ML17194A406

Proposed Tech Specs Re Reload Fuel & Core Design,Transients & Accident Analyses & LOCA Analysis

ML17194A406
Person / Time
Site:	Dresden
Issue date:	01/11/1982
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML17194A407	List: ML17194A406 ML17252A552 ML20039G953 ML20039G956
References
NUDOCS 8201190338
Download: ML17194A406 (47)

Text

Attachment 6 Dresden Station Unit 3 DPR - 25 Proposed Technical Specifications Revised* Pages Previous A.7::.

1 12 2

53 5

42 6

52

• 6A 7

52 10 42 11 42

• llA 13

_42 14 42 15 42

• 15A 16 42 18 42 19 42 20 42 21 42 22 42

• 22A 34 42 36 Original

• 36A 42 52 42A 42 46 42 58 17 62 42

~

_i_ *

. '*. 8201190.338 82011"1" :;:;J:=>J !.. ~

• PD~ ~DOCK. 05000249

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<" *--~-.......... * *.

c, '" PDR*L* :

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(".

• .~ -- :....

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No.

2 -

Revised* Pages Previous Amm. No.

62A 42 63 42 64 17 65 Changes 27 and 18 78 40 81B 42 81B-l 42 81C-l 42 81C-2 42 81C-4 42 81D 34 81E 22

• 81£-1 82 42 BSA 42' e

85B 42

• 85B-l 86A 42

• Deriotes a new page;

1.0 Dr:_f

.TIONS

)

The succeeding frequently ueed terms are ex-plicitly defined so that a Wliform interpretation of the specificatione may be achieved.

A.

(Deleted)

B.

Alteration of the Reactor Core - The act of moving any component in the region above the core support plata; below the upper grid and within the shroud.

Normal control rod move-ment with the control rod drive hydraulic ayetea ta Dot defined ae a core alteration.

c.

Critica! Power -~~0-~~) - Tlw c*ritical power ratio is the 1*at iu of that as~embh*.

power which causes some point in the assembly to expe1*icm:e transition boiling to the as,.embly power at the reactor con.Ii t ion of int crest as ca lcula~U:~

application of the Xl~-:-1 correlation., -

(Reference XN-Nl*'- 1)1:.,)

0.

Hot Standby - Hot standby means operation with the reactor critical, system pressure less than 600 psig, and the main steam isolation valves closed.

E.

Immediate - Immediate means that the required act1on will be initiated as soon as practicable considering the safe operation of the unit and the importance of the required action.

F.

Instrument Calibration - An instnunent cali-bration means the adjustment of an instrument signal output so that it corresponds, within ac-ceptable range, and accuracy, to a known value(s) of the parameter which the instrument monitors.

Calibration shall encompass the entire instrument including actuation, alarm, or trip.

Response time is not part of the routine instrument calibration, but will be checked once per cycle.

G.

Instrument Functional Test - An instnunent funct ionai test means the injection of a simu-lated signal into the instrument primary sensor to verify the proper instrument response alarm, and/or initiating action.

H.

Instrument Check - An instrument check is qualitative determination of acceptable oper-ahil ity by observation of instTW11ent behavior during operation.

This detennination shall inc luJe, where possi hie, comparison of the instrument with other independent instruments measuring the same variable.

Aaend*ent No.

. nm-1'1 I.

J.

.~

~l~l~*l!'U~"~! ~!"~- !!"_ 0p1_=r~1 ~nn_Jl.r.nl - 1'**

hai, *"0: ccmdll 11111s tur 0111:1.ll 11111 s1*cc1 fr lhe

...,... accr:l'l*l*lo levch ol

• t*~lt.* 1*e1t111*-

-* Rece111uy 10 assul"e 1;.fc sl.ul*tt* onJ ut**

el"allo11 of lhe f*clU1y.

ll"hcn 1lu:so cn11oli1*nn1

• • • -*. the rla11t c.a" be.., *.,.. atcd s;afr:ly and

.._,***I 11111.11 Ions c.an ho safely contrnl led.

ll*l*l*g_S~fc*r_~r~~r:~ ~***tln1_l~_~c;_c;). - 11.e ll*illnc Sa[~ty tytlr:* llettl~tll aro )Ctllllf!I On llhl "*cnl*t '"""""'ch 1111t11110 tho ;1111011.:t le peulcct he act Ion at

• lcv.,I s11ch that th* safe*r l!*lls ul 11 nut ho eacc.:1lt;1I.

lhe u:clon Q.

.. twee* tho 1efe11 ll*lt and thc:so '"" tn1s

~

.. *~u:nts....,... wllh 11011ul nrerallon.,....

..,_ lhc:to se11i"I'*

lhe *;n1:tn has been

• t..a.lhhed 10 that with 1*1"01*cl" ope... llon of ***

l1HlnS1cn1a1lon tho safety U*ll* wlU never be ea<e.:JcJ.

K. fi'rac ti on For le r.ca ec y C:E,

• 1e rac

• on of limiting power tlensity ts thr. rat.i.o of the Lin1~ar Heat Generatitin Hate (LIIGn) existing at a given location to the design LHGH !\\1r tltat bundle type. F'LPD does not apply to ENC fuel.

L.

tAil<_~:f~~"~-"~'~~unT~~ -

A loalc,,,.

i.r.runct i.. u.al test *c"'" " 1cs1 of :.I I rela11

-* (011t11cts nr II Inch: cil"Cllll rro* U:n9or

• • eel lwalc*I Jc:vlco to lniuro al I cosi1ionc:nt1

• r* orcr11lelo 1*c:1" Jc,! \\:*1 l11tcnl.

N11eu1.1*01tl*

• • • , 11c1 Inn "i 11 ro 10 cm*plct*lon, L*., p*-r*

-..111a b* uarlr:d 11nJ 11.ah.:s *'11cn.:J.

,,,..,.,..:1 r

• • ~- f'.ril.....,,......... ll;.t_i_~.:~!~..") - 11'*

1;;r.;,-;;1i\\-

CorCi--ci-i i 1 c-.:;*,-1..... ~..,.1 t To

s.

c*ncsl'on*Un1 to 1ho *ost I l1Jll ln1 fual a*sc:.a.1, l* tl1e core.

11. llode - l1'e reactor"°* 11 that 11hlr.h ls eii.il.*........., ttle e:>Ju-seloctor-1vllch.

!:'r!!~blo -

• ar*l-. **b*J!tl...., l***I '* c..,....."'* or ***Ice

"'**II "" 81'e**ehl*.. 1oc:n It lo c*fM'bl*.. r P*rfe*wh.. Ile

• 1*1:1:lfle" l\\tncllon(!I).

l"'t*lldt In lhl* d*flnltlun *hall he the **.. **f'l I on \\h*t "'I nu1... 11r.** r **lenci11ul h11:l*-*n-

.. al11.n, c-trnl*, nonoonl or.ti *-!"l!""'*.Cf *l*ct1*lc*I povor 1101u-co**. Efl"lhlfl or """' ";\\lor, hohrlc*llon or other M**l I l**'J *.,..1.,ocnt th*l **** rC:IJ'tlrr.*I for. lh* *rat...

"""'"'**** ti*atn, c*.....,.nenl or dftvlr. 1 lv perir..1,. ll11 l\\*ncllno(*) *ro *1110 c*1**hl* ol.,orlo,1*lne u.. ar r*lot..

..,.,port l\\lnc\\lon(al fl1**r10\\ln.11 - C'r****t.l"I "e*n* that * ***t...., *11t:1ratea.

l* 9in:c~nl *r ""*tc* I* rerfo""'"" lle lnlr.nolu*I f\\loc\\lon* Jn It* n1*11*l.-.1d -*1or.

!"'r~~.!.~"I ~tel! *lnlc:rHI bc:twee* tho e..

.,:* one u:T11elln1 onl*10 an*I eho end of *h*

ne*t suh1c11ucnl refuol 1111 o~ta1*.

Pl"l*~rr Cnnt*lr.ncnl l11te5rl*t - Prl**rr con"t"ilnl.;;-nnn.e1-..-fli*cans *"** lh* drywall and rrr.tt*1ro..... r.. 01slnn ch.....,...,. lnlat:t and nil of lh* followln1 con1lltlon1 are 1atldl.. 1 I. All aanual cont*h*enl holal Ion valwe1 **

I i11e1 conned Ina lo th* re*ctor coolant *JI*

t~* or containment w~lch are nol refl'llr**

lo bu open durln1 accld**t con*ltlons are t:losc*I.

J.

At lcnst one door In eech *lrlncl It closed a111I SCIO h:d.

J, All 11utu.atlc contaln*ent hohilo* walwes en: 01*cr.1ltle or dcact hated In lhe holot..

r"sh Ion.

4.

All lillnd Uu1** and *11nv*r* ere cles...

Protertlvo lnHni;a.1flt*llon lleflnltlons I.

lnttl"Ul8t;nl 'h*~*el - An ln,lrUlllcnt ch0111-ncl **,..n, an.,,.n1c*cnt or

• tensor.....

a111lli;ory e:tui~r:nt 1"cc1uhed lo aenorat*

anJ 1rnn1*ll t~ *trip *11tc*

• tln1l* trip slcn*I related *o tho rtanl rara*cter J

• onhorcd by tkl~t lnsll"**enl channel.

---

~---------------------:!11-----

1.1 SAFETY Lttirr

,.i

,.*,-I l\\mendment Ho.

2.1 LI'1UINC Sl..rrl'T SYST~~. SEiI'l.~C

1.

APRM Flux Sera"' Trlp Settlns {Ren "odel

\\."hen the nutor 11odP. avltch ls In the run rosltl~n, the APRl1 fluK scra* 1ettlng shall be'

~

s ~ [.ssw0 + 62]

vlth a ma.Kl~u* set polnt of 120% for cor~

flov equal to 98. JC 106 lb/hi' and greater, Wh8r-!I S

• set.Ung 1n per cent. or rated power V 0* por cent. or drive flow requhc4 to produce a rated core f lov of 98 "lb/hr.

In the event of operation of any fuel assembly fabricated by GE with a maximum

. fraction of limiting power density (MFLPD) p,reatcr than the fraction of rated power (FRI'), the setting shall be modified as follows:

Where; S ~ (.58W0 + 62)

FRP = fraction of rated (2527) MWt) ff'RP

]

LMFLPD thermal power MfLPU = maximum fraction of limiting power density for GE fuel The ratio of FRl'/MFLPD shall be

~*.*'

equal to l. 0 unl(~Ss the actual operating value is less than 1.0, in which case the actual operating value will. be used.

I I

1.1 Safety Limit Bases FUEL CLADDING INTEGRITY The fuel cladding integrity limit is set such that no calculated fuel dam-ages would occur as a result of an abnormal operational transient.

Be-cause fuel damage is not directly observable, a step-back approach is used to establish a Safety Limit such that the minimum criLical power ratio (MCPR) is no less than the MCPR fuel cladding integrity safety limit.

MCPR) the MCPR fuel cladding integrity safety limit represents a conservative margin relative to the conditions required to maintain fuel cladding integrity by assuring that the rucl does not experience transition boiling.

The fuel cladding is one of the physical barriers which separate radioactive materials from the env(rons.

The integrity of this cladding barrier is related to its relative freedom from perforations or cracking.

Although some corrosions or use related cracking may occur Juring the life of the cladding, fission product migration from this source is incrementally cumulative*

and continuously meas~rable.

Fuel cladding perforations, however, can result from thermal stresses which.

occur from reactor operation sip,ni-*

ficantly above design conditions and the prot-t:tion syst:l:*m safety settir)~S.

While J:.tD+RJl product mi~n1Lion frum cladding perforation is just as measurable as that from use related cracking, the thermally caused c 1 acl Jing 1w r. for a 1* ion signals a threshold, hl'ynnd which still greater thermal stresses may cause gross rather than incremental c Laddi~g dd.l!L*-

ioration.

Therefore, the ful'l ~ladding AmPnrlmPnf-Nn.

A.

Safety Limit is defined with margin to the conditions which would produce onset of trans-ition boiling, (MCPR of 1.0.)

These conditions represent a significant departure from the condition intended by desi~n for planned operation.

The MCPR fuel cladding integrity Si:lfety Limit assures that during hormal 01wration and during anticipated operational occurrences, at least 99.97. of the fuel rods in the core do not experience transition boiling.

See reference XN-NF-524.

Reactor Pressure > 800 psig and Core flow > 10% of Rated Onset of transition boiling results in a decrease in heat transfer from the clad and, there*fore, elevated clad temperature a~J the possibility of clad failure.

However, the existence of critical power, or boiling L~ansition, is not a directly observable parameter in an operating reactor.

Therefore, the margin to boiling transition is calculated

.from plant operating parame.ters such as core puwer, core flow, feedwater temperature, and coie power distribution.

The margin for each fuel assembly is characterized by the critical power ratio (CPR) which is the ratio of the b1111dle power which would produce onset of boiling divided by the actual bundle power.

The minimum value of this ratio for any bundle in the core is the minimum critical pnwer ratio (MCPR).

It is assumed that thl' plant operation is controlled to the nominal protective setpoints via the instrtimented variables.

(Figure 2.1-3).

The MCPR Fuel Cladding Integrity Safety I Limit assures sufficient *conservatism in the operating MCl'R limit that in Lhe event of an anticipated OfH'rat ion al occurrence from the limiting condition for operation, at 10

~,

I I

OP!\\-25

. 1.1 °S.Vt;TY LTI-11T J. C?re Th*~Mn1'1ovrr Lf*lt (R~actor.

'!.!.!.H*u~ l_eco e!iigJ

\\lh*n the rtactor rressurc is < 800 p!I i & o: core flow h hu th~ 101.

of rated, the core ther~al pover shall not e>:ued 25 perc.ent of rete4 th*:**l rower.

l. The neutron flux shell not uu:eed the scnai s~tttng tst*bllshed lh Sptc1f1c~tion. 2.1.A for lo"gcr than l. 5 eeconds as 1.ndlcattd by the proc*** co"'Puter.

2. '""*n the proc*ss coep~ter *i* out of set"Vlce, this 11fet1 11.. u shell be 8.S:J\\#M41 to b@

e*cteded l{ tht neutron flUI( ~*cttd..-: the ~or:un setting ~stabllshtd by Speclflcetlon 2.1.A and

• control rod set~* dots not o~cur.

.....,....... ~ '

(

)

D.

A*~et1ij!e,t.*r tevelShu\\d~~n Conditlo~

"'°'""ev.r thf! uactor h In the shutdcvn condltlon with lrradlAteJ 1~@1 In the rtector v*s~el, the

~At*r Jtv*l ~hall not be Its~ then that corrt~

ronAl"ft to 12 lnrhPS ~bove the top of the Active fu~a*~htn lt is S~At~d ln the cor~.

• Top of a~t,ve fuel is defined to be JfiO t nch1*s above vesse 1 zero (see l l*t"l"l.l '))

'I ~'

.) * (.

e

.~.,~

. ~

~

2.1 Lll'ilTI:iC Slt-&.rt STSTE!t smn-c

• ~

The U" flu* *ct*"' ISt!ttln!: shall be set et hu th*n or equal to 120/115 of full scale

• B. !!.!'" Ito:! !lock Set.Ung The APRfot rod b loc1' H ttlnc ""*11 1u1:

I S ~ [.58WD + 50]

. The de flnl tlons used *bove for 'the APPJt sen* tr1P1?ply.

ln the event of operation of any fuel assembly fabricated by GE with a maximum fraction limit in1~ power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:

l S ~ (.58W0 + 50) ~~~p~

The dt' f ini t ions used above for the APRM scram trip apply.

The ratio of FRP to MFLPD shall be set equal to l.O unless the actual operating value is less than 1.0.

Jn which case the actual operating value will be used.

The adjust-ntl'nt may al.so be performed by increasing 1 the Al'RM gain by the inverse ratio, MFLPD/

fRP, which accomplishes the same degree of protection as redu<:ing the trip setting by l."IJ I> /Mi;'f IJf) 7

1.1 SAFETY LIMIT

,~

Amendment No.

2.1 LIMITING SAFETY SYSTEM SETTING This adjustment may also be performed by increasing the APRM ~ain by the inverse ratio, MfLPD/FRPJ which accomplishes the same degree of protection as reducing the trip setting by FRP/MFLPU:-

2.. APRM Flux Scram Trip Setting (Refuel or Startup and Hot Standby Mode)

When the reactor mode switch is in the refuel startup/hot standby position, the APHM scram shall be set at less than or equal to 15% of rated neutron flux.

6A

Safety Limit Bases

1. 1. A Reactor Pressure> 800 ps ig and Core Flow.> 107. of Rated.

(cont'd) least 99.97. of the fuel rods in the core would be ~xpected to avoid boiling transition.

The margin between calculated boiling transition (MCl'R=l.00) and the.MCPR Fuel Cladding Integrity Safety Limit is based on a detailed statistical procedure which considers the uncertainties in monitoring the core operating state.

One specific uncertainty included in the safety limit is the uncertainty inherent in the XN-3 critical power correlation.

Refer to XN-NF-524 for the methodolpgy used in determining the MCPR Fu~l Cladding Integrity Safety Limit.

The XN-3 critical power correlation is based on a significant body of.

practical test data, providing a high degree of assurance that the critical power as evaluated by the correlation is within a small per-centage of the actual critical power being est ima.ted.

The assumed**

reactor conditions used in defining the safety limit introduce cons*:. v-at ism into the limit because boundingly high radia 1 power p;eaking factors and boundingly flat local peaking distributions are used to estimat~ the number of rods in boi~

.. tll'.'ansition.

Still further con~'V.Jrtsm is induced by the tendency of the XN-3 correlation to overpredict the n11mber of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 co~rclatiori provide a reasonable degrc~ of assurance that.during s~slained

• operation at the MCPR Fuel Cladding.

.e Integrity Safety Limit there Would be no trans-it ion boiling in the core.

If boiling transition were to occur, however, there is reasbn to believe that the integrity of the fuel would not necessarily be compromised.

Stgnificant test data accumulated by the U. S. Nuclear Hegu latory Connniss ion and private on~anizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very ~onservative approach; much of the data indicates that LWR fuel can survive fur an extended period in an environment of transition boiling.

If the reactor pressure should ever exceed the limit of applicability of the XN-3 critical power correlation as defined in XN-NF-512, it would be assumed that the MCPR Fuel Cladding Integrity Safety Limit had been violated.

This applicability pressure limit is higher than the pressure safety limit sp~cified in Specification 1.2.

For fuel fabricated by General Electric Company, or~~ration is further constrained to a maximum linear heat generation rate (LHGR) of 13.4 kW/ft by Specification 3.5.J.

This constraint is established to provide adequate safety margin to li. plastic strain for abnormal operational transients initiated from high power conditions.

Specification 2.1.A.l provides for equivalent safety niargin for. transients initiated from lower power conditions by adiusting the APRM flow-biased scram by the ratio of FRP/HFLPD.

Specification 3.5.J establishes the maximum value of l.HGR wh i.ch cannot be exceeded during steady power operation for GE fuel types.

For fuc*l fabricated by Exxon Nuclear Company, (ENC) fuel design cl'iteria have been established t(l provide protection against fuel centerline melting and cladding strain, ENC has performed 11

'!:iafety Limi.t Base_~

1.1.A.

Reactor Pressure> 800 psig and Core Fuel) 107. of Rated.

(cont'd) fuel design analysis which demonstrate that cen tPr 1 ine me 1 ting i.s nnl pre dieted to occur <luring transient overpower conditions throughout the life of the fuel *. Protection of the MCl'R ;md MAPLHGR limits and operation within the power distribution assumptions of the fuel design analysis will provide adequate protection against. centerlinl' melt anJ ensures compliance with ENC's clad overstrain criteria for steady state and transient operation.

Since ENC' s design criteria are more. coilservative than the 1% plastic strain limitation on GE fuel, the LHGR limitation and Al'RM scram adjustment for GE fuel established in specifications 3. 5.J and 2. I.A.. I.

respectively are unnecessary for the protect ion of ENC fuel.

The proce<lura 1 controls of spcci.fication J.l.*H wi.ll ensure that ope1-ation of ENC f11el remains wi.thin the power distribution assumptions of the fuel design analysis.

~*--f"I'~~

lla

1.1

~~r,.tv t'imtt n~~r'.1 1.1.C ro~cr Tran3lcnt (cont'd)

Th~ co~putcr provtdcd han e

~e~ucnc~ <:nnunc~nt1on pro~ram',,;hich

~ill tndtcetc the ncqucnce tn "'hich ocr~ms occur ouc~ e3 neutron flu~,

p~~:;surc, etc.

Thi~ pro~rC!m al:;o 1~*Jlcct~3 *,:hen th~ scr~m s~tpo1nt 13 cl~~r~d. This "'!11 prov\\1c lnfor~:it1on on ho1*; long a scr;.1:n coml1t1on <'llt:;l.s

~nd thu'.; provJdc ~omc rr.c:i::turc of lite cn~r::y adtll"d dt.:rtn~ a tr;!n:;t':!nt.

1'!1u3, r.-::>o..01~tcr information norm:illy "'*Ill be nv~tl~ble for. 2n~lyzt~~ ~cram3; ho~

cvc:-, 1f the cor1pulcr Jnfor:n1tton :JhouUJ

~ot he ev:ltlnble ror ~ny scram an~lyot~.

!:ipcc1!"1c;!t1on l.1.C.2... 111 be rcll<*tl on t-::>

tl~tcrr.ilnc tf ii oafcty l1m1t h;n been v:ol:?tcd.

DurJnc:r; periods when the rcoctor J:J shut clo.,.n, constclcre?tlon must also be r,tvrn t'l *..;.:,tcr l~vel :-r 11itr~nc:1t!l tluc to ~he c!'fe'.:t or clcc~j' hc:!t.

]f renctor "'2tcr l~v~l 3hould tlro~ bclo"'

l~c top of th~

<!C~~VC fu~l ourl1~~ th13 t~r.e, lh~

~b:llty to cool lhc core 13 rctluced.

1':-1i5 reduction tn core cooJ1n~ cap-

~~illtJ coul1 lc:~d to clev."!t~rl cbd1!1np; t~~.C'"!r.Jtu:-c~ nmJ cl.~d p~rror~tlon. 'fhr.

co:-~ ;;lll be coolr1J :Juff1ct~11tly to prr.-

,.~!l.(.J);;d ~cltln~ ~ho11ld the \\o;<!:.r.r lcvi;l b~t*.!'dtl!~cd to t:*;o-thlrd:1 the er.re hr.tr.ht.

!"::;tL":.>li5h!T".~nt ~r :.~w :i;ifct.1' llr.ill ;lt l~

Lich~~ ~bov~ l~l! tn;l 0f the !°'l!c 1* p1-0*1 ~des

... ilc*!ll.1tc n:... ::-~lri.

Thl3 lr.vcl 1.;!ll lw con-t.l*n;o1:"lly. mo:;tto!.*c*.! 1*.hcnn.vcr LI~*~ r*:c1t*-

c11l2L~o:\\ pu~~p:; :ir~ not o;>crfltt::*_;;.

• • Top of ilct. i ve ru~ l is dr fined to hr

)f10 inr.h~s above vr!s~~el zero (r,Pe -

11.iscs J.21.

2.1 L1mtttns Sof~ty Sl_~tcm Sctttng 8~~c 3 FUF.L CLADDING INTEGRITY The.cbnormal operat~onal transtcnts nppllcable to OP"!r~tton of the Un1ts h~ve been 2n2lyz~d lhrou~ho~~ the opcctrum or plonned oper~ttn~ con-d l t lons up to the r~ted thcrm1l power contl 1 t lon of ~r;-**7 l*:Wt.

ln odd 1t ten 2'j~'/ i*i'Jt 1s the 11cc113cd 1na:t!:num st~ad,..

st~te po~er level or the unlts.

This m~xlmum st~ady-st~te po~cr level ~tll never kno~ln~ly be cxc~cdcd. See referencel XN-NF-79-71.

lJ Amendment No.

Limiting Safety System Setting Kascs 2.1 FUEL CLADDING INTEGRITY (cont'd)

- Conservatism is incorporated into the transient analyses which define the MCPR operating limits.

Variables ~hich inherently possess little or no uncertainty or whose uncertainty has little or no effect on the outcome of th~ limiting transient are selected at bounding values.

Variables which possess significant uncertainty that may have undesirable effects on thermal margins are addressed statistically.

Statistical methods used in the transient analyses are described in XN-NF-81-22.

The MCPR operating limits are established such that the occurrence of the limiting transient will not result in the violation of the MCPR Fuel Cladding Integrity Safety Limit in at least 957. of the random statistical combinations of uncertainties.

In Reneral, the variables with the greatest statistical significance to the-consequences of antici.pated operational occurrences are the reactivity feedback associated with the formation and removal of coolant voids and the timing of the control rod scram.

Steod~-~tote operation wtthout ~orccd r~.uletton wtll not be pcrm.1ttr.tJ, e*x~cPF~aur ln~ D tart up test tr.r,.

The on~ly~13 to support O?Cr~tlon 3t various poKcr 2nd flo~ reJrtlon~hlp~

ha:J considered opcr~tton ~Ith c1th~r one or two rec1rculat1on puw.p~.

Th~ b~3es for tndlvlducl trtp ~~tt1n~s ar~ dtscu3scd 1n the followln~ para-grC>ph:J.

A.

1.

For an3lyses of the the1111al consequences of the tronslcnts, tho MCPR's stated *n p3r*ir3ph J.s.i,, 5 th~ ) imit inq cnnf1lt1011 or opr.rt'ltlon l

und those which are consP.rvatlveJy asnumed

~~ exist prior to initiation of the transients.

Neut*ron Flux Tr1p Sett 1nsr:s

~

l\\PRM Flux Scrun1 Tr1p Setttns_JRun Mode)

The ever2ge power r3n~c mon1tor1ng (APHM) system, whtch ls calibrated ustng hc~t b~lancc d3t~ t?ken durln!

9t:!ady-stutc conrl1t1o:-i:J, reads in pcrcc:lt or r3tcc.J. thcr~nl power.

8c-cau3c ftes1on ch~~b~r3 pro~:jc the baslo tnput ~ t~n:? l:J, t~~ /,Pni*l s1st~::1 rC'sponds

• d t rec tly to ~\\*cr~~.c :ieutron flu~.

Dur!n,,. trnn3lc:1t3, the ln3ta:i\\.aneous ratr.* ~f heot tr;;n:;fcr fror.t the ruel

( rcac tor thcr;11a l po\\*;cr) 13 lc3:J than the ln~tnntnncou3 n~utron flux due to the ttm:? con!itnnt or the rucl. 'there-fore, durtn~ nbnor~nl op~~t'ltlonnl trnns tents. t.hc therm:> 1 po~cr' or the fuel ~111 be le3~ thJn th3t lndSc~tcd by the neutron flux at thr. scrzm :ietttng.

An~ly~c~ d~mo~~tratc that. wlth a 120 pcrc~nt 3~ram tr!~ 5ct~ln~. non~ or the nhnorm~l opcr~tlon~l tr~n~tcnt~1nn~lyzcd vjoktc th~ fu<:?l

~;Jfct7 f,tr!~t :'lfld u~~re

,~ (1

~ub:;t<1nttal n1arr,tn r1*on rll'*l -!:-n:'!:;C?.

'J'hcrc fClrr.,

t.hc u~c of rJow

-.~ft*i**:nc-::-d

tc!"."!:1: i;.1*J1* pro\\".ldc:: C\\'~r. nd*J!t.l!'l:1-:l ~::r~:n.

14 Amendment No.

11.A. Neutron Flux Trip Settings

1. APRM Flux Scram Trip Setting (Run Mode) cont'd)

An increase in the APRM scram trip setting woul~ decrease the ma~gin present before the.f':1el.cladd1ng integrity Safety Lrnnt is reached.

The APRM scram tri.p setting was determined by an analysis of mar.gin~

required to provide a reasonable range for maneuvering during operation~

Reducing this operating margin.would increase the frequency of spurious scrams which have an adverse effect on reactor safety because of the resulting thermal stresses.

Thus, the APRM scram trip setting was selected because it provides adeCJuate margin for the f uc.: 1 cladding in ~egr it y Safety Limit yet allows operating margin that reduces the nossibility of unnecessary scrams.

The scram trip setting must be adjusted to ensure that the. LllGR transient peak f,,, G.E. fuel is not increased fur any combin-ation of Maximum Frac t i*nn of Limiting Power Density (MFLPD) and reactor core thermal power, The-k'rjam set ting is adj us te~i in

~i5'.i!.~ce with the forn11rla 1n specification 2.1.A. l. when the.

MFLPD is greater than the fraction of rated power ( FIU').

The adjustment m;-1y also be accornplishl'd by increasinp, the Al'RM gain by the reciprocal of FRl'/MFl..PD.

This provides the same degree of protection as reducing the trip set~i~g by FRP/

MFl.PD by raising the 1n1tial Al'RM rcac,tlng closer to the trip settinr such that a scram would be receiv~J at the same point in a Lr;insi.ent as if the trip setting had h~'1*n reduced.

APP.~ ttui 9ota~ Trsp Settln!

jf1.. !'u~l.al" Stl'r~ & Hot S~t~1db7 Mode) t&f' 81'f1Jl'tUon th the *tnrtup 1110de whtle lhf re1ator ti et low ~re1,urt, the APRM t11:f'1111 eettlng or 15 perc~nt or rated power pro~t~~* lda~uatt ther~3l Mnr~tn between the

~h' 9~~po1nt er.d the 1ntctr 11~1t, 25 p!r*

nut ot r:1 tcd.

Tho margin h ede-1uetc to

• e~~!lt'Tl,dote enttctpotcd -rn~neuvcrs *oococtated

~~~h "o';~f'.,hnt 1tortup. £rreots or tn-i~c~'sn~ pr~ouur* et 1c~o or low *voSd con-

~a: ttr-c 1dtaot*, eeld ""t~* rror.t 10.-rce~

,..,.*,:lrt,!1 d*::-!~!~ :;urtup 1:1 riot much _colder than that ~lready in the svstem, tempera-ture coett1ctent!r-"Bre emall, end oon-t.rol rod patterns 11:-e constrained to be.uniform by O?cret1r.~ procedures b~ck~d up by the ro~ ~orth mtnl~lzer.~

or ~11 possible so~rcc~ or rencttvtty input, unsro~m control rod wtthdrawal ts the most probabl~ cause or a1gn1fl**

c~nt po~er r1sc.

E~c~u3e the tlvx dS3trtbutlon es3oc:£:~d ~!th u~irorm rod ~lth1rr~~13 tio~s n~t Involve hl~h local pcak3, end b~c~use several rods must be "-Oved to ch~n~~ ?owrr bJ 1 nt~nlrtcnnt pe~cen~~;c or rated po~er, th~ :r:?t'! or po;;~r !"~~c 19 very 110~.

Gencr.i lly, the he:>;,; C lux ls 1n *ncitr cc;~lllb~lum *'!th tt:e fl:Jston rate. In an es:>u:r.cd u:i l l.>r::t rod,,. tthdrDtt!Jl ap..

p;-03ch to the :Sere::: leY'!l, the rate Ot P?~~r rt~c 1s no ~~~~ th~n 5 percent or r~tc~ power per ~!nute, an~ th~

A?rl!*i :;J:!te~.-ould lie more th3n Ddt'quate 15 Amendment

1.1.A.

Neutron Flux Trip Setting

2.

APRM Flux Scram Trip Sett in~_ ( R1~ [~1~ __ 1_, -

or Start and Hot Standby MtK e)

(cnt~~

to o~oure e acrn~ tcrore th~ po~cr could exceed tl~se!"ct1 ltmlt. The 15 per~~n: ~~~M scra~ r:~i*~' ec~lvc un~

t!l the :::ode a*;~t~~ ::; pll'ccd in th'."

RU~ p~~ltlon.

~h:3 ~~t:c~ occurs ~~~n rc11cto1" prCS:SUt'C

~ [;rf'Ot'!r than 850 l>:S lg.

JR:4 Ptux Scrrr.t Tr!'> Sct~!n5 The IJU1 117s te'" con= 1:! ts or 8 chamber**

' ln each or the re~ctor protection e1stem lo~!o chan~els. The J~K ts a 5-dcc~d~ 1n~tr~~~~~ ~htch cov~r' the rnn!C or po~er lev~l bct~ce~ th~t covered by th~ Si:'.*: am! the."?P.:*h The 5 d~codes er~ bro~~~ ~c~n Into 10 ra~ges, each belnc or.e-hal~ or a dec~de ln size.

lSa

t.t.A.

• ~

JCeut:-on nux Trtp.. setttng'

3. U~:4 Flux Screm Tr1p Set~lng {cont'd")

The JR:" ecre111,t:-1:> nett tng or* 120 H!~l::~on~ 13 ecttve ln eech ran&e.or th~ lfl~.

For c~om?le, tr th~ *1n3tru-

~~~~ were on r~r.~c 1, t~e scr3m Getting

~o~ld be

• 120 dlvlslons ror thnt rdn~el l!ke~l:>-:?, tr th~ 1n:>trur,,cnt ttere on r3nge 5, the scro~ ~?uld ~e 120 dlv1slo~= on th3t run~e. Thus, os the 1n~ ~:J ren~ed

\\:? to occo::u>dPJt'? the lncreaoe tn po~*;cr l~vcl, the acr~~ trsp settl"~ ts al~o rented up.

Th~ ~o8t ~l~ntricant.~ources or ren~

ttvtt:; ch~nf:>~ durln~ the ;>ower lncre:.se are ~~c*to cQnt!'ol ~o~ ~1thd:-nwel.

In 0!'1e:' to ensu!'e the t the 1Hi4 provided e1~~~ntc*prot~ctton egaln3t th~ oln~le rod ~*!thdr:?'l>a:l cr*.

4 or, o ran~'! o!'. roll "ltb!~c~at 1 ace hJ~a;t~ ""s !'n:-lyo:cd.

Thts e~~!;c~~ lMclcd~d 3tertlns the ccc!~ent P.~ "~rloc:: po... :cr l~vels. The :".!~~~ s~-

• v~::-e c:~~ 1a-1?l*1t-s _an 1n1t~~l co:'!<J ltlon tn tth~ch the re<?ctor ls Just cu~cr1t~ccl r~d :h~ I~~ crot~~ ts ~~t v~t on 9~~1e.

2.1.*

J.d-!!.~~011:!1 cor:oerv:1t1nm 'ftr.s tn~-:on 1n thts en~l)'Sl:J DY O!l~U!n,:u_; thot th~ 111i*1 Ch3nncl clQ3~~t t~ the ~~:h~r~~n rod 1~ ~rP~~3cd.

"&'h!!-l*esu!ts or \\.ht:J 0:1?.lysls 30~~-; th:1t the

~n!t 1~ a-;:r:w;*.:-::*J c:id per.;< po*,;r:r 11~; ':."?d

~? 0:1. pe:"C'!~t of rat~~ pOl*:<!r, ti1'J3 1r.alntalnlnq

~.CPR above the f.'CPR fl!'!?l cladding integ!: ity safety limlt.

Based on the above

?;!nl:;::e, th'.? !?t:i p:-o.,:tl~~ p:-ot~ctton a:;.:?lnst

!oc:- l co~1t!'ol red ~I th*J:-l'"a 1 cr:-o:-s nn1 con-

..,,.0*1- ***t'-* 1 -*****,1 or

\\;

l

J 0#

c?.~ ro ro* ~ J" :.:<?.*'-""tee

• 1?1~:; tJ&'OV iuc:J bu*.:!<up p~*u~~ct.lo:i ror t.h'! r.!'!':*l, Amendment No.

.'"'\\

APRM Aod Block Tr1p Setttns Reactor.po~er level m~T be **rted *7 rnov1n~ control rods or by ver1tng th1 r'!o!rcultt1on tlow r3te. The APR.~

syotc~ pro~1~ea o control rod block to prc\\*cnt qross rod vlthdraval st coa:Jtunt rcclrculctton hot:

rnte t, prot~ct ~~~tn:Jt

~roaaly exceed-ing the MCPR fuel claddlnq lnteqrlty safety limit.

This rod block tr1? ~cttlni, ~h1ch ta e~tc

~nt1cally v~rled ~lth rcotrculotlon loop Clo~ r~tc, prevent~ an !ncrc~ae 1n th~ r~a~tor p~w~r l~v~l -to e~crs-

1 lvc vnlu~:J c!ue to cnnt.rol r~ ~~:!t drD'l-~:ll.

'l'h'? flow vcr1abl~ t;olp s~ttlng prov tc!cn ~ u~:J t:t:l tie l :'!::' !"!; ln fro:w (\\;'! l da:~:ice, n::;3u::Jn~ o l'lt-:?CC:Y-fit:l:;o C"'e 000-t *o

, n ~..

.. 1l*~

r.~*;, ~r.::::nr:, over th~

-:?ut.lre r(!cJ!'calatSoil flow range.

'!°:1*,t ru:"l:-gln to the !iate~v Ll:n!t tr.er'!~::<-:> :!I th~ fl~~ dec~c~~c~ C:~ Lho o~~c~:**j trip s~t:!~~ ~~r3uS flow rel~t*cft;~4p*

• h

('

IL *'"'"-

f

.... '!re.01*e the ~H~:-:1t C:?!ie r.;c?!l ~;li1ch cc\\;jd occur <!u:-::1~ a\\:~~~7-ot:tte 0?*.. *3.

.t!cn I~ e:; lOC~ or r~te~ th~r~~l p;~cr bccru3e or th~ APRM ro~ block tr:;>

oc~t1~s. The oct~~l p~~'!r d!o:r!butlo~

1., t.1c eore :s c:>~n:,t:uhcd bv *~,.c*"*,.d t

l

.. t-**-*

co:i

~:>

re..,

DC'1U~ncc:; and !a r.:on!tO!'!!d C l"r. t * !'.!UO\\:~ 17 t>*1 the 1 n-CO...,.

1 n~** *..,. t*-

Q

,.3 i*; 1 t.;1

~~t' APRK ~cl':tr.1 \\.r1~.,....... :.

the ;,pRM ro~! :>l.,.**i< t.. 1p.... ~... 1.:.:";:.::..!d

• • ~..... -

lust('d clc1wnward or AT'HM gain increased I ii' the rnnxlmum frac.ti.on of lindting p11wcr d1~n

;i ty for G.1;:. f'uel exceeds tllP. fractlon or rated power, tt*:.1;; pre-Sl'rvlng the APT{M rod block safety murlJ,in.

16

~ ** Tut'blno S~t'J? Ve1¥e"'Scr.'.'n - Tho tur'blne eto;>> YalYO clo~t:I"! 11c:-aD trlp r:ttlclp.'\\tcs tho prcs:.uro, r.-:utro!1 fl*JX nl':" h!D\\ flux 1nc~!:::o th:it. could

~~U~t fro~ :r:'.?!d ~!C~U~3 Of the turb1no ~lop

.,;,lv*;:s. ;ilth c scr~" tl"l'O Gct.llnc or 10 Jl'!rCcnt. OC YDhO Cl?~Ut'U frOI:\\ full Oj)CR, tho re:oultmat tr.crease in O\\\\rf:ace heal flux ls ll:"\\1'.cd such U*.:t.t l:Ci>R rc:r.n1n' above 1 the* HCPR fuel clnddln9 lnteqrlty safety llmlt, even durln9 th~ worst case transient th~t ~ssum~s the turbine bypass is closed.

r. C:cnna:nr 1.oad 1'dectfon Scre11 -

The genera-tor lo:d rejection scra~ is provided to cntlcl1u1tc the rapid increase in pressure end ncutroa flux resulting from.

fa.:i~ closu~ of tho turbine.con~rol valns du~ to 6 1~ reject.lo~ end subsequent f~tlu-:"C or the b7J't'-"'5Z 1.o., ll prc'fent.s

-;

;;:;! frc:i ttco:-i~~ le:a3. lhM the MCPR fuel claddin9 lnte9rity safety limit for this transient.

For the load rejection without bypass transient from 100\\ power, the peak hC'at flux f and lhP.refore l.HGR) increases on the order of 15\\ which provides wid~ mac9in to the value corresponding to f°tk'l Cl~riticrlin1~

melting and 1% cladding :;tra:i n.

Amendment No.

c.

R~nctor Coolnnt tov Pr~99ure tnltlAte9 Mftin $ten~

lso~ntio_!'I Valve Closure - The low pressure 1soloti01l "t 8~0 rstg wns provided to give protection ar,ainst fo9t renctor dcrreuurizatton and. the resulting rnpld cooldown of the ves!iicl.

AJ,,ontoce vos to~en of the !lcrnm feature which occurs vhcn the ""1in steam llne tsolatton valves are clo~cd :to provide for reoc:tor 9hutdnvn so thot orcratJon at nu9sure*

lover t:111!1 tho!lc specif led ln the thC!r"al hydraulic sa(cty Umlt docs not *occur, althour;h operntlon ot o pre~sure io'-.?l" lf,on 650 pstc wo"'1d not necess11rtl1 constitute on unsa(e condition.

B. fl:iln Stcor.1 Line lsolntton Volve Closure Sera* - The low pressure isolation of the moln stco111 lines at 850 pslg V09 provided to give protection acafnSt rn?id reactor dcpressuritatlon and the.resulting r:ipld cooldovn of the vessel.

Ad'fant;,ge vas taken of the scram feAture vhlch OCCUt'S vhcn the ma.In steam line tsol3t1on volve9 ore closed, to provide for reactor shutdown so *that high rower operation At low re:ictor pr~ssurc doe9 not occur, thus pro'fldlng protection ror the fuel clodJlng lntc~rlty safety limit. Operation of the reactor et pressures lo~er

• than 850 p!tlg rcqulre9 that the rc3cror wnode 1tvltch

-be in the stttrtur.position "'he.re rroteetloh of the fuel clutldlnc intccrlty ufety Umlt ts provided b1 tlte. llUI hl:h neul ron flux scr.,...

Tims, the co111blnotlon of 11111111 !ltec111 line low pressure holntlon and lsobtlott V3l'fe clo9ure screm as9ures the avellablJlty of neutron flux scram protection over the entire rnnr,e of opplic:iblllty of the fuel c:Jnddlng lntegdtJ eafcty Umlt.

In acldltion, the tsolotlon vnhe closure scro111 nntlclpntes the pressure end flux tr:tnslcnts which occur durln& nor~:il or lnndvertent hol:itlon v:iJve closure. \\lith the scrr.*s set at lOZ valve closure,thcre h no appreciab1P incrP.Pse in neutrun flux.

I

1.2

~AFETY LIHIT J ** 2 REACTOR COOLl\\tlT SYSTEH

~pllcnbtllty:

Applies to limlte on reactor coolant system pressure.

Objective:

To cstablteh a limit belmi which the.integrity of the reactor coolant system ts not thleatencd due to an overpressure condition.

Speclflcatton:

1'1e reactor coolant system pressure ehall not excef!d 13115 pslg at any time when irradiated fuel is present in the reactor vessel.

Amendnent No.

1. 2 LIH1T1Nr. SAFF.n *svSnM SETTING

2. 2 f\\El\\CTOR COOIJ\\NT SYSTF:H

~cnhlJtty:

Applies to trip settings of the tnstrumente and

• kv!....:o whlch nre provided to prev.:nt the reactor RyAtcm safety limits from belng exceeded.

Ohjectlvc:

To dcrlne the level of the process variables at whlch auto1Mtic protective action ls initiated to prevent the ea(ety lilllita fro* being exceeded.

Spr.ctrlcatlon:

A.

Reactor Coolant High Pressure Scram shall be

~1060 p11ig.

B.

Prt111.1ry System Safety Valve H01Dtnal Settings shall be as (ollO\\ls:

1 v:ilve al ll 15 ps.lp:*

2 vnlvcs at 1240 J>Rf K 2 valves at 1250 pslp:

2 valves at 1260 pi; 1 g 2 valves at 1260 pelg 111e al luwable sctpulnt error for ench valve ahall be HI.

• Tarp,et Rock combination safety/relief valve 19

!!w.t 1.1 'ltt* n*ctoT eoolMtt *.,.,"' lntetrfty h "

tt11Por-tant b*rrlel' la th* PH"ntlon of anl:ontrolled r-le*** o4 fl**lon *roducte. It le ***entlal that tt.e lntetrlty o.f thl* *yne* be protected by **t11bll*hlnl

• pre11ure ll*tt to ~. obeerYed (or ell oper*tlng c~ndltlon* end vheneYer there I* Irradiated fuel lft the reactor Y****t.

'nte pr***ure **fety lllllt of ll2' p*ll ** 11e**ure*

t~* weeeel *t*.. *p*c* pre**ure lndlc*tor I*

e~ulYalent to ll7S P*l& *t the louwet eleY*tlon of the r*actor* coolan: *Y*t*~. The ll7S pslg Y*lue I*

dertY** froa the d***t~ Pr***uree of* the r*3ctor pr.1e111re Yeseel.

COOlan~ *Y*t** plplna: *nd hole-t!oa coadeneer.

The r**~ctlY* deslGn prC!se11r**

are 1150 ptla at 575*r. ll 1S peta *t S6o*r. end 1250.

psl1 *t s1s*r. Th* pre**ur* snfety ll*lt u~~ cho9ea

• • tn* lover of th* prrs*ure t.1'31\\Slrnt* p*r*ltted 111 ;~* *;>pllcahl* ~nip code**

A'>:iF. loller ond Pres~ure Ye~**l Code, Sectlcn 111 (or the pr*~~ure

~**sel *nd'lsaletlon conl*nsar *nd l"SASI 111.l Code fer :he reactor coolant *1*t*~ plplna.

The AS*:Z toiler *r.d Pressure Yeste 1 eo~. peralt* fl'f!SH*Jre transients u;..to IOI over <<!~sign press*1re (llCI Jt USO

• 1115 pdg)

• and t!tC! YSi\\51 C.,de pe~l t*

f:'l!!u.ure tr"n~l*nts up tet ZOl O\\"'!r th.. cir*!*....

rre1sure (120: X 1175

• 1410 ;>sic)~ The s~r*tr 1.t:slt Pressure of lJJS psta h rcfennced to the lowest elevation of the reactor vessel.

The design pressure for the recirc. suction line pipin~ (1175psig) was *chosen relative to the reactor vessel de.sign pi;-essure.

Demonstrating compliance of the peak vessel pressure with the ASME 0\\'1.~rpressure

• protection limit (1375psig) assures compliance of the sul'tion piping with the USASI limit (l410psig).

~~v.*1.uat:i.on met)hodology used to assure that this

~~~y limit pressun* is riot exceeded for any reload is ducumented in Hef erencc I

XN-NF-79-71.

The design basis for the reactor pressure vessel makes evident the substantial margin of protection against failure at the s;1fety pressure limit of 1375 psig.

ThP vessel has bt~en designed for a general membrane stress no greater

• th<.m 26, 700 psi at an internal pressure of 1250 psig:

this is a factor of 1. 5 below the yirild strength of 40,100 psi at 575°F.

At that pressure limit of 1375 psig, the general membrane stress will only be 29,400 psi, still safely below the yield sttength.

The relationships of stress levels to yield strength are compara,Ple for the primary system pipin~ and provide a similar margin of protection at the established safety pressure limit.

The normal operating pressure of the reactor coolant system is lOOOpsig.

For the turbine trip or loss of electrical load transients, the turbine trip scram or generator load rejection scram, together with the turbine bypass system, limit the pressure to approximately llOOpsig (?.).

In addition, pressure relief valves have been provided to reduce. the probability of the safety valves, which discharged to the drywellt operating in the event that the turbine bypass should fail.

Finally, the safety valves are sized to keep the*reactor vessel peak pressure below 1375 psig with no credit taken for the relief valves during the postulated full closure of all MSIV's without direct (valve position switch) scram.

Credit is taken for the neutron flux scram, however, The indirect flux scram and safety valve Rctuatjun provide adequate margin below the peak allowable vessel pressure of 1375 psig.

Reactor pressure is continuously monitored in the cnntrnl room during operation on a l 500 psi full sci le pressure recorder.

14> sAR,s-ec-tTun--rr-:z. 2 -

also:

"Dresden J Second Re load License Submittal," 9-14-73.

20 also:

"Dresden Station Special Report No. 29 Supolement B."

e

• e 2.2 In ~u... wtt'h ~tlon llt of the ASPCr °"'** the D*f~ty **l*re svat ~ *et to a"" *t no hlAhet' than lOll of *Hit!" pft*eunr. and lh"f ~t 11.. tt the Yf!*Ctol' prr**ure ~o nn wore th*n l IOt of d*nlr.**

prtn!lute.

llnt'- th* neut con f lux.,r.ra.~ *n-t **f*ty

• • lve ectu*tlnn ere r*~ul~ to prevent ov~f'l'r*e e\\:tfr.ln1 the rt'itc:tnr pr**,.ure -e~el encl thu*

• • c:*rdln& the pre~**re a~lety ll~lt.

11>~ pre*1n1re

• cram lo available ** a backup protection t.o the dLn.*ct valvP l.J<*:_;ititm tr*:Lp

~;c1*am:~

and.the high flux scram.

If the hl9h flwc acran1 were to fall.

• hl9h prea*uro *cram ~ould occur *t 1060 p*i9. * ~nalyses are performed os described in refPrcnce XN-N~-7~-71 fur each reload tu assur*e that the pi*essur*e safety limit lu not exceeded.

~*-*~,..,,

Amendment Ho.

e

-~

21

(""\\ e 0

0.!-.-----------------~0--------------------~-'--

. j, t LIM!tllfC CO..DtttOlt roll OP!MTIOlt l

'. 3.l

• r.ActOll r!IOTtCT1cr.t STST~t

!f Pllt1tltl I ltxt

.Aprtlt1 te the lttttrut1C!nUtlott,...

4190t!attd ~e*lce* ~hlch lnltl*t*

• tf1dor eero*.

~J!.~*

to a2t~r* th*.,.r.,t11t, of th*

r*actor *rotectlon *1*tea.

hedflc~I A. th* Ht:tolttUe.. tn!.CUD *nu'!"!>>er of tdp

.,,te.., ~J !'thtl:""\\.....,. n\\:~cr of* s~stru-r*!rtt duumel~ t!*n:: !'!'\\*!"It ba Ci'C!'t'Oble fnr ench ;o:ltlon of the reactor eode

,.,ttch sh::Jl be,., ~"'"" In T~lile 3. 1 :1.

The system response times from tbe opening of the sensor contact up to and including the opening of the trip actuator contacts shall not exceed 50 milliseconds.

I B. If durinP o.nP.rAtinn, thP mnximnm frac~:iton -o( limiting power density f~;~" fabricated by GI~ exceeds t*~e fraction of rated power when operating above 25% rated thermc.11 power, either:

I

a.

The APRM scram and rod bock scttinr.

shall be reduced tu the values given by the equrit i.ons *in Speci-fications 2.1.A.l and l.l.R.

Amendment Ho.

.l.1 SUP.VEtLLNlti uquuumtMT 4.1 R£,\\Cf0:\\ Pl:OTF.tTIO~ SY...llQ!

A(tpllu to the *urvctlhnce of the fftlUVlllC:D*

tnt!:n ftnd ~'socioted dovlcea vhlch tnltl*te.

l'cactor: 1cr:1*.

Ob *r. ct I vr.:

To specify the typo and frequency of

• urvcJll~nce to be opp\\led to the proteetloa Snstrutr.'!ntotlon.

!reel rtc,.,t ton:

• A.

ln~trc"'~ntetlon syste~ shall bo functfon~lly t~stcd L~J C3llorntcd m**

ln~lcatr.d In Ta.hies 4.1.l L,d 4.1.2, respect l\\reli.

I. Daily during reactor power* operation ~bo~e 257. rated thermal power, the core power distribution shall be checked for:

1.

Maximum fraction of limiting power density for *fuel fabricated by GE (Ml"LPI>) ;mu compared with the fraction of rated power (FRP).

2.

For compliance with assumptions of the Fuel Design Analysis of overpower conditions for fuel fabricated by ENC.

3.1 LIMITING CONDITIONS FOR OPEKATION Specifications (cont'd)

b.

f The power distrihution shall he chan~ed such that the maximum fraction of 1.imiting power density no lbnger exceeds the fraction of rated power.

For fuel fabricated by ENC, operation of the core shall be limiteJ to ensure the power distribution is cu11sistent with that assumed in the Fuel Design Analysis for over-power conditions.

~"

~

.. -.. !"'~..-.,

Amendment No.

4.1 SURVEILLANCE REQUIREMENT

n h:ttr Sl"r:im :tnd rod hloc:k condition. 11.us, I( lhe c:*llhr:illon were 1>cdormcd tlurlni: ovcr-

ilinn, nu~ !lha1Jlni: would nol he possil>lc.

U:iSt*d **n t*~:11t*ri<'ncc :il nlhcr t:l'nt* r!'ll lnJ; sl:illonl', ~lrHI ur lnstruml'nls, s11d1 :-19 those in the Fl1*w lli:tslni: l\\cwnrk, I!! nt*I sli!nHic!'lnl

uul lhcn*rurl', lo :froid s1mriu11s scrams, :1 C: :t 1ihn1I i1111 f l"l'f llll'IH:~* o( C:1c:h rd Ut:I i 11 I: Olll:'l gt.

Is c:zll:1lill:iht*d.

Gn1111* CC) 1lc,*lcc:s nrc :tclh*c onl>* tlurln;: n a:h*cn 1u*1*llun of t!u: 011crnll1111:1I cycle. Fur cx:a1111*lc, lhc 111!\\l Is :tcll\\*c tlurlns: st:nh*1* :"Ind ln:tc:Un* 1l11ri11:: f11ll-1mwer u1tc1*:11l11:1.

Thus, the onl)* h:sl lhal Is tnc:tnln::hil Is lite o:*c 1*cr-Cunuctl Jui:t 1n*lor tu 1tl1111down ur &l!'lrln1*: I.e.,

lhc le~h lh:1l :trl' 1*cdunncu just prior lo use o( lhe l11~1n11m:nl.

C:1tlhr:tlion frcr1ucnc1 or lhc lnsh*umcnt ch:in-ncl lit di\\*hlcd lnlo lwo s:rou9tll. These :nc :1~

follU\\\\*~:

I.

l'n~slvc ly11c lndlc:tllna: dc,~lc:u lh:tt c:an be t*mu11nrcd with like units on :a conl111u-O**!I l,:1tilS.

2.

V:acunm lube or scmlconduclor devices n:icl rk:t*clon th:tl drill or iose SCn!illh*ity.

F.xrc1*lcnt*~*w1th p:tssh*c l~-pc lnslrunlents In Co1nmfl11wr:slth Erllsnn J:l'nt*r:tllns: sl:tllons :snd s11bst:1lh*1tJ lnrJlc:tlcs lh!'ll lhc :<ipcciricJ c:-ilihra-li1*ns :trc: :u!!*r 1~:tlc. Fur those 11C'\\*lc1~s whir!l cm1tlor :*11111lificrs, etc., d:*lfl s;1rdfic:tlio:is c:1H Cur *!1'TIJ 1.. lJu less lh:tn O. -1*:,./inun:h; I. c.,

T' In the~*~

t 1!10nlh :s drift o~.. 1'; w1111hJ o<:c~u* :uul lhns pruvit.li11;: for :tl!cq11:1:c m~t*;:in.

For the APR;\\I s~*sl<'m drllt of clcclro:-tlc

ipp~r:1l115 Is not the onl)' consldcr:illun In <!e-lc r111lnln1: :t C!'ll llH":tlion frcriucnry.

Ch:i11~c In 1111\\H'r tlislrlhullo:i :1111.J l.. S!I or c::1:tmhcr SC!"l~i li\\*it~* *li,*1:-ilc :1 r:11ihr:ilion '-.,.**,.~. ~c.. *c:n rl:tys.

C:11i;,r;1llon un lhl ~ h-r1111cnq* :i_ssurc s Ill :ml opcr:iliun :tl or hduw lhcr111:1? llmlls.

,\\ ('Ufnl'!'ll"ISllll ur T:*hlcs.... l. l

11~t.I *l. I.?

h11llcah:s th:il !!ix inst n::iu:nl ch:in:lCls h:tve not hl'cn lndm!ctl In lhc l;1lh:r Tahh:. Tht*:ic :trc:

\\!111lc !>wilch In S!a11tia\\\\ n, ~!:1:iu;1l ~n:tm, lll~h

\\\\';1kr l.cn:I In Scn1m 1Jisch:1q~c T:u:k, ;\\lllln Slc:im I.Inc 1:-;ul:!llun \\"ah*c l:ln~urc, Gcm:r~lor J.o:itl lkjct*1i1111,

-in~ Turhi1:l' Slofl \\'ah*e Clos111"e.,\\!I ur lite dc\\*lcl'S UI" b\\.'n::urs :tssucl-

=-*nl wi1?1 lhl."sc sc1*:1111 hmt*lluns :.re lllm11h:

on-uH swill-hes :uuf. ln:nc.!, c:t!l11r:11i1111 Iii no\\

'1'1*lh::1!ilc, I.e., lhe :;witch ht either un oT~

uU. Further, these S\\\\ llcht*s ;u-e 11101mlt*d suli*lll I" lhc cle,*lcc :11111 !i:l\\*c n \\*cry luw 1trt*h:1hllil)" or 1110,*!n;, t.'. ::. the switches In the sc1*::t11 clisc:h:-iq~c,*11lumc 1:111k.

1:~sctJ on the :1hu\\\\:, *no t*:illhr:1I i1*n Is 1*cc1:1l rc*I fur lhc11c t1i x Inst r11111cnl d1:umd:t.

B.

The MFLPD for ~uel fab~icated by GE shall he checked once per day to determine if the APRM gains or scram requires adjustment.

This may normally be done by checking the LPRM readings, TIP traces, or process computer calculations.

Only a small number of cont.ri>l rods are moved

<laiJ y a11rl thus the peaking factors are not expected to change significantly and thus a daily check of the MFLPD is adequate.

For fuel fabricated by ENC, the power distribution will be checked once per day to ensure consistency with the power

<li~tribution assumptions of the fuel design unulysis for overpower conditions.

During p0riods of operation beyond these power distribution assumptions, the APRM gains nr scram settings may be adjusted to ensure consistency with the fuel design critew-;,

for overpower conditions.

34

3. 2 LIMITING CONDITION FOR OPERATION C.

Control Rod Block Actuation

l.

The limiting comlillons of operation for the Instrumentation that lnlllalefl control rod block are given In Table 3. 2. 3.

2.

The minimum number of operable Instrument chamwls Sllf'CHied in Tahle 3. 2. 3 for the Roel Block Monitor may be reduced hy one In one of lhe trip systems for maintenance and/or testing, provhled that this condition does not last longer than 24 houni in any 30-day period. In a 1 Id it ion, one ch~mne l may be by;>asscd above 307. rated power without a time restriction provided that a limiting control rod pattern does not exist and the remaining RBM channel is operable.

D. Steam Jet-Air Ejector Off Gas System

1.

E~cept as specified in J.2.0.2.

below, both steam-jet air ejector off-gas system radi:-1tion mon-itors shall be operable during reactor power operation.

The trip settings for the monitors shall be set at a value not to exc1.:.*ed the eq11ivalent of the stack release limit specified in Specif icaL inn 3. 8.

The time delay setting for closure

. of the steam jPt-alr ejector isolation valvl's shall not exceed 15 minutes.

4. 2 StJRVEILLANCE REQUIBEMENT 36

3.2 LIMITING CONDITION FOR OPERATION

2.

From and after the date that one of the two steam-jet air ejector off-gas system radi-ation monitors is made or found to be inoperable, continued reactor power operation is permissible during the next seven days provided the inoper-able monitor is tripped in the upscale position.

4.2 SURVEILLANCE REQUIREMENT 36a

r:ln!~'J~ 1:~. or C,!>~:r.b\\' In:at..*

Ch:anMlS Par

.-:-in ~ 3tc:":\\ 1 l

• 1 2*

l 1

l 3

l DPR-25 JKSTR01'L.~J\\Tt<.'lf. Tl!l\\T Im:TL'\\TJ:S 1lOD Bf.OC:lt Table l.2.l 1nstrum!!nt A?~i Up*cale (rl~~ bin~J(7)

Trlo JA* l s~~tlner

~

t'HI' 0.5~0 + 50

]Ji'LPIJ APNI up11c&1!* Creruol

• en4 8tGrtup/Rot:

Etant\\by r.:oo-?)

APr.Jt downecalo (7)

• Red block fftonl tor- "111'Cala ( fl<N b(asJ 17) nod bloc" l'IOnltor cl~-nscnle (7)

JP.. 'I dawnseele CJJ liVI upacale lfl=1 clet@ctor not £\\llJ.y in2ttrtad

• Ill tho cor*

sruJ detect.or not ln otartup posltlon

~ 12/12' full ocale

• 2 J/125 full *c:*l'!

0.65 w0 + 115

.? 5/125 f Gill *cal*

> 5/125 full scalw

~l08/12S full.seal' C4l 121 (2)

~* ~*~-~,2~1.~5~~G..._ __.._--'S~ltft---UJ!C~D~1~a=--------------------------_..--.::;;.:::..;.._.;:::;:;,;:~.::a..;::;..:a.;:~---

'1*-*'!""*r, o' cc~nts 11*c I I l\\mendmen*t No.

I (2

. \\

I

l,'..

u T~3~~ 3.2.3 (cont)

1.

Fe::- the St:?r'.:\\:?/~ot. stnnd~y a:id Run P<'!=itl.~:i5 of the r.cat:tor P~'Y!c Selector Switch.

there* !J~r.~. l be t~o c;><!rnbic or t.:::-:.pp~<\\ tr i? ~yst,c~3 f:>r c:ich ru~ctior.,. c::-:ce?~ t~c S~t rod blod:s.. Ir..M up!lc~l~. I~~ tio,.m:.~cr.lz ;l)r:d :?.:*! c!'.!t~ct:>r r.o~ ful1.y 1ns~!"t~4 !n th~ core n~~d nf)t b:! opcr\\\\ble in th'! "r.'Jn po:Jit.io:"I a:id ;,,.,~, C.cwr:scnle. l\\Pr'.:*l upscale (flow bhs),

and ROH dOMnscile neoa not be operable ln the Startup/llot/ Standby llN)de.

'Jbe RBM* upecele nee4 not be operable at le1111 \\hen 30J ratd thermal power.

One ~hftl'lnel..,., be bypnHe4 aboft )OJ rated thel"lllftl pover

• prorlded \\hilt.

a lhlltlng control rod patterD doe* not exht. For *1eten vlth *re than one channel Der trip *7*t*,

If t.he f ir:;t col\\,:nn collnoc oo m~;: z:or oo~n ~r1p sy*n:cJ:'~,. *t:nt:t systems shcill oc trippeci.

I

2.

ff,u n~rcea.,~ nr drive rto~., required to produce a rated core flow or 9u lilo/an.

MF'LPD=highest value of F'Ll-'D for* (]. E. fuel.

3.

IP.M ao-~nscol9.m~y be bypassed when it is on its lowe:;t r&nge.

4. Tni3 fu~ction ml!y be bypass~d.when th*! co~nt rate is ~l~O cps.

5.

One of the fo~r SP.H inputs r:tay ba byp'ls:;cd.

./

e

6. This SP.M func~ion mn;* b~ bypci:;3cd in the higher Irut ::am;cs when thC! IRM u;;>scale red block is opcrcibl~.

1. :t*ot required while performinq lo"' PO\\*r.r physics tests at atmospheric pressure during or after rcfucli:lCJ nt power levels not to exc~o:?d S f'!1f(tJ.

__,._,. I

~**-*~r.,

Amendment No.

42.A

_........... ~---------

1,. -

0 B:ts~s:

3.2 In addition to.reactor protection Instrumentation

"*hlch Initiate* a rnctor scram. protective Instru-mentation has been provided which lnll1:1tes action to mlllg:1te the consequences o( acclclcnls "'hlch :are bcyont~ the operntors :1bllil)* lo control, or lcrml-n:1le1 orcroator errors hclore lhcy rcsuU In Sl'rlous con1.1eq11cnct"s. This M.'l of ~clflcatlons Jlrovlcles li1c limlll11:: conclllions of O!'crallon for tht* prini:1ry system lsolallon (unction. lnlllatlon of thl' emcr-t;l*nc)' core cool Ing sysu.*n1 1 conlrol r01I IJlocl> arMI slancll)y ~39 lrc:tlmcnt li)'Stems. The ohjrctlvt*S or the speclflcallo:lS arl' (I) lo :a&Slll'l' the l'Hecth*l'ncsa ol the pa*otcclh*c Inst r1uucnt:1llon \\VI.en rN111in*1I hy prcscrvlni; lls C:IJ1:1blllly lo tolcrntc a slni;lc lailure.

-of :iny componc.-nl of such syslc-ms even 1h1t"ln.: p1*rl-0tl9 wlll'n roa*tlons of such sp;tcms :trr oul of sc*n*lce for molnt<'n:incc. arMI (II) lo prciicrlhc the h*li, tiCl-t111i;s requlrrcl lo assua*c a1lc1111:1lc rcrforn:uncc.

\\\\11cn nccesliar)*, one clwnnrl m:1y llc m:ule* lnoJl(!r-ablc for brll'f lnlcrv:11s lo cont.lucl 1*l'*iul rc*I luncllonal lesla onJ callhrallon9.

Some of lhc scllln~9 on the lnslrumcnt:lllon lhal Initiates or conlruls core and conl:1lnmcnl cooling hue tolernnccs e Kpllc Illy stated *here lhc hl~h :and low valacs :1rc holh erlllcal and m:ty ~a,*<? :1 subslan-llal ellccl on s:1fely. ll i;hould be nolcr.l:1!lon, where onl)' the hli:I\\

or lo\\\\' end uf the Bellini: has 11 direct l11"arin:; on s:1rl'l~*. arc chosen ot a lcn*I 3w;1r from the norm:1I nnrw;alln~ ran:;c lo 11rc,*e11t lnadv<?rh*nt at:l'.l:tllun or

_,\\~J'ar'!'t_,. p~*sl~*m ln,*oln*d and <'~~ur-:.- to ah~*,rm:al

~trttfttllo:19.

lsolallon vnh-P!J ore lnslall.-d ln tho~ 1111(?9 th:1l rx*n~tr:ate th4! primary cuntalnmenl arMI musl be lsnlatccl durln~ a loss of cool:int acclclr.nl so th3l the r:11ll;;t1011 closr limits :1r~ not cxccc*ll'1l 1lurlng an ac::ltlent conclillon. J\\ctuallcn or thrsc vah*cs h lnlll:1lc1I by prolt*cllH* l11slrm11cnl:1Uon shown In Amendment Ho.

0 0

Table 3.2. l which ecnece the condition. for Miich h1olallon Is required. Such lnslrumenl:1llon must be

av:1llablc,\\*hcncvcr prlm;iry cont:alnanrnt lnte;rlty 111 rcqulrccl. The objl'cllvc Is to lsolnte the prlm11r1 cont:1lnmenl so that the i:11hll'll~* or 10 CFR 100 are not exceeded t.lurlng an :acclclt':-il.

111r lnstrument11llnn which lnlll11tea prlm1r1 1y1tem h;r>hllon Is conll(?clcd In a dual bus arr*.::-:~mcnt.

Thus, the cllsc-1111slon r:t\\*t*n In the base* for ~clR c:allon 3. 1 Is attpllc:ablc llcrc.

IMM* -

.... l*HI l-U-ftt fl..... Mtln f-1 le........ te

.. 160 l..ch4*...... *****I **rot.... *fl** ea1.. a,.. t..,... f*ll

...,,., pre***r* ****.c**** *** ete.. **t** the a.. le.. a **** I*

• l 'o* 1.-chrr*..... *****I ****... 144 IM"9'** *._..,...,.. el

• ~**** *****

~.,,.tit*******"**** ac'I** ***I '*"'**I.I*

1...chre lo... el IMn **tllel' f*f'I ***lf!M *

..,_*****......, l*I*

• etpo6ftt9 -*** *~N I* tit* IAJCA *"*"'*****

T"** trlp a"ltl*l** cl***** ef Co.,.p I.... J P*caarr ~

holetlOft **IH* ** *-* -

telp,.,. recbcwl*U..,,_,. l**f*S*

• ...,., ** **ell.. '*'****,................. so*... ~.......

• • • • • • *er* **** a...,11w............., *c****......... *****c....

• • • • • clo****,._..,..,. **I*** *Ill M cl*............ *I****- el

.... c.......... _...,,.........,... -*-...... *... ""'.....

tMrelcne.. c.-****

..... *-............., ***tr-*****-........... -* **-

• • • W4l*f ******* t44................................ *f......

'"'"'........... *~*...... ect*** r *****

Thia trip lnlllatc11 closure or liroup I prln1:ary cont:ilru:1ent lsol:1llon V3lvc~s. ncr. Scc!lon 7. 1. 2. 2 SAR, ~nil :also ocllvatc*s th-:.- ECC suh!\\ys!*!ms, st:1rt9 the ein<'rt;ency clic*scl ::cncr:1tor :a111l lri11s the i'eclrcul:allon pu11111s.

1*1i1s trip scllln~ lc,*cl was chcs<'n lo he t:l=:h ennu;:h lo fll"l'\\'rnt !>111:.-lous 01"1".'r:illon bat lo'" rncu~h lo l:tl-ll:1tc ECCS o:it*rat!o:i ar...l r:rlnmn* ~nslc*m lsol:illo!t so lh::l no mrl:tr.:; of 1:1r. hel d::;lcJ1ni: wlli occ-1*r 11n~

so that l'"SI 3cc!tlent conlin~ can lie occom11llahrcl

.11111 the i;ui*ll'lin.. s o( 10 l: r:n too wlll not IK' \\*lol:ttMI.

l*"or lhc* Clllll(Jlrtt' eh-c:lll! rl*P*ntl;i I hrt*:ik or 3 211-lnch rccln:ul:1llon tin.- :in*I 11*ith t!ll' trip Sellin;: 1.rt,*e11 rlHn*.-, 1-:CCS ln!tlallon :-iml primary s)*t'l*tn hilll:atlon a re lnillall'I! In time to mt*cl lhl.' :1IJO\\"C crllt*rl:1.

J. J LHllTING CONDITION FOR 01'1'.lt#\\TION C.

Scram Insertion Times

1.

11rn avcrnre scram insertion time, based on the de-energization of the scram pi lot valve solenoids ns time zero, of all oper-able control rods in the reactor 11owcr operation condition shall be no greater than:

\\ Inserted f roin Fully Withdrawn s

20 so 90 Avg. Scram Insertion Times (sec)

0. 375 0.900 2.00 3.50 111e av~ra~e of tho scram insertion times for the three fastest control rods of all groups of four control rods in a two by two arr:.ay shall be no gre_ater than:

\\ Inserted From Fully Withdrawn s

20 so 90 Avg. Scram Insertion Ti mes (sec)

0. 398 0.954

2. 120 3.800

2.

111e maximum.scram insertion time for 90\\

insertion of any operable control rod shall not_qxceed 7.00 seconds.

~,....,.,I

....,........,,~i-r,

4. J SllltVE I Ll.ANCt: m:11111 m:~u:NTS r..

Scram Insertion Times

1.

After each refueling outugc and prior to power operation with reactor pressure above 800 psig, ull cont.-ol rods shall be subject to scram-time tests from the fully withdrawn po!. it ton.

The scram times shall be measured without reliance on the control rod drive pumps.

2.

At 16 week inte~vals, 50\\ of the control rod drives shall be tested as in 4.J.C.I so that every 32 weeks al 1 of the control rods shall have been tested.

Whenever SO\\ of the control roJ drives have been scram tested, Pn evalua-tion shall be made to provide reasonable assurance that proper control rod drive performance is being maintained.

3. Following cc*mpleti.un.uf each set qf scram t1~~;ting as dL~GcrlhcLl above, the results will be compared ai:~ainst the average scram

p1-~ed i.lir; tribu tiun used in tl-e

• tran:;i1~nt analysis to verify the appli-cald.lity of' tl11~ cur-rent MCPH Operating*

Limit.

Herer* t(I ~;pt~cil'ication 3.5.K.

,Amendment No, 58

c.

I 0

lnJlutlve oi a C~C'rlc cont.°"I r-n~: Jrh'c"

>robin n~d the r.

.*ctot t1l ll ~c shll

.:,,.,n.

Ano if da.*u1e wllflln the control roJ "* l.vc

~.:<~*nltQ *nd In r~rtlcular, c~~cks in ~rive tntcrnll ho~tlns~. ~~r."ot be rut~~ out, ~htn a CCftCl'iC nroblca a(fcctir.I a n;.?;~Cr o(.J~~~eS ci:1n?t bo ruled out. Circll!::frre:r.thl.u:i.:lrs

. 1'Ctultlnc froa stress asslstod t"tcr1nnubr corrcslc~ have cccu~rcd In the col let hcuslna of ch\\vct at sevcrd t~iib. ll'.ls t1J>c of cr1c\\in: cc-~ld occur ln

• nc.::~cr or drives end :.r '!~c cr.lds rrer-3,:itcd untl I se~*cronco.

o( t~e collct hc~sin~ occurred, s~Ta~ cocld

~t prevtr.ted In the *ffc~tcJ r>>cls.

Lialllnz t~c perlol of o~eratlon ~Ith a p~tentl*lly 1cvcred collct t.ousln1 and rcqulrln: incrcase4 su:"":cllbnce *f:cr d-:tect ln~ o:... :acct r:~.. ~ ** 3SSUTC thlt th~ :'C2Ct~r wlll not b2 o;u:ntccl wlth a hri1~ nu:.ber of rods vlth

!.dlel. colht :t-:..:::lniS*.

Control Hut.i Wlthdr*uwa.l

1. Control rod dropout accidents a.s di9 cussed in HeferPnre XN-.NF-BO-l'),

Volu1n0 l, cnn len.d tu rd.1*.nj f'icanL cr*re

.... re. I! couplinz lntc~r\\ty ls c..~l:itctned.

t~c ;~s~lh\\llt! '! 2

'~l ~~crc~t occ\\Jcnt ls cl\\r.h.:ucd.

r.,~ O\\"c::tr::vc.l r:>sltlcn futu:-c rrev\\~S :

pc!l:t~O check A~ GnJy UnCoup:o~

d~\\ves 1118Y ~**ch this f~Jl:lon.

N*~tro~

h11u-.11t*t l*n res,ens* to rod r.:v*,:-.::it ptov*~ei a v*rlflcatlon that the ro4 ts !ol*

lowl~I lt* drive.

Absence of such response to drl~e c2\\"c~ent ~ould provide cause for

• u1pectln' a rod to be uf!coup.led end !ltuck.

Rcstrlc*i~I reccu~llng verJf lr.otJons to povcr le~flal.'::!.1f.~~a 20: prowldcs assurance that a rod"dl;,,*'i~~ins a -:-<"coupll~G "*~lflc:*tlon vouid no:. res**lt In

• rod drop eccldent.

2.

~t-.o ronuol rod ti.u1\\n1 suppert restrlcts the outv*r* *oveatnt o{ a tG~trol rod to Jc~~ th~1; :S lr.:?io In ~he f>rtrc-:!!:',..,*.:ta.

c*.*c:-.:

c.~ a t:c*:..1:;~:-::= f..!l lt*ro.

Tit~ 8JI01&11t* ci u.:iul.,~t:t "'~.tc~ cu:i~d b4t ~*"*' ~1 lhh

.Amendment Ho. 42 0

J.

01

.' 0 s*:\\ll n.'flOunt.of nd v\\fh,*uw:1I. whlch h lH*

th~n ~ norr~I sln,.1e.vithdr~w:1l lncr4:"'1cn~ vlll

..>~ CG!\\tdt\\;tC to

my,b::.:.;c to t?\\~ ;-rh::ry cool~nt syst~~. The Jcslin b3~ts ts el*tn '"

!cctlon

• fo.6. l ol the !':\\:t. :and the.tcsl~n cvahtl*.

tlon i~ tivcr. ln Section 6.6.J. This ;u,p.>rt ts not re~cl red tr the nactor cool:lnt systc:.

Is at *t:t.os;:t1-:rlc pressure s.lnco thtre '"011ld thtn be no ~r\\wlic rorce t, r~~l~ly eject a drive hou~lnt.

A~Jltlon2ll7, t~e sup~o~t t1 nnt rcqul rl!d \\ f a 11 control roJs Ire fi:l ly tn~erlcd *r.d if an :adcGU3te shutdci."!\\ r.>rcln

"*ith c.ne co:'ltrol rod... t:l:Jui.-n. ~as been i~:t0n*

stri:t~d slncc the reactoT would re'!1aln subcrltlcal' even ln the eYent of COQpletc ejection of the strone~st control rod.

Cor.trol rod wlthdr:aval ar.d Insertion se~11~nce* ere cst*~lished to 3ssuro that the n1alOU111 lnsc~~ence lnli~ld~al con:r~l rod or control rod st~cnts

\\o!1ac~ r.:-c wh~C:!'i:*-n could r.ot be ~1:>rth cr.c.ufh to cou~e the rod drop accident design limit of l80 cal/gam to be exceeded l i t~cy W'!r~ t:> ~:-:tp c11t o! the c<<1r*

tn the ~*n~c~ Jeftncd for t~c ~tJ Or~p ~c~tdcnt.f'>

These. sc!lucnccs are dcvclcr:cd p:lor t'> \\;ti~hl c.rcrnticn of the un!t foll.,-.dr.z ;.:.y n!i.'!H:i: ou~:!l9

• n!I t*!c r"':~**'. rc~:~n~ that :?n o;>cratt'r ':>!10;1 th;~o

~c111.?ncc~ loa..::tcd l:? br th;,e,r.u.tlo:i c.f t~c r.:*.~t.

or a second qualified station employee.

These sequences are rl~veloped lo limit reactivity worths of control rods and together with the inte~ral r£>1

"~le.:!.:)' lir.ltcts :1:-:J t}.r e:t\\:>n or the CO't'!IOI ra~ lii~e srstc"* "'t r~:entlal re1ctlvlty l11sc:-t in:i s~ch :~u tt.e rcsu! :s o(

• c<<ir.t~*-.1 rOd dro? :sccloc~t wll l no~ e11cc:~ a a.nlr.!e fu'!l Cl'lt!:"'Jr c:>:l~cr:t of 2S:t ut/er.*.

T!:c r-u~ foci er.th:si~r cf *

~LO c~!/~*~ ii bclc~ th~ c~er~y co:i!e:it a: ~~i'~

T~p'd f~cl J!!,~rs!l tnl p~~~~rf SJS~CN da~aze hlVO been fcu:~d ;.c; l'ccar l:nc.! on ea;~rl::c:iUI tn:1 *s ls dlscu~scJ ln Rcfc:e~ce I.

lhc :.1~:::rsi s ::r t~c 't':i! ro! Ted iro:t ecc!dc:t ~~*

cri~~~a!ly prcsc~ted in S::ttc~s 7.~.J. 1:.1.1.!

Gn1I t.:.1.1.<I! of tl-.c S:trety..\\!'::Jlyils ?.cper:.

r:.:;rc:-.:-o-t'._~ts ii; :oulrti:::.l c:ip1~\\H:r b\\"o a!l~*c.d s,.,, **

rc(a~e-:J an,Jysls of the control :-!!d d'Of' sc**'6t' 62

Bases(cont'd)

Parametric Control Rod Drop Accident analyses have shown that for wide ranges of key reactor parameters (which envelope the operating ranges of these variables), the fuel enthalpy rise during a postulated control rod drop accident rem~ins considerably lower than the 280 cal/gm limit.

For each operating cycle, cycle-specific parameters such as maximum control rod worth, Doppler coefficient effective delayed neutron fraction and maximum four-bundle local peaking factor are compared with the results of the parametric analyses to determine the peak fuel rod.enthalpy rise.

This value is then compared against the Technical Specification limit of 280 cal/gm Lo demonstrate compliance for each operating cycle.

If cycle specific values of the above parameters are out~ide the ranoe assumed in the parametric analyses, an*exte~sion of the analysis or a cycle specific analysis may be requir.ed.

Conservatism present in the analysis, results of the parametric studies,

- and a detailed description of the methodology for performing the Control Rod Drop Accident analysis are provided in reference XN-NF-80-19, Volume 1 (Supplements 1 and 2).

(1) t*one, C.J., Stire, l.C. encl Voole1, J.A., "llod Drop Aecl~cnt Anolysle for L3rge tollln1 ~ater P.cactoro",

NED0-10~27, ttarch 1972.

(t)

Stl~"""A:!IC"/l Peone. C.J., e"'lf You"I*

n.n.,.,llod Drop A:cltcnt Anely:i1.a for t:u ca 11::1l' s", Sun le"'cnt 1 -

1-\\ L!J0-

10) J J, Jvly *1912

(]) Stlrn. ll.C., roone. C.J., end H*un, J.tl., ~*::o~ Drop f,cd<'.c:it A:tclyeh for L.nce l\\lll' 11 AJJcndua t:o. 2, [r.poeed Core:i"

• Sc;-ph~~n: 2-t:r.DO l0J2 7

• J~nu3r/ 1~1).

Th~ Rod \\..'orth Hlnlml:r.er 11rovl1lr~1 nulo111:1tlr.

uuporvleto" to nnnur~ thnt. out or !t'"*1u~nce t.ontrol rocl5 will not be vll111lrawn or ln~ertf!d:

I.e.* lt llmlls operator dcvlattom1 fro* pl11n."lecl wlthdraw3l sequr.nce!I.

Rcr. Section J.9 51\\R.

lt serves as 3 bnckur to proccclur3l control o[

control rod worth.

In the evc*nt th.1t the ltod Worth tllnt:nlr.cr ls out of service. wbcn rcquil'ed 9

1 !.lcenscd operator or oth~r qunllftr.d technical einployre c.... n mtll

u.:1lly (ulf l ll the control rod p3llcrn conform:tncc functlnn~ of the Rod Worth tllnlnl7.cr.

Jn this C3!'1'! t procedural*

control Js cKcrclsr.J by vr.rlfylnr. nil control rod poslt Ions a(tr.r the wlthduw.11 of each croup. prior to procccdlnc to the neY.t

&roup.

Allowln& substttutlo~ o[ a sc~ond Independent ~pcrator or en&incer ln c~s~.

o( R\\.IH lnopr.r:ihi l lty recocnlv'!S the C.lpabllltJ to adequately aonltor proper rod scq~~ncl~g in an alternate 11.:1ni1er without unduly rcsCrlct- '

Ing plilnt operations.

Above 20' power. thel'e l*.

no requlrencnt th."lt the R:,-:*I be opera~le si11ce the control rod drop accident wlth out-of-acquence rods will result in o peak fuel cncr&Y contt'flt o( lr.ss th:tn 2JJO cal.l&::t*

To nssure hl&h R\\lll 01v;illablllty, the !-~>::*: ls rcqurled to be opcrntlnc durln& a ~tartup for the wlthclr01w01l of 3 slcnlflcant nu~bel' control rods (or any st3rtup o(tcr Ju~e 1. 1974.

4.

• The Sourcf! Ranco tlonltol' ('SR.'t) systcQ perfonH no outom.,tic saFcty sy~tc~ functlon 0 i.e ** it has no scram £unction.

It docs provide the 62a

a

.J ll.

1 l

~

J

,,-~

\\. _,

opcr;~or v1th a visuol indication of neutron lc~cl.

thi~ is needed for kr.owledgeable and

~!iic~~nt rcactor~sto~tup ot low neutron level.

rr.c c~~~e~c~nccs of reactivity ~cciJcnts ore f\\!~Ct !.uns c-~ the in.lti.:Jl neutron flux.

The rcqukrc~~nt ~( at lc~st 3 counts per second

=s:*rnt-..*s th::t nn'J ~raa:;lent, should it occur~

bcr,liw ;lt ur above the inlti.al v:iluc of io-of r&J::cd j>c*.:cr used in the analyses of tr:msicnt9 frci~ cnlJ ccndf titm~. ' One 01>1.:rablc SIU*I channel

• .-ould be adequate to monitor the ~pproach. to c:-l: lr.:i! ~ ~;* :1s !r.;; hot!'lo;:e:ieous pntterns of.

sc&Jttc:-cd co:itrol roJ uJ lhJrn'l.11.

A ciinim1.:m of. t~u o~cr~bl~ S~M's or~ provlc.lcd es an added consct'v:it iia11.

The r.ud !Jloc:C. Monitor (R&t-1) is designed to auto-matic:!lly ;>rcvr.nt fuel Ja~nge in the event of erro:?cous t:od,,,.!thdt"a\\,;:11 from locations of high

'10\\.'l

r ccn!li ty durin~ hJ ch power level oper:ition.

'!'::o cl1a:mch arc prnvltlcc.I and one of these may be byp::s!:cJ (ro~ the console (or ll'aintcnance and/or tcsthr.. Triprlnc of one of the channels wlll block erroneous rod wlthJraYnl soon en~u~1 to prevent fuel d~1:0:1:: 0e.

11ils ~;ystc::1 baC:<s up the opcnJtor who wlth-

"r~r.. *-; rorfq ac!:ortlln~ to n written seq~*ence. The srcd.f icd restrictions wit)l one channel out of a.oorvlr<? con!>erv.itiv.:ly :1ssure th;it fu~l danmr.e

'-'ill r:(lt o:cur t!t:c to i*od withdr~t1:ll errors when t!1is condlt!on c::l~ts.

lu:1.-?nd1'1c11Ls 17/18 and 19/20 prcs~r.t ti1c tc!:ults r.f on ev<1luation of a rod block' r.:o:iitC't!' f:t!lurc.

'Che~;\\? omcnc.;ir.cnts show that during rc."!ctc-r oper:ition ".lilh certain lln1itlnc control roo '."'::ttcrr-:,_thc vltltdr:r~al or a de~i&nntcd sinr;le c'l-.~~<*t ro~~.Ht r,**:11lt in 1Jnc or more [1Jcl rot!,

vath i:C;>R:a i;;JtJ:'9W1'0I. t<C?R fu.,I cln<.iJlr.'I lnt*qrity 1af~ty ll*lt, Durlnq u&* nf auch f>.Jtll!;*;:~, ic 1s J*a1;;*.:i1 ll!:1t lt!:Jll!1g of the Rml sy:Hcr: ~rro:- to vltl11!r;:11;1l of such rods to assure its r:-1rr:!b i U :)* 11111

i~:;ure th~t improper with-Jr.:i;-:.11

'~oc'.l not ocr.cr. It is th~ re~pon!>ihl llty of.llw ::l:clc.ir t:1:)::cr.r to iJentify these 1J1:1ltlnc P~t :*!:-ns a~*J the d*:~* l1~n:i~cd rod9 cltlwr when the p:ittern~ l!TI!

~nltlally established *or as they dcv~l~? du~ to the occu~rcnce of inoperable control rods tn other than limitin~ paLLerns.

c.

'\\.

Scram Insertion Times The performance of the control rod insertion systl.'m is analyzed tu verify the system's abi 1 i ly to bring the re;1ctor subcrltical at a rate L.1st ennu~h lo p1*event violation of the MCPK Fuel Cladding Integrity Safety Limit an<l thereby avoid fuel damage:

The analyses demonstrate that if the reactor is operated within the limitations set in Specification J.5.~, the negative reactivity insertion rates associated with the observed scram performance (as adjusted for statistical variation in the observed data) result in protection of the MCPR safety limit.

In the analytical treatment of most tr,111~;ienrs, 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods.

This is adequate and conservative when compared Lo the typically observed time delay of about 210 millisecons.

Approximately 90 milliseconds after neutron flux reaches the trip point, the pilot scram value solenoid de-energizes and 120 milliseconds later the control rod motion is ~stimated to actually he~in.

However, 200 milliseconds rather than UOmilliseconds is conservatively assumed for this ~ime int~rval in the transient analy-ses, a~d ts ~lso included in the allowable scram 1nsert1on ti~~s soecifie~ in Soeci-f ication 3.3.C.

In the statistical treatment of the limiting transients, a statistical distribution of total scram delav is used rnther than the bounding value described above.

The performance of the individual control rod driv0s is monitored to assure that scram pc:rfurmance is nut de~raded.

Fifty percent of the control rod drives in the reactor are tested every sixt~en weeks to verify adequate nerfnrmanci>.

Observed nlant <lata were used to del.ermine the aver;1Pe scram :1erformance 63 used in the transient anslyses,

Scram Insertion Times (cont'd) and the results of each set of control rod scram tests during the current cycle are compared against earlier rest1lts to verify that the performance of the cnntrol rod insertion system has not changed signifi-cantly.

If an individual test or group of tests should be determined to fall outside of the statistical population defining the scram performance characteristics used in the transient analyses, a re-determinatil'll of thermal margin require-ments is undertaken (as required by Specification 3.5.K) unless it can be shown that the number of individual drives falling outside the statistical population defining the nominal performance is less than the allowable number of inoper-able control rod drives.

If the number of statistically aberrant drives falls within this limitation, operation will be allowed to continue without rede-termination of thermal margin require-ments provided the identified aberrant drives are fully inserted into the core and deenergized in the manner of an inoperable rod drive.

The scram times for all control.rods are measured at the time of e;1ch refueling outage.

Experience with the plant has shown that control drive insertion times vary little through the operating cycle; hence no reassessment of thermal margin r~~irements is expected under normal ~~~ns. The histor.y of drive performance accrn111i1 ated to date indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean,\\thich tends to lwcome skewed toward longer scram ti~es as operating time is aet;umulated.

The pn1bability of a drhe not exceeding the mean 90% i11sertion time by 0.75 RPrnnd i.R or,:>::itPr th::in 0,C}C}C} for rl normal distribution.

64

D.

E.

Control Hod Accumulator:;

The has is for this spccifil:;tl ion wns not dcs-crlhcd in the SAR and, lhcrdure, Is pre!'cnled in its entirely. Reriuiring 1111 more than one Inoperable accumulator in any nine-rod sriuare array is based nn a series nr XY PDQ-4 q1iarter core calculations of :t cold, dean core. The worst 1*ase in a nine-rod ilhdrawal sequcnl'c rc~ulll'd in a kcrr <I. 0 -- other repealing rod scqucnc.cs with more rods withdrawn resulted In kcrr **I. O.

Al reactor rn~i:;surcs In excess or :iOO psiK, C\\'cn those conl rol rods with ln-op1*rahle accumulators will he nhlc to meet rc-quircd scram Insertion timt:s due to the action or rc:-iclor pressure. In addition, they may he norrnal!.J,Inserted uslr\\g the cont rol-rotl-"dri\\'c hytl~tl.k.,. systcm. Proced111*a I control w i II asl:.~

1'"1'1Vlft'~onlt*ol rods ith inopcrahlc accu-m11l:ato1*s \\\\'ill he spact*d In :1 one-in-nine a1-r:-iy 1*:-it ht r 1 ha n ~l'OUflC'tl to15et her.

Heacllvlty Ano_malles DurinK each ruel cycle excess operating renc- *

( rlevlsed with Cl 1anges ?.7 and lf3 ls sued 1/29/7 ~)

G.

tlvlty varies as fuel depletes and aa any burnable poison In supplementary control le burned. The magnitude or this exces1 reRctlvlty may be Inferred from the critical rot! confl~ratlon. As fuel bumup progresses, anomalous behavior In lite excess reacllvlly

• may be detected by comparison of the crit~ _

lcal rod pattern selected base states to the predicted rod inventory at that state. Power ope rallng base conditions provide the most sensitive and directly interpretable data

• relallve to core rcactlvlly. Furthermore, using power operatln~ base conditions per-mits frequent reactivity comparisons.

Hequlrlng a reactivity comparison at the specified frequency assures that a compari-son will be made before the core reactivity chan~e exceeds l'l ~. Deviation& In core reactivity greater than t<<J, ~are not ex-pected and requl re thorough evaluation.

~ I Une percent reactlvlly llmll Is considered safe since an Insertion of the reactivity Into the core would not lead lo translent8 eicceed-lng design conditions of the reactor vyslem.

Economic Generation Control System Operation of the facility with the Eco~omic Generation Oontrol System with automati.c flow control is limited to the range of 65-*

100% of rated core flow.

In this flow range and with reactor power above 20% the reactor

• can safely tolerate a rate of change of load of 8 MW(e)/sec. (Reference FSAR Amendment 9"-

Unit 2, 10-Unit 3).

Limits within the Econo-

• mic Generation Control System and Reactor Flow control System preclude rates of change greater than approximately 4 MWe/aec.

When the Economic Generation Control System !

ig in operation, this fact will be indicated on the main control room console.

The results of initial testing will be provided to the AEC at the onset of routine operation with the.*

Economic Generation Control System.

---

T

~-----------------------------------------'""' ---....

J.S LlMlTl~'G CONDITION FOR OPERATIOM D. Automatic Pressure Jtollot Subsy1tems

• I. !xeept H 1poclfled In J.S.0.2 and S below,.

the Automatic Pressure Relief SU:,system 1hall be operable whenever the reactor

• pressure ls 1reater than 90 p1l1 and irrad\\ate4 l\\lel ii ln th* reactor ve11*l* *

2. Pro* and after the date that one of the flvo rellef vahu :-6(1.tbe* autor.iatlc pressure r~ller :subsyste:11 ls ~3c!o or found to bo inopernblo when tho* reactor h pressuriled ob,,~c 90 pslc..-lth irr:adlntcd !u'll in tho n:tctor vessel, rc11ctor opeution h pomlsslble only durlnc the 1uccccdln: seven doy~ ~nlest rcr3lrs arc ~3de and provided th3t durlnc such tlrr.o the 11rc1 Subsystem is operi.ble.

J. FroQ and crter the date* that Moro than one of flve.rtlie£ volv~s of tho outcmctlc pressure relief subsystt'r.1 r.1:1t1o or Councl to be lno;>cr~blo..-hen the re3ctor h prcssurl:~d above 90 pslg with 1rr:dl3tcd fuel ln tho reactor vesscl,-re3ctor oporatlon ls pet~lssl~le only ~u~ln& the succeeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> unless repairs arc mad~ anJ provided that durlnz such tlme the HPCI Subsyste*

ls oror:blo.

__,,_..~ I

~*-*~.-.,r; Amendment No.

4.S SURVEILLANCE REQUIREMENT D. Suncll hnco of th* Auto1Htlc Pressure

- Roller Subsyste* shall be perfor11ed H follows:

1. During each operatln1 cycle the followln1 shall be pcrfor11ed:

a. A simulated automatic lnltl1tlon whlch o~ens all pilot valvss, an4

b. Wlth the reactor 1t pressure each

• rcllef valve Sh3ll be *anumlly opened.

Relief valve opening *hall be verlfled '1

• a compensating turblne byp*** valve or control valve clo*ure.

c. A loglc syste11 functional test shall be performt'd each rofuellnc out*I**

2J. When lt ls determlnt'd that one relief valve of the aut~tlc pressure relief sub17stea ls inoperable, the llPCI shall be dc111onstrat..

to be operable lnnedlatel7 and weekly thereaft..

s. When it ls detort1lned that.ore than one

~elief valve of the *~to*atlc pressure relief subsystcn ls inoperable, tho HPCl subsystem shnll be demonstrated to be op~rabl* immedlatel) 71

---

()--------~-------------. ---------()~~~~~-------D_P_R~-2~S---------o~

J.S IJJUTINO CONDITXOR P0R OPERl\\TIOR 4.5 SURVEILLANCE REOUIR£MENT I. Average Planar LJICH During steady stHt.e pn1-rP.r npPrattnn, thr~

Average Planar Linear Heat Generation Hate (APLHGR) or all the rods in any l'ur*l n::::P.m-

.bly, as a functlun of averagr~ planar f:xposu e I

fur G.E. fuel and avPrage bundlr~ expoirn 1*e for Exxon fuel at any 1U<ial location. ehall not exceed the rnr.xt.z:n1m avcrn9a. planar UtGR nhown in Fir;uro l.5-1. If at any timo durin9 OPf'rfttlon it i* dctcrrnlnod by nonn~l sur-voil~cnce thnt tho lir:lit1ng vnlue for l\\?UICR is boin9 exceet1'1d. action shnll be lnitloted within 15 mlnutos to restore op~ration t.o Yi th ln tho pr-cAcr ibe<l l U\\ite.

If the APtJIGR in not roturned to within tha pro&cribod limit* within bilo (2) houra, tho rez:.ctor shall ho brouqht to tho Cold Shutdown condition within 36 houra.

SUrveill3nce and corresponding action ehall-continue until roRctor opera-tion lo vlthin the prescribed llmita.

hnendment It>.

I.

!'.Y~rn!J.o_~~~n_nr Lln-,nr llcat CcnC!ratlon lt"to {J\\PlJIGR}

,.he J\\PUIGR for e1u:h t:ype of f\\Jel a* a function of averRqe plnnar exp:>eure for G.E. fuel J and av1:!r::i1o;c l1undl1~ expo:~ure l'or ~xxon fuel shall 01~ d1!LP1*111incd dally during reactur operaticin at

• 2 ~~% rated thermal power.

818

3.5 LIMITING CONDITION FOR OPERATION 3.5.J LOCAL LHGR I

During steady state power operation, the linear heat generation rate ( LllGR) of any rod in any fuel assembly fabricated by GE at any axial location shall not exce~J the design value of 13.4 kw/~t.

If at any time during operation, it is determined by normal surveillance that the limiting value for LHGR for G.E. fuel is being exceeded, action shall be initiated within 15 minutes to restore operation to within the prescribed 1 imi ts.

If the*

LHGR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cnld Shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Sur-veillance and corresponding action shall.

continue until reactor operation is within the prescribed. limits Amendment No. 42 DPR-25 4.5 SURVEILLANCE REQUIREMENT J.

Linear Heat Generat.ion Rate (LHGR)

The LHGR shall be checked dai1.y during I reactor operation at 2 25% rated thermal power.

81B-l

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3.5 LIMITING CONDITIONS FOR OPERATION K.

Minimum Critical Power Ratio (MCPIO During steady state operation at rated co~~

f lo~, MCPR shall be greater than or equal to -

Unit 3 1.30 (All fuel types)

For core flows other than rated, the MCPR operating limit shall be as follows:

1. Manual Flow Control-the MCPR Operating Limit:*

shall be the value frum Figure J.5-l sheet 1 or the above rated core flow value, which ever is greater.

2. Automatic Flow Control-the MCPR Operating Limit shall be the value from Figure 3.5-2 Sheet l, Sheet 2 or the above rH~ed core flow value, whichever is greatest If at any time during steady state power operation, it is determined that the limiting value for MCPR is being exceeded, action s~all be initiated ~ithin 15-minutes to restore operation to within the prescribed limits.

lf the steady state MCPR is not.

returned to within the prescrihcd limits within two (2) hours, the reactor shall be brought to the Cold Shutdown conllition I

d within-lQ...hours.

Surveillance and correspon ing action~tra't'l'"'continue until reactor operation is within the prescribed limits.

In the event that thr control rod ~cram time results of specification 4.J.C.] fall outside the distribution usell in the transient analyses, the MCPR Opernling Limit will he incre;-1:-:;cd as specified hy the nuclear fuel V1~ndor if rl!£1llired to maii1tain adequate mat*gin to the MCl'R Safety Limit.

DPR-25 4.5 SURVEILLANCE REQUIREMENTS K.

Minimum Critical Power Ratio (MCPR)

MCPR shall be determined dail~ during a reactor power operation at ~257. rated thermal power and following any change in power level or <listributio~ that would cause operation with a limiting control rod pattern as described in the bases for Specification 3.3.B.5.

Amendment 34 d~ted 5-3-78 BlD

1.5 1.*

II -

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~ 1.Z 1.1 1.0

--.*-;I.

~'*-*'!"""~

orn-25 30 40 50 60 MANUAL AND AUTOMATIC FLOW CONTROL FLOWMAX '117\\

0 80 Total Core Recirculating Flow (S Rated, 98 mlb/hr)

Pl gure 3. 'r.2.U)heet l of 2)

~CPR Llmi t for Reduced Core Flow 90 100

1.7 AUTOMATIC FLOW CONTROL ONLY 1.6

~

at 1.5

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1.l.,._ ____.,._ ____ *-----4-----f-----1 JO 40 50 60 70 60 90 Total Core Redrculating Flow (I Rated, 98 ml b/hr)

Figure 3.5-2 (~heet 2 of 2)

NCPR Llmlt for Automatic Flow Control

I\\.

~.orr. ~Ot:_<1Y nnd r.rc_x~ -~~cl~-~f th_r.:_TI_!!'!

~tcm - This 9pcc i ( icat ion rrnsurc!I thnt adcqu~tc cmcrqcncy cooling c11pobility is ovoilnblc.

BRsed on the loss of coolant analyses included in Refcr~nccs (1) and (2) in nccordance with lOcrns0.46 ;md l\\ppcn-dix ~' core coollnq systems provl<lc suff icicnt cooling to the cor~ to dissipate the energy associated with the loss of coolant accident. to limit the calculated peak clad temperature to less than 2200°F, to assure that core geometry remains intact. to limit the core wide clad metal-water reaction to less than 1%, and to limit the cal-culated local metal-water reaction to less than 17%.

The allowable repair times are es-*

tabllshed so that the.avera9e risk rate for repair would be no 9reater than the basic risk rate.

The method an~

concept are described in Reference (J)._,Usinq the results Cl) *t.oss of Coolant Accident Analyses Report for Dresden Units 2, J anrl Qu~rl-Citles Units 1, 2 Nuclear Power Stillion~,*

Ntmo-241461\\, llevisionl,* l\\pcH 1979.

Amendment No.

0 e

0 I

• ?.

Should Ont" <'Or<" PJ'l':'~* iaub:o~*ttlc-m IN-C'nm* ln-opc rable. lh* r*m:1lnlng C'Orl" !ft':-:1\\" !'Incl lhl" enllr~ LPCI !1)"5l('m :art" :1uil;iblt' ~hould the (2) NED0-205G6. General Electric Company Analytical Model for Lose-of-Coolent Analysis in Accordance vith 10CFR50 Appendix K.

(J) APEo-*cuidelincs for Determlnln9 S~fc Tc5t Intcrvnls and Repair Times for Enqincercd S~(cquards* -

April 1969, I.M. Jacob~ an~

P.W. H:irriott.

{ll) X N - N 1;*-H l.-'( 1i "Uri *sden Un.lt 3 LOCA Model U!*; i ng the r*:NC EY.F:M Evaluation Model MAl'LllGH Hesul t!:;"

82

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Q.

J.5 t!.~ ~!n'. C~Hlon tor. OP?l':\\tlon Dn:1C?~ (C~:1l 'dl I. l.ve=ero Plr:.r:t.r 11;r.n.

r f z

0.

Th1D s;-ec!flca~ion e..eour.c:J U\\!lt tho l'(J!\\k

• c1A~dl11~ tcr.tp::rnturo folloutne a i:eolt:fat~

doclr,n b~:110 lo::n-o(-cool~nt ncctdonl "lll r.o~ c~c~cd tho 2200'-T 11~\\t opoclf1c1 in l<X:rR~O App:r.d1~ K 'ccn:>\\~ertr.e tho po::lul4te4 ortcct:1 or fuel F:>llot dc:tclflc:l!.ton, Tho )."l\\k dodtlnc to11r.nrn.luro follolfln~ a p3~lulclcJ los:1-0C-ccol.nnl cccJ,lcnt 1 'l

.1=rl11:.rll:r a runcllc,:t or u,., R\\""1*:*.~;3 1~:1.n or nll Ut:> rM!: 1n a fl!ol c~r.~:-*l>l1 c.! r.ny u.lol loc!\\llon ur:d lo 0.111 cl*~i:::?:dcn.t !!?cor.d-1\\r!ly 0:1 tho roJ to red r:::~:':.lr d1oll.~b.:l1on vltMn a f\\!cl O!>:.:c~bly.

Sine!> r.xpcle1 local

.,:rl!.tlC:'\\:J ln po1::r dlLlrlliuUon uHMn n fu3l ~~~o:i.bly a.ffccl !h'.> c~lcl:L"\\t.r.d r::ilt ol.n4 tonr.n~tu~ by lc=>:J thnn i2oc;-

r.113l1':~ to the ~ek \\C!l!f.~1'3l\\:ro for I\\ trrlc:!l ru~l design, th~ 11:-Jl\\ on tho 0*1or~~o rk!\\... r ua;n io su!tlcl'-ni to l!!l:ur.! th3l c~lculctcd t~~P*

c:rotur~:s M"9 bolc.w tho tccrn50, Apr-uudlx K llri.U.

The maximum average plannr LHGRs shown

ln Pigure 3.5.1 arP brn;r~d 11n calculations

. employing the models dp~;cribed in

• '.Reference (l) and in rcl'r:rr!nce XN-NF-f~l-75 *

rower op~ration with Ar'l.ll!:Hs at or below

. thm;~~\\jn in F'ig. 3. '.5. I w,;t;ur*'G thn t.

• the peak cTadding tr!mpet*ature fnllowlng a postulated loss-uf'-cuolant accillent will not exceed the 2:!CJ0°1°' lirni L.

(1) "l.us9 oC Coolant l\\ccirlent l\\nalyges Report for Dres~en Units 2, ) and Quad-Citles Units l, 2 Nuc~ear Power Stillions," NE00-241461\\, Revision 1,

/\\pr1l, 1979.

/

r:'"l.

u

'I'he maximum average planar LJIGRs for G.E. I t.°UP.l pl1itt*!d ln l*'ir,. j.').l at hir:h1~r P.x11n::ures re!;ult in a clllculb.t~d r.c:il< cl.1.d tc1:1r.cnt~ or 1-:::-~

tt::in 22ooor.

llc.~*cvcr tho ru:da:\\!:. av"=~

Jllnn:ir UIGRo nro oho:*n on Flc. ).,5, l e:.

ll~ll!J b=c~u~o conror.:-o'lnr.o c~lculnt1c~3 hsTe net b:cn ~~rrorr::d to justlf7 opor~tlcn ai Utcna 1n OXCC09 or tho:Jo ahour..

~

• Loc!\\l ll!t;J\\

Thlo opcclflcnUon ll!IOU?"S thd. tho

• rcxlmun llnc3r hcnt ~cncl'3~1on re!~ 1n atty Cw,*l r1*d L";ihr.icatccl t1y li.E. is lei..;s than the dei..;ign linear 8$A l

.5 Limiting Condition for Operation Hnses (cont'd) heat gener~tion rate even if fuel pellet den-sification is postulated.

For fuel fabricated by ENC, protection of the MCPR and MAPLHGR limits and operation within the power distribution assumptions of the Fuel Design AnRlysis provides adequate protection against cladding strain limits, herice the LHGR limitation for GE fuel is unnecessary for the protection of ENC fuel.

The steady-state values for MCPR specified in the Specification were determined ~sing the THERMEX thermal limits methodology described in XN-NF-80-19, Volume 3.

The safety limit implicit in the Operating limits is established so that during sustained operation at the MCPR safely limit, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The Limiting Transient f!l CPR implicit in the operating limits was cal-culated such that the occurrence of the limiting transient from the opera~in~ limit will not result in violation of the MCl'R safety limit in at least 957. of the ranJnm statistical combinations of uncertainties.

Transient events of each t:ype ant ici pated during operation of a BWR/3 were evaluated to determine which is most.restrictive fn terms o-f* therma I. margin r.ef~uirements. *The gen~n1iFt~t4oa<l reject ion/turbine trip without Lypass is typicall.y the 1 imiting event.

• The thermal margin Pffect:s of the evt*nt. are evaluated with the TllEKMEX Methodology and appropriate MCl'R l.i.mit s cons is tent wit: h the XN-J critical power co1-relati.1111 ;11*e de term i.ned.

Severa 1 factors influence which transient: results in the largest:

reduction in critical power ratio1 such as the cycle-s~ecific fuel loading, exposure and ftwl type.

The current cycle's reload licen-sin~ analyses identifies the limiting transien for that cycle.

As de~cribed in specification 4:,.C.3 and the associated Bases, observed plant data were used to determine the average scram perfor-mance used in the transient analyses for determininR the MCPR Operatin~ Limit.

If the current cycle scram time performance falls outside of the distribution assumed in the analyses, an adjustment of the MCPR limit may be required to maintain margin to the MCPR Safety Limit during transients.

Com-pliance with the assumed distribution and adjustment of the MCPR Operating Limit will be performed as directed by the.nucl~ar fuel vendor in accordance with station procedures.

For core flows less than rated, the MCPR Operating Limit established in the specifi-cation is adjusted to provide protection of the MCPR Safety Limit in the event of an uncontrolled recirculation flow increase to the physical limit of pump flow.

This pro-tection is provided for manual and automatic flow control by choosing the MCPR operating limit as the value from Fi~ure 3.5-2 Sheet 1 or the rated core flow valve, whichever is greater.

For Automatic Flow Control, in i!d<lition to protecting the MCPR Safety Limit during the flow run-up event, protection is provided against violating the rated flow MCPR Operating Limit during an automatic flow increase to rated core flow.

This protection is provided hy the reduced flow MCPR limits shown 1n Fir,ure 3.5-2 Sheet 2 where the curve corresponding to the current rated flow MCPR limit is used (linear interpolation between the MCl'R limit lines depicted is permissible).

85b

~.5 Limiting Condition for Operation Hases (cont'd)

Therefo~e, for A*1tomat ic Flow Control, thl' MCPR Operating Limit is chost*n as the va l.ue from Figure 3.5-2 Sheet l, Sheet 2 or the rated flow value, whichever is great.est.

It ~hnuld be noted that if the rat eel flow MCPR l.imiL must be increased due to degra<lati.011 of control rod scram times during the current cycle, the new value of the rated flow MCPR limit is applied when usini Figur~ 3.5-2 Sheet 2..

~1,,

......,lf"lt',.te".,

858-'

x.

A...,rece Pleft!l:O LllCR.

At. core thorml power te*ei. leee than ex:

eqcal to 2S.,er r.cnt, opcrat.lng pl.Dnt.

cxp-?rlcnce nnd thermal hydraulic: nnal711e*

lr.dlc~te lhot the rc:sulllnc ovcr:1gc pl.Dn:sr LilCR h belou the n>xlnnm ovcrnce plomr LRCR b7 n con*lder3ble rrorstn; thcrcCorc, cvelu*t.10D or t.he nvcrece plenar LllGR below thte povcr*

leYel ls not ne1ee11s~r7. The c1oll7 rcqulro-mnt. for clDculoUng...,erace planar LllCR abo<We 25 pitr cent r3 lec1 the 1"11131 power Ill 1ur:-1,1cn~1ln *.

po~cr 4hlrlbutlon ahl~~*

ere lit~ u ten... re have not been.icnttl-.

• cant. poue. or q.-trol rocl c,.ncce.

,,.. ~j: *tr,('.

11

!.ot:al t~IC!t

1

'11-.c! U:CR f(lr G. F.. fuel shall b~ ch"!c1Mc1 c1n0,7 4urtng re*etor operation et creeter.thon or eauol to 25 r-r cent po11er to cl~ terct "-' H r*.:'! 1 bur nu;> or cont.rol r;,! 1:ovenent

~~' c*u:sea ehr.n~e ln po-Jcr dl!ltrtbutlon.

A llmltin9 LHCR value is precluded by a considerable margin when employing a per

• ls.elble conttol rod pattern below 2S\\ rated

~~~,.+ pove r

• Amenctnent No.

.. i I i' *;!,~

l. Mini-Cdtlcol "'"' Aotlo !i9(lll :. '

1 i *:-r.::1* I,,,

At core thenul power Inell

• 1.~1 t~ ~!'.~.\\~*'. {:1JI to 25 per cent, the reactor vlll.9'e ~r.llln "* ; ~~~ l~~

~t ::1lnlr.nc recirculation.F.-, spt... pnd tflo.~'. '"/:,,;

• • ~

11<>derator vold content wll l be **i7 s*ll. *~l 1,-, ~1*f al I dcs11nated control rod patterns.Jhl~~llli;t ~:\\ :' :t*l ~

e111ployed 11t this point, operet~fti*"pl~t **,.~...C.*. *r, Hd the111al hydraulic ***rsi1 *iftdtcltet thair.tlljit, 14' ~

rcsul tln1 MCPR value ls in oaceu of, ;.,.11'~ :* :1:1.'

by a considerable aar1tn. With tills IOll wtcl'. ; * *' I'

content, any Inadvertent core no.I i~cr***.. ;.

would.only place operation in a 110re*con-

• ~

senatlve 110de reletlve to NCPR.

.,.. L '.*!'..

.1.n i' I*

. '. if" 1"e dally requlre.ent for calculetlna*

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MCPR atiove 25 percent rated thenal'.

• *.., '.;t;U~J ~*

0111,¥,

powor ts sufflclont since power dbtrlb~lelil"i(I". :\\t * *

~':;;l shifts an very slow when there hDYe not hen'..

sienUlcant power or contrvl rod chanaoa~,..,{".".'.!'~.; J~.*i

. i

.... ("' **~*'-t.1'**1fl1 *,..* i In addition, the I correction ""lied tJ.(:.::t*H~.~'.**. r:J.

the LOO provides.. Sr1in. for flew* Jncpaie*;. i., '.~*.;, :.:.tt frOll low flows.

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