Text

Application for Renewal of Certificate of Compliance 9191. W/Six Oversize Drawings.Fee Paid
	ML20248E739
Person / Time
Site:	07109191
Issue date:	03/28/1989
From:	Olsen A; BABCOCK & WILCOX CO.
To:	Macdonald C; NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
Shared Package
ML20248E744	List: ML20248E739; ML20259F740; ML20259F743; ML20259F748; ML20259F752; ML20259F757;
References
	25361, 89-043, 89-43, NUDOCS 8904120353
	Download: ML20248E739 (2)
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RETURNTOdEEd.

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l Babcock &h n v i nuci..r ru.i Division a McDerm,ott

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d' U.S. NUCLEAR REGULATORY,

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' U. S. Nuclear Regulatory Ccrt1 mission Mr. Charles E. MacDonald, Chie$

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Division of Safeguards & Transportati Mall SECTION f7 MNSS 9,

DOCKET CLERK Washington, DC 20555 4

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Reference:

Certificate of Compliance No. 9191 Gentlemen:

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'Ihis is to request the renewal of the referenced Certificate of Compliance.

'Ihe consolidated application for renewal is attached. This application does not contain any changes to the container or its use.

Your letter dated January 25, 1989, suggested that the renewal application follow the format of Regulatory Guide 7.9.

I have reviewed the Guide and the container file and found that following the Guide would result in changes to the original Safety Analysis Report that was developed by another company. 'In lieu of using the suggested fomat, this consolidated application contains all of the previously submitted and approved descriptions and test data far th6 container.

A check in the amount of $150.00 is attached to tieet the requirements of 10 CFR 170.31 tor this application.

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i If you have any questions regarding this application, please call me. '~~

Ycirs truly, o

Arne F. Olsen

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a McC,ew company P O Box 785 Lynchburg. Virgna 24505 0785 (804) 522 6000 U. S. Nuclear Regulatory Comission September 24, 1987 c.

Attn

d. E. MacDonald, Chief 87-165 Transportation Certification Branch Division of Fuel Cycle and Material Safety, tetSS Washington DC 20555 Gentlemen:

2 e Babcock & Wilcox Ccapany, Naval Nuclear Fuel Division is requesting a modification to Shipping Container USA /9191/AF, Certificate of Ccapliance Nmber 9191 in accordance with 10 CFR 71.31(b). De modifications'will enable this package to be used for transporting low enriched silicide/almine Petten Fuel elements (with loadings not exceeding 626 grame U-235 ;er element) as a Fissile Class I, II, or III shipment.

S e nuclear criticality safety analysis of the Petten fuel element transported in the 9191 Container is provided in the attached document fr a the NNFD Research Iaboratory to C. C. Boyd, dated September 2,1987 titled, " Nuclear Safety Analysis of the Unirradiated Fuel Shipping Container for Shipnent of

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Petten Elements".

Pending approval, please make the following changes to Certificate of I

Cmpliance Number 9191 dated April 23, 1984:

5 (b) Contents (1) Type.& form of materh1 Unirradiated uranim fuel element enriched in the U-235 isotope ccmposed of almine plates.

(2) Maxim m quantity of material per package ORNL-BNL Container Seven (7) uranuim silicide (U Si A1) fuel elements containing a maximm 3

of 626 grams U-235 per fuel eI nt with a U-235 enrichment of <20 weight t U-235.

A check for $150.00 is attached to cover the initial amendment fee.

NNFD requests a priority review by your office so that the necessary endorsements may be obtained for an international shipment.

Sincerely, l

ar ie C.' Boyd Jr.

Nuclear Safety & Licensing Officer CCB/sih Attachments UNCLASS1FIED

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References 1.

Certificate of Ccmpliance No. 9191 dated April 23, 1984.

l 2.

Safety Analysis Report for Packaging: h e Unirradiated Fuel Shipping Container ORNI/ENVIM-15, September 1979.

3.

Babcock & Wilcox Application dated April 5,1984.

4.

Nuclear Criticality Safety Assessment of ORR, NBS, and RFBR Fuel Element Shipping Package, J. T. Scmas, ORNI/CSD/'IM-77, January 1979.

I 5.

Union Carbide Imtter dated March 21, 1979 frca J. T. Bomas to R. W. Knight.

6.

Union Carbide Imtter dated September 10, 1979 frcm J. T. Scmas to' J. H. Evans.

7.

Babcock & WLicox Memo dated March 20, 1984 frcm M. N. Baldwin to N. R. Regan.

8.

Babcock and Wilcox Meno dated September 2,1987 frca M. N. Baldwin to C. C. Boyd.

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Dete NUCLEAR SAFETY ANALYSIS OF THE UNIRRADIATED FUEL SHIPPING CatfIAINER FOR SHIPMDff OF PETTEN ELEMDiTS September 2, 1987 This letter to cover one curio,ne, and one sutm only.

We analysis of the unirradiated fuel shipping container, designed by ORNL is complete.

The attached analysis shows that the container is suitable 626 grams U-235 per element.for shipping up to seven LEU, Petten fuel elements We attachment may be und as a basis for amending Certificate of Compliance #9191 to include the Petten elements.

h. N $Y M. N. Baldwin bp Q. A. Statement:

I have reviewed the work reported in this memo.

and output of five computer cases and compared this work to the previous workI have of Reference 3.

I concur with the methodology, the computer model, the results and conclusions.

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F. M. Alcorn UNCLASSIFIED b/7M./b s WEPl17 cuan,

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9/2A7 Page 2 Introdu$ tion The unitradiated fuel shipping container, designed by Oak Ridge National Laboratory and licensed by NRC (see Reference 1) for shipping R-2 fuel elements, is herein examined from a nuclear criticality safety point of view for its suitability for shipping Petten fuel elements. We overall safety analysis of the container has been reported in Reference 2 and the nuclear criticality safety analysis for R-2 fuel elements was detailed in Reference 3.

He Petten element is similar in geometry and composition to the R-2 element for which the container is already licensed.

Both are "MTR Type Elements".

Both contain proliferation resistant low enriched uranium (less than 20 wtt U-235), and both incorporate U 51/A1 as the. core material.

Each Petten element is made up of 20 fuel plales, and each plate is loaded 3

with up ter 31.3 grams U-235 (i.e., 626 grams for the entire element).

He fueled section of the element is illustrated in Figure 1.

the Petten fuel elements including end adapters is 35.748 inches. Total length of Method of Analysis h is criticality safety analysis, showing that the Petten fuel is acceptable in the shipping cask, is identical to that used in Reference 3 to show the acceptability of.the R-2 elements. TM only difference in the two analyses is the substitution of Petten elements for R-2 elements.

Both analyses follow that used by Thomas (Refs. 4, 5, and 6),

i.e., the KENO IV code and the Hansen-Roach cross section sets were used, nine element positions were assumed per cask rather than the actual seven, the steel forming the inner basket was associated with each element in an identical manner, the nature of the container was identienl, the geometric represen-tation of the container was identical, and a reduced density phenolic foam was assumed.

He validity of using the KINO IV code and Hansen-hoach cross-section set for LEU fuel (approximately 20% enriched) was demonstrated by a calculation of I

tk 'EU critical experiment which was conducted at the Ford Nuclear Reactor, University of Michigan, in December,1981. Wis calculation explicitly modeled each fuel plate and water gap as has been done in the present safety evaluation. mis benchmark calculation gave a result that exceeded the measured value by a aK of 0.027 0.007. Details of this calculation are reported in Reference 7.

Romas (Ref. 4) references the work of D. W.

Mugnuson for validation of the code and cross-section sets used with borate-phenolic foam.

Results cf Calculations

-Results of the calculations for casks containing Petten elements (each element loaded to 626 grams U-235 in a fuel matrix of U Si,3/A1) are presented in 3

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9/2/B7 Page 3 Tablesiand2.

Note that an infinite array of packages under normal conditions gives a very low K-eff of less than 0.3.

When the interiors of the casks are flooded and no water is present between the casks to provide isolation, K-eff reaches a maximum.

In this latter case, K-eff reaches 0.95 when nine elements per cask are assumed loaded and the cask is assumed damaged.

When the number of elements is reduced to seven, by filling two element positions with water (designated by the letter "W" on the arrangement diagram in Tables 1 and 2), K-eff of the infinite array of packages is well below 0.9 for all postulated conditions.

These results are almost identical to those calculated for the R-2 elements, for which the cask was licensed in 1984.

This is explained by a comparison of the K-eff for single, fully flooded and reflected Petten and R-2 elements.

As shown in Table 3, their K-eff values are virtually identical.

Table 3 also shows that reasonable differences in the Si and Al content of the fuel meat of the Petten elements has essentially no effect on the resulting K-eff value.

As a part of this evaluation ~ effort, eight cases run in 1984 to show the R-2 elements acceptable in this cask were repeated.

within two standard deviations of the previous results.All re-run results were Conclusions Calculations show that up to seven LEU /silicide/ aluminum Petten fuel elements with loadings not exceeding 626 grams U-235 per element, meet the requirements of a fissile Class I package in transport, when loaded unitradiated fuel shipping container identical to the ones licensed under to an Certificate of Compliance 9191.

Since a fissile Class I package is more restrictive than a fissile Class II or III (from a nuclear criticality safety viewpoint), the packace evaluation is acceptable for all class shipments.

References 1.

Certificate of compliance No. 9191 dated April 23, 1984.

2.

Safety Analysis Report for Packaging:

l Container ORNL/ING/m-15, September 1979.The Unitradiated Fuel Shipping

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Babcock & Wilcox Application dated April 5, 1984.

l 3.

t 4.

Nuclear Criticality Safety Assessment of ORR, NBS, and HFBR Fuel Element Shipping Package, J. T. Thomas, OPNL/CSD/*m-77, January 1979.

5.

Union Carbide Letter dated March 21, 1979, from J. T. Thomas to R. W. Knight, 6.

Union Carbide Letter dated September 10, 1979 f rom J. T. Thomas to J. H.

Evans.

7.

Babcock & Wilcox Memo dated March 20, 1984, from M. N. Baldwin to T. M.

Alcorn.

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a Table 1 COMPtfrED K-eff VAINES FOR 'IHE I.NDAMAGED CASK Number K-eff 2e & Run Identifier Arrangement of Casks condition 9 Ele W Cask 7 Elem/ Cask of 7 Elements Infinite No water present 0.286- 0.004 Array'

( AUG87 *PE'fA)

Infinite casks flooded and 0.808 0.008 1

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Array water between (AUG87*PETB) casks I

Infinite casks flooded and 0.911 0.006 0.842 0.008 EEW l

Array no water between

( AUG87 *PE'IC)

'(AUG87*PETF)

EEE casks EEW Table 2 i

COMPUTED K-eff VALUES FOR THE DAMAGED CASK hr K-eff 2e & Run Identifier Arr g m M of Casks Condition 9 Ele WCask 7 Ele CCask of 7 Elements

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Infinite Casks flooded and 0.813 0.008 Array water between (AUG87*PETL) casks Infinite casks flooded and 0.937 0.008 0.855 0.008 EEW j

1 Array no water between (AUG87*PETE)

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3 Petten 626 gm U Si,3/Al +10% Si 0.467 0.006 AUG87*PETL 3

3 Petten 626 gm U Si,3/Al -10% Al 0.467 0.006 AUG87*PETM 3

3 Petten 626 gm U Si /A1 0.463 0.006 AUG87*PETF 3

3 R2 500 gm U Si /Al 0.464 0.006 AUG87*R2A 3

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. a McD.rmott company P. OJ Box 785 Lynchburg, Virginia 24505 0785 (804) 522 6000 April 5, 1984 Docket No.'71-91 I

D. S. Nuclear Regulatory Commission Attus. C. E. MacDonald, Chief.

Transportation Certification Branch Division of Fuel Cycle & Material Safety, NMSS j

Washington, D. C. 20555 Gentlemen:

_ The Babcock and Wilcox Company, Naval Nuclear Fuel Division requested a modification to shipping container USA /9853/B( )F, Certif; este of Compliance Number 9853, in a letter from N. A. Regan of B&W to C. E.

MacDonald of the NRC and dated March 22, 1984. This letter replaces the above referenced letter in its entirety.

B&W, NNFD is-now requesting a different modification to the USA /9853/

B( )F shipping container which will involve the issuance of a new certificate of compliance and package identification number., Because this modification involves the use of this container for shipments of Type A quantities of material per 10 CFR Part 71, B&W is requesting the container be redesignated AF for low enriched fuel shipments. Per a' telephone discussion with R. H. Odegarrden of your office, the new Certificate of Compliance No. would be USA /9191/AF and the new model No. would be 9191.

This predesignation would e'nable this. container to be used more expeditiously for international air shipments. B&W intends to use'the ORNL or BNL redesignated container for transporting R2 Materials Testing

'and Research Resetor elements to European customers.

The nuclear criticality safety of the R2 elements transported in the 9853 type container'is provided in the attached Babcock and Wilcox Company, Lynchburg Research Center letter from M. N. Baldwin to N. A.

Regan, dated March 20, 1984; " Nuclear Safety Analysis for the Unirradiated j

Fuel Shipping Container for Shipment of R2 Elements".

i Upon approval, the following will be the only changes necessary to Certificate of Compliance USA /9853/B( )F for predesignation as Certificate of Compliance USA /9191/AF:

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T Babcock & Wilcox April 5,1984 5(b) Contents (2) Maximum quantity of material per package ORNL-BNL Seven (7) uranium silicide U3S12/A1 fuel elements containing no more than 500 grs:_s U-235 per element with enrichment 20 wt % U-235.

i References Safety Analysis Report for Packaging:

The Unirradiated Fuel Shipping ORNL/ ENC /TM-15, September 1979.

Nuclear Criticality Safety Assessment of ORR, NBS, HFBR FUEL Element Shipping Package, J. T. Thomas, ORNL/CDS/

TM-77.

Union Carbide letter September 10, 1979.

I ORNL letter dated September 18, 1979.

I Department of Energy letter dated November 1, 1979.

LRC letter dated March 20, 1984.

B&W has classified this request as an administrative amendment and a check for the applicable fee is enclosed.

We request a priority review by your office so that the necessary endorsements may be obtained for international. shipment.

Sincerely, BABCOCK & WILCOX, NNFD N. A. Regan Nuclear Sa ety & Licensing Officer NAR:ah Attachment

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U. S. Department of Transportation Attu:

R. Rawl Materials Transportation Bureau Office of Hazardous Materials Regulation (ORMR)

Washington, D. C. 20590

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Contract No. V-7!.05-eng-26 ORNL Engineering 1

SAFETY ANALYSIS REPORT FOR PACKAGING:

THE UNIRRADIATED FUEL SHIPPING CONTAINER J. H. Evans W. D. Shipley R. W. Mouring Date Published - Septe=ber 1979 l

OAK RIDGE NATIONAL LABORATORY cak Ridge, Tennessee 37830 operated by.

UNION CARSIDE COR.?ORAIION for the g j,g 4 - yQ }__

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.o COSTESTS Page ABSTRACT I

1.

GESERAL INFORMATION I

1.1 Introduction I

1.2 Package Description...

2 1.2.1 Container description 2

1.2.2 Operational features.

5 1.2.3 Contents.

5 2.

STRUCTURAL EVALUATION 6

2.1 Mechanical Properties of Materials 6

2.2 General Standards for All Packages 6

2.2.1 Closure 6

2.2.2 Container lifting device.

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2.2.3 Lid lifting device.

-7 2.2.4 Tie-down devices.

8 2.3 Standards for Type B and Large-Quantity Packaging.

8 2.3.1 Load resistance 8

2.3.2 External pressure 10 2.4 Compliance with Standards for Normal Conditions of Transport 10 2.4.1 Heat 11 2.4.2 Cold.

12 2.4.3 Pressure.

12 2.4.4 Vibration 12 2.4.5 Water spray 12 2.4.6 Free drop 13 2.4.7 Compression 13 2.4.8 Penetration 13 2.5 Compliance with Standards for Hypothetical Accident Conditions..

14 2.5.1 Free drop 14 2.5.2 Puncture.

15 2.5.3 Water i==ersion 16 3.

THERMAL IVA*_JA!!OS.

16 3.1 Nor:al Cenditions ef Transport 16 3.2 Thermal Accident 15 4

COSTAISMIST 13 5.

SHIELOISC 19 l

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QUALITY ASSURANCE 19 7.1 Fabrication, Inspection, and Acceptance Tests.

19 7.2 Routine Operating and Inspection Procedures...

'20 7.3 Periodic Maintenance and Inspections 20 7.4 Records.

20 REFERENCES 21 Appendix A - FA3RICATION DRAWINGS AND DATA SHEETS.

A-1 Appendix 3 - FUEL ELEMENT DRAWINGS B-1 Appendix C - ROUTINE PACKAGING AND INSPECTION PROCEDUPIS C-1 1

Appendix D - PENETRATION TESTS D-1 Appendix E - APPROVAL DOCUMENTS.

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LIST OF TABLES Table M

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Design sum =ary of unirradiated fuel containers 4

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Mechanical properties of container :aterials 7

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LIST OF FIGURES Figure Page 1,1 The unirradiated fuel element shipping container.,,,,

3 2.1 Container as a simple beam 9

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SAFETY ANALYSIS REPORT FOR PACKACING:

THE UNIRRADIATED TUEL SHIPPING CONTAINER J.

P., Evans W. D. Shipley R. W. Mouring A3STRACT The Unirradiated Fuel Shipping Container was designed at the Oak Ridge National Laboratory for the transport of unirradiated fuel elements. The container was evaluated analytically to determine its compliance with the applicable regulations governing containers in which radioactive and fissile materials are transported, and the evaluation is the subject of this report. Computational and test pro-cedures were used to determine the structural integrity and thermal behavior of the container relative to the general j

standards for normal conditions of transport and the standards for hypothetical accident conditions. Results of the evalua-tion demonstrate that the container is in compliance with

'the applicable regulations.

1.

CENERAL INFOR.V.ATION 1.1 Introduction The Unirradiated Fuel Shipping Container was developed at the Oak Ridge National Laboratory in 1978; three containers of differing di=en-sions were designed.

The containers will be fabricated, inspected, and filled with phenolic f oam in accordance with the drawings in Appendix A.

The primary use of the containers is to ship fuel elements. The containers provide impact and thermal resistance for the fuel elements during transport for both normal and accident conditions by rail, highway, air, and water modes.

The contents for which the design was evaluated are outlined in Sect. 1.2.3.

They ec= ply with Department of Inergy regulations contained in the Code of Federal Regulations, Title 10, Fart 71;l in the AEC Manual Chapter 0529;2 and in all :::ediate

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Action Directives in ef fect as of this report date.

The containers also comply with '.*.S. Department of Transportation regulations published in the Code of Federal Regulations, Title 49, Part 173. 3 Calculations, engineering logic, test results, and docu=ents demonstrating compliance are presented in the following sections of this report. Copies of the approval documents are reproduced in. Appendix E.

1.2 Package Description ~

1.2.1 Container description The features of the package are illustrated in Fig. 1.1 and listed in Table 1.1; fabrication drawings are in Appendix A.

The container consists of a basket and a cylindrical housing. Each container has an outside dimensien (OD) of 24-1/2 in., except for the National Bureau of Standards container, which has a 26 in. OD.

The basket for the Oak Ridge Research Reactor (ORR), Bulk Shielding Reactor (BSR), and Pool Critical Assembly (PCA) fuel elements is 39-7/8 in., with the overall length of the container being 56-5/8 in.

The basket for the Brookhaven National Laboratory (3NL) fuel elements is 38-3/4 in., with the overall length of the container being 75-1/2 in.

The basket for the National Sureau of Standards (S35) fuel ele =ents is 70-5/16 in., with the overall length of the container being 87-1/8 in.

The area between the basket and outer housing is filled with phenolic foam. The phenolic feas provides some structural support and is an ef f ective thermal insulator. Vent holes 3/16 in. in dia:eter are located in the sides and bottom of the external housing for release of gas pressure that may be generated by deco -

position of the phenolic in a fire. After the container is filled with foam, these vent holes are sealed against the weather by two coats of an

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epoxy paint that d:es not react with the phenolic foam.

The epoxy seals are designed to vent in a fire, since the epoxy paint vill decompose at i

j temperatures well below 1000'?.

This ability has been demonstrated by l

Mallet and Newlon on verv si:11ar epoxy seals.-

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The closure of the basket is afforded by 16-in.-0D, 1/8-in.-thi:k l

flanges. Eight 3/3-in. hex-head bolts secure the closure.

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Tne ur. irradiated fuel elecer.: shipping cor.:ainer.

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BM%%M2Lha51%M.5& MMMDi.42L%%L f2RE2XLhdid.haha:D M d 2 Laid.% 2.Ri8 E E h 2 n A 6.

5 structure of lumber 5-1/2 in. deep. A fea=-filled plug 5-3/4 in. high fits on top of the inner housing and is secured by 6-5/8-in. hex-head bolts with lock washers and nuts.

i The materials of construction, as specified on UCC-ND data sheets DS-XDE 10191-1,2,3, and Rev. 1 are stainless steels and alucinum to ASTM specifications.

The inner container is a basket that is permanently fixed in the top of the outer container by four 2 x 1/2 x =5 in.-long phenolic spacers bolted to the basket and container by eight 1/4-in.-diam bolts with lock washers and nuts. The bottom of the basket is secured by four threaded rods welded to the bottom of the container.

The basket has a capacity of seven fuel elements. Three differ' ent containers have been constructed to ship different size fuel ele =ents.

The 'only difference in the con-tainers will be the overall length, basket length, and basket cross section.

The weights of the baskets are 119, 92, and 66 lb.

The net weights of the packages (cylinder plus basket plus forklifting device) are 660, 550, and 450 lb.

Therefore, all calculations will be based on a gross weight of approximately 850 lb.

Compliance of this heavier package with design specifications demonstrates that the lighter packages would be well within design li=1ts.

1.2.2 Operational features The container is specifically designed to house seven fuel elements.

Three dif f erent containers are available to transport ORNL, N35, or BNL fuel elements.

1.2.3 Contents The containers are designed to transpert seven f resh or unirradia ted fuel elements from the fabricator to the applicable reactor site. These I

reactors are (1) the ORR, the SSR, and the PCA, located at Oak Ridge, I

Tennessee; (2) the National Bureau of Standards Research Reactor (NSSRR),

located near Washington, D.C.; and (3) the SNL High-Flux 3eam Reactor (HT3R), located at Upton, N.Y.

The elements are of si=ilar construction.

h, g

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i t

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They are curved plate with U 0 -A1. cores clad with type 6061 aluminum.

36 The mini =um 235U enrich =ent is 93%. Other characteristics of the elecents are listed in Table 1.1, and drawings of the elements are in Appendix 3.

l 2.

STRUCTURAL, EVALUATION The package complies with the structural requirements of the regulations.1 The c...alational results and engineering logic presented in the following subsections de.monstrate compliance with these performance criteria.

Additional evaluations considered pertinent to the safety and operability of the package are included.

The effects of both nor=al and specified accident conditions on the structural integrity of the package are considered.

2.1 Mechahical Properties of Materials The values shown in Table 2.1 are conservative, but are typical of container materials.5-7 The static properties shown provide a reasonable approximation of dynamic conditions if care is taken to ensure that the use of static values results in a conservative evaluation. For most structural materials the static properties are much less than the dynamic properties. Nucerous reports substantiate this, including ref. 8.

2.2 General Standards for All Packages The general standards fer all packaging cover the chemical and galvanic reactions of the =sterials of the package and closure of the package.

The containers will be constructed of type 300 stainles's steel filled with phenolic foam. There has been no evidence of any cerresive reactions between these materials and the contents.

l 2.2.1 Clesure The standards specify that the package be equipped with a positive closure that vill prevent inadvertent opening.

In this case the lid ef

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= Table 2.1 Mechanical properties ef container materials Type 304L Preperty Sy=bol stainless steel Carbon steel j

Yield stress (psi) 28,000 30,000 Ulti= ace tensile strength (psi) 75,000 48,000 Modulus of elasticity (psi) 2B x los 29 x 106 Elongation in I

2 in. (%)

50 20 f

Density o

I (1b/in.3) 0.290 0.283 Allowable shear

stress, Tmax =

/2 (psi) Y Imax 14,000 15,000 the basket is ;acured by eight 3/8-in. hex-head bolts with lock washers and nuts and qualifies as a positive closure. Closure of the outer lid is attained by tix 5/8-in. bolts with hex nut and lock washer.

2. 2. 2-Container lif tina device If there is a system of lifting devices that is a structural part of the package, the regulations require that tais system be capable of l

l supporting three ti=es the weight of the leaded package without generating l

stress 'n any eaterial of the package in excess of its yield strength.

This section is not applicable because there are no lif ting devices, and the c:ntainer is designed fer forklif t handling.

2.2.3 Lid liftine device The lids cf the outer housing and basket are of such size and shape

)

as to eske hand lif ting a;;11 cable and t: preclude the need for any inte;ral lid lifting device.

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4 2.2.4 Tie-down devices Eccause these containers have no integral system ef tie dovns, no calculations as to force-withstanding capacity are required by the general standards.

In transit the containers are secured by the same techniques used for conventional cargo without lifting lugs.

2.3 Standards for Type 3 and Large-Quantity Packaging The structural standards for large-quantity packaging cover load resistance of the packaging and the external pressure which the package must withstand.

Compliance of the Unirradiated Fuel Shipping Container with these requirements is discussed in the following subsections.

2.3.1 Lead resistance When regarded as a simple beam supported at its ends along any major axis, the container must be able to withstand a static load nor=al to and uniformly distributed along its length r. hat is equal to five times its fully leaded weight without genes. ting stress in any caterial in the shield in excess of the yield strength of that material.

The equivalent i

cross section of the container analyzed in this study is illustrated in Fig. 2.1.

The cross section of the container is composed of the outer 1/8-in.

stainless steel shell, the phenolic fea=, and the 16 gauge stainless steel (CA. SSI) basket as shova in Fig.1.1.

For a thin shell the secent cf inertia about its center is r/2 2

I=

v dA = 4 (r 2 sin 2 9)tr ds=fr I t

m m

m where r = radius (in.),

t = thickness (in.).

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= =.

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=

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2 t

2 82121.5

= 2121.5 (a) SHIELD AS SIMPLE BEAM

)

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(b) EQUIVALENT CR' OSS SECTION OF SHIELD i

Fi ;. 2.1.

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Neglecting the 'ef f ect of the phenolic foam and the basket,.the moment of

'j inertia of the container is I = st(r)3

= w(0.125) (13.0)3

= 862.8 in.".

The maximum bending stress is given by the expression o = Me/I - SWLro/8I; so a = (850)(87.125)(13.0)/6(862.8) = 697 psi,

~

where L is the length r> the outer housing. Since this stress is less than the yield stress for either housing material, it is concluded that the container compiles with the simple beam requirement.

2.3.2 External pressure.

The regulations require that the shipping package be adequate to ensure that the containment vessel will suffer no less of cencents if,

subjected to 25 psig external pressure.

For the purposes of stress calculations, it will be assumed that the foam-filled. cavity is at the external pressure. One potential consequence of c.:ternal pressure is buckling of the cylindrical shelfs of the shield.

Since'the outer shield housing is not airtight, there can be no pressure differential across it.

Since the inner housing (basket) is not scaled, there can be no pressure differential across it.

Therefore, this section is not applicable.

2.4 Conp11ance with Standards for Normal Conditions of Transport The regulations for normal conditions of tansport for a single package require that the effectiveness ef the package will net be sub-s:an:ially reduced by the normal crndi:i:ns :f transport, that there will Mw:rr c

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er 11 be no release of radioactive material from the centain=ent vessel, and that there will be no gases er vapors in the package which could, through any credible increase of pressure or an explosion, significantly reduce che effectiveness of the package.

There is no circulating coolant other i

i chan atmospheric air, and there is no =echanical cooling device required or provided. The container and basket are so designed that the contents will not be vented to the atmosphere under normal conditions of transport.

These conditions include the effects of heat, cold, pressure, free drop, and penetration.

The container should exhibit acceptable perfo'rmance with respect to all of the normal transport conditions.

2.4.1 Heat The package must be able to withstand direct sunlight at an ambient temperature of 130*F in still air without reducing the effectiveness of the packaging. A simplified one-dimensional heat transfer analysis (Sect. 3.1) was used to compute the steady-state temperature. distribution in the package under the specified conditions.

The surface temperature under these conditions would be approximately 208'T, and the temperature at the center of the package would be approximately 208'F.

i The temperature that would exist in the package will not adversely affect the container.

The materials of construction do not suffer signifi-cant loss of physical properties at these temperatures.

The maximum service temperature of the phenolic foam 3 is 300'F, which is also well above the temperature encountered.

The regulations 3 set ferth by the Department of Transportation further stipulate that the temperature of any accessible surface of the fully 1:aded shipping package shipped by c: mon carrier shall not' exceed 122*? when the pa:kage is in the shade in still air at an ambient tempera-ture.

?c: en s computati:n, an ambient temperature of 130'? is assumed, and since the containers have no internal heat 1:ad, the temperature of the sunface will be 100*?.

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Ndk2bI2NEMkb$$i1h$bbbhfIbdNbbbbb5Isb.htbb5[$kNt.NNbhb,

$)bbbbkbk-12 2.4.2 Cold i

The shipping package must be able to withstand an ambient temperature of -40*F (420*R) in still air in the shade. Because the containers have i

no internal heat lead, the temperature of the entire package will be

-40*F, and because the containers are not sealed, no pressure differen* 4 will result.

At -40'T the type 304L stainless steel to be used would not suffer brittle fracture because its transition temperature is considerably lower than -40'F.

In addition, the container closure bolts will be fabricated of stainless steel to avoid the low-temperature-embrittlement problem.

j 2.4.3 Pressure The regulations for normal conditions of transport specify that the package be able to withstand a reduction in atmospheric pressure to 0.5 atm (the resulting pressure being 7.35 psia) at the 130*F nmbient temperature condition. Because the basket is not sealed the pressure in -

i the basket will be the same as the atmospheric pressure.

Therefore, this section is not applicable.

2.4.4 Vibration The outer housing of the container is to be 1/8-in. stainless steel with a 26-in. outside diameter (maximum sized -- NBSRR - container);

transport vibrations would not be expected to affect the integrity of the container. The phenolic foam shoul'd further isolate the basket against vibrations.

The bolts that secure the lids are equipped with lock washers to prevent loosening.

2.4.5

a t er sprav The centainers will be fabricated of stainless steel and will not suf fer appreciable corrosive da= age during the water spray test.

The phenolic foan is also not affected by water spray.

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2.4.6 Tree drop The regulations f er nereal conditions of transport require that a package weighing less than 10,000 lb be capable of withstanding a free drop through a distance of 4 f t onto a flat, essentially unyielding

,i horizontal surf ace, striking the surface in a position in which maximum 1

1 damage is expected to result. Under this impact, there should be no 1

I release of package contents (Sect. 2.5.1).

The tests reported in these

]

sections are much more severe; therefore, it can be concluded that the package would survive the 4-f t drop without loss of containment or integrity.

2.4.7 Compression It is required that packages weighing less than 10,000 lb be capable of withstanding a compression load of five times the container weight, or 2 psi distributed unifor=1y across the top and bottom, whichever is greater.2 In this instance, five ti=es the gross weight of 850 lb is somewhat greater than that of the 2-psi condition. The maximum compressive stress in the walls is c - P / A = 5'a*/ rD : t = (5)(850)/r(25.875)(0.125) = 418 psi, where Do = mean dia=eter of outer wall (in.),

t = vall thickness (in.).

Thus the package can support five ti=es its weight with relatively low I

stresses generated in the walls.

2.4.S

?enetracien The regulations fsr normal conditions cf transport also stipulate that the package be capable of withstanding the impact of the flat end of a vertical steel cylinder which weighs 13 lb, has a diameter of 1-1/4 in., and is dropped fro: a height of 4 ft, normally onto the expesed surface of the packa;e that is expected to be the =ost vulnerable to pun: cure, T.is :es: vould n:t reduce the effectiveness cf the :entainer

_ ~

U$i,$$$$iW$5$$b?$hkNikA.$$

$$$Y A YO5 i

1 14 and would result in no more than a dent in the surface of the container.

Tests were conducted at ORNL to verify these conclusions (see Appendix D).

i j

2.5 Compliance with Standards for Hypothetical Accident Conditions The standards for the hypothetical accident conditions stipulate that a package used for the shipment of fissile or a large quantity of 1

i radioactive caterial shall be so designed and constructed and its contents so limited that if it is subjected to the specified free drop, puncture, ther=al, and water i=mersion (fissile material packages only) conditions, the reduction in containment would not be sufficient to increase the external radiation dose rate.to more than 1000 millirems /hr at 3 f t from the outside surface of the package; no radioactive material would be released frem the package except for gases containing total radioactivity not to exceed 0.1% of the total radioactivity of the contents of the package; and the contents would re=ain suberitical.

The following sections vill show compliance with the standards for hypothetical accident conditions.

2.5.1 Free drop The first in the sequence of hypothetical accident conditions to which the container =ust be subjected is a free drop through a distance of 30 ft onto a flat, essentially unyielding surface.

The container =ust strike the surf ace in the position expected to cause maximum darage.

A full-scale test model of a fire and impact shield of construction similar to that of the unirradiated shipping container and the ORNL Gas-Cylinder Fire and Impact Shield was designed, fabricated, and tested the Cak Ridge Gaseous Diffusion Plant as reperced by Mallett and at Sevion."

T.ese tests demonstrated compliance with the regulations after being subjected to the regulatory accident.1-3 The fire-resistant phenolic foa: insulation used in the test container was the same as tha.

to be used in the subject container. The volume of determination of the

- tested shield did not exceed SP. for this package and would indicate no damage to the inner c:ntainer.

l l

hen 2 a

m i

e e.a J

m nqrg.p.ym m ;r n?t a s;W A:

mop.w : guy.%q.;i.W:

1.:>.Kr wi.ah.::;.A.Exu2A&ClE.m.W[:n < :'V7:i,lra u/O;D ik.^.d&M:D'.aa %;;,;9 M D W %.A wit:. *W n/*Kgy,[MWTMC D.hWCWl3bh!n:!b 15 i

l The drums tested by Mallett and Newlon' had locking ring type closures whereas the ORSL gas-cylinder fire and i= pact shield and the unirradiated fuel shipping containers have a uniform closure design.

The gas-cylinder package utili:es 8-3/S-16 USC bolts, whereas the unirradiated fuel container has 6-5/8-11 UNC bolts.

The Mallett and Newlen closure utilized a single 3/S-16 UNC bolt in the locking ring closure."

The flanged and bolted closure employed ca the containers is designed to reduce the possibility of loss of lid in the event of an impact such as a 30-ft free-fall. Note that the closure is not sealed and does not form part of either the primary or secondary containment. To comply with regulations the lid need only remain in place in order to S

maintain thermal (fire) protection. Closure flanges and bolts are fabricated from ductile materials t;.at would deform locally near the point of i= pact, if subjected to a free-fall. We acknowledge that failure of the fastener directly at the impact point is possible, though not probable.

Such a failure would not change the magnitude of permanent deformation of the container from the model tested at CRGDP."

2.5.2 Puncture The second in the sequence of hypothetical accident conditiens to which the package must be subjected is a free drop through a distance of 40 in to strike, in a position in which maximum da: age is expe::ed, the top end of a vertical, cylindrical mild-steel bar mounted on an essentially unyielding horizontal surface.

The mild-steel bar shall be 6 in, in diameter, with the top hori: ental. and its eilge rounded to a radius of not ecre than 1/4 in., and of such length : hat it will cause maxinu damage to the container bu: not less than 3 in.1:ng.

The long axis of this bar shall be normal to the surf ace of the con:ainer upon i= pac:.

In their report en a similar package, Mallet: and Newl:n' presented the results of a pune:ure :es:.

Their centainer was d::pped en 1:s side and was inden:e2 abcu: 1.15 in, wi:heut rup:ure of the wall.

Ee:ause of the similar design, 1; is ::::1uded :ha: the subje:: shield w: ;d behace in s.;ee fashion and :ha: n: rup ure w:ul: :::_r.

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[b h

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16 2.5.3 Water i=ersion L

Thn final condition'in the specified hypothetical accident is i=ersion 'in water to a depth of 3 f t for a period of 24 hr.

The container is not sealed,. so the container cavity would be co=pletely filled with water. - The primary containment (alu=inus clad fuel) would not be breached as a result of any of the accident conditions; hence the unirradiated fissile material vould not be in contact with the water, and no radioactive caterial vould be transferred to the water.

The fully leaded container filled with water and fully reflected by water remains suberitical.10 The phenolic foam of the container does not degrade upon water immersion; thus the criticality calculations are valid during the entire time of the immersion test.

3.

THERMAL EVALUATION The package must remain effective af ter exposure to severe thermal environments. Applicable normal and accident environments are specified i

in the regulations -3 and are discussed below. ' Analytical evaluations and tests have been utilized to demonstrate the ability of the package to re'=ain effective after exposure to the specified environments.

3.1 Normal Conditions of Transport The package.nust be able to withstand direct sunlight at an ambient-temperature of 130*F in still air without reducing the effectiveness of the packaging. A simplified one-di=ensional heat transfer analysisil was used to co=pute the steady-state temperature distribution in.the package under the specified cerditiens.

To decer !.ne the te:;erature of the outside surface of the container, an expressien that includes convective and radiative heat transfer effectsU is used. The expression is of the for:

q=q"q r,

s q = h 3-(T - ! ) + 0.173 E -A f((T + !.603/(103)]'

: s a

s j

3

- ( (T, ' ' 6 : ' / (100 s."

/1)

t'17 & n :n u s? %. :.;1 q g, w %g <a n K:w k u : % q k r w. g,4 W ;l Wi. h w :

k. eg 2 an ha 0a w w-.u. w c. w w x ;: : w.w.4 w %:m. G e n. L F W 't ; & &.V M ; M M ; Q % T TM7Fa

.%uws% 4h W T c.:.u.su:Li.u.at

1 17 where q

convect heat flow (Btu /hr),

=

q = radiative heat flow (Btu /hr),

h, = convective heat transfer coefficient 12 = 0.19(T, - T )l/3 3

Btu /hr ft2.*y, A = convective heat transfer area = 52.1 ft,

2 T, = surf ace temperature (*F),

T, = ambient temperature, 130*F, 2

A = radiative heat transfer area = A (ft ),

q

= emissivity of the, painted shield surface (ref.12, Table 5.2),

c F12 = gray body shape factor 12 (%. 6).

The heat lead q on the package is only the solar heat load.

Because of the cyclic nature of solar heat load and the large time constant associated with containers of this type, the solar heat load per unit area can be approximated by numerically integrating the area under a normal incident solar energy curve (Fig. 5.3 in ref. 12) and is equal 2

to 144 Etu/ft.hr.

(This average solar heat flux is the same as was used in an example problem on p. 143 in ref. 12.) The solar contribution 1

is the only heat load in this calculation, and the exposed surface area vill be conservatively calculated as the convective heat transfer area, including the top and sides but not the bottom:

A = rr2 + 2rrl r

= -(15. 25)2 + (2)(n) (13.0) (82. 875)

= 7500 in.2 = 52.1 ft2, Tne solar heat load is q, = (144)(52.1) = 7500 Stu/hr.

j

hf[ NUN 5bbskbNbh[blfdb$bb.f kbbbbb bbh,.bhhbib b$hbbh I

18 The net heat load at the surface is q = 7500 Beu/hr.

The surface heat transfer equation (Eq. (1)] was solved via an l

iterative process to determine the surface temperature T, corresponding to q = 7500 Stu/hr.

In this case T, a 208'F.

Since there is no internal heat load, the maximum temperature of the container is the same as at the surface, 208'F.

3.2 Thermal Accident l

The third in the sequence of hypothetical accident conditions specified by the regulation to which the package must be subjected is exposure for 30 min within a source of radiant heat having a temperature i

of 1475'F and an emissivity coefficient of 0.9.

For calculation purposes the surface of the package is assumed to have an absorption coefficient i

of 0.8.

The package shall not be cooled artificially until 3 hr af ter the test period has ended, unless it can be shown that the temperature on the inside of the package has begun to fall in less than 3 hr.

Similar packages with the same fire-resistant foam insulation, as reported by.Mallect and Newlon, were tested according to the above j

requirements. The complete results indicate that at no point in the package inner liner will the temperature reach 200'F.

It is concluded, therefore, that the fuel shipping container will not exceed 200'F under the same conditions.

4 C ONTAI.W.ENT I

!: is requiradb3 that packages =aintain containment.' their radic-active centents during normal and specified accident conc: t.ons.

The contain=ent boundaries and capabilities of the fuel shipping container are outlined below.

i Il' In this package, the fuel ele =ent cladding constitutes the ;rizary containment cf the nuclear =aterial. The basket lid and outside housing lid will p:cvide further containment of the fuel unirradiated elenents.

i

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U 19 5.

SHIELDING l

This section is not applicable, since the container will not be used to transport irradiated fuel elements.

6.

CRITICALITY I

l l

The container loaded with any.of the five types of fuel listed in

{

Table 1.1' may be used as a Fissile Class I package.10 l

j The analyses 10 conservatively assumed that the package contained nine fuel elecents in 3 x 3 square matrix instead of the proposed seven in a triangular matrix.

It was also assumed that the principal da:aga to the package as a result of package test requirements (10-CRF-71, Ene :gy)*was

,i i

that resulting from a fire. Tetts 13 with the borated phenolic foam 4

insulation indicated 'a char depth of from 2 to 2.5 in. =ay be expected.

I The damage package evaluations were performed for a charred depth of 3 in.

The absence of water from the inner container results in a very low K., less than 0.2.

Water in the inner container, or fuel element baskets, results in a K. no greater than 0.9.

7.

QUALITY ASSUR.GCE The regulations require packaging to be designed, fabricated, and operated in conformance with an established for:a1 quality assurance pro-gram.. Compliance with this require =ent is discussed in the following subsections.

7.1 Fabrication Inspectien, and Acceptance Tests The fabrication, inspection, t.nd acceptance ceses are specified j

i en the drawings X3E10191-002, X3E10191-003, X3E10191-010, X3Il0191-011, X3E10191-100, X3E10191-101, and da:a shee ts DS-X E-10191-1, ;5-X E-10191-2, and DS-XDI-10191-3.

The welds will be made and inspected ac:ording ::

l the data sheets.

The fer cus caterials :o be used in the con:ainer are I

specified on the data shee:s :o AS~M standards.

The phen:11: f:a: vill l

2 4

-. i w e m s w _ r:

-=

.1s

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^ ^ ^

~~

~ ~ ~

f2Elfd3$$$$5E5Eb!E$kbkENiINIh5NNNNNEbb b

l 20 a

l be poured as specified on the data sheets and job specification JS-31536-1, Rev. 1, dated March 28, 1968, fire Resis$nn! P.eno!ic Foc.

Acceptance criteria are specified on the data sheets. Di=ension inspection and l

inspection of finishes shall be perfor=ed per the data sheet to ensure that they are within the specified tolerances. A certificate of com-pliance with the drawings and data sheets will be issued to the owner.

l l

\\

7.2 Routine Operating and Inspection Procedures I

1 The ORSL Operations Division has established routine operating and l

j inspection procedures and standard check lists to ensure that all ship-

=ents are safe, comply with the regulations,1-3 and co= ply with all ORNL procedures and regulations. Copies of the procedures and the check-lists are presented in Appendix C.

7.3 Periodic Maintenance and Inspections Prior to each shipment, routine inspections are perfor=ed with the owners' procedures and check lists. The ORNL procedures are presented in Appendix C as an exa:ple. Maintenance is required only when routine.

inspections indict.te damage. There are no ti=e-degradable padding caterials used in the construction of the containers except for gaskets, padding, and cement used in the construction of containers. Any da= age or degradation to these would be detected during routine inspection and replacement effected.

7.4 Records It is the responsibility of the owner to assure that his containers are f abricated to the specifications on the drawings and data sheets.

In addition, the evner shall =aintain fabrication and inspection records in an auditable form for the life of the container.

l 1

I

~

-w

kE@M@l3A2$NMOdfEME3!M[$$'32ifd!16.Nbb$h5h$[sbbNSh.bh$h$bbN$$$$bb 21-REFERE5CES 1.

Code of Feder:1 Rega!:tions, Title 10. Part 71, " Transport of Licensed Materials;" see also Ted. Regis;. 31, 9941-49 (July 1966).

'2.

U.S. Atomic Energy Com:ission, " Safety Standards for the Packaging of Fissile or Other Radioactive Materials," AEC manual, vol. 0000, General Ad= ministration: Part 0500, Health and Safety, chap.-0529 (June 16, 1973).

3.

Code of Feder:! Regulations, Title 49, Part 173, " Transportation;"

see also Ted. Regist. 33,-14920-31 (October 1968).

4. - A. J. Mallett and C. E.- Newlon, Fratective Shipping ?: k:ges for 9-and 12-Inch Siczster l'F Cylinders, ORGDP K-1714 (April 1967).

6 5.

The International Nickel Company, Froperties 'of Some Net:Is and Alloys, A-297 (1971).

6.

1974 Annu 13cck of ASDI Standards, ?:rt 3, Steel - Sheet, Strip, Wire.

7.

1974 Annu:1 Ecok of ASTM Stand: ds, ?:rt 5, Steel - Ears, Ecvgings.

8.

D. S. Clark, The :nfluence of Imp :t Velocity cn the Tensile Ch.~ c-

eristics of Scme Aircraft Metals ord Alloys, NACA TN-868 (October 1942).

9.

R. D. Seagren, Oak Ridge N :icnct Labcratcry ?eturnchie Type _3 Gas l

' Shipping ? ck:ge, ORNL/TM-3544 (October 1971).

10.

a J. T. Thomas, Nue: car Criticali:y Safety Assessment of CE?, N35, l

ni HT33 Fuel itemen: Shipping ?ack:;e> ORNLlCSD/TM-77 (January 1979).

l 11.

F. Kreith, Principles of Sect Transfer, International Textbook, Scranton, Pa., 1965.

12.

L. 3. Shappert, C:sk Designer's Caide, ORSL/NSIC-68 (February 1970).

f 13.

A. J. Mallett I

and C. E. Newlon, "New End-Loading Shipping Container fer Unitradiated Fuel Asse:blies," Fraceedings f See nd Jr. tern -i:n:;

syn;:siw~ cn ?::k: pin; :ni irans;:r:::icn :l ?:iic:o:i;e x: eri::s, con 631001 USAIC, 1968.

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APPENDIX A Fabrication Drawings and Data Sheets 1

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Fabrication Notes for Dwgs. X3E-10191-002, Rev. A and X3E-10191-003, Fsev. A 1.

Functional acrectance Critorion The %el element tasket assemoly (Part Number 2) will be acceptable if, when assembled and centered in the shipping container body (Fart Number B) an inspection tar 3 15/16'.1/16 Jnches secare and a straightness of 1/32 in. ever its full length of 40 in, can be freely irserted and withdrawn from each and every one of the 7 fuel cavi-ies. The inspection bar may have a r.axirum corner racius of 0.125

!.015 in.

2.

Fuel Eleetnt Easket Issembly Fabrication The fuel elerent tasket assently (Fart Number 2) may be fabricated by a combination of formed (bent) shapes and welded joints at the "abricator's option. The seats shall be joired by welds of 1/2 in. lone on 4 in. eenters or ecual. The functional acceptance criterion shall be oer note 1.

3.

Succested Fuel Basket Assembly Procedure A.

Cross tim:ers ', Fart Nummer 14) installed in the shipping container body (Part Number 8).

B.

The threaded rods, nuts and lockwashers (Part Numbers 8F, 5 and 6) installed and tigniered on the duel tasket assembly bcttom plate (Part Number 29).

C.

The fuel element basket assembly (Fart Number 2) lowered into the shipping centairer body (Part Number 8) with the threaded rods (Part Number 8F) passing througn the holes in the shiocing container's bcttom plate (Fart Number 8C).

D.

Tne enenolic spacers (Fart Number 10) will te f rstalled between the fuel element basket uceer clate (Part Number 2C) and the clip (Fart Number SE) on

ne inner sur' ace of the container. The holes in the phenolic spacers will te adjusted to locate the 'uel element basket in the center of the container body within :l/16 in.

E.

Tre thresced r:d (Fart 'br.ter BF) at tne bett:m cf the container body will be teas 40 ed through the channel fixture (see ?wg. X3E 101:1-lC1) by acplying a 10 f t/* b torcue. The threaded rods will be tack-welded, the channel fixture receved, 9e welcs com;leted and finally the excess rods will be removed and t,e b;ttom plate ground smoctn.

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05-t0E-10lil 1, 'ev, i' Fage 2 of 3 The fabricator shall cualify all welcing procedures used in accordance aith

!ection :X cf ite American Society c' "t tanical Engireers ',ASME) iciter and Press,re ressel Code. A 100% vis:.al ins;ecticn is re:Lired for all meles and

e v.eles shall co ply wi h ; ara; a:n !.15 of Ar:S1.1 - 1 :- 75. Visible de'ects ssch as cracts, pinholes, incomplete 'esion, c:Id lacs, ooresity and under cuts

{

shall be re cved and recaired by ;rincing, er by grincing and rewelcing. The re; aired wele shall meet the cuality requirements of :ne original weld. Written i

irs;tc fon reports are not reccired.

B.

Welcire and :ns:ection of Weld Jeir. ire Carbon Steel to Stainiets Steel t

The meicing process is OC Gas tungsten.are, arocn snieiced. !ne fii'er raterial I

shall be ERMCR 3 (Inconel S2T). The reheat temperature shall be 60 300'F.

A 100'. visual irs:ection is reovired ':r all welds and the welds shall co ;1y with ;aragraph 8.15 cf AWS 01.1 1975. Visible defects such as cracks, ;inholes, incomplete fusion, cold laps, porcsity and undercuts shall be removed and repaired by grinding, or by grinding and reweldine. The repaired weld stall meet the cuality requirements of the criginal weld. Written inspection reports are not required.

5.

Materials A.

Lum:tr shall be douglas fir or Southern pine, with moisture content 18'4 or less, No. 2 er tetter grade. The plywood may be any grade.

B.

Stainless sheet per ASTM A240 or A167, type 304 or 347 hot finished ar.d annealed.

C.

Stainless plate per ASTM A240,, type 304 or 304L or 347 hot rolled.

D.

Alurrinum sheet AA2024-T4 cr 6061-76 ASTM B209.

I E.

Steel bar and structural shapes, tyce 304L cr 347, ASTM 276 or A479, hot rolled, annealed and pickled.

F.

Screw, alley steel ASTM A4a9. SAE 5,105,000 psi min. tensile.

G.

Nut, stainless steel type 303, ASTM A194, Grade SF.

H.

Sheet, phenolic laminated brown or natural fine weave continuous filament wovenglass, fabric laminate melamine resin base nema Grade GB-112M.

I.

Sheet neoprene, 50 to 60 durometer hardness.

J.

Screw, stainless steel, any 300 series.

K.

Carbon steel plate per ASTM A-516 grade 60 hot rolled and annealed.

6.

Feam Filling Foam filiing shall be fire resistant phenolic foam and water proofed per job specification JS-31536-1 and as 'ollows.

A.

Fill the top section by featr,ing in one pour into top shell (container lic) and using a mold of scecified finished shape. Weignt of fcam filling shall be 2 1/ 2 lb

10'..

B.

Fill the bett:m section between tre fuel element tasket and container tody in two aopr0xicately coual weight pours with the exposed finish sur' ace being staped by fear.ing into a mold. Weight of finished fcam filling tre te t:m section shall be 35 ih

10':.

Mater'als may be 525;i%:ed :tr t/tL 729. Evalue:f cn of Materials in Fire-Resistant benclic Fca. C i. :ss;rerty, 3. E. Farris ard R. R. Wright, March 1,1973.

JS-31525-1 K/tL-729 Ori;iral

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CS IDE 10191-1, Rey, 1+

Fage 3 of 3

'7.

Dimensional Irsoection and Inseection of Finishes All cimensions, angles and fintsnes snell be snop inspected to insure they are within the specified tolerance. A certification cf co pliance with the-drawings and data sheets will be issued by the fabricator to the owner.

' General revisions to reflect the replacement of carbon steel components with stain-less steel components:

(1) Item 4, welding change to carbon steel to stainless steel welding requirement.

(2) Item 5, deletion of carbon steel specification and addition of stainless steel specifications. ~

(3) Deletion of painting requirements.

(4) Item 6 addition to foam filling requirements.

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HFER Unir*ndiated Fuel Element Shireino Cortainer Fabrication Notes for Owgs. X3E-10191-C10. Rev. A and x3E 10191-011. Rev. A 1.

Functional t:ce:tance Criterion ine fuel eierent tasket assem:1y (Fart Number 2) will be acceptable if. wnen assembled and centered in the shipping container body (Part Number 8) an inspection bar 315/16*1/16 inches souat e and a straightness of 1/32 in, over its full length of 40 in, can be freely inserted and withdrawn from each and every one cf the 7 fuel cavities. The inspection bar cay have a maximum corner radius cf 0.125 '

!.015 in.

2.

Fuel Ele ent Easket tssembly Fabrication Tne fuel element tasket assemoly (Fart Number 2) may be fabricated by a c mtination of forced (bent) shapes and welded joints at the fabricater's option. The seams shall be joined ty welds of 1/2 'n. long on 4 in, centers er equal. The f.netional acceptance criterion shall be per note 1.

3.

Succested Ft el Easket Assembly Frecedure A.

Cress tim:ers (Part Number it.) installed in the shipping container body (Part Number 8).

3.

The threaded rods, nuts and lockwashers (Part Numbers 8F, 5 and 6) installed and tightened en the fuel basket assembly bottom plate (Fart Number 25).

C. ~ The fuel eie: ent basket assem:1y (Part Nueber 2) lewered into the shiccing centainer body' (Part Number $) with the threaded recs (Part Nueter 8F) passing through the holes in the shipping container's bottom plate (Part Nurter BC).

D.

The prenelic scacers (Part Number 10) will be installed between the fuel element tasket up:er plate (Part Number 2C) and the clip (Part Nur.ber EE) on the inner surface of the container. The holes 'n the :henolic s*acers will be adjusted to locate the fuel element basket in the center cf the container body with'in *1/15 in.

E.

The threaded rod (Part Number EF) at the bettcm cf the contairer b:dy will be tersiened through the chant.el fixture (see Owg. X3E-10191 101) ty acclying a 13 ft/lb tersue. The threaded rods will be tack-melded, the char.nel fixture removed tr.e -elds completed and finally the excess rods mill be rem:ved and tre Dc;*cm plate ground smooth.

4 kleids -

A.

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1 CS xDE-10191 2, Rev. i' Fage 2 cf 3 The fatrica :e shall c'.alify all welding procedures used in ac:ordan:e with Section *X Of :6e American Society of Mechanical Engineers ( ASME) Soiler and l

Pressure Vessel Code. A 100% visual inspection is required for all elds and

ne helds snail comely with paragraen 8.15 of AS D1.1-1975. Visible defe::s such as cracts, pinnoles, inconpiete fusion, cold lars, cor:sity and under cuts shall re re ::ved and recaired by grincing, er by grincir; and re eiding. The repaired weld shall teet *.he c'.ality recuirerents cf the original weld. Written inspection re orts are not recuired.

I S.

Weidine and :rscettien of Weld joinine Carbon S* eel to Stainless Steel Ine welcing process is D; Gas tungs en-arc, accen sr.ielced. Tne fliier caterial shall be IRNICR-3 (Incorel 82T). The preheat tem:erature shall te 60-300*F.

A 1005 visual inspection is recuired for all welds and tne welds shall comply with paragrach S.15 of AWS 01.1-1975. Visible defects such as cracks, pinholes, incomclete fusion, cold la;s, porcsity and undercuts shall be remcved and repaired by grinding, or by grinding and rewelding. The repaired weld shall meet :ne cuality requirements of the original weld. Written inspection reports are not recuired.

5.

Materials A.

Lumter shall te douglas fir er southern pine, with moisture centent 18% or less, No. 2 er bet:er grade. The plywood r.ay be ary grade.

B.

5:ainless sneet per ASTM A240 or A167, type 304 cr 347 het finished and annealed.

C.

Stainless plate per ASTM A240, type 304 cr 304L cr 347 hot rolled.

O.

Alurr.inum sheet, AA2024-74 or 6051-76 ASTM B209.

E.

Steel tar and structural shapes, type 304L or 347, ASTM 276 or A470, hot rolled, annealed and pickled.

F.

Screw, alley steel ASTM A448, SAE 5, 105,000 psi rin. tensile.

G.

Nut, stainless steel type 303, ASTM Algd, Grade BF.

I H.

Sheet, ;henolic !aminated brown er natural fire weave continuous filament wevenglass, fabric laminate relarrine resin base nera Grade GB-112M.

I.

Sheet neoprere. 50 to 60 durore:er hardness.

J.

Screw, stainless steel, any 300 series.

K.

Carb:n steel ;1 ate per ASTM A-516 grade 60 hot rolled and annealed.

6.

Feam Filline fcam filiin; shall be fire resistant phenolic fcam and water proofed per fcb s ecifi:sti:n 05-31536-1 ar.d as follc*s.

A.

Fill he isp section by f amirg in ene pour into t p shell (container lid) and using a P:10 Of s:ecified finished snace. lleight cf f:am filling shall te 21/2 'b

Z.

B.

Fill ne b; ::= section between the fuel ele ent tasket ard contairer body in tac at:-:sira:ely e:Lal weiget :ours with the execsed finish surface be f rt; sra:ed y #:amirg into a 2:14. Weicht of firished f am fillinc

he :: : se:: ice shall be 45 'b t 10%. Materials cav te s.tst'tu.e: :er K/ L-729, Evalue: ten cf "aterials in Fire-Fesis*.ar,t Fhenclic

~

ar. 0. 5. *a s;terty. 'i. E. rarris, and R.R 3ri;tt, Ma~.1, is,3 JS-2 12! 1 t/~L 729

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05-XD -10191-2, Rev. l' Fage 3 of 3 7.

Dimensional Inspection and Insoection of Finishes All oimensions, angles and finishes sna)1 oe snop inspected to insure they are within the specified tolerance. A certification of conDliance with the drawings ' nd data sheets will.be issued by the fabricator to the cwner.

a

' General revisiers to reflect the replacement of carbon steel components with stain-less steel components:

(13 Item 4, melding change to carbon steel to stainless steel welding requirement.

(2) Item 5, deletion of carbon steel specification and addition of stainless steel specifications.

(3) Deletion of painting requirements.

(4) Item 6 addition to foam filling requirements.

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Fabrication Notes for Dwgs. X3E-10191-100, Rev. A and X3E-10191-101. Ttev. A 1.

Functional acce:tance Criterd:n Tne fuel element basket asser:iy (Part Number 2) will be accectable M wnen assembled and centered in the shieping container body (Fart Nue:er b) an irs;ection bar 4 7/16*1/* 5 inches set.are and a straightness of 1/32 in. Over its full lengin of 70 in, can be freely inserted and withdrawn from each and every one of the 7 fuel cavities. The inspection bar may have a maximum corner racius of 0.125 1.015 in.

2.

Fuel Element Easket Assemtly Fabrication The fuel elecert tasket assecoly part Number 2) may be fabricated by a combination of formed (bent) shapes anc welded joints at the fabricator's option. The seams shall be joined by welcs of 1/2 in. long on 4 in. centers or equal. The functional acceptance criterion chall be per note 1.

3.

Succested Fuel Easket Assembly Procedure A.

Crcss timcers (Part Num:er 14) Installed in the shipping centainer body (Fart Nur.ber 5).

B.

The threaded rods, nuts and lockwashers (Part Numbers SF, 5 and c) trstai'ed and tightened on the fuel basket assembly bottom piate (Fart Num er 25).

C..The fuel eierent basket assembly (Part Number 2) lowered into the shiccing container'b:dy (Part Nureer 2) with the threaded rods (Part Number 8F) passing through the holes in the sni: ping container's bottom plate (Fart Number 80).

D.

The phenclic spacers (Part Number 10) will te installed between the fuel element basket upper plate (Part Number 2C) and ne clic (Fart Number SE) on the inner surface of the container. The holes in the thenolic scacces mill be ad.hs*ed to locate tre ' gel element basket in tre center f the cer.tainer body within *l/16 in.

E.

The th eaced red (Part Nu-ter BF) at the bet *om of the certa'rer body dll be tersioned through tre :*ancel fixture (see ?wg. X3E-iclil-101) by a:elying a 10 ft/lb tercue.

~he u.readed rods will be tack-aelded, ne chanael future removed, ne welds com0leted and finally the excess reds will be re Oved and the bctt m : late ground s :c*h.

a.

Velds A.

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05 10E-10191-3, Rev ;-

Fage 2 c,f 3 The 'abricat:r shall qualify all welding procedures used in accordance with Section IX cf the American So:iety Of Mechanical Ergineers (ASWE) iciter and fressure Vessel C:ce. A 10Ct visual ins;ection is required 'or all welds and suen as cracks, :innoies, in:0mplett fusion, cold *4;s, porcsity and under cuts snail :e remosed and re; aired by grinding, or by grineing and rewelding. The re; aired weld shall meet the cuality requirements of the original weld. Written ins;ecti:n re;crts are not required.

l B.

Welcine and *eseection ef Weld Joinino Carbon Steel to Stainless Steel Ine aeicing :Po:ess is DC 3as turgsten-are, arcon snielcec. Tne f1Her material shall be ERNICR-3 (:nconel 82T). The preheat tem:erature shall be 60-300*F.

A 1005 visual inspection is required for all welds and the welds shall comply with ;aragra;n 3.15 of AWS 01.1-1975. Visible defects such as cracks, pinholes, incom;1ete fusion, cold laps, porcsity and undercuts shall be removed and repaired by grincing, or by grincing and rewelding. The repaired weld shall meet the quality requirements of the original weld. Written inspection reports are not required.

5.

Materials A.

Lumoer shall be douglas fir or southern pine, with morsture content 18t or less, No. 2 or better grade. The plywood may be any grade.

B.

Stainless sheet per ASm A240 or A167, type 304 or 30al or.147 hot rolled.

C.

Stainless oiste per AS m A240, type 304 or 347 hot rolled.

D.

Aluminum sreet, AA2024-T4 or 6061-T6 ASm 8209.

E.

Steel Ear and structural shapes, type 304L cr 347, AST, 276 or A479, hot rolled and pickled.

F.

Screw, alley steel AS N A449, SAE 5, 105,000 psi min. tensile.

G.

Nyt, stainless steel type 3C3, ASm A194, Grade SF.

H.

Sheet, phenolic laminated brown or natural fine weave continuous filament wovenglass, fabric laminate melamine resin base nema Grade GB-ll2M.

1 f.

Screw, stainless steel, any 300 series.

J.

Sheet neoprene, 50 to 60 duremeter hardness.

K.

Carbon steel olate ;er AS m A-51b grade 60 hot rolled and annealed.

6.

Form Fillino l

Fcam filling shall be fire resistant chenolic fcam and water procfed per job specification JS-31536-1 and as follows.

A.

Fill the top section by foaming in one pour into top shell (container lid) and using a mold of s:ecified finished shape. Weight of foam filling shall I

be 21/2 lb 10*

B.

Fill the oct*:m section tetween the fuel element basket and container body in two apsr:ximately ectal.eigrt pours with t.*e exposed finish su-face being sna:ed by foamirg into a mold. Weight of finished foam filling ine cottom section shall be 60 lb t10%. Materials may be sutstityted :er K/TL-729, Evaluation of Material in Fire Resistant Fhenolic Ecam, C. E. *av;* erty, 3. E.

'arris, and R, R. Wright, March 1,1973.

J! 3.53f-1 K/TL-729 Crigiral Substitute i

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-A.t-1 CS X E 101:1-3, Rev ;*

Page 3 of 3 7.

Dimensional !rsteetion and inseettien of ?"nishes All 01menstors, angles and finishes snall oe snop inspected to insure they are within the specified teierance. A certification of co.t;liance with the drawings and cata sheets will be issued by the fabricator to the owner, J

General revisic-s to reflect the replacement of carbon steel ec=ponents with stain.

less steel components:

(1) Item 4, welding change to carbon steel to stainless steel welding requirement.

(2) Item 5. deletion of carbon' steel specification and addition of stainless steel

. specifications.

(3) Deletion of painting requirements.

-(4) Item 6, addition to foam filling requirements.

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tis specifi:stion shall :ever =aterials and procedures f: r.1xit.g and

'7;1ying f:ased-h-place fire resista:t phenolic foam where specified on -he dravir.gs.

MA*"i'R*ALS AND ??C?ORNCNS All caterials used in the fire resistart foam shall be as listed telev and shall te prepared by blemiing, scree:.ing, etc., as defhed. &ah -

stitations of =aterials may be rade cr.ly where noted. Octal (;astity of :sterials shall be as refaired to fill the c: stainer cavities and

revide the finished foa.: veight as defined en the diavings. hiivid';al veight (;ar.tity of each czponent =Aterial shall te ra:h to provide a i

finished foa :sterial.1xt;re of the prepertic. ate blend specified belev.

Car;ccent Veight ;er i

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Material vender - Sade Na:e (1b s. ).,

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Saher sad Adarson 8.2 + 0.1 22k Anhyd sts No.1135, er egal

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Powiar,Reagar.t G:sda 4

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T.iberglass Rovir.gs Owens-C:rting Tiherglas 9 2 + 0.1 1/L' Chopped Lengths No. Sc5 (ES1), c; eral

~

mat'!210 S TEE R2:rJUt.DE3?S Phenclic Fesin 3

J 1.

Store 1.n a.irtight storage contal ar.

1 2.

Ma.xtra shelf life is three months fra date of macufseture if stored at room temperature.

4 or MLxdra shelf life is six scoths fr:st dtts of mardnettre if stored at a temperature of kCC + 507.

M: Inte :f manuttet;;e shocid be marked cc the stcrs4e c:ctainer.

Pre-blended Licuid C:impecents "ha tt:ree liyaid comper.ents may te ;rs-ble:ded or stored ae;arstely.

j If ;re 'lar.f ?d, ths lisaid mixture :Ay te stored is tittight storage I

c:c.ti.sers at a texterature of LC' 5'T fer a reriod of time r.cs ex-l ceedire c.e.;c:th ;;$ ct to fi:L1 ti:entir.g, ;revided this sititier.&1 tize d:es rc, ex eed the taxir a st :a4e ife for the ;hentii: resin.

If the 11t;.1 ecuper.ar. s a.re at:rea re s.rately, the same st:rtge re.

7: irs =a:.ta rust tc :tserted f:: ::e p.er.:'1: re sin :2p=er.t.

l 7:<ie r

-:p y.t s i

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All 7:. earc ar.t..t.L r e si:re d in tirtist.t s.. : tge c *.? tir.e 6 at issa l

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te:;tra.: e cf.,rc kt af t er pr e.tle: 1: *. t e 7.. ev::.: abscr;;i.,n cf ScA II Ibe

& 38 OF Qi

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eqg2kD&ymyLG5Npp;tes gpxqmqpp%viuil;W::el.MmLECWG57.3E!?f4fEWPWWnknWL4='MAdMLM Mi& M N yL. ew12R

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e STANDARD RT.rERENCE ' INrORMAW.,N

%-a=

JS-31$36-1 lev. 1 r...

I 3 23 l.......-6o

..u, 3 28 63 TIF.E P251S33" PE' 01.10 TOAM

- - a All other 4:erials store all attar ce=;=e t :sterials 2 airtight stcrs4e cor. takers at rom ter;erature ut.til ready fer bie:dir.g.

PP.I?A?ATI3 CT 7CI:'. G 00*:3*TG A.T: A00!SSOFIIS C:r: tai..er In prepri:g a cztaker fer re:eivhg the f:a ed is.;1 ace filli:g, all loose inner pieces shall to securely a chered in ;1 ace to preve: dis-

a

eze: t by the foa=, a:d vest h=les to preve:t voids in the fi ished f:nm and to provide gas relief when exposed to high heat shall be drilled as defted = the trad:gs. :te ze:e:sary tiochur fra.1..g. bracing, etc., shall be batalled to hold in ;; ace the co:Aner shell, eds, i' 'r.g, et:., as a;;11:able, ud preve:t disterti:n to crat-et-toleracce conditi:s d:.:ri.g the f:sabg eperstio:.

Nolds Melds to shape ex;csed surfaces cf foe.3 shall be 1: stalled and secured is positic: just after the fimal foam raw =aterial is placed so fosa cas ez;ud in o thg mo11. Air vest heles not ex:eeding 1/k" i: disester may be :Ade at Stervals in a acid to allcw fer escaping air and thereby prevent veid sp:es in the finished surft:e. Melds shall be z.sde of such strength to ;revent distortic: to eat-ef-t:1erence conditica durks the fes=1 ; :;ers-i=. A :old release (e.g. I erson ud C21:g acid release cce;=-d.N:.1223 cr :.Nc sf1=) sas11 :e a;;11ed to told sv.: faces whi:t vi*1 :osta t f:4m material to that the sold can to reseved without da.s4bg the thisted f:ma surface.

c' lear.ir.; 3efere 7:a:1 g All stvd:st, :its cf vood, :stal filir.gs, uter droplets, gresse, oil, ud cther f reign ;sr-icles sh1r.1 be recoved fr= the scids or ecstaber cavities tet:re foam rsw cateris.1 is added.

CX33 ??*iI-*. :IS P e-31er.d2 ;

Lir.id ? : 1..r.t s. ? e.tle:: 1

f uit :=;c: er.t s in a.y cider ty c:ctir.kg

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t..e.r.grt:57.~:s -.d nitrir.g to a.1 f: n ::.sist e::y. :::;trature cf t.*.s

..s:...a :ui r u.a.. : = iC3 : it? '. s - it is at'ed :a *.ht six. :f fu si t*er.11r.; is :: te delayed, reseal ::: t' ended :2:;or.e:ts so

-i a*::spheri ex;:s re is rek:e:i to a :Mi:r..:.

8 Se 3

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STANDARD P.EFERENCE INFORMAT10N

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... 5 t!* S 1 Fev. 1 J

8 3-2 -66 i.......

l 3 28-63 T:Zg RESISU.N PEDCLI: ?*.AX Pevier Oz.eerts Screes all empme:ts thre;4h a rieve (::stietal ?;reau et Sta.ndards No. LO) er.d pre-hie =d to a unifc = consistency. F.e seal h* ended em;=ents so atacapharic exposure is reduced to a =1:i=;=.

Fitti ?. lending 1.

Add 11 :11 ccampone:ts er pre-blended 117:11 cm;stests to mixer tub 7

a d stir to a uniform cc sistency, about 2 mi: rates ter pre. blended emp= er.s. Tampertrare of the phenclic resh cr pre-h'. ended lig;id czyc:ents shall be 60*

5 7 when it is added to the mix.

2.

Add fibegitas and six to a uniform ec=sistency.

2 Adi the pre-hiended pcvder cz;ccents it; idly Ltd mix to a unifem c:csis*e:cy, 30 to 60 secc.ds.

h. ' quietly tra..zfer the =ized rav :steriti to the receiver coutther so that the foe. air.g tetio: ta.kes phce withis the ecstair.er.

P!.ACL43 A_C C"JP!'C T*AM MAT'27.AL The mixed fram rav :steriti shall te spread evenly over the bett:: r:rface of the receiver ecettirer so fca.: expr.is in reittively unifer= layers.

ite first pear cf botta section shr:1d he ins;ed i=:ediately after fee:-

I ing tegir.s to ~'

-'ce sice of veid spe.ces. Air c;re at rom to ;ertree for at least e~e tra to allev for full ex7Ar.sico cf fona.

A.fter fcashg actics and curir.g is czpleted c. t peu, the next yrcr of rateriti ese te spread ever.ly es te; cf the ;;evices ex74:ded hyer. All set;estis.1 ;r.: s shall yroceed 4th a ritirm deity so tMt at=cs;teric e.xperce of.he i=ternal fiters is =hi= iced. ~he : ranter cf pears shall be as specified = the L-twings. Af*er the f u si pe r, the fca.s s M11 te alir.ed to e;re at least fear hears at roou *e:,ertrare tefere trachg, 2.cids, trd tecesscr' es tre re. cved.

NO I: Setae:titi pres are self to:iing to the previces prar with no t;3 rect *cids =r cleava;e ;hnes.

P.DC'Al CT I-1020, PC* 3, An ACTIE3:7:I! A2 7*N:52 ?:0 V075.

Oeteral All traci.4,==1.is, ar4 a.ny accessorie s not Art cf the thisted contaher, i

j te defhed c the d twisgs, shall te retoved after feazir.g a.nd curir.g opert-j *h times are ccupieted. Ary ext:aded foe.a mA:erita (e.g. ex :;si:r.s tarpagh 8

wae

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! $ [ N h d b I b b/$fi b b b b I k h I b i 5

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.l A-:2 STANDAD REF EhiHC.: INFO *iMATION i-*"

l.*3-31536-1F.ev.1 ie.es l 3-23-66 3 22 68 7:;.Iam rARPE R:C7%M air ver.: h:14s),*.11 te er.re.*:2.17 tri==ed a.:.d re=oved to ;:eviis a smooth rarface cver the c::teric aid interict rarft:e of the em: taker.

.Ma zAteria.1 splashed c.to ax;csed rarfaces of the ecc;tair.er shall to reseved.

All :ald release, e.rease, tast, et:., which verald prevent a coating from adheri.=g ;rc;erly shall te res:ved ir::s the fir.ished foa.m s'arfaces, prier tb c:st i..g, by vipimg vi-t a soft 1cth sata-ated vith tri chlcreethyle:e.

Pat: Fir _- icid: 5 Fi=isha1 T:4:ed Surfa:e w terial a

Voit spaces is the fini:hed fea.= sv.rft:e shall be Mtched by filling with adiitir.=.a1 f:ms ra.i sterial a:d all:vir.g the =Ateria.1 to exund and carre.

I2:sssiva f=sa or r.r gh s;o s left by this pro:ess ar.all be dressed sr.ooth 4.h the re7:a=11 g surface to give es evers11 'miftrm rcrf ace.

Vate-:r: ^ing Seal.st:: tight al beles is o;ter shell a:1 in=er liter wh&. a;;11:able, (e.g. ve=t heles, gaps ar:rmd :a;s for till holes, r.111 and screw he* es, etc., a.s a;;11:stle) vi.h two costs rf e;cxy++ ;rier to fisish c:st cf

?

pai:t. Ee:1.T.tartight all exposed rerft:as :f fcam 4th two coats of erxy.

C' TA.' 33 Ak. 2 Ce11:s cL*.v. a:e te:e, er c. er clessing age: s =ay be

sed f:r :*enci:-

ut assi.1.4 :r re.2::vi.r.g pr.1:les of splattered f:a= fr: a metal,4rts o-of ec.taf er r;rfaces.

C:iER

e f=1.1.-.bg t>,*. e.1 che:ical a.slysis of cured feta ray t e used as refere::e data fo: rualear : ale 1.2.asi:r.s.

Vt. (

Ca.:b:n 11.0 F.ydr :se.

k.5 3:r =

32 E 1.1 '. : ;r.

2.2

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A I; q shal.a te /. a;11 e 4g2 e;exy vi;h p..

37 i,.,31 a s usaf t:t;. ed by 8..

Cits ?rcea::s Cer.pny, Tairlaw.. Tew.7ers '

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e APPENDIX C Routine Packaging and Inspection Procedures 1
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PRE-USE INSPECTION CHECKLIST UNIRRADIATED FUEL ELEMENT SHIPPING. CASK Oa'k Ridge National Laboratory Oak Ridge, Tennessee Container Serial No, Reference Drawings: X'E-10191-002 X3E-10191-003 BY-DATE A
Visually inspect container exterior for shipping damage, rust and. general' state of repair.
B Visually inspect container interior for rust, condition of foam, moisture and general state of. repair.
C Visually inspect neoprene spacers and gasket.
1 D
Visually inspect. basket closure bolts, nuts and lock washers for condition and conformance to D-1 through 0-5.
o t
1 Number recuired - 8 each 2
Solts - Hex Md. 3/8"-16 UNC-2 1-1/4" long alloy steel AS M A-449, SAE 5, 105,300 psi min tensile.
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3 Nuts - SST 3/8"-16 UNC (welded to basket) 4 Lock washer - 3/8" carbon stael 5
';uts and bolts engage freely by hand i
E Visually inspect closure bolts, nuts and lock washers for condition and confcenance to E-1 through E-5.
Number reouired - 6 ea.
2 Bolts - Hex Md. 5/S"-11 UNC-2 1-1/4" sta'r'ess steel Tyoe 304 ASTM A-193 Gra:s 35.
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Nuts - SST 3/8"-16 UNC.(welded to container flange) 4 Lock washer - 3/8 carbon steel 5
Nuts and bolt engage freely by hand.
List deficiencies found and corrective action taken below, i
Approved by:
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' PROCEDURE FOR LOADING ORR UNIRRADIATED FUEL ELEMENT SHIPPING CONTAINER DATE NAME 1.
Record element numbers 1
2 3
4 5
6 7
2.
Clean inside of shipping container 3.
Inspect container.
No loose or foreign material to be-present inside shipping container.
Record container number 4
Visually inspect cask exterior for shipping damage, rust and general state of repair, i
5.
Visually inspect basket spacer and gasket.
6.
Visually inspect basket closure, bolts, nuts and lock washers for condition and conformance to 6-a through 6-c.
, a.
Number Required b.
Bolts - Hex Hd.
I c.
Nuts d.
Lock washers carbon steel 3/8"
'I nominal size e.
Nuts and bolts engage freely by hand.
1 7.
Visually inspect lid closure, bolts, nuts and lock washers for condition and con-formance to 7-a through 7-e.
a.
Number Required b.
Solts - Hex Hd.
c.
Nuts d.
Lock washers carbon steel 5/8" nominal size e.
Nuts and bolts engage freely by hand.
1 3.
Load elenents.
This operation to be witnessed by ORNL representative.
Heat seal one end of polyethylene tube.
Slide element into tube.
Heat-seal top end of polyethylene bag. Use vacuum cleaner to pull air out of bag.
Place element (s) in container (reximun 7 elements).
Replace ccver on cortainer.
Ee certain He element is bel;,e the flat Of the ccver and *.ht! **e colye!"y'e's !!;
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9.
Place container lid in place and bolt to container.
(Torque bolts to 25 ft-lbs.)
10.
Radioactivity inspect (10 CFR Part 71)
Write and apply radioactive Class I label;.
Accept Reject
11. Place - tamper - safe seal U.S. Treasury.
Type "E" (cup type) on shipping container.
Record Seal #
Witnessed by inspection or QA Representative Time Date I have witnessed sequence 11 and accept delivery of the fuel elements.
I certify that fuel element numbers listed in seq.1 and leaded in seq. 8 i's tamper-safe sealed in the shipping container with the above numbered tamper-safe seal.
ORNL Representative Enter seal number in daily inspection log.
12. Check shipping container for: this operation to be witnessed by ORNL Representative.
1.
All bolts in lid and tight.
2.
Tamper-safe seal in place on lid.
i 3.
Radioactive label.
d.
Shipping label and shipping papers securely attached.
Witnessed oper.12 ORNL-Representative I
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K. K. Chipley E. M. King J. M. Robinson 1,. B. Shappert on December 22, 1975, I perforned the " Penetration" Test described in Annex 1 of AIC MC.0529 or. the existing packages and models listed on the attached data sheet. In all cases the resulting damage was a small dent. It is now obvious that this test will not reduce the effectiveness of steel or stainle ss steel jacketed packages and that the resulting damage is insignificant.
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Page 2 - Certificate No 9853 - Revision 0
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June 7, 1979 Union Carbide Corporation Nuclear Division ATTN:
Dr. Herman Post:a Director Oak Ridge National Laboratory Post Office Box X Oak Ridge, Tennessee 37830 Gentle =en:
SAFETY ANALYSIS REPORT FOR PACKACING THE UNIRRADIATED FUEL SHIPPlNG CONTAINER Reference is given to the letter dated April 26, 1979, from M. E. Ramsey in regard to the above subject.
The draf t Safety Analysis Report for Packaging (SARP) has been reviewed by the staf f of the ORO Safety and Environmental Control Division. The SARP is approved for publication.
Enclosed is a copy of the ORO Certificate of Co:pliance.
The original Certificate of Compliance has been sent directly to the authors so that it may be reproduced for inclusion as an Appendix to the SARP.
The orig-inal should be returned to ORO af ter publication.
The closure arrange =ent for these containers is considered to be an im-provement over the standard drum closure ring used on the 12" dia eter UF6 overpack which was successfully tested to Type B package requirements.
However, sin.ce the closures are different and an engineering evaluation does net appear feasible, we request that a drop test be made en a proto-type.
NRC has raised ceestions on the closure of the CRNL gas cylinder fire i: pact shield.
Since licer. sees vill be using both the unirradiated fuel shipping cont ainer and the gas cylinder fire and i= pact. NRC ques-tions sho:Id be resolved.
Sincerely, b
l Joseph A.
Lenhard Assistant Manager
.r Ener&y MS ; AJ Eesearch and Des lopment i
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g.3 TC 79-1 INTRA LABORATORY CORRESPONDENCE
. C AK RIOGE NATIONAL LABORATORY February 28, 1979 To:
R. W. Mouring Subj ect : Approval of Safety Analysis Report for Packaging:
Unirradiated Fuel Element Shipping Container The draft (received February 2,1979) of the~ subject SARP has been reviewed by the Transportation Committee. The SARP is approved for :echnical content and approach -- for submission to DOE.
Coments I
from individual members of the Committee have been forwarc 3 for-
c. consideration.
E.nj. ld E. M. King Transportation Committee DO(:j r ec: Com=ittee Members G. H. Burger J. A. Ccx J. H. Evans R. V. McCord l
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E-5 Union Carbide Corporation-2 June 7, 1979 Nuclear Division
Enclosure:
As stated cc:
.C. A. Keller, MS-30, w/enei.
T. H. Hardin, AD-46, v/ enc 1.
J. E. Rounsaville, ER-111, w/ enc 1.
W. H. Travis, MS-33, w/cncl.
R.'F. Hibbs, UCC-ND, w/ encl.
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Cak Ridge National Laboratory Safety Analysis Report for Pack-Post O!! ice Sex X aging: The Unitradiated Tuel Shipping Cak Ridge. Tennessee M830 Container
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Report: CRNL/DG/ M-15
e. cCNDsTION:
The can.f.we.i cone.: oaa smo* ine f.itm.g of one reovu emeats of 5.costi O of 10 C7 A 71. si secosso *. eae ime coao't.oas to*sd*o
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packagin g:
(1) Model: OR5L Unirradiated Tuel Shipping Container (2)
Description:
Packaging for unirradiated fissile catorial at fuel elements. He fuel ele-nents are positioned in a basket consisting of seven square cavities fabri-cated from 16 gauge plate and a base fabricated from eleven gauge plate.
he plate is Type 300 stainless steel. Eight 3/8" nuts and bolts retain the basket lid, which is made from 0.125" thick slu=inu=. in place.
he basket is positioned inside a cylindrical outer shell.
he outer shell and lid are fabricated f rom eleven gauge plate and the base is 1/4" thick plate.
he plate for the shell is Type 300 stainless steel, ne outer lid is held in place by six $/8" nuts and bolts.
he basket is supported on 2" x 6" ti bars inside the outer shell.
~he re=sining space around the basket is filled with phenolic foam insulation.
2tre are different types of packages. Table I describes the details of each design.
b.
Contents:
(1) Type and fe m cf caterial M.
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E-7 Page 2 - Certificate No. 9853 - Revision 0 contained in fuel plates as reacter fuel elements.
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ORNL/ ENG/TM-15 Dist. Category UC-71 INTERNAL DISTRIBUTION I
1.
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A,. Ara =ayo 28.
E. M. King 2.
M. 3 ender 29.
R. W. Knight 3.
D. Box 30-34 R. V. McCord 4
G. Bewden 35.
J. R. McGuffey 5.
H. G. Burger 36.
J. D. McLendon 6.
C. D. Cagle 37.
R. W. Mouring 7.
D. D. Cannon 38.
F. H. Neill 8.
K. K. Chipley 39.
T. W. Pickel, Jr.
9.
H. C. Claiborne 40.
M. E. Ramsey 10.
R. L. Clark 41.
J. E. Ratledge 11.
J. A. Cox 42.
J. N. Robinson 12.
W. K. Croviey 43-47.
R. W. Schaich 13.
F. C. Davis 48.
R. A. Sch=idt 14.
H. R. Dyer 49.
R. D. Seagren 15-19.
J. H. Evans 50.
L. 3. Shappert 20.
R. E. Eversole
$1.
W. E. Terry 21.
F. C. Fitzpatrick 52.
J. T. Thomas 22.
C. S. Force 53.
J. W. Wacher 23.
J. H. Gillette 48-49.
Central Research Library 24.
K. W. Haff 50-52.
Document Reference Section 25.
S. S. Hurt 53.
Laboratory Records Depart ent i
26.
G. R. Jasny 54-68.
Laboratory Records, ORNL R.C.
27.
R. A. Just 69.
ORNL Patent Office EXTERNAL DISTRI3UTION 70.
E. W. Sailey, DOE, ORO 71.
E. L. Barraciough, DOE, Albuquerque Operations Of fice, Fest Office Box 5400, Albuquerque, NM 87115 72.
L. G. 31acek, DOE, Supply Div., ORO 73.
R. ?. Chitwood, DCE, Div. of Transportation and Fuel Sterage, l
DCL Headquarters 74 H. N. Culve r, DCI, CR0 l
75.
R. I. Elder, Chicago Opera'.10-ifice, 9500 South Cass Avenue, Argonne, IL 60439 i
76.
J. M. Treedman, Sandia Laboratories, P.O. Box 3800, Albuquerque, NM 57115 77.
R. E. Harris, DOE, ORO 78.
Assistant Manager, Energy Research and Develep=ent, 30F.-CR0 79.
W. G. O'Quinn, DCI, Savannah River Cperations Office, Post Cf fice Ecx A, Aiken, SC 29601 S0.
R. F. Odegaarden, ':EC, Of fice of Regulation Washington, DC 205 3 l
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R. W. ?eterson, Batteile Memorial Institute, 505 King Avenue, Columbus, OH 43201 82-86.
R. W. ?ovel, Brookhaven National Laboratory, Upton, NY 11973
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W. A. ?ryor, DOE, ORO S8-92.
Tavfic Raby, Chief of Reactor Operations, National Bureau of j
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Standards, Washington, DC 20234 93.
R. Rawl, U.S. Department of Transportation, 6th and D Street, SW, Washington, DC 20590 f
94 J..\\. Sissler, DOE, Div. of Transportation and Tuel Storage, 1
DOE Headquar:ers 95.
D. W. Templeton, DOE, Richland Operations Office, Post Office Box 550, Richland, VA 99352 96.
W. H. Travis, DOE, ORO 97-99.
Safety and Health Protection Branch, DOE, ORO 100-270.
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(
i Centra:t No. k'-7;C5 eng 25 I
{
)
COMPUTER SCIENCES DIVISION i
l
{
i Nuclear Criticality Safety Assessment of ORR, 535, and HF3R Fuel Element Shipping Package J. T. Themas (Sponsor: John H. Evans, Originator:
J. T. Thomas)
Date Published: January, 1979 NOTIC E This coewment contains information of a p el.mina'y retu e.
It i subject to revislon of Cor'eetion and the*efore does not rep *esent a l
r;ner report.
WION CAR 3IDE CORPORATION, STCLEAR DIVI 5:0N eperating the Oak Ridge Gase us Diffusion Plant Oak Ridge National Laboratory Oak Ridge Y-12 Plant Paducah Gaseous Diffusion Plant fer the
v..* +&
r., ;
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0 by%tl 2iaa i;.s> f k duwkl%
TABLE OF COSTESTS Pace LIST OF TABLES.
v A35!RACT.
1 I.
INTRODUCTION.
l' Fig. 1.
Proposed Package Design 2
I!.
MEZ403 0F ANALYSIS.
3 III. DESCRIPTION OF CODE Ih7UT.
3 IV.
RESULTS OF CALCULATIONS.
7 V.
CONCLUSIONS.
1,0 RITERENCIS.
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L:ST OF TA3 TIS Page Table 1.
Geometric Representstien of Package.
4
!able 2.
Description of Materials for Code Input.
9 Table 3.
Computed Neutron Multiplication ? actors fer the 1:ndamaged Package.
9 Table 4 Co:puted Neutron Multiplication Factors for the Damaged Package.
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Suelear Cri:icality Safe:y Assessmen: ef ORR, 535, and lit 3R Tuel R1enen: Shipping ?ackage J. T. Tnomas 1
ABSTRACT A fuel element shipping package empicying a berated-phenolic fcam as a ther=al insulating ma:erial is designed to transport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Seam Reac:ct, er the National Bureau ef Standards Reac:er.
This report pre-sents the criticality safety evaluation and demonstrates that the requirect.nts for a Fissile Class I package are l
satisfied by the design.
l i
I.
INTRODUCTION The nuclear criticality safety of a shipping package designed to transport as cany as seven plate-type fuel elements is examined by calculational techniques. Three distinct packages are proposed, one for each of three si:es having a light-wa:er reactor. These tre the Oak Ridge Ressarch Reac:or, the Brockhaven Tas: Seam Reac:cr, and the reactor at the Fational Bureau of Standards. The three packages have si=ilar neutrenic characteristics, the same materials of construction, but dif fer slightly in their dimensions.
Tne gress characteristics ef a typical package are shewn in 716 1.
The analysis is perfer=ed assuming there are nine fuel elenents present in the package rather than the seven preposed. Inis modification is made because it facilitates the goemetric descri;: ion of the rackage in the calculation, and resui:s in an everestima:e cf :he neutren ceu;11ng be:veen pa:Pages and :f the keff :f the arrays :: ;; kages.
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3e neutren multiplication factors of the shipping package and of
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d section sets has been validated by calculation of critical experi=ents
with fuel elements' of the same fissile materials and configurations.- The results of the validation with fuel elements are reported in Refs. 5 and 6.
The results of calculations of critical experiments with the borated-phenolic foam are reported in Ref. 7.
The conclusion of the comparison of-calculations and. experiments is that systems calculated to have a k,ff of 0.98 should be regarded as having. a potential for criticality. The analysis is performed with HTER fuel ele =ents containing 350 g U(93.2)' per element.
III. DESCRIPTION OF CODE INPUT Each plate in a fuel element is described in the ecde by a bex type and these are arranged to form a fuel element within a region of the steel grid. One-half the thickness of the steel forming the 3 x 3 matrix is associated with each element. When the matrix of fuel ele =ents is described in the code, the correct steel thickness is specified between fuel elements; h vever, enly one-half the steel thickness is represented fer the outer surface of the = atrix.
~'his description is :ensarvativa-in that it vill result in larger calculated keff's than vould be ceasured.
The ge :etry description is given in Table 1.
A fuel ele:ent and its associated section f the matrix is ferred by stacking btx types in the erder 9. ~. 5. 3, twelve l's, 2.
, 6. and 3.
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1 1
I
'~'
'_E-.m-L.--
'l-h-
"---A
p l M a ? N R 3G s 2s5 E is s f f! E M a c M B M # J n s n fi 2 R n F r h W I 0a? 2 E M & M s. N S J f M i ?1 EsisHM 7
types are required to preserve the different wa:er channel thicknesses and the nonfuel-bearing end plates of the HT3R element. The materials occupying the gec=etric regions specified in this case are representative of a calculation in the da aged package evaluations.
The caterials and their number densities are 1'dentified and reproduced in Table 2.
The.
and boxes of the fuel elements are represented as a sceared density which preserves the = ass of aluminum. This is material 6 in region 4 of boxes 1 through 9 in Table 1.
The reflector region description of Table 1 specifies the geometry in the region between the 3 x 3 steel matrix to the outer container of carbon steel: The material mixture in region 4 normally would be 8, the undamaged borated-phenelic foam insulation, but in the casa depicted region 4 represents the result of damage to the insulation by exposure to fire.
The principal damage to the package vill be a charring of the berated-phenelic foam insulation. Actual tests show that an average char depth from the' outer surface will net exceed 6.4 cm.
An over-estimate of neutron coupling between damaged package would be observed if a larger char depth is assumed; theref re, a value of 7.6 cm was used in the evaluacirn of the da= aged package.
IV.
TuSL1TS OF CA1.CL1ATIONS The computed multiplication facters for the undamaged package are
esented in Table 3.
It is evident that the absence f water from the fuel region of the package results in very lev values f r keff.
This is 5
consistent with previous results with these fuel elements. There is a large increase in kef f when water eccupies the fuel regien : h:vever, k, cerains well ':elow a value :f unitv.
l e n d w_._e w __-
-n.-=2-_
_-_x_n._
u.
L
,:x 1
.1,
.m w_
r.a c a m s +
an- --,
[S!N O!$5 bli[ N d $5'E i S $$ $ bid $ $ N3 b5I$'b dTN N d b E bE! $ M b d 5' N b b 5 N b.$ 5 D N N b btAO N b INfS N 5
>~
Table 2.
Description ef Ma:erials for Code Input.
I l
w! X T U.7 E NUCL!DE OE N SI T Y 4
1
-02502
1. 3 0 73 ? E-03 '
)
1 92503
1. 3 0 70 3 E -3 3 a l.Q l.0 1 E_-1.5 Uranium N.de and Aluminum 1
qa1U a
1 02834
1. 41241 E -3 4 1
8100
7. 4 7 30 0 E -33 1
13100
5. 2 3 86 7 E - 0 2 e 2
1210_p A. o 2 723
-c ?
a1...
3 6100
1. 7 3 76 5 E -3 2 '
3
!,1101 Not used in stud 7
?.60647E42 3
?10 0
3. 6 8 92 5 E -0 3 4
5.0.?
% e.0.0.0 3.0 E.=0.2 >
5 502
1. 0 0 00 0F 00)
Water 6.
13100
1. 81 19 0 E -0 4 )
Aluminum 7
13100
1. 8 1 19 0 E M 4 'l 7
11mi e.3 2 ~:C_4 * -9 2 f.
A h-4 *" ~ 2 M W2 * '"
7 N130 3 16 19 7 E -3 2 )
j 8
6100
4. 4130 0E -03 '
S 1101
5. 6 3 E 0 0 5 _0 3Scrated o,nench.e feam 03 a
a i e,o 3,,qg.go o =
8 5100
3. 5 3 60 0 E-04 S
14100
1. 5 0 400 E-3 4 8
11100
1. 5 3 60 0 E -0 5
__ A IJ.10.0
2. 3 2 a0.0 5.-0 5 6
13100 0.905002-36 A
12100
9. GS 300 E-35 8
20100
6. 0 5 60 0 E-75 e o
6.1.0.0
$_Q.21".C.E _0.?
9 1101
7. 2 0 010 E -03 9
8100
'.8947 # -3 3
3. I 100 5"E -3 4 ocrated p'nene.,0. cam and 1
4 9
5100 o
13 1 ^ o
1. 2?2.20.0 e -o 4 9
11100
1. 4 4 30 0 E -0 5 WOM o
17100 2.053005-05 9
13100 7.93400E-05 o
1 2.1 0.0
.*.2120.0 " =3 5
)
9 20100
5. 3 4 10 0 E -0 5 10 200
1. 0 0 00 0 E 0 0 Stainless Steel 11 100 3 0 0 00 0 E 00)
Car' on Steel S
i i?
A 10.0 6 O 3.9.3 0 E :-0 3 '
12 1101 00 1?
S100
0. 0
!?
5130
?.1 1 oS O E -7 4 I
i?
tc,La.O 1 32.70;E-34_
y _
i 12 11130 1.44200E-35 "a.. e.... m;. f m e o....
j 12 17130
0. 0 12 13130
7. 9 3 4 0 3 E
,6 f 0a m 1.2 12.1.0 o
21.10 ; r : ';_
12 20130 0.26 130E "Si l
e AtA.?
t 1
Am
'^- - '-'
^^E*-
~ -
E
^ -~ ^'

~"f

Lpgu.gr,u,ewn$y ~.:x sM:. v%e. :m.:Rm WTin w - y M.ka&'&sm.Ex='. m%'mt.. ~ M.M, "?WBWW&
m ":
> iHGMD'W e
.ca: Et.2.@!L k &L2 ?si M 1 A M L%W%uM L%Laktum &ta
- u.
o.
s' 9
Table 3.
Computed Neutron Multiplication Factors for the t'ndamaged Packase Nu:Ser ef Conditions kefs + :
ackages r
Single Water in fuel region 0.669 0.008 Single Water in fuel region, package 0.670 0.00S closely reflected by water Infinite array Water in fuel region 0.812 0.007 Infinite array Water in fuel region, water 0.812 0.007 filling void between packages Table 4 surmarizes calculations of the damaged package condition.
Again, one finds k,well below unity.
Table 4 Computed Neutren Multiplication Factors for the Da= aged Package Number of Conditions k,ff 4 c 3ackages Single No water present 0.033 0.002 Single Waar in full region 0.666 0.006 Single Water in full region, pa:kage.
0.683 0.009 closely reflected by water
nfinite array No water present 0.092 0.033
nfinite array Water in fuel regien 0.739 0.);S
niini:e array Water in fuel regi:n, water 0.652 0.G'.C filling veid between pi ckages l
l l
l l
l
. ce y : y ~. -,,.
..,s
..,_,m..
4
h"fhb N d ibi b$[$ b N 2 b kf2b b b b bh b hf'bbbIllI bb i b b k b b b '
5 b
+
10
(
1 1
V.
CONCLUSIONS The evidence of this study shows that the package leaded with seven F.T3R fuel elements meets the nuclear. criticality safety require-i cents of a Tissile Class I package..In view of the comparative calcula-l tions of-various si=ilar fuel elements with different fissile caterial 1
loadings reported,s,s it =ay be concluded that the package may also be used as a Tissile Class I package for the ORR and the NBS fuel elements with fissile =aterial leadings at least as large as 350 g 23s / element.
U Tuel ele =er.ts of similar construction used at the Oak Ridge National Laboratory, such as the PCA reactor (140 g 2ssU/ ele =ent) and the SSR reactor (200~g 2 3 st/ element), may also be shipped in the container.
)
.,._.--__=
~.
=--=-- - -
A--
E'S$NIUbi$$$$$$!5N?; 0$$$$NW!N$5l$$bbWib.N$b5$$.$$bN$$NN$$?Nb$b$b$$$$?$.$N$bN 5
. o..
t 11 REFERENCES 1.
L'. M. ?etrie ~ and N. T. Cross,
.3li:-71*:, An :. pr:rei l-l:v.:e ::r::
Iri:d::".?:y Pr:gr~., ORNL-4938, Oak Ridge National Laboratory.(1975).
2.
.G. E. Hansen and W. H. Roach, Si: :nd !l.: teen Gr:u: lr:ss !a:-i:v.s f:r ?:30 :nd :r.:gr edi::e Critic:L Asse..h!ies, LAMS-2543, Los Alames Scientific Laboratory (1961).
3.
J. K. Tox and L. W. Gilley, " Critical Experiments with Arrays of ORR and 3SR Tue1 Elements," !.*eu:::n ?hysics :ivisi:n Ar.nu:2 ?r:gress
.?e?::: f:P ?eriod T';iir.; !!?: ember 1, :Si?, ORNL-609, Oak Ridge National Laboratory (1958).
4 E. B. Johnson and R. K. Reedy, Jr., Critical E :eri 'ents :15 :4 5?I.?T-3 Tue! I~emer.:s, ORNL/TM-1207, Oak Ridge National' Laboratory (1965).
5.
J. T. Thomas, llu:2e~ ~~' -lity Safety :f the Tue; i:e cr.: ?:bri::-
i

n I::ili:y :: At::ebers, !!:ssachusetts, ORNL/CSD/TM-55, Oak Ridge -
National Laboratory (1978).
6.
J. T. Thomas, luele r Cri;i::lity Assessmer.: ' f 3:h.?iige.Tese:r:h
.?e :::r Tuel Storage, ORNL/CSD/TM-58, Oak Ridge National Laboratory (1978).
7.
D. V. Magnuson, Critical :hree-Dimer.sion:2 Arr:ys :l l.'eu:r:n :nter-
ting Ur.its: ?:rt III Arrays of C(23.2) liet:1 Separ::ed by Various lf::grd:!s, UCC-ND Y-12 Plant (1972).
8.
A. J. Mallett and C. E. Nevlon, ?r:tective Shipping ?: k ge f:r E-inch-Ji:~eter UTs Cylinder, K-1716, Oak Ridge Gaseous Diffusion Plant (1967).-
9.
A. J. Mallect and C. E. Newlen, "New End-Loading Shipping Container for Unirradiated Fuel Asse=blies," ?rs:eedings of Se: nd _~e. ern::ict:T, 5,~p:sium :n ?::k:ging :nd Tr:nsp:rt: i:n of.?:iic:::ive '.l::eri:~s, 5
CONF 681001 USAEC (1965).
l
% M*WLHb' n L&-'d 'E Y ? 2 LL A'-
--
i !
..~
,,--^
b,b" C,7 * ". P 4 $
'"f**-
'J S
3 i* "' O N ^ k
~
23323NIIE353152517:UIA/$NS8/IsdE5S$NOdDI$0$5Id$$5YdEh$$$[b22bh$f3D$b ((bbbr.
.n
~ :,
13 ORNL/CSD/IM-77 INTERKAL DISTRIBUTION 1.
F. T. Sinford 7-11.
J. T. Thomas 2.
G. H. Burger 12.
G. E. Whitesides 3.
H. P. Carter /A. A. 3 rocks /
13-14.
Central Research Library (2)
CSD Library 15.
Document Reference Section, 4
R. Gwin Section, Y-12 5.
R. W. Knight 16-18.
Laboratory Records (3) 6.
R. V. McCord 19.
ORNL Patent Of fice i
EXTER.
J., DISTRIBUTIGH 9
20.
C. Hopper, Texas Instruments, Inc., Metallurgical Materials Division, Attleboro, MA 02730
21. Nor=al Ketzlach, Nuclear Regulatory Commission, Silver Springs, MD 20910 22.
J. N. Rogers, Division 8324, Sandia Laboratories, Livermore, CA 94550 23.
F. L. Sherman, Texas Instruments, Inc., Metallurgical Materials Division, Attleboro, MA 02730 24 D. R. Smith, Los Alamos Scientific Laboratory, MS-560, P. O.
Box 1663, Los Alamos, NM 87544 25.
R. L. Stevensen, Office of Nuclear Y.aterial Safety and Saf*e-guards, Nuclear Regulatory Commission, Washington, DC 20555 26.
Chief, Mathematics and Geoscience Branch, DOE, Washington, DC 20545 27.
Office of Asst. Manager for Energy Research and Development, Department of Energy, ORO, Oak Ridge, TN 37830 28-54 Technical Information Center, Department of Energy, Oak Ridge, TN 37830 i
l l
wwaxx _,
x_
+
e :-.<
w r~'-
~
~ -
l
$ b b V ~ M 0 $ $ $h$!~lx i 00$?b5N!$YN$.$Nb kNb0b$$ b$
b Y{id$ Y a.
i ORNL/CSD/TM-77 Contract No. W-7405 eng 26 COMPUTER SCIENCES DIVISION i
1 Nuclear Criticality Safety Assessment of ORR, NBS, and HFBR Fue'. Element Shipping Package J
T. Thomas (Sponsor: John H. Evans, Originator:
J. T. Thomas)
Date Published: January, 1979 NOTICE This docurnent contains infortnetton of a prehrr.inery nature.
It is suelect to revision or correction and therefore does not represent a final report.
UNION CARBIDE CORPORATION, NUCLEAR DIVISION operating the Oak Ridge Caseous Diffusion Plant Oak Ridge National Laboratory Oak Ridge Y-12 Plant l
Paducah Caseous Diffusion Plant for the DEPARi ENT OF ENERGY l
l
/,/
1 J10 / #1 A K A / / Ll 1
II
~_ N Y 7 ' Y " '.,F
- =
s.- ~ ~ ~..
~-
?,: m.m..n. ? : r?r.".m:r.v.n;..,%...~vmW. ;4.; ~w, Wnr ?.v:W~s.".. ~ 'r.; * *. i ? ' F
T -YE.'#fF;C'$^@.W 3-f
.m s
..g n.., c.,
1,.
..,.e p'Anssl.L. 't? WLMld r.il.>9 U
'U, NGW~OWlY *.L4 s5*\\ 3 %>.'+4E Q hn. '. A*J = d-J. i~dh 's !C 5:L ' =E~iE*<Ad=*UE;h, t, e
a
' W "" =
l 6
i i i'
a 1
LIST OF TABLES Page Table 1.
Geometric Representation of Package.
4 l
Table 2.
Description of Materials for Code Input.
8' Table 3.
Computed Neutron Multiplication Factors for the Undamaged Package.
9 Table 4.
Computed Neutron Multiplication Factors for the Damaged Package 9
I V
!?bbhbb N $I5b$.YiNENNY I.
.5 n L
..;,+'
Nuclear Criticality Safety Assessment of ORR, NBS, and.HFBR Fuel Element Shipping Package J. T. Thomas ABSTRACT A fuel element shipping package employing a borated-phenolic. foam as a thermal insulating material is designed to transport as many as seven fuel elements for use in the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, or the National Bureau of Standards Reactor. This report pre-sents the criticality safety evaluation and demonstrates that the requirements for a Fissile Class I package are i
satisfied by the design.
-l I.
INTRODUCTION The nuclear criticality. safety of a shipping package designed to transport as many as seven plate-type fuel elements is examined by calculational techniques. Three distinct packages are proposed, one for.each of three sites having a light-water reactor. These are the Oak Ridge Research Reactor, the Brookhaven Fast Beam Reactor, and the reactor at the National Bureau of Standards. The three packages have similar neutronic characteristics, the same materials of construction, but differ slightly in their dimensions. The gross characteristics of a typical package are shown in Fig. 1.
The analysis is performed assuming thet e are nine fuel elements present in the package racher than the seven proposed.
This modification is made because it facilitates the geometric description of the package in the calculation, and results ir an overestimate of the neutron coupling between packages and of the k gg af the a' trays of packages.
e
MWW,bOi&QM A,"0Y:
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,4 L]1+.SGbA.'l 'Aa[$.?D $'
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o i
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-r 0
l g/cp w s +l l;
1 *.y
. wet eas.se 1
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/
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l '
- (
- Y.
- g s
uo r
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o
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Fig. 1.
Proposed Package Design
_m___m__
__.___m__-______._____mm
_m___U
.Y?lS!iNl Nbh0b$ YE $AY bb$Y$Yh$bb Eb Y
Y 3
II.
METHOD OF ANALYSIS The neutron multiplication factors of the shipping package and of l
arrays of packages were calculated by the KENO IV code and the Hansen-Roach neutron cross-section sets.2 This combination of code and cross-4 8
section sets has been validated by calculation of critical experiments '"
witn fuel elements of the same fissile materials and configurations.
The results of the validation with fuel elements are reported in Refs. 5 and 6.
The results of calculations of critical experiments with the borated-phenolic foam are reported in Ref. 7.
The conclusion of the comparison of calculations and experiments is that systems calculated to have a keff of 0.98 should be regarded as having a potential for criticality. The analysis is performed with HFBR fuel elements containing 350'g U(93.2) per element.
III. DESCRIPTION OF C1DE INPUT Each plate in a fuel element is described in the' code by a box type and these are arranged to form a fuel element within a region of the steel grid. One-half the thickness of the steel forming the 3 x 3 matrix is associated with each element. When the matrix of fuel elements is described in the code, the correct steel thickness is specified between fuel elements; however, only one-half the steel thickness is represented for the outer surface of the matrix. This description is conservative in that it will result in larger calculated k gg's than would be e
measured.
The geometry description is given in Table 1.
A fuel element and its associated section of the mi.trix is formed by stacking box types in the order 9, 7, 5, 3, twelve l'
, 2, 4, 6, and 8.
The different box 1
j k
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p Table 2.
Description of Materials for' Code Input.
MI XT URE NUCLIDE DE N SI T Y 1
-92502
1. 3 0 703 E-03 '
1 92503
1. 3 0 70 3 E -0 3 1
921t3 4 y._Q132E_-0.5 ranium Oxide and Aluminum 3-92834
1. 4 1 241 E -0 4 1
8100 7.473003-03 1
13100
5. 2 3 86 7 E 4 2 d 2
1310_0
6. 0 2 72.6 E -0 2 A 1,,,,, i n,-
3 6100
1. 7 3 76 5 E -0 2
3 1101 P. 6 0 64 7E -0 2 Not used in stud 7 3
17100
0. 6 8 925 E -0 3 4
Sf.?
5..OS.040 E -0 2 2 5
502
1. 0 0 00 0F 00)
Water 6.
13100
1. 81 19 0 E -04 )
Aluminum 7
13100
1. 81 19 0 E -0 4 '
7 lini 6 37. 39.4 Eg 7 Aluminum a nd Wa f or 7
N100 3.16 19 7 E -0 2,
8 6100
4. 4130 0E -03 '
8 1101
5. 6 3 eO O E -0 3 sa m i o.o
3. np.eo o E -0 3 Borated phenolic foam 8
5100
3. 5 3 60 0 E -04 8
14100
1. 5 0 400 E-04 8
11100
1. 6 3 60 0 E -05 a
17.10.6
2. 3 2 20.0 E_-o M B
13100
0. 9 9 50 0 E -0 6 8
12100 9 8 8 30 0 E-06 8
20100
6. O M 60 0E -05 d o
6.Lp_O 6.cQ29egE 0.3*
9 1101
7. 2 0 010 E -03 9
8100 3 8 9 47 0E -0 3 P enolic foam and Borated h
9 5100 3.1 1 00 0 E -0 4 o
1.9 1 o n 1,22 70.0.E -QA 9
11100 1 4 4 300 E -05 W oca 0
17100 2.05300E-05 9
13100
7. 93 400 E-06 o
12.10.0 A. 2.1_7.0.0 f.=0 $
9 20100
5. 3 4 10 0 E -0 5 10-200
1. 0 0 000 E 0 0 Stainless Steel 11 100
1. 0 0 000 E 0 0 )
Carbon Steel 17
^ 10_0 6.07090E.=03' 22 1101 0.0 1P 8100
0. 0
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12 17100
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12 20100
5. 3 4 10 0 E -0 5 4
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Table 3.
Computed Neutron Multiplication Factors for the Undamaged Package Number of Conditions k gg i a Packages e
Single Water in fuel region 0.669 0.008 Single Water in fuel region, package 0.670 0.008 closely reflected by water Infinite array Water in fuel region 0.812 0.007 Infinite array Water in fuel region, water 0.812 0.007 filling void between packages Table 4 summarizes calculations of the damaged package condition.
Again, one finds k,well below unity.
Table 4 Computed Neutron Multiplication Factors for the Damaged Package n
ns keff i a Pa kage Single No water present 0.033 0.002 Single Water in full region 0.666 0.006 Single Water in full region, package 0.688 0.009 closely reflected by water Infinite array No water present 0.092 0.003 Infinite array Water in fuel region 0.739 0.008 Infinite array Water in fuel region, water 0.682 0.007 filling void between packages G
e
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CONCLUSIONS L
the evidence of this study shows that the package loaded with seven HFBR fuel elements meets the nuclear criticality safety require-ments of a Fissile Class I package.
In view of the comparative calcula-tions of various similar fuel elements with different fissile material loadings reported, it may be concluded that the package may also be-used as a Fissile Class I package for the ORR and the NBS fuel elements with fissile material loadings at least as large as 350 g 23s / element.
U Fuel elements of similar construction used at the Oak Ridge National Laboratory, such as the PCA reactor (140 g 23sU/ element) and the BSR reactor (200 g 2 sU/ element), may also be shipped in the container.
i
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REFERENCES 1.
L. M. Petrie and N. F. Cross, RENO-IV: An Im' proved Monte Carlo Criticality Progrcm, ORNL-4938, Oak Ridge National Laboratory (1975).
2.
G. E. Hansen and M. H. Roach, Si and Sixteen Group Cross Sections for Fast and Intermediate Critical Assemblies, LAMS-2543, Los Alamos Scientific Laboratory (1961).
j 3.
J. K. Fox and L. W. Gilley, " Critical Experiments with Arrays of ORR l
and BSR Fuel Elements," Neutron Physics Division Annual Progress t
Report for Period Ending September 1,1958, ORNL-2609, Oak Ridge National Laboratory (1958).
4 E. B. Johnson and R. K. Reedy, Jr., Critical Experiments eith SPERT-D Fuel Elcments, ORNL/TM-1207, Oak Ridge National Laboratory (1965).
5.
J. T. Thomas, Nuclear Criticality Safety of the Fuel Element Fabrica-tion Facility at Attleboro, Massachusetts, ORNL/CSD/TM-55, Oak Ridge National Laboratory (1978).
6.
J. T. Thomas, Nuclear Criticality Assessment of Oak Ridge Research Reactor Fuel Storage, ORNL/CSD/IM-58, Oak Ridge National Laboratory (1978).
7.
D. W. Magnuson, Critical Three-Dimensional Arrays of Neutron In' ter-acting Units: Part III Arrays of U(93.2) Metal Separated by Various Materials, UCC-ND Y-12 Plant (1972).
8.
A. J. Mallett and C. E. Neu1on, Protective Shipping Package for S-inch-Diameter UFs Cylinder, K-1716, Oak Ridge Cascous Diffusion Plant (1967).
9.
A. J. Mallett and C. E. New1on, "New End-Loading Shipping Container for Unirradiated Fuel Assemblies," Proceedings of Second International Su posium on Packaging and Transportation of Radioactive Materials, CONF 681001 USAEC (1968).
l
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INTERNAL DISTRIBUTION 1.
F. T.-Binford 7-11.
J. T. Thomas 2.
.C. H. Burger 12.
G. E. Whitesides 3.
11. P. Ca r t e r/ A. A. Brooks /
13-14.
Central Research Library (2)
'CSD Library
.15.
Document Reference Section, 4.
R. Gwin Section, Y-12 5.
R. W.. Knight 16-18.
Laboratory Records (3) 6.
R. V. McCord 19.
ORNL Patent Office EXTERNAL DISTRIBUTION 20.
C. Hopper, Texas Instruments, Inc., Metallurgical Materials Division, Attleboro, MA 02730 21.
Normal Ketzlach, Nuclear Regulatory Commission, Silver Springs, MD. 20910 22.
J. N. Rogers, Division 8324, Sandia Laboratories, Livermore, CA 94550 23.
F. L. Sherman, Texas Instruments, Inc., Metallurgical Materials Division, Attleboro,.MA 102730 24.
D. R. Smith, Los Alamos Scientific Laboratory, MS-560, P. O.
Box 1663, Los Alamos, NM 87544 25.
R. L. Stevenson, Office of Nuclear Material Safety and Safe-guards, Nuclear Regulatory Commission, Washington, DC 20555 l
26.
Chief, Mathematics and Geoscience Branch, DOE, Washington, DC 20545 27.
Office of Asst. Manager for Energy Research and Development, Department of Energy, ORO, Oak Ridge, TN 37830' 28-54 Technical Information Center, Department of Energy, Oak Ridge, TN 37830 l
1 d_ _
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'ml NUCLEAR DIVISION Of INTERNAL CORRESP0?idENCE September 10, 1979 J. H. Evans, Building 1000 (4-6381)
Re-evaluation of ORR Shiocing Packaoe In March of this year it was pointed out that the density of the phenolic foam used in nuclear criticality safety evaluation of the ORR package was that corresponding to a confined faming, whereas, free feming was implied by the package weights. The change in foam density from 0.201 g/cc used in the analyses reported in ORNL/CSD/TM-77 to 0.05 g/cc, typical of free forming, does not affect the conclusions given in the referenced document. The density reduction does result in an increase in the infinite neutron multiplication of packages but the value remains below a k-eff of 0.9.
The package, there-fore, meets the requirements of the Fissile Class I category. The attachment sumarizes the benchmarking and the calculations perfomed-If I can be of further help, please let me know.
Y :=-
J. T. Thomas. Chaiman Criticality Comittee JTT/bbf Enclosures cc:
H. M. Butler, Bldg. 3546 (4-4340)
G. H. Burger, Bldg. 3546 (4-4339)
-R. Mouring, Bldg.1000-C215 (4-6397) m i
4 I
m mm mm _,_ mmm m.
,-.. _ ~._-.
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l
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. SPe81*i 0F CALCULATIC*lf -. - -.----..a5 0F ORA P4COGE 1
E
[
. Description of code inout MetaH al number densities:,
i
. M,ix,ju_my, Nuclide Density
Mixtum, NuclQe consity 1
92501
'1.52678-4 2
13100 6.02726 2 1
92503 1.52678-4 5-502 1.0 1
92833-B.49960-6 7-13100 1.8190-4 1
92834 1.84990-5 7
1101
6.32394 1-8100 2.34536-2 7'
8100 3.16197-2 i
1 1102 1.86761 10 200 1.0 1
13100 4.51613-2 11 '
100 1.0 Pheno 11e Foam Mixture Mixtum Mixtum 3!S.I.!M, 8-Nomal 9-w/ wood 12-charred 6100 1.10880-3 2.72390-3 1.50995-3 1101 1.39150-3 3.04120-3 0.0 8100 7.78390-4 1.75120-3 0.0 5100 8.8844-5 1.44660-4 7.79750-5 14100 3.77890-5 5.98470 5 3.31750-5 11100 4.11060-6 6.50780-6 3.60750 4 17100 5.84920-4 9.25890-6 0.0 13100
' 2.26010 4 3.57820-6 1.58350-6 12100 2.48320-6 3.93130 4 2.17930-6 20100 1.52160-5 2.4C870-5 1.33530 5 Geometry description of fuel mgfan of par 4-a9e Box tves for 3=3=1 arrance_ meg jggign,, ~ 6eonstry Mg.7,
.x tv sz 1
Cuboid 1
3.1115 4.0259 30.1(2
'2 Cuboid-
'3 Cuboid
~ 2 3.5814 4.0259 30.162 7
3.5814' 4:0259 8f.027 4
Cuboid 5
3.7700 4.2145 88.027 5
Cuboid 10 3.8460 4.2905-98.344 Package description sun ound 3=3=1 arrang. -
-nt
.Rggign, Georetrv Mixture
\\
1 Core body 0
2x = 11.538 ky = 12.371 er = 88.344 3
Cylinder 8
r = 23.380 3-Cy!inder 9
r = 23.380 -
-< = 88.344,
+s - 08.344
+: = 95 400, 4 = 95.400 4
Cylinder 12 r = 31.000
= 10L11
-c a 102.94 5
Cylinder 11 e = 31.320
+t = 103.43, e = 103.58 6
Cuboid Va M able ex = 31.380
y = 31.380 ar = 113.80 l
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a L
l' t-i Benchma'rk The typical fuel element considered is the HFBR' loaded to 350 g U(93.2) per element. The interior of the region defined by the aluminum side plates was a homogeneous mixture of the aluminum, r anium oxide and water which resulted in number' densities given above. Confirmation of the adequacy of this fuel element representation was had by recalculation of two of tb configurations calculated in ORNL/CSD/TM-58 (1978) where the element was described.in detail.
The two calculations were those of Table 4 in TM-58. A comparison of the calculations is made in Table 1.
i Table 1.
Comparison of Homogeneous and Detailed representation of HFBR Fuel Elements l
Arrangement TM-58 Detailed Homogeneous 4x4xl 1.32 e 0.008 1.137 2.0.006 4x3xl 1.054 0.010
'l 059 e 0 006 This comparison with the validated calculational method raported in ORNL/CSD/1M-58
shows good agreement. An additional calculation of a 3x3x1 array in water gave a k-eff of 1.004 t 0.005..The latter arrangement is that used in the package evaluation and gives an indication of the effectiveness of the package as a reflector.
Package calculations A submerged single package with nine elements as described above 'gave a k-eff of-0.715 1 0.005.
The k-eff of in-infinite number of packages without water between packages was 0.889 1.0.006.
The k eff of an infinite number of packages with water between the packages was 0.794. 0.007.
These calculatir s with nine fuel elemen.s may be considered as producing an m
upper limit to any k-eff that may be achieved with seven elements.
The lower k-eff for seven elements is sufficient to consider allowance in the mass per element as + 10 wt% uranium.
==
Conclusion:==
Any finite number of packages in transport cannot result in a k-eff larger thcn that corresponding to the k of 0.9.
The nu-lear criticality safety assessment of the ORR shipping package meets the requirements of Fissile Class I package in transport.
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r-OAK RIDGE NATIONAL LABORATORY OPEMATED SY UNION CARSIDE CORPORATION NUCLEAR DIVISION O
Po$T oPPICE SoE X OAK elo4E. TENNESSEE 37E50 Sep. : e ar 18, 1979 To: Recipients of Subje:t Report Report No.: ORNL/ ENC /TM-15 Classification:
U Authors:
J. H. Evans, W. D. Shipley, and R. W. Mouring
Subject:
Safety Analysis Report for Packaging: The Unitradiated Fuel Shipping Container i
The enclosed arrattra (pp. E-5 through E-8) should replace pages E-5 E-6, E-7, and the blank page that precedes the distribution list. Pages of this erratum have been reproduced on adhesive-back paper for your convenience.
Laboratory Records artment Information Division l
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Un!.en Carbide Corporation 2
June'7, 1979 Nuclear Division i
Enclosure:
As stated
.l cc C. A. Keller, MS-30, w/ enc 1.
3 T
R. Hardin, G-46, v/ encl.
J. E. Rounsaville, ER-111, w/ enc 1.
W. B. Travis, MS-33, w/ enc 1.
R. F. Hibbs, UCC-ND, v/ enc 1.
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{1) PW== e esais e ee eeneeueJJ (Il I;tle ene lesatieaet.se se reserv er es s at (31 one-Oak Ridge National Laboratory Safety Analysis Report for Fack-Post Office los I aging: The Unitradiated Fuel Shipping cak Ridge, Tennessee 37830 container 1
Report: ORNL/DIC/TM-15
4. CoNotTiosas N eece e amiee.am i. sea ime #wem.a se ene re.re mag ee saan o es to C7 m 71, a see sme, eae ine amame=a Gemenes
.a e e emme.
5 Denseessa se Pasmepas one Avinorisse Geacoas, neese mweemr. Aaw.e casus. Otner ceae.e.eas, sas Aeteeasse.
a.
Packasinz (1) Mode 12 CRNL Unirradiated Fuel Shipping Container (2) Descriptions Fackagios for unirradiated fissile material as fuel elements. ne fuel ele-aants are positioned in a basket consisting of seven square cavittee fabri-cated from 16 gauge place and a ban fabricated from eleven gauge place.
The plate is Type 300 stainless steel. Eight 3/8" outs and bolts retain the bashr.t lide which is made from 0.115" chick aluminum in place. De basket e
is positioned inside a cylindrical outer shall. The outer shall and lid are fabricated from eleven gauge plate and the base is 1/4" thick place.
The plate for the shall is Type 300 stainlass steel. The outer lid is held in place by six 5/8" nuts and bolts. De basket is supported on 2" x 6" timbers inside the outer shall. The remaining space around the basket is filled with phenolic fosa insulation.
nere are different types of packages. Table 1 describes the details of each design.
b.
Contents:
(1) Type and form of material m una 4 u.
n ec Mehad es s93T 235U and is in the oxide form. It is es.ce.sei= a.e-JUN 4 1979
! e, a.
,m.a oe.e.
som vna v s. ospaavuent es aa amov fs. = _ see oot s ar on.see.
7m. s.s,wwe. n e.a t. see oos,ses,.
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.di n wt U. S. Department of Energy William H. Travis Director s
Post Office Box E Safety and Environmental Control Cak Ridge Tennessee 37830 Division e
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f g.7 Fase 2 - Certificate so 9853 - savision 0 contained in fual plates as reactor tual elemmats.
(2) Fissile Class I
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l U.S. Nuclear Regulatory Ccmnission ATIN: Charles E. MacDonald,. Chief Transportation Certification Branch Division of Fuel Cycle & Material Safety
(
Washington, D. C.
20553 1
(
Gentlemen:
i Reference is given to your letter dated October 1,1979, to Larry Blalock in regard to the NRC review of the Report ORNI/ENG/IM-15, j
" Safety Analysis Report for Packaging: The Unirradiated Fuel Shipping Container," FCIC:RHD, 71-9853.
In reply to the specific additional infonnation which you requested, ORNL has prepared the enclosed supplemental data. The data have been reviewed by the staff of the DOE-ORO Safety and Envircanental Control Division and we fully concur with the conclusions.
We request a priority review by NRC so that the containers may be used at an early date.
Sincerely.
William H. Travis, Director MS-334:WAP Safety & Envirorrnental Control Division
Enclosure:
Supplemental Data (9 copies) cc w/ encl.:
T.H. Martin, AD-46 D. M. Ross, G-135, H(>-GTN
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C.A. Kaller, MS-30 s -
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