Text

Proposed Tech Specs Allowing for Use of Exxon Nuclear Co Fuel Assemblies for Ensuing Cycle 9 Reload & Future Reloads
	ML17194B412
Person / Time
Site:	Dresden
Issue date:	12/21/1982
From:	; COMMONWEALTH EDISON CO.
To:	;
Shared Package
ML17194B411	List: ML17194B410; ML17194B412; ML20071Q036; ML20071Q059; ML20071Q087;
References
	5039N, NUDOCS 8212290193
	Download: ML17194B412 (58)
	v • d • e

ATTACHMENT* 6 Dresden*Station*Unit-2 DPR..:.19 Proposed* Technical Specifications Revised*-Pages Previous ** Amm; No~

License Cond ~

3~M 63 l

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62 5

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59

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59 10 58 11 58

llA 13 58 14 58 15 58

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18 58 19 43 2d 58 21 58 22 58

22A 34 58 36 Original.

36A 42 67 42A 67 46 58 58 19 61A 67 62 57 62A 58 628 43 63 58 64 58 65

. Changes 27 and 18 78 46 818 58 818-1 58 Delete 81-C (7X7 curve)

(

8212290193 821221 1

PDR ADOCK 05000237 p

PDR i ; ___.

.:.. 2 -

Proposed-Technical*Specifications Revised**Pages 81C-2 81C-3 81C-4 81C-5 810 81E

81E-l 82 85A 858

858-1 86A

Denotes a new page Previous-Amm~ -No~.

58 58 58 58 58 21 58 58

58 58 NOTE:

Page 8lt-l remains *as is; it is not being deleted or revised~

5039N

\\,

Am.

5039N DPR-19 M.

Provisions to allow operation with one recirculation I o op o u t o f s e r vice :

1.

The steady-state thermal power level. will not exceed 50% of rated

2.

The Minimum Critical Power Ratio (MCPR) Safety Limit will be increased 0.03 (TS 1.1.A and 3.3.B.5.C)

3.

The MCPR Limiting Condition f::>r Operation (LCO) will be incieased 0.03 (TS 3.5.K and Fig. 3.5-2)

4.

The Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits will be reduced to 70% of current values for all fuel types.

(T.S. reference 3.5.I)

5.

The APRM Scram and Rod Block Setpoints and the RSM Setpoints shall be reduced by 3.5% to read as follows:

T. s. 2.1.A. l s <.58 WD + 58.,5 T.S. 2.1.A.l* S ~ (.58 WO+ 58.5) FRP/MFLPD T.S. 2.1.B S < (.58 WD + 46.5)

T.S. 2.1.8*

S ~ (.58 WD + 46.5) FRP/MFLPD T.S. 3.2.C (Tabre 3.2.3):

APRM Upscale' (.58 WD + 46.5) FR/MFLPD RSM Upscale~-(.65 WD + 41.5)

6.

The suction valve in the idle loop is closed and electrically isolated until the idle loop is being prepared for return to service.

7.

APRM flux noise will be measured once per shift and the recirculation pump speed will be reduced if the flux noise averaged over 1/2 hour ex9eeds 5% peak to peak, as measured on the APRM chart recorder.

8.

The core plate delta p noise will be measured once per shift and the recirculation pump speed will be reduced if the noise exceeds 1 psi peak to peak.

In the event that MFLPD exceeds FRP for General Electric fuel.

1.0 JE:..r iTIONS The succeeding frequently used terms are ex-plicitly defined so that a uniform interpretation of the specifications may be achieved.

A.

(Deleted)

B.

Alteration of the Reactor Core - The act of moving ony component in the region above the core support plotl!*, below the upper grid and within the shroud.

Normal control rod move-ment with the control rod drive hydraulic system is not defined as a core alteration.

rw~-1 <l r.'\\

... ~\\.

C.

Critical Power Ratio (CPR) - The ~ritical power ratio is the ratio of that assembly power which causes some point in the assembly to experience transition boiling to the assembly power at the reactor condition of interest as calculated by application of the XN-3 correlation.

(Reference XN-NF-512)

D.

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

E.

Inunediate - inunediate means that the required action 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 instrument cali-bration means the adjustment of ~n 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 instrument functional 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-ability by observation of instrument behavior during operation.

This determination shall include, where possible, comparison of the instrument with other independent instruments measuring the same variable.

Aaend*ent No.

1*

,r--.

I

1.

~i~~l!'t Condl!~~"?-'!!!"_~r~!"~'~~U~*~'!!. - 11'*

hai ~ tll0: CC*nJ i I lltllS for 011c ral Holl ll'CCI r, I fac

o.

J.

...... eccc1*l.JltlO levels or S)"IHC.M rerlnr*-

mce *ccessary to essuro 1;.(c U;u tn1* anJ 011-entlon of the facllll1.

hlu:n these cnn11itlon1

- - 11e1t. the pla11t can be n1*cratcd sarcly and

..._,.. a 11tu.u lon1 can lio ufel1 controlled.

U*~~ '~IL~.."!r~ty_~r~!!.!.!i!'~~ 1~1. ll~~-~! - 11ae lfailinc sa(*!l1 sruc11 sc1d11a:s aro ut1h1cs on llbl,..,cut at Inn*'" lch I nit I 1110 tho -illll ou.~* le

,...1cct ho act Ion at

levr.I such that th* 1a(etr 11*111 ul 11 nut bo eacc.:.tc:d.

lite nclon Q.

... wee* tho 1arc1r ll*lt and these 1cttln1s

~*e~cn*s ****** vlth nor*al or*r*tlon "'

klow the: so sen in11.

lite *a rein has been

Obi hhctl 10 Iha& with 1*ro11cr opuallon of the ltitn.. cnt*tl* tho u(ot1 U*IU wUI never be eace~dcJ.

K. Fraction of Limiti For f'Uel fabricateaoy-GE, ~he rrac't1on of limiting power density is the ratio of the Linear Heat Generation Rate (LHGR) existing at a given location to the design LHGR for that bundle type. FLPD does not apply to ENC fuel.

L.

la&lc_~!~~~f.!.'.!'~t Ion "!.~~ - A lo1lc srs*

let9l**nct ion.-1 test *can:a 11 lcH of al I rel*y*

_. co11tac11 n( a Incle circull fro* s-.osor

- aclhatcd device to Insure all c0111110ocnlS are opcr11hlo.. er dc1l1:n intent.

Nhcra a*ossl*

- - . act Inn will co to c*111111*l11tion. Le ** p*-r*

Ill be s&artecl anJ valvtis 11pcn.:J.

M.

!!_i~lir** '-!.!.!.~~"..L!:.c:"..:"!'.' R;ot_!_n_J~)

The

lnl..... ln-coro critic* l'"""' r.11To cerrcsl'on*Un1 to lite *ost H1.1lUn1 fuel nae.a.Ir ** t:1e core.

s.

I. 11Dde - 1"* reactor.odo ls that which ls e~ll*h** Lr t~c *~u-1oloctor-1vltch.

DP~-19 r\\

.~erabl* - A iiJ*l.,.. *1b11Jl'l-, l1*et '* CClllpnhent, or dewlce l'h*ll he orea*ehlo when It lo C*(N'bl~ nl P*rlo1wl11g lte

1*re:lflr" f\\*ncllon(n).

l"'l*Jldt Jn thll do!lnlllun ahell be the ****pllon that *II noc11*::1n., *ltcnci*nt h1c;lt-*n-

- llon, contrnl*, norwnl or.cl eaarct:t.cr ol-trlc*l power

ou1*coil, ccu*IJnc: or acnl ";1ler, h1brlc*llfln or other

- - II 1**7 4P'IUl1aa~1t lhet **** rcfJ'*lre*I for tho *rat..,

""!IJllt..., lroln, co.,.*1..nenl or d"vlc I to perl*n,. Ila l\\*ncllnn(*) *re *lno c*p*bl* or.,.rlo,ntlne their...... t..

aupport Alnc:llon(

l.

~.!!'~~!.!'.& - ClporAtl"C saeana that

9'111tet1, **bQatea.

lraln, c~nl nr dt!vlc* I* r-erlonainll Ile lnlr.nilutl ll\\N1ellona Jn Ile re1p.1h*ud...... or.

~.:1*a1in&_!r.~le -lnterw*l bctueea the encl o:* one rl:Ttlcllh1 ouuco and tho encl of the neat **thscqueat refuel ln1 outa10.

Prl*a~ Cnntalr.ncnl lnte5rltt

Prl*arp contilnacnt lntcirTiilM:ans th*t tho drpell end rrc:ssuro IUf'proutnn challlher are Intact and all of tho follovln1 contlltlons aro sat lsfledi I. All aanu.-1 conUl*enl lsolatloa Yalwca on lines connect ln1 to the reactor coolant 111-tc* or contaln*ont wblch *r* not required to ho open durtaa accldoat con4ltlon1 are c:loscel.

J.

At lcnst ono Joor In e*cb alrlocl 11 close*

au*I sc11led.

J. All auta.atlc contelnaent lsolatloa walwe1 a1*c n1*cultle or dcact lvated In the holat..

potlU011.

4.

All bllnd lla,.1** and manwa11 He close4.

Prourt ha Inst "":'~talion ltcflnltl_!!!!

I.

lnstrwac:nt r.h11r.*cl -

An l.nstr.. cnt chan.

nc I *cnns an a' r*n1t.-.cnt of

jcnsor au*

aual1'ary cqut~ent rc*auhcJ to aenerate and trons*lt *~

trlr *rstc* a sln1I* trip 1icnal related lO tho plant para*cter J

110nhorcd by tln1tt lnstremcnt channel.

r.~ent No.

e

I

r.

~l~~~*~~I Condi~ l~n~- f!!~- ~'.: ~.~t!~n_J~:~'~}. - ll**

hai,.11.:.c:cmdil **,... lor 011cratu111 sa*corr 11.e

o.

J.

- - - 8C:c:c11t.1l1lo levels of S)"IHC:* rerlnr*-

mce necessary to *ssuro 1;.fc st*u 11111 11nJ oa*-

...,,.,. of the foc:llllr.

t.1u:n tltc:so con1litions

- - .,,

the pl1111t can ho C1pc:ra1cJ safely *nd

........ a sllu;a1 luns can ho safelr controlled.

u*~~ t_ng __ ~_"!f~1 1_~r.~!~~ ~~:~~ a~ 1r_ !!~~-~! - 11.e lf*il &nc sa1*.:1 r syuc.. sci dues ;aro :u:fl '" on IMlrt111tcn1111 Inn*.. a.tch lnh lino tho a111ou.~t lc

,_..1ecl ho act Ion at

levr.I snda Iha& th* safotr ll*ll* vi I I not 110 eace.:dcd. lite nclon Q.

... wee* 1ho_s11fe1r ll*lt and these sc11ln1s iN:t*Hcnu **r1i* vhh nor*al oreratlon lylnt tlclOlt 1he:so sell la1s.

Tlao *arc In has been

U.bl hhcd so Iha& wllh l'roa*cr opcr*tlon of Ila*

a..1n111cn1a1ID11 lho safetr ll*lU will never b*

cau.cJcJ.

K. Fraction 6f Limitin For fuel fabricated by GE; the rract1on of limiting power density is the ratio of the Linear Heat Generation Rate (LHGR) existing at a given location to the design LHGR for that bundle type. FLPD does not apply to ENC fuel.

L.

t.slc_~!~!-'.:S_~.!_*~*~l lun "!'.o:s~ -

A lo1lc s1s-let9t11nct inn.al test *c:;anli

lest of 311 rel*r*

_. co.alacts nf a loeh: circull fro* sc.nsor I* Hllvatcd Jewlco to IMura all c:omponcnll

- - opcrt1hle l'er dcsl.:n Antcnt.

Nlacro 11oul..

- - . acalun wi II io au co111ph:l ion, Le ** 1"-r*

I II be s la rUJ and v;a lv.:s.111cn.:J.

N.

~~lirt* '-!.!.!.~~~':"..:"!'.' A*t_!_n_J!!Cl'R) - The

IA(... ara ln-c:oro c:r it i ca I I""'"' rJ tTo

5.

carrcsl'on;Un1 to tlao.o:at Uu1Un1 fuol

.. sc.a. tr 1* t:ae core.

I. W.cle - Th* nactor 90Jo ls that Nhlch ls i~llsh** br th~ e:>Ju-soloc:lor-*vAtch.

DP~-19

~

\\

.~~

- A *r*l-, aub11r:tt-. l1**l '* c....-.11l. or d**lce ah* 11 1111 opera hi o when It Io c;apf'ble rel p*rlo1wh11C I to

1*.,t:lflcct l\\onctlon(n).

1,..1JtrJt In lhl1 dart11lt1un ahall be the ****pllon lh*l *1 l nuc1111r.-1y *Ucncil'11t h1cb-*n*

.;11llun 0 cont ml** no.,... I or.cl eiM. ~Cd.CJ el-trlcal power eou1*co11, ccu*llnc or 11.:nl "°'ler. lubrlc*llnn or other

..... 1111117 *'flll1>0r11t lh*t **** requlrr.*I for the ar*t***

"91h:.r11t..., t1*01n, c.,.,..,,nant. or d1tvlr.

lu perl*n* lta runcllon(*) aro *lno capable of.,orfo,1*lne their r*l*tffl oupport Alnc:\\lon(ol f'l**r..!!_~!.n..& - C'por*tlnc Ae*n* that * *1'11tee,

911h:11atea, ii=i"in, cnmpnncnl nr dP.vlc;e la r-erro... tn11 lte lnlt:t1dvtl f\\111cUon1 Jn Ile re11uh*ad..... or.

~~1*a1 lnLf1.C:I!. *lnlcrval bclveea lh* encl o:* one rt:Tllcitln1 onuco and the end of lh*

neat subscquenl rofuell1t1 oata10.

Prl*a!J Contalr.r.1cn1 ln*esrltr - Prl**ry contilnlacn1 ln1e1rltr *4:*ans that lho drpell

nd rrr.ss11ro suppro11tnn ch11.. 1or *re Intact and all of lhe follovln1 contlltlons ero sat ls fl eds I. All aanual c:ontehment lsol*tln vahres o*

lines connc.:tlna 10 the roac:lor coolant s11-1c.* or conlaln11ont wblc:h are nut required to bo open durJn1 eccldeat con411tlons *re*

closeol.

J.

At lcnst ono Joor In e11cb *lrlocl Is close4

. *n*I 1e11 l cd.

J. All au10.atlc conteln*ent Isolation **lwes arc na*cublo or dcac:1 lv111ed In lhe holal_,.

poslU011.

4.

All bllncl fl*AJ*S end 11anw*r* *re clos.,..

Protertlvo lnstreA:'?>tollon Ucflnltl..!!!!_

I.

lnstr&m1~nt r.ha~Gel - An lnstl'Ulllcnt ~han ncl *cnns 11n *rr11n1e91cnt of a :1cnsor a1'4I

11* II i ;uy e:tui ~ent rcc1uhcd lo 1cnerate and trnns*i* l~

trip 51tlc* a sln1I* trip slcnal related ~o tho plant para*e&er 2

110nilorcd by tk\\ltl lnnnmcnt channel.

1'4111~ent No.

e

.. --*~****-*

1.1 SAm'T LDf.lT 1.1 111t*t. C!.* *,n;tt~:c UITl-:Clt tTY

~i'i' 1'I c:i?t ll l.!I, I

n1c S~(ctt LiMits established to pr~~cr~c th~ fuel cl:iddin& intetrity A?;tlt to these variables t1hich*

r.onll.or the fuel thermal behavior.

Obtt"ctive The objective of the Safety Limits 1~ to establish limits"belov vhich.

the intccrity of the fuel claddin*

u h preserved.*

Spccif ic~tlons A.

r.c~ctor ?rcssurc >800 psig nnd Core FlC'1: > lOl. of R:iti:d.

The CY.istencc of a ~inimum critical power r~t!o (l*:cpn; less than 1. Ot) shJll constitu:c vlolaticn of the HCPR fuel cl~dJinc ihtcgrity ~3fcty limit.

AnJp~-fment No.

I.

DPR-19 2.1 LD11TL'lC SAFETY STSTEK SE'ITINC 2.1 FUl*:t. CLMJDJ::c:

l~Tr.r.t~lTY "rrlicnbl 1t tr.

~

~:.....,

'l'hc t.laltlnc S3f ety Syst~I:' Set tines apply to trip settincs of the instTU-nicnts and devices vhich ue provided to prevent the fuel claddin; lntcc-rlty S~fcty Liolts from bein& c~

cccded.

~bjcctivc n1e objective oC the Limitin& Safe-ty System Scttin&s is to def in~ the level of ihc process variables at vhich auto~atic protective action is lnitl3tcd to.prevent the Cu~l clod-dir.& intccrity Safety Limits (ro~

bcinc exceeded.

Spcc:I f lent.tons A.

Neutron Flt:X Trlp SetUn~

':he limitinc s~f ~ty systc~ trip

~ettints sli01ll be :is specified below:

s

..~------------..---------~---------------------

/

1.1 SAFETY Lltin" DPR-19

- 1t1 I

Amendment No.

2.1 Llt1IrINC S&t"ETY SlSTt:'a smi.~G *

1.,APRH Flux Sera"' Trlp Set tlng (Rt:n Mode)

\\."hen the Teactor aiodP. svitch. is tn the run roslti~n, the APRff flux scra~ setting

. shall be*

s' [sawn + 62]

~ ol vith a maxi~u* set polnt or 120% for cor~

flov equal to 98_ x 106 lb/hr and gnater, 11h8P-I S

aettlng 1n per cent or rated power Vo* ]:M!r cent. ot drlve flow required to pr~uce a rated core f lov or 98 Hlb/hr.

In the event of operation of any fuel I

assembly fabricated by GE with a maximum fraction of limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting shall be modified as follows:

Where* S ~ (.58W0 FRP = fraction of (2527) MWt)

/FRP

]

+ 62 > LMFLPD rated thermal power MFLPD = maximum fraction of limiting power density for GE fuel.

The ratio of FRP/MFLPD shall be set equal to 1.0 unless the actual operating value is less than 1.0, in which case the actual operating value will be used.

1.1 SAFETY LIMIT Amendment No.

DPR-19 2.1 LIMITING SAFETY SYSTEM SETTING This adjustment may also be performed by increasing the APRM gain by the inverse ratio, MFLPD/FRP, which accomplishes the same degree of protection as reducing the trip setting by FRP/MFLPD.

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 APRM scram shall be set at less than or equal to 15% of rated neutron flux.

6A

.. -. -'\\

I DPR-19

. ; 1.1

s.ve:rr tr.*tlT J. C~re Th1rMnl'?ovcr Lfmlt (R~actor.

'!!.!.ssure $.~ 8CO p:; ig)

When the reactor pressure is < BOO psi& o: c~rc f lov is les:1 th~ 107.

of rAttd, the core thermal po~er shall not exceed 2S perc.ent of rated th1:-*a1. power.

1. 11tc neutron flux shall not exceed the scraa setting establl9hed in Speclflc~tion* 2.1.A for* longer than l.S eecunds as indicated by tha proc*** co81puter.

%. 'When th* process compll.ter 11 out of service, thls 11(et7 llalt shall be e.s::urM.t to be e*ce1d1d 1( the neutron flux exceed!: the ~or:un setting *stabllshed by Speclflc*tion 2.1.A and

control rod scr** does not occur.

D_,

At~chr Val*r Level (Shutd,,~n CondlUonl

\\lkene"T the reactor l9 in the shutdown coridltlon with lrradlat1J f~el In the reactor v*s~el, the

"At*r. level f.hall not be less then that corre!O-pondlng to 12 Inches ~bove the top of the active f~e**~hen lt is s~~ted in the core.

Top of a~ttve fuel is defined to be 360 1nch~s above vessel zero (see l1:-t:.1C:l.J. 2).

. I 2.1 Lil'ilT:UiC SA!*.r.rt STSTE:t smn'C The UK flu* sctalft ~'!ttln~ shell be set *t lts~ th*n or equal to 120/125 or full scale

B.

~M ltcrl Block Set.ting Tho APR~ rod block 1ettlnc *h*ll \\e:

I S ~ [.58w0 + 50]

. 11\\e.ddJ.nltlons used above for 0the APP.It sens trip':Ji'ply.

In the event of operation of any fuel assembly fabricated by GE with a maximum fraction limiting power density (MFLPD) greater than the fraction of rated power (FRP), the setting~

shall be modified as follows: tt

_l s '(.58WD + 50) L~~p~

The definitions,used above for the APRM scram trip apply.

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

In which case the actual operating value will be used.

The adjust-ment may also be performed by increasing 1 the APRM gain by the inverse ratio, MFLPD/

FRP, which accomplishes the same degree of protection as reducing the trip setting by FRP/MFLPD.

J\\mPnrtmP.nt 7

1.1 Sa' ~Y Limit Bases FUEL CLADDING INTEGRITY The fuel cladding integrity limit is set such that no calculated fuel dam-ag~s 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 crit*ical 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 I

integrity by assuring that the fuel does not experience transition boiling.

The fuel cladding is one of the physical barriers which separate radioactive materials from the environs.

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 during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation signi-ficantly above design conditions and the protection system safety settings.

While fission product migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforation signals a threshold, beyond which still greater thermal stresses may cause gross rather than incremental cladding deter-ioration.

Therefore, the fuel cladding Aµi!=nqµient Nq.

DPR-19 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 design for planned operation.

The MCPR fuel cladding integrity Safety Limit assures that during normal operation and during anticipated operational occurrences, at least 99.9% of the fuel rods in the core do not experience transition boiling.

See reference XN-NF-524.

A.

Rea~tor Pressure > 800 psig and Core Flow > 10% of Rated Onset of transition boiling results in a decrease in heat transfer from the clad and, therefore, elevated clad temperature and the possibility of clad failure.

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

Therefore, the margin to boiling transition is calculated from plant operating parameters such as core power, core flow, feedwater temperature, and core power distribution.

The margin for each fuel assembly is characterized by the critical power ratio (CPR) which is the ratio of the bundle power which would produce onset oftransltlo'lim..

boiling divided by the actual bundle power.

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

It is assumed that the plant operation is controlled to the nominal protective setpoints via the instrument eel variables.

(Figure 2.1-3).

The MCPR Fuel Cladding Integrity Safety Limit assures sufficient conservatism in the operating MCPR limit that in the event of an anticipated operational occurrence from the limiting condition for operation, at

s~ ;ty Limit Bases 1.1.A Reactor Pressure> 800 psig and Core Flow> 10% of Rated.

(cont'd) least 99.9% of the fuel rods in the core would be expected to avoid boiling transition.

The margin between calculated boiling transition (MCPR=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 methodology used in determining the MCPR Fuel 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 estimated.

The assumed reactor conditions used in defining the safety limit introduce conserv-atism into the limit because boundingly high radial power peaking factors and boundingly flat local peaking distributions are used to estimate the number of rods in boiling transition.

Still further conservatism is induced by the tendency of the XN-3 correlation to overpredict the number of rods in boiling transition.

These conservatisms and the inherent accuracy of the XN-3 correlation provide a reasonable degree of assurance that during sustained operation at the MCPR Fuel Cladding DPR-19 Integrity Safety Limit there would be no trans-ition boiling in the core.

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

Significant test data accumulated by the U. S. Nuclear Regulatory Connnission and private organizations indicate that the use of a boiling transition limitation to protect against cladding failure is a very conservative approach; much of-

~

the data indicatesthat LWR fuel can survive

~

for 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 specified in Specification 1.2.

For fuel fabricated by General Electric Company, operation 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 margin for transients initiated from lower power conditions by ad.iustirH? the APRM flow-biased scram by the ratio of FRP/MFLPD.

Specification 3.5.J establishes the maximuni value of LHGR which cannot be exceeded during steady power operation for GE fuel types.

For fuel fabricated by Exxon Nuclear Company, (ENC) fuel design criteria have been established to provide protection against fuel* centerline melting and cladding strain, ENC has performed 11

Safety Limit Bases 1.1.A.

Reactor Pressure> 800 psig and Core Fuel> 10% of Rated.

(cont'd) fuel design analysis which demonstrate that centerline melting is not predicted to occur during transient overpower conditions throughout the life of the fuel, Protection of the MCPR and MAPLHGR limits and operation within the power distribution assumptions of the fuel design analysis will provide adequate protection against centerline melt and ensures compliance with ENC's clad overstrain criteria for steady state and transient operation.

Since ENC's design criteria are more conservative than the 1% plastic strain limitation on GE fuel, the LHGR limitation and APRM scram adjustment for GE fuel established in specifications 3.5.J and 2.1.A.l respectively are unnecessary for the protection of ENC fuel.

The procedural controls qf specification 3.1.B will ensure that operation of ENC fuel remains within the power distribution assumptions of the fuel design analysis.

DPR-19 lla

1.1

~~r~tv Ltm1t n~sc3 1.1.C Po~cr Transient (cont'd)

Q.....

c..J

~

w.;;.

~

c**

~h~ ~o~puter provtdcd hao e

~e~ucnc~ ~nnunc~atton program*~h1ch

~111 1nd1c~te the 3equence in which

.acr~ms occur such e3 neutron flux, p~c~sure, etc.

This pro~r2m also 1~11cct~3 ~hen th~ scr2rn s~tpotnt 13 cl~ar~d. Thls ~111 prov1~e tnfor~~t1on on ho\\*; long a scrnm cond 1t1on ext:; ts 2nd thus provJde some mca3urc of the cn~r~t added durin~ a tr~nst~nt.

Thu3, co~puter ~nformatlon normnlly ~111 be av~ t lc:b le for 2nz lyz In~ 3cr,,m3; ho"*-

cvc~, lf the computer 1nformJtton 3hould no~ be ~v~tlable for any scram an~ly31s,

~pcct!tc~tlon 1.1.C.2 ~111 be rel1ed on to d~tcrmtnc 1f a a2fcty llmlt han been v ~ol? ted.

DurJn~ periods when the rcoctor ts shut do~n, con~1dcrat1on must also be ~1vcn to *.;;_,tcr l~vel :-c:~u1rer:lent3 due to the crtect or decey hc~t. If reactor ~etcr l~v~l 3hould drop bclo~ the top or th~

2C~~VC fuel ourlr.~ thl3 t!me, th~

eb!llt~ to cool the core 13 rcduc~d.

Th1~ reduction ln core coolln5 c~p-

> llit] coul1 lc1?d to elev~t~rJ cl:iddtng t~~~~ratu~c~ an~ clcd p~rfor~tlon. The co~~ ~111 be cooled 3uff1c1ently to prc-v~nt clad melting should the ~~ter level b~ r~duccd to t~o-thlrds the ccre he1~ht.

!::.; tz~ l.ish!r.~nt or the :Jfl re ty llm t t "t 12 l:1ch~z ~bov~ the to;> of the rue J* pro*1 hies i'*lc*!Uatc n:~:-~tn.

'fht:> level \\"ttll be con-tl!lt;Ot;'lly montto!'c1 hhcn~vcr tl!c rt~cir culct.!o:t pu:i!p:>

~r'! not oµcrilt1:~:;.

l

- Top ~f active fuel is defined ~o he 360 1nch~s above vessel zero (see Bases 3.2).

2.1 DPR-19 Ltmtt1n$ Sofcty Sl_~tcm Setting B~~c~

FUEL CLADDING INTEGRITY The.cbnorm3l operational trans1ents applicable to op~r~t1on or the untts have been an2lyz~d throu~hout the spectrum or planned oper~t1ns con-d it1ons up to the rated therm~l pow~r condltlon or ~~*:*7 1-:'"'t*

An odditlon,

2'j~"1 i'i~*lt 1s the ltcensed max!:num stcad~

sta tc power level or the units.

This maximum st~ady-st3te po~cr level wtli

~

never knowln~ly be cxc~cdcd. See referencel ~

XN-NF-79-71.

13 Amendment No.

Limiting Safety System Setting Bases 2.1

_FUEL CLADDING INTEGRITY (cont'd)

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

Variables which inherently possess little or.no uncertainty or whose uncertainty has little or no effect on the outcome of the 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 95% of the random statistical combinations of uncertafnties.

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

S~eady-~tote operation without ~orccd r~otrculat1on wtll not be-pcr~1tted, except dur1~~ atartup test1r.~.

The an~ly313 to support O?Cr~tlon ~t various po~er 2nd flo~ rel~t1on~hlp3 h~3 considered opcrct1on wtth c1thcr one or two rec1rculatton purnp3.

Th~ ba8ea for 1nd1vldual trip !etttn~s are discussed 1n the followtn~ para-graph3.

A.

1.

DPR-19 For an3Jyses of the thermal consequences of the transients, tho MCPR's stated j.n p3ra.1raph l.S.IC as the limiting conc1it1011 or opr.ration b

d those which are conservatlvely asnu~ed t~u~xist prior to initiation of the transients.

Neutron Flux Trip Sett1n~s APRM Flux Scram Trip Setttn5 (Run Mode)

The average power ran~e mon1tortng

{APRM) sy~tem, which ls calibrated.

ustng heat belance data_ taken during

9 st~ady-statc dond1t1o~s, reads 1n pcrce~t or rated, thcr~nl power.

Bc-cau3c flgslon cha~bers pro~:je the bas1o tnput :sl~n~ls, th~ /,Pfti-1 s:1st~::t rcs9onds d trectly to t1\\*cr:igc :icutron flu~.

Dur!n~ trans lc:tt.3, the 1n3 tau~:>neous rate* or heat tr<insfcr rror.t the fuel (reactor thcr;na l po\\*;cr) ls lc3:J than the lnntantancous neutron flux due to the ttm:? con:;tant of the fuel. *there-fore. durtn~ abnorr.-ml opcrat lonnl trilnslcnts, the thcrm:>l po'-ier'or the fuel will be lc:;s than th3t lnd1c~tcd by the neutron flux at the scr2m sctttng.

nn~ly~c3 dcmor.~tratc th3t wtth n 120 percent sr,ram tr!~ sett.in~. none or the abnorm~l opcr~t1on~l tr~nstcnto1an~lyzcd vjolute tl14? fu-?l S:.1fct1 Mr.!~t :tnd tt'!~re

~:.

19 a

~ub3t<tntJal ni3r~tn rror.t rw*l ~:"r.t~;e

0J'hcrcfClrc, t.hc unc or floH :"t"!fc:r*:nc!'ll sc~~m trip prov1dc5 cv~~ nd~!tJ?n~l n~r~~*

14 Amendment No.

J.l.A. Neutron Flux Trip Settings

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

An increase in the APRM scram trip setting would decrease the margin present before the.f~el.cladding.

integrity Safety Limit is reached.

The APRM scram trip setting was determined by an analysis of margin~

required to provide a rea~onable range for maneuverin~ 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 adequate margin for the fuel cladding in~egrity Safety Limit yet allows operating margin that reduces the ~ossibility of unnecessary scrams.

The scram trip setting must be adjusted to ensure that the LHGR transient peak for G.E. fuel is not increased for any combin-ation of Maximum Fraction of Limiting Power Density (MFLPD) and reactor core thermal power, The scram setting is adjusted in accor.dance with the formula in specification 2.1.A.l. when the MFLPD is greater than the fraction of rated power (FRP).

The adjustment may also be accomplished by increasing the APRM gain by the reciprocal of FRP/MFLPD.

This provides the same degree of protection as reducing the trip setting by FRP/

MFLPD by raising the initial.

APRM reading closer to the tn.p setting such that a scr~m would be received at the same point in a transient as if the trip setting had been reduced.

APP.~ t1ui SetaM Trtp Setttn5 jft.. ~11!1.ot ltl'I'~ 81Jiot S':ie11db7 Mode) re~ *~!~t,ton th the 1tnrtup mode while lht re1ator II 1t lo* pret,ute, the APRM icra~ 1etttn1 or 15 percent or rated power

~ro~ldt1 1dt~u1t1 thenr.31 mar!ln between the th* 9,~~11nt tr.d the aRrctr li~lt, 25 p!r-

"' or r3ted.

The margin h *de-1uete to tef~:ti:n,dote en,tctpated moneuwcrs *u~eoctated

~~~fl ito';~f' t>>hnt etortup. £rreots or in-i~c3,1n~ ptfmuure et zero or low *votd con-t':!:a& er-e tdtaoa*, eold wat~* rror.t aource~

f'*:*dltt1!1 d*.:r!~~ :.tertup ts hot 11111ch colder than that already in the svstem, tempera-ture coerr1c1ents-:-are emall, end oon-t.rol rod patterns ere constratned to be.unttorm b7 operet1r.s procedures b\\lcked up by the ro~ worth 1111nlr:!izer.*

or ~11 possible so~rcc' or renctivlt~

input, unSfo:-m cont~ol rod withdrawal ts the most probabl~ cause or aign1t1**

c~nt power rtsc.

E~cau3e the tlux d13tr1bution 2sooc:e~t'd ~ith u~srorm rod with1r~~313 do~s not involve hl~b local pcak3, end bec~use several rods aust be moved to ch~nc~ power by a a1gn1Clcnnt p~rcent~~e or rated power, th~ r~te or po~~r rl~c ts very 110~.

Gencr~lly, the hc~t flux ls tn -ncnr C<t~111b=-1una *:!.th tt':e r1:1aton rate~ In an as3u~cd unlf~r~ rod ~1thdrawol ap-pro3ch to the sere::: level, the rate ot p~~~r ri~e is no ~~~c than*5 percent or re?ted po~cr per ~!.nute, and the Attn:.; :;7l!te::: would be moro th!n adequate 15 Amendment

2.1.A.

Neutron Flux Trip Setting

2.

APRM Flux Scram Trip Setting (Refuel or Start and Hot Standby Mode) (cont'd) to a9aure a Gcrn~ terore th~ po~cr could exceed tll'! sa!'ct7 ltm1t. The 15 p~1*cam: f,;>;U*i scr_,!'!': r:~:l i:l9 act1\\*c un-t!l the ~ode a~!t~~ 1~ pl~ccd tn th~

RU~ PQ~1t1on. Th!3 ~~1:c~ occurs ~hftft reactor prcs:aure :~ cr~at~r than 850 D:l1g.

JRM Plux ScrP~ Tr!~ Sct~1n5 The IRM a7stem con=tsts or 8 chambers.

in each or the re~ctor protection system lo~!c chan~e!s. The l~K ts a 5-dcc~de 1n~tr~~~n~ ~htch cov~r3 the ran~c or po~cr levzl bct~ccn th~t covered by th~ s;;:.; :tm! the.A?P.;*t" The 5 dr.codes ere bro~'!:I <!c~n into 10 ro!'lges, eoch betnc one-hal!' or a dec~de 1n sl~e.

DPR-19 15a

. \\, *.. r,.... *~ j.**

-~

2.t.A.

JCeut:-on Flux Trip* *Setting*

3. u::4 Flux Screm Tr1p Set~tng (cont'd')

The lRH acrem t~tp aetttng ot*l20

~1~l~~on~ 13 ecttve ln eech ~an&e.or th9 1PJ4.

For c~O$?le, tr the 1n3tru-De~: were on r~r.cc 1, the scr3m ~etttng

~ould be e 120 dlvtstons tor that ran~e~

11kcwl3~, lt th9 1n3trument were on range 5, the acra~ w?uld ~e 120 dlvislon= on th~t run~e. Thus, os the lR~ !3 ran~ed

\\:? to occo=odtJt'? the lncrea:ie tn power l~vel, the acr~a trSp aett1n~ ts al~o rented up.

Th~ r.ost st~ntt1cant.~ources or rea~

t1*1t:; ch~nge durln~ the power lncreose are ~'-=c*to control :-o~ ti1thd:-awnl. *Jn orde:- to ensure thet the IRM provided ed~~\\!ntc* prot'lctton aga1n3t the. ainGle rod w!thdra,;cl cr*~or, o ran~e or rod "lth~:-c~;ol accid~mts "PS Pn;ily~cd *. Thta e~::!1c::i tnch:c!ed 3terttns.the ccc!dcnt

~= v3rtou~ po~cr levels.

The ~o~t s~-

vcre cc:ae 1n~?l 01c's a:i 1n1t~~l concJtt1on Sn tth!ch the re~ctor la Just eubcrtt!cel c~d ~h~ lH~ crot~~ is co~ v~t on s~~le.

/.cJ'!!t!ou:!l cor:oervatinm ~r.s tnk~n 1n this e:a~lysi~ bJ ~1s:;ut11h:~ that the llli-1 ch:tnncl clo:s~:.t to the w:.:t:*Jrn"° rod 1!J byp~~:Jcd.

'A"h~ 1 esul ts. or th ta o:aP. l)*s Is 3no~*: tho t the rc::ct.or ta ac?":u:*.'..:c1 :?:id per.~ po~er l l~j :"!d 2.1.11

~., 0:1~ pe:-C'!:'!t. or rate~ po\\*:er, ti1~3 111aintaininq

~:CPR above the ~CPR fuel cladding integrity safety limit. Based on the above

~;V'!l:;:::s, th'! !B:*i p:-ovldes p:-otect1on a:;~1nst loc:o 1 co:1t:-ol ro~ \\>t lth~:-e~'a 1 cr:-o:-s nnd con-t*J 110*1:::

-..!th*J:o::~.;:Jl Of Ct>ntrol ro:J:J Jn ~c::'-"'.'!lCe c.~:! i'l'Oviuca bu~l<up p1*utcci;10:1 1*or the 1'?!~:*1.

Amendment No.*

DPR-19 APRM Rod Block Trip Setttns Reactor po~er level m~7 be *erted b7 movtn; control rods or by ve17tng th~. r~o!rcul2t1on flow rate.

The APR:4 syatcm prov1~ea a control rod block to prc\\*cnt qross rod withdrawal at coa:> tont rec lrcul:?tton i'lo~

rnte to prot(!ct ~r:"J*1n:it grossly exceed-ing -the MCPR fuel claddinq lnteqrlty safety limit.

This rod block trip ~cttlnJ, wh1ch ta e~tc-n!nt 1cally vcrled w 1th rcct.rculat1on loop Clo~ r:!te, prevent~ an !ncrc~ae*

1n th~ reactor p~wer l~v~l to e~crs-o lve vnluc::J c!ue to control r~ ti~th dra~~l.

'l'h'? flo.,; vGri:lbl~* tr lp setting*

provt~c!i ~ub!l t:1:1t Se l :"!l..,r!; tn tro:a f.&.:e l dDr.i~ce, n:s:su: ~J.ns o a-teo~1-1:t:1ta c;>c-a-t~on ~t ~ho t~i? ~e:tl"~' over th~ *

'!nt lre rec J re a lat Soil flow range.

'!"Zl*~

n:t:-g In to the Sate tr Ll:n!t tncr'!:>~C-:J :!8 th~ Clo~ dec:-c:~~c~ t~:- the D"'*'c~-~**J trip sett1:~:: '"er:n:s Clo~ rel~tte:.:~*pJ

..:-:'!re. 01*e the ~-:orot c:!~e HC?!l ~;htch cc~jd occur du~~~5 a~e~~7-::Jt~tc o;>e~3-

. t!cn ts l!t lOC~ ol' r~tctJ th~r;:::-1 po~:er bec~u3e or the APRM rod block tr~;>

ac:t!~~. The nct~al p~~'!r d!o:r!butlo~

1:i t.1c core is c:Jtn:,Uuhed by t:i'ecH'!ed co:itr:>~ red DeiJu~nces and ts r.:o:iH.o:-~d cc~ t !!'!~O\\::; 11 b;,' t.he in-core tr~:: s*:: te::s.

~3 ~iL~ ~~e APRM ~cr3~ ~r~n *~***~~

ti

, 6..

.,... _ ""***~*

~ A nn r~~ ~lQck t~1p =ettlns S~ ~d-just~d downward or APRM gain increased I if the maximum fraction of limiting power density for G.E. f'uel exceeds the fraction of rated power, thus pre-serving the APRM rod block safety margin.

16 r..~

. r--

~. Turblno S~<'J? Vclvo Sen.n - Tho ttirblne ete>A> Yal.Ye clo~\\:l""! nc::aa trlp ri.:lt1cl~tcs tho prcs:.uro, r.~u\\ro~ fl~ or.~ h~A\\ flux 1ncrc~~o th~t could re:;u1t. rrc:'.I =i>>1cl e!.cau~s of' the turblno :;top

~lv~o. n1t.h A scr.:.A t.1:11> Gct.Unc or 10 p!rccnt. or wbe cl?~uru fron full open, tho re:;ul~t. lr.creose 1n aurr:ice heat. flux ls llm \\cd such lM.t. MCPa rc,..aln:t above 1 the* MCPR fuel cloddiftCJ integrity safety limit, even during th~ vorst case transient th~t ~ssum~s the turbine bypass ls closed.

F. CcncTa:nr 1.oad Rdectit>n Sera" - The geaera-tor lo=d rejection scraa is provided to onticlpatc the rapid increase in pressure and ncutroa flux Tesulting from.

f &:Jt ClOSU:'°l of thct turbine Con~rol Y&l'YeS due to A l~ r:ject.lo~ end subsequent.

r~ll\\a~ or the 'b7!'2'-":;: 1.o.' lt. prc*ent.s i*~:r:! :rc:i tieco~r.~ le.:;:s *than the MCPR fuel cladding integrity safety limit for this transient.

For the load rejection without bypass transient from 100\\ power, the peak heat flux (and therefore LHGR) increases on the order of 15\\ ~hich provides wide margin I

to the value corresponding to fuel centerline melting and 1% cladding strain.

Amendment No.

DPR-19 C.

Rcnctor Coolant Lov Pre9suTe lnitifttes Mnin Stcnm Isolation Valve Closure - The low pressure isolation nt 850 psig vas provided to give protection ag3inst fast reactor dcpressuriz3tion ond. the resulting.

rnpid cooldown of the ves.!'lel.

AJ,,*ont3ce vos to~en of the scrnm fC?aturc which occurs vhen the Min steam line isolation valves ore clo~cd.to provide for reactor shutdown so that operation ot pressures lover t:um those spc:ciflcd in the thcrr.tal hydraulic safety limit docs not ~ccur, althouth operation ot a pre~suce lo'-.?l" tl1on 650 psig wo\\lid not necessarily constitute an unsafe condition.

B.

H:tln Stenr.s Line Isolation Volve ClosuTe Scram - The low pressure isolation of the moln steam lines at 8S0 psig VOS provided to give protection against rapid re3ctor dcpressurization end the.Tcsulting rapid cooldovn o( the vessel. Advantage was.tokea of the scrom fC?11ture which occurs vhcn the main stcom line isolation valves are closed, to provide for reactor shutdown so "that high power operation at lov re~ctor pr~ssure does not occuT, thus providing protection for the fuel*clodJlng integrity safety limit. Operation of the reactor at pressures 101a"er than 850 psi& requires that the re3cror lllOJe switch be in the st1irtup,position vhe 0re protectloh of the fuel cluddlnt lntccrity s01fety Umlt ls provlJ~J by the llt>I lll:;h neutron ( lux scr01111.

Thu9, the combination of main ste~m line low pressure lsol:ttlon 01nd isol3tlon valve closure scram assures the avallablJ.ity of neutron flux _scrom protection over the entire ronce of oppllc:tbllity of the fuel cJndding integdty

~fety limit.

In addition, tltc isolation vnlve closure scrom anticlpo_tes the pnssun and flux transients which occur durln& nora.,1 or inndvertent hobtlon vnJ.ve closure. With the scrr.*s set at 10? valve closure,thcre ls rio appreciablP incrP~se in neutron flux.

18 I

1.2 SAFETY LIMIT 1~2 REACTOR COOLANT SYSTEM 1

~plicebtlity:

Applies to limits on reactor coolant system pressure.

Objective:

To establish a 111111.t below which the.integrity of the reactor coolant system is not threatened due *to an overpressure condition.

Sped fication:

11le reactor coolant system pressure shall not exceed 1345 psig at any time when irradiated fuel is present in the reactor vessel.

Amendment No.

DPR-19 2.2 LIHITING SAFETY 0SYSTf.H SEl'TING 2.2 Rf.ACTOR COOLANT SYSTEH App licabll ity:

Applies to trip settings of the instrumC?nts ond dcv!....:o which are provided to prev.:nt the re:ictor.

syBtem safety limits from belng exceeded.

Objective:

To define the level of the process vari:ibles :it which automatic protective action-is initi:itcd to prevent the safety limits fro11 being exceeded.

Specification:

A.

Reactor Coolant High Pressure Scram shall be

~1060 psig.

B.

Primary System Safety Valve Nominal Settings shall be as follows:

1 valve at 1115 psig*

2 valves at 1240 psig 2 valves at 1250 psig 2 valves at 1260 psig 2 valves at 1260 psig TI*e allCNable setpoint error for each valve shall be +U.

Target Rock combination safety/relief valve 19

. e

t.t DPR-19

'nle nectoT coot.at *79tftl lnterrtty h llft ft11por-tant barrier Sn the Pt'eWnUon of Wltontrolled t'e-le*** of fl**ion product*.

It t* eeeentlal that the.

lntesrtry of tht* ~J*tem.. protected b7 **tabltehinl a pressUTe llat.t to he obeerved for ell operating ccndltlons and vhene*er there I* Irradiated fuel *~

the reactor..... t.

The pressure safety limit of 1345 psig as measured by the vessel steam space pressure indicator ensures margin to 1375 psig at the lowest elevation of the reactor vessel.

The 1375 psig value is derived from the design pressures-of the reactor presse1 vessel and coolant system piping. The respective design pressures are 1250 psig at 57SCF and 1175 psig at 56CPF.

The pressure safety limit was chosen as the lower of the pressure transients permitted by the applicable design codes:

ASt-E Boiler and Pressure Vessel O:lde,Section II I for the pressure vessel and USASI Code for the reactor coolant system piping. lhe ASt-£ Boiler and Pressel Vessel Code permits pressure transients up to lrJA; power over design pressure (llrJA; x 1250 - 1375 psig), and the USASI Code permits pressure transients up to 2~

over the design pressure (12~ x 1175 ~ 1410 psig).

The Safety limit pressure of 1375 psig is referenced to the lowest elevation of the reactor vessel.

The design pressure for the recirc. suction line piping (1175psig) was chosen relative to the reactor vessel design pressure.

Demonstrating compliance of the peak vessel pressure with the ASME overpressure protection limit (1375psig) assures compliance of the suction piping with the USASI limit (1410psig). Evaluation methodology used to assure that this safety limit pressure is not exceeded for any reload is documented in Reference XN-NF-79-71.

The design basis for the reactor pressure vessel makes evident the su~stantial margin of protection against failure at the safety pressure limit of 1375 psig.-

The vessel has been designed for a general membrane stress no greater Amendment No. 47 than 26,700 psi at an internal pLcssure of 1250 psig:

this is a factor of 1.5 below the yield 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 strength.

The relationships of stress levels to yield strength are comparable for the primary system piping 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 drywell, operating in the event that the turbine bypass should fail.

Finally, the safety valves are sized to keep the reactor vessel peak pressure J 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 actuation provide adequate margin below the peak allowable vessel pressure of 1375 psig.

Reactor pressure is continuously monitored in the control room during operation on a 1500 psi full seal~ pressure recorder.

(4) SAR, Section 11.2.2 -

also:

"Dresden 3 Second Reload License Submittal," 9-14-73 20 also:

"Dresden Station Special Report No. 29 Supolem~nt B."

2.2 IA mmpllnce with Section IU of the ASttE ODde. tt.e oafr.ty **l*H SU8t k *et to open at no hlRhe" thon lOll of 4esllP' pnHure. and th~.._t ll*tt the t'f!actor pr**eure :o no 910re than 1101 o( d"9111' pttn!lure.

llnth thf' neutron flux.,cra:a an" earety

- lve actuatlnn are requl~ to pre*eqt OY~rl'r~.-

9\\:rlr.lng the rc"ct.,r prea.. ure weo!lel and tbu9 exceedin& the pre~**re o~(et~ ll~lt. The pren..ure scram ia available aa a backup protection to the direct valve position trip scrams and.the high flux scram.

If the hi9h flux scram were to fall, a hi9h ~resauro scram ~-nuld oeeur at 1060 pd9 *. Analyses are per formed as described in reference XN-NF-79-71 for Amendment No.

each reload to assure that the pressure safety limit is not exceeded.

DPR-19 21

. 0

. '* t LIKrftltC CCl'fDttto* roa OPEMTtOIC l

\\. 3.l tr.Ac:tOR Pt.OTtCTIO~ SYSTF.:t

!f!lhnl>>llh%1

.Ar,.tltt to th* lnttl'U11cnutton 11nd

~1soc!at~d-~evtce,~hlch lnltlat* a te1ctor aero** *

~

J!.tl!:!! I to uttJro tht operabllltJ of th*

reactor *rot*etlOft *J*tea.

ttteel flc:ttlont A. the !let:wot11tt,..-tn!~ 1\\u:o?ler of trip tJflteM, ~.s !'llnl:"\\!:9 n"~cr of inst~

t*!ttt duumeU t!tnt !'\\tnt bo. Ci'enble fnr ench ;o:ltlon of the reactor 9!0de

.. 1ltch 1h~J1 be "' ~lven in t11ble 3.1 :1.

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

If durin~ oneration, thP. m::iximnm fraction-of* limiting power density for fuel fabricated by GE exceeds the fraction of rated power when operating above 25% rated thermal power, either:

r I

- a.

The APRM scram and rod bock settinr shall be reduced to the values given by the equations *in Speci-fications 2.1.A.l and 2.1.B.

Amendment No.

DPR-19 4.1 SUP.VEILl.Nfet ll.EQVllm~ttN'f 4.1 Rri\\CTO:\\ PF.OTF.CTIO~ SYSTEM t;ro l !Eb ll ttv 1 Afipllcs to the survcUl.3nce of the tnstTUftlCD*

tntl~n ftnd ~'soclated devices vhich lnltlate.

reactor scram.

~r.ctlvr.:

To specify the t7pe 11nd frequency of eurvetll:ince to be applied to the protectloa Snstruir.~ntotlon.

Spccl Uc:'!t lon:

A.

Inr.tr~-n:~ntction systeirs shall ba (unctlon:illy tf!stcd anJ C3Ui>rnted :as ln~tcatcd ln Tables 4.1.1 L,d 4.1.z, r.:spcct lvely.

I. Daily during reactor power ()peration ~hove 257. rated thermal power, the core power distribution shall be checked for:

1.

Maximum fraction of limiting power density for fuel fabricated by GE (MFLPD) and 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.

22

3.1 LIMITING CONDITIONS FOR OPERATION Specifications (cont'd)

b.

[

The power distribution shall be changed such that the maximum fraction of limiting power density no longer exceeds the fraction of rated power.

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

Amendment No.

" DPR-19 4.1 SURVEILLANCE REQUIREMENT 22a

n h:itr sc.-r:im :tnd rod hlock condition. 111us, 1r.lhc c:1llhr:tllon were performed durlns: oper..

iltnn, nu~ !';h:11>lni: would nol be possilJlc.

U!1st*d "" ~:11l 0 rll'ncl' nl olhcr 1:cnc:r:tllni; st:tllon!'l, ~lrUI of Instruments, such :is those In the 1:1'" Ui:"tsln,: t\\cwork, II' not sli:nific:mt nncl lhcn*forc, to :froid s1mrluus scrams. :t c:1lihr:ttinn frcf1ucncy or c:ic:h rcfuclins: oitl:t&;t.

Is c:i.cl:1hll:ihl*d.

Crou1* (q cfc,*lccs nrc :wth*c only durln:: n s:h*c:n ltttl'llun nr l!ac OllCrnllon:tl cycle. Fur cx:11111th!, lhc 111~1 Is *:iclh*c clurlns: sl:1rt111> :tnd in:tcti\\*c eh1ri11:: full-1>uwer utter:illcm. Thus, the only h:st lhal Is 111c:1nln::ru1 Is the 0:1e 1ter-furmcd Jut:l 111*lor tu tilntldo\\\\*n or st:u*tu1*: I. c.,

the tests th:it :ire 1K:rformcd just l>l'lor lo use or the lnslrnmcnl.

Callhrnliun rrcf)t*Cner or the lnsln1mcnt ch:tn-ncl Is clh*trlcd Into '"'o s;rou11s..111esc are :a~

follw*s:

l.

1,Mslve? ly1te lndlcallns: dc,~lccs th:it c:an lJC l 0011111nrcd with like units on a conllnu-

° h:utis.

2.

V:icmnn lube or semiconductor devices

n:ul 1h.*:ccturs lhnl drill or iose scn~Hh*il)'.

Experlcnl'C wllh p:issh*c lypc Instruments In Co1n1nunwrnllh *:dlsnn J:l'ncr:tllns: sl:tllons :ind subst:11h***-' inrJlc:ilcs th;,l thc specified c:i1i1Jra-lic*ns :i1*c :u!N1u:itc. Fur lhnsc tlc\\*lccs which cnt11loy :*111pliricrs, etc., d:*lft s;1t'<:Uic:ilio:1s c:1H for *!rilt to bo lcss.lh:tn U.*t*:~./innnth: I.e.,

In lhc 1wrlnd er a month :e th*irt o!..t"i would o<:c."tlr :uul thus providin;:- for :u!cqua:e m::r;;in.

For: the APnM systC'n1 drift of clcctro:ilc

spp:tr:ttu' Is not the only consideration In c!c-tcr1>>lnlni: :\\ c;,libr:illon frcqucnt'y. Ch:m.:c In 1u1wc.*r clislrlhullo:i :uu.I loss or ci*:imbcr sc~sl l h*H~* rlh:l:tlc :i c:-i lllll':tt i un c\\*c.*n* ~c\\*cn rl:iys.

C::tllhr;illun on lhls rrcc,ucnc.:y :i.~:;urcs 1>l:111l opcr:elion*nl or bl'luw lhcrm:1l l1111its.

,\\ l'Or.tp:UISUl1 or "l~i1hlCS."I, I. l

11~tJ *I. l.:?

lndh:alc.:s lh:il !lix instn::n..:nt ch:1n:1cls h:wc. not IJccn lnl'lm!cd In lhc l:1tl.:r l';ihlc. These :ire: *

~!u1lc Switch in :>!mtcia\\\\ n, )!:1:1u:1l Scr:tm, lll:;h

\\\\';ilcr Lc,*cl In Scl':IOI ltisch:iq~c T:u~k, M:1ln Slc:un I.Inc l:;ol:!llun \\'ah*c Clo:nirc, Gcncr~lor l.o:ul ICcjcction, :tm! 'l'urhinc ::;1111t \\':1h*c Closure. A!I ur tile dc\\*icc$ UI' !tt.*n:;un nssuc:l-

.h:cl wil!1 tht.*sc scr:11u funcliuns :.re i.lmph:

on-orr swill-hes :ind, lu.~nc.:\\!, ca!l!ll':1tiu11 hi nol

tpplic:"'k, I. c.,.lhc :.witch I~ dthcr un ur.

uU. Ftirlher, these swllchl*:s urc mmmtl-d

ol icll1* lo the dc,*h: c :uul !ul\\'C n \\'Cl'Y low IH'1th:1hllily or 1110\\*ln;, c. :;. the swilc.:hcs In the scr::111 disch:..q~c:,*olumc t:ink. l::1scd on the ;1ho,*c, *no c.*allhr:1liun Is n~c1:1lrcd for these asix lns11*mnc:nl l'l*:im\\r;l:s.

B.

The MFLPD for fuel fabricated by GE shall be checked once per day to determine if AmPnrlmPnt" Nn.

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 control rods are moved daily and 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 distribution assumptions of the fuel design analysis for overpower conditions.

During periods of operation beyond these power distribution assumptions, the APRM gains or scram settings may be adjusted to ensure consistency with the fuel design crit e"""&'

for overpower conditions.

DPR-19 34

~.

3. 2 LIMITING CONDITION FOR OPERATION C. Control Rod Block Actuation

1.

The limiting conditions of operation for the instrumentation that initiates control rod block are given in Table 3. 2. 3.

2.

The minimum number of operable instrument channels specified in Table 3. 2. 3 for the Rod Block Monitor may be reduced by one in one of the trip systems for maintenance and/or testing, provided that this condition does not last longer than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in any 30-day period. In addition, one channel may be bypassed above 30% 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. Except as specified in 3.2.D.2.

below, both steam-jet air ejector off~gas system radiation mon-itors shall be operable during reactor power operation.

The trip settings for the monitors shall be set at a value not to exceed the equivalent of the stack release limit specified in Specification 3.8.

The time delay setting for closure of the steam jet-air ejector isolation valves shall not exceed 15 minutes.

DPR-19

4. 2 SURVEILLANCE REQUffiEMENT 36

3.2 LIMITING CONDITION FOR OPERATION

~*

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.

DPR-19 4.2 SURVEILLANCE REQUIREMENT 36a

l*:in!::\\~-: i:'1. or C.!J=i'ab\\s In:at.*

Chai\\MlS Per

~:-11> :.ty:stc:'.\\(l) 1

1 2

1 l

l J.

l 2(5) 2(5) (Gl 1

fvnendment No.

OP~-19

.I~STRt'H-tirrATIO~ TUl\\T. J:NITL"\\TJ:S 'ROD Btoa Table 3.2.l

'Inst rur.:~ r. t*

'lr 1 v JAve l

..;....;.~~------------~~---*---'-"--~~~~~-

A?i'Ji upscale (fl-o.i b1a:1J (7) 0

5~D + SO FLPD UJ. j AP~I up:acal'!t (refuel en~ 8t-!lrtup/not: -

~12/125 full acale stanaby coa-1)..

APJUI. downt;cala (7)

Rod block lft0n1~or uprcala (flow b!os)(7)

Rod bloc~ monitor d~-nscola (7)

. JNl'dawru:~cla (3) lM upaciato* *.

XRM dat~cto~ no~ Lully J..n~urtc4 ln t.he cor*

SiVJ deta~t~~* not in otartup position Sl'.11 uosriata 5cram Discharge Volume Water Level.- High D.65 w0 + 45 1

( 2>1

~ 5/125 f \\\\11 'ace.le l!_S/125 full scala

~108/125 full.scal9

. (4).

sio5 c: c-.ants/s~~

25 gal.

c2*.

TAB~2.3 (Cont.)

DPR-19 NOTES:

1. For the Startup/Hot Standby and Run poeitione of the Reactor Mode Selector Switch, there ehall be two operable or tripped trip systems for each function, except the SRH rod blocks, *1RH upscale* IRH downscale and IRH detector not fully ins.erted in the core need not be operable in the "Run" position and APRH downscale, APRH upecala (flow biae), and RBH downecale need not be operable in the Startup/Hot Standby mode.

The RBH upecale need not be operable at leaa tba~ 30% rated thermal pover.

One channel may be bypaeeed above 30% rated thermal..

paver prowided that a limiting control rod pattern doee not exi*t. ¥or *Y*teme vith.ore than ane channel

per trip *J*tea, if the firet column cannot be met. for both trip eyeteme, the *ystems shall be tripped. For the Scraa Di*charge VolU11e water level high rod block, there i* one in*trument channel for one trip systf:lll.

~--

\\ -

2.

Wo percen~ of drive flow required to produce a rated core flow of.98. Hlb/*.*

MFLPD=highest value of FLPD for G.E. fuel.

IRH dovn.cale.., be bypa**ed vben it i* on it* loveet range.

3.

4. Tbh function.., be bypa**ed vben the count rate h

~100 cp*.

5 *. One of* tbe four SRH lnPut* u7 be bypae*ed.

6. 1:hi* SRH function aay be bypassed ln the higher IRH range* vben the IRM up*cale rod block i* operable *

. *1. Not required while performing l0v pove~ pbyelc* teet* at atmoepherlc prea*ure durln1 or after refueling at

over level* not to exceed S KW(t).

Amendment No. 67 42a

\\

I. ;

r

..... ~-

... 0 lbSt'I:

J.2 In addlUon to"reoctor protection lnstrwnentatlon

\\\\'hlch tnltl11te1 a reactor scrom, protective lnstru*

mentallon h111 been provided which lnlll:ates action lO mitigate the consequences o( llCClclcnts which llte bey~ tho opcrotora :abllll)' lo control. or lcrml-n:ates operator crrora before they result In &crloua consequcnet"s. This set of Spectrlc:atlons provides the limltln:; conclltlona of ot>erallon for the primar1 system Isolation funcllon. lnlllullon of the emcr-G'-'ncy core cooling ayslcm, control rocl block and stancP*y sr.as trc:atmcnt &)"Stems. 1"he ohjl'c:llvcs of the spcclflcathms arl.' (I) to assure the crreclh*cncsa of the protecth*c lnstrtunentatlnn when rr(luirt*cl by prcscrvln,; Ila c:ip:tbtllty to lole1*alc :i slns:lc failure

. or :any component or 1meh syslcms even clurlns: 1-*rl-ocls \\\\*lwn f'Ol'llons of such &)"Stems arc out o( sc*nlce for ntlllntcn:.ancc, llnd (II) to prc&crlbc the trl1> a;ct-t111a;s required to as&Ul'C :adcf1u:itc pcrfornrnncc.

When ncce&Hry, one clrnnncl may be ma1il! Inoper-able for brief lnterv:ils to conduct rcf.. lre1I functional teals 11nJ c:allbratlons.

Some or the aelllnga on the lnstrument:itlon lh:it Initiates or controls core llnd cont:tlnmcnt cooling ha,*e tolernncca expllcllly &lated **here the high and low values :arc hoth erlllcal and may h:we a substan-tial cHect on safctv. It should be noted that the sct-polnta or other lnsirumer.l~a!lon, where onl)' the hlgl\\

ur low end uf the 8eUln1t has a direct l1e:1rin:; on safct~*, arc chosen al a lcn~l away from the normal operatlnit ron:;e tu pre,*e11t Inadvertent acl\\1:tllon of ti*,. ~aft>l~* Pystt-m ln,*oh'ed and cxpo!un.- tr.- alm*.Jrmal altu:allons.

Isolation vnh*es ore lnsl:tl1ed In those lines that pcn~lr:ale the primary containment and musl be lsnlatccl during a loss or coolant acclclenl so that the r:u.ll:;llon dosr limits :tr~ not exceeilcd tlurlng an

tc::hlcnt condition. Aclu:tllcn of lhrsc \\*:tlvcs Is lnlll:alecl b~* prnll.'cth*c lnstriunent:attori shown In Amendment No.

0 0

DPR-19 Table 3. 2.1 which senses the condltton1 lor which lsolnllon Is requtrl'd. Such Instrumentation must be

available whcnc,;cr primary conl:ilmncnt Integrity Is rcqulrccl. The objective Is to laoh1tc tho prlmol'J cnnt:alnmenl so that the guldt*linca of 10 CFR 100 are not exceeded during 11n acchk'~l..

The lnstrumenlllllnn which lnlllatea prlma.., *1*lem lfiolittlon Is conncclccJ In a clu:il bus arr::~~mcnt.

1'hus, the discussion.;,*en In the biaar* for ~cm c:itlon 3.1 ta 111.pllcablc here.

""9 *-*reectw **tn l***l l**tr-tetl* le -

te td* et>*

lnct..* -

,.,. **-* a.. u-"t llotl of ecu" 1.. 1 I*........ **

..,.0 IM:he* *- **u*l urot etlll *ll*I' ellowl., lw,.,. loll et ~o* h*che* *- *****I *no. -

l** 1.-cll**...... ti* t.. *I ectl** foel. **troflt l*I f**l II** ** eetl** 1.. 1 le.. t* l.I*

lnc""* lOftll*I' t... 3 ***lier roel d***~*.......,.... fS***tOl trl*

eetpotnta wirr* *ie-tl IR tlhe LOCA an.,lr****

Tll** trlp l"ltl*tc* eloa*r* ef G*°"P I... J.....,, eowtal......

hobt *- **I*** lllut doe* -

trl* the reckcoletl*.,_.. lrafe.-

ence IA* Sect*- J. J.11. ror

Ul* Hlll., *C SCN 111e.,.* *..._

- U*I *c*o II** I*""* *-*,.,.of ecU** r.. u... *II*._...

- h* cl*-**,....,.,. **l-* will.,. el*........ ~fwetl*..

the claddl.,. occvu *-* fM tM -*'-- keell1

,.. *Ul.. I*

'IN *- *- HMtW....., l*atr-*l*tl* le... to tdp...... H-tOI W*t*I l~wel le *4* lllC!we...... **eael tel'e lwtt*... ef ectl..

- - '... "*' ** *** a.a.. **....,. *****& **,.. -st tioca.a l*..

,..... *"'"*,...,.. ** ectt** r.. u.

This trip Initiates closure or Group 1 prln1ary cont:ilnrncnt Isolation valves, Ref. Section 1. 7.2.2 SAR, nntl :ilso ocllv;1trs th~ ECC suhs~*st~ms, sl;arts the emc.-ri;encr diesel ::cncralor :ind lri1>s the rcclrcul:tllon pumps.

1"hls trip sellln;: lc\\*cl was ehcscn lo be t:l:;h ennu;h to J'lfC\\'enl ~pt:rlous oa~r:itlon but low cncu:.rft lo 1:11-ll:ate ECCS o:*l'ral!on ar.d prlmar\\' nslt*m lsol:!tlcm 80 th;:l no mc1t1r.:; or foe foci cl::*~ld1n~ wlli ocrl*r 11n!I 80 that post :icclclcnt conlin;; can be accompllshrcl

aml the i;uitlcl in~s of 10 C FR lf>O will not I.It' \\*lol:tlC"tl.

For th" complete clrc~m fot.. _.nll:al hrcak of a 211-lnch recirculation lllH' :11111 \\\\*ilh t!ic* trip scllln:; gt\\*cn rhm*r, ECCS ln!ll:1tlon an1I prhn:try syle1n lsol:allon 11rc lnlllakc! In time lo mt*l.'t the :tbo\\*e crlll.'rlia.

46

3. 3 LIMITING CONDITION FOR OPERATION C.

Scram Insertion Times

1.

The average scram insertion time, based on the de-energization of the scram pilot valve solenoids as time zero, of all oper-able control rods in the reactor power operation condition shall be no greater than:

\\ Inserted From Fully Withdrawn 5

20 so 90 Avg. Scram Insertion Times (sec) 0.375 0.900 2.00 3.50 The average of the scram insertion times for the three fastest control rods of all groups of four control rods in a two by two array shall be no greater than:

\\ Inserted From Fully Withdrawn 5

20 so 90 Avg. Scram Insertion Times (sec)

o. 398 0.954 2.120 3.800

2.

The maximum scram insertion time for 90\\

insertion of any operable control rod shall not exceed 7.00 seconds.

DPR-19 4.3 SUllVEILLANCE REQUIREMENTS

c.

Scram Insertion Times

1.

After each refueling outage and prior to power operation with reactor pressure above 800 psig.

all control rods shall be subject to scram-time tests from the fully withdrawn position.

'J1:ie scram times shall be measured withciut reliance on the control rod drive pumps.

2.

At 16 week intervals. 50\\ of the control rod drives shall be tested as in 4.3.C.l so that every 32 weeks all of the control rods shall have been tested.

Whenever 50\\ of the control rod drives have been scram tested. an evalua-tion shall be made to provide reasonable assurance that proper control rod drive performance is being maintained.

3. Following completion.of each set ~f scram testing as described above, the results will be compared against the average scram speed distribution used in th:

transient analysis to verify the appli-cability of the current MCPR Operating*

Limit~ Refer to Specification 3.5.K.

58 Amendment No.

i

~

r --*::

~

llPR-25 lndlcetlwe of* 1enerlc cnntrul rod drlwe*

proble*.... the reectur wl 1 l be ehutduwn.

Aleo, If d... 1* within the cnnt rol. rod clrl** Mchenh* end, In perlicular, crack*

ha drlwe l1tternal houein1*, c111mot be ruled out, the* e 1enarlc pruble* arrectln1 e I,..

... 11er of dr hH cennot be ruled oul. Clr-

,'~.:*.cuaferentlal creek* reeultln1 frCM etr***

- - hte* lnter1rem1ler corroelon have

~. occurred In the col let laouein1 of Jrlwee et Hwerel Mil'** T11le t7pe of cncU111 coultl occur la

nueeber of ddwee and lf the

crack* propagated unt I l eewerance of the

col lal hou.&111 occurred, ecr** could be pre-wented In the effected rode. Ll*lt Ina the

,.rlod of Oftentloea vhh e potentlellr H*ered collet laouelt11 and requlrlna In-creHucl 1urwel llance after detect Ina one etuck

,.,. wlll Heure that the reactor will nut be o~nl** whla

hr1*.......... of rode vltb felled collet laouelnt**

J,.*Th* operabllltr of the ecr.. dlechera*

wah,.. wunt end drain **h** ***nrae the proper vc11tl111 end dralnl111 of the wolu... *

r '* ene11ree thet v*ler *ccWM1latlon doee 11-.~ occur wblcla would ceuee *n earlJ ter*la**

t Ion of co1ttrol rod MOVe*nl durln1

full core ecr-. Theee epeclflcatlune prowlde lor the periodic we.-lflcet ion that the **h** are open *n* for Leet ln1 of theee wahee under reactor 1cr* cu11Jltlon1 Jurina each ae-fuellna Outage.

8~

Coatrol lad Uhhdnvel

-. I I. Control rod dropout acctdents as 1ltscusseit tn Ref-erence XN-NF-80-lq. Vol. 1. can lead to slqnlftcant

....,.. II couplln1 l11tqrll1 h aalntalned, lhe poHlbllltr of

rud d.-opuut acclde11t h ellala*t*** The owertrawel 1*oeltlon faatuir*

~.. ""

z.

l.

core DPR-19 prowldee

poeltl** check ** oalJ llllC0'9pl..

driwee **J reach thh poaltlon. hutron I**

elruaentellon reepnn** to rod.o.e*at pro-

ldoe

werlflcatlan thet the irod la lollowhia he drive.

Abeence of eucla reepoa** to drlwe 110*eaaent vould prowlde cauee for eu1pectl111

l"od to be uncoupled end *tuck. le1trlctln1 re-

'coupllng **rlf lcatlon1 to power level** *llow**

201 pl"ovlda* ***urence th*t

roll *rop._rlna

recouplln1 werlflcetlon WOYld ao~-reeult l*

rod drop ecc Ideal

The con*rol rod hou1la1 eupport r*etrlcte ta..

outward 110wement of e control rocl to I*** t....

l lnchee In t~** eatr.-lr r-t* ewent of *

bouelna failure. Tl**...,....t of reectlwhJ *ldl could be added *1 thle... 11.....,..t ef rotl vltll-drawal, Vl1lch 11 le** than

nor.el elaale vltll-drawal lncr.-nt, will not coatrlbuta to *J

......,. to th* prl*rJ cool*t *J*t*. The tle*lp beele le 1lven I* lectlon 6.1.I of ta.. IAI,...

the dHl1n ***luetlon le "* l* leer.I°" ***~>.

Thie euppart 11 Rot re~ulrad tr th*... ector coolant *1*t.. le et et.uepharlc praeeure el...

there woul* then be ao drlwlaa fore* to HpllflJ eject e drlwe houeln1.* AddltloaellJ, the *upport h not required U all coatrol rotle *n f*llJ !o:-

A

\\~~!;*

eerted, anJ If en.ctequete ehutdow *ral* vlth..

control rod vlthdrawn hae been *-*r.r*t.. el**

the reactor would re.. ln eubcrltlc*l.,,.. bl U..

event of complete *J*ctlon of the.,.,....,, e..-

trol rod.

Control rod wlthdrewal *nd laeertlOla.........

are eetebl hhed ta ***ure tliet' th* **'- I**

ee*1uence lndlvldual control,.,. or control IOll..

eequencee which are vllhdrawn coul* ltOt lie wrU.

I I!

.I II

~fl J.i t

H:

enuu1h to cause the rod drop accident design 1.. lt of 200 cal/<Jlll to be exceeded If they Wl'e ta drnp out of.

f the axe In the 11111nner defined for the Rod Drcp llccldlrit. CJ lheee eequ~ are developed.J1r!or ta lnJtW.. -*---*-......

1 O(leratJon ila

~

e

'*.0.

I

'. ?*:

of the unit following any refueling out~ge

- <**\\and the requirement *that an operator follov

theee 11equencea la backed up by the opera~.

__._ _ _,_ __ aEion of the i&M.o~a.. secoRJ qualffieci.

station enploye. 'nlese sequences are I

developed to llmit r'eoc.t ivlty vorth1 of control rude and* together with the integral

'rod velocity limiter* and the action of the

control r~ drive 11y11te111, limit potential re-activity ineertlon 1uch that the result* of

0 a control rod drop accident vil l not exceed a maximWI fuel energy content of 280 cal/gm.

The peak fuel enthal1*Y of 280 cal/gm h belov the energy content, 42~ cal/g.., at which rapid

.... * *.* fuel dhperaal and primary ayatea da.. age have

.*.!: ~**

JJeen found* to occur baaed on experimental dau

aa la diacua1ed in Reference 1.

The anal1al* of the control rod drop *ccldent waa originally presented *in Section* 7.9.3,

.

14.2.l.2 and 14.2,l.4 of the Safety Analyll1 0 leport. l*provement1 in analytic*l capability 1.... *,_have allowed

110re refined analyala of the

control rocl drop accident.

Amendment No. 59 DPR-19 I

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62

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B*!!,!(cont'd) rara*trlc Control Rntl Ornp Accltlrnt analy&l'B have ehovn that fur wl 1lc ranr,cs or key tl.'lll:tor para11eters (which envl'ln1u~ the 01*erallng ranKeS of these varlahle11,, the fuel enthalpy rise during a postulated control rod drup accident re111alns conslderubly lnwt*r than the lHO c1tl/t1.ffl ll*lt.

For euch operating cycle, cycll'-Rpeciflc Parameter* e11d1 aa.. a*lmum cunt.rot rotl Wurth, llc*1*pler coerrlch*nt crfectivc clclityctJ nt!lll ron fraction and.. axlmu.. fnur-luandle local pcnklng fact.or are cn111p11tt*d with the resul ls of the parametric analy&l'B to tletermlne the pt.>uk ruel rod enthal11y rise.

Thls value ls th1*n compared against the Technical S11ectrlcatlon I lmll of l80 cal/fl.* lo 1h.**11n11trate co1111*llan1:e for each operating cycle.

tr cycle s11cclr&c valUr!I of the above parameters are outside the ranr.e

a11eumed In the paramt.*l rte annl yses. an ext 1*n11ion of the analysis ur a cycle 1111eclflc analysl!I

.., be required.

CunscrvatlRm ptt*!tent In the analysts, results of the parumet.rlc stmfil's, and a detal h*d deacrlpt Ion of the llll!th111lolugy for performlnt1. the Control Mud Drop Acc.:ltl1.*nl analysts 11re 1*rovlclt~1I In n*r1*rence XN-N..--Ho-19, Volt~ I (Sllflpleml.'nl.s I an1I l,.

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'lhe Rod Worlla Hlnl*her 19ro.,t.lu *ul01L1tlr.

hUltOl"V ie1un to AnllttU! lhftt Ollt or lt<!'IUf!HCe c:ont.rol rods wll l not be vllhdrnvn or lnnerte*:

I.e., It ll*lla opec-~tor devlatJon11 (rOll pla1U1**

vlthdrowial acque11cu. lei. Section J.t SM. ;

It eervea **a bnckup to procedurnl.control oC control rod worth.

ln the event that the 1ct41 I North Hlnl~l&er la out of *e*vlce, when require*,

' !lcenaed operatnr or othu qunllllt!d technical eii.plo7et!' ~Ith -mudl7 full Ill the control rod patle~n'l!onfor*nnce f"'1C'tlo11a of the Rud Wot th Hlnlal i.ei~' In thh cane* procedural*

control' Ill* c\\&erc.. aed' by'*vr.rHJlng alt. control

rod poatl: lona dt~r' 'the wlthdr-vd of **ch 1

group, prior to proe~edlng to lhe... xt 1"

&l'oup.' *'Allowing *\\lbatltutlon of a second

11. ' 1 independent 0019eo111toa'*.01"englneer ln e*H..

of RWH

tnopP.Tlabl l*l ~, orecognlaee the c.1pabllltJ to adequately aonltor* proper. rod *'e*..a:nclna l*

an altunate"11111t11e** wlothout umh*lr ccdrlct* *

Ing plant oparatlona.

Above 201 power, there le.

no requlreaont that the a-.r.1 be,operable aipce the control rod drop accident with out-of-eequence rod11 wl ll result ln a.,pen.. fuel enal'g7 content ol lP.1111 thnn Hq eallc:a. to n1rnure lllth RWH a.aUebllltJ, ~he ll\\,"!*I la r<<!*1urled to be operating d11rlnf! a atertup for the vlthdrov1'11 of

elr,nlf leant nUClbe' contTol rods for an7 atart.., *ftcr J*ne 1, lt*

4.

111e Source Ranco tlonltor (SIDI) *Jet** perform no outoi.atlc eafr.tr *J*tat1 f.ectldtta I.e., lt boa no *en* function.

It doa prowl4'e.the 61b

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oi)er;~or 1'-ith a visual indiCD.tion of neutron level. this is needed for knowledgeable and

~!fic!~nt reactor sta~tup ot low neutron level.

T"r.e c?~se~ccnccs of reactivity ~ccidcnts are fc:ict ions e>f the in.ltial neutron flux.

The rcqulrc~ent of at lc~st 3 counts per second

~~s~r~s thnt ariy traa$fent, should it occurA bc$;1&W :it or above the initit'.11 v;,lue of io-of r.i:ed pc*.:cr cscd in the.in<1lyscs of tr~nsients fr~ cnlcJ ccnditiom:. ' One operable SRH ch.innel

\\.*ould be adequate to rnoni tor the <?pp roach to c:-l:ic::1~~::: t1slr.:; hor.iu~l!ncous pnttcrns of.

scat tc:-ed control rod wJ Lhdi-nw:tl.

A oinimum of t:.-o oj)er.:!>le Sin&'s ar~ provided as an added conservatism.

The r.od Block Monitor (RGH) is designed to auto-

a<1tic:!lly prevent fuel d<1~<1ge in the event of erro:?cous ~od "*ithdr.i,.*31 from locations of high l'01o'l?r licnsity during high power level operation.

T:~o clrn:mc ls ore provi,lcd *:md one of these may be bypo:s::cJ fro::a the console for rraintcnancc and/or tcstbc. Triprinc of one of the channels will block erroneous rod vlthdr<iw:il soon cn~ugh to prevent fuel d;i::a::t;c.

'nais systc::a boc:Cs up the operator who with-

"r:i*...-s

-orf~ occortlln:; to a vrittcn scqttcnce.

The srcclficd restrictions wit~ one cltanncl out of

~oru!ro com;erv.1tiv~ly ;ts sure th<tt f U'?l d<1mnr.e

... 111 r.ot o:cur ch:c to rod wlthdr~w.il errors when this condlt!on c::l~ts.

Ar.1enth'lcnts 17/18 and 19/20 prcs~r.t the rc~ults of on evnluation of a rod block' r::onitor f:tllurc.

'£hc!:e amcnclmcnts show that during rc<?ctt"r opcrntion vllh certain limitinc control roe; :-~ttcrr~, the wi thdrm.. al of a dcsicnnted single C'l"i!.~<11 ro*j <"'>t1ld r,:~111 t in one or mllre fuel rods with ~:Ci'R:a le*:.:i then the ll(Ci'R fuel chdJir.'J lnt~rity **f~ty li*lt.

Durin9 use of such p.Jtlcn:~, 1t is J*aJ;;*.;il ll::it tcsl1?1g of the RB~I sy~tcr: :'rfo:- to wltlu!r;:\\.ral of such rods to assure iu r.r*r.r~bilit>> \\lill n!>::;ure that improper with-Jrn~Jl ~oc~ not oc~cr. It is the rc~ponsibility of. tl1c r:"ch::ir Eu~;l::ccr to identify these limiting P~tt*.!:-ns o~*J the J~s it!lotcd rods either when the patterns ere initially established 'or as they develo~ due to the occurrence of inoperable control rods in other than limiting patterns.

DPR-19 Scram Insertion Times I.

The performance of the control rod insertion system is analyzed to verify the system's ability to bring the reactor subcritical at a rate fast enough to prevent violation of the MCPR Fuel Cladding Integrity Safety Limit and thereby avoid fuel damage.

The analyses demonstrate that if the reactor is operated within the limitations set in Specification 3.5.~, 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 transient;s, 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 to 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 estimated to actually begin.

However, 200 milliseconds rather than 120 milliseconds is conservatively assumed for this time.interval in the transient analy-ses, a~d is ~lso included in the allowable scram 1nsert1on ti!'l~s specified.. in Soeci-f ication 3.3.C.

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

The performance of the individual control. rod drives is monitored to assure that scram performance is not degraded.

Fifty percent of the control rod drives in the reactor are tested every sixteen weeks to verify aaequate oerformance.

Observed olant data were used to determine the averaP.:e 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 res~lts to verify that the performance of the control rod insertion system has not changed signif i-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-determination 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 each 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 requirements is expected under normal conditions.

The history of drive performance acc~mulated to date indicates that the 90% insertion times of new and overhauled drives approximate a normal distribution about the mean uhich.tends to become skewed toward longer scram ti~es as operating time is accumulated.

The probability of a dri~e not exceeding the mean 90%* insertion time by 0.75 Rernnrl iR ureRte~ thRn n.QQQ for a normal distribution.

DPR-19 64

D.

Control Hod Accumulators The basis for this specification was not des-cribed in the SAR and, therefore, Is presented in its entirety. Requiring no more than one Inoperable accumulator in any nine-rod square array is based on a series or XY PDQ-4 quarter core calculations of a cold, clean core. The worst case in a nine-rod withdrawal sequence resulted in a kerf < 1. 0 -- r)ther repeatin~ rod sequences with more rods withdrawn resulted In kcff > 1. O.

At reactor pressures In excess or :JOO psig, even those control rods with In-operable accumulators will be able to meet rc-quir<:>d scram Insertion times due to the action of reactor pressure. In addition, they may be normally Inserted using the control-rocl-clrl\\'e hydraulic systeJn. Proced111*al control will assu1*<:> that control rods w i lh inoperable accu-mulators will be spaced In a one-in-nine a1*1*ay ralhC't' than grouped together.

E.

Hcacllvlly Anomalies During each fuel cycle excess operating reac-(Revised with Changes 27 and 18 issued 3/29/74)

DPR-19 Uvfty varies as fuel depletes and as any burnable poison In supplementary control ts burned. The magnitude of this excess reacllvlty may be Inferred from the crlllcal rod configuration. As fuel bumup progresses, anomalous behavior In the excess reactivity

may be detected by comparison of the crtt-teal rod pattern selected base states to the predicted rod inventory at that state. Power ope rating base conditions provide the most sensitive and directly interpretable data relative to core reactivity. Furthermore, A

using power operating base conditions per-W mits frequent reactivity comparisons.

Requiring a reactivity comparison at the spectfled frequency assures that a compari-son wlll be made before the core reactivity change exceeds 1 % 61<. Deviations in core * ' '

reactivity greater than 1% ~are not ex-pected and requf re thorough evaluation.

One percent reactivity limit ls considered sare since an insertion of the reactivity into the core would not lead to transients exceed-ing design condltlons of the reactor system.

G. Economic Generation Control System Operation of the facility with the Ecoz:iomic.*

Generation Oontrol System with automatic 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 !

is 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 Syste'!'*

65

(~

, ____..... ________________________________ -r S.S Ll~lTJ~'G CONDITION FOR OPERATION

o. Automatic Pressure Rollof Subsystems I. except ** 1pocltled ln 3.S.D.2 and 3 below, tho Automatic Pressure Relief S~system shall bo operable whenever tho reactor pressure ls greater than 9~ psl1 and 1rrad\\ate4 tuel 11 la the reactor vessel* *

I. Prom and after the date that one of the fivo relief vahes ;!>fl.the auto1:1atlc pressure r~Uef :aubsyste:n is ~3~e or found to bo inopcrnblo "hen tho* reactor ls pressurized ab~~C 90 psig ~lth lrrAdintcd fu~l in tho r~~ctor vessel. reactor operation ls pcn:alsslbl*

only durinc the succccdln: sc\\*en doys unless rcr3lrs arc lll3do and provided that durinc such tln:o the 11rc1 Subsystem is operuble.

S. From and cfter the date that more than one of flve.rtlief valves of tho outcmctlc pressure relief subsyster.1.

r:1adc or round to be lno;u:r:iblc l\\hen the reactor ls prcssud:ttd above 90 psl1 with lrr:dlatcd

!uol ln tho. reactor vessc1,*reactor operation ls perzlssi~le only ~urlna the succeeding 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months unless repairs arc mado and provided that 4udnz such Umo th* HPCI Sub=-ystea ls opar:iblo.

Amendment No.

1"' -- -----

4.S SURVEILLANCE REQUIREMENT D. Survcllhnco of th* Auto11111tlc Pressure Rolle! Subsystem shall bo performed as follows:

1. During each operating cycle tho followln1 shall be performed:

l. A simulated automatic lnltlatlon which O?ens all pilot valvss, and

b. With the reactor at pressure each relief valve shall be manually opened.

~

Relief valve opening ehall be verified '1 ' ~

a compensating turbine byp*** valve or control valve closure.

c. A logic system functional test shall be performed each rofuellnc outa1e.

23. When it ls determined that one relief valve of the autom3tlc pressure relief subsystca is inoperable~ the llPCI shall be dccionstrated to be operable lnnedlatolr and weekl7 thereaftea

s. When it ls detemined that *re than one

~elief valve of the *~tomatic pressure relief subsysteo ls inoperable, tho HPCl suhsyste11 shall be demonstrated to.be operable lmedlate~*

DPR-19 78

.~

DPR-19

0----------------------o~---------------------o J.S LJ'JUTINO CONDXTIOB FOR OPERATIOlf 4.5 SURVEILLANCE REOUIR£MENT

. ~

I.Average Planar LHGR During steady state power operation, the Average Planar Linear Heat Generation Rate (APLHGR) of all the rods in any fuel assem-

.bly, a_s a function of average planar exposu e I

for G.E. fuel and average bundle exposure for Exxon fuel at any axial location. shall not exceed the rnr.xir.\\um average planar UIGR shown in Fic;uro l.5-1. If at any timo durin9 opP.rntion it ia dctcrminod by normal sur-voillcnce thct tho limiting vnlue (or l\\?L1tGR ie beinCJ excee<lfld. action shall be initioted within 15 mlnut~s to restore op~ratlon to within tho prcecribed limits.

If the APIJIGR in not roturned to within tha prcscribod limits within two (2) houra. tho re&ctor shall bo brought to tho cold Shutdow~ condition within 36

. hours.

survelilance and corresponding action ahall continue until roactor opera-tion lo within the prescribed limit*.

Pmemment ~.

I.

1\\V~_!ag_e___R.~i'.".'_ftr Lin~nr lleat C_,n<!ratlon Rate lAPIJIGR)

'l'he 1\\PUIGR for eBch type of f\\Jel a* a function of avern9e _plcnn.r exposure for G.E. fuel J and average bundle exposure for Exxon fuel shall be determined daily during reactor operation at 2 25% rated thermal power.

818

DPR-19 3.5 LIMITING CONDITION FOR OPERATION 3.5.J LOCAL LHGR I

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

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 limits.

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

Sur-veillance and corresponding action shall continue until reactor operation is within the prescribed limits Amenc!IJ1ent No~ 42 4.5 SURVEILLANCE REQUIREMENT J.

Linear Heat Generation Rate (LHGR)

The LHGR shall be checked daily during I reactor operation at 2 257. rated thermal power;-

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.. - ~~;~:-~;L~~J.f~~~~:,t:)T~Trti:~~~1;;~~~~-~~~i~~t~~-:,.. -***

.::Jttf~-~~~-=;T~:j;;~0.:;=1~Y#~::it~~t~1.S:-*~i*=~~~T.~i~::#

Dresden Unit 2 Fuel T,ype: n* 8D3. 02 10,000 20,000 30~000 40,000 Figure 3.5-1 (Sheet It or 5)

Bundle Average Exposure (MWD/MT)

.Narfmum Avenae Planar Linear Heat Generation Rate (.NAPIBGB)

Versus El:poeure 81 C-4

z D

~

<(

If c II.

If c.

u-.

z.

e*

04

~..

0 Ir

c.

<(

II.

t II. c Ir c Ir w

z..

Cl :f

~....

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ii i J: c

Pl ci z 10.5 10.0

_9.5

~

r...

~ --

~ 9.0 m

c:;

c 0

~

m

~ 8.5 c

ell 0

~

m Ill = 8.0 I-m Ill c

...:l I-d c m r-i 41

~

s..

Ill >

9 E

~

E DPR-19

t-H--T::t-:h~t+1,:: J ::i.-

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i;.:::: :.:!::,, 'l _:::.r:.,,l ::.:::~t'*::,

1:£:'j:c:::=--~~:-::-:-:::r==,~=-=-r.-=:-t: :.._;;:...-:t.:::;.;:...j':':::_:~:.. :~::.;:_

IT:..;... :J.: L:+: +:~ :: :', r ~~L+:~~; :i::;:.::::f::<;::, :=:~,~~:;~~:..::+::~:+~~i'~~h~:::_: :::::..:.:+~~::::q::*

f:::::. ::....;: :: : : : : :~:~:~-t:.:j::w::~::::i,~=~::~n~;,:~i:o::::~:.~~4~;~~~t~~4,;~=:J:*=~;~=-=-;.3:-;

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. :... 1::*

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~i..~:+/--=:::=- ~~~~::::;:~!:~

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~?-+/-:~~~4_=::::*-=;:_:-*.::_~*:~~

Fuel T-ne:

- -*1

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I

_;~#~~:=:~:;i::~:::-:L:~j$:::

~~:~~::;:=-:~::::!'::: ~1--:~~:: --:::-::::,:::,-:::i:::~~:f:::-:~J:~.::_~:=r-~s:~~;:~f~~~r-:~~~-::::=:.?>*

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.... *:~;* -.:L :=..

.c:*:c:rc~~:':::f~0=:~~~;;:-:::=-:J~~=J-:~~=-~-~~~,~-;~-;i:~~~:f~~'=::-]0:-:~::-=*~-

l 10,000 Bundle 20 000 Average ~osure i

~

.. ':) /r~ 30, ooo 4c t-;,*JJ~=i:--_~f:~;i.;£~5i~i~-~FJ-~~4.~-~--~~

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='=L::~:---:r:*::~:::.;:.:::.:~:~L::* ~.:i:: _:;:,;*: ::.:-.:=.==

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=.:...=:~_--;-:I:::-....:.:..:.

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.:;i:~-~~:~;.-:_. :

Figure 3.s-1 (Sheet 5 o~ 5)

Maximum Average Planar Linear Heat Generation Bate (MAPLBGR)

n. ~dle A-nnp bpo*ure Bl C-5

3.5 LIMITING CONDITIONS FOR OPERATION K.

Minimum Critical Power Ratio (MCPR)

During steady state operation at rated core flow, MCPR shall be greater than or equal to -

UNIT 2 1.31 for All SX8 Fuel Types 1.35 for XN 9X9 LTA 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 from Figure 3.5-2 sheet. 1 or the above rated core flow value, which ever is great_er.

2. Automatic Flow Control-the MCPR Operating Limit shall be the value from Figure 3.5-2 Sheet 1, Sheet 2 or the above rated 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 shall be initiated within 15 minutes to restore operation to within the prescribed limits.

If the steady state MCPR is not returned to within the prescribed limits within two (2) hours, the reactor shall be brought to the Cold Shutdown condition within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months .

Surveillance and corresponding action shall continue until reactor operation is within the prescribed limits.

In the event the average 90% scram insertion time determined by Spec 1.3.C for all opera~le control rods exceeds 2.74 seconds, the MCPR limit shall be increased by the amount equal to (0.092T-0.252) where T equals the.average 90% scram insertion time for the most recent half-core or full-core surveillance data from SpeG. 4. 3. C.

DPR-19 4.5 SURVEILLANCE REQUIREMENTS K.

Minimum Critical Power Ratio (MCPR)

MCPR shall be determined daily during a reactor power operation at ~25% rated thermal power and following any change in power level or distribution that would cause operation with a limiting control rod pattern as described in the bases for Specification 3.3.B.5.

Amendment 34 dated 5-3-78 81D 9.'

l.S 1.*

r 1.l

.. j

§ 1.2 l.l l.O 30 DPR-19

*--. **-**-****-----~~-----------.

40 so 60 70 80 90 100 Total Core Recirculating Flow* (I Rated, 98 mlb/hr)

Figure 3.5-2 (Sheet 1. of 2) MCPR Limit for Reduced Core Flow 811 I I i

I I.

I

DPR-19

e.

.A r 1.5

.-..

j

§ 1.4 l.J Total Core Recirculating Flow (I Rated, 98 mlb/hr)

Figure. 3°5-2 (Sheet 2 of 2) MCPR Limit for Automatic Flov Control 8U.-J

L I

3.5

~lmitinq*Conditlo'!!..,_~£~rntion Banen A.

~~rr. spi:nv nnd_J,rc_J: __ ~O!:lc ~f th<<!_R.!!l!.

System - This 9pccif ication a5surcs thnt adequate cmcr9cncy cooling capability is available.

Based on the loss of coolant analyses included in References (1), (2) and (4) in nccordance with 10CFRS0.46 ;ind l\\ppcn-dix ~' core cooling systems provide sufficient coolin9 t~ the cor~ to dissipate the cner9y associated with the loss of coolant accident. to limit the calculated peak clad temperature to less than 2200°F, to assure that core 9eomctry 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-*

tablished so that the ~verage risk rate for repair would be no greater than the basic risk rate. The method an~

concept are described in Reference (l). Using the results (1) *Loss of Coolant Accident Analyses Report for Dresden Units 2, J and Quad-Cltles Units l, 2 Nuclear Power St~tlonR,"

N~00-24146A, Revision!, April 1979.

-ndment No.

I I

DPR-19 I

~

Should one Cort" 11p1":\\.\\

0 ttub~~*,.C'm ~C'Om* In-operable. lh~ r*m:llnlng cort" !"p:-:n* ~nd lhl' entire LPCI syst('m are :aull:ible ;hould thC' (2)

(l)

I (4)

NED0-20566, General Electric Company Analytical Model for tose-of-coolant Analysis in Accordance with 10CFR50 Appendix K.

APEo-*cuidelincs for Determining s~fc Tc5t Intcrvnls and Repair Times for Enqincercd S~fcquarda* -

l\\pril 1~69, I.M. JacOb9 an~

P.W. M:irriott.

XN-NF'-81-75 "Dresden Unit 3 LOCA Model Using the ENC EXEM Evaluation Model MAPLHGR Results" 82

. Q..

.J. 5 L!::l! ~1n-r C~1U.on tor. <T?rntlon Dn:>~:J (C~:ll 'd)

I. l.ve::ero Plro.r.:lr lll(':R '

r t

~..

1'h1n s;ec1flco.Hor. uourc:s tM.t tho 1-onk

cl4d41n,; tc~p~rnturo follo\\lln~ e. i:ent.t:l.Ated doclr.n ~:s1o l~n-o(-cool:-.nt nccldonl 11111 r.ot cxc~cd tho 2200'T 1111!\\t opoc1f1c1 1n tW.R~O AJ*~r.c11~ K 'ccn:>l~erlr.e tho po:;t.ulated o!'Ccct:1 QC fuel r.=>llot. dcnclflc:lt..\\on.

Tho p:nk clodtlnr: to1Dr.1>rn.t.uro follollln~ o.

po:.tulclcd loss-ot-ccolr.nt. cccltlcnt 1-;

prlu:.rlly 0. funcllCI:\\ or lh1, CW\\ll"~'-!;:> H:r.n or all tJ13 rC'iz 1n a r.col c~r.6=-..1111 c~ r.ny u.lnl loc:LUon or.d 1o 0.111 tl*~~=!:dca\\.t. !!-:>cor.d-

~r1ly o:i tho ro.5 to red r:;~:~r dlch'ibl.:Uon wltMn a reel o~::c~bly.

Slnc~ cxr-:1clt~d local v::-1!.~lc;i:s ln po1::r d1r,lrlbut.1on ul th! n n Cuol =~~o:t.bly affect ~h-3 c~lcl:l.'\\t.r.il p:ilt olAd t.onr.:tr~tur.> by lc:s:a t11nn t2oc; r.lbll *:,, to the J:.:lck tei.'f.~r:ltt:TO for a t.:;rlc::.1 fu:tl design, th~ 11:i1 t on tho a*1or~~o rl~~.. r lJiGR lo su!!1c1r.nt to c:1:u~ th3t c~lculntcd t~Pp CrcLt.t!r~a are bolew tho 1CCFRSO, Apr-uudlx K Urilt.

The maximum average planar LHGRs shown

tn Figure 3.5.1 are based on calculations

.employing the models described in

':Reference (1) and in reference XN-NF-82-sa

power operation with APLHGRs at or below

those shown in Fig. 3.5.1 assures that

the peak cladding temperature following a postulated loss-of-coolant accident will not exceed the 2200°F limit.

(1) *1.oss of Coolant ~ccident Analyses Report for Dresden Units 2, l and Quad-Cities Units 1, 2 Nuclear Power Stations," NE00-24146A, Revision 1, April, 1979.

DPR-19 I

The maximum average planar LHGRs for G.E~ I,

fuel plotted in Fig. 3.5.1 at higher exposures result in a calculo.lcd f.C3k cL'.!.d tc~r..o~t~ or 1~~3

e tt::in 2200~.

Hc.~*cvcr tho n.ula:u:t.,,v~=~

plnn:ir UIGRo o.ro oho:rn on Flc. ).,51 1 o:s A

11Rll3 b~cau5o conror.:.~nr.o calcul4~1c~3 hsYe net. b!cn p~rrorr.::d to just1r1 opor~t1cn at lJICJta in OXCCDS ot tho:Jo aho'rr.*

J. Loc!ll lJ!t;R Th1o opcctric*at1on MDUX"S Uu.t tho

tux1mun linc3r heat Bencr.i~1on rato 1n any fuel rod fabricated by G.E. is less than the design linear 65A I

I I

.5 Limiting Condition for Operation Bases (cont'd) heat generation 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 Analysis provides adequate protection against cladding strain limits, hence the LHGR limitation for GE fuel is unnecessary for the protection of ENC fuel.

The steady-state values for MCPR ipecified.

in the Specification were determined using 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 safety limit, at least 99.9% of the fuel rods in the core are expected to avoid boiling transition.

The Limiting Transient A CPR implicit in the operating limits was cal-culated such that the occurrence of the limiting transient from the operating limit will not result in violation of the MCPR safety limit in at least 95% of the random statistical combinations of uncertainties.

Transient events of each type anticipated during operation of a BWR/3 were evaluated to determine which is most restrictive in terms of thermal margin requirements.

The generator load rejection/turbine trip without bypass is typically the limiting event.

The thermal margin Rf f ects of the event are evaluated with the THERMEX Methodology and appropriate MCPR limits consistent with the XN-3 critical power correlation are_

determined.

Several factors influence which transient results in the largest reduction in critical power ratio,such as L

I DPR-19 the cycle-specific*fuel loading, exposure and fuel type.

The current cycle's reload licen-sing analyses identifies the limitin8 transien for that cycle.

As described in spec~fication 4.3.C.3 and the associated Bases, observed plant data were used to determine the average scram perfor-mance used in the transient analyses for determining the MCPR Operating 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 in accordance with Technical Specifications 4,3,C,3 and 3,5,K, 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 val¥e from Figure 3.5-2 Sheet. l or the rated core flow valve, whichever is greater.

For Automatic Flow Control, in addition 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 by the reduced flow MCPR limits shown in Figure 3.5-2 Sheet 2 where the curve corresponding to the current rated flow MCPR limit is used (linear interpolation between the MCPR limit lines depicted is permissible).

85b

4.5 Limiting Condition for Operation Bases (cont'd)

Therefore, for Automatic Flow Control, the MCPR Operating Limit is chosen as the value from Figure 3.5-2 Sheet 1, Sheet 2 or the rated flow value, whichever is greatest.

It should be noted that if the rated flow MCPR Limit must be increased due to degradation of control rod scram times during the current cycle, the new value of the rated flow MCPR limit is applied when using Figure 3.5-2 Sheet 2.

DP.q-19

... l*

858-'

surlelll.ance *Re9u1re~nte lJaaee (cont*4)

I.

A"fl!rece Plamr LllCR.

At core therml power levele leee than er.*

eq\\!al to 25 per c:cnt, operating plant*

cxp-:rlence Dnd thermal hydraulic onol71ee lr.dlc~te that the re~ult1nc aYcr~ge plnnsr LilCR ls bclov the mxlmuo avernce plo.nir LRCR b1 o conllder:.ble r.1rsln; there!ore, evcluation of the DYCrace planar LUCR belov thle power*

le*el 1s not nccees3r7. The 40117 requlro-r.cnt for claculot.lng aYerace planer LllGR above 25 snr cent rat.eel thcnn:>l power b aur~tclent atnco power 4Sot.rlbut1on ehtrta ere alov vhcn there ha"le not been li&nltl*

cant. pouer or control ro4 chance*.

J.

Lot:a l t. !fCR

'l't.a Lt:CR for G.E. fuel shall b~ cmck'!c1 aat'-y during reactor operation at ere* ter than or eauol to 25 p-.?r cent po~cr to cl"!t.erc1~ tr f*.:-:1 b\\:rnu;> or control r~! ~ove::ient

~31 cauDcd chr.n~* ln po-.ter 41Dtribution.

A limitin9 LHGR value is precluded by a considerable mar9in when employing a per missible conttol rod pattern below 25\\ rated thermal power.

Amendment No.

~:...: :' '*

DP!?-19

~

~

I. Mlnl11U11 Critical Power Ratio (MCPR)

At core the.rmal power levels less than or equal to 2S per cent, tho reactor viii be operatin1

=t ~lr.iou:: recirculation ~\\Jltlp speed and the moderator vold content will be very s*nll. For all designated control rod patterns which 11ay be employed Rt this point, operating plant experience*

and thenaal hydraulic analysis indicates that the resulting MCPR Yalue ls in excess of require*ents by a considerable.-argin. With thls low YOid content, any inadvertent core flow Increase would only place operation ln a 110re con-serYative mode relative to ~R.

'lbe daily require111ent for calc~latlna MCPR a~ove 25 percent rated theraal power is sufficient since power distribution shifts are very slow when there have not been significant power or contrul rocl changes.

In addition, tho I correction applied to the LCO provides *f rgin for flow increase fro11 low flows.

86A

NUCLEAR ENIRGY BUSINESS GROUP* GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA 99125 GENERA~. ELE.CTRIC APPLICABLE TO:

PUBLICATION NO.

NED0-24146A 79NED273 T. I.E. NO. -----------

TITLE LOCA ANALYSIS REPORT FOil ERRATA Ancl ADDENDA SHllT 7

DRESDEN UNITS 2, 3 AND QUAD NO.

July 1982 CITIES 1, 2 NUCLEAR POWER STATIONS DATE ISSUE DATE April 1979 REFERENCES ITEM (SECTION, PACilE l"ARACilRAPH, L.INE)

1.

Page 4-11

2.

Page 4-1~

NOTE: Conwct *II coPi* of the applir:MJ/t1 publication a rpecifitld below.

INSTRUCTIONS (CORRECTIONS AND ADDITIONS)

Replace with new page 4-11.

Replace with new page 4-14

PAGI l of l

NED0-24146A Table 41 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

Dresden 2,3/Quad Cities 1,2 roEL TYPE:

P8DRB282 Average Planar Exposure MAPLHGR PCT Oxidation

~MWd/t2 (kW/ft)

(oF)

Fraction 200 11.2 2132 0.029 1,000 11.2 2132 0.029 5,000 11.8 2183 0.033*

10,000 12.0 2189 0.032 15,000 12.0 2199 0.033 20,000 11.8 2181' 0.032 25,000 11.3 2110 0.025 30,000 11.1 2075 0.043 35,000 10.4 1981 0.035 J

40,000 9.8 1886 0.021 NOTE:

Credit taken for the effects of pre-pressurization of the fuel rods Table 4J MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

Dresden 2,3/Quad Cities 1,2 FUEL TYPE:

P8DRB265H Average Planar Exposure MAP LB GR PCT Oxidation

~MWd/t2

~kW/ft2

.L!l Fraction 200 11.5 2163 0.032 1,000 11.6 2171 0.032 5,000 11.9 2192 0.033 10,000 12.l 2198 0.033 15,000 12.1 2200 0.033 20,000 11.9 2190 0.032 25,000 11.3 2116 0.026 30,000 10~7 2018 0.018 35,000 10.2 40,000 9.6 1934 0.022 J

1835 0.009 NOTE:

Same as Table 4I 4-11

NED0-24146A Table 4K MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

Dresden 2,3/Quad Cities 1,2 Average Planar Exposure (MWd/~)

200 1,000 5,000 10,000 15,000 20,000 25,000 30,000 NOTE:

Same as Table 41 MAPLHGR (kW/ft) 11.3 11.4 11.7 11.8 11.8 11.4 10.9 10.4 Table 4L FUEL TYPE:

P8DRB239 PCT Oxidation

.m Fraction 2117 0.027 2129 0.028 2160 0.030 2141 0.028 2145 0.028 2114 0.025 2041 0.020 1968 0.015 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

guad Ci ties 2 FUEL TYPE:

P8DGB284 (Barrier Fuel)

Average Planar Exposure MAPLHGR PCT Oxidation (MWd/t)

(kW/ft)

.rn Fraction 200 ll.5 2152 0.031 1,000 ll.5 2150 0.030 5,000 ll.9 2185 0.033 10,000 12.0 2189 0.032 15,000 12.1 2199 0.033 20,000 ll.9 2180 0.032 25,000 ll.3 2109 0.025 30,000 10.9 2042 0.020 35,000 10.3 1956 0.030 40,000 9.7 1846 0.010 NOTE:

Same as Table 4I 4-12

)

NED0-24146A Table 4M MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

Quad Cities 2 FUEL TYPE:

P8DGB263L (Barrier Fuel)

Average Planar Exposure MAPLHGR PCT Oxidation (MWd/t)

(kW/ft)

(oF)

Fraction 200 10.2 1991 0.018 1,000 10.3 2003 0.018 5,000 11.5 2133 0.027 10,000 12.0 2185 0.032 15,000 12.l 2199 0.033 20,000 11.9 2187 0.032 25,000 11.4 2116 0.026 30,000 10.8 2020 0.019 35,000 10.2 1919 0.013 40,000 9.5 1829*

0.009 NOTE:

Same as Table 4I Table 4N MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

guad Cities 2 FUEL TYPE:

P8DGB263H (Barrier Fuel)

Average Planar Exposure MAPLHGR PCT Oxidation

~M'.;d/ t~

(kW/ft)

(oF)

Fraction 200 10.l 1978 0.017 1,000 10.2 1982 0.017 5,000 11.2 2102 0.025 10,000 12.l 2196 0.033 15, 000 li.l 2198 0.033 20,000 11.9 2188 0.032 25,000 11.3 2116 0.026 30,000 10.8 2022 0.019 35,000 10.2 1921 0.018 40,000 9.5 1830 0.009 NOTE:

Same as Table 4I 4-13

NED0-24146A Table 40 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

Quad Cities 2 Average Planar Exposure MAPLHGR (MWd/t).

(kW/ft) 200 9.9 1,000 9.9 5,000 10.4 10,000 11.7 15,000 12.0 20,000 11.9 25,000 11.6 30,000 10.9 35,000 10.2 40,000 9.5

. NOTE:

Same as Table 4I

Table 4P

. FUEL TYPE:

P8DG:S298 (Barrier Fuel)

PCT Oxidation f:ll Fraction 1962 0.016 1959 0.016 2000 0.018 2148 0.028 2198 0.033 2194 0.033 2153 0.029 2074 0.034 1963 0.032 1827 0.009 MAPLHGR VERSUS AVERAGE PLANAR EXPOSURE PLANT:

Quad Cities 1,2 FUEL TYPES:

P8DRB265L, P8DGB265L (Barrier Fuel)

Average Planar Exposure MAPLHGR PCT Oxidation (MWd/t)

(kW/ft)

(OF)

Fraction 200 11.6 2171 0.032 1,000 11.6 2177 0.033 5,000 12.o 2198 0.034 10,000 12.1 2195 0.033 15,000 12.1 2199 0.033 20,000 11.9 2187 0.032 25,000 11.3 2115 0.025

\\

30,000 10.7 2018 0.018 35,000 10.2 1927 0.022 40,000 9.6 1834 0.009 NOTE: Same as Table 4I.

4-14 J

ML17194B412

Text

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