ML19310A882
| ML19310A882 | |
| Person / Time | |
|---|---|
| Site: | 07106581 |
| Issue date: | 06/30/1980 |
| From: | SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19310A880 | List: |
| References | |
| 16596, XN-052-01, XN-052-R01, XN-52-R1, NUDOCS 8006300623 | |
| Download: ML19310A882 (74) | |
Text
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SUMMARY
OF XN-52, REV." 1 CHANGES (June 1980) f Original Section No.
Page No.
Change Description N.A.
iv thru xii Modified Table of Contents, List of Tables and List of Figures due to changes result-ing from reduction in the number of M0 fuel i
types.
t; 2.0 2-1 Added reference to a listing of ifcensing drawings.
2.1.2 2-6 Reduced the number of MO fuel types from six g
to four and renumbered accordingly.
2.4.1 2-12 Reduced the number of M0 fuel types, added fuel rods to identified package content and referenced location of identified
[
package limits per Fissile Class III shipment.
2.4.2 2-13 Modified reference to table number changed by deletions resulting from reduction of MO fuel types.
Table 2-I 2-14 Reduced number of M0 fuel types and renumbered accordingly.
Table 2-II 2-15 Reduced number of M0 fuel types and renumbered-accordingly.
Table 2-IV 2-17 Reduced number of M0 fuel types, renumbered accordingly, and added identification of Fissile Class and number of packages per each Class III shipment.
Table 2-V 2-18 Added limitation on the number of packages i
per Fissile Class III shipment.
Tables 2-VI 2-19 &
Modified to reflect reduction and renumber-thru 2-IX 2-20 ing of M0 fuel types.
Table 2-X 2-21 Renembered from original Table 2-XII as a result of deletion of two M0 fuel types.
Table 2-XI New Added new table which gives a sumary list-ing of the appropriate licensing drawings and identifies the applicable containers.
Table 7-I 7-2 Deleted two M0 fuel types and renumbered accordingly.
8006300h g J
a 11.0 11-1 thru Dcleted discussion of XN Fuel Types I and 11-3 II as not applicable due to deletions and fuel type renumbering.
12.0 12-1 &
Deleted two M0 fuel types and renumbered 12-2 accor.dingly.
12.1 12-2,3,4, Deleted reference to methods used for 5, a 6 analysis of the deleted M0 fuel types, renumbered remaining fuel types, sub-sections, and tables appropriately.
12.2 12-7, 8 Deleted reference to calculations for 9 & 10 two deleted M0 fuel types, renumbered atmaining fuel types, subsections, and tables appropriately.
12.3 12-10, 11, Same as for Section 12.2.
12 & 13 12.4 12-14, 15, Same as for Section 12.2 16, 17 & 18 12.5 12-27 Renumbered M0 fuel types and referenced tables due to deletion of two fuel types.
Table 12-I 12-28 Deleted Table 12-II 12-29 Deleted Tables 12-III 12-30 thru Renumbered and changed references to tables thru 12-XVII 12-44 and sections as appropriate.
Figures 12.1 12-45 thru Revised pagination and referenced tables.
thru 12.9 12-53 Figure 12.10 1 2-54 Deleted - applicable only to a deleted fuel type.
Figure 12.11 12-55 thru Revised pagination and numbering system due thru 12.14 1 2-58 to deleted fuel type.
13.0 13-1 Added references 8 and 9.
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GUIDE FOR INCLUSION OF REVISED PAGES OF XN-52, REVISION 1 (June 1980) l.
Replace pages iv through xii with pages iv through xii dated June 1980.
i.
2.
Replace page 2-1 with page 2-1 dated June 1980 3.
Replace page 2-6 with page 2-6 dated June 1980.
I Replace pages 2-12 through 2-15 with pages 2-12 through 2-15 4.
dated June 1980.
5.
Replace pages 2-17 through 2-21 with pages 2-17 through 2-21(a)
I dated June 1980.
6.
Replace page 7-2 dated April 1980 with page 7-2 dated June 1980.
7.
Replace pages 11-1, 2 and 3 with pages 11-1 and 2 dated June 1980.
8.
Replace Section 12 in its entirety with the new Section 12 dated June 1980.
9.
Replace page 13-1 with page 13-1 dated June 1980.
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iv XN-52 Rev. 1 Y
g June 1980
?
CONSOLIDATED LICENSE APPLICATION FOR EXXON NUCLEAR COMPANY, INC.
MODEL 51032-1 AND -la SHIPPING CONTAINERS Certificate of Compliance 6581
(
Docket 71-6581 h
TABLE OF CONTENTS I
SECTION TITLE PAGE NO.
1.0 INTENT.......................
1 -1 2.0 PACKAGE DESCRIPTION................
2-1 2.1 Model 51032-1 Container..............
2-1 2.1.1 Container Description..
2-1 2.1.2 Fuel Element Clamps, Shock Mounts & Separator Blocks 2-5 2.2 Model 51032-la Container..............
2-8 2.2.1 Container Description...............
2-8 l
2.2.2 Fuel Element Clamps, Shock Mounts & Separator Blocks 2-9 2.3 Both Models 51032-1 and 51032-la Containers....
2-11 2.4 Package Contents..................
2-11 2.4.1 Model 51032-1 Container..............
2-11 2.4.2 Model 51032-la Container..............
2-13 3.0 PACKAGE HANDLING..................
3-1 3.1 Package Loading........
3-1 3.2 Transport Controls.................
3-6 3.3 Unloading.....................
3-7 4.0 PROCEDURAL CONTROLS...............
4-1 5.0 GENERAL STANDARDS FOR PACKAGING...........
5-1 6.0 STRUCTURAL STANDARDS FOR LARGE QUANTITY PACKAGING.
6-1 7.0 CRITICALITY STANDARDS FOR FISSILE MATERIAL PACKAGES _
_ 7-1 8.0 EVALUATION OF A SINGLE PACKAGE...........
8-1
'N' y
XN-52, Rev. 1 June 1980 t
TABLE OF CONTENTS (Continued)
SECTION TITLE PAGE NO.
9.0 STANDARDS FOR NORMAL CONDITIONS OF TRANSPORT....
9-1 10.0 STANDARDS FOR HYPOTHETICAL ACCIDENT CONDITIONS...
10-1 10.1 Model 51032-1 Package...............
10-2 10.1.1 Free Drop Tests..................
10-2 10.1.1.1 Summary of Model 51032-1 Drop Tes ts........
10-2 10.i.2 Package Component Tests and Evaluations......
10-5 10.1.2.1 Model 51032-1 Separator Block Integrity......
10-5 10.1.2.2 Integrity of the Aluminum Clamps..........
10-6 10.1.2.3 Short Strongbacks Used in Some Shipments......
10-6 10.2-Model 51032-la Packages..............
10-7 10.2.1 Model 51032-la Container-End Drop Evaluation....
10-10 10.2.2 Model 51032-la Container - 75? Cover Corner Drop Evaluation..................
10-11 10.2.3 Model 51032-la Container - Horizontal Cover Drop Eval ua tion..................
10-12 10.2.4 Model 51032-la separator Block Integrity......
10-14 10.3 Fuel Rod Drop Tests................
10-14 10.4 Summary......................
10-15 11.0 EVALUATION OF AN ARRAY OF PACKAGES.........
11-1 12.0 SPECIFIC STANDARDS FOR FISSILE CLASS I AND III PACKAGES........'..............
12-1 12.1 Method, Discussion, and Verification........
12-2 12.1.1 XN Type I, II, III, IV, AA and Generically Characterized Fuel Elements............
12-2 12.1.1.1 KENO II (18 Energy Group) Calculational Method...
12-2
- 12.1.1.2 KENO IV (123 Energy Group) Calculational Method..
12-4 12.2 Results of k, Calculations.............
12-4
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vi XN-52, Rev.1 June 1980 h
TABLE OF CONTENTS (Continued) i t
SECTION TITLE PAGE NO.
i 12.2.1 XN Types I, II, III, and IV Fuel Elements.....
12-4
/
12.2.2 Generically Characterized Fuel Elements......
12-6 12.2.3 XN Type AA Fuel Elements..............
12-7 L
12.3 Single Package Evaluation.............
12-7 12.3.1 XN Types I, II, III, and IV Fuel Elements.....
12-7 12.3.2 Generically Characterized Fuel Elements......
12-8 1
12.3.3 XN Type AA Fuel El ements..............
12-9 12.4 Demonstration of Compliance with 10 CFR 71.38 and 71.40.......................
12-10
{
12.4.1 Undamaged Fissile Class I Package Arrays......
12-10 12.4.1.1 XN Types I, II, and IV Fuel Elements........
12-10 12.4.1.2 Generically Characterized Fuel Elements......
12-12 12.4.1.3 XN Type AA Fuel El ements..............
12-14 12.4.2 Undamaged Fissile Class III Package Arrays.....
12-15 12.4.2.1 XN Type III Fuel El ement..............
12-15 12.4.2.2 Generically Characterized Fuel Elements......
12-15 12.4.3 Damaged Package Arrays....~...........
12-16 12.4.3.1 XN Types I, II, III, and IV Fuel Elements.....
12-16 12.4.3.2 Generically Characterized Fuel Elements......
12-17 12.4.3.3 XN Type AA Fuel El ements..............
12-18 12.4.3.4 3hipments of Individual Rods.............
12-19 12.5 Summary......................
12-20
13.0 REFERENCES
13-1 9
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vii XN-52, Rev. 1 June 1980 LIST OF APPENDICES APPENDIX NO.
TITLE APPENDIX I APPLIED DESIGN COMPANY, INC. LIFT EYE ANALYSIS 4
r APPENDIX II STRUCTURAL ANALYSIS OF MODEL 51032-1 PACKAGING TIE-DOWN SYSTEM 8
APPENDIX III APPLIED DESIGN COMPANY, INC. REPORT 2526A
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APPENDIX IV 30-F00T DROP TEST PROCEDURE AND REPORT PACKAGING MODEL 51032-1 APPENDIX V PACKAGE COMPONENT EVALUATIONS APPENDIX VI FUEL R0D DROP TEST REPORT i
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viii XN-52, Rev. 1 June 1980 LIST OF TABLES 4
TABLE NO.
TITLE PAGE NO.
2-I Revised Fuel Element Identification Numbers 2-14 2-II Radioactive Material Limits (Mixed 0xide Fuels)...................
2-15 L
2-III Radioactive Material Limits (UO Fuels)...
,2-16 2
2-IV Fissile Material Limits (Mixed Oxide Fuels).
2-17 2-V Fissile Material Limits (U0 Fuels).....
2-18 2
6 2-VI XN-Type I..................
2-19 2-VII XN-Type II.................
2-19 2-VIII XN-Type III.................,
2-20
[
2-IX XN-Type IV.................
2-20 2-X Limiting Fuel Element Physical Characteristics...............
2-21 2-XI Sumary Listing of Applicable Licensing Drawings..................
2-21(a) 7-I Individual Package Reactivities.......
7-J 10-I Energy Dissipation Accounting for Model 51032-la Packages Containing XN-Type AA Fuel Elements Relative to Drop-Tested Package...................
10-17 12-I Calculated k, for Unmoderated 5 wt% U-235 Enriched UO 12-22 2................
12-II Theory-Experiment Correlations.......
12-23 12-III Comparison of Computed Infinite Media Multiplication Factors...........
12-24 12-IV Mixed Oxide Fuel Element Single Package Evaluation.................
12-25
ix XN-52, Rev. 1 5
June 1980 LIST OF TABLES (Continued)
TABLE NO.
TITLE PAGE NO.
12-V Single Damaged Package Evaluation......
12-26 12-V' Fuel Element. Description..........
12-27 12-VII Reactivity of Undamaged Fissile Class I Package Arrays...............
12-28 12-VIII Undamaged Arrays of BWR Sized Fuel Elements.
12-29 12-IX Undamaged Array--Unmoderated Fuel Elements.
12-30 12-X Fuel Assembly Description..........
12-31
{
12-XI Two Undamaged Shipments...........'
12-32 12-XII Mixed Oxide Fuels - Damaged Package Arrays.
12-33 12-XIII UO Fuel Element - Damaged Package Arrays.. '
12-34 2
12-XIV Sumary of Model 51032-1 and -la Packaging Limtis..............
12-35 12-XV Sumary of Computed Reactivities for XN Fuel Types................
12-36 4
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x XN-52, Rev. 1 June 1980 LIST OF FIGURES 4
FIGURE NO.
TITLE PAGE NO.
2.1 Containment Vessel (Isometric View)....
2-22 2.2 Containment Vessel Layout.........
2-23 2.3 Base Assembly - Model 51032-1 and -la...
2-24 2.4 Cover Assembly - Model 51032-1 and -la..
2-25 2.5 Standard Model 51032-1 Strongback.....
2-26 2.6 Four Fuel Element Packaging 8
(Isometric View).............
2-27 1
2.7 Short Strongback.............
2-28 2.8 Instrumented Square Fuel Element g
Shipping Arrangement...........
2-29 2.9 Instrumented Triangular Fuel Element Shipping Arrangement...........
2-30 2.10 Instrumented Fuel Element Strongback Modifications...............
2-31 2.11 Model 51032-1 Component Details......
2-32 2.12 End Thrust Bracket - Model 51032-1 & -la.
2-33 j
2.13 BWR Fuel Packaging Arrangement -
Model 51032-1...............
2-34 2.14 Model 51032-la Container General j
Arrangement................
2-35 2.15 Model 51032-la Container Strongback....
2-36 2.16 Model 51032-la Separator Block......
2-37 i
2.17 Model 51032-la PWR Fuel Element Clamp Assembly.................
2-38 2.18 Model 51032-la BWR Fuel Element Clamp Assembly.................
2-39 2.19 Type AA Fuel Element Thrust Brackets...
2-40 2.20 Honeycomb Energy Dissipation Components -
Model 51032-la..............
2-41 1
a
.Y' I.
l xi XN-52 Rev. 1 2
_ l June 1980 LIST OF FIGURES (Continued)-
?
FIGURE NO.
TITLE PAGE NO.
l 3.1 Package Tie Down System..........
3-10 3.2 Shipping Record Sheet 3-11 3.3 Department of Energy Regional Coordinating Offices for Radiological Assistance and i
Geographical Areas of Responsibility...
3-12 3.4 Radioactive Material Shipping Inspection Record..................
3-13 l
10.1 Steel and Aluminum Clamp Assembly Force Deflection Curve Comparison........
10-19 12.1 Infinite Media Multiplication Factors for Low Enriched 0.5 inch UO Rods in Water as 2
a Function of the Water-to-Fuel Volume Ra ti o.......
12-37 12.2 5.0 wt.% U-235 Enriched UO Rod-Water 2
Lattice Infinite Media Multiplication Factors..................
12-38 12.3 4.0 wt.% U-235 Enriched U0 Rod-Water 2
Lattice Infinite Media Multiplication Fa cto rs..................
12-39 12.4 3.0 wt.% U-235 Enriched U0 Rod-Water 2
Lattic. Infinite Media Multiplication Fa c to rs.................
12-40 12.5 Single Package, Damaged, Fully Flooded..
12-41 12.6 Assumed Configuration of Damaged Package Arrays..-............
12-42 12.7
' Single Cell of Infinite Array of Undamaged Packages... T........
~12-43 w
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xii XN-52, Rev. 1 l
June 1980 LIST OF FIGURES (Continued)
FIGURE NO.
TITLE PAGE NO.
r 12.8 Assumed Geometrical Configuration of
. Undamaged Packages (Model 51032-1 )....
12-44 12.9 Assumed Geometrical Configuration of Undamaged Packages (Model 51032-la with Type AA Fuel Elements)..........
12-45 12.10 Undamaged Package Arrays.........
12.
12.11 Array of Damaged Packages.........
12-47 12.12 Fuel Rod Packaging (Typical)........
12-48 I
12.13 Individual Fuel Rod Packaging.......
12-49 4
4 4
2-1 XN-52, Rev. 1 June 1980 2.0 PACKAGE DESCRIPTION As specified in 10 CFR 71.22, the Model 51032-1 and
-la packages and their respective contents are described herein. For ready reference, a listing of the safety /licen-sing related drawings is provided in Table 2-XI.
2.1 Model 51032-1 Container The gross weight of the Model 51032-1 packaging is 4000
+ 100 pounds. Specific materials of construction, weights, dimensions, and fabrication methods of the packaging components are as described below.
2.1.1 Container Description The containment vessel is a 43 inch diameter (nominal dimension) right cylinder 216 inches long, fabricated of 11-gauge (0.1196 inch) steel (see Figures 2.1 and 2.2).
The containment vessel is fabricated in two sections-
-base and cover assemblies (see Figures 2.3 and 2.4).
Continuous 2 x 2 x 1/4-inch closure flanges are welded to the base and cover assemblies and a 1/2-inch rubber "0" ring gasket is fitted between the mating flanges. Using ten 1/2-inch steel alignment pins permanently fixed in the closure flange of the base assembly, the two halves of the containment vessel are mated and sealed together with 58,1/2-inch 13UNC-2A steel closure bolts; steel washers (9/32 inch thick) are inserted between the mating flanges to prevent excessive distortion of the "0" ring gasket; 1/2-inch 13UNC-28 steel-nuts tightly seated complete the closure.
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2-6 XN-52, Rev. 1 June 1980 of the clamps that contact the fuel element are lined with 1/4 inch thick Buna-N rubber pads. The BWR fuel element clamps (see Figure 2.13) are fabricated of aluminum with ethafoam (low density expanded polyethylene at approximately 6 pounds per cubic foot density) pads,
- 3/4 and s 1/2 inch thick, added between the fuel element i
and the strongback and clamps, respectively. Fuel elements supported in this manner may contain tight-fitting corru-gated polyethylene shims interlaced between adjacent rows of fuel rods within the fuel elements. A typical corrugated polyethylene shipping shim, and a schematic diagram showing the clamping method with associated shims' and ethafoam pads in place, are shown in Figure 2.13.
XN Types I, II, III, IV, and some of the generically characterized fuel elements will be packaged with molded corrugated polyethylene shims between adjacent rows of fuel rods within the fuel elements. When such shims a'e used in the packaging, ethafoam (low density expanded polytheylene at 6 pounds per cubic foot density) pads.75 and.50 inch thick will be added between the fuel elenent and the strons ack and clamps, respectively. These pads, used in conjunction with the clamping procedure described
^
above, provides support for the fuel elements while retaining the structural integrity of the shipping package.
The generically characterized UO fuel elements with 2
which such shims and pads are included, are identified in Section 12.5.
A comparison of the energy absorption capabilities of the alternative support methods indicates that the method using ethafoam pads will absorb at least 1.2 times the
2-12 XN-52, Rev. 1 June 1980 Currently licensed mixed Pu0 -UO fuel element identifi-2 2
cation numbers and the corresponding numbers used in this document are tabulated in Table 2-I.
Design characteris-tics for these four specific mixed Pu0 -UO nuclear fuel 2
2 elements and for generically described low-enriched U02 fuel elements are sumarized herein.
Identification of these fuel elements, along with the maximum number of elements and the maximum radioactivity of the radioactive constituents contained in a single package, are tabulated in Table 2-II for specific mixed-oxide (Pu0 -UO ) I"'I 2
2 elements and in Table 2-III for generically characterized UO fuel elements.
2 The identification and maximum quantities of the fissile constituents contained in a single package are tabulated in Table 2-IV for specific mixed-oxide (PUO -UO ) #"'I
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2 2
elements and in Table 2-V for generically characterized UO fuel elements.
In addition, the Fissile Class for 2
each different fuel type is identified and, where appro-priate, the limiting number of packages per shipment is given.
Although Tables 2-III and 2-V address the contents of fuel elements, these same geometric and fissile material limits -
are requested to apply to the shipment of fuel rods. A description of the packaging method for individual rods and groups of fuel rods and the safety evaluation of those con-tents is given in Section 12.
All fuel elements contali. pelletized and sintered. UO r
2 Pu0 -UO encapsulated within stainless steel or zircaloy 2
2 tubing. The physical characterist es of the various fuel elements are tabulated in Table 2-X.
In all cases, individual rods are held in the respective arrays by
2-13 XN-52, Rev. 1 June 1980 upper and lower tie plates and intermediate spacers. The contained UO r Pu0 material is uniformly distributed 2
2 throughout the active length of the individual fuel rods.
Note that for the generically characterized fuel elements (XN Types A through F) the following conditions were assumed:
1)
The fuel is uranium-dioxide (UO ) at 95 percent of 2
theoretical density.
2)
The clad is zircaloy 2 or 4, conservatively modelled
~
as pure zirconium.
3)
The clad thickness assumed was 0.020 inch, a value which is conservatively less than any present Exxon Nuclear Zr clad thickness.
4)
The gas gap was assumed to be 0.005 inch.
As previously noted, some Exxon Nuclear fuel elements contain gadolinium, cobalt, or other neutron poison roos.
In all cases, these poisons are conservatively neglected in performing the criticality safety calculations.
2.4.2 Model 51032-la Container In addition to the contents described in Section 2.4.1, the Model 51032-la container may be used to transport larger fuel elements. The' maximum content weight for the Model 5102?-la container is 3700 pounds. One such fuel element design (XN-Type AA) is--currently licensed for -
transport in the Model 51032-la container. Details rel-ative to that fuel element are also given in Table 2-X and Section 12.
t-2-14 XN-52, Rev. 1 June 1980 TABLE 2-I REVISED FUEL ELEMENT IDENTIFICATION NUMBERS Licensed Fuel Element Revised Fuel Element Identification Number _
Identification Number VII I
VIII II XIV III XVII IV t
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2-15 XN-52, Rev. 1 June 1980 TABLE 2-II 4
RADI0 ACTIVE MATERIAL LIMITS (MIXED OXIDE FUELS)
Maximum Number of Maximum XN I.D.
Radioactive Elements Curies (Type)
Materials per Package per Package I
Pu0 -UO 2
13,400 2
2 II Pu0 -UO 4
49,200 2
2 III Pu0 -UO 4
51,300 2
2 IV Pu0 -UO 2
13,500 2
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.2-17 XN-52, Rev. ?
June 1980 TABLE 2-IV FISSILE MATERIAL LIMITS (MIXED OXIDE FUELS)
Maximum-Fissile Constituents" XN I.D. Fissile Packages Total Maximum (Type)
Class per Shipment I.D.
Quantity I.D.
Quantity (kg/ Package)
(kg/ Package)
I I
N.A.
U 362 U-235 7.6 Pu 2.42 Pu 2.00 f
II I
N.A.
U 51 0 U-235 16.0 Pu 6.00 Pu 4.80 f
III III 8
U 51 0 U-235 23.0 Pu 6.25 Pu 5.0 f
IV I
N.A.
U 240 U-235 5.0 Pu 1.8 Pu 1.4 f
- A summary of the fuel rods contained. in each specific mixed-oxide fuel element is presented in Tablet 2-VI through 2-IX.
t r-
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TABLE 2-V l
FISSILE MATERIAL LIMITS (00 FUELS) 2 XN Maximum Fissile Constituents fuel Fissile Packages Maximum Total Maximum Y /Y Type Class Per Shipment w f Enrichment I.D.
Quantity I.D.
Quantity.
(kg/ Package)
(kg/ Package).
A I
N.A.
1 2.1 3.5 U
700 U-235 24.5 3
B I
N.A.
1 2.1 3.5 U
1500 U-235 52.5 o$
C III 8
1 1.8 4.0 U
1500 U-235 60.0 D
II;I 8
1 2.1' 4.0 0
1500 U-235 60.0 E
III 8
1 2.3 4.0 U
1500 U-235-60.0 F
III 8
11 5.0 0
2 1500 U-235 75.0 R
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2-19 XN-52, Rev. 1 June 1980 TABLE 2-VI XN-Type I Number of Rods Fuel Rod Description 5
1.59 + 0.05_w/o U-235 12 2.42 + 0.05 w/o U-235 15 2.87 f; 0.05 2/o U-235 4
2.87 + 0.05 w/o U-235--l.0 f;0.05 w/o Gd 023 7
2.19 + 0.05 w/o Pu in Natural U 5
3.05 j; 0.05 w/o Pu in Natural U 1
Solid Zircaloy-2 (No SNM)
J TABLE 2-VII XN-Type II 4
Number of Rods Fuel Rod Description 16 2.30 j; 0.05 w/o U-235
+
32 3.20 j; 0.05 w/o U-235 40
- 4. o j; 0.05 w/o U-235 4
4.60 + 0.05 U-235--l.2 f; 0.05 w/o Gd 0 23 24 5.45 f; 0.05 w/o Pu in Natural U 4
.1 Solid Zircaloy-2 (no SNM)
I.N
].
2-20 XN-52, Rev. 1 June 1980 TABLE 2-VIII XN-Type III
~ Number of Rods
-Fuel Rod Description 16 2.30 1 0.05 w/o U-235 32 3.20 1 0.05 w/o U-235 36 4.60 1 0.05 w/o U-235 4
4.6010.05 U-235 - 1.210.05 w/o Gd 023 25 5.45 1 0.05 w/o Pu in Natural U 4
Cobalt targets (no SNM) 4 Solid Zircaloy-2 (no SNM)
TABLE 2-IX XN-Type IV Number of Rods Fuel Rod Description 20 2.64 1 0.05 w/o U-235-6 1.79 1 0.05'w/o U-235 8
2.74 1 0.05 w/o Fissile;-Pu in Natural U 2
2.24 1 0.05 w/o Fissile; Pu in Natural U i
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TA8LE 2-1 LIMITING FUEL ELEMENT PHYSICAL CHARACTERISTICS Fuel Nominal Nominal Nominai Nominal Nominal XN Rod ~
Array Rod Clad Nominal Fuel Pellet,
Active Fuel No. of Array Dimensions Pitch Clad Thickness Fuei Rod Diameter Fuel Length Type fuel Rods (Square)
(inches)
(inches)
Material (inches) 0.D. (inches)
(inches)
(inches)'
I 49.
7x7
-5.35 x 5.35 0.738 Zircoloy
.034 0.570 0.478 144.
II 121 11 x 11 6.52 x 6.52 0.577 Zircaloy
.034 0.449 0.371 70.
III 121 11 x 11 6.52 x 6.52 0.577 Zircaloy
.034 0.449 0.371 70.
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IV 36.
6x6 4.50 x 4.50 0.700 Zircaloy
.0302 0.563 0.481 116.
I A
< 100
< 10 x 10 < 5.2 x 5.2
< 0.806 Zircaloy
.020
< 0.550
< 0.500
< 192.
or S$*
or.015 8
1 289 1 17 x 17 1 8.55 x 8.55
< 0.806 Zfrcaloy
.020
< 0.550
< 0.500
< 192.
Y or $5 or.015 U
1 60 x 8.60
< 0.769 Ztrealoy
.020
< 0.550
< 0.500
< 192.
C 289 1 17 x 17 8
.or SS or.015 1 48 x 8.48
< 0.806 Zircaloy
.020
< 0.550
< 0.500
< 192.
D 1 289 1 17 x 17 8
or SS or.015 1 40 x 8.40
< 0.830 Zircaloy
.020
< 0.450
< 0.400
< 192 E
< 289
< 17 x 17 8
or SS or.015 1 00 x 8.00
< 0.806 Zircaloy
.020
< 0.550
< 0.500
< 192.
F 1 289 1 17 x 17 8
or SS or.015 a g 0.424 0.3565 154.
Ef AA" i ;36 16 x 16 9.01 x 9.01 0.563 Zircaloy
.030 1
49 y.
w..
May be Zircaloy-2. Zircoloy-4 or stainless steel. For criticality safety E :::
calculations a 0.20 inch zirconium clad was assumed.
O Type AA fuel elements may be transported only in Model 51032-la containers'.
,I
l.
2-21(a)
XN-52, Rev. 1 June 1980 TABLE 2-XI
SUMMARY
LISTING OF APPLICABLE LICENSING DRAWINGS Ref. Figure Applicable Number Drawing Number Package Model(s) 2.3 XN-NF-303,359 (Sheet 1, Rev.1) 51032-1 & -la 2.4 XN-NF-303,360 (Sheet 1, Rev.1) 51032-1 & -la 2.5 XN-NF-303,898 (Sheet 1, Rev. 0)
~ 51032-1 2.6 JN-300,607 (Sheet 1, Rev. 0) 51032-1 2.7 XN-NF-303,891 (Sheet 1, Rev. 0) 51032-1 2.8 JN-600,843 (Sheet 1 Rev. 3) 51032-1 i
2.9 JN-600,844 (Sheet 1, Rev. 2) 51032-1 2.10 XN-NF-303,890 (Sheet 1, Rev. 0) 51032-1 2.11 XN-NF-300,609 (Sheet 1, Rev.1) 51032-1 2.12 XN-NF-303,364 (Sheet 1, Rev.1) 51032-1 & -la 2.13 XN-303,897 (Sheet 1, Rev. 0) 51032-1 & -la 2.14 XN-NF-303,357 (Sheet 1, Rev.1) 51032-la 2.15 XN-NF-303,354 (Sheet 1, Rev. 2) 51032-la 2.16 XN-NF-303,357 (Sheet 2, Rev.1) 51032-la l
l 2.17 XN-NF-303,356 (Sheet 1, Rev.1) 51032-la l
2.18 XN-NF-303,818 (Sheet 1, Rev. 0) 51032-la 2.19 XN-NF-303,355 (Sheet 1, Aev.1) 51032-la 2.20 XN-NF-303,358 (Sheet 1, Rev. 1) 51032-la i
.g w,,. +.
7-2 XN-52, Rev. 1 June 1980 TABLE 7-I INDIVIDUAL PACKAGE REACTIVITIES XN I.D.
Single Package k,ff, I
< 0.634 +.013 II
< 0.765 +.009 III
< 0.762 +.011 IV
< 0.557 +.012 Generically
< 0.97 Characterized (UO )
2 AA
< 0.880 + 0.010 9
- h
l.
11-1.
XN-52, Rev. 1 June 1980-11.0 EVALUATION OF AN ARRAY OF PACKAGES, The effect of hypothetical accid.nt conditions specified in Appendix B of 10 CFR 71 on the nuclear safety of an i
ocray of Model 51032-1 or 51032-la packages has 'been evaluated by comparison to previously tested and licensed packages of similar design and construction, supported by additional drop tests perfomed on Model 51032-1 packages, component static tests where appropriate, and engineering evaluations where necessary (see Section 10).
It is assumed that each package in each shipment is daniaged to the extent described below.
XN Types I, II, III, IV, AA and Generically Characterized (UO ) Fuel Elements 2
For the purpose of nuclear safety calculations required by 10 CFR 71.38 and 10 CFR 71.40 (Paragraph 71.39 is not applicable since Model 51032-1 and 51032-la packages are transported as Fissile Class I or Fissile Class III),
each package is assumed to be damaged to the extent described below as the result of being subjected to the hypothetical accident conditions specified in Appendix B of 10 CFR 71.
1)
The fuel elements remain intact; fuel rods do not break nor are the fuel elements disarranged. A minimum separation of six (6) inches between adjacent fuel elements in a single package is maintained. The fuel elements will remain 7n the strongback, the -
strongback retains its geometrical shape, and the strongback remains enclosed within the packaging
.=
1 11-2 XN-52, Rev. 1 June 1980 (cover and base assemblies remain secured together).
The strongback breaks loose from the base assembly and the minimum distances between fuel elements in adjacent packages are:
Fuel Element Separation Model 51032-1 Model 51032-la cover-to-cover:
8 inches 8 inches side to side:
6 inches 5 inches cover to side:
8 inches 8 inches The effective volume of the containment vessel is not significantly reduced.
2)
Complete water moder,ation and full water reflection of the package contents is assumed in all nuclear safety calculations performed on packages subjected to the hypothetical accident conditions.
4
1 12-1
.XN-52, R2v. 1 June 1980 12.0 SPECIFIC STANDARDS FOR FISSILE CLASS I AND III PACKAGES Model 51032-l' packages containing XN Type I, II, and IV fuel elements are transported as Fissile Class I shipments on, or in, multiple use vehicles. Model 51032-1 packages containing XN Type III fuel elements are transported as Fissile Class III shipments on, or in, exclusive use vehicles.
l Model 51032-1 or 51032-la packages containing generically characterized (UO ) fuel are transported either as Fissile 2
Class I shipments on, or in, multiple-use vehicl,es; or as Fissile Class III shipments on, or in, exclusive use vehicles, as identified herein. Model 51032-la packages containing Type AA fuel elements are transported as Fissile Class I.
To demonstrate that shipments of these packages remain subcritical under all credible conditions, nuclear criti-cality safety evaluations have been made for each of the specific mixed-oxide (Pu0 -UO ) XN-type fuel elements 2
2 described in Section 2, and for Type AA fuel slements.
Furthermore, conservative limits on the physical dimensions, enrichment, fuel pellet diameter, and water-to-fuel volume ratio of generically characterized UO fuel elements, 2
from the viewpoint of nuclear criticality safety, have been established for both Fissile Class I and Fissile Class III shipments. The results of these evaluations are presented herein and a summary of the derived limits is given in Section 12.5.
The criterion used to derive limits on fuel element parameters was that for both normal conditions of transport and accident conditions (damaged package arrays), k,ff + 3a 5 0.970.
Fuel element parsmeter l
I
l-12-2 XN-52, R:v. 1 June 1980 limits were established by applying the criterion to both conditions with subsequent selection of the more conserva-tive limitations.
12.1 Method, Discussion, and Verification 12.1.1 XN Type I, II, III, IV, AA and Generically Characterized Fuel Elements 12.1.1.1 KENO II (18 Energy Group) Calculational Method The KENO-II Monte Carlo code was used to calculate reactiv-ities of interacting arrays of well moderated pa.ckages.
Multi group cross section data (18 energy groups) used in the Monte Carlo calculations were averaged by the GAMTEC-II and CCELL codes, respectively.
Extensive theory-experiment correlations have been pe'rformed using cross section data averaged by the GAMTEC-II code.
These evaluations, although primarily for plutonium fueled systems, demonstrate the self consistency of the GAMTEC-II code.
To demonstrate the adequacy of the GAMTEC-II code for undermoderated slightly enriched uranium systems, the infinite media multiplication factor was computed for 5 w/o U-235 U02 p wder using cross section data obtained from the ENOF/8-III library. The resulting value of k,,
as well as the values obtained using the JERBEL and HAMMER codes, are given in Table 12-I.
J
12-3 XN-52, Rev. 1 June 1980 The computed value for k, for unmoderated 5 w/o enriched U0 shows that the GAMTEC-II code, utilizing cross section 2
data obtained from ENDF/B-III library, is conservative with' respect to the other calculational methods by at least 2 percent.
Theory-experiment comparisons have been made for small water-moderated critical arrays of fuel rods. Such critical experiments have been evaluated using the KENO Monte Carlo code with 18 energy group cross section data averaged using the CCELL code.
The upper boundaries of the 18 energy groups used to average cross sections for these calculations were as follows:
10 Mev,,'.79 Mev, 6.07 Mev, 4.72 Mev, 3.68 Mev, 2.87 Mev, 1.74 Mev, 1.35 Mev, 183 Kev, 24.8 Kev, 3.36 Kev, 454 ev, 101 ev, 37.3 ev, 13.7 ev, 5.04 ev, 1.86 ev, 0.683 ev.
The results of these calculations are shown in Table 12-II and are presented with the results of other theory-experiment correlations in Reference 7.
Inspection of the results indicate that the calculational method yields conservative results relative to the experimental data.
In addition, the KENO calculated reactivities given in Table 12-II agree with the previously performed DTF-IV transport theory calculations within the statistical unce"tainty of the Monte Carlo calculations.
L
'f.
XN-52,~ Rev. l 4-June 1980 12.1.1.2 KENOLIV (123 Energy Group) Calculational Method
.In addition to the method described above, the KENO IV
' Monte Carlo code was utilized to calculate the reactivity _
of various undermoderated and moderated package arrays.
Multigroup cross section data from the XSDRN 123 group data library were produced for input into KENO IV using the NITAWL and XSDRNPM codes.
Specifically, the NITAWL code was used to obtain cross section data adjusted to account for resonance self-shielding by the Nordheim Integral Method. The XSDRNPM code, a discrete ordinates, one-dimensional, transport theory code, was then used to prepare cell-weighted cross section data representative l
of the fuel region for input into KENO IV.
Theory-experiment correlations have been performed for UO rod-water lattices using the 123 energy group XSDRN 2
cross section library data in KENO IV.
Results of these calculations are summarized in Table 12-II and are presented with the results of other theory-experiment correlations in Reference 7.
12.2 Results of k, Calculations 12.2.1 XN Types I, II, III, and IV Fuel Elements For the specific XN-type mixed-oxide fuel elements covered herein, values of k, have been computed, assuming full water moderation, using the CCELL code. The results of those calculations are shown in Table 12-III.
- b a.
12-5 XN-52, Rev. 1 June 1980 For comparison, Table 12-III also gives values of k, computed using the JERBEL code, or a two-dimensional diffusion theory code JOT.
The two-dimensional code used cross-section data averaged either by the CCELL code or by the JERBEL code, indicated as CCELL/ JOT or JERBEL/ JOT, respectively.
The two-dimensional code gives values of k, for the entire fuel element while values of k, computed using the CCELL and JER8EL codes assume a fuel element averaged pin.
This assumption has been shown to be conservative by comparisons with detailed design calculations (see Table 12-III for typical comparisons).
It is also apparent from data given in Tabl 12-III that for XN Type II fuel elements, the CCELL code is approximately 7 percent conservative with respect to the more detailed CCELL/ JOT method; and that, for XN Type III fuel elements, it is conservative by about the same amount relative to the JERBEL/JDT method.
This conservatism for the XN Types II and III fuel elements results due to the significant quantities of gadolinium which were neglected in the CCELL calculations.
Other calculations indicate that the actual degree of-conservatism is approximately 1 percent in reactivity for unpoisoned cases.
It is readily apparent that neutron poisons included in the fuel elements have substantial influence on criticality safety, and that the practice of neglecting them in criticality safety calculations introduces a significant degree of conservatism.
Note that XN Types I, II, and III mixed-oxide fuel elements all contain Gd 02 3 p isoned fuel rods which were ignored in the CCELL calculations reported in Table 12-III and in subsequent criticality safety calculations.
n l
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XN-52, Rev._1 12-6 June 1980 12.2.2 Generically Characterized Fuel Elements Infinite media multiplication factors for UO r d-water-2 lattice systems were calculated using the CCELL code for low U-235 enrichments as a function of enrichment (< 5 wt percent U-235), pellet diameter (< 0.5 inches), and fuel rod pitch (square lattice) or, equivalently, water-to-fuel volume ratio (< 2.3).
Results of k, calculations for rod-water lattices with limitations on the U-235 enrichment, pellet diameter, and water-to-fuel volume ratios as noted above, are summarized in Figures 12.1, 2, 3, and 4.
Examination of these data indicates that:
1)
Figure 12.1--The maximum values of k,for 3, 4, and 5 wt percent U-235 enriched UO rods in water occur 2
at water-to-fuel volume ratios of greater than 2.1.
2)
Figures 12.2, 3, and 4--The maximum value of k, for.
3, 4, and 5 wt percent U-235 enriched UO rods in 2
water occurs at a pellet diameter of 1 0.5 inches for water-to-fuel volume ratios of i 2.1.
At a water-to-fuel volume ratio of 2.3, the maximum value of k, occurs at a pellet diameter of 1 0.4 inch.
Calculational results summarized in Figures 12.1-12.4 indicate that for fully moderated fuel elements h'aving 4
enrichments of < 5 wt.% 235-0 and water-to-fuel volume ratios of 12.1, it is conservative to assume a pellet diameter of 0.5 inches. At a water-to-fuel volume ratio of 2.3 it is conservative to assume a pellet diameter of O.4 inch.
12-7 June 1980 XN-52, Rav. 1 12.2.3 XN Type AA Fuel Elements For'the XN Type AA fuel element the value of k,was computed assuming full water moderation, using the CCELL code. The calculation assumed a fuel element averaged fuel-rod-cell and resulted in a v'alue of k,of 1.434.
12.3 Single Package Evaluation 12.3.1 XN Types I. II, III, and IV Fuel Elements For the XN Type I, II, III and IV fuel elements described in Section 2, the reactivity of a single package is less than those computed for the fully flooded array of damaged packages. The reasons for such a decrease are:
1)
No fissile material will be interacting with the single package, and 2)
Two sides of each fuel element are separated from the water reflector by the 1/4 inch thick steel strongback rather than one side as assumed in the fully flooded array of damaged packages (see Figure 12.5 for geometrical details). Additionally, for the Type IV fuel elements, it was assumed that the ethafcam pads between the strongback and the fuel elements were totally cr.shed.
This assumption increases the reactiv'ty of the array by approxi-mately 2 percent.
Maximum reactivities for the single packages of fuel elements assuming full water moderation, reflection, and infinite fuel element length [~are shown in Table'12-IV-
12-8 XN-52, Rev. 1 June 1980 These values are all based on the 95 percent confidence level (k,77 average 1.96o), and were computed using the KENO-II Monte Carlo code with multi group data averaged by the CCELL code as previously described. These results demonstrate compliance with accepted criticality safety criteria.
12.3.2 Generically Characterized Fuel Elements To satisfy the requirement of 10 CFR 71.36(b), it must be shown that a single damaged package will be subcritical under conditions of full water reflection and optimum credible moderation.
The package was assumed to be fully flooded with water, and where applicable, the ethafoam (expanded polyethylene at approximately 6 pounds per cubic foot density) pads located between the fuel element and the strongback were (conservatively) assumed to be crushed. The resulting geometrical configuration is as shown in Figure 12.5.
Note that the results of these calculations are nonconservative when compared with those presented in Section 12.4.3 of this report which assumed an infinite array of damaged packages. This results because of the larger portien of the strongback considered here and the absence of surrounding regions of fissile material with which each package in the infinite array may interact.
However, the elimination of significant quantities of steel (i.e., the spacer blocks and other package structures) indicate that these calculations retain a fair degree of conservatism.
Since this config-uration is not limiting when compared with the require-ments for subcriticality of interacting arrays,.only a few cases were examined.
~
W
12-9 XN-52, R:sv.1
-June 1980 In these and all subsequent cases evaluating generically characterized fuel elements, unless otherwise noted, the fuel material was UO at 95 percent of theoretical density; 2
the clad was 0.020 inch thick zirconium; and the diametrical gas gap was 0.010 inch. For all KENO-II calculations, water cross sections and steel epithermal cross sections were averaged by the GAMTEC-II code, and the steel thermal group se'.f-shielded cross sections were calculated using the BRT-1 (Battelle-Revised THERM 05-1) code.
The results of KENO-II Monte Carlo calculations based on the geometrical arrangement shown in Figure 12.5 are summarized in Table 12-V.
12.3.3 XN Type AA Fuel Elements For the XN-Type AA fuel element described in Section 2, the reactivity of a single package is less than that computed for the fully flooded infinite array of damaged packages. The reasons for such a decrease are two-fold:
1)
No fissile material will be interacting with the single package; and 2)
Two sides of each fuel element will be separated from the water reflector by the 1/4 inch steel strongback rather than one side as assumed in the fully flooded array of damaged packages (see Figure 12.6 for geometrical details).
As a consequence, the reactivity of a single package when fully flooded and reflected by water is less than 0.900 m
.008 which was computed for tlie damaged package array. -
y 12-10 XN-52, Rev. 1 June 1980 12.4 Demonstration of Compliance With-10 CFR 71.38 and 71.40 12.4.1 Undamaged Fissile Class I Package Arrays l
l 12.4.1.1 XN Types I, II, and IV Fuel Elements Under normal conditions of transport, fuel elements contained within undamaged packages can be considered to be moderated only by the materials used for packaging.
(There is no leakage of water into the packaging during.
the water spray test; reference Section 9).
Specifically, some moderation of the fuel elements results from the addition of corrugated polyethylene shims within the fuel elements as previously described. These shims may be included in the packaging of XN fuel alement Types I through IV.
In addition, ethafoam (low density expanded polyethylene at approximately 6 pounds per cubic foot density) pads may be included around these fuel elements.
A summary description of the Fuel Types to be shipped as Fissile Class I packages (Types I, II, and IV) is given in Table 12-VI.
Since these fuel types may be shipped with or without the inclusion of polyethylene shipping shims, the analysis examined both the totally unmoderated and slightly moderated configurations.
For the unmoderated configurations, each fuel-bearing region was assumed to coatain the Pu0 -UO at the pellet density reduced by the 2
2 volume fraction of the oxide contained within the fuel element (see Table 12-VI).
This volume fraction was computed based on the volume of the fuel element surrounding the outside fuel pellets.
Reactivity calculations, however, assumed the oxide to be homogeneously spread throughout the maximum fuel efement size.
These ' assumptions
~
~
l.
12-11 XN-52, Rev. l' June 1980 i
result in an excess of. fissile material present within the package'of between 8 and 27 percent.
In addition to the unmoderated configuration, calculations were performed -
for the alternate configuration which includes the use of plastic shipping shims and ethafoam pads around the fuel ei t:ments.
A single cell of an infinite array of such undamaged t
packages would appear as shown in Figure 12.7.
To simplify the Monte Carlo calculations, conservative assumptions:
were made regarding the geometry of an array of such undamaged packages. The assumed geometric configuration is shown in Figure 12.8.
i i
The ethafoam region shown in Figure 12.8 is 0.75 inch thick between the fuel element and the steel strongback, and 0.50 inch thick elsewhere. When the ethafoam pads are not included in the packaging, the fuel element region is located 0.50 inch from the steel strongback (spacing is preserved by rubber-backed steel pads).
The carbon steel region is nominally. 0.125 inch thick on top and 0.375 inch thick on the sides and bottom. The water region was varied in thickness from 0 to 1 inch to determine the optimum thickness.
e i
Results of the infinite array calculations for XN Fuel Types I, II, and IV are shown in Table 12-VII. As can be seen, an. infinite, array of undamaged packages of Types I, II,.or.IV fuel elements is subcritical and thereby satisfies.the requirement of 10 CFR 71.38(a).
.\\
s
,o 12-12 XN-52, Rev. 1 June 1980 12.4.1.2 Generically Characterized Fuel Elements As previously noted for Fissile Class I shipments, an infinite array of undamaged packages, with optimum inter-spersed moderation, must be subcritical in any arrangement.
Under normal conditions of transport, fuel elements will be either unmoderated or slightly moderated by the inclusion of plastic shipping shims.
Consequently, both of the alternative packaging methods have been evaluated.
A.
Slightly Moderated Systems For fuel elements packaged with poly ^thylen'e shipping shims, the reactivity of each array was determined assuming a ifmiting effective water density within each fuel element. The effective water density is determined based on the total hydrogen content of the contained mass of polyethylene shims.
Typically, the effective water density is between 0.12 and 0.17 g/ca*.
To determine the fuel element size which may be transported as a Fissile Clasr I package, reactivities were computed u' sing the KENO-IV code for the conservative geometrical arrangement shown in Figure 12.8.
These calculations were performed for U-235 enrichments of 3.2, 3.5, and 4.0 wt. percent using the limiting fuel pellet diameter and water (void)-to-fuel volume ratios which result in the highest package reactivity.
~
3 Effective water densities of 0.15 and 0.20 g/cm were examined.
The results of those calculations assuming optimum interspersed moderation, are-given in Table 12-VIII. These data were, utilized to establish
~
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12-13 XN-52, R;v. 1 June 1980 a limiting fuel element size of 5.2 inches, at an enrichment of 3.5 wt.% 235-U, when the packaging includes plactic shipping silims within the fuel element and surrounding ethafoam pads. The limiting conditions for the shipment of fuel elempnts contain-ing corrugated polyethylene shipping shims as Fissile Class I packages are indicated as XN Type A fuel in Table 12-VIII while other configurations which have no fuel type identified are for comparison purposes to indicate the effect of variable parameters.
B.
Unmoderated Systems For fuel elements that are not packaged with corrugated polyethylene shipping shims (i.e., no moderating materials internal to the fuel element) analyses were performed with the KENO-IV code using the geometric arrangement shown in Figure 12.7 (ethafoam pads were not included adjacent to the fuel element).
For this particular evaluation, however, the fuel element size was fixed ac 8.55 in::hes square.
In the case of unmoderated fuel elements, the maximum reactivity of the array occurs at the maximum 002 density within the fuel element. Consequently, the following limitations were assumed:
I 1) 002 pellet density--100 percent of theoretical; 2)
Pellet diameter--0.5 inches (maximum); and 3)
Water-to-fuel volume ratio of the fuel assembly-
-1.3 (minimum).
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l.
12-14 June 1980 XN-52, R:v. 1 The results of these calculations are given in Table 12-IX. These data show that the maximum reactivity occurs when there is approximately 0.6 inches of water between adjacent packages.
(Note that there is no ethafoam included around the fuel elements thereby resulting in optimum conditions occurring when moderation is included external to the packages.)
As for Part A above, applying the criterion that k,ff + 3a 5 0.97, it is demonstrated that Type B fuel elements packaged in Model 51032-1 or -la containers meet the requirements for normal conditions of transport as Fissile Class I packages.
12.4.1.3 XN Type AA. Fuel Elements Under normal conditions of transport, XN Type AA fuel elements (see Table 12-X) contained within undamaged Model 51032-la packaging can be considered to be unmoderated.
The packaging method for XN-Type AA fuel elements does not include the use of ethafoam (low density ext,Ltded pclyethylene) pads around the fuel element or any materials interspersed within the fuel elements.
To simplify the Monte Carlo calculations, conservative assumptions were made regarding the geometry of an array of such undamaged packages. The assumed geometric configuratior. is shown in Figure 12.9.
With optimum interspersed moderation (0.55 inch) between the packages, the reactivity of an infinite array of Model 51032-la packages containing XN Type AA fuel elements was computed to be < 0.913 at the 95% statistical confidence level. This value was computed using the KENO-IV computer code with 123 group cross section data ~obcained (fom the NITAWL/ASDRNPM codes as described in S6ction 12.1.1.2.
(
+ -
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12-15 XN-52, Rev. 1 June 1980 12.4.2 Undamaged Fissile Clas-III Package Arrays 12.4.2.1 XN Type III Fuel Element For undamaged packages containing XN Type III fuel (Fissile Class III shipments), infinitely long and wide arrays of two-high packages were examined.
The geometric arrangement of fuel elements within each package was assumed to be as shown in Figure 12.10.
For these packages, however, each fuel eleme'nt was conservatively considered to be fully moderated by water and the two-high array of packages was reflected by an effectively infinite thickness of water (6 inches) on both the top and bottom.
The reactivity for this array, computed using the KENO-II code, was 0.530 +.014 for the M Type III fuel element.
These calculations conservatively' demonstrate that touch-ing identical shipments of undamaged packages containing XN Type III fuel would be subcritical when fully reflected on all sides by water.
12.4.2.2 Generically Characterized Fuel Elements For undamaged arrays of Fissile Class III packages of generically characterized UO fuel elements the geometric 2
arrangement and calculational methods summarized in Section 12.4.2.1 were used.
Restalts of the calculations for the various fuel element parameter limitations are given in Table 12-XI.
These results, when ccmpared to those presented in Section 12.4.3, show that Fissile Clas. III packaging limi.tations must be established on the basis of the damaged package arrays (i.e., arrays of damaged packages containing
12-16 XN-52, Rev. 1 June 1980 identical fuel elements are more reactive than the two undamaged shipments placed edge-to-edge and reflected by water).
12.4.3 Damaged Package Arrays 12.4.3.1 XN Types I, II, III, and IV Fuel Elements Packages have been subjected to a series of drop tests (see Sections 10 and 11) and supporting analyses were performed to ascertain the maximum package damage under hypothetical accident conditions.
The tests demonstrate tnat the minimum spacing between fuel elements in adjacent, stacked, damaged packages, is 8 inches. At least 3 inches is provided by the assembly clamps in each of the l
adjacent packages, anc a minimum of 2 inches is provided by the package stiffener rings.
(This results in a total minimum sep> ration of 8 inches, and assumes that the stiffener rings overlap and do not meet when stacking damaged packages.) Also, the separator blocks between
~
fuel elements within individual packages have been shown 4
to maintain a 6-inch separation between fuel elements.
The most reactive possible arrangement of fuel elements within four adjacent damaged packages is shown in Figure 12.11.
The assumed geometric arrangement of fuel elements used in the nuclear safety calculations is shown in Figure 12.6.
As can be seen, the effect of the containment vessel walls, and portions of the steel strongback, have J
been conservatively ignored. Also, the assumed geometric
~
configuration postulates both the minimum vertical and horizontal separations siaultaneously, a situation that is impossible to achieve :nder hypothetical accident conditions.
~
12-17 XN-52, R v. 1 June 1980 Reactivitifes calculated for the XN Types I and II mixed-oxide fuel 'lements in this assumed configuration, and for XN Type >.II and IV fuel elements in a similar configuration in which the one-half-inch spacing between the steel and fuel element regions has been eliminated, are given in Table 12-XII.
These results conservatively demonstrate compliance of XN Fuel Types I, II, III, and IV with the requirements of 10 CFR 71.40(b).
12.4.3.2 Generically Characterized Fuel Elements The specific standards for licensing of Fissile Class I packages includes the requirement that (see 10 C,FR 71.38(b))
250 damaged packages remain subcritical in any arrangement with optimum credible interspersed hydrogenous moderation when closely reflected by water.
Furthermore, for Fissile Class III packages a single shipment--when subjected to j
the effects of the hypothetical accident conditions as specified in 10 CFR 71, Appendix B, with optimum credible interspersed hydrogenous moderation and close-water reflection-- must remain subcritical.
Both of these requirernents are conservatively satisfied by consideration herein of an infinite array of damaged packages.
As discussed in Section 12.4.3.1, the most reactive possible configuration of damaged packages, as determined by drop test results, can be conservatively represented by the configuration shown in Figure 12.6.
As previously noted, the geometrical configuration in Figure 12.6 allows both the minimum vertical and horizontal separations simultaneously, a situation which cannot be achieved under hypothetical accident conditions.
Also, portions
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12-18 XN-52, Rev. 1 June 1980 of the steel.strongback are conservatively ignored. The packages were assumed to be fully flooded.
Homogenized cross sections were generated by the CCELL code (HRG/ THERMOS) for the fuel elements while the cross sections for water were generated by GAMTEC-II.
GAMTEC-II was also used to generate the epithermal cross section data for steel and the BRT-1 code (Battelle-Revised THERM 05-1) was used to generate the thermal group self-shielded cross sections for steel.
Results of the KENO-II Monte Carlo calculations, for the geometric configuration described are given in Table 12-XIII as a function of fuel element size for various combinations of U-235 enrichments and water-to-fuel volume ratios.
The criterion utilized to establish packaging limits for damaged package arrays is that k,ff of the array be $ 0.97 at the 99 percent statistical confidence level (i.e., k,ff(1verage) + 3o 5 0.97).
Using this criterion, data presented in Table 12-XIII were used to derive Fissile Class I and III package limits based on the reactivity of infinite arrays of damaged pacPages.
Limiting fuel element characteristics are summarized in. Table 12-XIV.
12.4.3.3 XN Type AA Fuel Elements The assumed ge1 metric arrangement of damaged fuel elements in the nuclear safety calculations is shown in Figure 12.6.
As for the generic fuel elements, the effect of the containment vessel walls and portions of the steel strongback have been conservatively ignored and.both the minimum vertical and horizontal separations are assumed to occur simultaneously.
12-19 XN-52, Rev. 1 June 1980 The reactivity calculated for an infinite array of fuel element packages in this assumed configuration is 0.900 t
.008.
This value was computed using the KENO-IV computer code with _123 group cross section data obtained as summarized in Section 12.1.1.2.
12.4.3.4 Shipments of Individual Rods Analyses presented in Section 12 demonstrate compliance of various generic UO fuel types under a variety of 2
limits which are not dependent on the metho1 of confining the fuel rod arrangement.
Since optimum interspersed moderation is assumed for all Class I shipments, and full moderation with water is assumed for all Class III shipments, the results of these evaluations are not affected by minor additions of materials between adjacent fuel element:
Consequently, it is requested that generic packaging limitations derived in Section 12 be applied to permit shipment of fuel rods in wooden boxes.
Such boxes would be constructed as indicated in Figure 12.12.
(Dimensions and packaging methods shown in Figure 12.12 are intended to be typical of those actually used.) Individual packaging limits on U-235 enrichment, rod diameters, assembly size, and water-to-fuel volume ratios and internal moderation would be the same as those established for generic U0 2 fuel elements.
In addition to permitting the shipment of fuel rods in wooden boxes, it is requested that a single fuel rod enriched to < 5 wt percent U-235 and having a UO2 pellet diameter of 5 0.5, be permitted within individual packages as shown in Figure 12.13.
As can be seen, the rod will n
12-20 XN-52, Rev. 1 June 1980 be wrapped in a protective material and enclosed within either a steel pipe or an angle iron protective cover.
If a pipe cover is used, it will be closed with threaded pipe caps at both ends to prevent rod escapement during normal and accident conditions of transport.
If-an angle iron is used, end plugs will be welded on each end.
Packaged as described above, the single fuel rod will be positioned on top of the clamps used to clamp fuel elements within the strongback.
Four (4) U bolts having a diameter of 1/4 inch will be used along the length of the rod to securely position the rod package on the strongback framework.
The addition of a single fuel rod, located and packaged as described above, is considered to have a negligibly small affect on the reactivity of the package under both normal and accident conditions of transport.
Additionally, the total weight of a loaded package will be limited to the licensed maximum gross weight of 7,400 pounds and 8300 pounds for the Model 51032-1 and -la packages, respectively. Hence, the addition of this single rod does not alter the previous evaluation of the package performance under hypothetical accident conditions.
12.5 Summary The results of criticality safety evaluations reported
~
herein demonstrate that the XN Types I, II, and IV mixed-oxide fuel element packages using Model 51032-1 shipping containers satisfy the requirements for Fissile Class I packages.
1he rr.ilt$ also demonstrate t$at X5
12-21 XN-52, Rev. 1 June 1980 Type III mixed-oxide fuel elements contained in Model i
51032-1 packaging satisfy the requirements for Fissile
~
Class III packages.
Table 12-XIV contains a summary of the limits on fuel element parametirs determined for both Fissile Class I and Fissile Class III shipments of generically character-ized low-enriched uranium fuel elements transported in either Model 51032-1 or 51032-la packages.
The criterion which was applied to determine.these limits was that k,ff
+ 3a $.970 for both normal and accident (damaged package) conditions.
Fuel element parameter limits were determined by using the criterion which imposed the more cqnserva-tive limitations. A summary of the reactivities computed or interpolated for the various fuel types under both normal and accident conditions is given in Table 12-XV.
i l
?
~
12-22 XN-52, Rev. 1 June 1980 TABLE 12-I CALCULATED k_ FOR UNMODeRATED 5 WT %
U-235 ENRICHED 002 Calculational Method
=
JERBEL 0.835 3
HAMMER 0.80 4
GAMTEC-II 0.855 6
4 e
i
12-23 XN-52, Rev. 1 June 1980 TABLE 12-II THEORY-EXPERIMENT CORRELATIONS Exp'tl.
Results NITAWL/SXDRN/
Moderator-Cylind.
CCELL-DTF-IV CCELL-KENO KENO-IV to-Fuel Core Calculated Calculated Calculated Exp't.
Volume Radius Reagtivity Reactivities Reactivities No.
Ratio em eff
__{k,f7to) k_f7ta) 1 1.405 26.820 1.016 1.008 1.006 1.007 i.005 2
1.853 24.294 1.015 1.014 i.005 1.013 2.005 3
3.357 23.600 1.011 1.003 i.005 1.008 i.004 4
4.078 24.771 1.009 1.010 t.005 1.002 2.004 5
5 4.984 27.172 1.005 1.005 i.005 1.013 *.004 e
f 4
-.,--n.
,* 1
^
12-24
.XN-52 Rev. 1 June 1980 TABLE 12-III COMPARISON OF COMPUTED INFINITE MEDIA MULTIPLICATION FACTORS II)
XN Type JERBEL/ JOT JERBEL CCELL/ JOT CCELL I
1.320 1.323 1.262(2) 1.361 II III 1.26(2) 1.358 IV 1.286 (1) The CCELL code is a combination of the HRG and THERMOS codes.
(2) Calculations include consideration of Gd 02 3 poison rods contained in the fuel element.
e e
i l
k 12-25 XN-52, Rev. 1 4
TABLE 12-IV k
MIXED OXIDE FUEL ELEMENT SINGLE PACKAGE EVALUATION
\\
j XN Type Single Package off D
f' I
0.659 II 0.783 4
III 0.784 IV 0.581 i
e f
E.
T a
. f.'
12-26 XN-52,.Rev. 1 j
June 1980
- TABLE 12-V SINGLE DAMAGE 0 PACKAGE EVALUATION
't Enrichment (wt % U-235 Pellet Diameter Y/Yf Assembly Size k,ff 6
3.5
.50" 2.10 8.50
.923 i.010
~ 4. 0
.50" 2.10 8.50
.927 i.009 8
5.0
.50" 2.10 8.25
.959 i.0115 i
6 a
J
)
)
i e
~
m y
--e_m
' '~~-
- ~ ~...
.,o TABLE 12-VI FUEL ELEMENT DESCRIPTION Outside Maximum Nominal Pellet-to-Volume Assembly Lattice Pellet Rod Pellet Fraction of XM Fuel Dimension Pitch Dimension Number Diameter Diameter Contained Types (inches)
(inches)
(inches) of Rods (inches).
(inches)
Oxide I
5.50x5.50 0.738 4.906 49 0.570 0.478 0.36533 II 6.62x6.62 0.577 6.141 121 0.449 0.371 0.34685 g
b IV 4.60x4.60 0.700 3.981 36 0.563 0.481 0.4128 1
z, see J
~
" ' ~ -
"~p.
TABLE 12-VII REACTIVITY OF UNDAMAGE0' FISSILE CLASS I PACKAGE ARRAYS k of Fuel Element k of Fuel Element PIckagesContaining k of Fuel Element PIckages Containing Polyethylene Shins, Revised Water Density ETement Packages With Polyethylene Shims &
Ethafoan Pads and XN Fuel Within Element Optimum Interspersed Optimum Interspersed Optiumum Interspersed
~
8 Type (a/cm )
Moderation Moderation Moderation 1
0.1622 0.702 1 005 (a)
(a)
II 0.1670 0.883 1 006 0.936 1 005 (b) 0.955 1 005 (c)
IV 0.1624 0.733 1 006 (c) cm (a) Not computed; by inspection of the relative sizes and enrichment of the fuel elements, it can be seen thpt the reactivity of this M0 fuel element will be less than the value obtained for the Type IV element.
(b) Optimum interspersed moderation occurs when there is 0.25 inches of water between packages.
(c) Optimum interspersed moderation occurs when there is no moderation between packages.
oE 5h i
$$3 and
l
~
~
TABLE 12-VIII l
l:
UNDAMAGED ARRAYS OF BWR SIZED FUEL ELEMENTS Fuel XN(j)
Average Pellet Effective Element KENO-IV Type Enrichment Diameter y/y(2)
Water Density Size Calculated Generic (w/o 235-0)
(inches)
F (g Hy3/cm )
(inches)
Reactivity 3
A 3.5 5.50 1 2.1 5 0.20 5.20 0.948 i.004 3.2 5 0.50
$ 2.1
$ 0.20 5.20 0.913 i.005 4.0
$ 0.50 5 1.8 5 0.20 5.20 0.975 i.004 y
O!
4.0
$ 0.50
$ 1.8
$ 0.15 5.20 0.962 i.004
'l 4
(1) Where no fuel type is designated, calculations are to indicate variations in reactivity with changes in parameters.
Parameters indicated for Type A fuel indicate limits for fuel elements packaged with polyethylene shipping shims with the fuel elements, p5
- s m (2) Ratio of available moderator volume to assumed fuel volume within the
- f*
fuel elements.
yg
?
~ c
~-
TABLE 12-IX UNDAMAGED ARRAY--UNM00ERATED FUEL ELEMENTS (XN TYPE B FUEL)
Enrichment Pellet Diameter Assembly Size Interspersed H O y !y.
2 k
(wt % U-235)
(inch) w f (inches)
(inch) eff
.3.5
.50
> 1.3 8.55 0.0 0.652 i.003 3.5
.50
> 1.3 8.55 0.25 0.898 i.004 3.5
.50
> 1.3 8.55 0.50 0.948 2.005 3.5
.50
> 1.3 8.55 0.75 0.951 i.005 3.y
.50
> 1.3 8.55 1.00 0.919 i.066 1
1
- Represents void fraction of fuel element rather than moderator content.
E G
b
. o<
emed 4
L
~
- ?:.
O%-
TABLE 12-X FUEL ASSEMBLY DESCRIPTION t
Outside Maximum Nominal Pellet-to-Volume Assembly Lattice Pellet Rod Pellet Fraction of Type Dimension Pitch Dimension Number Diameter Diameter Contained Fuel (inches)
(inches)
(inches) of Rods (inches)
(inches)
Oxide AA 9.01 x 9.01 0.563 8.802 236 0.424 0.357 0.2911
'?S y
e
!b o <
I emme e
=
Y 12-32 XN-52, Rev. 1 f-June 1980 TABLE 12-XI l
TW UNDAMAGEO SHIPMENTS (Class III Requirements)
I
'~
Enrichment Pellet Assembly y jy k
(wt % U-235)
Diameter w f Size off 3.5
.50 2.1
'8.75 0.909 1 010 4.0
.50 2.1 8.75 0.947 i.008 4.0
.40 2.3 8.25 0.893 i.008
[
5.0
.50 2.1 8.25 0.932 i.0085
^
i 5.0
.50 2.1 8.50 0.947 i.0095 1
5.0
.50 2.1 8.75 0.978 i.0095
- em J
4 y
-_e
12-33 XN-52. Rev. 1 i
l June 1980 TABLE 12-XII i
MIXED OXIDE FUELS - DAMAGED PACKAGE ARRAYS XN Type Array Reactivity t
I.
0.634 1 013 II 0.765 1 009 III*
0.762 1 011 1
IV*
0.557 1 012
- These cases assumed a slightly more conservative geometry than that shown in Figure 12.6, in which the spacing between the steel region (Region III) and the Fuel Assembly region (Region I) has been eliminated. Note that in all actual cases, this spacing is assured by the presence of rubber-backed steel pads on the sides of the strongback.
t
+
g i
i 1
12-34 XN-52. Rev. 1 l
June 1980 TABLE 12-XIII 00 FUEL ELEMENT - DAMAGE 0 PACKAGE ARRAYS g
(Fissile Class I and III) i Pellet Assembly f'
Enrichment Diameter Size
.I (wt % U-235)
-(inches) w! f (inches) eff 3.5
.50 2.1 7.75 0.858 i.007 3.5
.50 2.1 8.0 0.891 i.007 3.5
.50 2.1 8.25 0.895 i.0085 I
3.5
.50 2.1 8.5 0.930 1.008 3.5
.50 2.1 8.75 0.946 i.008 4.0
.50 1.8 8.25 0.928 i.008 4.0
.50 1.8 8.50 0.936 i.008 4.0
.50 1.8 8.75 0.951 i.008 4.0
.50 2.1 8.0 0.922 t.0075 4.0
.50 2.1 8.25 0.935 i.008 4.0
.50 2.1 8.5 0.938 i.008 4.0
.50 2.1 8.75 0.959 i.006 4.0
.40 2.3 8.0 0.922 i.0075 4.0
.40-2.3 8.25 0.933 i.009 4.0
.40 2.3
- 8. 5 0.948 i.009 5.0
.50 2.1 7.75 0.927 i.0095 5.0
.50 2.1 8.0 0.940
.0085 5.0
.50 2.1 8.25 0.961 i.0065
=~
=.-
t
. TABLE 12-XIV
SUMMARY
OF MODEL 51032-1 AND -la PACKAGING LIMITS Average Polyethylene Effective Water Fuel XN Nominal 00 Pellet Fuel Element Shipping Shims Density of Element 2
Fuel Enrichment Diameter Water-to-Fuel Included In Contained Shipping Fissile Size Type (wt.% 235-U1 (inches)
Volume Ratio Fuel Element Shims (g/cm )
Class (inches) 3 A
5 3.5 5 0.50 5 2.1 Yes
$ 0.2 I
$ 5.20 1 2.1 No Not Applicable I
$ 8.55 j$
8 5 3.5 5 0.50 1.3 i V,/V7 IN C
1 4.0
$ 0.50 1 1.8 Yes Not Limited III 1 8.60 1
D 5 4.0 5 0.50 1 2.1 Yes Not Limited III
$ 8.48 E
5 4.0 5 0.40 5 2.3 Yes Not Limited III
$ 8.40 F
5 5.0
$ 0.50 5 2.1 Yes Not Limited III
$ 8.00 c, g -
h" AA 1 3.3 0.3565 0.563 inch No Not Applicable I
$ 9.01 gj$
square pitch c3 j
-.a 1
4 e
' ~~~
~
.. ~,.,
TABLE 12-XV
SUMMARY
OF EOMPUTED REACTIVITIES FOR XN FUEL TYPES Single Undamaged Damaged Fuel Fissile Package Reference Package Reference Package Reference
- Type, Class Reactivity Section Array Reactivity Section Array Reactivity Section I
I
< 0.659 12.3.1
< 0.955 +.005 12.4.1.1 0.634 +.013 12.4.3.1 II I
< 0.783 12.3.1 0.955 +.005 12.4.1.1 0.765 1 009 12.4.3.1 III III
< 0.784 12.3.I 0.530 1 014 12.4.2.1 0.762 +.011 12.4.3.1 IV I
< 0.581 12.3.1 0.733 +.006 12.4.1.1 0.557 +.012 12.4.3.4 A
I-
< 0.858 +.007 12.3.2 &
0.948 +.004 12.4.1.2(A) < 0.858 +.007 12.4.3.2
~
Table 12-XIII
~
B I
0.930 +.008 12.3.2 &
0.950 +.005 12.4.1.2(B) 0.930 +.008 12.4.3.2
~
Table 12-XIII C
III 0.936 +.008 12.3.2 &
< 0.947 +.008 12.4.2.2
< 0.936 +.008 12.4.3.2 a
~
Table 12-IIII
~
~
7 w
D III
< 0.938 +.008 12.3.2 &
< 0.947 +.008 12.4.2.2 0.938 +.008 12.4.3.2
~
Table 12-XIII
~
~
.I
~
Table 12-XIII
~.008 12.4.2.2 0.942 +.009 12.4.3.2 E
III
< 0.938 e.008 12.3.2 &
< 0.947 a:
~
F III
< 0.940 +.009 12.3.2 &
< 0.932 +.009 12.4.2.2 0.940 +.009 12.4.3.2
~
Table 12-IIII AA I
< 0.900 +.008 12.3.3
< 0.913 12.4.1.3 0.000 +.008 12.4.3.3 C 3 k ~ biP
$ <F e
M s
6 k
s
- 8 he s
12-37 I
-52, Rev. 1 FIGURE 12.1 Infinite Media Multiplication Factors for Low Enriched 0.5 inch UO R June 1980 aFunctionoftheWaterkoodsinWateras Fuel Volume Ratio i
l l
~
-~
1.50 8
1 5.0 wt.% 235-U k,
1.45 i
4.0 wt.% 235-U 1.40 i
~~
3.0 wt.% 235-U
~
i 1
i l.5 VgVf 2.0 ey
12-38
~
XN-52, Rev. 1 j
FIGU"E 12.2 5.0 wt.% U-235 Enriched U0,, Rod-Water June 1980 Lattice Infinite Media Multipitcat.fois Factors I
i i
i i
i 1.52 f
I i
1,51 I
t.
i VgVf = 2.1 1.49 VgVf = 1.8 1.48 i
i e
e i
i
.30
.40
.50 Pellet Diameter (inch)
T m
12-39 XN-52, Rev. 1
{
FIGURE 12.3 4.0 wt.% U-235 Enriched UO Rod-Water une 1980.
Lattice Infinite Media Mul$1 plication Factors f
I I
I I
I I
l 1.48 r
1.47 4
vjVr 2.3 1.46 VgVf 2.1 VgVf = 2.0 1.45
'~
VgVf = 1.8
~
1.44 I
i i
i I
.30
.40
.50 Pellet Diameter (inch) 12-47
4
//
12-40 XN-52, Rev. 1 FIGURE 12.4
-3.0 wt.% U-235 Enriched UO Rod-Water Lattice June 1980 7
f Infinite Media Multiplication Factors I
i i
i e
i 1.41 t
u 1.40 t
VgVf=
.3
- v.._
1.39 VgVf = 2.1 1.38 VgVf = 1.8 i
\\
\\
1.37
.30
.40
.50 Pellet Diameter 12-48
l 12-41 XN-52, Rev. 1 June 1980 e
i l'
6" II
[
7 III 6 " -.
I I
~ 6"
{
24.875*-
6" Region Number Description I
Fuel Element II Water III Steel 4
- Dimension given is for the Model 51032-1 container strongback.
The Model 51032-la container strongback dimension is one (1) inch larger.
FIGURE 12.5 Single Package,1amaged, Fully Flooded l
^
3 3
m
^
,,.r.
i l
12-42 XN-52, Rev. 1 June 1980 A = 1.0 a
f II 6"
l a
+ 6"*-.
3" I
7 A=1.0
- I i
8" a
,_ A = 1.0 3"
1 I
t t
6" III f
~
A = 1.0 Region Number Description _
I Fuel Element II Water III Steel (0.25")
- A separation of 6' inches is assured in the Model.51032-1 container and 5 inches is assured in the Model 51032-la container.
FIGURE 12.6 Assumed Configuration of Damaged Package Arrays
12-43 XN-52, Rev. 1 June 1980 '
V IV r
b II
' IV 4
I I
i III f
Region Material Description I
Fuel Element II Ethafoag (expanded polyethylene at 6 lb/ft density) pads or void III Void Region IV Car'mn Steel Strongback and Outer Containment Vessel V
Optimum Interspersed Water Moderation FIGURE 12.7 Single Cell of Infinite Array of Undamaged Packages
~
y.
,1 12-44 XN-52, Rev. 1 June 1980 d
z'**
+
II g
V 12.5" o
IV
/
a 12" y**
I I
III a
x*
- y --- -
o u
'I
[
24.875" r-25.625" r
r Region Description I
Unmoderated or Slightly Moderated Fuel Element II Ethafaam or Void l
III Void IV Carbon Steel V
Water i
Dimensions variable to provide optimum thickness of interspersed hydrogeneous moderation.
Dimensions x, y, and z are variable and are dependent on the cross sectional dimension (y) of the fuel element.
FIGURE 12.8 Assumed Geometrical Configuration of Undamaged Packages (Model 51032-1)
s.
5
.12-45 XN-52, Rev. 1 June 1980
~~III a
2.992
II i
.-_IV
- 12.5" 12.0" 9.008" I
- 7.86" o
P 25.875" 26.625" r
Dimensions variable to provide optimum thickness of interspersed hydrogeneous moderation.
See region description and dimensions below:
Atomic Densities Region Material (atoms / barn-cm)
I Fuel Element (see Table XII II Void III Carbon Steel 0.08479 IV Optimum Thickness 0.033368 Water
=--
Assumed Geometrical Configuration FIGURE 12.9 of Undamaged Packages (Mo' del 51032-la with Type AA Fuel Elements)
. ;t.
(
12-46 f
XN-52, Rev. 1 June 1980 4
V 6"
4 a
I I
I 38" 4
4---
25.375" A=1.0 *
- A = 1.0 IY II I
I 38" III 13.813"
'l V
6' s
Region Number Description I
Fuel Element II Ethafoam or Void III Steel Strongback IV Void V
Water Reflector FIGURE 12.10 Undamaged Package Arrays
y, e.
f' 12-47 XN-52,'Rev. 1 i
June 1980 e
I t
II II l-J I
I I
I 4
s J
[
I I
II I
I II I
I ll IV II Region Number Description I
Fuel Element II Water III Steel Strongback IV Steel Containment Vessel Wall FIGURE 12.11 Array Jf %inaged. Packages 1
l
~
y.,
13-1 XN-52, Rev. 1
,l June 1980 eI
13.0 REFERENCES
1.
CONF-710801 (Volume 2) Health and Safety (TID-4500),
" Proceedings, Third International Symposium, Packaging, and Transportation of Radioactive Materials", August 1971, pp 873-885.
2.
Exhibit P " Application for Licensing of Combustion Engineering, Inc., Shipping Container Model 927A",
July 3,1969, License SNM-1067, Docket 70-1100.
6 3.
Exhibit P (including Appendi." P-1), " Application for Licensing of Combustion Engi.1eering, Inc., Shipping Containers Models 927B and 527C", February,23,1971,
{
License SNM-1067, Docket 70-1100.
4.
V. O. Uotinen, G. L. Gelhause, U. P. Jenquin and C. R. Gordon, " Calculations ~of Power Distribution and Reactivities", Reactor Physics Quarterly Report, April, May and June, 1970, BNWL-1381-2, August 1970.
5.
Nuclear Safety Guide, TID-7016 Revision 1,1971.
6.
L. E. Hansen, et al., " Critical Parameters of Plutonium Systems. Part I: Analysis of Experiments", Nuclear Applications, Volume 6, 381-390, April, 1969.
7.
XN-NF-499, " Criticality Safety Benchmark Calculation for Low-Enriched Uranium Metal and Uranium 0xide Rod-Water Lattices", April 1979.
8.
P. W. Davison, et al., " Yankee Critical Experiments -
Measurements on Lattices of Stainless Steel Clad Slightly Enriched Uranium Dioxide Fuel Rods in Light Water",
YAEC-94, April 1, 1959.
9.
V. E. Grob., et. al., Results of Critical Experi ients in Loose Lattices of UO Rods in Water, WCAP-1412, 2
March 30,1960.
(
l L
l l
i j
16596 l