ML20198P094

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Criticality Safety Analysis for BU-7 Shipping Package Without Boral Liner - Scrap Matl Content
ML20198P094
Person / Time
Site: 07109019
Issue date: 12/30/1998
From:
GENERAL ELECTRIC CO.
To:
Shared Package
ML20198P091 List:
References
NUDOCS 9901060307
Download: ML20198P094 (27)
v • d • e


Text

'

APPENDIX J

" CRITICALITY SAFETY ANALYSIS FOR BU-7 SHIPPING PACKAGE WITHOUT BORAL LINER - SCRAP MATERIAL CONTENT", DATED DECEMBER 30,1998 9901060307 981230 5

PDR ADOCK 07109019 d

PDR LICENSE SNM-1997 DATE 12/30/98 PAGE DOCKET 71-9019_

REVISION 0

J

ilU-7 USA /9019/AF December 30 1998 0

l Criticality Safety Analysis Model BU-7 Shipping Package without Boral Liner Scrap Material Content i

l GE Nuclear Fuels Page 1 of 26

BU-7 USA /9019/AF December 30; 1998 l

6. CRITICALITY SAFETY EVALUATION l

l 6.1 GENERAL DESCRIPTION The BU-7 packaging consists of a 16 gallon steel drum that is the containment system and a 55 gallon steel drum. The 16 gallon steel drum permanently installed in the 55 gallon drum and foam insulation fills that space between the containment and 55 gallon drum.

A removable insulation plug is installed between the inner drum lid and 55 gallon drum cover.

The package shall be used to transport unirradiated uranium oxide compounds of various physical and chemical forms. The contents are not to exceed of 15 kg UO2 (13.22 kg U) 2 at 5.00 weight percent enrichment in "U.

This mass limit for the contents allows for efficient loading and shipment of the scrap material generated by low enrichment uranium fuel manufacturing operations 6.2 PACKAGE DESCRIPTION 6.2.1 CONTENTS The package shall be used to transport solid uranium in various chemical forms. The source of the uranium is the feed material to conversio, which meet the requirements for Enriched Commercial Grade Uranium defm' ed in ASTM LW6-96. The uranium isotopic uranium distribution is shown in Table 6.1.

Table 6.1 Uranium Isotopic Distribution Isotope wt %

29U 0.0 - 0.05 235U 0.1.

5.00 23%J 0.0 - 0.025 238U 99.3 -95.00 6.2.2 PACKAGING The packaging consists of an inner containment drum,55 gallon drum, and foam insulation between the two drums (GE drawing number 112D1592).

l 6.2.2.1 inner Container Assei'nbly The product containers are normally two 5 gal %n steel pails or three 3 gallon steel pails.

The contents or product container may be contned in a polyethylene bag. The steel Page 2 of 26 l

BU-7 USA /9019/AF December 30,1998 pails and polyethylene bags are used for normal handling of the contents outside the packaging.

6.2.2.2 Inner Container The inner containment drum is a 18 gallon,18 gage minimum to 14 gage maximum, steel drum. The inside diameter of the inner container is 14.05 maximum to 13.75 minimum inches. The height of the drum is 27.25 maximum to 26.75 minimum inches required to accommodate the product containers. An inner lid is bolted to the inner container flange.

A silicone rubber container gasket is the seal between flange and cover.

6.2.2.3 Drum The BU-7 transport package uses a 55-gallon, DOT 17-H, UFC Rule 40, or UN/l A2/X400/S/ YEAR / COUNTRY /MFGX/1.2MM steel drum (22" inside diameter x 35-3/16" inner height, nominal dimensions). The drum body and bottom am fabricated from a 18-gauge (0.0495", nominal) carbon steel or stainless steel sheet. The drum lid is j

fabricated from a 16-gauge (0.%25", nominal) carbon steel sheet. The rernovable head is closed by means of a bolt-locking ring.

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1 Page 3 of 26

BU-7 USA /9019/AF December 30,1998 l

i Table 6.2 Material Specifications Material Density Constituent Atomic density (g/cm')

(atoms /b-cm)

Carbon steel 7.82 C

3.92100E-03 l

Fe 8.34910E-02 1

Phenolic resin 0.12 H

3.01400E-03 B10 0

B11 0

C 2.30500E-03 0

2.05100E-03 j

Si 5.28900E-05 Uranium oxide 10.96 U235 1.23780E-03 U238 2.32200E-02 0

4.89160E-02 Water 1.00 H

6.68660E-02 O

3.34330E-02 l

l Page 4 of 26

BU-7 USA /9019/AF December 30,1998 6.3 CRITICALITY SAFETY ANALYSIS MODELS 6.3.1 GENERAL MODEL 6.3.1.1 Dimensions Figure 6.1 represents the vertical elevations of the single package models as seen along the vertical centerline of the package. A radial cross section through the contents is also displayed in Figure 6.2. The figure dimensions were used in the package calculations.

6.3.1.2 Materials Figures 6.1 and 6.2 show cross section of the single-package model use in the analysis.

Table 6.3 identifies the region materials, material densities, and masses as used in the calculations, and the actual masses typical of the package. The water materialis included in the model and density varied to determine the effect of reflection at the boundary or interspersed moderation.

Table 6.3 Material Specifications Material No. Material Density (g/cm3)

Model mass (kg) Actual mass (kg) 0 VOID 1

UO 10.96 15.0 0.0 - 15.0 2

2 Water 0.00 - 1.00 0-70 0

3 Carbon steel 7.82 24.5 42.2 4

Water 1.00 0-118 0

5 Phenolic 0.32 ( 20 lb/cu. ft.)

0 5.7 resin 0.11 ( 7-9 lb/cu. ft.)

15.1 14.7-18.9 6.3.1.3 Models - Actual Package Differences The single package model of the BU-7 differs from the actual drum in the contents, l

msulating foam, and inner container and drura dimensions.

Contents A maximum quantity of uranium is used in the model. The chemical and physical form of the uranium in the model have most reactive credible for loading originating from a low enrichment uranium fuel fabrication facility. Actual contents will generally be less i

I Page 5 of 26

l l

BU 7 USA /9019/AF December 30 1998 0

l than the maximum quantity allowed and the material may consist of uranium oxides, uranium fluorides, and uranium nitrate compounds mixed in a waste matrix ofincinerated material, residues, or dirt.

The contents is normally loaded in 3 gallon or 5 gallon steel pails used to handle material

. outside the packaging. The model does not represent the steel pail or any polyethylene bag material that is actually loaded into the inner container. The contents modelis uranium material moderated with water. An optimum moderation condition represents the effect of any other credible packaging materials or waste matrix material.

Representing credible moderator as water is conservative because actual packaging materials or waste matrix material are less effective neutron moderators or more effective

- neutron absorbers.

The actual physical form of the material may vary from irregular shape granules to uniform cylinder shape pellets. The material model is a uniform cylinder of uranium oxide in a cylinder shape rod. Other model geometry for the uranium material, such as sphere shape granules, may be more reactive. However, it is considered sufficiently conservative to represent actual contents as a uniform lattice arrangement of cylinder shape rods.

Insulatine Foam The actual insulating foam is phenolic foam that is fire retardant. The space between the inner container and drum is filled with phenolic foam 7-9 lb. / cu. ft. A removable insulation plug placed between the inner lid and drum lid is phenclic foam 20 lb. / cu. ft.

minimum density. All insulating foam, including the insulation plug,is represented by a material composition that is a minimum 7 lb. / cu. ft. Actual phenolic foam contains Boron, but any this neutron absorber is omitted from the material composition used in the model.

Representing the insulatiug foam as a minimum material density accounts for normal fabrication variance in density. Actual hypothetical accident tests resulted in some

. charring of the insulation plug and insulating foam. Any degradation observed during the accident tests is considered in the accident array model by replacing the insulating foam with optimum interspersed hydrogenous moderation.

55-eallon drum and inner container The nominal drum dimensions are use to model the geometry of the drum. In the.odel, the drum wall is a straight wall cylinder without the rolling hoops. The drum model also does not include the top and bottom drum inset into the drum wall, bolts, and locking rings. The carbon steel drum wall, bottom, and lid have also been reduced by 12.5% to allow for manufacturing tolerance on gauge thickness. Reducing amount carbon steel from the model is conservative because there is less material to compete with the uranium for neutron absorption reactions.

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d t

Page 6 of 26

-... ~

l BU-7 USA /9019/AF December 30 1998 0

6.3.2 CONTENTS MODEL l

Figure 6.1 and 6.2 show the package contents configured for the normal transport l

l condition single-package and package array calculations.

Solid UO2 al a theoretical density of 10.96 g/cc is used to represent scrap material that contains uranium. The mass of UO2 is limited to a constant value of 15.0 kg. The uranium material model is UO2 is in cylinder shaped rods oriented perpendicular to the vertical axis of the drum packaging.

Each rod is enclosed in a cuboid shape cell with dimensions that determine the pitch of l

the UO2 rods. Moderation of the UO2 is defined by a water-to-fuel volume ratio (WTOF) that is the ratio of the volume of water in a unit fuel cell and the volume of the UO2. The water-to-fuel ratio is defined by the pitch and UO2 rod diameter, D, as follows:

PITCH * "

WTOF =

A 4 x PITCH'

=

xD' KD 2

A j

The pitch is calculated for the required water-to-fuel ratio. The UO2 rods are arranged in an array with the cylinder axis oriented along the y-axis. The number of UO2 rods arranged along the x-axis is an integral number determined by dividing the diameter of the inner containment by the pitch.

l The height of the UO2 material is determined by calculating an average UO2 compound density for a given WTOF ratio. The average density, p.g. is defined as follows, assuming a water density of 1.00 g/cc:

i WTOF + 10.96 1 + WTOF The height of the UO2 rod array in water inside the inner containment is defined as follows:

i MASS (UO2)

HEIGHT =

, where p,,(l - WTFRu2o)xRL,,

comfamntergt i

WTOF l

WTFR u2o = WTOF+ 10.96 This calculated height of the UO2 rod an ay is divided by the pitch to determine an integer number of UO2 rods required in the vertical z-axis for the required mass of UO2. ~he l

height used in the model is the integer number of UO2 rods times the pitch.

1 Page 7 of 26

BU-7 USA /9019/AF December 30 1998 2

6.3.3 SINGLE PACKAGES To meet the general requirements for fissile material packages,10 CFR 71.55, a package must be designed and its contents so limited that it would be suberitical under the most reacty; configuration of the material, optimum moderation, and close reflection of the containment system by water on all sides or surrounding materials of the ekaging.

The gap region between UO2 rods is completely flooded with water, as is the void region in the inner container. Full density water is placed in bo:L regions because the pitch and diameter of UO2 rods in the fissile content of the package is varied to determine the most reactive configuration. Full density water composition is added to the insulation foam material composition to consider both reflective conditions.

The package was subjected to the tests specifed in 10 CFR 71.71, Nonnal conditions of transport, and the gcometric form of the package was not substantially altered, no water leakage into the containment occurred, and no substantial reduction in the effectiveness of the packaging was observed. The damage incurred will not affect the technical evaluation, and the package contents under normal conditions of transport no more reactive thari the contents under the general requirements.

To address the requirement of 10 CFR 71.55(e), a single package is considered with optimum internal moderation and a least a 30 cm water reflector on all sides. The damaged package experienced less than 5 % reduction in diameter due to impact testing.

Any packaging diameter reduciion is not considered in the analysis. Limited packaging material loss occurred as result of fire testing. The insulation plug exhibited extensive charring during the fire test and there was some charring of the insulation foam between th inner containment system and 55 gallon drum. This region was modeled as insulation foam reduced to one-third the normal density and full density water. The water from the immersion test is assumed to optimally moderate and fill any void space in the inner container. The minimal damage resulting from the crush and puncture tests will not influence he reactivity of the package. The damage incurred will not affect the technical evaluation, and the package contents under Hypothetical Accident Conditions of transport will be no more reactive than the conten+s undec &c Normal Conditions of Transport or general requirements.

6.3.4 PACKAGE ARRAYS Cylinder transport packages such as the BU-7 may be shipped in a tightly packed triangular array or may be shifted to that configuration because of hypothetical accident conditions. This arrangement is more reactive configuration than a square-pitch arrangement because the triangular pitch provides absolute minimum center-to-center spacing of the fissile contents, the maximum density of fissile units, and thus the greatest potential for increased neutron interaction between fissile contents. The triangular array model is developed by embedding the single package model in a repeating pattern that produces a tightly packed triangular array. An infinite array model represented by a Page 8 of 26

BU-7 USA /9019/AF December 30c 1998 single package with mirror reflection on all sides. The infinite array is a square lattice arrangement, but is used only to determine the effect of interspersed moderation.

Two array models are evaluated, one for condition of the single package for normal transpoit conditicas and another for hypothetical accident conditions. The model consists of a finite array of close-packed, triangular-pitch, single packages consistent with the normal conditions of transport. From 10 CFR 71.59, standartis for arrays of fissile c

material packages, undamaged package arrays are evaluated with void between the packages. As required by 10 CFR 71.59, the damaged packages must be evaluated as if each package was subjected tot he tests specified in 10 CFR 71.73, Hypothetical Accident Condit*ans, with optimum interspersed hydrogenous moderation. Funher, the finite array of packages is reflected by 31 cm of water on all sides.

Various finite array sizes with dimensions to minimize neutron leakage at the boundaries were investigated in ortler to ascertain the number of subcrtical packages under normal transport conditions and hypothetical accident conditions. The condition of each package in the array is that described in Section 6.3.3 for the single packcge, except that the void space in the containment is not filled with water to allow credible interaction between the packages. Also, the foam insulation packaging is not included in the accident anay packages to allow credible interaction between packagi.ig subject to hypothetical accident conditions.

Page 9 of 26

BU.7 USA /9019/AF December 30,1998

)

VOID SPACE. MATERIAL 2. 0.9525 cm COVER, MATERIAL 3. 0.1087 cm L

a...
u lNSULATION PLUG VOID SPACE

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MATERIAL 5 MATERIAL 0 OR MATERIAL 2 40

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5.080 5.687 4.....

MATERIAL 3 l

0.4763 cm l

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. MATERIAL 5 10.767 cm VOID SPACE MATERIAL 0 OR MATEIAL 2 1

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MATERIAL 1 AND MATERIAL 4 1,ct

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MATERIAL S MATERIAL 3 8

l 7.62 cm 0.1087 cm 40

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-20 MATERIAL 2 0

20 1.905 cm Figure 6.1 Axial cross section of BU.7 package 4

Page 10 of 26

BU-7 USA /9019/AF December 30,1998 t'

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10 20 Figun: 6.2 Radial cross section (A-A) of BU-7 package contents Page 11 of 26

BU-7 USA /9019/AF December 30 1998 0

6.4 METHOD OF ANALYSIS GEMER, a proprietary General Electric company standard criticality analysis computer codes was used in the analysis of these computational models. All calculations were performed using Pentium processors running under Windows 95 or Windows NT.

6.4.1 COMPUTER CODE SYSTEM GEMER is a Monte Carlo program which solves the neutron transport equation as an eigenvalue or a fixed source problem including the neutron shielding problem. GEMER adds an advanced geometry input package to the problem solving capability of the Monte Carlo code which is very similar to KENO. (Refemnce 1) 6.4.2 CROSS SECTIONS AND CROSS-SECTION PROCESSING GEMER uses cross sections processed from the ENDF/B-IV library tapes. These cross section are prepared in 190 group format and those in the resonance region may have the form of the resonance parameters or Doppler broadened multigroup cross section.

Thermal scattering of hydrogen in water is represented by the S(cz, ) data in the ENDF/B-IV library. The types of reactions considered in the Monte Carlo calculation are fission, elastic, inelastic, and (n,2n) reactions; the absorption is implicitly treated by reducing the neutron weight by the non-absorption probability on each collision.

6.4.3 CODE INPUT l

All problems were started with a flat initial neutron distribution over the fissile material l

regions only. Calculations were nominally run with 65 or 105 generations at 500 or 2000 neutrons each, skipping the first 5 generations before starting the statistical output processing, for a total of 30,000 or 200,000 histories. The smaller batch size is used for single package calculations and the larger batch size is used for array calculations.

Figures 6.3(a),6.3(b),6.3(c), and 6.3(d) used represent the input files. The files shown in figures 6.3(a) corresponds to cases describing a single-package, and figun: 6.3(c)

{

corresponds to cases describing package arrays. Figum 6.3(b) lists the parameter values i

to describe the contents, and figure 6.3(d) lists the parameter values for describing the l

arrays. Material numbers correspond to the materials described in Table 6.3.

i i

Page 12 of 26 i

)

1 I

\\

- BU-7. USA /9019/AF December 30,1998 KENO CEOM 0 /* # OF REGIONS OR ZERO 01 /* # OF BOX TYPES OR ZERO 1' /* # OF BOXES IN X DIRECTION 1' /* # OF BOXES.IN Y DIRECTION 11 /* # OF BOXES IN Z DIRECTION O /* BOUNDARY CONDITION OPTION l

0. /* STARTING SOURCE OPTION

.l /* COMPLEX EMBEDDED OPTION O. /* # OF PRINr PI4TS j

BOX TYPE

.1'

/* LOWER PORTION OF BU-7~

CYLINDER 3-17.8080 -35.1000 -35.2087 16*.5 CYLINDER 5 28.5750 -35.1000 -42.8287 16*,5 CYLINDER' 3.28.5755 -35.1000 -42.9374 16*.5 l

- CYLINDER

-2 28.5760 -35.1000 -44.8424 16*.5 l

CYLINDER 3: 28.6840 -35.1000 -44.8424 16*.5 l

. BOX TYPE.

2' /* SINGLE P"-7 CYLINDER

.2 28.6850 45.3692 444.8424 16*.5 CUBOID 4~ 28.6850 -28.6850 28.6850 -28.6850 45.3692 -44.8424 ~ 16*.5 BOX TYPE

-3 ~/* UPPER PORTION OF BU-7 CYLINDER 0- 17.7000 35.1000 -35.1-HEIGHT 16*.5 1

[

' CYLINDER 3 17.8080 35.5763 -35.1-HEIGHT 16*.5

, CYLINDER

.5,23.4950 35.5763 -35.1-HEIGHT 16*.5

-- CYLINDER

'2 23.4955-36.G880 -35.1-HEIGHT 16*.5 CYLINDER 5 23.5750 44.3080 -35.1-HEIGirr 16*.5 1

CYLINDER-3 28.5755 44.4167 -35.1-HEIGHT 16*.5 CYLINDER-

'2 28.5760 45.3692 -35.1-HEIGHT-16*.5

[-

CYLINDER 3 28.6840 45.3692 -35.1-HEIGHT 16*.5 BOX TYPE 4 /*' FUEL CYLINDER CELL-WATER-TO-FUEL =12.00 YCYLINDER 1 RADIUS 17.7000 -17.7000 16*.5 i.

- CUBOID 4-PITCH /2 -PITCH /2 17.7000 -17 7000 PITCH /2 -PITCH /2 16*.5

!=

BOX TYPE' 5 /* IlfrER DISK FUEL CYL-NO BORAL CYLINDER

'4 17.7000 PITCH 0.00000 16*.5 1

l CYLINDER 3 17.8080 PITCH 00 0.00000 16*.5 l=

CUBOID 5 17.8080 -17.8080 ' 17.8080 -17.8080 ' PI'ICH 0.00000 16*.5 j'

BOX TYPE.

6 /* INTER DISK FUEL CYL-NO BORAL l

CUBOID 5 '17.8080 -17.8080 17.8080 -17.8080 HEIGHT 0.0000 16*.5

CYLINEER

-5 28.5750- HEIGHT 0.0000 16*.5 CYLINDER 3 28 '8140 HEIGHT 0.0000 16*.5 4

l.

CORE O. 28 e J -78.6850 28.6850 -28.6850 45.3692 -44.8424 16*.5 i

CUBOID.

4 Ss

'e.6850 59.6850 -59.6850 76.3692 -75.8424 16*.5 2

1 1 1 1 1 1 1 1:1 1

BEGIN COMPLEX

' COMPLEX 5 4 -17.7000 0.00000.0.30400 NX' 1

1 PITCH 0.0 0.0 COMPLEX 6-5 0.00000 0.00000 0.00000 1

1 NZ 0.0 0.0, PITCH

/* PACKAGING AND CONTENTS COMPLEX 2"1 0.00000 0.00000 0.00000 1

1 1

0.0 0.0 0.0 f

COMPLEX 2 6 0.00000 0.00000 -35.1000 1

1 1

0.0 0.0 0.0 l:

' COMPLEX 2 3 0.00000 0.00000 0.00000I i 1

1 0.0 0.0 0.0

(

END CEOM l

. DEFAULTS =YES END GEMER

!-~

l i

l F:

I i

l 1

l J

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f Figure 6.3(a)- BU-7 single-package 1

Page'13 of 26

' BU-7 USA /9019/AF -

December 30 1998 0

WTOP WTFR RHOAVG PITCH CALCULATED NZ NX HEIGHT HEIGHT a/cc em em em RADTUS=0.06350 (0.050 inch rod diameter) j 7

0.39 2.25 0.3183 11.124 35 111 11.1405

)

9 0.45 2.00 0.3559 13.905 40 99 14.2360 11 0.50 1.83 0.3899 16.687 43 91 16.7657 13

.0.55 1.71 0.4211 19.468 47 84 19.7917

- 16 0.60 1.59 0.4641 23.639 51 76 23.6691 20 -

0.65 1.47 0.5158 29.201 57 69 29.4006 RADIUS =0.09530 (0.075 inch rod diameter) 7 0.39 2.25 0.4775 11.124 24.

74 11.4500 9

0.45 2.00 0.5339 13.905 27 66 14.4153 11

.0.50 1.83 0.5848 16.687 29 61 if.9592 13 0.55 1.71 0.6317 19.468 31 56 19.5827 16 0.60 1.59 0.6961 23.639 34 51 23.6674 20 0.65 1.47 0.7737 29.201 38 46 29.4006 RADIUS =0.12700 (0.100 inch rod diameter) 7 0.39 2.25 0.6367 11.124 18 56 11.4606 9

-0.45 2.00 0.7118 13.905 20 50 14.2360 11 -

0.50 1.83 0.7798 16.687 22 45 17.1556 13

.0.55 1.71 0.8423 19.468 24 42 20.2152 16 0.60 1,59 0.9281 23.639 26 38 24.1306 20 0.65' 1.47 1.0315 29.201-29 34 29.9135 RADIUS =0.15880 (0.125 inch rod diameter) 7 0.39 2.25 0.7959 11.124 14 44 11.1426 9

0.45

.2.00 0.8898 13.905 16 40 14.2368 11 0.50 1.83 0.9747 16.687 18 36 17.5446 4

13 0.55 1.71 1.0528 19.468 19 34 20.0032 16 0.60 1.59 1.1601 23.639 21 31 24.3621 l

20 0.65

'1.47 1.2894 29.201 23 27 29.6562 I

  • .1DIUS=0.19050 (0.150 inch rod diameter) 7 0.39 2.25 0.3550 11.124 12 37 11.4600 9

0.45 2.00 1.0678 13.905 14 33 14.9492 11 0.50-1.83 1.1697 16.687 15 30 17.5455 13 0.55 1.71 1.2634 19.468 16 28 20.2144 16 0.60 1.59 1.3922 23.639 17 25 23.6674

)

20 0.65 1.47 1.5473 29.201 19 23 29.3987 i

RADTUS=0.2223 (0.175 inch rod diameter) 7 0.39 2.25 1.1142 11.124 10 32 11.1420 9

0.45 2.00 1.2457 13.905 12 28 14.9484 11 0.50

?.83 1.3646 16.687 13 26 17.7398 13 0.55 1.71 1,4739 19.468 14 24 20.6346 16 0.60 1.59 1.6242 23.639 15 22 24.3630 20 0.65 1.47 1.8052 29.201 17 20 10.6884

.1 -

I i

l i

l Figure 6.3(b) BU-7 single. package contents parameters (a0152F??.in)

]

I 1

i Page 14 of 26

l J

EU-7 USA /9019/AF.

December 30,1998 1

[_

KENO GEOM

'O

/* # OF P.EGIONS OR ZERO O'

/* # OF BOX TYPES OR ZERO

1. /* # OF BOXES IN X DIRECTION

.:1-~/* # OF BOXES IN Y. DIRECTION-NZ /* # OF BOXES IN Z DIRECTION O /* BOUNDARY CONDITION OPTION O /* STARTING SOURCE OPTION 1- /* COMPLEX EMBEDDED OPTION i

t 0 /* # OF PRI!fr PIOTS BOX TYPE.

1 /* LOWER PORTION OF BU-7 CYLINDER-3 17.8080 -35.1000 -35 1087 16*.5 OYLINDER 5 28.5750 -35.1000 -42.8287 16*.5 CYLINDER 3

28.5'455'-35.1000 -42.9374 16*.5 CYLINDER 2 28.5760.-35.1000 -44.8424 16*.5 CYLINDER 3 28.6840 -35.1000 -4f.8424 16*.5 BOX TYPE

-2'

/* SINGLE BU-7 C1(LINDER 2 28.68b0 45.3692'-44.8424 16* 5 i

BOX TYPE 3 /* UPPER PORTION OF BU-7

)

CYLINDER 0 17.7000 35.1000 -16.8570 16*.5 CYLINDER 3 17.8080 35.5763 -16.8570 16*.5 CYLINDER 5 23.4950 35.5763 -16.8570 16*.5 CYLINDER 2 23.4955 36.6880 -16.8570 16*.5 CYLINDER 5 28.5750 44.3080 -16.8570 16*.5 CYLINDER.

3 28.5755 44.4167 -16.8570 16*.5 CYLINDER 2 28.5760 45.3692 -16.8570.

16*.5 CYLINDER-

3. 28.6840 45.3692 -16.8570 16*.5

. BOX TYPE 4 /

  • FUEL CYLINDER CELL-WATER-TO-FUELS W'IDF l

YCYLINDER l' O.09530 17.7000 -17.7000 16*.5 l

CUBOID 4 0.30436 -0.30436 17.7000 -17.7000 0.30436 -0.30436 16*.5

' BOX TYPE 5 /* INTER DISK FUEL CYL-NO BORAL CYLINDER 4 17.7000 0.60870 0.00000 16*.5 CYLINDER 3 17.8080 PITCH 0.00000 16*.5 l

CUBOID 5 17.8080 -17.8080 17.8080 -17.8080 0.60870 0.00000 16*.5 BOX TYPE 6 /* INTER DISK FUEL CYL-NO BORAL CUBOTO 5 17.8080 -17.8080 17.8080 -17.8080 15.1944 0.00000 16*.5 i

CYLINDER -

5 28.5750 15,1944 0.00000 16*.5 CYLINDER-3 28.6840 15.1944 0.00000 16*.5 BOX TYPE.

7 /* PROBLEM BOX FOR TRIANGULAR ARRAY - ONE LAYER DEEP i

CUBOID 2'X

-X Y

-Y 45.3692 -44.8424 16*.5 CORE 0.X

-X Y

-Y Z

-Z 16*.5

' CUBOID 4 X+31

-X-31 Y+31

-Y-31 Z+31

-Z-31 16*.5 7

1 1 1 1 1 1 1 NZ 1 1

BEGIN COMPLEX,

COMPLEX 5 4 -17.7000 0.00000 0.60870 58-l' 1 0.6087 0,0 0.0 l

COMPLEX 6 5 0.00000 0.00000 0.00000 1

1 36 0.0 0.0 0.6087

/* PACKAGING AND CONTEPTTS COMPLEX-. 2 1 0.00000 0.00000 0.00000 1

1 1

0.0 0.0 0.0 i

i COMPLEX 2 6 0.00000 0.00000 -35.1000 1

1 1

0.0 0.0 0.0 f

-COMPLEX 2 3 0.00000 0.00000 0.00000 1

1 1

0.0 0.0 0.0

/* TRIANGULAR ARRAY COMPLEX 7 2 -387.248 -373.125 0.00000 NX NY 1 57.3700 99.5000 0.0 COMPLEX 7 2 -358.563 -323.375 0.00000 NX NY 1 57.3700 99.5000 0.0 END GEOM DEFAULTSsYES L

END GENER L

L t-Figure 6.3(c) BU-7 package a Tay input file for single package with 0.075 h h UO2 rod diameter, WTOF=12 contents (a7152007.in) l r.

.5 Page 15 of 26

__m BU-7 USA /9019/AF December 30 1998 0

NZ /* # OF BOXES IN Z DIRECTION

-ARRAY SIZE LENGTH WIDTH HEIGHT NZ cm em em 4x 5x3 258.165 256.370

'270.635 3

7x 9x4 430.275 455.370 360.846 4

]

7x 9x5 430.275 455.370 451.058 5

9 x 11 x 6 545.015 554.870 541.270 6

i 11 x 13 x 7 659.755 654.370 631.481 7

-12 x 14 x 8 717.215 704.120 721.693 8

14 x 16 x 9 831.865 803.620 811.904 9

]

l BOX TYPE 7 /* PROBLEM BOX FOR TRIANGULAR ARRAY - ONE LAYER DEEP CUBOID 2

X

-X Y

-Y 45.3692

-44.8424 16*.5 CORE O

X

-X Y

-Y Z

-Z 16*.5 4x 5x3 129.083

-129.083 128.185

-128.185 136.108

-134.527 7 x -9 x 4 215.138

-215.138 227.685

-227.685 191.477

-179.370 7x 9x5 2'5.138

-215.138 227.685

-227.685 226.846

-224.212 1

9 x 11 x 6 2/2.508

-272.508 277.435

-277.435 272.215

-269.054 11 x 13 x 7 329.878

-329.878 327.185

-327.185 317.584

-313.897 12 x 14 x e 358.563

-358.563 352.060

-352.060 362.954

-358.739 14 x 16 x 9 415.933

-415.933 401.810

-401.810 408.323

-403.582

/* TRIANGULAR ARRAY i

COMPLEX 7

2 X

Y 0.00000 NX NY 1

57.3700 99.5000 0.0 1

4x 5x3

-100.398

-99.500 4

3 7x 9x4

-186.453

-199.000 7

5 7x 9x5

-186.453

-199.000 7

5 9 x 11 x 6

-243.823

-248.750 9

6 11 x 13 x 7

-301.193

-298.500 11 7

12 x 14 x 8

-329.878

-323.375 12 7

14 x 16 x 9

-387.248

-373.125 14 8

COMPLEX 7

2 X

Y 0.00000 NX NY 1

57.3700 99.5000 0.0 4x 5x3

-71.713

-49.750 4

2 7x 9x4

-157.768

-149.250 7

4 i

7x 9x5

-157.766

-149.250 7

4 9 x 11 x 6

-215.138

-199.000 9

5 11 x 13 x 7

-272.508

-248.750 11 6

12 x 14 x 8

-301.193

~273.625 12 7

14 x 16 x 9

-358.563

-323.375 14 8

l i

?

Figure 6.3(d). BU-7 package array parameters for triangular array 4

,1.

1 i

t 1

4 1

4 Page 16 of 26

BU-7 USA /9019/AF December 30,1998 i

l 6.4.4 CONVERGENCE OF CALCULATION l

Problem convergence was determined by examining plots of k,gby generation run and skipped, as well as the final keff edit tables. No abnormal trends were observed to indicate non-convergence of the eigenvalue solution. A sample convergence plot for the l

. a package array case a4152007 shown in Figure 6.4 is representative of the convergence evaluations. A k,g is determined as the maximum value of k,p(i) minus 3c(i) where i equals the number of batches skipped.

i 1.11 LEGEND L

e A4152007 1.07 l

1.03 e

0.990 K-EFF i3 er 0.950 llkh

(({o"oooo no no no o

O.910 l

8.870 i

l 8.830

-a 6

40 80 120 160 200 240 I

BATCHES / BATCHES SKIPPED X10

't 1

e Figure 6.4 Sample k,gconvergence plot, case a4152007 Page 17 of 26 i

BU-7 USA /9019/AF December 30,1998 6.5 VALIDATION Validation of GEMER consists of performing calculation of benchmark experiments including the area of applicable to the uranium oxides. Bias for GEMER and the ENDF/B-IV library has been established for the area of applicability for the BU-7 package. The bias determined is no greater than 0.0246 (&,- ) at a 99% confidence lew 1. The area of applicability for the 39 benchmark calculations are enrichment ranges from 2.3 h 7.00 weight percent U-235, H/U-235 ratio 93 to 400, and square lattice UO2 metal. The.naximum allowed mass UO2 for BU-7 scrap content is calculated in the II/U235 ratio 400 to 800. The benchmark calculations for uranium oxide material in the l

H/U235 ratios from 400 to 800, but not square lattice UO2, have a bias that is less than the square lattice benchmark calculation in the lower H/U235 area of applicability. The larger bias is applied to account the physical form of the uranium material in the model use to determine the mass limit. (Reference 2)

Using a general equetion for the upper safety limit (USL) and requirements of 10 CFR 71, calculations are considered suberitical,if the following condition is satisfied:

keg + 2a S0.9500 - &, + p kg + 2a S0.9254 i

Page 18 of 26

BU-7 USA /9019/AF December 30 1998 0

6.6 CRITICALITY CALCULATIONS AND RESULTS This evaluation demonstrates the suberiticality of a single package (Section 6.6.1) and an anay of packages (Section 6.6.2) during nonnal conditions of transport and hypothetical accident conditions. The determined trarsport index (TI) for criticality control of damaged and undamaged shipment is given in Section 6.6.3.

6.6.1 SINGLE PACKAGE Calculations show that a single package remains suberitical under general requi'ements for fissile material packages. To meet the general requirements for fissile material packages,10 CFR 71.55, a package must be designed and its contents so limited that it would be subcritical under the most reactive configuration of material, optimum moderation, and close reflection of the containment system by water on all sides or surrounding materials of the packaging.

Calculations demonstrate safety under the general requirements by considering evaluating the effect of varying the physical fonn from a homogeneous mixture of UO2 and water through a range of solid fuel rod diameters that results in a maximum reactivity. The maximum reactivity for any physical form of the fuel occurs at a moderation equivalent to about 50 to 55 weight pen:ent water ( l I to 13 water-to-fuel ratio or 600 to 700 H/U235),

because the mass is limited to 15 kg of uranimn oxide. The maximum effect of the physical form of the material estimated to be abo st 0.02 Ak,g over a range of moderation that results in a maximum reactivity for this mass limit. The maximum k,g value does is not differ significantly between 45 to 55 weight percent moderation. Figure 6.5 shows that the most reactive contents is a fuel rod diameter in the range of 0.05 to 0.125 inch (95 to 320 mm). The maximum k,g value does is not significantly different over in the 45 to 55 weight percent moderation.

Case a0153f07 of Table 6.4 represents the optimally moderated containment reflected on all sides by the container reflection provided by surrounding materials of the packaging and 31 cm of water. The contents for this case 4 defined by a fuel rod diameter of 0.075 l

inch tnd spacing on a pitch for a water-to-fuci ratio of 12. The maximum cingle package keff of 0.797 0.005 is considered subcritical, that is,0.797 + 2 x 0.005 = 0.80 < 0.92, i

The calculations for the general requirements also demonstrate safety under Normal Conditions of Transport and under Hypothetical Accident Conditions, because the contents is assumed to be moderated to the most reactive credible extent for both conditions and close reflection by water and insulating foam on all sides is more reactive than reflection by packaging alone. All calculations were performed at the maximum 3

allowable U235 enrichment (5.00 wt %) to ensure maximum reactivity and eliminate the need for calculations at lower possible enrichments.

Page 19 of 26

J l

BU-7 USA /9019/AF Deceanhcr 30 1998 0

)

- Table 6.4 BU-7 Single Package Caleclation Summary, (15 kg UO2 and close reflection by water on all sides)

Case Description keff a 1

a01540fh Homogeneous UO2-H2O (40 wt %)

0.73823 0.00439 a01545fh UO2-H2O (45 wt %)

0.76089 0.00450 a01550fh UO2-H2O (50 wt %)

0.77920 0.00382 a01555fh UO2-H2O (55 wt %)

0.79607 0.00459 a01560fh UO2-H2O (60 wt %)

0.79324

  • 0.00410 a01565fh UO2-H2O (65 wt %)

0.76739 i 0.00329 1

a0157f05 0.050 inch diameter UO2-H2O (WTOF=7) 0.72315

  • 0.00397 a0159f05 UO2-H2O (WTOF=9) 0.77876 1 0.00449 a0151f05 UO2-H2O (WTOF=11) 0.79353
  • 0.00488 a0153f05 UO2-H2O (WTOF=13) 0.80397 0.00319 a0156105 UO2-H2O (WTOF=16) 0.79563
  • 0.00364 a0152005-UO2-H2O (WTOF=20) 0.78469
  • 0.00500 a0157107 0.075 inch diameter UO2-H2O (WTOF=7) 0.73477 2 0.00490 a0159f07 UO2-H2O (WTOF=9) 0.76699 0.00408 a015!!07 UO2-H2O (WTOF=11) 0.79421
  • 0.00383 i

a0153f07 UO2-H2O (WTOF=13) 0.79672 1 0.00460 a0156f07 UO2-H2O (WTOF=16) 0.79250 1 0.00371 a0152007 UO2-H2O (WTOF=20) 0.78347

  • 0.00493 a0157f10 0.100 inch diameter UO2-H2O (WTOF=7) 0.73477
  • 0.00490 j

a0159f10 UO2-H2O (WTOF=9) 0.75366

  • 0.00454 a0151f10 UO2-H2O (WTOF=11) 0.79199 0.00391 a0153fl0 UO2-H2O (WTOF=13) 0.79886
  • 0.00384 j

a0156fl0 UO2-H2O (WTOF=16) 0.78410

  • 0.00307 a0152010 UO2-H2O (WTOF=20) 0.72551
  • 0.00406 a0157f12 0.125 inch diameter UO2-H2O (WTOF=7) 0.74693
  • 0.00389 a0159f12 UO2-H2O (WTOF=9) 0.76946
  • 0.00458 a0151fl2 UO2-H2O (WTOF=ll) 0.79696 0.00400 a0153fl2 -

UO2-H2O (WTOF=13) 0.78960 t 0.00349 a0156f12 UO2-H2O (WTOF=16) 0.76702

  • 0.00352 a0152012 UO2-H2O (WTOF=20) 0.75570 1 0.00397 a0157fl5 0.150 inch diameter UO2-H2O (WTOF=7) 0.72776 1 0.00459 a0159f15 UO2-H2O (WTOF=9) 0.78610
  • 0.00486 a0151fl5 UO2-H2O (WTOF=ll) 0.78422 i 0.00411 i

a0153f15 UO2-H2O (WTOF=13) 0.78415

  • 0.00457 a0156fl5 UO2-H2O (WTOF=16) 0.77107 0.00389 a0152015 UO2-H2O (WTGF=20) 0.74072
  • 0.00360 a0157fl7 0.175 inch diameter UO2-H2O (WTOF=7) 0.73941
  • 0.00388 a0159f17 -

UO2-H2O (WTOF=9) 0.77198 1 0.00380 a0151fl7 UO2-H2O (WTOF=ll) 0.78373 0.00469 a0153fl7 UO2-H2O (WTOF=13) 0.76243 0.00395 a0156f17 UO2-H2O (WTOF=16) 0.75394 t 0.00373

r.0152017 -

UO2-H2O (WTOF=20) 0.73123 0.00396 Page 20 of 26 i

t BU-7 USA /9019/AF December 30 1998 0

d 4

i 0.968 2

LEGEND

- MAXIMUN ALLOLIED

)

LINEAR FITS ORDER = 2 9.930 1

4 9.900 e.87e

.K-KFF 12cr j

8. 84 ts 1

i l

l l

9.816

=

- e"

~*

4

)

o o

o g

0.780 e.rse 0

40

-80 120 160 200 2160 3

FUEL ROD DIAMETER X10 Figure 6.5 Single package maximum reactivity of fissile material l

i i

Page 21 of 26

L, BUo7 USA /9019/AF December 30,1998 6.6.2 PACKAGE ARRAYS The calculation results of Table 6.5 show that an finite array of packages is subcritical under nonnal conditions of transport and hypothetical accident conditions. For the normal conditions of transport, case a4152po7, a 12 x 14 x 8 array of undamaged packages, the calculated a k,f s 0.91681 i 0.00220, which is suberitical because 0.9168 i

+ 2 x (0.0022) = 0.9212 is less than the USL of 0.9254. Therefore, five times "M undamaged packages with nothing between the packages that is suberitical is equal to 1344.

Figure 6.6 shows the effect ofinterspersed moderation on the calculated k,g for an infinite array of packages.. The effect of the insulating foam is equivalent to about 0.05 fraction of full density water interspersed between the packages. This is consistent with the hydrogen content of the ir sulation roam that is about 0.05 times the hydrogen content of full density water. Optimum interspersed hydrogenous moderation is occurs when there is no insulating foam and ncihing herween the packages.

)

For hypothetical accident condir. ion, case a2152007, a 9 x 11 x 6 array of damaged packages, the calculated k,gis 0.9057610.00233, which is suberitical because 0.9058 +

)

2 x (0.0024) = 0.9106 is less than the USL of 0.9254. Insulating foam that is part of the L

packaging material decreases the interaction between single packages. Therefore, two 4

times "N" damaged packages that is suberit.ical with optimum interspersed hydrogenous I

moderation is equal to 594, i

Figure 6.7 shows the effect of array size on the calculated keg for both the norrnal condition of transport and the calculated hypcthetical accident condition.

J 4

L Page 22 of 26

BU-7 USA /9019/AF December 30 1998 0

Table 6.5 Results for package _ array calculations _

Case Interspersed Description keff

  • c H2O density (g/cm3)

Normal Conditions of Transport all52po7 0.0 7 x 9 x 5 array 0.8647110>00.208

- r2152po7 0.0 9 x'11 x 6 array 0.89757 i 0.00236 a3152po?

0.0 11 x 13 x 7 array 0.91182 i 0.00208 adl52po7 0.0 12 x 14 x 8 array 0.91681 i 0.00220 a5152po7 0.0 14 x 16 x 9 array 0.93358 i 0.00133 Hypothetical AccidentCondtion ail 52007 0.0 infinite array 1.15713 i 0.00202 i

ail 52027

'0.025 infinite airay 1.01491 i 0.00253 ail 51057 0.05 infinite array 0.93684 1 0.00258 ail 52po7 Phenolic foam infinite array 0.99508 f 0.00437 ail 52077 0.075 infinite array 0.89089 i 0.00314 ail 52107 0.10 in. finite array 0.86133 i 0.00297 ail 52207 0.20 infinite array 0.80178 1 0.00339 ail 52507 0.30 infinite array 0.77527 i 0.00359 ail 52507 0.50 infinite array 0.76144 i 0.00383 ail 52807 0.80 infinite array 0.75834 i 0.00348 4

ail 52fd7 1.00 infinite array 0.75699 i 0.00333 0

all52007 0.0 7 x 9 x 5 array 0.87910 i 0.00435 p

a2152007 0.0 9 x 11 x 6 array 0.90576 1 0.00233 a3152007 0.0 11 x 13 x 7 array 0.93723 i o.00252 a4152007 0.0 12 x 14 x 8 array 0.95685 1 0.00209 a5152007 0.0 14 x 16 x 9 array 0.98019 i 0.00201 1

1 3

(

Page 23 of 26

BU-7 USA /9019/AF December 30,1998 1.40 LEGEND N0DERATOR e WATER x INSULATING FOAM 1.30

=

1.20

? >

1.10 K-EFF i3a 1...

1 9.900 8.800 I

i t

e.7ee

-se se Se se 7e' 9e axe WEIGHT FRACTION HATER X10 Figure 6.6 Infinite package array k,g vs. Interspersed moderation Page 24 of 26

. _. ~ _ _.

l BU-7 USA /9019/AF December 30,1998 l

e.998

+-. -

LEGEND k,,,f = 1.16

'~

CONTENT URANIUM MASS

):

1 e NORMAL CONDITION l

x ACCIDENT CONDITION l

8.970 HAXINUM ALLOWED 1

t u

I e.950 k,,,f = 1.0!

n o

0.930 i

K-EFF 12cr w.

s ri o l

e.91e

[

i o

i 8.898 i

l 2N = 600 SN = 1250 Tl=0.2 l

e.s7e i

l 1

e.050 30 70 110 150 190 230 ARRAY SIZE X10

~

\\

l J

4

!t i

j Figun: 6.7 Package array k,g vs. array size Page 25 of 26 5

a BU-7 USA /9019/AF December 30 1998 2

6.7 TRANSPORTATIONINDEX The transportation index (TI) for cr!*icality control is determined by the number of packages that remain below the upper safety limit. For normal conditions of transport, a finite array of five times 268 packages is suberitical. Under hypothetical accident conditions, two times 297 packages is suberitical. Thus a maximum of 268 packages may be shipped for a nonexclusive shipment, and a TI = 0.2 is assigned to tl.c BU-7 package with a 13.22 kg uraniam content.

REFERENCES i

1. Taylor, John T., "GEMER, Microcomputer Version User Guide", April 20,1994,
2. Peters, William C.," Validation of the GEMER.4 Monte Carlo Code for Applications with Enriched Uranium Fuels at GE Nuclear Fuels in Wilmington, North Carolina,"

December 8,1997.

t 4

Page 26 of 26

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