ML042030127

Safety Analysis Report for Extended Power Uprate

ML042030127
Person / Time
Site:	Browns Ferry
Issue date:	06/30/2004
From:	Hayes R General Electric Co
To:	Office of Nuclear Reactor Regulation
References
DRF 0000-0011-1328 NEDO-33047, Rev 0
Download: ML042030127 (183)

Text

ENCLOSURE 14 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS - 418 -

REQUEST FOR LICENSE AMENDMENT FOR EXTENDED POWER UPRATE OPERATION NON-PROPRIETARY NEDO-33047 BROWNS FERRY UNITS 2 AND 3 SAFETY ANALYSIS REPORT FOR EXTENDED POWER UPRATE See Attached:

- . * ; 0 * -

GE NuclearEnergy NED0-33047 DRF 0000-001 1-1328 Class I June 2004 Browns Ferry Units 2 and 3 Safety Analysis Report for Extended Power Uprate R L. Hayes

V GE Nuclear Energy 175 CurnfrAve., San Jose, CA 95125 NEDO-33047 Revision 0 Class I DRF 0000-001 1-1328 1u= 2004 BROWNS FERRY UNITS 2 AND 3 SAFeTy ANALYSIS REPORT FOR EXTENDED POWER UPRATE Pmpaid rby- T L HJye.

Apjp0vct1 b Apptmw by.

To I, TnolIuPL 1 1tty TmmVdlleyJ4oft

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.- 4 NEDOS04S *lRevidon O

IMPORTANT NOTICE REGARDING CONTENTS OF THUS.REPORT Please Read Cuofiully Tle only undertldngs of the General Electric Company (GE) rspctiag inforannon in this doumern are contained in the cont-nt betwen TVA and GB, Contract Order No. P-92NNP-82068D-ODI, effective 14 May 2001, and nothing contained in this document shall be constaued as 6banig the contract. he, ust of thin InforutiOt by anyone other ta TVA, or for any purpose other than that for which it is intanded, is tot authorized; and, with respect to any uma'uthodzed us, GE makea no representation or wauranty, express or implied, and assumes no liability as to the completenesi, acuracy, or usefilneas of the infbnnuiion contained in this document, or that its use may not infringo privately owned rights.

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NED0433047.. kerkion B Tabia Of Contentls 1.1 rI UM A ...... . ....... ... .**. ~...4**U....~b~d. i bbS4........., 1 1.2 Purpose and Approach ........... ................ ....... .. ........ 1-1.

1.2.1 Uprmts Anialysie Badsi. .......................................................... 1-2 1.2.2 Computer Codas.,............................. . ....................... ....... -2 1.3 VUprated Plnt Operating Conditions ....................................... . 1-4 1.3.1 Reactor Heat Baac..... . .................. 1-4 1.3.2 Reactor Performance Improvement Features ............................. 4 1.4 Summaroy and Conclusions ........- ................... ........ . ........ 1-5

2. REACTOR CORE AND FUEL PBRFORMANCB ....... .... ................... 2-1 2.1 Fuel Desig And Oprd n ................. ............. ............... 2-1 2.2 Thera Limt Aassent ................................................. . 2,2 2.2.2 Minimum CrdticPower Ratiao peraingLimit ........................... 2-3 2.2.3 MAPLEOK and Maximum LHGR Operating Limits ................................ 2-3 2.3 EReactisityr cJlaa~cteristics ....... . .. ...........

,.................... . ....... ...... ... 2~-3 2.3.1 PowVer/FlwO peratC811,31ing .... . ...... . ............. qgt...I.....t.....I.Ib..d.....2 2.5 Reactivity Control ..... . ...... ..... .............................................. 2-5 2.5.1 Control Rod D)rive Sysutemr. ... ,..........................................,.........2-5 2.5.2 Contrcl Rod DrivecPositioningeand Cooirrg .............................

2.5.3 Control Rod Drive Integrity Asme&nm~nt .........- 1...... ............... 2-6 2.6 References.. -....................... ................... 2....

3. REACTOR COOLANT AND CONNECTED SYSTEMS .......-............... 3-1 3.1 NUtCLEAR1~ SYfSiTEMv PRLESSUREL RE.LIEF. ................ I.-I .~.............3-1 3.1.1 IASILV Setpoint 1olexfa=..ce.......... -................................... 3-1 3.2 1Reatotor Overpressuire Prtoto Aai.yuis ........... .~.......................... IbI*Iq~I 31 3.3 Reactor Vessel andlInternals ................ ..........................

.. .. ......... .3-2 33.1 Reactor Vessel Fracture Toughness........ . ................ ............. 3-2 3.3.2 Reactor Vessel Sfructurai Byuluation.. ... .................. ............ 3-3 3.3.3 Reator lIntnnalPr'surefliffarncam (RIPDs) ...... 6... ........................... 3-4 3.3.4 Reactor lnternals, Structxaral Eivaluation ......................... ....... . . 3-4 3.3.5 FlowInduced Vibration..................I...... ................................ 3-9 3.3.5 Steait Seprator and D)ryer Perftbrpwamoe..........,.... ....................... 3-11 3.4 Reactor Reciroulatin System .... .................... -............................. -

3.5 Reoactr Coolant Pmeasure Boundary Piping .................. ............. . . .. 3-11 3.5.1 Recirculation SyuternBvaluation............... ..........

. .................. . 3-12 3.5.2 Wui Stcamn and Associated Piping SyuteznEvaluation (musds cotainment) ..3-12 liii

-7; --- I - -- Tr -: - - --.- .- . - - - I.. - - .

NED 0433047 - Revislon 0 335533 F oe wW a EtaruEvalua io .. ............................. ..................

h. . .......................... 3-11 3,5.4 Other RCPB Piping Ev uluation ........... .............. ..... ................ .. .... 3-14 3.595 Piping Flow Induced Vibration ........... ..

. .--.-............... . ...... . .....3-i5 3.6 2v1air Stearti Ine Flow PRestrictora ....................... d..... -,999**t.......... 9.. 94~*999*31 3.8 Reactor Core IzolationOCco]n ....... . 6 9....... .............. ..I..........-.1.....

3.9.1 Shutdown Cooling Made ..... . ................. 3-18 3.9.2 Suppressioin Pool Cooling Mode .......... .................................... 3-18 3.9.3 Coti nn pa o ligM d .... .................. ........... 3-19 3,9.4 Supplemnental Spent Fuel Pool Coln ... ......... ............. 3-19 3.9.5 Steam Condensing Mode .... ...... ,................................................ 3-19 3.9.6 Standby CoormngjCronstiee .............. . ............ ~...................39 3.10 EReactor WVutcr Cleanup Sy'sterx .................................... .99..........9I9...............3

• 3.11 Balance-tX-Plant Piping Evaluation.......................................... 3-20 3.11.1 Pipe Stress es........................................... ............ 3-21 ..

3.11.2 Pipe Supports .... ... . .. ............

,......I.... ........... 3-22 3.11 .3 Flow Acclerated Corrosion.................................... ................ 3-22 3.11.4 Main Steam end Assoziated Piping (Guiside Containment) ................... 3-22 3.12 References. .. ..... .........................

,............................ . 3-23

4. BNGINE3RBD SAP=TYFEATURES......... .......................... . 4-1 4.1 Contament System Perfrm~ance...............................-............4-1 4.1.1 ContaInment Pres~urrand Temperature Response ......c..... ........... .....4-2 4.1.2 Containment Dynamic Loads....................................................4-4 4.1.3 Conlaimnnirt Isolation..............................., ,....... ... ~ .......... ........... 4,5 1 t~

4.1.4 Generic L~etter 89-10 Progr'amn....................... - .. d1 . ................ 4.5 4.1.5 OenedicLettrr89-16 ..........- ...... . .......... ................. 4-5 4.1.6 Generic Letter 95-07 . . ....... ~............................... ............. 4.6 4.1.7 Genieric Letter 96.06............................................... ... 4-6 4.2 Bme-rgbnzy Core Cooling Systems ............... ...............-...

. ........... 4-6 4.2.1 Iligh Pressue.Coolant Injection System ........................-. 4-6 4.2.2 Low Pressure, Coolant Injection.. .. ................... ............ 46 4.2.3 Core Spray System ........... ~....................... .................. 4.7 4.2.4 Automatic. Depressurizatian System. ...... . ..............- ....... -.......... 4-7 4.2.5 BOGS Net Positive Suction Head............ ........................... ...... 4-7 4.3 Emrgncxay Core Cooling Systm Pefformnanc.. ,....................... .............. 4-9 4A4 Man ControlRown.AtosplierContrcl System .............. ... ............. 41 4.5 Standby GasTreatment Systemi.,........................-......................41 4.6 Mfain Steam Isolation Valve Leakage Control Syuiem....................... ....... 41 4.7 Post.LOCA Comnbustible, Gas Controlb.........................................41 4.8 .Refbrences .........................- . ...... . ........................... 4-12 4TPUIEN41AMFOIfAN}D CONTRO' - .O1....................................II~

)1'JNS ................... 5-1 19*9 5.1 NSSS Monitoring and Control Systems . ~.........-...........................,,.,5-1 5.1.1 Control Sy'tern Evaluationa .................... ..............

, .................... 5.1 5.1.2 Neutron Monitoring Syutem........................................ -

5.1.3 Rod WIorth . ................................................................. -- - 5-2 iv

N~EDO-33D47 - fnlrfon 0 5.2 HOP Mofnitorln and Control Systems ................................. .............. n..... 5.2 5.2.1 Pressure Control Syatoiri~.................... ~..................... I4.ec4e . e t ...... .. _5.2 5.2.2 Feedwete Control Sytm................................................ 5-3 5.2.3 Leakc Petaeonir System ..I*.~ ....... ..............

4. ~ 4.4 ..................... 5-3 5.3 Inistruzment Setpoinits........... .. ....... . ..... .die~ie 1441 ..... . .e.....e....I..... .......... *e .4 5.3.1 Ilgh-Preaaurc. Scram ..... . ..... . ...... . .

.... ........ e 5.3.2 High-Pressurro ATWS Recirculation Ptum Ti.......,, p . ....................... . '5 5.3.3 M m{in. Steam RH ief "ulveIolto ........... .......... .... .... .............. . ,. ...... e...S-5 5.3.5 Neutron Monitoring System ..............................-....... ..... _$6 5.3.6 Main Stzam i~n;,High Radiation Scram .. ................ ....... . .... .. 5-6 5.3.7 Low Steam Line Pressure MSIV Closure (RUN Mode)................ ..... .. 5-6 5.3.8 Reactor Water LevellInstruments.............. . .............. ............ 5-6 5.3,9 Mzin Steam Tunnel High.Temperamure Isolation ...................-

5.3.10 Low Condenser Vacuum ....................... I...............-.-.........5-7 5.3.11 TSV Closure and TOCl Fast Closure Scram BypaAs................-7 5.3.13 Pressue R~egul~ator ... .. ......... ......................... ................ S5-S 5.3.14 Feedwaer Plow Setpoint for Rzcircutation Cavitation Protection ......... .

5.3.15 RCIC Steam Line. Higb Flow Isolation . ....................... ... ........ -

5.3.16 J1PCI Stearzi Line High Flowv Isolatioun ................ ........... .............-. ..... 5-8 5.4 RWelezices ......... .eD..............Hc* .,....ee........ 5-

6. BL.2.C'RCALI POVt2R ANDrI AUXLIJIARIY' ShkSTBI..................e~...*... 6-1

6. 1.1 ,, POffSier. P.o w ~ y eer se m.. . .. i. ... e ,.... 4 ... Ie........ e......... b .e........ . ..e.~.,

d e . pc...

. . . . . . . 6 i1 6.1.2 On-site Power Distribution System ........................................ 6-2 6.2 DC Power .. . ..... ~. .... . . .......... ................ 6-3 6.3 F~uelIPool . .c........ .................... .............. _

6.3.1 Fuel Pool Cooling ,. ...................................................... . , 6-3 6.3.2 Crud Activityand Corrosion Products ...... ........................... 6-4 6.3.3 Ra-diation Levels ........... .. .... I......................... . .................. 6-4~

6.3.4 F'Me iRdks .............. .. e....e.... e............p....,. ~c. ........................... ,-

6.4.1 Service Water Systems............................................... ... .........-. 6-5 6.4.2 Main Condonued/Circulating WztedNormal Heat Sink Performance . ..... 6-6.

.6.4.3 Reactor Building Closed Cooling W..t& System . ... .... .........

e....... c. ..... 6-7 6.4.4 lRaw C~ooling W'ate Systemn.................... . . . .............. ............ I 6-7 6.4.5 ultimate Heat Sink_.4...... ........ 6-7 6.5 Standby Liquid Control Syetem..I.............................. .. . .......... 6-8 6.6 Powtr flependent HVA.C..H....,.,H. .. ..... ............. ..... ............ 6-9 6.1 Flire Protection........ .c....e....e..................................................... 6.9 6.7.1 10 CFR 50 AppodkxR FibeEe r......4..4...................................

nt 6-10 6.8 Systems Not InipacWe By Extende Power Uprate.. ...................

6.8.1 Systeis With Nojmpct ..e.e. ........... c,........................................6411 6.8.2 Systems With InsignIficant Impact.............................................. 6411 6.9 References ..... . ............... ..... ................

............ .. .... 6-1I V

., -.. . I . - I NEDO.33047 - RevylWon 0

7. POWER CONVERSION SYSTEMS . ................... . ...

. ................ .7-I 7.1 Tubine.(je iertor *.., . .. ." .... . . ....... ,,......... . ............... 7.1 7.2 Condenser And SteamnJet Air Ejacbn... ......... . .................. ........... 7.2 7.3 2Urbirie Steam By'puss . ... .......... ............ .... . ............. -.. .... ... ..... ........ 7-2 7..4 Feedwate And Condensate Sys:temns ................ .......................q . , ....

........ . ..... 7-2 7.4.1 Norm~al Operation ......................... . ............. ** ... ................. 4I ....... M 7.4.2 Tranmient Operation,.. ........................... ......... ..............

S.1I Liquid And Solid Waste~Maae ........................ ................ 8-I 8.2 Giseous Wgaste Mvlnagw ent...... ..... ................... . ...... .- ............. ...... 8.1 8.2.1 Offgu System.................... ..................... . ....... ........ 8-2 8.3 Radiation SoinceasnTbe Reactor Core ......................... ,...... -2...

8.3.1 NJorrmal Op eint ion. .......... *............................................i .... 8-20...........

8.3.2 Normal Post-Operation ....................... .......................... 8-3 8.4 Radiatloc. Sources In RPeotor Coolman.......... ...................... . ....... 8-3 8.4.1 Coolant Ac.Uivation Products ....................... . ...................... 8-3 8A.2 Activated Corrction madi Fission Product u... .......... ................. . ... .......... 8-4 8.5 Radiaion. Level ..............s ... . .. ... ......................................... 8-4 8.5.1 Norimul Opezation ...... W - ................. .. .......... .............. 8-4.

8.5.2 Normzal Post-Operation ........ . .......

I. 1.0.......... 8-5 835.3 PostAcoident.................... . ......................... ................... 8-5 8.6 Nomua1Opaatian Off-Sift Doses. . .. ..................... ...... .... -

9. REACTOR SAFETY PERFOXmANCE EVALUATIONS ....... ...... 9-I 9.1.2 Power and Flowv Decpendent Limits ................. 4*4*i. ..... ............... . 9-2

'9.1.3 Loss of Feedwater Flow Event~.............. .. . ..........

. ...........  :. 9-3 9.2 Design.Basis Accidents ................................................. . 94 9.3 Special Evenats ...-..... ~..............,...................

, .-...... ........... .... 9.4.

9.3.1 Anticipated Transient Without S==...................am........ .-....... 9.4 9.3.2 Station. Blackcout............................................ . . . .... 9.5 9.4 PRcfere~nces . ............................

,. ,.....p.,..q.. ......... .... ....... ... ,..... 9-6

10. OTHREvALuATIo ............ NS . .................. .......... ... .. 10-1 10,1 High Energy Line Break.................... ................ ..... .... . 10-1 0.1.1 ¶f*Wceaure, Prcsuro aad Htimidityr Prodc fl#n. ......................................... 10-1 10.11 M2z?4ir SrAmi ILine Breaks.......................... -...................., 10-1 10.13 Feedwatar Line Break,...... .. ............. . ... ............... .. 10-1 10.1.4 HPCI Steam Line Breaks.. . ...... . . . .................................... 10-2 10.1.5 RCIC Steam Linr Braks .........-........... .................. .......... 10-2 10.1.7 Pipe&Whip andJelitnIpingement .. ............ ................... ....... 10-2 10.1.8 Internmal 1FIoodmg frNm IIEIB L.ine Bresic.9 .... ..... .......... . ,.,..........,.... 10-2 10.2 Mode=a~ Bnerg Line Brea...........U. b... . I.............................. 10-2 10.3 Environmental. Qua~lificatian ........... ................. .. . ............. 10-3 10.3.1 Electrice! Equipment................................. ..... .... .... 10-3 Ai

NEDU-33047 - RevysLon 0 10.32 MechmniculBquipmeatWitNon-Me ikComp onents .......................

s........ 10.4 10.33 Mechanical Component Desmga Qualificaion.... ........ ................ 104 10.4 Testing .. ............. . . ,,,,,..... . ... ..... 10-4 10.4.1 Recirculation Puwp TtEDng ......... ............ ............. 10.6 10.4.2 I0 CFR 50AppendkiJTcsing . .. __ 10-6 10.4.3 Mn SteamLine1 Peedwator, andRetorReiruIslion PipingPFlw Induced Vibrntio Teouim..... ........... ............. _. ..... ....... 10-6 10.5 Individuil PIlu utB.ioT ....................... ..... ........ ,,,.,,.,,,..

, , , I. 10-7 105.1. I nitiating Byent En ~quency,................................. ,.....n. . ..................... 10-E 105.2 Coiponent andc System Reliability ,........................,.,. ._ . 10-9 10.5.3 Opertor Res o.n .. .. . .. ...... , ,.. ....... , 10-9 10.5.4 Succeas Criteria ......... .,.,._.....,,,e...,.......,...................................,,..,,,.... 10-13 10.5.5 Bftcml Events .........

_ . .. . 10-14 105.6 Shutdown Risk _ ........... _ . .- . . 10-14 10.5.7 ProbabilisticJRU Assement ..............-_..... . 0 10-I5 45.......

10.6 Opheztor TrsirigAnd HumanF tors

• dh............... .......... . .It.........,.,,..,..,,.,.,. 10-16 10.7 Plant Life ....... ..... .. .10-17 10.7.1 RPV Internal Components . . ... 10-17 10.7.2 Flow Accelerted Corrotion . . .. _ . _._._.,,.0-i 10-is....................

10.8 References ... . ... 10-19

11. IlCENSINGOBVALUA T IONS .... i..........,.,g.. .. . ..................

q ., . . 11-1 11.1 Oiher Applicable Hoquirements . .......

.............. ...... 11-1 11.1.1 NRC and Indugky Communications . . .. ...... -

11.1.2 Plant-Unique Items ... . . ..

................ .. , _. 11-1 11.2 References . .......... . ._ _ . .. 11-3 vii

. * . -* *. ." . ... - * .7 R*

NRDO3047

• Reavion 0 Tables Tablo 1-1 Olossary of Terms Table 1-2 Browns Fery Cmtrmnt and lPU Plant Operating Conditions Table 1-3 Computer Codes Used For EPU Analyses Table 3-la Brawas Fory Unit 2 Adjusted Reference Tenmperatues Tulal 3-lb Browns Perry Unit 3 Adjuted Refereneo Toeratures Tsble 3.2a Browns Ferry Unit 2 CUts of Liiting Components Tablt 3-2b Browns Pony Unit 3 CUFs of Limiting Components Tabl 3-3 Browns FerryRIPDs frNoral Con on (psid)

Table 3-4 Browne Ferry RIPDs for Upset Conditions (psidf Table3-5 BroxvnsPerryRlPDs farFaulted Canditions (p6id)

Table 3-6 Browns Ferry Rctor Internal Components - Summary of Stresses Table 3-7a Browns Ferry 30? Piping Table 3-7b, Browns Ferry BOP Piping CS and RHR Table 3-7c. Browns Ferry BOP Piping Main Steam System (Outside Containment)

Table 4-1 Browns Ferry Containment Pcrformance Results Table 4-2 Browns Ferry Short-Term Containment Input to NPSH Analysis Table 4-3 Browns FerryLong-Term Containment Inputto NPSH analysis Table 4-4 Browns Fary EPIUDBA-LOCA NPSH Margins ad Containment Overpresnuro Crdit Table 4-5 BrownS Ferry ECCS Perfonnance Analysid Reults Table 5-1 Drowns Ferry Analytical Limits For Setpoiuts Table 6-1 Browns Ferry EPU Plant Electrical Characteristis Table 6-2 BrwnS Ferry Offsite Electric Power System Tablt 6-3 Browns Ferry Spent Fuel Pool Parameers Table 6-4 Browns Ferry Efflunt Discharge Comparison Table 6-5 Browts Ferry Appondix R Fir Event Evaluation Resulsd Table 6-6 Browns Fcrry Systems With No Impact Table 6-7 Browns Ferry System With Insignificant Impact TableV-.l Browns FOrry Parameters Usod fbr Transient Analysis Table 9-2 Browin Ferr Transient Analysi Result Table 9-3 Browns Ferry Key Inputs for ATWS Analysis Table 94 Browns Ferry ATWS Analysis Results Table 10-1 Browns Ferry Hgh Energy Line Breaks Table 10-2 Browns Ferry Equipment Qualification for EPU Table 10-3 Summasy Comparison of Baseline and Updatod CDF and LIE Table 10-4 Summary of the Initiator Contributions to COF and LBRF for Brown Ferry Viii

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NEDO-33047 - RbInlan 0 Table 10. Frequency Weighted Fractional nportance to Core Damp of Operator Actions Used in Browni Ferry PRA Table 10-6 Results of Browns Ferry PRA Pereview Table 10-7 Browns Ferry FAC Parameter Comparison for BPU Figures pigure 1-I Browns Perry BPU Heat HalaeCO-Nominal Figure 1-4 Browns Perry EPU Heat Balhnce - Ovcrpressure Protection Analysis Figure 2-1 Browns FerryPowar/Flow Operating Map Figure 3-1 Browns Perry Response to MSIV Closure with Flux Scrun Figure 3.2 Browns Ferry Response to Turbim Trip with BYas Failure. and Flux Scram Figu= 4-! Browns Ferry Tine-integrated Coatainent Hydrogen Generation Figure 4-2 Browns Perry Uncontrolled Hi and O2 Concenralions in Dywell and Wetwel1 Figure 4-3 Browns Ferry Drywell Pressr oRespnse to CAD Operation without Venting Figure 4-4 Browns Perry CAD System Nitogen Volume Rcquiremcnt Figure 9-1 Browvs Ferry Turbine Trip with Bypass PFailure Figure 9.2 Browns Perry Generator Load Rejectioa with Bypass Failur=

Figure 9-3 Browns Perry Feedwater Controller Failure Maximum Demand Figure 9-4 Browns Perry Feedwter Controller Palbre - Maxinum Demand vvt Bypass OOS Ix

W-NED043047 -ReisIon O EXECUTIVE

SUMMARY

This report su=maie to esults of all dignifict safety evaluation. pcrformed that justify upoti th licnsed themnal power at Browns Ferry Units 2 and 3 (hereafter, Browns FPrry unlegs explicitly noted). The requested license power level is an ice to 3952 MWt fiom the current licensed reactor themal power of 3458 MWt.

This rzport folows the NRC approved generic format and content fbr BWR BFU lioensing reports, documented in NRDC-32424P-A, "Generic Guidelines For GCenera! Electric Boiling Water Reactor Extended Power Uprato," (ttorerncol) coamonly callod "BLTRI," Per 3Lfl1, every safety issue that slould be addr&esc in a plant-specific EU licensing report is addressed in thi report. For issues that have been evaluated generically, this report usually only roferences the (NRC approved) generic evaluations in either ELTRI or NEDC-32523P-A, "Generic Evaluations of Gencral Electric Boiling Water Reactor Extended Power Uprate," (Reference 2) which is commonly called "ELTR2."

It is not the intent of this report to address all tho detaiIs of the analyses and valuations reported herein. For example, only previously NRC-approved or industry-aoopted methods were used for the analyses of accidents and transients, as documented in ELTRI. Therefore, the safety analysis methods have been previously addressed, and thus, are not addessed in this report Alro, event and anysis descriptions that re already provided in other licensing reports or the UPSAR re not repeated vwithin this report. This eport summarizes the renits of the significant operational and safety evaluations needed to justitr a licensing amendment to alow for EPU operation.

Uprating the power level of nuclear power plants can be done safely within plant-specific limits and is a cost-effective way to increase installod electrical generating capacity. Many light water reactors have alrady been uprated worldwide.

An increae in electrical output of a. BWR plant is accomplished priarily by generation and supply of higher Btian flow for tho turbine generator. Browns Ferry, as orginally liccesed, has en is-designe cquipment and stem capability to accommodate steam flow rates at least 5%

above the cursrnt rating. Also, Browns Ferry has sufficient deeign margins to allow the units to be safely upruted up to 120% of its OLTP.

A higher gtear flow is achieved byincreasing the reactor power along slightly modified rod and core flow contro lines. A limited number of operafgparaneters are changed, wmnBtpoints are aute and instruments a recalibrated. Plant procedures ar revised, and tests simlar to son of the original startup tests uo peformed, x

NEDG43047 - Revison 0 Detailed evaluations of tho reactor, engineered safcty features, power convenion, emergency power, support systems, environmental issues, desi~gn basis accidents, and previous licesing evaluations wero performed. This report demonstrates tht Browns Ferry cmn safely operate at the requested licese power level of 3952MWt. However, non-safety power gearation modifications must be implemented in order to obtain the electrical power output associated with 1004 of the EPU RIP level. Until these modificados are completed, the non-eafety balance of plant may limit the electrical power output which in-tum maty limit the operating thermal power level to less than the licensed EPU TP level. These modiflcations have bee evaluated and they do not constituto amraterial alteration to the plant, as dicussied in 10 CPR 50.92.

The evaluations and revievs were conducted in accordance with the criteria in Appendix B of ELTRI. The results of the following evaluations end reviews, presented In this report, wer found to be acceptable:

• All safety aspects th are effected by the incrase in thrmal power and prating pressure were evaluated;

• Ealuations wore performed using NRC-approvod or industry-accepted analysis methods;

• No change, requiring comrpliance with a more recent industry code andtor standard, is being requested;

• The UFSAR will be updated forthe EPU related changes, after EPU is implemented, per the requirement. in 10 CPR 50,71(e);

• Modifications will be implemented under 10 CFR 50.59;

• Sytenms and components affeotd by EPU were reviewed to ensure there is no significant challenge to any safety syste;

• Compliance with cuitn plant environmental regulations were reviewed;

• Potnthally affected commitments to the WRC have been reviewed;

• Planed ohaangesnotyet implemented have el? been reviewed for the efcs of EPU;

• All EPU-related Technical Specfication changes are identified and justified; and The Browne Ferry licensing requiremeonw have been reviewed, and it is concluded that this EU can be accommnodated (1) without a signifcant increae in te probability or consuence of an accident previously evaluated, (2) without creating tho possibility of a rnw or different kind of accident fro any accident previously evaluated, and (3) without exceeding any existing regulatory limits applicable to the plant, which might cause a significant reduction in a margin of safety.

xl

NEDO-33047- Revision 0

1. INTRODUCTION 1.1 REPORT APPROACH Uprating the power level of uudcar ponWr plants can be don saoly within oertaia plani-speciflo limis. Most GE BWR plants have the apability and mains for an uprmti of Sto 20% witht major NSSS hardware modifcatios. May light water reactors have alre be= uprated worldwide. Over a thousand MWe have alreedy been added by uprato in the United States. Several BWR plants are among tbose that have already bee uprated The following evaluation upports au EPU to 3952 MW?, which corresponds to 120% of the OLTP. Tbh OLTP level is 3293 MWt.

This report follows the NRC approved generic format and content for EPU licensing reports, as described in Section 3.0 and Appendices A & B of HLTRI, and the NRC staff position letter reprinted in ELTRI. The analytical mathodologies used for ECCS-LOCA valuations, containmnnt evsluations, transient evaluations, and piping ections are dooumnted in ULTRI,Section I and Appendices D, B, 0, and KC The limitations on use of these method; as defined in the NRC staff position lettor reprinted in 0LTR1 woev followed for tis BPU analysis.

Many of the component, system and perfornance evaluations contained wtin this report have been generically evaluated in ELTR2, and found to be acceptable. The BLT2 genric evaluations ore basd on (1) a.20MA th power incnasz, (2) an increased operating dome pressure to 1095 psia, (3) i rector coolant tesuperature increase to 556WF, and (4)steam and FW hirras of aboW 24%. The plant conditions ssumed in the ELTR2 evaluations bound the conditions for this EPU.

A glossary of terms ieprovided in Table I-t.

1.2 PURP08E AND APPROACH An increase in electical output of a BWR is accoplished primarily by geneCrtion and supply of higher steam flow to the turbine generator. Most BWRs, as originally licensed, have an as.

designed equipment and system capability to accommodate steam flow rates at least 5% above the original rating. In addition, continuing improvenenta in the analytical techniques (computer codes) based on several decades of BWR safety tchnology, plant perforace feedback, operating experience, and improved fuel and core designs have reculted in a significant inease in the design and operating margin between the calculaW safety analyses results and the current plant licensing limits. The available margins in caculated results, combine with the as-designed excess equipment, system, Bad comnponent capabIlities (1) have allowed many BWRs to Increase their thenmal power ratings by 5% without any NSSS hardware modiflication, and (2) provide for power icreases -up to 20% with some non-safety hardsre modifications. These power inocease. involve no sairificazt increase in the hards presented by the plants as approved by the NRC at the original license stage.

The method for achieving higer power is to use the MELLIA power/flow mp, and increas core flow along the MELLL4A flow control linos. Hownvro, ther is no increas in the maximum allowable recirculadon flow value.

1.1

NED-33047 . Rzvsllon 0 1.2.1 UpraticAnalyus Bais Units 2 and 3 are currently licensed at 3458 MWt, and most of ft current safety anatysee are based on this value. Tle BPtU RW level included in this evaluation is 120% of the OLTP. Plant Bpocific B;PU parameter arr listed in Tablc 1-2. The BPU safet analyses are based on a power level of 1.02 times the ?PU power level unless the Regulatory Guide [.49 two percent power factor is already accounted for in the analysis metod. consistent with the methodology deribed in Refe 3.

1.2.2 Computer Codes NIRC-approved or industry-aocepted cn'uter codes eand calculational techniques are used to demonstrate compiane with the applicable regulary acceptance critteri The application of these codes to the EPU analyses complies Atith the lhictaions, restricons, and conditions specified n sa approving NRC SBR whete applicable for each oode. Any excptions to the use of the code or conditions of the applitzble SE are noted in Table 1-3.

1.23 Approah Ihe planned approach to aheving the BPU RTP level consists of'. (1) an increase in the core thermal power with a more uniform (flattened) power distribution to create increased steam flow, (2) i corresponding increase in the PW system flow, (3) no increz in maximur allowable com flow, Lnd (4) reer operation pritnily along an extension of the MELLLA rod/flow control Lines.

This approas6 is based on, and Is consistent with, the NRC-approved BWR generic M gidelines that are presented in. ELTRI. The plant-uniquo evaluations are based on a reiew of plant design and operating data, as applicable, to confirm excess design capabilities, end, if necessary, Identify amy items which may require modifications associated with the EPU. For some items, bounding analyses and evaluaions in ELTR2 demonstrate plant operability and nfety. The gencric analyses and evaluations in BLTR2 are based on a 20% of original licensed power inrease. Por the Browns Ferry BPU, the conolusions of "ystem/component acceptabilit stated in ELTRZ ae.

bounding. The scope and depth of the citation results provided ertin are established bosed on do generic BWR EPrU guidelines and unique fauea of the plant. Te results of the following evaluations, presented in this report, were found tobe acceptable:

(a) Reactor Core and Fud Performance: Specifio analyses requited for EPU have been performed for a representative fuel cycle with the reactor core operating at EPU conditions.

Specific core and fuel perforniawe is evaluated for each operating cycl, and will continue to bc evaluated and documented for the operating cycles ftat implement EPU.

(b) RCS and Connected Systems; Evaluations of the NSSS compon=ts and systems have been porforned at EPU conditions. These evaluations confim the, acceptability of the eff ots of the higher power and the associated change in process variables (i.e., increased pressure, trzpembzre, and steam and FW flows). Safety-rolate equipment performanc is the pnim:ry fbou; in this report, but key aspt ofreator operational capebility are also included 1-2

, I

NED0-33047 - oeviion 0 (c) Enginered Safety Feature Systems: The effects of EPU power opraion on the Containment, BOCS, Standby Gas Treatment syste and other ESF; have been evaluated for key events. Th4 evaluations include the ontainment responses during limiting AOOS, specie] eves, ErCOS-LOCA, and MSRV contaoinmt dynamic loads.

(d) Control and Insfrumentstlnn: The control and instnentatinn tignal ranges and amalytica limlits fbr setpointm hvo bce evaluated to establish the effects of the changes in various process p ometers such as powerpresso, neutron flux, stean flow and FW flaw.

As required, sepoint evaluaions have been performed to determine the need for any TS allowable value changes far various fuctions (e.g., mak steam Ino hish flow isolation setpoints).

(e) Electrical Power and Auxliary Systems: Evaluations have been perfored to establish the operational capability of the plast elecicel power and distribution system and auxiliary systems to ensure that they are capable of supporting safe plant operation at the BPU RTP power level.

(f) Power Conversion System: Evaluations have been performed to establish the operational capability of vrious non-safety BOP systes and components to ensure that they sre capable of delivering the increased power output, andfor the modifications nrcassary to obtain fllI EPU power.

(j) RadwaS Systems and Radiation Sources: The liquid and gaseous waste management system. have been evaluated at limiting conditions for EPU to show that applicable rclease limits continue to bi met dwing opemafion at hig power. Tho radiological consequonces have been evaluated for BEU to show that applicable regulations have been mat fb the EPU power conditions. This evaluation includes the effect of higher power level on source temn, on-site doses and off-site dose;, duinng nonmtl operation.

(h) Reatfor Safety Performance Evaluations: Th limting UFSAR nalyss for design bas events are perfonned as part of the EPU evaluation. ELTRI identifies the limiting a yses that require reanalysis for EPPU. The EPU results in no new lliiting event beyond those identified in BLTRI. All limiting accidents and transients are analyzed based upon Uimiting conditions for thie EPU and show continued conpliance with regulatory requirements.

(1 Additional Aspects of EPU!: HELB and EQ evaluations arm perforned at bounding conditions for the EPU to show the continued operability of plant equipment under tho E U conditions. The effects of the ENU on the plant IE are analyzed to demonstrate tae there nrc no new vulnerabilides to severe accidents, G) LcnsingEvaluations: Te applicableplant liwisig coitments, BBulletins, Crc3iar Notices, etc. are evaluated fortheeffects of heEPU. ((

)) Item unique to the plant are shom to be ac ptable for BPU operation.

1-3

r :

NEDO-33047 - Revlsion 0 1.3 UPRATSI PLANT OPERATING CONDITIONS Tho following evaluations justify icreasing the licened thermal power to 120% of the OUT value This new RTP valu provides an increase of steam flow to approximatey 123% of the original vahe, and a corresponding increase in elecotrical power ouwtp+/- To accoraplish is performance increase the ruted thermal power is increased to 3952MWt. The following deocriptions provide information on the original and the EPU plant operating conditions.

1.3.1 Reactor HeatBalance Tho cprating pressure, tho tot coro flow1, ad tho coolant t1 rodynamic stato chbarcterize to theml hydraulic performance of a BWR rector ora The EPU values of those paramers are used to establish the steady state operating conditions and as Initl and boundary conditions for the required safety analyse. The BPU values for hese parameen arm determed by perfing heat (eneg) balanc alcuions for the reactor system at EPU condito The reactor heat balaace relates the thennal-hydraulic parameters to th. plant stm and FW fow conditions for th selected core thermal power level and opeating pressure. Opeamiozml parameters from aotual plant oprtion aro considered (e.g, seam line pressure drop) when determining The expected Bt conditions. The thelmal-hydraulic parameters define the condifions for evahuating the opation of the plant at EPU conditions Te thecma]-hydrlic parameters obtained for t EBPU ccnditaio also define the steay ste operating conditions for equipment evaluations. Heat balances at apppately seleeted conditions define the inhiial and boumdaq conditions for plant sfty analyses.

Figure 1-1 shows the EPU heat balanct IWYo of B}PU and 100% rntd coe flow. Figu 1-2 ishow. he BPU heatbalance at 102% of EPIJ and 10%O core flow.

Table 1-2 provides a summary of te mactor thenal-hydraulic parameter; for the current rated mad BPU conditions. At EPU conditions, the maximum nominal operating reaor vessel dome pressure is maintained at the current value, which minimizes the need for plan: and licensing tanges. With the increased steam flow and associred non-;afety DOP modifications, the current dorne pressure provides gufficint operating trbine iWlet presreo to assure good prssure control characteristics.

1.3.1 ReRctor Perfotn ncelmprovenent~eatures tho UFSAR, core fel reload evaluations, and the Technical Specicationo rrently include aUlowcs for plant opeaio with the perhnance improvement features and the equipment 00S

• lsted in Table 1-2. When limitig, the input parameters related to the peformanc improvennt fetre or tie equ ent OS have been inluded in.the xafety anyses for EPU. The usn of these peubnao e lmrveet featu and allowing for equipment OOS is contiud during EPU opcraaion. The evaluation that are dendent upon cycle length are. peormed for EPU assuming a, 24-monthcycle,

i I NEDO-33047 - Reviion B 1.4

SUMMARY

AND CONCLUSIONS This evaluation encompasses an EPU to 120% of OLTP. Th straegy for ahieving hMA power is to extcnd the MELLLA power/flow map region ilong the upper boundary exteision.

The Browns Penry licensing requiements have been rviewed to donstrate that bis uprte can be acorrmodated without a uipificant inerae in te probability or consequene of an accidtt p iously evaluted, without creati the posibility of a new or different kind of aident from any accident preiously evaluat4d, and without xcding any wedsting regulatory limits or desin allowable limhz applicable to th. plantwhkch might cause a reduction in a margin of safity.

1.5 REFERENCES

1. GE Nuclear Energy, "Generio Guidelnes for CGntra! Eletic Boiling Water llnztor EPU,"

(ELTRI), Licensing Topical Report& NEDC-32424P-A, Class III (Propriedty), February 1999; and NEDO-32424, Class I (Non-prdprietary), April 1995,

2. GE Nuclear Energy, "Gencric Evaluation. of General Electric Boiling Water Reactor EPU,"

(ELTh2), Licearing Topical Reports NBDC-32S23P.A, Class III (Proprietry), February 2000; NBDC-32523P-A, Supplement I Volume I, February 1999; and Supplement I Volume Ii, April 1999,

3. GE Nuclear Energy, "General Electric Standard Application for Reactor Fuel, GESTAR-II,"

NEDE-24011IP-A- 14, April 2000.

I-5

W NEDO33047 -Revddion 0 Table 1-1 Glanary of Terms TelM DefInition AC Alternating current ADS Autorntic Depnsaudzation System ADHlR Auxiliary Decay Heat Removal AL Analytical Litrit ALARA As Low as Reasonably Achievable ANS Ameican Nuclear Society ANSI American National Standard. Institute AOO Anticipated operational occurrences (moderate fequency transiet events)

AOP Abnormal Operating Procedure AOV Air Operated Valve APLEOR Aversge Planar Linear Heat Generation Rate APRM Average PowerRange Monitor ARI Alternate Rod Insetdon ARTS APRM&BM/Teclmical Specifications ASMB American Society of MechEicial Engineers AST Alteniate Source Term ATWS Anticipated Transient Without Scm AV Allowable Value BHP Brake horse power BUT Baron injection initiation temperature SOP Balanoe-of-plant BPWS Banked Position Withdrawal Sequence BTU Brltilh Thera Unit BWR Boiling Waer Reactor BWROG BWR Owns Group BWRVIP BWR Yessel end Internals Project CAD Containmet Atmoaphere Dilution CBDT Causebased decision tree CD? Care damage frequency aD Condente filter demi eralizer CFR Code of Federal Rsgulations 1-6

NEDO.33D47 - Rg'IuIoJ 0 CLTP Cust Ucca60d Thermal Power CO Condensation oscillation COLR Co Operating Liznits~Rpot CP. Critical Power Ratio CRD Control Rod D[ive CRDA Control Rod Drop Accident CREVS Contzo Room Emergmqy Vontilation System CRHZ Control Room Habitability Zone CSC Containment Spray Cooling Cs Cora Spray CSS Ca support structurr CST Condensate Storage Tank CUP Cumlaltveusagpefctor DBA Deaig basis ccident DC Direct current DHR Decyheatremoval DLO Dun] (recirculation) loop operation ECMS Errergency Core Cooling System EECW Emergenqy Equipment Cooling Water MFPY Wffective full power years EHC Elcftrom-ydraulic control ELITR Extended (Power Uprate) Licensing Topical Report ELTR1 Gmeric. uidcdnes for General ElctricBoiling WaterRumtorEPU ELTR2 Generic Evaluations of GCneral Electric Boiling Water Reactor EPU EOC End of cycle EOP Emergecy Operating Procedure RPRI Eloctric Power Reoearh Inltute EPU Extended Power Upratc ESF Enginered Safety Feature EQ Environiental quelillcation

.FAC Plow Accelerated Corrosion FWR Final IFeedwoau Tenperature Reduction FHA Fuel Handling Accident FIV Flow induced vibration PLIM Failure Ukelihood index methodology 1.7

NEDO.33047 - Revislon O Term finhi PPCC Fuel Pool CoUng and Clea.p FW Feedwaer FWHIOOS Feedwaterlheater out of service 00C G*erkicComnmunivnlion GDC vnerOalDeaignCriteri4 G3B General Electric Company GENE General Electric Nuclear Energy GEL Generc Lcttcr HCR- Human cognitive reliability HELB High Energr Line Break HE? Human error probability HEPA High EfficiancyParticulato Air H& IlChes of mercwy absolute HPCI HighPressuro CoolantInjection HPT High'prcseurc turbine HRA Human Reliability Assessment UVAC Petting Ven:tiling and Air Conditioning HWC ydroe Water Chemistry HWWV Hardened Wctwell Vent HX Het exchange IASCC Irradiation-assistcd stress corrosion cranking ICE lnreased Core Flow ICs Intagrated computer system E3 Inspection and Enforcement BEa Inspection and Enforcanent Bulletin IEEE Institute of Eleoorical and Electronics Einetrs IGSCC Intergranular csass corrosion cracking IL13A Inshiunent Line Breac Acident IORV Inadvrtent Opening of a Roeief Valve 1PE Individual PlmntEvtluation IRM Intermediate Range Monitor ISP Integrated surveillmnco program LCO Liziting Conditions for Opeutaoa LDS Leek Detection Systemn LERP Large early relea ftequenoy 1-S

-s NEDO-33U47 - Revlison n rim Linear Heat Generation Rate LOCA Loss-Of&Coolant Accident LOFW Lossg-of-foedwater LOOP Loss of OWhitoPcow LPFC Low Rossura Coolant Jjcction LPRM Local Power Range Monitor LPSP Low Power Setpoint LTIP Long Term Torus Integrity Program MAAM Modular Accident Analysis Program MAPLHGR Maximum Average Planar Linear Hea Generation Rate MBTU Millions DfBTUa MCPR. Minimum Critical Powerkatio MELB Moderate Bnergy Line Break MELLLA Maxim um Extnded Load Line Limit Analysis MeV Million Electron Volts Mib Millions of pounds MOV Motor operated valve MS Main steam MSIV Main Steam Isolation Valve MSIVC Main Steam Iholation Valve Closure MSIV-LCS Main Steam Isolation Valve Leakage Control System MSL Min steam line MIJLB Main Stoamline Broak MSLBA Main Steamline Break Accident MSRV Main steam relief valve MSRVDL Main steam relief valve discharge lime MSVV Main steam valve vault MVA Mcgs, Volt Amps MYar Megavar MWO Megawattselectric MWt Megawatbermal NA Not Applicable NDE Non-destructive exnmination NPSJ Net positive sctiort head NRC Nuclear Regulatory Co~rnission 1.9

NEDO-33047 - Revislon 0 Tflm NSSS Nuclear steam supply system NTSP Nominal Trip Secpoint NUREO Nuclar Regulatns OLMCPR Operting Limit Minimum Critical Power Ratio OLTP Original Licensed Thermal Power

• 005 Out-of-servico OPRM Oscillation Power Range Monitor ORAM Outage Risk Assessment M gement AP Differential pressure - psi P's 25% of BPU Rated Thermal Power PCS Pressure Control System Pcr Peak cladding tempenturo PRA Probabilistic Risk Assessment PRFO Preasure RegulatorFailure-Opcn PRNMS Power Range Neutron Monitoring System PSA Probabilistic Sa&ty Analysis PSF Performance-shping factor psi Pounds per square inch pu. Pounds per square inch b absolute psid Pounds per square inch - differential psig Pounds per square Inch - gauge P-T Pressure-tzmperaturc RBCCW Reactor Building Closed Cooling Water RBM Rod Block Monitor RIC Reactor Core Isolation Cooling RCPB Reactor Coolant Pressure Boundary RCS Reactor Coolant System RCW Raw Cooling Water RHLn Residual Heat Removal RHRSW Residual Heat Removal Service Water RIPD Reactor internal pressure difference(s)

RPS ReactorProtectio System RPT Recirculation Pump Trip RPV Ractor Pressure Wese RRS Reactor Reirculation Systemn 1-10

T, 9 . - . ..- -

NED033047 - hevidon 0 Term RSLB Recirulaotin sytm line break RTP Raw Themal power RTNmr Rzfbrcnec tonepratre of nil-ductility tranition RWCU Reaor Water Clemup RWE Rod Withdrwwalirror RWM Rod Worth Minimnizer Sat EPTU alternaig Stres intensity Sm Code allowable sftrs limit SAR Safety Analysis Report SBO Stationblackout SDC Shutdown Cooling SER Safety Evaluation Report SFP Spent fiue pool SOTS Standby Gn Tratment System SJAB Steam let Air Ejector SHB3 Shroud head bolts SIL Service Infornnaion Letter SLCS Standby Liquid Control System SLMCPR Safety Limit Minimum Critical Power Ratio SLO Singlo-cop operation SOV Solenoid Operated Valve SP Selpoint SPC Suppression Pool Cooling SPDS Safety Paramieter Diaplay System SR Surveillance Requiment SRM Source Range Monitor SRP StandudReview Pan.

SRSS Square Rootof the Sum of the Squares SRV Safety relief valve SC0 Struchue, systen, component SSDS Safe Shutdoun System SSP Supplenentas Surveillance capsule program TAF Top of active f.el TB Tubine bypass T/O Tuzbino Generator 1-11

NEDO-33047 -Rcvidon 0 IR MfIUldon TBV! Tuibine Bypass Valve TFSP Turbine first stage pressurc TWV Turbine Control Valve TMG Turbine/gentuaor TLD Thennolunineseent dosimeter TTeohicl. Requirements Meniul TS TcChnical Specification(S)

TSV Turbine Stop Valve TVA Tennessoe Vallcy Authority TW Thi BmIlbIB UFSAR Updated Fina Safety Analysis Report UHS Ultimate beat sink USE Upper shelf mnrgy VFD Variable Frequecy Drive 1-12

- .. t .

NEDO-33047 - Revison C Table 1-2 Broswni Ferry Curent and EPU Plant Operating Conditions Cunwnt*

Lktnend EEN Pameter Value Thra Power (MWt) 3458 3952 4

Veacl SteamFlow Cvlbhr) 14.153 16A40 Full Power Coam Flow Range Mlb/hr 83.0 to 107.6 101.S to 107.6

% Raed 81 to 105 99to105 Mximum Nominal Dome Preare (psia) 1050 No Change Maximum Nominal Donc Tlmperture (T) 550.5 No Change Presure at upstrarm side of TSV (psla) 985 962 Full Power FW Flow (Ml/hr) 14.103 16.390 Temperaure (IF) 381.7 394.5 Core Inlot Fathalpy Btlb) 524.7 3' 523,2

• Based on curent reactor heat balace.

• At normal FW heating.

• At 100% core flow conditon.

Currently liceased perfonmance inwrovement featres and/or oqipment OOS tht are included in EPU evaluations:

(I) AtFl-MELLLA wich ?RNMS (2) BOC Coutdow (OESTAP. fOtric Anulyiis)

C3) SLO (4) FFWT (s) PWH00 (6) One MSRV 008 (7) 3% MMV8tcpnttotcra (E) 1CF

@9) EOCZRPrT00 (10) TSOOS (11) 24-moot fl cycle (12) ReacorLvcel3 Reductlon 1-13

NEDO-33047- Revilon n Table 1-3 Computer Codes Used Far EPU Analysem Txck Cumpoue Vemion or NRC Commenb TkCode Revhlin Approw!Cniet ReaaorHaLBaance ISCOR o9 Y(1) NEDE-24011P Rov. 0 SER Reamor Core snd Fuel TOBLA 04 Y NEDE-30130.P-A Prforma= PANACEA 10 Y NIDME30130-P-A ISCOR 09 Y(1) NEDE-24D1 IP Rev. 0 SER RPVFIuence DORTGOIV 01 N (2)

TGMLA 06 Y (3)

RPYVnItrnl3 SAP4007V 01 NA NEDO-10909 (4)

ShturgInteritY Evduatlon I Rearoor Inhtmg ISCOR 09 Y(t) NED&2401 IP Rcv. 0 SHR Prcss Dlfrnces LANS 07 (5) NEDB-20566-P-A Transient Analysis PANACEA 10 Y NED-30130-P-A (6)

ISCOR 09 Y(1) NEDE-24D1 IF Rev. 0 SER ODYN 10 Y NEDO-24154-A SAFER 04 (7) NEDC-32424P-A, NEDC-32523P.A, (8), (9),

(10)

TASC 03A Y NEDC-32084P-A. Rcv. 2 ATWS ODYN 10 Y NEDB-24154P-A Supp. 1, Vol. 4 STEM? 04 ( 1)

PANACEA 10 Y NEDB30130.F-A ISOOR 09 Y() NED-2401 1P Rev. 0 SE TASC B_ 03A Y NEDC-32084P (12)

Contkment Syttm SHBEX 05 Y (13)

RNsD M3 - 05 Y NEDO.10320, April 1971 LAM3B 08 (5) NBDE-20566-F-A

._ Sptenba 1986 Appendix R Fi OBESTE 08 (7) NEDE-2378.1-PA, RI. 1 Protctim S 04 (7) (8) (9) (10)

_ SHEX 04 Y (13)

US BILSO 04V NA (4) NBDE 235045, Februay 1977 [RLH6j 1I14

NEDO-33047 - Revision 0 Ta Computer Venon or NRC CodeC Rhdon Approved Commqit ECCS-LOCA LAMB 08 Y NEDE 20566-P-A ESTh. 0o Y NEDB.237B5-l-PA, Rev. 1, (B), (9),NEDC.

23785P-A. VolM, Supp 1, SAFER 04 Y Rev. 1 NEDE-23785-I-PA,

. Rev. 1, (8), (9), (10),

NEDC-2378SP-A, Vol m, ISCOR 09 Y(4) Supp 1, Rev. I TASC 03A Y NEDB-2401 IP Rev. C SER

_ _ NECD32084P-A, Rev. 2 Fission Product ORIGEN2 2.1 N Isotope G mtion and Inventory Depetin Code MS Piping Aitalas 7[PB 16 Y SttntlAnalyslmfrogra ME I50 17 (14) SuhretlAnajlsyagPrangm HYTRAN 1,6 (14) HydrauLic: Trmsmiat Analyim RBLAP5 3.a Y Usedforhydraulio modeling of two-phase flow.

HELD-OPC Mgmand BLAPS MOD 3.2 Y HELB.OPCwm flow rate EnergyRelsesa Vmd entapydata.

BELB-OPC GOTHC 6.1i Y EW4-OPCtmperaturo, SuWbrpurtnt pressure and reladve Analyuis humidity profiles for

. . reactor building areu.

Individual Pltant RISKMAN Windws NA RISICMAN is used us thD Evakation 5.02 Code fora many I submlttals toNRC mA 4.0.4, NA MAhP is used fr ft Rev 3 Ihorm-hydraAio analysis for may [ subitzais tD NRC.

1-is

NEDO3047 .Rovlulon 0 Competer VerdYm or NRC Codt RRvIsor BOP Performnco PEPSE 64A (15) Used to develop the turbine cyclehaablance.

Multd-flow 1.10 (13) Used for hydraulic modeling of tho condenate and FW stems.

RELAPS MOD 3.2 Y Usod for hydraulic modelIn of the heater drai sytem (two pbase flow).

Condenser Evaluation PEPSE 64A (is) Used todevdlop te turbine cycle host balace.

Raw Water Cooling Muldiflow 1.10 (15) Used for hydraulic EvaluetiarL modeling of the w water cooling system, Reactor Recirculation HYTRAN 1.6 (15) Usod to develop force tIme Vibration Monitoring hisoy fles ibrpresure pulsfition occurring during stcady sate operationoF the reactor recirculadon pumps and piping.

TPIPE 16 Y Used to evaluateu Seady stake vibrtion of the recirculation and attached piping.

Condensate Pump, TPIPE 16 Y Used to evaluatepiping Condcnsaeo Booster stresses and to determine PT, midFW Punp . support loads for modifcations, and rnodification; made to the Stem Packing eonot satapiping 5therB8puS Multi-flow 1.10 (15) Used for hydXauliO ValvC modifications. n1odelb ofthO codnae andd FWsystems FWHeatr SAP 2000 (15) Ue tODvstu age stresn in EvaluatonS the pa partition plates inside the PW heatek TurbineBuilding WRICOL 11 (15) Used to evalute the HVAC Evaluation performance of the cooling 1cSlu for hnwreased

_ ftmpewturea dueo EPU 1-16

NED-0.33047

• Revision 0

• Th application of these codes to the tU analyse omplies with the limitatons, rwtricti'v, and condios specifled Inthe approving NRC SER where applicable for each code. The applioation of the codes also comples with the SEMI fr the BPU programs.

Th1 ISCOR code is not approved by zwne. Howevn, the SER supporting approval of NBDE32401 1?P Ro. 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridloy (OR) finds thc models and methods acceptable, and mentions the use of a digital computer codde The ro*eced digitl computer code is ISCOR. The use of ISCOR to provide core thennal-hydraulic information in reactor internal pressure differences, transient, ATWS, Stability, and LOCA applioations is consistent with the approved models and methods.

2. WC(C543, "TORr-DORT Two. and Three-Dimensional Discrete Ordinates Transport Version 2.8.14," Radiation Shielding Information Conter (RSIC), January 1994.

3. Letter, S.A. Richard (tUSNRC) to 0. A. Watford (GE), "Amendment 26 to CIE Licensing Topical Report NEDBE-2 LI-P-A, GESTAR I - Irmplementing Improved GE Steady-State Methods (TAC No. MA6481)," November 10, 1999.

4. Not a safety analysis code that requires NRC approval. The code application is reviewed and approved by GENE for rLeveb2" application and is put of GENE's standard design process. Also, the application of this code has been used in previous power uprute submittals.

5. The LAMB code is approved for use in ECCS-LOCA applications (NDE-205P-A and NEDO-20566A), but no approving SER exists for the use of LAMB in the evaluation of reactor inteal pressure diffreces or containment system response. The use of LAMB for these applications it consistent with the model description of NEDE-20566P-A.

6. The phyuics code PANACEA provides inputs to the transient code ODYN. The improvrmects to PANACEA that were documented in NEDE-30130-P-A were incortporated into ODYN by way of Ameadment I! of GESTARII (NEDB.24011-P-A).

The us. of PANAC Version IO In this application was initiated following epproval of Amendment 13 of GESrAR II by letter from G.C. Lains (NRC) to J.S. CQrnnlEy (GE),

MEN 028-06, Subject "Acceptance for Referencing of Licensing Topical Report NEDP-24011I-P-A Amendmt 13, Rev. 6 General Blectric Standard Application kr Reactor Fuel,t March26, 1998.

7. Th ECBCS-LOCA codes are not explicitly approved for tranient or Appendix R usage.

The ataff concluded that SAFER is qualified as a code for best estimate modeling of loss-of-coolant accidents and loss of inventory events via the approval letter and evaluation for NEE-23785P, Revislon 1, Volume II. (ltter, C.O. Thomas (NRC) to J.F. Quirk (GE),

"Review ofNEDE-23785-I (P), "GESfl-LOCA and SAFERModels for tho Evaluation of the Loss-of-oolant Accident, Volumes I and ir', August 29,1983.) In addition, theuew of SAFER In the analysi of long term Loss-of-Feedwater events is specified in the approved LTRs for power uprate: "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," NEDC.32424P-A, February 1999 and t Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," NBDC-32523P-A, February 2000. The Appedix P. events are similar to the loss of FW and small broak LOCA avert, 1-17

NEDO-33047 - Revisoan n

8. Letter, JY. K 1pproth (GB) to USNRC, Tnmiittal of GE Proprietary Report NEDC-32950P "Corpilution of Improvements to GONE's SAFER ECCS-LOCA Evaluation Model," dated Jmnuazy 2000 by letter dated January27, 2000,

9. Letter, S.A. Richards (NRC) to J.F. Klapproth, "Gentral Electric Nuclar Energy (GENB)

Topioal Reports GENE Q43DC)-32950P and GEN (NEDC).32084P Acceptability Review," May 24,2000.

10. "SAFER Model for Evaluaion of Loss-of-Coolant Accidents for Jet Pump and Non-let Pump Plants;' NEDB30996P-A, General Blectric Company, October 1987.

11. The STBMP code uses fimdamenW mass and enery conservation laws to calculate the suppression pool heatup. VTe use of SThMP was noted in NEDE-24222, "Ausesment of BWR Mitigation of ATWS, Volume I & II (NJ4RBO-(60 Altemnate No. 3) December 1, 1979." The code has been used in ATWS applications since that time. There is no formal NRC review and approval of STEMP or the ATWS topic&l report,

12. The NRCapproved the TASC-03A code by letter from S. A. Richards, NRC, to S.F.

Klapproth, GE Nuclear Energy,

Subject:

"Review of NEDC-32084P, TASC-03A, A Computer Code for Transient Analysis of a Single Fuel Chnnel," TAC NO. M50564, Marh 13, 2002. The acceptance version has not yet been published.

13. The Lpplication of the methodology in the SHEX code to the containment response is approved by NRC in the latter to G. L. Sozzl (GI) from A. Thadani (NRC), iUse of the SHEX Computer Progran and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Prenure and Temperature Analysis," July 13, 1993, 14, Code provides input to TPIPE. TIPWE has been used by TVA to support aubmittals to NRC.

15. Code used to determine nosafety related paramete and infbrmation.

1.18

I I .4l NEDO33047 - RnhlSon 0 Ah- 1.1 H Carndifia ea uprutrm Oide afSV

.CowThwmsIPwsur Pump HeltIns 10.6 Ciwnup Loom -4.4 rave- Iface A I TurblnsCydo USA 3956.7 MWt Figure 1-1 Brown: Ferry EPu Heat Bounce - Nominal

(§ I100% Power and 100% Core Flow) 1-19

NEDW43047 - RevisIon )

Ai 1.i H

're1m aoJPevwt 4030.6 Pump Huftn [GA Cleunup Lcksa -4.4 Ohers Sem Logle -I .1I TurbinaCylo Urs 4W35.7 MMWt Figure 1-2 Browns Ferry EPU Heat Blance - Overprenure Protection AnRlysis

(@ 102% Power and 100%

IBoreFlow) 1-20

NEDO-33947 - RtvWaon U

2. REACTOR CORE AND FUEL PERFORMANCE 2.1 FUEL DESIGNAND OPIkTION EPU ireies tO averaP power density proportional to the power increase. Browns Ferry is currently licensed with n average bundle power of 4.53 MW/bundle. The averge bundle power far SFU is 5. 17 MW/Uundle. The ESFU average bundle power is within the rangO of other operating BWRH.

The avrage power density has some effects on. operating flexibility, rcutivity charactnistics end energy requirements. The additional energy requirements for E3PU are,met by an incrcave in bundle enricement, an increae= in th reload fuel batch size, and/or changes in fuel loading pattern to naintain the desired plant operating cycle length, The power distibutioti in the core is changed to achieve increased core power, while limiting the IACPR. LIGR, and MAPLHOR in any individual fuel bundle to be within it allowable value as defined in the COLR.

At the OLTP or the EPU RTP conditions, all fuel and core design limits continue to be met by plained deployment of fuel enrichment and burnable poison. This is supplemented by core management contr! rod pattern and/or core flow adjustments. ((

)) However, revised loading patterns, larger batch sizes and potentially now fuel designs may be used fo provide additionail operating flexibility and maintain fuel cycle length.

Tho EPU evaluations assuma a refrmc ecyilbtrim core of (3B14 fuel. No new fiel product 1ne, designs are inoduced for BFU, and EPU doe notrequire a change to any fiel design limit. The fuel design imits arm established for all new fuel product line designs as a part of the fuel introduction and reload analyses, if The reactor core design power distribution usually represents the most limiting themal operating uteat derign condifion. It icludes allonuve, for the combined effects on the fiel heat flux End temperature of the gross and local power density distributions, control rod pattern, and reactor power level djustmnt ding plnt operation. NRC approved core design methods were used to analyze coro pszfonnano at the EPU 1P level, Detailed fel cycle calculations of a repeseentativo oore dosign for this plant demonstate the feasibility of SFU RTP operailon while maintaining bfel desiga limits. Thernnalhydraulic design and opeting limits ensure an acceptably low probability of boiling tranition-induced fuel cladding faile ocurzig in the core, even for the mosL geere postulated peations] transents. As neoded, limits ame also placed on fuel APLRGR and/or fuel rod LEGRs in ordar to meet both peak cladding ter catre liaibt for TIe limiting OCGA and fuel mechaical desigabues.

Tho subeque reload core designs for opeon at the SF1 RTP level will take, into accouzt tho above limits, to ensure acceptable differences bewen tle licensing limit, and their corresponding operating valum, 2.1

NEDO-330347 - Revision 0 EPU may result ia small chamo in fuel buup, the =om of fuel to be used, and isotopic concentrations of the madionuclides in the irmdaed tE relative to thz original level of burnup.

NRC-approved limits for burnup on the fuel desgp are not exceeded. Also, due to ths higher Bteady-st Ctoperting power associated with tho BEP, the shot-tom2 curie cont of the reactor flul inreases, The eftots of higher power operation on radiation source; and design basin accident doses are discussed im Sections B and 9.2 repectively. EPU has some effects on operating flexibllil%, reactivity charateries, and energy requirements. These Issuw are discussed Inthe following secliona based on GE experiece and f chractcriftics.

2.1.1 Fuel Thermal Margin MonltDring Tbreshold The power level above which fuel thermal margin monitoring is required changes with BPU.

The original plant operating licenses EOt this monitoring ftesbold tt a typical value of 25% of a1P. rr Thc fuel thermal margin monitoring threshold is scaled down, if nocemssry, if

3) lherefbore, the Browns Ferry fuel thernal monitoring thrshold is Iowered to 23%

CE 1))

A chango in the Mel thermal monitoring threshold also requires a corresponding change to the TS reactor core safety limnit for rcduced preesure or low core flow. The above discussion is consistent with the TS related discussion in Section 9.1.

2.2 THERMAL LIMITS ASSESSMENT Operating lihits sure that regulatory and/or safey lmits are riot exceeded for a rangs of postulated events (e.g, transients, LOCA). This section addresses the effecti of BPU on thermal limits, A rerco equflibrin ce of GE14 The! is used for the PU evalustion. Cycle-apecific core configumtions, evaluated for each reload, confim. EPU capability, and establish or confirm cycls-apecifio limits, as is currently the practioe.

2+/-1 Safety Limit Minimum Critical Power Raio The SLMCPR can be affected slightly by EPU due to the Hater power distribution inherent in the incasd power level. 1[

2-2

'- . I NEDO-33047 - Revislon B

)) The SLMCPR analysis reflects thE actual plant or loadig patten and is pefncd ft each plant reload c . [

J))

2.27 Minimum Critical Power Rato Operating Limit The OLMCPR is detmincd On cy*lo-pecific bsis fro doe ~ult of th reload trasiet analysis, u described in Sections 5.32 and 5.7+/-.1 of ELTRI and Section 34 of BLIS (Reerence2). Tis roeachdoes not chang forPU. Theorequird OLMCPRis notexpBed to sigificantly change (< 0.03) as shown in Table 3-1 of ELTRI and Figur: 5-3 of ELTR2 and from expedenca with otier uprated BWRs. For the reference core of 0B14 fael, the OLMCIPR for EPU FTP operation is shown in Table 9-2.

II 1] The OLMCPR is calculated by adding the change in MCPR due to the limiting AOO event to the SLMCPR, and is determined on a cycle spncific basis. EPU does not change the method used to determine this limit. The effect of EPU on AOO events is addressed in Section9.1. [

2.2.3 MAPLUOR and Maximum LEGR Operaflng Limits The MAPLHGR end maimum LRGR limibt r maintained as described in Section 5.7.22 of ELTRI. No s.gnifcant change in openilon is aticipat due to the BPU bued on experience fi=

other BWR upratrn, Tho ECCS peformane is addressed in Section 4.3, and uses a reference equilibriumcoreofGEl4ftmel brEPU. E(

J))

2.3 REACTIVITY CHARACTERISflCS All minimum shutdown margin requirements apply to cold conditions, and a maintained without change.

Operailon at higher powtr could reduce the hot excess reactivity during the cycle. This IoSS of ractivity does not affe safety, and is not expected to signifcty affoct dhe ability to nmnage the power distribution through the cycle to abieve the target power level. However, the lower hot excess wctivty can result in aiving an earlier all-rods-out oondition. Through fuel cyle redesign, sufficient excess reactivity can be obtaind to match the desired cycle length. Increasing hot reactivity may result in less hot-to-cold reactivity diffirnces, and therefre, mmaler cold shutdown margins. Howover, thi potentia loss in margin can be accommodtad trough cmo design, and crent design and TS cold shutdown margin requircmes are notaffetd. Ifneeded, a bwidlc design with Improved shudwn mrn chctezistics can be used to prev tht flOXubilhH between hot and oold reativity requiremnents for fiture cycle.

2-3

NEDO-33047 - RvaWon 2.3.1 Power/Flow0peratiughMap The BPU aluyses and evaluations conservatively usuflit he, current licensed MELLLA and ICF operating domains. The EVP power/flow operating map gre 2-1) includes the operating dowain changet for BPU, and also shows the applicable Browns Ferry porformace improvement features (e.g., MELLA and 1CE) addressed in Sectin 1.3.2. Th ciangcs to the power/flow operating map ar [(

J3 The maxium thermal operating power and maximum core flow shown on Figurc 2-1 correspond to the EPU RI? and the previously analyzed cor; fow range when resealed so that EPU RTP is equal to 100% rated.

The power/flow operating zap changes, inoorporated into Figure 2-1, ar consistent with the changes shown it Fgure 5-1 of ELTRI.

The details of th reactor operating domain for BPU conditions ar provided in Fiure 2-1. The operating domain for EPU isdefined by the following boundaries:

• tho MELLLA uppor load line- extended upto thEBPU RI? lcvd; s the maximum EPU RTP corresponding to 1200% of ihe OLTP; and

• thcICFlineuptoEPURTPBtjOS%ofratdcarecflow.

Consistent with ELITR 1, these boundauies define er increas in the extent of the operaing domain above the OLTP between the extended (relative to OLT?) MELLLA upper load line and ICE line, Thernal hydraulic instability wcolusion regions are not shown on FiWe 2-1.

Analyses and evaluations have been performed to demonstrate that Browns Ferry may incrase core flow to opera within tie region of the operating map bounded by tho con*tant speed line between lOOP/105F and 52.SP/112.6F for EOC coatdown at constant maxiinum pump speed line.

EPU doea not affect SLO because the maximi stainable thermal power during SLO is leas t1 CLTP, and is limited by the available recirculation flow. SL is bounded by the MLLLA domain in terms of absolute thermal power versus core flow. Therefore, a separate SLO power/flow operating sap is not needed forEPU.

2.4 STABILTY Browms Ferry has installed a PRNMS with OPRMs to implement 1he BWROG Long-Term Stability Solution Option III. OptionlII ovaluations are Core reload dependent and are performed for each reload fUel cycle.

Option III is a detect-and-suppross solution, which combins closely spaced LPRM detectors into "Cells" to effectively detect reactor instability. Browns Ferry, having imuplemrnted Option III has demonswated that the Option INUip setpoint Isadequato to provide SLMCPR protection for anticipated reaor instability. Tnis evaluation is dependent upon te core and fael design 2-4

C. _ _ _ _ _ '_. * . - ** *

* .

NSDO-33047. RevWon 0 and Is performod for each reload. Thefore. the effect of EPU wili be analyzed for tho firss reload wi~dl inorporaws the new rated powerlvelv The OPRM indesigned to provide the Option 2II automnatio scram.

The Optio III rip Is armed only whe plant operation is within the Option 1I trip-mabled region, Tbh Option III trip eabld region is defined as the region on the pow/rflow map with power >30%Q OLTP and core flow <60% ratmd core flow. The actual flow setpoint is doetrmincd by recixulation drive flow at Browns Feuy. Fox EU, the Option III tip-enabled region is rescaled to maintn the same absolute power/flow region boundaries, Becaus. tht rated core flow is not changed, the 60% core flow bounmduy Is not scaled. The 30% OLT. boundary changes by ithe fbllowing equation:

EPtJ Region Boundary = 30 OLTP ((100% + BPU (%OLTP)

Thus, for a 120% of OLTP EPU:

EPTJ Region boundary - 300% OLTP (1S *F120D%) - 25% EP The OPRM setpoint will be evaluated for the uprated reload core prior to EPU bmplekentatiot 2.5 REACTIVO CONTROL 1.5.1 Control Rod Drive System The CRD system is used to control core reactivity by positioning neutron absorbing control rods within tho reactor and to scram the eactor by rapidly insorting withdrawn control rod; into the core. No change is made to the control rods due to the EPU. Thz effect on the nuclear characteristics of the fuel Isdiscussed in Section 2.3.

FarBrowns Perry, the scram limes aro decesed by the tansient press8ro resposel ((

.JJ At nonnal operating conditions, tis aocumiilamor supplies the initial wram preasure and, in the scrn continues, the reactor bcomes fth primary sorce of prcssux to complete. th scram. ([

.1]

2.5,2 Control Rod Drive Poitoning and Codllng and the autoatic operation of the system flow control valve maintains the required drive wzter presmure and cooling wate flow rate. Therefore, tec CRD positioning and cooling fnotions are not 2-

NEDO-33047 - Revtsion 0 fliocted. The CRM cooling and normal CRt posidwdzg fimfons ge opeatioial wCrsidons, not frty-nlatd fnctions, and we not aebted by EPU operuting conditions.

Plant opeantng data has confirnied that t CRD system flow control valve operating position has nfficient operating margin.

2.5.3 Control Hod Drive Integrlty An ent The postulated abnormoal oprting condition for the CRD design asumee a failure of th8 CRD Bystem preesure-rogualatiog valve that applies the maxim pump discharge pressure to the CRD mochnnism internal components. This postulated abnorrmal pressue bounds the ASM reactor ovcxpressure limit. [(

1 Other mechanical loadings a addrssed in Section 3.32 of this report.

2.6 REFERECS

1. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," (ELTRI), Licensing Topical Reports NEDC-32424P-A, Class III (Proprietay), February 1999; and NEDO-32424, Class I (Non-proprietary), April 1995.

2. GE Nuolear Energy, "Generio Evaluations of General Electric Boiling Water Recor Extended Power Uprate," (LTR2), Licensing Topical Reports NEDC-32523P-A, Clasp Im (Proprietary), Februauy 2000; NEDC-32523P-A, Supplement I Volume 1, February 1999; and Supplement 1 Volumo fl, April 1999.

2-6

WEDO-33047 - Rteio S core FW (thr) a 6 M6 2& M t 0 *f to I" M It. tin I3jIIIIIIIIIIIII.I i-i.---. j----------- _ _

.450 iLa:

mn- M Iac4

+/-fl CFl flw l 9 9 - *-r~ 9~ - --~~~ ~~r~64~

,_ _ ,. 'r------  ! XArhI t g

-nsne usaL Mhlfca

-- t 1 _ -_-- ~'rI; .~; ;4 Os I

.. - - '- -r .M --- '-- . .

WMfltwi4 I

.am1 '--

I-, X I

g:¢ la 9

sot 4 30 6 10 Mb fl A*

• S t id M MO M18 1il Coreflw(%K Figre 2-1 Browns Ferry Powerflu Operanting Map 2-7 I

NEDO3047 - RsIulon D

3. REACTOR COOLANTAND CONNECTED SYSTEMS 3.1 NUCLEAR SYSTEM FRRSSM MEIRP The nuclear system pressure relief system prevents overpreamuzition, of the nuclear Dsytem during AGOs, te plant ASMB Upset overpressure protection event, and postulated ATWS sveas. The MSRVn along with other finctions provide this protection. An evaluation was performed it order to confirm the adequacy of the pressur relief system for BPU conditions.

nTe adequacy of the pressure relief systm is also demonutrated by the overpressure protection ealuation perforned for each reload coro and by to AMfS evaluation performed for BFU (Section 9.3.1).

For Erowns Ferry, ONI RV .etpoint incre;e is needed becausc ther is no change in the do=

pressure or simmer margin. Therefore, there is no effect on valve functimnaity (opening/closing).

3.1.1 MSRV SetpolntToieznce MSRV uetpoint tolerance is independent of U. EFL evaluations are performed using the existing MSRV setpoint toleranoe analytical limit of 3% as a basis. Actual historical in-service surveillance of MSRV setpoint perfomnance test rults are monitored separately for compliance to the TS requirements.

3.2 REACTOR OVRRPRESSURE PROTION ANALYSIS The design pressure of the reactor vessel remains at 1250 psig. Th accmptnce limit fbr prssurlzatio events is the ASME code alolanble peak pramsure of 1375 psig (I10% of design iou). The overpressw profttion analysis description and analysis method are provided In Section 5.5.1.4 and Appendix REof ELTRI (Reference I). As shown in Table E-1 of ELTR1, the limiting presurization eonts are t MSWV olonu and turbine tip with turbine bypass failure.

Both events ar (ccnseatively) analzed assuming a filure of tho valve position scrmr. Tho anatl o tat de m a ror ospssure oF pi l055 (Whiis higher thn the nominal EPU do= preasare), the MSRV analytical limi in Tibla 5-1, and one MSRV (with lhe lowest setpoinf) OOS. Startg from 102% of EPU RTP, the calculat peas RPV pressure, located at the bottom of th vesscl, is 1342 psig. The corresponding calculted maximum reactordomprnssureis 1314psig. he peakcalculatrdRPV preaureremainsbelow the 1375 psig ASME limit, and the mdximum calculated dome pressure remains below the TS 1323 psig Safey LimiiL Theroe, there is no decrease in marg of safety. The results of the HPU overere protection anlysis tre given in Figrs 3-1 and 3-2 and are consistent ((

]1 3-1

f w s v s -

NEDO-33047- Reviion D 3.3 REACORVESSELAND INTERNALS The RPV striture and support componzts form c pressUre boundary to contain the reactor coolant and moderator, and form aiboundary against leukage of rdroactiv materials into the drywel. Tho RPV also provides stuctural support for the reactor core and internals.

Conpwrehnaivo rviews have asessed the effects of fineased power conditions on the eactor vessel and it. intenals. Ibese reviews and awoclated amlyses show continued oompliance with the orinl design and licensing criteria for the rear vessel and internais.

33.1 Reactor Vessel racture Toughbue The netun fluence is both reanalyzed for EPU, based on capsule flux wire data, and relcakued using 2-dimensional neutron transport try (Reference 3); she neutron transport methodology is consstent with Regulatory Guide 1.190. Te Regulatory Guide 1.190 fluxs are conservaively applied for the entiro 60-year plantlife. The revised flc is used tD evalue the ves a stthe requi emnfof I) CFR 50, Appadix 0. Tbe results of thso ervaluaions indicate that (a) Tho USE reI0 ns bounded by the BWROG equivalent margin analysis, thereby demonstrating compliance with Appendix G.

(b) The beitline material RThn-r mairn below 200F.

(c) Tho Technicel Specification P-T curves have been revised in accordance with the 1998 Edition of the ASME Boiler and Pressure Vessel Code including 2000 Addenda (Reference 4). The hydrtst pressure for EPU is the minimum nominal operating pressure.

(d) The 40 year life (34 effective full power year (EFPY)) shift IS increased, and consequently, requires a change in the adjusted reference temperature, which is the initial RTNDT plus the Shift. This shift v, used to revise the P-T curves for EPU (Reference 4). Thess values and the 60-year life (52 EPY)b ift providd in Tables 3-lnd 3-lb, for Browns Fery Units 2 and 3, respoctvely.

(e) The surveillance program wnsists of three capsules for each unit One ¢apsule containing Charpy specimens was removed from the Browns Ferry Unit 2 vessel after 8.2 EFFY of operation (end of Fuel Cycle 7). tested, reconstituted, end placed into the vesel during the Unit 2 Cycle 8 Refueling Outage. The remaining two capsules have been in the reactor vessel since plant startup. BPU has no effect on The existing surveillance schedule. 'Th first Browns Perry Unit 3 capsulEwas remved from the vessel during the Fuel Cycle 8 outage, but was not tete. Browns Ferry is part oftho BWRVIP ISP (SSP Program and complies wit the withdralvsl schedule specified for representative or surrogate surveillance capsules that now represent each unit Therofore tt I0 CFR 50, Appendix H surveillance capsule schedule for the ISP/SSP governs. Implementation of EPU has no effct on tebo BWRVIP ithdawal .chedule.

The maximum nominal operting dome pressure for EU is unchanged fro that for currnat power operatio. Therefore, the hydrostatic and leakage test pressureR ar accetable for the MEPt. Because the vessel is in compliance, with the regulatory requirements, operation with BPU does notbave an adverse effect on the reactor vessel fracturc toughness.

3-2

NEDO-33047 Revision 0 3.3.2 Reactor Vessel Stuctrural Evaluation The efEct of RPU wau evaluzted to enure that the reactor vessel components continue to comply with fth existing structural requirements ofthe ASME Boiler and Prssure Vessel Code.

For the compoents under Onsidcrtion, th 1965 code with addenda to and inchiding swner 1965 for Unit Zand summer 1966 for Unit 3, are the codes of construction, and were used as the governing codes. However, if a component'; design has been modified, the governing code for that compo t wa the code tued ln the strest analys of tho modified component New stresses w*ere deteined by scaling the "loriginal" ute;Ces based an the BPU conditions of temperature and flow. The analyses were performed for the design, the Norm and Upset, and (hoEmergencyandFaulted conditione. Any increein annulum pressurization, jetreactionpipe restraint, or fuel lift load,, was considered in the analysis of the components affieted for Normal, Upset, Emergency, and Faulted conditions.

3.32.1 Design Couditions Because there are no che s in the design conditions (vessel pressure and tmperaure) due to EPU, the design stessme are unchanged and the Code requirments are mnet 3.3.2.2 Normal and Upset Condilons Th rcwr coolant temperaturo and flows (exsct core flow) at BYU conditions are only ligtly changed from those at crent rated conditions. A CLUP analysis for Browns PFey was performed In 1997. Only cianges in temperaur and flow since tat time, arm considered I rein, The only other loads, which oduld affeot the EPU RPV evcluaion are mehani loads such as seismic, fieol lift and Rechculation LOCA loads which do not chage with EPU. Therefore, only changes in twpemrature and flow are considered. Evaluaios were performed at condition that bound the slight hane ia operang conditions. he type of evaluations is minly reconcililion of the stresses and usage &ctos to reflect BPU conditions. A primary plus secondary strss analysis was performed showing BPU stresses still meet the requirements of the ASME Codc,Section III, Subsection NB. Lastly, ffth Iigue usage was evaluated for the limiting location otoornpa with a usage f&otor greter tan 0.5. The Browns Ferry fittigue analysis results for the limiting components are provided in Tables 3.2a and 3-2b. The Browns FPey analysis rcul] forEPU B Eshw tha all components meet their ASMiE Code requirments.

Browns Perry FW nozzles with the triple-lecvn. double-seal, thenral slcove design ure qualified by UT inspetion iethods based on ASME Section M Code, which ar approved by NRC.

FFWTR is consid ldin the fatigue usage evaluation end included with the sytemr cycling effects shown in Tables 3-2a and 3-2b.

3.3.2.3 Emergency and Faulted Conditions The stresse due to BEmergency and Faulted conditions are based on loads such as peak dome pressur which o unchanged. Thcrefore, Code rtquireennts are met for all RPV components.

3-3

NEDO-33047- Rrbvloa 0 3.3.3 butr lntnenal PreureDferen (R l)

The inease in core average power aone would result imhigher co loads and RPDs due to the higlrr core exit steam quality.

The RIPDB are calculated for Normal (.teadysitte opertion), Upset, and Faulted coditic for all maorretorinternalooroxezti. [(

Tables 3-3 trough 3-5 compar results for the various loading cmdiions between curent analysis results and opmaton PI for thevcssel intnal tAeaffected by the changed RIPDs.

3.3.4 Reactor Internals Structural Evaluation The reactor internals consist of the CSS components and menmCSS components. The reactor interals are not oerdfied to the ASME oode; except the control rod drive as noted, however, the roquiremornnt of the code are used u guidrliace ia their design basis analysi. Th' evaluations and stress recorwiliation in support of thz thermal power ncreas arc performed consiftent wit the design basis anslysis of the.components. The reactor internal components evaluated are:

Core Support Structure Component.

• Shrud

• Shroud Support

• Core Plate

• Top Guide

• Control Rod Drive Housing

• Control Rod Guide Tube

• OrficedFuol Support

• Fuel Channel Non-Core Support Stmcture Components

• Steam Dryer

• FW Sparger

• Jet Pumps

• Core Spray Line and Sparger

• Acce aKHole Cover

• Shroud Head and SteamSeparator Assembly

• Iacoro Housing and (uide Tube

• Vessel Head Cooling Spray Nozzle

• Jet Pump hnstrument Penetration Seal

• Differential Pressure and Standby Liquid Control Line

• Control Rod Drive 3-4

NEDO-33047 - Revidon O The orinl configurations of the intnal components arc considered in the EPU evsluation unless a component has undergone permanent structiral modifications, in which case, the modified conflguration Isused as the basis fbr &t eviluation, The efect on the loads as a result of the mal-hydramulic change duc to EPIU reevaluated for th reactor internal;. All upplicable loads and load combhvaiwis arm considred coosiatnt with the xisting dcsig basis nlyiui. These loads inlude the RIPD4s eimic. loads, flow induced and mous; louds due to RSLB-LOCA, nd thermal Ioed.. The RIPDs increase lbr sotnme co onts/loading conditions as a result of EPU. The flov conditions and thernal effwts were considered in the evaluation, as applicable. The seismic response is unaffected by EPU. The acoustic and flow inducod load. in the anmulus as a regult of the RSLB-LOCA are included in the tvaluation and are bounded bypre-EPU valute.

The EPR loads are compared to tdom inthe existing deign basis aailysIs. If the loads do not Increase due to EPUT, the the existi analysis results bound lho EPU conditions, and no fitatr evaluation is required or performed. If the loads incrse due to tie EPU, then the effect of the load increase is evaluated frther.

J))

Table 3-6 presents the governing stressae for the various reactor int<el cowonenzs. All sesses ere within allowable linits Bad the moctor intmnal components ire dmonstrated to be struct4rully doquatc for EPU.

The following reactor vessel intenals am evaluated forthe eftm of ehanges in loads due to EIPU a) Shroud: IL

)) Therefore, thc stuctural integrity of the Shroud is acceptablc for EPU.

b) Shroud Support: (

1] TMefore the structuml integrity of the Shroud Support is acceptableC for EPU.

3.5

99 v @ J *o *s r -@

tr NEDO-33047 -Revision a c) CorePlate; [E

)) Therfore, the core plate remains structurally qualifld for EPU.

d) Top Guide: I[

]1 Therefore, tb stnutural integrity of the top guide iS acceptable for EPU.

c) Control Rod Drive Housing: ((

))

Therfore, the structural iniegrity of the CRD housing is acceptable for WPU.

f) Control Rod Guide Tube: ((

)) Therefore, the stnctucral integrity of the control rod guide tuba is awcptable for EPU.

g) OrIficed Fuel Support. C(

J] The&reo, the struotural integrity of the oriflood fiel suppart is acceptable for EPU.

h) Fuel Channel: [f

]1 Therefore, the sttural egit of tde eIecbannels I acceptable forRPU.

i) Steam Dryer: ((

)) In response to tho recent dryer Miures obseved t another BWR site durmg EPU operation, a detailed evalution wiIl be performed to exanine dyer components susceptible to faIlure at EPU conditions. Th results of the quatitative evaluation will be used to identify any additional modifications needed to maintain steam dryer tructurul 3-6

NEDO-33047 - Redvdon 0 intrity at BPU conditions. If any steam dryer components requiring modification ae Identified, these modificatiosm will be implemoatd prior to opeation at theo PU conditions. Referto Section 3.3.5.

j) Feadwuter Sparger: ((

JI Therefore, the stiuctural integriy of the FW spaer is ncctable for EFU.

k) Jet Pumps. if

)) Threfore, the structural integrity of the jet pump assembly is mcceptable for EPU.

Jet Pump Riser Brace Repatr (Unit 3): Because the load conditions pertaining to the jet pump and tho riser brac repair remain unaffectod by EPU, tho existing reair design basis rensims valid for BPU. Repair inspection, however, should continue using the rcommendations curently in plac.

1) Care Spray Ltnes mnd Spargers: a Therefore, the stuctural integrity of the core spray line and the spargers is acceptable for EPU.

Core SprRy Line T-box And Downeoner Modifitltonv (Unit 3): Because the applicable loads for the care spray system remain unaffectsd by EPV, the Com Sray Line T-box end Downcomor modifications remain qualfied in the repaired conditon Repair inspections should continue using the curret recommendations.

m) Access Hole Cover: l[

3 Thereflor; the structural integrity of the Acss Holo Cover is acceptable for BlJ.

3-7

NEDU-33047 -Revision a n) Bhroud Head and Steam Separator Assembly (inMluding Shroud Head Bolts): ((

)) mh~fbre, the rtructural integity of the shrud bead and stewn separator asembly Is acceptable for EPU.

a) In-Core Hoaling and Guide Mebt: Et j1] Thlfore, thz struohtral itegrity of fhe In-core Housing ed Guide Tube is acceptable for EPU.

p) Vessel Hnd Coollng Spray Nozzle: ((

JJ Therefore, the structural integrity of thr vessel head cooling spray nozzle Isacceptable for BPU.

cx) Jet Pomp instrument Penstratin ;Seal:]:

3] Therefore, the structural integrity of the jet pump instrument penetration seal is acceptable forEPU.

r) Core Differentlal Pressure and Liquid Control Mne ((

3] Therefore, tbe structIral integrity of thc differentiaI pressure and standby liquid control line is accptablo for BPU.

s) Control Rod Drive: [

)) Thefo, tho structaml intogity of the control rod drivo is acceptable for EPU.

3-8

NEDO-33047 . Relsoln 0 3.3,5 Fllow lduced Vibraflon The core flow dependent RPY irtemals (in-core guide tube aad control rod guide tube componenta) are acceptable for EPU operation beouse the maximum oom flow does not change.

EPU operation increases the steam production in the oor, resulting in an increase in the core pressure drop. Ther is only a,slight increase (0.1%) in maximum drive flow atEP!) oondifions.

The incrsa in power may increase the level of reactor inlater. vibration, Anulyses were performod to evaluateIfI effects of FIV on tih reactor inte l at EPU conditions. This evaluation used a bounding reactor power of 3952 MVt and 105% of rated core. flow. This assessment was based on vibration data obtained during startup testing of a prototype plant (Browns F=y Unit 1). For components requiring en evaluation but not instrumented in the prototype plant, vibration data acquired during the startup testing from sPiilar plants or acquired ouide the RPV is iwed. The expected vibration levels for EPU wer estiumated by odrapolating the vibration data recorded in the prototypo plant or similar plants and on GE BNWR operating expienen. These expected vibration levels ware then compared with the established vibration acceptance limits. The following components were evaluated:

a) Shroud b) Shroud head and moistur separatr c) Jet pump d) FW sparger e) In-core guide tubee f) Control rod guide tubes g) Steam drye h) Jet punp Sensing lines The results of the vibration evaluation show that continuous operation at a reactor power of 3952 MWt and 105% of rated core flow does not result in any detrimental effects on tho safety-related ractor internal components CE 3.9

.r. , - - . . , *. * - .

NEDO-33047 -Revislon 0 During fIU operLtion, the coponenta in the upper zone of the reactor, such as the moisture separators and dryer, am mowtly affected by the increased stem flow. Components in the core region and components such as the core spray line are primarily affected by the core flaw.

Components in the annulus region such sa the jet pump ar primarily affected by the recirculation pump drive flow and core flow. For the BPU cmnditions, there is no change in the miaximum licensed cor flow in comparison to the CLTP condition, resulting in negliible changes in PIY on the components in the wnnulus and core regions. Only the moisture separator and dryer are significantly affected by EPU conditions. The steam dryer and moismure separators are not safety-elated coponients. However, the moisture separator loads act on the shroud through the shroud head. Becaud the shroud is a safcty-related component, the scparatort6hroud atruntre ws tested at various power conditions up to rated power during startup. The sepsator/sbroud struciure was evaluated from the test datam Recent uprate experience indicates that FIV at EPU conditions may lead to high cycle fgue failure of some dryer components. A detailed evaluation will be peformed to examine dryer components susceptible to failure at EP2 conditions. The results of the quuttitative evaluation will be used to identify any additional modifications needed to maintain steam dryrtsanctural integrity at EPU) conditions. If any steam dyer componnts requiring modification are identified, theso modifications will be implenmeted prior to operation at the 13FU conditions.

The calculations for BPU conditions indicate that vibrations of all safety-related reactor internal components arm within the GE acceptnce criteria. Tbh analysis is consvative for the following reasons:

- The GE criteria of 10,000 psi peak stres intensity isles. than the ASNM Code criteria of I3,600 psi;

• The modem an absolute summed-, and

• The maximum vibration aplitude in each mode is used in the absolute sum process, whereas in rality the peak vibration anplitudes are unlikly to occur at the same time.

In addtl to the above components a supplemnmtal evaluation was perfonned for addition!a components per the requirment of NRC Regulatory CGuid. 1.20 evaluated for PIV. The following components were evaluated and found to be accetable for NIV effects at E2PW conditions: guide rods, top head inxtnacot nozzle, head spray nozz!c, top head vent nozzle, CS sparger, CS piping. tel assermbly, shroud head bolts, MSL nozzle, water level instrument nozzle, and top guide, Based on the above, it in concluded tat Fly effects are expected to rerain withn wac table limits at EPU conditions, 3-10

NEDO-33047 -Revidoa 0 33.6 Steam eparator andDryer Performance The prfonance of th etcrun upftrs and dryer has been evaluated to ensure that the quality of the steam levin&th reactor pressure vassal remaidn acceptable at EN) conditions. EPU increases the saturated stem generated in the reactor core At constant core flow, this results in an increase in the separator inlet qualty and dryer facc velocity and a decrea in the water level inrside the dryer skirt. These factors, in addition to the coro radial power distribution affect the rteam separator-dryor perfonuace. Steam sepaatordryw performance was evaluaxtd la PEU equilibrium cycle limitihg conditions of high radial power peaking and the applicable core flow range shown on the power-flow map (Figure 2-1). The predicted steam moisture content was found cceptable.

3.4 REACrOR RECIRCULATION SYSTEM II The EPU power condition is accomplised by operating along extensions ofcurrent rod lines on the powenfflow nap (Figu 2-1) with no increase in he maximm oor: flow at E]PU RIP. The core reload analyses are performed th the mrost conservatie allowable corm flow. The evaluation of the reactor recirulation system performance at th4 BFO RTP wilt ens that adequate cor flow can be~ manaine&

SLO is unchanged by BPU.

The systen piping has bean reviewed for operation at the uprated conditions and found to rwet its deslign rvquiren (see Section 3.5). System componeits (e.g., pups and valves) will be evaluated .+/-EPU conditios to ensure that safety and design Objectiven are met 3.5 REACITOR COOL4NT PR1ESSURE BOUNDARY PIPZNG Tho effects of the BPU have been evaluated for the Recirculaion, MS (inside containme MS Drains, RCIC, HPC1, FW (inside containment), RWCU, CS, SLC, RHR, RPV Head Vent line, MSRVDL and CRD)piping systems using the present codeds) of record. The effiects of pressure, flow, vibratio= end thenral expanion displacernents (where apliloable) were evaluated. The evaluations of the above piping systems are either summarized in the following subsections or in Section 3.11. The original Codes of record (as referenced in the appropriat calculatIons), Code allowabIe and analytical techniques ware used and no new assumpfions were introduced.

3-11

- . t . ... .. Pe s ..

NEDDO33047 - Revision 0 An Lterate pipg evaluason process wus used for dhe MS piping evaluation for the Browns Fery EPU instead of the gaer promen described in Appendix K of BLTRI. A description of th Brown Ferry pip evaluaion process is incded in Seton 35.2.

Flow auceleratsd corrosion for all potantislly affected piping systems is addressed in 3.11.3.

3.541 Recirculatfon Sytem Eviulution The Rcifculation system was evaluated for compliance with the B31.l Power Piping Code stream criteria for pipe and pipo componanIs, and for the effects of vibration and thermal expansion displacements on the piping snubbers, hangers, strs and pipe whip restraints. Piping Interfaces with RPV nozzles, penetrions, pumps and valves woro also evaluated 3..1.1 Pipe Streses The effects of power uprate have bcen evaluated for te recirvulation loDp piping using the preset code of record: B31.1.0 Power Piping Code, 1967 Editon. Tbe piping was evaluatod for complimce with the B31.1 Code stress criteria and for the offects of thermal expansion displecents on the piping snubbers, hangem, and struts. Piping intorfactes with RPV nozzles, penetraions, pumps and valve, were also evaluated.

A review of the changes in pressuro, temperature, and flow associated with EPU indicate tha pipig load changes do notresult in load limits being exceeded for the recirculation piping or for RPV nozzles. The original design analyses have sff t differences (xces design mrgins) betwe calculated stresses and B31.1 Code limits to justt operation at the E;PU operating flow, pressure, and temperature.

The design adequacy evaluation results show tat die requirmonts of B31.1 Power Piping Code requirements are satisfied for the recirculation piping systems. Therefore, the S13U does not have an advers effect on the Recirculation piping design. No new postulated pipe break locations were identified.

3.5.12 PIpe Supporta Tho Recirculation systen was evaluaed for he cficts of vibaton and thwrmal expansion displacements on te pping snubbers, hangrs. end struts. A rovievw of the changes in temperature and pressur associated with the EPU indicates thatpiping load changes do not result in any ruppet load limitben exceeded.

3.5.1.3 flowAcelerstedConroslon Tho Recirculation system piping and cornponms are mnde of stainless steel and are not subject FAC depadation.

3.5.2 Main Steam and Associated Piping Systw Evaluation (Inslde contaiment)

The MS png system and associated branch p4ping (inside containment) was evaluated for onpliancewithc theUSAS-B31.1.1O, l967 Code strescriteri. a 3-12

NRDO-33047 - Reylston 0 The MS system flow will increase by approximately 20% for EPU. An result of the increases in flow, the TSV closure forces will increase sglificantly. Due to these increases in transient foces, Malyses f the TSV clDsure transient were perfornmd for the MS piping. The TSV fluid transient loads were generated utilizng the bounding closing time for th1eTSV.

The MS piping and pipe supports were evaluated for the TSV fluid transient loads in combination uith pressure, deadweight, avd thermal loads. Because a seismiC event may cause a unit trip and a ISY closure, the IiV transient loads were also considered concurrent with applicable soisrmic loads. Due to th time rlationships bctween the loads resulting frDm TSV, MSRV discharge, and pipe break eovnts (3.;., LOCA); no combination of these loads is requircd.

The tvaluation of the supporting stcure is being rrvewed. Wher sqqxrts frm differet main steam lines or differcnt systems load the &s=e member of th drywell steel, the sceisic and TsiV loads of these diffrrmt lines will be combined by the SRSS method, on an s needed basis.

Combination of tess loads by ths SRSS method is acceptable because the scismic rmeponse of diffrrent lines and tho fluid transient forces for different lies are out-of-phase, wt& peak leads occurring at difrent die.

Tho branch piping connected to the MS hoads (MSRVDL, RClC, HFCI, RPV Vero, and MSIV Drab) was evaluated to detmine the effec of the incrased MS flow on the lin.' This evaluatin conctuded that the branch lines arc accptablc for the increased MS systm flows following EPU.

As with to MS piping, the pressures and tempeatres for these brunch lines do not change a a.

result oVEPU.

Any modifications required to rmitigate the effects of increased tragient loads vill be completed prior to BPU implemantaicm.

3.5.2.1 Pipe Stresses Analyses evaluating the increased tbino stop valve closure transien loadin due to increases in MS flow indicate tat piping stresses remain within the oods allowables for the MS System. The oiginal design analyses have sufficient design margin betwee calculated stsses and the USAS-B3 1.1.0 code allowable limits ojustify operation at BPU conditions Similrly, the branch ppelines QSRVDL, RCIC, HPCL RPV aet, and MSIV Drain) connected to the MS bes were evaluated to determie the effect of the increased MS flow on the lines. Tlhis jasluaton concludod that pipe strsses will remain wthin e code allowables for the MS branch lines.

33+/-2 Pipe Supports The pipe supporte for the MS piping systm havc been evaluated for increased loading associated with the limitn transiout at EPU conditions for adequate design margin to accommodats the 3-13

NED043047 a RevIslon 0

!nesed support loads. Ay pipe suport modificaion, demed ncssary due to EPU increased ramniont loads wiL bcomnpleepdriorto EPUimpleoentaion.

The supporting infruore for 1he MS pipin syrn is curently b g evaluated for increaed loading associated with the limiting transient at EPU canditios. Any suppodiog sfructwe modifications decmed nvceesry due to BPU incresed trnsient loads will be completed prior to BU bnpletatimn.

3.5.23 Flow Accelerated Corromlon PAC for all potentially affected piping systems is addressed in Section 3.11.3.

3.5.3 FeedWaterEvaluatIon ThPoW systcm(inside cotinet) was oaluated for complianocwithfteUSAS-B331.)l-1967or equivalent Code stress criteria, andfor h efbcts of vibration and therral expansion displacements cn the piping snubbers, bangers and struts. Piping intefas with RP nonles, penetrations, flangcs end vaIves wer ao evaluated The results of tis evaluaton are prvided in Table 3-7a.

3.5.3.1 Pipe Strmmes A rcevw of the smral inteas in preisure, temper and flow asociated with EPU indicates that piping load changes do not result in load limits being exceded for tha FW piping system or for RPV nozzlew. The original design analyses have suficient design margin between calculated strees end USAS-H31.L.0-1967 Code allowable limits tojustify operation ut EPU conditions.

The design adequzcy evlution show, that th rquirements of USAS-B31,1,O-1967 Code requiremets rerain satisfied. Therefore, EPU does not have an advenr efIct on VthFW piping design. No nv postulated pipe break locations were identified.

3.5,3.2 Pipe Supports The FW systam ws evaluated for the effets of vibration and Othmal expansion displaments on the piping snubbes, hangers, and stmts. A review of the lacuses in tIezpue and EW flow associated with EPU indicates that piping toad changes do not result in ay loud limit being

exceeded, 3.5.3.3 Flow-Accelerated Corrosion Flow-acclrated corrosion for a11 potntally affected pipi systes is addressed in Section 3.11.3, 3.5.4 Other RCPB Piplg Evaluation hi section addresses the adequacy bf the oth= RCPB pipin3 dsigs,for operation at the EPU conditions, Thnmina opcratingpsureand tempenntrs oftoreaaorar notchangedbyEPU.

Aside fro MS and FW, no other system oncted to dte RCPB wcperienwc a,significant increased flow rate at EPU conditins. Only minor change; to fluid conditions are experinced by thes yhtuns due to higher sta flow fom te reactr and the subsequent chane in fhid conditions within te reactor. Additionally, dynmiopiping loads for MSRV EPU conditions marbounded by 3-14

NRD-33047- Revldon 0 ts used in the ext nalys. TheBPU effcb have been evaluated for the RCP portion ef the RPV head vet line, SC, CS, RPV bottom dmin, MSRV discharge piping and RWCU piping, u required.

3.5.4.1 Pipe Streues Thee systems were evaluated for compliance with the USAS B3 I. or ASMB Code strss critria (as pplicable). Because none of these piping mystens eVeriec any ignifai chang in operating conditions, they are all ptable as curently designed 345.A2 Pipe Supors The systems listed above were evaluated for die effcts of vibration and thermal expansion displacements on the piping snubbers, hangs and struts. A review of the changes in pressure, temperature and flow associated with EPU indicates that piping load change. do not result i any load limit being exceeded.

3.4A3 OtherRCPB Piping flow-Accelerated Corroulon Flow-acceerated corrosion for all potentially affected piping systems is addressed in Section 3.1I..3 3.5.5 PipingowInduced Vibration Key applicable structures include the MS system piping and suspension, the PW system piping and suspension, and the RRS system piping and suspension. In Bddition, branch lines attached to do MS sysfemr piping ore considered.

RRS drive flow is not significantly increased (c 5) during EPU operation. ((

The MS and FlW piping have increasd flow rates and flow velocities in orde to accommodate EPU. As Bresult, the MS and FW piping experience inrcrased vibration levels, approximately proportional to the square of the flow velocities. The ASME Code and nuclear regulatory guidelines require somne. vibration test data be taken and evaluaLed for these high-encrgy piping systems during initial operadon at BPU condition.. Vibration dim fbr the MS and FW piping inside containmet will be acquired using remote sensors, such as displacement probes, velocity sesors, and accelerometrs. A piplig vibration stastp Ut program will be performed and the results will be reviewed for acceptability. The FI testing will be performed during RPU power ascension.

The sbftyrelstd temoweJls and probes in the. MS and FW piping syst=s were evaluated and found to be adequate for the fncreued MS and 7W flow as a result ofEPU, 3-IS

NEWO-33047 -Revision O 3.6 MAiN STEAM LINE FOW REgIRICroRS The irea in steam flow iato bas co uignificant effec on flow eatrictor eosion, n is no effct on th stuctural integrity of the MS flow element (restrictr) due to the incrsed diffretial presure because t. rtstrictors wercdeaigmd and analyzed for the choice flow condition, Folloing a postulated stea lin. break outside contaimet, the fluid flow in the broken steam line increases until the MSL. flow restrictor Inita the fd flow. Becaus the madm operating dome pressur does not cnge, the rouiting brook flow rat is uncmanged from tho curret analy6is and te operational stessce are not ffrctod, Therefor;, the MSL flow restrictors are not significantly affectodby EPU.

3.7 MAINSTEAM ISOLATION VALVES The MIVs are part of the RCPD, and perform to safety finction of steam lin isolation during certin ulmnonnal events. Thz M9LVs nmst be able to close within a specified tirm range at all design and operating conditions. They r designed to saesfyileake limits set fiorh in the plant TS.

The OIVu have been [( )) evaluated, as discussed in Section 4.7 of ELTR2. Thi evaluation cosve both the effects of the changes to the strtr capability of te MSW to moot

.pressue boundary requiremcts, and the potential effecs of BPU-rlated changes to te safety funotions of the MSIVs. ((

)) The MSIVs will be modifiod to accommodate the higher valve stem forces caused by the increased steam flow ruts. Therefore, ((

J] and the MSIVs are acceptable fbr EPU operation.

3.8 REACTOR CORE ISOLATION COOLING The RCIC system eaation scope is provided in Section 5.6.7 of ELTRI.

The RCIC systen is required to maintin siont water invtoy In the reaotor to pemdt adequato cor cooling following a reactor vessel isolation event accompanied by loss of flow :fom the FW systun. The system design injection rate mst be sufcient for corzFUauoo with the syst limiting critria to aintin the reactor waW level above TAF at the EPU conditions. Thi RCIC system is designed to pump water into the rectr vessel ove a wvide range of opering pressures. As descrbed in Section 9,1.3, this event is addressed on a plant specife basi. Tho resls of te Browns Ferry plant specific evaluation indicabt adequate watu level margin above TAP at the EPU conditions. Thus, thb RCIC ine rate is adequate to med ths design basis event.

An. operational require t is that the RCIC syste can restore the reaor wter lvel while avoiding ADS timer initiation and MSIV closure activation functions mocissocd wth the loow-i*-

low reactor water level seqoin+/- (Level 1). This rcquement sisnteod to avoid unnoeauy 3-16

NEDO-33047 - Ervlan 0 initiatios of these safety sytemn. Th results of Ike Browns Perry plant specific evaluation indicates tat the RCIC ystom is capable of mainbtining the water level outside the shroud above nom l Level I setpoint &flhting LOPW emit at the BPU conditiwn;. Thus, the RCIC hijection rate is adequateto meet tberequhrines foriaventrty wakeup (set Section9.l).

For the BVU, ther is no chag to the nonnal reactor operating pressure and the MSSR setpolnt remain the some. There is no change to the rmim specfied reator pressur for RCIC system

]) because there are no physical changes to to punp suction configuration, and no changes to the system flow rate or rnimum atmospheric pressur in the shppression chamber or CST. EPU does not affect the capability to transfer the RCIC pump sumdon on high suppresicc pool level or low CST lewl from its normal alignment the CS]T, to the suipression pool, and does not change the eisting requirneois for the nnsfer. For AIYS (Section 93.1) andAppendixRSection 6.7.1). operation oftheRaClrystematsuppression pool temporats greate than the opratiomal linil may be accmplished by using the dedicated CST volum as the soavo of vim . Thrfore, the specified opeational teoperatum limit for the process water docs not change wit the EPU. ((

] The effect of EPU on the operation of hec RCIC systn during Station Blackout evenas is discussed in Section 9.3.2 The eacorrystemresponseto aloss ofFWtransiontwith RCICisdis6cused in Seetion9.1.3.

For Browns Ferry, a portion of the CST volume (135,00 gallons) is reserved for RflC operation by IhB usae of a BtbakIPs in the tank, The increase iC ractor decay heat duO to EPU reduces the amount of time that RCIC can mainain reactor ves;el level in hot shutdown conditions utilizing this reserve volume from greater hn 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> to a little less than 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

This is not a safety-related funatiDn and procedures ire in plaIe to direct the establishment of additional sources of water if the CST level approwhes the top of the standpipe. Additionally, (haI UFSAR provides a discussion of suppression pool tempceratre following reactor vessel isolation with RCIC operation, Operation of RCIC during this time would not be affccted by EPU conditions, however, the energy added to the suppression pool from the MSRV discharge would increase due to the inaeased decay heat associated with BPU.

] Therefore, the RCIC system is acceptable for EPJ.

3.9 RESIDUAL HEAT REMOVALF SEM The RER system evaluation process is descibed in Section 5.6.4 of BELTR1. Tho following results for thc RHR system evaluation [t JJ 3-17

NEDO-3047 - Revlson 0 Th RER system is designed to restoru md maintain the eator coolant inventory following a LOCA and renove rectfor deay beat following reactor shutdown fox normal, flnsien and accident condifis. lbo EU efU t on the HR system is a result of the highoer decay heat in the cors corresponding to tha uprated power d fth increaded amoint of reactor heat discharged into the containment during at LOCA. Thc RHR system is designed to cprzt in the LPCI mode, SDC mde, SPC mode, CSC mode, Supplemental Spent Fuel Pool Cooliog and Standby CoolinglCrosutis. mods.

The LPCI mode, as it rlates to the LOCA response, isdiscussed in Section 4.2.2.

The SF0 mode is manually initiated following isolation transint or a postulaed LOCA to maintain the containment pressure and suppression pool temperature within design limits. The CSC mode reduces drywell pressure, drywell tnpevatre, and suppression chamnber pressure fbllowing an accident. The adequacy of those operating modes is demonstrated by the containent analysis (Section 4.1).

The higher suppression pool temperature and conztainment pressuxt during a postulated LOCA (Section 411) do not affect hadware capabilities of RUR equipment to perform the LPC1, SFC, end CEC functions.

The Supplemental Spent Fuel Pool Cooling mode, usn existing RHR heat removal cetaokiy, provides supplemental fuel pool cooling capability in the event that the fue pool heat load exceeds the heat rernoval capability of the FPCC syflem, The adequacy of fuol pool cooling, inclading use of the Supplemental Spent Fuel Pool Cooling mode, isaddressed in Section 6.3.

3.9.1 Shutdown CooUng Mode

[1?

3.9.2 Suppresmlon PoolCoollng Mode TIhe fmctional design basis ea stated in the UPSAR for the SPC Mode during nonnal plant operation is to contol the inia pool tempertaru below the TS limit to so that the pool temperature immediately after a blowdown does not exceed the condensadon limit in the evnt of a deig bas LOCA, and o esur the [lng-tam pool t penturr does not caceed tie tons. athed iing mnalyis limit The EPU maximum suppreasson pool temperature (Section 4.1.1) ik utilized as the 3 I8

t - i NEDO-3047 - Reyielon n tons attached piping onlysis temperature limit in (he tors attaed piping analysis therefore, this objective is met for BPU.

The increase in decay heat due to EPU inceses the hut input to the suppression pool resultig in slightly higher containmont temperature and pressure during the initial stages of a LOCA.

The EPU effeot on the containment (drywell and torus) tpemperture, pressure, and condensation limit after a design basis LOCA is described in the containment analysis (SectIon 4.1.1).

As shown in Section 4.2.5, there is adequate NPSH margin during the RHK pump operation under the poat-LOCA operating conditions.

3.93 Contalbment Spray Cooltng lode Tho CSC mode provides water from the suppression pool to spray headers in the drywell and suppression cIamber to reduce conuinment pressue nd temperature during post-accident conditions Following EPU, increases in the post-LOCA contment spray temperature correspond to the increase in suppression pool temperature. The rate of increase has a negligible effect on the calculated values of dMwell pressure, drywll temperane, and suppression chamnbor pressure since these parametors reach their hghest values prior to actuation of the containment tpray as shown in Section 4.1.1.2 ad 4.1.1.3.

The CSC mode is used to reduce containment pressure following a LOCA, which cart affect lbs available NPSH. The adequacy of NPSH margin during the RHR pump operation under the post-LOCA operating conditions is discussed In Section 4.2.5.

3.9.4 SupplementalSpantFuelPool Coolng The RHR Suppleneta Spent Fuel Pool Cooling Mode, using ih existing R heet remova ]

capacity, provide; suppleimmtl Ml pool cooling mn the event that the fuel pool hea load exceeds the beat removal capability of the FPCC system due to off loading of the entire core. This mode operates along wit the FPCC system to maintain the Fuel Pool tempate within acceptable limits during a reactor cold shutdown. The increased ubort-term fel pool heat load due to EPtJ does not exoeed th combined heat rmal capacitiesof this mode and PPCC system. (See Section 6.3) 3.9.5 Steam Condensang Made Steam Condensing mode ofRtHR is not inst at Browu Ferry.

3.9.6 StandbyCoollng/Crosidea Standby CoolingfCrssties utilizes the stanby coolant supply connection and the RHR crossties to provide additional long-term redundancy to the emergency core and containment Cooling systems. This futotion is not affected by EPU because the performance requiremzents for the emergency core and containment cooling systes are not chaged.

3-19

NEDO-33047 - Rebion 0 3.10 REACTOR WATER CLEANUWrSYEM RWCU syteM opeaion at ft EPU R7T level slightly deura tha terperature within the RWCJ systm. This system is designed to :mcve solid and dissolved impuritis from recirculated reactor coolant, thereby rnducing the concenfration of radioactive and corrosive species in the reactor oolant The sem is capabl ofperforming this fumionat the PUR leveL Based en operating eperience, the PW iron input to he. ctor increase s a zesilt of the inreased FY flow. This input Ire s the calculated reactor water iron concentration from 20.4 ppb to 23.7 ppb. However, tis change is oonsidered insignificant, and does not fect RWCU.

Thl effects of EPU oa the RWCU system f&ncdonal capability have boen reviewed, and the system can perform adequately during BPU with the original RWCU system flow. This RWVCU system flow results in a slight increase in the calculated reactor wafler conductivity (from 0.10 pS/cm to O.l1 pS/cm) bccaus ao the incrase in FW flow. The prceent reactor water conductivity limits are unchanged for EPU and the actual conductity remain within thesc limits.

The system piping and components have been reviowed far operation at the rmuted conditions (pressure end temperature) and found to meed its safety and desig objcctives, including antinning structural integrity during nomrml, upset, emrtgency, and faulted conditions. In the event of a HEL in the system piping, appropriate isolation shall be achieved (see Section 4.1.3). fr tn Sections 3.5 and 3.1 fbr evaluatio of pipe and support adequac, and Secdon l0o1 for the HELB evaluation, 3.11 BALANCE-OF-TLAN'PIPING EVALUATION The DOP piping systems evaluation consists of a number of piping subsystens that move fluid through systems outside th RCPIB piping.

For soene BOP piping sys , the flow, proe , temperature, and mechatical loads do not inccase. ((

Large bore and small bore ASME Clue 1,2 and 3 equivalent piping and supports not addressed in Section 3.5 were evaluated for accetability Lt EPU conditions. The evaluation of the BOP.

piping and supports %ts performed in a manner similar to the evaluation of RCPB piping systems and supports (Section 3.5), using applicable ASME Section III, Subsection, NCZ'ND or B31.1 Power Piping Code equations. The origins! Codes of record (as referenced In the appropriate calculations), Code ullowableg, and analytical techniques were used and no new assumptions were introduced.

3.20

NEDa-33047

• ReYision 0 The LOCA hydrodync loads, including tie pool Aswll loads vent rust loads, CO loads and chugging loads were riginlly defined and evaluated for Browns Fersy. The atructu attached to the torus shell, such as piping system, vent penetarIons, and valves are based on these LOCA hydrodynaznc loads, For BPU conditions, thb LOCA torus shell resporme loads were re-evaluated using momr realislic RPV depresszatioon to within the capability of the available number of MSRVa. These loads were found to be acceptable and thero ae no resulting effects on the tonn shell attched structure.

The effects of the BPU conditions have been evauated for the following piping systeu:

• MS - Outside Containment (specifically addressed in Section 3.11.4) a Extraction Stewa, Heater Vents and Drains 0 MW and Condensate

• RWCU - Outside Containment

• ERR- Outside Containment

• RPR Seivice Water - Outsi Containmeat

.* CS - Outside Containnent-Pump Suction /Punp Discharge a HPCI - Outsido Containmnt

• RCIC- Outsids Containment

• SLCS - Outside Containment

• CRD

• EBCW

• RBCCW

• SPC/ADHR

• ROW/Stator Cooling Water

• SGT

• Off GM

• Torus Attached Piping including ECCS Suction Strainers 3.11.1 Pipe Sfrees Opertion at the EPU conditions increaes stream03 piping and piping syun components thto slighily higher operating temperatue and flow rates internal to the pipes. For those ytems with analysis, the maxinwm stres levels and f6igue analysis results wer reviewed based on specific increases in tmperature presame and flow rate. (see Tables 3-7EL and 3.7b). Por those systems that do not require a detailed analysis, pipe routing and flexibility ws evaluated and datermined to be acceptable. Thse piping systs hve been evaluated, ung th prcess dfined in Appdix K of ELTRI and foud to meet to appropriate code criteria for te EPU conditions, based on the design margins between actal stremas and coda limits in the original design. All piping is below the code allowables of the peeset code of record: USAS B31.1.0 - 1967 Power Piping Code and ASMlE Boiler and Pressuru Vessel Code - Section I1, Division I through the summer 1977 Addenda for torus ettached piping. No new postulated pipebreak locations were idenefled.

3.21

w r e h. . e w NEDO.33047 - Redlilon a 3.11.2 Pipe Supports Operation at th EPU conditions allghtly increases te pipe support loadings due to increases in the tonepete of theo fcted piping systm (se Tables 3-7a, 3.7b, awd 3-7c).

The pipe supports of the systes affected by BPU loading increases (RER, CS, and Tom attached piping systems) were reviewed to deteanine if ther in sufficent margk to code acceptane teisa to aootommodato the incased loeings T rview shows that in most caes thr is adequate design margin between the orignal design sresse and code limits of the suports to awmodate the load incease. A very limited number of pipe wupports will recuiie a more detiled ealaton to show that the support structure is cceptable for the increased loading. Should any dotaileld evaluation show tht the code limits cannot be inet modifications will be made pror to BPU nwlernmrtation.

3.11.3 Flow Aceelerated Corrosion The integrity of high energy piping systems is assured by prop& design in rdance with the applicable codes and standards. A consideration in. assuring proper design and maintaining system operation within the design is the allowable piping thickness values. Piping thickness values of carbon steel components can be affected by FAC. Browns Ferry has an established program for monitorn pipe wall thinning in single phase and two-phase carbon steel piping, Process variables that infuvence PAC at Drowsns Ponry ar moisture content, water chemisty, femeratvre, oxygen, flow path gometry and velocity, and material composition.

EPU operation results in some changes to pammetars affecting FAC in those systems associad wit the turbin- cycle (e.g., condensate, FW, MS, extaction ston). The evaluation of and inspection for FAC in BOP systems is addressed by compliance with NRC Geeric Letter 39-8, "Erosion/Corosion in Piping." The Browns Fery PAC program currently monitors the tffected system (see Section t0.7). Contined monitoring of the systems provides a. high level of confidence in the integrity of potentially susceptible piping systems. Appropriate changes to piping inupection fequency will be implemented to ensure adequate margin is maintained for those systems where process conditions change. This includes adjustments to predict material logs rates to projeot the need for mainrenanoreplscst prior to reaching minimum wall thickness requirerents. The program provides assurarce that the EPU does not adversoly effect piping systems potentially susceptible to pipe wall thirming due to FAC.

3.11.4 Main Steamnad Asociated Piping (Outslde Containment)

To MS piping systm (outside confaimnnent) was evaluated for compliance with Browns Ferry criteria, Included in the evaluation w the effects of EPU on pipin stresses, piping supports and the associated building struature, turbine nozzles, and valves.

The MS piping pressures and tenipcratures outside containrment ae not affected by kPU; there was no ffect on the analyses for thcse parameters. The incroase in MS flow results in increased forces from tho htubie ntop valve closure transient. The turbine stop valve closure loads bound thz MSIV vave loads because ths MSIV cloture time is significantly longer than the stop valve closure tine, The MS analysis results ae proided in Tablo 3-7a 3-22

NEZDO.3043f Revlska 5 3.11A.1 PipeStreses A eview of ft increse in flv associaed with EPU indicates tiat piping load changes do not result in load liits being exceeded for te MS pping sytm outd containmnt he oiginal deigm ha ufficfent deafSn margin to justify operation at the EU conditions. The press and tecprar of the MS piping is imhanged for EPU and the pipe strsses am, ueptblo.

3.11.4.2 Pipe Supports The pipe support (pmimarily srng type supports) and trbine nozzles for It lAS piping system outside containm t were evaluated for the increased loading and movements associatd wit the turbine stop vali clos turnsiet at EPU conditiom. The evsluaions demcstate that the supports and birinae nonhs have adeqate design margin to aicommodute to hreased loads end movements resulting from EPU.

3.12 R.EFERENaS

1. 1GE Nuclear Energy, "Generic Ouidolines for Genrual Elcctrio Boiling Water Reactor Extcndod Power Uprate," (ELTRI), Licensing Topical Reports NEDC-32424P-A, CAs III (Proprietawy), February 1999, and NEDO-32424, Class I (Nonaproprietay), April 1995.

2, 08 Nuclear Energy, "'Generie Evaluations of General Electrc Boiling Water Reactor Extended Power Uprate," (ELTR2), Licensing Topical Reports NEDC-32523P-A, Clsus 111 (Prpritay), Februay 2000; NEDC-32523P-A, Supplement I Volume 1, February 1999; and Supplernat I Volume 1, April 1999.

3. 0ENuclear Energy 'GE Methodology to RPV Fait Neufrn Flux Evuation," Licensing Topical Report NEDC-32983P, Class Di (Proprietary), August 2000 and NBDO-32983 -A, Class I (Nonllpripdetary), December 2001.

4. NRC, March 10, 2004, "Browns Forry Nuclear Plant Units 2 and 3 - Issuance of Amcndments Regarding Proesre-Tcmnpcnrxe L}imit Curves (TAC Nos. M0:0807 and MC080S)", Anendments Nos. 288 and 247.

3.23

NEDO0-33H7 - RcvIfloaD Table 3-1l BrDOnt F. Pry Uit2 Adjusted Refereane Temperatures 40 Year Life (34 EFPY):

naf nThAca. Lft mhiu L.5*Il M1dfYis&2MTfl mms- [.362

?" pPnhmAMTrbjwaa. LaBFli Wift T7m . 613 SCAU M.211 2A=m2 M1r4 MlmYeit2flncts"- S2.-I n prYl tnr thdlO- 1.CSX 4~IM1 0

..L1 _*? .34)y a VA 34 V" M MY CN:Jm ? uA ?CR AUTLC"T Ka %WI cr hl 1 India A SloE Urim SW ART IF _9C IF __ .SP

_T __-lr

• -U272ItC24 - 0.t6 052 ll1 .20 IWON 40 D 12 St f1 6.21t.1is Ct4r9.1 OA? A II 111I *n0 (.0 E s a 17 14 34 6i 6.33.1 7 C2440-2 0.1) 05t III t 1.1.18 n p 0 It 4 17 11 DtS 7.4 A D 9] -1 0.14 0 53 t o .1 i I.Co 41 a Ii m 1 e a I.127.26 C E 47it t I.E 0.12 222 *go I M OM 42 t 1J U el 72

• 1174D0 C2449.I 0.12 9.50 7t -ID 12D Sl D 25 31 6 5 bt IS V'o~

ZI 4 0.24 0 237 141 23.1 lA G4S do 13 0 01 I22 14o Vt vbn w O.D0.0> 12? -to i ID JoS2 D 1 so DP Ss Is R AW mbsklyad on BSAW.221+/-1t*&

g Sp itU h iL%$% nbnf uunn al sW nbb .

60 Year Life (52 EPPY):

lab Taih A . .12 5 e Or O I.D.fln w- 132.11 tMe St orb rtar rfn s'9. limt li m 1 Wdd T W 6 2 fft S2 Ifllnk CD.b mnw - S MW 1ll f wle 3TY IF uM TMo LIB-IL bmi SR mnYlnlnTjMeW- 31.I'1I Van caowjtr t mrxHArttr %Q Y WIT 5m'P t t l 5 2 '3PIlF M

%WI crF M 11oo A RM "I Mr flATZ&

.6 OIU 0.10 0.52 20 -to iamo of § n t Sf 3S 1 44U 7.11 026 9 -I 0.2 0.42 2317 .2o L 1 (* so a I4 3I2 34 14 4 4127 4t CC46:2 0.21 0.1 1 DS a ILdtfl 45 0 27 14 7S IV

£4274 A n I-1 0.14 oi l Ps .IDl 1 I S JI p 27 0 is '7' C-l:

IC467-7-25 0.25 0.5 212 -ID 1,04I1 i3 3 27 34 2 C 041740 C2au*6r 0.Ir tin t1n Ts .A14i t 2t 0 1 Y 4 71 a L ng odkaI 29 s' OJ A 21S7 I2 It 3.1 1tz +(v I 1 13 It Kt 0 11D M mia l" O. ESIII 0.02 0.AS 227 .44 hasH1h a P 2 54 Id 7*

'hIUThamhosyhbod cm EAW.U2lftl1 Fpl 4Jhn hata oumbflh anmb i u bh.

3-2.4

W-T . ..

v- .

• l -z . - . r--- -

NEDG.33047

• Reililon 0 Table 3-lb Brois. FePn Unit 3 Adjusted Retenee Temperature.

40 Year L. (34 EFPY):

?[Mt 11fak. 6,27 Ir, 343mlvtatLaI. L 3.f432 ?tft 3 OW7YPs&2D. - 14 294211 mhftf

'WE hdorw 4.23 btX )IWTtN6W.N - 1.01342 W t 34 Of IhmL3 Twsu HIS I Wote tbhid WT 34Wlr O -A r DWV 34 c r tm MT V EAlT= Wsl 5I3 o MAZyh*

.7 _Irm

. ALiW

.7 LfrjN

.7 AM

.7 Ah1 651-SA C1252-2 DtS 0.57 205 6LO 2. ! 45 O 17 2 5 19

.1 507 Olls-I D ) 0.l Fs 40 239-1 2 5 0 I7 54 1 5 1445-t2 032174 C0l4 066 2 10t4 -4 l 2 6 43 C 31 4 r 7 23 mummlkbrll 6-34. 2 C0 14t 0 ) 0.64Ito ot 4D 0 U.0-i* 1 0 17 3 7t . U 6-24532 02514F; 0ID0 0.44 6S Z-20 1 to3, 0 14 2F 53 U 6.144 31207.1 0.11 0.92 01 .20 t3 ll-li 17 0 12 3 t7 *1 wIS(r..

Lnhkwd fl BMW 024 6.37 t41 .l3 LOS'-I a 21 2 l a III

__ eCeV lswc& US rt7 .40 _ S-lI 52 0 25 SO II _

4 38W bmmhbybtmud on tw4szznazt 60 Year Lfe (52 EFPY):

fl r.wx e- 8.25 n- tl" tapk s 222 w~lek 1 Vzfr PSA t~b.ft-hse I 41W1 2-304 obr;m wcl sirypiALTshan-k 143411 nhen r ~Yt uI U'N i u¶ 1 63423 nftn f S* %94pmft MTflw - 2A33.6 zhwfl

_bhl IYT IlWr el @ r DY 52 fl SPy tUPO4M.r uFMl0RIcATfmo ICo ml4 0P 1~ud Flu n A tlha Mug~u na. AK?

pr %fr Ir r a LA ir1I461 .4 0224 R.33 0-2 too to IS 2I1 UO 4 IT 14 09 59 6-14f.7 r322134 01 a so .W 1WB21 64 a IT 14 la P 644I -It C 21.11 0.34 t.46 LOIS .4 IA&8Il 51 0 IT J4 id 14 6-243 032D2.1 DM 080 61 .I A411 Da7 n 4 It S a

6-4.44s 02114 9.230 03 do L SnU 84 a IT U n 47 m t4 0l jnrw . CA 1SI Iii Jz 10 l l a n 72 15 8 1? 141 23,1 Q1s 13 11 3: 42 31 15T Lnmtolk BltW 014 ciumbsnps DOWlS Om 8.4i 114 .a ij+Ii Qq c sr i1rs 7S

• W Abnahr ehmi e b&w.nstC23 I sa t a bnal vbmk a as usaue uI6.

3.25

av I ... .- -- - . - -: -- ---

v- ~ r * *- t- - 7 . , -

NEDO33047 . Revidon 0 Toble 3-2a Browns FeMy Unft 2 CUFa of lUmntin Components (1)

P + Q Stren (ka) CUP Current EPU Allowable Current EPU Allowable Component (ASME Code iAmit)

FWNO0Z1 47.2°' 49.2° 69.9 (3S.) 0.984 0.997(4) 1.0 (Blead raditus)_ _ _

Main Clonure Said CEross setion . 49.2 49.2 73.4 (3S) 0.762 0.762 1.0 Maximum pripherut 103.3 103.3 1 10.1 (3S _

SupportSkift I 15.9P 115.$0 80.1 (3SX) 0.904 0.904 1.0 Recircuaion OUtlet 75.5 75.1 80.1 (3S.,) 0.779 0.779 1.0 NozzlC .__.

Notes:

1. Only components wM usage f&cors greater du 0.5 areincluded in this table

2. E*, alternating stelsa in mcordance with ASME code, Section l1Subsection NB is shown.

3. TheInnl bending has been included. P + Q Stresses are acceptablo per CLTP elastio-plastio analysis, wh is valid for BPU conditions.

4. The conbined usage Uactor for system cycling + rapid cyclig Is0.9966 for normal duty Mad 0.997 forFFWTR.

3-26

NEDO-313047 -Revisiun Q Table 34b Brownis Ferry Unlt S CUFs oLfm tlng Components PQSimi (kd) CUP!

Component Current EPU Allowable Current EPU Allowable (ASAR Code Limit)

FWNozzle 47,2°) 49.2v 69.9 (3SO 0,984 0.997(4) 1.0

( lend reAdus) _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

L*mClosureStud Irns section 49.2 49.2 73.4 (3S.) 0.762 0.762 1.0 xirmperiphenl 103.3 103.3 110.1 (3S.)

S ot SkIdr lI5.90) 115.99 80.1 (3S, 0.904 0.904 1.0

• lation Outlet 75.5 75.1 80.1 (3S, 0.779 0.779 1.0 o71e Notes:

1. Only components with usage factos greater than 0.5 art included in tis table,

2. Sje. Liternating stresm in accordance wMi ASME code,Section III Subsection NB is shown.

3. Themal bending has been included. P + Q stresses are acceptabli per CUI elastic-plastic Lnalysis, whict is valid for EPU conditons.

4. The combined uimag factor for system cycling + rapid cycling is 0.9966 for aonial duty and 0.997 for FFWTR.

3-27

!I--- --- - - .. - - -:- -- - -

NEDO-33047- Redulon 0 Table 33 Browns Ferry RIPDs for Normal Condtons (pild)'

Parameter Q=

Coro Plate ad Guido Tube 2284 24.40 Shroud SupportRingand Lower Shroud 31.06 32.89 Upper Shroud 8.23 8.55 Shroud Head 8.42 9.43 ShroudHead to Water Level (rrevemible**) 108 12.24 Shroud Head to WaterLevel (Elavatin**) 1.07 0.94 Top Guide 0.61 0.61 StamDryer 0.33 0.42 Fuel channl IWall 11.67 13.31

' 105% core flow

• t*re velsib toss is the loss across the patof; the seevationloss or re.rsible head loss i teo los;s between the insiide shroud to the exit of the separators.

3-28

- i

• NWO-33047 - Rflyllca B Table 3-4 Broan Ferry RIPDs for Upuet Condtlona (psid)

Parmeter fLT EPU CoroPlateandGuide Tb . 25.24 26.80 ShroudSupportRing andLoez Shroud 33.46 35.V Upper Shroud 12,34 12.82 Shroud Head 12.63 14.14 Shroud Head to Waterlel (Lrevesible* 16.20 18.36 Shroud Head to 'WaterLevl (Blevation*#) 1.61 1.41 Top Guide 1.10 0.92 Steam DrS y 0.50 0.62 Fuel Channel Walt 14.57 16.21 IM5% Corm flow Irreversible loss is the logs across the separators; the elevation loss or revemible head logs is the loss between thd inside shroud to the exit of the separators.

3-29

NED 043047 - Revlidon 0 Table 345 Browns Frry RIPMs for Faulted Conditon, (POW Core Plate and Guide Tlbe 30 28.5 Shroud Suppott Ring and Lower Sroud 52 5l Upper Shroud 30 29 Shroud Head 30 295 ShroudHead toWatr Levetlrevemible") 32 32 Shroud HeadtoWaterLevewl (Elevon**) 2.1 1.4 Top Guide 2.8 1.1 SteamD*,u*** No Cange Fusl Channl Wall 14.6 15.5

• I05% Coreflow

" Irreversible loss is the loss across the separators; the olovation loss or rversible hoad loass is the loss botween the inside shroud to the exit of the separtors,

• • • These pressuro drops ir for an MSLE outside primry containmcot The steam dryer pressure drop is greatest for the high flow, low power condition (interlock point). The interlock conditon has not chunged with the EPU.

3-30

NEDG-33047- RzVluon 0 Table 3-6

• Browns Ferry Reactor Internal Components - Summary of Strewe Item Component Category! Strsintsod CLTP l EU Alowable Location Service Category value Value

._ Condition I Shroud NrmulfUpst Bbuydcd by ProEPU design basisLoads/Strse; Emerugenoy Yaulted 2 Shroud Support Desip SteCssp(si) 24,500 30,062 34,930 Operati I I 1 3 Shroud Support Faulted Bounded by piEP. de"lwi haWs Loada/Stresges 4 Core Plate Norma/Upset Buclding/Sldlsg l 25.2 26.8 28.0

. AP (psid) II I 5 Cor Plate Emergency) Bnded byPe Ud dsign bais >oads/strests

• 2 F_

_. .ted 6 Top (ulda NomUpset BoundedbyPrm-aU desigbasis Loads/tressea BEaergancy Faulted 7 CRD Housint Qualitative Au csmect (See Section 33 .fel_

8 Control Rod Guide NonnalJ~pt AP Bwkding 0.24 0.26 0.40 Tube . (Pc I I

! 9 Co31rolRod Guido Emergency) Boundad byPreEPO designbals Loada/Stressa ba Tu_

10 . Oriflced Fuel Support Fwalted Normal/ps Siren (pi) 12,413 lZS27 J 15,580 11 Orliced Puel Emergencyl Bounded by Pro-EPtI design basis Loads/Sitesse, Support Fauted 12 Fuel Channels Qualifiedper Propdetary Fuel Dloalgm Basis 13 Sam Dryer (Hood) Nranal/Uppet Siren ni l 4,054 l 5,027 l16,950 14 Stoam Drye Faulted Boundd by PreEgPUdsi basis Loads/Strewa 15 FWP Spa NOnnuJIpm Pm 4Pb + Q 70,800 70,910 76,500 Slottd Ring Therm. Bendiq 16 FW Sparer No=alUpset Pm +Pb (psi) 5,190 6,990 21,450 Heade PipetctIeI 3-31

NEDO-33U47 ERevison 0 Item Component Category/ Stre_W _ad CLTP EPU Allowable Locatlon Service Category Value Value Condition _

17 _WSpLr EmaerSency Pm+Pb(pl) 6,020 7,820 28,600 Header kiPe4 I I I I8 FWSparge Faulced Pm+PA(Psi) 33,690 35,490 42,900 Header P&iTee . I_

19 *JetPump (including N IaUpsot Bound dbyPre-EPU dee1gnbasIs Loada/1tesea riser bravo repair - Bmorgnfcy BFN 3) Faulted 20 Core Spray Une and Qualittive Asessmnt (S Sootion 33.4(1))

Spar (includes T-box and dowcono er Repsire

- EBFN 3) 21 Aocess HoCover NormlUpset Pm4 Pb (Pei) 6,756 7,093 349s0 22 Aocces Holo Cover Brngencyf Boided by Prem-ElU deilga bails LoadSerses Faulted 23 Shroud Head ard NormultUpset Pm +Pb (pi) 33,993 34,M9 34,950 SM Separator Assembly ( ) _ .

24 Shroud Head and Bwergency Pm + Pb tspi) 31,348 34,671 52,425 Seam Separator Assembly (SIB).

25 Shroud Head and Faulted Pm + Pb (psi) 41,432 41,758 69,900 Steam Separator Aesentbly (SH) .-

26 In-Cor Housing end Qualitatie Assessment (Se Section 2.3.4(o))

Ch. Tbeo 27 Vessl Head Cooling QxalltatvsAsesmont (See Sxtion 33.4(p))

Spray Nozzle 2B JetPump lttrnunent Qualtatve Amssesm t (See Seoton 3.3.4(q))

Peantrtioon Seal 29 Core Differeatial Qualitative Assessmet (8ee Section 33.4(r))

Pressre and Standy LUqgd Contol Line 30 CRD Qualitative Atcfngme t (Seo Secdon 33A(s))

3-32

NEDO-33D47- Revision U Tible3-7a Browns Ferry BOP Piping FW, Extraken Steam, FW Heatr Dratins and Vents, and Condensate maximum pipe itess inoreas:

Tempemrutre 5.5%

Sn Pressure0%

Fluid Tranisient 0%

baximum pipe support hading increase (due to thernal expansion loang): 5.5%

Tble 3-7b Browns Ferry DOP Piping CS and RHR (Ousideo Contanment)

Maximumpiponres increase:

Terwaeure Cexpansion 14%

Pr==ue 0°,;

Fluid Transients 0%

Mximm pipe support loading increase (due to thermal expansion loading): 14%

3.33

F.~. - : -- - -- -* ---

NEDO.33047 - Revlislon 0 Table 3-7c Browns FPry BOP Piping Main Stam sytem (Outilde Containment)

Maximum ppe stres5 at BPU:

Teneratur expansion No change PretSir NOu Fluid Trnsients Aceptable' Maximum pipe support loading FSU (due to thonnal expansion Ioading3: No change EPU (due to fluid transient loadingY! Acceptable' Notes:

1. Pcroattae increases for 1ho MS piping (outside containmwt) emnot provided because the turbine stop valve trmnsient was not previously anlyzed for Browns Ferry.

mheofbre, no comparison of the stresses or lands including turbine stop valve load cane can be made. However, the results of the evaluations show that the piping stresses and the supports meet the acceptance criteria for the EPU conditions and no modifications ar required for thelMS piping outside contairnent 3-34

-,C -

NEDO-33f47 - Revison 0.

" 1* l PA PA *4 P "u 11/

Lw.Infi omilUp 641 I'll +*Finelyt mlw rTh cm,* Timn mIP.

figure 3-1 Browna Ferry Rzponse to MSIV Caosure with V1ux Scram (102% EPU power, 105% core flow, and 1055 psig initial d~uneprassme) 3-35

.. v . . . . . a - *rwr .. e * - -

. . ...." . - ..-I. .. .. - .. . - - . ,

INEDO.&337 -Roylilon V lkhr-*-D kR~

-Ail &WmNwf 1XI INN 1I BUN 9:i 41.1 Ii

• 4 * * *  ? i Ii 5

• it I adL~s~~mup~

Ai 40Q Figure 3-2 Brownr Ferry Response to TurbIne Trip with Bypass Fifflurt and Flux Scram (I102% BPUpower, 105% cort flow, and 1055 psi& bNital domiepreaaun) 3-36

-+ o - ! w w w s x s z

- . -- . ~.--. -  :- - .., S- . .* I NEDO-33047 - ReQan U

4. ENGINEERED SAFETY FEATURES NURBG-0800, "Stadurd Revlew Plan for tho Rvw of Saf:ty Analysis Reports for Nuclear Power Plants," Section 6.1t, subsection I stu, "Engineered safety &eatre (1SF) 3xo provided in nuclear plants to rmtigute tho consequenoes of design basis or loss of-ooolant accidents." The Browns Ferry featres evaluated within this section ur designed to (directly) mitigte, the ccnscquoecm of postulated acoidets, ond thbs, are classified in ftc plant UPSAR a. engincred safety *ature.

4I CONTAINMENTSYSTEM PERFORMANCE This section addreses the efect of the EPU on various apsets of th. Brovns Pony containment syntem perforrmnoce.

The UFS&R provides the contaimnent responses to various postulated accidents that validate the design basis for the continment. Operation at the EPU RTP causes changes to sone of the condition. far the containment analyaes. For example, the short-term DBA LOCA containment response during the reactor blowdowvn is governed by the blowdown flow rate. This blowdawn flow rate is dependent on the reactor initial thennal-hytdratlic condition;, such as vessel domne preisure and the mass and energy of the vessel fluid inventory, which change slightly at the EPU RTP. Also, the long-term heatup of to sppression pool following a.LOCA or a trasient is governed by the ability of the RUR system to remove decay heat. Becaue the decay heat depends on the initial reactor power level, the long-ter containment response is affcted by MPU. The containment pressure Bad terperature responses have bean reanalyzed, as described in Sectio 4.1.I, to demnstrate the Browns Fery acceptability for operation atEPU RTP.

The analyses wcro performed in accordance with Regulatory Guide 1.49 and ELTRi (Reference I) using GE codes acd models (eerecs 2 through 5), Theo E methods have been reviewed and approved by the NRC sReferce. 6 atd 7). Confirmatory calculatioas Hvtb the SHEX code and The NRC-accepted HXIZ code show a difference of less than 10? in peak suppression pool temperatur betwe the two codes, Therefore, the use of the SHEX code for Browns Ferry complies with the NRC requirments for vse in the EPU anelyses preeentod in Reference. B.

The major difference eween the curret USAR and EPU containment analyses is senrice water teperaturo, which was increased from 92VP to 950w for all analyses.

The effect of BPU on the containment dynamic lodsa due to a LOCA or MSRV discharge baa also been evaluated as described in Section 4.1.2. Tiese loads were previously defined generically during the Mark I Containment LIMP as described in Refrenoe 9 and accepted by the NRC per Refreaces 6 and 7. Plant-specific dyoic loads irealso defined (Refece 10),

and were accepted by the NRC in Reference 1I. The evaluation of the LOCA conwtanent dynamic loads la based primarily on the results of the short-term analysis described in Secdon4.1.1.3. The MSRV discharge load evaluation is baed on no changes in the MSRV opening setpoints for EPU conrditimns.

4.1

- - ;- 171'

'..:. -7,111: r --. -.

NEDO-33047 -Reviion U 4.1.1 Contalnment Pressure and Temperture Response Short-tenn and long-term containment lyzes results are reported in the UPSAR. h short-ter analysis is directed primarily at determining thc drywell pressre responso during the initial blowdown of the tactr vcasbl invnory to the containment following a large break inside the d4ywell The long-term analysis is dircted primarily at the suppression pool tenperature; response, considering th dcay heat addition to the suppression pool. The celaht Qf BPU on the events yielding the limiting containment pretsure and temperture responses are provided below.

4.1,1.1 Long-Term Suppresinon Pool TemperatureResponue Short-term and long-term containmwt anaysis results are reported In the UISAR. The long-teoxn analysis is directed prinazily at the pool temperatre raeponse, concidering the decay heat addition to the pool.

(a) Bulk Pool T peratumre The long-term bulk pool tmpemurae response with EPU was evaluated for the DBA LOCA.

The analysis was performed at 102% of BPU RMiP, Table 4-I compares the calculated peak values for LOCA bulk pool temporarure. To current analyses have been perfored using the same RHR containment cooling capability used in the UFSAR Section 14.6.3.3.2.3 analysis (K-223 BTU/sec-0 F/X), but with a higher service water temperature (95'? versus 92?), The BPU analysis was performed using a realistic deay heat modbl (ANSIANSI 5.1 with 2a uncertainty),

similar to the current UFSAR analysis. Benchmark calculations vre made as requested by the NRC in Reference S. The HrownsFery calculated peak bulk suppression pool tepemtures are provided in Table 4-1 for both 102% of CLP and 102% of BPU RTP. This comparison shows that EPU results in an incroas of 7.7WF in peak bulk suppression peal ttmparautre, based on munrent methodology.

Based on the analysis and limit values shown in Table 4-I, VD* peak bulk pool temporature with EPU is acceptable from a structul design standpoint.

The containment response used for NPSH evaluaions is calculated using Browns Ferry specific Inputs to naximize suppression pool tempeuatreand mini containment pressure, similar to the DBA-LOCA analysis using the Sne methodology. The sppression pool Tenpure and oorrespoading wetwell pressure fbr the short-term and long-term NPSH containment analyse am used in the evaluation of tho available NFSH for the CS and the RHR pumps. The results of that evahaion re provided in Section 4.2.5.

(b) Local Pool Tesnperatire withMSRV Discharge The local pool teprtmr limit for MSRV discharge is specified in NURBG-0783, becu of concerus resulig fom unstable condensation observed at high pool temperes in plants without quendms. The MSRV discharge quenchers a Browns Ferry r dightly below the elevation of the ECCS suction line penetrnton. The peak local ueppression pool temperature at Browns Feny has been evaluated for EPU and meets the NURBG-0783 criteriL Therefore, the peak loca suppression pool teinempre at Browns Ferry is acceptable for BPU conditions.

4.2

NEDO-33047 - Retvion 0 However, it is neceassary to ensurt that Am inestio in the EGCS aucion line is not of concern duin steam MSRV discharge it high suppression pool tenmertur. because The top of the, EG uction stniners at Browns Ferry are located above the T-quencbr. Per Refier 12, TVA addresned ECCS suction separation. TVA evaluated the physical configuration of th suppression pool, MSRV T-Quencrs, nd ECS sction stainer utilizing fte infnmalion contained in NBDO-30832 (Reference 13), the NRC SER and t aociated Brookhaven report Based on this evaluaton, the BECS suction piping would not ingest steam bubbles hat could later collapse and induce waterame load. TesWcochisiu innain valid rhEh Ptconditions.

4.1.12 Short-Term ( aTremperatureResponse The drywell airsace temperature limit is specified in Table 4-1. This limit is basd on a bounding analysis of TIe supeoheated gPs temperaure reaced during the steam blowdown to the drywell during a LOCA. The changes in the reor vessel conditions at EPU increase the xpcod peak drywel gas temperature following B LOCA by 1°F. Therefore, the dryvell gas temperature response with EPU does not exceed the limit Short-term containment response analyse for DBA.LOCA demonstrat fth:t operation at EPU RTh does not reauIt in excoeding the contain t design limits. These analyscs cover the bloattwn period wr the mnxinmm drywell airspa= temperatu occurs. The analyses were performed at 102% of 8PU RI?, using the methods reviewed and accepted by the NRC during the. rk I Containment ITI. The calculated peak drywull airspace tmperaWes are provided inTable 4-1.

Tzble 4-1 also shows the values frm calculations Bt CLTP usn the same methods. The total time that the drywell airspace temperatue exceeds th containment stucuraI deign basis tmperature of 281 is lessIthan one minute. This short dion isnot sufTfient for tho avag shall temperatro to exceed the contaimnent structural design temperature.

The wetwell gms space peak temperature response is calculated assuming thermal equilibrinn between the pool aid wetwell gs space for fth short-term cortainment responsi. Table 4-1 shows that the calculated bulk pool toemperturo icreass slightly at the EPU condiion, Thccefore, the wetwelI gas space increases by the samn amount The, short-term wetwLl gas spaCe tempra s at the EPU conditions ar below the suppression chamber the design temperatures. Therefore1 the abort-term wetwell gas temperature respones at EPU are acceptable.

4.1.13 Short-Term ContainmentPresaureResponse Short-term ontainment response analyses wer performed for tho limiting DBA LOCA, which assumes &double-ended guillotine break OfE reciroulation suction Iine, to deornostrate that EPU does not result in exceeding the containment design limts. The short-term anlysis ccvca the blowdown period during which the maximum drywell pressure. and differential pressures between the drywell and wetwell occur. These analyses were perihrmed at 102% of EPU RTP, using methods reviewed and accptd by the NRC during the Mark I Containent LITIP with Ihc break flow calculated using a more detailed RPVimodel (Reference 5) previously approved by Ihe NRC The results of these short-term analyses are snmmarized in Table 4-1 for comparison to the drywell design pressure. As shown by these results, th maxium dryweil pressure values at the EPU conditions are bounded by the UPSAR analysis value and by fl=

design pressure.

4.3

T ':, ' , '

NEDO-33047 - Revi or O 4.1.2 Contalnment Dynamic Loads 4.1.2.1 Losof-Coolant Accident Load.

The LOCA coniainmt dynamic loads analysis for BPU i based primarily on the shortterm LOCA analyses. These analyses were performed as desbed in Section4.1.1.3, using the Mark I Cantainment LTIP method, except that the break flow was calculated using a more detailed MIP model (ference 5). The application of this model to BPU cwtainment evaluations is identified in ELTRI. These analyses provide oalulatd values for the controlling paraeters for the dynamic load; throughout the blowdown. The key parameters are drywell and wetvell pressure, vent flow rates and suppreuion pool temperm . The LOCA dynamio loads for EPU include pool swell, CO. and chuggig loads. For Mark I plants like. the Bromns Peryvunits, the vent thrust londs are also evaluated.

Thoe sart-term cotalinm t rssponso conditions with EPU are within the rge of test conditions used to define the pool swell and CO loads for Browns Pery. The peak dzywell pressure fhom these analyees is given in Table 4-1. The long-twim responlS conditions at EFU conditions when chugging would occur ar within the conditionr used to define the chugging load.. The vent tst loads at EPU conditions arm caleulated to be less than the plant-specific values calculated during the Mark I Containment LTTIP. Therefore, the LOCA dynamic loads are not affected by EPU.

4.1.22 Main Steam Relief Valve Loads T1e MSRV ai-clearing loads include MSRVDL loads, suppression pool boundary pressure loads, and drag loads on submerged strucfts. These loads re influenced by MSRV opening epoint pressurej the initial water leg in the MSRVDL, MSRVDL geonmely, and suppression pool geometry. For the first MSRV actuations following an event involving RPV pressurization, the cotrolling parametric change introduced by EPU, which can affect the MSRV loads is an increase imMSRV o g setpoint pressure. However, this EPU does not include an increase in the MSRV opening setpoint pressures. BPU may reduce the timn between subsequent MSRV actuations, wich may affect the load definition for subsequent actuations.

The MSKV oping load walues, which ame the basis for tho MSRVDL loads and the MSRV loads on te suppression pool boundary and submorged struces ame not changed . The ect of BPU on the Icd definition for tibsoquent MSRV atuations ha been evaluated. The load denition for subsequent MSRV aetuftions is not affeted becaeo the MSRVDL roflood heig -used for Browns Pony is the maximum refiod height (Refmeace 10), which is controlled by th MSRVDL gemetry and the MSRVIDL vacumn breaker capacity. Becauseo all thee parameters, i d fg theSRV soWeis, do not change, loads due to subsequent NSRV actusdons are not affected by SWP.

Therefoe~ PIdoes notasfcttheMSRVloadsorloeds dafinitions.

4.1.23 SubcompartmentPressurizadon Thz annilus pressure load on te biological shield wall due to a postulated break in a 4-inch jet pump instruent line nozzle is evaluted at BPU oonditiornS. The annuls prossur load (2.4 paid) evaluated inUFSAR Section 12.2.2.6 (at 102% of CLTP) renains bounding compared 4-4

NEDO-33047 .yReydon 0 tothe 102% of BPU nnumus rsurelondof 2.3 psid fornormal.FW temperatures. ForPFWrR at 102% ofBPU conditions, the nulus; pessure load is 2.6 pdid. The biological shield wall and component designs remain adequate, because there Is subsantial nmrin to the structural design value of I9psid.

4.1.3 Continmant1botation The system designs fbr oentainment isolation are not affcted by EPU. The capabilties of isolation actuation dvice; to perform during norma operations d under pout-accident condition, have- been deteined to be acceptable. Therefore, the Browns PeTry containment isolation capabilities are not adversely affected by the EPU.

The AOV and SOV parameters (tmperature, pressure, flow) were reviewed and no changes to the fimetlonal requireents of any AOV/SOV were identified as a result of opecrat at EPU conditions.

Operaion at the EW) condit ions is w!ihn hliIpressu- and temperature capabilities of the AOVs and SOVs. Therefore, tho AOVs and SOVa remain capable of perforning their design basis fUnction.

4.1A GenerIc Lefter 8910 Program The MOV process pvarnmtrs (temperare, pressure, flow) wore reviewed and no significant changes to the fimctionul requirements of the OL 89-10 MOVs were identified aa a result of operating at EPU conditions.

Operation at the BPU conditions increases post-accident room temperatures (< 100 F) where the MOVc are located. Operation at th imcreased EJ conditions is within tho pressure and amnbient temperature capability ofthe tL 89.10 MOw. Thertfore, the OL 89-10 MOVs reanin capable ofperfonming their design basis fiuctiona.

4.1.5 Generfc Letter 89-16 In response to Generio Letter 89-16, Browns Ferry installed a HWWV system. The current design of the HWWV was based on 1.05% of 3293 MWt (OLTP). Therefore, at the EPU RTP conditions, to .xistIng HWW' exhausts a smaller pcontago of RTP. Based on the as-built design, th HWWV would exhauit aproxirmately 0.88% RiP at 3952 MWt (EPLI RTP) and is designed to be operational during a,$BO.

The primary objetive of the hardened wetwell vent is to preclude primuy containment failure due to overpressurization, given a lose of decay heat removal CMW sequence) wenL Using the ANSI/ANS-5.1-1979 decay beat (nominal) cunre, 0.88% RTP is reached at approximately 5.6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. From EPU conditions, the containment pressure at 5.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Is 46A paig, which IS below the contaiment design prssure end primary containment pressure limit of 56 psig. At thee EPU conditions, decay heat will be below the relieving capacity oF the hardened wetwell vent before containent pressuro reache the design prssure IinT therefore, the existing HWWV Mees te intnt Df Generic LAeur 89-16 fbr EPU conditions.

4-5

NEDO-33047 . Revlalon 0 4.1.6 GenericLefter 9507 MOW used Bs contzinment or HELB lsolation valves have been reviewed for the effects of operaions at EPU conditions, including trzmal binding nad pressure locking (Generic Letter 9507). The operability ofMO Vs is documnted as part of th plat GL 89-1O program.

4.1.7 GenerfcLetter96-06 The Browns Ferry Waluations for Generic Letter 96-06, "2dsurtmceof Equpmenz Operabilty and ContaimentIntegrItyDufingDesugn-Basbs Accident Conditions,"were acoomplishedusing the peak drywell temperature (336"F) for a. MSLB inside containment. The equipment and containment remain within their design aloawablks for EPU conditions.

4.2 EMERGENCY CORE COOLING SYSTEMS Ewci ECCS is discussed in the following subsections. The effect on the fimctional capability of each systerm due to EPU is addessed. Tho ECCS porfbrman;. evaluation is contained in Section 4.3.

4.2.1 Wgli Pressure Coolant Injectlon System The HPCI systm is deignd to pump water into the reactor vessel over a wide range of operating pressures. The primary purpose of the HPCI is to maintain reactor vessel coolant inventory in the event of a small break LOCA that does not immedeiately depressurize the reactor vesal. In this event the HPCI system maintains reactor water level and helps depressurize the mnctorveseal. Tho adequacyofth HPCI systenmis demonstrated in Section 4.3.

[E j] the HPCI pump and turbine remain within their allowable operating envelopes, the HPCI "yt is capable ofdelivering its design injection flow rate, and the turbine has the capacity to develop the required horsepower and speed.

Therofore, the, PCI system is acceptable for EPU.

4.12 LowPresure Coolaat Injetion The LPCI mode of the RHR sydem is automatically initiated in tho event of a LOCA. Whtn operating in counction with other ECCS, the LPCI mode provides adequate core cooling for LOCA events.

The incmrse in decay heat due to EPU could increase the calculated PCI' following a postulated LOCA by a small amount. The ECCS performance evaluation presented in Section 4.3 deronstrates that the exiting LPCI mode performance capability, in conjunction with the other ECCS, is adequate to meet the post-LOCA core cooling requirement for EPU RIP conditions.

1E 4-6

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X NEDO-03347 - Rvson 0 4.2,3 Core Spray Systen The CS system is automatically initiated in the event of a LOCA. When opeing it conjunction with other ECCS, the CS system provides adequate core cooling for LOCA events.

Therc iitno change in the reactor premim at which the CS system is required to operate.

The increase in decay heat due to EPU could increase the calculWed PCT following a postulated LOCA by a anall amnount Thz ECCS pefformnee evwaation presented in Section 4.3 demonstatee that the existing CS syste performance capability, in conjunotion with the otter ECCS as required, is adequate to meet the post-LOCA core cooling requirement for the EPU conditions. [

J))

4.2.4 AutomaticDepremsurizati System The ADS evaluation scope isprovided in Section 5.6.8 of ELTRI.

The ADS uses MSRVs to reduce reetor pressure following a small break LOCA, when it is asmned that te high-presue ECCS has failed. ThIs fimction allows LPCI and CS to inject coolant into the vessel. Plant design requires a zninimum flow capacity for tho M8RVa and ta ADS iitiates following confirmatory signals md anociatd time deliy(s). The required flow capacity and ability to initiate ADS on appropriate signals are not affcced by BPU. The ADS initiation logic and ADS valve control [( are adequatz for EPU conditions.

4.2.5 ZCCS Net Positv Sucdon Head Following a LOCA, the RHR and CS pump. opeate to provide the required core and containmexnt cooling. Adequate margin (NPSH available mrinus NPSH required) is required duming this period to ensurC the essential pump operationL The limiting NPSH condition occur during either short-term or long-term post-LOCA. pump operation and depend on the total pump flow rabs, debris loading on, the suctlon strainers, and suppression pool temperature.

WA previously requested containment ovcrprossure oredit for Browns Perry Units 2 and 3 (Refore 16). In Refmrnce 16, TVA indicated that thceneed for containment overprusmure credit in the short twn was based on RHR requirements, and in the long term was based on CS requirements. The pre-EPU analysis indicates that up to 3 psi of overpressura credit (considering whole nunber value) is required for the short-t=e case forRHR pump operation to maintain adequalz NPSH. One (1)psi of overpreasure credit is currently required and pproved for the long-term case for CS pump adequate NPSH.

Far both the pre-EPU Pad the EPU anlyze, aimaximized suppression pool terupemture and a mnimized contaimnt pressure wer. assumed. EU RTP operation incrtases the reactor decay 4-7

-..... , . - r.- . . . .. .

NEDO-33047 T Revioa 0 heat, which increases the heat addition to the SppreIHion pool following c LOCA. Therefore, changes in vapor pressure cowrreonding to the increase in supprossion pool temperatrs afect the NPSII margin. After 10 minutus, operation of the RHR pumps far containment waling in the containment spray modo with continued operation of a CS loop for SCS injecton is also assed.

nTo NPSH margins woe calculated bsued on conservatively assuming RHR maximum flow rates and CS desi flow rates during the shot-tc m, and REIR and CS design flow rates during thdlong-term. Tho system flow ratu for the shiort-torrn caso r 42,000 gpm total RHR flow and 12,500 gpm total CS flow. The system flow rates for the long-term ase are 13,000 gpm total RXR flow and 6,250 gpm total CS flow. The methodology used to deterie the amount of debris generatd and transported to lbe ECCS strainers is generaly based on NEDO-326E6, the BWROG Utility Resolution Guidance for EMCE Suction Strainer Blockage. The minimum quantity of paint chips recommnded by Ihis guidmac is 85 Tbs. Browns Ferry has identified a maimmurn Furfae area of 157f1 for unqualified coatings within the primary containment which reprosents an additional I lbs. Therefore, a total of 103 lbs of strainer paint debris wa used for sizing the strainers. This quantity did not change with EPU. Because the ECCS pump flow rates vere unchanged for EPU, strainer approach velocities were not alfected. Therefore, the debris loading on the suction stainers for BPU is the same as the pre-EPU condition. The assumptions in th, ECCS NPSH calculations for fricton loss, statio head, strainer loss, flow, and NPSH required have not been changed since the issuance of the amcndment related lo NRC (L 97-04 (Reference 15).

Tho short-torm EPU NPSH analyss (0 to 600 seconds) indicates that with a containment overpressure (suppmsion chamber air space pressure) credit of 3 psi the RHR pumps have adequate NPSH margin. The short-term analysis also indicates that greater than 3 psi of overpressure is available fom the beginning of the eventuntil approxfmately 35 0 sconds. From 350 seconds to 600 seconds, the short-term analysis (using inputs that conservatively maximized suppression pool teperatue and minimized containment pressure) indicates an available overpressure of less than 3 psig. For the brief time that the short-tenn analysis indicates the less than 3 psi is available, the RHR pumps only require 245ps. In addition, historical plant testing has demonstrated that the RER pumps ar capable of operating for short priods of tim at NPSH values lWs than (approximately 9 feet) the manufacturr's required NPSH wthout degradation or substantial loss of flow. Therefore, RUB pump operation is not adversely affected by containment pressure less than 3 pui. Tlis was previously presented for pre-UPU conditions and approved by the NRC in Reference 15. In Reference 1S, the NRC stated that 'the use of 3 psi of containment overpressure above the initial uirspace pressure is accepable for the first 10 minutes after a,LOCA." Reference 15 also concludes that CS pump operaton is not affectod by tis lower containrmt overpressure during the short tern.

The long-term EPU NPSH analysis (0 until the end of the event) indicates that up to 2 psi (considering whole number value) containmen overpressure credit is required when the suppression pool tenpterut exceods 191F to obtain adequate NPSH margin for the long-term operation of tie CS pumps. This is an increase from the I psi of overpreusure credit currently approved for pre-EPU conditions. The long-term analysis demontrates that greatr than 4 psi of containment ovepesmure is available during this period.

4-8

NEDP033047 - Revlsiln O Tables 42 and 4-3 provide the results of the short-term and long-term containnt responses Table 4-4 provides the suppression pool temnpeuture and the required containment overpressuro to mainta NPS Hmargins during the DBA LOCA for EPU conditions.

Based on the above, Browns Eerry it requouting approval of 3 pi of oveupressure credit to meet both the short-tenn and long-term NPSH requireme. A ingle contaimnt overpresaure credit value is requested both to account for potential future contingencies and to provide consistcy betw the inpur to the shaort and Iong-texm analyses. Other means to increore the NPSH mugin wcre found unfeasible.

One RHR pump is required to operate during either the. SBO or an Appendix Bfire event. BPU RTP operation inc=aes the reactor decay het which incme s the heat addition to the suppression pool following thest events (see Sections 6.7.1 and 9.3,2). As a result, the long-term peek suppression pool water teperature and peek containment premare increas. The NPSH evaluation at these peak pool temperature shows adequate NPSR margins during the SBO and the Appendix R events with containment overpressures of 1 psi and 10 psi, respectively.

The HPCI system primary function is to provide reactor inventory makeup water and assist in depressurizing the reactor during an intrmediate or smalI break LOCA. The HPCI gystem can operate with Suction from the suppression pool at a temperatre below 140fl during the first 10 minutes after initiation of the event. BPU hfs an insignificant effect on the time for the suppression pool temperat= to reach 140QF. If the HPCI pump operates beyond the first 10 minutes following the event, the reactor operator may terminate HPCI pump operaton when the suppression pool temperature reaches 140°F. The HPCI pump NPSH margin remains adequate as long as the suppression pool temperature does not excoed 140°F during HPCI operation.

HPCI system operton is credited during ATWS, Appendix R., and SBO events. The suppression pool temperature docs not affet the NIPSH main, because the HPCI pump takes suction from the CST during these events.

4.3 EERGENCY CORE COOLING SYSTEM PERFORMANCE.

The Browns FDey ECCS for each unit is designed to provide protection against postulated LOCAg-caused by ruptureS in the primary system pip4 end the ECGS perfomance charactristicL do not change for EPU. The ECS-LOCA performance analysis demonstrate.

that 10 CPR 50,46 requirenmets continuw to be net following EPU operating condition,.

1E

i 4-9

NEDO033047 . Reviuion 0 Tho E PI efflot on PCr for small recirculation line breaks i largor than the BPU bffct on PCT far large line breaku. ThE incrasd decy heat assoiated with EPU results in a longer ADS blowdown time leading to a lur BCS system injection and a higher PCT for the nall break LOOA, As as.retul the limiting LOCA case that define. the Brown. Perry Licening Basis PCT utBPU far GE14 fuel is asmallrecirculaffon discharge line breakwithlaBafy failure.

The effects on cwnpliance with to othe acceptanc oriteria of 10 it: 5OA6 for the limiting large and zmall breaks re evaluated fbr the Browns Ferry EPU. For power uprats, there is ,

negligible effect on compliance with the oer ceptance criteria of 10 CFR50.46 (local claddlag oxidation, coro-wido metal-water reaction, coolable geontry and long-term cooling).

The local cladding oxidon and core-wide metal-water reaction were calculated and detenrined to be within the 10 CPR 50.46 acceptance criteriaL Coolable geometry and long-torm cooling have been disposlioned generically for BWRs. These generic dispositions are not affected by EPU.

The Licensing Basis PCT is deminod based on the calculated Appendix KC PCT at ratd core flow with en adder to account for uncertinties, For th BPU, the OI3 Licensing Basis PCT is 17800?F at rated core flow. The comparable 01313 Licensing Basis PCT for the CLTP conditions is 1810 0F at rmted core flow. For the EPU, tho GB14 Lioensing Basis PCT is 1830 0 F at rated corc flow. The comparable CR14 Licensing Basis PCT for the CLT? conditions is 1760°F at rated core flow. At EPU conditions, the limiting break size is the large break for GE1S and the 0.06 t small break for GE 14. The result. of these analyzses are provided in Table 4.5. The changes in PCT are small when compared to the PCI main to the 10 CFR 50.46 licensing limit of 2200 0F.

Rhfece 17 provides justification for the elimination af the 16000? Upper Bound PCT limit and generic justification tha the Licensing Basis PCIT will be conservative with respect to the Upper Bound PCI. Tho NRC Sf in Roferonoe 18 accepted this position by noting that becauso plant-scific: 'Upper Bound ?Cr calculations have been perfornmd for all plns, other means may be used to demonstrate compliance vift th original SER limitations, These oth means ar acceptable providcd there are no significant changes to a,plant's configuration that would invalidate the existing Upper Bound PCT calculations. The canges in magnitda. of the PCI due to EPU demonstrate that this plant canfiguration does not invalidate the existing Upper Bound PCT calculation. Afltr the implawntation of EPU, te Licensing Basis PCT will continue to bound the Upper Bound PCT. Therfore, the Licensing Basis PC1' is sufficiently conservativz.

For SLO, a, multiplier is applied to the Two-Loop Operaion PLHGR and MAPLHGR limits.

This multiplier insures that the Two-loop Upper Bound PCT is also bounding for the SI)O case.

The SLO PC? valuea remain well below the 22000F limit 4.4 MAIN CONTROL ROOM ATMOSPHERE CONTROL SYSTEM The CREVS processes outside air needed to provide vntilation and pressurization of the CRHZ during accident conditions, The CRE VS units arm trd and the CRHZ is isolated on receipt of 4-I0

- --- -- --- . -- --. I-NED433047 - RoYudon 1 r primary coinimant isolation signal or high radiation signal in the Control Building itake duct When the CRIZ is isolated, a fixd umount of outside air is ftered.

[VA has submitted a rqet for an amendment to the plai-operating license that supports Te M1 scope implementation of an AST for Units 1, 2 and 3 (Referenwo 17). The WA request Includes the radiological dose consequences for the design bases acidents and includes the CREVS operational parametert at BPU conditions.

4.5 STANDBYGASTREATMENTSYSTtM The SG0T8 is designed to maintain secondary containment at a,negative pressurc and to filter the txhauat air for removal of fission produs potentiilly present during abnormal condition. By limiting the releas. of airborne particulates and halogens, the SOTS limits off-sits and control room dose following apostulated design basis accident.

TWA as submItted a reque for an amendment to the plant-OPerting Uces that supports the full scope ierpleientation of an AST for Units 1, 2, and 3 (Referenoc 19). Thc TVA rcqucst includes the radiological dose conseqences for the design bascs accidents and includes the SOTS operationalparameters at BPU conditions.

4.6 MAIN STEAN ISOL4TION VALVE LEAKAGE CONTROL SYSTEn Browns PFrry does not use a.MSIV-LCS.

4.7 POST-LOCA COMBUSTBE GAS CONTROL The Combustible Gam Control Systen is designed to maintain the post-LOCA oonceufration of oxygen or Lydragen in the containment atmosphere below the lower flammabilify UlmL As a rosult of EPU, te post-LOCA produodcm ofhydrogcn and oxygen by radloIysis increases proportionally with power level. This increase In raiolysis has an effoc on dhe tim available to start the system bshre reching procedw ally controlled lnit but does not affect the aity of the system to maintain oxygen below the lower flammability limit of 5% by volume as specified in Safety Guidc 7. The required star: tim for tho CAD syswte decrcases from 42 hours4.861111e-4 days <br />0.0117 hours <br />6.944444e-5 weeks <br />1.5981e-5 months <br /> to 32 hour3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />sas & result of Esu. This reduction in required CAD initiation tim does not affect the ability of the operators to respond to the postulated LOCA. The integrated hydrogen production rates frt radlolysis and motae-water reactiot aro shown ia Figure 41. U0ortrolled hydrogen and oxygen concentratiomn in tht drywell and wetwell are shown in Figure 4-2 and tIh Dry-well PressureResponse to CAD Operation without Venting is shown in Figu 4.3.

The TS require sufficiet on-sita storage of nitrogen in each of t. tvo 4000-gallon sorg tanks to maintain contanent oxygen below 5% during the 7-day period fbllowing the postulated LOCA. For CLTP, this requirement is satisfied by maintaining a minimum of 2500 gallons of liquid itrogen in ch tank, equivalent to t volume of 191,000 swf (20'C and 14.7 psia) per tnkc.

As a result of increaed production rte ofradiolytic gas following BPU operation, the required 7-dy volume of nitrogen hinue9 flum 155,000 motto 197,0flsao which exceeds the available 191,000 sce nWply required by th TS. An evitlon was perfomed to determine it amount 4-11

.NEDO-33G47 . RTovilon 0 teeded to mahitain a 4-day supply following the posuated LOCA. Thin resulted in a nitrogen volumo of 104,834 sof tat is less han the available 191,000 ucf lupply required by the pl TS.

The TS BDas liquid nitrogen 7-day rquiremet is conservative, because additional liquid nitrogen can be delivered within one day or less. Two liquid nitrogen dittribution facilities rer located within 1-day travel distance ftom Browns Pary. Each facility is capable of delivering 5000 gallons or more of liquid nitrogen to Browns Ferry with less ihan 4 days notice. The historical average delivery timo is I day, Thc TS re not changed, however, the TS Base. will be revised to a 4-day ntrogen storage requirmnet to accommodate EPU operations. The CAD systemnitrogen volume requim ts are shown in Figure 4-4.

4.8 REFERENCES

1. GE Nuclear Energy, "Geteric Guidelines fur Gcenral Electric Boiling Water Reactor Extendod Power Upra%" Licensing Topical Reports NEDC-32424P-A, Class I11 (Proprietary), Febrary 1999, and NEDO-32424, Class I (Non-proprietary), April 1995.

2. GE Nuolear Energy, OThe GE Presure Suppression Containmect System Analytical Model,'

NEDMI-10320, March 1971.

3. GE Nuclear Energy, "The General Electric Mark mII Pressu Suppression Containment System Analytical Model," NBDO-20533, June 1974.

4. GE Nucler Energy, "Maximum Discharge of Liquid-Vapor Mixturs from Vessels,' 1NEDO-21052, September 1975.

5. GE Nuclea Energy, Genral Electric Model for LOCA Analysis In Accordance Wit I0 C5R S0 Appendix K," NBDE-20566-P-A, September 1936.

6. NURBG-0800, U.S. Nuclear Regulatory Commission, Standard Review Plan, Section 6,2.1.1,, "Pressue - Suppression Type BWR Contahuments," Revision6, August 1984.

7. NUREG0661, "Mark I Containment Long-Term Program Safety Evaluation Report," July 1980.

8. Lettertto GaryL. Sozzi (GE) from Ashok Tadai (NRC) onto Use of tht SHEX COmputer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis, July 13, 1993.

9. GE Nuclear Energy, "Mark I Containment Program Load Defnition Rport," NEDO-21888, Revision 2, November 1981.

10. SFN Report CEB-B3-34, "Browns Pony Nuclear Plant Toms Integrity Lang-Term Program Plant Unique Analysis ReporV', Rev. 2, dated 12/10/1984 It . Letter fom USNRC to H. G. Paris, WVA, entitled "Mark I Conlaiznent Progrnm - Browns Foery Nuclear PlantUnits 1,2 and 3," dae May 6, 1985 (A02 850513 002).

12. TVA Letter, R03 991201 679, orom T. E. Abuny to USNRC, "Browns Ferry Nuclear Plant (DPN) - Unit 2 and 3, Corrected Inrmation for Technical Specification Change Request TS-384, Power Uprato - (TAC NOS M997 11 and M99712 ' dated December 1, 1999.

4-12

NRDO-33047 -Redilon 0

13. GE Nuclear Energy, `Sf1t kli ion of Limit on Loal Suppression Pool Temperature for SPY Discharge with QueanoherlT NEDO-30832, Class 1,December 1984.

14.0E Nucloar Eaergy. "Genetic Evaluations of General Electric Boiling Water Reactor Extended Power Uprat' Licsing Topical Reports NEDC-32523P.A& Clus II4-Februay 2000; NEDC-32523P-A, Supplement 1 Volume 1, Febnmty 1999; and Supplement I Volume 11, April 1999.

15. NRC Letter, "Browns Ferry Nuclear Plants, Units 2 and 3 - issuanwc of Amendments Regarding Crditing of Containment Ovcrprssiie for Net Positive Suction Head Calculations for Emergency Core Cooling Pumps (TAC Nos. MA3492 and MA3493),"

September 3, 1999.

16. WA Letter, "Browns Ferry Nuclear Plant (BEN) - Units 2 And 3 - License Amendment Regarding Use of Contairancat Overpressure for Emergncy Core Cooling System (ECCS)

Pump Net Positive Suction Head (NPSl) Analyses," Septembcr 4, 1998.

17. 03 Nuclear Energy, "GBR-LOCA and SAFER Models For Evaluation of Loss-of-Coolant Accident, Additional kfbnrafion For Upper Bound PCI CalcuLaion," NEDB-23785P-A Volume III Supplement 1, Revision 1, Mrh 2002.

18. Stuart A. Richards (NRC) to James F. Klapproth (GENE), Review of NEDE-23785P, Vol. HI, Supplement 1, Rsvision I, "GESTR-LOCA and SAFER Models for Evaluation of Loss-of-Coolant Accident Volume III, Supplemewt 1, Additional Information for Upper Bound PCI Calculation," (TAOC No. MB2774), February 1,2002.

19. TVA Ldttor, "Browns Perry Nuclear Plant (BFN) - Units 1, 2, and 3 - License Amendnent -

Alternative Source Term," dated July 31, 2002, ROB 020731 649, including Tech. Spec. No.

405 (TVA.BFN-TS-405).

4-13

s- -t @@

@ X-@

Nam-aMa4? -ftlitoatt B Table 4-1 Brawn* Ferry Contfaiment Performance Resulft Current Rad Power EPI LiUit UFSAR Current Method)

Parmeter Peak Drywoll 50.6 47.703 48.50) 56 Pressur (psig Peak Drywell 297.0 294.3c) 295.2P 340/281 Temperaturo (

Pek Bulk Pool 177.0 179.0) 187.3(') 281 Temperature (DE)

Peak Wctwvll 36.3 29.9 30.5 56 Pressure (psig)

1. hec Current Rated Power, Curent Method adalysis uses the BPU Power analysis method with CLTP inputs.

2. The acceptanoe limit for drywell airspace temperature is 340 0 F, %whilethe shell design value is 281CPO 'Telistedpeakvalues are for airspace temperature.

3. Bounding mass end energy release data points were seleced for input to M3CPT that more closely match the LAMB output for the EPU Analyse cases as compared to the previous power uprate. This technique resuls in lower mass and energy release to the drywoll, which produces a lower peak drywell pressure and mmnperawre at the sam power level.

4. Service wator tenperaturo was inreased fu 92W to 95°P.

414'

r - -} S . cowl

. . .. l-.. , .. . . -.. .- . . .- . . .

NEDO33047 -RtvilonOe Table 4-2 Browns Ferry Short-Torm Containment Input to NPSHI Anlyslt 0 14A 95 54.34 36.73 12S,9 101 31I 37.98 136A 151.47 29.67 ,. 139.0 201.84 22.73 143.1 304.94 17.95 148.3 3S1.75 17.23 149.9 399.94 16.92 151.2 500.87 16.81 153.4 600.1216.81 . 155.4 4.15

NEDOD33047 RMvion 0 Table 4-3 Browns Perry Long-Term Containment Imput to NPSH analyin Ang.

0 14.4 95 99.63 *38.45 141.0

.197.82 .36.34 142.6 297.76 34.35 143.7 40S 31.00 146.2 607 24.43 152.8 4,134 19.90 175.8 7,105 20.69 181.9 14,682 20.99 186.6 37,426 20.05 181.9 50,180 19.27 176.7 4-16

i NEDO.33047 -RevWi I Tabl4-4 Browns Ferry EN!1)U3LOCA NEST MrGob aid CoiAlnhMat Overpresue Credi i

II 600 IS 2-46 0 6.75 S tmn analyis. Ovayp u r ied t meetRHRNPSH

___ _ _requirements

'i 601 152.4 0 12.95 6.55 Lo- Adzulysb 4I 4,150 . 175Z3 0 632 0 Greatrt OAOf ovnpreamnro I

-eqrd for loog-tcmfor CS 7,090 181.85 1 6232 0 Greaw dmntpsiufovvxpresmr I required fit long-trm for CS I

14,700 186.6' 1.90 632 0 PeokSuppxtssionPool 37,500 181.85 1 6&32 0 Less tia 1 psi ofovprase r.ired for longterm f CS 4-17

NEDO-33047 - Rerblon 0 Table 4.5 Browne Ferry 1ECCS Performance Analyui Reoult 10 CFR5.46 Parameter OLT? MXU LI=

Method SAFBR/GBSTR SAFHRIGESTR Power 105% OLTP 120% OLTP

1. Licensing Basis c IS8I0 (G;EI3)(1) < 1780 (GE13) c 2200 PwakClad < 1760 (GE14)(1) < 1830 (GE14)

Tomperatre, (PCM OF

2. Cladding <2.0 <c30 517 OildaloM %

Original Clad Ihickness

3. Hydrogen CO. *cQ0 <1.0 Generation (Core wide Metal Water Rectfion) %

4. Coolablc OK OK MIeet I and 2, above Geomatry S. Core Long OK OK Core flooded to TAF Term Cooling or Core flooded to jet pump suction elevation and at least one CS systeml is operting at reed flow.

(1) An update of the Licensing Bais PCT at 105% OLTP was calculated for the EPU analysis. This alows frCIrmpiso wi the BPU Lc ing BWs PCT results.

4-18

N&DO.33047 - ReYion i lI a I 10 flrs Figure 4-1 Browns Ferry Time-Integrated Containment Hydrogen Generation 4-19

. .: .-- .. - S - .. 7r.-r. .-. -.- -. ...- r Jo NEDO-43047. Rgvkoln 0

%p5

, I O ,I; I A 0.01 m1 I 10 100 Tirm After LOGA (days)

Figure 4-Z Browne Ferry Uncontrolled HE and Oz Concentrations in Drywel mnd Wetwel 4-20

.*)****

NEDO-33047 -RWoul 0 I

I too Tofr. Aftr LOCA (dayu)

Figure 4-3 Browns FerryDrywcll Preaura Responmse o CAD Operation without Venting 4-21

NEDO.33047 - RetbIon 0 1,200Q.

, i00_ W _

II tPW.OW I2

1

800,000 I 1:3.M /Vf _____ ___

200,000 I + +

0- ,,. . ............ . ....

0 20 40 8o 100 Tima AftSP LOCA (dyus)

Figure 4-4 Browns Ferry CAD System Nitrogen Volumne Requirement 4-22

NEDO-33047 -Revidon 0

6. INSTRUMENTATION AND CONTROL The safety-reated and major (nonusafety) proc monitorng instnres, controls and trips (alytio linits for eetpointa) that could be afifoted by the BPU am addresso d below, The following evaluadons u based on tho NRC approved guidelines in Appendix F of ELTRI (Reference 1).

5.1 NSSS MONITORING AND CONTROLSYSTEMS The instruments and controls that directly interact wit or control the reactor ar usually considered within the NSSS. Thc NSSS process variables, insaunmnt setpointu and Regulatory Guide 1.97 instrumentation that could be effected by the EPU were evaluated. As part of the ETU implementation, the NRC approved VA sectpoint methodology (Reference 2) is used to genera the allowable values and (nominal trip) setpoints related to the analytical limit changes shownin Table 5-1.

The following summarizes the rerults of the NSSS evaluadona.

5.1.1 Control Systems Evaluatiou Changes in process vnriables and their efets on instrument satpointn were evaluated fbr the BPU operation. to deterine any related changes. Process variable changes are inpleInnted through changes in plant procedures.

TS instrument AVe and/or se6ointa are those sensed variables, which initate protective actions.

To determination of insfrumsnt AVs and etpoints is based on plant openg experience and the conservative ALs used in specific licensing safety analyses. The setnis are selected with sufficiet margin to preclude inadvertent initaion of the protective tation, while asUring (hat adequate operating margin is niaintained between the system settings znd the acual limits.

Increases in the core thermal power and steam flow affect name instrument setpointn, as described in Section 5.3. These setpoints were adjusted to maintain comparable differences between system settings and actual linits, and were reviewed to essure that adequate operational flexibility and necessay safety functions are maintained at the EPU RT? lvcl.

5.:2 Neutron Montoring System The APRM power signls ame recalod to the EPU RTP level, such that the indications read 100% aethe :nw licensed power level.

EPU imnplementation has little efiect on the IRM overlap with the SRM and the APRMs. Using nornal plant survillsace proedures, the IRMs may be adjuste as requird, so that overlap with the SRMs and APRM6 remains adequate. No cbange is needed in the APRM downscale At EPU IT, the average flux ecxprienced by the detectors incrases due to the average power increase in the cor The maximum flux exerienced by an LPRM rernin apprdat the sam becaus the poek bundle powers do not appeciably inrease. Duo to the inmso in ttrom flux eAperlenced by the LPRMs znd TIP., th neutronic life of the LPRM detetors mey be reduoed and S.1

• p -* r - .. -w  ?*

• I . - -. . * -I - .  : -. *.. * . - . - .

NE0-33047 - RnIslen 0 radiatdn levels of te TWPs msay bc inca. LPRMu are desied es rqCheeable componertt.

The LPRM acouracy at the Icteted flux I. witin specied limits, nd LPRM lftin lis an operational casderation that is handled by routine rplacanent. TIPs are stored in shielded rooms, A sall increase in radiation levels is ooommodated by the adiation, prot*otioprogra for normal plant operation.

The incruec in power level at the same APRM reference level mults in increasd flux at the LPRMs that am used as inputs to the RBE The RBM instmeation is referenced to en APRM channel. Because the APRM haa been rescaled, there i only a small efeot on the RBM porformanco duo to the LPM parformance at the higher average local flux. The change in performance does nothavc a significant effoct an the ovenall RBM perfomance.

The Neutron Monitoring Syste installed at Browns Perry are in accordance with the requirements established by the GE design specifications, 3.1.3 Rod Worth Minmizer The RWM isBa normal operating system that does not perform a safety-related function. The function of the RWM is to suppot the operator by enforcing rod patterns until reactor power has reached appropriate levels. ((

]The power-dopendent instrument setpoints for the RWM are included in die TS (see Section 5.3.12).

52 BOP MONiTORING AND CONTROL SYSTEMS Operatiot of tho plant at the EIU R1P levul hua minimal efioct on the BOP system instrumentation and control devioes. Based on the EPU operating conditions for the power conversion and auxiliary systems; most process control values and intrumentation bave sufficient rarteadjustment capability for use at the expected EPU conditions. However, some (non-sxfoty) modications may be needed to the power convermlon sysomas to obtain full BPU RIP.

5.2.1 Pressure Control System The PCS is a normsl operating systoem.1hat provide. fat and stable responses to system disturbances rlaed to sueam pressure and flow changes so that recator prssure is controlled within Its normal operating range4 This system does not perform a safety function. Pressure cntrol operational testing is included in the EPU implemenfation plan as described in Section 10.4 to ensure that adequatz turbie ontrol valve presure control and flow margin is available.

If ]n 5.2

r.. 4o .7rt -. . * .w~

@* -* -.ev*@v-*-e-vX--

l .. ,. ^, .

NEDO-33047 - Revision 0 5,2,1.1 EHC Turbine Control Syitem The turbine EEC system was reviewed for the increase in core thermal power and the associated increase in rated steam flow. The control wtomi are expected to perform normally for BPU RIP operafion.

ND modificao to the turbine cotol valves or the turbine bypass valves are required for operation t thc EPU conditiot. Confirmation teting will be performed during power asenion (see Section 10.4).

5.2.1.2 TurbIne Steam Bypas Systam The Turbine Steam Bypass System is a norrnal operating system that is used to bypass excessive steam flow. The bypass flow capacity is included in some AOO evaluations (Section 9.1).

These evaluations dmonstrate the adequacy of the bypass system. Some ofthe limiting events in the reload analyses take credit for tho availability of the bypass sytem. Thelreload analyses are used to establish (he oore operating limits.

5.2.2 Feedwaftr Control Sstenm Thc PW control system controls reactor water level during normal operations. (The capacity of the FW pumps to adequately support EPU RTP operation is discussed in Section 7.4.) The minimum excess flow capacity requirement for adequats reactor water level control is approxhnately 5% of the operating point flow rute. The control signal range is capable of accessing as much of the flow as needed. Therefori, the capacity is sufficient for acceptable control.

The control system itself is adjusted to provide acceptable operating response on the basis ofunit behavior. It will be set up to cover tho curent power range using start and periodiD testing.

An expansion of the steam flow signl range (part of the three-element control mode) is planned to ensure fll coatrol near the EPU RTP event wiE one MSEV closed. No changes in the operating water level or water level trip setpoints ame required for the BEU; Therefore, margin for trip avoidance is maintd . Fox BPU, the FW flow control system device settings have the sufficent adjustenet ranges to ensure satisftoty operaion. However, this will be confEired by pertfoinlng unit tests during tho power ascension to the. EPU oonditions (Section 10.4),

11]

Failure of this ,sytemIs evaluated [j th ft FtW controller failurei-maxlnm demand event. An LOFW trnsient event can be caised by downscale.

foilurofthDcontrols. TheLOFW eventis discussed inSection 9.1.3.

5.23 Leak Petectdon System The only cffct on the LDS due oBPU Is a sihti e- i b W tmeat and steam flow,

[t J] Th reased FW temperatr results in a mall 5-3

e NED -33f47 - Revidon O increase in tho MS tbnne tempec(urec lFI. [I

)) MSLtighflow is disiacd in Section 5.3.4.

5.3 INSTRUMENT SRTPOINTS TS instrument AVs and their associated NTSPs ae provided for those senscd variables that initiate protective actions and nre generally asociated with the safety analysis. TS AVs ar highly dependent on tie resuls of the safety anaalyu The safety analysis pencrally establishes the ALs. The deterruinaion of the TS AVs and Vle NTMPs includes consideration of measurement uncertaindes and is derived fom the AL9. The settings are selected vith sUificient margin to minimize inadvertent itiation of the ptotaive action, while assuring that adequate operating margin is maintained between the system settings and the actual limits. Thero is margin in the safety analysis process that is considered in establishing the stpointprocess used to establish the TS AVff and selpoints.

Increases in the cor thernal power, FW flow, and steam flow affect some instumeat sefpoin&.

Those setpoints are adjusted to maintain ccmparable differences between system settings and actal limits, and are reviewed to nsure that adequate operational flexibility and necessary safety fmtions are maintained at the EPU RIP level. Whore the power increase results in new instrumets boei emnployed, an appropriate sotpoint cekulation is perforned and Ts changes r irpplemented, as required.

All TB insenta were evaluated for effccya from EPU using the cdsing WA setpoint methodology (Refacen 2). This methodology is onsistent with NRC Regulatory Guide 1.105, ead has been previously reviewed by the NRC, This evaluation included a review of evironmnental (i.e., radiation and tempeaure) effects, process (i.e., meaured paramete) effcts and analytical (i.e., AL and margins) effects on tho subject instruneats.

Th instrument fucton AL is the value used in the safety analyses to demonstrato aceptablo nuclear safety system perfotnnce is maintainsL The AV and NTSP are then chosencalculated such that the instrument funCtions boeir reaching te AL under the wost-cae eivironmentalevent conditions. Instrument NTSPs account for measurable instrument ChSbscterisios (e.g., dri, accuracy, repeatability).

Table 5-1 summanizes the current and EPU ALs.

1U S4

. .. I..

NEDQ-33047 - Rneetloe o 5.341 High-Pressure Scram Duri a pressum increase transient that is not terminated by a dirct scram or high neutron flux scam, the high-pressure scram tenninsat the event. The reactor vessel high-;pssure scram signal settings are maintaincd slightly above the reactor vessel waximum normal opering pressr and below the specified AL. The setting permits normal opmtion without spurious scrm, yet provides adequate margin to the maximum allowable reactor vessel prsmure.

1]

5.3.2 HIgh-Pressure ATWS Reclrulaion Pump Trip The ATWS-RPT trips ths recirculation pumps during plant transients associated with increases in reactor vessel dome pressure and/or low reactor water level. The ATWS-RPT iEdesigned to provide negative reactivity by reducing core flow during the initial pan of an ATWS. The ATWIS-RPT bigh-pressure setpoint is a sigficant factor in the analysis of hbe peak reactr vessel pressure fom an ATWS event. The ATWS-RPT low reactor water level setpoint is nat a significant fctor for the limiting ATWS events. The low reactor water level stpoint is not affected by BPU.

The major consideration for the ATWS-RPT higb-pressure setpoint is an increase in the calculated peak vessel pressure during a hypothetical ATWS event because of the higher initial power. The ourrent ATWS-RPT high-pressure setpoint was included in the ATWS evaluation discussed in Section 9.3.1. This valuation concludes that the calculated peak vessel pressure remains below its allowable limit for on AlTWS event Therefore, the current ATWS-RPT high-pressure setpoint is acceptable for EPU.

5.3.3 Maot Steasm Reller iVuve Because there is no increase in reactor operating domne prassure, the setpoints for the MSRVs are not increased. Thus ALs for setpoints do not need to be updated. The cuzrent values were used in the ovrrvesue protecton and transicet wanlyses discussed in Sections 3.2 and 9.1.

5.3A MaIn Steam ligb FlowIsoladon The MSL high. flow isolation is used to initiate the isolation of th Group 1 primary containment isolation valves. The only safety analysis event that credits this trip is the MSLB accident For this accident, there is a diver trip from high area teniperafe. There is sufficient margin to choke flow, so the AL for EPU is maintained at theo current 144 peroent of rated toam flow in each MSL.

No new instrumentation is required because the existing instumentation has the required upper range limit to re-span the inrnmnt loops for tho highr steam flow conditi;O A new setpoint Is calculated using the methodology noted In Section 5.3, and no TS change is required. This will ensure that sufficient margin to the trip seroint exists to allow for normal plant testing of the MSIVs and turbine sto nd control valves. This approach is consistent wit Section F.4.2.5 of ELTRI.

Sd

NEDO-33047- Revison4 53.5 Neutron Monitodring SyYkmt The AL for fte APRM Neutron Flux Scram remains the same in tei of pecnt power, and thus, the percent power values for the TS AV and the NTSP do not change.

J]1 For DLO, the clamped AL, AV, sad NTSP retain the MELLLA domain values in percent RTP.

Tho SLO AL for the fixed (clamped) APRM scram is evaluated to be the same Ls for DLO.

A new nominal tip seipoint and AV are calculatod for The APRM setdown using Browns Ferry current design basis nethodology. This methodology is bssed on GE NEDC.3 1336, wbich has been evaluated and accepted by the NRC (Referoe. 3).

The severity of rod withdrawal error durig power operaion, event is depondent upon the RIBM rod blok setpoint. This uelpoint is only Bpplicable to e contrl rod withdrawal err. if

)) Tch flow biased REM is clamped based on its power value at IOYe core flow and 100% power. The REM sstpoints are based on the cyle-spvciflc RWB transiont analysis, ozd ths, ore confirmod or revised (as nooded) via the reload core desiM review and approval process.

The ALs for the above trips are provided in Table 541 5.3.6 Main Steam Line Hllh Radiation Scram BroWns PoFrydoes not have a MSL radiation level scram.

5.3.7 Low Stea=lJ Pressuri MSIVClosure (RUN Mode)

The purpose of thin uetpoint is to iniate MSIV closue on low steam line pressure whn the reactor is in the RUN mode, This nstpoint is not changed for EPU, as discussed in Section F.4.2.7 of ELTRI.

5.3.8 Reactor WaterLevelIstrnuneats The reactor water level trip values used in the safety analyses do not require cbanng os a result otSPU.

5-6

- 5 -Z -

• so s

)IEDO-33047 - Revision .0 5.3.9 MaIn SteamTunntelHigh Temperaturc lrautalon At EPU conditions, the fncrease in ambieat tcmprtrare isnot signficant (c ID) and no change to theMSL Tunnel High Tcmperasro Isolaton setpoiat is required.

5,3.10 Low Condsuer Vacuum Browns Pury de not havn a low condenser vaccuM MS1V isolilion or scram trip.

5.3.I TS Cloanre and TCV Fast Closure Scram Bypam UhBTSV closure and TCY fast closure scam bypass allows these sort to be bypassed, whon reantor power is ufficiently low, such that th scram function is not needed to mitigate a TIG trip. This power level is the AL for determining the actual trip sctpoint, which coesr from the TFSP. The TFSP sotpoint is chosen to allow oporational margin so that scrams and recirculation pump trips can be avoided, by transferring steam to the turbine bypass system during TIG trips at low power.

Based on the guidelines in Section F.4.2.3 of ELTRI, tOe ISV Closure and TV Fast Closare Scram Bypass AL is reduced (see Table 5-1). [1

)) The new AL is bad on a reator steam flow within approxImately 1Y of the original steam flow. Due to changes to the turbine, anew first stage pressure setpoint will be determined.

F2U results in an increased power level and the HPT inodifications result in a change to tIe rlatlonship of twrblc first-stage pressure to reactor pow-er level. The TFSP etpoint is used to reduce scram end reirculation pump trips at low power levels where the turbine steam bypass syutem is effective for turbine trips and generator load rejcetions. In the safel analysi, this trip bypass onIy applice to cveta at low power levcls that result i. a tuibinc trip or load rejmcdon.

Mainlining the AL at the same absolute power as for the current setpoint, maintains the ame fransient analysis basis and scram avoidsace range of the bypas valves.

Because the HPT is modified to support abieving the uprated le1ml, a now AL (in psig) corresponding to tha same absolute power as the unt AL is established. Threfore. a new setpint is calculated using the metodology as noted in Section 5,3, and the TS applicable condition in percent RTP has been changed. The AV (in psig) for Browns Ferry will be revised prior to BPU iMpleMontion.

To ensure that the new value is appropriate, RPU plant ascension staup test or noml plant surveillace will be used to validate that the actual plan interlock Is cleaed consistent with the safty analysis 5.3.12 Rod Worth Minlmizer The Rod Worth Minimer LPSP is used to bypass the rod pattern consitints established for the control rod drop accident Ut greater than a pre-established low power level. The rneasurent 5-7

w

. 1.

NEDU-33047 -Revision O parameterg are FW and Steam flow. [

5.3.13 PIure:Regulator The PCS i diuscu ssed Section 5.2.1. The PCS provides the men by whioh theoperang wifl pressure setpoint of the reactor is established, provides for loading of the main turbine generator relative to reactor pow4er and provide. for control of the mro inrbine bypass valves. The PCS cootrvllhig pressure signal is reactor pressure.

The reactor dome prssure is not changed for EPtM. However, the incoased steam fow result in a somewhat greater steam line pressure loss. Therefore, the pressure reguator operatonal setpoint must be adjusted to achieve the desircd reactor pressure, The small differences n tuning parameter vlues will be reconfirmed during the power ascension testing. Specdflo EHC and steam bypass control system tests will be performed during the initial EPU ascension phase, as summarized in Section 10.4.

5.3.14 Feedwater Flow Seipoint for Recircnlatlon Cay~tion Protectlon The current value of the FW flow setpoint remains uimeanged ii terms of actual FW flow rate, because the oavitation interlock requirement is not based on the percentage of rated flow.

However, thg.relative setpint. as it appears on the powerflow map, is reduced slightly to account for EPU RTP. This is consistent with Sedon E.42. of ELTR..

5.3.15 RaCC Steam Line High Flow isolatIon For EPU, the AL for steam line high flow indications remain based on 300% of the maximum rated steam flow to the RCIC turbines Because there is no incea5 in the maximum reactor pressure as the result of EPU (based on the upper analytical pressure for the lowest group of MSRVs), tere is no change in tho RCIC turbine maximum steasn flow rate or in the RCIC steam line high flow differental pressure values.

5.3.16 [PCI Steam Une High fow Isolation For EPU, the AL for the steam line high flow isolation remsns based on 225% of tho maximum rated ;team flow to the HPCI turbk. Bcause there s no increaso In the m imum reactor presusro as a result of EPU (based on the upper analytical pressure for the lowest grtup of MSRVa), there is no change in the HPCI turbine maximum stem flow rats or in the HPar steam lino ih flow differential prmsure values. Tho Hacr stem l AV forhigh flow isolaton does not change.

54 REFERENCES I. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Powea Uprato," (ELTRI), Licensing Topical Reports NEDC-32424P-A, Clans m (Proprietary), February 1999; and NEDO-32424, Class I (on-proprietary), April I 995.

'.8

NEDO-33047 - Revidon C

2. 7VA Branch Technical Instruction, EEB-TI-2B, Sctpoint Calculatons, Revision 5, Febm7y 25, 2000.

3. Safety Evaluation by tha Office of Nuclear Reactor Regulation Topical Report Infunmntation SetpointMethodology Geanral Electric Company NEDC-31336, Revision 1, November 6, 1995.

5$9

• - *:YrS t ::- r ,, _, . - . --- - -- - .- - .. .. I.. . -- --- - - .-- - -

NED033047 - Revisloa 0 Table 5-1 Browns Ferry Analyical Limits For Setpotnts Y

Aunslv fftal L1mit.

Ar.ter Current lZEPU APR34 CaIibrmdon aoi, OM ) .338 3952 APRM Somratbd neormal Power Scram DLO Fixed No chop LO FIxed No change DLO Flow Blsued (%RT?) I') 0.66WD +58% 0.55W%+67.5%

SLO P0low Based (R.66V - A)+ O.55 (Wu.- A) +

S LO Flo w B ia __ ___ ___ ___V___ ____T_ ___ 6___ ___ ___67.5%

APRM Neutron Flux Sr= No change APRMI Sctdown Scrm (%RTP) 25 J 23 Rod Block Monitor DLO Flow Biased (%RTPZ) 1 0.66WO + 64% 0.55WD + 63.5%

SLO Plow Biasod (%RTP) (QM(f) O.66(WD - AW)+ 0.55Oc - AW) +

64% 63.5%

Rod Block Monitor Upcale Function Range.

Low Power Range No Chop ntermado Power Range No Chansg V High Paw Range NoChme Typical Law Trip StpoInt<) No Chunge Typic4a I orinodlato Trip setpoiLt '3 No Ciange Typinal Hgh Trip Setpointt{) No Chanp Rod Worlh Mini_ _ No Change Ve"OlHighptW Scram No CBhnge High Preasur AlWS RPT No Change MSRV Setpointa No Chang&

TSV & TC Scrm Bypas (YoRI) ) 30 l 26 510

---

-- .. . - - - I - - I -- - - - -

• NEDO-33047 -Revlon 0 Table 5-1 Browns Ferry Antlytical Limits For Setpoints (continued)

Analytical Limits Parameter Current J EPU MWL High Flow Isolation No Chan I MSL Eigh FVow Isolation (diffcrtialprc6suo) 131.7 pald l 196.6 psid MSL Tunnel High Tcmpom 6eholkdon No Cha e PW Flow C.\tat!on Intzrlok Saepoint, 'No Change Low Steam Line Preasure MSIV Closure (Run Mode) No cbanga RCIC Stewm Line High Flow Isolation No change HPCI Steam Lino High Flow Isolation INo han£g J' .

1. NDOCnditistZken inaY safetyanalysi.

2. WD Is Y recircuatlon drive flow whtee IOO0 drive Low is fth required to achieve 100ya corC flow at 100% power, and A\W Is the differoncc botwm the DLO and SLO drive flow at the uame core flow.

3. TS AV pvidod.

4. Changed on a cyclo-pecific basb and doctmnedn In the COLIL 5Sti

I NEDO-33047 - Revison 0

8. ELECTRICAL POWER AND AUXILIARY SYSTEMS I AC POWER Tno Browns Ferry AC power supplies each include both ofWlsits and on-site power. The on-site power distnbbution systzm consists of transformer, buses, ad switchger. AC power to the distribution system is prodided f-ro te transmission system o from onste Diesel Generators.

The Browns Ferry HPU plant electrical ohotareticss are shown in Table 6-1.

6.1.1 Off-Sta Power System The existing off-uite electrical equipment %sNu determined to be adequate for nonmai operation wvith the up rated electrical output as showvn in Table 6-2. 'Te only significant change in electrical load demand is due to the replacement with lager motors fbr the Condensate Booster and Condensae Pumps due to increaed flow demand a+/- EPI conditions, The reiew concluded the following:

• The Main Isolated Pha6e Burs Duct is to be modifiedup-ratcd to have a continuous crnt rating of 36,740 Amperes and an asymmetrical current rating of 346,989 amps to support the Geneator outut at !PU conditions.

• The T ap Isolated Phase Bus Duct is to be modifiod/up-rated to have an asymnmetrical curr rating of 602,143 sanps to support the Gonerator output at KU) conditions.

• The Generatr breaker is to bc modified/up-ratod to have a continuous current rating of 36,740 Amperes and zn asymmetrical current rating of 204,529 amps to support the Generator output at E! con didons,

• The exising main power traunforxsrv tmd assoiated relaying amrbeing upgraded as a.

materiel condition improvement due to obsoloscence. The replaoemet transfomers ars adequate for operation wit the HPU-related electrical output of the generator.

• Changes will be required to plant operating procodurea to prevent automatic tr fer to the 161-kV system when any unit is in backfaad or when say 'USST B is out of service to avoid overloading any of the 161-kV power supply circuit;,

• The existing 5OkV mwtchyard bums, breaker, and switches am adequate for EPU operations. Howovor, additional breakers and associated relaying ar being added to increase operting flexibility of the 500-kV owitchyid.

• The protective relaying for th min geneaor, transfrmer, and switchyud is adequate for the BPU generator output However, relays may be upgraded or added as necessay to help ensure grid stability under curtain closo-in fault conditions.

A Transmission System Study has been performed, considering the increase in electiical output to dernontrate conformnce to General Design Criteria 17 (IOCFRs0, AppendixA) and to analyze for unit/grid stability, Tho study documeted that no additional changes arc rquirod for 6-1

NED0433047 . Revislon 0 the Browns Ferry ofEBito power system to continue to meet GDC-17 requirents. Analyses in the study also determinod that operation at EEPU Electricl outputs will not have a signifoant adverse effect on reliability of the offaite electical system or on the stability of the Brows Ferny

units, 61.2 On-OtePowerlDisfrlbution System The on-site power distibution system loads were reviewed under nommal and emergency operating scenarios for EPU condifions. Loads were computed based on equipment nameplate datu or brako horsepower (BlHP) as applicable. These loade were used zs inputs for the computation of load, voltage drop, and ahort circuit currnct values. Operation at the EPU conditions is achieved for normal and emergency conditions by operating equippment within the nameplate rating running kW or applicable BHP.

The only significant change in electrical load demand is associated wit power generation sstem motors for the condensato and condonsate boostr pumps. These system pumps experience increased flow demand at EPU conditions and will be replaced with higher capacity pumps and motors. To support those load inrerases, modifications to the onsitt electrical system wvill be performed prior to EPU operation. Load flow and short ciuit calculations wr performed to verify the adequacy of the on-sito AC system for the proposed change.. The existing protective relay titangs are adequate to accommodate the incwsed load on the 4kV power syst.

Selective coordination is maintained between the pump motor breakers and the 4kV Unit Board main feeder breakers.

Sigaifoant ohange to the on-sitt power analysis include:

• The BB? of the RectorRcirculalion Punp motors increases 17% for BPU, but remais within its motor uprate analysis capability.

• The Recirculation MG sets have been replaced with VDs. IlTe capability of a VED is 9000 HIP, which is adequats for the expected Renctor Recirulation Pump motor load of 8550 BHP. '

• The electrical load demand associated with power generraon system motors for the condensate pumps and corldeosate booster pumps incroase for EPIT. lTese system pumps experlemce increased Mow dmand at EPU conditions and will be replaced with higher capacity pumps and notors.

BPU coditons are achieved by utilizing xsting equipment operating a. or below the nwmplate rating and within the calculated BHlP for the required pump motors for both normal and emergency operating conditions.

Units I ad 2 share four independent diesel generator units coupled, as an alternate sourc of safet-rclated power, to four indepondent 4160-V boards, ((

J] The systems have suffioient 6-2

w b - 6 .w. P @ @ * @ w  : w-NEDO-33047 -Redoen 0 cpacity to support OI required load. to achieve and maintain safe shutdown conditions end to operato the BCCS equipment following postulated enoidwts and transients.

6.2 DC POWER The DC power distrbution system provides control and motive power for various systmslcomponcnts iMthin the plant In normal and emergency opntaing conditions, Icad are computed based on equipment nameplate ratings. These loads are used as inputs for the computation of load, voltge drop, nd shrt ciruit cumntvalues. Te load addition for control logic relays associated with on site power system changes are within existing margins.

Operation at the EPU conditions does not inreaso any load beyond nameplate rating or reise any component operating duty cycle; therefore, the DC power distribution system remains adequate.

63 FUELPOOL The fuel pool systems consi9t of storzge pools, fuel rack,, the FFCC system, end the ADfl system. The objective of tho fuel pool system is to provide specially designed underwater storage apace for the spent fuel assemblies. Tho objective of the fuel pool systems is to remove the decay heat from the fuel assemblies and maintain thbe fuel pool watcr within specified temperature limits.

6.3.1 Fuel Powl Cooling The Browns Feny 5FP bulk wata tempemturo must be maintained below the licensing limit of 150O. The limiting condition is a full core discharge with all remining storage locations filled with used fuel from prior discharges. A normal batch offload (approximately 332 fuel bundles) is assumed for outage planning with the additional assumptions in either case (batch or full core) of only one of two htainS of the FPF0 systm and only one oftwo trains of tho non-sdety ADHR system available, 24-month fuel cycle, ANSL'ANS 5.1-1979 + 2a, and GE14 fuel. The RHR system supplemental fuel pool cooling mode may be uked to augment the capacity of the FPCC system when the ADHR system is unavailable. The batch and full core ofllozd scnmios were also analyzed with only one of two trains of FPCF system and one train of RHR in the supplemental fuel pool cooling mode. Te key results of these analyses are presented in Table 6-3. The temperature requirement ures operator comfort (an operationl requirement),

and provides ample margin against an inventory loss in the fuel poot due to evaporation or boiling.

The EPU SFF heat load it higher thian the pree-EPU heat toad. The EPU heat loads at the limiting full core offload condition and the nornal batch offload are calculated and ten the bulk pool temperaure is determined to evaluate the FPCC systen adequacy. BPU does not affect the heat removal capability of the FPCC system, the ADHR system, or The supplemental fuel pooI cooling mode of tbh PRHR system. BPU results in sligtly higher core decay heat loads durng refueling. Each reload affects the decy heat generation in the SFP aftr a batch discharge of fuel fom the reactor. Thb full core offload heat load in the SFP reaches a maximmn 6-3

< _s wb wsr ;v w * ' ' v M vS S r;

.1- .. .... - .... -. -.

NEDO-33047- Reviort 0 immedately after the full core discharge. Based on the h..t load evaluations, the SFP bulk temperature r omain less than 1501F for eitheroore offlosd can; and thus, is acceptabls far EPU conditions Thb SFP normal makeup sourc is from the Seismic Category 11 Condensate Storage systen, wiith capacity of 100 gpm and is not affctd by EPU and remains adequate for EPU conditions.

In thc unlikely event of a oomplete loss of SFP cooling capability, Table 6.3 shows that the SFP could reach the boiling temperatur eud produce a maximum boil-oft rate of 104 gpaIL Two Scismic Category I emergency maklup sources, te RHRRHR Service Water cronstie gad the EOCW system, each have a.nakeup capability of at least 150 gpn.

Prior to each refueling outagc, calculations e performed to determine the actual pool heat load and determine Which equipment must be placed in service to mainitai pool temperature.

Adminiutrative controls are used to ensure tha the fucl pool cooling capacity is not exceeded during core ofllosd Eistdng plant instumntation and procedures provide adequate indications and direction for monitoring and controlling SEP temperature and level during normal batch offloads and the unexpected case of the limiting full core offload, Symptom based operating proceduros exist to provide mitigation strategics including placing additxcmal cooling trains or systems in service, stopping fuel movement, and initiating make-up if necessay. The symptom based entry conditions and mitigation strategies for thene procedure do not requirs changes for EPU.

6.3.2 Crud Acvity .nd Corroflon ProductX The crud in the SFP would increase by approximately 2%, assuming tha all residual crud in the RCS is transported to the SFP, This is based on a RWCU system removal efficiency of 90% and approximnately [6%Increase in PW flow for EPU. However, the i easis insipnificant, and SFP wer quality is maintained by the FPCC sygecn.

6.3.3 Radlatdon Levels Thc normal radiation levels around the SEP may incr slightly primarily during fuel handling operation. Cunt Browns Perry radiation procedures and radiation monitoring program would detect say changes in radiation levels and initiatt appropriate actions.

6.3.4 uel Racki The increased decay heat 1omm tha EPU results in a bigher hat load in the fuel pooaduring long-term stage. The fl racks am designed for hier teperatures I2) th=n tbe licensing limit of IS0". There is no effect on tho design of the fbel racks because tho orginal fuel pool design temperature is not exceeded.

6-4

at\ o & @: : s NEDO.33047 -Revidon 0 6.4 WATER SYSTEMS The Browns Peny water systemS are deigned to provide a.reliable rupply of cooling water for normal opeaion nd desip basis accident conditions.

6.4.1 Senicc Water Syste The Browns FPenr water systems consist of eafety-related and nonsafety rclated service water systems, th circulating watcr system and main oondenser, and thc ultimate heat sink. The safety-related service water system include the BECW system, the RHRSW system, and the UHS. The non-safety-iolated serice, water systems include the RBCCW cyteim and the RCW Bystem.

6.4.1.1 Safety-Related Loads The safety-related service wator systenu are designed to provide a reliable supply of cooling Water during and foilowing a design buis accident for the following essential equipment and systems:

• RH s;

• SF HXs, as needed for supplemena cooling;

• SF? emergency make-up, if necessazy;

• Staby core Pnd oontinment cooling emergency barkp, if neoesanry; Lad

• EECW system.

The evaluation of the systems performance is given in the following subsections.

6.4.1.1.1 EmergencyEqiprnwentCooling Water ystem TIhe Lafety-related performance of the EECW systm during and following the mos demanding design basii event, the LOCA, for the foUowing oqipment and systems is not dependent on RTP:

EDO Enginz Coolers; ERR Pump Scal Coolers; Diesel Generator Building Chiller; Eleotric Board Room ACU Condents and Chillfn; Control BELy Chillers; N - OzAnulyrsn; Conrol Air CGmpressors; RBCCW is; and Unit Lequiprmt geing a baclp toUnit 2 and Unit 3, 6-5

NEDO-33047 -Revidon O The diesel generator loads, RBCCW liXa, control air compressor loads, RER pump seal lods, 142.02 Analyzer loads, and Unit 1 equipment loads serving a backup to Unit 2 and Unit 3 tmanin unchanged for LOCA coditions following upraftd operation. Thc building coaling loads (Area Cooling Units) also remain the sam as that for rated opeation becase the equipment performance in theae ta bas remained unchanged for post-LOCA conditions. 'TM RHR and CS Room Cooler post-LOCA hat loads incase slightly becuas ofroom temperature incaes at ERU conditions (<2YF for ERR and c 3T fbr CS), but remain within the current design limits.

The EBCW system is a shared system with the capacity to supply cooling water all threc units.

EPU does not significantly increase equipment cooling water loads, and thus, the capacity ofthe BECW system remains adequate.

6.4.I.i2 ReY&Azd eatRemovaServe Wafr S3Yste The containment cooling analysis in Section 4.1.1 shows that the post-LOCA RHR heat load increases due to an increa in thc maxn suppression pool termperatuz that ous following a LOCA. The post-LOCA containment and snppression pool response; have been calculated based on an energy balance between the post-LOCA heat loeds and the eaisling heat rmwval capacity oftheRHR andRHRMV systems. As dasussed in Sections 3.11 and 4.1.1, the existing suppression pool structure and associated equipment hav been reviewed for acceptability based o thisic aseod post-LOCA suppression pool temperatur, Therefore, the cantainnent cooling asalyis and equipment review demonstrate that the suppression pool tomperaur can bec maintained within acceptable limits in the pout-accident condition at BPU based on. the existing capability of the RHRSW system. With BPU. the RHRSW system has sufficient capacity to supply adequate cooling end makeup to the spent fuel pool heat oxchangera and spent fuel pool, respectively. In addition, the RIRSW Bystem haa sufficient capacity to serve as i standby coolant supply for long term core and containment cooling Bs required for BPU conditions. The RHRSW system flow rate is not changed.

6.4.2 MaIn CondenuerIClrculf1ng Water/Norml Heat Sink Performance T maim condenstc, oirculaing water, end hea snk system re designe to remove the beat rojectod to thc condenser and threby maintain adequately low condenser pxesure as recommended by th tinbine vendor. Maintaining adequately low condonser presure entures the efficiont opeation of thturbine-gpnermtor and miniies wear on the turbine last stagebuckets.

BPU operation Increafes the heat reqected to the condenser mad, iheforbm reduces the difference betwi the operating preure and the recomnmeded mi condmenser pressure. If condenser presures approach Go main turbine baclkpisu limitafin then meactor horuA power reduction would be required to reduce tho heat rejected to the condensor and maintain condenser pressure within to min turbine requirements.

The performance of the nmin condenser was evaluated for BPU. Thi evaluation is baud oa a design duty over the acual range of circulating water Wet temperatures, and confirms that the comdenser, circulating water system, and heat sink a adequatz for E'U operation. Curent 6-6

NEDO 33047 - Rcvnsion 0 ran turbine backpressurc limitations may require load reduotIons at the upper range of the Mticipad circulating water inlet temperabIre.

6&4.2 Dlihucrge Limts The state discharge limits were compared to the curt discharges and bounding analysis diohazrges, as shown in Table 6-4. ml, comparison demonstrates that the plant rerrans within the ate discharge limit during operation at RPU. Based on rocorded historical dta, the administrativo control procedurs presently in placo remain valid to ensure EPU operation remains within state discharge limits.

6.4.3 Reactor Building Clsed Coolng Water System The heat loads an the RBCCW system increase <0.1%. The RBCOW heat loads are mainly dcpenadt on the reactor vessel temperAturc and/or flow rates in the systems cooled by the RBCCW. The change in Yewsel tomperatur is minimal and does not result in any significat increase in drywall cooling loads. The flow rates in the systems cooled by the RBCCW (e.g.,

Recirrulation and RWCU pumps cooling) do not change due to EPU and, therfore, are not affected by EPU. The operation of the remaining equipmet cooled by the RBCOW (cg.,

seample coolers nddrain mump coolers) is notpower-dependent and is not affected by EPU. The RBCCW system contains sufficient redundancy in pumps and heat exchange to ensure that adequate heat removal capability is available during normal operation. Sufficient heat removal capacity is available to accommodatc the smill increase in beat load due to EPU.

6.4.4 Raw Cooling Water System The temperature of RCW syste discharge results from the heat rejected to the RCW system via componets cooled by the sytem. The power dependent heat loads on the RCW y tem, that are increased by BPU, are those related to the operation of the RBCCW system, the condensate pumps, condensate boosie pumps, and the isolated phae bus duct air HX. The increase in RCW system discharge temperature from these sources due to EPU is < IF, which is minimal end within equipment tolnces.

6.4.5 Ultimate Heat Sink The UHS ia the Wheeler Reservoir/reunesseo River. The upstream tyeprature of the river is unaffectd by operations at EPU conditions. The existing UHS system provides a sufficient quantity of water at a tempeature.less than 95T (design temperre) to perform its safety-related functions for EPU, As discussed in Seotion 4.1, the serice waer CUHS) temperature assumed in thc DBA analyses was increased from 922F to 95R. Therefore, the TS for UHS lirnits zre changed to reflect tih new EPU alyses.

mh. UFA includes a dicussion relative to heatup of the downstream portion of the pool thda would exist following the loss of the downstream dam on dte Tennessee River. The river thennal rise post-shutdown would incres due to the increase in decay heat associated with EPT conditions but would not silfcantly affect iis event.

6-7

NEDO-33047 - Revidon 0

&.5 STANDBY LIQUID CONTROL SYSTEM Tho SLCS is designed to shut down th reactor ftom rated power conditions to cold shutdown in the postulated situation that sore or all of the control.rods cannot be iserted. This manually operated system pumps a highly ecriched sodium pntaborato solution into the vessel, to provide neutron absoipdon and achieve a subckitcl rctor condition. SLCS is designed to inject over a wide range of reactor operating pressures.

J) The TS minimum avadable volumo of sodium pentaborate solution associated wit this increase is bounded by the volurme requested in Reference . [

. U The boron injection rato requirement for the limiting ATWS event with SLCS injection, is not increased for BPU.

Tho SLCS is designed for injection at a maximum reactor pressure equal to the upper analytical limit for hiblowest group of MSRVs operating in the safety relief mode. For the EPU, the nominal reator dome pressuro and the MSRV skpoints are unchanged. Therefore, the capability of the SLCS to provide its backup shutdown fUnction is not affected by the EPOt The SLCS is not dependent upon any other MSRV operating modes.

Based on the results of the plant specific ATWS analysis, the maximu reactor lower plenum pressure following the limiting ATWS event reahed 1204 psig during the time the SLCS is analymd to be in operation. Consequently, there is a corresponding incrsIe in the maximum pump discharge pressure and a decreasr in the operating pressure margin for the pump discharge relief valvis. The operation of the pump dischargo system was analyzed to confimrm that the pump discharge relief valves rc-clie in the event that the system is initiated before the fime that the reactor pressure recovers from the first transient peak. The evaluation compared the calculated maximam.reactor pressure needed fbr the pump discharge relief valves to re-close with the tower reactor pressure expected during the time the MSRVs are cycling open and closed prior to the time when rated SLCS injection ft assuned in the ATWS anlysis. Consideation, was also giYen to system flow, head losses for filll injection, and cyclic prWsso pulsations due to the positive displacement pump operation in detenilning thc pressure margin to the opening set point for to pump dlsoharge roliefvalves. Tho pwup discharge relletvalves are peoiodically 6-8

- .7. . 'C . -. .

NED-33047 - Rwvidoa O tasted to maintain this tolerance. Therefore, thc current SLCS procmes parameters associated with 1ho minmum boron injectionrate are not changed The valuation shows tha EPU has no adverse effect on the ability of the SLCS to mitigate in ATWS event.

6.6 POWER DEPNDENT LIVAC The HVAC systems consist mainly of heating, cooling supply, haust, and recirulation units in the turbine building, reactor building, and the drywell. EPIU results in ulightly highs pross temperaures and s=lI increases in the beat load duo to higher electrical currnt; in sore motors and cables.

The affected reas are the drywall; the steam tunnel and the BcOS rooms in the reactor building; and the FW heater bay, condenser, and the condensatelPW pump areas in the turbine building.

Other arem in the reactor building and the turbine building are unaffected by th EPJU because the process temperatures remain relatively constant The increased heat loads during norma plant operation result in c 0.50 F increase in the drywall, tho MS hunel, end ECCS rooms frma increased decay heat In the turbine building, the

• maimum teperature inorese in the FY heater bay, comdeasat/W puny ares, and condenser are is < 2.

Based on a reviow of design basis doctnients, thc design of the HVAC is adequate for the BPU with the exception of the condensate and condensate booster pump motor coolers. Replacement of or modification to these pump motors, described in Section 7.4; may require modifications to their coolers.

6.7 FIREPROTECTION This ectiSon addresses th effct of EPU on the fire protection program, fire suppression and detection systms, and reactor and containment system respones to postulated 10 CPR50 Appendix R fire events.

1 ]JAny changes in physical plant configuration or combustible loading as a result of modifications to implement theo fU, will be evaluated in accDrdace With the plant modification and fire protection programs. The safe shutdown systems and equipment used to achieve and maintain cold shutdown conditions do not ebange, and wre adequate for the EPU conditions. The soop of operator actions required to mitigate the consequence; of a fire ar not affected. Therefore, the fi= proteSon systwus and analyses ar not afibod by EPU.

The reactor and containment responses to the postulated IO CFR 5O Appendix R fire event at EPU conditions ae evaluated in Section 6.7.1. Tbe results show that the peak fuel cladding temprature, reactor pressure, and containmnmnt. pressures and temperatures are below the acocetance limits and demonstrate that there is sufficient time avaulbila for the opertors to 6-9

NEDO-33047- eviutan D perfom the necessary actions to chieve nd main cold shutdown oondition, Therefor, the fire protection systems and analyes are not adversely affectod by EPU.

6,7.1 10 CF50Appead tR Fre Evmt A plant-speciec evaluation was peformed to demonstrate safe shutdown capability in compliance with the requirements of IO CPR 50 Appendix R as=uming BFPU vonditios. The limiting Appendix R fire evet from the current analysis was reanalyzed asuing EPU. The fiul heatup anlysis was perfrmed using the SAER/GEGSTRLOCCA analysis model. The containment analysis was perfonned using the SHEX model. Justification for usin SAFPBRGBSTR-LOCA md SHEX models for EPU calculations is procented in Section 4.

Thee are the name analysis rrrthqdologics that were used for the axisting Appendix R Fire event nyalyis. This evaluation determined the effect of EPU on flel cladding Integrity, reactor vessel integrity, and containmt integifty as a result ofthe fire event The postulated Appendix fire event uing the minimum SSDS was analyzed for the three cases described below:

Cas 1: No spurious operation of plant equipment occurs and the operator initiates threo MSRs 25 minutes into the event. -

Case 2: OQu MSRV opens immediately duo to a spurious opening sinal generat as a result of the fire. The MSRV is reclosed 10 minutes into the event by operator action. The operator initiates three MSRVs 20 minutes into the event, Canc 3: Ono MSRV opens immediately as in Case 2, but remnains open throughout the event.

Tho operator initiates three MSRVs 20 minutes Into the event The above are the sarne cases as those described in the Browns Pery Fire Protection Report (aeferenoe 2) excoept as described below.

These cases were evaluated for BPU with some reduction in conservats In the analytical assessweat.

Fcr the pr-EPU analyses, for all cases it was conservatively assumed that the LPCI in'ection does mot ocr until reactor pressure is S 200 ps, instead of the standard injection point of 319.5 paig, which delays LPCI injection into the vessel. For thz BPU assessment the analysis is based on the reactor vessal presaure reaching 385 paig, and thea the LPCI injection valve is opened by operator action. LPCI flow to thevesuel begins at 319.5 paig. Thin DdjUusfrnThto the analysis does not affect any operator soteon because the curreat procedures dirot the operations stafltopen the LP nJoction valve when RPV pressure is 5 450 psig.

The bounding PCr cme i Cue 1. For this cGae, time availablo to iheopeor t opentree MSRVs is 25 mninutes at the EPU conditions. The pre-BPU analysis dctermined the three MESRB were required to be opened *within30 minutes. This reduction in the time evallable does not have my effect because the current procedures require is action to be completed within 20 610

. - .. --- -r- :. - - ... - - . .: .. ..I- . ...

NEDO-33047 -Revbloa.O minutes. For CLTP mnd BPU, the PM ae calculated using consnmntivv LPCI performance characteristics (.g, min flow rats as fiunctions of vessel pressure).

En addition, spurious opuration of the HPCI system was reviewed in acrdance with Rcference

2. The HPCI systen was assumed to initiste at the onset of the Appendix R event, and flow at its nominal flow rat. The time for the reactor vessel water level to reach the MSLs is greater than 6 minute. Therefore, plant procedures will require HPCI isolation prior to 6 minutes during an Appendlx R event.

The results of the Appendix R evaluation for EPU provided in Table 6-5 demonatrate that the fuel cladding integrity, ractor vessel integrity, and containment integrity are maintained and that sufficient time is available for the operator to perform the necessary actions. The current exemption for the omentary core uncovery during depressurization remains necessary for EPU.

BPU does not affect mny other exemplios.described in Refrrece 2. No changes are necessary to the equipment required for safe shutdown for the Appendix R event. Onz train of systems remains available to achieve and maintain snfc shutdown conditions from either the main control roor or the rcmote shutdown panel. Thcrefore, EPU hau no adverse effect on the ability of the systems and personatl to mitigate the effects of an Appendix R fire event, and satisfies the requirements of Appendix R with respect to achieving and maintaining safe shutdown in the event of a fire 6,8 SYSTEMSNOTIMPACrED BY EXTENDED POWERUPRATE 6.8.1 Systems With No Jmnpct Sismir to the systems listed in Table J-1 of ELTRI (Referec 3), the systes in Table 6.6 are not affected by opezrion oftho plant at the EPU power level.

6.8.2 System. With InsignificantImpact The v ystem anfcted in vcyminor way bycperaion of te plant a the uprated power level are listed in Table 6.7. This listing is similar to te systeas listed in Table J-2 of ETRI. For these systm, the effocts ofEPU aro isignificantwithrespecto their deaiwi and operation, 6.9. REFERENCES

1. TVA Letter, 'Srowns Perry Nuclear Plant (BFN) - Units 1. 2. and 3 - License Amendment

- Alternatve Source Term," datd July 31, 2002, ROP020731 649, including Tech. Spec.

No. 405 (TVA-BFN-TS-405).

2. Tennessee Valley Authority, "Fire Protection Report," Vol. 1, Revision 16, January 2001.

3. GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate," (TEIR 1), Lioensing Topical Reports NEDC-32424P-A, Class m (Proprietary), Februay 1999; and NEDO-32424, Class I (Non-proprietary), April 1995.

611

Z . 1. I. I ...

NEDO-33047 - R°vion 0 Table 6-1 Browns Ferry RPU Plant ElectrIcal Chareteristies Parameter Value Guuanteed GCeneator Output (MWc) 1265 Rated Voltage O(V) 22.0 Power Factor 0.98 Guanteed Genrator Output (MVA) 1280 0___ntouW __ __ _ . 33.591 Isolated PhAs Bus DucttRadtng A) 36.740 Main Trunumiers Ruling (MVA) U2IU3 I h00M1344 Transfor Output MA) 125.

(1) 1280 MVA Gentor rating - 30MVA Staton Load Table 6-2 Brownms Ferry Oflflte Electric Power System Component Rating EPU Output GOnerator (MVA) 1280 1280 Isolated Phseo Bus Duct (k) 36.740 33.591 Main Traneformers (NIVA) 1500 (Unit 2) 1250 )

1344 (Unit 3)

.AfuliutyTrmnzfonner (MVA) 72 )

Switchyard (imitng) (MVA) 1750(& 1250')

1. 1280 MVAprojected ultinate Unit2/3 gensratoratingg-3OMVA Station Load

2. Two auxiliary transformers rated 40 MVA mnd 32 MVA.

3. Detzmineduiing actual plant datawith estimated aMtional leading due to EPU.

4. Seven SOD kV lines each rated at 1750 MVA.

6-12

-__ . . ; - w NEDO-33047- Ruvlalox 0 Table 643 Browns Ferry Sp ent Fuel Pool Pnrametcm Condilons J Parameter DAi Limiting Ful Limit te QMfad ConfigurationIlt One train each ofPPCC and ADHR in service o]

Peak SPP Tenpraturoe (F) 99.1 121.5 125 (a to) 150 (Full Core)

Time to Peak SFP Temperature (hr) so 109 NA Time to boil from los of all oowling at 14 5 NA peak temperature (hr)

Boil off rate (gpm) 4B 104 150 Configuratlan 2:

One train each of FPCC and RHR supplementaI fuel pool cooling mode in service Peak SP? Temperature (I) 124.9 149.8 125 (Batch)

I50 (Full Core)

TIl to Peak SFP Temperature (lir) 13io 22901 NA Time to boil from loss of all oooling at 13 4 NA peac temperatre (h)

Boil off rate (gpm) 42 80 150

1. Adguncs core offload begins 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after racwtor shutdown to allow for cooldown, vessel head removal, refileing cavity fill1, and other rofellng proparations.

2. Assumes core offload bogins 95 hour0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br />s- after reactor shutdown and includes 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of in-vossel stay time boauso the RHR suppkimntal fuel pool cooling mode l8 less heat removal capacity than the ADHR systen and SO hours to allow for cooldown, vessel head rornoval, refueling cavity fillng, and othr refueling preparation.

3. Assumes core offload begins 165 hous after reactor shutdown eand inoludes 115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br /> of in-vesed sty time because the RHR supplemental fuel pool cooling mode has less heat removal capacity tn the ADHR system and SOhoum to allow for cooldown, vessel head removal, refueling cavty fitIing, and other reueling prepaxmrin.

6-13

r - ..... ....... ,, _, _

NED043047 - tevldon 0 Table 64 Brawn. Ferry Effluent Dhchzrge Comparivon Parameter State Mwdmum EPN)

Lit Current' Flow (mllion gallonrdawy) Nonw 1008 No Change Downstream Tenperature 24-hour avg. (F) 9O.0 90.0 No Change DcwnvsiemTemperabire 1-hour avg, (OF) 93.0 93.0 No Change In-stream AT. 24-hour avg (OF) 10.0 8 <10 Chlorine (average/day) (mgLfper day) 0.064 Notdoetablc No Change Netheat addition (MBTIJhr) None 8128 9284

• Most consevative. value for each observed parameter, and does nt represent ccmourrendy observed conditions.

6-14

rr-J'--..- Bl NEDOw33047 -Reyision 0 Table 6-5 BrownR Ferry Appendix R Fre Event Evaluation Reualts Parameter . TP EPU App, R Criteria Claddig HeoaPp (POT), bf 1485 1428 s1 S0U PrimaySyutcmPresure,psig 1150 _- 1150 1375 Primary Contminm Pressuro, psig 18,6 13.6 !56 Suppreseion Poo! Bulk Tempeaturc, 'F 212 227 c27 S22'7M NPSH (C) Yes Yes Adequate for system using iuppression pool water sourme

1. NPSH dezonsfrued adequate, ice Seodon 4.2 5.

2. ContaInent structe design firnit

3. Tozus attached piping limit.

6-15

-*i.*--***r'---- -- . I.,@< -

NEDO-33D47 - Revion O Table 6-6 Browns Ferry Syitami With No Impact smdem unikat 012 Auxliary 1Boiler 018 Fuel Oil 020 Lubricating Oil 027 Cond Circulating WOateruTbe Cleaning 028 Water Treatment 029 Potable Water 032 Control Air 033 Servico Air 034 Vacuum prii 036 Auxillary Boiler FW Secondary Triccent 037 Gland Seal Water 038 ingulating Oil 040 Drainage 043 Sampling and Watcr Quality 044 Building Hcatng 049 Breathing Air 050 Raw Wate Chemrnical Treatment 051 Chernical fteatroent 052 Seismi; Monitorin 056 Temperature Montoring 058 BiothemialFacility 064C Secondary Cantainment 079 Fuel Handling and Storage 086 Diceol Generc Starting Air 099 Reactor Pratection I11 112 Shop Equipment 200 200 Series display boards 244 Communication 247 24D VAC LijMing Systcm 258 Opertions Recorder 260 Security 301 Sewage Disposal 302 Bovators 315 Microwave 327 Flood Proteti n 417 Meteorological Tower 6S16 L

• o*- r ea . **.~O.Io . - . *- .- .

NEDG-33847 lRtton 0 Table &7 Browns Ferry Syptms With Inignlficant Impact Sytem Number fnaIf Tia 004 Hydmogn Water Chemistry 009 control BayPnels 030 Nonalw Ventilation 053 Demineralize Backwash Air 0S, Annunciator 080 Drywell Tomperawnr Monitoring 925 25 Series Panels (Local Panels) 6&17

NEDO.33047 -Re#vion 0

7. POWER CONERSION SYBTEMS 7.1 TURBINE-GENERATOR The turbine wd generator was otiginaly designod with a maximum flow-passing capability and generator output in exoess of rated conditions to ensuro that the original rated stemm-pasaing capability and generator output Is achieved. This excess design capacity ensures tha the Wrbitne and geerator meot rated conditions for continuous operating capability with allowanc for variations in flow coefficients from expected values, manufacturing tolerances, and other variables that may adverscly affect the flow-passing capability of the uwits. Tho differenc in the steam-passing capability between (be design condition and the rated condition is called thMe fow margin.

The nuine-genzr w originally dceined with a flow margin of 5%, however, for operation at CLT? the turbine was redesigned and nly operates wit a flow margin of approximat 3%.

The current rated thottle stem flow is 14.12 Mlh/br at a thtrottle preaur of 980 psia. The searaoris rated at 1280 MVA, which re6ults in a rated electical output (gross) of 1265 MWe at a power 2ctor of 0.98.

At BPU PIP? and reactor dome pressure of 1050 psia, the turbine operates at an increased rated throttle steam flow of 16.44 Mlh/hr and at a throttle pressu of 962 psia. To maintain control capability GE uses a minimum taget value of approximately 3% throttle flow ratio, with controllability confirmed by unit testing am described in Section 10.4. For operation atEPUE tho high pressure turbine has been redesigned with new diaphragms and buckets for at least the minimum target throttle flow margin, to increase its flow passing capability.

The expected environmental changes, Buch } diurna heating end cooling effects changing cycle efficiency, periodically require management of reactor power to remain within the generao rating. The required ariations reactor power do ndt approach the magnitude of change; periodically required for surveillmnce testing and rod pattern lignments and other occasional events requiring de-ruting, such as equipment out of service for maintenae A rotor missile analysis was performed at the EPU conditions based on the NRC-approved methodology in NUREG-1048, which applies to units with shrik-on wheels. Based on the calculated results of he mnissile analysis, the missile failure pmbability is acceptable.

The overspeed calculation compares the entrapped steamn energy contained within the turbine and the associated Piping, after the stop valves trip, and the sensitivity of the rotor traia for the capability of overspooding, The entrapped energy Increases slightly for tho BPU conditions.

hMe hurdware modification design and implementation proeces establishes the overspeed trip settinge to provid, protection for a.turbine trip.

7-1

NEDO43047 -RvlAlon 0 7.2 CONDENSER AND 8TAMJT AIREaCTOs Th condenser converts the stam discharged frm thectbine to water to provide a source for the condensate and FW systems. The SJAB renmove nnoondwnsable gases ftm tke condensr to improve ffimal performance.

The oandenser and SJAB functions are required for Dormal plant peration md are not safety-related.

The coadecers were evaluated for peffonnanc at EPU conditon, based on a mriimum cold water tmperarur of 90T and current circulating water system flow. Additional analysis at BPU conditions also determined the condenser back presnure would be below the 5" Hg. design limit, assumning cleanliness levels as low as 85%.

Due to the increase in condensate flow rate associated with }BPU conditions, the retention time of condensate in the condenser hotwell is slightly reduced to 1.7 minutes. Condenser hotwelI capacities and ievel instunientation are adequate for EPU conditions. Periodic eddy current testing and water chemistry monitoring ae performed to monitor the effects of EPU RIP operation on the condenser tubes.

The design of the condenser air removal system Is not adversely afeeted by BPU and no modification to the system is required. The physical size of the primary condenser and evecuation time are lhe main f*ctors in establishing the capabilities of the vacuumpumps. Theso pameters do not cbange. Because flow rates do not cb ,g;there is no change to the two-minute holdup time in the pump discharg line routed to the reactor building vent Stac. The design capacity of the SlEABs is not fected by BPU, because they were originally designed for operation at greater than warranted flows.

7.3 TURBINE STWAM BYPASS The Turbine Steam Bypass system provides a means of accommodading eXcess Ste=m gSerated during norrnal plant maneuvers and transients.

The trbino bypase valves weore initially arted for a total steas flow capacity of not less than 25% of tb original rated reactor steam flow, or -3.56 Mlb/hr. Bach of nino bypass valves is designed to pass a steam flow of -400,000 lbmfhr and does not change at EPU RIP. At EPU conditions, rated reactor stern flow is 16.44 Mlbfhr, resulting in a bypass capacity of 21.71% of EPU rated steam flow. Thc bypass c city at Browns Perry remain adequato for normsal operational flexibility at B3PU RTP.

The bypass capacity is used as an iwut to the reload analysis process for the evaluation of transsient events that credit the Turbine Steam Bypass System (se Section 9. 1).

7.4 FEEDWATER AND CONDENSATE SYSTEMS The FW and condensate aystau= do not perform a systen level safety-related fiotion, and are designed to provide a reliable supply of FW at the trpepatr; presure, quality, and flow rats aS required by the rectar. Therefore, these sstnsare not safety-related. However, dweir 7-2

NED(333047 - Revision 0 perfomc he a,moor effect on plant availability and capability to oprate at the EPU colditon.

Modiiaons to som nonafety-reated equipment in tie 7W and condenrte systems are necessay to ttain uI EPU core tral power. bipleentation of Ihesomodifications is reviewed per the site 10 CER 505.9 procs For DPU, the FW and condmsate systems mee the following perfnanc criteria with modifications to onte nonsafet-relead equipmt:

1. The system provide E reliable ripply of 7W at the EFt dome pressure wit sufficient capacity to supply the stady-stats MI flow demaded at the EPU condition.

2. The systes have the capacity to provide at leat 105% of the EPU FW flow. This ensures that Broswis Ferry remiins avilable during water level transient, avoids 6ms, and

• minizes challenges to plant safety sytems.

3. The FW system is capable of providing adequate MW flow Lt the expected operating pressura, and to provide umit trip avoidance when one 7W pump is tripped.

4. The rimaut capacity of the FW system in the limiting pump eimnnt does not exceed 1he perfcmanc cvaty assumed in the transient analy6es.

7.4.1 Normal Operation System operating flows at EPU inrease approximately 20%,l of rated flow at the OLTP. The condensate and F/W system will be modified to assure acceptable peformance wit the new syskte operating conditions.

The FW heatus will be analyzd and vtrifled to be acceptable for the higher PW heater flows, temprtturee, and pressures for the PIU, and renre prior to implementaelon of EPU. The perfornance ofThe FWhear will be monitored during the EPU power ascensionproaranL 7.4.2 Transient Operation To account for FW demand transients, the 7W system was evaluated to ensure that ELminimum of S%margin above Rho EPU FW flow was avallable, For system opera with ll sye pumps available, the predicted operating parametrs were acceptable and within the compone capabilities.

The FW sster post feed pump trip capacity was evaluated to confirm that with tle Modifications to the PW and Condensate system configurzatons, the capability to supply the tinnsient flow requirements is maintaed or inceased. A transient analysis was performed (soction 9.1.3) to determine the reacor level response following a single FW pump trip. 'he rAuts ofto suaalysis show that the syste response is adequate duming the EPU conditions.

7.4.3 Condensate Demlneralzenr The effect of EPU on the GEDs was reviewed. The system requires modification to support CFD IWI flow operation during backwasbk nd pro-coating without requiring a plant power reduction. The system experiences slightly higher Ioading; resulting in slightly reduced CFD 1-3

NEDO.33047 - Rniiloa 0 run time.. However, the reduced rm times am acoeable (refar to Section 8 for the effects an the mdwstt .yutez).

7-4

1- * * -~ -.

NEDa43047 -Revsidon 0

8. RADWASTE AND RADIATION SOURCES B. LIQUIM AND SOLID VASTEI MiANAGELMENT The Liquid and Solid Radwaste systems collects, monitors, processes, stores, and return.

pocssed radioactive waste to the plant for reuse or for discharge, The single largest source of liquid and wet solid waste i5 from the backwash of condensate demnineralizers. .EPU results in an increased flow rate through the condensate demineralizers, resulting in a reduction in the mverage time between bokwushes. This reduction does not affect plant safety. Sinilarly, the RWCU filfereineralizer requires more frequent boknashes due to hier levels of Impurities as a result ofthe inceased PW flow.

The floor drain collector subsystem and the waste collector subsystem both receive prodic inputs fom a variety of sources. EPU does not affect system operation or equipment performce. Therefore, neither subsystem is expected to experiecc a.large increase in the total valums of liquid and solid waste due to operation at EPU oonditions.

The incremod loading of soluble and Insoluble species increases the volume of the liquid processed wass by 4%and the volume of the solid processed wutes by 15%lo. The total volume of liquid and solid processedwaste does not signmfictly increase (as comparod to thc Radwastp.

System capacity) b=caue the only increase in processed wasto is due to morm fequent ibackwashesa of the condensate demineralizers and RWCU flter denineralzera. The total liquid and solid increase. ar within the Radwaste System capaci Therefore, EPU does not have an adveme effect on the processing of liquid and solid radwast, and there arc no significant enviroumrental effects.

The inreases in the liquid and the solid processed waste r based on the incse duo to the FW flow increase. The percentage bounding value for the inres in Liquid and solid procesed waste is equal to or less than that of the FW flow percentage increase.

8.2 GASEOUS NASTE MANAGEMENT The gaeous wast managenct systems coltec controL process, store, and dipose of gaseous radioactive wast gensated during nonnal operations. The gaseous wase manrtagent systms include the offgas sysen and various building ventilation systems. Th syskns are designed to meet tho rcirments of IOCFR 20 and 10 CFR SO, Appendix I, Non-condensable radioactivo gas frn the main condenser, along with air inleekc normally contains activation gases (principally N-16, 0-19 and N-13) and fission product radioactive noble gasue This isthe rmaor source of radioactive gas (greater ta al other sources coambned). These non-condewsable gases along with non-radioeolive air ire cotnuously removed forn the main condense by the RIA~s, which dishage into &a offgss s-sten Building ventilation system cnrol arboe radioactive gases by using combinstions of devices such as IPA nd charcoal filters, and radiation monitor.s at signal autonatic isolation damperm or B.I

NEDO-33047 - Revhon O tip supply and/or exhaust fans, or by maintaIning negative air prwqzre. where requpired, to linit migration of gases. ((

)) ThIs Il because the amount of fisson products released into the coolant dqcnds on te number and nate of the *el rod defitts, end Is approximately linear wi& respect to core thermal power. The concentration of coolant activuan products in the steam remans nerly constnt. The release limit il et administntively controlled varable, and is not a fiucton of core power. The gaseous effluents are well within limits at original power operation and remain well within limits following implementation of BPU. There re no sinificant environmenal cffects due to lPIU.

.2.1 Oftgas Syetem The primary function of the Offgas system is to process and control the release of gaseous radioactive effluents to the site environs so that the total radiation exposure of persons in offaite aroas is within the guideline values of 10 CPR 50, Appendix I.

The radiological release rate is administratvely controlled to remain within existing mfite release rats limits and is a fiuction of felo cladding performance, main condenser air inleakag, charcoal adaoxber inlet dew point, and chucoal adsorber temperature.

Becuse E3PU affects the flow rate of radiolytic hydrogen and oxygen to the Offigs System, tle catalytic recombiner temperature and offgas condenser heat load ar affeted. nTe Browns Ferry radiolytic decomposition rate is based upon Browns Ferry design specifcations adjusted forBPU power level. The EPU analysis for the Offgaz System utilized a bigher decomposition rats that in more conservative than the Browns Perry plant specifia decomposition rate. The BPU hydrogen flow ratu and concentrations are still within the design limits of the Offgas System. The catalytic rocombiner and offkas condenser, as well as downstream component, have sufficient design margin to handle the incroase in fthrmal power for EPU without exceeding the systen design limits of tenperature, flow rates, or beat loads.

In addition; HWC operation when used will cause a reducti= in cm radiolysis. The combination of the HWC injeced hydrogen plus the reduced radiolysis is expeoted to prodou a, lower net hydrogen flow to the Offgas System.

83 RADIATON SOURCES IN TH REACTOR CORE B.3.1 Normal Operation During power operation, the radiation soUrces in the care are directly related to the fission rate.

Tese sources include radiation ftom the fission process, accumulated fission products sand neutron reactions as a.seconday result of fission, Historically, these sources have been defined in tcrms of energy or activity released per unit of reactor powver. Th fsre, for EPU, the percent increase in the operating source terms Is no greathan the percent increase in power 5.2

.P 6 z. . . e e e . f -

NEDO-33047 - RevIdion L3.Z Normal Pout-Operation The post-operation radiation soUres in the o arc pririly the result of axunmlstedfissin products. Two separate fcrs of post-opecation source data ae normally applied. The fst of these is the core gamm-ray source, which is used in ibielding calculation; for the core and for individual fuel bundles. Thi ources term it defined in terms of MeV/sec per Wat of reacor thermal power (or equivalent) at varioug times after shutdown. The total gam enry source, thetefom, increasec in proportion to reactor power.

ne second set of post-operation source dat consists primarily of nuclide activity inventories for fission products in the fel, These data rc needed for post-accident end pent fuwl pool evaluations, which are performed in compliance ith regulatory guidance that applies different release a-d transpogr assunption; to different fission products. The core fision product inventories for these evaluations are based on an assumed fuel irradiation time, which develops "equiliuriurn" activities in the fuel (typically 3 years). Most radiologically significant fission products reach equilibrium within a 60-day period. [{

))The radionuclide invenaories are provided in terms of Curios per megawatt of reactor thermal power at various times after shutdown.

The rults of the plant specific radiation sources e= included in the LOCA, FM, and ORDA radiological analyse. presented in Section 9.2. Plant specific analyses for NUREG-0737, Item ll.B.2, post-accident mission dos"e havc ben prforne The results of this assessment Bre accounted for in the plant radiaion protection program.

8.4 RADIAfION SOURCES IN RXACrOR COOLANT Radiation sourmcs in the reactor coolant iclude activaiion products, activated corrosion products and fission productes 8.4.1 Coolant Activaton Producd During emoter operation, the coolant passing through the core region becones radiotive as a result of nuclear reactions. The coolant activation. especially 2.16 activity, is thc domniniant sourCe in tho turbine building and in the lower regions of the drywell. The activation of the water in the core egion is in approximate proportion to the increase in therma power.

1E

)J The activation products in the ateamJ from EIU arm bounded by th exiting deasi basis concentation. Th margin in the design basis for reaoctr coolnt activation concentrations significantly exceeds potential iocreases due 6-3

P

- . - .: - T .. ' . - . ... @ z NEDO-33047 - Reviden O to BPU. Thereore, no ohage is required in the activaticn design basis reacfnr coolant omncentratio for BPU.

.4.2 Activated Corrosion and FMion Products The reactor coolant contn activated corrosion products, which are the recult of metallic materials caring the water and being activated in the rator region. Under BPU conditions, the W flow ireases 'with power and the Activation ruts in the rmtor region increases with power. The netresult is an increase in the activated corrosion product production.

The total aetivatd corrosion product acivity is approximatly 3% higher then the original dcsign basis activity as a cansequence of EPU. Howver, the sum of the activated corrosion product activity and the fission product activity remain acsmall fration (C3%) of the total design bas activity.

FPsiion products in the reactor woolent ar separable Into the products in the steam and the products in the reactor water. The activity in the stemn consists of noble gases released from.te core plus can'yover ectivity ftom the reactor watr. This activity Is the noble gas offgas that is included in Browns Ferry design. The calculated offgas rates for BIU asfer thirty minutes decay are wel] below the original deaigj basis of 0.35 curies/secc Therfore, no change is required in the design basis for offlgs activity for EPU.

The fission product activity in th reaotor.water, like the activity io te steam, is ft result of minute releases ftom the fel rode. Fission product davity levels in the reactor water were calculated to be higher an previous calculated data, increasing £ 13% from current values due to BPU. These activity levels rematin a &action (<2%) of the design basis fissioa product activity. Terefore, the activated corrosion product and fission product activities design bases are unchanged for EPU.

Forte EPU, normal radiation sources increase slightly. Shielding aspects of Browns Perry %were conservatively designed for total normal radiation sources. Thus, the increase In radiation sources does not affect radiation zoning or shielding and plant radiation area procedural controls will compensate for Increased normal radiation sourccs.

8S RADIATION LEVELS 8.5.1 Normal Operation For EPU, normal operation radiation levels increase slightly. Por conservatisn, ray aspects of Browns Ferry were originally designed for higher-thanfexpected radiaton s=CCs hus, th Increase in radiation levels does not affct radiation zoning or shielding in the various areas of Browns Farny because it is oftset by conservatism in the original desga, source terms used and analytical techniques.

The nonmal operating doses am generally based on dose rate measuremts at various locations during plant operation at OLTP conditons. The normal doses specified for OM' conditions are Increased by 20% with the exoeption of four zones where additional data was available to 8-4

NEDO-33047 Revidson 0 demonstrae that the normal doses apeificd for these areas contained Dufflicint margin to account for a 20% Increase In the observed dose mta~ Th increaed normal radiation dose are evaluated and determine to have no advors. effecton safetyrelated plantequipment; u indica in Seotions 10.3.1 and 10.32, Individualwocerexpures can be maintained withinacceptablo limits by controlling ac css to radiation ares using th itst ALARA prowu Procedural controls compensate for inoed radiM=on levels. In addition, Browns Peny has pruviously implneated zin injection and noble metal chaeical addition to limit the inoean inno radiation l doses from the implementation ofHIWC, 8.5.2 Normial Post-Operafton Post-operation radiation levels in most areas of Browns Ferrya expected to inrurco by no more than the percentage increase in power leve.1 In a few areas ar the rator vwater piping and liquid radnsate equipment, the lncre could be slightly higher, Regardless, individual worker exposures can be maintained within acceptable limits by controlling access to radiation areas using the site ALARA program. Procedural controls compensate for increased radiation levels. Radiation measureents will be made at selected power levels to ensure the protection of personnel.

8.53 Pot Accident The 100-day post-accident radiation doses are expectd to increase by 12% or less at EPU RTP

&ornpared to the post-accident dooes for CLTP conditions. For some areas, the post-accident doses speoified for CLTP conditions are bounding for the EPU conditions. The increased post-accident radiation doses have no adverse effect on safety-related plant equipment aS indicated in Sections I0.3.1 and 10.3.2. Plant specific analyses for NUREBG0737, Item ll.B.Z,post-accident mission doses have been perfiomed.

Section 9.2 addresses the accident doses for the Main Control Room.

8.6 NORMAL OPERATION OFF-SITE DOSES Tho primary soures of normal operation of fsit doses are (I) airborne releases fromn the offgas system and (2) gamnma shine from Browns Ferry turbines..

The increase in activity levels is proportional to the percentage increae in core thermal pocer.

The TS limits implorntt tho guidelines of 10 CFR 50, Appndix L EPU does not involve significant increases in the offuito dose from noble gases, airborne particulates, iodine, tritium, or liquid efflunts. Present offeite radiation levels fonnr a ngligibla portion of background radiation. Thereibre) the mnoal of ibe dsos 8ro net significantly affected by peation atEPU ad moainbelowcdoelhnioflOCER20, WOCFRSOAppendixL and 4OCFR 190.

Browns Perry has implemented zinc jection and noble mete] chemical additon to limit the increase in normal radiation doses from the irmplementation of HWC. The EPU increase in

• stoa flow results in higher levels of N-16 and other activation producta in the tubines. The increased flow rate and velocity, which rezult in shorter travel times to the twtiea and lass radloactive decay in trunsit, lead to higher radiation levels in and aound the turbines and offtito 8-5

- .- I' . .

NEDO-33047 - ReniNlod a ikyshine dose, Any disernblo inavao in radiafion e astemat of increased N-16 would be measured on thz site environmel ILD staions. Past history from these TI) utations for the bnplemenbaion of HWC and the 5%power increase has t shown any disceible Increase in raditon at offaite locations. Therfore, it is unlkely thatthe inucrease in N-16 source term duo to EPU results in any measumrble dose to the public.

8*6

. - : . . r ._ ___ _ _

gr NEDO-33047 - Revidon O D. REACTOR SAFETY PERFORMANCE EVALUATIONS 9.1 REACTOR TRANSIENTS The UPSAAR evaluate. the effics of a wide range of potential plant AOOs, commnonly referred to as transients. Disturbancs to Browns Ferry caused by a malfunction, Bsingle equipment failure or an operat e=ror art investigated acoording to the type of initiaing event per Chper 14 of the Browns Feny UFSAR Appendix B of BLTRI (Reference 1)idontifies the limiting events to be considered in each category of events. The generio guidelines also identify (he analytical metods, tho operating conditions tat are to be sssumed, and the criteria that are to be applied.

Thc following pamrgrphs address each of the limiting events and provide a murnmary of the rosulting traientsafety analysis. The result give here arcfor a representativ core, and show the overall capability of the design to meet all transient safety criteria for RPU operation.

Table B-1 of ELTRI provides the speiflo events to be analyzed fbr EPrU, the power level to be assumed, and the computer models to be used. The UnIents that are not listed in Table B- are generally milder versions of the, analyz events.

I]

The reactor operating conditions thft apply moat directly to the transient analysis are summarized in Table 9-I. They ore compared to the conditions used for the UPSAR analyses and a typical Unit 2 reload core analysis. An equilibrium care of GB14 fuel was used as the representive fuel cycle for the EPU transient analyses. Most of the transient events are analyzed at the full power and maximum allowed core flow operating point on the powcr/flow map, shown in Figure 2-1. Direct or statistical allowance for 2%V power uncetainty is included in the analysis. ((

JJ The SLMCPR in Table 9-1 Was used to calculate The OLMCPR value(e) required for the analyzed events. For all pertinent cvvnts, one MSRV is considered to be 00S, and the MSRV .tEoin+/- tolence is considered to be +314. A discussion of other equipment OOS options ib provided in Section 1.3.2.

Tc transient events of each category fom Table B-1 of BLTRI were nalyzrd Their iWnps Knd results revise the licensing basis for the transient analysis to the BPU RTP. The ovnpressurlratlon analysis is provided in Section 32. Otler transient analysis results for the full EPU RTP condition are provided in Table 9-2. The most limiting transient event results are shown in Figures 9-1 through -4, A. shown in tO table and figures, no change to the basic clmuaoterisics'of any of the liifing events is caused by BPU.

The severity of transients at less than rated power arm not significantly affected by the EPU, because of the protection provided by the ARTS power and flow dependent limits.

The historical 25% of RTP value for the TS Safety Limnit, some thermal limits monitoring LCOs rold, and somc ER; twholds is bWsed on (( )) analysce (cvaluated up to '-0%of original RTP) awlicable to the plant design ,ith highest average bundle power (the BWR6) for all of the SWR1Vproduct lines, As originally lioensed, the higlest sveraeo bundle power (at 100%

95-

NEDO-33047 -Rcvklou 0 RTP) for my BWR6 is 4.8 MWtbundlz, The 25% RT? vul is iLaoneervava buis, as described in the TS, [t 11 9.1.1 Fuel Thermal MargIn Events 1]

9.1.2 Power and flw Depeudt LimIts The operating limit MCPR, LHGR1, sndior MAPLHOR thermal limits ase modified by a,flow frtn whzn Browns Ferry is operating at less than 100%/a core flow. This flow fador it primarily based upon an ervaluation of the slow red ulation iise, uvet ((

Simnilarly, the th-nnal limits are modified by a power factor %tenBrowns Pery is operating at lesa than 100%D power. ((

92

NEDO-33047 - RevYs3o1 0 9.1.3 Lou .ffedwaterFlowEvent For the LOFW event, adequate nent core cooling is provided by mzninaining the water level inside the core shroud above the TAF. A plant specific analysis was peforned at EPU codition.. This analysis asued failure of th HPCI wystem and 'used only the RCIC system to restore the reactor water level. Becauso ofthe extra decay heet from EPU, slightly more time Is required for the automatic systems to reetore water level. Operator action is only needed for long-tern plant shutdown. The results of the LOEW analysis show that the minimum war level inside the shroud is 58 inches above the TAF at EPU conditions, After the water level is restored, the operator manually controls the water level, reduces reactor pressure, and initiats RHR shutdown cooling. This sequence of events does not require any new operator actions or shorter operator response times. Therefore, the operator actions for a LOPW transient do not significantly change for BPU.

As discossed in Section 3.8, an operational requirement is that the RCIC system restores the reactor water level while avoiding ADS timer initiation and MSIV closurc activation functions associated with the low-low-low reactor water level setpoint (Level 1). This requiremet Is intended to avoid unnecossary initiations of those safety systems, and is not a safety-related fcmti.on The results of the LOFW analysis for Browns Feny show that the nominal Level I setpoint trip is avoided.

The loss of one MWplnmp event only addresses operational consideations to avoid reactor scram on low reactor water level (Level 3). Tis requiremnet is intended to avoid unnecessary reactor shutdowns. Because the MBLLLA region is extended along the existing upper boundary to the EPU RTP, there isno increase in highest flow control line for EPU. Therefore, the results of the loss of one FW pump cvnt are insignifcant. A plant-spedflc evaluationcfis that the level it maintained above Level 3.

9.2 DESIGN BASIS ACCIDENTS This section addresses the radiological consequences of DBAs.

Plat pecific radiological dose consequence analyses havc been performed for thc DBAs at EPU conditions utilizing AST in accordance with 10 ClR 50.67. The results of thema analyses for the LOCA, the CRDA , the FHA, and the MSLB are provided in the AST license amendment submittal (Reference 3). 1Th calculated doses remain within applicable regulatory aceptance criteria.

The ILBA analysis was also pwrformed at EPU conditions utilizing AST. The radiological consequences of this event remain bounded by the other postuiated line brekt.

943

.V-. . ..-- . .. .----.- - , -:-

Vw... v- ~. - .. . . -r : -

NED033047 - Revison 0 9.3 SPECIALSEVTS 9.3,1 Anipated Transient Witbout Scram The overpresuure evaluation includes consideration of the most limiting RPV overpressure casc, 1[

A,LOOP does not result in a reduction in the RHR SPC capability relative to the MSIVC Wd PRPO cases. With the m1e, RHR SPC capability, fth containment response for the MSIVC and MRO cases bound Om LOOP case.

Browns Perry mets the ATWS mitigation requirements defined in 10 CFR 50.62:

a) Istallatioa of an ARI system; b) Boron injecion equivalent to 86 gpm; and c) Installation of antomatic RPIlogic (i..,ATWS-RPT).

In addition, plant-specific ATNVS analysis is pformned to ensure that the followg ATWS acceptance criteria are met a) Peak vessel bottom pressur. less than ASME Service Level C limit of 1500 png; b) Peak suppression pool temperature less than 28 IT (Wetwell shell design temperature); end c) Peak containent pressur less than 56 psig (Drywell design pressure).

Te limiting events for the acceptance criteria discused above are tIho PRFO event and the MSIVC event The ATWS analyses have been peormed for 105% OLIP and for BPU RTP to dmostrt the effect oftht fPU on the ATWS deeprance criteria. There Is no chosge tothe assumed operator actions for the BPU AWS analysis, and thre is no change to the required hot shutdomn boron weight The key inputs to the Browns Ferry ArWS analysis ra provided in Table 9-3.

The andlysis wu performed using the ODYN code. The results of th& analysis are provided in Table 9-4.

The results of the ATWVS analysis meet the above ATWS acceptance criteria. Therefore, the response to an ATWS event atBPU is acceptable.

CoolIble core geometry is ensured by meeting the 22000 peak cladding ten rusre and the 17% local oladding oxidation acceptance criteria of I0 CER 5046d ((

9.4

.1* - . ..  :

NEDO-33047 - Reviion. 0 U

9.3.1.1 ATWS with Core Instablity The effects of an ATWS with core instability event occur at natural circuIBtiOn folloWing a rccirculation pump trip. It is initiated at approxirnatoly tOe amnc power level as before EPU, because the MELLLA upper boundary is not increased, The coro design necessary to aclieve SW operations may affect the susceptibility to coupled thermal-bydraulic/neutronic core oscillations at the natural cculadon condition, but would not significantly affect the event progression.

Several fretors ffect the response of an ATWS instability event inluding operating power and flow conditions and core design, The limiig AWS core, instability evaluation presented in Refiernces 4 and 5 was performed for an assumed plant initially operating Ea OLTP and the MELLLA minimum flow point. R D BFU allows plants to increase their operating thenmal power but does not allow an incroase in control rod line, (

1] The conclusion of Reference 5 nd the associated NRC SBR that the analyzed operator actions effectively nitigate an ATWS instability event are applicable to the operating conditions expectcd for EPU.

ntiaI operating conditions of FWHOOS and FFWTR do not significantly affect the ATWS inability response reported in References 4 and 5. The limiting ATWS evaluadom assumes that all EW heating is lost during the event and the injected FrW temperature approaches the lonvest achievable main condenser hot well tomperaure. Th minimum condeet hot well temperature is nOt affected by FWHOOS or F<WR. Thus, as compared to the event initiated from a norma1 FW temperature condition, the event initiated from either the PWHOOS or FFWTR condition would have less modeator reactivity inertion band on a smaUlr temperture difference between the initial and fina FW temperstures. Theref the power oscillation for FWHOOS or mvmTR is expectd to be no worse than for the nomal temperate condition.

9.3.2 Station Blackout SBO was reevaluated using the guidelines of NtMARC 87-00. The plant reonse to and cping capabilies for an SBO event ar affected slightly by opertion at EPU RTP, due to the Increase ia the initial power level and deay heat. Decay heat was conwvrtively-evauated 9-5

I@ . I NEDO.33047 -Reyislon 0 mnung end-of-cycle (24.month) and GE14 fuel. There re no changoes to the systems and equipment used to respond to an SRO, nor is the required coping tim cbage&.

Areas containing equipmet necossiry to copc with an SBO cevot were tvaluated for thee effect of loss-of-ventilation due to an SBO. The evaluation shos tat equipment operability is bounded duo to conservatihm in thz rxisttng desagn and qualification bases. Th battery capacity remains adequate to support HPCRCIC operation afQer EFU. Adeuate compressed gas capacity exists to support the MSRV actuations.

The current CST inventory reserve (135,000 gal.), for HPCIVRCIC uso, ensures that adequate water volume is aveilable to remove decay heal, doprosmurize tb reactor, and raintain reecior vessel level above the top of active fhat (approximately [22,000 gal, required). Peak containmrnt pressure and temperature remain within design bases. Consistent with the DBA-LOCA condition, tbe required NPSI nagin for the RFf pumps has been evaluated (sia Section 4.2.5) end a component acceptability revewbag been completed (see Section 3.9).

Based on the above evaluations, Browns Ferry continues to meet the requirments of 10 CFR 50.63 after tbeBPU.

9.4 REFERENCES

1. GE Nvolear Energy, "Geneio Guidelines for General Electric Boiling Waler Reector Extended Power Uprate," (ELTRI), Licensing Ibpical Reports NEDC-32424P-A, Class 111 (Proprietay), Februazy 1999; and NEDD-32424, Class I (Non-pxoprietary), April 1995.

2. GE Nuclear Fnergy, "Generic Evaluations of General Bleockic Boiling Water Rator Exteded Power Upat;"(ELTR2), Licensing Topical Reports NEDC-32523P-A, Class lf, 1?ebruary 2000; NBDC-32323P-A, Supplement I Volume I, February 1999; and SupplmRent I Volume II, April 1999.

3. InA Letter, "Browns Ferry Nuclear Plant (3N)- Units 1, 2, and 3 - License Amendinnt -

Alternative Source Term," dated July 31, 2002, ROB 020731 649, including Tech. Spec. No.

405 CIVA-BFN-TS405).

4. GB NtroarC Energy, "Qualifiction oftheOne-Dimenaional Core Transient Model (ODYN) for flailing Water Reactors (Supplement 1 - Volume 4)," LicenSing Topical Report NEDC-24154P.A, Revigion I, Suppleent 1, Class 1H, February 2000.

5. GE Nuclear Energy, "Mitigation of BWR Core Thermal-Hydraulic Inmtabilities in ATWS,"

NEDO-32164, Decembe 1992-9-6

NEDO.33047 . Re~ion 0 Tabl*9-1 Brown, Fery Parameters Used for Tmnslent Analysis Parameter Cycle EPU Aulyis RAW ThermDa Powe (MWL) 3458 3952 Analsis Power (%Rated) 100110J2 100/1D22 Antlysis Dom Preure (psi6) 1050 1050 Analysis Turbine Pressure Cysix) 974 ' 9734 Rated Wave StemFlow(Ovblar) 14.150 16A40 Analyis Stesm Flow (5 Rsted) 1.000 10.0 Rtsd Core Flow (Mlbhzr) 102.5 102.5 Rated Fower Ccr FlowRangem 1-105 99-105

(%Rid) _

Analysis Core Flows(Mblb) 107.6 107.6 NormalPWTapectue 381.7 394.5 FW TemruReducwdon (AT ¶') 54.7 54.7 Steam Bypas Capacity 25.2 21.7 (V Rated Stnmflow)

ReactorLowWaterLeven3ScSnC S18 SIB (inches above vessel raw) _____ ______

NumberofMSRVs anued inthe 12 12 anlylis.

MC:R Ssfty Limit* 1.08 1.08 I Unit 2 Reload 12 (Cyci 13) results provided far coniparion.

2 GEMII analymls at 100%, RBDY anaysBs at 102%.

3 Reload analysIs value based on currat vneasured team lino pzesure drop.

4 EPU Iput value represets the conservatve value (lowest sftmlino pressure loss for any unit).

5 All analys at maximum win flow u nexplicitly noted otherwis.

6 A Ic-pressur bunk aelpofr MSRV is usumid OOS for Wrient nalysis.

9-7

.. w.*----- I ...

NEDO-33047 - Revkon 0 Table 9-2 Brows Ferry Trandent AnM it RenIt MCPR OpertlngUamf Pek Peak HAEl Event Neutron Flux (% of. ACPM Opao A Optlonm Flux (% of Ratkd EPU)

Rated SPU)

Generator Load Rcjecfion with 680 133 0,26 I.42 1.39 Bypasm FeIure _. . _

Turbine Trip With Byp as Failure 673 132 0.26 1.42 1.39 FWController Fallure MaxDem d 629 136 0.26 1.41 133B FWControllerFailuro Max Dm=ad 742 141 0.31 IA7 1.44

. mio os _ __ _ _ _ _ _ _ _

Pressure Regulator Downsie (1) (1) (1) (1) (1)

Faigure Lonof rWlati (2) (2) 0.13 I.!

InadvalfontHC [Actuation 112 109 0.06 (3) l (0)

Rod WitldmwalError (2) (2) 0.19(i - 127 Slow Recirculalio In=rcse (5) (S) (5) MCPRe PautRecirculation Incnum 181 94 0.14 (6) (6)

Ganeator Load Reectiokm with 590 129 0.22 137 1,34 Bypass MSIV Clonin - AII Valves 123 100 0.03 C7) (7)

MSIVCoMrM - One Vslve 130 106 0.06 (7) (7)

LostofPWFlow 100 .o _ (S) (5) (S)

Lon of OneFA Pvmp 100 100 (5) (5) (5)

1. Not required based ontUSAR 145.2.8,

2. Peaknutron Gfux andpeakbeatfluxaro notreportd (orthelow (nsimens.

3. HPCI sboudd byLon ofFWHeafing.

4. With rod block monitor 5etpointof l l f.

S. Notapliodblo.

6. Fat reofrculation inmsc It bounded by off-rned IlMitb.

7. Boundedby the UcneutorLoadRRectciowithBypus PaFlule 9-8

g- . o NEDO.33047 -Rodslon Q Table 9.3 Brownt Ferry Key Inpats for ATWS Analyst Input Variable CLUP EPU Reaotor power MWt) . 3458 3952 Reactor dome prensure (psi) 1050 1050 MSRV capacity (13 valves) (Mbr) 11.31 11.31 Highpro9iureATWS-RPr(p!Sig)

• 1177 1177 Number ofMSRVs Out-of-servloe (ODS) '1 1 Table 9-4 Browas Ferry ATWS AnslysIs Result#

Parameters CLTp EPU Peak vmse1 bottom pressur= (puSg) 1368 1484 Peak suppresaion pool tzmperatur CF) 214.6 214.1 Peak oontalamnt pMSsurc (psig) 21.7 21A Peak uladding terpeture (IF) 1476 1453 Local cladding oxidation(%) <17 <17 9.9

NEDO.33047 - Rcrluon D EU 4 -4,

-e-lwuidwh Pha mm

____M~W~

_CWWR

%a I

u rm s is e *U U rU U U mnow W~ to ~ to v. W mu - , flw" 14A me flit

.K, 4-",

,. 4. , ,. 44 a U4 ,.

WPeQNdl Figure 9-1 Browns Ferry Turbine Trip with Bypass Failure

(@ 10OO E RT and 105% Core Flow) 9-10

. .. " - "I- ,.! - --- -- - -- - - - - .. --. --

NE1DO-33047 - RevIhon 0 muu 04 Mni f PA Il Eu aMe U d sU A Ad Uor AFlow) ro P1 i I

Fiue 9-2 Brownus FirrtyGenerator Load R.J0cflon wlth Bypara Failure

((@ 100% EPU RiT and 105% Core Flow) 9g11

NED0433047- RnvWon U mi

-Oonilu mm

-Carb iMk

_.. .*=. A en SM l j D £4 3d aL no

'I' to 4:- -

14 1;

._ I- -

• 11

4. _

O. *A *4 MW SW

'a'.-,"

Figure 9-3 Brown FerryFeedwater Controlier Fallure - Mnimum Demand

(@ OO% mU RTP, 105% Coro Flow nd 394TFW rcp.)

9-12

NEDO-33047- Rzyldon n we,-

_1>D" hIel Ohjft

-&-Nmotp U.

mun

_. = _IL so Ii. .

Us (lla no Su TW Pa UMOW FOsure 9-4 Browns Ferry Feedwatgr Controller Falfure -Maximunt Demnitd with Bypass 00S

(@ OOO/o EPU RTP, 105% Core Flow azd 394 STFW Tenp.)

9-13

- X NEDO 33047- Rrvioa 0

10. OTHER EVALUATIONS 10.1 HIGHENERNGYYLNEBREAK HlB s arm evaluated for their effecte om EQ. Operation at the EPU level requires an increase im the taam and FW flows, which results in a slight inorease in downomor subcooliQt. This, in turn, result in L sa11 icrCas, in the man and energy release rates fllowing L HELH.

Evaluation of those piping systems detemined that There i9 no change in postulated break locations. The EPU affects oa the HELB mass and energy releases are documented in Table 10-I.

10.1.1 Temperature, Prmusrend Humldity Froflles The HELB analysi3 evaluation was made for all systems evaluated in the UPSAR. The equipment and systems that support a safety-related ftmtion are also qualified for the environeontal conditions imposed upon them, At the BPU power lovo, some of the mase and cnergy release for HELBs outside the, primary contaniment increase, potentially causing the eubcompartment pressure and temperature profiles to incroase, as shown in Table 10-1. The relative humidity change is negligible. In most oases, Ihe increase in the blowdown rate ix mail and the resultin profiles are bounded by tho existing profiles due to the conservatism in the current HELB analyses, as discussed for each break 10.1.2 Maln Steam Line Breaks

)) However, the intermediate size MSLB is defmed a the largest break that the MEL high flow sernsos do not detect and isolato (144% of rated flow). Becse rated steam flow incroases, the mass flow rate for the intermediate size MSLB also increases. The mnas and energy releases for the intermediate size MSLB wee sxaalyzed at BPU conditions. Becaus (he Reactor Building pressures and temperatures for the double-ended MSLD remain bounding, there is no effect on ths MS HELE evaluation due to the increased flow rae for The intermediate Eize MSLB.

10.13 FeedwaterLne Break The CLTP mass and energy releases for FW line breaks are zffeted by change. in the FlY system including Increased FW flow rate and modifications to the condensate, condensate booster, and PW pwups. Th mass and energy releases for double-ended breaks and acitcal coka in the FW lines were re-analyzed at EPU conditions. The reactor building preasue, temperature and relative humidity prfiles used for EQ were detennined to be bounding for BPU conditiono.

10-I

-.-..... I.

NEDt043047 -Revinlon 0 10.1A UC7 Steam Line Breaks Because there imno increase in the reactor dome pmssur relative to the CLTP analysis, the mass flow rates for HPC stem linc breaks do not increase. Terefbre the CLUP analysis of Ite HPCI steam line breaks Isbounding for EPU conditions.

10.1.5 RCICSteamueBreRks Bocauseo therm is no increa it dt reactor dome preasuo relative to the curreat analysis, the inn Row rafes for RTCIC seam line breaks do not inmrme. Therefore the curat analysis of the RCIC ste lina breaks is bounding for.EPU conditions.

10.1.6 RWCU Syftem Line Break An evaluation of the mass and energy releases for ZWCU line breaks at CLTP and EPU conditions indicated that the EPU mass releases for RWCU line breaks increased by 4.4% based on K compadson of the critical flow characteristics at CLTP and EFU conditions. The enthalpy of the fluid released decreased by 1%due to itcreased subooling in the reactor recirculaion fluid. Tez reactor building preisuro, tenperanim and relative humndity profiles at EPU conditione were evaluated for the effect on equipet qualification as discused in Section 10.3.

10.1.7 PlpeWbip end Jet Implngement Pipe whip and jet impingement loads resulting from bigh energy pipe breaks are directly proportional to system pressure. Operatin at the BPU 105% core flow condition requires a small increase in RRS pump discharge pressure. The effect of this pressure increase on the RRS discharge piping has been evaluated and confirms that the oxisting pipe whip or let impingement loads on HELD targets or pipe whip restraints arm bounding for the EPU conition.

Additionally, a review of pipe stress calculations determined tler the P temperature increase Usociated with EPU conditions does not result in pipe stress levels above the thresbolds required for postulating HELBs, except at locations currently evaluated for breaks. As a reult EPU conditions do not result in now HEL locations nor ae the existing HELB evaluations of pipe whip restraints and jettargeis anoted.

10.1.8 IntealF loodiag from HELB Line Breaks The only higb energy liquid filled lines in the rezctor building are RWCU and 7NW. The mass releases for the critical break flow for tie RWCU breaks were calculated using the break flow increase due to the lower EPU RWCIU onthalpy. Tho resulting inreaes in flood level datanmined for the affected flood ars were in ificant (c 1"). VYsystem hardAre changes have beeanevaluated and the flooding rate from .FFWline break is aoeptalle. Te water level in the hotwells, the existing dranage systems, and the flood barriars are ot changing; therefore, the eisting fW line break flooding analysis is valid for the EPU conditions.

10.2 MODERATE ENERGY LINE BREAK MELB5 are evahluaed fbr ther cffct. on EQ.

10.2

.- l.- s-e NED0-33047 - Rayion 0 System design limits (design pressur=) used as input to the MEBR flooding analyses "O changed by EPIJ. Therefore, the MELB internal flooding evaluations are not affcted by th EPU and tho design change process ensures wntinued evaluation of all changes for ffct on MELB flooding.

10.3 ENVIRONMENTAL QUALIFICATION Safety-related comnponents are required to be qualified for the envlronment in which they are required to operate. Tabl 10-2 provides a listing of the parameters used in EQ.

10.3I ElectilalEqulpment The safetyrxlatcd electrical equipment was reviewed to ensure the existing qualification for the normal and accident conditions expected, in the ara where the devices are located, rewait adequate.

10+/-1.1 InsIde Contarnment EQ for safety-related electical equipmenrt located inside the containnimit is based on MSLB andlor DBAILOCA conditions and their resultant temprature, pressure, humidity and radiation consequences, and includes the environments expected to exist during normal plant operation.

Normal temoeratures are expected to increase lightly, but remain bounded by te nornal temperatures used in the EQ analyses. Tho pos.t-acident peak temperatr and prossure do not significantly increase for EPU. The long~term post-accidcnt temperatures inside containment increase. However, the increase in long-term post-acoident teanperetures was determined not to adversely affect the qualification of safety-relatd electrical equipment.

Te current radiation levels under normal plant conditions were conservatively evaluated to Increase in proportion to the increase in reactor thermal power. The accident radiation levels increaae by < 12% above the CLTP levels. The total integratod doses (nomal plus accident) for EPU conditions were determined not to adversely afflct qualification of tht equipment located inside containment The increased radiation doses resulted in a reduction oftho radiation life of some solenold. located inside contaisment. However, the qualified life based on tharnad aging.

is shorter than thelradiaton life for these solenoids. Therefor the equipment qualified life W11 not reduced due to the increased radiation doses.

10.3.1.2 Outalde Containment Accident taeperature, pressuro, humidity environments, and flood levels used f6r qualificution of equipment located In harsh envirownents outside containment result from an MSLH, or other HELDB, whichbver is limiting for each plant area. The ed]sti HELB pressure profiles were determind to be bounding for EPU conditions. The peal HBLB tempeatrs at BPU RTP zxo bounded by the existing values used for equipment qualification. The temperature and humidity profiles that arm not bounded by the existin conditions werc re-evaluated and do not adversely affect the qualification of safety-related electrica equipmt. The acident temperature resulting from a LOCA/MSLD inside contnirmmt increased for som reactor building armas due to the 10-3

• . 1' - - -

NEBDW33047 - Reviloan 0 additional heat load resultig fom th inreaese in drywell and wetwll temperstur. However, the inormse in Iong-term post-accide te eratures, was evaluated and determined not to advemely affeot the qualification of uafety-roluted eleofrical equipment ThI notral temperature, prossuro, and humidity condidons do not chang as a result of EPU. The current radiaion levels under normal plant conditions wore Consrvatively evaluated to increase in proportion to the increase in reactor thmnal power. The accident radiation levels increase by c 12% above the CLTP levels. The total integrated doses (nonnal plus accident) for BPU conditions were evaluated and determined not to adversely affect qualifiction of the equipment located outside of containment.

10.3.2 Mechanlcal Equipment With Non-Metallic Componenti The chnges to normal and post accident ambient conditions for safrty.related equipment, as a result of EPU conditlon, aw discussed in Section 10.3.1. Reevaluation of th stafety-rlated mhanical equipment with non-mctallic components identified som: equipment potentially affected by the BPU conditions. Thee effects were evaluated and determined not to have an advcrse effect on the functional capability of non-metallic componnts in the michanical equipment both inside and outside containmt.

10.3.3 Mechical Comp-nent Design Qualfflcation The process fluid operatng conditions of equipmertntoomponents (pumps, heat exchange, eto.)

in oertain system are affeted by operation at E'U due to Btighy inraed temperatures, pressur, and in some cases, flow. The effects of increased fluid induced loads on .afcty-relatcd components are described within Sections 3 and 4.1. Increased nozzle loads and comnponent support loads, duo to thz revised operating conditions, wero evaluated within the piping assessments within Section 3. These increased loads are insignificant and become negligible (i.e., remain bounded) when combined with the governing dynamic loads. Therefore, tho mecanical components and component supports are adequately designed for EPU conditions.

10.4 TESTING Compared to the initial tartup program, [

1], EPU requires only a limited subset of the original strtup test program As aplicable to this plants design, tering for EPU is consistent with the descriptions in Section 5.11.9 and Appeadix L, Seotion L.2 of BLTRI. Specifically, the following testing will be performed during the initial power ascension uteps for EPU:

I. Testing will be perfonmed in accordanco with the TS Surveillance Requirements on Instrmnentation that is re-calibrted for EPU conditons, Overlap btween the IRM and APRM will be assured.

2. Stoady-slate data vail be taktn during power ascension beginning at 90% CLTP power and continuing at each EPU power increase ina This data will allow system perfcrmanoe parameters to be projectad through the EPU power ascension.

10-4

NEDO-33047 - Royaou O

3. EPU power inreasee above the 100% CLTP will be made along an established flow oontro1rod line in increments of equal to or less than 5% power. Stead-nate opeting dta, includiog fel thermal margin, will be taken and evaluated at cach step. Routine measr ts of reactor and systm presur, flows, and vibraton will be evaluated from each merasurment point, prior to the next power inereet Radiation =hr wtill be made at selected power loel to ensure the protection of pesomel.

4. Control 5ystem telts will be performed for the reactor Weaor wat Level contrls, pressure controls, and recirculation flow controls, if applicable. Theae operational test. will be made at the appropriate plant conditions for tht teat at oea of tbe powvr incrnents, to show acceptable adjustrents and operational capability.

S. Testing will be done to confirm the power level near the turbine first-stage scram bypass sctpoint

6. A test specificution is being prepared which identifies the EPU tests, the associamd acceptance criteria and the appropriate teat conditons. AU tsting will be done in accordance to written procedureeasure red by 10 CFR50, Appendix B, Criterion XI.

The same perkrniance criteria will be used as in the original power ascension tests, unless they htve been replacd by updated criteria since thc initial test program. U JJ For BPU, Browns Porny does not intend to perfbrm larg transient teutin involvig an automatic scram from a high power. Transient expezience at high powere at operatiq BWR plants baa shown a close correlation of the plant transient data to the evaluated events. The oprating history of Browns Ferry demonstrates thatprevious tranient events from full power are within cxpeted peak limitg values. he trwiont analyses demonstrate tat safety criteria are met, and the: tis uprate does not cause any previous non-limiting events to become limiting. Based on the similarity ofplants, past transient testing, past analyses, and the evalualon of test results, Ih effects of the EPU RTP level cam be analytically determined on L plant ipocific basis. No new systms or features were installed for mitigation of rapid pressurization anticipated operational occucos for this EPtU. A scram from high power level results in an unnecessary and undesirable transient cycle on the primary system. Therefore, additional tnient testing involving a scram from high power levels is not justifiable. Should any future large transients occur, Browns Perry procedures roquire verification that the actal plant response is in accordance with the predicted response. Plant event data recorders are capable of acquiring thc necessary data to confirm the actual versus expected response, Further, the important nuclear characteristics requird for transient analysis are confw ied by the steady state physics testing. Transient mitigation capability is demonstrated by other tests required by the TS. In addition, the limiting transiet analyses ar included ze part of the reload licensing nalysis.

10S5

NEDO-33047 - RevWson O 104A1 Recirculailon Pump Teting Vibration testing of the reciroulsion pumps is not required, bocauss ther is no chnge in the 3anximum core flow. To maitain the same cort flow with the increased core pressure drop (due to an increase Insteam production), rr tion flow (drive flow) increases slightly (< 3/.)

The "contairnent noise" observed in a BWRI4 - 25! in 1994 i not expected at Browns Ferry, At that plant an inortese in containment noise and vibration levels during plant operation was observed at increased recirculation pup speds. Bused an test results, the utility concluded that the increased noise was a direct rwilt of the RHR check valve not being properly seated. The testing demonstrated that the containmnt noise levels were greatly attenuated when the RHR chock valve was properly seated. Thus, this phenomenon is unrelated to EFU and no containmt noise is expected due to BPU.

10.4.2 10 CFRSO Appendix3Testg The plant 10 CFR 50 Appendix J test progran is required by tho Technical SpociScations. This test pogram periodically pressurizes the containmet (Cypt A tes), the containment penct:rtions (Type B teat), and the containment isolation valves and tes boundary (Tpo C teats) to the calculated peak contaimuent psure (P.), and measures leakage. Resulting from tho EPU, Pa changes to 485 psig, as shown in Table 4-1. Therefore, the 10 CPR 50 Appendix Yt..t pzogr is rovised to reflect this calculated peek containmet prese value.

10.4.3 Malin Steam Line, Feedwater, and Reactor Recdrculatlon PJping Flow Induced Vbration Testing Piping for the MS, FW and RRS will be monitored for vbrations during initial plmnt operation for the new EPU operating conditions. This test progr will show that the vibration of these piping systems isacceptable at the EPU conditios.

The MS and F}W piping systems are normally affected by an BPU, because their mass flow rates and operating pressures usually increase at EFU. The mass flow rates in these systems typically increase in proportion to the EPU power level increase. The flow induced vibration level.

simultaneously increase in proportion to the icresse in the fluid density and the square of the fluid volocity at these higher mass flow rates.

There is a small recirculation drive flow increase for EPU, and thus, vibraton monitoring will alsobeperforted on the RRS piping.

The MS, FNY and RRS piping inside oontainment will be monitored with romote vibration sensors. Also, the MS and FW piping outside containment will be monitored with remote sensors or hand-hild insirments. The vibration monitoring devices will be located on portions of the piping system determined to be most susceptible to vibration.

Actptable vibration criteria vill be established for these locations prior to testng, Vibration monitoring of those pipin systems will bo performr4 at power levels below the final, maximum extended power level. Vibration data is typically colleted at 50%, 75%, 100%, 105%, 110%,

10-6

NEDO-33047. RPvIuon O I 15% and 120% of OLTP. The masured vibration loewl at each power level will be compard to the acceptance criteria to verf the piping is below the acceptance criteria.p1ior to moving to the next power level. In this mnner, te vibration monitoring ting can proceed as the plant operates for the fist time at each new power level, and at the mne time avoid the rmmote possibility of incuring high vibrations and damging the plant equiprmnt (piping), before appropriate correetive actions can tace place.

105 INDIVDUAL PLYAT £VALUATJON PRAs are performed to evaluate the ri of plant operation, The individual Brosns Ferry Unit 2 and Unit 3 PRAs for Unit 2 and Unit 3 operation were evaluated and updated as a result of the analyses inputs, results, and modifications associated with the EPU.

To ensure that all risk-significant effects of the EPU are reprented in the revised PRAB, all EPLJ 2nalyses and associated plant modiflcatians vt systenatically reviewed to ideatify their effect on the elements of a risk assessmcat. Specifically, the modifications associated with the EPU were reviewed with respect to theirpotential effect on the PRA models.

Regulatory Guide 1.174 provides the guidance framework for using PRA in risk-informe decisions for plant-specific changes to the licensing basis. The acceptance guidelines consider both the magnitude and size of the changes to CIPF and LRER. The baseline and the EPU CDFs for both units ae below 10 5 events pcr year and the change in CIP for both units is sligly greater than 10.6. The baselin and EPU LERFs forboth units are below IQ4 events per year and the change in LERF for both unit is slightly greater than i0a7.

Based on the aceptance guidelines of Regulatory Guide 1.174, the plant changes reflected in the updated PRAs are in Region 1], the calculated increase in CLIF is betwe en t 4 and 10'5 and the calculated increase in LERF i betveen lO7 per year and 17' peryear. As stated in Regulatory Guido 1.174,

• *'Wenthe calculated increase in CLP is in to range of I0' per reactor year to 104 per reactor year, applications will be considered only if it can bo reasonably shown that the total ODF is lean than 10 per reactor year (Region II)." As shown in Table 10-3, the total CDP for Browns Ferry rrnains below the guideline value of Io4 per reactor year.

• "When lbe calculated increase in LERF is in the range of 10 7per rectr year to I0V per reactoyear, applicaions would be considered only if it can be reanably shown tha the total CDE is less than IO' per reactor year (Region II)," As shown in Table 10-3. the.

total LOBRB for Browns Perry renains bolow thz guideline value of Iff5 par roectr year.

The evaluations of the uncetainty from previous evalustionas of the xisk at Bro'wm Ferry have indicated that the range fictor of the CDF distribution was approximately 4.6. The current evaluation is a modification of those analyses. No new sources of uncertainty we introduced.

Therefore, the range fiator is not expected to increase significantly for this evaluation.

10.7

NEIO-33G471 - Rlsion 0 The arno analytical codes used to support thebasolln PRAs were used Inthe updates for EPt).

The plant-specific MAAP models were used to tupport the system suces critera deoterminion and sequence timing. Tbh RISKMAN integrated computer code was uwed to perform the lCCCeIHasy data and system analyses and to rprsent the response of the operators and plant systems to the initiators considered.

10.5.1 jnititng Event Frequency Thirty-five initiang eent categories are considered in the baseline and updat PRAs for each wi;t aixten transient initiator categories; Thirteen LOCA initiator categoris; and aix internal floodn initiator categories. The initiator ctegories for the EPU are the same, s those in the baseline models.

The EPU does not result in plant equipment operation beyond thi design ratings and conditions.

The tranient categories with the most potential to be affected by the EPU are those associated with trip set points, such as reactor scrLm, syste isolations, and operating equipment trips. A review of these conditions concluded diat the operational margins remain within values consistent with the baseline models and that changes regarding initiating event cat-gory frequencies are not required to reflect EPU conditions.

The frequency of loss of offsite power events (either "losa of 5OD kV to the plant' or "loss of 500 kV to one unit") due to grid instabilities is not affected by BPU. In addition to the loss of offilte power represented by these two in'itiatos, there is the potential that thD grid 1B lost following the trip of a unit. Por the PRA analysis approach, rapid separation of a large generating unit from te grid has the potential to cause Browns Ferry grid instability and loss of ofiuite power. This possibility is represented by a unique top event A grid stability analysis has beern perforned, considering the incrcase in electicaI output, to deonstrafe confonuance to Goneral besign Criteria 17. In addition to the nonrml configuration, the analysis considen various transmission system contingencies.

The baseline PRAs identify five initiaors as a remit of the loss or degradation of support systeas: loss of I&C bus A; loss of I&C bus B; loss of plant air; loss of RBCCW; and lose of RCW. The duty on these systems i cssentiallyunchangod as a result of the EPU. Therefore the frequency of theso initiators does not change under EPU conditions.

Tho nornical RPY pressure does not change at the EPU conditions. The RPV and piping-monitoring programs do not change. Therefore the frequency of LOCA and internal flooding initiators does not change at the EPU conditions.

ATWS is not modeled as a separate initiating even in the Brawns Ferry Unit 2 and Unit 3 PRAs.

Distead, the scram fimetion is queried for oed sequence model as an inteogral part of the response of the plant and the opeatrs to ai initiator. RPS reliability is taken from NUREGICR-5500.

The total frequency of ATWS is detemined by multiplying the frequency of tho tansiont events by the likelihood of failure to scram. As discussed above the frequency of occurrence of the transient initiators is conservatively usumted not to chang aftr the modifications ar 10-s

.. I- - .. .. .. -

NEDO-33047 -Rvldon 0 impletented. No modiftcations are anticipated to affect the likelihood of scram. Therefore, the frequency of ATWS ii not expected to change.

The frequencies of the initiator categories in the baseline PRAs for Browns Ferry do not change in tho updated Unit 2 and Unit 3 BPU PRA models. For the mSajoity ofthe initiator castegories, the EPU has no affect on the fequency of occcurence. The three initiator categories potznially ffectd (turbine trip with bypass; LOFW; and loss of l condmensate) are.ompcnsated farbytthe exected decrease in thb frequency of the caegories "loss of FW' and 'loss of all condensate,"

Use of the same Initiator data in the updated models is conservative. The sunuay of the initiator contributions for the baseline and EPU risk models for CDF and LBRF end-stas Is presented in Table 104.

10.5.2 Component and SystemneRalabwty No increase in component failure rates is anticipated as a result of EPU conditions. Under EPPU conditions, equipment operating Units, conditions, and/or ratings are not exceeded. Existing plant component monitoring programs detect degradation if it occurs and corrective action is taken in a timely manner. It is possible 1hat EPU conditions may result in seloeod components requiring refirbinent or replacaent .more fnequently; however, the bnctionality and reliability of components and systems is maintained at the current standard.

10.5.3 OperatorResponse The effect of the EPU on operator response Mapabilities following an initiating event requiring safe plant shutdown was evaluated. Tbe evaluation addressed all post initiating event actions by considering two fadts: 1) the rution in the amount of time available for the operatrs to completa an action, and 2) whether the reduced time available was a.significant influmice on the error rates estimated for the human actions in the baseline (pre- EPU) PRAs.

The decay heat produced following a.reactor scram is proportional to the inventory of fssion products, Which is directly proportional to thc power level, A reactor operating at 12% OQLTP power compared to the same reactor operating at pro- EUP RTP will contain 120/1OS - 8/7 of the Inventory of fissiOn products. Co nquently, the RPV inventory and/or nupprcssion pool hbet capaity available at the tim of the -fnilriang event reach their respective limits for succesDM action in approximately 7/1 of the amount of time available for the baseline case, which amounts to a less than a 15% reduction in available response time. This reduction in available time applies to operator actions srtly after the initiating event to respond to eiterno Injection into tho RPV andor no suppmion pool cooling by manual initiation of these ctiOns.

Bocause the decay heat doecrases to less tlan 2% of RTP within 20 minutes after shutdown, the increased decay heat at EPU becomes less significant regarding the time available for operator actions. The effect on the time available fbr operator actions in response to an ATWS from the EPU power level is strongly influenced by th intermediate power levels assmoiated with the ATWS. However, for purposes of ostimang M18?s, it is reasonable to ass e that the inriternediate power levels during an ATWS arm also proportional to tbe initial steady atate 10-9

NEDO-33047 - Resvlon 0 power. This assumption is supported by MAPP calculations performed in support of the evaluation of the EPU.

The operatr reponse evaluation for EPU evaluated the 61 post-initiating event actions used in the baseline models to ideftify the early event actions required to respond to either the loweing of x*V level or an increse in suppression pool tenpcrature durin shutdown or ATWS conditions. Fifteen early response actions and ATWS-reltd actions post-initiating event human enos were identified in the baseline (pro- EPU) PRAm for further evaluation of tho relative importance of the time available for operator response. A sixteentaction, "failure to backup scram" (which would reduce the frequency of the ATWS sequences), was identified, but no credit is taken for it in the baseline models.

For the fifteen actions, the time aNailable for compleing the action was reduced by 1/8, and each actio was ovaluated to determnine the inuence of tis reduction on the error rates estimated for the human actions in the baseline (prE.EPJ) PRAm. An useszanmet of the relative importance of the time available for operator response is explicitly included in the models used to estimate HENs in earlier releases of the Browns FPrty PRAs RPefeiac 2 used licensed plant opciator ratings of PSFs to quantify human errrs using a. FLIM, in which the relative weight and difficulty mociated with tim costraints was rated by three operating crews for each post-initiating event acion. During the update to the Browns Ferry Unit 2 and Unit 3 baseline PRAs, the HEPs for some actions were modified using the EPRI approach for the analysis of human actions, BPRI TR-lOD259. The BPRI approach uses the HCR correlation foractions where time-reliability related fictors are considered most imporant and the CBDT method for actions where PSE rolaed factors other than time are considered most important.

As a result of those evaluations, the HEN for five or the fifteen actions were changed to reflect the reduced time available due to the EPU. Four of the actions are those In which time-reliability related factors are oonsidered the most sigificant EP contributor. These actions bad been quantified using tbl HCR model in the baseline models. The IEZP for these actions were updated for EPU by incorporating the reduced available time diroly into the HCR model. For eleven of the fifteen actions, time constraints were not ELprimary contributor to the evaluation of the HE?. In the baseline PRAs, these actions had been quantified using either the FLIM or CBDT mehodology, both of which evaluate PSFB that may not bc strongly influenced by the time available. For these actions, the effect of the reduction of 15% in available time was evaluated considering the reasoning for selecting values of the PSFs ji the original evaluations.

If the reduction in tim availtblo did not significantly change the assessment of the baseline PSM's of an otion, its HEP was not modified Using this approach, the HBPs often of the eloven actions were determined to be unaffected by the EPU, and one action was updated.

The HENPs that were changed are:

HOAD is the failure to inhibit the ADS with an un-isolated RP1V. HOAD)! applies to sequences in which the RPV is not isolated and the injection from the FW system is available. During these conditions, the necessity to reduce level is dictated by the heat up of the suppraBsion pool, because the suppression pool must absorb Thef energy That it not dissipated ViL afih turbine ste 10.10

NEDO-33047 -Revidlon 0 bypass valves. The tim availablo before havitg to reduce the RPV level lo .122 inches is dependent on the tuppression pool heat up rate and 10 minutes is used for the baseline case. In addition, it was consrvatively assumed that a hgb drywell pressure ignal is also present.

Given these oonditims, tbe ADS timcr proidce four minutes after reaching -122 inches, and the total time available is thea 14 minutes for the baseline case. This action is not time-critical (other PS?. such as preceding and concurrent actions influence the action) and is modeled wing the CBDT methodology. For the BPU case, the suppression pool heat up rate incrases, and the total time available decreues to 12.5 minutes for the required action, The aotton is still nottime-critical, and ths CEUT model remains applicable. However, using the EPRI methodology regading minimum time needed to aply compensating factors for HBPs, the reduction of tirm is assumed to restrict the time available to review the operating crew acdons, thus removing credit for one of two revisitatlon checks forn the cognitive portion of the ME. This rmsultod in an increase to the estimted mesa HEP from the baseline value of 3A45EO3/dand to a EPU value of 4.89EO03tdemand HOAD2, Inhibit ADS actuation, during ATWS with an isolated RPV, is the failure to inhibit the ADS. HOAD2 applies to sequences Where the RPV is either isolated or the FW system is unavailable as P source of water for the RPV. In addition, it was conservtively assumed that a high dry well pressure sigaal is also present Under those conditions the Jovel of the RVP rapidly decreass, nd the time ailabi is limited by the 115 sooond ADS timer. HOADZ is a time-

.critical action, and the. estimato of the HE? is based on the EPRI 11CR correlatIon model. Tho ADS tImeout i6 de sinam frboth th baseline and EPU power levels. However, the mean REP for this action increases from a baseline value of 4.64E-03/demaid to a EPU value of 9.52B-03/detnand in the HCR model affecting the rate at which the H1? changes when the tine required to complete the action is compared to *h median tin obtained ftom operator simulator experience. For the EPU, the natural logarithm of the standard devialion of T im, the median crew completion time for a time-related task, was raised from 0.4 to 0.45 to reflect the BPRI guidelines for the high stress anticipated during this event.

HOAL2, allow RPV level to drop and control at TAF, during ATWS with isolated RPV, reduces reactor powor Bs the SLCS is injecting. The operator allows lavel to drop to the TAF, but once the TAF-is reached action must be initiated to and gain control of injection within 2 minutes (120 seconds) to proven the RPV level fromn decreaing to < -190 inches where core damage is assumed to begin. HOAL2 is considered a tineicalaction with hi serss, and the estimate of the HEP is basod on the EFRI HCR correlation model. For the EPU estimate, the available timo (TW) in the NCR model is decreased ftom 120 to 105 seconds to reflect the reduced time availeble. In addition, for the EPU, the natural logarithm of the standard deviation of Tr, the median crew completion time for a ime-rsetod task was inc d from 0.4 to OAS to reflect the EPRI guldelines for the high stress Wficipated during this nativity. As a result the HE? for this action increased frm a bueline value of 3.91BO2deniand to aEPU value of I.293-Oldemand.

ROSLI, activate SLC with an tm-isolated RPV, the System time-window for operator response is based on avoiding having to deorease the RPV level to the top of fihe coresThis is a time-critical action wit high stress, and the estimate of the HEP is based on the EPRI HCR correlation 10-11

NEDO-33047 - Revslion 0 model. Thz time aaiable for the action depends on the point at which the suppession pool rosche the BIT, estimated to bo 3-5 minutes for the baseline case, and is taken as 240 econds for the Tw in the HCR correlation. For te EPU estimate, the ilabl, time is dece d frm 240 to 210 seconds to reflec thc increased heat up rnt As a raub, the HEP for this ation increases from a baselinrvalue of 6.71-03ldeanad to aEIU valueof l.6102/dcm;nnd.

HOSL2, activate SLC with an isolated RPV, the system line-window for operator response is based on avoidin having to docroase the RPV level to the top of the core. This is a itime-iticul action vwth high stress, and the estimate of the ?P is based on the E1 HC. corrlation model. The time available for tbe action depends on the point at which ghe suppression pool reachs BIIT, which ocurs in Ios time for the case of an isolated core, because heat cannot be released vis the turbine steam bypass valves. For the bseoline case, the T. in the HOE correlation was estimated at 3 minutes (180 seconds). For the EPU estimate, the available time is decreased fim I80 to 153 second; to reflect the incTead heatup. As a result, teo HB? for thls action incases from a baseline value of 3.30E-O21dtnand to a BPU vahl of 7.71-Odemand.

The HEPs developed by the operator response evaluation apply to both the Browns Ferry Unit 2 and Unit 3 PRAs. Table 10-5 presents a nmure of the relative importance of the operator actions changed by the BPU for each of t.e PRAs. Table 10-5 abows the frequency-weighted frtion of thes EPU CDF wit sequences containing failures of any of the huMan aotions in the PRA models. Table 10-5 is oidered by fictional Importuc for Unit 2 and sextnd; down to the point where all five of the actions when the HEN wre inrea ed in the models. It can be seen that (with the exception of two cross tie actions - these values differ duo to the cross tie differences between units) the magnitude end order of the frquency weighted fractional importance of the operator responses for both units is very similar, It can also be seen from Table 10-5 that the human actions whos HIEP increased appar in sequ=c;ce that comprise a higher frequency-weighted fraction of the CDF than in the basel~ne. However, the REP of an acton does not have to incremse for it to have a larger effect on CDP, as discussed below.

The largest frctional increase was observed in HORVD2, manual depressurization of the RPV using MSRVs, whero the HBP did not increase as a result of the reduction in the dim available.

Failure of this action is now part of the sequences comprising 55%° of tho Unit 2 CDF and 43%

of the Unit 3 CDP, up fiom 10% and 7% rospectively. This Is due to the elimination of ethanced C )RD iR-ection as a suc=cesful alternative for msaintainin RPV level under certain conditions when other high-pressure injection systems have failed and the vessel remains at high pressure.

Consequently, the frequency of core damage sequenoes where the operators are required to, but fail to depresunnize the RPV has increased, Three of the fifteen human actions with the highcst fractional CDP importce have larger HEPs as a result of the reduction of time evailable for completing actions resulting from the BPU, All the of these actions, HOSLI, HOSL2, and HOADI appear in ATWVS related smarios. The largest fractiona increase oocur for HOSLI, activato SLC with an un-isolated RPV, which incrOease from 1.2% to 2,2% CDF for Unit and from 0.8% to 1.7% CDF for Unit 3. Thus, it is 10-12

NEDO-33047. Revlon 0 ooncluded that there Bro no nignificant Iceases in CDP due to the reuced time available for operators to complete post initiating Ovent actiots.

To provide support for (bo cstimzst of the EPU influene on huma action earro evaluations, threeocniios were performed on the. Browns Ferry aimulator with io reactor initially opeating at the BP TF level, Most of the acion. with changed error rates involved ATWS scenarios, The ATWS scenarios were jpmrmed both with and without MSLV closure.

Additionally, a motor scram with the loss of bigh-pressure injection was perronred to verify the operator actions using manual depressudrzaton and level control using low-pressure jection.

The simulator results indicate that the operators v-ere capable of performing the required actions wi'tn the time frames Ithe human reiability calculations.

10.5.4 Succs Criterla The response to an initiator is repsented in the PRA models by a set of discrete requirements for the operation of individual sytan and the performano of specific operator actions. These soecario-apecfic requirements define the success criteria for system operation and operator actions to fulflll the critical safety functIons necessary to maintaii the reactor fuie in a safe.

condition. The critical safety functions are reactivity control, RPV pressure control, containment pressure control, and RPV inventory makeup. These individual criterda were reassessed at the BSFU R1? conditions and increased decay heat.

The scram function is not affected by EPU because operation of the CaD system and the RPS is not affectod by the EPU.

Because of the increased nominal initial power level, the tm vailable to perform selected operatr actions decreases. Examples include initiation of the SLCS in ATWS senarios and control of th RPV level following a tanient These effects amesummarized in Section 10.5.3.

The thermal hydraulic analyses supporting the 6uccess criteia used in the baline (prie EPU)

PRAs were performed using MAAP. These analyses include consideotion of scenario. using MAAP 3.083 to support the 1992 IPE m well as analysee using MAAP 4.0 to support the baseline PPAs representing operation at 105% of the OLTP. To invetigate the potenti effect of EPU, th entire set of thermal hydraulic analyses used to support the PRA success criteria were reanalyzed using MAAP 4.0.

The results of the reanalysis of supporting thermal hydraulio calculations indicate that enhanced CRD system operation (i.e., operation of both pumpi for vessel injection) alone is not sufficient to prevent fuel dantage if tbh RPV remains at high pressure. If te,vessel is successfully depresurized within six hours following a nuccessfil scram, enhanced CR1 system operation is sufficient to provide successful RPV level control. The scenao models in tho FRAs wer updated to efleot this change In suecess criteraL All otte system success criteris were confirmed by the new analyses.

10-13

NEDO-33047 - Revifon O 0AS5 EsternalEvents Thz eflbct of the EPU was reviewod to detenino whether any ncw plant vulncrabflities exist from the occurrence of internal fires, seismic event, and other cxernal events. Equipment changes assCoItod with the BPU are rinor and do not affect reliability. ThE EPU does not affect any existing sructures, fire loading, or fire zonwe. and thcrefore no rmn vulnerabilties arm introduced.

The Browne Ferry PEBB (tefrrenoe 3) and Seismic PBEE Report (Reference 4) were reviewed to determine whether there were any existing conditions where the BPU could introduce new vulnerabilitbes.

The PEEB review concluded that thire are no new firc-induced vulnenbilitic asociawd with the EPU. The firc zones, fire loading, and safe ahutdovn paths for Browns Ferry do not change for BPU; therefore there is no Icrease in the vulnerability to internal fires.

Bocauso the Efl) modifications do not affect the structures or component anchoring, no aew vulnerabilities are introduced am rmsult of a.seismic event.

The IPEEB states that the Browns Ferry Plant/Faciides design is robust in relation to the 1975 SRP Criteria anda walk-down did not reveal any potential sgnificat vulnerability tht were not included in the original design bsis analysis. Becaus there are no external or other structural changea associtted with the EPU, there are no new vulnerabilities introduced from wind or flood events.

There re no changes in the EPU that could be affocted by transportation or nearby facility a mdent.,

thus there are no new vulrablitie introducod from transportation and nearby facility acidents.

10.56 Shutdown Rilsk A PRA model to evaluate shutdown risk, spcoifically CDF or LERE, has not been developed for Browns Ferry; however simplified risk evaluation tools are utilized Browns Ferry utilizs a defanse-in-depth approach to managing risk during plant shutdowns. To assist in the management of risk during shutdowns, TVA uses EPRI's conputer coda, ORAM. This process specifically monitors the safety functions: sbutdown cooling; fuel pool cooling; inventory control; offsite AC powtr; onsite AC power; primary and secondary cona ent; and reactivity control.

EPU increases the anount of decay beat following shutdown, which has the greatest offect on RHR capability. Prior to each cutage, a decay heat prediction baed on best-estimate values (i.e.

no statistical uncertaity or added conervaismn applied) is detenninect This decay heat prediction is used to create a,"time to boil off curve' which is then used by th. outo planning team to ensure that heat removal 6ystems arc available and that contingency plans are rade for muantennce and testing of systers. The icremental decay heat due to the BPU will ulightly extend the tim that the existing DHIK systerm will need to remain in service during plant shutdown and remain available right aftr shutdown.

20-14

- - - ---- ---. I .- . - ...- ".1 ---- "  % '-'. :.-. I,.-,:.,-- .. T. .- .' ... ... .1 '. , . .. I . -1 .1 NEDO-33047- Rcviion 0 The Brown; Ferry TS and TRM adrs the above requirernet regarding shutdown risk umnameat concepts. Brown. Parr procedures provide complete and consisrt implementation of the steps required to ensur that shutdown risks attributes an controlled.

ThrebM EPU has no effict on te process controls for sbutdon risk managemetnt ad Lnegligible effect on th overall ability ofBrowns Furyto adequately manae shutdown risk.

10.5.7 ProbablrtlcRiskAmieamieat The PRAs are the end products of over 10 yean of analysis efort The Browns Ferry PRAF are living doGUmants PRAR and were updated as recently as early 2003.

TVA procedures provide the details describing the use of the PRAs at Browns Perry to support the Maintenance Rule. They assis in cstablishing performanic c riteria, bancing unavailability and reliability for risk sgnificant SSC and goal setting and provides input to the onste Bxpert Panel for the risk significance determination process when. rvisions to the PRAs take place.

Futotions are potentially considered risk significant itf any of to following conditions aro setisfied:

• Functions modeled in the level I PRA are found to havt P risk vAbievemeot worth greater than orequal to 2.0; Funcions modeled in the level 1 PRA re found to ha^v a nsk reduction worth of less than or equul to .995; or

• Functions modeled In the level I PRA are found to have a cumulative contribution of 90% of the CDP.

Becaus the PRAs aro actively used at Browns Perry, a formal process is in place to evaluate and resolve PRA model relstCd issue; as they arM identified. The PRA Update Reports are evaluated for updating every other refueling outage. The adinstrativc guidanc for NOis activity is contained inTVAProcoduret During November 1997, TVA participated in a PEA Poer Revim Certification of the Browns Fony PRA adminrsted under the auspices of the BWRDG Peer Certification Committee. The purpose of the peer review process is to setablisb a method of assosin the technical quality of Lthe FRA for thzir poteutial applicatlons.

The Peer Review evalwation process utilized a tiered approach using standardized cbecldists allowing for a detailed review of the elemnts and the sub-elements of the Browns Ferry PRA to idetify strengths Lad area that needed improvement ho review system used allowed the Peet Review ten to focus on technical issue6 and to issue their assesment resulf in the fbrm of a "grade" of 1 through 4 on a PRA sub-elemet level. To reasonably span the spectrum of potetial PRA applications, the four grades of cetification as defined by the BWROG docment

'¶Report to the Industry on PSA Peer Reviev Certification Process - Pilot Plant Resuls" were employed.

• 1015

NEDO-33047. Rehlan O The BrownS Ferry Per Review summaried in Table 10-6 resulted in 5 conastoent valuation across lII the elements and sub-eletnenls, Also, during the most recent PRA update, all significant firnings (i.e., designated as Level A or B) from. t Peer Certification were resolved, resulting inal PRA elements now having aminimum certification grade of 3.

10.6 OPERATOR TRAIING AND HUMA FA ORS Some dditional training is required to enable plant operation at the EPU RTP level. For EPU conditions, operator responses to transient, accident and special ovents are not affected. Most abnormal events rcsult in automatic plant shutdown (scram). Some abnormal events result in automatic RCPB pressure relief ADS actuation ndfor ntomatic ECCS actuation (for low water level events). BPU does not chango any of the automatic safety finolons. Alter the applicable automatic resposes have initiated, the subsequent operator aotions (e.g., nmintaining saf shutdown, core cooling, and cotainmont cooling) for plant safety do not change for BPU.

Thc analog and digital inputs for the ICS and SPDS will be reviewed to detcrmine the ceects.

from HPU. This iacludes required changes to monitored point, calculatincs, lert and trip setpointa. Various changes in EOP curve; and limits, if required, will also require am update of the SPDS. Any changes required to the ICS and SPDS comTuter will be completed prior to startup at EPU conditions.

Following a review of the BPU modifications and identfied key procedure changes, recommendations for operator training and simulator changes and a final dekrrnination of the opertor training needs will bc made, consistent with thie Browns Ferry training progran for sleotion of modifcations for operator trainin. Any modifications required for EPU will be dvalusted for its effect on the ICS and SPDS and any reqired changes (including any new monitoring points) will b: addressed as a part of tho modificion. Any changes roads will be discussed ns a part of the operator-training program far EPU Training required to operate the plant following EPU will be conducted prior to restart ofthe unit at the EPU conditions. Data obtained during stamp testing w11 be incorporated into the training as nceded. The classroom training will cover various aspects of EPU including changes to parametrs, setpoints, scales, procedures, systems and startp est procedures. The classroom training will be combined vith simulator training. The simulator training vill include, as a minimum, a dmnstrtion of transients that show the greatest change in plant response at EPU RTP compared to CLTP.

Installation of the BPU changs to the Browns Ferry Simulator is planned for approximately six months prior to EPU impleientation. The firt cycle of each traning yer includes taining on the nfications to be installed during the upcoming spring outage, including simulator training as required. The two training cycls prior to EFU implementaion Wilt cotwleb the reormended operator cluaroom and simulator training for EPU implementation. The simulator changes will include hardware change. for new or modified control room instrumentation and controls, software update fox modeling changes due to EFU setpoint changes mad re-tuning of the core physics model for cycle specifo data The Simulator ICS will 10-16

NEDO-33047. Rabidon.

rlso be updated far EBU modificafioe as part of tho simulaoroutoge prior to BPU implementatioL Soe differee mayt exist until EIJ is inplemened on both units. Tese diffeces will be inluded on th4 SimulntortUnit 2oUnit 3 diffzrencs listing whicb Isreviewed in each training cycle.

Operating data will be collected during EPU implementston and start-up testing. Ths data will be conqred to seimulator data as required by ANSIJANS 3.5*1985, Section 5.4.1. Simulator aeccptanoo testing will ba oducted to benchmark the simulator perfrmance based on design and enginering analysis data as required in ANSIJANS 3.5-1985. ANSI/ANS 3.5 is endorsed by Regulatory Guide 1.149, revision 3 and 10 CFR 55. The simulator acceptance testing for EPU is planned to be caomlete w10iin 30 days ater steady stale oporation at 120% of OLTP.

10.7 PLANT LNE 10.7.1 RPV Internal Components The plant life: evahltion identifies degradation mechanisms influenced by incruses in fluence and flow.

Browns Ferry has a procedural controlled program for the inspection of solected RPV internal componenta in order to ensure their continued structural intcgrity. The inspection techniques utilized are primarily for tho detection and churwterization of service-induced, surface-onnected planar discontinuities, such as IGSCC and IASCC, n wolds and in the adjacent base marial. Browns Perry belongs to the BWRVJP Orgaization and impleenioation of the procedurally controlled program is consistent with the BWRVIP issued dowments. The inspection strategies recommended by the BWRVIP consider the effects of fluence on zpplicable compDnents and zm based on carnponent confgumation and field experience.

Components selected for inspection include tose that am identified u suscptible to in-service degradation end augmented examination is conducted for verification of structural integrity.

These componnts have been idcetified troug thec roview of NRC 1EBs, BWRVIP documents, sad rmcommendations proyided by GB SILs. The inspection program provides perfornance ftequency for ND1 and associated aoceptanoc criteria. Components inspected include the following:

• CS piping

• CS spargers

• Coe hroud and core shroud support (inoludes access holo covrs)

• Jet pumps and associated components

• Top guido

• Lower plenum

• Vessel ID attachment welds

• Inktrunentation penetrations

• Steam dryer drain channel welds

• FW Rpargorn 10-17

e - @- - -. - . - . - -- . . - - -I--- . -- -

NEDO-33047 - RCV#onI

• CarepILte

• SLC piping (internal to RPV)

Continued implementaion of the currt procedure progrpm assures the prompt identifloation of any degradation of reactor vessel internal comlpo2e0t experienced during BPU operating conditions. Browns Ferry utilizes mitigation teerlquqg, such as HWC and NMC, to mitigate the potentlil for IGSCC snd IASCC. Rectorvessel-watr chemistry conditions an tals maintained ccnsistent with te EPRI guidance (Reference 5) and other established industry guidelines.

EPRI periodically updates the itwr chemistry guidelines, u new information bocomes avilable. As EPRI updates ocur, they arm reviewed for possible incorporation into the Bron.s Ferry Cemistry Program.

The peak ffuence increae experienced by the reactor Internals does not exceed the ftreshold value, which reflects c characteristic rise in intargranular cracking Refcrenc 6). The curtcnt inspection strategy for the reactor internal components is adequate to manaeg any poteial effcs of MPU.

10.7.2 Flow Accelerafsd Corrosion The Browns Fery procedurally controlled FAC program ctivities prodic detect, and monitor wall thinning in piping and components due to FAG. The PAC program is based on the EPRI guidelines in NSAC-202I, R2, 'Recommendations for n.Effective Flow-Accelerated Corrosion Progrmn'. TIh FAC program speclficatfons and reuireme ensura that the FAC program is implemented as required by NRC Generic Letter (GL) 89-O, "BrosionlCorrosion-nduced Pipe WaIl Thinnimg. The FAC program uses selective component inspections to provide a measure of confidence in the condition of systems susceptible to FAC for each unit Tmese selective inspections are the basis for qualifylng un-inpected oomponeat for futher service. in additica to ftis aggressive monitoring program, selacted piping replacements have been porfonned to maintain suitable design margins and PAC resistant replacement materials are used to mitigate occurrences of FAC.

ACEICWORKSTM FAC model (in accordane with tbe CHRECWORKS1h FAC users guide and EPRI modeln guideline) ham been developed for Browns FOrIy to predict the FAC wear rate (single and two-phase fluids) and the remaining serieo lift for each piping component The controlled CHECWORKSTM FAC model Is updated after each reIeliog outage, The FAC models ase al6o used to identify FAC examination locations for the outage examination list and uses empirical datz input to the model.

Process variables that influence FAC at Browns Ferry:

• Moirtum content

• Water chemistry

• TeOerature

• Oxygen 10olB

NEDO-33047 - Revidog 0

• Flowpalgemetry andvelocity

• Malaria! composition Browns Fey bus predicted EPU systen operating conditions that are used Rs inputs to the CHECWORKSTN FAC model. BPU affob moist content tepertO, ymgen and flow velocity but these remain within the CHECWORKSW FAC model parameter bounds. TUble 10-7 compares key parameter values -Cdetig FAC BPU paramnte values result in changes consequential enough to beat balance modal. tant certain systeus or portion of systems (extaction steam) see a disproportional Increase in wear compared to the percet power increaae, i.e., FAC ;tar rate will increase in some systems and decrease in others.

The CEWWORKS' PAC program targets MAC suscptible piping and components and includes the installation of FAC rosistant material. Based on experience at pre EPU operating conditions and previous FAC modeling- results, it is anticipated that the EPT opnting conditions will result in canges for the CHECWORKST model. The changze may then result in additional inspctiona scope, unless carbon stwl piping and components have bean replaced with FAtC resistant material.

The Browns Ferzy FAC progrm will continue to adequately manage lose of material due to flow accolerated corrosio, such that the piping and components will continue to perform their intended fimction at PU conditions.

10.8 REFERENCES

1. GE Nuclar Energy, "Generic Guidelines for Geceral Eletic Boiling Water Reactor Extended Power Upratre: Liensing Topical Reports NEDC-32424P-A, Class III (Proprietary), fPbrary 1999; and NEDO-32424, Class I (No-proprietazy), April 1995.

2. Tennessee Valley Authoity, "Browns Ferry Unit 2 Individul Plant Examination,"

Septeber 1992.

3. Tennessee Valley Authority, '¶Browne FCrJY Nuclear Plant Indnidual Plant Examinstion for Extral EMmts pPBEE3)," July 1995.

4. Tennessee Valley Authority, "Seismic IPBB Report, Browns Fery Nuclear Plant,"

Revision 0,June 1996.

5. EPRI Report, 'BWR Watcr Chemistry Guidelints -2000 Revision," TR-1035 15, Revision 2, Februay, 2000.

6. Ensincering Materials Strss Corrosion Handbook, Peter L Anderson, 1992, Chaptr 6, "l-mAdiation-Assistod Stress Corrosion Cckin."

10-19

NEDO.33047

• Rev~don.O Table 10-1 Browins Ferry Mgh EnegY Line Brezki Ia =cr(Chgnet) Due to EPU Beak locaton MSLB in Steam Tunel or No vbage *No chaage No change MSVV FW Line Breaks in Steam Tunel or MSVV + 12.3% No change S5O.5*P RCIC Steam Line Breaks iih Reactor Building or MSVI No obango No change No ohange HPCI Stewa Line BreAs in Reator Building No cmSge No ohange No Change RWCU Breaks in Rewctor :9.31 Building 44.4% <0.07 psi 10.20

- ' - . - .- - .-- *.- r . -. - *- . -. - . ... ..

• f ' -'-;.

NEDO-33047 RaAvWm O Table 10-2 Browns Ferry Equipment QuisWIcktien for EFU Paramer Noranl EPU plant opuation radiation increase inside containnt for EQ 143%

Accident EPU radiation increase inside ontuinmeni for EQ C 12%

Accidcnt EPU peac tperatur inside conltaient for BQ No Change Accident EPU peak pressure inside containment for EQ NoD 0hn Accident EPU ternperaturc caWside contaminant for EQ No change Accident EPU presmure oubide contaisnmt for EQ No change Nonnal EPU radiation level increase outide pontanment for EQ 5 14.3%

Accident EPU radiation outside containmet for EQ < 12%,

10-2I

P, .- a. - -- .- , Ir - - -. - -. -:?r - ru --.- ,: '.r -.. , - .- - . - --. - -.

NEDO-33047 - RYla1lon 0 Table 10-3 Summry Compariton of Bualine ad Updated CDF antd LERF Baseline EPU Browns Fery Unit 2 .

Total CDF (yr', mean value) 1.255 E-6 2.624 E-6 LERF (yf', mee value) 2A55 E-7 3.927 E-7 Browna Fery Unit33 total CD y, mea n valUe) 1.907 E-6 3.361 E-6 LERF (yr, mean vnIue) 2.688 E-7 4.532 E-7 10-22

NW1}043047 -Rilsuoe 0 Table 10-4 Summury of the InItiator Contributlow to CDF ud LX (or Brown Ferry UD1tZ Iitttor catgoyconbrlbutm to CF and L F Initaor Category Imanbuucy Darelin PU!

cDF LERF CDF L=U In addvctcntvpctb4 of 4.36EB0 4.52E-08 11923-08 4,43P-08 199E-OB oIO MSRV dvertcnt aPmif 3.42E-04 237239 6.18B.1 2.62-09 6.64E-II tw or more MRVe Isdvetit SCRAM 2.57E-01 2.66B-0 4.41B-10 8.993-08 3.93E-O9 Lose Df 500 kV toplant 9.32E-03 4.35B-09 1.013-09 3.913-VS 3.06E-09 Loss DfSOOkV to one 3.42-02 1.82PA08 S.32B-09 1.513-07 1,35E'08 Lois ofI&C But A 4.10E03 6.2509 5.51B-10 2.113-O8 1.87E-0 Lam DfI&CBusB 4.103-03 6.25E-09 5.56E-I0 2.1 IE4O 1.87E-09 Lou ofall condeniate 1.24E-02 9.84B-09 1.82B-09 5,58-08 4.6B3-09.

Lose oteoodnw heat i.20301 8.81-08 2.92M08 5,. 7 7.16B03-sink M[V c1ue, bypus, loss of condener vgtcn).

Lose of PW 4.816-D 1,10- 08 7.76B-09 S.1 IE-B 1.05E-08 Lom of plantair 1.201-02 5.t7B-09 2.8-09 5.17B-OS S.733-09 TDtd less of ofditb 7.153-03 5.05-07 L.E-03 4.823-07 1.1 5E.08 pow er_ _ _ _ _ _ _ _ _ _ _ _

Low of RBCC'V 1.10E-02 1.87"BS 2.05B-09 6.01-038 5,97B.09 Los ofmrw cooling 7.953-03 8.84U-08 1.35B-08 6.77E-08 9.84B.09 M=Mncsy los of 7.56E-03 4.31E-10 8.453-11 2.05S-09 3.12ID10 affsi power .

Turblneg wlihL= 1.43B400 2.543-07 a.22z-08 6.75B-07 1.53B-07 Break outside 6.67B04 3.19B-09 1.423-ID 3.72-8 2.32B-09 coutsinment ExcaivscLOCA 9.39-09 939B-9 9."B-69 9.39O..09 9.39B-09 LnaCingays _ 4.641348 4.643-08 4.49-08 4.64, 4.64B.08 LOCA __ _ I__ _ _ I__ _ __ _ _ _ _

U SpruylineAbreak 1.573-06 12.32H-09 1 L47B-11 2.32B-09 E.7-1I 10-23

]EDO.33047 -RevIylon 0 Table 10-4 Summary of theitiator Contributions to CDF atd LBRF (oonfnued)

U&I: IntUtosr utenor coifbum to CDP nad LERF InIator Catqory Men ePU CDF LZR CF LZRF

. jit C B brik Srlines 1,57E-K 4.05M09 1.5Si-10 4.05B-09 1.52,10 Rech don drchwe. I.10B-05 6.53E-09 1.595-10 6.S38-o9 1.59E&0

.li Abnk ___ .

Redmuatic diaebarp 1.103-05 5.24E-09 4.66E-11 5.243-09 4.66B-11 li B bAk RedIalonosuadcn 785E-07 3.55B-10 o.oOE+V0 3,55B-10 O.004B00 li=n biank RecirclaIon =uton 7.85E-07 3.5M-0 0.00G400 3.55-10 o.ao0+o00 lint B breakc Od aIorgeLOC& 1.573-06 7.563-10 3.14B-12 73-1BB0 3.14B-12 Medium LOCA 4.OE-05 221FB08 4.'36B-09 221B-08 5.3DE-09 Small LOCA 5.0E-04 1.1 &09 9.54-1D 1.15B-09 9.44s-10 VeYr small UOCA 338E-03 2.713-10 7.1IB11 1.05E-09 1.77B.10 (RaCi:UImiDGon pufp seal E lood in Reactor lCW 1.203-05 1.94-10 L.9909 3.45B-09 5.43E-10 Dufld1g - siutdown unit _

E8 flood in RatC 1.70-06 3.113-10 4.48E1 1.115-10 0.00B4X00 Buildir~g - upuiting tnit Flood ftomhle 9.103-05 2.322-09 '7.07B-11 2.29B09 7.03B-11 czsad r ao stank ._ _

P1od ft= the fsm 1.3430- 4.55E-09 1.75-10 4A02s.09 1.2E-09 Largc tbtlnc buftdlnt 2.Z0E-03 2.06B-08 3.43B-09 1.623-08 2.97349 iffood Small tmbie bufld& IA4E02 2.52-08 2.0B3-09 7.533-08 5.643-09 flood 10-24

NEDO-33047 - Revidon O Table 10-4 Browni Ferry Summiry of the Inittator Cotributions to CDF and LERF (conUnued)

Unit 3 Inlator category contribution t CDF and LXER InItior Catnory Muem frtqu.acy Budase (6VuW pe yuar) I J CDF LERF CDF LERF rAdvert oopezigooMSRY oE- 4.36E.2 S5 1.3o0- 5A3RSD8 2.04BO Iadvestvt opeing ortwo orsore 3.42E 04 Z.94E-D9 7.4411 4F-D9 7.92E-1l MSRV_

Inadveztent SCRAM 2.576-01 2,93-Q 4A7B4O 9.42B-D5 4.35EB9 Lon.of 500 kV toplant 9.32E-03 1.ISE-Q08 230E-09 4,461208 4.33B-09 Loe of 500 kV ton=imit 3.42B-02 4.15E.OS 5.13B 09 1.69E-07 1l41,i-0 Locs of I&C Bus A 4.10E-03 633E&09 4.64B- 10 2.12E8-O 1.75-09 Loss of l:C Bua B 4.I0.03 6.33B&Q 4.648-10 2.1ZZ-09 I.75E-09 Lics af all cordante 1.24202 1.448-08 1.908-09 B.13E-05 1.306-05 Lows DfoondomobWt snk (MSWV 12=01 5,74-08 2,961-08 5,673-07 7.29B -8 closui tabine trip wiout bypur, Io"odcxT=Y)

Low of FW 4.81B-02 1.135-08 7.93E-09 5.15-08 I.1OB OS LDII ofplAntair 1NU2 4.09-09 1.86U-9 5.158-05 5.3 1B-0 Toul lose of oitiit power 7.153-03 1.078-06 3.210-08 1.058-0 3.IRE-O8 Lms of RBOW 1.10-02 1.89E-D3 I.188-09 6.02E-O8 5.668-09 LOSI Of MWoDoingw SItT 7.95E-03 7.74E-D8 126S-08 8.03E-O 1.46E-08 Mo=ntyloloofottbpowe' 7.56E-3 497B-IO 8.44E-II 2.13E-09 3.1M780 Turbli tdp'weithbaypose 1.43B+O 2.70B-07 8.378-08 7.05E-07 1.58B.07 Brea cuxdcanfoJnment 6.676.04 4.058-D9 1.453-1D 3.73E-09 2AOB-09 ExomfatLDCA .9209 9.39-09 9.39B-09 9.39B-09 9.39B Og htoroing symm LOkCA 4.646-08 4.64E6-8 4.646-08 4.64BM0 4.64B-08 Cmrn Spray line A brosk 1-7-6 244B-09 4.72B-11 2A4E9 4.7211 CMoPMYUnoR bn~k 1.578-C6 355-9 1.08-i 3.55B-09 I,04B-ID Recirculaon disebaugo BineA break I.10805 6391309 4.252-11 6.38B-09 4.235-11 10-25

NEUDO33U47 -Rev~alan f Table 10-4 Summary of thU Initlator Coutributlons to CDP and JAR? (continued)

Unit3: Jatatoi cauory conafrbudon to core damsp.frequeacy xad 123 InIeor Catery Mean frapencyq B.lla. EPU Recrlation dUchdre 1.103OS 6.42E-09 4343-11 6.42B-09 4.341311 lin t Bbr k_ _ _ _ _ _ _ __ _ _ _

Reairulation rucdon line 7.85P-7 3.69310 0.003400 3.682-10 0.001+40 A brak Recirulaon auction line 7.85B-07 3.688-10 0.003400 3.683-10 D.OS-fO4 3 break Otbo Wgp LOW A 1.573-06 7,9711-10 OW 7.97510 0.0OEK+

Medium LOCA 4.0B-05 2.232-08 5.223-09 2.23E-D8 S22-09 Small LOCA 5.00304 1.25B-09 9.39B-10 1.26EOg9 9,29B-10 Vzy rual LOCA 3.35-03 3.09B-1D 7.123-11 1 l.lO0-9 1.79E-10 QRsiroulaton pump ueal lBCW lood In Rzaco 1.20B-02 8.99D10 1.953-10 3S95-09 5.651-10 Building - sliudowt unit EBC folod in Rector 1.70-06 2.82I-09 8.31 -1 I 2.59-D9 3.22E-10 Building - cpenting unit . .

Flood fom the 950305 2.36B309 9.7-1 1 2.345D9 9311-21 conden storage tak Flood frman thru 12343.S 2.95B-03 8.74&10 2.726-0 8.79B-09 Luge Wnebuilding 2.20303 lE77B-08 3.16-09 1.952-OS 4.101-09 flood Small tinbine building 1.443-02 4.312OBS .163-09 1M.?07 1.54M-08 flood . ....

10-26

... .. ;- .- r-*. *:r~r .  %.- r - *- *.-

NEDO433O47 - Rhvlon 0 Table 10-5 Frequency Weighted Fractional Importance to Core Damage of Operator Actions USed in Browns Ferry PRA Difab t I gpp FfKlPqW ItvdFc VFrnquncy.We FlPrudb ViAu~latla puratrActlnaDecrplJo Irwin? Fradkonallmporanuce ["ewn Ifracdonudlmporlunco IDMlIRS to (bra Damap to Coro Dmmags aulDU2tsrzdna nuvj PA EPU D VBate U3 ZPU HOVD Drwui1t.05f Wui b.01 3.58-01 4.52-0 6.28-02 4.3r-01 34.dE01 MUPIPgMSRVw I __

HOLP2 OpruiorPalls tblaldesWo 19E 0I a3S-02 -1.03-01 1.7B41 9.1t.02 .742.02 1Nt{ui Suppenftoa Pool

_ __ Cooling _ _ _ __ _ _ _

HOSPI AliF RMR for Bcppreuion 1.2-01 5.D0-2 -6.7F-02 9.6002 4.6B-D2 -S.0DBD2 Fool Cooling UI2 Alp Alteae bIrIOon tD. I D.01 &7E.C2 .5.72,M 0.0E+00 0=+00 NIA RPV'vat l~oUnit I/Vn~tl RCvvj~t641ttML Coaoiea _

HOU12 MainLiaRFVLvyo 23B 02 3.78-0: 1.3"-0 1.38-02 0.0l+00 -1.38-02 W/Art~ue Sozuzc, SP RINO HDR Flood HOSLI Iaikto LCS CIven ATWS YES 1.2I-02 2.2e-02 9.4B-D3 B.1B03 IE-f2 _.130 WKUnjeolstw RPV .

HORP2 SutFRaIMMPump tbrPC 4.1.-D3 1.3802 9.D003 3.D8D3 14542 73-03 Li 9Sir Not AiDWpaepd I HOSLI Ida SLCS, Gien m AWS Y 6.41-03 1.28-02 5.98-03 4.1203 9.7B-03 5.6E 03 vwlh RPV h td .o_ .

HOSV1 Ddc MhVClom Logi 2.4B302 1.18-02 -1313502 16.BD-02 8FO -7.-0

___ATWS wiETn*he Trip _ __ ... ....

HOED! DqruAniwith te Ttrblne 6.3304 tAB-O3 7.78-03 4.5-04 6feDU3 6,2F3-0 3

yPw8 Valvw t u ,of IltHpad RCIC HQADI Indbk1ADS Duing ant ATWS YES 3J4-03 50-O03 24A-03 23-03 4E-UD3 2313-43 HOLA1 Manul Cozvolt oftow 9.5-03 4.7B-03 -4.03 5.26W3 3.7503 .2.5B-03 Prwur IrowronDuri HOSW1 TunsfctrModoSwlflhlo 3.B0303 2A-3 *7.3E-04 2.4E-O3 2.0B403 -3.5-04 Rflrllunwdown HOSP2 AlIgn RHRIb: Suwrenaa 3.303 1,9BD3 *2.IB43 2.5B-03 1.03 -1.1M-03 Pool Co Am . . .. .

HOLPI CDnmo1 RPV Level at Lw 7.03-0D3IA 3 *S.43-03 733E-0 2.13.03 -.5.703 PrruroUinJ RHR for 0r SPAy IHO---- I~e and oomirl Vewxl 6,913-0 1,5B-W I.6B 04 4.3B-04 I2 0IM 7.6B-44 LCrl .C_.

10-27

. --. - . - ..-- :-r: *n* -. -*- -. -- - '.~~? - - -.

n .. . . .

NEDO 33047 REevhlou 0 Table 104 Frequency Weighted Fracdonal Importnce to Core Damage of Operator Actions Used In Browns Pery PREA (continued)

DahbuAk H Pwsq~umntyWtgi Frudoan Fraern,'iWdghtd Fr~acin Virlable Oporatr Aotln DmfnlpflurBP Fr h ce lom l imperUm" l p ]

-__tzonu Dmininto bCor Damq a

• lR l 8Br IW UBData l3 IEPUl l HOOU11 Cm-tict I Pump. dAXX 3.1B-03 1A483 *1.78-03 L.1B01 0.oE4) .1.13-01 pto Urs2TormU9 .. I AM MbtI ADS, ATWS, Isoated YES 2t3S 4 9.0A9 6.6304 1B-04 7.IB-04 5.5B-04

_ __ vendl_ _ _ _ _

HOBB1 umnd Sta RlRSW Sing 7.DS-04 3M3F4 4.7E-04 53B-04 3.OE-04 -233-04 PpAferALOS wilt eptial 2CW 0ak4 Lkvd ccmfl usdg ATWs U&0M4 6zev5. 4ss.05 7.7305 4.F-O05 -2.98-5 HQRP1 StutPRn &CS FW Ibr O.CBm 0.O+t 0.50450.oalo 2n/os -L7005 LPCL L1 SIpl Noj

_ _ ttdpatad Table 10-6 Results of Browns Ferry PRA Peer Review PRA ELEMENT CERTIFICATION GRADE INITIATING BVENThS 3 ACCIDENT SEQUENCE EVALUATION 3 THERMAL HYDRAULIC ANALYSIS 2 SYSTEMS ANALYSIS 3 DATA ANALYSIS 2 HUMAN RBLIDI1fY ANALYSIS 3 DEPENDENCY ANALYSIS 3 STRUCTURAL RESPONSE 3 QUANTIFICATION 3 CONTAINMET PERFORMANCE ANALYSIS 2 MAINTENANCE AND UPDATE PROCESS 3 10.2e

' 's ^ .-

• a_. @ , . . ..

NEDO-33047 - Reybian 0 Table IN07 Browns Ferry FAC Parameter Coomparhmon for EPU CRECWORKW 105% OLTP CPU Paramearw AIlowble Input TypIal Rouge TypeW Rwp of Vues of Values Stem Flow 1.100,000,000 767,000 to 930,000 to 13,B40,000 (Ibm/hr)i 11,800,000 VClo0ity(ft/se) Calculatedin 122!to 169 132 to 171 Stnm Quaaity 0 to 100 92.6 to 97.9 92.8 to 98.2 Opetadn 0 to 7SO 312 to 388 314 to 403 t=mpetum m I _______II 10429

NEDD-33047 - R4vWon 0 11, LUC:ENSING EVALUATIONS 11.1 OTHER APPLCABIJE REQUIREMENS Tho analysis, design, and implementation of BPU were rviewed for compliance with the current plant licensing basis acceptancs criteiL and for complisnce with nrw regulatory requirernzub and operating experience in ths nuclear industry. J[

)) The associated tables identify the issues that are generically evaluated, and issues to be evaluated on a plant-unique basis. Te applicable plant-unique evaluations bhve been performed for the subjects addressed

below, 11.1.1 NRC and Industry Ctosmunlcations All of the issues from the following NRC and industry communications are either gencrically evaluated in ELTR2 (as supplemented), or arc evaluated on a plant-specific basis as part of the EPU program. Those evaluatious cozclude that every issuo is either (i) not affected by the BPU, (2) already incorporated into the generic EPU program, or (3)bounded by the plant-npccffic EPU evaluations. The NRC and industy comnunications evaluated cover the subjects fisted below.

• CPRs a NRC IMI Action Items v NRC Action Iters (Formerly Unresolved Safety Issus) and Now Generic Issues a NRC Regulatory Guides

• NRC Generic Letters

& NRC Bulletins a NRC InformationNotioces a NRC Circular:

• INPO Significant Operating Event Reports (aplicable to theEPU)

• GE Services Information otters a GE Rapid Information Co 0mm icafin Service Iformation Letters 11.1.2 Pant-Untque Items Plant-unique items whose previous evaluations could be affected by operation at the EPU RTP lovel have beon identified. These are (1) the NRC and Industry communications discussed above, (2) the safety evaluation for work in progress and notyet integrad into the plant design, (3) the temporary modifications that could have been reviewed prior to the EPU and still exist 11I

NEDO-33G47 - ReyIslca 0 after BPU implementation, and (4)the plant BONs. Thoes ites will be rviewed for possible effccty the EPU, and items acted by the BPJU will be revised prior to EPU implementation.

11.1q,11 Commitments tob.s NRC Prior to BPU impleneaation, the potntially power dependent NRC coinnmitmenta are reviewed for required changes due to BPU conditions prior to BPU implementtion. T1e commnitents that ae &Ad by EPU wiIl be update to account for the effects of BPU.

Il.I.2.2 10 CFR 50.59 E~valuatdons 10CFR 50.59 evaluations perfroned for work in progress and 10CFR50.59 evaluations completed but not yet included in the UFSAR are reviewed prior to EPFL implementstioa for trquirod changes due to BPU conditions. No 10CFR 50.59 evaluation process change is required for BPU.

11.1.2.3 TemporaryModtfncatons Pre-existing Temporary Modifications, Technical Operability Evaluations, Open Work Orders that will be in ffct afler EPU implcmentation will be reviewed and revised, if necessary, to include EPU conditions.

11.1+/-4 Emergency and Abnormal Operating Procedurem EOPs and AOPs can be affected by EPL. Some of the BOB. variables and limit curves depend upon the value of rated reactor power. Some AOFs mnay be affeted by plant modifications to support the bigher power level.

EOPs include variables and limit ourves, defining conditions where operator aotions are indicated. Soano of these variables and limit curves depend upon the RTP value. Changing Earm of the variables end 1imit curves requires modifying the values in the EOPs and updating the support documentation. EOP curves and limits may also bo included in the safety parameter display system and will be updated accordingly.

The charts ndWtables used by the ope S to pedorm the BON are reviewed for my required changes prior to each core reload. Th EOPs wer reviewed for any changes required to implement EPtJ. The operators will receive trainig on these procedures as desoribed in Section 10.6.

AOPs include event based operator actions. Same of these operator action. may bf influenced by plant modifications required to support th increase in rated reactor power. Changinig some of th1 operator actions may require modifications to the AOPs and updating the support documentaion. The plant AOl's wert reviewed for any ffecs of power uprate and no change to t11L event-based actions arm required. Some of the seipoint. used in the AOPs chango duo to EPU. The operators will recei've training on these procedures as described in Section 10.6.

11-2

5:

• NEDO-33047 - ReisyIo 0 The plant BOP are reiewed for any effmcts of the EPU, and the BOPs will be upduted, as necessamy. This review is based on Section 2.3 of ELTR2, which includen a list of operator action levels, which are sensitive to the BPU.

11.2 REFPRENCES

1. GE Nuclear Energy, "Generic Evaluations of General Blectrio Boiling Water Reactor Extended Power Uprac," (ELTR2), Lioensing Topical Reports NEDC-32523P-A, Clus III (Proprietaiy), February 2000; NEDC-32523P-A, Supplement I Volume 1, February 1999; and Supptement I Volume la, April 1999.

11.3

ENCLOSURE 6 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 2 AND 3 PROPOSED TECHNICAL SPECIFICATIONS (TS) CHANGE TS - 418 -

REQUEST FOR LICENSE AMENDMENT FOR EXTENDED POWER UPRATE OPERATION BFN EXTENDED POWER UPRATE UFSAR REVIEW MATRIX This enclosure provides a matrix identifying sections in the UFSAR that are currently under evaluation for change for EPU implementation. TVA will complete the final UFSAR changes following approval of this change.

See Attached:

Browns Ferry Extended Power Uprate UFSAR Review Matrix

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4 X N N X X X N N N X X AppendixR All Sections lComments lNotes

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