ML20147E836

New Fuel Storage Rack Criticality Analysis for Wolf Creek Generating Station

ML20147E836
Person / Time
Site:	Wolf Creek
Issue date:	12/31/1987
From:	Marsico P PLG, INC. (FORMERLY PICKARD, LOWE & GARRICK, INC.)
To:
Shared Package
ML20147E795	List: ML20147E788 ML20147E810 ML20147E825 ML20147E836
References
PLG-0590, PLG-590, NUDOCS 8803070170
Download: ML20147E836 (14)

Text

1 PLG-0590 NEW-FUEL STORAGE RACK CRITICALITY ANALYSIS .

FOR THE WOLF CREEK GENERATING STATION Prepared by PETER J. MARSICO I-Prepared for WOLF CREEK NUCLEAR OPERATING CORPORATION i Wichita, Kansas December 1987 8803070170 880226 PDR ADOCK 05000482 P DCD l Pickard,Lowe andGarrick,Inc.

Engineers e Applied Scientists e Management Consultants Newport Beach, CA Washington, DC L _

INTRODUCTION The new fuel storage racks for Wolf Creek accommodate 66 fuel assemblies in 6 rows of 11 assemblies. Although new fuel assemblics are always stored in a dry condition in these racks, the racks are enclosed by a reinforced vault, therefore, the potential exists for the racks to be flooded or exposed to an aqueous mist or foam under accident conditions.

Consequently, the analysis includes both the fully flooded assumption and the condition of optimum moderation at low water densities.

DESCRIPTION OF THE ANALYSIS The computer codes used for the criticality safety analysis of the new fuel storage racks have been extensively tested and benchmarked against experimeny,al data as described in Reference 1. The LEOPARD (Reference 2) program is used to generate macroscopic cross sections for all materials utilized in the analysis, and the P00-7 (Reference 3) program is used to evaluate the ef fective multiplication factor for the geometric arrays used to represent the new fuel assemblies as they are stored in the rack.

There are three new fuel storage racks each of which provides space for 22 assemblies (two rows of eleven) with a center-to-center spacing of 21 inches. A sketch of a rack is shown in Figure 1, and a plan view of the rack and storage pit arrangement is shown in Figure 2. The two-dimensional basic cell geometry shown in Figure 3 was used to calculate the infinite multiplication f actor for an infinite array of the rack stainless steel tubes containing the new fuel assemblies. The basic cell was also used to calculate flux weighted macroscopic crosi sections to be used in P00-7 synthesis calculations to determine the ef fects of neutron leakage on the k,of the bas.c cell.

The fuel assembly characteristics utilized for the analysis are shown in Table 1. A uniform axial enrichment distribution of 4.50 w/o U-235 was

< assumed for sach fuel rod in the assembly.

8462C111787 l

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The k ,of the fuel pin cell used to generate cross sections for the rack criticality analysis is shown as a function of water density in Figure 4. For the normal dry storage condition, the fuel assembly k, will be essentially the same'as or lower than the fuel pin cell k ,,

and therefore an upper limit for the k,pp of the . rack for dry conditions may be obtained by Ixtrapolation of the fuel pin cell k ,

to zero water density. The resulting k, is about 0.75, and the k,pp of the finite rack will be substantially less than that value for dry conditions.

If it is assumed that the entire rack volume is enclosed by the concrete walls and floor of the pit, then the water density in the resulting enclosed area can be varied, and the k ,of the rack can be determined from edits which combine the void-water, stainless steel tube and fuel assembly regions of the model shown in Figure 3. The resulting fuel rack k ,is~also shown in Figure 4. At water densities below 0.1 gm/ce, an optimum moderation condition is approached. The apparent optimum moderator density is quite low because of the large volume f raction of the void-water region as compared to that of the fuel region. However, these large k,s at low water densities are not of concern because of the high neutron leakage at low water densities from these finite racks in both axial and radial directions.

To determine the effect of neutron leakage on the multiplication factor, flux weighted cross sections representing the fuel assembly, stainless steel tube and void-water regions of the basic cell model were used in one-dimensional calculations. Figure 5 shows the one-dimensional models used to determine neutron leakage effects in the north-south, east-west, and axial directions. The concrete walls and floor were represented as

-separate regions, and vacuum boundary conditions were applied at the outer boundaries of the concrete and at the top of the variable water density region at the elevation corresponding to the top of the new fuel storage pit.

8462C111787 . ,- -

, 1 The neutron multiplication factor, which includes neutron leakage effects l

in the axial direction only, is shown as a function of water density in i

' Figure 6. The maximum neutron multiplication factor is seen to be about ,

l 0.g6 at'an optimum water density of about .05 gm/cc. Hwever, these l calculations assumed zero neutron leakage in both the east-west and north-south radial directions. One-dimensional calculations using the [

models shown in Figure 5 were also performed in the north-south and  !

east-west directions to determine th'e neutron leakage effects in both of these directions. From these two calculations in the radial plane, group independent bucklings representative of the neutron leakage in both east-west.and north-south di*ection's were determined. The sum of these two group independent bucklings, which then represents the total neutron leakage in the radial plane, is then utilized in the one-dimensional axial model shown in Figure 5. These axial calculation: then become synthesized three-dimensional representations of the new fuel storage i racks. The k,ffs from these three-dimensional representations are compared to the k,s from the basic cell calculations in Figure 7.

The maximum k,f, at low water densities is now only about 0.81, and it occurs at a water density of approximately 0.06 gm/cc.

'In addition'to the other conservatisms in the calculations, no credit was

taken for neutron streaming ef fects at low water densities. Consideria; the large distances separating fuel assemblies in the rack, such effects l would be expected to significantly increase neutron leakage from the racks and thereby further reduce the neutron multiplication factor.

Figure 7 also demonstrates the acceptability of the new fetl storage racks in the flooded condition which corresponds to 6 water density of 1.0 gm/cc. As shown, the neutron multiplication factor is somewhat less l than 0.88 for the fully flooded condition and therefore clearly acceptable, 1

Therefore, even with the unrealistic assumption that a uniform water

! density cf anywhere between 0 and 1.0 gm/cc can be obtained throughout the entire new fuel storage pit, these results demonstrate that new fuel 8462C111787 l l -

assemblies with an initial enrichment of up to 4.50 w/o may be safely stored in the Wolf Creek new fuel storage racks. Becatse'of the conservative techniques and assumptions used to evaluate the madmum Possible neutron multiplication factor, there is more than reasonable Jssurance that no significant hazards based on critic 41jty safety are involved in storing fuel assemblies of up to 4.50 w/o U-235T4n the Wolf Creek new fuel storage racks. ,

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• REFERENCES

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1. F.J. Marsico, Pickard,'l. owe'and Garrick, Inc., Spent Fuel Storage,, , ,

fj Rack Criticality Analystp for the Wolf Creek Generating Station, / fi

' ,4 -

December 1987. 3 i h~

2. R.F. Barry, "LEOPARD--A Spectrum DependcQ Non-Spatial Depletion Code for the IBM-7094," WCAP-3263i September 1983. ,/

W.R. Caldwell, "PDQ-7 Ref'rence"fah'ual," WAPD-TM-678, January 1967.

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!e p TABLE.1. FUEL ASSEMBLY JECHNICAL.INFORMATION FOR WOLF CREEK-b

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Q:: . Assembly- Description 1 R' o d array .

17 x 17 0verall dimensions, in. 8.426 x 8.426 "y.

Fuel Rod-Description .

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Fuel rods- per. assembly 'j 264 i -Outside diameter, in. O.374 Cladding wall thickness, in. 0.0225 t Cladding material , Zircaloy-4 i Pitch, in. y 0.496 Active fuel length, in. -

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Fuel,Peilet Ocscription-Pellet'matifial 002 Pellet density',,gm/cc 10.28

-Outside diamet.#r, in. 0.3225 UO2 W';ight (ib/ft.of fuel rod) 0.3643

Coatrol &gd Cell ~0escription  ;

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)) Contrbi Guide thimble rod cells mai.ertal per assembiy Zircaloy-4  !

Guide thimble outside diameter, in. 0.482  ;

' Guide thimble inside diameter, in. 0.450. 1

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, Instrument Cell Description  ;-

1, Instrument cells per assembly 1 Guide thimble material Zircaloy-4 Guide thimble .outside diarneter, in'. 0.482 Guide thbable'inside diameter, in. 0.450 Girds,N,ffectiveWeightinCore(ib)

Inconel 2324 s

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