ML19211A495
| ML19211A495 | |
| Person / Time | |
|---|---|
| Site: | 07106581 |
| Issue date: | 11/30/1979 |
| From: | Ehlers R, Hansen L, Sofer G SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER |
| To: | |
| Shared Package | |
| ML19211A494 | List: |
| References | |
| 14897, XN-052, XN-052-R01, XN-52-R1, NUDOCS 7912200011 | |
| Download: ML19211A495 (90) | |
Text
XN-52, Rev. 1 CONSOLIDATED LICENSE APPLICATION FOR EXXON NUCLEAR COMPANY, INC.
MODEL 51032-1 AND -la SHIPPING CONTAINERS Certificate of Compliance 6581 Docket 71-6581 1631 137 November, 1979 8
7 912200 OI/
XN-52, Rev. 1 ii CONSOLIDATED LICENSE APPLICATION FOR EXXON NUCLEAR COMPANY, INC.
MODEL 51032-1 AND -la SHIPPING CONTAINERS Certificate of Compliance 6581 Docket 71-6581 Prepared by:
Ibb
///.76/77 L. E. Hansen Accepted by:
-M
=
R. J. EhJhrs, Manager Purchasing and Logistics Concurred by:
(d
- -h-7#/
G. A. Sof.e anagqY Nuclear uels Engineering Approved by:
r R. N Kso'n, pedager' Corporate Licensing and Compliance 1631 138 e-
XN-52, Rev. 1 iii Distribution Copy
- l-8 NRC, Transportation Branch
- 9 W. J. Cooley, USNRC, Region V 10 L. P. Bupp 11 W. T. England 12 R. K. Robinson
- 13 R. Nilson 14 E. R. Astley
- 15-16 L. E. Hansen 17 H. P. Estey 18 C. O. Brown 19 R. H. Schutt 20 G. Dressler (Lingen) 21 T. L. Davis 22 D. L. Cornell 23 R. J. Ehlers 24 J. 8. Edgar 25 W. E. Niemuth (Lingen) 26 D. K. Hauptfleisch (Lingen) 27 H. Wessling (Lingen) 28 P. Kruys/F. deWaegh (Brussels) 29 J. W. Long (Lingen) 30 J. A. Perry 31 R. B. Stephenson "32 G. A. Sofer
- 33 P. G. Sarafian 34 J. F. Patterson 35 Document Control (Lingen) 36-40 Document Control (4 extras)
- Indicates limited distribution pending completion of review by the US NRC.
b f
f39
XH-52, Rev. 1 iv CONSOLIDATED LICENSE APPLICATION FOR EXXON NUCLEAR COMPANY, INC.
MODEL 51032-1 AND -la SHIPPING CONTAINERS Certificate of Compliance 6581 Docket 71-6581 TABLE OF CONTENTS SECTION TITLE PAGE NO.
1-1 1.0 INTENT..
2-1 2.0 PACKAGE DESCRIPTION.
2.1 Model 51032-1 Container..........
2-1 2-1 2.1.1 Container Description.
2.1.2 Fuel Element Clamps, Shock Mounts & Separator Blocks 2-5 2-8
- 2. 2 Model 51032-la Container.
2-8 2.2.1 Container Description.
2.2.2 Fuel Element Clamps, Shock Mounts & Separator Blocks 2-9 2.3 Both Models 51032-1 and 51032-la Containers....
2-11 2-11 2.4 Package Contents..........
2-11 2.4.1 Model 51032-1 Container....
2-13 2.4.2 Model 51032-la Container.......
3-1 3.0 PACKAGE HANDLING.
3-1 3.1 Package Loading.
3.2 Transport Controls............
3-6 3-7 3.3 Unloading..
4-1 4.0 PROCEDURAL CONTROLS.
5-1 5.0 GENERAL STANDARDS FOR PACKAGING 6.0 STRUCTURAL STANDARDS FOR LARGE QUANTITY PACKAGING.
6-1 7.0 CRITICALITY STANDARDS FOR FISSILE MATERIAL PACKAGES 7-1 8-1 8.0 EVALUATION OF A SINGLE PACKAGE......
1631 140
XN-52, Rev. 1 v
TABLE OF CONTENTS (Continued)
SECTION TITLE PAGE NO.
9-1 9.0 STANDARDS FOR NORMAL CONDITIONS OF TRANSPORT.
10.0 STANDARDS FOR HYFOTHETICAL ACCIDENT CONDITIONS...
10-1 10.1 Modes 5103;-1 Package.
10-2 10-2 10.1.1 Free Dror tests.....
10.1.1.1 Summary of Model 51032-1 Drop Tests.
10-2 10.1.2 Package Component Tests and Evaluations...
10-5 10.1.2.1 Model 51032-1 Separator Blo.k 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 Evaluation.
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 11-1 11.0 EVALUATION OF AN ARRAY OF PACKAGES..
12.0 SPECIFIC STANDARDS FOR FISSILE CLASS I AND III 12-1 PACKAGES........
12.1 Method, Discussion, and Verification.
12-2 12.1.1 XN Type I Fuel Elements......
12-2 12.1.2 XN Type II Fuel Elements.
12-3 12.1.3 XN Type III, IV, V, VI, AA and Generically 12-4 Characterized Fuel Elements........
12.1.3.1 KEN 0 II (18 Energy Group) Calculational Method.
12-4 12.1.3.2 KENO IV (123 Energy Group) Calculational Method..
12-5 12.2 Results of k, Calculations.
12-7 1631 14i
XN-52, Rev. 1 m
vi TABLE OF CONTENTS (Continued)
SECTIOM TITLE PAGE NO.
12.2.1 XN Types I and II Fuel Elements..........
12-7 12.2.2 XN Types III, IV, V, and VI Fuel Elements...
12-7 12.2.3 Generically Characterized Fuel Elements.
12-9 12.2.4 XN Type AA Fuel Elements..........
12-10 12.3 Single Package Evaluation.
12-10 12.3.1 XN Type I and II Fuel Elements.....
12-10 12.3.2 XN Types III, IV, V, and VI Fuel Elements...
12-11 12.3.3 Generically Characterized Fuel Elements.
12-12 12.3.4 XN Type AA Fuel Elements......
12-13 12.4 Demonstration of Compliance with 10 CFR 71.38 and 71.40..
12-14 12.4.1 Undamaged Fissile Class I Package Arrays.
12-14 12-14 12.4.1.1 XN Types III, IV, and VI Fuel Elements.......
12.4.1.2 Generically Characterized Fuel Elements.
12-16 12.4.1.3 XN Type AA Fuel Elements.....
12-18 12.4.2 Undamaged Fissile Class III Package Arrays..
12-19 12-19 12.4.2.1 XN Type I Fuel Element.
12-19 12.4.2.2 XN Type II Fuel Element.
12.4.2.3 XN Type V Fuel Element.
12-21 12.4.2.4 Generically Characterized Fuel Elements.
12-21 12-22 12.4.3 Damaged Package Arrays...
12.4.3.1 XN Types I and II Fuel Elements.
12-22 12.4.3.2 XN Types III, IV, V, and VI Fuel Elements..
12-22 12.4.3.3 Generically Characterized Fuel Elements.
12-23 12-25 12.4.3.4 XN Type AA Fuel Elements.
12.4.3.5 Shipments of Individual Rods.
12-25 12-27 12.5 Summary.
13-1
13.0 REFERENCES
1631 142
XN-52, Rev. 1 vii LIST OF APPENDICES APPENDIX NO.
TITLE APPENDIX I APFLIED DESIGN COMPANY, INC. LIFT EYE ANALYSIS APPENDIX II STRUCTURAL ANALYSIS OF MODEL 51032-1 PACKAGING TIE-DOWN SYSTEM APPENDIX III APPLIED DESIGN COMPANY, INC. REPORT 2526A APPENDIX.V 30-F00T DROP TEST PROCEDURE AND REPORT PACKAGING MODEL 51032-1 APPENDIX V PACKAGE COMPONENT EVALUATIONS APPENDIX VI FUEL R0D DROP TEST REPORT 1631 143
XN-52, Rev. 1 viii LIST OF TABLES TABLE NO.
TITLE PAGE NO.
2-1 Revised Fuel Element Identification Numbers 2-14 2-II Radioactive Material Limits (Mixed Oxide 2-15 Fuels)...................
2-III Radioactive Material Limits (UO Fuels).
2-16 2
2-IV Fissile Material Limits (Mixed 0xide Fuels).
2-17 2-V Fissile Material Limits (U0 Fuels).
2-18 2
2-VI XN-Type I................
2-19 2-19 2-VII XN-Type II......
2-VIII XN-Type III............
2-19 2-IX XN-Type IV.
2-20 2-X XN-Type V.
2-20 2-XI XN-Type VI...........
2-20 2-XII Limiting Fuel Element Physical Characteristics.
2-21 7-I Individual Package Reactivities.
7-2 10-I Energy Dissipation Accounting for Model 51032-la Packages Containing XN-Type AA Fuel Elements Relative to Drop-Tested 10-17 Package.
12-I THERM 05/HRG/2DB Benchmark Calculations.
12-28 12-II Comparison of Measured Criticals and Calculated Multiplication Factors Using JERBEL.
12-29 12-III Calculated k for Unmoderated 5 wt% U-235 12-30 Enriched UO2 12-IV Theory-Experiment Correlations..
12-31 1631 144
XN-52, Rev. 1 ix LIST OF TABLES (Continued)
TABLE NO.
TITLE PAGE NO.
12-V Comparison of Computed Infinite Media 12-32 Multiplication Factors..
12-VI Mixed Oxide Fuel Element Single Package 12-33 Evaluation...............
12-34 12-VII Single Damaged Package Evaluation......
12-35 12-VIII Fuel Element Description.
12-IX Reactivity of Undamaged Fissile Class I Package Arrays........
12-36 12-X Undamaged Arrays of BWR Sized Fuel Elements.
12-37 12-XI Undamaged Array--Unmoderated Fuel Elements.
12-38 12-XII Fuel Assembly Description.
12-39 12-XIII Two Undamaged Shipments...........
12-40 12-XIV Mixed 0xide Fuels - Damaged Package Arrays.
12-41 12-XV UO Fuel Element - Damaged Package Arrays..
12-42 2
12-XVI Summary of Model 51032-1 and -la Packaging Limtis 12-43 12-XVII Summary of Computed Reactivities for 12-44 XN Fuel Types.
1631 145
XN-52, Rev. 1
~
x LIST OF FIGURES 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 2-26
- 2. 5 Standard Model 51032-1 Strongback.
2.6 Four Fuel Element Packaging (Isometric View).
2-27 2.7 Short Strongback.
2-28 2.8 Instrumented Square Fuel Element 2-29 Shipping Arrangement.
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 2.13 BWR Fuel Packaging Arrangement -
2-34 Model 51032-1..
2.14 Model 51032-la Container General Arrangement.
2-35 2.15 Model 51032-la Container Strongback.
2-36 2-37 2.16 Model 51032-la Separator Block.
2.17 Model 51032-la PWR Fuel Element Clamp 2-38 Assembly..
1631 146
XN-52, Rev. 1 xi LIST OF FIGURES (Continued) 6 FIGURE NO.
TITLE PAGE NO.
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 3.1 Package Tie Down System.
3-10 3-11 3.2 Shipping Record Sheet
- 3. 3 Department of Energy Regional Coordinating Offices for Radiological Assistance and Geographical Areas of Responsibility.
3-12 3.4 Radioactive Material Shipping Inspection 3-13 Record.
10.1 Steel and Aluminum Clamp Assembly Force Deflection Curve Comparison.
10-19 12.1 Infinite Media Multiplication Factors for a Function of the Water-$o ods in Water as Low Enriched 0.5 inch U0 R Fuel Volume 12-45 Ratio.
12.2 5.0 wt.% U-235 Enriched UO Rod-Water LatticeInfiniteMediaMul$iplication 12-46 Factors.......
12.3 4.0 wt.% U-235 Enriched UO Rod-Water LatticeInfiniteMediaMul$1 plication Factors..............
12-47 12.4 3.0 wt.% U-235 Enriched UO Rad-Water LatticeInfiniteMediaMul$iplication 12-48 Factors.
1631 i47
XN-52, Rev. 1 xii LIST OF FIGURES (Continued)
FIGURE NO.
TITLE PAGE NO.
12.5 Single Package, Damaged, Fully Flooded.
12-49 12.6 Assumed Configuration of Dams,ed Package Arrays.......
12-50 12.7 Single Cell of Infinite Array of Undamaged Packages............
12-51 12.8 Assumed Gecmetrical Configuration of Undamaged Packages (Model 51032-1).
12-52 12.9 Assumed Geometrical Configuration of Undamaged Packages (Model 51032-la with Type AA Fuel Elements).
12-53 12.10 Assumed Configuration for Evaluating XN Type II Fuel Element Shipments.
12-54 12.11 Undamaged Package Arrays...
12-55 12.12 Array of Damaged Packages.........
12-56 12.13 Fuel Rod Packaging (Typical).
12-57 12.14 Individual Fuel Rod Packaging.
12-58 b
f f48
2-1 XN-52, Rev. 1 2.0 PACKAGE DESCRIPTION As specified in 10 CFR 71.22, the the Model 51032-1 and
-la packages and their respective contents are described herein.
2.1 Model 51032-1 Container The gross weight of the Model 51032-1 packaging is 4000 t 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 ll 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-2B steel nuts tightly seated complete the closure.
1631 149
2-2 XN-52, Rev. 1 Seven steel stiffening rings (five rollover angles and two end rings) are welded to each of the base and cover assemblies to strengthen the containment vessel shell.
Rollover rings are fabricated 2 1/2 x 2 1/2 x 5/16-inch angles and end rings are fabricated of 3 1/2 x 2 1/2 x 3/8-inch angles.
Four 7 gauge (0.1793 inch) steel skids are welded to the base assembly. These skids support the package and are designed to permit bolting the stacking brackets when packages are stacked for storage or transport.
Stacked packages, however, are not normally bolted together during transport.
Four sets (two per set) of stacking brackets fabricated of 7 gauge (0.1793-inch) steel are welded to the cover assembly.
Welded to each set of stacking brackets is a steel lift-ing lug.
These lugs are fabricated cf 3/8 inch steel and may be used to support the loaded package.
Use has been shown not to generate stress in any material of the packaging in excess of its yield strength with a minimum safety factor of 3.4.
Two fork lift pickup channels are welded to the base assembly to facilitate package handling.
These channels are fabricated of 1/4 inch steel.
Fourteen (seven per side) shock-mount support brackets fabricated of 1/4 inch steel are welded to the interior side of the base assembly shell. The weight of the fuel elements and the related support mechanism is transferred 1631 150
2-3 XN-52, Rev. 1 to these brackets through up to 14 shock mounts.
(The actual number of shock mounts included in each package is dependent upon the weight of the fuel elements being transported.)
The shock-mounted strongback supports and protects the fuel elements. The standard strongback (see Figure 2.5) is designed to securely hold two long (or four short, see Figure 2.6) fuel elements in place with a minimum spacing of 6 inches between the two fuel element cavities formed by the strongback components.
The main strongback member is a single "U" shaped channel formed of 1/4 inch steel.
The standard strongback channel is about 196 inches long, 25-3/8 inches wide, and 12-1/2 inches high.
Alternate strongback channels that are shorter or have other minor design variations are used interchangeably with the standard strongback. Alternate strongback channels are structurally the same as the standard ones except for the dimensional differences.
All are fabricated of 1/4 inch thick steel.
See Figures 2.7, 2.8, 2.9 and 2.10).
Side and bottom steel angle (2 x 2 x 1/4-inch) supports are welded to the exterior of the strongback channel in seven locations on the standard strongbacks and five on the short strongbacks to provide rigidity and additional strength.
Separator blocks (3/8 inch thick channels, 6" wide x 8" high x 9" long) are bolted (two 5/8-llVNC-2 bolts each) to the strongback channel such that the centerline of the spacer blocks corresponds to the centerline of the strong-back channel.
The number of blocks used in each package 1631 151
2-4 XN-52, Rev. 1 is dependent upon the weight of the fuel element to be transported.
The minimum number required as a function of fuel element weight in pounds is specified in Section 2.1.2.
Fourteen 4 x 3 x 3/8 inch steel angles are welded to the exterior sides (seven per side) of the strongback channel (five for the short strongback).
During shipping, these angles secure the strongback to 2 x 4 x 1/4-inch support tubes by a 5/8-11UNC steel bolt, nut, and lcck washer system (one each per lock-down angle).
Seven strongback support tubes (five for the short strong-backs) provide support and hold the strongback issembly in place during shipping and storage.
These support tubes are fabricated of 2 x 4-inch steel channels (1/4 inch wall thickness) and are 29-5/8 inches long.
Tte support tubes are attached to the interior of the contain-ment vessel through shock mounts (two per support tube),
to the shock mount support brackets. The shock mounts minimize vibrational effects on the fuel elements during transport and handling.
In the event of a fire severe enough to destroy the natural rubber portion of the shock mounts, the fuel elements ramain in essentially the same position within the package as the result of the steel bolts, washers, and nuts incorporated into the shock mount assemblies (see Figure 2.11).
The effectiveness of the shock mount system is not fully realized unless the trunnion assembly is disengaged prior to sealing the containment vessel.
Consequently, the trunnion assembly contains a blocking feature that will not allow the cover and base assemblies of the containment 1631 152
2-5 XN-52, Rev. 1 vessel to be mated while it is engaged.
The trunnion assembly has no other transport significance; it is merely a device to aid in the loading and unloading of fuel elements.
Steel end thrust brackets (see Figure 2.12) are bolted to the strongback at both ends of the fuel elements to prevent longitudinal movement. When shipping four (4) fuel elements, the two short steel center thrust brackets (see Figure 2.6) are bolted into the strongback between fuel elements in each cavity.
A handle is attached to the center thrust bracket to facilitate bracket removal from the strongback during unpacking operations.
There are no materials specifically used as nonfissile neutron absorbers g moderators in this packaging.
2.1.2 Fuel Element Clamps, Shock Mounts and Separator Blocks Fuel elements are clamped in place within the strongback and restrained from lateral or vertical movement (see Figures 2.1 and 2.6).
These clamping devices hold the fuel elements against the bottom and sides of the strong-back channel such that the maximum fuel element separation distance is achieved.
The adjustable clamps are mounted on 2 x 1-1/2 x 1/8-inch steel angle brackets that extend laterally across the top of the strongback channel. These brackets are clamped (using two 5/3-inch steel bolts per bracket) to the top of the strongback channel.
There are two types of clamps, one designed to clamp on the spacers of PWR fuel elements and the other designed to clamp between the spacers of BWR fuel elements.
PWR fuel element clamps (see Figure 2.11) are steel and the surfaces 1631 153
2-6 XN-52, Rev. 1 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, s 3/4 and s 1/2 inch thick, added between the fuel element 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 ethaf am pads in place, are shown in Figure 2.13.
XN Types I, II, III, IV, V, VI, and some of the generically characterized fuel elements will be packaged with molded corrogated polyethylene shims between adjacent rows of fuel rods within the fuel elements. When such shims are 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 strongback 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 niethod using ethafoam pads will absorb at least 1.2 times the s-1 1631 154
2-7 XN-52, Rev. 1 energy of the originally designed and tested support system.
As a result of comparisons between the two support methods, it has been concluded that under maximum credible accident loading conditions either support system meets all struc-tural requirements (i.e., either the basic system tested, or the system using ethafoam pads, polyethylene shims, and clamps over the fuel rod spans between spacers).
When transporting fuel elements weighing in excess cf 800 pounds, restraint bars are included in the package.
Restraint bars consist of 2 x 1 1/2 x 1/8-inch steel angle brackets that extend across the top of the strong-back channel and are clamped to the strongback flanges in the same manner as are the full clamps.
The restraint bars are provided for additional restraint in the event of an accident.
Strongback components required for each package vary with the size and weight of the fuel elements shipped. The limiting criterion is that the components used to hold the fuel element in place in the strongback (i.e., the full clamps) do not fail at a lower force than the shock mount system.
(The fuel elements must be retained within the strongback).
The specific criteria applied is that the number of full clamps and separator blocks per unit weight shall be equal to or greater than the number of clamps and separator blocks cmployed in the Model 51032-1 30 feet drop tests.
The number of full clamps, shock mounts, and separator blocks to be included in the package "all satisfy the following equations:
1631 155
2-8 XN-52, Rev. 1 N 1 W/187.5 b
Nc1 N ~
S "s I "c
c Where:
N = number of separator blocks required; b
N = number of full clamps required; W = weight of the fuel element (pounds); and N = number of shock mounts.
s The number of restraining bars employed for transporting fuel elements weighing in excess of 800 pounds shall be one fewer than the number of full clamps, (i.e., N 'I)*
c In addition, half clamps are normally applied at the end of each fuel element but are not taken into account in this calculation.
These half clamps provide some degree of conservatism. When four short fuel elements are transported in one container W shall be t."e combined weight of the two fuel elements.
2.2 Model 51032-la Container The gross weight of the Model 51032-la packaging is 4600 100 pounds.
Specific materials of construction, weights, dimensions, and fabrication methods of the packaging components are as described below.
2.2.1 Container Descriotion The outer container vessel of the Model 51032-la con-tainer is identical (interchangeable) to that used for 1631 156
2-9 XN-52, Rev. 1 Model 51032-1 packages and described in Section 2.1.
The overall arrangement of the Model 51032-la container is shown in Figure 2.14. The strongback (see Figure 2.15) is also basically the same as the Model 51032-1 standard strongback except that the interior width is increased by one (1) inch and the thrust plate locations are changed to accommodate slightly larger fuel elements while main-taining at least a six inch separation between adjacent fuel elements.
Separator blocks used to assure a minimum separation between fuel elements within each container were modified by additic-of a gusset plate for increased strength (see Figure 2.16).
Significant differences between the two models occur in the shock mounts (see Figure 2.14), full clamps (see Figures 2.17 and 2.18), separator blocks (see Figure 2.16), and some of the associated bolts.
In addition to these differences wnich characterize the packaging model, additional components are employed when Type AA fuel elements are transported in the Model 51032-la containers.
These are 1) special strongback thrust brackets (see Figure 2.19) at each end, and 2) aluminum honeycomb impact limiters (see Figure 2.20) at each end between the thrust brackets and the end of the outer containment vessel.
(The special thrust bracket and honeycomb material are retained at the lower end (ar ir the trunnion) of the package for all fuel element shipments but the upper thrust bracket and honeycomb may be replaced with the bracket shown in Figure 2.12.)
2.2.2 Fuel Element Claacs, Shock Mounts and Seoarator Blocks The fuel element full clamps for Model 51032-la packages have been strengthened relative to those for Model 51032-1 1631 157
2-10 XN-52, Rev. 1 package as explained in Section 10.
The full clamp angle bar has been replaced by a 2-1/2 x 2-1/2 x 1/2 inch angle bar and the clamp that fastens the angle bar to the strongback has been revised for greater strength.
The design is shown in Figure 2.17.
The steel clamp used to fasten PWR fuel elements in the strongback is shown in Figure 2.17 and the alumir.um clamp for use with BWR fuel elements is shown in Figure 2.18.
This packaging also requires half clamps, one at each end of each fuel element and not less than one fewer restraining bars than full-clamps.
The number of full clamp assemblies (N ) required shall be sufficient to provide strength greater than that of the net strength of N shock-mount bolts, calculated s
at 13,000 lb force per bolt.
The strength of the clamp assemblies has been determined by experiment to exceed 23,000 lb force per assembly. The specific criteria for determining the required number of full clamp assemblies, shock mounts and separator blocks within each package are:
W "b 1 231; N 1 2 0; and
$N
-2 S N, 5 N e
c Where:
N = number of separator blocks required; b
N = number of full clamps required; c
N = number of shock mounts required; and s
W = weight of the fuel element (pounds).
1631 158
2-11 XN-52, Rev. 1 2.3 Both Models 51032-1 and 51032-la Containers There are no sampling ports or tie-down devices.
There are two valves on the containment vessels; one is used for pressurizing (with dry air or nitrogen) the con-tainment vessel prior to shipping (or storage), and one for relieving the containment vessel pressure prior to unsealing the vessel.
As such, both valves are located in one end of the containment vessel.
These valves are not of safety significance and, indeed, are not normally used (i.e., the containment vessel is not normally pressur-ized except for leak testing prior to shipment).
There are no structural or mechanical means provided or required for the transfer or dissipation of heat and there are no coolants utilized in the packages.
(Decay heat for the unirradiated fuels to be transported is negligible, < 20 watts).
2.4 Packace Contents 2.4.1 Model 51032-1 Container Each fuel element is enclosed in an unsealed polyethylene sheath.
The ends of which are neither taped nor folded in any manner that would prevent the flow of liquids into or out of the ends of sheathed fuel elements.
The maximum content weight for the Model 51032-1 package is 3400 pounds.
1631 159
2-12 XN-52, Rev. 1 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 six specific mixed Pu0 -UO nuclear fuel 2
2 elements and for generically described low-enriched UO 2 fuel elements are summarized herein.
Identification of these fuel elements, along with the maximum number of elements and the maximum radioactivity cf the radioactive constituents contained in a single package, are tabulated in Table 2-II for specific mixed-oxide (Pu0 -UO ) f"*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 ) 1"'I 2
2 elements and in Table 2-V for generically characterized U0 fuel elements.
2 All fuel elements contain pelletized and sintered UO 2
Pu0 -UO encapsulated within stainless steel or zircaloy 2
2 tubing.
The physical characteristics of the various fuel elements are tabulated in Table 2-XII.
In all cases, individual rods are held in the respective arrays by 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.
1631 160
2-13 XN-52, Rev. 1 2)
The clad is zircaloy 2 or 4, conservatively modelled as pure zirconium.
3)
The clad thickness assumed was 0.020 inch, a value whi,h 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, same Exxon Nuclear fuel elements contain gadolinium, cobalt, or other neutron poison rods.
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 51032-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-XII and Section 12.
1631 161
2-14 XN-52, Rev. 1 TABLE 2-I REVISED FUEL ELEMENT IDENTIFICATION NUMBERS Licensed Fuel Element Revised Fuel Element Identification Number Identification Number I
I III II VII III VIII IV XIV V
XVII VI 1631 162
2-15 XN-52, Rev. 1 c
TABLE 2-II RADI0 ACTIVE MATERIAL LIMITS (MIXED OXIDE FUELSJ Maximum Number of Maximum XN I.D.
Radioactive Elements Curies (Type)
Materials per Package per Packace I
Pu0 -UO 2
20,500 2
2 II Pu0 - 0 4
20,000 2
2 III Pu0 -UO 2
13,400 2
2 IV Pu0 -UO 4
49,200 2
2 V
Pu0 -UO 4
51,300 2
2 VI Pu0 -UO 2
13,500 2
2 1631 163
TABLE 2-III RADI0 ACTIVE MATERIAL LIMITS (U0 FUELS) 2 XN Humber of Maximum fuel Fissile Maximum Radioactive Elements Curies y jy TyLe Class w f Enrichment Material per Package per Package A
I 5 2.1 3.5 UO 2 or 4*
1.5 2
B I
$ 2.1 3.5 00 2 or 4 2.0 2
ro h
C III 5 1.8 4.0 UO 2 or 4 2.3 2
G III
$ 2.1 4.0 U0 2 or 4 2.3 2
E III
$ 2.3 4.0 UO; 2 or 4 2.3 F
III
$ 2.1 5.0 00 2 or 4 2.7 2
- Two fuel elements of standard length or 4 short fuel elements of equivalent weight.
f E
U asemame b
2-17 XN-52, Rev. 1 TABLE 2-IV FISSILE MATERIAL LIMITS (MIXED OXIDH FUELS)
Maximum Fissile Constituents
- XN I.D.
Total Maximum (Tvoe)
I.D.
Quantity I.D.
Quantity (kg/ Package)
(kg/ Package)
I U
247 U-235 6.1 Pu 3.0 Pu 2.46 f
II U
236 U-235 5.3 Pu 1.70 Pu 1.35 f
III U
362 U-235 7.6 Pu 2.42 Pu 2.00 f
IV U
510 U-235 16.0 Pu 6.00 Pu 4.80 f
V U
510 U-235 23.0 Pu 6.25 Pu 5.0 f
VI 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 Tables 2-VI through 2-XI.
1631 165
TABLE 2-V FISSILE MATERIAL LIMITS (u3 FUELS) 2 XN Maximum Fissile Constituents Fuel Fissile Maximum Total Maximum V /V TyjLe Class f
Enrichment I.D.
Quantity I. D.
Quantity (kg/ Package)
(kg/ Package)
A I
$ 2.1 3.5 U
700 U-235 24.5 B
I
$ 2.1 3.5 U
1500 U-235 52.5 C
III
< l.8 4.0 U
1500 U-235 60.0
?
D III
$ 2.1 4.0 U
1500 U-235 60.0 C'
E III
$ z.3 4.0 U
1500 U-235 60.0 F
III
$ 2.1 5.0 U
1500 U-235 75.0 E
b u
5 w
2-19 XN-52, Rev. 1 TABLE 2-VI XN-Type I Number of Rods Fuel Rod Description 4
Cobalt largets 16 2.55 + 0.05 w/o U-235 12 3.30 1 0.05 w/o U-235 21 4.20 + 0.05 w/o U-235 4
3.30 + 0.05 w/o U-235 - 1.0 + 0.05 w/o Gd 0 23 24 3.65 + 0.05 w/o Pu in Nat. U TABLE 2-VII XN-Type II Number of Rods Fuel Rod Description 21 2.95 + 0.05 w/o U-235 6
2.00 + 0.05 w/o U-23b 9
2.84 + 0.05 w/o Pu in natural uranium TABLE 2-VIII XN-Type III 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 + 0.05 2/o U-235 4
2.87 + 0.05 w/o U-235--1.0 + 0.05 w/o Gd 0 23 7
2.19 + 0.05 w/o Pu in Natural U 5
3.05 + 0.05 w/o Pu in Natural U 1
Solid Zircaloy-2 (No SNM) 1631 167
2-20 XN-52, Rev. 1 TABLE 2-IX XN-Type IV Number of Rods Fuel Rod Description 16 2.30 1 0.05 w/o U-235 32 3.20 + 0.05 w/o U-235 40 4.60 + 0.05 w/o U-235 4
4.60 + 0.05 U-235--l.2 + 0.05 w/o Gd 023 24 5.45 + 0.05 w/o Pu in Natural U 4
Cobalt targets (no SNM) 1 Solid Zircaloy-2 (no SNM)
TABLE 2-X XN-Tyce V Number of Rods Fuel Rod Descriotion 16 2.30 0.05 w/o U-235 32 3.20 + 0.05 w/o U-235 36 4.60 + 0.05 w/o U-235 4
4.60 + 0.05 U-235 - 1.2 + 0.05 w/o Gd 0 23 25 5.45 + 0.05 w/o Pu in Natural U 4
Cobalt targets (no SNM) 4 Solid Zircaloy-2 (no SNM)
TABLE 2-XI XN-Type VI Number of Rods Fuel Rod Description 20 2.64 + 0.05 w/o U-235 6
1.79 + 0.05 w/o U-235 9
2.74 + 0.05 w/o Fissile; Pu in Natural U 2
2.24 + 0.05 w/o Fissile; Pu in Natural U 1631 168
TACLE 2-XII LIMITING FUEL ElfMENT PHYSICAL CHARACTERISTICS Fuel Nominal Nominal Nominal Nominal Nominal XN Rod Array Rod Clad Nominal Fuel Pellet Active Fuel No. of Array Dimensions Pitch Clad Thickness Fuel Rod Diameter Fuel Length Jyge Fuel Rods (Square)
Jinches)
Unches)
Material (inches) 0.0. (inches)
(i nc hes)
(inches) 81 9x9 6.22 x 6.22 0.707 Zircaloy
.0375 0.563 0.478 69.
11 36 6x6 4.72 x 4.72 0.770 Zircaloy
.0345 0.570 0.489 60.
111 49 7x7 5.35 x 5.35 0.738 Zircaloy
.034 0.570 0.478 144.
IV 121 11 x 11 6.52 x 6.52 0.577 Zircaloy
.034 0.449 0.371 70.
V 121 11 x 11 6.52 x 6.52 0.577 Zircaloy
.034 0.449 0.371 70.
VI 36 6x6 4.50 x 4.50 0.700 Zircaloy
.0302 0.563 0.481 116.
A
$ 100
$ 10 x 10 $ 5.2 x 5.2
< 0.806 Zircaloy
.020
< 0.550
< 0.500
< 192.
m nr 55*
or.015 d3 s
B
$ 289
$ 17 x 17 $ 8.55 x 8.55
< 0.806 Zircaloy
.020
< 0.550
< 0.500
< 192.
or SS or.015 C
$ 289
$ 17 x 17 $ 8.60 x 8.60
< 0.769 Zircaloy
.020
< 0.550
< 0.500
< 192.
or SS or.015 0
$ 289
$ 17 x 17 $ 8.48 x 8.48
< 0.806 Zircaloy
.020
< 0.550
< 0.500
< 192.
or SS or.015 E
$ 289
$ 17 x 17 $ 8.40 x 8.40
< 0.830 Zircaloy
.020
< 0.450
< 0.400
< 192.
or 55 or.015 F
$ 289
$ 17 x 17 $ 8.00 x 8.00
< 0.806 Zircaloy
.020
< 0.550
< 0.500
< 192.
or SS or.015 16 x 16 9.01 x 9.01 0.563 Zircaloy
.030 0.424 0.3565 154.
2
~
AA**
< ?56 N
W May t,e Zircaloy-2, Zircaloy-4 or stainless steel. For criticality safety
~
calculatio-a 0.20 inch zirconium clad was assumed.
y Type AA fuel elements may t;e transported only in Model 51032-la containers.
~
O a
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5-1 XN-52, Rev. 1 5.0 GENERAL STANDARDS FOR PACKAGING The materials from which the packaging is fabricated (steel, rubber padding, and gaskets), along with the contents of the package (zircaloy or stainless steel ciad fuel rods, stainless steel and inconel fuel element hardware, polyethylene wrapping, and desiccant material),
will not cause significant chemical, galvanic, or other reactions in air, nitrogen, or water atmospheres.
The positive closure system has been previously described in Section 2.
In addition, each package will be sealed with Type E, tamper indicating seals. These features prevent inadvertent and undetected opening.
The lifting system (four steel lugs welded to the cover assembly stacking brackets) was analyzed to be capable of lifting an 8300 pound package without generating stress in any material of the packaging in excess of its yield strength with a minimum safety factor of 3.4 (see Appendix I).
Alternatively, two forklift pickup channels (1/4 inch steel), are welded to the bottom of the containment vessel base assembly to facilitate forklift handling.
Administrative controls are used to prevent the lifting of stacked packages.
If the lifting system were to be subjected to an excessive load and fail, continued containment of the contents would not be jeopardized since the containment of the radioactive materials is not dependent upon the packaging.
There are no shielding considerations involved.
1631 190
5-2 XN-52, Rev. 1 Administrative controls prevent the use of any structural part of the package as a lifting device.
There is no identified system of tie-down devices on the packages.
However, a combination of shoring, positioning studs, axial, and transverse chokers (chain or cables) is employed to secure packages to the transport vehicles.
The " tie-down" system used to satisfy the criteria set forth in 10 CFR 71.31(d) is as shown in Figure 3.1.
The only structural part of the packaging which could be employed to tie the packages down are the stacking brackets and stiffening rings.
There are eight stacking brackets per package and the analyses in Appendix II show that a minimum of two of these (per package) could be used in a " tie-down" arrangement (along with shoring and cross chokers). The stiffening rings are normally used as tie-down points.
These are heavy members that can easily support the tie down loads.
If the stacking brackets were to be subjected to an excessive load and fail, continued containment of the package contents would not be jeopardized since the containment of the radioactive materials is not dependent upon the packaging.
I631 19l
10-6 XN-52, Rev. 1 Where W is the weight of one fuel element expressed in pounds or, if four fuel elements are contained, the combined weight of two fuel elements.
10.1.2.2 Intecrity of the Aluminum Clamos Due to the excessive weight of steel fuel clamps for packaging BWR fuel elements, Exxon Nuclear has designed the aluminum clamps shown in Figure 2.13.
These clamps would be loaded most severely in a hypothetical drop on the container cover. As described in Appendix V, the clamp loading and deformation is limited by the early tensile failure of the shock mount bolts and contact of the clamp angle bars with the container cover.
Tests on the aluminum clamps have shown that the aluminum clamps will behave as well as the steel ones used in the drop tests and will retain the fuel elements within the strong-back.
(See Figure 10.1 for the comparison of the force deflection curves for the steel and aluminum clamp assem-blies.)
10.1.2.3 Short Stronabacks Used in Some Shipments Some fuel elements are significantly shorter than the standard strongback for Model 51032-1 containers.
A shorter strongback (see Figure 2.7) has been designed which will be used interchangeab'y with the standard strongback for those fuel elements which can be accomo-dated.
Except for length, it is structurally the same and would be equally effective in retaining the fuel elements in the hypothetical accident.
In addition, the shorter fuel elements have a corresponding decrease in weight wnich results in reduced loads under hypothetical accident conditions.
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10-10 XN-52, Rev. 1 When used for shipping fuel elements which result in a package gross weight of 7400 pounds or less, the upper thrust plate and honeycomb material shown in Figures 2.19 and 2.20 may be replaced with the thrust plate shown in Figure 2.12.
In this configuration energy dissipation at the upper end of the strongback is quite similar to that of the Model 51032-1 package except that the increased energy dissipation capability of shock mount and other bolts within the package is preserved.
10.2.1 Model 51032-la Container-End Drop Evaluation In the Model 51032-1 drop test, the shock mount bolts sheared with little energy dissipation when the container impacted with the ground.
The container crumpled only two inches at impact with the only evicent damage being to the container end where the ring was torn loose at the weldments and the end pushed in slightly at the flange which was crumpled over.
Those distortions represented the conversion of the container kinetic energy to strain energy.
Then, following shearing of the shock mount bolts, the moving strongback impacted the end and caused both further container end damage and crumpling of the end of the strongback.
Except for localized damage, the package was not significantly damaged and the result demonstrated compliance with the Part 71 packaging stand-ards.
With respect to the vertical drop model, the main differ-ences with the new package design are the shorter strong-back extensions beyond the thrust plates, the increased weight of strongback and fuel eleme,nts, the added aluminum honeycomb, and the change in bolts as outlined above.
Damage to the end of the container in the initial imoact 1631 193
10-14 XN-52, Rev. 1
~
Due to the increased energy dissipation of bolts and clamps in the Model 51032-la package, the impact energy would be significantly reduced.
Otherwise, the nature of the impact of the stringback against the cover would be very similar to that observed in the drop test.
The approximate energy balance is presented in Table 10-I.
10.2.4 Model 51032-la separator Block Integrity The integrity of the separator blocks in the Model 51032-1 package was not tested in the Exxon Nuclear drop tests.
In a hypothetical drop on a closure flange, the separator blocks would be loaded by deceleration of one of the fuel elements.
It is assumed that the fuel element clamps would not be effective in supporting this load and would fail if the separator blocks were crushable.
To assure that the blocks will withstand the required force, a gusset plate is welded within the blocks as shown in Figure 2.16.
The blocks were tested as described in Appendix V and assure a minimum spacing of six inches between fuel elements within the container. The number of blocks required for 1850 pound fuel elements is eight (8).
10.3 Fuel Rod Droo Tests To supplement information obtained from the package drop tests and assess the capability of fuel rods to withstand dynamic loads similar to those experienced under hypothetical accident conditions, drop tests were also performed with individual fuel rods.
Details relative to t:sse tests 1631 194
12-1 XN-52, Rev. 1 12.0 SPECIFIC STANDARDS FOR FISSILE CLASS I AND III PACKAGES Model 51032-1 packages containing XN Type III, IV, and VI fuel elements are transported as Fissile Class I shipments on, or in, multiple use vehicles.
Model 51032-1 packages containing XN Type I, II, and V fuel elements are trans-ported as Fissile Class III shipments on, or in, exclusive use vehicles.
Model 51032-1 or 51032-la packages containing generically characterized (U0 ) fuel re transported either as Fissile 2
Class I shipments on, or in, multiple-use vehicles; 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 elements.
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 + 30 <_ 0.970.
Fuel element parameter 1631 195
12-2 XN-52, Rev. 1 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 Fuel Elements The methods and nuclear data utilized to calculate k,of XN Type I fuel elements are consistent with the methods and data used throughout the nuclear industry for water moderated systems.
The analysis utilizes the HRG code to obtain multi group epithermal cross sections and :e THERM 0S code to obtain cell-averaged trarmal group pars.r.-
eters for each rod. The two-dimensioral diffusion theory code 208 is used to compute the k, cf the fuel elements.
To verify the accuracy of this method, it was used to compute the k,ff of a series of experimental arrays of mixed oxide (Type I) fuel rods surrounded by water reflec-tors.
Comparisons of calculated and experimental k gff values are shown in Table 12-I and are discussed in more detail in Reference 4.
As can be seen in Table 12-I, the calculational method predicts k well within 1 percent ak/k in each case.
eff The calculated k, for a fuel element is applied to the criticality evaluations by use of the one group buckling calculation:
k, eff I+B M g
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12-3 XN-52, Rev. 1 2
The value of the migration area (M ) is also obtained from the HRG/ THERMOS model, and is internally consistent with the calcu-lated k,.
2 The geometrical buckling (Bg ) is obtained from:
2 2
B = (x.
)
+( " 3) g Whera x and y are the lateral dimensions of the fuel element and the fuel element is assumed to be infinitely long.
The augmentation distance (A) for light water moderated and reflected fuel rods normally falls in the range of 6 to 7 cm; a value of 7 cm is used in these criticality calculations.
Use of the one group buckling calculation rather than the two group model, 2
k,EXP (-B t) q k
=
eff 22
).g L
9 results in a calculated k,7f for these fuel elements which is approximately 10 to 20 percent ak/k conservative (high).
12.1.2 XN Type II Fuel Elements The methods and nuclear data utilized to calculate the k, of the XN Type II fuel elements are consistent with the methods 1631 197
12-4 XN-52, Rev. 1 and data used throughout the nuclear industry.
The analysis utilizes the JERBEL code (an improved version of LEOPAhD) to obtain multigroup cross sections and k, for fuel rod cells, and tne P0Q-7 code for two-dimensional rod array calculations.
Confirmatory rod array calculations and reactivity calculations for the fuel element arrays (c'escribed in Section 12.4) were performed with the Monte Carlo code KENO, using cross sections derived from the CCELL (HRG/ THERMOS) code.
The JERBEL code has also been used by Exxon Nuclear to compute k,ff of a series of experimental critical arrays of fuel rods which represent wide variations in fissile isotope type and content, wide variations in moderator-to-fuel ratio, both zircaloy and stainless steel cladding, and various concentrations of soluble poison in the water moderator.
The results of these calculations are tabulated in Table 12-II.
The average difference between the computed k and 1.000 is less than 0.2 percent.
eff 12.1.3 XN Tyoe III, IV, V, VI, AA and Generically Characterized Fuel Elements 12.1.3.1 KENO II (18 Eneroy Gr A Calculational Method The KENO-II Monte Carlo code was used to calculate reactiv-ities of interacting arrays of well moderated packages.
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 performed using cross section data averaged by the GAMTEC-II code.
1631 198
12-5 XN-52, Rev. 1 These evaluations, although primarily for plutonium fueled systems, demonstrate the self consistency of the GAMTEC-II code.
To demonstrate the adequacy of the Gl.MTEC-II code for undermoderated slightly enriched uranium systems, the infinite media multiplication factor was computed for 5 w/o U-235 UO2 p wder using cross section data obtained from the ENDF/B-III library.
The resulting value of k,,
as well as the values obtained using the JERBEL and HAMMER codes, are given in Table 12-III.
The computed value for k, for unmoderated 5 w/o enriched UO 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 KEN 0 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, 7.79 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 s
ev, 0.683 ev.
1631 199
12-6 XN-52, Rev. 1 The results of these calculations are shown in Table 12-IV 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 KEN 0 calculated reactivities given in Table 12-IV agree with the previously performed DTF-IV transport theory calculations within the statistical uncertainty of the Monte Carlo calculations.
12.1.3.2 KEN 0 IV (123 Eneroy Grouo) Calculational Method In addition to the method described above, the KEN 0 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 KEN 0 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 represtatative of th9 fuel region for input into KENO IV.
Theory-experiment correlations have been performed for UO r d-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-IV and are presented with the results of other theory-experiment correlations in Reference 7.
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12-7 XN-52, Rev. 1 12.2 Results of k Calculations c
12.2.1 XN Tyoes I and II Fuel Elements The fully moderated k, of the nominal Type I fuel elements is 1.148 (cobalt and gadolinia included), and 1.34 (neither cobalt nor gadolinia included).
The fuel elements will be shipped with both types of poison rods installed.
For nuclear safety evaluations, it is assumed that the poisons are cmitted, and a k, of 1.34 is used in related calculations.
The uncertainties in the fissile isotope content of the fuel elements introduce an uncertainty in k, of less than 1 percent.
For fully moderated XN Type II fuel elements k,is calculated to be 1.32 using the JERBEL/PDQ-7 method, and 1.34 using the HRG/ THERMOS method.
Since the calculations performed using the JERBEL/PDQ-7 method explicitly recognize the actual distribution of Pu0 within the fuel element while 2
the HRG/ THERMOS method assumes a uniform distribution of Pu0, the difference between the two calculated values of 2
k,is to be expected.
The uncertainties in the fissile isotope content of the fuel elements introduce an uncer-tainty in k, of less than 1 percent.
12.2.2 XN Types III, IV, V, and VI 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-V.
1631 201
12-8 XN-52, Rev. 1 For comparison, Table 12-V also gives values of k, computed using the JERBEL code, or a two-dimensional diffusion theory code JDT.
The two-dimensional code used cross-section data averaged either by the CCELL code or by the JERBEL code, indicated as CCELL/JDT or JERBEL/JDT, respectively.
The two-dimensional code gives values of k, for the entire fuel element while values of k, computed using the CCELL and JERBEL codes assume a fuel element averaged pin. This assumption has been shown to be conservative by comparisons with detailed design calculations (see Table 12-V for typical comparisons).
It is also apparent from data given in Table 12-V that for XN Type IV fuel elements, the CCELL code is approximately 7 percent conservative with respect to the more detailed CCELL/JDT method; and that, for XN Type V fuel elements, it is conservative by about the same amount relative to the JERBEL/JDT method.
This conservatism for the XN Types IV and V 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 percant 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 III, IV, and V mixed-oxide fuel elements all contain Gd 0 Poisoned 23 fuel rods which were ignored in the CCELL calculations reported in Table 12-V and in subsequent criticality safety calculations.
1631 202
12-9 XN-52, Rev. 1 12.2.3 Generically Characterized Fuel Elements Infinite media multiplication factors for U0 rod-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)
Fiaure 12.1--The maximum values of k, for 3, 4, and 5 wt percent U-235 enriched UO r ds 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 r ds in 2
water occurs at a pellet diameter of 1 0.5 inches for water-to-fuel volume ratios of < 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 having enrichments of < 5 wt.% 235-U and water-to-fuel volume ratios of < 2.1, 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 0.4 inch.
1631 203
12-10 XN-52, Rev. I 12.2.4 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 value of k, of 1.421.
12.3 Single Package Evaluation 12.3.1 XN Type I and II Fuel Elements The Model 51032-1 package will contain two XN Type I fuel elements.
The Model 51032-1 packaging was designed.to accommodate four such fuel elements, but current needs require that no more than two be loaded per package.
The two Type I (short) fuel elements will be secured at opposite ends of the strongback and on opposite sides of the separator blocks.
In order to simplify calculations, the two fuel elements are assumed to be secured at the same end of the strongback with a separation distance equal to the width (6 inches) of the separator blocks (the actual separation distance will be approximately 12 inches).
Complete water moderation (k, = 1.34) and full water reflection are also assumed.
The isolation provided by the water assumed to be between the fuel elements is ignored.
Based on the above information and assumptions, k eff single Model 51032-1 package containing two XN Type I fuel elements was calculated to be less than 0.84.
1631 204
12-11 XN-52, Rev. 1 Model 51032-1 packages may contain four XN Type II fuel elements (two positioned end-to-end on each side of the separator blocks).
Complete water moderation and full water reflection are assumed.
Based on these assumptions, the k,ff of a single package is calculated to be < 0.74 using the JERBEL/PDQ-7 method, and < 0.75 using the CCELL/ KENO method.
12.3.2 XN Types III, IV, V, and VI Fuel Elements For the XN Type III, IV, V and VI fuel elements described in Section 2, the reactivity of a single package is less than those ccmputed 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 VI fuel elements, it was assumed that the ethafoam pads between the strongback and the fuel elements were totally crushed. This assumption increases the reactivity 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-VI.
1631 205
12-12 XN-52, Rev. 1 These values are all based on the 95 percent confidence level (k average i 1.960), and were computed using the gf 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.
12.3.3 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 portion 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.
1631 206
12-13 XN-52, Rev. 1 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 self-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-VII.
12.3.4 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 twc-fold:
1)
No fissile material will be interacting with the single package; and 2)
Two sides of each fuel element will le 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 0.886
.008 which was computed for the damaged package array.
1631 207
12-14 XN-52, Rev. 1 12.4 Demonstration of Comoliance With 10 CFR 71.38 and 71.40 12.4.1 Undamaged Fissile Class I Package Arrays 12.4.1.1 XN Types III, IV, and VI 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 element Types I through VI.
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 III, IV, and VI) is given in Table 12-VIII.
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 contain 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-VIII).
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 element size.
These assumptions 1631 208
12-15 XN-52, Rev. 1 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 elements.
A single cell of an infinite array of such undamaged 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.
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.
Results of the infinite array calculations for XN Fuel Types III, IV, and VI are shown in Table 12-IX.
As can be seen, an infinite array of undamaged packages of Types III, IV, or VI fuel elements is subcritical and thereby satisfies the requirement of 10 CFR 71.38(a).
I 209
12-16 XN-52, Rev. 1
/
12.4.1.2 Generically Characterized Fuel Elements As previously noted ft, 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 polyethylene shipping shims, the reactivity of each array was determined assuming a limiting 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/cm.
To determine the fuel element size which may be transported as a Fissile Class I package, reactivities were computed using 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-X. These data were utilized to establish 1631 210
12-17 XN-52, Rev. 1 a limiting fuel element size of 5.2 inches, at an enrichment of 3.5 wt.% 235-U, when the packaging includes plastic shipping shims within the fuel element and surrounding ethafoam pads. The limiting conditions for the shipment of fuel elements contain-ing corrugated polyethylene shipping shims as Fissile Class I packages are indicated as XN Type A fuel in Table 12-X 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 at 8.55 inches square.
In the case of unmoderated fuel elements, the maximum reactivity of the array occurs at the maximum U02 density within the fuel element. Consequently, the following limitations were assumed:
1)
UO pellet density--100 percent of theoretical; 2
2)
Pellet diameter--0.5 inches (maximum); and 3)
Water-to-fuel volume ratio of the fuel assembly-
-1.3 (minimum).
1631 2ll
12-18 XN-52, Rev. 1
.>.;~
...: -... ~
The results of these calculations are given in Table 12-XI. 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 keff + 3a < 0.97, it is demonstrated that Type 8 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-XII) 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 expanded polyethylene) 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 configuration 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 < 0.905 at the 95% statistical confidence level.
This value was computed using the KENO-IV computer code with 123 group cross section data obtained from the NITAWL/XSDRNPM codes as described in Section 12.1.3.2.
16,31 212
12-19 XN-52, Rev. 1
'~
..~.O...
m,, -
..,.-.,. n eg -...y_. w, ~
~
12.4.2 Undamaged Fissile Class III Package Arrays 12.4.2.1 XN Type I Fuel Element A shipment of XN Type I fuel will consist of a single package containing two Type I fuel elements.
For purposes of evaluating a double shipment, it was assumed that two undamaged packages were stacked on top of one another.
Thus, the four fuel elements would be in a rectangular array with minimum horizontal and vertical edge-to-edge separation distances of 6 (assumed) and 19 inches, respect-iveiy.
Since the packages were assumed undamaged, there would be no water inside the containment vessel.
Neverthe-less, in this evaluation, complete water moderation of the package contents was assumed.
The system was evaluated by the solid angle method (described in Reference 5).
The k,ff of each fuel element, in this case air reflected to permit interaction, was calculated to be less than 0.59 (augmentation distance = 4 cm).
The subtended fractional solid angle of three units, entered on the fourth unit, was calculated to be 0.191.
This value falls within the guideline in TID-7016, Revision 1, which assumes a closely fitted reflector around the array.
Thus, a double shipment of undamaged ;]ackages containing XN Type I fuel has been demonstrated (conserva-tively) to be subcritical.
12.4.2.2 XN Type II Fuel Element A shipment of XN Type II fuel consists of a maximum of five packages containing up to four XN Type II fuel elements each.
Two shipments would contain 40 fuel 1631 213
12-20 XN-52, Rev. 1
.. w7 _.--. u
..,,g. 7,. -
.-.,,...~,+._._...
r, elements.
Packages are shipped in an array, two packages wide by two packages high.
Compliance with 10 CFR 71.40(a) and (b) was evaluated using the KENO code as described below.
To assure a conservative evaluation of the nuclear safety of the shipment, it was assumed that the shipment was disarrayed and crushed so that the separation between fuel elements provided by the outer container was lost.
It was further assumed that the separation between adjacent fuel elements in a single strongback was reduced to only that provided by the separator blocks (6 inches), as opposed to the as-loaded separation (15 inches).
Under these assumptions, the shipment becomes an array of four-fuel element cells, as shown in Figure 12.10.
The steel shown in Figure 12.10 represents the sides and bottoms of the strongbacks plus the steel of the outer shell.
The calculations assumed a horizontal infinite array of such cells, and the cells were assumed to be infinitely long.
The multiplication factor of this infinite array was calculated using cross sections (18 energy groups) averaged by the CCELL code in the Monte Carlo code KENO.
The resulting k for the array is 0.803 +.009.
The gff result conservatively demonstrates compliance with 10 CFR 71.40(a) which requires that subcriticality be maintained for two undamaged shipments placed side-by-side and closely reflected by water.
This result also conservatively demonstrates compliance with 10 CFR 71.40(b) with respect to the criticality safety of a single shipment subjected to the hypothetical accident conditions (see Section 12.4.3.1).
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12-21 XN-52, Rev. 1 n u-n - m..,,: n.c.,
.ms.(~ -.,;.... m-
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12.4.2.3 XN Type V Fuel Element For undamaged packages containing XN Type V 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 12.11.
For these packages, however, each fuel element 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 XN Type V fuel element.
These calculations conservatively demonstrate that touch-ing identical shipments of undamaged packages containing XN Type V fuel would be subcritical when fully reflected on all sides by water.
12.4.2.4 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 12.4.2.3 were used.
Results of the calculations for the various fuel element parameter limitations are given in Table 12-XIII.
These results, when compared to those presented in Section 12.4.3, show that Fissile Class III packaging limitations must be established on the basis of the damaged package arrays (i.e., arrays of damaged packages containing 1631 2l5
12-22 XN-52, Rev. 1 u ? ~. rc.m- -* L w - u n.~,-
,,*+ec v.nr v.. - 4+i.,..
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 and II Fuel Elements A shipment of Type I fuel will consist of a single package containing not more than two Type I fuel elements.
The calculations and evaluation presented in Section 12.3.1 are applicable to this case.
Thus, a shipment subjected to the hypothetical accident conditions has been demonstrated to be subcritical.
As stated above in 12.4.2.2, compliance with 10 CFR 71.40(b) for XN Type II fuel elements is demonstrated by evaluation of the array considered therein.
12.4.3.2 XN Tvoes III, IV, V, and VI 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 that 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 adjacent packages, and a minimum of 2 inches is provided by the package stiffener rings.
(This results in a total minimum separation of 8 inches, and assumes that the stiffener rings overlap and do not meet when stacking damaged packages.) Also, the separator blocks between b
b
12-23 XN-52, Rev. 1 fuel elements within individual packages have been shown 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 12.12.
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 been conservatively ignored.
Also, the assumed geometric configuration postulates both the minimum vertical and horizontal separations simultaneously, a situation that is impossible to achieve under hypothetical accident conditions.
Reactivities calculated for the XN Types III and IV mixed-oxide fuel elements in this assumed configuration, and for XN Types V and VI 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-XIV.
These results conservatively demonstrate compliance of XN Fuel Types III, IV, V, and VI with the requirements of 10 CFR 71.40(b).
12.4.3.3 Generically Characterized Fuel Elements The specific standards for licensing of Fissile Class I packages includes the requirement that (see 10 CFR 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 the effects of the hypothetical accidact conditions as specified in 10 CFR 71, Appendix B, with optimum credible 1631 217
12-24 XN-52, Rev. 1
,, _......... _ - _. _.. _,.. ~.............,.. - _.,. - _., -
interspersed hydrogenous moderation and close-water reflection-- must remain subcritical.
Both of these requirements are conservatively satisfied by consideration herein of an infinite array of damaged packages.
As discussed in Section 12.4.3.2, 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 of the steel strongback are conservatively ignored.
The packages were assumed to be fully flooded.
Homogenized cross sections were generated by the CCELL code (HRu/ 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 gecmetric configuration described are given in Table 12-XV 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 eff of the array be 5 0.97 at the 99 percent statistical confidence level (i.e., keff(average) + 30 < 0.97).
Using this criterion, data presented in Table 12-XV were 1651 218
12-25 XN-52, Rev. 1
.- - ~ -.-.. --,....,,., -.
-.--n.-~.
used to derive Fissile Class I and III package limits based on the reactivity of infinite arrays of damaged packages.
Limiting fuel element characteristics are summarized in Table 12-XVI.
12.4.3.4 XN Type AA Fuel Elements The assumed geometric 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.
The reactivity calculated for an infinite array of fuel element packages in this assumed configuration is 0.886
.008.
This value was computed using the KENO-IV computer code with 123 group cross section data obtained as summarized in Section 12.1.3.2.
12.4.3.5 Shioments 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 method 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 elements.
Consequently, it is requested that generic packaging limitations derived in Section 12 be applied to permit b
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12-26 XN-52, Rev. 1
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shipment of fuel rods in wooden boxes.
Such boxes would be constructed as indicated in Figure 12.13.
(Dimensions and packaging methods shown in Figure 12.13 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 UO 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 U0 pellet 2
diameter of < 0.5, be permitted within individual packages as shown in Figure 12.14.
As can be seen, the rod will be wrapoed 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 strangback.
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 effect on the reactivity of the package under both normal and accident conditions of transport.
1631 220
12-27 XN-52, Rev. 1 m,
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 Modsl 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 Summarv The results of criticality safety evaluations reported herein demonstrate that the XN Types III, IV, and VI mixed-oxide fuel element packages using Model 51032-1 shipping containers satisfy the requirements for Fissile Class I packages.
The results also demonstrate that XN Types I, II, and V mixed-oxide fuel elements contained in Model 51032-1 packaging satisfy the requirements for Fissile Class III packages.
Table 12-XVI contains a summary of the limits on fuel element parameters 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 keff
+ So <.970 for both normal and accident (damaged package) conditions.
Fuel element parameter limits were determined by using the criterion which imposed the more conserva-tive limitations.
A summary of the reactivities computed or interpolated for the various fuel types under both normal and accident conditions is given in Teble 12-XVII.
1631 221
13-1 XN-52, Rev. 1
13.0 REFERENCES
1.
CONF-710801 (Volume 2) Health anti 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 F. Jel 927A",
July 3,1969, Licenca SN:1-lC57, Docket 70-1100.
3.
Exhibit P (including Appendix P-1), " Application for Licensing of Combustion Engineering, Inc., Shipping Containers Models 9278 and 927C", February 23, 1971, License SNM-1067, Docket 70-1100.
4.
V. O. Votinen, G. L. Gelhause, U. P. Jenquin aid C. R. Gordon, " Calculations of Power Distribution and Reactivities", Reactor Physics Ouarterly 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 Aoplications, 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.
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V-1 XN-52, Rev. 1 APPENDIX V PACKAGE COMPONENT EVALUATIONS The Model 51032-la package includes design changes to assure that the main shock mount bolts yield and dissipate energy in any drop configuration, except for a drop in the normal upright configuration which is the least subject to failure potential.
Clearances between the assembled strongback and the containment vessel are about two and one-half inches and may limit the combined distortion of bolts, full-clamps and shock mounts.
The static tensile tests result in shock mount bolt failure at about 1.7 inches of bolt distortion, which is sufficient for the desired energy dissipation.
When the bolts are loaded transverse to the shock mount, the net distortion at failure, including bolt and mount, is about 3.2 inches in static tects.
To assure that the distortion occurs in the shock mount bolts, other bolts which could fail and relieve the stress on the shock mounts have been strengthened.
These have also been tested stat-ically to verify that their strength exceeds that of the shock mount bolts.
Full-clamp assemblies consist of 2-1/2 x 2-1/2 x 1/2 inch angle bars which span the strongback, clamping to its lip, and sliding clamps which bolt to the angle bars and hold the fuel elements in the corners of the strongback (see Figure 2.17).
These full-clamps were strengthened, by about a factor of three, relative to the drop-tested package to assure that they retain the fuel elements within the strongback during the time required to distort and fail the shock mount bolts.
In the drop-tested package, abrupt failure 1631 223
V-6 XN-52, Rev. I honeycomb is not sensitive to this uncertainty and is described here assuming that the gap is closed only by relative motion between strongback and container.
The two ends of the shipping package differ in honeycomb absorber design because at the fuel element nozzle end the nozzle projects two inches into the honeycomb.
The honeycomb is cut back in that area and additionally cut back to facilitate assembly.
The design is shown in Figure 2.20.
As the strongback and fuel elements move forward, the nozzle impails into the honeyccmb and it is assumed that the honeycomb area interior to the nozzle is unavailable for energy dissipation.
Crushing of the raised section of honeycomb material begins when the 1/2 inch clearance gap is closed.
The area of the raised 2
section is 135 in and begins crushing first. With the exception of the nozzle area, the depressed section begins to compress when the strongback has moved an additional 2.25 inches toward the container end.
The 2
depressed area crushed is 77 in The honeycomb thick-ness is 7.75 inches in the raised area and 5.625 inches in the depressed area.
The manufacturer has provided test data on the production run for the honeycomb which shows that the honeycomb will crush 80% with an average force of 1310 psi.
The energy absorption capability is therefore:
E = 1310 x 0.8 x {7.75 x 135 + 5.625 x 77} = 1,550,000 in-lb E = 120,000 ft-lb 1631 224
V-8 XN-52, Rev. 1 crashes uniformly and has a restraining force of 400,000 lb.
Complete compression from 8.25 inch thickness to 1.65 inch thickness could absorb 220,000 ft-lb of energy.
Since the shock mounts provide 30,000 ft-lbs and the total needed is.nly 159,000 ft-lbs, the honeycomb will only crush 4 inches.
The strongback will not reach the container end and will not crush.
V.4 Integrity of the Full Clamos A lower limit for the strength of the full clamps was determined by loading one in a near prototypical manner on the Tinius-Olsen te. ting machine.
Preliminary tests demonstrated that small design changes would greatly improve the performance and, therefore, part no. 5 of Figure 2.17 was replaced by a similar part 3/4 inch thick and 2 1/2 x 4 inches.
This provides 4 inches of bearing length on the lip of the strongback.
The bolts for the sliaing clamp have been replaced by similar but high-strength grade 8 bolts with 150,000 psi ultimate strength.
In the final test the 2 1/2 x 2 1/2 x 1/2 inch angle bar began yielding at 17,000 pounds force and was bent 1 inch at 23,000 pounds force.
At that point there was some slippage in the test jig linkage and the bolts, part 10 of Figure 2.17 appeared near to failure.
The test was run with a weaker SAE grade 2 bolt rather than the specified grade 8.
Because the test had demonstrated sufficient strength the test was terminated prior to failure.
The total deflection of the beam resulting from combined beam bending, bolt distortion, clamp distortion, and strong-back lip distortion was 2.3 inches.
The distortion at 23,000 lb would have been less and the strength higher with the high-strength bolt.
1631 225
V-9 XN-52, Rev. 1 The test also determined that the sliding clamp is self-locking under the applied loads and will not slip.
There are nine full clamps in the Model 51032-la shipping package with the Type AA fuel elements.
These provide a total restraining force in excess of 207,000 pounds of force (9 x 23,000).
This is sufficiently larger than the 180,000 pound strength of the 14 shock-mount bolts and assures that the shock-mount bolts would elongate to failure and prevent failure of the full clamps in a 30 ft drop on the cover.
Tests conducted on the aluminum clamp assemblies shown in Figure 2.18, result in a deflection of 0.267 inch at 10,000 pounds force.
For BWR fuel elements this indicates smaller deflections, at equivalent "g" loadings, than were obtained in the drop tested Model 51032-1 package.
The force deflection curve is shown in Figure V.4.
V.5 Integrity of the Separator Blocks The Model 51032-la package separator blocks have been tested on the Tinius-Olsen compression machine. The test established that buckling strength of the gusset plate was greater than the 30,000 pound limit of the machine.
The plate did not buckle and there was no significant block deformation.
Without the gusset plate, significant deformation occurs at 16,000 pounds force.
The attachment of the separator blocks to the strongback was also strengthened.
- Notably, Grade 8 bolts with a shear strength of 90,000 psi are used instead of carbon steel bolts with a shear strength of 38,400 psi and 3/8 inch thi :k washers are added in place of 1/8 inch thick washers to distribute the load over a larger area of the strongback channel.
1631 226 14897
.