non-safety related control system and are only modeled in Chapter 15 non-LOCA analyses if doing so results in more severe results in the analysis.For transients analyzed for peak RCS pressure, sprays are not modeled.For DNB related transients, where it is conservative to minimize pressure, the sprays are modeled.The Pressurizer sprays are not modeled in the Chapter 15 LOCA or steam generator tube rupture (SOTR)analyses.Guidance within the procedure will ensure that operation of the auxiliary spray flow with an RCP in operation will have a minimal impact on the charging header pressure and flow, and will maintain parameters similar to that experienced during normal plant evolutions such that the activity does not change the ability of the charging header to maintain pressurizer level, supply of normal charging and the supply of RCP seal injection flow.The effect of an inadvertent auxiliary spray operation is bounded by the failure of a pressurizer safety valve in the open position, therefore use of the auxiliary spray valve does not create the possibility for an accident of a different type than any previously evaluated in the UFSAR or create the possibility of a malfunction of an SSC important to safety with a different result than previously evaluated.

Operation of auxiliary spray is controlled within the established thermal fatigue guidelines for the system.Therefore, the procedure being revised in this activity has no effect on any design basis limit for fission product barriers nor is it used to perform any evaluations that are described in the UFSAR.Based upon the above, the procedure revision may be implemented as proposed without prior NRC approval.Attachments:

Attach all 50.59 Review forms completed, as appropriate.(NOTE: if both a Screening and Evaluation are completed, no Screening No.is required.)(NOTE: if both a Screening and Evaluation are completed, no Screening No.is required.)

Forms Attached: (Check all that apply.)o Applicability Review o 50.59 Screening 181 50.59 Evaluadon 50.59 Screening No.SO.59 Evaluadon No.BRW*E*2011*100 Rev.Rev._0=--_

StationJUnit(s):

50.59 REVIEW COVERSHEET FORM Braidwood Units I&2 LS-AA-I 04-1 ()(Revision 3 Page 1 of 4 ActivitylDocument Number: EC 380047 , EC 380048 Technical Specification Basis Change Requests12-001 and 12-003, DRP 14-020 and DRP 14-021 Revision Number:---'lu,....

_Title: Installation of Backup Power Supplies for the Steam Generator Power Operated Relief Valves (SG PORV's)NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CPR 50.59(d)(2).Description of Activity: (Provide a brief, concise description of what the proposed activity involves.)

The proposed activity will install a battery backed-up 480V AC uninterroptible power supply (UPS)for two of four SG PORV's (I ,2MSOI8D and C)on each unit.The UPSs will be physically installed in the 414'EI.Electrical Penetration Rooms for Division 11121 and in the general area of the 426'El.for Division 12122.The current primary 120V AC power supply to 1 (2)PA33J from MCC 1(2)31X2 will be replaced with an instrument bus power supply from Bus 1(2)1P03J.

The backup power supply for I (2)PA33J from MCC 1(2)3IXl is unaffected by this change.I (2)PA33J provides control power to the pressure control loop for the SG PORV's.Instrument Bus I (2)1P03J is normally powered from an uninterroptible power supply via the 1(2)13 Inverter.The diverse power supplies to the SG PORV's will provide 90 minutes of uninterrupted power supply to the selected SG PORV's in the event that their normal power supplies are lost.UFSAR Change Request 14-020 (Unit l)and 14-021 (Unit 2)will update the Section 7.2.2.4.2 Page 7.2-44, Section 10.3.1 Page 10.3*1 to reflect the installation of the UPS for two of the four SG PORV's and instrument power supply to IPA33J.Technical Specification Bases Pages B3.7.4-1 and 3 will be updated to reflect the installation of the UPS on two of the four SG PORV's with a 90 minute battery backup in the event of a loss of the Class IE buses.(Reference Basis Change Request 12-00 I for Unit I and 12-003 for Unit 2)Reason for Activity: (Discusswhythe proposed activity is being performed.)

The SG PORV's hydraulic control units are powered from MCC's on Division 11 or Division 12 on Unit 1 and Division 21 and 22 for Unit 2.Each division supplies power to two SG PORV's.Therefore a potential single failure of an electrical bus would adversely affect both SG PORV's powered from that division.During a Steam Generator Tube Rupture Accident, two SG PORV's are required to support cooldown and depressurization of the Reactor Coolant System (RCS)for the Margin to Overfill (MTO)event The ruptured generator is isolated and the SG PORV is closed to mitigate the consequences of the accident per the emergency operating procedures.

Note the SG PORV fails closed on loss of power.If a single failure affects the power supply to two of the remaining SG PORVs, only one SG PORV would be available for cooldown and depressurization.

The current licensing basis for the SGTR Accident (MTO)assumes the worst case single failure is a failure of a single SG PORV to open.However, the failure of an electrical bus would be a more limiting case.Therefore, in order assure two SG PORV's are available to support cooldown and depressurization of the RCS back up power supplies will be provided to one of the two SG PORV's on each division.The instrument bus power supply to 1 (2)PA33J will ensure the SG PORV controls on Division II and 21 are available to operate the valves in the event of a loss of power to the electrical bus to the MCCs on those divisions.

Instrument Bus I (2)1P03J is normally powered from an uninterruptible power supply via the 1(2)13 Inverter.I (2)PA34J already has diverse power supplies, one from a Division 12122 instrument bus and one from a motor control center Division 12/22 120V AC ESF MCC distribution panel.Effect of Activity: (Discuss how the activity impacts plant operations, design bases.or safety analyses described in the UFSAR.)Plant operation will not be changed by the installation of UPS's for the SG PORV's and repowering 1(2)PA33J from an instrument power bus.The SG PORV's will continue to function in the same manner and the safety function as described in Technical Specification Basis B3.7.4 will be maintained.

StationlUnit(s):

50.59 REVIEW COVERSHEET FORM Braidwood Units I&2 LS-AA-I 04-1 00 Revision 3 Page 20f4 ActivitylDocument Number: EC 380047 ,EC 380048 Technical Specification Basis Change Requests12-001 and 12-003.DRP 14-020 and DRP 14-021 Revision Number:

_Title: Installation of Backup Power Supplies for the Steam Generator Power Operated Relief Valves eSG PORV's)The current design basis for the SG PORV's for SGTR accident MTO analysis assumes the worst case single failure is the failure of a SG PORV to open on an intact stearn generator.

By installing this change, this input would remain valid even if an electrical bus failure adversely affected the normal power supply to two SG PORVs on the same divsision.

In other accident or transients, one or two SG PORV's are credited.These transients are not adversely affected.The safe shutdown analysis in Section 2.4 of the fire protection report is not affected.Various fires scenarios adversely affect the SG PORV power supplies.The fire protection report credits the safety valves as a means to remove decay heat until local control of the SG PORV's can be accomplished using the hand pumps located near the valves.Summary of Conclusion for the Activity's 50.59 Review: (Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

I.UFSAR Chapter 15.1.4 discusses the inadvertent opening of a SG PORV.Given that the new equipment is qualified consistent with original design basis requirements, is highly reliable, and is expected to have similar reliability as the existing equipment, the valve will continue to fail close on a loss of power, the backup power supply will reduce the likelihood of a loss of the power supply, and all the control loops will have instrument power supply, the installation of UPS's and repowering 1(2)PA33J from an instrument power bus does not result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR.2.The likelihood of occurrence of malfunction has not been increased because thenewequipment is considered highly reliable and qualified for the application.

The UPS is also designed to fail safe such that a failure of the UPS inverter will allow the normal power supply to continue to feed the SG PORV affected.In addition, by adding a backup power supply to two SG PORV's the likelihood of a loss of power is decreased.

The control system is not affected by this change.Therefore, the installation of the UPS's on two of the four SG PORV's and repowering l(2)PA33J from an instrument bus will not result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.3.The SG PORV's are discussed in various UFSAR Chapter 15 accidents as a release path for various dose calculations.

The function of the valves in the accident response is unchanged.

The release path is not affected nor is the flow from the valve affected.The valve will continue to fail closed on a loss of power.The operator actions for a stuck open SG PORV are unchanged and the plant response as modeled in the safety analyses is unaffected.

Hence, the offsite mass releases used as input to the dose calculations are unchanged from those previously assumed.Therefore, it is concluded that the installation of UPS's and repowering 1(2)PA33J from an instrument power bus does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.4.The consequences of a malfunction (SG PORV)stuck open in the Steam Generator Tube Rupture Accident (Dose Case)and the Reactor Coolant Pump Shaft Seizure are unchanged since the flow rate and controls are not affected by this installation of the UPS or repowering I (2)P A33J from an instrument bus.The actions to mitigate this malfunction are unchanged as an operator will be dispatched to close the SG PORV upstream isolation valve.Therefore, it is concluded that the installation of UPS's and repowering I (2)PA33J from an instrument power bus does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.

StationlUnit(s):

50.59 REVIEW COVERSHEET FORM Braidwood Units I&2 LS-AA-I 04-1 00 Revision 3 Page 3 of 4 AetivitylDocument Number: EC 380047 ,EC 380048 Technical Specification Basis Change Requests12-001 and 12-003, DRP 14-020 and DRP 14-021 Revision Number:

_Title: Installation of Backup Power Supplies for the Steam Generator Power Operated Relief Valves eSG PORV's)5.The UFSAR already reviews an inadvertent opening of a SG PORV and a failure of the valve to reclose upon opening, The installation of UPS's will not create any new failure modes.The valves will still fail close on a loss of power from both the current and proposed back-up power supplies.Since these failure modes are already reviewed, the possibility of an accident of a different type is not created.Therefore, it is concluded that the installation of UPS's and repowering I (2)P A33J from an instrument power bus does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR.6.The UPS's will not affect the function or operation of the power supply to the SG PORV's.The UFSAR already evaluated a SG PORV's failing open and failing to close.Installation of the UPS and instrument bus power supply for I (2)PA33J will not affect the results.The SG PORV valve will continue to fail closed on a loss of power from both the current and proposed back-up power supplies.Failure of the UPS inverter will allow the normal MCC to continue to feed the SG PORV.If the UPS fails such that the power supply to the SG PORV control panel is not available, this is similar to a failure of the SG PORV control panel and or a failure of the normal power supply.The result would be the same, Therefore, the installation of the SG PORV UPS's does not create the possibility for a malfunction with a different result.Therefore installation of the SG PORV UPS and instrument power to I (2)PA33J does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR 7.Implementation of the UPS power supplies for the 1(2)D and I (2)C SG PORV and providing instrument power to I (2)PA33J does not affect the integrity of the fission product barriers utilized for mitigation of radiological dose consequences as a result of an accident Plant response as modeled in the safety analyses is unaffected and no parameter which impacts a fission product barrier is changed, Hence, the mass and radioactivity releases used as input to the dose calculations are unchanged from those previously assumed.Therefore, it is concluded that the installation of UPS's and repowering I (2)P A33J from an instrument power bus does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.8.The assumptions and methods used in the plant accident analyses are not affected by the installation of the UPS.The current SG Tube Rupture Accident MTO Analysis assumes two SG PORV's are available for cooldown and depressurization of the RCS.A bus failure on either division could render two SG PORV power supplies inoperable.

This would cause the SG PORV's on that bus to fail closed, The margin to overfill case for the SG Tube Rupture Accident models the ruptured SG as isolated.If the ruptured SG is on the opposite division of the failed division, only one SG PORV would be available for cooldown and depressurization.

By installing the UPS on one valve on each division.a second SG PORV would be available for cooldown and depressurization.

By installing the UPS, the SGTR design and licensing basis (Le., ruptured SG does not overfill)can be met considering a single failure of a bus adversely affecting the power supply to two SG PORV*s.The equipment is designed and qualified consistent with the original licensing basis requirements.

Therefore, it is concluded that the installation of UPS's and repowering 1(2)PA33J from an instrument power bus does not result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.

StationlUnit(s):

50.59 REVIEW COVERSHEET FORM Braidwood Units I&2 LS-AA-l 04-1 00 Revision 3 Page 4 of 4 ActivitylDocument Number: Ee 380047 , BC 380048 Technical Specification Basis Change Requests 12-00 1 and 12-003.DRP 14-020 and DRr 14-021 Revision Number:--'1.....,....1.""'0,""'0...,..0...,,0'--

_Title: Installation of Backup Power Supplies for the Steam Generator Power Operated Relief Valves (SG PORY's)Attachments:

Attach al1 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)o Applicability Review o 50.59 Screening l8J 50.59 Evaluation 50.59 Screening No.50.59 Evaluation No.BRW-E-2011-122 Rev.Rev.1

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Revision Number: Nt A LS-AA-I04-l00 1 Revision 3 50.59 REVIEW COVERSHEET FORM Page.!of.l.StationlUnit(s):

Braidwood Unit 1 Activity/Document Number: TRM Change#12-003 Title: Change In-Core Decay Time for AIR16 NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity: (Provide a brief, concise description of what the proposed activity invol ves)The proposed activity makes the following changes to the Technical Requirements Manual (TRM)to reduce the minimum required In-Core Decay Time (ICDT)for A1 R16 from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 95 hour0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months s:*Braidwood TRM Section 3.9.a,"Decay time," states"The reactor shall be subcritical forthe last 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months s84 hours for A1 R15)." This activity will revise this statement by replacing84 hours for A1R15)" with95 hours for A1R16)."*Condition A under TRM 3.9.a states"Reactor subcrltlcal for<100 hours<<84 hours for A1R15)".This activity will replace"<<84 hours for A1 R15)" with"<<95 hours forA1 R16)".*Surveillance requirement TSR 3.9.a.1 will be revised by replacing84 hours forA1 R15" with95 hours forA1 R16".Reason for Activity: (Discuss why the proposed activity is being performed)

It is anticipated that during AIRI6, work activities wiD be completed and the required plant configuration wiD be established to support commencing movement of irradiated fuel from the reactor vessel to the Spent Fuel Pool (SFP)prior to the current TRM fuel movement ICnT constraint of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after reactor shutdown.Effect of Activity: (Discuss how the activity impacts plant operations, design basis, or safety analyses described in the UFSAR.)The proposed changes will allow starting AIR16 reactor core offloading activities at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after the reactor is shutdown.The Byron and Braidwood spent fuel pool cooling design basis analysis (#BRW-OO-OOIO-M Revision 0)is based on the minimum ICDT of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to starting fuel transfer, however an outage specific evaluation has been performed to support a reduced ICDT for AIRI6.This is in compliance with UFSAR Section 9.1.3.1, which states that outage specific evaluations may be performed, in support of shorter fuel decay times in the reactor, by taking credit for existing margins in the design basis analysis.The limiting case for the design basis analysis for the Spent Fuel Pool is based on one train of Spent Fuel Pool Cooling in operation.

Starting core offload at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after shutdown wiD not result in increasing the design basis heat load for the Spent Fuel Pool (SFP)Cooling System.Moving fuel at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after the reactor is shutdown does result in an increase in the heat load input to the Spent Fuel pool, from the offloadedassemblies,wben compared to starting core offload at 100 hrs or later.However, the cumulative heat load to the spent fuel pool will not be greater than the design basis heat load, since the actual heat load in the SFP due to previously stored fuel assemblies is less than the heat load that was included in the SFP design basis analysis.A significant margin of nearly 3 H:\ICD1\AIRI6\AIRI65059.DOC Revision Number: Nt A LS-AA-I04-1001 Revision 3 50.59REVIEWCOVERSHEET FORM Page..Lo f;l Activity/Document Number: TRM Cbange#12*003 Title: Cbange In-Core Decay Time for AIR16 MBtulbr or about 5%of tbe design basis beat load for tbe SFP has been calcnlated between tbe total beat load tbat is added to tbe SFP during AIR16 and tbe design basis heat load.Tbe resulting maximum bulk water temperature wiD be below tbe design basis maximum bulk water temperature of 162.7 OF.Tbe impact of the bigher heat load on the temperature in the Spent Fuel Pool, witb one train of cooling in operation, will be minimal.In fact, the total heat load increase due to moving fuel at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after shutdown is<3%higher than the heat load that is added moving fuel at 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .This increase is minimal and would result in a minimal temperature increase (estimated to be at 1.5 of with a Component Cooling temperature of 105 OF)for the Spent Fuel Pool.This temperature increase does not have a degrading effect on the Spent Fuel Racks, inclnding the boralpoisonpanels.

Limitations and alarms related to Spent Fnel temperatures have not been changed and are in place during core omoad.In addition, Operating Procedure BwOP FC-l includes a precaution to monitor Spent Fuel Pool temperature frequently and make adjustments as necessary to spent fuel pool cooling, during periods of high heat loads, such as core omoads.The radiological design basis analysis for the Fuel Handling Accident is based on a minimum decay time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to movement of irradiated fuel assemblies (Reference design analysis#BRW*04-0041-M/BYR04-047 Revision 2).Sections 83.9.4 and B3.9.7 of the Braidwood Technical Specification Bases are not being revised as part of the AIRI6 activity since the minimum ICDT for radiological considerations is not being revised and the revised ICDT of 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months for AIR16 still meets the constraint of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .Occupational Radiation Dose Occupational radiation dose will remain within limits.Access to the areas that are affected by the defueling operations is controlled in accordance with station procedures.

Electronic dosimeters are required to continuously monitor the incurred radiation dose in the areas in order to limit personnel exposure to below 10CFR20 limits.These existing controls are not affected.Summary of Conclusion for the Activities 50.59 Review: (Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable.

is not required.)

This activity does not increase the frequency of occurrence of a Fuel Handling Accident or a Loss of Spent Fuel Pool Cooling event, or increase the likelihood of occurrence of a malfunction of an SSC important to safety.The proposed change does not increase the failure rate of refueling equipment or increase the risk of a fuel handling accident due to human error.Spent fuel handling tools will not change, nor will the method or procedures for handling spent fuel assemblies.

Existing administrative controls and precautions remain in effect, only one spent fuel assembly is lifted at a time, and the fuel is moved at low speeds, exercising caution that the fuel assembly does not strike anything during movement.An outage specific evaluation (Design analysis#BRW-OO-OOIO-M Revision OOOW)has concluded that the total aetnal heat load in the Spent Fuel Pool as a result of the reduced ICDT is bounded by the total heat load specified in the design basis analysis.Thus, there is not an additional demand, beyond the heat load considered in the design basis analysis, on the cooling system.This activity does not result in an increase in the consequences of an accident or in the consequences of a malfunction of an SSC important to safety.The otTsite dose resulting from a Fuel Handling Accident with a minimum In-Core Decay Time of 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months is bounded by the design basis Fuel Handling Accident dose witb a minimum ICDT of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .This activity does not create a possibility for an accident of a different type than any previously evalnated in the UFSAR as tbere is no new equipment being introduced, and aU existing fuel transfer equipment is heing operated using existing procedures.

H:\ICD1\AIRI6\AIRI65059.DOC LS-AA-I04-1001 Revision 3 50.59 REVIEW COVERSHEET Page.2....

of:l Activity!Document Number: TRM Change#12-003 Title: Change In-Core Decay Time for AIR16 Revision Number: N/A This activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR.The increase in heat load in the Spent Fuel Pool has been evaluated; although an input parameter (oftload start time)has been changed, the resulting impact on the SFP bulk water temperature analysis is bounded by the design basis analysis.In addition, the local water temperature, fuel cladding temperature, and maximum heat flux have also been evaluated and have been found to be acceptable.

The adequacy of the reduced leDT for AIR16 is based on margin in background decay heat since the SFP is not filled to its capacity.The reduction in ICDT does not result in a change in the internalcontainmentpressure that would represent a challenge to the containment design basis limit internal pressure of 50 psig (Reference UFSAR Table 3.8*4).The maximum cladding temperature for the spent fuel is well below the design basis limit of 2,200 OF from 10CFR50.46 (Reference UFSAR Section 15.6.5).This activity does not change the method of evaluation for the Spent Fuel Pool Cooling System described in the UFSAR or in the SER for the Power Uprate Project.Decay heat input to the spent fuel pool was calculated for the earlier ICDT of 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months using the method described in NRC Branch Technical Position ASB 9-2.This is the same method that was used to calculate the decay heat values that are evaluated the design basis temperature analysis for the Spent Fuel Pool (BRW-OO-OOlO-M).

Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.50.59 Evaluation No.BRW-E-2012-85 Rev.Rev.H\ICD1\AIRI6\AIRI65059.DOC 50.59 REVIEW COVERSHEET FORM StationlUnit(s):

Braidwood 11&2 LS-AA-I04-IOOI Revision 3 Page I of2 ActivitylDocument Number: TRM Change Request No.12-007 Title: Eliminating Action 3.3.y.D from TRM 3.3.y Revision Number: """00"--_NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity: (Provide a brief, concise description of what the proposed activity involves.)

The proposed activity will revise the Braidwood Technical Requirements Manual section 3.3.y, Engineered Safety Feature Actuation System (ESFAS)Instrumentation.

Specifically, TRM Condition 3.3.y.0, ("One or more Individual Steam Line Isolation-Manual Initiation channels inoperable", is being eliminated.

The TRM revision will remove the TRM requirement to restore individual steam line isolation manual initiation channels to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .In addition, multiple Station Procedures are being revised to reflect the revision to TRM 3.3.y.Reason for Activity: (Discuss why the proposed activity is being performed.)

Current TRM Condition 3.3.y.0 is overly restrictive in that it requires a non-functioning Individual Steam Line Isolation Manual Initiation Channel, which is associated with the active-side train, be restored to operable status within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months or the associated valve must be declared inoperable and applicable Conditions and Required Actions of LCO 3.7.2,"Main Steam Isolation Valves (MSIVs)" must be entered.The capability of isolating individual MSIV's is beyond the requirements of the design bases analyses.Thus, in order to provide flexibility for the Operators, this activity will eliminate the restrictive TRM Condition and associated Required Action regarding the Individual Steam Line Isolation-Manual Initiation Channel inoperability (TRM Condition 3.3.y.0).Effect of Activity: (Discuss how the activity impacts plant operations, design bases, or safety analyses described in the UFSAR.)Incorporating this activity eliminates TRM Condition 3.3.y.0 and revises the aforementioned procedures to reflect that.It does not change how plant SSC's are operated and how they function during normal operation and the transients evaluated in UFSAR Chapter 15.The design bases of the plant's SSC's have not been changed nor the design inputs and assumptions in safety analyses described in the UFSAR.The MSIV's will still function as designed under all conditions described in the UFSAR.Summary of Conclusion for the Activity's SO.59 Review: (Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)

Eliminating TRM Condition 3.3.y.0, ("One or more Individual Steam Line Isolation-Manual Initiation channels inoperable" does not:*Result in more than a minimal increase in the frequency of occurrence of an accident previously evaluated in the UFSAR,*Result in more than a minimal increase in the likelihood of occurrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR,*Result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR, 50.59 REVIEW COVERSHEET FORM StationlUnit(s):

Braidwood 11&2 LS-AA-I04-1001 Revision 3 Page 2 of2 ActivitylDocnment Number: TRM Change Request No.12-007 Title: Eliminating Action 3.3.y.D from TRM 3.3.y Revision Number:_*Result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR,*Create a possibility for an accident of a different type than any previously evaluated in the UFSAR,*Create a possibility for a malfunction of an SSC important to safety with adifferentresult than any previously evaluated in UFSAR,*Result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered,*Result in a departure from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses Attachments:

Attach all 50.59 Review forms completed, as appropriate.

Forms Attached: (Check all that apply.)o Applicability Review o 50.59 Screening I:?SI 50.59 Evaluation 50.59 Screening No.50.59 Evaluation No.BRW-E-2012-120 Rev.Rev.00-'------

50.59 REVIEW COVERSHEET FORM StadoWUDit(s):

Braidwood{Units I" 2 LS*AA-I04-IOOI Revision 3 Page I of4 Adil'ity1Docw:neft4 Number: Be 388161 (UFSAR DRP 14-087 Tide: LAKE SCRW HOUSE T8AYfiI lNG SCBW LEVEL CONTROL RerisioD Number:....,:Q"'-

_NOTE: For 50.59 Evaluations.

information on this fonn will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of IQ CFR 50.59(d)(2).

Descrlpdon of Adiftty: (Provide a brief, concise description of what the proposed activity involves.)

The proposed activity eliminates the automatic Lake Screen House (LSH)traveling screen 12" differential level trips associated with each Circulating Water (CW)pump.the 15 second time delay associated with each CW pump's trip.and upgrades the existing traveling screen level measuring instrumentation for both units.The following changes are included in this activity:*Removal of six Hagan Ring Balance differential pressure (dIP)level instruments (lLDR-SWOOI, ILDR*SWOO7,SWOI3, 2LDR*SWOOI, 2LDR-SWOO7, 2LDR*SW0l3) located in 1(2)PL02J and replaced with six new dIP Level Transmitters.

six new dIP Level Recorders, and a new power supply to power the loops, which will provide the same functions as the removed obsolete instruments.

Removal ofsix Differential Pressure Switches (ILDS-SWOOI, ILOS*SWOO7, ILOS-SWOI3, 2LDS-SWOOI,SWOO7, 2LDS-SWOI3) and six Time Delay Relays (lPL02J-TORA, IPL02J-TDRS, IPL02J-TDRC, 2PL02J-TORA, 2PL02J-TORS.2PL02J.TORC)located in 1(2)PL02J, which are associated with the CW pump trips.<<I Modify the Lake Traveling Screen Wash Control Panels 1(2)PL02J to reflect the removal and addition of the above instrumentation.

<87 TWe: LAKE SCRHBN HOUSE IMYEIING SCBHfiN LEVEL CONTROL current procedtu:e BwOP CW-26.DEFEATING CIRCULATING WATER PUMP TRIP ON TRA VEUNG SCREEN DELTA P.to defeat the CW pump trip.BwOP CW-26 eunently contains a Limitation and Action (B.2)to trip the affected CW pump with an observed 30 inch differential level across its associated traveling screeD.This design change permanendy removes the trip such that this procedure is no lonser necessary.

This activity will not affect: the other traveling screen modes of operation due to a timer.differential level (normal backwash at 6" AI..and rotate in slow speed or emergency backwash at 10" AI..and rotate in fast speed).or manual actuation.

This activity will not affect any other existing CW pump trips other than the traveling screen 12" differential level.The existing level instrumentation will be replaced with new reliable transmitters and recorders providing the same functions as the existing instruments which include control.indication.

recording.

and alanns.There are two LSH traveling screens pet CW pump (12 screens total).The purpose of the traveling screens is to filter any small pieces of matter and debris that have passed through the LSH bar grills before it can enter the suction for the CW pumps.The purpose of the differential pump trip is to protect: the structural integrity of the travelinl screens and to protect the CW pump under low suction pressure conditions.

The proposed modifICation will not affect the design functions of the traveling screens or the CW pumps.Although not explicidy stated within the Braidwood Design Basis (UFSAR).the existence of the 12" water differential level CW pump trip would limit the buildup of debris and resulting traveling screen differential level.thereby reducing the potential for traveling screen over-loading and failure.This condition would also reduce forebay water level such that it minimizes impact on the performance of the Essential Service Water (SX), Fire Protection (FP).and Non*Essential Service Water (WS)Systems.The implementation of this activity will have no adverse impact to the CW System for condenser operation.

the FP System.the WS System for non.-safety reblted cooling loads, and the SX System for safety-reblted safe-shutdown cooling loads.While the traveling screens and associated differential level instrumentation serve no safety-related function.they are designed to support General Design Criteria 2, Design bases for protection against natural phenomena.

in that the traveling screens will not interfere with the successful operation of the safety-related SX System through collapse and blockage of the SX intake structure.

The cooling pond intake structure (fol'ebay walls.floors and ceiling.including the traveling screens)is seismically qualified to protect the function of the SX System.The SX System design basis does not credit the traveling screens to perform any function in support of the SX System mission.The SX System is utilized to reject: core decay beat and safety-related equipment heat to the Ultimate Heat Sink (UHS).The UHS is a part of the cooling pond located in front of the Lake Screen House.The significant heat loads on the SX System include core decay heat removal via the Residual Heat (RR)Removal and Component Cooling (CC)Systems and post Loss of Coolant Accident (LOCA)containment cooling.Other loads include safety-related equipment cubicle coolers and oil coolers.As such.the SX System does receive the appropriate Engineered Safety Features (ESF)actuation signals and is powered by ESF power supplies.The SX System draws water from the UHS.through the traveling screens.and returns the water to the far end of the UHS.The traveling screens are designed to ensure they will not fail in a way that affects the operation of the SX System.The design of the traveling screens includes allowances for seismic loading.both Operational Basis Events (ODE)and Safe Shutdown Events (SSE).The traveling screens have no safety-related design basis criteria.The operational requirements of the SX System and the UHS are provided in Technical Specifications 3.7.8 and 3.7.9, respectively.

The SW System is not subject: to any Technical SpecifICation requirements.

The redundancy and separation at the LSH maintains the required suppression water source for the FP System.The FP System can be supplied by the SX System.if needed in an emergency.

If the traveling screens and SW System function as designed and remove debris.the likelihood of an impact on the entire FP System is not more than minimal.The WS System is non-safety related and not required for the safe shutdown of the reactor (UFSAR Section 9.2.1.1).The WS pumps do not have a trip function on high traveling screen differential level.The loss of a WS pump or pump performance could potentially result in reaching temperatUre limitations on Balance of Plant equipment that warrants power reduction.

pump trips.or manual reactor trips.Although the absence of the CW pump trip function on high differential level could impact the WS System in this circumstance the likelihood of this occurring is no different than if the existing non-safety related CW pump differential level failed to function.

50.59 REVIEW COVERSHEET FORM StadoDlUDit(s):

Braidwood I Units I" 2 LS*AA-I()4...IOOI Revision 3 Page30f4 AdivkylDocammt Number.Be 3881611 UFSAR DRP 1+087 Tide: LAKE SCBf:FN HOUSE IRAYfllNQ SCRFffiN LEVEL CONTROL Revision Number:---10'--_The elimination of the traveling screen differential level trip function does reduce the defense in depth for a fouling event at the LSH as described in the Braidwood Station response to SOER 07-2.1nJake Cooling Water Blockage.However.existing procedure guidance in conjunction with annunciator alarms and operator response will minimize the potential for a gross fouling event affecting plant equipment.

The failure of the CW System controls can fail in such a manner to prevent traveling screen operation.

prevent SW System operation.

and prevent CW pump trip.A review of the UFSAR revealed that the turbine trip event.as described in section 15.2.3.is the only accident that may be affected as a result of the proposed modification to eliminate the LSH traveling screen differential level trip function for the CW pumps.The trip of a CW pump or pumps is an initiating event that results in a loss of condenser vacuum which can cause a turbine trip.Therefore.

the end result of the failure is the same as the results currently described in UFSAR section 15.2.3 and the results are unchanged for any crediblelpotential malfunction due to the failure of the CW System controls.The elimination of the traveling screen differential level trip will not increase the frequency of a loss of condenser vacuum or a turbine trip.Based on the above.the design basis and safety analyses of the traveling screens.SW.CW.Sx.fP.and WS Systems are not adversely impacted.Summall')'

of CondusIoD tor the Activity's SO.59 Review: (Provide justification for the conclusion.

including sufficient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 Screening.

50.59 Evaluation.

or a License Amendment Request, as applicable.

is not required.)

This activity eliminates the automatic LSH traveling screen 12" differential level trips associated with each Circulating Water (CW)pump.the 15 second time delay associated with each pumps trip.and upgrades the existing traveling screen level measuring instrumentation for both units.Existing CW pump dIP indication and indicated pump motor amps will still be available in the Main Control Room.The operator(s) can continue to manually trip the CW pumps.if conditions are warranted.

upon receiving an alarm so the proposed modification will not have an impact on plant operation from the Main Control Room.The existing level instrumentation will be replaced with new reliable transmitters and recorders providing the same functions as the existing instruments which include control.indication.

recording, and alarms.While the traveling screens serve no safety.related function.they are designed to support General Design Criteria 2.Design bases for protection against natural phenomena.

in that the traveling screens will not interfere with the successful operation of the safety-related SX System through collapse and blockage of the SX intake structure.

The SW System traveling screens installed at the LSH are designed to falter cooling pond debris larger than approximately 3/8".The Circulating Water System Controls is a control system not required for safety and umelated to reactor safety as described in UFSAR section 7.7.1.15.Although not explicitly stated within the Braidwood Design Basis (UFSAR).the existence of the 12 64 water differential level CW pump trip would limit the buildup of debris and resulting traveling screen differential level.thereby reducing the potential for traveling screen over-loading and failure.This condition would also reduce forebay water level such that it minimizes impact on the performance of the SX.fP and WS Systems.The design function of the traveling screens is not being changed by the implementation of this activity.As described above in the effects section and detailed in the attached evaluation.

the proposed change does not alter the UFSAR described design function of the SW.CWo SX.fP and WS Systems.Based on a fouling event and indications in the MCR.Operators will take action upon receiving a CW Pump Low Delta P alarm(BwAR 112-17*BI3) and Traveling Screen Trouble alarm (BwAR 112*17-EI3) of tripping the CW pump(s).Since this activity installs better and more reliable instrumentation.

and the alarms are fast enough for operator action on tripping a CW pump.enough margin is ensured and available for plant equipment in the forebay (SX.WS.and fP).The 50.59 evaluation concluded that the design basis function of the SW.CWo SX.FP and WS Systems are maintained by this change since existing procedure guidance in conjunction with annunciator alarms and operator response win minimize the potential for an event affecting plant equipment, which provides an adequate degree of protection.

Since the frequency of accidents or malfunctiollS are not increased.

the consequences of accidents and malfunctions remain bounded.no new accidents Revision Number:___50.59 REVIEW COVERSIlEET FORM StadonlUDit(s):

Braidwood 1 Units 1&2 AdtritylDocumem Number: EC 3881611 UFSAR DRP 14-Q87 Tide: LAKE SCRW HOUSE TRAm INO SCRBfN LEVEL CONTROL LS-AA*I04-IOOI Revision 3 Page4of4 or malfunctions are created, and there are 00 changes to fission product barrier protection or evaluation methodology, the proposed activity may be implemented without prior NRC approval.Att.lldmleats:

Attach all 50.59 Review forms completed.

as appropriate.

Forms Attached: (Check all that apply.)o Applicability Review o 50.59 Sc:neaiag I8J so.s9 Evamadoa 50.59 Screeaiag No.50.59 Evaluadoa No.BRW*E*2012*155 Rev._Rev._0::..-_

StationlUait(s}:

50.59 REVIEW COVERSHEET FORM Braidwood Unit 2 LS-AA-I04-1001 Revision 3 Page 10f4 Revisiolll Number:..-.!L-Tide: STEAM GENERATOR MARGIN TO OVERFILL (SG MTOl PORV TRIM REPLACEMENT AND VALVE BLOCK INSTALLATION NOTE: For SO.59 Evaluations.

information on this form wiD provide the basis for preparing the biennial summary report submiued to the NRC in accordance with the requirements of 10 CPR 50.59(d}(2).

Deseriptioa of Activity: (Provide a brief.concise description of what the proposed activity involves.)

The proposed activity will install a new valve internal trim in the Steam Generator (SO)Power Operated Relief Valves (PORVs).2.'\fSOI8A-D.

and a valve block which will limit the flow to the current design basis flows consistent with the Steam Generator Tube Rupture (SOTR)Margin to Overfill (MTO)and Dose Case analyses.This cbange will also increase the actuator setpoints wbich is required to support opening and closing the valve with the increased trim size.This change will also modify the support installed on the 2C MS line downstream of the 2MSOlSC to improve stress margins for the bigher valve thrust loads with the modified actuator setpoints and increased trim size.UFSAR Draft Revision Package 14-027.included with EC 380046 Revision 2.updated the maximum stresses listed in UFSAR Section 5.2.2.5.3 for the SO PORV piping associated with the Unit I valves.UFSAR Draft Revision Package 14-092 included with EC 389664 win update the maximum stresses listed in UFSAR Section 5.2.2.5.3 for the SO PORV piping associated with the Unit 2 valves.Reason for Actinty: (Discuss why the proposed activity is being performed.)

IR 1358008 documented unexpected vendor flow testing results which demonstrated that the existing valve capacity is actUally less than originally assumed in the accident analysis.As a result, the margin to overfill bas been significantly reduced.To increase the margin to overfill.the station bas elected to restore the capacity of the SO PORVs on Unit 2 to be consistent with the original design basis capacity.This change will install a new trim in the SO PORVs which will increase the valve capacity at full open.This modifICation will also install a valve block to limit the stroke of the valve which will limit the valve capacity to be consistent with the flow rate utilized in the current design basis analysis.The maximum flow rate of the modified valves will not exceed the flow assumed in the SGTR Dose Case analysis.The setpoint change for the actuator is required to increase the actuator capability with the new trim installed.

The support change on the 2C MS line is required due the increased loads on the support as a result of the higher valve thrust loads with the modified actuator setpoints and increased trim size.meet of Activity: (Discuss how the activity impacts plant operations.

design bases.or safety analyses described in the UFSAR.)During normal operation the SO PORVs are closed.In the event the pressure increases.

the valves automatically begin opening in response to a pressure controller to maintain SO pressures below the safety valve setpoints if possible.The modified SO PORVs will essentially perform the same.IR 1358008 documented unexpected vendor flow testing results which demonstrated that the existing valve capacity is actually less than originally credited in the accident analysis.Therefore.

by installing the new trim and valve block the actual valve capacity will increase which will improve the cooldown capability of the SO PORVs.The stroke time to open and close will decrease with the reduced stroke with the valve block installed.

The piping has been conservatively analyzed for the maximum flow through the valve assuming the valve block is removed.This condition is bounding for the configuration with the valve block installed.

During a SO tube rupture event.operator actions require the operators isolate the ruptured SO by closing the feedwater.

auxiliary feedwater and main steam isolation valves md the affected SG PORV.The intact SG PORVs are then used to cool down the ReS below the saturation StacioDlUait(s):

50.59 REVIEW COVERSHEET FORM Braidwood Unit 2 LS*AA*I04-1001 Revision 3 Page 20f4 Tide;STEAM GENERATOR MARGIN TO OVERFILL (SG MIO)PQRV TRIM REPLACEMENT AND VALVE BLOCK INSTALLATION temperature of the ruptured steam generator.

Once the cooldown is complete.the RCS is then depressurized to equalize pressure with the ruptured steam generator and terminate the primary to secondary leakage.Given the capacity will be consistent with the current design basis.there is no adverse impact on the SGTR Dose Case and the actual margin to overfill will improve.The MTO Analysis will not be affected since this modification will restore the valve capacity to the capacity currently credited in the MTO Case.The operator action times associated with the S6TR event are not cbanged.Summuy 01 COIldusioa for the Ac:thity's 5G.S9 Review: (Provide justification for the conclusion.

including sufficient detail to recognize and understand the essential arguments leadiDg to the conclusion.

Provide more than a simple statement that a SO.59 Screening*.50..59 Evaluation.

or a License Amendment Request.as applicable.

is not required.)

1.The S6 PORV (otten described as a S6 Relief Valve)valve control system is not cbanged.so the likelihood of a failure of the control system which could cause an inadvertent opening of the S6 relief valve remains the same.The installation of the valve block does not contribute to failing the valve open since it does not directly operate the valve.The valve block will limit the open travel of the valve.The valve is a flow over the seat.Therefore.

pressure helps maintain this valve in the closed position.This would tend to decrease the frequency of occurrence of a S6 PORV valve failing open.Section 15.6.3 discusses the Steam Generator Tube Rupture Event.The S6 PORVs are credited in this analysis but are not the accident initiator.

Therefore.

the installation of the new trim.actuator setpoint change.support modification, and valve block does not increase the frequency of occurrence of an accident previously evaluated in the UFSAR.2.UFSAR Seaion IS.I.4discusses the possible malfunction oftheSG PORV (SG Relief)to open inadvertently.

As discussed above the likelihood of this malfunction has not increased.

UFSAR Sections 15.6.3 and 15.3.3 discuss the possibility of the valve to fail to reclose once opened in response to a transient.

The valve is a flow over the seat.therefore flow will tend to close the valve.The required fon:e from the actuator to close the valve will be decreased due to the increase trim size and as a result.decrease the likelihood of the malfunction.

Various sections in Chapter IS discuss that the valve opens in response to transients.

The modified valve will require increased thrust to open the valve due to the increase in the trim size.The valve actuator has been evaluated by the manufacturer and is capable of operating the valve with sufficient margin with the hydraulic actuator setpoint change described above.Furthermore.

flow testing has demonstrated that the modified valve is capable of opening under maximum differential pressure conditions.

Therefore.

the likelihood of a failure to open malfunction is not increased for the modified valve.In addition.the UFSAR sections often do not credit the S6 PORV function but instead rely on the main stearn safety valves to open to provide the heat removal function.Therefore.

the installation of new trim.actuator setpoiDt change.support modification.

and valve block will not result in more than a minimal increase in the likelihood of occutrence of a malfunction of an SSC important to safety previously evaluated in the UFSAR.3.Various accidents discuss the SO PORV as a potential release path that is evaluated as part of the dose calculations.

These accidents will not be affected since the new flow capacity of the valve is bounded by the flow assumed in the safety analyses.Also many of the accidents are not based on the flow rate from the valve but the primary to secondary allowable tube leakage which will not be affected by this modification.

Failure of the valve block is not considered credible such that the valve would travel full open since the valve block design has significant margin to failure compared to the actuator stem.Therefore.

if the actuator thrust were increased.

the valve stem would fail prior to the valve block failing StatioalUDit(I):

50.59 REVIEW COVERSHEET FORM Braidwood Unit 2 LS-AA-l04-100J Revision 3 Page 3 of4 Revision Number:_0_Tide;STEAM GENERATOR MARGIN TO OVERfILL (SG MfO)POSY TRIM REPLACEMENT AND Y AL VB BWCK INSTALLATION which would cause the valve to go closed.Therefore, it is concluded that the installation of the new trim.actuator setpoint change, support modification and valve block does not result in more than a minimal increase in the consequences of an accident previously evaluated in the UFSAR.4.The SO tube rupture accident and Reactor Coolant Pump Shaft Seizure dose analysis assumes a single SO PORV sticks open.This is still considered a credible failure mode.The consequences of this accident are not affected by the new trim and valve block installation since the new flow capacity of the valve is bounded by the flow assumed in the safety analyses and failure of the blocking device is not considered credible.The actions to mitigate this malfunction are unchanged as an operator will be dispatched to close the SO PORV upstream isolation valve.Therefore.

it is concluded that the installation of the new trim.actuator setpoint change, support modiftcation and valve block does not result in more than a minimal increase in the consequences of a malfunction of an SSC important to safety previously evaluated in the UFSAR.5.The new valve trim and valve block do not affect accident initiation sequences or response scenarios as modeled in the safety analyses.No new operating configuration is being imposed by the proposed change.The actuator setpoint change wiD allow the valve to continue to operate the valve similar to the existing operation but with increased capability and within the design limits of the actuator.Therefore.

it is concluded that the installation of the new trim.actuator setpoint change.support modification and valve block does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR.6.The current malfunctions evaluated are failure of the SO PORV to open and a failure of the SO PORV to close when required.The valve failing closed will have the same result.The valve block will be set such that the valve capacity will be consistent with the current design basis flow rate.When the valve fails open, the result would be the same because the flow rate assumed the accident analysis would not be affected.Failure of the valve block is not considered credible due to the robust design, significant margin of safety and the actuator shaft would fail prior to the block failure which would cause the valve to go closed.Therefore, the installation of the new trim.actuator setpoint change.support modifICation and valve block does not create a possibility for a malfunction of an sse important to safety with a different result than any previously evaluated in UFSAR.7.Installation of the new trim.actuator setpoint change, support modification and valve block installation does not affect the integrity of the fission product barriers utilized for mitigation of radiological dose consequences as a result of an accident.Plant response as modeled in the safety analyses is unaffected and no parameter which impacts a fission product barrier is changed.Hence, the mass and radioactivity releases used as input to the dose calculations are unchanged from those previously assumed.Therefore.

it is concluded that the new trim.actuator setpoint change, support modification and valve block installation does not result in a design basis limit for a fission product barrier as described in the UFSAR being exceeded or altered.8.The assumptions and methods used in the plant accident analyses are not affected by the new trim.actuator setpoint change, support modification and valve block installation.

The current SG Tube Rupture accident analysis margin to overfill ca...e assumes two SO PORVs are available for cooldown and depressurization of the ReS.This is not affected by this change.The dose case assumes one SO PORV fails open.This input is not changed nor is the evaluation methodology.

Therefore, it is concluded that the new trim, actuator setpoint change, support modification and valve block installation does not result in a departure StationIVnit(s):

50.59 REVIEW COVERSHEETFORM Braidwood Unit 2 LS*AA*l04-IOOt Revision 3 Page 4 of4 Revision Number:_0_Tide: STEAM OENERA TOR MARGIN TO OVERfILL (SO MIQ)POKY TRIM REPLACEMENT AND VALVE BLOCK INSTALLATION from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses.Based on the above.the replacement of the SG PORY valve trim.actuator setpoint change and suppon modifICation does not result in an unreviewed safety question and maybe implemented without NRC approval.Attadtmeats:

Attach all 50.59 Review forms completed.

as appropriate.

Fol'IDS Attadaed: (Check all that apply.)o Applicability Review o 50.59 Seneaiq tg:J 50.59 Evaluation 50.59 SereeaiDg No.50.59 EvaJuatioD No.BRW*E-201Zat61 ReY.ReY._0::..-_

50.59 REVIEW COVERSHEET FORM LS-AA-l()4..IOOI Revision 3 Page 1 of 3 StadonlUnit(s):

Braidwood.

ByronlUnit t anell AcdvitylDocument Number: Abtp'mal'omponent Pwitiog Shut Cor'min, the Isolation V.,.tg*MSIY Hydquli(Accumulator Manifolds' Belief Vjllyes RevloJioo Number: Hla Title: Close the Isolatiog Valves for the Relief ValVes on each of the MSIY Mumu1ator MjUJifulds NOTE: For SO.59 Evaluations, information on this fonn will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR SO.59(d)(2).of Activity: (Provide a brief, concise desaiption of what the proposed activity involves.)

Chanp the normal pesidoa 01 the ftliel valve isoIadoa vah'es (1I2MSOOIAlB/ClD-WHl)

OD the hydrade acc:umuJator IMaito" assembly lor the MaiD StaID Isomdoa Valves (MSIVs).lbe normal poskioo of the ftliel valves loJo.tioo valves loJ dIuIed froID"Opeta" to"Closecl".

Theposidoa 01 the tnMSe3IAfBICID..WHI va"es_indicated OD dna m..t0:z39 for BraWwood, M*152 SIL 57 for ByroD and OD the PASSPORT 0031....for the valves.R for Acdvity: (Discuss why the proposed activity is being performed.)

Thecriteria iD pl"CJrCedure OP-AA-I08-IOI,"ComroI of Equipment and Systeaa Statua", AttadlftIftt

%,"'AbDonnaI COlDpOl'lellt Posidoa Sheet (ACl'S)",loJ appIW to activities dUat nsaIt iD abDonnalof,..nt COIDpODeDD.

RevIew illwith thfs procedure a.....to the activity to dose thell1MSOOIAIBICID-HlHl reW valves loJoladoD valves.A nmew of the AttaduneDt 2 respoases IDdkates tbat a 10CFRS0.59 review Is requInd.for this adtrity.A coDdidoD was recently noted whereby, duriDrla..__break a<<ideat, the MSIV acauDDlators iCOUJd experience a higher temperature tbaD pnvlODslJ c:onsidend.iD........t temperature will cause the acaunu.Iators' temperature to rloJe wbidl causes the accumulators' hyclnuHe and ,...elk pressures to rloJe above tbe pressure ranae of 4,.to 51.....nonnally maiDtaiDed ia the acaunuIator.

At approm-tely 5400 psil iD the accumumtor, the hydraulic manifold relief valve actutes..bteeda hyclnuHe oil back to tbe rae"oir'.II aD escersive amouDt 01 hydraDik oD loJ nIievt4 to the.-rvoir'lIlS MSIV roolD teIDperatures rise, theft*may not be a suflldeDt.veatol'y 01 hy....oil...to fu.Hy dose the MSIVof the acaunuIator pressure w_dema"'-to respond to a MaiD Steam LiDe Break eveat (Refennc:e IR 1H)1409899 for Byron and lR 1H)I....for Bnklwood).

Eft'ed of Activity: (Discuss how the activity impacts plant operarions, design bases.or safety analyses desaibed in the UFSAIl)lbe reUef valves were added to relieve l.ncreased accumulator pnssure caused by RUduatiODs iD ambient temperatures

......_I plant operadon.Prior to the IDStalladoD of the reUef valves (and reIW valves by,..valves)it was to reduce pnssure below ad-ktrative IbniD by performiIJI a partial strub 01 the MSIVs.The frequency of the partial stroM noludoDS ebaHeaged adUtor elutomrers and potndalIy aeeelerated wear of the equiplDeDt thus redueiDI compoDent reliability.

The loJoIadoa 01 the reUef devic:e will result ill the need for manual actioa to reduce ac:cumulator pressure below the administrative limit of 5400 psi, but a partial stroke of the MSIVs wiD not be needed.Steps to take thfs acdOD are iDduded iD proadure BwOP MS-S Rev.28 for Braidwood and BOP Ms.!Rev.17 for Byron.The supporting doc:umeDtadon for the addldoIl of the hydraDik reUef valves (Braidwood EC , 415so.Byron ECs 77503 Ie 775(4)sbows dUat the potential for reUef valYes failure WIllSthus, the modifkatioo induded a mamw loJoIadon valYe to mailltaiD the MSIV safety fcmc:tioa while_Iud", a reW vah'e bypass valve to provide a manual method to eontroWUlit actutor pressure.Isolating tbe bydraDlk_nifold reUef valves wHi not impact the asswnpdoDS made ia the safety a...lyses.The adioa to isomte the hydraDik reUet valves will eDSUJ'e dUat the hydraDik acanIluiators maiD the oil inventory that is requirecl to dose the MSlVs whea demanded.The MSlVs wiD thus he able to dose IllS demallded OD a MaiD SteamiiDe Isolation signal geaerated automaticaly or_nuly.As diseussed iD the Byron SER (SupplelDeat S)and the Braidwood SER (Supplement 1), the MSIV's aecumulatot' WIllS tested to a lBialmum mternal pressure of 1500 paL The NRC WIllS conc:emed about the valve aecumulators exceed'" their proof pressure under hiP temperature ambient aeddeot conditions.

The MSIV actuator was successfully tested, under simulated aeddent conditions, as part of the Eaviroamental Quaiifkadon activities without exceeding the 50.59 REVIEW COVERSHEET FORl\1 LS*AA*I04-IOOI Revision J Page 20f3 StationlUnit(s):

8pld;n04.Urmn/UniU...a ActivityJDocumrent Number: AbrtormII CPomoBII Pmitipn Shed fpt Clsnina me 1so1a.1m VII.m me MSIV Hydmdi$Ag;umulato[

Manifolds' Relief Valyes Revision Number: tfLA Tide: Close me lsgJatioQ\'alm for the Re1ief YaIYeli on cadi of me MSlY Hydraulic Accumulator Manifolds accwnuJator interul pressure Udt (The......._aca..18tor pressure reported ia EQ Binder EQ-BB-024 is 5,SOO psfa).The tested did not Indude die hydra" reIel valYes.The hydraulic reIel valY.were not Hckd to pro.1de overpressure protec:don.

The relief.alv.were adcIed by ByroalBnldwood StadoD toprasun ca.-d by f'luctuadoal ia aaableDt tempera......dnriDi.nonmd plat........Prior to..........

doD oldie nlW'valv.(and reliefvalYe by,.valves)It was neces8aI'1 to prasun by performiDI a paI"CW strou san'" of the MSIVs.The relief val"es were not iastalled to Ibnftdle aa:umufaton below their"'1__dowable wOl'ldDl pressure or to satisfy a spedftc ASME code deUp requimaent.

The reIel valv.pro.1de a pressure controllfDl or 1Inddn1 fbac:doa to p""'e die 1011I&ernl1'I!UabIIty ofia the actuator-mblJ.The desip Includes a 0.013 meb orifice to limit die bydraulic nUeYed......abe systHa suda diet It is wfthia the capadty of abe sy....hydra.pump to maiatafD pressure ia the of relief valYe or....opea.Tberet'o....the deUp of die reIel.alve was In....as an opendoul con"enieac:e to presene die&ernI reIUl" of die MSlVs wilbont creatIDI a potendaJ loa of fuacdon due to reIel valYe talan.Summary offor die ActI"ity's 50.59 Review: (Provide justifICation for the conclusion.

including sufficient detail to recognize and understand the essential arguments leading to the conclusion.

Provide more than a simple statement that a 50.59 S<:reening.

50.59 Evaluation.

Of a License Amendment Request.as applicable.

is I'IOt required.)

This activity does not iacreae tileofof a Main Steam Line Break Acddent or 0 Inadverteelt MSIVe"ent, or lac....tile liUUIaood ofof a..1futM:doa of..sse important to diety.The cba.does not impact die pressure boudar7 fntepity of the MldD Steam System.The actIYity does not died tile seismic and environmental of tile MSIVs, nor does it impact tbe Maia S..pipina lodlnI conditions.

This activity does not result In aftIn theof an xddent or In die conseqMllCt!S of aot an sse important to safety.The Larp SLB outsWe contaimnent upstreaDI of MSIV is iiDIftinI tor 0ftWte Dose consequences.

This actJyity ensures that die MSIVs dose_1lISSUDIed in die accident uaafyses to isolate tile DIllin steam lines on a MSLB.This verities that tileof a MSLB are not iDIpIded..

In a Stea.Generator T.Rupmte acddent, tile MSIVs are closed by operator action to isolate the rupture Stea.Genentor.The SGTR event does not d'td the conditloDs In the MSIV room, thus, isoladnl the hydraulic relief for the MSIV by".DIUiIoki does DOt bave any impact on tile assumed MSIV closure in the SGTR event.The MSIVs are 8liIJWDed to close on a main steaDlline isolation sipalln a feedwater system pipe hreak event.As discussed in tbe UFSAR, tile t_water line break willa tile mostconseqnences would be tbe one that occ:urrecI inside containment between a steam senerator and die f_water dIeck val.e.In this cue, the contents of the steam generator would be I"deased to contaiaDlent.

For this scenario, environmental conditions ia the MSIV rooms will not be aft'ected and tbe MSIVs'function to close will be maiatainecL The enviromnental conseqnenca of less severe feedUne breaks In tile MSIV rooms are bounded by main steamIine breaks.nus activity does not create II possibility for an acddent of a dilrerent type tUn oy previously e.ahatted In tbe UFSA.R.The MSIVs are maintained in the open position dW"ing normal plut operation.

A conseqnence of a..mmcdon of the hydraulic dn:uit could only be closU.re of tile MSIVs.lnad.ertent MSIV closU.re is evafuated In UFSAR section 15.2.4.The proposed cbange does not change tbe results of postulated events wbere an MSIV is assumed to fail open.This activity does not ereate a possibility tor a..Ifancdon of an sse important to satety with a diflerent result than any prevlonsly evaluated ia UFSAR.Inadvertent do:sure of a MSIV bas been evafuated in the UFSAll.The isolation of tile hydraulic relief valves does not cbange tbe results evahatted In the UFSAR for this taDure.The MSIV inadvertent dosan is bounded by a Turbine Trip and dosure of tbe MSIVs is not u.swned for this event.

50.59 REVIEW COVERSBEET FORM LS-AA*I04-IOOI Revision 3 Page 3 of3 StadonlUnit(s):

Qgisbmod, QJnmlU Dit'a-;;AedvltylDoN_

NumbIIr: 61Jmtmi&1 COmp2Dl:m PositjgO SbC em: "05ina lbe IspWion VllwllO thl WN Uydmulir Accumulator Maoifo!ds' Belief Yaha RevidoD Number: tfLA TIde: Cls>se the Isplation Valves for till Relief valves go each of the MSIY Hydraulic Accumulator Manifolds 1'bIs activity does not result Ie..dump tUt would cause any system parameter to dIaa(Ie.Tberefo the activity does not nsuIt Ie a DeslpIBasII Limit FIssIon Product Barrier..described Ie the UFSAR beiDa or attend.11Ie results of the......stea"'" break acddentI an not.tcted by tbIs activity.1'bis activity does DOt itnpad the cakulated coat,,"'" temperature aDd ,........upon a MiLD.Tberefo tbIs actIYlty does not uc:eed or alter desIIa buis IlIDhs related to tud daddl..RCS pnsIUft bouDclary"ad con......1'bIs aedvity does not result Ie chanainl the method that bas beft used to evaluate the MSLD or other acddentll that....analyzed Ie the UFSAR.Based on the results or the SO.59 Evaluation, the acdvity may be ilDplemeuted without a Ucense llIDItDdIDent.

AuadtmeutII:

Attach all SO.59 Review forms completed, as appropriate.

FOrDIIII Attacbed: (Check all that apply.)o Appllcabmty Review o SO.5'ScneftIaI 1&'1 SO.5'Evaluation SO.5'Scl"ftllinl No.SO.5'Evaluation No.6(;..1%-009 BRW*E-201:&-186 Rev.Rev.0 o LS-AA-l 04-1 00 1 Revision 3 50.59 REVIEW COVERSHEET FORM Page..!of...J.Station/Unit(s):

Braidwood Unit 2 Activity/Document Number: TRM Change#12-010 Title: Change In-Core Decay Time for A2R16 Revision Number: N/A NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).

Description of Activity: (Provide a brief, concise description of what the proposed activity involves)The proposed activity makes the following changes to the Technical Requirements Manual (TRM)to reduce the minimum required In-Core Decay Time (ICDT)for A2R16 from 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months to 95 hour0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months s:*Braidwood TRM Section 3.9.a,"Decay time," states"The reactor shall be subcritical forthe last 100 hour0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months s95 hours forA1 R16)." This activity will revise this statement by replacing95 hours for A1R16)" with95 hours for A2R16)."*Condition A under TRM 3.9.a states"Reactor subcritical for<100 hours<<95 hours for A1R16)".This activity will replace"<<95 hours for A1R16)" with"<<95 hours for A2R16)".*Surveillance requirement TSR 3.9.a.1 will be revised by replacing hours for A1R16" with hours for A2R16".Reason for Activity: (Discuss why the proposed activity is being performed)

It is anticipated that during A2R16, work activities will be completed and the required plant configuration will be established to support commencing movement of irradiated fuel from the reactor vessel to the Spent Fuel Pool (SFP)prior to the current TRM fuel movement ICDT constraint of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months after reactor shutdown.Effect of Activity: (Discuss how the activity impacts plant operations, design basis, or safety analyses described in the UFSAR.)The proposed changes will allow starting A2R16 reactor core offloading activities at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after the reactor is shutdown.The Byron and Braidwood spent fuel pool cooling design basis analysis (#BRW-00-0010-M Revision 0)is based on the minimum ICDT of 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months prior to starting fuel transfer, however an outage specific evaluation has been performed to support a reduced ICDT for A2R16.This is in compliance with UFSAR Section 9.1.3.1, which states that outage specific evaluations may be performed, in support of shorter fuel decay times in the reactor, by taking credit for existing margins in the design basis analysis.The limiting case for the design basis analysis for the Spent Fuel Pool is based on one train of Spent Fuel Pool Cooling in operation.

Starting core offload at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after shutdown will not result in increasing the design basis heat load for the Spent Fuel Pool (SFP)Cooling System.Moving fuel at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after the reactor is shutdown does result in an increase in the heat load input to the Spent Fuel pool, from the offloaded assemblies, when compared to starting core offload at 100 hrs or later.However, the cumulative heat load to the spent fuel pool will not be greater than the design basis heat load, since the actual heat load in the SFP due to previously stored fuel assemblies is less than the heat load that was included in the SFP design basis analysis.A margin in excess of 1.6 MBtu/hr or Revision Number: N/A LS-AA-104-1 001 Revision 3 50.59 REVIEW COVERSHEET FORM Page..,Lof

i Activity IDocument Number: TRM Change#12-010 Title: Change In-Core Decay Time for A2R16 about 2.6%of the design basis heat load for the SFP has been calculated between the total heat load that is added to the SFP during A2R16 and the design basis heat load.The resulting maximum bulk water temperature will be below the design basis maximum bulk water temperature of 162.7 of.The impact ofthe higher heat load on the temperature in the Spent Fuel Pool, with one train of cooling in operation, will be minimal.In fact, the total heat load increase due to moving fuel at 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months after shutdown is<2%higher than the heat load that is added moving fuel at 100 hours0.00116 days 0.0278 hours 1.653439e-4 weeks 3.805e-5 months .This increase is minimal and would result in a minimal temperature increase (estimated to be<1 of with a Component Cooling temperature of 105 OF)for the Spent Fuel Pool.This temperature increase does not have a degrading effect on the Spent Fuel Racks, including the boral poison panels.Limitations and alarms related to Spent Fuel temperatures have not been changed and are in place during core offload.In addition, Operating Procedure BwOP FC-l includes a precaution to monitor Spent Fuel Pool temperature frequently and make adjustments as necessary to spent fuel pool cooling, during periods of high heat loads, such as core offloads.The radiological design basis analysis for the Fuel Handling Accident is based on a minimum decay time of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months prior to movement of irradiated fuel assemblies (Reference design analysis#BRW-04-0041-M/BYR04-047 Revision 2).Sections B3.9.4 and B3.9.7 of the Braidwood Technical Specification Bases are not being revised as part of the A2R16 activity since the minimum ICDT for radiological considerations is not being revised and the revised ICDT of95 hours for A2R16 still meets the constraint of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .Occupational Radiation Dose Occupational radiation dose will remain within limits.Access to the areas that are affected by the defueling operations is controlled in accordance with station procedures.
Electronic dosimeters are required to continuously monitor the incurred radiation dose in the areas in order to limit personnel exposure to below 10CFR20 limits.These existing controls are not affected.Summary of Conclusion for the Activities 50.59 Review: (Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion.
Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
This activity does not increase the frequency of occurrence of a Fuel Handling Accident or a Loss of Spent Fuel Pool Cooling event, or increase the likelihood of occurrence of a malfunction of an SSC important to safety.The proposed change does not increase the failure rate of refueling equipment or increase the risk of a fuel handling accident due to human error.Spent fuel handling tools will not change, nor will the method or procedures for handling spent fuel assemblies.
Existing administrative controls and precautions remain in effect, only one spent fuel assembly is lifted at a time, and the fuel is moved at low speeds, exercising caution that the fuel assembly does not strike anything during movement.An outage specific evaluation (Design analysis#BRW-00-001 O-M Revision OOOX)has concluded thatthetotal actual heat load in the Spent Fuel Pool as a result of the reduced ICDT is bounded by the total heat load specified in the design basis analysis.Thus, there is not an additional demand, beyond the heat load considered in the design basis analysis, on the cooling system.This activity does not result in an increase in the consequences of an accident or in the consequences of a malfunction of an SSC important to safety.The offsite dose resulting from a Fuel Handling Accident with a minimum In-Core Decay Time of 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months is bounded by the design basis Fuel Handling Accident dose with a minimum ICDT of 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months .This activity does not create a possibility for an accident of a different type than any previously evaluated in the UFSAR as there is no new equipment being introduced, and all existing fuel transfer equipment is being operated using existing procedures.
LS-AA-104-100 1 Revision 3 50.59 REVIEW COVERSHEET FORM Page2...of J.Activity/Document Number: TRM Change#12-010 Title: Change In-Core Decay Time for A2R16 Revision Number: N/A This activity does not create a possibility for a malfunction of an SSC important to safety with a different result than any previously evaluated in UFSAR.The increase in heat load in the Spent Fuel Pool has been evaluated; although an input parameter (offload start time)has been changed, the resulting impact on the SFP bulk water temperature analysis is bounded by the design basis analysis.In addition, the local water temperature, fuel cladding temperature, and maximum heat flux have also been evaluated and have been found to be acceptable.
The adequacy of the reduced ICDT for A2R16 is based on margin in background decay heat since the SFP is not filled to its capacity.The reduction in ICDT does not result in a change in the internal containment pressure that would represent a challenge to the containment design basis limit internal pressure of 50 psig (Reference UFSAR Table 3.8-4).The maximum cladding temperature for the spent fuel is well below the design basis limit of 2,200 of from 10CFR50.46 (Reference UFSAR Section 15.6.5).This activity does not change the method of evaluation for the Spent Fuel Pool Cooling System described in the UFSAR or in the SER for the Power Uprate Project.Decay heat input to the spent fuel pool was calculated for the earlier ICDT of 95 hours0.0011 days 0.0264 hours 1.570767e-4 weeks 3.61475e-5 months using the method described in NRC Branch Technical Position ASB 9-2.This is the same method that was used to calculate the decay heat values that are evaluated the design basis temperature analysis for the Spent Fuel Pool (BRW-00-00I0-M).
Attachments:
Attach all 50.59 Review forms completed, as appropriate.
Forms Attached: (Check all that apply.)Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.50.59 Evaluation No.BRW-E-2012-201 Rev.Rev.--=-0_
50.59 REVIEW COVERSHEET FORM StationlUnit(s):
_LS-AA-l 04-1 00 1 Revision 3 Page 1 01'2 ActivitylDocument Number:
_Revision Number: Title:
_NOTE: For 50.59 Evaluations, information on this form will provide the basis lor preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).
Description of Activity: (Provide a brief, concise description of what the proposed activity involves.)
This Special Procedure, Test, or Experiment (SPP)is being created to provide the instructions required to roll the Unit 2 main turbine from zero speed to the 450 and 600 l'pm plateaus for ventilation testing following replacement of the generator hydrogen cooler.Main turbine speed will be controlled using the installed Ovation Digital Electro-hydraulic (DER)system in the same manner as a turbine roll during unit startup.The difference between the normal unit stal'tup and the turbine roll for the generator ventilation testing is that the reactor will notbecritical.
The plant will be in Mode 3 at Normal Operating Pressure (NOP)and Normal Operating Temperature (NOT)with the steam dumps control Reactor Coolant System (RCS)temperature in the steam pressure mode of operation, The steam created from Reactor Coolant Pump (RCP)heat and decay heat that would normally be dumped to the main condenser through the steam dumps will be used to roll the turbine to the test speed plateaus.The proposed activity (SPP 12-006)only provides the directions for operation of the DEH system in order to roll the main turbine and establish the conditions required for testing of the main generator and hydrogen cooler.Performance of the actual ventilation testing will be evaluated separately.
Reason for Activity: (Discuss why the proposed activity is being perfol111ed.)
This SPP will allow the Unit 2 main turbine to be rolled to 450 and then 600 rpm so that internal generator air flows can be checked following replacement of the hydrogen cooler.The new hydrogen cooler is being installed during A2R16 and must be checked to ensure proper air flows through the cooler is obtained to ensure proper cooling of the hydrogen in the main generator.
Effect of Activity: (Discuss how the activity impacts plant operations, design bases, or safety analyses described in the lJFSAR,)There is the potential for an RCS cooldown, however the SPP providesguidancein the form of abort criteria to prevent an adverse plant effect.The SPP will roll the main turbine to the designated speed plateaus to allow Engineering to obtain the required data in order to ensure proper air flow through the new cooler and therefore adequate cooling of the hydrogen in the main generator.
The main turbine will not be operated near the resonant speed range of 950-1150 rpm.Abort criteria in the form of minimum Pressurizer (PZR)level, RCS temperature, Steam Generator (SG)level and pressure, as well as a loss ofDER control is provided.The abort criteria will ensure an adequate margin exists to prevent inadverteut Engineered Safety Features (ESF)actuation (Safety Injection, Auxiliary Feedwater, Main Steam Isolation, Letdown Line Isolation), excessive cooldown or turbine trip.Summary of Conclusion for the Activity's 50.59 Review: (Provide justification for the conclusion, induding sufficient detail to recognize and understand the essential arguments leading to the conclusion.
Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
The activity being implemented (SPP 12-006)does not operate the Main Steam, Turbine, or DEH systems in a manner that is outside of the reference bounds of their design basis or in a manner that is inconsistent with the descriptions found in the UFSAR.The implementation and performance of the SPP does not result in more than a minimal increase in the frequency of the Inadvertent Operation of the Emergency Core Cooling Systems.There are no physical changes required to any UFSAR described SSCs as a result of the performance of the SPP.The systems are being operated in the same manner for which they were designed.Performance of the turbine roll for ventilation testing is not the initiator of any new accident and no new failure modes are introduced.
Station/Unit(s): 50.59 REVIEW COVERSHEET FORM LS-AA-l 04-1001 Revision 3 Page 2 01'2 Activity!Document Number: Revision Number: Title: The actions performed in the SPP for rolling the main turbine to the required speed plateaus are tbe same as those perfOl'med dudng a normal unit startup therefore the activity does not create the possibility for a malfunction of an sse that is important to safety with a different result that has been evaluated in the UFSAR.There is no effect on any design basis limits for fission product barriers, any UFSAR described evaluation methodologies as a result of the performance of the SPP.Therefore based upon the above, the SPP may be implemented liS proposed without pdol'permission fl'om the NRC.Attachments:
Attach all 50.59 Review forms completed, as appropriate.
None Forms Attacbed: (Check all that apply.)Applicability Review 50.59 Screening 50.59 Evaluation 50.59 Screening No.50.59 Evaluation No.BRW-E-2012-210 Rev.Rev.
LS-AA-104-1001 Revision 3 Page lof2 50.59 REVIEW COVERSHEET FORM Station/Unit(s):
Braidwood Unit 1&2 ActivitylDocument Number: Design Change/EC 383592, DRP 14-022, Revision Number:
_Title: INCREASE RESIN VOLUME IN MIXED BED DEMINERALIZERS TO 35 CUBIC FEET NOTE: For 50.59 Evaluations, information on this form will provide the basis for preparing the biennial summary report submitted to the NRC in accordance with the requirements of 10 CFR 50.59(d)(2).
Description of Activity: (Provide a brief, concise description of what the proposed activity involves.)
This activity revises the volume of resin in the Unit 1 and 2 Mixed Bed Demineralizers from 30 to a maximum of 39 cubic feet.The Mixed Bed Demineralizers (l/2CVOlDAIB) are designed for a capacity of 39 cubic feet, Reason for Activity: (Discuss why the proposed activity is being performed.)
The resin volume is being increased as part of a source term reduction initiative.
Effect of Activity: (Discuss how the activity impacts plantoperations,design bases, or safety analyses described in the UFSAR.)The increased volume and type of resin will improve the efficiency for removing ultra fine particulate.
This in turn, will reduce the source term (dose)within the plant.There will be no effect on the performance or operation of the demineralizers since they are designed for 39 cubic feet of resin.An insignificant increase in system pressure may occur due to the higher pressure differential across the vessels.The Chemical and Volume Control System is designed to compensate for changes in vessel differential pressures.
Summary of Conclusion for the Activity's 50.59 Review: (Provide justification for the conclusion, including sufficient detail to recognize and understand the essential arguments leading to the conclusion.
Provide more than a simple statement that a 50.59 Screening, 50.59 Evaluation, or a License Amendment Request, as applicable, is not required.)
The increase in resin volume for the Mixed Bed Demineralizers does not create any adverse effects to an UFSAR design function.The demineralizer vessels were designed to contain 39 cubic feet of resin.Therefore, an increase from 30 to 39 cubic feet is acceptable.
There is no change in Dow rate, no increase in the radioactivity contained in the inDuent and the pressure drop inside the vessel is not increased beyond design limits since vessel is designed for 39 cubic feet of resin.The seismic qualification and interface with the structure were originally performed for the fuJI weight of the vessels.Therefore, there is no adverse affect on the seismic qualification of the vessels or on the Category I structure.
The use of different resins is in accordance with the UFSARandany resins being used will be evaluated by the Station or Corporate Chemistry Departments, Calculation BB*A*03 has determined that there is no increase in dose in the general areas surrounding the vessels due to increased radioactive isotopes collected in the vessels.Based on this, the activity does not create any adverse effects to an UFSAR design function.No UFSAR procedures that describe design function are adversely affected by this activity.The demineralizers will continue to remove particulate from the water as per the original design.Based upon the use of MircoShield software in lieu of the original outdated ISOSHLD*III software, this change in methodology was addressed in Question 8 in a 50.59 evaluation and found to be acceptable.
A new computer program was used to evaluate the shielding provided by the existing walls.This program (MicroShield) is used throughout the nuclear industry and is similar to the program (ISOSHLD*III) used originally at Braidwood.
MicroShield has been used in calculations reviewed and approved by the NRC.The use of this software has been evaluated and determined that this is not a change to any element of an UFSAR described evaluation methodology used to establish the design bases or in the safety analyses.The activity does not constitute a new test or experiment not described in the UFSAR because the operation and control of the Mixed Bed Demineralizers is unaffected by the increase in resin volume.Therefore, no SSC will be used outside the reference bounds of design or in a way that is inconsistent with analyses or descriptions in the UFSAR.Chapters 9, 11 and 12 of the UFSAR will be revised to document the change from 30 cubic feet to 39 cubic feet of resin.The increase in resin volume in the Mixed Bed Demineralizers is consistent with Technical Specification and the Operating License and, therefore, does not require these documents to be changed.Based on the above, the conclusions reached in the screening and evaluation, NRC approval is not required prior to implementing this change.
L';-AA-104-1001 Revision 3 Page 2 of2 50.59 REVIEW COVERSHEET FORM StationlUnit(s):
Braidwood Unit I&2 ActivitylDocument Number: Design Change I EC 383592.DRP 14-022.Revision Number:-"""0"'-0...,1,-"0""-00"'--
_Title: INCREASE RESIN VOLUME IN MIXED BED DEMINERALIZERS TO 35 CUBIC FEET Attachments:
Attach all 50.59 Review forms completed.
as appropriate.
Applicability Review 50.59 Screening Forms Attached: (Check all that apply.)o C8J 50.59 Screening No.BRW*S*2011*58 Rev.!50.59 Evaluation 50.59 Evaluation No.BRW*E*2012*245 Rev......;;..0
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