ML18024A891

Amend 46 to License DPR-52 Changing Tech Specs to Permit Facility Operation in Cycle 3 Following Refueling Outage

ML18024A891
Person / Time
Site:	Browns Ferry
Issue date:	05/25/1979
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML18024A892	List: ML18024A891 ML18024A894
References
NUDOCS 7907130411
Download: ML18024A891 (52)

Text

UNIT 0 "-YA es NUCLEAR REGULA ~ CRY v ts'iiV>ISSICiI ivASYilNGTQI'LD. C. 2~"""a TENNESSEE VALLEY AUTHORITY OOCKET NO. 50-260 BROllNS FERRY NUCLEAR PLANT, UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 46 License No.

OPR-52 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Tennessee Valley Authority (the licensee) dated February 9, 1979, as supplemented by letters dated May 15,, 1979 and May -16, 1979, complies with the standards and requirements of the. Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; 8.

The facility will operate in conformity,with the application, the provisions of the Act, and the rules and regulations of the Commission; C.

There is reasonable assur ance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; 0.

The issuance of this amendment will not be inimical to the common defense and'ecurity or to the health and safety of the. public; and E.

The issuance of this amendment is in accordance wi th 10 CFR Part 51 of the Commission's regulations and all applicable requi rements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in,the attachmen to this 1 cense amendment and paragraph 2.C(2) of Facility License No.

OPR-52 is hereby amended to read as follows:

{2)

Technical S ecifications The Technical Specifications contained in Appendices A and 8, as revised through Amendmen 'o.

, are hereby incorporated iin the license.

The Iicensee shall operate the facility in accordance with the T chnical Specifications-qoovzso I(( '

l5

3.

This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION

Attachment:

Changes-to the Technical Specifications, Thomas A. Ippol to, Chief Operating Reactors 8ranch 43 Oivision.of Operating Reactors Date of Issuance:

May 25, l979

ATTACHMENT TO LICENSE AMENDMENT NO. 46 FACILITY'PERATING LICENSE NO.

DPR-52 DOCKET NO. 50-260 Revise Appendix A as follows:

1.

Remove the following pages and replace with identically numbered pages:

vii/viii 7/8

+910

~15 16 17/18 19/20

~21 22

~23 24

~25 26

+29 30 71/72 101/102

~113 114 13+11 32

%33/134 139/140 159/160 167/168 169/~70

~181 182

~219 220 329/330 2.

The underlined pages are those being changed; marginal lines on'hese pages indicate the revised page.

The overleaf page is provided"for convenience.

3.

Add the following new page:

172a

~

~

I

'I

LIST OF TABLES Cont ad

'Pa b1e 4.2.F 4.2.G 4.2.H Title Minimum Test and Calibration Frequency for Surveillance, Instrumentation Surveillance Requirements for Control Room Isolation Instrumentation Hinimum Test and Calibration Frequency for Flood Protection Instrumentation P~ae No.

105 106 107 4.2.J.

3.5.I 3.6.H Seismic Monitoring Instrument Surveillance 108 Y~LHGR vs Averse P'anar

+posure q~,y7p, 172a Shock Suppressors (Snubber

)

90'.6.A Reactor Coolant System Inserv ice Inspection Schedule 209 3.7.A 3.7.B Primary Containment Isolation Valves Testable Penetrations with Double 0-Ring S ea 1 s

~

~

~

~

~

~

~

~

~

~

~

~

~

~

~

250 256 3.7.C 3.7.D 3.7.E 3.7.F Testable Penetrations with Testable Bellows Primary Containment Testable Isolation Valves Suppression Chamber Influent Lines Stop-Check Globe Valve Leakage Rates Check Valves on Suppression Chamber Influent Lines

~

~

~

~

~

~

~

~

~

~

~

~

257 256 263 263 3.7. M 4.8.A 4.8aB 3.ll.A 6.3.A 6.8.A Tes tabl e E 1 ec trica 1 Pene tra tions Radioactive Liquid Waste Sampling and Analysis Radioactive Gaseous Waste Sampling and Analysis Fire Protection System Hydraulic Requirements Protection Factors for Respirators Hinimum Shift Crew Requirements 265 287 288 324 343 360 Amendment No.

g$,

46 vii

LIST OF IL'liUSTRATIONS

~FI Ure 2.1.1 2.1-2 4.1-1 Title APRM Flow Refer. ence Scram and APRH Rod Block

Settings,

~

~

~

~

~

~

APRH Flow Bias Scram Vs. Reactor Core Flow Graphic, Aid in the Selection, of an Adequate Interva'I Between Tests Page Ho.,

26 4.2-1 3.4-1 3.'4-2 System Unavailability Sodium Pentabo rate Solution Requirements Sodium Pentaborate Solution Requirements Volume Concentration Temperature

~

~

~

~

~

~

~

~

~

119 139 3 5'.2 f Factor K

173 3.6-1 3.6-2 6.1-1 Minimum Temperature

'.F Above Change in Transient Tempera ture

~

~

~

~

~

~

~

~

~

o 188 Change in. Charpy V Transition Temperature Vs.

Neutron Exposure 189 TVA Off'ice of Power Organization for Operation of t(uclear Power Plants 361 6.1-2 6.2-1

6. 3-1.

Functional Organization........

~..

~..

362 Review and Audit Function

~

363 In-Plant Fire Program Organization 364 Amendment No.

35 vii 1

10.

~Lo Lc - A logic is an, arrangement of rclsyg, contacts, and other components that produces a decision outout.

(a)

~zniciacin

- a logic chan rscaiva

~ ignals rr'on nhannals and produces decision outputs to the actuation logic.

(3)

Actuation - A logic that receives signals (either frocg initiation logic or channels) and produces decision outputs to accocgplish a protective action.

M.

Functional Tests -

A functional test is t'e manual operation or initiation of a s7stemg subsystem, or component to verify that i functions vithin design tolerancas

( ~,g.,

the annual s art of a cora spray ~ to veri!y hat it runs and that it pumps the required volugsa ot'atar).

X ~

Shu'td~

The reac tor is '

a ahutdovn co"di rien Mhen tho r"ac tor uode Wtch is in the shutdcvn aoda posit'o" and no cora alterations ara being performed.

Y.

En ineered Sa!eeuard - An engineered safeguard is a sa!ety systan the actions oE vhich are essential to a: safety action required in response to accidents

~

Cumulative DoMntime - The nw~~tive da-ztgme !or those safety opponents and systems Mhoae dcvn issue is 1Ldted to 7 consecu ive days prior to requir'~ reactor ahutd~ shall be

~g ted to any 7 days M a oc"secutive 30 day period.

SAFETYLlilIT'IH ITIN SAFETY SYS E~f S~i. TI'i'0 1 ~

'riEL CLADDD!G INTEGRITY A licabilit Applies to the interrelated; vari'-

ables associated Mich. fuel thermal behavior.

2 1

FUEL CLABDIhC Il~cEGRI~ Y

~Aplicalii1i~t Ap'plies'to'trip settings ot, the instruments and devices uhich ar' prov',decl tc prcvcnc: the r=actor systec: <<afety limit~ from being exceedecl

~Gb ective To establish limits ~hich ensure the integrity of the fuel -clad-d icg.

To def ine the level of the process varia'bles at cihich hutoriatic pro-tective action iu initiated to pri'-

vcnt the fuel claCdinc~ h;.tcgrity safety licit froti oeir:g, exceeded<<

S ecifications A.

Reactor Pressure.

> 800 psia and Core Flov > 10X of Rated.,

Vnen the reactor pressure is greater than 800 psia, the existence of a ainirca cr'ti-tal po;er rctio

(~,"CPR) 1 ss.

c.hac 5.'.07 liall constitute violation o f th a fuel c,l.add ic integri~ s-fety li&'.

S ecification The luaiting.,<<Erty systeo setting" 'hill be as specified beloM:

A.

Neutron Fl.ux Scca 1.

APRN Flux, Scram Trip Setting (Run linda)

,ll.'lien ch Hode SMItch i'" !n th.

RD?l position, the A"?.".

,'flux scram trip setting shall be:

S<(0 66V + $44) eche re:

S Sate:ing in peccant of rated the~1 poplar (3293 lÃt)

M '~ Loop recircuiaticn Elm'a e in pe".cant of rated (rated loop recirculation finis rate equal,s 34.2-106 )Lb/hr)

Amendment iso.

3~,

46

SA<'i".i'7 ).I ~1JT FII.-:l. Cl.hnl) Ill<, 1NTVf:RTTY P. I FVEl. CJ.hnDIfJC I NTFm);ITY

~ Ie the event of oI.aration w'th the core maximum fraction of limiting power density (CHFLPD) grenrer than fraction of rated thermal power (FYJ')

the setting shall be modified as follows:

S~ (0.66W + 54Z);RP CMFLPD For no combination of loop recircu-lation flow rate and co"e the~

'os~r shall the APRM flux sera.= :"..'"

setting be allowed tn exceed 170'f kated thermal power.

(Note:

These settings assume operation within the basic thermal hydraulic design criteria.

The. e criteria art I,IIGR <- 18.5 kw/ft for 7,.7 fuel ard-13.4 kw/ft for SX8 and SXSR fuel, MCPR within limits of Specification 3.5.k. If it is determined'hat either of !hest design criteria is being violated during operation, action shall be initia ed within 15 minutes to restore operation within presc ribnd 1 imit,:.

Surveillance requirements for,lpf'cram setpnint are given in spec ifi ca t i on 4.1.B.

2, APLI Vhen !hc reactor mode sW!ch is in the STARTVP POSITION, the APRM scram shall be set at less than or equal to 15X of rated power.

B.

Core "heaial Pove.

Limit (Pear t

". Pressure

<800 psia)

';.~ e.; the reactor pressure is less tho..n or equal to 800 psia, 3.

IF~I--The IRH scram shall be se! a!

less than or equal to 120/125 of full scale.

B.

APRM Rod Bio=I. Trio Settin~

The Apl'Ui Po" b I ck trio settp an ~n;? I be:

'mendment No.

3~

3B 46

':iAFFTY LIIGT-',YNYYIN0 YIIY'I;YY NY.",,YNII Sill!TZYNI",'.1 FUr.L CLADDIHG INTEGRITY 2.1 IFVEL CLAIDDIIIG I?P'i'EGRITY'r core coolant flov's'han

'10$ of rated, the core thermal pover shall not ex-ceed 823 t&t (about 25+ of rated thermal pover).,

RR< (0.66M + 42~~)

vhei c:

Bod blocR setting is percent c!f rated thermal pover (3295 FQt)

V

= Iu~p recirculation flov ante

.'in percent of'uted (rated loop recirculat ion flcv rate e,equals 34 2 y 10 lb/hr)

In the event of operation vit]i the cor" maximum fraction of limitin" po"er density (C"ZI.PD) grca ter than frac t "on o F rated thermal pover

(,Fi:1.') the set tin~, hall, be mcidificd as follows:.

S Y (0.664 +6'22)

FPZ R.B CYIFLPi)

C.

4'henever the reac or is in the shutdovn condition vith irradiated fuel.;In the. rea'c-tor vessel, the vater level shall not -be 1 ss than.17.1 in. above the top of. the normal active fuel zone.

C.

Scram ud isc equation > 538 in. above reactor lov vater vessel zero levY S"rar 'ur'>Yn~ stop

< 10 percent valve clcsure valve closure E.

Scram--turbine control va'} re 1.

Fast. closure IJpon t.rip of the fast actin;"

solenoid valve Y.

2.

Ix!s.s of control

> 550 psic.

oil pressu e

IF.

Scram lov con-de,nser vacuum 23 inches Hg vacilum G.

Scrammain steam.

< 10 percent line isolation valve closure H.

Yain s'earn i:solation

> 825" paid Y alVC C1OSure nuClear SyStem lOv pressure 10 Amendment Ho.

32

hASF5:

. IIFL CLADDI!IC II:TFCRITT SiiEETY LIMIT The fuel claddtng represencs one of the physical bar act tve mater ia Ls Eraa environs.

The inc egr icy of th t c1 add

< q s ca arr ers vhtch eparate zadfo-relaced co ica relative freedom frau erfo a i f

corros tan ar use-related cracking, may occur du i h

1'f fsstan product migration from chts source is ur ur ng t e ixfe of ehe claddtnng

~

u ce is ncremeneaU.y tv~lative aad conctnuously measurable.

Fuel cladding per!or c

h-h p r.arations, ovevur.

can result

!rom t crawl stresses'vhfch occur from reactor aperacio i tf on s ~

tcanely above des,t n

and ehe pzocectian system setpoincs.

Mhtle ftssfon product af raci cladding perf ormatfan fs )use as measurabla th f

c ch 11 sed claddi rfo tio i

al h

h 1

ura a as t at from use-related crackfn gn a t res old, beyord vhich still greacer thermal scresses may cause gross rather than incremental claddfn dete eton.

Therefore, the fuel cladding safety lfmf't f operaCfng conditions vhich can resulc fn cladding pei forat tan.

The fuel cladding integrity Ltmfc is sec such th c

1 1

ed ac no ca culaced fuel damage vould occur as.a result of an abnormal operational transi t

B noc directly observable, the fuel cladding, Safet L'

t d

rans eat.

ecause fuel dna 8 co the conditions vhfch vould produce onset cransftf b fli

(

'ng, a ety imit ts defiaed vith margin This establishes a Safety Limit such that the mtafm rans t on oflfag (KCPR af 1.0).

ia o

1 e

m a um critical paver ratfo (NCPR}

a no ess chan g,07, MCPR >1.07 represencs a conserv ci Che conditfans required co maiataia fuel claddfag fat i

a ve marg relative to Cegrity.

Onset of transition botltag results in a decrease f

h a

d c

crease n

eat transfer f"om the clad an, therefore, elevated clad temperacure and Che pos iblit f

Since boiling cranotcfan is noc a directly observ bl pass ty o c ad failure, serva e parameeer, eha margin to botl ing transition is calculaced fram pLant op n

operating paraueecra sucn as core

pcrier, core flov, feedvater temperature, and core pover distribution.

The mar for each fuel assembly is characterised by che critf 1

c ca paver ratio (CPR) vhtch fs the ratio of the 'bundle paver vhich vould produce divided by che actual bundle paver.

The mfntmum 1

f h'ro uce onset of transitfon boilin8 n

mum va ue o

c is ratio for any bundle cho core fs the mfnimum crfcfcal paver ratio (L'CPR)

L i plan operation ts concrolled to che nominal procecti

. e t

s assumed that the o ec ve secpotncs. via che tn tru-menced variables, t.e.,

normaL plane aperacion pres t

d Ft i.

en e

an tgure i.l.l by the n~Ina I ey~er cpA fLnv control

Liuke, ae'a f~tv I t<tt Q'IC t

QICPR of l.07)ha<

~u<<<ctene

<<onservac sm eo assure ehac in the event of an abnormal f

arma operational transient inic tace~.

fram a normal operatfng condition

(.'ICPR > limit ifi in s s speci ied n specification 3.5.k) more than 99.9~

of the fuel

. on.

e margin betveen rods tn che core are expected to avoid boiling transition Th

~CP:I af L.O (onset of cranstt tan botLing) and che safety. limit 1 07 i,d s,d erived fran a detailed staefsti al analysts canstdertng all of the uncereatnt tea tn monf-tortag the care operaetng state including uncertainey tn the b tli n

e a

ng cransxcton correlgeton as descrtbed tn Reference l.

The un~ertatntte l

ed t d i che safety limtc are provtded aC the begtnntng of each fuel cycle.

Amendment No,~ 46 15

.l.I hag gg Because the boiling transition correlation ie biased cd e ln"ge qucudc,ity full scale data there is e very high confidenc that operation of Is i'u~

assembly at the condition of MCipH =1'.07 v'os,d isot pz'oduce boiling tran--

sition.

Thus, although it is not required to cata'blish the safety li't.'dditional margin exiata hetveen the earettr limit and the antnal Qeufenee of loss of cladcling integr.ity.

However, if boiling transition vere to occur, clad perforation "ould no be expected.

Cladding temperatures waul'd increase'to. approxbuately 1100oF vhich i.e below the perforation temperature of the cladding material.

This ha'een veri fied by tests, in,thc General @ectrztc Test Reactor (GETR) where fuel similar in design to BFHP operated above

'he critical heat. flux for a significant period of ti.,e (30 minutes'ithout clad perforatione Zf reactor pressure should ever exceed 14'00 psia,duz-ing ao. '>1 pover operating (the limit of sipplicebility of 'thc'. bibiling transition corz-e-lation) it vou.ld be assumed that the ft~el, ciadding integrity'afety Li zit hao been violated.

In addition to the boiling transition lieiit'M'CPR ae 1.07) constrained to a maxir;um L1lGR of 18.5 kw/ft for 7x7 fuel aud 13.4 tcw/ft for 8x8 and 8x8R fuel. This lirzit is reached when the Core Maximum Fraction of Liiniting Power Density equals 1.0 (CMFLPD 1.0).

For the case where i,ore Maximum Fraction of Liriting Power Density exceed>>

the Fzactiorc of Rated

'herzoal

Power, operaLion is permitted only at l,ess than 100K of rated power ond only with zeduced APRM scram 'setti'ngs, as req i d b

. if

2. 1.A. 1.

u re y spec ic',ar.ion, At pressures below 800 psia, the core el~ati'on pzeosure drcrp (0 0 flow) io greater than 4.',56 psi.

At lov povers. and flovs this pressure differential is maintained in the bypass region of the core..

Since the pressure drop in the bypass region is eseez>titslly all, elevation

head, the cora pres, sure drop at lov povers and flow vi11 alva s* be greater than 4.56 psi.

Analyses shov that with a filo~>> og Q8K,O~ lbe/hr bund1e flov, bundle prc cue.e drop is nearly ind~epe~ndc.nt of bundle pen. r and kiae a value of 3.5 psi.

Thus, the bund1e flow with a 4,.56 poi dri~ing 'heed vill be greater than c!8x10 lbs/lu..

Full scale ATLAS test data taken

't press<uco from 14.7 psia to 800 psia indicate that the, fuel aeaezsbl'ritical power at this flov is approximately 3.35 NA.

Pith the desi~

peaking factors this corresponds to a coz"e therme1 powmr of, r."c than 5 g..

Thus, a core thermal pover lizrdit of 20$'or reactor pressures beLot>> 800 peia io conservative.

For the fuel in til~e core during periods when the zeector ie shut down, con-sideration must also be given to water lci>>el requirements duc to the effec:t this t me cif decay heat.

If ve.ter level shout,d drop below the top of the fu 1 d ime, the ability to remove ciecey heat is redluced.

This rcd

,e Llr ng coolin ca abilit p

y could lead to elevated cladding temperatures and clad

', ce o rc uction'n A. long as the fuel rerd~ins,covered

~with water., sufficient,,

perfcration.

Ao 1 cooling is available to prevent fuel clad pex-foz ation.

Amendment NO.Q2, 35 d

'<6

The safety limit has been established at 17.7 in. above the top of the irradiated fuel to provide a point vhich can be monitored and also pro-vide adequate margin.

This point corresponds approximately to the top of the actual fuel assemblies and also to the lover reactor lov vater level trip (378" above vessel sero).

RZ$ 7'.ROICE l.

Con'eral Electric bLR Thermal Analysis basis (CKTAB) Data, Correlation and Design Application, HEDO 10958 and NFDE 10958.

2.

General Electric Reload Licensing Amendment for BFNP Unit 2 Reload No. 2, NEDO-24169.

January 1979 as amended by NED0-24169A.

Amendment No.

35, 46 17'

i>

Jl PAGE DF.L":.7F.

"~ 1 Bh5K3:

f I ITDic. zgv sY."T..N sETTINcs R..LATFD To FUFL cL~0 ~ yo ItlTicgii(

able ta opecraccon 0

c e Bra rc V bucfear Plant have been analyzed throughout the spectr of c re o

p armed operacfng ran-dftfons up to the, design thermal paver rardfclon af ~!,Ir 'g!-

g

~

<nalyses very based upon plant operation fn accordance vith Che ap. acing map given of tne FSAf(.

In addition 3293 tfuc i th lf I

t e

censed aax faun pover level oj Bravns Ferry Nuclear Plant, aad this z'epresents the waxf ad paver vhich shall nat knavfugly be exceeded.

e max cue ste y-state Qanservatism is incorporated in the transient analyses in estimating the controlling Eactoc's, such as void reactivity coef Eici nt, conrrol rod scr~

vorth, scram delay time, peaking factors, and axial pover shaped.

factocs are selected conservatively vith resoect to their effect on the applfcable transient results as determined by the cure ant analysis model.

Thf s trans fenc modal, evolved. over many years, has been substantfated

'n opera-tion as

-a conservative cool 'or evaluating reactor cyrdm'.c performance.

Result" obtained fram a cancral Electrfc boiling vacer rc=ctor have been compared uich predfctions made by the radcl.

The comparisions and zesi its are summarfa~d in Reference l.

The absolute value of che vafd react fvfty caef f'fcfent used fn the analysis fvcly cscfmaccd to be about 25K great value expected cv occur durfng thc core lffetiac.

The sera~ vorth used has be<<n dcraced to bc equivalent to dpproxf.-..ately sfy CF che total scram bort~ ot chc control rods, The scram delay cfce and race af rod fnscrtion allover

~1>>>>> vc~i are consac vh tivcly.c t equ.il co thc langcs c dclgy ar d sfov<<

,est fneertian race acceptable by Technical Specfffcatfpna.

The effect af scc sm orth, scree delay time dad rod fneerCfon rate, all'onservatively applied dr f

are o

greatest efgnfffcance fn the early portion of the negative react'vity fuse tf Th o

negative reactivity fs assured by the time requfr~

C f

5" f

v user on, e rapid insertion e ".equ rmencs ac 5>> and Zog insertion.

By the C lac the rods, are 60K inserted, approximately fo " d ll f

tfvf,ty has been inserted vhich strongly turns the tr f

d y our o lars of negative reac-de<<fred effect.

The tf=es for 50r. and 90K fnsertf e

cans ent, and acconplfshes the completion of the expected performance in the earliez'ortion of the transient, and to aatablfah the ultimate fully shutdovn steady-state condition Far haalyses of Che thermal consequences of th>> transf specified in, specification 3.5.k is conserve to initiation of the transien s.

conservhtivelv assumed to exist nrior ransxents.

oncro) Lin arz ecc.s Jld inf c

~

Craasfencs at che design paver level, producr" aare pes f'-,

~

th a

e pess azc.c hn.".vers than voulJ re<<of by using expected values of control parve:

C

~

d L

paver I;vcls.

rdmecers on anaLyzfng ot hi "her steady-scaco operacfan vfchauc forced recfrc'ufac fan vill ant bc pemftc d

for wore than 12 hc an ours.

and the start of a recirculation pur,o irom th circu)ation conc)ition. wi)l not be oeri itted un)ess th ter n rs~

b t th lo t.o b

t

• t d d

h This reduces the oosi ti ve reacti vi ty insert '

ar e

an t e core coo)ant temoeraturc is less h

inser ion to an acceotabl y low va) ue.

Amendmert No..BL 35, 46 19

In surrinary 1.

Tbe liceased.

maximum.pcrver level is. 3,293 ?St.

2.

Analyses of transients employ adequately conservative values of the, controlling reactor parameterst 3,

The abnormal operational transients verite analyzed,to a,pover level, of 3440 H~:T.

4, The analytical procedures nov used result ia a more logical ansver than the alternative tcechod of assuming s higher ~scsrciag'aver ia co0$ u4cH tion vith the expected values for the parameters.

Tha, bases for individual set points are',discu'ss0d belovt A.

Neutron Flux Sere=a 1.

APRM High Flux Scram Trip Setting (Run Hode)

The avsrage pover range. nonicoring

('APiN), system, vhich is calibrated using heat balance data, taken duriag steady-state coadicions, reads'a percent of raced pover (3,293 NMc).

Because fissioa chambers pro>>

vide the basic input signals, the Q'1N'ystea responds directly to average nc:u"ron flux.

During transients,,

thc i"scantaaeous rate of heat transfer from. t'e fuel (reactor) the~1 pbver) 's less thaa the.

instantaneous neutron flux, due to the tMe constant of the fuel.

i Tharcfore during transieacs induced, by disturbances, the thecmaIL paver of the fuel vill be less than hit indicated by, the neutron flux at the scree setting.,

Analyses reported

~in, Seccioa 14 of the Pma1.

Safecy Analysis Repo=t deaoast "aced

,'ch'at,'vich,a 120 percea.

scram trip setting, noneof che abnormal operational transients analyzed v'ola =,

the fuel safety limit aad there is a subscsntisl,margin from fue'l damage.

Th refore, use oi a flov-biased sedan provides even additional Figure 2.1.2 shams the flou biased scram as a function of core flow.

An increase in the APE't sera=, set',cing vould decrease che margin pre; sent before che fuel c'cddin", it>>egrity safety linc is resch<<d.

The APRN scram-setting vas d<<e>oined.

by an analysis of margins required to prov'de a reasonable ran:;e for asaeuveriag during operation.

Reducing cnis operating aargLa vould increase t'.-.e frequency of spurious

~crsas, vhich have-an adverse e."'ecc on reaccor safecy becsus<<

oi

~lhe'esulting ther-.~l stresses.

thus, che Al??8 setting vss selected bacauss it provides ad'quate

.=argin ior the fuel cladding iat<<gravity

~afsty limit ye allcvs operating cLargia that reduces the possibilg ~, o<,

11aaocossa'ry'cr~s

~

20

2.1 enszs

\\

The scram trip setting must be adjusted to cns>>re that the LRCR transient peak is not increased for any combination of CMFLPD and FRP.

The scram setting is adjusted in accordance arith the formula in specification 2.1.A.l Mhen the CMFLPD exceeds FRP.

Analyses of the lmcKa8 cransienrs shoM chat no scram a" juscmsnc i co assure HCPR

> 1 ~ 01

+hen the transient is initiated from HCPR s limits specif ied in specif ication 3. 5. k.

APRN Flux Scram Tri Settin (Refuel or Scaft f!, Hoc Scandb gods)

For cperecton in ch". ecarcup aade vhile the rcaccnr is et lov pressure che APRi! ecrrm ee c inS af'5 percent of raced po-cr prov'.dcs adeuusce thermal cargtr.

becveen che eecpaint snd che safety 1inic, 15 percenc of raced.

The marRin ke adequate to accommodate. antic!oated maneuvers associated vich poier plant etartup.

Effects of increasing pressure at sero or loM vo!J cnncent are r irar, cold vater from sources avail-able du. ir8 scarc<

itch is placed in che RUE!,position. This cvicch occurs chen rcactoc'ressure ie greater than 850 peig. 3. IR.'f Fl<<.x Sera~ Tri Seccfn Thi IRif Syscea cnnsfscs nf 8 chambers, ~ in each a: cue reactor rnt,c [ (.ui sysc n: logic <<horme!s. Thc !R'I ! s a 5"decade fns ruaient;;hich cn<<ers che 'rang zf pave< 1 ~ve 1 bet>>ten chat cc <<ared by: ie Si~'l a>>J chi>> AZR.'i. T=.~ dee.ides are rovered br the fR.'f by means o: a raric s~fcch and che 5 d codes nre broken down incn 10 ranges, each be!ng one-hali n'. dic:.d c n si;.e. ~c 1R.'I scram sect!n8 of 1 0 divisfnns fs active in each r~np~ oi che 1."~4. Fur Amendment No.B< 35~ "6 21 BASFS 3', Ill Flux Serve Trir Section" (Contin<i<ed) example, if the instrument were on range 1, t'e. scram settir<g would be at 220~ divisions for chat range; likewise if the inset~ment was on 'tan<-,e 5, Che scram setting would be 120 divisions on that range, Thus,, as the I'lP. 's ra<age'd up accommodate the increase in power jl.evel, the scram setting is ulao ranged up, A, scram at 120 divisions on che IBt instrv<<<encs remains in effect as long as. Che reactor is in the startup <<<<ode. In addition, 'the APR1N 15Z scram, prevents higher power operation without being in the RUN mode, The IPJ'4 scree. provides pro ec on roceccion for changes which occur both locally and over the entire core, The. most significant sources of -'eactivity change during Che power increase are due to control rod'withdrawal'. For 'insequence confro) rod wighdrawa1, the rate of change of power is slow enough due to the physical limitation of withdrawing control rods, that heat: flux is in equilibrium wil:h *the neutron flux and an IRM scram would result, in a reactor shutdown, well before any safety ~ limit is exceeded. Por the case of a single control 1rod, with<!rawel error, a range of rod withdrawal accid< nts was analyzed. This'analyris inc'uded starting~ the accident at various power levels. The most severe ca~e involves an initial condition in which the reactor is gust subcricical and C1i<e IM! ystea is not yet on scale. This condition ex'sts at quarter rod, density. Quarter rod density is illustrated in paragraph 7.,5.5 of che FSAR. Addicdonal conseamatism was taken in this analys'is by assucing that the iR'.i chan~>el closest to the withdrawn.rod is bypa sad. The results of this analysis show that the reactor is scramrted and'peak power limited to on< percent of raced power> thus maintaining MCPR above 1'07 ~ Based on the above analysis, the I~ provides protec'tic<n against local'ontrol r'od withdrawal errors and. continuous wichdrcwal of control rods in sequence, B. APRH Control Rod Block Reactor power level may be'aried by moving control rods <or by varyinj the recirculation flow. race ~ The APRM system provides,a contro', rod block co prevent rod withdrawal beyond a given point ac conscanc recir-cuclacion flow. rate, and, thus to protect,against Che condition of a MCPR. less -chan 1.07 ~ lhis rod bl<ack trip secting; which is aucoioitical,ly varried with recirculation lo'op flow rate, prevents 'an increase in che reactor power level co excess values clue co control red wich-

drawal, The flow variable trip setting provides substantial unr,"<<n Amendment No. g~,. 46

2.) a~szs fron fuel dmage, assurcing i steeidy-state orlerat ion, at rhe trip sctcinrc. over the encire recirculation f lou range. The nargin to the Sa(ary Light increases as the floM decreases for che spl R !fied rrip seccing versus flou relationship; therefore, che uorsc case HCPR Mriioh could occur during steady-state operation is at )ORX of rated cheraa ) pouc r bec ause of the APRH rod b lock c r ip sect ing. Thc ~ccval poi'er distr)but ion ! n chc core is established by speci fird control rod sequences and ii elonlcorcd conc inuously by the in-core LPRH system. AR'ich the ApRH scracl trip ser ring, the d'PY.'I.rod block trip..et t ing is ad justcd 'dovnuarl) if che CMFLPD exceed ppp thus preserving 'the RJ~:r co'r block 'afety roargin. e Reaernr rlarer ln la: el Reran and l nl alen ~(lnee r I eln -lr e lineS) point for the )ou level scr~ ! s above the bo'c tou o: thc separator skirt js level has been used'!n cransicnc analyses dealing vi rh coolant invcnto~ decrease. The rc u'lcs reporred in PSAR subsection )<.5 sbcld Chat sera= and !solar )on of all process lines (except rRRain steacR) ac this level'dequately proteccs chc fuel and the pressure barrier, because HCPR is greater than 1.07 cn a)l cases, and syst ecr pressure does noc reach the safety v lve settings. T)Rc scrau sect ing is approxiuace)y

3) inches beloM the norcra) operating range and is thuc adequace co avoid spurious "cr~

The turbine stop valve closure trip anticipates the pressure, neutron flux and heat flux increases that would result from closure of the stop valves. With a trip setting of 10,. of valve closure from full open, the resultant increase in heat flux is such that adequate thermal ma~gins are maintained even during the worst case transient that assumes the tu.bine bypass valves remain closed. (Reference 2) E. Turbine Control Valve Scrag 1. Fasc Closure Scracr This turbine control valve fast c1osure scram anticipates the pressure, neutron flux, and heat flux increase that could result from fast closure of the turbine control valves due to load re)ection coincident with failures of the turbine bypass valves. The Reactor Protecti'on System initiates a scram when fast closure of the control valves is initiated by the fast acting solenoid valves and in less than 30 milliseconds aiter. the start of control valve fast closure. This is achieved by the action of the fast acting solenoid valves in. rapidly reducing hydraulic control oil pressure at the main turbine control valve actuator disc dump valves. This loss of pressure is sensed by pressure switches whose contacts for m the one-out-of-two-twi ce logic input to the reactor protecti on sys tear. This trip settinq, a nominally 50" qreatpr closure tinre and a diffelent valve character istic from that of the turbine stop valve, combine tu produce transients very similar to that for the stop valve. No signifi-cant change in t<CPR occurs. Relevant transient analyses are discussed in References 2 and 3 of the Final Safety Analysis Report. This scranch is bypassed when turbine stcam flow is below 30.". of rated, as measured by turbine first state pressure. Amendment No. 52, , 46 23 g. Scram oa, lose of! coatrol oil pressure The turbine. hydraulic control ay<<tea operates using high pres<<uric oil. There are several poin'ts in, this:, oil <<yetcm v'here a los<<of oil pressure could result in a fast,closure of the tuzbine control valves. This fast closure. of thc tuzbiaa control valves ia apt, protected by th>> generator load r<<j<ctgop ecran, since failure af the oil system vould aot result ia z<he feist, closure,, solenoid valves being actuated. Por a turbine poqtrol valve fact closure the core vouId bc protected 'by the, A3'RH end high reactor prc>>Ieurc <<creme. Hovevcr, to provide thc acme margins ae provided for the Ieaerator load, rejection scram,on, feet closure. of the turbine, control valves ~ acl am hJLa been added, to the, reactor protect)op

system, vhtch seneca failure of control ctil prceeure to the tur-bine control ay>>tern.

Thie ie an anticipatory scram and result>> in reactor ehutdovn before any tfgaiflc:aat iiacpcsec ia pz'eeeurc or aeutroa flux occurs. The transient rc<<ponce Xa very aimilar to that rceultiag from ths-generator load rcjcction. t,. Hain Condcneez Lov Vacuum Scram To protect the main condenser against ovcrprceeurc, a lose of coa; dancer vacuum initiatce automatic closure of the turbine atop valve<< ~ad. turbine bypass valve ~. To anticipate thy traasicat aad automatic scram, resulting; freya the closure': of the, turbine-stop valves; lov con-denser vacuum initiatee a, <<cram. The lov iva<<uua <<cram,eet point ie ~elected, to initi<<c a <<crea befc

e. tha closure of the turbine atop valves ie initiated.

~ ~ C. 4 H. Hain Stcam Line Ie action on Lov Pz'caeurc,and Hain Steeim L~ic Isolation S< rsm Thc lov prceeurc isolation of the mein <<team linc>> et 825 peig vie provided to protect against rapid reactor dcprceeurizetion <<nd tbie resulting rapid cooldovn of the vceecl. Advilntagc ie taken of thc ~cz'am feature that occurs. vhcn thc main stcam line isolation valve>> are closed, to provide for reactor ehutdova eo that high pover opcr<<L-, tion at lov rcactoz prcosur docs not occur, thus providing protection for'he. fuel clidding, intcgr'ity safety limit. Operation of the reac-tor at prcseuz'ce lover than g25 peig rcquirc>> chat the reactor

mode, avitch be in the STARTUP position, vh'czc protection of the fuel cladding, iategz'ity saf ~ ty limit ie Irrovidcd by the IKW and APRH high ncutZon f faux

- acrams.

Thus, the combination of'.miin stcam line lov~ pressure ieo1.et)on and isolation valve closure scram eeeurei.

the availability of neutron, flux scram protection over the entire range of applica'b'lity of the fuel cladding integrity safety limit. In addition, thc isolation valve, closure eczaa anticipatee the prcssure and flux. transients. that occur duriag, norma) or inadvertent isolation val.ve closure ~ Mith the. >>ere~ <<et at 10 percent 'of valve closure, aevtroa flux doe ~ not incrceec. 2L><

2. I shsKi 4

Y,. Reactor }a~ voter level set otnt for inf t fat ion o( HPCI and and core s ra umos. These systems mafntain adequate coolant inventory and provide core cooling Mith the objective of preventing excessive clad temperatures.. The desfgn. of these systems co adequately perform the intended func-tion is based on the specified lou 'level scram set point and fnit'fa-tfon set points. Transient analyses reported in Sectfon 14 of the FSAR demonstrate that these conditions result fn adequate safety aargfns for boch the fuel and the system pressure. References 1. Lfnford, R; 8., "Analytical Methods of Plant Transient Evaluations for the. General Electric Bofling Mater Reactor," NEDO-L0802, Feb., 1973. 2. Genera1. Electric Reload Licensing Amendment for BFNP Unit 2 Reload No. 2, NED0-24169, January 1979 and NED0-24169A. Amendment Ho. M 46 i' TI ~ I ~ I I ~ ~ ~ I ~ I ss ' I ~ ~ 'I i". )".i!I ~ "I ". il ~ ~ I<< I I I ~ ~ ~ > ') . ~ '.lI;Il: i...:,);;..; t 'L ..Il!<<)I ~ s I .I... !- FTA'>> A'PR' ~ ~ I I ~ I 'I I s Bi~ SCRA!t I >>lsll 1 I I I ~ I I I C 5' I I ~ Q ~ ~ ':::U>>.-'.-:i:::.: 'I-:I.:'!':-: 'Ql'. J: ~ I ~ ~ I ~ ).S:~~l'LC'>> Coll,~OL LKi I')7'c ~ i r I*", i P) f'.'I) 'I 6o )) i'0 ).;I')"I l "'! ~ t"t I 'I "'III.l I ~, I tQ ~ >I: ~ > ~ ~ llss' ~ " '<< 's ).I><<". '.'-)!I I' ~ > ~ NA UR'IL CHCbi~~T.u>'i.;.,I s ~'l'I ~ ~ << i):: l s ~.'I 0 I' l":):~ ~ ~ I ~lt ~ ~ I I ~ II ":.'r:.i.' 'EQ: t t)!), r sl ~llt 10 .. <<t.>. ')lt li: 'Iii'. ~ ~ I't I'. '0 s ~lilt}I ll,', I I I I ~ I I > I~,I I

J).L:;

0 ,!it; [" st ' s' << I), l ]* <t. 50..), kO::! 70t -)8I)..*.CW. it100.,!I3 ~ 0't 120 I!' ~ i) ) I ' t l 'kl'. I " ' I . I li~i I :t".I. ! I s 'I I' !CQH". cQQI'v" FD, 'Q,. ITQ> QEEIQ,'g" ~ II I APL~."1Z" BRAS SC,'P~ttl Vs. HZ'Mi)R GOi i LC'a F'XG. 2 s 1-2 I , ~ t~, ~ 1.2 BASES pressure moni or.higher, in the vesse'1. There o".e, ."oLiovin~ any transient that is severe enough to cause concern that this. safety limit ~alas viol~to"', a cal'culation vill be performed using aL'1 avaiiuble inor.-ation o d.'..cr-m~~e if the safety limit -~as violated. REP:.cali C" S l. Plant Safety Analysis '(B:-.P'."-M Section LL.Q) 2, .~S:Z Boiler and Pressure Vessel Code Sec:i"n Zi: 3. U~ 3 Piping Cede, Secticn B31.1 i!esc or tzsel end. hppurterances t! cnenicel m pJ'n ('ri.P ."3A.'". Su'ect icn 4,2) 5 ~ General Electric Supplemental Reload'icensing Submittal for ,:Bzowns: Perry Nuclear Paver Station Unit 2 Reload No. 2, NEDQ-24169;. January 1979 and NED0-24169A., Amendment No. 35~ 46 Ac

2. 2 BASES REACTOR COOLANT SYSTEM lNTEGRXTY The pressure re.'Lief system for each unit at the,Browns Ferry Nuclear Plant has been sizied to meet two design bases.

First, the total safety/

reliex valve capacit'y has been established to meet the overpressure pro-tection critexia of the ARK Code. Second,, the distri'buxion of this zequized capacity 'between safety valves and relief valves has been set to meet design basis 4.4.4-1 of subsection 4.4 which states that

the, nuclear system, relief valves shall prevent opening of the safety valves.

during normal plant isolat.'Lons and load rejections. The details. of, the analysis which shows compliance with the ASNE Code requirements is presented in subsection 4.4, of the FSAR and the Reactor Vessel Overpressure Protection Summary Technical Report submitted in, response to question 4.1 dated December .1, 1971. To meet the saxety design basis, thirteen safety-relief valves have been installed on unit 2, with a total capacity-of 84.2Z of nuclear boiler: rated steam flow. The analysis of the worst overpressure transient, (3-second closux'e of all main steam line isolation valves) neglecting

the, direct scram (valve position.scram) results in a maximum vessel p~zessure of 1299 psig if a neutron flux scram is assumed considering one relief valve, is inoperaible.

This reSults: in an 76 psig margin of the code allowable over-pressure limit of 1375 psig. To meet, the operational-design basis, the total safety-relief capacity of 84.2Z'f, nuclear boiler rated has bee'n divided i'nto 70% relief ,(ll valves) and 14.2Z safety (2 valves).~ The analysis of the plant, iso- 'lation transient (turbine tzip with b'ypa'ss"valve failure to open),',aasuzilng, a turbine trip scram is prese'nted in, Ref~erence 5 on,page 29. This atnalysILs 'hows thatlO of llrelief valves limitlpressure at the safety valves ,to 1226 psig, well,'elow the setting'f the, safety valves. There~fore,~ the safety valves ~.1 not open. This analysis shows that peak system pressure is limited to '.L250 psig which is 125 psig below the allowed vessel overpressure of 1375 psig. Amendment No. g$, 46 30 ~ ~ NOTES FOH TABLE 3.2'.8 Mhenever

any, CSCS System is required by seccfon 3.5 to be operable there shall b>> tvo operable trip systems except as noted.

If a requirement of che first column is reduced by one, the indicated action shall be taken. If'he same function is inoperable in more than one trip system or tha first column reduced by 3sore than one, ~ction B shall be taken. Action: A. Repair in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. If the function is noc operable in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, take secfon B. B. Declare the system or component inoperable. C. Imnedfscely take action B until povar is verified on the trip systems 2. 3. D. No action required, indicators are considered redundant. In only'one tr,ip system. Not considered in a trip system. Requires one channel from each physical. location (there are 4 loca-tions) in the steam line apace. 5. Pith diesel paver, each RHRS punrp ia scheduled to start fmnedfately and each CSS pump is sequenced to start about 7 aec later. 6., Mich normal pover, one CSS and one RHRS pump is scheduled to start fnscsntaneously ~ one CSS, and one RHRS puep is sequenced to start after about 7 sec vith ~ imilsr pumps starting after about 14 sec and 21 sec, at vhfch time the full complement of CSS and RHRS pumps vould be operacfng. 7 ~ The RCIC and HPC I dceam line high flov trip level sect fngs are g,iven terms of differencfal p~essure. The RCICS set ting of 450" of H 0 corresponds to 300I of raced steam tiov at 1140 pais a d 210K 16 s an at 5 psfs. The HPCIS setting of 90 psf correspands to 225Z of rated flov ac 1140 psfa and 160Z at 165 psia. 8. Note I does noc apply to this item. 9. The head tank is designed to assure chac the discharge piping from the CS and RHR pumps are full. The pressure shall be maintained at or above the values listed in 3.5.1, vhich ensures vater in the discharge piping and up to the head tank. ItOTES FOR TABLE 3.Z.B (Continued) 10. Only one trlLp syotel~ for each cooler fAn. 11. In only two of tha four: 4160 Y. ahutdoMn..boar'dg.'eo note. 13. 12 ~ In only onc of the four 4160 V shutdown booar'ds.'ee'ota 13, 13. Ari eaergency, 4160 V ihutdiwn board.* ig; con<sidaiad a trip oyer<a. 3LBRSM p<alp wou1Ld -be ihoparable. Refer 'to'uction 4..5.C for tc..a roquiromantsi o1! a R33RSW pump boinF, iaoporable. 15. The accident sigrial is-the satisfactory ccapletion cf a e:i-c<c)t-of-tm taken tvice logic of the dryvall high praoaure plus low rogatory, pros>> gare or the vosocl Icw water lavol (s 3'lg" a1av'a v'eu )cl'oro) o<rigiaatixig, in the core spray oyster trip syetaa. 16. The ADS circuitry is capable of accomplishing its pzotectivo action Mith one operable trip oyster. Therefor'a ono. trip cyst"'a c sy bs za3<cn out of service. for fuactiona'L, testing and calibration for,a, p"riod not to exceed

8. hours 17.

Two RPT systems exist, either of which ~'~il'1. t'ri3) both recirculation pumps. The systems will be individually 'functionally tested monthly. if the test period for one, RPT system exceeds 2 consecutive

hours, the.

system will be declared inopera1ble. If both RPT systems are:<noperable or i'f 1 RPT,system i. inoperable for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, an orderly power reduction shall be initiated and reactor power sha.'Ll be less than 85/'ithin 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. 4>>nor'nt No ~, 46 4.2.B (Continued) t40 Function RHR Area Cooler faa Logic Core Spray brae Cooler Pan Logic Functional Teat Teated during functioaal teat of inetruacat

channels, RHR motor start and thermostat (RHR area cooler fan).

Ho other test required. Tested during logic ayaten fuactional test of instrcuaeat

channels, core spray motor start and therao-atat (core spray area cooler fan).

No other teat required. Calibration H/h N/A Xnatnawmt Check H/A inatruaeat Channel >> Core Spray Kotore h or D Start Teated during functional N/A teat of core spray puap (refer to section 4.$.A). H/A lnatruacnt Chaanel-Core Spray Hotora 5 or C Start Tested during functional H/k teat of core <<pray punp (rofer to section 4.$.A). H/A Eaatruaent Chaaael-Coie Spray Loop l Accident Sigaal Teated during logic oysteca functional teat of core spray syateac. Hlh H/A luatruacnt CI~el.- Core Spray Lcop 2 Accident Signal 'fact)RSM initiate Logic uPZ initiate logic } c>l'f breaker Yestod durius logic aye tea functional teat of core spray

ayatcna.

once/6 uonths onro/month ~inca/operni. t ng rye l e Mlh H/h H/A H/A H/h H/A TABLE 4.2.C ,SURVEILLANCE RE()VIRtAclfIS FOR INSTRUHiTATIO!1'TllATINITLLTE ROD SLOCKS'unction APRN. Upscale (FloIi Blas) APRN Upscale (St'artup Node) APRH. Doiiascale APRN Inopeiative llltV ff ~ des ~ a s \\ ho 1 UPScd Jc '1r 10v'1ad/ 'M,Dounscale Functional Test (13) (1) '(13) (1) (13) (13) I ~ 0\\ 11 J/ (13) Calibration '(ll) once/3'.nonthe once/3 eoaths once/3 a)amphi '.f/A oned/6I lalntlld once/6 aonths Instrument C!Ieck once/day (8) ooce/day (8) oace/day (8) opec/dal (8) /A~I. /RI Vill 'CI ue/ l>l once/day (8) SSV t aun aOOPCtdC1VC IRH'Vp*cale I~ \\ L<J I~ %1 J'4>>/ (13) uld ~II dl once/3 IdIaths ~ca~a/Aald /Al vll% %,I 'UQI once/day (8) IR l DNAdcale (1)(2) (13) once/3 IRntha 'once/day (8), IRH Detector not in Startup. 'Podittan IRH Inoperative (2) (once/opeia-- ting cycle) (1)(2) (13)'once/operating cycle (12) H/A . SRH'pscale SRH 'ovaac'a1.e.

0) (2)

-(13) (1)(2) (13) once/3,.deaths oace/3 aaatha once/day (8) oace/day (8) SRH Del.eetnr'nat 1rI otartup Position SRH Inoperative Plov Bias Cospsrator P?ou Bias.-Upscale Rod Black:Logic. R"CS 'Ad/a trhl'nl (2) ' (ouce/0 p e ra= ting "cycle) (l)(i) (13) (1)(1$ ) (1) (IS) (16) (1) Qllcc/ OP'cc dl ills c/L@%I. (hal I '. ~ 1 ltd'\\ V/A once/operating 'cycle (20) once/3,antha H/A once/3 months M/A H/A n/A'li/A ti/R. ~2, SJLSF.S The XPCf high i:n<<and temperature inst runentat lun art provided co detect a breai, in the hPCI steam P!Pinit Trfrrfng Of Chil fnatr~lentatiOn te-sui t s fn ac t va t ion of XPC 1 fsv! ~ '. ion valves Tripp l>>g logic for che high flov is a 1 ov( o! ? log,fc,, and a l 1 sensors arc rcqui rcd to be opec able. llfgh t<<npecatvr~ ln th<< ~1 lofty n! the XF l equlpnent ls sensed by 4 Seta O! a bfne(a'i:Ic te.p<<rature SvftrheS. Yh<< lG t<<.~perature SufteheS are arranged fr. ? trl, systems vfth 8 te~peratur. svf('chc'n each trip system. The XPCi trft . <tl>>.: n.'! Psf fs l;rt tin ..d ?":C l fur peratut c ~re svt h th; t ccrc i rr nvci y 1 s prr< mt rd a< ' 1>sfon product release fa Mfth:n.': its. The RC!C hfl 1 1 in~ and I ~~, +f Jt VI c fn"trv~eni et $ n.i srr arr ngcd the san t that for th xpcl. Thc trip sett fng n' 50" h.,n fo: hfch flov and 200'F fot tenpcratvt e ri baal d on tl c race crf tc: 'fa as l e HPCl. llff;h t<<npcrat>>rr w.'hr F(actor Clcanun Systro . !nor dra ln could indicate ~ break fn thc c lca. up syste-. t'!.cn hfgh temper tvrc occurs, the cleanup system is isolated. The fns(runentatfnn vhfth fnf tlat<>rcd fn thc Specf!ltattc n p.caerves thr c.'!cct.'vcness o( th>>..ysten perfoda vh<<n .-a !n< c >>a~ca or test fnP, fs befog perfarned An thfs fs vhcn !~,".'.c ! unct iona'. testing fa be in', per!ornr fn a dual this fashion, even during exception to The control

cr'! iti tun n>>s are pm"fded tn pyric nt excessive control rod vf'tldraua) o th:t

.'lCl'l; d~es nn: dcctews~ t( 1.0 Thc trip logic for thfs (unctlo." fs 1 out of n. c.s , nny trip or. nne o! s.(x Apg.'f's, eff,ht IL".'s, or !ou; SR.'1's vf1 1 result fn a rc l b!oci. The mtr f-.v~ t -.; '~;.n'. charm<>Rfc fa'fvrc triter.'a.'; oct. Tuo RBM channels are pro-vided an'nly one of these may be bypassed from the console, for mainte-ra..ce an~/cr testing, prcvided that this condition does not last longer than 21'ours in ary thirty day perioc. This time period is only 3$ of the operating tire in a.".onth and d"e not significantly increase the risk of preventing a i inadvertent contre'od vithdraual. The APR't . o" t ivt', !<<oct inn 1s f lou b!ascd ana p< rv"n'is a sf gnf fleant reduc-tion in XCPR ~ ~ s~,c'. ~ 1!y c r.nf; opcrat fon at reduce 'lcv. The APR.'f pro-vide s Crrss corr >rotc:t f<<>>,

1. e,,

! 1 "1 ta the grnss rc re pouer increase fror. vf th 'rajya i o.'rr.: rol rnds In thc normal vithcrava1 sequence. The I trfps are set a: t~at X:pp 'is oafn.afned f rcater than l 07 ~ The RBH tod hint~ (v: '..'i n pro<<fr'cs ts ~..a'. prot res f~n nf thr core; 1 e., the prevent 1 cn n.' r 1 t ical p, er ln a local rcc ion o.'he core, for sfngle rod vfth 'ra~~i. error 1 ron a 1 fnftfrS, ron't tol md pat tet n. Amendment No..40, 46 )13 5ASFI If the IRK channels are in the i~rat, condition of'lloMtd bypasai

the,

~ e J) Lng arran(Ment i s qunh t hac 'nr unbias seel i 4". i bann \\s ~ a, rod block! signal is generaci d hi inr!i~ th!'i ce<<t<<4. neiitrons,flux bn Lncreaaei) <<ore than a factor oi )0. A dovnsca)e* indicaticn Ls, an indication the~ Lnstrusienc haa. failed or the inscrumenc is nnc aensicivt enough. In eit)~er; cyst, the iristrusiest vill noc respond to chances in control gpd uiotioll and, thus, cpntro), tod setion is prevented. The refue)ing inter)ocks alan op>race onc logic channc', and art required for sa fee v on) v vhrn 'ht cocle svL c ch is. Ln tht. re.ueling pciaition. For c f f ace ivc erne rgencv core cnolint fcr ama)) pipe brtaks, the eCI bysi'tet<< aust function since r.actor pressure does not decreast rapid enough to a))oi'ither core spray cir ll PC) co operate in time. Tbt automatic 'pres iure relief func c los is pr ovi 'lcd as a backup co,thr. HPCI in the event tbe HPCI does nc c opi ri(r. The arrangement of chc tripping, contacts Ls such aa to prnvidn this fu>>i' inn vhe n ineccsaarv and minimixe spurious operation.

Vbe, trip sectiii 6 riven iii the spec ificatioii are -adequate to assure tbe aboyc, criccria err met.

1hc specification preserves tbt effectiveness of the, ayoccm di>r inc, per in)s nf nalncenance, tescilng, pr p ~ ) ibratLon, and also minimixes che risk nf inadvertent operation; i.e...,only, one instrument channc) oiic of service. No pose treat~mt nf:-.".os radiation -.:nicn. s are prcvidcd and, vhen their trip pninc is re cbed, cause an Lso)scion o'. the of.'-gas lint. Isolation is initiate" uhcn boch instr~..cncs reach their, high, trip point or one has an upsca)e trip and cht other a doiwsca'e trip, or, both have a downscale 'trip ~ Boch '.>>scrumcncs are required fo: trip buc the Lnscriimtnta ar set so tha - any instruments are sec so chac che Lnacan:aneous stack re)east rttt limLc jlvtn in Specification 3.6 is noc excetded., Four r id l>>c,'nn v i>>1 i us i iirc pi iv!~ 1 ~ ~ l . inr t>aclii upi j <I'lieth Wni c iact Primary Concaiiimnnc isolation (Crojp 6 iao!ac ion va)ytsi) RcaI:cor Building lsolytiipn and operation n( chc standby Gas 'free.; ent sysccm. These instrument c)}antoe)s <<onicor tiie radiac ion in the Reactor xone iveli>tii)a(ion exhaust ducts and in the Refue) in'.one. Trip ai'ctiny, of 100 mr/hr for the m>n!corp Lp, cht;Re.fut)in', 2onz are bpse~d upcn initiating norma! vt nc i!ac io>>, iso!ation and SG.S opcratiori ao:hit none of the activity re)eased during cht refutling, accident leaves tbci Reactor Bui)ding via the nor..al venti)ation -path but rather all ths activity ia processed by chc SGTS. flov inceg. a cora and sump f L 11 dcccrciine )cakalce in the droit fi)1 a 'kno~ vo)ume v'll be uc ay%cern ib a1ao provide tc dict (Se e Tab)'e

3. 2.,'E).

rate-and pump out rate tiotr>> sr used:o A'ystem vherebj the tim interval :o to proviide a backup. (in air sarplirg ecc )tskagt inside tbt primary concaiamenc 114 ,r,> ~ ~ i ~>r t!.r nprrarnr ul:h n vis >ai r>>d.r tlo.'l ~ r>~ l. T!>c cn>>sn>i>>ri> r>> u ' c ~: ! v.' y ~(r iden tn ere n f>>net f>i>>n o(;,>c if', > il >>ctin > (l i... ":h" rcqc> rc>r, nc of ar lrns'. 3 cnu>ir s i ~<r nr(<>nd znaurr) thit.>;.<<rr ns fune, aio<<ld l < occur, r>"sf>>a nr. vr zl>ovc thc l>>:r! >I vafu>. nf 1O o( rnid l l nw:r ~ d l i L'. nnsly >i" nf '.rir>.'.cnr s I r>>>> cord r.'>><<'.(! l>>nn. 0>.; o>> rnbi>> .'.>>H er>>nnc l vn d be a>lcquiec >>>>nl r nr the approach <<r r f r '.:a I l ty us.'n:. ncrao>>cncous aeecrn.> n( scattered cori rnl md %':.Br-"r>l..A mini;.cs '>c'we ol>ersbl>: 5: f'a are provfded as an added corrrctvutiJ 5. Tha Red> Rfoc'~;(onitor (RB,".r fs dcs'.-nrd ro auto=at'.cally p avc lw (uc d'"njr': f > 'nc cv>>'f ni.-u. rud vf;h-..i >oc". fc"-a o: high g vcr rfei f ty Cur( 'q 'ifgh ro"c ~ cvcl oprrar.ion. Tvo RBM channels are provided, and one of these ray be bvpassed from the console for maintenance and/or testing. hutoraatic rod withdrawal blocks from one of the channels will blcck erroneous rod withdrawal soon enough to prevent fuel damage. The specified restrictions with one channel out of service conservatively assure that fuel damage vill not occur duc to rod vithdrawal errors when this condition ex'sts. A li'r f tfng.cnntrol red pattern fs a pattern whiclr rP, u]l in the core being on a thermal hydraulic I fr>ft, HCPR r f vcn by SpeCifiCatian 3.5.k Or LHGR Of 18.5 fOr 7X7 Or 13.4 f)ur inp. use of such pit tcrns, f e i s ]ud, >>', that tr stfnr, ref tlic RB(f system prior to -f el>- drawn!l nf such rods to assure fts opcraof lfty w: 1] assrrr> that improper wf.thdrawal does not occur. It f s norma 1ly the respons ib'. 1 ity of the tluc 1 ea r Fngfneer to fdentf fy these limfting patterns and r'r designated rods either when thc patterns are initial ly cstablf shed nr as they develop duc occurrence of inoperable cenerol rods in otl nr th lfmfefng patterns. Other Persnnnel qualified eo form these functions may-be designated by the plant super f nt endent to per form these func t iona. Scram Insrr t'on T'imes The cone ro1 rod sys tera is designated to bring the reactor subcrf tfca': at the race fast enough tn prevent fuel damage; f e, to prevent the 'ICPR f rom becoming less than 1.07, Tire ) imiting pn>rer transient is given in Reference

1. Analysis of this tran fcnt shnws that tne ncgat fvc reactiviey rites resulting frnr. the scram with the average respnnse of all the drives as given fn the above specification provide th

~ r equf red pre e eat f nn, and HCPR remains grei ter than 1.07. On an rii'.';;l:P, snr>> dcg rada t f nn o f con t ro 1 rod sc rar. pci (nrr,><< " ~ ":<<red r!ur f ng nl ant st art up and was de terr f ned r r he r.<< Amendment No. 40, 46 part I.ulate mater fal ()~rohably construccfop debris) p,'urging;; an fnccrnal control rod drfve filter. The-desi'gn,of'he present cont ro'od-drive (Hade) 7RDB144B) fe pro<el'y fmproved by the. rcln ac inn nf rhe filter to ~; )o-. ~ t5on,'oui of the'c'ra6. drfve path: l.e., ft Can>>n longel" interfere wfkn scltaoi parforleence, even l rhmlilccel v blocked. The dry radrd pet I ormence of the; origfnal drive (CRD7RDB144A) unde~ dirty operating conditions and the fns'ensftivi'ty 'af the. redr ~Irnec drive IcR.".7RDB]~4B) has been dema'nscrared by e ~ er I

a. o.'nginerrfnR eeet s under sinu)aced

'reactor operating conC:c ione The euccesef ul perfoMance o! the nev drive under actu ~1 aperacfng condf c fons hae. also beer., 'demonstrated,by cnn>> scent.y Roar fn-s< rvi=e test results For plants uefng the new !rfvr end rw) be fi Ferrcd I,'res plant>>'s'fng', the alder model Jrfv wfch a modif ied ~il ~ rRer sc.een site, internal filter which fs 1 se prune co plugging. Data hae'een'c'cueenced'y surveil-lanc: repcr:e in various operacf lg planta'. these include Oyster Creek, I<oncfcello, Drest!en 2 and Dlesden 3. Approximately SOOO drive teece,have been recorded to 'dace.'ailnwIng identification of the 'plugged ffl'cer" problem., vel;y freq~cnt ceram testa were >>ecessary to 'ensure proper perfomancc.

Hovever, the ~ire f'requeac scram tests art now-conafderec tarelly unnrceeserv and cnwfee for the fo.[loving tee'sons:

1. Crrac Ic scree perforaa>>ce hes been identified as due Lo en obstructed Crivc filter fn type "A'" drives.'he drives In BF! are of l he new "B" type design hose screw. performance ie-uneft'ected"-by ffleer condition. Thr d:rc load ie primarily released d ~rfnp ~ tal tup of the rene:or Mhen chc rciccor end its evscrne are first sub)ected to flows end preen ire end cherrml stree ~ es. ipecial attcn-c fon and neei urce i r>> now being taken to assure cleaner system~ ~ Rei ccore vfch drives ident fca.'r eimiIlar (shorter ~tlokc, sma)) cr pf eton areas) have ~op&raked'hrough many refuel fng cyc les wcth no sudden or lerPatfc exchanges in scram ierformnce, This preoperacfonal and ' tart'up 'tea cihg ie 'ufffc cent to detect anoms'ious, fIrfve perfo. mancc ~ 3, he 72-hour nucegc lfaft whfch fnftfeked'h'e e'CaI't bf 'the frequenc ccrc m tenting fs arbl.crarv. Iievfn-no lngical basis other chan quancifving e- "me]ar Out~age" whfch 'might'easons bly be ceueeC bv an event so severe oe ta poeefblv'ffecc drive perfarr.lnce. Thfe requfremen'c fe 6nw'fee because fc provicec an Incentive for shortcut act,iona ta haetej returnf l~ "on,line" to avoid the ~ ddf tfonal test fnlI d~ue ~ i'2-hour outage. 132

3. 3/4. 3 aisles:

The eurvef 1 lance requirement for scca'esting of all the cnntrol rods after each refueling outage and lOX of the control rods ac 16-veek intervals fe adequate for deternfnfng, the opera-bfifcy of the control rod systcn yet ie not eo frequent as to cause exces ~ ive veer on the control rod ayecea components. The numerical values assigned to the predicted acran perfor-nance are based on the analyefs of data from other Mt'e Wch conc rol'od drives the sane ae those on Brovns Fer< Nuclear F lnnt. The occurrence of scram tines vithin the linits, but sfgniff-cancly lobster chan che average, should be vievcd ee an indica-tion of eyscemat fc problem vfth concro) rod drives especially ff the number of drives exhibiting such scren cfnee exceeds

eight, the slloveble number of inoperable rods.

1n che analytical treatment of the transients, 390 ufllfeeconds ~ re allotted betveen a neutron sensor reaching the ecren pofnt end the start of negative reactfvity insertion. This ie ade-quate and conservative vhen compared to the cypically observed tine delay of about 270 nfllfseconds. hpproxinately 70 milli-eecondn after neutron flux rcachee the trfp poinc, the pilot ecrsn valve solenoid pover supply voltage goes to zero en epproxinarely 200 tafllfeeconde later, control rod notion begins. Thc 200 ofllfaeconde are fnc)uded in the allocable ecren inser-tion tine~ specified in Specification 3.3.C.

• In order to perform scram time testing as required by specification 4.3.C.l, the relaxation of certain restraints in the rod sequence control system is required.

Individual rod bypass switches may bc. used as described in specification 4.3.C.1. The position of any rod bypassed must. be known to be in accordance with rod withdrawal sequence. Bypassing of rod's in the manner described in specification 4.3.C.I. vill allow the subsequent withdrawal of any rod scrammed in the 100 percent to 50 percent rod density groups; however, it wi'l maintain group notch control ovez'll rods in the 50 percent density to preset 'power level range. In

addition,

.RSCS vill prevent movement of rods in the 50 percent density to preset power level range untiL the scrammed rod has been withdrawn. 133 Amendment No. 32

3. 3/4

~ 4 BAS ES D. Reactivitv Anomalies Dut ing each fuel cycle excess operative reactivity varies, as fuel d apl,etes and as any burnable poison in suppleimentary ccintrol is burned. The magni tule 'f this excess reactivity may be inferred from the critical rod conf igurat Lon. As fuel burnup pro-

gresses, anoma ious beha~tic r in'he excess reacti~>i t'y may be detected by comparison of the critical rod pattern ar selected b,ase states to the predicted

'od inventorv at that state. Rower operating base conditions provide the most sensi,tive and directly interpretable data relative to core reactivity. Furthermore, using power operating base conditions perini ts frequent reactiv'it'y eompa'riso<cs. Requiring a react ivity c'om'parison at the specif ied frequency assures that a comparison wi)l be made before the core reactiviry change exceeds lX R Q Deviations in core reactivity greater than lM+are not expected and require thorough evaluation. One percent reactivity into the core would not lead o transients exceeding design. conditions of rhe reactor system. References Ge'neral Electric Supplemental'eload Licensing Submittal for Browns Perry Nuclear Power Station Unit 2 Reload No., 2, NED0-24169, January 1979 and NED0-24169A, Amendment No. , 46 134 90 \\ SOLUTION TEMPERATURE MUST BE EQUAL TO OR GREATER TRAV. THA INDICATED BY THE CU}tVE 8o 0 vo tI) P~(i f60 4 ( ~ 50 10 15 SODi 'M P>> NTABORATK SO!,'l"t ION (aa, w/o N~ 8 0 H 0) 2 10 I6' 8RO'HH5 FERRY NUCLEAR PLAHT FIHAL 5A'F ETY AHALY5I5 R jPORT SODIUM PENTABORATE SOLUTION TEMPERATURE REQUIREMENTS FIGURE 3.4 139 OASES: 'STANDBY t.f~l> LD CONTROL SYSTEM A. 1( nu mora than onc operable control rod ks vfthdcsvn, tha i basic, shutdovn rasctkvicy requirement for the cora ks satisfied and,tha Standby. L!quid Control. System ks not required.

Thus, Che basic reactivity raqu.lremant for the core is the primary determinant of vhen the liquid control sya tern is required.

The purpose of the lkqukcl control system ks to prqvfpe <he capability of bringing tha reactor from full paver to a cold,'enon-ftee',shutdavn'ondi-tion essumfng that nona of the vkthdravn control rods can be inserCad. To meet this ob]actkv"., the liquid control system ks designed to inject e quantity of boron that pi.oducas a concentration greater than 600 ppm of boron kn the reactor core Ln lass than 12S minutes. The 600 ppm con-centratlon 1n tha reactor core 1s required to bring the reactor from full paver to a aubcritfcal condition, considering the not to cold reactivity difference, xenon poisoning,

atc,

"(he time requirement for insartkng tha boron solution ves selected to override rhe rata of reactivity,'Lnsartfon caused by cookdovn of the, rcacto: fol- ~ovfng tha xenon poison peak. T a minimum Lkmktat fon on Cha relief valve setting is intended to prevent the lose of liquid control solution via the lifting of a relief valve at .too lov a pressure. The upper limit on tha relief valve sectings provides. system protection from ovarprassure. B. Only one of cha tvo standby liquid control pumping loops Ls needed for operating tha system One inoperabla puripkng circuit docs not frrnad-1ataly threaten shutdovn capability, and reactor operatfon, can continue vhfla tha circuit Ls being repaired. Assurance that,'tha remaining, system vill perform ,'l,ts. intended fun"Cion, and that the Long-terms average svafknbfkfty of tha system is not reduced ks obtafnarl fTo S one-out-of- ~vstam bi an sllovsbla equfpmant out-of-sara Laa rfma of ona-thfrd of the normal surveillance frequency. Th1s method determines an equip-mrnt out-of-iarykca t(mu of tan days. Additional conservatism Le introduced by reducing tha ekkov~bka out-of-service tfma ro *evan days, and by fncreased tastfna of the,operable redundant component. C. l.avcl indication and alarm indicate vhethar rhe solution volume has

changed, vh feb might indic ~ te a possible solution'concentration,change.

The cast interval has been established in const larat.'Lon of these factors. Temperature and liquid level skarms for the system are annunciated kn the control -room. The solution Ls kept st least 10'F above the saturation'emperature. to guard against boron prackpLtatfon. The margin ksifncludedi in Figure 3.F 2 The volume concentration requirement of tha solution'art abch'hat should evaporation occur 'from sriy point vithfn the cur'va~ a lov level alarm vill annuna1 ate before the temperature-'concentration requirement ~ sre exceeded. 140 Amendment No. 46 ITIHC COHtiITIC)HS VAR OPFRATfdIv .H bhintenan.e of'illed Df"char e.. Pip>> ~e suction of the RCIC an" Iid t I pumps gAIall be al imed to tice condo..". iate storage

tank, and the prcssur e.'upor>>s-sion chatiber head tank shel'. normally be aligned to'>>rve the d)srhlirg. pipin,",

f Ihc RJPi end CS pum 's. Ti>> condensate head tank may be used to se<<r>> -h>> PHJI and CS discharge piping if the PSC iieed tank is unavailabl>>. '!ne nressu c indicators on the dischc "g. r". 'he RJLR and CS pimps "hell indfc"'" Pe~', legs than listed belnv. Pl-75-20 48 psig Pl-75-46 48 psig Pl-74-51 48 psig P>- 4-"-q fill I sig

l. ~avera e planar L' fnraR.ar

~ ere.ar len Race During steady scace Rover np>>ration, che tfaximum Average Planar Heat Generation e (MAPLHCR) for each type o: fuel as unction of average planar exposure r snail noc exceed tl e lie'.ting value shovn in Tables 3.5.1-'1,-2,-3>-4pazzd -5, Tf'c any c !me during operation ic is A&ermined by nnriiial surveil lance chat the limitfng valve for APf ffCR i being

exceeded, action shall te inic ia ed vith-in 15 minutes to restore eperacion co

'dichio the prescribed limits. !f che APLHGR is noc returned to v!chfn ch>> prescribed lfcifcs vtchin tvo (2) hours, th>>a reactor shfli he brought tc the Cold <<hu tdovn c onif 1 t ion vithin 36 hc v rs. Surveillance and ccrresponding ac ion shill continue until reactor operation is vfthin che Prescr!bed ':-fcs. Linear Heat izeneri t ion Rate (LHCR) During st eady s tat e paver cpera c ion, the 1'near heac generation race (LHGR) of any rod in ar~ fveJ isserably at any axiaL location shall not exceed the maximum allovahle LHGR as calculated by che follcving eqvat fon: RefeRVF:LLLd'uMCI: Pn, u I.': Lil".dTS 4.5.H Msinzcnunce of Pil'ed l:ischir Pcs l. Every sionch prior to che c.sting of the RIfnS (LPCI and Cont ainmen Spray) and core spray

systems, the discharge,!Pfping of thd.>> systems shall be vera ted freu the high point and vater flov determined.

2. Folioving any period vhere che LPCI or core spray systems have oot been r!",iir, J npe rdble, Cha'C JS- "barge p.'p! ig of che inoperil la sys-t~ shall be vented from tlie high point prior to the return of the systeci co service. 3. Whenever the HPCI or RCIC system is lined vp co take suction free th condensate storage

tank, ch! Jis-charge piping of the HPCI and RCIC sha'1'e vented froci che high poise of the system and vacer flcv observed on a monthly basis.

4'. I'hen the PJ!RS and the CSS az'e r-- quired to be operable, ch. prtssii "e indicators idhich oonitcr the disc-hargee lines shall be monitored daily and che pressure recorded X. Hax&un Aver-."e P'ansz'icear Heat Gener~- tion Rate (:!APLHGR) The e'fPPLHCR I or <<ach type oN f;e! tion of ~v>>rage pfirvr <<".Posvc>> s>al.i. bv. detamined daily dur'ng reactor ope.ation at a 25Z rated thermal pover ~ J, Linear Heat Generation Rice (LHGR) as a fvc c ion be checked daily during reactor'".eeoc.'on ac 25'ated thermal pave., paeendeeent Nn. P5 45 1.IYITlhC'ONJITIOHS FOR OV<'RATION , bLfRVP,ILLAhLR RK~UIRFRRHTS. LHCR LHCRd t ] 5 P/P) (L/LT)j LHCRd Design LHCR LS.5 kw/it. for 7x7fue] r]3,4 kw/fe for 8xgfue] t} P/P) " Maximum po'~a sc~fkfng pe.nalcy 0 026 ear Ix/ iue'0.022 far Sx8 and 8+SR fuel LT << Total core length<< 12.0 feet for 7tc7 and Sx:8 fuel <<12.5 fe.et foi: 8rrSR fuel L << Avfal position above bottom of core. If ae aay efme durfng opera-,lan ie is deter-mined by rormal surve' Lance thae che limiting 'alue .or LHGR is being

exceeded, action shall

'e init'ated vithin 15 minutes co rescore operation co vithfa che prescribed 1!Lairs. If the LHCR is noc recurned ea vichfn che prescribed limits vichin rva (2) hours, ehe reac tor sha' be b roughc ca che Cold Shutdavn condftion vithin 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Surveillance and 'orresponding action shall cant!Lnue uncf1 reactor operation is v'chfn the prescribed limits. K. Minimum Critical Power Ratio (MCPR) K. Mir..imam Critical Foyer Rat fr (MCPR) The HCPR operating limit, for BPNP 2 cycle 3 is 1.33 for 7X7, 1.30 for.SZS, and 1. 28 for SZSR fuels. These limits apply to steady state po-. wer operation at rated power arid 'flow, 'For core flows other than rated, the NCPR shal3., be greater than the above limits times Kf. Kf is the value shown:Ln Figure. 3,.5.,2. MC<'R shall be determfned daily during reactor p~r o]peratfon ac 25X eared thermal paver and fol- }ovfng nny change-in paver level ot distr fbuefoon that would cause <opera: tfou vfth a lfrafting concrol rod pattern as'described in the 'bases fz Spec ifica'cion

3. 3.

If'ar. any c]me durfng operation fc ia determined by nor al surveillance chat the lfmfting value iar .1CPR is befng

exceeded, ac cion sha11 be infe fac ed vichin 15 minutes co restore operacfan to vfchin the prescribed

}fmfts. IE the steady scace MCPR fs noc returned ca vithin che prescribed ]i!sfc s vich jn cvo (2) bones,,che reactor shall 'be brought the Co)d Shucdavn condic'an vfch'n 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. Survei] }arce and inrrespard!ng action she l L cnntinu< unt 11, t r.<; car oi'er )e Lon 1s ~ x chin the prescribed l iaics. L'enarc n<< Sec'u 1 c'"Jeuc s any ni che .'.= ting vaLweS iden:: <ki~d'pec '.f1ea fans

3. 5. I, J, or K, are

<'vr ceded and che spicif1ed rewed]a', accian 's taken. chq evenc shaiL be '}agged.ar" tepncced 'n a 3C'-day vrftcen report. Amendment No. 46

3. 5 msrs

).5,C Auteeastit'e ressurization S stem (AnS) This specific ~ Cion ensures thc operability of the. ADS under all condi-tions for vhlch the depressuritation of the nuclear systaa ia an essen-tial response to station abnormalities. The nuclear system pressure relief ayatca provides autocratic nuclear system depressuriration for small breaks in the nuclear system ao that the loM-pressure coolant in)ection (LPCI) and the coro spray subsystem can oprratc to protect the fuel barrier. Note that this specification applies only to the automatic feature of the pressure relief ayetca. Specif )cation 3.6.n specifies the requirements for the pressure ral'ief 4sction of the valves. lt ia possible for any nutbar of the valves aeeipncd to thc AQS to be incapable of performing their ADS functions hecaurc o( instruacntation failures yet be fully cioabla of perforaina their prcssure relief function. because tie automatic dcpreeaurization system does not provide aakeup to the reactor priiary veeael, no credit, ia taken for the. steam cooling of the core caused by the,system actuation to provide further conservative to the C"CS. With two ADS valves known to be incapable of automatic operation, four va1ves remain operable to perform their ADS function, The ECCS lass-os-coolant accident analyses for smal.l line breaks assumed that four of the six ADS valves were operable. Reactor operation with three ADS valves inoperable is allowed to continue for seven days provided that the HPC1 system is demonstrated to be operable. Operation with more than three of the six ADS valves inoperable is not acceptable. Amendment No. 35 167' l.$.II 'Iolntcnancc oI I'llad IIlscl~ia.ir e l;L~e I i f Ch COCC spray, LPCI. IIPCIS, and,IICLCS, ~ rr nipt fille4~:vac.r hs~aier can develop in Ch'is 'pfpinI". vhen the pusrp-snd/or 4 To ainisiitc daniatp tip the dischsrxe piping and ca ensure aJJod siaricln Ln the opec'ation of c)Iesic sist'esis', this Technical spy~el I'lcpci'on 'equirna Che'iSCharge lin>>S CO b>>, fi)led vhenevC r Che'yeteai iS fn qn, aper>>iblc cond'.cIon. lf a discharge, pIpes~

noC, I i 1 Lcd, the puiaps l Chjit jsupo y,

that 1 I ne <<us c bc assuocd to be lnopei iiblc for'kch*ical Speci ficytiqn pur-poses'~ Th>> core spray and RHIC syste<< iilscItarrc pipins hlxh painC vent is,visually, checked for vaccr flov once a ~~ch nrior.o C stini; to en>>ur>> that

che, lines <<rc f I 1 lcd.

The visits I'hecIt.inq vlll avn(4 start tnt Che core porgy,or, RIIR system vlch a dischsrR>> linc not fill>>4. In addition ta e Qs~ be,ti <<nd ta ensure a tilled,diucharge line, other than'r a~. tp testing, a.pressure suppression chamber heaiIj t~ is located appraxisia ely, the dischairge. line highpaint to supp~i makeup vatjer for these syspei<<s., TlIe condensate head tank located apprasIisssttep 100 f'eet <<beve, the disqhaIrge, high point sers~s as s backup chugging system, vhen 'the'i'eseure suppression chamber head tank is not in service. Systems pispharge preaiuris indi'cators are IIsed Ca d rsai e the vater level abave thy, pscIsarpe line high pipint. The Indicator,'s viUref1ect appraxiIaately 30 psig far a inter level <<t'the gh pc peig. far a, vater level in the pressuresuppressian, clubber head tax'nd, are sIc: itared daily ta ensure that the. discharge 1'nes are filled. Mhen ln their nonenl standby cond ition, the suet ion for the IIPCL and-, RC'fC, pewps arc z nn i n N a 1 I nncii cn thi'onilrnsa tc, sco'rene t wn>. ~ MIitch is. phy'>> I ca 1 ly ac hI III'CI, higher clrvat inn ciinn thc IIPCLS -and APLCj pin lnr,. This assiires itM and ICCIC CIischnrxc pipln>> rcr ains filled>> Purtiicr.sssujrsnce ls nrnyMca$ by observing vnice r flov fra<<cheae sysc~s IjlixIi points iiinnthly~ Kixiaua Iveregs Planer Linear, Hesc Ceperytion 'Rate'HAPLHGIC) This specification,assures thee the peak cladding tcsiperature folloving cha, postulated dciign basis loss-of-coolant eccidjsnt, vill,not exceed this liait spe:ified in chc 10CFIC5I2, Appendix X, The peek cladding 'cesipersturs fallowing ie postulated loss>>of-coolnnc ~cci-Jeent is pri<<arilv a Ilunccion of che ajvorjagp hoot, ge'ncc'ation rotc of'll the y rode of s. fscl asscsibly at any axial llociaclon,and Is only dcpcndcnc seco'rd-il, Ch rod Co rod power dlscribution vichin an asse~bly. Since cx-, sib 'ectcd loca.'ar iatianS, ln power discpibucIon,vithin a, fueL'ssi<<> Ii affccC Che calculated peek c Lad ce<<pcracurc py ~legs (lian 20 F relative co.chc peak,cc<<peracurc for a typical fuck design, chc Liislc on che average linear hojet generation race is sufficient to, azure I'hajc col'culaccd ccnIiccacurcs ~re.vithin the LOCFR50 AoocndLx K Liaii.c ~ TIie limit:Lng value Car NAPLH('R is shown in Tables 3.5.X-1,-2,-3,-4, &-5. The,ane,'yses supparting these limiting 'rsl;es L.I presented n -'f&0-24088 and NED0-24169,'mendment No. Btt 46 II5$ 3 ~ 5 J, ~ I I ne<<r Hest Ccnez'at'.on Rata I.IICR Tlzfs specification assures thse tha, Linear hest gener<<efoo rsec Ln any zod is less chan thc dcsfgn Linc<<z bene Ccncrsefon ff fuel pallet dcn<<fffc<<tfan is postulated. Thc pover spike penalty specified fs ha<<cd an the <<n<<L-ysfs pres! nead fn Section 3.2. 1 cf Reference 1 as modified fn References

2. and 3, ond a<<su<<!es a

Lln~ nzly incres in'arf stfo!! Ln e" Isl gaps bc-tvecn. core bottom and

zap, aud assures vith o 952 eonffdcncc, that no morc thon one fuel rod cs!.rcds the de,.itn linc<<>> hcoe CencrstIon r<<te duc eo povcr spLLLng.

Thc LIICH oa a function of core I!eight sh-11 1!c clicct'ud daily iur-ing rcsctaz'pcz:!tioo at 25Z pov!r eo dcecmfoc if fuel burnup, oz con-tzof rod mavemene has c<<u cd chonFcs fn paver d fstrfbuef on., For LIIGR to bc a lfafefng value bclov 25K rated therm<<L paver, thc ~ ITVF vould I!<<vc to greater eh<<n 10 !:hfeh L pz'cclu!Icd by a considerable r:erg fn vhen cmployfng sny~cr!ofssfbic contra'od pattern. Mfa~a Crit'c<<1 Paver Ratio MCPR Ae core eh<<ra<<L paver levels less than or equal t 25 h o e e reactor vfll be ,operating at afnfaua recirculation p~p speed and the aoderator .vofd content vlf bs ve~ !n<<all. Far sll designated canez'ol rod patterns vh'ch uay be em-ployed: ae.this

paine, operating plane experience

<<nd ehera<<L hydr<<ulfc <<n<<L-ysls &dfcaeed eh<<e the resulting MCPR value ~s Ln exc f n excess o requirements by a considerable a<<rgfa. ~fth this lov. void content n ! <<ny M<<vc tent co're flcv facrease voufd only place operatioa in a aoz a aoz'e. conserv<<efve aade rela-tive to MCPR. The daily zequf eaenz for calculati 'IC+" b 25 ng ..!, a ave 52 rated ther <<L hsv paver Ls sutficLeat since paver dLstrLbuefan shifts are ve y s Lav vhen there sve aoe been sfgnff Leone paver oz control z'ad cha. h aged ~ s e rtqufreaene f0'r cslcul<<ting !.CPR vhen s LLMefn control rad e lfCPR vill be ka g ra pettera fs approached easuz es that v e ova follavfag a change Ln paver or paver shape {reg<<rdless of aagaitude) thac could place oper<<tLaa at a therzssl 1Laft. Re ortfa Re ufreaeaes The LCO's ossacfseed vfeh aonftarfng the fuel od operat!ag condLtions are required to be n! t <<t <<11 efaes, L.e., there fs no allavable tlae ir vtich rhe pl<<ae e<<a kaavingly exceed ehe Lfafefng values for,H'CR,

ULCR, MCPR.

lt fo s requizeaeat, as stated in Specif fcsefoas 3.5.1 tha ' f ce <<ay t "e duriag steady state pe'er opez<<tica, L: La determined that the L~tfag values or &oiSCR ~ LHCR, ar 'MCPR are exceed<<I act'an Ls then inft'<<ted to restore operation eo viehLn the prescribed if=its. This action fs iaiti<<ted as oooo <<s noraal surveillance Ladfcaees ehs-an cperaeing - it 1as been reached. Fach event involving steady <<tace c <<tace ope.stion beyond a specified Lfxtfe sho'1 be logged <<ad reported quarterly. Lt aust oe reccgni-ed e? e thez e fs alvays an <<ctfon vhich vauld. return any of t,.e p ra=et ~ rs (NPDCR, LifCR, or MCPR) eo viehia pz'escribed

ladies, aaaeiy paver reductioa.

Izader aost circumstances, eh's vill'oe be the aaly <<Lterasefvc, Re er aa es I~~Fuel !Ieasf! Lest'oa Fffects oc Ceneral Electric 8aff'.aq ua cr Pu!1," Supplements 6, 7, <<nd 8, h~>>10735, P.gus '.9.3. Suppf caeat. 1 eo.echnfcsl Repaze cn "ens'f fest'cas af Ceaersl glee rfc Reactor

Fuels, Dece be.

L4, 197< (USA Rzgu'<<cozy Sta."f) 3. C~aicaefca: V. A, Nacre to l. S. Mitche'1, '~calf ied CR.".wel for Puel Oensfficaefaa," Docket 50-321, Inarch 27, 197>>. General Blectric BUR Reload 2 LicensingAmendment for BRF Unit 2, '8ED0-24169,January 1979 and NED0-24169k.. Amendment Ho. 3~~ 46 169

4. 5 Core an<<l Cent ainmcnt Ceel in~~Sstems Sut ye illa.,c Fre~uencies Tl>e testing,, interval for'he cere

<<nd containment c ~Ling <<ystems. is. based'n industry practic:c, quantitative reliability analysis5udgement and. practicality. The. core cooling sysrems have not been designed to be fully'e<<table during operation. For example, ih t'e 'case" 'of 'he. RPCl,, <<ueoaati'c initiation during pever operatien vould result in pumping, c'old vater 'into 'he reactor vessel. vhich is net desirable. Complete ADS testing during poser operation causes an undesirable lesswf-coolant inventory. To in<<re'ase tnc availability of the core,and containment cooling system, the cemponents. vhich make up thc system; i.e., instrumentation, pumps, valves, etc.,'We tc ~ted frequently. The pumps and moter operated in/ection valves are also tested each month to assure their operability'. 'A !simulateid automatic'dtu4- 'ion test once.each cycle combined vith mehthly tests of the; pumps-and in)<<c-tion valves is deemed to bc. adcquarc testing of these systems. Mhen cempene>>ts and.s>>bsyst.ems are eut-'ef-cervvicc, overall core and c'enr'aih- '<<nt cool In', r<< I iabil ity iei maintained b'y dcmbnstrating the operability of the remeinin"equipmcnt. Tbc degree ef 'epkrab11'ity to be demenstr!ate'd depends on the nature ef chic rcasori for the out-'ef 'ervice 'quuipii'ent. I or routine'ut-ef-service periods caused by preventative maintenance, etc., the 'pu<<iIp hnd'alve operability checks vill be performed to demonstrate operability of the remaining components. Hovcvcr, if a failur'c,'esign deficiency, caus'e t'e

outage, then the demonstration of opcrabilf'.ty should bi e thorough enough to assure that, a: generic preblem does not e1xist.i For 'example if an out~of-

<<ervice perio<<l vas ca>>scd by failure of a pump to deliver rated capacity'ue to a design deficiency,,the ether pu'mps of this type might be sub'jec'ted to a flov rate test in addition to the operab:ility checks. Mhencvcr a: CSCS system or loop is <<c,ade'nop test or calibration, the other CSCS syst'emsl operable shall be con'iidercd operable if thi lance testinr, frequent,y and there is ne rca If the tuner.ion,.aystcm, or loop undci; t'est'r exceeds the trip level sc.tting, the I.CO tcstinR for thie systesc or loop sh~ll apply.'rable because of a requirhd 'I; lbeps rhn't <<re'e'quired t'e be cy are ~iithin the required'u'rveIil* sen te suspec't they, a'e ineperablc. ol. calibration is found inoperable and the required surveil.lance Redundant operable components. are sub)ected'6 ihcz'ea<<ed'e'sting during equip-ment out-ef-i<<ervice times. This adds, further conservatism and-increases assurance that adequate cooling is available should the ireed arise. The MAPLHGR, LHGR, and MCPR shall be cliecked daily te determine if fuel burnup, or control rod movement has caused changes in poMer distribution. Siince changes due to burnup are sloM, and only a fev control rods are moved d<<ily, a daily check of peMer distz,'ibutien ia. adequate. 170 TABLE 3.5.I-5 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE AVERAGE PLANAR EXPOSURE MWd/t 200 MAPLHGR kW/ft 11.2 Euel Type: 8DRB284 PCT ( F) 1685 1,000 5,000 10,000 15,000 20,000 25,000'0,000 ll.3

11. 8

12. 0 12.0

11. 8 ll.2

10. 8 1667 1671

'.647 1669 1672 1633 1596 Amendment No, 46 172a t.tttr S.ttt.". CnVntTIONS FOII OPI RATtntt 3.6.C Coolant Lcakasfd. SURUF II.Lh.tCF. 'Rt OUIR~ENT D. 3., If che coadicion in 1 or 2 above cannot be

mec, an orderly shutdown shal'e ia'iaced and thc reactor shall be shut-doMn in the Cold Condition vithin 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Safer and Relic! Valves l.. When more than one relief valve or one or more safety valves are known to be failed, an orderly shutdown aha13 be initiated and the reactor depressurized to less than 105 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. D. ~Safe aed Sel'e. Valves l. Ac least one sa!cty valve aad approximately 'one-ha'! o! all relief valves shall be bench-checked or replaced vich a beach-checked valve each opera-tiats cycle. All 13 valves (2 safety and 11 relief) vill have been checked or replaced u-=.. the coma'ction of every second cycle.. 2. Once during each operacinE

cycle, each relief valve shal:

.be manually opened unt. chc..o-couples dovascream of che va:ve indicate sccam is flovin~ from the valve. 3. The integrity of che relief! safety valve bellovs shall be continuously monitored. 4. At least one ".elicf valve sba'1 bv disassc. bled aad inspected each operating cyc'. E Jet Pecos E Jae Pas.es 1. Mhenevcr the reactor is in the scartup or rua modes, all*get pumps shall be operable. If it is determined that a get pump is inoperable, or if tvo or more )et pump flov instru-menc failures occur and can-noc be corrected vich'n 12

hours, an order'.y.shutdown shall bc anicd.aced anc che reactor shall be shu dova 'n the Cold Cond'ion vichia 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

l. Whenever there is rccircu'acicn flov vit'h the reacco. 'n che starcup or rua mod'es vith boch recirculation Pu Ps <<ueeee nss s )et,pu. p opc. ab ld v sha ~ ' checked da'ly by verifying ch". the folloviag cond'ions co ncc occur sim 'aaeous'y: a. The tvo recircu'ioa .'o""s have a fiov imba'ance c: 152 or =ere vhcn che pu=-s are 0 c a cd at <<"c sd e speed. Atttendment No. 35'6 I.Iv 1 T I Itr. i".nkf>I I I l)'ls Fnh'PF RAT to!I suIIVI.'LLA'IcI RPI)u ',I ItÃ~rIIT >.e.t: Jr t I'uw~i 4.6.E Jct Pumps 3 ~ The reactor shall not be operated with one re"ircu'.icn loop out. of scapi,ce or more than 2I'our". With the reactor operatitte if one recircitlation loop is out, o ser vice, the plant shall. be place-" in-a hot shutdown conditio;. with zi 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless t'ne loop is sooner retu n c to service. 3.6.F Jet P.."2.oi Yismatch l. W!ien bo h recircul t'on pu:zps. are in steady stat. operation, the speed of the feist,:-r puzzp shal3. be maintained w,ithin l2c~>~ the -sp 'ed of the slower piunp when core potter .is Gv';~ or nore of.rated poez or 135'$ the speed o the slo~ er p. ri when, core powe" 's below i)0~I,I of rated power. 2. lx specification

3. 5.'."..1 cannot be m t, one recizculation pump shrM be tripped.

'b.'he in'dirated value of core t'lov rate varies, fro. thi ~ialue derived from 1nop t'lov-ci.-r surcmcnts by nor<<. than 101 c. The diffuser to lcn'cr pli au> differential pres'sure rc".d-ing one an individual ]iet pusp v'aties f;c.'ei th" incan of all ~et p~ d' Erron ~ tial pressirros by aor>> than 10X. 2. Whenever there is t ceil'culat. on flow with th>> reactor in the Startup or Idun Hodr and rne -c- 'i'reals'cion pump is operas'.n I 'i'th'he Cquc15,z'cr vrlv.. C)r ~ed, the diffuaer te lover plenu~ diffIsrcntl,al pzcssurc shnll )~ chccfced diilly and the dl ffe;:n-tinl pressure of an indivi<'.u~1 Jct puep in a lunp shall noc 'ary from the acean of all ~e 'iIap'-@ifEcrentiai'rI.saures in tlat lo'op ',by'a'rc than 10'.. FoU.owitilz one ptimp oper;"ion, the discha pe valie of the loir speed. pimp may not be opened uq.'ess the speed of the faster ptuip is less than 50',>'f its z ated speed. 5. Steady stat "operation with both, recirculation pumps out of'er-vice for up to 12 hrs is per-mitted. During such -int fval r'estart of the recirculation ~ ~ umps is oermitted,. provided the oop discharge temperature is within 75oF of the saturation temperature of.the reactor vessel water as determineIi by dome pressure; The total elapsed time in natural circula-tion and o'ie punIp oferation must be no oreatei thah 24 'Irs. (:. S true tura 1,'1ht. arity l., 'I'he stziictuzul inteI;rity of the prinary system shall be

1. 'eci'rculatio'n punp epeeds shall be checked'nd-logged ac least once pet day, G.

'Struc'cu".'a 1 lfitecritv 1. Table 4.6.A together vich sup-plvsientery':not.es-, spec ties L'hc Amendment No. 32 detected .cosonably <<n a mat e cf few hou s u ili2inP t..e available leakage detection schemes and, if he crig'n cannot be detez mined in a reascnaoly snort time the uni should be snut down to allow further investigaticn and corrective action. The total leakage rate consists of all leakage, identified and unidenti-fie", which flows to the drywell flocr drain a.-.d eouipment drain sumps, The capaci y of thc d"ywell floor sua:p ~u. p is 5'3 gpm and, the c paci " of the dr.;~~ell e<,ui".-:ent su=p pump is also 50;p=. Henoval of 25 gpn frc" ei h r of these sumps can oe acco.=pl'si:e-'ith cons<<idcrable m'-in. BI r.M.!i~S l. Nucl ar System Leakag Bate Limits (Br.<<P PSAH Suosecticn 4.10) 3.6.D/4.6,0 Sa fet and ?el'ef Valv s The safety,and relief valves are reo" ired to oe c arable aoove the ores-sure (l95 psig) at wh'ch the cor s ray sys e=.s is not desi<<-,red to del'vcr full flow, Tn pressure relic. system for eaci. un't at tho Lrowns Perry Nuclear Plant. has be~n sized to meet two desi.-n bases. ?irst the ote'afety/relief valve capacity has been es a "lish d " neet he overpress'e protcc.icn criter'a of the ASLZ Ccd

Second, ti:e distribution of his reauired capacity 'oetween.safety velv s and

." 1'ef valves has .been se: o meet design bas's 4.4.4-1 of subsection 4,4 wi'<<ch.states ".hat the ruclesr system relief valves shall prevent opening of the sa ety.valves during normal plant isola ions and load reJect<<ons. The details of the analysis wh'ch shows ccmpliance, as modified by Be.erence L, with the ASAP. Code reoui"en nts is pr s nted in subsection 4.4 of the .=SAR en" the Reactor V ssel Overpressure P."otecticn Su-mary Technical Report su"n'tted in Amend=ent 22 in respcrse to question 4.1 dated, Decem'oer 6, 1971 To meet the saEety design basis, thirteen safety-relief valves have been installed on unit 2 with a total capacity of 84.2Z of nuclear boiler rated steam flow. The analysis of the worst overpressure. transient, (3-second closure of all main steam line isolation valves) neglecting the dizect scram (valve position scram) results in a maximum vessel pressure oE 1299 psig if a neutron flux scram is assumed considering one relief valve is inoperable. This reSults in an 76 psig margin of tne code al'lowable over-pressure limit oE 1375 psig. To meet the operational design, basis, the total safety-relief capacity of 84.2/ of nuclear boiler rated has been civided 'nto 70/ reliei (11 valves) and 1'4.2/ safety (2 valves). The analysis of the plant iso-lation transient (turbine trip with bypass valve failure to open) assuming Amendment No. 35~ 46 pic 3'. 6/4. 6 'BASKS: a turbine trip scram is presented in Reference 5 on page 29. This.-analysi,s shows that10 of llrelief valves'imit pressure.at the safety, valves to 1226'sig, well',below the. setting of the safety iralves. ThereEoze, the safety valves will not open. This analysis, shows chat peak system-pressure is limited,'to 1250 psig which is'/5'sgg below the allowed vessel overpressure of 137.i psig. Experience in relief and safety valve operation shows that a, testing of 50 perce'nt oE the val'ves-per year is adequate to detect 'failures or deteriorations. The relief and safety valves are.,be!>chtested every .second::operating cycle co ensure that their SecI paints, are within the + 1. percent, tolerance.. The. relief valves, are tested 5.n place, once per operating cycle to establish thar. they trilll C>peh and. pass'.,steam. The 'rcquiremencs escablished'bove apply when the nuclear system can 'be pressurized above ambienc conditions. These requiremencs are applicable at, nuclear system'ressures below normal operating pressures because abnormal operat:La'nal transients, could possibly start ac these candicions such chat eventual'verpressure relic E would be needed.

However, these; transients are much less

severe, in terms of'ressure, than those starting at rated. condit:Lans.,

The valves need nbt be fuhctional when the vessel head is.'ema'ved, -since the nuclear system;cannot.be pressuiized. REFERENCES'. Nucl'ear"System Pressure -Relief System (BFNP FSAR Subsection 4.4). 2. A'mendment 22'. in respo'ns'e .Co 'AEC 'Question

4. 2',of'ecember 6,

197IL. 3; "Protection','Against. Overpriessure".,'(ASIDE Boiler, and Pr'esaure Vessel Code; Sect:Lan ITT, Article 9) 4." Brawns: Ferry Nucleaz: Pl'ant'esign Def'iciency Report-,<<Target Rock Safecy-RelLef Valve's:,.transmitted by-J.:E. Gillelan'd.Co F;. E; Kruesi; August 29, 1973. 5. General Electric, BWR.Reload 2 LicenSing.A>!>endm'ent Eoz 'BFNP Unit 2, NED0-24169,January 1979.and'.'VL LD0-'24169A. 3'. 6. E/4. 6; E Je t Puree s Failure. oE-a jec pump nozzle asscmblg h41dIia~ gechenishi, 'noz" le assembly and/or,ri,ser, would inc..'ease. the cross-sectional Elm~ area for 'blawdown fo11awing,the design:.baiis dbuble-ended 'line break. Also, failure of'e di ffus'er would. eli'minace-che capabilicy co.r6 i&ad'he core co;Cwa.-thirds. heighc level fo.'Llbwing a, re'circulation:line-b'zeak. Therefore, if a Eai Lur~e occur'red, .repairs:e>use .be mad,e. We decection cecani'qua 's as follows. Wf,th,che cwa recirculac:".on,put>ps balanced in speed to within + 5 percent> the, E'aw races in bach z'ec'ula-l tion loot>s, vill b'e verified:.bv. control,room monitor'ng inscruments. ~ =~ che flaw z'ate"v:!lues do, not dif.'!er by. more chan.10 pe, canc, z'iser and, noz='.e ass'embly.iniegricy, has been verified.

220, Amendment No; 35

Daily tests of,innunciation lights and audible devices are per t Grmed as a r out ir e oper:i tion function.

hE CQg, syst..m manufacturer recommends semiannual testing of CO, system fir~ d=tection, circuits.,

Fi'gure

6. 3-1 describes the in-plant fire protection organization including the roving fire watch.

In addition, other operating personnel periodically inspect the plant dur ing their normal operating activities for fire hazards and other abnormal conditions. Smoke detectors w'1 be tested "in-place" using inert freon gas applied by a pyro."onics type applicator which is accepted throughout the inlus rial fire protection indu try for tes ing product= o'ombustion detectors or by use >of the tlSA chem.cal smoke genera ors. At the present time the manufacturers have only approved the use of "punk" for creating smoke. TVA w'll not use "punk" for testing smoke detectors. 329 5.0 HAJ0R ncslnH FRATURcs 5 1 Sing FL'ATHRL'S Broms Ferry unit l? i,s located at Browne Ferry Nuclear Plant site on property o~ncd by the United States and in custody of the TVA, The site shall consist of approximately 840 acres on the north shore of Wheeler Lake at Tennessee River Rile 294 in Limestone

County, Alabama.

The minimum distance from the outside of,the secondary containment b'uilding to the boundary of'he exclusion arear's'efined in 10 CFR 100.3 shall bc 4,000 feet. 5.2 RCACT!OR A. The core shall consist of 364 fuel assembl.ies of 49 fuel rods:

each, 168 fuel assemblies of 63 fuel rods each, and 23.'2 fuel.

assemblies of 62 fuel rods each. B. The reactor core shall contain 185 cruciform-shaped control: rods. The control material shall be boron carbide povder'84C) compacted to approximately 70 percent of theoretical density.

5. 3 REJECT()R VESSCL The reactor vessel shall be as described 3n Table 4.2-2 ot the FSAR.

'T'e app!licable design codes shall be As described in Table 4.2-1 of the FSA'R. 5.4 COHTAlHHCHT A. The ]principal design parameters for th!c primary containment shall be. as given ice Tab)e 5.2-1 of th!e'SAR. The applicabl'e design codes shall be as described in Section 5.7 of the FSAR. i B. The secondary~ co'ntainment shall 'be as described in Sectioh 5.,3 of the FSAR 'C. Pc!netrations to the primary containment. and piping passing through such penetrat iona shajL1 be designed in accordance .vith the et'andards set forth in Section 5',2,.3.4 of the,FSAR. 5.5 FuCL STORAr.C A. Th!e arranngement of fu'el in the, neM-Fuel storage facility-sh!al1. be such th!at k F, for dry conditions,, ia less than 0.90 and flooded i!i f!..ss than 0.95 (Section 10.2 of FSAR). Amendment No. 35, 46 >~ c)