ML20217N306

Criticality Safety Review of 2 1/2-,10- & 14-Ton UF6 Cylinders

ML20217N306
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Site:	07106553
Issue date:	10/31/1991
From:	Broadhead B OAK RIDGE NATIONAL LABORATORY
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ORNL-TM-11947, NUDOCS 9910280171
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Text

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ORNL c a s m,,.,,,

VIASTER COPY OAK RIDGE NATIONAL LABORATORY Criticality Safety Review of 21/2, ammrs===sr=

10, and 14-Ton UFe Cylinders B. L. Broadhead l

1 MANAGED BY MARTIN MARIETTA ENERGY SYSTEMS,1NC.

FDR THE UNITED STATES aR 3

ORNUTM-11947 CRITICALITY SAFETY REVIEW OF 2%,

10, AND 14-TON UF CYLINDERS 6

i B. L. Broadhead Computing and Telecommunications Division at Oak Ridge National Laboratory P.O. Box 2008 Oak Ridge, 'IN 37831-6370 Date Published-October 1991 1

Prepared by the OAK RIDGE NATIONAL LABORATORY Oak Ridge, Tennessee 37831 managed by MARTIN MARIEEA ENERGY SYSTEMS, INC.

for the U.S. DEPARTMENT OF ENERGY under contract DE-AC05-840R21400

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TABLE OF CONTENTS

' LIST O F FIG UR ES.................................................... iv LIST OF TABLES...........................

v ACKNOWLEDGMENTS...............................................

vii 1

AB STRACT.......................................................... ix

1. INTRO D UCTION..................................................

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2. ANALYSIS METHODOLOGY........................................

3 2.1 ANALYSIS TOOLS............................................

3 2.2 VALIDATION STUDIES..........................................

3

- 2.3 ANALYSIS OVERVIEW..........................................

5 2.4 SENSITIVITY STUDIES..........................................

6

3. 2 %. TON CYLINDER ANALYSIS.......................................

7 3.1 MODEL DESCRIPTION..........................................

8 3.2 ANALYSIS RESULTS FOR 2%-TON CYLINDER...................... 12 3.21 Infinite Array Results without Overpacks........................

13 3.2.2 Infinite Array Results with Overpacks...........................

13 3.2.3 Single ' Unit Results.........................................

16 3.2.4 Sensitivity Results..........................................

16 3.2.5 Validation Results 16 3.2.6 S umm a ry................................................

18

4. 10- AND 14-TON CYLINDER ANALYSIS................................

18 4.1 MODEL DESCRIPTION.........................................

18 4.2 ANALYSIS RES'ULTS FOR 10- AND 14-TON CYLINDERS.............

18 4.2.1 Infinite Array Results without Overpacks........................

24 4.2.2 Infinite Array Results with Overpacks...........................

24 4.23 Single Unit Results.........................................

29 4.2.4 Sensitivity h.:sults..................................,........

29 4.2.5 Validation Results 29 4.2.6 S u m m ary................................................

31

5. REFERENCES....................................................

31 til...

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I LIST OF FIGURES l

Figure Page 1.kn versus average energy group causing fission (AEG) for 51 benchmark calculations.

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2. Fuellocations analyzed to determine most reactive configuration for a co nst a nt mass....................................................

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3. Model for 2%. ton cylinder without overpack................................

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4. Model for DOT 21-PF-1 overpack......................................

10 S. Plot of k,y versus water specific gravity for infiaite arrsy of 2%-ton UF6 cylinders (square-lattice, full-diameter model)...........................

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6. Plot of k,g versus water specific gravity for infinite array of 2%-ton UF6

)

cylinders for original and 7% reduced pitch case.........................

15

7. Model for 10. ton cylinder without overpack..............................

20 l

8. Model for 14. ton cylinder without overpack...............................

21

)

9. Model for Paducah Tiger overpack......................................

22

10. Plot of k,y versus water specific gravity for infinite array of 14-ton UF6 cylinders (square-lattice, full. diameter model)...........................

25

11. Plot of k n versus water specific gravity for infinite array of 14-ton UF e

6 cylinders for original and 7% reduced pitch case.........................

26

12. Plot of k,y versus water specific gravity for infinite arrays of 10- and 6

14-ton UF cylinders (square-lattice, full-diameter models)..................

27

13. Plot of k,g versus water specific gravity for infinite array of 10-ton UF6 cylinders for original and 7% reduced pitch case.........................

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1

.iv

LIST OF TABLES i

Table Page i

1. Constituent material mass and density (2%-ton cylinders).....................

11 i

2. Constituent material number density data (2%-ton cylinders)..................

11

3. Different types of calculations and their use in this study.....................

12

{

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.4. Tabulated results for various models of 2%-ton cylinders......................

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5. Constituent material mass and density data (10- and 14-ton cylinders)............

23

6. Constituent material number density data (10- and 14-ton cylinders).............

23

7. Tabulated results for various models of 10- and 14-ton cylinders................

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ACKNOWLEDGMENTS The author wishes to acknowledge the help and guidance given by H. R. Dyer and C. V. Parks in support of this work. Also, appreciation is expressed to C. M. Hopper and H. R.

Dyer for their review of this manuscript. The assistance of the following persons is also acknowledged:

J. C. Turner, who performed many of the calculations; and J. W. Insalaco, W. C. Jordan, and R. I. Reynolds, who provided data for the calculations. J. T. Thomas and R. R. Rawl are acknowledged for recognizing the need for this work and developing the funding to support it.

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ABSTRACT Currently, UF cylinders designed to contain 2% tons of UF are classified as Fissile 6

6 Class II packages with a transport index (TI) of 5 for the purpose of transportation. 'Ihe 10 mn UF cylinders are classified as Fissile Class I with no TI assigned for transportation. The 14 ton 6

cylinders, although not certified for transport with enrichments greater than 1 wt % because they have no approved overpack, can be used in on-site operations for enrichments greater than 1 wt %. The maximum usU enrichments for these cylinden are 5.0 wt % for the 2%-ton cylinder and 4.5 wt % for the 10- and 14-ton cylinders.

This work reviews the suitability for reclassification of the 2%-ton UF packages as Fissile Class I with a maximum U enrichment 235 6

of 5 wt-%. Additionally, the 10 and 14-ton cylinders are reviewed to address a change in maximum 235U enrichment from 4.5 to 5 wt %.

Based on this evaluation, the 2%-ton UF cylinders meet the 10 CFR.71 criteria for 6

Fissile Class I packages, and no TI is needed for criticality safety purposes; however, a TI may be required based on radiation from the packages. Similarly, the 10- and 14-ton UF packages 6

appear acceptable for a maximum enrichment rating change to 5 wt %

U.

235 1

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1. INTRODUCTION

~

The 2%-ton UF cylinder is currently in wide use for both national and international 6

transport of UF.. Use across national boundaries necessitates licensing and certification activities within each country of transport. Recently, the Japanese attempted to arrange a shipment of 2%. ton UF cylinders with an assigned transport index (TI) of 0. The U.S. Department of 6

Transportation (DOT) currently assigns a TI of 5 to such shipments. Based on a rigorous Japanese supporting analysis and a known conservative approach of the U.S. analysis, the shipment was permitted. This criticality review is meant to provide a rigorous U.S. analysis to determine the TI for 2%-ton UF cylinder shipments.

6 The U.S. regulations governing the packaging and transportation of fissile radioactive materials are contained in the publication 10 CFR.71.

Under the current 10 CFR.71 regulations, packages are classified according to Fissile Class I, II, or III and a corresponding TI is determined for each package design. Fissile Class I packages (TI not assigned, but effectively equal to zero) can be transported in unlimited numbers without any criticality safety controls.

Fissile Class II packages (0.1 < TI s 10) are normally limited to a cumulative TI (sum for all packages) of 50.* A Fissile Class III shipment (packages with a TI > 10) requires special arrangements for control of each shipment. Under proposed rule changes, the fissile class designations are discontinued, while the TI value ranges, TI = 0,0.1 < TI s 10, and TI > 10, are still used to prescribe controls during shipment. 'Ibe discussions in this report use both the old and the new package designations where possible, with greater emphasis on the Fissile Class designations since the proposed regulations are not in effer. at the time the report is written.

This criticality safety review focuses on three UF packages currently in use: 2%-ton,10-ton, and 14-ton UF cylinders. Because of the varied nature of each cylinder's use, they are 6

treated separately.

The goal of the first phase of this work is to provide a review of the suitability of the 2%-

ton cylinder and overpack for a subsequent reclassification as a Fissile Class I (TI = 0) package.

Currently, the 2%-ton cylinder with the overpack is classified as a Fissile Class II (TI = 5) package' for the purpose of transportation. While the maximum U enrichment that can be n5 placed in this cylinder is 5.0 wt %, shipments in excess of 1.0 wt % s5U require the cylinder be

• See Ref.1 for definition of transport index.

1

placec; in an overpack. The overpack design for the 2%-ton cylinder has been granted approval froiii the U.S. Nuclear Regulatory Commission (NRC) [ Certificate of Compliance (CoC) 4909 and CoC 9196] and DOT (DOT 21-PF-1A and 21-PF-1B, referred to generically as the 21-PF 1).

The technical results for the 2% ton cylinder are presented in Sect. 3.

The second phase of this work assesses the impact on both the 10-ton and 14-ton cylinders of a change in maximum msU enrichment from 4.5 wt % to 5.0 wt %. Specifically, for

.the 10-ton cylinder, the question to be addressed is what the new TIis for 5.0 wt % product.

For the 14-ton cylinder, the impact of such a change should~only be felt for on-site operations and only an assurance of criticality safety is needed.' In physical terms, the 10-ton and 14-tori cylinders are vey similar. Both cylinders have the same diameter, with the 14-ton cylinder being longer than the 10 ton cylinder. ney are both' limited to a maximum of 4.5 wt % s5U.

However, only the 10-ton cylinder has an approved overpack; thus, the 14-ton cylinder cannot be shipped with greater than 1.0 wt % enrichment. De 14-ton cylinder is therefore used primarily for on-site operations rather than for transport, ne 10-ton cylinder and overpack (the Paducah Tiger) are classified as a Fissile Class I (TI = 0) package,' having received approval for transport from the U.S. Department of Energy (DOE) (DOE 6553, currently under renewai) and the NRC (NRC 6553). He methods used in the analysis for both phases of tlO work are described in Sect.2. The technical results for the 10- and 14-ton cylinders are described in Sect. 4.-

De amount ofinternal moderation is very important for all three cylinders since a single cylinder is critical given sufficient moderation. Subaiticality is maintained through the use of moderation control, both by limiting the H/U ratio to 0.088 and assuring the cylinder is a " leak-tight" container, he justification of a " leak-tight" container is based on the physical and chemical characteristics of UF, under transport conditions and the rigorous quality assurance used during package filling and preparation for transport. Therefore, a premise of no water in-leakage into the UF, cylinder is made for each of the above analyses.

2

2. ANALYSIS hfETIIODOLOGY 2.1 ANALYSIS TOOI.S The criticality calculations necessary for this review were performed using the CSAS25 control program of the SCALE-4 computer system.5 He functional modules executed by this program include BONAAfI, NITAWL-II, and KENO V.a. The neutron cross sections used in this project were obtained from the SCALE 27-group ENDF/B-IV criticality library. Both the cross-section library and the SCALE-4 system are publicly available from the Radiation Shielding Information Center (RSIC). At Oak Ridge National Laboratory (ORNL), the SCALE-4 system is maintained under configuration control on an IBhi mainframe. The SCALE 27-group library validation is discussed in the next section.

2.2 VALID TION STUDIES References 6-7 provide a basis for the validation of the analytic tools used for this project.

The original validation effort applied to an early SCALE-3 version of the CSAS25/ KENO V.a system on an IBhi 3033 computer ystem. Reference 7 documents the updating of this validation effort for the SCALE-4 version of SCALE on the IBhi mainframe at ORNL Both validation efforts used the SCALE 27-group ENDF/B-IV cross-section library.

This latest version was used to perform the calculations in this study.

The code and cross-section validation performed in ref. 7 consisted of determining ker for a series of 51 benchmark critical experiments. These benchmarks consisted of a full range of possiblo experiments including 11 highly enriched cases and 40 low-enriched cases. The resulting k,a values were analyzeu statistically to determine the single-sided, uniform width, 8

closed-interval, lower tc?crance band such that 99.9% of the distribution of calculated k Will ar fall above the tolerance band with a 95% confidence level. He two curves shown in Fig.1 give the least-squares fit and corresponding hmit curve for the calculated kar values of the 51 benchmark critical experiments as a function of the average neutron energy group causing finia AEG. The top curve represents the least-squares fit to the data, and the bottom curve gives the lower limit of kur such that 99.9% of the distribution of calculated kar values are within the tolerance band with a confidence level of 95%. This bottom curve, along with the range of AEG values for this problem, is used to establish the suberitical maximum ker value for this study.

3

K eff vs. AEG Causing Fission For 51 Benchmark Calculations 1.03 Least-squares fit.to data 1.02 -

........J i.o i.

~l-'.....,

j i.co -

2 n

• o.99 -

i x

o.98-99.9% lower tolerance band with a 95%

t 0.'7 -

confidence level l

.... " ' -..,,, g o.96-o.95 oo S.'o to'.o 15'.o 20.0 25.o l

Average Energy Group l

Fig.1. k,,versus average energy group causing fission (AEG) for 51 benchmark calculations.

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2.3 ANALYSIS OVERVIEW The Fissile Class I regulations in 10 CFR.71.57 require that suberiticality be assured during both normal and accident conditions. The regulations for normal conditions require that an infinite array of packages with optimum interspersed hydrogenous moderation be suberitical.

The regulations for hypothetical accident conditions state that for 250 packages with optimal moderation between the packages, suberiticality must be assured. The analysis procedure described below should yield conservative estimates of k,g for accident and normal conditions.

The procedure used in this study begins with an infinite array model of 2%- and 10-ton UF. cylinders in their overpacks. The cylinder overpack is then replaced with variable-density water. The pitch or spacing between the cylinders in the array is determined by the overpack size. This pitch is such that the packages, if the overpacks were present, would touch. The removal of the overpack increases k,y due to the removal of a neutron-absorbing interstitial material. The variable-densitywater region allows evaluation of a full range of moderation (from void to full-density water). In the absence of gross deformations in the geometry (see Refs. 2-4 for a discussion of protective packaging performance), the resulting curve of k,g versus water density spans both the accident and normal conditions. The range of possible water densities is physically bounded on the low end by dry, burned insulation and on the high end by flooded conditions after a fire test where the insulation could possibly saturate with water. This range of possible water densities is bounded by the use of void to full-density values.

For the 14-ton cylinders, a similar approach was used in which the UF cylinders were modeled with variable density interstitial water moderation. Cylinder-to-cylinder spacing was set at the same value as the 10-ton cylinder because both cylinders have the same diameter.

If the k n values remain in the suberitical region for an infinite array with all possible e

water densities, then both accident and normal conditions of criticality safety for Fissile Class I

~

have been met. Additional array calculations with the overpack present can then be used to assess the change in k,y from the array results with no overpack. Calculations of single cylinders without overpacks and with infinite water reflection are used to provide calculational checks on i

some of the array results.

The arrays as described above were all modeled as square lattices. The use of a j

triangu'. pitch allows for a denser packing; however, the geometry is much more difficult to model in the computer code. A triangular pitch array is possible only for the 2%. and 14-ton cylinders, since the 10. ton cylinder has a square overpack. However, pitch reduction cases were 5

evaluated for all three cylinders to verify the expected lack of sensitivity to cylinder.to-cylinder spacings. Portions of the k versus water density curve corresponding to near-peak conditions err were regenerated for all three cylinder sizes assuming a 7% reduction in the cylinder-to-cylinder spacings (i.e., pitch). The 7% pitch reduction accounts for the difference in packing factors for the two lattices (0.79 for square pitch versus 0.90 for triangular pitch) because the cell volume varies as the square of the pitch. For these additional runs, the peak value of k n is not expected e

to differ from the previous runs. However, due to the differing interstitial volumes, the water density at peak k,g is expected to shift somewhat.

2.4 SENSITIVITY STUDIES The final set of calculations investigates temperature and fuel location effects. The temperature effects investigate the reactivity consequences of high UF6 temperatures (corresponding to the fire test conditions) and low UF temperatures ( 40*C as required in 6

10 CFR.71.55 part d). The fuel location studies are necessary because the actual UF6 configuration in each cylinder may vary. Thus, the most reactive fuel configuration must be determined. The configurations studied are shown in Fig. 2. The models are not drawn to scale, but simply to indicate the gross fuellocation patterns. Models (a)-(c) each have approximately 40% void space. The fuel density for model (d) is reduced, such that all models have the same UF loading. Model (c) represents the configuration due to the filling of a cylinder with liquid 6

UF, which then cools uniformly from the outside with a corresponding decrease in volume.

6 The resulting solid Uh typically has a void in the center. Model (a) represents the opposite configuration from model (c), and model (b) represents the fuel configuration expected from preferential cooling of UF on only one side of the UF cylinder. Model(d) approximates the 6

6 low-density UF fuel configuration. Experiments' have shown the most likely configuration to 6

be a combination of model (b) and model (c), with the central void in model (c) shifted toward the cylinder edge slightly. Base case calculations were performed with model (c).

Additional calculations will investigate the reactivity effects of varying the UF and j

6 moderator temperature during accident conditions. Temperature effects can be broken into density variations (both UF and moderator) and resonance-capture variations. The density 6

variations are accounted for in the fuel locations studies (e.g., model (d) above for low density UF ] and the <ariable-water density calculations to obtain the optimal interstitial moderation.

6 The resonance capture effects will be studied by performing three calculations, one at normal 6

O = fuel

= void g

l

%.?

(a)

(b)

OO (c)

(d)

Fig. 2. Fuel locations analyzed to determine most reactive configurations for a constant mass.

temperature, one at elevated temperatures corresponding to an accident case, and one at -40*C as required in 10 CFR.71.55(d). A standard temperature of 20*C was used for the base case.

3. 2%-TON CYLINDER ANALYSIS Phase I of the analysis evaluated the 2%-ton cylinder and its corresponding overpack, DOT 21PF-1, following closely the general description given in Sect. 2. This included the use of moderation control and the use of a " leak-tight" container to limit the hydrogenous material within the cylinder. The fuel region model was based on the fuel configuration shown in Fig. 2(c). An infinite array model incorporating an internal H/U ratio of 0.088 was then developed to allow the determination of the optimal interstitial moderation. In the 2%-ton cylinder model, the overpack is removed and replaced with variable-density water, while maintaining the same package-to-package spacing as a square-pitch array with the overpacks touching. The neglect of the overpack makes the results conservative (i.e., increases k,y) 7

4 because of the removal of neutroa-absorbing materials. The variable-density water region allows for the determination of optimal interstitial moderation.

3.1 MODEL DESCRIPTION The 2%-ton UF cylinder model was developed from the actual cylinder and overpack 6

dimensions contained in ref.10. The steel cylinder radius was taken directly from the description in ref.10. The curved lid and bottom surfaces of the cylinder were not modeled exactly; instead, the internal height of the cylinder cavity was determined from the volume as reported in ref.10.

The resulting model has a flat head and bottom rather than the actual curved one. The volume of the UF inside the cylinder was obtained from the total UF weight and a UF density of 5.1 6

6 6

2 g/cm, as given in ref.10. The inner radius of UF [see model (c) in Fig. 2] was then determined from the resulting volumes assuming a uniform UF thickness on tne sides and ends of the 6

cylinder. The single cylinder model is shown in Fig. 3 and was reflected on each of the six faces for infinite array calculations. Also, the single unit calculations used this model with the variable-water density region replaced with an effectively infinite water reflector. Similarly, the calculation with a 7% reduction in the pitch (approximating a triangular pitch array) used the same model except the outermost dimension (i.e., the outer boundary of the variable-water density region) was reduced by 7%. The cylinde, was assumed to be centered axially within the variable-water density region.

He model described above and shown in Fig. 3 is approximate due to the replacement of the overpack by variable-density water. The model shown in Fig. 4 corresponds to the 2%-ton cylinder in an overpack (DOT 21-PF-1). Calculations performed with this model can be directly compared with those using variable water density at a water density which is equivalent to the hydrogenous content of the overpack. His comparison allows the change in k,g for the model shown in Fig. 3 to be determined.

He materials contained in each region specified in Figs. 3-.4 are described in detail in Tables 1 and 2. Table 1 presents each material, its total mass in the model, and its actual mass.

Table 2 gives the material constituents and their respective atomic densities for completeness.

The models described above give the details for only one unit (i.e., a complete cylinder with and without its overpack). For infinite array calculations, these models were reflected on each of the six faces to represent an infinite array. No finite arrays were included in this study.

8

NOT TO SCALE 0.0

- 54.81 Region 1 -

l t....

. 0.0 Region 2 %

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Region 3 Region 4 %N N

0.0 54.81

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.. 41.16 a siaa d N

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.. 27.91

.. o,0 j

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Fig. 3. Model for 2%. ton cylinder without overpack. (All dimensions in cm.)

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NOT TO SCALE Region 13

- y

(

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Re gion 14 ~~~

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Re gio n 15 --

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'l 54.6l 39.21 ;

39.37 Fig. 4. Model for DOT 21.PF-1 overpack. (All dimensions la cm.)

10

4:

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Table 1. Constituent material mass and density (2%-ton cylinders)'

Shown in Model mass Actual mass Region Fig.

Material Density (kg)

(kg) 2%-ton cylinder

-1 3

Void i

2 3'

UF.

5.1 2,281 2,277 3-3 Steel 7.8212 495 635 4

3 Water or void Variabic Variable DOT 21 PF1 '

i 13 4

Void 14.

.4' Stee!

7.8212 80 NA 15 4-Phenolic foam 0.029 33 NA 16 4'

Steel 7.8212 154 NA l

Table 2. Constituent material number density data (2%-ton cylinders) i Density Atom density

)

Region Material / Reference (g/cc)

Constituent (atoms /b-cm) 2 UF, /ref.10 5.1 S U 4.4168E-4 SU 8.2860E-3 F

53134E 2 H

7.6800E-4 3,14,16'

' Carbon stee!/ref. 5 7.8212 Fe 83498E-2 SCALE s'tandard C

3.9250E 3 composition i

4

~ Variable-density water /

0.9982*

H 6.6751E-2*

ref.5 O

33376E-2*

SCALE standard composition 15 Phenolic foam /

0.029 C

5.9669E-4 ref.11 H

7.7981E-4 2*B 1.0221E 5 8 B 4.1465E-5 Si 13680E-5 C1 2.4630E-6 O

53063E-4

• Corresponds to full-density water.

11,

r 3.2 ANALYSIS RESUL'IE FOR 2%-TON CYLINDER l

Calculational results are given below for l

1.

infinite arrays with variable density water replacing the overpacks (3.2.1),

2.

an infinite array with the DOT 21-PF-1 c;erpack modeled (3.2.2),

3.

a single 2%-ton cylinder surrounded by an effectively infinite water reflector (3.2.3), and

4. '

fuel location and temperature sensitivity calculations (3.2.4).

These latter calculations allow the quantification of reactivity effects of fuel configuration inside i

the UF, cylinder and the effects oflow and high fuel temperatures corresponding to accident l

conditions. An overview of these calculations and their purposes are given in Table 3.

l Table 3. Different types of calculations and their use in this study o

Calculation type Purpose Infinite array with variable interstitial water density (without overpack) l Original outer radius Seek optimal interstitial moderation

'/% reduced outer radius Verify optimal conditions at different radius, estimate squarc. pitch versus triangular pitch effecu l

Infinite arrays with overpacks modeled Estimate change in k,, due to neglecting overpack I

i Single unit, infinite water reflection Meet regulatory requirement, consistency check l

on full. density water values from infinite array calculations Temp :rature effects at near Estimate temperature effects for accident optimal conditions conditions Fuel configuration (see Fig. 2)

Identify most reactive fuel configurations 3.2.1 Infinite Array Results without Overpacks j

A plot of the k.tr versus interstitial water specific gravity (SG) for the 2% ton UF6 cylinder is shown in Fig. 5. The k rr values are plotted versus water SG for convenience; e

3 however, the corresponding water density at a SG of 1.0 is 0.9982 gm/cm (water at 20'C, standard pressure). Thus, the abcissa label could be replaced with " water density." General 12

features of the curve include a peaking of ken for low water SG, followed by a steep decrease with increasing water SG,: nd ending with a slight increase near unity SG. De peak k value en of 0.817 0.003 at an SG of 0.015 represents the point of optimum interstitial moderation. The rapid decrease in k results for larger water SG indicates an overmoderated condition. He eu slight increase in k,g at SG values near unity arise from a change in role of the water from a moderating material to a reflecting material.

Details of the region around the peak k,g values are shown in Fig. 6. Here, the original curve for square lattice pitch is shown along with the curve for the 7% reduced pitch (associated with triangular pitch). The original curve peaks at a slightly lower SG value (0.015) than the 7%

reduced pitch case (0.020). However, the corresponding peak values of 0.817 0.003 and 0.816 0.002 are statistically indistinguishable. This is the expected behavior since the interaction between neighboring packages is governed by the total mass of moderating material.

~

For differing separation distances, but equal masses, the densities (SG) should change while the k,y values remain constant. The k n value of 0.817 should therefore represent a maximum for e

the 2%-ton cylinder for up to 2,277 kg of UF at 5 wt %.

6 3.2.2 Infinite Array Results with Overpacks The previous infinite a. ray calculations were all performed with models that replaced the overpacks with variable-density water. Additional calculations were then performed for the 2%-

ton cylinder with its overpack to assess the degree of conservatistn in the preceding results.

These calculations, as before, were for an infinite array of these units. The k,y value for.the 2%-ton cylinder is 0.655 0.002. Le effective water SG for this case is 0.01. Comparing these k,y values with the values fiom Fig. 5 indicates that the cylinder overpack decreases ken by 0.15 Ak for the 2%-ton cylinder.

j 3.2.3 Single Unit Results The single unit model for the 2%-ton cylinder consisted of the cylinder without an j

overpack with an effectively infinite water reflector. This case should be essentially identical to the infinite array k,y value with full-density water (SG = 1) because the water acts as an infinite reflector at full density. The single unit result is 0.453 0.003. This result is nearly identical with the SG = 1.0 value shown in Fig. 5.

The single unit result is primarily used as an i

I independent consistency check on the infinite array results.

13

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0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 H2O Specific Gravity Fig. 5. Plot of k aversus water specific gravity for infinite array of 2%-ton UF cylinders c

5 (square-lattice, full-diameter model).

0 14

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+

0.55-

~

"'\\'s "

G.

4 0.50-

"i"

'#.g.,

i

""--i-...........g 0.45-l Legend

^

4 i

E 2.5 Ton t

O.7.%.R.e.d.u.c.e.d..P

' ",o o

o.:

0'.2 0'3 0'.4 0.5 H2O Specific Gravity Fig. 6. Plot of k,,versus water specific gravity for inGnite array of 2%-ton UF, cylinders for original and 7% reduced pitch case.

j 15

3.2.4 Sensitivity Results De temperature for all calculations thus far has been 20'C. The final set of calculations quantified the effects of temperature and fuel location on the k,g results. The temperature effects were estimated by analyzing the 2%-ton cylinder at optimal moderation (an SG of 0.0150) with temperatures of 65 C,20*C, and -40*C.

The k,y values for 65'C and -40*C ate 0.817 t 0.002 and 0.818 0.003, respectively. These values are equivalent to the base k,g at 20*C of 0.817 0.003. The temperature effects are thus extremely minimal.

The second sensitivity area investigated was that of fuel configuration. The infinite array model used for this study was the 2%-ton cylinder at a ' vater SG of 0.02 (see Fig. 5). He fuel location studies analyzed all the cases shown in Fig. 2. Case (c) is the fuel geometry chosen for all cases thus far. This geometry was chosen since it was a likely physical configuration and also was expected to be the most reactive. The array results for cases a, b, c, and d are 0.774 i 0.002, 0.782 0.003, 0.811 0.003, and 0.812 0.003, respectively. Cases (c) and (d) are statistically indistinguishable when the standard deviations are taken into account and represent maximum reactivity conditions.

3.2.5 Validation Results The complete set of supporting results for the plots shown in Figs. 5-6 are given in Table 4.

In addition to the k n and water SG values, the table reports the AEG parameter e

values used for correlation to the lower limit of k as discussed previously in Sect. 2.3. The ca values of AEG given in the table range from 9 to 16. Over this range, in Fig. I the lower tolerance limit (suberitical limit) falls between 0.953 and 0.961.

For conservatism and convenience, the single value of 0.95 is chosen as the upper suberitical limit of k,g for this study.

Thus, the acceptance criteria for the calculational results presented above are the reported k,g plus two standard deviations must be less than 0.95, and the AEG is within the range of 9 to 16.

3.2.6 Summary The maximum k,y value for the conditions of optimal interstitial moderation with the premise of no water leakage into the UF cylinder, has been shown to be 0.817 0.003 for the 6

2%-ton cylinder with 5 wt % *U enrichment. Applying a 2a safety margin yields a k n value e

of 0.823. Since this is a peak value, the 2%-ton cylinder has a k,g less than the 0.95 upper suberiticallimit criterion at allinterstitial moderation conditions. These k values have been en 16

F i

Table 4. Tabulated results for various models of 2%-ton cylinders

+

i Case k,,

Std. dev.

Water SG*

• 'AEG*

Inficute array-square lattice pite 2%-ton results UF1 0.452 0.003 1.0 14.4 UF15 0.443 0.003 0.8 14.4 UF3 0.444 0.003 0.5 14 3 UF14 0.467 0.003 03 14.9 UF13 0.500 0.003 0.2 15.4 UF4 0.634 0.003 0.1 16.7 UF5 0.759 0.003 0.05 16.6 UF6 0.809 0.003 0.025 15.4 i

UFS 0.811 0.003 0.02 14.8 j

UF9 0.817 0.003 0.015 14.1 i

UF10 0.814 0.002 0.012 13.5 UF7 0.801 0.002 0.01 13.0 UF2 0.720 0.002 0.0 93 Infinite array-7% reduced pitd:

2%-ton results UF20 0.455 0.003 0.5 14.5 UF22 0.488 0.002 03 15.1 UF21 0.538 0.003 0.2 15.8 UF19 0.668 0.003 0.1 16.7 UF18 0.809 0.003 0.03 15 3 UF17 OE16 0.002 0.02 14 3 UF16 0.810 0.002 0.015 13.5 Overpact array results 2%-ton result UFD1 0.655 0.002 0.01 103 Single unit-infinite H 0 reflection 3

2%-ton result UF59 0.453 0.003 1.0 14.4 Temperature cHects 2%-ton results UFT1 (65*C) 0.817 0.002 0.015 14.1 UF9 (20'C) 0.817 0.003 0.015 14.1 i

UFD ( 40*C) -

0.818 0.003 0.015 14.0 Fuel location cirects 2%-ton results UFAI (Fig. 2a)

' O.774 0.002 0.020 14.2 UFBI (Fig. 2b) 0.782 0.003 0.020 14 3 UFB (Fig. 2c) 0.811 0.003 0.020 14.8 UFF1 (Fig. 2d) 0.812 0.003 0.020 14.8

' Specific gravity.

• Average energy group causing fission.

17

shown to be insensitive to cylinder spacing and temperature effects. This final k,,, value corresponds to an infinite array of optimal interstitially moderated cylinders; thus both norraal and accident conditions for Fissile Class I have been met. These final calculations should be i

conservative due to the neglect of the overpack materials ne degree of conservatism in kerr has been estimated at 20% for the 2%-ton cylinder.

j Based on this evaluation, the 2%-ton UF cylinder with 5 wt %

U enrichment meets S

6 the 10 CFR.71 criteria for Fissile Class I packages, and has a TI of zero for criticality safety purposes.

4.10- AND 14 "IDN CYIJNDER ANALYSIS Phase II of the analysis evaluated the 10- and 14-ton UF cylinders in a very similar 6

fashion to the 2% ton analysis described in Sect. 3. The use of moderation control (with H/U of 0.088) and the assumption of a " leak-tight" container were again employed in the analysis of the 10- and 14-ton cylinders. The fuel region model was again based on the fuel configuration shown in Fig. 2(c). Infinite array models of the UF cylinders with variable-density water 6

interspersed were used to determine the optimal interstitial moderation. These infinite array models were analyzed with cylinder spacing corresponding to a square-lattice arrangement followed by a reduced-pitch square lattice to approximate a triangular pitch lattice arrangement.

4.1 MODEL DESCRIPTION The calculational models of the 10- and 14-ton UF cylinders (Figs. 7-8) were based on 6

the physical descriptions given in ref.10. Approximations were made in these models primarily in the head and bottom regions. These regions were modeled as flat, where the actual geometry was curved. He cylinder internal volume was conserved, while the amount of steel in the cylinder wall was underestimated for conservatism (steel is an absorber, the removal of which should increase k,rr). The UF was assumed to adhere to the cylinder walls with a central void 6

space as shown approximately in Fig. 2(c). For both the 10- and 14-ton cylinders, the void space was approximately 40%.

Calculations were performed for the models shown in Figs. 7-8 in three different ways.

Single packages were analyzed with the models shown in Figs. 7-8 surrounded by an effectively 18 l

infinite water reflector. Infinite array calculations were also performed by specifying reflective or mirror type boundary conditions for each face of the single package models. The final type of calculation consisted of placing the single UF cylinder models inside the 10-ton overpack 6

(Paducah Tiger overpack) shown in Fig. 9. This calculation allows determination of the increase in k rr when the overpack is removed.

e He materials specifications for each region shown in Figs. 7-9 are described in detail in Tables 5-6. The materials number density, actual mass, and mass in the model are given for completeness.

4.2 ANALYSIS RESUL'IS FOR 10- AND 14 ' ION CYLWDERS Calculational results are given below for -

1.

infini,te arrays of these cylinders with variable-density water replacing the overpack (4.2.1),

2.

an infinite array of fhe Paducah Tiger overpacks containing 10-ton UF cylinders (4.2.2),

6 and 3.

single 10- and 14-ton UF cylinders surrounded by an effectively infinite water reflector 6

(4.2.3).

The infinite array calculations with variable-density water constitute the major portion of the analysis. These calculations allow the optimal interstitial moderation and, hence, peak k rrvalue e

to be determined. He single package results are essentially a check on the array calculations since the array calculations for full-density interstitial water, and those for the single package with an effectively infinite moderator thickness, should be virtually identical. The calculations with the overpack modeled (only for 10-ton cylinder) allows the conservatism in the approximate model to be evaluated.

4.2.1 Infinite Array Results without Overpacks Plots of k rr versus interstitial SG for both the 10- and 14-ton cylinders are shown in e

1 Figs.10-13. nese k.tr values are plotted versus water SG for convenience; however, the 3

corresponding water density at an SG of 1.0 is 0.9982 gm/cm (20*C, standard pressure). Thus, the abcissa label could be replaced with " water density." The same general trends seen for the 2%-ton cylinder are seen for the 10- and 14-ton cylinders. Generally, the curves peak for low water SG, followed by a steep decrease with increasing water SG, ending with a slight incre: e 19

NOT TO SCALE o.0

-- 96.84 Region 5 -

• l......................... o.o Region 6 %

Region 7 Region 8 0.0 91.76 I

-- 389.26

.. 328.30

..-326.71 Region 5 N

..- 307.13 NAi Region 6 j

N g

i Region 7 i

N

. - 82.13 Region 8 j

i

-- 62.55 g

.. 60.96

. 0.0 j

i ; 62.55

' 41 38 I-- 60.96 l

Fig. 7. Model for 10-ton cylinder without overpack. (All dimensions in cm.)

20

e.

NOT TO SCALE 0.0

- 96.84 i

i Region 9 -

i l

L e....

. 0,0 Region 10 i

Region 11 Region 12

}

l 0.0 91.76

-- 471.30

.. 4io,34

.. 4oe,73 Region 9 N

.. 388'42 NA!

Region 10 j

N

' s 5

Region 11 l

..- 82. 8 8 l.

Region 12

. 62.55

.. 60.96

l l

l ll

. 0,0 i

• 62.55

• 40.64

-- 60.96 Fig. 8. Model for 14. ton cylinder without overpsck. (All dimensions in cm.)

21 l-

NOT TO SCALE 0.0

- 96.84

-- 96.52 n

..r..,

Re gion 17 -

(

i.

.. o,o

......r...r..

Region 18

~

Region 19 f.

I Region 20

/

10 Tr n{yh 0.0 91.76 from Fig. 7

' - 389.26 N

i

. 3 g8,94

. 353.38

\\

s%..... ;f

-- 352.90 Region 17 N

Region 18 %

N Mw i

Region 19 Re9 on 20 i

-- 26.20 25.72 A

?

.. 0,32

. 0,0 I

75.72 i 91.44 76.20 Fig. 9. Model for Paducah Tiger overpack. (All dimensions in cm.)

22

(

e Table 5. Constituent material mass and density (10- and 14-ton cylinoers)

Shown in Model mass Actual mass Region Fig.

Matenal Density (kg)

(k)

E 10-ton cylinder 5

7 Void 6

7 UF.

5.1 9,555 9,539 7

7 Steel 7.8212 1,580 2,041 8

7 Water or void Variable Variable 14-ton cylinder 9

8 Void 10 8

UF.

5.1 12,527 12,501 11 8

Stee!

7.8212 1,976 2,359 12 8.

Water or void Variable Variable Paducab Tiger 17 9

Void 18 9

Steel 7E212 722 NA Polyurethane 0.087 674 NA 19 9

20 9

Steel 7.8212 910 NA Table 6. Constituent material number density data (10- and 14-ton cylinders)

Density Atom density Region Material / Reference (g/ce)

Constituent (atoms /b-em) 6,10 UF./ref.10 5.1 WU 4.4168E-4 80 8.2360E-3 r

53134E.2 H

7.6800E-4 7,11,18,20 Carbon steel / ret 5 7E212 Fe 83498E-2 SCALE standard composition C

3.9250E-3 Variabic density water /

8,12 ret 5 0.9982*

H 6.6751E 2*

SCALE standard O

33376E.2*

composition Polyurethane foam /

- 19 ref.4 0.08654 C

1.7750E.3 H

3.5500E-3 N

5.9200E-4 0

1.*1830E-3

• Corresponds to full density water.

23 I

~

l e

l l

l near unity SG. The peak value of k n (see Figs.10-11) for the 14-ton cylinder is 0.768 0.002 e

l for a water SG of 0.005. In Fig.11, curves of k nversus water SG are given for both the square-e lattice pitch and a 7% reduced pitch. The 7% reduced case gives essentially identical results, 0.766 0.002 versus 0.768 0.002.

The results for the 10-ton cylbder are shown in Fig.12, together with those for the 14-ton cylinder. These are shown together since the 10 and 14 ton cylinders are the same radius, only diffenng cylinder lengths. The overall trends of the 14-ton cylinder can thus be expected to duplicate those of the 10-ton cylinder. For the water SG values shown, the 10- and 14 ton curves are very similar. The 14-ton k n values are expected'to continue to be slightly larger than e

those of the 10-ton values since for high water densities, the infinite array k,g should approach the single unit k,g. In Fig.13, k,y curves for the original radius versus a 7% reduced radius are given. The same characteristics as the 2% ton and 14-ton cylinders are seen. Here, the peak values are 0.769 0.002 at an SG of 0.005 for the original radius and 0.763 0.002 at an SG of 0.005 for the reduced radius.

4.2.2 Infinite Array Results with Overpacks j

As stated earlier, only the 10-ton cylinder was analyzed with an overpack. This calculation, as before, was for an infinite array of packages with overpacks,10-ton cylinder and Paducah Tiger overpack for this case. The k nvalue for an infinite array of 10-ton cylinders and e

overpacks is 0.547 t 0.002. The overpack has an equivalent water SG of 0.05. By comparing the k,a value at an SG of 0.05 in Fig.12 (approximately 0.63).with 0.547, a difference of 0.08 Ak is seen. His represents a conservatism in the infinite array without overpack calculations.

4.2.3 Single Unit Results i

ne single unit models for the 10- and 14-ton UF cylinders consisted of the cylinder without an overpack surrounced by an effectively infinite water reflector. These values are used primarily as an internal consistency check on the infinite array results for an SG of 1.0. The single unit results are 0.526 0.002 and 0.533 0.003 for the 10- and 14-ton cylinders, respectively. These two values are very nearly the same, as expected, since the 10- and 14-ton cylinders are very similar in size. Comparing the 14-ton single unit result with the k n at an SG e

of 1.0 in Fig.10 also yields excellent agreement.

24 i

l l

L

o e

0.80 e

5 0.75 4 l-a l III i

0.70-,.

+

?

i

=u

• 0.65 -

e

\\

e 1

i M

i 4

0.60-i t

0.55-I i

41

=

i 4

t I

i 6

i i

3 t

0.50 0.0 -

0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1.0 4

H2O Specific Gravity Fig.10. Plot of k.aversus water specific gravity for infinite array of 14-ton UF cylinders (square. lattice, full-diameter model).

+

0; s

1 1

1 i

o.6o j

?

4_

1

)

r o

o.75-

-t" 4

?

e c

b r

1

+

i t

o.Fo -

e>

e u

s o',

i \\.,

l w'

I.

.,G i.

1 o.65-

^

1

+

, N -(

o.60 -

4

• t...' a Legend i

i i

j i

I E 14 Ton O 77. Reduced P o.55 o.co o.01 o.o2 o.o3 c.c4 c.os o.o6 o.or o.oe o.co c.io H2O Specific Gravity i

Fig.11. Plot of k, versus water specific gravity for infinite array of 14-ton UF cylinders 6

for original and 7% reduced pitch case.

26 V

0.80 i

i

' o, 0.75-

+

t

'. i il i

i i

i i

.i...

~

o,7o.

.t

-i t

e-

.2 i

i 1

s w

0.65-g.

?

a

^'

0.60-Legend

..,, l

..'.......a u to Ton O.14..To.n...

i i

0.s5 0.00 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.05 0.09 0.10 H2O Specific Gravity Fig.12. Plot of k.aversus water specific gravity for infi ite arrays of 10. and 14-ton UF.

n cylinders (square-lattice, full-diameter models).

1 27

• ,* s e

0.60 i

I l

4 Y

p' 5

0.7$ -

.j -

l 1I i

4 0.70-i

=

u 2

--e l

g

~

M 0.65-

.,~.

G.

i t

0.60-j I..

+

..~.7-Legend i

j i

• -E)

E 10 Ton i

t i

i i

[] 77. Reduced P 0.55 f

0.00, 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08 0.09 0.10 H2O Specific Gravity Fig.13. Plot of k nversus water specific gravity for infinite array of 10-ton UF cylinders e

6 for original and 7% reduced pitch case.

28 i

4.2.4 Sensitivity Results The sensitivity calculations described in Sect. 3.2.4 for the 2%-ton cylinder were not repeated for the 10- and 14-ton cylinders. The conclusions for the 2%-ton cylinder included the insensitivity of k to temperature effects. The peak cylinder temperatures for the 2%-ton and en 10-ton fire tests were very nearly the same, assumed to be 65'C for this analysis. The extreme temperature of -40*C is specified by the regulations. Also, the normal condition temperature of 20*C was used for all other calculations for the 2%,10- and 14 ton cylinders. The similarity of cylinders and temperatures encountered should allow the temperature sensitivity conclusions from the 2%-ton analysis to be applied to the 10-ton ar.alysis (the extreme temperature conditions apply only to cylinders with overpacks).

The fuel location sensitivity results generated in Sect. 3.2.4 should also be applicable to the 10- and 14-to. cylinders. Based on similar geometry arguments, the configurations with maximum ken hould again be models (c) and (d). The geometry corresponding to model (c) was s

used in the analysis of both the 10- and 14-ton cylinders.

4.2.5 Validation Results he complete set of supporting results shown in Figs.10-13 are given in Table 7. The table reports k,yvalues, water SG values, and AEG parameter values used for correlation to the lower safe limit of k,tr as discussed in Sect. 2.3. The values of AEG given in the table range from 9 to 14. Over this range, in Fig. I the lower tolerance limit (suberitical limit) falls between 0.953 and 0.961. For conservatism and convenience, the single value of 0.95 is chosen as the upper subcritical limit for k,y in this study. Thus, the acceptance criteria for the calculational results presented above are the teported kert plus two standard deviations must be less than 0.95, the upper suberitical limit.

4.2.6 Summary The maximum k,g values for the conditions of optimai interstitial moderation with the premise of no water leakage into the UF cylinder are 0.768 0.00: ad 0.769 0.002 for the 10- and 14-ton cylinders, respectively. Applying a 2a safety margin yields corresponding k,g l

values of 0.772 and 0.'773. Since these values represent peak reactivity, both the 10- and 14-ton I

cylinders have a k,y less than the 0.95 upper suberitical limit criterion at all interstitial moderation conditions. These k,g values should be insensitive io fuellocation in the cylinder, I

i 1

29

Table 7. Tabulated results for various models of 10 and 14-ton cylinders Case t,

Std. dev.

Water SG*

AliG' I

Infinite amy-4quare4attim pitti

10. ton resulu UF40 0.553 0.002 0.1 12.7 UF39 0.628 0.003 045 13 7 UF38 0.730 0.002 0.02 13.8 UF42 0.745 0.002 0.015 13.5 UF46 0.756 0.002 0.01 12.8 UF44 0.769 0.002 0.005 11.6 UF45 0.755 0.002 0.002 10.4 UF36 0.726 0.002 0.0 9.3

14. ton resu!u UF7.3 0.532 0.002 1.0 11.7 UF24 0.527 0.002 0.5 11.5 UF25 0.529 0.003 03 11.8 UF26 0.531 0.003 0.2 11.6 UF27 0.568 0.003 0.1 12.5 UF23 0.636 0.003 0.05 13.6 UF29 0.701 0.003 0.03 13.9

~

UF32 0.727 0.002 0.02 13.5 UF33 0.750 0.002 0.015 13.3 l

UF31 0.761 0.002 0.01 12.6 UF34 0.768 0.002 0.005 11.4 UF35 0.750 0.002 0.002 10.2 UF30 0.726 0.001 0.0 9.2 Infinite amy-7% ralucxxi pitch 10-ton resulu UF58 0.569 0.003 0.1 12.9 UF57 0.732 0.002 0.02 13.6 UF56 0.762 0.002 0.01 12.6 UF55 0.763 0.002 0.005 11.3 UF54 0.748 0.002 0.002 10.1 UF53 0.726 0.002 0.0 93

14. ton resulu UF52 0.581 0.003 0.1 12.8 UF31 0.737 0.002 0.02 13.3 UF50 0.766 0.002 0.01 12.4 UF49 0.764 0.002 0.005 11.2 UF48 0.746 0.002 0.002 10.0 UF47 0.728 0.002 0.0 9.3 W any resulta 104on result UFU2 0.547 0.002 0.05 11.2 Single unit-infinite 11,0 refkx2ma 10-ton result UF60 0.526 0.002 1.0 11.9 s

144on result

]

UF61 0433 0.003 1.0 11.8

'SpeciGc gravity.

' Average energy group causing Gr:s.on.

l 30 l

cylinder spacing, and temperature effects. These final k,rrvalues correspond to an infinite array of optimal interstitially moderated cylinders. Thus, the 10-ton UF. cylinder should meet both the accident and normal conditions for a Fissile Class I (TI = 0) cylinder with 5.0 wt % 25U 2

enrichment. These results also indicate that the 14-ton cylinder should be able to accommodate an increase in enrichment from 4.5 wt % to 5 wt % for on-site operations.

These final calculations should be conservative due to the neglect of the overpack materials. The degree of conservatism has been estimated at 12% for the 10-ton cylinder.

Based on this evaluation, the 10-ton UF cylinder with 5 wt % : U enrichment meets 35 6

the 10 CFR.71 criteria for a Fissile Class I package with a TI of zero for critic:.lity purposes; however, TI may be required based on radiation from the packages.

I

5. REFERENCES 1.

"Part 71-Packaging and Transportation of Radioactive Material," 10 CFR.71, Code of Federal Regulations 10,195 (Revised a. of January 1,1990).

2.

A. J. Mallett, C. E. Newlan, Protective Shipping Packages for 30-Inch-Diameter UF4 Cylinders, K-1686, Union Carbide Corp., Nucl. Div., Oak Ridge Gaseous Diffusion Plant (April 1967).

j 3.

Unica Carbide Corp., Nucl. Div., Internal Correspondence, Approvals Committee in Nuclear Safety to ORGDP Nuclear Safety Committee, May 19,1969.

4.

D. H. Stitt, SafetyAnalysis Report on the "Paducah Tiger" Protective Overpackfor 10-Ton Cylinders of Uranium Hexaf7uoride, KY-669, Union Carbide Corporation, Nuclear j

Division, Paducah Gaseous Diffusion Plant (June 1975).

5.

SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, Vols. I-III, NUREGICR-0200, Rev. 4 (ORNUNUREGICSD-2/R4), Vols. I-III (draft February 1990). Available from Radiation Shielding Information Center as CCC-545.

l 6.

W. C. Jordan, N. F. Landers, and L M. Petrie, Validation ofKENO V.a Comparison with Critical Erperiments, ORNUCSD/I'M-238 (December 1986).

7.

W. C. Jordan, SCALE 4 27-Neutron Group ENDFIB-IV Based Cross-Section Library Validation (to be published).

31

\\

,< s

. S.

H. R. Dyer, W. C. Jordan, and V. R. Cain, "A Technique for Code Validation for Criticality Safety Calculations," to be published in Proceedings of American Nuclear Society Meeting, held in Orlando, Florida, June 1991.

9.

Union Carbide Corporation, Nuclear Division, Internal Correspondence, " Voids in UF.

Cylinders," M. G. Otey to C. W. Walter, February 24,1971.

10.

Uranium Herofluoride: Handling Procedures and Container Descriptions, OR'O-651 (Rev.

5), U.S. Department of Energy (September 1987).

11.

USDOE Material and Equipment Specification SP-9, Rev.1.

1 j

i I

32 i

F e,]

ORNIIDd.11947 INTERNAL DISTRIBUTION 1.

S. M. Bowman

24. L B. Shappert 2.

M. C. Brady 25.

G. R. Smolen 3-7.

B. L Broadhead 26.

R. Stachowiak 8.

H. L Dodds 27.

W. C Stoddart 9-13.

H. R. Dyer 28.

J.S. Tang 14.

C. M. Hopper 29.

J. C Turner 15.

J. W. Insalaco 30.

M. J. Welch 16.

W. C. Jordan 31.

B. W. Welles 17.

V. S. McCauley 32.

R. M. Westfall

18. L F. Norris 33.

G. E. Whitesides/R. P. ieinius

19. C V. Parks 34-35. Laboratory Records Dept.

20. L M. Petrie

36. Laboratory Records, ORNL-RC

21. R.R.Rawl

37. Document Reference Section 22.

J.-P. Renier

38. Central Research Library 23.

C. H. Shappert

39. ORNL Patent Section EXTERNAL DISTRIBUTION 40.

D. M. D'Aquila, Martin Marietta Energy Systems, Inc., Nuclear Criticality Safety, P.O. Box 628, MS 2213, Piketon, Ohio 45661 41.

J. W. Bennett, DOE Field Office, Oak Ridge, P.O. Box 2001, Oak Ridge, TN 37831 42..

M. Bennett, DOE Field Office, Oak Ridge, P.O. Box 2001, Oak Ridge, TN 37831 43.

L Blalock, Nuclear Materials Transportation R&D, U.S. Department of Energy,

- EM, Washington, DC 20545 44.

R. A. Boelens, Martin Marietta Energy Systems, Inc., P.O. Box 628, Bldg. X.

344a, MS 4018, Piketon, Ohio 45661 45.

W. Carriker, Radioactive Materials Branch, Office of Hazardous Materials Transportation, U.S. Department of Transportation, 400 7th Street S.W.,

Washington, DC 20590 l

46.

C A. Caves, U.S. Department of Energy, Washington, DC 20545 47.

R. Chappell, Transportation Branch, Office of Nuclear Material Safety &

Safeguards, U.S. Nuclear Regulatory Commission, MS WF1, Washington, DC 20555 43.

R. N. Collier, DOE Field Office, Oak Ridge, P.O. Box 2001, Oak Ridge, TN 37831 49.

R. H. Dyer, DOE Field Offic :, Oak Ridge, P.O. Box 2001, Oak Ridge, TN 37831 50.

E. P. Easton, Transportat'an Branch, Office of Nuclear Material Safety &

Safeguards, U.S. Nuc'aar Regulatory Commission, MS WF1, Washington, DC 20555 l

51.

D. Edmund.;n, Westinghouse Electric Corp., Nuclear Fuels Div., P.O. Box 5906, I

~ Colum%, SC 29250 33

\\

y,,

52.

F. P. Falci, Manager Nuclear Materials Transportation R&D, U.S. Department of Energy, EM-51, Washington, DC 20545 53.

A. S. Garcia, Argonne National Laboratory, Fuel Cycle Division, P.O. Box 2528, ANL-W, Bldg. 765-A, Idaho Falls, ID 83403-2528 54.

R. F. Garrison, Nuclear Materials Transportation R&D, U.S. Department of Energy, EM-51, Washington, DC 20545 55.

B. Hook, Paducah Gaseous Diffusion Plant, P. O. Box 1410, Paducah, KY 42001 56.

D. R. Hopkins, Transportation Branch, Office of Nuclear Material Safety &

Safeguards, U.S. Nuclear Regulatory Commission, MS WF1, Washington, DC l

20555 57.

W. R. Householder, Nuclear Container, Inc.,1410 Strawberry Lane, Johnson City, TN 37604 5S.

J. Huffer, Paducah Gaseous Diffusion Plant, P.O. Box 1410, Bldg. C-743, Paducah, KY 42001 59.

W. Jackson, Advanced Nuclear Corp., P.O. Box 130, Richland, WA 99352 60.

F. Kovac, Martin Marietta Energy Systems, Inc., P.O. Box 628, Piketon, OH 45661 61.

C. E. MacDonald, Transportation Branch, Office of Nuclear Material Safety &

Safeguards, U.S. Nuclear Regulatory Commission, MS WF1, Washington, DC 20555 62.

R. Miller, Combustion Engineering, Inc., P.O. Box 107, Hematite, MO 63047 63.

T. S. Needels, U.S. Department of Energy, EH, Washington, DC 20545 64.

R. Newvahner, Martin Marietta Energy Systems,Inc., P.O. Box 628, MS 2210-A, Piketon, OH 45661 65.

W. A. Pryor, PAI Corporation,116 Milan Way, Oak Ridge, TN 37830 66.

F. Punch, Chief, Packaging Certification Program, Division of Quality Verification and Transportation, U.S. Department of Energy, MS EH-321, Washington, DC 20545 67.

R. I. Reynolds, Paducah Gaseous Difft...,n Plant, P.O. Box 1410, Bldg. C100, Paducah, KY 42001 68.

J. Russell, DOE Field Office, Oak Ridge, P.O. Box 2001, Oak Ridge, TN 37831 69.

D. C. Thomas, Manager, Gaseous Diffusion Operations, Office of Uranium Enrichment, U.S. Department of Energy, MS NE-33, Washington, DC 20545 70.

J. T. Thomas,16 Laurel Place, Norris, TN 37828-0296 71.

M. Wangler, Chief, Radioactive Materials Branch, Office of Hazardous Materials Transportation, U.S. Department of Transportation, 400 7th Street, S.W.,

Washington, DC 20590 72.

L D. Williams, Martin Marietta Energy Systems, Inc., P.O. Box 628, Bldg. X-100, Piketon, Ohio 45661 73.

C. J. Withee, Transportation Branch, Office of Nuclear Material Safety &

Safeguards, U.S. Nuclear Regulatory Commission, MS WF1, Washington, DC 20555 74.

J. Zidak, General Electric-Nuclear, P.O. Box 780, Wilmington, NC 28402 75-84.

Office of Scientific and Technical Information, U.S. Department of Energy, P.O.

Box 62, Oak Ridge, TN 37831 85.

Office of Assistant Manager for Energy Research and Development, DOE Field Office, Oak Ridge, P.O. Box 2008, Oak Ridge, TN 37831 34