ML18025B230

Amend 58 to License DPR-52,changing Tech Specs to Permit Operation in Cycle 4

ML18025B230
Person / Time
Site:	Browns Ferry
Issue date:	11/12/1980
From:	Ippolito T Office of Nuclear Reactor Regulation
To:
Shared Package
ML18025B231	List: ML18025B230
References
NUDOCS 8012080295
Download: ML18025B230 (40)

Text

t;~

gD,$ E R~G0 Cg~o

~o~

W~*~+

UNITED STATES NUCLEAR REGULATORY COMMISSiON WASHINGTON, D. C. 20555 TEI'INESSEE '/ALLEY AUTHORITY DOCKET NO. 50-260 BROMNS FERRY NUCLEAR PLANT UNIT NO.

2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 58 License No.

DPR-52 1.

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

A.

The application for amendment by Tennessee Yalley Authority (the licensee) dated July 14, 1980, as supplemented by letters dated August 29 and October 7, 1980, 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; B.

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 assurance (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 COIImIission's regulations; D.

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

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

2.

Accordingly, the license is amended by changes to the Technical Speci-fications as indicated in the attachment, to this license amendment and paragraph 2.C(2) of Facility License No.

DPR-52 is hereby amended to read as follows:

(2) Technical S ecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.

58, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

80g2II80 295

0 f

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. Ippolito, Chief Operating Reactors Branch jj2 Division of Li'censing Date of Issuance:

November 12, 1980

ATTACHMENT TO LICENSE AMENDMENT NO'. 58 FACILITY OPERATING LICENSE NO.

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

1.

Remove the following pages and replace. with identically numbered pages:

F910 15/]j

~17 18

~25 26 27/28 29/30

~79 80

~131 132 133/1 34

~159 160 167/168

~169 170 172a 219/220

~249 250 269/270'29/330 2.,

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

The overleaf page is provided for convenience.

l5 a

l.. T

)'))::I. 0).r,n))I)I~ TNTIf:IITn P, T FVEI. CI.AT)DTIIC T hTF~I:T'fY Tn the event of operation with the core maximum fraction of limiting po~er density (CMFLPD) greater than fraction of rated thermal power (PYZ) the setting sha11 be modified. as follows:

6 + (0. 660 + 66Z)

PDD CMFLPD For no combination of loop recizcu-lation flow rate and coze the.

<<1 power shall'he A~RM flux scram ""'.,"

setting be allowed to exceed 120K of rated thermal power.

(Note: These settings assume operation within the basic thermal hydraulic design

critezia, These criteria are LHGR+ 18 ~ 5 kw/ft for 7x7 fuel and ~13.4 kw/ft for 8x8, 8x8R.

and Pgx8R>

and iICPR within li=its of Specification 3.5.k.

Ef it is determined that either of thcso design criteria is being viol tcd during opezation, aetio.)

sha11 bc initiated within 15 minutes to rc tore operation within pzescribnd limit:.

Surveillance requizemears for /)PF.:.:

scram setpoint are given in spec ification 4.1.B.

Core ".be-..~<<l Pove." Li)~'t

(!>q<<c t.." Pressure

<800 T)sia) ii) c.-. 'I)e reactor pressure is 1css th,)) or equal 'to 800 T)sia,

'l Amendmeni No. gg, 3$, g$,

58 2.

APR:IVhen thc rca-tor mode switch is in the STARTUP POSTTION, the APRM scr-m shall be set at less than or equal to 15Z of rated

power, 3,

IR'IThe ZR"I scram shall be set at less than r equal to 120/125 of full scale, B.

APRM Rod Blo"'r. Trio Settini The APRI'. Ro" block trip setting sh 1) be:

'lArr.Pf LINTY c

i,qgTq IRD SAFI~<

CYSTS !'ITTTNC.

X.l FlJZL CLADDIH0 INTEGRITY or core coolant flou is l=ss

~ than l(4 of rated, the core thermal pover shall,not ex-ceed 823 Ydt (about 25'i of rated theta'overj.

2.l HEEL CL!V)DIII0 IIITBBRITY

'RR<

(O.ki6V + 42~i')

vherc:

'> = Rod block setting is percen'of.'ated t.herrrral paver (325~3 I1Mt)

Lo~p recirculation f1<a,&nt,e in percent of rated (rat;crl loop recircula)ion flc,u rate erlu ls 34.2 X 10 lb/hr)

Ir. the event of operas'.ion r."itir I.ne cor" raxirrur-, fraction o'. linitin~~ po'er density'C.'ZLPD) greater than fraction of."a,ed therrral p"ce.

(Fr:P) "the set:in",shal'be f'rodificd's follows

~

S

<(0.66" + 42K)..

C.

Vh n'ever he reactor is, in the ahutdovn condition vi h irrariiatei tuel in the reac-tor vessel, the vater.

1 vel

.".hall. not be less than 17.7 in. above the top of he norrrel active fuel xone.

C., Sera."s

";-." isr.'iuaticn

> 538-in. ab. ve

, re'ac!';or lov i'ater vessel

-ero levr S 'rar-.,-- urbinc stop

< }0 pe. cent v ilve closure

'valve closure e E.

Serac-turbine control valve 1.

Fa. t closure Upon trip of

.the-fast actr r,-

soIcnoid valve!.

2.

Loss o: control

> 550

'ps:q oil p".es,su:;e F.

Scram--lov.con-23 inches denser vacurw H!g vacuum Gi Scr:mnain stez.-,'

10 per cent line isolation

-valve closure H.

Vain s'earn isolation

> 825 psig i alve closure nuclear systc~ lou IH'essurc l0

!jp5RS:

~ URL CLM01t C Ihi FCR 1 i j Sr S~if LLMl~

The fuel c'dding represents one of the physical barr ters vhfch separate rad'o-at( fve rjac <<rials

~ re-env frona

~

The fat<<gr i(y o E th f= cladd ~~og barrier is related co ita relative fre<<dec freej perforations er cree'fa".

Ai(hough ojjjc corros fen or use>>relat ed cracL lag may occur during the 1 ffe of che cladding, ffs sic product, aigrat fee froc (his source is increaen cally cuaxlativc cad continuously eeasurable.

Fuel c'addiag per'era(iens, he~cv~r.

can result froc str<<sses vhfch occur frca reactor operatfea sig.".Lf 'f can ly above design fons and Chc protec"oa systca se(oeiacs.

4hfle fiss'on product a'gration froc (1add fng. per oz mtfon fs g us t as aeasurab Le as thatf roe u <<-r<<L>(ed c. ack~~g

~

ther-ally-caused cladding perforations e fgnal a threshold, be>wad vhich st&I

~rcater" the=-cl s(resses rLay cause groes rather Chan facrcacatal clMdiag dc(czfora-

ioa.

Therefore, the fuel c'add'ng saf cty Lf~t is de fin<<d fa revs oE Che, reactor epe a fng cendi( oas vhich can rcsu' Xa cladding pcrforatfon.

Th<< fu<<L cladding ia(egri(y lf f( 's se(

such Chat no calculated

!ucl daaage veuld occur as a resuLr. ef an abner=aL operational trans i<<at.

8<<cause fuel dzzagc fJ ne(

d i(ectly observable, Che fueL c'dding Saf ety Lwft is defiaed vith aargw to chc cond f fons vh'h vould produce ense" transition bofliag (HC7R of 1.0).

estab Lfshes a Safety Li-'

such that the afawuc crfticaL pave. ratio (NCTR) no 'ess (haa 1,07.

HC-R > 1.07 represents a consc~acivc cargfa relative to condft'oas requited to asfataia fuel claddi"g iategrity.

t ef trans'ion bo'i1f ag results ia a decrease in heattraas fez f-oej the clad therefore, elevated clad'caperature aad the possfblity of clad failure.

Since boiLfag cranef cfen fs noc

a. d!rcctly ebscrabl e paraec(cr, cha jsargfn c.ansf tfon fs'alculaccd frea pLant operating paracetere such aj core eev<<r

~ core fL~, fe<<dva(er Ceejpcratu re, and core pev<<r dfs(rfbu(ioa.

The e argin fuel ass<<ably

'is charactcri=cd by che crit'cal pover r-(fo (CPR) v hach at '

of the bundle pcwer vhich veu Ld produce onset of (ransit" on boilfng I

by Chc actual bundle pcver.

The a'aftra va ue of Chis rat o for any buad e

ch co. c '

thc efaiaua cricical'aver ratfo (M').

It is assumed chat the ra(fen is controlled co che ncafaal protcctfve st(pain(s via chc fn d var fables, i.e.,

normal plant operation presented en Figurc

~.l by t..c wjna] j'i"e( ci" fir M cnn(rnl 1fac.

nc Sw <acv Lf>>fC OQp, of 1.07}ban c>>f j'<(feat (o assure cha( fn the even( of an abnormal operational (raasienc (ons<<rva('5 co

!.:jit ac e..

roe

' f c e.. froe a nona I opera c fag cond i en (qCpjL y i>if s spec+ tied ia seeeificat'ea

3. 5. g) core

.hen 99.9i e1'he fuel

!a (he cere are

<<xp<<c(<<d Co avoid bof '

ng transition.

The atr g'w betvcca pl o f I. 0 (ense(

of rr=ns ftion bo'!ng)

=nd cne safety 'mf(

1.07 is derived f!on a d <<caf led sta('

(i=a 1 aaalys is cons fd<<rfag all ef, the uac sr afzt( fes fn rjonf-( fag srac<<

. Lnc)ud ing uncertain(y in the boil ing C

ransom

( ion ce((<<lg(fan-as d<<scrfbed in R<<far<<ace L.

Th<<unc<<(ca fn(i<<s <<>>ploy<<d fn derf vfng Che -zfncy Lfaft are provided at thc b<<gf..ning of each fu<<L cycle.

15

.1.1 BASKS Jzccause thc boiling transition coz'relation is based on a lc"gc quantity o<

full scale data there is a very high confidence that operation of a fuel assembly at the condition of MCPR ~1.07 vould not produce boil.Lng tran-sition.

Thus, slthough it is not required to establish the safety~ 1.i~mit additional margin exists bctveen. the'sa!Cct'y 3$mit and the sctusJt occurencc

-. of loss of'ladding integrity.

Hovcver if boiling transition vere to occur, clad perforation vou3d not be expected.

C3.adding temperatures vould increase to approximately 1100oF vhich is belov the perfora ion tempI..ratuz'e of the cladding material.

This has been verified by tests in the Ocnersl Mectric Test Reactor (GlZR) vhere fuel similar in dcsigzz ~o BFliP oversteps

<<boLre

'he critical heat f3.ux for a significant period of tizze (3G minu~es) vithout'lad perforst:Lon.

If'eactor pressure shou'ld ever exceed 1400 psia during no..'".3, pover'perating (the, limit of applicability of'he boiling transition corre-lation) it vould be assumed that th'e t'uei( clsddizzg intc ity Safety Li"it has bcca violated.

In addition to the boiling transition liteit (HCPR 1.07) cpcration is I

constrained to a maximum LBOR of 18.5 kvlft fdr 7x7 fuel and 13.4 kv/ftfor Sxg,

.t oz Sx8,

SxSR, and PSxSR fuel.

This liad.t $s yeached when the Core Mxi!Z!um Px'actiotz of Limiting Power Dens.ity equal's 1.0 (CHFLPD 1.0)!

for the case vhere Core Haximum Fraction of Limiting Power IDensity exceeds the'4'sc ion e!f Rat'ed'Thermal Povcry operation is, permitted on/y st less than 100~i of rated

2. 1.A. 1.

power and only vith reduced APRH scram settings as required b'ifi:i y spec, cation At pressures bclov 800 psia, the cor'c e';?era%ion pressu"

'drerp (0 pover 0 f3.ov) is greater than 456 psi.

At lov povsrs and flovs this pressure differential is maintained in the bypass region of'he: coze..

Since the pressure-drop in the bypass region is essentially all'levation head,,

the core prcssure drop st lov povers and f1m'iU, alvsys bc greater than 4.56 psi.

Analyses shov that'. vttth' flezv of 28)Q03 lbs/hr 'bundle flov, bundle pressure drop is neaz'ly'independent of 2!undle pester a'nd hag a value of 3.5 psi.

Thus, the bundle f'lov vith a 4.56 psi d iving, head, vi3.1 be greater than 28xl0 3bs/hr.

Full scale ATLAS test data taken at pressures f."om 14.7 psis -to 800 psia indicate that the fuel assumably critical pover at this flov is approximzIte'ly 3.35 M'A.

Vith the design peaking factors. this corz'esponds to a coze thermal povr of rorc tlat 504..

Thus, a core thermal paver 3.:Lmit of 25$ for reactor presmu c."

bclov 800 psia is conscrvzLtive.

For the fuel in the core during periods vhen the react;.or. is. shut dovn ration must also be given to vates',icvcl z'equireme'nts due to th ff of decay heat.

Zf vater level should 'dry'vp 'bel.ov the to of'he fuel d c ability to remove decay heat is reduced..

This reductio'n in cooling capability could lead to elevated cladding tcmpcz'aturcs and clad, perforation.

As lo, long as th fuel rankins covered vi h vatcr, suffi~<ic t ccnling is available to prevent fuel clad perforation.

16 Amendment No. 3g, 3$, l$,

88 2

The safety'i~it has been established at l7.7 in. above the top of the irradiated fuel to provide a point vhich can be uonitored and also pro-vide adequate oar@in.

This point corresponds approx~ately 'to the top of the actual fuel asseublies and also to the lover reactor lov vat r level trip (378" above vessel sero).

aznR..wcZ 1.

Ccneral Electric NR Thernal Analysis Basis (CKTAS) Data, Correlation and Design Application, NEDO 10958 and NFDE 10958.

Amendment No.'3$,

Ajar, 58 l7

0 FAG 3F' 7".

1Q

2. l BAcES J.

c R.

Reactor

) oM iwcc>>'

evel set oint for ini t fat ion o(:HPC1 and RCIC c)osin> ~ nein scca+ isolation valves and scarc'n L'PCl and core s ra ouaoe.

These system maintain adequate coolant inventory and provide, core cooling vich the objective nf prevent L~g excess ive ciao temperatures.

The design of these,syste=s co adequately per fort>> che intended func-tion is based on the specified Lou level scras>> set point and initia-tion set poincs.

Transient analyses reported in Section 14 of the FSAR demonstrate thac these conditions result in adequate safety aargins for boch the fuel and che system pressure.

L.

References

.1.

Llnford, R. g., "Walytical Ywthods of Plant Transient Evaluations for the General Electric Boiling Mater, Reactor,".REDO-10802, Feb.,

1973.

2 ~

Generic Reload Fuel Application, Licensing Topical Report NEDE-20411-P-A, and Addenda.

Amendment i'fo.

P$. Q>> 58 25

I I

f I

l I

~

o

(

~ I

}.

~ "I

}

~ '

I I

I

'rom: rwrroor rouro (j or orrrn>:::

l I '

V = -tO '~'I>>

Or-0-" O... 0

- t))I

..V'>>'

~

~

I

-: O.

..'0-. O=.n;"

O

~

I I

~ r n)O

~rt (r)

>I t')

W ;II 0 j.

(O m

~

D n

II Irt

~ 'C)

In

~ rrJ['i[

'.Iril

~>>IA I

~ ~

O AI0 IV ~

0 I.

~Ii'

=rj-

-r}l n'}j Ir)

~

~ E:

--I (.I

it t Lr

~ ~ I

~

. ~.".VI

~err}wrr c'=(-...

I.A.'

II I I

.:I~l r I-~

~ h I \\,

l&lJ

~r

~

~

rk lr) '

e

~ l.=-=

~L~

,L e-l.'

IL In JC) l.I O

I tl I..

I I

I

}

r j I

r.~

-i:.. "I I

l

}".':...

~..

~

~

r

}

i

~

~

B~FZTY L ill.

LIH?TIXO GAY/ fT ST'>T'AI St'TTIX~i

a. 2 REACTOR COOLANT SYSTEM INTEORlrY 2.2 REACTOR COOlANT STSTVf INTEGRI'ooixcabilit~

A,plies to lieiita on reactor coolsnc

~ oyuceis pressure hpplieo co trip settings of the instr~ence and devices vhich are provided co prevent the reactor system safety limits fzon being, exceeded.

Object ive Tv escaolfsh a lieiit belcv rhich the integrity of the reactor coolant system ia noc chreaccned due co an overp: esaur'e condicinn.

0~bc c t i v c

. To defi>>e the level of the pcoceas variables at vhich automatic.pro-tective action is initiated to prevent che pressure safety li~it from being exceeded.

S ec Ifica c,inn 5 deification A.

The pressure ac the lovest point of the reactor vessel shall noc

. exceed 1,375 psig vhenever irradiated fuel i.a in tha reac>>

cor vessel

~

Protective Action Iitiiting Safety Th Iinicing safety sysce~ settings shall be aa specified belov:

A.

Nuclear ayatea safety valves open nuclear syatee pressure 1250 psig

+ 13 pai

{2 valves) 8.

Nuclear system relief va)vca open--nuc1 ea r system pressure 1105 pai'g +

ll psi (4

valves) 1115 psig +

Ll pai (4

valves) 1125 paia

+,

ll psi

{3 valves)

C.

Scram--nuclear systea high pressure c I,O55 pai ArrePdnent NO. 35

<'ub 'I 27

1.2 BASES

REACTOR COOLANT SYSTE'f INTEGRITY The safety limits <<or the reactor coolant system pressure have been selected such that they are below pre'ssures at which it can be shown~

that the integrity of the system is. n'ot 'endangered.

However,,the pressure safety limit:- are set high enough such that no foreseeabl'e circumstances can cause the system pr'essure to rise over these limits.

The pressure safety limits are arbitrarily'eleCted to be the lowest transient overpressures allowed by th'e app3.ic'able codes,;

ASHE Boil'er

'nd Pressure Vessel Code,.Section TII, and USAS Piping Code, Section B3l l" The design pressure such that, when-the Boiler and Pressure added to the design is established.

(1,250 psig) of the 'react'or 've'ssdl is established lO-percent allowance (125 p i) allowed by the AS>ifE Vessel Code Secti'on II'E f'r pressure transients

is

pressure, a trans'ient pressure limit of 1,375 psig Correspondingly, the design pressure

'(1,148 psig for suction and 1,326 psig for discharge) of the reactor re'circu3ation system piping are such that when the 20-percent allowance (2'30'and 265 psi) allo~ed by USAS Piping Code, Section B31-l for pre. sure transients are added to the design pressures, transient pressure limits 'of 1,378 and 1,591 psig are established.

Thus, the pressure safety limit appli'cable'o pOwe'r operation is established't 1,375 psig (the lowest transient, overpressure allowed by the pertinent codes),

ASNE Boiler and Pressure Vessel'Code,'ection ITI, and USAS Piping

Code, Section B31.1.

The current cycle's safety analysis concerning the most severe abnormal'perational, transient resulting directly in a reactor coolant system pressure increase is given in the supplemental reload,- licensing submittal for the current cycle.

The reactor v'eschel pr'essure code limit of 1,375 psig given in subsection 4.2 of the safety'nalysi's report: is well above the'eak pressure produced oy the overpre'ssure'r'ansient described abdvel

Thus, the pre. sure safety limit appli'cable'o power operation is well above the peak pressure that can.result due'o reasonably expected overpre0su're

'ransients.

Higher design pressures have

'been established fOr 'piping~ within the reactor coolant system, than for the reactor vessel.

These increased design pre'ssures create a consistent design which assu're<

that, if 'the', the pressure with'in

'he reactor vessel does not exceed 1,375 psig, t:he pres ures withi'n the piping cannot exceed, their respective'zlansient pressure limits du'e i'.o

'tatic and.pump heaids

~

The safety limit of 1,375 psig actually applies to any point in the reactor vessel;

however, because of the stati'c water head, the highest pressure point will occur at: the bottom of the'essel'Because the pressure is not monitored at this point, it: cannot be'eere~'tly determined if this safety'im:it has been violated.

A:iso, because of 'tha pote'ntially varying head level'nd flow pressure

drops, an equivalent pr'essur~

c'annot be, apriori determined for a Amendment'o.,'g, 58

1.2 BASES

~~'p>j:4

,~'ugg pressure monitor~'hi'gher in the vessel.

Ther'efore;.following any transient that is severe enough to cause concern. that this safety limit was violated, a calculation will be performed using all available information to deter-mine if the safety limit was violated.

REFERE'ACES Plant Safety Analysis (BFHP FSAR Section 14.0) 2.

ASiK Boiler and Pressure Vessel Code Section III 3.

USAS Piping Code, Section B31.1 4.

Reactor Vessel and Appurtenances Hechanical Design (BOP FSAR Subsection 4.2) 29 Amendment iNO. g$, g5, 58

2.2 BASES

4 REACTOR COOLANT SYSTEM Zi%TEGRiTY The pressure relief system for each unit at the Browns Perry Nuclear Plant has been sized to meet two design bases~,

Pirst> t'e, total safety/relief valve capacity has been established t'o meet the overpressure protection criteria of the AS~iK Code.

Second, the distribution of this required capacity between. safety va]Lves and reliqf 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 dnring normal plant isolations and load rejections.

The details of the analysis which shows compliance. with the AS'i'K Code requirements is presented in subsection,4.4 of the PSAR and the Reactor Vessel Overpressure Protection Summary TI echngcal Report, submitted in xesponse to question 4.1 dated December 1,

1971.

To meet the safety design

'basi.s>>

thirteen safety-rel:Lef valves have

~been, installed on uni.t 2 with total capacity~ og 8Q.2j.'f nuclear boiler rated steam, flow.

The analysis of the worst overpressure transient,i (3-second closure of al]L main steam line iso.'Lation valves) neglecting

the, direct scram (valve pos:Ltion scx'am) results gn a mazimum vessel pressure i

which, if a neutron flux scram is assumed considering one relief valve is inoperable'

.-,has, adequate-margin to the code allowable over-pressure limit of 1375 psig.

To meet the,operatic>nal design basi.s, the total safety-re.Lief. capacity ox 84.2/ of nuclear boiler rated has been divided into 70X rel'ief (11 valves) and 14.2Z safety (2 valves).

The analysi.s of the limiting plant isolation transient is presented in the supplemental reload ].icensing submittal for the. current cycle.

This analysis shows that 10 of ll'elief;valveS limi,t pxessure at the safety valves to avalue whLch is below the, setti.ng, of the safety valves.

There-

fore, the safety valves will not open.

This, analysis shows that peak system pressure i.s, 1:Lmited to a value. which,is well below tbe allowed

~

vessel overpressure of 1375 psig.

30 Amendment No. 3$ >>

A$ >>

58

TABID'.2.F

&iryeillance Inatpuaent'ation O

~t

~t CJl 00 HlnlauAi //

of'pposable Instniiaent Channels Instrument //

lf2H - 'l6 -

94 lf~H - '76 -

104 Insti~iment Drywoil aiiil'orus llydrogen Concentration Type Indication und lian e O.l - 20) lfotes Pdl-.64-13'7 PdI-64-130 Drywell to Suppression Cliu'uiber Differential preeoure Indicitor 0 to 2 paid (1) (2) (5)

NOTES FOR TA5'Lf.'.2.P (R

(I.)

tron and~ rer the dace that one'f'hese pTFanecer.s is. reduced to

/ (

one indication, continued operacion's pernissibJe during chs succeeding thirty days unless such instrumentation is sociner cade operable (2)

From and after the date tha one~ of~ these parameters is not indi-cated in the concrol roon, continued operation

's pernissibl'e during, the succeeding seven days unless such instrunentation is sooner sade operablc.

'(3)

?f the requirelsents of notes (1) and (2) cannoc be net, either the requirements of 35.H sha3.1 be cnplied vith'or an orderly shutdovn sha3.1 bc initiated and the'eactor shall be in a Cold Cond i'cion vith kn 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

(4)

These surveillance instrunehts a:.e considered to be redundant to each iother.

Xf the requirements of one of the indications orderly shutdown shall a Col,d Shuedmm witELn notes (1') and'2) cannot be metand

'if'annot b'e restored

~~ shan,(6)

hours,

'an'e initiated Md the reactor shall be in 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />s~.

80

O.

3/4

~ 3 BASES:

does provide the operator with a visual indication.of neutron level.

The consequences of reactivity accidents are functions of-the initial neutron flux.

The requirement of at least 3 counts per second assuzes that any traysient, should it occur, begins at or above the initial value of 10 of rated power used in the analyses of transients from cold conditions.

One operable SRM channel would be adequate to monitor the approach to criticality using homogeneous patterns of scattered control rod withdrawal.

A minimum of two operable SRf's are provided as an added conservatism.

5.

The Rod Block Monitor (RBM) is designed to automatically prevent fuel damage in the event of erroneous rod withdrawal from locations of high power density during high pex level operation.

Two RBM channels are provided, and one of these may be bypassed from the console for maintenance and/or testing.

Automatic rod withdrawal blocks. from one of the channels wi1l block erroneous rod withdxawal soon enough to prevent fuel damage.

The specified restrictions with one channel out of service conservatively assure that fuel damage will not occur due to rod withdrawal errors when this condition exists.

A limiting control rod pattern is a pattern which results in the core being on a thermal hydraulic limit, (i.e.,

MCPR given by Specification 3.5.k or LHGR of 18.5 kw/ft for 7 x 7 or 13.4 fox 8 x 8, 8 z 8R 6

P8 z 8R fuel). During use of such patterns, it is judged that testing of the RBM system pxior to withdrawal of such rods to assure its operability will assure that improper withdrawal does not occur.

It is normally the responsibility of the Nuclear Engineer to identify these limiting patterns and the designated rods either when the patterns are initially established or as they develop due to the occurrence of inoperable contzol rods in other than limiting patterns.

Other personnel qualified to perform these functions may be designated by the plant superintendent to perform these functions.

Scram Insertion Times The control rod system is designated to bring the zeactor subcritical at.the rate fast enough to prevent fuel damage:

i.e.,

to prevent the MCPR from becoming less than 1.07.

The limiting power tzansient is given in Reference 1.

Analysis of this transient shows that the negative reactivity rates resulting from the scram with the average response of all the drives as given in the above specification provide the required pzotection, and MCPR remains greater than 1.07.

On an early

BWR, some degradation of control rod scram performance occurred during plant startup and was determined to be caused by 131 Amendment Nr. gg, P$,

58

.~ i part fculace c<acarfal

()Irahably consctuctfop Jebtfs) p'<sigfnj an internal control rod dtfv< ffleer.

The deafen OIf the ptcsenc conc ro'odi drive (Rode)

VICBSlcr B) is pro. sly fr<rptoved by the relr cr<i)c n nf the ffl cer co a lo-.ation our of the scraa drive parh:

C.e.,

fit Can nr languor fncerfere v!cn scrao pstforaance, even C:

r~<<<p) cce) v blc eked.

The dr trade J petf or~ance of the otf Bfna' dr f "e (c-<CD'>)CDBlc"~)

unde< dfrtv operatfng conditions and the insensitivity of the tedr ~'Igned drilve (CR.,"R>B) ~4B) has been deo'onst rated by 4

sar f.s o! <nginerrfnIr te ~ c s under sfnu) aced tractor operat fng cond;r fons The auccessf u) pe. fornance o.'he nev dtfve un<fet actual operating, cond f r fons haa 4).so beer.

der<<anscraced by cons acrnc'y Icoor fn sr.vf e I'est tesu) cs fior ip an a in che net Irf vc and w) be 1>> f er rcd fr w plants us inc rhe older mde.'riu Vfth 4 nadf f CCd <larger Se.een Cfire.

Iintetnal f'lter Vh'Ch fo 1.sa prune tc pluming.

Daca has been dccucoented by su. reil-lanc: tepcr:4

<n vittbua opetatf":p. planta.'Theaa i-elude Oyster Creek, Nontfcello, Dresden 2 and Dc esden 3.

• pprox'mately SQOQ drfve Ctscs have baen fecot ed to date'.

Fol) r vfng, fdencf.'f atf on of'hc

'plugged,

. fit et" ptable=,

very frequent srr ~i test~ v<.rc>>ecess~rv to enr<urr proper perf'a~ance.

Havever.,

the i~ere f'requeOC Sera~

CeSCS

a. 4 nOV Canaideted tOta unn~cescar

< and invfae fct the follov<<<nF, reasonsrtac fc saris.

pe. Ifor"a>>ice has been fdenc f f fed as due co rn obacrucced Ct'vc,'!fleer frr type "A" drfivea.

Th>> dtfves EF!Pr are io. the neV "B'" type deaf gn Vhaae

'SCt<<.'CtfomanCC fs unaI, rE'c<<ed by if 1ce candf<<io'nJ 2.

Th<

C:rc 1oad reactor vhen to.flovs and c fon cnd aea<

~

syater<<<i.

Re<

~ trokc

~

4!'Aa

)

re! ue1 IInIt cyr

<erfor;inc e.

'iu!ffc I enC to prfinetf ly relrasrIC dirfngi scattuo of'hc CIIic reactor and f '4 s vs'<.r:4 art ffirst sob]cc cd pr ean <te and cher.-a! 't teal ~ ea.

"Spwcfa) atccn-

~rreC

< re niOv be fnB taker< ta aisutb C'leaner ctora vith drf ves ident k< a1'or'fnilar (shot'ter-cr'f<<or: steat

) h4ve opera c ed through conj les v? th no suddcr. ot circa't fc'hantea fn acta=

This preopetat Canal ~and atsttup tI sting fs detect anoia<rloua )tive 'e".!o~nce.

3.

he 72-hour nutate<< lfr:ft vhfch fnfcise ed che scarc of the taquefrt acrr m cene fnr, f 4 ar'bi ctarIY, 'avcni-no logical basis other chan quancf I,'vfng, a "na)ot oucafre"'h'fch night r'eaaof'4-bly be cau ~ eC bY ILn event 40'evere As Co poco b

y 0

feet.'tf ve petlfot'vance Th f4 I equi tensnc 's 'nvfse because f C provfdec ar: fncent fve for shor cu." adtfbna'O tisat'en 'terutnf sq "on ifr<e" to avofd the addi tfonal 'ceacfhB

'due a 72-haut outage 132

3 3/4 S

sass@

The surveil lance requirement for scram testing of all the c'nntro?

rods after each refueling outage and lOZ of the control rods at l6-Meek intervals ia adequate for determining the opera-bility of the control rod ayatca yet is not ao frequent aa to cause excessive vear on the control rod system components.

The numerf cal values assigned to the predfcted acraN perfor-mance are based 'on the analysis of data fros other'MR'a with control rod drives the came ae those on groans E'crry Nuclear Fiant.

h The occurre'nce of acron tieea Mithin the limits, but signifi-cantly ion-..er than the average, should be viewed aa an indica>>

t ion of systematic problem Mith contre]

rod drives especially if the nur)er of drives exhibiting ouch scran tines exceeds

eight, the atlovable nuceber o'f inoperable rods.

ln the analytical treatment of the transient ~,

390 milliseconds are alloved betveen a neutron sensor reaching the scram point and thc start of negative reactivity insertion.

This ia ade-quate and conservative vhen compared to the typically observed tine delay of about 270 nilliseconda.

Approxiuately 70 milli-seconds atter neutron flux rcachea the trip point, the pilot acran valve solenoid pover supply voltage goes to zero an approximately 200 milliseconds later, contzol rod cotion'beg" na.

Thc 200 ailliseconda are included in the allovable scram inser-tion tines specified in Specification 3.3.C.

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

Individual rod bypass svitches may be.

used as described in specification -4.3.C.1.

The position of any rod bypassed must be known to be in accordance with rod vithdraval sequence.

Bypassing of rods in the manner described in specification 4.3.C.1 vill allow the'ubsequent withdraval of any rod scrammed in the 100 percent to 50 percent rod density groups; however, it vill maintain group notch control over all rods in the 50 percent density to preset pover level range

~

In

addition, RSCS vill prevent movement of rods in the 50 percent density to preset pover level range until the sera'mmed rod has been vithdrawn.

133 Amendment tto; 32 V

C

,3,: 3/4. 4 BAS.ES::

,During. each:fuel cycle

~ exce~ss~ nperative reactivity varies as f'cie'.1 depl'ates a,nd, as any burnable'oison in. supplementary contro'1 is bpr~xep.

The magni.tude of 'this'xcess r'eactivi ty mayi be inferred from the,

.critical roc! conf'igu'r'at,ion..

As 'fuel burnupi pro-

gresses, ancimalous behavior.in 'the excesi, reactivity may,be detect ed 'by.compa rison; of the criti cal-rod pat tern at, selected base, states, to the pred'icted rod j.nventory at that sta te.

Po~er oper'a't'ing'ase con'dit'ions 'j<rovide the ~most~ sensitive,and di'rect ly

,i n t e r pr e t a'b 1'e d'a t a r e I a tiv' t o co r e r e a c'tivi t y.

"Furthermore, using power iopiereting base, conditions

per'mi'ts'requent'eactivi~tyi comparisons.

Vi V

.Requir'ing a.reactivity icompiarison,at the specif'ied f'requenh.>

ass.ures that a icomparison Mill be made

,before the c'ore react ivi t~y exchange, exceeds

.kX A Deviations in.core 'react'iiviity, greater than '1ZB/ are

,not expected.

and require thorough evaluation.

One

.percent

'reac't iyity'nto the, core would not lead to

.transients, e'xteeding,des iign cond itions of thai'es c tor sYS:,cern.

.References 1,.

iGeneric,R'eload Fuel Appl<Lcation, Licensing 'Topjc4alI Report,;

NEDE-Rv~~Ol 1-)?-AB 'and Addenda.

4 V

e 1

E S

i4 134 Amendment, Ho.~3$ gp,,

58

(T I tlc CO>ul I los'S VOA OPFaAT IO.'4 ileekvr. ~ LLd'4MCI.

lee

~ u l.e:.u'ITS

.4 Maintop nan.c otiliad '(hi "char

l. Pine

'e suction of t'h e RCIC and hPCI. pumps befall bc alimed to the condensate s'.oragc

tank, and the I>res "urc supprfa-'.on chamber head tank "hall. nora:ally, br aligned tn erve l.he dl sc-hnrpe piplin,".

.of Ihc Bllll and'S

'pi@"."s.

Tl c ccnden-ate he~d tnn". mny bc used to s>>. Ie the R!III, an]

C. fllncharg>> piping il'.hf PSC nead took I s unnva ilab'.

'1 n nr: s.:urc Indicator" nn thc di scharg:

o." the RK+

'nd CS p<imps shall indicate no less l th.zn list e l belou.

l Pl-75-20 49 psig PL-75-46 49 psig Pl-74-51 49 ps>g Pl-f 4-f"f lfn ps'l I.

Av ~ rape Plarar Lin>>ar Heac Generation Pl d

Ouelnr, s cade scute pouar operation, the Ma> imum,Average Pl'anar Heat Generation Rate (MAPLHGR) for each type of fuel 'as a function of average planar exposure shall not exceed the, limiting value o'

Tal Les 3.5.l-1.-2.-3.-4.

and -5.

Tf.it any t lmf. during n'peratlon it is Ari crmined by norma'i" survei I 1'ance,that.

t lie 1 iml c ln,". va lue foi APl HGR I s;be lng

'x>>edged, at e ion shi'I l be

. init iatcd uith-in 15 minvern tc r>>store operation to ilrnin rhc Prencr i bed 1 imi's.,

If che API HCR is nor returned to ulthln the pr ~ acr ibad 1 lmlts ul thin tuo (2) hours;,

the reactor eih >I'c brough to the Col'd 4

hutdof-~ cone!ic.'on uit'hin 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Survd I 1]ance and corresponding action shall cont fnu>> until reactor operation i is ~lrhln the>>rcncrlbed

.'amies.

tf I

Liniar Heat:"neration Rat" (LHGR)

Dur.ng steally state pouer operation, che 1'n.ar heat generat ior. race (LHGR) of any rod in ar; f <el zsserab'ly at any i

axi.hl loess lun shall noc ecceed the max;num al',oval ld I.HGP as calculated by)

Che fo 1 1 nhd lug

~ quar inn:

4.'D.ll ffefn en: ee ef F'1'ed Dfeehe.df". Pf.

l.

Every month: prior to the c!stl~g of the RIBS (LPCI and'oncai,"~en Spray) and-co're spray systems, the discharge piping of these systems shall be vented from the high point and vater flou determined.'

2.

Folloving anv per'iod uherc the LPCI or cere=spray systems have not been require'd to be operable, the'di's-charge-piping of. the inoperabls sys-te= shall be vented from the high point prior to the.return of the system to service.

3.

,Whenever the HPCI or,RCIC system lired up,to take, suction frofa the condensace storage tank,, che dis-charge piping of the HPCI and RCIC shall be vented from che high po'inc of the syst'm and uater flou observe.'

on a monthly basis.

4.

Men th R!IRS,and" thc CSS are t>>-

quirexl to be operable, th pressure iqdicators uhl,ch monitor the d is-

" charge 1'ines sha'll'be monitored daily and'he'ressure record%

e J.

'Linear Heat'eneration Rate (LHGR>

be The LHGR as a fusee ion of ccrc '. e'.'"

checked'aily during reactor cperat lan at 25X rated thercal pouer.

Maximum." Ayers" e P'ansr Linear Heat Sererw-tion. Rate

('IAPLHGR)

The 'dIPPLHGR for each type o! Fuel a~

a fu:c-tion of average planar exposur shall be deteanined daily during reactor operation at 0

25Z rated charm'al power.

4 Amendment No. g$,

P$, 58

l.vlITINO CONDITIO EPOR OPERATION L

LLANVE RF~VIRFITFNTE LHGR < LHGtt,,(1 - (Ap/p).(L/LT)3 LllGlt

~ Design LllGR NE 18.3l kW/ft for 7x7fuel d

t RR 13.Cii klan/Ct for SXS,

8xSR, nnd PSxSlt.

fuel (dp/P)

~ Maximum power spiking penalty i

~ 0.026 for 7x7 fuel

~ 0.022 for SxS,SxSR,and PSxSR f'ucl LT Total core lengthen

~ 12.0 ft foi 7x7

& 8xS NN 12.5 ft for 8x8R & PSx81)

L ~ Axial position above bottom of core If at any time during operation it is deter" mined by normal surveillance thiat the limiting value for LHGR is being exceeded,

.action shalli be initiated within 15 minutes to restore op>>ration to within iche prescribed limits.

If the LHGR is not returned to vichin the prescribed limits within two (2!) hours the reactor shall be brolughit to the Cold Shutdown condition within 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />sK Surveillance and corresponding accion shall cont:inue until reactor operation is within che prescribed 1imits.

K.

Minimum Critical PcnFer Ratio (MCPR)

The MCPR operating limit: for BFNP 2 cycle 4 is 1.32 for 7X7, 1,27 for SXS, SxS]R, an 25K paced thermal power and fol-Tovlpg any chang>> in j~over level. or distribution eh t would cause opere cion with a lbaicing control rod pattern as described in the bases fc Specification 3.3.

If'at any thee during operation it ia determined by normal surveil1ance chat the liTFLiting value for MCPR is being exceeded, action shall be initiated vithin 15 minutes co restore operacion co vithln the prescribed limits. lf the steady state

."lCPR,'is not returned co within the prescribed limits within cvo (2) hours, che reactor shall be brought co che Cold Shutdown cond iclon Wchiis 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Surveillance and cnrresponding accfnn shall cont tnuP dTTt t'l ! r.RL cnr

@pe rac inn is vachln cbr prrpr r ll>ed l lmlco.

L'~lolf lgl II Nlll~ c K l'(.

Ef any nf the lim'.ting valcFog ldenc inflect lni Spec'.f tcac tons 3.5. I, J, or ll, are;exeeedeL!

chc opec'fled remedial accion ls taken, chy event shall be logged and ze;Forced'n a 30-day written report.

.'Amendment No. g$, Pp, 58

W~rS 5.S.C Autaaatic Oe reesurisation S <<tea (ADS)

Thie opecif ication ensure<<

thc operability of the ADS undet'll condi-tione for vbfch the depree<<urination of'he nuclear

<<yeten ia an essen-tial re<<ponae to etation abnoraalitieo.

The nuclear ey<<tea pre<<aura relief ay<<tea provides autocsatic nuclear ay<<tee dcpre<<surisation.

for small breaks in the nuclear ay<<tea ao that the ]ev-.pres<<ure coolant in)ection (Ltcl) and the coro spray eubeyateaa can oprrate to,pro(ect,the fuel barrier.

leyte that this apecificat&n

~ppliea only to the automatic feature of the pressure relief <<yatea, Speciffcation 3'.6.0 apecifia<<

the requiraoenta for the,prQ<<sure relief function ef the valve<<.

Zt ia possible for any number of the valves a<<signed to the AOS to.be incapable of perfonaing their ADS functiana

~

hecauee of in<<truaentatinn failurea yet be fully caoabla of perforaing their pres<<urs relief function.

Secauae tl~ automatic dcpreeaurization

<<y<<taa does not provMe makeup to

~ the reactor priaary veeael, no credit ia taken'or the otean cooling of the core caused by the ay<<tan actuation to provide further conservative to

@he C"CS.

hfith two ADS valves known to be incapable of automatic operation, four valves remain operable to perform their AOS function, The ECCS loss-of-coolant accident analyses for small line breaks assumed that +0<

of the six ADS valves were operable.

Reactor operation w)th three ADS valves inoperable

)s allowed to continue for seven days provided that the HPCI system is demonstrated to be operable.

Operation with more than three of the six ADS valves inoperable is not acceptable.

Amendment No.

35 167

SA PS 5'.5;II" ~ifntcnancc of Filfid'isc~Is.ir e Pf~e f

of the core spray.

LPCI. HPCIS.

and ItCICS are not

tilled, a vater hailer can develop fn thf ~ pfpfnr. eben the Instep and or d

T afnfnf'tc da~aac to the discharge, pfpfng and to ansu;c achlod Isarcfn fn the operation of.,these systems,.this Technical Speci ffcatfo~

requires the discharpi lfnes t'o be ff'fIIed vhenever the system fs fn ondftfon. lf a'ischarge pion is not fflied, the pumps thit <<ups Iv

.th'at line adust be assuaged to be fIiopIert~blIs for Tec!hnfcal Spec posva.

ii The core spray and I'CHR syste~ dfschqrre pfpfna hiah poi'nt 'vent

)s visually ch'ackc4 foe'ster flov once a svsnth,prf'or to testing to enoure that the If+is are f I fled.

The"- vis>>a'1 checkfng, Mill 'avdid 'tartina the core spray or NIIt'yat'ee.vith n disc!iarge 1 fnc,riot, f(11'ed.

In addition to the ~usual observation'nd to ensnare a'11.led gach'arge line other.than prfo to testing, a'..presae suppression.-ch'aib'er'eIuf,t'axjk

$'s locate'.d approximately ZO feet abo~

the'ischarge line bfghpofnt t'o supp~i, makeup'ister for-these systeeuI.

The t,

he'ad tupik,'located approximate~

100 feet abcfvo the discharge high con'd<<nse,,e e'a su reasfos c5a"be".

pointczars as a bec&ap charging'yst'm vbeo the pressure supprea

}iead t'ank. f'i: not fn serrf'ce'.

'Systest Q'scl&rge pressure imdfcatprs, a"e Imed tc det'eiiaine'- the: vat'er le'rel'bove t'e df'schILrge linc high point.

The

!fndLca 0 '5

'nt vilIre&'t~t'pproxfmat'sly 30 psi'g for a vater level at the high po.nt and

~

pvf'g for a'. vater: 1ciml.'n. the'repsureepxppressfon chImbe-.

heed teak aod, are

=~

itored'aily to ensure'ha't the-discharge li'nes are. fiU.e4.

Ken'n'hcfr nonenl st'zndhy cnndftinn the suet I'c>n for tbe'IPCI

.ind RCTC

< are -el I j;nc Il'n; the. co.-Ai.nsa'te stqriI.e tank,,vhfch is phvi fc~11y ic hfah'ri'li'vat'Inn th;in'h'c'IPC1S'nd'CICS'lnlnn.

This assures that th<

0'ndi aCTC. dficharad" pipin~ rcr afns~ ffiled.

Furtlrcr assurance is nrnvtxled by nb'aervfn!<:, vote t'loM. fr~'- these'ystoaq',h'$'ahi points ~nthly ~

tfaxfaua Ivorage'Planar Linear Heat C;ana,rat,fus Rita" (HAPILnCX)

Th'is: speoiEfcatfon assures that tbe peak': cladding ti~pera,ture follovfng tbo postulated dcafgn bzsf'i: Ioss-of-coolant'ccLdent vill not exceed tha 1',imit specified in the

'IIOCFR50;46 AIppendix K.

The. poak= cliddfn(('emperature'olloyfng': ~ot'ul'ated loss-oL-coolant acci-denta'rf barfly a:. funct'fbn of'h., averqgc, heat','Ler'icratfon rotc of aII the rode of" a. f'ucl asseinb'ly. at'ny: axf".L'I 1'ocatfon an'd f'<< only, d."'.pendent second-arily on'h..

rod. to rod poMer distr)bu~tipn yfthfn an asse~bly

~

Since ex-pected Ibex 1 var fat'fons 1'n pointer di"tribution v) third a'uel assesilily aif feet th't< ca'fcufa:cd peak clad teuipcra'tvr'q b'y lbss tIian 20 P rela Ivc co.the pea'k'c~psraturc for' typi

< a 1 (ucl, desi pn,, 'the 1 fnft on the average I ines r Inst'acner'ation rate" is suf f'fcicr t go pssurg chit, caIculatcd ten pctiturcs a'e.'within th'e lOCI=.R50 Appendix K liiIIit.The limiting value for MAPLHGI< is

.;hovn'n'a)]es-,3.).g-.l,-p,-3,-4,.

g-5, The Mlclyses Cupp~rtrng Cne"e 1'i+itin'g'alues. is pre'sent'ed

'n reset'enCe

'4.'mendment',

No'; p$;, gg', 5S

r~ (,f

.Lfnchr Hest Gcn>>rac Rate I.IICR Tl:fs op>>cfffcacfon assures Clast tho linear heat gencrhtfoh rote in any i's less 'chan the desfgn linear bent g ncration ff fuel pellet denhfffcotfon is postulated.

Thc pouer spike penalty specified fs based on (he anal-yofs pre's<<ncod'n Section 3.2.1 of Rcfcrencc 1 as modf fied fn. References

~

2 snd 3, ond assumes e lf>>tnrly fncrea inC vai fstion i>> o"Ial gaps bc-tvecn core bottom,and

cnp, e>>d assures vl th' 951: co>>f fdc>>cc, that>>o morc thon onc fuel rod cxc.<<cds the desi'>> lf>><<ar hcoc" Cencrhcfon race duc to povcr spfl:fnt.

Tnc L)ICR a a funccfon of coi'e licfght shall bc checLud daily dur-fnG reactor operation ac

> 25Z pover to determine if fuel burnup

~ or co>>-

trol rod. movcmenc 'hos cau cd changes fn povcr dfstrfbucfo>>.

For LIIGR to bc a 1'faf tf>>g value bclou 25/ raced thermal povcr, the R factor would have to be 0 24] vhfch fs precluded by h considerable margin vhen employing loess than sny ocrmissfblc control rod pattern.

3.5.K.

Hfa~m Critical Pover Rseio FCPR F

Ac core thermal pover 1>>vela lese than or equal to 25, th>> reactor vill be opsrstfag at mfnfmum recirculation, pump speed and the moderator vofd content vill be very small.

For all designated concrol rod patterns vhich may be em-ploysd at this pofat, operating plant experience and thermal hydraulic anal-ysf ~ fndfcac>>d that the resulting MCPR value is in excess of r'equfren>>nts by s cons id>>rable mhrgfa.

4fth chio lov void content, any inadvertent core flou increase vould only place operation in a more conservative mode rela-tive Co HCPR.

The daily requirement for calculating HCPR above 25X rated thermal pov>>r is sufffcfeat since pov>>r distribution shifts are very slav vhen there hsv>> aoc been sfgnfffcant pcver or control t'od changes.

The requfrem>>nc for calcqlatfng HCPR vhen a limiting control rod pattern fo approached ensures that MCPR vill be kaova folloviag a change in pouer or.pover shape (regardless of macaftude) th~ r. could'lace operatioa at a chermal limit,

~R>> ortm

.Re ufremeats The LCO's ossocfsted vfth monitoring the fuel rod operating condicfnns are required Co be met oc all cfmes, f.e., there fs no allovable time in,vl.ich ch>> plane caa kaovingly exceed the lfnftfng values for MAPLHCR, LHGR, and HCFR.

Ic fo s requiremeat, as stated in Specifications 3.$.1,.J, chat ff ac aay cfme duriag steady state pover operatfcn, ft is determined chat che limiting values for HPgLHGR, LHGR, or HCPR a-..o exceedal act'on fa then 'fnftfated to restore operation ro vithin the prescribed lfmfcs.

Tnis sctfon f e iaitiaCed as soon as:"normal surveillance fndicates tha-an operating ft b*s been reached.

Fach event involving steady state operation beyond a specff fed 1fmfc sbh11 be logged sad reported quarterly.

?c must ae reccgnf-ed ch-t cher>>

fs alvays an action uhfch vould return aay of the p react ~ =o (HAPUfGR, LIIGR. o:

MCPR)

Co vfthfa prescribed limits, namely pover reduction.

ader most circumstances, this vill aot be the only alternatfve, H

Ittfef>>nw'ts 1.

Pusl Dens fffcat ion FffecCs oa General Electric Boff'g Matc RR J tor Fuel," Supplements 6, 7, and 8, bECY.-10735, Avguat 19.3.

Supplement 1 co Technics'eport on.Densif icacicas of General Ef>>ccrfc Reactor

Rsels, December 14, 1974 (US* Regulatory Staff).

Ccnmuofcacfoa.'.

A. Hoot'e to 2, S. Mitchell, "Hoofffeo CE Model for Pu>>l Densification ~" Docicec. 50;321f March 27>>1974.

4.

Generic Reload Fuel Appliqation, Licensing Topical

Report, NEDE>>24011-P-A,, and Addenda.

169 Amendment No., PK. N ~

58

i4,5 'd Cont a'men Co n

5

~ 'ees Bur ve L~lla..c~

Fr@ uencies Core a

1 t

ol~f~y

";Vite. testing interval 'for the core and containmen~r c ~lfng <<ysteias fs based on industry'ractice, quantitative relfabflfty analysis, fuddlement and

.practtcal'fty.

The'core cooling systems hav'e Hot'een designed to be

'fuLly'est'ab'lc

-during operitfon..For example, in the case. of'he RPCX, sutoaatfc

.inf'tfatfon during power atperation vould <<esult fn pumping cold water into the reactor vessel which fo not desirable.

~ Complete ADS testing durfng pover operation causes an undesirable loss-of'-coolant, inventory.

To increase thc svaflabflf'ty of the cori! and containment Cooling system, the components vhich make up thc system; i,.e., fnstrumehtaltfhn, pumps, valves, etc., are tested'requently.

The pu'mps "and motor operated fnfectfon'valves

<<re also tested each. month to assure their operability; 'h simulated automatic actua-

'tion test once each cycli., combined v'ith monthly tests of the panspi and injec-tion valves is deemed to be adequate testing of ~these systems.

Mhen components and subsystems, are out-of-service, ovii.'rail core,and contain-ment coolfnr. rclLihILltty is maintained by dccmonstra ting the.operability of thc rene inin"cqu1 pmcn t.

Thc degree of ope'rab L 1 icy to be demo@stra ted depends on thc nature of the reason for the out-of-service equlp~ent.

For routine outwf-'servfcc perfods caused by prcventati've'aaintenahce, etc.. the pump and vol've operabflLty checks.vfll be performed to-demonstrate operability of the

'remaining comjoncnts.

Hov'ever, ff a faflur'c,

'e'sing'n deficiencycause the outage,, then'he demonstration of opcrabfli'ty'ihould be thorouFh enough to as'sure that a gener!Lc problem does noe. exis~t. 'or example if an out-of-

servfce, period, vas caused',

by failure of. X,p'urn) tb deliver rated capacity due, 'to a design deficiency, the other pumps of this type mfght be subJected

'to a.'flov rate test in addi tfon.to the operability checks.

Mhencvcr-a 'CSCS system or loop is made inoperable because of,~ required teat: or cal'ibration t'h e other CSCS systems or loops.that <<re.required to be operable: shall be considered operable if they ark within'he required surveill-

,ancee.'testfnr, frequency'nd there fs no i>ca'sor'i to suspect they are inoperablc.

?f:th'e function, system, or loop under test or calibration is found inoperable or exceeds'he trip level scttfng, the LCO 'and the 'rei'quired surveillance, t'eating for the system or loop shall apply.

fi Redundant operable components are sub)coked tk fhci'eabed testing durfng. equip-aent out of-'. service times.'his adds fuirth'er ',cohse'rv'it f'sm,'ind increases assurance, that adequate cooling is avaflabl'e,.should the reed arise.

Maximum Ayers e Planar LHI~~R. UtGR~and HCPR The NPIAGR, U(GR, and MCPR shall be checked daf)y to determine ff fuel burnup, or,.control, rod. movement has caused changesi i'n poMiar 'distribution.

Since chsiges due: to burnup are slov, and on3.y a fev contr'ol 'rodi are moved daily, a daily check of power distz fbutfoo is adequate.

170

TABLE 3.5.I-5 KQ'LHGR VERSUS AVERAGE PLANAR EXPOSURE Fue1 Type 8DRB284 and P8DRB284 AVERAGE PLANAR EXPOSURE

~".fd/t 200 1,000',

5,000 10>GOO 15,000

~ 20,000 25,000 30,000 HAPLHGR

. kV/ft

11. 2 11.3

11. 8 12.0

12. 0

11. 8

11. 2 10.8 PCT

('F) 1685 1667 1671 1647'669 1'672 1633 1596 172a Amendment No. Q, 68

0 I

~

il

0 6 4.6 BASES detect'cd reasonably in a matter of few hours u ilizing the available leakacc detection

scnemes, and if 'the origin. cannot be determined in a reasonably snort time tne unit should be snut down to allow further investigation and, corrective action.

The total leakage,;rate"consists.of all leakage, identified and.unidenti-

.fieD, iwhich flows to the dryw'ell floor drain anted: eouipment drain suros.

The capacity of thc drywell floor sump puro is 50 gom,and,the capacity.

of the drywell eru'ip:ient su."p:,pump is, also 50

".p...

Removal of 25. gpn from either of these sumps can be accomplished with; considerable mrpin.

REr.=Xi,!i~S 1.

Nuclear System Leakage Hate Limits (BFl2'SAB, Suosection 4.,10).

3.6.D/4,6,0 Safety end Relief Valv s The safety and relief valves are reouired to b" o,,erable above

. h pres-,

sure

(.1'35 psic),,at which the core spra", sys e=.s is not.des'igned to deliver fu11 flow.

Th pressur reli. system for each un't at th Drowns Ferr" Nuclear P1ant

?as been sized to meet two desi~n, bases.

First,. the total safe y/relief valve caopecity has been estaclished to meet the ovcrpress""c protccticn criteria of the AStZ Cod

Second, tne distriouti'on of this required caoacity between safety valves and r lie". valves has oeen se to meet design basis 4.4.4-1 of subsection 4.4 which sta es that the nuclear system. relief valves shall prevent opening of the safety valves during norcel plant isolations and load, regections, i

L The d tails of the analysis which shows compliance, as modified by Reforence 4,

with thc ASlK 'Code requirem nts is presented in subsection 4.4 of the CESAR end thc Reactor V ssel.Overpressure Protection Su:,mary Technical Report su"m'tted in Amendment 22 in response to question 4,1. dated December '6, 1971 To meet the, safety design basis, thirteen safety-relief valves.have been installed on unit 2 with total, capacity of 84.2% 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 direct scram (valve position scram) results in a maximum vessel pressure which, if a neutron flux scram is assumed.considering one relief valve is inoperable,

-.has adequate margin to the code allowable over-pressure limit of 1375'sig.

'To meet the operational design basis, the total safety-relief capacity of 84.2% of nuclear boiler rated 'has been divid'ed into 70% relief (ll'alves) and 14.2% safety (2 valves).

The analysis of the limiting plant isolation transient is presented in the supplemental reload licensing submittal for the current cycle.

Thi's analysis shows that 10 of ll relief valves limit pressure at the safety valves to a va'lue which is below the setting of the safety valves.

~ 'There-

fore, the safety valves will not open.'his analysis. shows that, peak system pressure is limited po a value which is well 'below.the, allowed vessel overpressure of 1375 psig.

Amendment No. N, Q, 58

~

~

/

219'

', ~

3.6/4.6 BASES:

l t

Experience in relief and safety valve operation shows that

'a testing of 50 'percent of the valves per year iI adequaite i to detect. failures or de'teriorations.

]the relief and safety valves are bienchtested every second operating cycle to eriIsur'e; that their, set. points are, within'he.

+ 1 percent tolerance.

The 'relief valv'es are:tested in pl'ace once,per operating cycle to estaiblish that they will open arI<d pass steam.

The requirements established above apply when the nuclear'ystem can be pressurized above ambient co'nditions.

These requirements are applicable at nuclear system pressures

'.below normal operating pressur'es because abnormal operational transients could possi'bly star<t at these conditilIonls such that eventual overpressuri.'relief would be needed.

However, theses transients are

<much less sev'ere, in terms of p'ressure than those sta<rting at rated conditions.

The valves need not b<e functional when the vessels head is removed, sin<ce the nuclLear system cannot be. pressurized.

REFERENCES 20 3.

4.

Nuc1eir System.Presiure Relief System (BFNP FSAR.Subsection 4.4)~

Amendment 22 in response to AEC 'Ques'tion 4.2 of'ecember 6,

1971".

I "Protection Against Overpressure" (ASME Boil.er and Pre'ssure Vessel

Code, Sect!Lon III, Article. 9)

Browns Ferry Nuclear Plant, Design DefiCie'@cjoy Report-,-Target Rock Safety-Rel!Lef Valves, transm!Ltted by< J>

E. Gilleland to F.

E. Kruesi, August 29, 1973.,

> ~

Generic Reload Pue.'L App]iication, Licensing Topical

'Report, NEDE-24011-P-A,. and Addenda.

3.6.E/4.6.E Jet Pumps Failure of.a )et pump nozzle assembly holddown mechanismnozzle assembly and/or riser,'would increase the cross-sectional flow area for, blowdown following.the design basis double-ended line break.

Also, failure of'he di ffuser would eliminate the capabi.lity,tb deflodd the. core to two-third's heiglit level following a recirculati<on line break.

Therefore, if a failure occurred, repairs'ust be made.

The detection technique is as follows.

With, the 'two recirculation: pumps balanced i'n spied to within;+ 5 percent,

<t:hd flow rdte's i'n boch recircula-tion loops wi,ll be veri fied:.by control rodm 'mohitbring indtruments, l:f the two flow rate valu'es'o not 'differ by more than 10 per'cer<<t, riser and nozzle assembly integrity has been verified.

220 Amendment; No.

58

LIMITING CONDITIONS FOR OPERATION

,SURVEILLANCE REQUIREMENTS

~

3.7 CONTAINMENT SYSTEMS 4.7 CONTAINMENT SYSTEMS H.

2.

With one hydrogen analyzer inoperable, restore at least two hydrogen analyzers to OPERABLE status within 30 days or be in at least HOT SHUTDOWN within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

3.

With no hydrogen analyzer OPERABLE the reactor shall'e in HOT -SHUTDOWN within. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

.Containment Atmos here Monitorin (CAM) Svstem-H>~Anal zer l.

Whenever the reactor is not in cold shutdown,two independent gas analyzer systems shall be operable for monitoring the drywell and the torus.

H.

Containment Atmos here Monitorin (CAM) S stem-H>~Anal zer l.

Each hydrogen analyzer system shall. be demon-strated OPERABLE at least, once per quarter by, performing

a. CHAlKEL CALIBRATION using standard gas. samples. containing a nominal, eight volume percent hydrogen balance nitrogen.

2.

Each hydrogen analyzer system shall be"demonstrated OPERABLE by performing a, CHANNEL FUNCTIONAL TEST monthly.

Amendment No. N, 58 249

~ThB'tE -3.7 h PRLKCRY ~CONThl~ "ISOthTIOi Vhl.VZS t

t3

%PCS oI taA>C.

LLSCSSL Lo.LCdLLOG T

tii LIIVV' V '. ~ S.~ o ~ A Nuaber 'of:Pover NaiUiun

Operating-tedn' oo'Iv>>O

'!loraal too'IIIon "hction 'on ln it'ia'tinj

Sl'SIII

'1ain iteaialine isola't ion valves (FCV 1 14 ~ 26's 37 ~ &5Ijl 15 t 27't 38s

~& 52)

Naih steasmline dra'n.,isolation PCV-1-55

& 1-56 3crc5 A7

'0 GC

'SC Reacto'r Mate'r sanple line isola=

t-'ion valves SC 2E RltRS shutdovn cooling supply

~

o

~

o

=

tviot tl. Ln o lt Aaolaclon vaAves s'L" I s 'Io I lUNS <<LPCL 'to 'ieacto'r 'PCV-74-53, 67 30 SC SC Reactor vessel head spray isola-at

~

~

PS t7 7!. 17

'tR LOI~ VaLVes

~ vo I 1 I ( S (v SC RllRS flulls and drain vent to suppreuoion chafAber FCV-74-10i, 103, -119,

& 120 e.op re>>>>io t'tribe>> Q>>nip PCV-74-57'8 Drys.-ll equlpaent dra'in discharge iuLslstion valves YCV-77-15h,

& 15B SA SC Dfyv'el 1 fLoIir--drain discIIargn-isolatlun valves PCV-77-2h

& 28

In conjunction with the Mark I Containment Short Term Program, a plant unique analysis was performed ("Torus Support System and Attached Piping Analysis for the Browns Ferry Nuclear Plant Units 1, 2; and 3," dated September 9 ~ 1976 and supplemented October 12,, 1976) which demonstrated

a. factor of safety of at least two for the weakest element in the suppression chamber support system and attached'iping.

.The maintenance of a drywell-suppression chamber differen-tial p'ressure of 1.3. psid and a suppression, chamber water level corresponding to a.downcomer submergence range;of'.0 feet to 4.60 feet will'ssure the integrity of the suppression chamber when sub]ected to post-LOCA suppressi'on pool'hydrodynamic forces.

~Inert&

I The relatively small containmentlvolume inherent in the GE-BWR,pressure suppres-tion containment

.and the large amount of zirconium. in the core are such that the occurrence of a very limited (a percent or so) reaction of the zirconium

'and steam during a loss-of-coolant. accident could lead to the liberation of hydrogen combined with an air atmosphere to result in a flammable concentration in the containment.

If a sufficient amount of hydrogen is generated

.and oxygen is available in stoichiometric quantities the subsequent

'ignition of the hydrogen in rapid recombination rate could lead to failure of the containment to maintain a low leakage integrity.'he

<4X hydrogen concentration minimizes the.possibility of hydrogen combustion following;,a loss-of-coolant,.accident.

Amendment No.,

42 269 Q

8, 878

BASES iO

'The.'occurrence of.primaxy system leakage following a major. refueling o6taige

.or.'other sche'duled shutdiown is much nIore probable, then the occurrence

of'.the.1'oss-of-coo]Lant aiccident upon which the specified oxygen concentration limit.is based.

Permitt'.'ing access to the drywell for leak inspections, duiing:a startup is judged prudent in terms of the added pl'ant safety

offered without 'significantly'reducing the margin;of, safety.

Thus, to

.'pr'eclude;the *possibility o,f starting the reactor and, operating fox extended,

,,;per'iods 'of 'time with s'ignificant leaks in,the primary system, leak inspections;

are:scheduled during s'tartup periods,,

when the primary system is at or near

'rated operating temperature and pressure.

The,24-,hour peiiod to provide

, inerting is judged to be 'suffic:Lent to,'perform the leak inspection.,and B

'establish 'the 'requ:Lx'ed oxygen concentration.

'To 'ensure that the hydrogen concentration is: miintained less than 4% 'following 'an 'accident; liquid nitrogen~is~ maintained on-site for

'conta'inment

.atmosphere diluti'on.

About 2260 gallons would be

sufficient:as

'a 7-. day supplys-and repleniEhment facilities can

.deliver liquid nitrogen to the site within.one day; therefoxe,

.a"requirement.of

.2500 gallonis is. conservative.

Follow Lng a loss

.of 'coolant -a'ccident the -Containment Air Monitoring (CAN)'ystem

continuously'onito'rs the'hydrogen concentration of the containment volume.

'Two.iniiependent systems

( a syst'e m consists of one hydrogen

sensing.circuit) are installed in the drywell and the torus.,

',Each sensor

arid associated, circuit is periodically checked by a calibration gas to verify operation.

Failure'f one system does not reiduce the ability to monitor

'system atmosphere as a second indeperident and redundant system willstill be.operable.

Xn:terms.of separabi!Lity, redundancy for a fai]Lure of the torus

.system is based, upon at least one operable drywell systetn.

The

,drywell '.hydrogen conce'ntration can be used to.'Limit the torus hydrogen

,concentration during post LOCA conditions.

Post LOCA calculations show that.the 'CAD system initiated within two-hours, at a flow rate of 100 scfm will limLt the peak. drywell and wetwell hydrogen con-

centration to 3.6% (at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) and 3.8X (at

32 hours3.703704e-4 days <br />0.00889 hours <br />5.291005e-5 weeks <br />1.2176e-5 months <br />), respectively.;

This is based upon purge initiation after 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> at,a flow rate of 100 scfm to maintain containiinent pressure below 30 psig.

Thus, peak

,torus.hydrogeri concentration cain be controlled below 4.0 pex'cent using

either 'the direct torus hydrogen monitoring system or the drywell hydrogen monitoring,system with appropriate conservatism

(~ 3.8%),

,as a guide for CAD/Purge operations.

270i Amendment 'No. Pg,:58

4.

Daily tests

.of.innunciatiori lights and audible devices. are pertcrmed a=-

a rnutir e nperat'glori functior>.

,: h+ CO, syst..m manufactur'er recommends semiannual tes in@ of CO: system fi r > d=t ection circuits.

Figure

6. 3-1 d scribes the in-plant fire,protection organization including the roving fire watch. 'n addition, other operating personnel periodica3;ly inspect. the plant. duri'ng their normal operatinq activities for fire hazards and ot'her abnormal'ond'itions.

.Smoke detectors wi}l be testeo "in-place" using inert freon gas applied by a pvro.=onics type applicator which is accepted-throughout the indust ial fire protec

.ion indu try for testing prOduCt=

O.

COmbuStiOn deteCtOrS Or by uSe.Of,th= rlSA Chemioal smoke generators.

At the present time the manuf acturers have only approved the use of "punk" for creating smoke.

TV@ will not us "pun'k" for testing smok'e detectors.

329

S HhJOR'CS trial FEATURES S iTC, FL'ATORCS'rans Fer'ry unit,'

i's locat;ed aQ Bpovlne lFepry Nuclear Plant iite or'> property oMnid by thr United States and in custody of the. TVA.

'Thc site slliall consist qf ypproxima,tc)y,B40 acres on the nonth shore of; 'Wheeler Lake at Tennessee River Rile 294 in-Limcstonle County, Alabama.

The l>>inimym distance from the outside of'he se'condary containment building to the

- boundiry of the: e>cclusion area as.,definpd

~in 10, CFR 100.3

.shall be 4,000 feet.

2 5;3 REACTC)R A.

The reactlor core may contain 764. fuel assemb3.ics consisting, of 7x7 assemblies having 49 fuel, rods each, Sxg assemb3.ies having 63 fuel rods each, and Sx8R (and PSxSR) assemblies having 62 ftlel rods eacll.

B.

The reactor cor'e shall contain 185'ruciform-shaped contr~ol~

rods.

The control material shall be boron carbide powder (B4C) compacted to approximately, 70 peicent of theoretical density.

REACSOII VESSEL

'The reactor vessel shall be as described in Table 4.7-'2 ot thc FSAR.

Thle applicable design coded shall'ie as dcs'cribed =in Tabl'e 4.2-1 of thc FSAR.

5. 4 COHTAltaEiNT A.

The p'rinc'ipal design parameters for thc primary containment, shall bc 'as given in Table 5.2-1 of the FSAR.

The, applicable design codes shall be as described in,Section-5.2 of the FSAR.i B., The secondary containmcnt shall be as described in Section 5;3 of the FSAR..

C.

Pci>etratlons to the primary containment and piping passing through sich penetrat fons shall be designed in accordance vith the standards set forth in Sec'tion 5.2'.3.4 of the FSARl S, S FUEL S'IIOILlil:E A.

The arrangement of fuel in the nleu-.fuel storage facility shall be such that k, for dry co'nditions, is less 4han IFf ~

0.90 and flooded's fess than 0.95 '(Section 10.2 of FSAR).

Amendment No;, P$ '; QE 68