ML20053C818
| ML20053C818 | |
| Person / Time | |
|---|---|
| Site: | 07002949 |
| Issue date: | 04/23/1982 |
| From: | Andogini G, Andognini G ARIZONA PUBLIC SERVICE CO. (FORMERLY ARIZONA NUCLEAR |
| To: | Ketzlach N NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 20693, NOS-82-326, NUDOCS 8206020597 | |
| Download: ML20053C818 (15) | |
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April 23, 1982
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hv:s Director of Nuclear Material Safety and Safeguards g Matt
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U. S. Nuclear Regulatory Commission Washington, D. C.
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Attn:
Mr. Norman Ketzlach d
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Subj : Special Nuclear Material License Response to NRC Questions for
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'7 Docket No. 70-2949 (SNM 1887) y g
Attached please find Combustion Engineering's res ns s t;LNRC44bEs-tions 10 and 11 as requested in NRC letter to G. C.
ognlini.2dsted January 8, 1982. Also attached are amended pages to our submittal dated February 17, 1982, in support of the Arizona Public Service Company application for authorization to receive, store, inspect and transport Special Nuclear Material.
The criticality analysis per-formed by Combustion Engineering on the dry storage of new fuel in the spent fuel pool will be transmitted to you upon our receipt of this information.
Should you have any questions or require additional information, please contact Mr. Steven R. Frost at (602) 271-3348.
Sincerely, GCA/JRP/jc
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Director, Region V, USNRC Aq a
6 NRC Resident Inspector - PVNGS ro NRC Project Manager - E. Licitra 8'
section data-for representative. fuel rod cells, and material between and Amendment 8 9.1-4 March 1982
Attacliment NOS82-326 Aprl! 23, 1982 8 C?" t 1/16 (TYP),l'iSIDE DIMENSION OF THE 80X c
a BOXES MADE FROM 0.109* (MIN) STAIN L ESS STEL L SHEO slut F.
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j l'alo Vertle Nuilear (;cneratin;: Stalient l'S All NEW FUEL STulGE PisCK (150" ACTIVC FUEL LE:GT!!)
rigure 9.1-1 M.u ch I U A:1 in nt H
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Attachment 4
- NOS32-126 Aprl1.23, 1982 Question 10 - Page 25-26, Section 7.?.?
1.
/
Provide the results of the' nuclear criticality saf ety analysis for the fuel assembly arrays in both the new and the spent fuel pool storage aFeas.
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Answer The result of the nuclear criticality aialysis for the spent fuel pool with
]
- 0. 393.
This, nalysis the fuel stored in a checkerboard arrty :15
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>;as done using the follo'.nny input.
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4 Fuel enrichment 4.3 wt'% U-23L (4.0 wt % U-23.> is the license limit)
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l'onolith 30'y stainless steel Q nches,
j in11 thickness
/
"L" insert'304 stainle'ss steel 0.175 inches thittness center-to-center. monolith J, - M 16 i rlhas---- --
~~
_ _ _j _..,_ -
The resu'.t'of' the nuclear critical.ty analysis using KEt:0 for the new fuel
.acks is Meff = 0.7767 i.0070.
Inis analysis was done using the following
./
input.
Fuel richment
'4.3 wt % 'l-235 (4.0 it % U-235 is license limit)
Stainless steel can Thickness' around each fuel assembly 0.10 irches 3
flist density (110) 0.0525'gm/cm 2
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.'* :.0S82-326 April 23. 1982 2.
Question 11 Page 26, Section 2.2.3, Provide a demonstration that shcus the effect of all degrees of moderation on'Keff of the arrays.
Describe the method (s) of analysis used and their validation.
Answer The only rack which' is currently analyzed for mist conditions is the new
.f t,el rack.
The results of the KEU3 analysis for the new fuel rack for 4.3 wt % U-235 (4.0 wt % U-235 is license limit) wi hout the 0.10 inch stain-less steel can around the fuel assemblies are as follows:
3 II O density, gms/cm ggrf 2
'0 030 0.9144 i.0067 0.0525 0.9277 t 0080
' O.045 0.8932.t.0070 3
A single case was run ~ for the 0.0625 9ms/cm water density including the 0.10 inch stainless steel can surrounding the fuel assenbly and resulted in a Kef f of 0.7767 i.0070.
Since this is well below the design limit of Keff 0.93 further variations of the mist density were not deemed
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necessary.
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The method of analysis and its validation is enclosed.
Since no critical
. experiments are available for mist condit'ons it is C-E's gicy to provide larger margins between the calculated Kef f of the rack and thp NRC lesign limi t ifTe ffT O M: -
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Attachment NuS82-326 April 23, 1982 Qualification of Analytical Methods Used in Spent fuel Storage Rack Analyses 1.
Purpose
~
The purpose of this paper is to provide qualification of the calculational model and evaluation of calculational uncertainties and/or biasfactors used in analyzing spent fuel storage racks, especially the llI-CAP rack s employing steel boxes and super ill-CAPS containing boron carbide poison.
This is based on the analysis of a variety of reactor and 'aboratory experi-ments.
The methods of cross section generation are essentially those of C-E's physics design procedures modi fied appropriately for use in four-group transport, discrete ordinate method criticality calculations, and Monte Carlo codes.
- 11. Calculaticnal Uncertainty and Bias The results of i.he analysis of a series of UO2 critical experiments are summarized in Table 1.
Table I includes the mean and standard deviation.
Although the spatial solution for the flux distribution was obtained by use of a diffusion theory code such as PDQ-7, transport corrections fo'r the reflector apd heterogeneous lattice ef fects were employed.
These calcula-tions suppp use of the differential cross-section data base, and. broad group cross section generation codes.
Since fuel storage ar-js do involve the spacing of the fuel assculies at larger separation di<.nces than in typical PWR reactor lattices, the pre-dictive capability the calculational model was tested on the following experiments.
In these analyses done for this memo, the spatial flux solution was obtained directly with the transport code, Arils't.
To assess the accuracy of the calculational model in predicting the multiplication factor of fuel assemblies having a separation distance suf ficiently large so as to be isolated, analyses were carried out for a group of subcritical exponential experiments on clusters of 3.0 w/o 00 fuel pins clad with 2
type 304 S.S. and moderated by !!7 (page 165 of Reference 7).
The cluster sizes analyzed vary from 181 to 301 fuel rods so as to enccapas-the range of sizes typical of current Ph'R fuel assemblies.
The cultiplication factors for the lattices analyzed using axial bucklings deduced from the reported relaxation lengths are tabulated below, flo. of fuel Rods Vse f f 181 0.99C5 211 1.0011 235 0.9966
' 265 0.9988 301 O.9984
At L.ichment Nos32-326 April 23. 1982 These results indicatc that the calculational model predicts the multipli-cation factor for small clusters of fuel rods in a water environment to a.
high degree of accuracy, i.e.
a bias of
.0017.
To ascertain whether the calculational node can predict the reactivity characteristics of thick stainless steel plates and boron poisoned plates an analysis was made of ptul experimental (Reference 8) critical separations of 2.35 w/o U-235 00 subcritical clusters.
The results using the Monte 2
Carlo code KEti0 IV are shown in Table 11.
Method of Calculation
.The calculation methods for these experir. ental comparisons which ' arc also used to determine reactisity for fuel rack s tora ge,
fuel shipping containers plus other fuel configurations founc in fuel manufacturing areas
.are based on CEpAr, cross sections.
Using an appropriate buckling value and taking proper account of resonance absorption,, three fast groups are col-lapsed from 55 fine energy mesh groups' in FORM and the one thermal group is collapsed from 29 thermal energy groups in. TllERMOS.
In addition, each component such as water gap, or peisen plate has its therm 31 cross section determined by a
slab TilERMOS calculation enploying the proper fuel en v i ronn.en t.
FORM and TiiERNOS are sub-programs of CEPAK.
For one dimensional analyses such as BNL exponential experiments the discrete o.rdinates code ANISti (Reference 9) is used.
For two dimensional analyses DOT-2W (Reference 10) is used.
For three dimensional analyses (such as the critical separation experiments) KEf 0.IV (Reference 11) is used.
Results
' The above analyses indicate a mean error between predicted and measured multip!ication factors of +.00135 and a calculational uncertainty of.00714 at the 95/95 cor.ficence level for the complete series of U02
' experiments.
Thus, using CEPAK cross sections we conclude the following -
~
Total fiumber of Results 41 Mean Value (D) 1.00138 Standard Deviation = 6 0.00337 6 Multiplier for 9/95 confidence 2.118 l
95/95 Confidence Level Uncertainty 0.03714 Bias (4 -1.0)
+.00138 Uncertainty Minus Bias
'.005/5 l
l g
,.t t achmen(
Nos82-326 April 23, 1982 It will be noted that the seven no boron steel cases have a bias of 0.00207 (i.e. the calculated value is.00207 grea'ter than the critical Kef f value
~
of unity) which is greater than thn mean bias.
The three boral cases have a bias of -0.00435 with unity particle sel f-shi el ding factor for the 8 C.
Because of the size and distribution of the boron carbide particles 4the boron allous more transmission than an equivalent hci.t.ogeneous boron carbide mixture.
t'eutron transmission experiments conducted by the University of flichigan for Brooks & Perkins Inc. (Reference 12) are consis-tent with using a 0.9 self-shiciding factor in the third of four CEPAK neutron group and a 0.75 self-shiciding factor in the thermal group.
These self-shielding factors which are used in designing
- boron containing fuel racks make the bias for these boral cases +0.00008.
Re ferences :
- 1. T. C. Engelder, et al, " Spectral Shift Control Reactor, Basic Physics Program", B 'W-1273, fioven.ber,1963.
- 2. R. II. Clark, et al, " Physics Veri fication Program Final Report", B&W-3647-3, March, 1967.
- 3. P.
W. Davison, et al, " Yankee Critical Experiments", YAEC-94, April,
- 1959,
- 4. W. J. Eich and W.
P.
Rocacik, " Reactivity and ficutron Flux ' Studies in Multi-Region Loaded Cores", WCAP-1443, 1961.
- 5. F.
J.
Fayers, et al, "An Evaluation of Some Uncertainties in the Comparison Between Theory and Experiments for Regular Light Water
- Lattices, Brit. fluc. En. Soc. J., 6, April,1967.
- 6. J. R. Brown, et al, " Kinetic and Buckling Measurements on Lattices of Slightly Enriched Uranium and UO R ds in Light Water",. WAPD-176, 2
1958.
- 7. G.
A.
- Price,
" Uranium Water Lattice Compilation Part I,
Bril Exponential Assemblies", B!il-50035 (T-449), December, 1966.
- 8. S.
R.
Bierm3n, ' E.
D.
Clayton and B.
M.
Durst, " Critical Separation Between Subcritical Clusters of 2.35 w/o U-235 Enriched UO R ds in 2
Water with Fixed lieutron Poisons", Pi;L-2438, October, 1977.
- 9. Ward W.
- Engle, Jr.,
"A Users Manual for Afil Sil, A One Dimensional Discrete Ordinates Transport Code With Anisotropic Scattering K-1693",
March 30, 1967.
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Attachment
. ful:1 82-126
- April.23 1982 TABLE I' Results of Analysis of Critical U02 Systems 2
- gerr, B tot NO-Lattice
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1 B&W(1)
.88-2 1.00121 2
11
-.172-2 1.00534 3
X
.79-2
.99838 4
XIII
..701-2 1.00419 5
XX
.202-2 1.00550 6
B8U (2) 1
.861-?
1.00269 7
2
.420-2 1.00443 8
yankee (3)~
1
.408-2 1.00088 9
2
.531-2 1.00115 10 3
.633-2 1.00136 4
.688-2 1.00244 11
' Yankee (4),
12
'Winfrith (5)
R1-20
.660-2 1.00214 13 R1-80
.626-2
.99942
-14 R3
.510-J 1.00422 15 Bettis.(6) 1
.326-2 1.00053 16 2
.355-2 1.00046 17
'3-
.342-2 1.00106
~
Average 1.00208
+.00206
- Using calculated radiai bucklings and measured axial bucklings.
P
L Attachment SOS82-326 a'*
April 23, 1982 TABLE !!
Calculated Keff Values For Separation Experiments s
Monte Carlo Expt !
Type Poison Plate Keff (STD Deviation) 15 None 1.00227
,.00534 04 None 0.99912
.00540 49 None 1.00221
.00473 18 None 1.00813
.00489 21 None 0.99589
.00461 28 304 S Steel 0.0 w/o Boron 1.00393
.00308 05*
304 5 Steel 0.0 w/o Boron 1.00329
.00303 29 304 S Steel 0.0 w/o Boron,
1.00271
.00302 27 304 5 Steel 0.0 w/o Doron 1.00418
.00273 26 304 S Steel 0.0 w/o Boron 0.99811
.00279 34 304 S Steel 0.0 w/o Boron 0.99793
.00297 35 304 S Steel 0,0 w/o Baron 1.00436
.00290 s
32 304 S Steel 1.05 w/o Boron 0.99970
.00524 33 304 S Steel 1.05 w/o Baron 1.01173
.00491 38 304 S Steel 1.62 w/o Boron 1.00289
.00512 39 304 S Steel 1.62 w/o. Boron 1.00208
.00506 20 Boral 0.99585
.00301 16 Bocal 1.00020
.00288 17 Boral 0.99519
.00286 l
Mean K rr Value 1.00157 e
Standard deviation
.00419 t
l 1
e 9
.