ML20076J158

Amend 74 to License DPR-59,revising Min Critical Power Ratio Operating Limits to Accommodate New Fuel,Adding Addl Rod Block Monitor Trip Level Settings to Facilitate Control Rod Withdrawals & Deleting Refs to Fuel Types

ML20076J158
Person / Time
Site:	FitzPatrick
Issue date:	08/26/1983
From:	Vassallo D Office of Nuclear Reactor Regulation
To:	Power Authority of the State of New York
Shared Package
ML20076J159	List: ML20076J158 ML20076J161
References
DPR-59-A-074 NUDOCS 8309060452
Download: ML20076J158 (25)

Text

,

gaa'%g'o, UNITED STATES

+

8 or ' i NUCLEAR REGULATORY COMMISSION 5 j.. ", i '

E WASHINGTON D. C. 20555 k

POWER AUTHORITY OF THE STATE OF NEW YORK

)

DOCKET NO. 50-333 JAf1ES A. FITZPATRICK NUCLEAR POWER PLANT AMENDf1ENT TO FACILITY OPERATING LICENSE Amendment No. 74 License No. OPR-59 1.

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

A.

The applicatfor for amendment by the Power Authority of the State of New Ycrk (the licensee) dated f!ay 25, 1983, complies with the standards and requirements of the Atomic Energy Act of 1934 as amended (the Act), and the Commission's rules and reguiations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of I

the Commission; C.

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

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

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

1 i

8309060452 830826 PDR ADOCK 05000333 i

P PM l

I

. 2.

Accordingly, the license h amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and paragraph 2.C(2) of Facility Operating License No. DPR-59 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 74, are hereby incorporated in the license.

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

3.

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

FOR THE NUCLEAR REGULATORY C0t1 MISSION

'[fggp,)tt./$

Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Attachmer.t:

Changes to the Technical Specifications Date of Issuance: August 26, 1983

i ATTACHMENT TO LICENSE AMENDMENT NO. 74 FACILITY OPERATING LICENSE NO. DPR-59 DOCKET NO. 50-333 Revise the Appendix "A" Technical Specifications as follows:

Remove Replace vii vii 6

6 6a 9

9 10 10 13 13 31 31 31a 31a 43 43 47b 47b 47c 47c 47d 47d 73 73 123 123 124 124 130 130 135a 135a 135b 135b 135c 135c 135h 135h 245 245 246 246 I

(.

LIST OF FIGURES FIGURE

"'ITLE PAGE 3.1-1 Manual Flow Control 47a 3.1-2 Operating Limit MCPR versus Y 47b 4.1-1 Graphical Aid in the Se'.ection'of an Adequate Interval 48

, Between Tests 4.2-1 Test Interval vs. Probability of System Unavailability 87 3.4-1 Sodium Pentaborate Solution Volune-Concentration 11 0 Pequirements 3.4-2 Saturation Temperature of Scdium Pentaborata Sclatien l'.1 f

3. 5-6 MAPGGR Versus Planar Average Exposure 135 d Reload 2, SDRB283.

3.5-7 MAPLHGR Versus Planar Average Exposure 135 e Reload 3, PSCR3265L 3.5-8 MAPLHGR Versus Planar Average Exposure 135f Reload 3, PSCRB28 3

3. 5-9 MAPLHGR Versus Planar Average Exposure 135 g Reload 4, P8DRB284L 3.5-10 MAPLHGR Versus Planar Average Exposure 135 h Reloads 4&5, P80RB299 f

3.6-1 Reactor Vessel Thermal Pressuriration Limitations 163 4.6-1 Chloride Stress Corrosion Test Results at 500 F 164 6.1-1 Management organiration Chart 259 6.2-1 Plant Staff Organiration 260 AmendmentNo./,[,

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1 JAfflPp i

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Z.

Top of Active Fuel The Top of Active Fuel, corresponding.

t to the top of the enriched fuel column of each fuel bundle, is located 352.5 inches above vessel l

zero, which is the lowest point in l

the inside bottom of the reactor l

vessel. (S)e General Electric l

drawing No. 919D690BD.)

{

AA.

Rod Density

}

j Rod density is the number of control I

rod notches. inserted expressed as a

'l fraction of the total number of i

control rod notches. All rods fully -

inserted is a condition representing j

100 percent rod density.

1 1

1

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i f

I i

l

-: ~ -

l a

I 2

Amendment No. 74 6a

.1 AFilPP 2.1 (cont'd) 1.1 (con t ' d )

D.

Iteactor Water Level (llot or Cold)

In the event of operation with a

~

maximum fraction of limiting power Shutdown Conditions) density (WLPD) greater than the Whenever the reactor is in the shut-fraction of rated power (FRP), the down condition with irradiated fuel setting shall be modified as follows:

in the reactor vessel, the water level shal1 not be less than that corresponding S

$(0.66 W + 54%) x PRP

.-')

MFLPD to 18 inches above the Top of Active l

Fuel when it is seated in the core.

Where:

PRP = fraction, of rated thermal power (2436 MWt)

MFLPD = maximum fraction of limitin9 power density where the limiting power density is 13.4 KW/ft.

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

(2).

Pixed Illgh Neutron Flux Scram Trip Setting When the Mode Switch is in the RUN position, the APRM fixed high flux scram trip setting shall be:

S 6,120% Power Amendment tio. 14, 3 0',

,4 3',

64~, 74 9

l

.lAMIPP 1.1 (t unt 'al) 2.1 (cont'd)

APlHIhlBlockShipSetting A.I.d.

11m AP!H Int block trip setting shall in:

S 0.66'W + 42%

deres.

4 S = Int block setting in percent of tiummal power (2436 >Ht)

H = foop recirculation flow rata in percent I

of ratal p~

In the event of operation with a maxistan fractio limiting power density (Wiro) greater than tJe fraction of ratal gnaar (ETP), the setting shall le mxlifiod as follows:

l S ( (0.66 W + 421)

FIP i

~

[Fleiro

s. tere FIC = fractjen of ratal Octmal pawnr (2436 FHt)

FFIPD = unxinssa fraction of ljraiting poer density Were tia 11sniting prAar density is 13 4 KW/ft, s

Sie re.tlo of Ele to MPIED shall km set equal to 1.0 unless the actual operating value is J<(, ;6,,4/f, p(, 7:[, 74 10 Amencimen t rio.

less than tim design value of 1.0, in Witch case tJe actual oterating vahm w!!1 km used.

t

jar:4PP e

1.1 (cont'd) provided at the beginning of each At 100% power, this limit is reached with

~

fuel cycle.

Because the boiling a maximum fraction of limiting power density transition correlation is based (MFLPD) equal to l.0.

In the event of large quantity of full scale operation with a MFLPD greater than the fraction on a data there is a very high confidence of rated power (FRP), the'APRM scram and rod that operation of fuel assembly at block settings shall be adjusted as required the Safety Limit would not produce in speci fications 2.1. A.1.c and 2.1. A.1.d.

-)

boiling transition. Thus, although

~

it is not required to establish the B. Core Thermal Power Limit (Reactor Pressure safety limit, additional margin

< 785 psig) exists between the Safety Limit and the actual occurrence of loss At pressures below 785 psig the core is greater,.than 4.56 psi.(0 power, O flow) elevation pressure drop of cladding integrity.

At low powers and flows this pressure differential is flowe rve r, if boiling transition were maintained in the bypass region of the core.

to occur, clad perforation would not Since the pressure drop in the bypass region be expected.

Cladding temperatures is essentially all elevation head, the core would increase to approximately 1100 F pressure drop at low powers and flows will which is below the perforation temperature of the cladding material.

always be greater than 4.56 psi'3 Analys es show that with a flow of 28 x 10 lbs/hr This has been verified by tests in bundle flow, bundle pressure drop is nearly the General F,lectric Test Reactor independent of bundle power and has a value (GETR) where fuel similar in design to PitzPa trick opera ted above the of 3.5 pai.

Thus, the bundle flow with

)

critical heat flux for a significant a 4.56 psi griving head will be greater period of time (30 minutes) without than 28 x 10 lbs/hr.

Full scall ATLAS test data taken at pressures f rom 0 psig clad per fora tion.

to 785 psig indicate that the fuel assembly critical power at this flow is approximately If reactor pressure should ever exceed 3.35 MWt.

With the design peaking factors 1400 psia during normal power operation this corresponds to a core thermal power of (the limit of applicability of the boiling transition correlation) it more than 50%..Thus,.*a core thermal power limit of 25% for reactor pressures below would be assumed that the fuel clad-dingintegrity Safety Limit has been 785 psig is conservative.

violateds In addition to the boiling transition 1imit (Safety Limit) operation is constrained to a maximum LIIGR of l 13.4 kw /f t.

}/, ;ri, jf, pf, f. 74 I

Amendment rio.

d 13

JAFNPP f

3.1 (CONTINUED)

MCPR Operating I.imit for Incremental C. MCPR shall be determined daily during.

Cycle Core Average Exposure reactor power operation at? 25% of rated thermal power and following any change in power level or distribution At fun Ili-trip DOC to mC-2Go/t to EOC-lGO/t that would cause operation with a limiting level setting _

IXC-2Go/t FOC-lGo/t to IIC control rod pattern as described in the bases for Specification.3.3.B.S.

,I S=

.66W + 39%

1.21 1.25 1.29 D. When it is detetmined that a channel has failed in the unsafe condition, the S=

.66W 4 40%

1.22 1.25 1.29 other RPS channels that monitor the

~

same variable shall be functionally S=

.66W t 41%

1.24 1.25 1.29 tested immediately before the trip I

system containing the failure is tripped.

S=

.66W + 42%

1.25 1.25 1.29 The trip system containing the unsafe failure may be placed in the untripped S=

.66W & 43%

1.27 1.27

1. 2 ')

condition during the period in which surveillance testing is being performed S=

.66W & 44%

1.33 1.33

'l.33 on the other RPS channels.

E. Verification of the limits set forth in specification 3.1.B shall be performed as follows:-

)

1.

The average scram time to notch posit a

~

38 shall be: 77 AVE <- trB 2.

The average scram time to notch position 38 is determined'as follows:

t AVE " b 161 i=1 where: n = number of, surveillance tests l

performed to date in the cycle, Ni =

number of active rods measured in Amendment No.,%d, 74 31

)

J M 1JI'p 2.

If s oluitoient 4.1.E.1 1u int. iret (l.c. tn tje ith surveillance, arvifi = '

(t. AVE) tien tlus OpcrdtlawJ I.Imlt tY1'll averago scram time to notch valikru6:n a fiinct lusi of t) are isa givent lui position 38 of all rods l'i uio 3.1-2 neasurell in the Ith sitrveillance 9

(T ~To\\

lacre l' a (TAUG 'U

^

A B

arel y = tlas averago ucram tino to notch 3.

'Ihe adjustal analysis mean scram g

toultion 311 an defInal in ripect-tino la calculatal as follows:

fication 4.1.E.2, Tm tie auljuulot aimlyulu nean nerarn o

N Linn au definal in a,lecificatioin 1

4. l. c.1, Q(sec)=' j( +1.65 cr t = tlus t; cram tino to notch goultion

{N 4

30 an sletinol in ogccification g

1 3.3.c.)

• Int e:

Stoiht Llo eteratiivj limit in3'n obtaliol fann thin flgine le Icon tIwin tbo operaLing 1Imit senere J4.= mean of the distril2Flon it3*It foiuvl in Specification 3.1.n.1 for tie average scram for tic a plicable Itirl trip level insertion time to notch position 38 = 0.723 sec.

setting tien specification 3.1.11.1 nhall a[ ply.

6= starvlant deviation of the distrilmstion for average scram insertion time to i

notch position 38=0.054 sec. '

If anytine durf rvj reactor operation greater tien N=h MW W d MM 2S% of ratal gxwaar it in deterinirxxl timt (lio limit-g da measum! In specifi-livj valie for tYJ'11 la leisyj exceuled, action shall cation 4.3.C.1 then le initiatal within fif teen (15) minutes to.

neutoro operation to within tic prencrilxx1 limita.

gg g

Il tie it.I'It 10 int actuivel to within the prescrilxxl arvi ie test intervals are given in limit u within tw (2) toiru, an onderly reactor specific,ation 4.3.C.

puur soluction shall le comencal innelintely,

'ite scactor guer ulul1 le ralucal to lean tion 25%

of :atal prur within the nexL four touro, or until tlum it'I'It is retutent to will.in the proscrilol limits.

tur anu f lows Lller ilun ratal, tivs tY1'R operat_ityJ limit blull ic aintt ipliol Isy the agpropriate kg 10 au nisun in fluuiu 3.1,1.

th.Jg',j,rf,74 hmn bent 31a

~

JAFNPP TABIE 3.1-1 (cont'd)'

~

l&7ClOH PIOPIUPIOll SYS'1121 (SCIW1) INSTIME27PATICH IEUJIIDETP

~

tbtes of Table 3.1-1 (cont'd)

C.

liigh Flux 1111 Scram Discharge Veltine Ili.gh Icvel when any control rod in a control cell containing D.

fuel is not fully inserted.

E.

APIt! 15% Power Trip 7.

tht recpiiral to be operable when prinnry containment integrity is not requiral.

8.

tbt required to be operable when the reactor pressure vessel head is not bolted to the vessel.

9.

'1he AProi downscale trip is autenatically bypassed when the IIM Instrunentation is operable and not high.

10.

An APlot will be considered ' operable if there are at least 2 LP1H inputs per level and at least 11 IPIM inputs of the nonini cciopienent..

11.

See Section 2.1.A.l.

'Ihis equation will be used in the event of operation with a nuximum fractim of, limiting power density 12.

(MFI.PD) greater than the fraction of rated power (FRP).

.\\

/

Where: FRP

= Fraction of Rated 'lhennal Pouer (2436 MWt)

= Maxinnun Fraction of I.imiting Power Density where the limiting power density is 13.4 IGi/ft. l MFI.PD

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

W

= Icop Pecirculation Flow in percent of ratal S

= Scram Setting in percent of initial 13.

'Ile Average Ibwer Range Monitor scram function is varlal as a function'of recirculation flow (W).

'Ihe trip setting of this function nust be maintained in accordance with Specification 2.1.A.l.c.

46, f,,6(, ff, f,'I, J2', 74 I

Anendnent Ib.

43 I

i Figure 3.1-2

{

9 Operating Limit MCPR Versus T (defined in Section 3.1.3.2)

FOR ALL FUEL TYPES 1

1 1

I l

l l

1.40

( 1,1. 4 0 ) _, 1. 4 0 I,

(1,1.37' 400 1.35 LO 7

1.35 go '

c

=

( 1,1. 3 2 )

w U*

1.30 E

(0.2,1.304) e 1.30 (0.667,1.

7)

[

(0,1.29) e

=

O2

~

C g

S e

ca 1.25 g

1.25

( 0,1. 25)

SOC-2 (0.6,1.236) go 1.20 1.20 (0,1.20) l i

i 0

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 E

Amendment No. fr4, 74 47b

. - _.. _ _ ~.

This page deleted 4

Aht tb. [, 74 47 c

i This page deleted 470 Amendment No.

, 74

JAFNPP l

TABI.E 3. 2-3 (Cont'd) 3 INSTRUMENTATIOff THAT INITIATES CONTROL ROD BLOCKS NOTES FOR TABLE 3.2-3 (Con t ' d) l The APRM and RBM rod blocks need not be operable in start-up mode.

From and after the time it is found that the first column cannot be met for one of the two trip systems, this condition may exist for up to seven days provided that during that time the operable system is functionally tested immediately and l

dai1y therenfter; if this condition lasts longer than seven d'ays, the system l

shall be tripped.

From and after the time it is found that the first column

]

cannot be met for both trip systems, the systems shall be tripped.

2.

IRM downscale is bypassed when it is on its lowest range.

3.

This function is bypassed when the count is t 100 cps.

4.

One of the four SRM inputs may be bypassed.

5.

This SRM function is bypassed when the IRM range switches are on range 8 or above.

6.

The trip is bypassed when the reactor power is i 30%.

7.

This function is bypassed when the Mode Switch is placed in Run.

8.

S= Rod Block Monitor Setting in percent of initial.

i N= Recirculation flow in percent of rated K=

Intercept values of 39%, 40%, 41%, 42%, 43% and 44% can be used with appropriate l

MCPR I.imits f rom Section 3.1.B.

9.

When the reactor is subcritical and the reactor water temperature is less than 212 F, the control rod block is required to be operable only if any control rod in a control cell containing fuel is not fully inserted.

10.

When the control rod block function associated with scram discharge instrument volume high water level is not operable when required to be operable, the trip system shalI be tripped.

d Amendment No.

fd, f 2, J, 74 73 4-e i

i t

JAE1H'P 1.5 (mnt 'd) 4.5 (cont'd) un litlon, tlat t=siti bhall le conalderal 2.

Eu110 wiry any gerlod ulu2m tJe IFCI 1:w gciable for tantonen sattufylivj slect-subsystuna or core spray subsystans fleatlonu 3.S.A, 3.5.C, aivl 3.5.1-:.

Imve set toen raguired to le operable, t!e discharge p)phy of the inoperable synton shall be vental fmn the high 11.

Average Planar I.inear lleal. Generation tule inint prior to tie retum of tie

( Al'illGIO syntun to service.

' live Al'illGit for eacts tyge of fuel as a 3.

law 2never the IllCI, ICIC, or core fusu:t ion of average planar exposure alull Spray Systan is Ihnt up to take ag>L exccol the limit iswJ value uluun in miction frun the ooixlensate storage l

l'iguien 3.5-b t hiuujh 3.5-10.1 f anytine tank, tie disclanje piping of tie doilivj reactor guur operation gieater litti, ICIC, aivl core Spray shall than 251 of ratol gamr it la determinal le vental fran the high point of iluL tic limitierj value for APillGit lu (in systan, asul Water flow observed l

leisvj exceolut, action stall tlwin Ic on a nonthly basis.

initiatul within 15 minuten to restore osciation to within tius giscucrital limit u.

4.

'lin level switches locatal on the '

i 11 ihe Al'IllGit lu not actursol to within 03re Spray avvi leilt Syntan discharge the pieseritol Iimie a utthin two (2) Iv>aro, piping high points which nonitor an onleily s cactor guer roloction alall le these lincu in insure they are full ainicucol iniioliately. W2 reactor grwr shall le functionally tental each blull Ic solocal to lesu than 251, of ratul nonth.

gomr within tiv: next four touro, or until tic Al'IllGit in s eturnal to within tin pte-II. Averag'e Planar I.incar lleat Generation Date l

u:rlinsi limita.

T5filiiGji) l

\\

'lle APillGR for each tyge of fuel as a function of average planar exposure shall Ic detennined daily durits reactor gerat_ ion at > 254 ratal thennal gewer.

d hicininent Ini. jf,J, 74 123

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3.5 ItASI:i (cont 'd) togulinient u for the useigenw:y dicuel generatnro.

are within the la UR SO A[penulix K limit.

'lho limillivj value for APillGil la sliown in

(;.

.Iti.i.n.t enaeum of Fillol llincharge pipe Figure 3.5-6 thtuFJh 3.5-10.

l l'

t If tic dinchange pipiavj of the ooie ripray,11CI, I.

I.inear lleat Generation llate.(IllGil)

It:I(?, aint 11101 ate not fillol, a water ILumer can develqi in tiniu piptivj 54ien tlus gainp(s) are

'this upecification assurca that the linear ut ai t al.

'lu minimize danage to the dinchanye Icat generation rate in any rod is less than pipinj anl to ensure adslal margin in tie ogoration tie design linear heat generation.

)

of these nystuna, thin technical sgecification soluinen tio diteluive llams to le fillol SAusn-

'the IJIGR shall be checked daily dtu;ing reactor ever the nyutun la majuinal to le gerable.

If operatior,at 3.25% rated thermal power to l

a dischange pige la sw)L fillal, the insips that.

detennine if fuel burnup, or control rod movanent, unggily llut llem unmt le austainst to le lieperable has caused changes in twer distribution. For for tecimical specification purloses. Ilmever, IJ1GR to be a limiting value below 25% rated if a valer luumer tere to occur, tie synton thennal twer, the ratio of local IllGR to umbt ut ill gerform its design functlon.

average IJIGR would have to be greater than 10 which is precltrled by a considerable margin II.

Avet age Planar I.inear llent Generation Itate (A_PillGit) when enploying any permissible control rod pattern.

'thiu unecification anunica that. lic geak clavilleyj tuip::ature followlivj the postiilatal denign lasts loun-of-coolant. accident will int exceal tic limit. upecifial in 10 Wit 50 Aggenlix K.

4

• 11 w teak cliutilsyj tuigerature followlivj a gostu-latal lous-of-coolant. accident la prinurlly a function of tie average leat generation rate l

of all tie inis of a fuel ausuibly at any axlal locat ion ant lu only de[civlent necorolarily on j

tlie sul to sul g<wr diutrilution within an annuibly. Since expectol local variations in lumr dist rilait ion ulIhin a fuel assuibly af fcct tium calculatul teak clad tuigurature Isy Jens llun i 20*F selative to the guak taigerature f or a typical fuel design, the limit on the i avenage linear heat generat. ion rate tu unf-ticient. to auunte t hat calculatal tuiperaturca I lbm niinient flo. 64,74

-130-2

i i

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AMut No. [,,d, 74 135a 1

'Ihis page deleted i

Amend:mnt No. 30, 64, 74 135b

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, 74 135c Ar:erdent No.

.1M111'P 4

Fitfire 3.5-10 13 -

40es n al so It. loa <tu 4 & S PDDial299 iC *I 4

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10 15 20 25 30 35 40 Planat Averago Exposure (GWD/ t')

[taxlinnin Average l'1anar I.inear IIcat Generation Rate (MAPT.llGH) l Versus Planar Averago Exposure

Reference:

HEDO-21662-2 (As Anunended August 1981)

L

,bI, 74 13Sh Ann:nilms:n t. tio.

g

JAFl!PP i

S.0 tildini Flwitllus n.

• 11ie reactor core contains 137

~

cmcifonn-shatxxl control rois as descrlini in Section 3.4 of 5.1 S I'IE-~

ile ESAll.

A. 'I1wn.Lmm:n A. Fi t zPatrick IAiclear 1

lux 1r Plant lu locatent on the PASilY 5.3 IU7CIOlt PluGStilE VISSEI.

l lan Lion of the title flile 14)lut uite, aggstoxinutely 3,000 f t. cast of tius

'the reactor pressure vessel 'is as Ilise flile li>lnt thiclear Station, tinit 1.

descrilxxl in Table 4.2-1 arvi 4.2-2

'll u !!!P-JAP site 1u on lake Ontario of tic MIAll. 11e alplicable design in Oaxso O unt ty, ILv Yor k, approxi-axles are descrilxx] in Section 4.2.

nat ely 7 miles avullutant of (kaxxjo.

of t he FSAM.

• 11n plant lu locatal at coonlinates s w ir th 4,1119, 545.012 m, east 3116, 968.945 m, 5.4 OcurAltrinrr on tic linivernal Transverse threator I

Syst,sn.

A.

'lic principal design parameters azul characterlatics for tin

11. 'llw sv arent tolut on tic progerty prinury contairment are given in liin f rian tle reactor fulldlin arvt Table 5.2-1 of the ISAR.

any Ioint u of giotential gaseous cl t lu'entu, with tivi exception of tle H.

'lte accorslary aintinirnent is as lake t.loreline, f u locatul at tic descrlini in Section 5.3 asvl the swa tlcant coriu:r of the property.

agg>11 cable codes are as described

'this dial,mce lu alpioximately in Section 12.4 of tie FSAll.

3,200 f t.,

anal in tjus riulius of tiv exclusion areau au definnt in 10 CFIt C.

Penetrations of the prinary con-i 100.3.

taliment arvi piping passing through auch penetrations are desigivx1 in 5.2 low lost aca>nlance with starulards set forth In Section 5.2 of tle ESAlt.

~

A. 'the reactor core ccusists of not 5.5 tuct.SIolw;l; j

nine than 560 fuel assenblies. Por A.

'lle new fuel storage facility design the current cycle, two fiel types are presept in the core: 8x811 arvl criteria are to sinintain a 1(gg dry PHx811. 'these fuel types are des-4,0.90 and flooda14 0.95.

criled in IMn-240ll. Itoth 8x81t Ocupliance shall be verified prior to and P8x811 fuel types have 62 fuel Intraluction of any new fuel design inis aix) 2 water rods.

to this facility.

l wolne:nt t>>. )D;[ y, y[ ps, 74

/

245 a

4 l

JAFtJPP i

h 5.5 (u>nt'd) l 11.

'the stent fuel storage pool is

}

designal to maintain 1(gp less I

tinin 0.95 tuxler all coixlitions as descrilul in tie AtiUcrity's a[>1ilication for stent ftel storage nrnlification transntittel to Llie flIC July 26, 1978.

In order to assttre Umtd.12 criterion is nut, new fuel wiLi 1;e limital to an axial loading of 16.28 <pn U-235/ axial an l

Y i

or ajuivalenL.

(Ibr une present

)

fuel design, <tescribed in NElo-24011, l

this axial laxling is equivalent to an average lattice enrichnent of 3.3 w/o IF235.) '1he nunber of slutt fuel assaiblies stored in tic stuit j

fuel pool shall not exceal 2244.

i I

5.6 Joismic tiesign i

l

'the reactor l>uilding and all engineeral safeguards are designed d

on a basis of dynamic analysis using i

acceleration response spectrtun curves which are nonialized to a gmtuxl notion of 0.08 9 for tie Operating 11 asis Earth-t qinke arul 0.15 g for de 11esign Basis l

Earugluake.

l l

1 d

y!,f, 74 246 j

hie xinent tb.

t n