ML20148M602

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SAR for Packaging (Oak Ridge Y-12 Plant Model Vcd Package for Enriched U Solution.) Concludes That Package Satisfies All Applicable Req.W/Encl Review of Rept by Wa Pryor of DOE
ML20148M602
Person / Time
Site: 07105560
Issue date: 09/03/1974
From: Dyer H
UNION CARBIDE CORP.
To:
Shared Package
ML20148M407 List:
References
Y-DD-155, NUDOCS 7811220079
Download: ML20148M602 (23)
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Text

{{#Wiki_filter:, ~ O REVIEW OF SAFETY ANALYSIS REPORT FOR PACIMGING (OAK RIDGE Y-12 MODEL VCD) PACKAGE FOR ENRICHED UPATIUM REPORT Y-DD-155 WILLIAM A. PRYOR General The packaging may be considered to be a special application of DOT Specifica-tion 6L - 110 gallon size (49 CFR 178.104). The principal differences are as follows: Requirement VCD DOT 6L l 1. Container: a. Outside Drum - DOT Spec. 17H 6J or 17H b. Containment Vessel (s) 1. Primary 13.54 Polyethylene None Specified Bottle 2. Primary Dimen-5 1/8 x 47 9/16 sions (in.) 3. Secondary DOT Spec. 2R 4 Secondary Dimen-6.065" dia x 50** 5.25 dia x 50*** sions (in.) c. Insulation Wood and Vermi-Vermiculite culite d. Insulation Thickness - 4" (top) 3.75" (end) Min. (in.) c. Spacers End wooden discs Steel Rods f. Gross Weight - (1b) 500 480 2. Contents: a. Fissile Materials 235 233, 235 U U U, Pu b. Form Solution Metal & Compounds 235 c. Mass of U - Max. (kg) 7.86 13.5

  • Equivalent to DOT Spec. 2R
    • Plus top flange and bottom plate
      • Includes caps

(( y) 781122 c>o7g

. Other design requirements are essentially the same. Thus, this review is concerned with those differences which are considered to be signifi-cant as well as the results of actual tests and evaluation. Inner Containment Inner containment is provided by the polyethylene bottle and the flanged and gasketed pipe. Since UO solutions are involved, only new polyethylene 7 bottles are used for each shipment. This eliminates the problem of degrada-tion of the polyethylene by long term exposure to NO ns. Any leakage 3 from the bottle, which will be minimal, will be into the stainless steel pipe. Stainless steel is compatible chemically with the solution. Test results indicated that the flanged and gasketed pipe did not leak in the Type B test series. Since the actual exposure in the thermal test was more severe, both in the maximum temperature and the time of exposure, the polyethylene bottle should readily survive the prescribed test conditions. physical Testing The physical testing confirms that the design requirements are satisfactory for meeting the hypothetical accident test series for Type B containers. The use of the laminated plywood discs to center the steel pipe appears to be a satisfactory substitute for the steel rod spacers required for the DOT Specification 6L Container. Inleakage of moisture does not decrease the effectiveness of the packagings; thus, the requirements of 10 CFR 71.33 are fully met. Nuclear Criticality Safety The evaluation for nuclear criticality safety is based upon direct experi-mental data validated computer codes. For the prescribed loading in the normal and accident conditions of transport; a. the individual packaging will remain suberitical, and l b. the arrays for Fissile Classes II and III will be suberitical. The assumptions and conclusions are consistent for the types of material transported.

. Quality Assurance The quality assurance requirements for fabrication and maintenance were reviewed by the ORO Quality Assurance Branch and were found generally to be satisfactory. Two changes in the quality assurance requirements will be made as follows: 1. Fabricators will be required to certify that packages are fabricated and assembled according to design and specifications. 2. The container inspection form will be revised to indicate that the packaging is assembled according to design prior to shipment. The SARP and drawing will be referenced. Distribution The SARP was distributed under TID-4500, UC-71. Conclusion The requirements of AECM-0529 have been met. Therefore, it is recommended that the use of this packaging be continued until June 17, 1978, at which time all packages using vermiculite insulation will be no longer be approved for transport. The ERDA-0R0 Certificate of Compliance will expire on this date.

=, p l Y DD 155 1 5 T l SAFETY ANALYSIS REPORT FOR PACKAGING (Oak Ridge Y 12 Plant Model VCD Package i For Enriched Uranium Solution) l l H. R. Dyer i 1 I j l 1 j l i1 .y. September 1974 i i l OAK RIDGE Y-12 PLANT 3 OAK RIDG E. TENNESSEE f I f prepared for the U.S. ATOMIC ENERGY COMMISSION under U.S. GOVERNM EN T Contract W.7405 eng 26 1 19 i y i I e._..

l o Fleference to a company or product name does not imply approval or recommendation of the product by Union Carbide Corporation or the U.S. Atomic Energy Commission to the exclusion of others that may meet specifications. Printed in the United States of America. Available from National Technical Information Service U.S. Department of Commerce 5285 Port Royal Road, Springfield, Virginia 22151 Price: Printed Copy $4.00; Microfiche $1.45 This report was prepared as an account of work sponsored by the United States Government. Neither the United States nor the United States Atomic Energy Commission, nor any of their employees, nor any of their contractors, subcontractors, or their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness or usefulness of any in formation, apparatus, product or process disclosed, or represents that its use would not infringe privately owned rights. Y t

=. Date of issue: September 3,1974 Report Number: Y DD 155 Distribution Category: UC 71 i SAFETY ANALYSIS REPORT FOR PACKAGING (Oak Ridge Y-12 Plant Model VCD Package For Enriched Uranium Solution) H. R. Dyer Y 12 Radiation Safety Department Technical Division e s 't l Oak Ridge Y 12 Plant ' P,0, Boa Y. Oak Ridge, Tennessee 37830 e Prepared for the U S Atomic Energy Commission Under US Government Cont ract W 74054ng-26

4 l 2 I 1 . CONTENTS -

SUMMARY

3 I N T R O D U CT I O N...................................................... 4 SAF ETY AN ALYSIS O F TH E VCD PACKAGE............................... 5 Description of the Shipment 5 Description of the Package................ 6 Outer Containment Vessel Description................ 6 inner Container Description 6 Inner Containment Vessel Description............................. 6 Wooden Centering Spacers.............. 6 Thermal Insulation.............. 8 Additional Package Description Considerations. 8 Safety Analysis of the Model VCD Package...........,........... 9 Evaluation of Normal Conditions of Transport.......................... 9 Hypothetical Accident Evaluation................................... 10 Nuclear Criticality Safety Evaluation...................................... 14 Single-Package Analysis................. ......................15 Arra f An alysis..................... ...........................15 Quality Anusance. .................................17 Fabrication and Assembly of Packages. .............17 + Routine Inspection of in Use Packages........................... .. 18 Co nc lu sion s........................... ..............................18 f I l 'l I l 9 4.

-- ~.- l.l' 3 l-i l. l-SUMM ARY - s An evaluation of the _Y.12 Model VCD shipping container for enriched uranium solution was ,f made to demonstrate,.its compliance with federal regulations governing the interstate transport of fissile materials. Destructive testing and engineering evaluations were made to demonstrate the structural integrity and thermal resistance of the package in meeting the physical testing described in the AEC regulations.(a) Reactor transport theory computer codes were used to analyze the criticality safety of the package containing enriched uranium solution. Results of this evaluation show that the package complies with all applicable regulations. Analysis of the Y 12 Model VCD shipping container indicates,that aqueous solutions containing any enrichment in the 2as U isotope meets the requirements for Fissile Class 11 shipments. Limitations on each package consist of a maximum volume of 13.5 liters of true solution and a contained 2 a s U mass not to exceed 7.86 kgs, e 1 i I (a) US Atomic Energy Commission regulations AECM-0529 and 10 CFR 71, " Safety Standards for the Packaging of Fissile and Other Radioactive Materials",' and " Packaging of Radioactive Material for Transport". ~

4 i INTRODUCTION A package used for the off site shipment of fissile materials is subject to evaluation under federal regulations which ensure the safety of personnel and radioactive materials while in 1 i transit. These regulations govern the structural integrity, thermal resistance, radiation shielding, nuclear criticality safety, and quality assurance of the package. The criteria are set 4 forth in USAEC Manual Chapter 0529 and Title 10 Code of Federal Regulations, Part 71. To ensure safety and secure approval for the use of a package for shipment, it must be j shown by physical testing and/or validated computational methods that the package complies with these regulations. The UCC ND Oak Ridge Y 12 Plant Model VCD package ~ was evaluated in accordance with these requirements, based on the methods and results of a j previous analysis.(b) The Model VCD package was originally approved under the provisions of Bureau of j Explosives Permit 2137 and BA 524 in 1965. The package was later reevaluated and 1 approval reissued under the provisions of the Hazardous Materials Regulations Board of the Department of Transportation, Special Permit 5560, dated March 15,1968. i + e 1 i 6 I i 4 1 4 's (b) Crume, E. C., Handley, G. R., Mee, W. T., and Pletz, R. H.; Evaluation of a Shipping Container for Enriched Uranium Solution, Y KC-109: Union Carbide Corporation-Nuclear Division, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee (1967). 14

l 5 SAFETY ANALYSIS OF THE VCD PACKAGE DESCRIPTION OF THE SHIPMENT / The Y-12 Model VCD shipping container, illustrated in Figure 1, is used for shipments of aqueous uranium solutions of any enrichment in the 2 3 s U isotope. The contents of each package shall be limited to 13.5 liters of true solution (stable solution, with no crystallization or slurry) containing not more than 7.86 kg of 23sU. The package meets the requkements for Fissile Class 11 shipments, with a transport index of 4.2 to be assigned to each package (unless, however, external radiation levels dictate a higher assignment). Fissile Class ill shipments are approved for no more than 24 such packages. 'N Drum lid Bolt-Type Sealing Ring Bolted 304 Stainless Steel Standard ( Blind Flange w,th Vitron Fluorelastomer s i s "O" Ring Lominate Fir Plywood Plugs with Lif ting Holes (16" D x 4" H) g j; q I fg >[, lamino'ed Fir Plywood Top Spacer / 'i (22 3/8" OD x 7" H, Upper Section,161/8" ID x 4" H, Y [. $! N lower Section,11 1/8" ID x 3" H)

hj wording on label

l Fissile Rodlooctive Material Container ={" AEC-O R USA /5560/BF 3 Union Corbide Corp. Nuclear Division i h<

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. s Y-12 Plant Ook Ridge, Tennessee <[ Container Model VCD Serial No Y-12-VCD-000l* d l: j \\ Vermiculite 85' Min / Drum \\ j$ f [ } h \\ Schedule 40 Type 304 Stainless Steel Pipe 4 :, w. p (6 5/8" OD x 50" lH) j !"s \\ x/ o \\ Polyethylene Bottle !N eb. 'l MI Steel Drum (16 gage carbon steel c -1 i,3 ;, 221/2" 10 x 60" lH) > / lominated Fir Plywood Bottom Spacer N k (22 3/8" D x 6" H with 6 3/4" x 13/4" cavity for oice) Figure 1. Y 12 FISSILE CLASS 11 ENRICHED URANIUM SOLUTloN CONTAINER.

6 DESCRIPTION OF THE PACKAGE Outer Con'ainment Vmel Description The outer drum shall consist of two 55-gallon drums.of DOT Specification 17H or equivalent weided bottom to top to form a 110-gallon capacity. The body, bottom head, and removable head sheets shall be of 16-gauge (0.0533 inch, minimum thickness) steel. The drum may be either a single sheet of steel or rnay be produced by welding together two appropriate lengths of such drums. The removable head is attached to the drum body by a 12-gauge (0.0946 inch, minimum thickness) bolted ring clamp with drop forged lugs. One lug shall be threaded for a 5/8 inch bolt for locking. The bolt shall be four inches long, which provides for tightening the ring clamp. A jam nut shall also be placed on the four-inch bolt to secure the closure. A 1/16 inch-diameter hole shall be drilled in the threaded end of the bolt for a wire type security seal. The removable head sheet shall be gasketed with rubber, or an equivalent material, which will prevent inleakage under normal transport conditions. The tare weight of the package is nominally 450 pounds; the gross weight, about 500

pounds, inner Container Description The inner container consists of a 50-inch-long, nominally 6 inch diameter, Schedule 40, Type 304 stainless steel pipe. The bottom end is permanently closed with a 5/8 inch-thick, C

Type 304 SS plug, welded into place. A 1/2 inch thick, Type 304 SS, slip on flange is welded to the top end of the pipe with eight,5/8 inch nuts tack welded to the back side of the flange to permit bolting a blind flange. The closing fixture is a 5/8-inch thick, Type 304 SS, blind flange with eight equally spaced hoks provided for attaching to the slip-on flange with 5/8 inch bolts. The blind flange is also grooved to accept an O ring (Vitron fluoretastomer or equal). Flange bolts shall be tightened to at least a 15-foot pound torque during closing. Details of the inner containment vessel are outlined in Figure 2. Inner Containment Vessel Description The inner containment vessel consists of a nominally 13.5 liter polyethylene bottle (51/8 inches in diameter by 47 9/16 inches high, including the cap). A drawing of the bottle is given in Figure 3. A new polyethylene bottle shall be used for each shipment of solution. Wooden Centering Spacers Laminated, fir, plywood spacers are used to center the inner container in the outer drum. The bottom spacer is.22 3/8 inches in diameter and 6 inches thick, with a I 6 3/4 inch diameter by 13/4 inch deep hole for centering the bottom end of the pipe. The l top spacer is 22 3/8 inches in diameter and 7 inches thick, with an 111/8-inch-diameter by 3 inch deep hole to accommodate the flange and a 161/8 inch-diameter by 4 inch-deep hole to accept a laminated, fir, plywood plug. The plug is 16 inches in diameter by 4 inches thick and has two 1 inch diameter finger holes in each side. The top spacer is held in position by 1

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8 four equally spaced wood screws. The top spacer has three 2-inch-diameter holes drilled through for inspection of the vermiculite insulation. Details of the top and bottom spacers are noted in Figure 2. Thermal Insulation N n [ l gs Thermal insulation is provided by filling all the void space between the inner and outer containers with vermiculite. Newly constructed containers shall be filled with a minimum of 85 pounds of vermiculite. 9/32 '- Continued use of the package will result in some settling of the insulation. Prior to each shipment, any N 2 5/166 void space resulting from settling shall be filled with ); .( o a vermiculite via the 2-inch diameter inspection holes in L$ iJ the top spacer. f l l Additional Package Description Considerations S i gj USAEC regulations specify additional descriptions of l I the package; however, the Model VCD package l neither has these additions nor are they needed. I A k Following is a list of descriptions that are not a part i of the package and the reason why each was not { { l considered. j i 2 l i k l l

1. " Materials specifically used as nonfissile neutron 4

l absorbers or moderators"- While the materials of l-construction do act as neutron moderators and ~ 47/8' - l i ~; absorbers, no materials are needed in the package l which have the sole function of moderating or j absorbing. l l

2. " Valves, sampling ports, lifting devices, and l

l 1 tie-down devices"- None of these devices are I / necessary. ' D' 5 1/8"

3. " Structural and mechanical means for the transfer Figure
3. SPECIFICATIONS FoR THE and dissipation of heat", and " identification and POLYETHYLENE BOTTLE.

volumes of any coolants and of receptacles containing coolant"- Since there will not be any internal heat generation from the\\ unirradiated uranium solutions, no means of heat transfer or coolants are required or prov;ded. I

i 9 l t SAFETY ANALYSIS OF THE MODEL VCD PACKAGE Evaluation of Normal Conditions of Transport / USAEC regulations (AECM 0529, Annex 1, and 10 CFR 71, Appendix A) set forth criteria that are to be considered in the evaluation of the package to withstand normal conditions of transport. Nine'specified conditions of the regulations are listed with the analysis for the Model VCD package.

1. " Heat; direct sunlight at an ambient temperature of 1300 F" Direct testing of the package under these conditions is not avai!able. However, based on heat transfer calculations (as presented in Cask Designers Guide - A Guide for the Design, Fabrication, and Operation of Shipping Casks for Nuclear Applications, ORNL NSIC-68) it can be shown that there is about a 150 F temperature increase in the outer drum temperature when it is subjected to the pres (.ribed conditions. However, with the vermiculite and wood insulation in place, this temperature increase would not be realized in the inner container. The recommended upper temperature limit for continued use of the polyethylene bottle is approximately 2000 F; so, even if the bottle were to reach the 1450 F temperature, neither the integrity of the container nor its contents would be affected.
2. " Cold an ambient temperature of -400 F in still air and shade" Direct testing of the container under these conditions is not available. Exposure of the package to this j

i temperature would have. no detrimental effect on the package; however, extended periods of exposure could conceivably result in freezing the solution. Since the / polyethylene bottle will remain pliable to temperatures lower than -400 F, freezing and subsequent expansion of the contents will have no effect on the bottle.

3. " Pressure atmospheric pressure of 0.5 times the standard atmospheric pressure"- A sudden drop in atmospheric pressure may cause a slight expansion of the drum head; however, this pressure decrease will have no effect on the inner container or contents.

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4. " Vibration vibration normally incidental to transport" Simulated vibration tests have l

not been conducted. However, past experience with this package has shown that some settling of the vermiculite will occur after the package has been used repeatedly. inspection holes are provided in the top wooden spacer for inspecting the level of insulation; and, if needed, additional vermiculite will be added. 5, " Water Spray-a water spray sufficiently heavy to keep the entire exposed surface i continuously wet during a period of 30 minutes"- Although not tested, water inleakage would not affect the safety of the package, as water was assumed to be in the insulation for the criticality analyses. However, significant amounts of water would increase the weight of the package, making it more difficult to handle. Past experience with this package stored outside in all types of inclement weather has never revealed water inleakage. t.

6. " Free Drop between 1 1/2 and 21/2 hours after the conclusion of the water spray test, a free drop through a distance of 4 feet onto a flat, essentially unyielding, horizontal l

10 surface in a position for which maximum damage is expected"- The four foot drop test i was not conducted since the 30-foot drop test for the hypothetical accident condit ons caused only superficial damage,

7. " Corner Drop - a free drop onto each corner of the rim from a height of one foot onto a

~ flat, essentially unyielding, horizontal surface"- Based on the limited damage suffered in the 30 foot drop test, a drop of only one foot would cause no damage to the package.

8. " Penetration impact of the hemispherical end of a vertical steel cylinder 11/4 inches in diameter and weighing 13 pounds, dropped from a height of 40 inches onto the exposed j

surface of the package which is expected to be most vulnerable to punc-l ture"- Penetration tests resulted in only cursory damage to the drum surface, with no j effect on the package or its contents.

9. " Compression a compression load equal to either five times the weight of the package or two pounds per square inch multiplied by the maximum horizontal cross section of the package, whichever is greater, for a period of 24 hours"- Five times the gross weight (2000 pounds) is the greater of the two forces. The maximum stress in the drum walls was calculated to be 410 pounds per square inch. Since the allowable stress in steel is in the order of 25,000 pounds per square inch, the compressive load will not affect the package.

l Hypothetical Accident Evaluation USAEC Manual Chapter 0529, Annex 2, and Title 10 CFR Part 71, Appendix B, define the hypothetical accident conditions to which a package is to be subjected. The test is to be s performed sequentially in the order indicated to determine the cumulative effect on a package or array of packages.

1. " Free Drop" a free drop through a distance of 30 feet onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected" Two prototype containers were drop tested. The container suffering the most damage (Figure 4) struck with its longitudinal axis nearly parallel to the drop test pad. The radius of the container, at the point of impact, was reduced by about one inch, with a somewhat less reduction along the axis. The other,less damaged package is seen in Figure 5. Neither the bolted ring clamps nor the lids came loose, and the ef fectiveness of the ring clamps was unimpaired.
2. " Puncture - a free drop through a distance of 40 inches striking, in a position where the maximum damage is expected, the top end of a vertical, cylindrical, 6 inch-diameter, mild steel bar, mounted on an essentially unyielding horizontal surface"- The containers were dropped such that the edge of the weld seam, where the two 55-gallon drums were joined, struck in the 6 inch diameter bar. Views of the damaged areas are presented in

) Figures 6 and 7. As seen, the resultant damage was about an inch indention at the point of impact, but there was no tearing of the container wall. s

3. " Thermal exposure to a thermal test of 14750 F for 30 minutes"- The two containers were fire tested in an electrical furnace for one hour at 17000 F. This more stringent test

11 was conducted prior to the implementation of AECM-0529, Annex 2, and 10 CFR 71, Appendix B, which require the less severe thermal test. ' h%g y ,7 . g ~f / .Y .g '/ S ~ ^ L.7 a lr' 2 /- l h ' \\. .,f -f,.y ,X ?,. .N. f'., . ~ ~ g. _ g. **. . -~y % . m ,T - '* Lg W... .p y - . 9,., 1 .... + $.,m.a s 9 ~q m:,, + 4 .] .. r. ,. ; ;_gppg'llL i . G. 'la<; f. r

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g. a 4.s ' 430lh h, WIQh ' l9 yyWQ. % Z, h a k C c, 5 +. Ap-gu(, ll:% ;W% g d y.g ; f;/ g p. <] A eif R fs,C.7:{d p :.;;._ < n ;' ' L.{ r b,s /. - f '.p^ ~"t '4 QGQV,y '4>c %.> O *b WV u - 4 w ~.. + 4 ~~ .. +. ;y, y; ;g Q', y.-* 3 7 y: ( ly _3g..yQ'*f PH 65 531 Figure 4. DAMAGED CONTAINER AFTER THE 3o-Foot DROP. Internal temperature monitoring of one container indicated that the thermal insulation limited the temperature of the inner stainless steel pipe to a maximum of 3500 F. However, localized melting of the polyethylene bottle occurred, releasing the contents to the inner pipe, as noted in Figure 8. No other changes in the vessel integrity resulted from this test. The internal temperature monitoring of the second container failed; however, the inner pipe temperature must have been much less since the polyethylene bottle did not melt, as can be seen in Figure 9. In each test, the gasket for the steel flange did not fail, thus retaining the solution in the steel pipe. in light of the fact that the prescribed test conditions are much less severe, both in temperature and time, the thermal insulation of the package would sufficiently protect the contents. However, for the purposes of criticality safety it will be assumed that the polyethylene bottle does melt and that the contents will be retained in the stainless steel pipe.

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6 y,, '[:l;, qf jf}4~ t w, e;,h!.,, \\ g a .ifll$Af } A ~ kl, ' {. , c/ -, +, t; .s c:' ' ' '. ;,f , g gy h W _p s t i 'g h, .a

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. fh; ~ ' ..., ti-l ' Y#/ . : 7.* iK*g. y n.' u,1 fy,,g:_y, g ;,p. q g,,,., o?. ~.; ., g,yyk, y&y.pQj a ['J [.fbdQ ) p ;;3_; g 'l ;:f$j!(jf}f)i) 4 t k &y & @? S Mjedi s j @> h' $$@26 x vy' fg u. .a ct e .a , 'g l ] Fipare 9. INTERNAL D '4-P w,,. First Test since there was . [r, , R.'j J <. Iy e f' $;%gp cQ, [, ' s .s '[ ', N ) %fw tM g . -[i, QD*O*c, ' E [' 'z;..q, y ... :n D.*

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116247 o B. INTERNAL DAMAGE AFTER A FIRE TEST AT 17000 F.

i ? 13 l l Q':.. ' 4.2 ,e, A' a:., :;a*%;,~;.. 'p/,e}p?(g,..-a'p;Qg; e '- l-.e,*t y ,, > f;;g:>p ,, ~'. p 'x' t + a v; g/- h', y.v,. v:yy fl[yj Jjfqj?Dfh(W):f.\\ ^l*b. _) y QW' kblkj}f; {h%hl*S W s;d d Pi! %l* W d IV d; \\l. k d [ !-$$i h li'l(@ D Y l-Y) YP, y-b23y e.b.Q%

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i :;tm i pwi m 7.l y;'s,' ^c vW' K ^ 08l %O r;a. N ws_ W 3'. 4 )ht Nin [;;c[, f,< ' k k:h ind t. 4 e i,ZCi f.e u,, i x.., v 115591 % MAGE TO A SECOND CONTAINEFl. (inner Pipe Temperature must have been less than for the no Damage to the Polyethylene Bottle) l l l

14

4. " Water immersion - immersion in water to the extent that all portions of the package are under at least 3 feet of water for a period of not less than 8 hours"- After cooling, the containers were submerged in water for 24 hours. As expected, water leaked into the outer container, saturating the vermiculite (Figure 9); however, there was no nieakage into the steel pipe inner container. The effect of water saturating the thermal insulation of the package does not compromise the nuclear criticality safety of the package or an

~ array of packages. NUCLEAR CRITICALITY SAFETY EVALUATION Calculations were performed using the S transport theory code, ANISN,(c) and the Monte n Carlo transport code, KENO.(d) Hansen Roach sixteen-energy group cross sections were used for all materials except two of the constituents of the vermiculite insulation. Sixteen group cross sections produced by the GAM Il(e) code were used for these constituents: magnesium and silicon. The calculational codes and cross sections are i considered to be well established on the basis of their success in calculating a large variety of experimental critical assemblies, as reported by Handley and Hopper.(f) j l The analysis was made for uranium enriched to 100% in the 2 3 s U isotope; therefore, the results are applicable to any enrichment. The hydrogen-to 2 3 s U atomic ratio was chosen to be in the range in which the critical volume is minimum; therefore, the results are applicable for any stable aqueous solution. Uranyl fluoride was chosen as the compound in solution; j therefore, the results can be conservatively applied to nitrate solutions or any other common aqueous solutions of uranium, i Material compositions and densities for the various materials used in the calculations are [ given in Table 1. An infinite cylinder mockup, used in the ANISN calculations, is illustrated in Figure 10. Finite cylinder mockups, used in KENO calculations, for normal and accidental conditions, are presented in Figure 11. Diameters of the mockups have been reduced by 7% to equate the triangular pitched array from square-pitched array calculations. The right-hand side of Figure 11 has the diameter reduced an additional one-half inch to account for damage resulting from the 30 foot drop test. 1 (c) Engle, W. W.; users Manual for ANISN, K-1693; Union Carbide Corporation-Nuclear Division, Oak Ridge Gaseous Diffusion Plant, Oak Ridge, Tennessee (1967). (d) Whitesides, G. E. and Cross, N. F.; KENO-A Multi-Group Monte Carlo Criticality Program, CTC 5; Union Carbide Corporation-Nuclear Division, Oak Ridge Gaseous Diffusion Plant, Oak Ridge, Tennessee (1969). (e) Joanov, G. D. and Dubek, J. S.; GAM II AB3 Code for the Calculation of Fast Neutron Spectra and Associated Multi Group Constants, GA-4265; General Atomic Division, General Dynamics Corporation, San Diego, California (1963). (f) Handley, G. R. and Hopper, C. M.; Validation Checks of the ANISNand KENO Codes by Correlation with Experimental Data, Y-1858, Union Carbide Corporation-Nuclear d Division, Oak Ridge Y-12 Plant, Oak Ridge, Tennessee (1972). r i l

15 Single Package Analysis Table 1 MATERI AL oEsCRIPTION FOR CALCULATIONS The criteria of USAEC regulations AECM-0529 and 10 CFR 71 specify that the single uranium solution (uo2 2) F package be so designed and constructed and

  1. ' '"i' "*' '

4 its contents so limited that, under the ""l"* normal conditions of transport and the gg, 33 .se hypothetical accident conditions, the pack-age will be subcritical. Polyethylene (CH )n 2 Density (9/cm3) 0.89 Under normal conditions of transport, the contents will be contained in the 4.875_ Tm a 4 stainiess sesei inch inside diameter polyethylene bott;e. oensity (g/cm3) 7.90 The minimum critical diameter of an infinite cylinder has been calculated (9) to carbon steel be about 10% more than the inside diameter censity (g/cm3) 7.82 of the bottle. Laminated Fir Plywood Under accidental conditions it will be Taken to be celtuiose at 1.0 g/cm3 density. assumed that the contents leak out of the polyethylene bottle into the steel pipe. KENO calculations of the package, as Packed censity,85 Pounds / Container (g/cm3) 0.134 shown in the right-hand side of Figure 11, had a ketf of 0.8610.02 when fully water Normal Composition in Terms of Atomic Densities 0' reflected. 24 atoms /cm3) Element Atomic Density (10 Based upon this comparison and calcula-Ai 0.00023: tions, the Model VCD package meets the Fe 0.000091 criteria for subcriticality for the single-H 0 000307 K 0.0 0120 package analyses. Mg 0.000421 o 0.00242 Array Analysis si 0.000524 1 The criteria of USAEC regulations AECM-0529 and 10 CFR 71 specify that a Fissile Class ll l package be so designed and constructed and its contents so limited that: (1) five times that number would be subcritical in any arrangement if closely reflected by water, and (2) twice that number would be subcritical in any arrangement if each package were subjected to the hypothetical accident conditions and the array was fully water reflected with optimum interspersed hydrogenous moderation. (g) Webster, J. W.; Calculated Neutron Multiplication Factors of Uniform Aqueous Solu-tions of 233U and 235, ORN L-CDC-2; Union Carbide Corporation-Nuclear Division, U Oak Ridge National Laboratory, Oak Ridge, Tennessee (1967).

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17 11 was assumed that, under accidental conditions, the polyethylene bottle does melt and that the contents are dispersed into the stainless steel pipe. Since the diameter of the pipe is greater than the critical, water-reflected, infinite cylinder diameter, the analysis of the array under accident conditions will be considered first. A series of ANSIN calculations were performed by Crume, et al, to search for the optimum i moderating conditions of an infinite array of infinite cylinders, shown in Figure 10. Results of the calculations show that, as the amount of hydrogenous moderation increases, the array reactivity decreases, with the most reactive state at normal dryness, as given in Figure 12. The decrease in reactivity is attributed to the neutron-absorbing qualities of the thick layer of stainless steel as the neutrons become more thermalized. 0

s-$-1 1

-2 '5 11 1.E -3 B $ ~<a ~.8 l l l 1 3 10 25 X Times the Normal Water Content of Vermiculite Figure 12. CALCULATED EFFECT OF WATER AS AN ARRAY MODERATOR. (k*, was Calculated for Normal Verrniculite Cornposition) A s~ies of KENO calculations were performed by Crume, et al, to search for the maximum number of containers which could be transported under the established criteria. The keff of a 4 x 3 x 2 high array of damaged containers, illustrated in Figure 11, was calculated to be 1 0.9310.02; the keff of a 9 x 9 x 3 high array of undamaged containers was calculated to be 0.89 10.02. Therefore, based on the 50-unit rule, each container must be assigned a minimum transport index of 4.2 to meet the criteria for subcriticality of a Fissile Class il shipment. OUALITY ASSURANCE Procedures are established to assure quality and compliance to the specifications in the fabncation and assembly of each package. In addition, assurance that the package continues to meet the requisite standards of quality in routine use must be maintained. Fabrication and Assembly of Packages . Fabrication of the new Model VCD packages to transport solutions may be performed in the plant of the shipper or by an outside vendor. A detailed listing of sizing, materials, and specifications has been prepared, and is provided in Figures 1,2, and 3. The fabricator is expected to meet these specifications and state in writing to the consignee that the package has been assembled accordingly.

18 The consignee, upon receipt of new packages, shall inspect each package, employing standard nondestructive methods. Tests must be performed by competent personnel to y assure that the package meets the proscribed specifications. Specifically, the inspection shall include, but not be limited to, the following:

1. Drum. The drum wall, lid, and ring clamp thickness shall be checked. Measurements must be within the prescribed specifications. If two 55-gallon drums are welded to make the 110 gallon container, the weld seam shall be visually inspected for quality and soundness.

Poor quality welds shall be reason for rejection.

2. Insulation. Visual inspection shall be made through the inspection holes in the top wood spaces. All void space between the inner container and outer 110-gallon drum shall be filled with vermiculite insulation.
3. Inner Container. Flange thickness, number and size of flange bolts, presence of an O ring, and inside diameter and inside height of the inner steel pipe shall be verified to be within specifications.

Measurements, data, and observations of the Inspection Department personnel shall be recorded for each package. The documented information shall be retained by the consignee for a period that is not less than the expected lifetime of the package or that required by applicable AEC manual chapters and Code of Federal Regulations. Any packages rejected are to be so labeled until proper disposition is made. Routine Irispection of in Use Packages Packages which have been accepted and placed in use for the interstate transport of fissile materials shall be inspected each time prior to the next use. Figure 13 is a reproduction of a typical form which may be employed when such inspections are made. The form requires identification of the package being inspected along with nine specific actions that are to be taken by personnel preparing to load the package. No specific department or individuals in the plant are required to perform this inspection. However, immediately prior to loading, personnel assigned the responsibility for loading the package must perform the inspection. If the package is rejected, it is to be labeled, removed from use, and returned for maintenance or further determination as to the feasibility of repair or destruction. ) CONCLUSIONS The nuclear criticality safety of loaded Model VCD packages was investigated using ) reactor theory calculations. The standards for safety used in this investigation are those given in AECM 0529 and 10 CFR Part 71 for Fissile Class ll packages. The analysis was done for uranium enriched to 100% in the 235 U isotope; thus, the results are applicable to any enrichment. The ratio of hydrogen atoms to 2 3 5 U atoms was chosen to be in the range in which the critical volume of aqueous solutions is a minimum; thus, the results are applicable to any stable aqueous solution. The particular compound in the aqueous solution was uranyl fluoride; thus, the results can be applied to any nitrate solutions or any other common aqueous uranium solutions.

19 OFF-SITE FISSILE MATERIAL CONT AINE R INSPECTION FORM se Model Number i Serial Number -- Date

1. Evidence of external cracks?
2. Does serial number and model number appear?

3, Note any obvious drum domoge and describe on back. 4. Is scaling ring intact?

5. Is retaining ring and bolt complete and serviceable ?

6, is inner filler domoged?

7. Verify presence of all inner parts.
8. Remove old shipping and/or empty labels.

9, Final container disposition:

a. Acce pte d
b. Rejected Packaging Foreman (sign)

Distribution: Copy 1 - Disp. Container Group' Copy 2 - Users File 'For fermanent file - to be maintained for audit purposes Figure 13. OF F SITE FISslLE MATERI AL CONTAINER INSPECTION FORM, Other conservative assumptions were incorporated in the calculations to determine the safe number of packages permitted for a single vehicle shipment, such as reducing the drum diameter by 7% to equate the square-pitched array calculations to triangular-pitched arrays and searching for the optimuni amount of hydrogenous moderation. Engineering analyses, comparative data, and destructive testing methods were used to demonstrate the structural integrity and thermal protection of the package with respect to the applicable standards for normal transport and hypothetical accident conditions. n Results of the Y 12 Model VCD package analyses show that the hypothetical accident conditions will not affect the capability of the package to maintain subcriticality and will prevent the loss of fissile radioactive materials.

20 DISTRIBUTION Atomic Energy Commission - Oak Ridge Yaggi, W. JJGoogin, J. M. Hickman, H. D. Y 12 Central Files (5) Pryor, W. A. Y-12 Central Files (master copy) Travis, W. H. Y-12 Central Files (route copy) Zachry, D. S.,' Jr Y 12 Central Files (Y-12RC) Paducah Gaseous Diffusion Plant Oak Ridge Gaseous Diffusion Plant Wilcox, W. J., Jr Edwards, A. K. Winkel, R. A. Oak Ridge Y 12 Plant Alvey, H. E. I Burditt, R. B. Briscoe, O. W. Burkhart, L. E. Butturini, W. G. Denny, A. Dyer, H. R. Ebert, J. W. Ellingson, R. D. Fraser, R. J. Gambill, E. F. Gritzner, V. B. Hensley, C. E. Johnson, C. E. Kahl, K. G. Keith, A. Kite, H. T. Lundin, M. I. McLendon, J. D. Mee, W. T. (10) Oliphant, G. W. Perry, A. E. Phillips, L. R. Smith, H. F., Jr Smith, R. D. Snyder, H. G. P. Weathersby, W. E. Whitson, W. K. in addition, this report is distributed in accordance with the category UC 71, Transportation a of Property and Nuclear Materials, as given in the USAEC Standard Distribution Lists for Unclassified Scientific and Technical Reports, Tl D-4500. l}}

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