ML19260C064
| ML19260C064 | |
| Person / Time | |
|---|---|
| Site: | 07105395 |
| Issue date: | 10/17/1979 |
| From: | Dipiazza R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 14769, NUDOCS 7912180304 | |
| Download: ML19260C064 (37) | |
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PbR Westinghouse Water Reactor Box 355 Electric Corporation Divisions PittsburghPemsy!vania15230 October 17, 1979 WRD-LS&S-853 U. S. Nuclear Regulatory Commission Office of Nuclear Material Safety & Safeguards Division of Fuel Cycle & Material Safety Washington, D.
C.
20555 Attention:
Mr. Charles E. MacDonald, Chief Transportation Branch
Subject:
Renewal of CC Shipping Package, Certificate of Compliance 5395, Docket 71-5395 Gentlemen:
The Westinghouse Electric Corporation hereby requests renewal of Certificate of Compliance 5395, Docket 71-5395 in accordance with the information attached to this letter.
The package information referenced by the current NRC certifica-tion has been consolidated into a single application per your request of June 6, 1979.
A check in the amount of $150.00 for the renewal of this certifi-cate will be transmitted under separate correspondence as soon as it is received from the Westinghouse Accounting Department.
Please send the renewed certificate to me at the above address.
If you have any questions regarding this matter,please write me at the above address or telephone me on 412-373-4652.
Very truly yours, onald P.
DiPiazza, M
.m NES License Administration lk
/slw lg Attachments a
14769 7912180 30y t
f 1.0 C.C.
Packaging 1.1 Papkaging Description Designation - C.C.
(Cruciform Control) Shipping Container Gross Weight - 4500 pounds Fabrication - The design and fabrication details for the CC Shipping Container are given in Champion Co.
drawings 10410,.10536, 10538 and 10541, which are attached as Appendix A to the application dated October 15, 1979 on Docket 71-5395.
Drawing SKA-219, which is also attached as part of Appen-dix A to this application, demonstrates the use of Top and Side Mounting Assemblies to clamp the contents in place.
Coolants - Not applicable.
1.2 Contents Description - CC Radioactivity - Not applicable for enriched uranium.
When mixed oxides are loaded, the maximum activity per package shall be 400,000 curies.
Identification and Enrichment of SNM - The SNM will be unirradiated uranium enriched < 5 w/o in the iso-tope U; or mixed oxides (PuO2-Uo2) containing up to 8 w/o total PuO2 in UO2 derived from natural or depleted uranium.
Form of SNM - the SNM will be in the form of clad fuel assemblies.
Specific data on maximum assembly parameters are included in Table 1.
The reactivity values listed in Table 1 equal the computed reac-o
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tivity adjusted by a factor to provide for the 1637 240 Docket 71-5395 Date: 10-15-79 Revision No.
Date:
Page 1
^
e 1.2 Contents Description - CC '(cont. )
probable error in the calculations.
The contents will be loaded in such a fashion that if the pack-age were to be flooded and subsequently drained, any water which may have penetrated the contents would drain simultaneously.
In the clad form, the assemblies will not disrup-tively react or decompose at the Accident Thermal Tests have established that the clad condition.
fuel rods meet the criteria for special form mat-erial.
No chips, powders, or solutions will be offered for transport in this packaging.
Neutron Absorbers, etc. - Not specified.
No decrease in reactivity resulting from this type of material is included in the nuclear safety analysis.
Maximum Weight of Fissile Content - Listed in Table 1.
Maximum Net Weight of Contents - Listed in Table 1.
~
Maximum Decay Heat - Not applicable.
2.1 General Standards for All Packages (Compliance with Subpart C of 10 CFR 71)
General Standards - The materials specified for the pack-age will not produce significant chemical or galvanic reactions.
The closure devices specified must be deliberately unfastened.
Each of the four lifting lugs will be capable of supporting the loaded container individually, so the system of four lifting lugs will support three times the weight of the loaded container.
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In addition, uhm design of the lifting lugs will be such 1647 241 Revision No.
Date:
Page 2 k f 71-539 Date: lo_15_79
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that, under an excessive load, the lug will fail across the hole before it would transmit disruptive stresses to the container.
Similarly, the tie-down devices, which will accomodate 1" diameter steel cables, will have adequate strength to meet the static load require-ment.
The requirements cf 10 CFR 71.32 are readily fulfilled by this package.
Calculations of the stresses which would occur in the packaging shell when transporting fuel assemblies result in a computed maximum stress of 6900 psi for a 36,000 lb. (5 x 7200 lb.) uniformly N
distributed load.
This load is conservative, even when transporting boxed fuel rods.
The stress value is well below the accepted yield strength value of even a mild steel, and is conservative since the calculations do not consider the effects of the rest of the packag-ing structure, such as the very rigid support members that carry the shock mounts.
Basic computations are supplied in Appendix B.
2.2 Normal Conditions of Transport Single Package Evaluation - Westinghouse requests that the test conditions specified in paragraphs 4.4.2, 4.4.3, 4.4.5, 4.4.6, 4.4.7, and 4.4.8 of specifica-tion MIL-C-5584B be approved as normal conditions of transport in lieu of those specified in Appendix A of Part 71.
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16@f 243 0 k t 71-5395 Date: 10-Revision No.
Date:
Page 4
2.3 Hypothetical Accident Conditions The accident conditions specified in Appendix B of 10 CFR Part 71 will not credibly produce an arrangement more reactive than that analyzed under General Criticality Standards.
The clearance between the massive vertical member of the internal components and the shell of the container will be such that a fuel assembly will not pass.
Any crushing due to the Accident Free Drop would only decrease the clearance.
The necessity for evaluating the Accident Puncture and the Water Immersion conditions is obviated by.the assumption of maximum moderation and full reflection.
The necessity for evaluating an Accident Thermal condition is obviateu by the limitation of the contents to clad fuel assemblies.
3.0 Criticality Evaluation 3.1 General Criticality Standards The contents of each package will be so limited that, for a single container, with the contents maximumly moderated and fully reflected, the adjusted eff of the contents will be < 0.90.
No consideration of dispersable material is required because the contents will be limited to clad components.
The criticality calculations are presented in Appendix B.
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3.2 Package Arrav Evaluation 161k7 244 Any number of undamaged, unmoderated CC packages will be nuc1carly safe, since uranium enriched to 5 w/o or less 3
in U is nuclearly safe in any quantities under any D ket 71-5395 Dofe: 10-15-79 Revision No.
Date:
Page 5
3.2 Package Array Evaluation (cont. )
conditions.
Any number of undamaged but flooded packages will also be nuclearly safe, since a single package will have a eff of < 0.90 and there will be a minimum of 12 inches of water between the contents of any two packages.
If the water drains away, the contents will also drain, so that the array returns to an unmoderated condition.
For the low-enriched assemblies under consideration, studies using LEOPARD and PDQ-03 demonstrate that a rise in eff (above the single package value) for an unlimited array does not cccur at any reduced water density, due to the parasitic neutron absorption by the container walls and internal structure.
Consequently, no degree of interspersed partial moderation can produce an array eff k
in excess of the single package eff resulting from complete flooding.
The maximum credible accident condition is conceived to involve only two packages, crushed top-to-top so that the spacing between the pairs of assemblies will be.500 inches, and aligned parallel to each other.
This array is'then assumed to be flooded.
The heavy structural members of the base and the internal component support structures of the packagings will provide sufficient spacing so that any other package in the shipment will be siolated from this combination by a minimum of 12 inches of water.
The Nuclear Engineering Department will use the calculational s
procedures specified under General Criticality Standards
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16M 245 eooe e
71-5395Date: 10-15-79 Revision No.
Dofe:
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3.2 Package Array Evaluation (cont.)
to assure that the adjusted eff docs not exceed 0.98 for the four assemblies in the maximum credible accident (MCA) array.
Since only two packages will combine to form the MCA
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k array with a eff < 0.98, and any additional packages can only form similar isolated arrays, any number of the CC packages will be nuclearly safe under hypothetical accident conditions.
3.3 Fissile Class II Limits Thirty-two (32) packages will be offered as a maximum Fissile Class II shipment.
Each package will be assigned 1.2 radiation units.
The nuclear safety evaluations indicate that any number of packages would be safe.
The limitations specified in this section have been based on considerations of vehicle capacity, shipment value and other non-nuclear considera-tions.
3.4 Fissile Class III Limits - CC A maximum of sixty (60) packages will be offered as a kk Fissile Class III shipment.
1647 246 4.0 Procedural Controls
.Each packaging will be visually inspected prior to each use.
If components, damaged or deteriorated so as to compromise the effectiveness of the package, are detected, they will be repaired or replaced to restore the packaging to essentially "like-new" condition.
The Manager, Production Control, will be responsible to maintain the required records on each shipment.
Docket 71-5395Dafe: 10-15-79 Revision No.
Date:
Page 7
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r s
e APPENDIX B CC PACKAGE NUCLEAR CRITICALITY 1.0 SAFETY MODEL This report describes the nuclear model used for the Model CC Shipping Package nuclear criticality safety analysis.
Figure 1 details the geometrical model.
Table 1 gives the element atom densities and Table 2 the nuclear. constants used in the two-group criticality calculations.
The model is given for a single flooded assembly; for two flooded assemblies (one package) ; and for the MCA, which is two packages crushed to a separation of.0.5 inches between assemblies (the maximum reactivity configuration).
The model is conservative in two respects:
1.
The container material is deleted from most of the calculations.
The MCA was run with and without the steel and a eff less than 0.98 was obtained in both cases, and 2
2.
The transverse leakage, B3, was set equal to zero.
Since the maximum eff computed using these conservative assump-tions was less than the specified maximum for each case, the CC Shipping Package nuclear criticality safety has been amply demon-strated.
Ik 1647 248 o
71-5395Date:
10-15-79 Revision No.
Date:
Page 9
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k 71-5395Date: 10-15-79 Revision No.
Date:
Page 10
TABLE 1 SAFETY MODEL ATOM DENSITIES (1024)
Assembl1 (flooded),
Enrichment = 4.5 w/o Water (n =1. 0g/cc)
Element
.0667
.03993 Hydrogen
.0333
.03451 Oxygen
.006415 Stainless Steel 235
.0003292 0
6
.00000218 0
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.006938 U
.00000195 U
N 160>7 250
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Revision No.
Date Page _11 Dafe: 10-15-79
e TABLE 2 SAFETY MODEL NUCLEAR CONSTANTS (All Temperatures = 68
- F)
Assembly Water Steel A. Group 1 D (cm) 1.16104 1.2333 1.18 E
(cm
).012944 0.000471 0.00055 a
E (cm
).02144 0.04760 0.000331 r
vf (cm-1).010846 0.0 0.0 E
B.
Group 2 (0.0 - 0.625 ev)
D (cm) 0.21435 0.165465 0.337 E
(cm-1) 0.264446 0.022210 0.22 a
VEf (cm
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16M 251 71-5395Date:10-15-79 Revision No.
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Page 12 p
APPENDIX B 2.0 CALCULATIONS 1.
INTRODUCTION These calculations are in support of an application for an NRC license to ship cruciform core SS clad, SS can uranium assemblies enriched up to 4.5 w/o.
It is concluded that such assemblies can be safely shipped.
A more detailed description of the assembly design is given below, together with details of the calculations performed and their results.
2.
ASSEMBLY DESCRIPTION AND ASSUMPTIONS A cross-section of the assembly appears in Figure 1.
There are eleven thick clad rods in corners as shown in the figure.
The average rod pitch is.518", the maximum fuel enrichmen;., 4.5 w/o and the density, 95% of theoretical.
The shipping container in which the fuel will be shipped in des-cribed in the following Champion Company, Springfield, Ohio drawings:
Drawing No.
Title 10410 Container - Shipping, Reusable, Metal 10541 Mounting Assembly Bottom 10536 Bottom Assembly (two sheets) 10538 Top Weldment (three sheets)
The manner of placing the two assemblies in each container is such that the sides of the assembly that are recessed to accommodate the cruciform rods, are away from each other as shown in Figure 2.
The assemblies have been represented in the spatial x-y calculations accordingly.
8 The following assumptions have been made which are all conservative from the standpoint of criticality safety:
16I>f 252 Docket 71-5395 Dofe: 10-15-79 Revision No.
Date:
Page 13
2.
ASSEMBLY DESCRIPTIONS AND ASSUMPTIONS (cont. )
The assemblies have been assumed to be of infinite length, i.e., they have a zero axial buckling.
None of the structural material of the cask has been included in the calculations, except the wall thickness in the maximum credible accident (MCA).
As the fuel is under-moderated in the assembly design, the density of the water has been assumed to be at its maximum value of 1.0 gms/cc.
3.
CALCULATIONAL MODEL Calculations were performed using modified versions of the LEOPARD and PDQ03 codes.
The PDQ03 calculations were performed for each of the following cases:
A single assembly, flooded.
Two assemblies in a shipping cask, flooded.
Two shipping casks crushed together and flooded - the MCA.
The two assemblies in each cask are assumed to remain intact and separated by their usual distance.
The casks are assumed to be crushed together so that both pairs of assem-blies are perfectly aligned opposite each other.
Onl'y the wall thickness of the two casks are assumed to lie in between the pairs of assemblies.
The two pairs of assemblies are then assumed to be separated by that distance which gives the highest multiplication factor.
The geometry, mesh intervals and boundary conditions used in the PDQO3 calculations for each of the above three cases are shown in
'k Figures 3 through 5 respectively.
161>7 253 No.
Date:
Page 14
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3.
CALCULATIONAL MODEL (cont. )
For the MCA the distance separating the two pairs of assemblies was varied to determine that which gives the maximum multiplication factor.
Calculations were performed at the separations of.324",
.384",
.5",
1" and 2".
Initially no cask wall steel was placed in between.
The.384" separation gave the highest multiplication.
Cask wall steel was then inserted.
4.
CALCULATIONAL RESULTS The results of the above calculations are tabulated in Table 1.
k A plot of eff as a function of the distance separating the two pairs of assemblies in the MCA appears in Figure 6.
The criticality safety criteria for the single assembly flooded and the two assemblic.s in the cask, flooded, is a eff of 1 90.
For the MCA, with cask wall steel included, the required eff is 1 98.
The corresponding eff values obtained in the calculations appear at Serial Nos. 1, 2 and 8 respectively, in Table 1.
They are all below the criteria and therefore the package is safe for shipment.
The following additional information is included:
Atom Densities Table 2 Macroscopic Cross-sections Table 3 5.
CONCLUSIONS Cruciform core SS clad, SS can uranium fuel as described in this report and enriched up to 4.5 w/o can be safely shipped in the 16}
254 Champion Company shipping container.
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[1
- Barry, R.F.,
"The Revised LEOPARD Code - A Spectrum
'7 dent Non-Spatial Depletion Code" WCAP-2759 (March 1965)
[2
- Caldwell, W.R.,
"PDQO3 - A program for the Solution of the Neutron Dif fusion Equations in Two Dimensions on the IBM 704,"
WAPD-TM-179 (May 1960).
D ReWsion No.
Date:
% M Docket 71-5395 ate: 10-15-79
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Table 1 Mu'tiplication Factors Serial No.
Case K,ff-
.8109 1
Single assembly flooded
.8642 2
Two assemblies in cask flooded MCA (two pairs of assemblies in two casks 3
crushed together) with no cask wall steel in between.
Separation between
.9996 pairs of.324".
Same as 3 wi'.h.384" separation 1.00228 4
Same as 3 with.5" separation 1.'00195 5
Same as 3 with 1" separation
.9936 6
Same as 3 with 2" separation
.9583
~
7 Same as Serial No. 3 with cask wall 8
.9712 steel in between
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16M 255 Docket 71-5395Date: 10-15-79 Revision No.
Date:
Pope 16
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Table 2 Atem Densities 24 (Atoms /cc,10 units)
Element Flooded Fuel Water Steel Assembly, 4.5w/o Hydrogen
.03800
.06688 Oxygen
.03439
.03344 Zircaloy
.0001513 i
Stainless Steel
.007299
.0897 U-234
.000002854 U-235
.0003506 U-236
.000002184 U-238
.007342 N
16M 256 s
D9 kf 71-5395Date: 10-15-79 Revision No.
Date:
Page 17
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Table 3 Macroscopic Cross-Sections (Temperature = 60*F, Water density used = 1.0 gms/cc)
Water Steel Cross-Section Flooded Fuel Assembly,4.5w/o Grour, 1 1.2855 1.18 1.1306 D (cm)
.0004409
.00055 E,(cm-l)
.01119
.04888
.000331 ER(cm~l)
.02146 uEf(cm-I)
.009583 Group 2
.1642
.337
.2077 D (cm)
.0222
.22 E,(cm-l)
.2353 urf(cm-l )
.3796 6
16M 257 s
D eket 71-5 39 5pote.10-15-79 Revision No.
Dafo:
Page is
Tabic 4 MCA - Row Mesh Intervals From Row 27 through Row Boundary
, Applies to Figure 5)
(
Separation Between Boundary Mesh Intervals Pairs of Assemblies Row No.
Row 27 through Inches Row Boundary, cm
.324 28
.4115 s
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.384 29
.0762,.4115
.5 29
.2235,.4115 1.0 30
.4293,.4293,.4115 2.0 31
.7095,.7095,.7095,
.4115 NOTE: The.4115 cm distance rep-esents half the thickness of the cask wall steel between the two pairs of assemblies.
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16s1 258
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Docket 71-5395Date:10-15-79 Revision No.
Date:
Foge 19
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INSTRUMENT WATER HOLE hic g(
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'd LLS Zr FILLER BAR Tiliti CLAD RCDS SS CAit AROUND ASSEMBLY Figure 1 Assembly Regular and Thick Clad Rod Pattern
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1601 259 Docket 71-5395 Dofe: 10-15-79 Revision No.
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Page 20
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- Sides recessed to accomodate cruciform rods.
Clamps press assemblier down on these sides f'
against.a rigid base and stu::
separating them.
f e
6 Figure 2 Manner of Placing Cruciform Assemblies in Shipping Container
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n 1601 260 Docket 71-5395 Date: 10-15-79 Revision No.
Date:
Page 21
t 0
8-27 -
35 0
Nl 8 -
27 --
35 1
Boundary conditions:
Zero flux on all four sides.
Column mesh intervals in cm:
Columns 0 thru 8; reflector:
3.0, 3.0, 3.0, 3.0, 3.0, 2.0, 1.0,.5 Columns 8 thru 27; assembly:
.1791,1.0922,.2235,13 times 1.3157,.2235, 1.0922,.1791 Columns 27 thru 35; reflector:
same as columns 0 thru 8 in reverse
. Row mesh intervals:
same as column mesh intervals Figure 3 Geometry and Mesh Intervals for a Single Assembly Flooded O
16M 261 Docket 71-539 Bate: 10-15-79 Revision No.
Defe:
Page 22
8' 27 '
35 0-8 27 --
35 -
Boundary conditions: Zero current on left, Zero flux on the remaining three sides Column mesh intervals in cm:
Columns 0 thru 8: gap between the two assemblies in cask,.51,.51,.51,.51,
.51,.51,.50,.25 Columns 8 thru 35, assembly and reflector:
same as for columns 8 thru 35 in Figure 3 Row mesh intervals:
same as in Figure 3 Figure 4 Geometry and Mesh Intervals for Two Assemblies in Cask Flooded
\\1 16M 262 Docket 71-5390afe: 10-15-7 9 Revision No.
Date:
Page 21
0 8
27 35 0-8--
27 --
Variable (see below)
Boundary conditions: Zero current on left and lower sides, Zero flux on remaining two sides Column mesh intervals: Same as in Figure 4 Row mesh intervals in cm:
Row 0 thru 27, same as in Figure 3 Row 27 thru row boundary, see Table 4 Figure 5 Geometry and Mesh Intervals for the MCA
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16M 263 24 Docket 71-5395Date: 10-15-79 Revision No.
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Page
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Date.16M 264 W69s71-539 5 Date: 10-15-79 Revision No.
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