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::. No.

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

• ~

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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

• 85B-l 42 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 c. Critica! Power -~~0-~~) - Tlw c*ritical of the specificatione may be achieved. power ratio is the 1*at iu of that as~embh*.

power which causes some point in the .

A. (Deleted) 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 B. Alteration of the Reactor Core - The act of to verify the proper instrument response moving any component in the region above the alarm, and/or initiating action.

core support plata; below the upper grid and within the shroud. Normal control rod move- H. Instrument Check - An instrument check is ment with the control rod drive hydraulic qualitative determination of acceptable oper-ahil ity by observation of instTW11ent behavior ayetea ta Dot defined ae a core alteration. 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. ~l~l~*l!'U~"~! ~!"~- !!"_ 0p1_=r~1 ~nn_Jl.r.nl - 1'**  !:'r!!~blo -

• ar*l-. **b*J!tl .... , l***I '* c..,....."'* or ***Ice hai, *"0: ccmdll 11111s tur 0111:1.ll 11111 s1*cc1 fr lhe "'**II "" 81'e**ehl* ..1oc:n It lo c*fM'bl* .. r P*rfe*wh.. Ile

. . . , . . . accr:l'l*l*lo levch ol *t*~lt.* 1*e1t111*- *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-

-

• Rece111uy 10 assul"e 1;.fc sl.ul*tt* onJ ut** ..al11.n, c-trnl*, nonoonl or.ti *-!"l!""'*.Cf *l*ct1*lc*I povor el"allo11 of lhe f*clU1y. ll"hcn 1lu:so cn11oli1*nn1 1101u-co**. Efl"lhlfl or """' ";\lor, hohrlc*llon or other

• • • - * . the rla11t c.a" be .. ,*., .. atcd s;afr:ly and M**l I l**'J * .,..1.,ocnt th*l **** rC:IJ'tlrr.*I for. lh* *rat ...

.._,***I 11111.11 Ions c.an ho safely contrnl led. """'"'**** 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 J. ll*l*l*g_S~fc*r_~r~~r:~ ~***tln1_l~_~c;_c;). - 11.e Sa[~ty tytlr:* llettl~tll aro )Ctllllf!I On fl1**r10\ln.11 - C'r****t.l"I "e*n* that * ***t....,

ll*illnc llhl "*cnl*t '"""""'ch 1111t11110 tho ;1111011.:t le peulcct he act Ion at

• lcv.,I s11ch that th* safe*r

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

• 11t:1ratea.

l!*lls ul 11 nut ho eacc.:1lt;1I. lhe u:clon Q. !"'r~~.!.~"I ~tel! *lnlc:rHI bc:twee* tho e . .

. . twee* tho 1efe11 ll*lt and thc:so '"" tn1s .,:* one u:T11elln1 onl*10 an*I eho end of *h*

~ .. *~u:nts .... , ... wllh 11011ul nrerallon ., .... ne*t suh1c11ucnl refuol 1111 o~ta1*.

. . , _ 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 ** Pl"l*~rr Cnnt*lr.ncnl l11te5rl*t - Prl**rr con"t"ilnl.;;-nnn.e1-..-fli*cans *"** lh* drywall ea<e.:JcJ. and rrr.tt*1ro ..... r .. 01slnn ch .... ., . . . , . lnlat:t and nil of lh* followln1 con1lltlon1 are 1atldl . . 1 K. fi'rac ti on For le r .ca ec y C:E,

• 1e rac

• on I. All aanual cont*h*enl holal Ion valwe1 **

of limiting power tlensity ts thr. rat.i.o I i11e1 conned Ina lo th* re*ctor coolant *JI*

of the Lin1~ar Heat Generatitin Hate (LIIGn) t~* or containment w~lch are nol refl'llr**

existing at a given location to the lo bu open durln1 accld**t con*ltlons are design LHGH !\1r tltat bundle type. F'LPD t:losc*I.

does not apply to ENC fuel.

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

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 J, All 11utu.atlc contaln*ent hohilo* walwes

• • eel lwalc*I Jc:vlco to lniuro al I cosi1ionc:nt1 en: 01*cr.1ltle or dcact hated In lhe holot . .

• r* orcr11lelo 1*c:1" Jc,! \:*1 l11tcnl. N11eu1.1*01tl* r"sh Ion.

• • • , 11c1 Inn " i 11 ro 10 cm*plct*lon, L*., p*-r* 4. All lillnd Uu1** and *11nv*r* ere cles ...

-..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 1c-.:;*,-1..... ~ .. , .1 t To s. --.---**--.-- .. -----------

Protertlvo lnHni;a.1flt*llon lleflnltlons c*ncsl'on*Un1 to 1ho *ost I l1Jll ln1 fual a*sc:.a.1, l* tl1e core. I. lnttl"Ul8t;nl 'h*~*el - An ln,lrUlllcnt ch0111-ncl **,..n, an .,,.n1c*cnt or

• tensor .....

11. llode - l1'e reactor"°* 11 that 11hlr.h ls a111lli;ory e:tui~r:nt 1"cc1uhed lo aenorat*

eii.il.* ........., ttle e:>Ju-seloctor-1vllch. anJ 1rnn1*ll t~ *trip *11tc*

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

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

J

---

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

, .i 1.1 SAFETY Lttirr 2.1 LI'1UINC Sl..rrl'T SYST~~. SEiI'l.~C

... 1. APRM Flux Sera"' Trlp Settlns {Ren "odel

,.*,-I \."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 10 6 lb/hi' and greater, Wh8r-!I S

• set.Ung 1n per cent. or rated power V0* 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)

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

ff'RP ]

Where; S ~ (.58W 0 + 62) LMFLPD FRP = fraction of rated thermal power (2527) MWt)

MfLPU = maximum fraction of limiting

-* power density for GE fuel I 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.

l\mendment Ho.

1.1 Safety Limit Bases FUEL CLADDING INTEGRITY The fuel cladding integrity limit is Safety Limit is defined with margin to the set such that no calculated fuel dam- conditions which would produce onset of trans-ages would occur as a result of an ition boiling, (MCPR of 1.0.) These conditions abnormal operational transient. Be- represent a significant departure from the cause fuel damage is not directly condition intended by desi~n for planned observable, a step-back approach is operation. The MCPR fuel cladding integrity used to establish a Safety Limit such Si:lfety Limit assures that during hormal that the minimum criLical power ratio 01wration and during anticipated operational (MCPR) is no less than the MCPR fuel occurrences, at least 99.97. of the fuel rods cladding integrity safety limit. in the core do not experience transition MCPR) the MCPR fuel cladding integrity boiling. See reference XN-NF-524.

safety limit represents a conservative margin relative to the conditions A. Reactor Pressure > 800 psig and Core required to maintain fuel cladding integrity by assuring that the rucl flow > 10% of Rated does not experience transition boiling. Onset of transition boiling results in a decrease in heat transfer from the clad The fuel cladding is one of the physical and, there*fore, elevated clad temperature barriers which separate radioactive a~J the possibility of clad failure. However, materials from the env(rons. The the existence of critical power, or boiling integrity of this cladding barrier L~ansition, is not a directly observable is related to its relative freedom parameter in an operating reactor. Therefore, from perforations or cracking. Although the margin to boiling transition is calculated some corrosions or use related cracking .from plant operating parame.ters such as core may occur Juring the life of the cladding, puwer, core flow, feedwater temperature, and fission product migration from this coie power distribution. The margin for each source is incrementally cumulative* fuel assembly is characterized by the critical and continuously meas~rable. Fuel power ratio (CPR) which is the ratio of the cladding perforations, however, can b1111dle power which would produce onset of result from thermal stresses which. boiling divided by the actual bundle power.

occur from reactor operation sip,ni-* The minimum value of this ratio for any ficantly above design conditions and bundle in the core is the minimum critical the prot-t:tion syst:l:*m safety settir)~S. pnwer ratio (MCPR). It is assumed that While J:.tD+RJl product mi~n1Lion frum thl' plant operation is controlled to the I

cladding perforation is just as measurable nominal protective setpoints via the as that from use related cracking, the instrtimented variables. (Figure 2.1-3).

thermally caused c 1 acl Jing 1w r. for a 1* ion signals a threshold, hl'ynnd which still The MCPR Fuel Cladding Integrity Safety greater thermal stresses may cause gross Limit assures sufficient *conservatism in the rather than incremental c Laddi~g dd.l!L*- operating MCl'R limit that in Lhe event of ioration. Therefore, the ful'l ~ladding an anticipated OfH'rat ion al occurrence from the limiting condition for operation, at AmPnrlmPnf- Nn. 10

I

~,

I .~.,~

. . ~

e

~

OP!\-25

. 1.1 °S.Vt;TY LTI-11T 2.1 Lll'ilTI:iC Slt-&.rt STSTE!t smn-c J. C?re Th*~Mn1'1ovrr Lf*lt (R~actor.

• ~

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

\lh*n the rtactor rressurc is < 800 The U" flu* *ct*"' ISt!ttln!: shall be set et hu th*n or equal to 120/115 of p!I i & o: core flow h h u th~ 101. . full scale

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

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

l. The neutron flux shell not uu:eed the scnai  : I s~tttng tst*bllshed lh Sptc1f1c~tion. 2.1.A for lo"gcr than l. 5 eeconds as 1.ndlcattd by S ~ [.58WD + 50]

the proc*** co"'Puter.

. The de flnl tlons used *bove for 'the APPJt

2. '""*n the proc*ss coep~ter *i* out of set"Vlce, sen* tr1P1?ply.

this 11fet1 11 .. u shell be 8.S:J\#M41 to b@ ln the event of operation of any fuel assembly e*cteded l{ tht neutron flUI( ~*cttd..-: the ~or:un fabricated by GE with a maximum fraction setting ~stabllshtd by Speclflcetlon 2.1.A limit in1~ power density (MFLPD) greater than and

• control rod set~* dots not o~cur. the fraction of rated power (FRP), the setting shall be modified as follows: l

_, S ~ (.58W 0 + 50) ~~~p~

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

D. A*~et1ij!e,t.*r ( .

tevelShu\d~~n Conditlo~

)

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

"'°'""ev.r thf! uactor h In the shutdcvn condltlon with lrradlAteJ 1~@1 In the rtector v*s~el, the The ratio of FRP to MFLPD shall be set equal

~At*r Jtv*l ~hall not be Its~ then that corrt~ to l.O unless the actual operating value ronAl"ft to 12 lnrhPS ~bove the top of the Active is less than 1.0. Jn which case the actual fu~a*~htn lt is S~At~d ln the cor~. operating value will be used. The adjust-ntl'nt may al.so be performed by increasing

• Top of a~t,ve fuel is defined to be 1 the Al'RM gain by the inverse ratio, MFLPD/

JfiO t nch1*s above vesse 1 zero (see fRP, which accomplishes the same degree of l l*t"l"l

' 'I ~'

.l

.) *

'( .) )

• protection as redu<:ing the trip setting by l."IJ I> /Mi;'f IJf) 7

1.1 SAFETY LIMIT 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.

Amendment No. 6A

.e 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 Integrity Safety Limit there Would be no trans-core would be ~xpected to avoid it ion boiling in the core. If boiling transition boiling transition. The margin between were to occur, however, there is reasbn to calculated boiling transition (MCl'R=l.00) believe that the integrity of the fuel would and the.MCPR Fuel Cladding Integrity not necessarily be compromised. Stgnificant Safety Limit is based on a detailed test data accumulated by the U. S. Nuclear statistical procedure which considers Hegu latory Connniss ion and private on~anizations the uncertainties in monitoring the indicate that the use of a boiling transition core operating state. One specific limitation to protect against cladding failure uncertainty included in the safety is a very ~onservative approach; much of limit is the uncertainty inherent the data indicates that LWR fuel can survive in the XN-3 critical power correlation. fur an extended period in an environment of Refer to XN-NF-524 for the methodolpgy transition boiling.

used in determining the MCPR Fu~l

• Cladding Integrity Safety Limit. If the reactor pressure should ever exceed the limit of applicability of the XN-3 The XN-3 critical power correlation critical power correlation as defined in is based on a significant body of. XN-NF-512, it would be assumed that the practical test data, providing a MCPR Fuel Cladding Integrity Safety Limit high degree of assurance that the had been violated. This applicability pressure critical power as evaluated by the limit is higher than the pressure safety limit correlation is within a small per- sp~cified in Specification 1.2. For fuel centage of the actual critical power fabricated by General Electric Company, being est ima.ted. The assumed**

• or~~ration is further constrained to a maximum reactor conditions used in defining linear heat generation rate (LHGR) of 13.4 kW/ft the safety limit introduce cons*:. v- by Specification 3.5.J. This constraint is at ism into the limit because established to provide adequate safety margin boundingly high radia 1 power p;eaking to li. plastic strain for abnormal operational factors and boundingly flat local transients initiated from high power conditions.

peaking distributions are used to Specification 2.1.A.l provides for equivalent estimat~ the number of rods in safety niargin for. transients initiated from boi~ .. tll'.'ansition. Still further lower power conditions by adiusting the APRM con~'V.Jrtsm is induced by the flow-biased scram by the ratio of FRP/HFLPD.

tendency of the XN-3 correlation Specification 3.5.J establishes the maximum to overpredict the n11mber of rods value of l.HGR wh i.ch cannot be exceeded during in boiling transition. These

• steady power operation for GE fuel types.

conservatisms and the inherent accuracy of the XN-3 co~rclatiori For fuc*l fabricated by Exxon Nuclear Company, provide a reasonable degrc~ of (ENC) fuel design cl'iteria have been established assurance that.during s~slained

• t(l provide protection against fuel centerline operation at the MCPR Fuel Cladding. 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 2.1 L1mtttns Sof~ty Sl_~tcm Sctttng 8~~c 3 1.1.C ro~cr Tran3lcnt (cont'd) FUF.L CLADDING INTEGRITY Th~ co~putcr provtdcd han e The.cbnormal operat~onal transtcnts

~e~ucnc~ <:nnunc~nt1on pro~ram',,;hich nppllcable to OP"!r~tton of the Un1ts

~ill tndtcetc the ncqucnce tn "'hich h~ve been 2n2lyz~d lhrou~ho~~ the ocr~ms occur ouc~ e3 neutron flu~, opcctrum or plonned oper~ttn~ con-p~~:;surc, etc. Thi~ pro~rC!m al:;o d l t lons up to the r~ted thcrm1l power 1~*Jlcct~3 *,:hen th~ scr~m s~tpo1nt 13 contl 1t lon of ~r;-**7 l*:Wt. ln odd 1t ten cl~~r~d. This "'!11 prov\1c lnfor~:it1on 2'j~'/ i*i'Jt 1s the 11cc113cd 1na:t!:num st~ad,..

on ho1*; long a scr;.1:n coml1t1on <'llt:;l.s st~te po~er level or the unlts. This

~nd thu'.; provJdc ~omc rr.c:i::turc of lite m~xlmum st~ady-st~te po~cr level ~tll cn~r::y adtll"d dt.:rtn~ a tr;!n:;t':!nt. 1'!1u3, never kno~ln~ly be cxc~cdcd. See referencel r.-::>o..01~tcr information norm:illy "'*Ill be XN-NF-79-71.

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 lJ

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

• -. 11.iscs J.21.

• Amendment No.

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

- Conservatism is incorporated into the For an3lyses of the the1111al consequences of the tronslcnts, tho MCPR's stated *n p3r*ir3ph transient analyses which define the J.s.i ,, 5 th~ ) imit inq cnnf1lt1011 or opr.rt'ltlon MCPR operating limits. Variables ~hich l und those which are consP.rvatlveJy asnumed inherently possess little or no ~~ exist prior to initiation of the transients.

uncertainty or whose uncertainty has little or no effect on the outcome A. Neut*ron Flux Tr1p Sett 1nsr:s

~

of th~ limiting transient are selected at bounding values. Variables which 1. l\PRM Flux Scrun1 Tr1p Setttns_JRun Mode) possess significant uncertainty that The ever2ge power r3n~c mon1tor1ng may have undesirable effects on thermal (APHM) system, whtch ls calibrated margins are addressed statistically. ustng hc~t b~lancc d3t~ t?ken durln!

Statistical methods used in the 9t:!ady-stutc conrl1t1o:-i:J, reads in transient analyses are described in pcrcc:lt or r3tcc.J. thcr~nl power. 8c-XN-NF-81-22. The MCPR operating cau3c ftes1on ch~~b~r3 pro~:jc the baslo limits are established such that the tnput ~ t~n:? l:J, t~~ /,Pni*l s1st~::1 rC'sponds occurrence of the limiting transient *d t rec tly to ~\*cr~~.c :ieutron flu~.

will not result in the violation Dur!n,,. trnn3lc:1t3, the ln3ta:i\.aneous of the MCPR Fuel Cladding Integrity ratr.* ~f heot tr;;n:;fcr fror.t the ruel Safety Limit in at least 957. of the ( rcac tor thcr;11a l po\*;cr) 13 lc3:J than random statistical combinations the ln~tnntnncou3 n~utron flux due to of uncertainties. In Reneral, the the ttm:? con!itnnt or the rucl. 'there-variables with the greatest statistical fore, durtn~ nbnor~nl op~~t'ltlonnl significance to the-consequences of trnns tents. t.hc therm:> 1 po~cr' or the antici.pated operational occurrences are fuel ~111 be le3~ thJn th3t lndSc~tcd the reactivity feedback associated with by the neutron flux at thr. scrzm :ietttng.

the formation and removal of coolant An~ly~c~ d~mo~~tratc that. wlth a 120 voids and the timing of the control pcrc~nt 3~ram tr!~ 5ct~ln~. non~ or the rod scram. 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

• Steod~-~tote operation wtthout ~orccd :tc!"."!:1: i;.1*J1* pro\".ldc:: C\'~r. nd*J!t.l!'l:1-:l ~::r~:n.

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. 14 Amendment No.

, . reducing the trip set~i~g by FRP/

MFl.PD by raising the 1n1tial 11.A. Neutron Flux Trip Settings Al'RM rcac,tlng closer to the trip settinr such that a scram would be

1. APRM Flux Scram Trip Setting receiv~J at the same point in a (Run Mode) cont'd) Lr;insi.ent as if the trip setting An increase in the APRM scram trip setting woul~ decrease the ma~gin had ----,,_________ _

h~'1*n reduced.

APP.~ ttui 9ota~ Trsp Settln!

present before the.f':1el.cladd1ng jf1 .. !'u~l.al" Stl'r~ & Hot S~t~1db7 Mode) integrity Safety Lrnnt is reached.

The APRM scram tri.p setting was t&f' 81'f1Jl'tUon th the *tnrtup 1110de whtle determined by an analysis of mar.gin~ lhf re1ator ti et low ~re1,urt, the APRM required to provide a reasonable . t11:f'1111 eettlng or 15 perc~nt or rated power range for maneuvering during operation~ pro~t~~* lda~uatt ther~3l Mnr~tn between the Reducing this operating margin.would ~h' 9~~po1nt er.d the 1ntctr 11~1t, 25 p!r*

increase the frequency of spurious nut ot r:1 tcd. Tho margin h ede-1uetc to scrams which have an adverse effect *e~~!lt'Tl,dote enttctpotcd -rn~neuvcrs *oococtated on reactor safety because of the ~~~h "o';~f' .,hnt 1tortup. £rreots or tn-resulting thermal stresses. Thus, i~c~'sn~ pr~ouur* et 1c~o or low *voSd con-the APRM scram trip setting was .  :~a: ttr-c 1dtaot*, eeld ""t~* rror.t 10.-rce~

selected because it provides adeCJuate

,..,.*,:lrt,!1 d*::-!~!~ :;urtup 1:1 riot much _colder margin for the f uc.: 1 cladding in ~egr it y than that ~lready in the svstem, tempera-Safety Limit yet allows operating ture coett1ctent!r-"Bre emall, end oon-margin that reduces the nossibility t.rol rod patterns 11:-e constrained to of unnecessary scrams. be.uniform by O?cret1r.~ procedures b~ck~d up by the ro~ ~orth mtnl~lzer.~

The scram trip setting must be or ~11 possible so~rcc~ or rencttvtty adjusted to ensure that the. LllGR input, unsro~m control rod wtthdrawal transient peak f,,, G.E. fuel ts the most probabl~ cause or a1gn1fl**

is not increased fur any combin- c~nt po~er r1sc. E~c~u3e the tlvx ation of Maximum Frac t i*nn of dS3trtbutlon es3oc:£:~d ~!th u~irorm Limiting Power Density (MFLPD) rod ~lth1rr~~13 tio~s n~t Involve hl~h and reactor core thermal power, local pcak3, end b~c~use several rods The- k'rjam set ting is adj us te~i in must be "-Oved to ch~n~~ ?owrr bJ 1

~i5'.i!.~ce with the forn11rla 1n nt~nlrtcnnt pe~cen~~;c or rated po~er, th~ :r:?t'! or po;;~r !"~~c 19 very 110~.

specification 2.1.A. l. when the. Gencr.i lly, the he:>;,; Clux ls 1n *ncitr MFLPD is greater than the fraction cc;~lllb~lum *'!th tt:e fl:Jston rate. In of rated power ( FIU').

• an es:>u:r.cd u:i l l.>r::t rod ,,. tthdrDtt!Jl ap ..

The adjustment m;-1y also be p;-03ch to the :Sere::: leY'!l, the rate Ot accornplishl'd by increasinp, P?~~r rt~c 1s no ~~~~ th~n 5 percent the Al'RM gain by the reciprocal or r~tc~ power per ~!nute, an~ th~

of FRl'/MFl..PD. This provides the A?rl!*i :;J:!te~ .-ould lie more th3n Ddt'quate 15 same degree of protection as 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") 2.1.*

APRM Aod Block Tr1p Setttns The JR:" ecre111 ,t:-1:> nett tng or* 120 H!~l::~on~ 13 ecttve ln eech ran&e.or Reactor.po~er level m~T be **rted *7 th~ lfl~. For c~om?le, tr th~ *1n3tru- rnov1n~ control rods or by ver1tng *.

~~~~ were on r~r.~c 1, t~e scr3m Getting th1 r'!o!rcultt1on tlow r3te. The APR.~

~o~ld be

• 120 dlvlslons ror thnt rdn~el syotc~ pro~1~ea o control rod block to l!ke~l:>-:?, tr th~ 1n:>trur,,cnt ttere on r3nge prc\*cnt qross rod vlthdraval **

5, the scro~ ~?uld ~e 120 dlv1slo~= on st coa:Jtunt rcclrculctton hot:

th3t run~e. Thus, os the 1n~ ~:J ren~ed rnte t, prot~ct ~~~tn:Jt ~roaaly exceed-

\:? to occo::u>dPJt'? the lncreaoe tn po~*;cr ing the MCPR fuel claddlnq lnteqrlty

• l~vcl, the acr~~ trsp settl"~ ts al~o safety limit. This rod rented up. block tr1? ~cttlni, ~h1ch ta e~tc

~nt1cally v~rled ~lth rcotrculotlon Th~ ~o8t ~l~ntricant.~ources or ren~ loop Clo~ r~tc, prevent~ an !ncrc~ae ttvtt:; ch~nf:>~ durln~ the ;>ower lncre:.se 1n th~ r~a~tor p~w~r l~v~l -to e~crs-are ~~c*to cQnt!'ol ~o~ ~1thd:-nwel. In :1 lvc vnlu~:J c!ue to cnnt.rol r~ ~~:!t 0!'1e:' to ensu!'e the t the 1Hi4 provided drD'l-~:ll. 'l'h'? flow vcr1abl~ t;olp s~ttlng e1~~~ntc*prot~ctton egaln3t th~ oln~le prov tc!cn ~ u~:J t:t:l tie l :'!::' !"!; ln fro:w (\;'! l rod ~*!thdr:?'l>a:l cr*. or, o 4

ran~'! o!'. roll da:~:ice, n::;3u::Jn~ o l'lt-:?CC:Y-fit:l:;o C"'e 0- 00 "ltb!~c~at 1 ace hJ~a;t~ ""s !'n:-lyo:cd. Thts t *o ~ *

, n .. *.

.. 1l*~ r.~*;, ~r.::::nr:,* over th~,,,

• e~~!;c~~ lMclcd~d 3tertlns the ccc!~ent -:?ut.lre r(!cJ!'calatSoil flow range. '!°:1*,t P.~ "~rloc:: po ...:cr l~vels. The  :".!~~~ s~-

• ru:"l:-gln to the !iate~v Ll:n!t tr.er'!~::<-:> :!I v~::-e c:~~ 1a-1?l*1t-s _an 1n1t~~l co:'!<J ltlon th~ fl~~ dec~c~~c~ C:~ Lho o~~c~:**j tn tth~ch the re<?ctor ls Just cu~cr1t~ccl trip s~t:!~~ ~~r3uS flow rel~t*cft;~4p*

r~d :h~ I~~ crot~~

ts ~~t v~t J.d-!!.~~011:!1 cor:oerv:1t1nm 'ftr.s tn~-:on 1n thts on 9~~1e.

• h ('

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

IL * ' " ' " - f en~l)'Sl:J DY O!l~U!n,:u_; thot th~ 111i*1 Ch3nncl

.t!cn I~ e:; lOC~ or r~te~ th~r~~l p;~cr clQ3~~t t~ the ~~:h~r~~n rod 1~ ~rP~~3cd. bccru3e or th~ APRM ro~ block tr:;>

"&'h!!-l*esu!ts or \.ht:J 0:1?.lysls 30~~-; th:1t the oc~t1~s. The oct~~l p~~'!r d!o:r!butlo~

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

• ~,.c*"*,.d

.. t-**-*

~? 0:1. pe:"C'!~t of rat~~ pOl*:<!r, ti1'J3 1r.alntalnlnq co:i t ~:> l re.., DC'1U~ncc:; and !a r.:on!tO!'!!d

~.CPR above the f.'CPR fl!'!?l cladding integ!: ity C l"r. t * !'.!UO\:~ 17 t>*"'1 the 1n-CO ... ,. .1. .n~** *.. ,. t*-

I

, * * .... .**** Q * ..., -****

,.3 i*; 1 t.;1 ~~t' APRK ~cl':tr.1 \.r1~ .,. ...... :.

safety limlt. Based on the above

?;!nl:;::e, th'.? !?t:i p:-o.,:tl~~ p:-ot~ctton a:;.:?lnst the ;,pRM ro~! :>l.,.**i< ..,..... t ..

lust('d clc1wnward or AT'HM gain increased

....... ~__ ...

.. 1p .... ..... -

.., 1.:.:";:.::..!d

• • ~

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

.. ,,.0*1-

..

• 0# ***t'-*

... " . .-*****,1 1

. . . .-.. or c?.~ \; ro l ro*:J ~ J" :.:<?.*'-""tee

• ii' the rnnxlmum frac.ti.on of lindting 1?1~:; tJ&'OV iuc:J bu*.:!<up p~*u~~ct.lo:i ror t.h'! r.!'!':*l, 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 Amendment No. murlJ,in. 16

~

• • Tut'blno S~t'J? Ve1¥e"'Scr.'.'n - Tho tur'blne eto;>> YalYO c. R~nctor Coolnnt tov Pr~99ure tnltlAte9 Mftin $ten~

clo~t:I"! 11c:-aD trlp r:ttlclp.'\tcs tho prcs:.uro, lso~ntio_!'I Valve Closure - The low pressure 1soloti01l r.-:utro!1 fl*JX nl':" h!D\ flux 1nc~!:::o th:it. could "t 8~0 rstg wns provided to give protection ar,ainst

~~U~t fro~ :r:'.?!d ~!C~U~3 Of the turb1no ~lop fo9t renctor dcrreuurizatton and. the resulting

.,;,lv*;:s. ;ilth c scr~" tl"l'O Gct.llnc or 10 rnpld cooldown of the ves!iicl. AJ,,ontoce vos to~en Jl'!rCcnt. OC YDhO Cl?~Ut'U frOI:\ full Oj)CR, tho of the !lcrnm feature which occurs vhcn the ""1in re:oultmat tr.crease in O\\rf:ace heal flux ls steam llne tsolatton valves are clo~cd :to provide ll:"\1'.cd such U*.:t.t l:Ci>R rc:r.n1n' above 1 the* HCPR for reoc:tor 9hutdnvn so thot orcratJon at nu9sure*

fuel clnddln9 lnteqrlty safety llmlt, even lover t:111!1 tho!lc specif led ln the thC!r"al hydraulic durln9 th~ worst case transient th~t ~ssum~s sa(cty Umlt docs not *occur, althour;h operntlon the turbine bypass is closed. 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

r. C:cnna:nr 1.oad 1'dectfon Scre11 - The genera- 850 pslg V09 provided to give protection acafnSt tor lo:d rejection scra~ is provided to rn?id reactor dcpressuritatlon and the.resulting cntlcl1u1tc the rapid increase in pressure r:ipld cooldovn of the vessel. Ad'fant;,ge vas taken end ncutroa flux resulting from . of the scram feAture vhlch OCCUt'S vhcn the ma.In fa.:i~ closu~ of tho turbine .con~rol valns steam line tsol3t1on volve9 ore closed, to provide du~ to 6 1~ reject.lo~ end subsequent for reactor shutdown so *that high rower operation f~tlu-:"C or the b7J't'-"'5Z 1.o., ll prc'fent.s At low re:ictor pr~ssurc doe9 not occur, thus pro'fldlng

-;

;;:;! frc:i ttco:-i~~ le:a3. lhM the MCPR fuel protection ror the fuel clodJlng lntc~rlty safety claddin9 lnte9rity safety limit for this limit. Operation of the reactor et pressures lo~er transient. For the load rejection without *than 850 p!tlg rcqulre9 that the rc3cror wnode 1tvltch bypass transient from 100\ power, the peak -be in the stttrtur .position "'he.re rroteetloh of the hC'at flux f and lhP.refore l.HGR) increases on fuel clutldlnc intccrlty ufety Umlt ts provided b1 the order of 15\ which provides wid~ mac9in tlte. llUI hl:h neul ron flux scr.,... Tims, the co111blnotlon to the value corresponding to f°tk'l Cl~riticrlin1~ of 11111111 !ltec111 line low pressure holntlon and lsobtlott melting and 1% cladding :;tra:i n. 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 Amendment No.

1.2 ~AFETY LIHIT 1. 2 LIH1T1Nr. SAFF.n *svSnM SETTING J ** 2 REACTOR COOLl\tlT SYSTEH 2. 2 f\El\CTOR COOIJ\NT SYSTF:H

~pllcnbtllty: ~cnhlJtty: .-

Applies to limlte on reactor coolant system Applies to trip settings of the tnstrumente and pressure. *kv! ... .:o whlch nre provided to prev.:nt the reactor RyAtcm safety limits from belng exceeded.

Objective: Ohjectlvc:

To cstablteh a limit belmi which the.integrity To dcrlne the level of the process variables at of the reactor coolant system ts not thleatencd whlch auto1Mtic protective action ls initiated to due to an overpressure condition. prevent the ea(ety lilllita fro* being exceeded.

Speclflcatton: Spr.ctrlcatlon:

1'1e reactor coolant system pressure ehall not A. Reactor Coolant High Pressure Scram shall be excef!d 13115 pslg at any time when irradiated fuel ~1060 p11ig.

is present in the reactor vessel.

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 Amendnent No.

!!w.t

• th<.m 26, 700 psi at an internal pressure 1.1 'ltt* n*ctoT eoolMtt * .,.,"' lntetrfty h " tt11Por- of 1250 psig: this is a factor of 1. 5 tant b*rrlel' la th* PH"ntlon of anl:ontrolled r -

le*** o4 fl**lon *roducte. It le ***entlal that tt.e below the yirild strength of 40,100 psi lntetrlty o.f thl* *yne* be protected by **t11bll*hlnl at 575°F. At that pressure limit of

• pre11ure ll*tt to ~. obeerYed (or ell oper*tlng 1375 psig, the general membrane stress c~ndltlon* end vheneYer there I* Irradiated fuel lft will only be 29,400 psi, still safely the reactor Y****t. below the yield sttength.

'nte pr***ure **fety lllllt of ll2' p*ll ** 11e**ure* The relationships of stress levels to

' ' t~* weeeel *t* .. *p*c* pre**ure lndlc*tor I* yield strength are compara,Ple for the e~ulYalent to ll7S P*l& *t the louwet eleY*tlon of the primary system pipin~ and provide a r*actor* coolan: *Y*t*~. The ll7S pslg Y*lue I* similar margin of protection at the dertY** froa the d***t~ Pr***uree of* the r*3ctor established safety pressure limit.

pr.1e111re Yeseel. COOlan~ *Y*t** plplna: *nd hole-t!oa coadeneer. The r**~ctlY* deslGn prC!se11r** The normal operating pressure of the are 1150 ptla at 575*r. ll 1S peta *t S6o*r. end 1250. reactor coolant system is lOOOpsig.

psl1 *t s1s*r. Th* pre**ur* snfety ll*lt u~~ cho9ea For the turbine trip or loss of

• • tn* lover of th* prrs*ure t.1'31\Slrnt* p*r*ltted electrical load transients, the turbine 111 ;~* *;>pllcahl* ~nip code** A'>:iF. loller ond Pres~ure Ye~**l Code, Sectlcn 111 (or the pr*~~ure trip scram or generator load rejection

~**sel *nd'lsaletlon conl*nsar *nd l"SASI 111.l Code scram, together with the turbine bypass fer :he reactor coolant *1*t*~ plplna. The AS*:Z system, limit the pressure to approximately toiler *r.d Pressure Yeste 1 eo~. peralt* fl'f!SH*Jre llOOpsig (?.). In addition, pressure transients u;. .to IOI over <<!~sign press*1re (llCI relief valves have been provided to Jt USO

• 1115 pdg)

• and t!tC! YSi\51 C.,de pe~l t* reduce. the probability of the safety f:'l!!u.ure tr"n~l*nts up tet ZOl O\"'!r th .. cir*!*.... valves, which discharged to the drywellt rre1sure (120: X 1175

• 1410 ;>sic)~ The s~r*tr operating in the event that the turbine 1.t:slt Pressure of lJJS psta h rcfennced to the bypass should fail.

lowest elevation of the reactor vessel.

The design pressure for the recirc. suction Finally, the safety valves are sized line pipin~ (1175psig) was *chosen relative to keep the*reactor vessel peak pressure to the reactor vessel de.sign pi;-essure. below 1375 psig with no credit taken Demonstrating compliance of the peak for the relief valves during the vessel pressure with the ASME 0\'1.~rpressure

• postulated full closure of all MSIV's protection limit (1375psig) assures without direct (valve position switch) compliance of the sul'tion piping with scram. Credit is taken for the neutron the USASI limit (l410psig). ~~v.*1.uat:i.on flux scram, however, met)hodology used to assure that this The indirect flux scram and safety valve

~~~y limit pressun* is riot exceeded Rctuatjun provide adequate margin for any reload is ducumented in Hef erencc I XN-NF-79-71. The design basis for the below the peak allowable vessel pressure of 1375 psig.

reactor pressure vessel makes evident the substantial margin of protection against Reactor pressure is continuously monitored failure at the s;1fety pressure limit of in the cnntrnl room during operation on 1375 psig. ThP vessel has bt~en designed a l 500 psi full sci le pressure recorder.

for a general membrane stress no greater 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 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.

~*-*~,..,,

21 Amendment Ho.

(""\

e..

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

t LIM!tllfC 0 .. ..

l

. j, CO..DtttOlt roll OP!MTIOlt .l.1 SUP.VEtLLNlti uquuumtMT

'. 3.l *r.ActOll r!IOTtCT1cr.t STST~t 4.1 R£,\Cf0:\ Pl:OTF.tTIO~ SY...llQ! .* *

!fPllt1tltl I ltxt

.Aprtlt1 te the lttttrut1C!nUtlott ,... A(tpllu to the *urvctlhnce of the fftlUVlllC:D*

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

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

tf1dor eero*. l'cactor: 1cr:1*.

~J!.~* Ob *r. ct I vr.:

to a2t~r* th* .,.r.,t11t, of th* To specify the typo and frequency of r*actor *rotectlon *1*tea. *urvcJll~nce to be opp\led to the proteetloa Snstrutr.'!ntotlon.

hedflc~I !reel rtc,.,t ton:

• A. A. ln~trc"'~ntetlon syste~ shall bo

.,,te r*!rtt th* Ht:tolttUe ..tn!.CUD *nu'!"!>>er of tdp

~J !'thtl:""\.....,. n\:~cr of* s~stru-duumel~ t!*n:: !'!'\*!"It ba Ci'C!'t'Oble functfon~lly t~stcd L~J C3llorntcd m**

ln~lcatr.d In Ta.hies 4.1.l L,d 4.1.2, fnr ench ;o:ltlon of the reactor eode respect l\reli.

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

The system response times from tbe I. Daily during reactor power* operation ~bo~e opening of the sensor contact up to 257. rated thermal power, the core power and including the opening of the trip distribution shall be checked for:

actuator contacts shall not exceed 50 milliseconds. 1. Maximum fraction of limiting power B. If durinP o.nP.rAtinn, thP mnximnm density for *fuel fabricated by GE frac~:iton - o( limiting power density I

(Ml"LPI>) ;mu compared with the fraction f~;~" fabricated by GI~ exceeds t*~e of rated power (FRP).

fraction of rated power when operating above 25% rated thermc.11 power, either: 2. For compliance with assumptions of the Fuel Design Analysis of overpower

a. The APRM scram and rod bock scttinr. conditions for fuel fabricated by ENC.

I 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.

3.1 LIMITING CONDITIONS FOR OPEKATION 4.1 SURVEILLANCE REQUIREMENT Specifications (cont'd)

b. 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 f assumed in the Fuel Design Analysis for over-power conditions.

~"

~ .. - ,,'

. !"'~..-.,

Amendment No.

,\ ('Ufnl'!'ll"ISllll ur T:*hlcs .... l. l :11~t.I *l. I.?

n h:ttr Sl"r:im :tnd rod hloc:k condition. 11.us, h11llcah:s th:il !!ix inst n::iu:nl ch:in:lCls h:tve not I( lhe c:*llhr:illon were 1>cdormcd tlurlni: ovcr- hl'cn lndm!ctl In lhc l;1lh:r Tahh:. Tht*:ic :trc:

ilinn, nu~ !lha1Jlni: would nol he possil>lc.  :\!111lc !>wilch In S!a11tia\\ n, ~!:1:iu;1l ~n:tm, lll~h U:iSt*d **n t*~:11t*ri<'ncc :il nlhcr t:l'nt* r!'ll lnJ; \\';1kr l.cn:I In Scn1m 1Jisch:1q~c T:u:k, ;\lllln sl:illonl', ~lrHI ur lnstruml'nls, s11d1 :-19 those Slc:im I.Inc 1:-;ul:!llun \"ah*c l:ln~urc, Gcm:r~lor in the Fl1*w lli:tslni: l\cwnrk, I!! nt*I sli!nHic!'lnl J.o:itl lkjct*1i1111, :-in~ Turhi1:l' Slofl \'ah*e

uul lhcn*rurl', lo :froid s1mriu11s scrams, :1 Clos111"e. ,\!I ur lite dc\*lcl'S UI" b\.'n::urs :tssucl-C: :t 1ihn1I i1111 f l"l'f llll'IH:~* o( C:1c:h rd Ut:I i 11 I: Olll:'l gt. =-*nl wi1?1 lhl."sc sc1*:1111 hmt*lluns :.re lllm11h:

Is c:zll:1lill:iht*d. on-uH swill-hes :uuf. ln:nc.!, c:t!l 1 1r:11i1111 Iii no\

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

Gn1111* CC) 1lc,*lcc:s nrc :tclh*c onl>* tlurln;: n uU. Further, these S\\ llcht*s ;u-e 11101mlt*d a:h*cn 1u*1*llun of t!u: 011crnll1111:1I cycle. Fur suli*lll I" lhc cle,*lcc :11111 !i:l\*c n \*cry luw cx:a1111*lc, lhc 111!\l Is :tcll\*c tlurlns: st:nh*1* :"Ind 1trt*h:1hllil)" or 1110,*!n;, t.'. ::. the switches In ln:tc:Un* 1l11ri11:: f11ll-1mwer u1tc1*:11l11:1. Thus, the sc1*::t11 clisc:h:-iq~c ,*11lumc 1:111k. 1:~sctJ on the onl)* h:sl lhal Is tnc:tnln::hil Is lite o:*c 1*cr- the :1hu\\:, *no t*:illhr:1I i1*n Is 1*cc1:1l rc*I fur lhc11c Cunuctl Jui:t 1n*lor tu 1tl1111down ur &l!'lrln1*: I.e., t1i x Inst r11111cnl d1:umd:t.

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

• B. The MFLPD for ~uel fab~icated by GE shall C:1tlhr:tlion frcr1ucnc1 or lhc lnsh*umcnt ch:in- he checked once per day to determine if ncl lit di\*hlcd lnlo lwo s:rou9tll. These :nc :1~

the APRM gains or scram requires adjustment.

follU\\*~:

This may normally be done by checking the LPRM readings, TIP traces, or I. l'n~slvc ly11c lndlc:tllna: dc,~lc:u lh:tt c:an process computer calculations. Only a small number of cont.ri>l rods are moved be t*mu11nrcd with like units on :a conl111u-O**!I l,:1tilS. <laiJ y a11rl thus the peaking factors are not expected to change significantly and thus

2. V:acunm lube or scmlconduclor devices a daily check of the MFLPD is adequate.

n:icl rk:t*clon th:tl drill or iose SCn!illh*ity. For fuel fabricated by ENC, the power distribution will be checked once per F.xrc1*lcnt*~*w1th p:tssh*c l~-pc lnslrunlents In day to ensure consistency with the power Co1nmfl11wr:slth Erllsnn J:l'nt*r:tllns: sl:tllons :snd <li~tribution assumptions of the fuel design s11bst:1lh*1tJ lnrJlc:tlcs lh!'ll lhc :<ipcciricJ c:-ilihra- unulysis for overpower conditions. During li1*ns :trc: :u!!*r 1 ~:tlc. Fur those 11C'\*lc1~s whir!l p0riods of operation beyond these power cm1tlor :*11111lificrs, etc., d:*lfl s;1rdfic:tlio:is distribution assumptions, the APRM gains c:1H Cur......,....... .. lJu less lh:tn O. -1*:,./inun:h; I. c.,

• !1'TIJ 1T' nr scram settings may be adjusted to In the~*~ :t 1!10nlh :s drift o~ ..1'; w1111hJ ensure consistency with the fuel design o<:c~u* :uul lhns pruvit.li11;: for :tl!cq11:1:c m~t*;:in. critew-;, for overpower conditions.

34 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-r1 111cnq* :i_ssurc s Ill :ml opcr:iliun :tl or hduw lhcr111:1? llmlls.

3. 2 LIMITING CONDITION FOR OPERATION 4. 2 StJRVEILLANCE REQUIBEMENT 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. . .

36

3.2 LIMITING CONDITION FOR OPERATION 4.2 SURVEILLANCE REQUIREMENT

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.

36a

DPR-25 JKSTR01'L.~J\Tt<.'lf .Tl!l\T Im:TL'\TJ:S 1lOD Bf.OC:lt Table l.2.l r:ln!~'J~ 1:~. or C,!>~:r.b\' In:at..*

Ch:anMlS Par

.-:-in ~ 3tc:":\ 1 Trlo JA* l s~~tlner

.\

1nstrum!!nt l A?~i Up*cale (rl~~ bin~J(7) ~

- 0.5~0 +

t'HI' 50 ]Ji'LPIJ 121

• 1 APNI up11c&1!* Creruol

• en4 8tGrtup/Rot:

Etant\by r.:oo-?)

~ 12/12' full ocale

• 2* APr.Jt downecalo (7) 2 J/125 full *c:*l'!

l 1

• Red block fftonl tor- "111'Cala ( fl<N b(asJ 17) nod bloc" l'IOnltor cl~-nscnle (7) 0.65 w0 + 115

.? 5/125 f Gill *cal*

(2)

I l JP..'I dawnseele CJJ

-> 5/125 full scalw 3 liVI upacale ~l08/12S full.seal'

• • l lfl=1 clet@ctor not £\llJ.y in2ttrtad

• Ill tho cor*

sruJ detect.or not ln otartup posltlon C4l

-~*'1*-*'!""*r, o' 11*c

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

cc~nts I

I I

l\mendmen*t No. (2

e T~3~~

u 3.2.3 (cont) l,'

• • 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.

2. ff ,u n~rcea.,~ nr drive rto~., required to produce a rated core flow or ..

I 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.

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 0 0*

B:ts~s: Table 3.2. l which ecnece the condition. for Miich h1olallon Is required. Such lnslrumenl:1llon must be 3.2 In addition to.reactor protection Instrumentation :av:1llablc ,\*hcncvcr prlm;iry cont:alnanrnt lnte;rlty

"*hlch Initiate* a rnctor scram. protective Instru-111 rcqulrccl. The objl'cllvc Is to lsolnte the prlm11r1 mentation has been provided which lnll1:1tes action cont:1lnmenl so that the i:11hll'll~* or 10 CFR 100 are to mlllg:1te the consequences o( acclclcnls "'hlch :are not exceeded t.lurlng an :acclclt':-il.

bcyont~ the operntors :1bllil)* lo control, or lcrml-n:1le1 orcroator errors hclore lhcy rcsuU In Sl'rlous 111r lnstrument11llnn which lnlll11tea prlm1r1 1y1tem con1.1eq11cnct"s. This M.'l of ~clflcatlons Jlrovlcles h;r>hllon Is conll(?clcd In a dual bus arr*.::-:~mcnt.

li1c limlll11:: conclllions of O!'crallon for tht* prini:1ry Thus, the cllsc-1111slon r:t\*t*n In the base* for ~clR system lsolallon (unction. lnlllatlon of thl' emcr- c:allon 3. 1 Is attpllc:ablc llcrc.

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 .,.. *-***-*- ................_ ................... >*

IMM* - .... l*HI l-U-ftt fl ..... Mtln f-1 le ........ te ol the pa*otcclh*c Inst r1uucnt:1llon \VI.en rN111in*1I hy .. 160 l..ch4* ...... *****I **rot .... *fl** ea1.. a,.. t .. , ... f*ll prcscrvlni; lls C:IJ1:1blllly lo tolcrntc a slni;lc lailure. ...,,., pre***r* **** .c**** *** ete . . **t** the a.. le .. a **** I*

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

-of :iny componc.-nl of such syslc-ms even 1h1t"ln.: p1*rl- *~**** ***** ~.,,.tit*******"**** ac'I** ***I '*"'**I.I*

0tl9 wlll'n roa*tlons of such sp;tcms :trr oul of sc*n*lce 1...chre lo ... el IMn **tllel' f*f'I ***lf!M * ..,_***** . . . . . ., l*I*

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

for molnt<'n:incc. arMI (II) lo prciicrlhc the h*li, tiCl- T"** trlp a"ltl*l** cl***** ef Co.,.p I .... J P*caarr ~ ** ,... ..

t111i;s requlrrcl lo assua*c a1lc1111:1lc rcrforn:uncc. holetlOft **IH* * * * - * - telp ,.,. recbcwl*U.., , _ , . l**f*S*

\\11cn nccesliar)*, one clwnnrl m:1y llc m:ule* lnoJl(!r- * ...,. , ** **ell . . '*'**** , . . . . . . . . . . . . . . . . . so* ... ~.......

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

ablc for brll'f lnlcrv:11s lo cont.lucl 1*l'*iul rc*I luncllonal ***** clo**** ,._.. ,..,. **I*** *Ill M cl*. . . . . . . . . . . . *I****- el lesla onJ callhrallon9. .... c .......... _...,, . . . . . . . . . , ... - * - ...... * ... " " '.... .

tMrelcne . .c.-****

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

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

• • • W4l*f

'"'"' ........... *~*

...... ect*** r *****

• * * * * *

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

low valacs :1rc holh erlllcal and m:ty ~a,*<? :1 subslan- Thia llal ellccl on s:1fely. ll i;hould be nolcr.l:1!lon, where onl)' the hli:I\ lsol:1llon V3lvc~s. ncr. Scc!lon 7. 1. 2. 2 SAR, ~nil :also or lo\\' end uf the Bellini: has 11 direct l11"arin:; on ocllvatc*s th-:.- ECC suh!\ys!*!ms, st:1rt9 the ein<'rt;ency s:1rl'l~*. arc chosen ot a lcn*I 3w;1r from the norm:1I clic*scl ::cncr:1tor :a111l lri11s the i'eclrcul:allon pu11111s.

nnrw;alln~ ran:;c lo 11rc,*e11t lnadv<?rh*nt at:l'.l:tllun or 1*1i1s trip scllln~ lc,*cl was chcs<'n lo he t:l=:h ennu;:h

_,\~J'ar'!'t_,. p~*sl~*m ln,*oln*d and <'~~ur-:.- to ah~*,rm:al lo fll"l'\'rnt !>111:.-lous 01"1".'r:illon bat lo'" rncu~h lo l:tl-

~trttfttllo:19. 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~

lsolallon vnh-P!J ore lnslall.-d ln tho~ 1111(?9 th:1l so that l'"SI 3cc!tlent conlin~ can lie occom11llahrcl rx*n~tr:ate th4! primary cuntalnmenl arMI musl be  :.11111 the i;ui*ll'lin .. s o( 10 l: r:n too wlll not IK' \*lol:ttMI.

lsnlatccl durln~ a loss of cool:int acclclr.nl so th3l the l*"or lhc* Clllll(Jlrtt' eh-c:lll! rl*P*ntl;i I hrt*:ik or 3 211-lnch r:11ll;;t1011 closr limits :1r~ not cxccc*ll'1l 1lurlng an rccln:ul:1llon tin.- :in*I 11*ith t!ll' trip Sellin;: 1.rt,*e11 ac::ltlent conclillon. J\ctuallcn or thrsc vah*cs h rlHn*.-, 1-:CCS ln!tlallon :-iml primary s)*t'l*tn hilll:atlon lnlll:1lc1I by prolt*cllH* l11slrm11cnl:1Uon shown In a re lnillall'I! In time to mt*cl lhl.' :1IJO\"C crllt*rl:1.

Amendment Ho.

SllltVE I Ll.ANCt: m:11111 m:~u:NTS J. J LHllTING CONDITION FOR 01'1'.lt#\TION 4. J C. Scram Insertion Times r.. Scram Insertion Times

1. 11rn avcrnre scram insertion time, based 1. After each refueling outugc and prior to power on the de-energization of the scram pi lot operation with reactor pressure above 800 psig, valve solenoids ns time zero, of all oper- ull cont.-ol rods shall be subject to scram-time able control rods in the reactor 11owcr tests from the fully withdrawn po!. it ton. The operation condition shall be no greater than: scram times shall be measured without reliance on the control rod drive pumps.

\ Inserted f roin Avg. Scram Insertion Fully Withdrawn Times (sec) 2. At 16 week inte~vals, 50\ of the control rod drives shall be tested as in 4.J.C.I so that s 0. 375 every 32 weeks al 1 of the control rods shall 20 0.900 have been tested. Whenever SO\ of the control so 2.00 roJ drives have been scram tested, Pn evalua-90 3.50 tion shall be made to provide reasonable assurance that proper control rod drive .*

111e av~ra~e of tho scram insertion times performance is being maintained.

for the three fastest control rods of all groups of four control rods in a two by two 3. Following cc*mpleti.un .uf each set qf scram arr:.ay shall be no gre_ater than: t1~~;ting as dL~GcrlhcLl above, the results will be compared ai:~ainst the average

\ Inserted From Avg. Scram Insertion scram ::;p1-~ed i.lir; tribu tiun used in tl-e Fully Withdrawn Ti mes (sec)

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

s 0. 398 Limit. Herer* t(I ~;pt~cil'ication 3.5.K.

20 0.954 so 2. 120 90 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, 58

,Amendment No,

0 01 .' 0 0 s*:\ll n.'flOunt .of nd v\fh,*uw:1I. whlch h lH*

lnJlutlve oi a C~C'rlc cont.°"I r-n~: Jrh'c" th~n ~ norr~I sln,.1e.vithdr~w:1l lncr4:"'1cn~ vlll

>robin n~d the r.

.*ctot t1l ll ~c shll

.:,,.,n.  ; . .>~ CG!\tdt\;tC to :my ,b::.:.;c to t?\~ ;-rh::ry Ano if da.*u1e wllflln the control roJ "* l.vc cool~nt syst~~. The Jcslin b3~ts ts el*tn '"

~.:<~*nltQ *nd In r~rtlcular, c~~cks in ~rive !cctlon

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

tntcrnll ho~tlns~. ~~r."ot be rut~~ out, ~htn a tlon i~ tivcr. ln Section 6.6.J. This ;u,p.>rt

• CCftCl'iC nroblca a(fcctir.I a n;.?;~Cr o( .J~~~eS ts not re~cl red tr the nactor cool:lnt systc:.

ci:1n?t bo ruled out. Circll!::frre:r.thl .u:i.:lrs Is at *t:t.os;:t1-:rlc pressure s.lnco thtre '"011ld

. 1'Ctultlnc froa stress asslstod t"tcr1nnubr thtn be no ~r\wlic rorce t, r~~l~ly eject a corrcslc~ have cccu~rcd In the col let hcuslna drive hou~lnt. A~Jltlon2ll7, t~e sup~o~t t1 of ch\vct at sevcrd t~iib. ll'.ls t1J>c of nnt rcqul rl!d \ f a 11 control roJs Ire fi:l ly cr1c\in: cc-~ld occur ln

• nc.::~cr or drives tn~erlcd *r.d if an :adcGU3te shutdci."!\ r.>rcln end :.r '!~c cr.lds rrer-3,:itcd untl I se~*cronco. "*ith c.ne co:'ltrol rod ... t:l:Jui.-n. ~as been i~:t0n*

o( t~e collct hc~sin~ occurred, s~Ta~ cocld stri:t~d slncc the reactoT would re'!1aln subcrltlcal'

~t prevtr.ted In the *ffc~tcJ r>>cls. Lialllnz even ln the eYent of COQpletc ejection of the t~c perlol of o~eratlon ~Ith a p~tentl*lly strone~st control rod.

1cvcred collct t.ousln1 and rcqulrln: incrcase4 su:"":cllbnce *f:cr d-:tect ln~ o:... :acct J. Cor.trol rod wlthdr:aval ar.d Insertion se~11~nce* ere r:~ .. ~ ** 3SSUTC thlt th~ :'C2Ct~r wlll not cst*~lished to 3ssuro that the n1alOU111 lnsc~~ence b2 o;u:ntccl wlth a hri1~ nu:.ber of rods vlth lnli~ld~al con:r~l rod or control rod st~cnts

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

c. Control Hut.i Wlthdr*uwa.l l80 cal/gam to be exceeded

1. Control rod dropout accidents a.s di9 l i t~cy W'!r~ t:> ~:-:tp c11t o! the c<<1r*

cussed in HeferPnre XN-.NF-BO-l'), tn the ~*n~c~ Jeftncd for t~c ~tJ Or~p ~c~tdcnt.f'>

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

.... re. I! couplinz lntc~r\ty ls c..~l:itctned. 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 t~c ;~s~lh\llt! '! 2 :'~l ~~crc~t occ\Jcnt ls *n!I t*!c r"':~**'. rc~:~n~ that :?n o;>cratt'r ':>!10;1 th;~o cl\r.h.:ucd. r.,~ O\"c::tr::vc.l r:>sltlcn futu:-c ~c111.?ncc~ ' ' loa..::tcd l:? br th;,e,r.u.tlo:i c.f t~c r.:*.~t.

rrev\~S : pc!l:t~O check A~ GnJy UnCoup:o~ or a second qualified station employee.

d~\ves 1118Y ~**ch this f~Jl:lon. N*~tro~ These sequences are rl~veloped lo limit h11u-.11t*t l*n res,ens* to rod r.:v*,:-.::it reactivity worths of control rods and ptov*~ei a v*rlflcatlon that the ro4 ts !ol* together with the inte~ral lowl~I lt* drive. Absence of such response r£>1 "~le.:!.:)' lir.ltcts :1:-:J t}.r e:t\:>n or the CO't'!IOI to drl~e c2\"c~ent ~ould provide cause for

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

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 Rcstrlc*i~I reccu~llng verJf lr.otJons to povcr dro? :sccloc~t wll l no~ e11cc:~ a a.nlr.!e fu'!l Cl'lt!:"'Jr le~flal.'::!.1f.~~a 20: prowldcs assurance that a c:>:l~cr:t of 2S:t ut/er.*. T!:c r-u~ foci er.th:si~r cf

• rod"dl;,,*'i~~ins a -:-<"coupll~G "*~lflc:*tlon ~LO c~!/~*~ i i bclc~ th~ c~er~y co:i!e:it a: ~~i'~

vouid no:. res**lt In

• rod drop eccldent. 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

2. ~t-.o ronuol rod ti.u1\n1 suppert restrlcts ls dlscu~scJ ln Rcfc:e~ce I.

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 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.!

u.:iul.,~t:t "'~.tc~ cu:i~d b4t ~*"*' ~1 lhh 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 ,.,,** 62

.Amendment Ho. 42 rc(a~e-:J an,Jysls of the control :-!!d d'Of' sc**'6t'

Bases(cont'd)

Parametric Control Rod Drop Accident analyses Th~ Rod \..'orth Hlnlml:r.er 11rovl1lr~1 nulo111:1tlr.

have shown that for wide ranges of key reactor uuporvleto" to nnnur~ thnt. out or !t'"*1u~nce parameters (which envelope the operating ranges t.ontrol rocl5 will not be vll111lrawn or ln~ertf!d:

of these variables), the fuel enthalpy rise I.e.* lt llmlls operator dcvlattom1 fro* pl11n."lecl during a postulated control rod drop accident wlthdraw3l sequr.nce!I. Rcr. Section J.9 51\R.

rem~ins considerably lower than the 280 cal/gm lt serves as 3 bnckur to proccclur3l control o[

limit. For each operating cycle, cycle-specific control rod worth. In the evc*nt th.1t the ltod parameters such as maximum control rod worth, Worth tllnt:nlr.cr ls out of service. wbcn rcquil'ed 9 Doppler coefficient effective delayed neutron ;1 !.lcenscd operator or oth~r qunllftr.d fraction and maximum four-bundle local peaking technical einployre c .... n mtll:u.:1lly (ulf l ll the factor are compared with the results of the control rod p3llcrn conform:tncc functlnn~ of the parametric analyses to determine the peak fuel Rod Worth tllnlnl7.cr. Jn this C3!'1'! t procedural*

rod.enthalpy rise. This value is then compared control Js cKcrclsr.J by vr.rlfylnr. nil control against the Technical Specification limit rod poslt Ions a(tr.r the wlthduw.11 of each of 280 cal/gm Lo demonstrate compliance for croup. prior to procccdlnc to the neY.t each operating cycle. If cycle specific values &roup. Allowln& substttutlo~ o[ a sc~ond of the above parameters are out~ide the ranoe Independent ~pcrator or en&incer ln c~s~.

assumed in the parametric analyses, an*exte~sion o( R\.IH lnopr.r:ihi l lty recocnlv'!S the C.lpabllltJ of the analysis or a cycle specific analysis to adequately aonltor proper rod scq~~ncl~g in may be requir.ed. Conservatism present in the an alternate 11.:1ni1er without unduly rcsCrlct- '

analysis, results of the parametric studies, Ing plilnt operations. Above 20' power. thel'e l*.

- and a detailed description of the methodology no requlrencnt th."lt the R:,-:*I be opera~le si11ce for performing the Control Rod Drop Accident the control rod drop accident wlth out-of-analysis are provided in reference XN-NF-80-19, acquence rods will result in o peak fuel Volume 1 (Supplements 1 and 2). 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' (1) control rods (or any st3rtup o(tcr Ju~e 1. 1974.

t*one, C.J., Stire, l.C. encl Voole1, J.A., "llod Drop Aecl~cnt Anolysle for L3rge tollln1 ~ater P.cactoro", 4. *The Sourcf! Ranco tlonltol' ('SR.'t) systcQ perfonH NED0-10~27, ttarch 1972.

no outom.,tic saFcty sy~tc~ functlon 0 i.e ** it has no scram £unction. It docs provide the (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 \ 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 62a Core:i"

• Sc;-ph~~n: 2-t:r.DO l0J2 7

• J~nu3r/ 1~1). *

,,-~

\. _, .

opcr;~or v1th a visuol indication of neutron c. Scram Insertion Times ,.

lc~cl. thi~ is needed for kr.owledgeable and '\.

~!iic~~nt rcactor~sto~tup ot low neutron level. The performance of the control rod insertion rr.c c~~~e~c~nccs of reactivity ~cciJcnts ore systl.'m is analyzed tu verify the system's f\!~Ct !.uns c-~ the in.lti.:Jl neutron flux. The abi 1 i ly to bring the re;1ctor subcrltical at rcqukrc~~nt ~( at lc~st 3 counts per second a rate L.1st ennu~h lo p1*event violation of the

=s:*rnt-..*s th::t nn'J ~raa:;lent, should it occur~ MCPK Fuel Cladding Integrity Safety Limit bcr,liw ;lt ur above the inlti.al v:iluc of io- an<l thereby avoid fuel damage: The analyses of r&J::cd j>c*.:cr used in the analyses of tr:msicnt9 demonstrate that if the reactor is operated frci~ cnlJ ccndf titm~. ' One 01>1.:rablc SIU*I channel within the limitations set in Specification

• .-ould be adequate to monitor the ~pproach. to J.5.~, the negative reactivity insertion c:-l: lr.:i! ~ ~;* :1s !r.;; hot!'lo;:e:ieous pntterns of . rates associated with the observed scram sc&Jttc:-cd co:itrol roJ uJ lhJrn'l.11. A ciinim1.:m performance (as adjusted for statistical of. t~u o~cr~bl~ S~M's or~ provlc.lcd es an added variation in the observed data) result in protection of the MCPR safety limit.

consct'v:it iia11.

In the analytical treatment of most tr,111~;ienrs,

'* 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 290 milliseconds are allowed between a neutron sensor reaching the scram point and the start of motion of the control rods. This

'10\.'l

r ccn!li ty durin~ hJ ch power level oper:ition. is adequate and conservative when compared

'!'::o cl1a:mch arc prnvltlcc.I and one of these may be Lo the typically observed time delay of byp::s!:cJ (ro~ the console (or ll'aintcnance and/or about 210 millisecons. Approximately 90 tcsthr.. Triprlnc of one of the channels wlll block milliseconds after neutron flux reaches erroneous rod wlthJraYnl soon en~u~1 to prevent fuel the trip point, the pilot scram value solenoid d~1:0:1:: 0 e. 11ils ~;ystc::1 baC:<s up the opcnJtor who wlth- de-energizes and 120 milliseconds later the "r~r..*-; rorfq ac!:ortlln~ to n written seq~*ence. The control rod motion is ~stimated to actually srcd.f icd restrictions wit)l one channel out of he~in. However, 200 milliseconds rather than a.oorvlr<? con!>erv.itiv.:ly :1ssure th;it fu~l danmr.e UOmilliseconds is conservatively assumed

'-'ill r:(lt o:cur t!t:c to i*od withdr~t1:ll errors when for this ~ime int~rval in the transient analy-t!1is condlt!on c::l~ts. lu:1.-?nd1'1c11Ls 17/18 and 19/20 ses, a~d ts ~lso included in the allowable prcs~r.t ti1c tc!:ults r.f on ev<1luation of a rod block' scram 1nsert1on ti~~s soecifie~ in Soeci-a r.:o:iitC't!' f:t!lurc. 'Che~;\? omcnc.;ir.cnts show that during f ication 3.3.C. In the statistical treatment

.J ll. rc."!ctc-r oper:ition ".lilh certain lln1itlnc control of the limiting transients, a statistical 1 roo '."'::ttcrr-:,_thc vltltdr:r~al or a de~i&nntcd sinr;le distribution of total scram delav is used "l c'l-.~~<*t ro~~.Ht r,**:11lt in 1Jnc or more [1Jcl rot!, rnther than the bounding value described

~ vath i:C;>R:a i;;JtJ:'9W1'0I. t<C?R fu.,I cln<.iJlr.'I above.

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 J

sy:Hcr: ~rro:- to vltl11!r;:11;1l of such rods to assure The performance of the individual control rod its r:-1rr:!b i U :)* 11111 :i~:;ure th~t improper with- driv0s is monitored to assure that scram pc:rfurmance is nut de~raded. Fifty percent Jr.:i;-:.11 '~oc'.l not ocr.cr. It is th~ re~pon!>ihl llty of the control rod drives in the reactor are of.llw ::l:clc.ir t:1:)::cr.r to iJentify these 1J1:1ltlnc tested every sixt~en weeks to verify adequate P~t :*!:-ns a~*J the d*:~* l1~n:i~cd rod9 cltlwr when the p:ittern~ l!TI! ~nltlally established *or as they nerfnrmanci>. Observed nlant <lata were used dcv~l~? du~ to the occu~rcnce of inoperable control to del.ermine the aver;1Pe scram :1erformance 63 rods tn other than limitin~ paLLerns. 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 64 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.

. 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 ~ I D. Control Hod Accumulator:; reactivity greater than t<<J, ~are not ex-pected and requl re thorough evaluation.

• The has is for this spccifil:;tl ion wns not dcs- Une percent reactlvlly llmll Is considered crlhcd in the SAR and, lhcrdure, Is pre!'cnled safe since an Insertion of the reactivity Into in its entirely. Reriuiring 1111 more than one the core would not lead lo translent8 eicceed-Inoperable accumulator in any nine-rod sriuare lng design conditions of the reactor vyslem.

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 G. Economic Generation Control System rc~ulll'd in a kcrr <I. 0 -- other repealing rod Operation of the facility with the Eco~omic scqucnc.cs with more rods withdrawn resulted Generation Oontrol System with automati.c ,

In kcrr **I. O. Al reactor rn~i:;surcs In excess flow control is limited to the range of 65-*

or :iOO psiK, C\'cn those conl rol rods with ln- 100% of rated core flow. In this flow range op1*rahle accumulators will he nhlc to meet rc- and with reactor power above 20% the reactor quircd scram Insertion timt:s due to the action *can safely tolerate a rate of change of load or rc:-iclor pressure. In addition, they may he of 8 MW(e)/sec. (Reference FSAR Amendment 9"-

norrnal!.J ,Inserted uslr\g the cont rol-rotl-"dri\'c Unit 2, 10-Unit 3). Limits within the Econo-hytl~tl.k.,. systcm. Proced111*a I control w i II asl:.~ 1 '"1'1Vlft'~onlt*ol rods ith inopcrahlc accu-

• mic Generation Control System and Reactor Flow m11l:ato1*s \\'ill he spact*d In :1 one-in-nine a1-r:-iy control System preclude rates of change 1*:-it ht r 1ha n ~l'OUflC'tl to15et her. greater than approximately 4 MWe/aec.

E. Heacllvlty Ano_malles When the Economic Generation Control System !

ig in operation, this fact will be indicated DurinK each ruel cycle excess operating renc-

• 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.

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

---

T

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

J.S LlMlTl~'G CONDITION FOR OPERATIOM 4.S SURVEILLANCE REQUIREMENT D. Automatic Pressure Jtollot Subsy1tems

• D. Suncll hnco of th* Auto1Htlc Pressure

- Roller Subsyste* shall be perfor11ed H follows:

I. !xeept H 1poclfled In J.S.0.2 and S below, . 1. During each operatln1 cycle the followln1 the Automatic Pressure Relief SU:,system

• shall be pcrfor11ed:

1hall be operable whenever the reactor

• pressure ls 1reater than 90 p1l1 and irrad\ate4 a. A simulated automatic lnltl1tlon l\lel i i ln th* reactor ve11*l*

• whlch o~ens all pilot valvss, an4

2. Pro* and after the date that one of the b. Wlth the reactor 1t pressure each flvo rellef vahu :-6(1.tbe* autor.iatlc pressure

• rcllef valve Sh3ll be *anumlly opened.

r~ller :subsyste:11 ls ~3c!o or found to bo Relief valve opening *hall be verlfled '1

• a compensating turblne byp*** valve or inopernblo when tho* reactor h pressuriled control valve clo*ure.

ob,,~c 90 pslc ..-lth irr:adlntcd !u'll in tho n:tctor vessel, rc11ctor opeution h pomlsslble c. A loglc syste11 functional test shall be only durlnc the 1uccccdln: seven doy~ ~nlest performt'd each rofuellnc out*I**

rcr3lrs arc ~3de and provided th3t durlnc such tlrr.o the 11rc1 Subsystem is operi.ble. 2J. When lt ls determlnt'd that one relief valve of the aut~tlc pressure relief sub17stea J. FroQ and crter the date* that Moro than one ls inoperable, the llPCI shall be dc111onstrat..

of flve.rtlie£ volv~s of tho outcmctlc to be operable lnnedlatel7 and weekly thereaft..

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.

s. When it ls detort1lned that .ore than one

_, ~elief valve of the *~to*atlc pressure relief

__,,_..~ I subsystcn ls inoperable, tho HPCl subsystem

~*-*~.-.,r; shnll be demonstrated to be op~rabl* immedlatel)

Amendment No. 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~ I. !'.Y~rn!J.o_~~~n_nr Lln-,nr llcat CcnC!ratlon Average Planar Linear Heat Generation Hate lt"to {J\PlJIGR}

(APLHGR) or all the rods in any l'ur*l n::::P.m-

.bly, as a functlun of averagr~ planar f:xposu e ,.he J\PUIGR for e1u:h t:ype of f\Jel a* a fur G .E. fuel and avPrage bundlr~ expoirn 1*e function of averRqe plnnar exp:>eure for G.E. fuel J I for Exxon fuel at any 1U<ial location. ehall not exceed the and av1:!r::i1o;c l1undl1~ expo:~ure l'or ~xxon fuel shall 01~ d1!LP1*111incd dally during reactur operaticin at

• rnr.xt.z:n1m avcrn9a. planar UtGR nhown in 2 ~~% rated thermal power.

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>.

818

DPR-25 3.5 LIMITING CONDITION FOR OPERATION 4.5 SURVEILLANCE REQUIREMENT 3.5.J LOCAL LHGR J. Linear Heat Generat.ion Rate (LHGR)

During steady state power operation, the I I

The LHGR shall be checked dai1.y during linear heat generation rate ( LllGR) of . reactor operation at 2 25% rated thermal any rod in any fuel assembly fabricated by power.

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 81B-l

DPR-25

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IS. rLMAR AY!RIGt l11'D~t Amendment No. 42

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DPR-25 3.5 LIMITING CONDITIONS FOR OPERATION 4.5 SURVEILLANCE REQUIREMENTS K. Minimum Critical Power Ratio (MCPIO K. Minimum Critical Power Ratio (MCPR)

During steady state operation at rated co~~ MCPR shall be determined dail~ during a f lo~, MCPR shall be greater than or equal to - reactor power operation at ~257. rated Unit 3 thermal power and following any change in power level or <listributio~ that would 1.30 (All fuel types) cause operation with a limiting control For core flows other than rated, the MCPR rod pattern as described in the bases for operating limit shall be as follows: Specification 3.3.B.5.

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 within-lQ...hours.

- I Surveillance and correspon d 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. Amendment 34 d~ted 5-3-78 BlD

orn-25 1.5 MANUAL AND AUTOMATIC FLOW CONTROL FLOWMAX '117\

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1.0 30 40 50 60 0 80 90 100 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

1.7 AUTOMATIC FLOW CONTROL ONLY 1.6

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~ 1.4 1.l .,._____ .,._____ * - - - - - - - - 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

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• 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-* Should Ont" <'Or<" PJ'l':'~* iaub:o~*ttlc-m IN-C'nm* ln-tabllshed so that the .avera9e risk rate opc rable. lh* r*m:1lnlng C'Orl" !ft':-:1\" !'Incl lhl" for repair would be no 9reater than enllr~ LPCI !1)"5l('m :art" :1uil;iblt' ~hould the the basic risk rate. The method an~

concept are described in Reference (2) NED0-205G6. General Electric (J)._ ,Usinq the results Company Analytical Model for Lose-of-Coolent Analysis in Accordance vith 10CFR50 Appendix K.

Cl) *t.oss of Coolant Accident Analyses Report (J) APEo-*cuidelincs for Determlnln9 for Dresden Units 2, J anrl Qu~rl-Citles S~fc Tc5t Intcrvnls and Repair Units 1, 2 Nuclear Power Stillion~,* Times for Enqincercd S~(cquards* -

Ntmo-241461\, llevisionl,* l\pcH 1979. April 1969, I.M. Jacob~ an~

P.W. H:irriott.

{ll) XN - N 1;*-H l.-'( 1i "Uri *sden Un.lt 3 LOCA Model Amendment No. U!*; i ng the r*:NC EY.F:M Evaluation Model

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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 .

• 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

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.,: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. .

'I'he maximum average planar LJIGRs for G.E.

t.°UP.l pl1itt*!d ln l*'ir,. j.').l at I

r The maximum average plannr LHGRs shown

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

h1~r P.x11n::ures re!;ult in a clllculb.t~d r.c:il< cl.1.d tc1:1r.cnt~ or 1-:::-~

f,..., . employing the models dp~;cribed in

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

. tt::in 22ooor. llc.~*cvcr tho ru:da:\!:. av"=~

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

rower op~ration with Ar'l.ll!:Hs at or below ll~ll!J b=c~u~o conror.:-o'lnr.o c~lculnt1c~3 hsTe

.z0 . thm;~~\jn in F'ig. 3. '.5. I w,;t;ur*'G thn t.

• the peak cTadding tr!mpet*ature fnllowlng a net b:cn Utcna 1n

~~rrorr::d OXCC09 to justlf7 opor~tlcn or tho:Jo ahour..

ai postulated loss-uf'-cuolant accillent will not exceed the 2:!CJ0°1°' lirni L. ~* 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 (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, lei..;s than the dei..;ign linear l

/\pr1l, 1979. 8$A

.5 Limiting Condition for Operation Hnses (cont'd) heat gener~tion rate even if fuel pellet den- the cycle-s~ecific fuel loading, exposure and sification is postulated. ftwl type. The current cycle's reload licen-sin~ analyses identifies the limiting transien For fuel fabricated by ENC, protection of the for that cycle.

MCPR and MAPLHGR limits and operation within the power distribution assumptions of the Fuel Design AnRlysis provides adequate As de~cribed in specification 4:,.C.3 and the protection against cladding strain limits, associated Bases, observed plant data were herice the LHGR limitation for GE fuel is used to determine the average scram perfor-unnecessary for the protection of ENC mance used in the transient analyses for fuel.

• determininR the MCPR Operatin~ Limit. If the current cycle scram time performance The steady-state values for MCPR specified falls outside of the distribution assumed in the Specification were determined ~sing in the analyses, an adjustment of the MCPR the THERMEX thermal limits methodology limit may be required to maintain margin to described in XN-NF-80-19, Volume 3. The the MCPR Safety Limit during transients. Com-safety limit implicit in the Operating pliance with the assumed distribution and limits is established so that during adjustment of the MCPR Operating Limit will sustained operation at the MCPR safely be performed as directed by the.nucl~ar fuel limit, at least 99.9% of the fuel rods in vendor in accordance with station procedures.

the core are expected to avoid boiling transition. The Limiting Transient f!l CPR For core flows less than rated, the MCPR implicit in the operating limits was cal- Operating Limit established in the specifi-culated such that the occurrence of the cation is adjusted to provide protection of limiting transient from the opera~in~ limit the MCPR Safety Limit in the event of an will not result in violation of the MCl'R uncontrolled recirculation flow increase to safety limit in at least 957. of the ranJnm the physical limit of pump flow. This pro-statistical combinations of uncertainties. tection is provided for manual and automatic flow control by choosing the MCPR operating Transient events of each t:ype ant ici pated limit as the value from Fi~ure 3.5-2 Sheet 1 during operation of a BWR/3 were evaluated or the rated core flow valve, whichever is to determine which is most.restrictive fn greater. For Automatic Flow Control, in terms o-f* therma I. margin r.ef~uirements. *The i!d<lition to protecting the MCPR Safety Limit gen~n1iFt~t4oa<l reject ion/turbine trip without during the flow run-up event, protection is Lypass is typicall.y the 1 imiting event. provided against violating the rated flow

• The thermal margin Pffect:s of the evt*nt. are MCPR Operating Limit during an automatic flow evaluated with the TllEKMEX Methodology and increase to rated core flow. This protection appropriate MCl'R l .i.mit s cons is tent wit: h is provided hy the reduced flow MCPR limits the XN-J critical power co1-relati.1111 ;11*e shown 1n Fir,ure 3.5-2 Sheet 2 where the curve de term i.ned. Severa 1 factors influence corresponding to the current rated flow MCPR which transient: results in the largest: limit is used (linear interpolation between reduction in critical power ratio 1 such as 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 ..

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At core thenul power Inell

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

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

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

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

cxp-?rlcnce nnd thermal hydraulic: nnal711e*

lr.dlc~te lhot the rc:sulllnc ovcr:1gc pl.Dn:sr 11<>derator vold content wll l be **i7 s*ll . *~l ~1*f 1,-,

al I dcs11nated control rod patterns .Jhl~~llli;t ~:\ :' :t*l ~

LilCR h belou the n>xlnnm ovcrnce plomr LRCR e111ployed 11t this point, operet~fti*"pl~t **,.~...C.* *r, .

b7 n con*lder3ble rrorstn; thcrcCorc, cvelu*t.10D Hd the111al hydraulic ***rsi1 *iftdtcltet thair.tlljit , 14' ~

or t.he nvcrece plenar LllGR below thte povcr* rcsul tln1 MCPR value ls in oaceu of, ;.,.11'~ :* :1:1 .'

leYel ls not ne1ee11s~r7. The c1oll7 rcqulro- by a considerable aar1tn. With tills IOll wtcl'. ; * *' I'

mnt. for clDculoUng .. .,erace planar LllCR content, any Inadvertent core no.I i~cr*** .. ;. , , '.

abo<We 25 pitr cent r3 lec1 the 1"11131 power Ill would.only place operation in a 110re*con- .. , ;*~ ,,

1ur:-1,1cn~1ln*.;po~cr 4hlrlbutlon ahl~~* senatlve 110de reletlve to NCPR. . , .. L '.*!'.. .1.n i ' I*

ere lit~ u ten ... re have not been .icnttl-. ... *: . '. if" ': " * :*: *.

• cant. poue . or q.-trol rocl c,.ncce. ** 1"e dally requlre.ent for calculetlna* ... i

• tr ,(' . .

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MCPR atiove 25 percent rated thenal' . :* *.., '.;t;U~J ~* 01 11,¥,

,,.. ~j: 11 powor ts sufflclont since power dbtrlb~lelil"i(I".  :\t ** * ~':;;l t~IC!t ;1 shifts an very slow when there hDYe not hen' .

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

!.ot:al

'11-.c! U:CR f(lr G. F.. fuel shall . .. . ("' **~*'-t.1'**1fl1 * , ..*i b~ ch"!c1Mc1 c1n0,7 4urtng re*etor operation et In addition, the I correction ""lied tJ.(:.::t*H~.~'.** r:J. . .:.tt the LOO provides ..Sr1in. for flew* Jncpaie*;. i., '.~*.; , :

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.

frOll low flows.

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

~~~,.+ pove r

• Amenctnent No.

86A