ML19253A646

Package Description,Technical Operations Model 650

ML19253A646
Person / Time
Site:	07109032
Issue date:	08/03/1979
From:	Marzilli D TECH/OPS, INC. (FORMERLY TECHNICAL OPERATIONS, INC.)
To:
Shared Package
ML19253A641	List: ML19253A640 ML19253A646 ML19253A691
References
NUDOCS 7909100617
Download: ML19253A646 (60)

Text

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Radation Prodsets Diws.on 40 No@ Avenue Burnngon, Massachusetts 01803 Te+ phone (617) 272-2000 PACKAGE DESCRIPTION TECHNICAL OPERATIONS MODEL 650 USA /9032/B Prepared by DAVID MAR 2ILLI 7909100 M N

- c (i ') )

Ri^J U

'u"Of

i 1.

General Infor=ation 1.1 Intrcducticn The Tech / Ops Model 650 is designed for use as a source changer and ship-ping container for Type B quar.:ities cf radicactive =aterial in special fem. The Model 650 confoms to the criteria for Type B packaging in acecrdance with 10CFR71 and satisfies tha criteria for Type B(U) packaging in as:cordance with IAEA Safety Series No. 6 1973 The scurces to be used in conjunction with Model 650 are Techf0ps sealed sources Models No.

A424-1, A424-9, A84101 and 369701. The Model 650 vill contain a =aximu=

of 240 curies of iridium-192 as special fer=.

1.2 Package Descriction 1.2.1 Packaging The Model 650 is 13 25 inches (337==) high,10 inches (254==) Icng, and 8.25 inches (21Ct:=) wide la overa'l dimension. The gross weight of the package is 70 pcunds (32kg).

TF.- cadicactive source assembly is housed in a titaniu= "U" tube.

"he tube is crimped in the middle cf the "U" to prcvide a positive stop fcr the source assembly. The "U" tube is cast inside a depleted uranium shiold asse=bly. The weight cf the uranium shield is 35 pounds (16kg).

The shield asse=bly is enclosed in a stee_ bottc= housing consisting of a rectangular shell and tcp and bottoc. cover plates of 0.135 inch (3 4==)

thick cold rolled steel. The rectangular steel shell is further enclosed in a round steel t _tbe with a vall thickness cf 0.047 inch (1.W). The void space between the shield an the shells is filled with a castable rigid polyurethane fcam. The stsel-uranium interface is separated with a 0.010 inch (0.25k==) thick copper shim.

Mcunted on the tcp cover plate is a source holddovn assc=bly. This asse=bly is used to secure the radicactive source in a shielded position during transport.

An cuter package lid, fabricated frc= 0.135 inch (3.h==) thick eclt Sl..e d steel, is bolted to the package to provide protection to the holddo -

asse=bly.

Tamper-prcof seals are provided during shipment of these sources. Two vent holes and asse=bly joints which are not leak tight provide passageways for the escape of any gas generated frc= decc= position of the potting fca in the event the scurce changer is involved in a fire accident. The outer packaging is designed to avoid the collection and retention of water.

The package is painted and finished to provide for easy decenta=ination.

The radicactive material is sealed inside a stainless steel ;urce capsule.

The capsule acts as the containment vessel fcr the radicactive material.

1-1 EVISION O

, ', ; O U

AUG.

3 373

i The Model 650 has been previously approved for use as a Type B package under USNRC Certificate of Ccmpliance No. 9032, Rev. 1 (Section 1 3).

Drawings of the Model 650 are enclosed in Section 13 1.2.2 Goerational Features The source asce=bly is secured in the proper storage position by means of the source holddown assembly. With the source in the storage position, the connector portion of the source assembly is located inside the source holddevn asse=bly. The holdlovn asse=bly is bolted to secure the source asse:bly in position. The bolt is seal vired.

1.2 3 Contents of Packaging The Model 650 is designed for the transport of Iridiu=-192 in quantities of up to 2h0 curies as Tech / Ops source asse=blies Models Ah2h-1, Ah2h-9, A8kl01 and B69701. The Iridiu=-192 is in special for= as prescribed in 1CCFR71 and IAEA Safety Series No. 6,1973 (Section 2.8) 1-2 REVISICN O AUG.

3 G73

i 1.3 AFFEYDLT US:aC Certificate of Ccepliar.ce No. 9032, Eev. 1

- Ibscriptive Assembly Dravirg, Model 650 1-3 REVISION O AU3.

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Fuem N AC 618 U.S. NUCL r M4 HE GUL & 'Uriy comirnioN U

CERTIFICATE OF COMPLI ANCE 10 C R 71 For A useoactive Maireials Pada9es 8

1.(aj Certefecate Nurrber 1.(u) A cvision Nn.

1 (c) Pack ane 1.icnialecation No.

1(d) Paues No 1 le! Tctal No. P.

9032 1

USA /9032/B( )

1 2

2. PRE AM8LE 2.fa)

This certificate is issued to satisf y Sections 173 3'13a.173 394.173.395. and 173 39G of the Departrrent of Transportation Hazar Materials Regslations (49 CFR 170189 and 14 CFR 103) and Sections 146-19-10a and 146-19-100 of the Department of Transportation Dangerous Cargoes Regulations (46 CF R 146 i49), as amended.

2.(b)

The packaging and contents described in item 5 betow. meets the safety standards set forth in Suboart C of Title 10. Code of Federal Regulations, Part 71. *"Packag:ng of Radioacteve Materials for Transport and Transportaison of Aad oactive Material Unde' Cer ta n Cond.tions.

2.(c)

This certificate does not relieve the consignor from compliance with any reuussement of the regulations of the U.S. Department of Transportation or other applicable regulatory.vjer'cses, including the government of any country thrCugh or into v hicle the packac will be transparted.

3.

This certificate is issued on the bases of a safety analysis report of the package design or application-3.(a)

Prepared by (Name and address).

3. (b)

Title and identification of eeport or soplication.

Technical Operations, Inc.

Technical Operations, Inc. application Radiation Products Division dated flovember 1,1974, as supplemented.

Northwest Industrial Park Burlington, Massachusetts 01803 3 (c)

Dock et No.

71-9032 4.

CONDITIONS This certifecate is conditional upon the fulfill.ng of the reau,rements of Subpart D of 10 CF R 71, as appi. cable, and the condlisons specif e-in item 5 below.

5.

Descripinon of Packaging and Authorized Contents, Model Number. Fissile C1.iss. Other Co.xfiteens. arv1 Relesences-(a)

Packaging (1) flodel No. : Model 650 (2) Description A steel encased, uranium shielded, Iridium-192 source changer.

Primary components consist of an outer steel shell, polyurethane potting material, uranium shield, Titanium "U" tube, and source holdown assembly.

The source holdown assembly secures the source assembly in position within the crimped "U" tube.

Tamper-proof seals and a padlock are provided on the packaging.

Total weight of the package is approximately 70 pounds.

l (3)

Drawings The packaging is constructed in accordance with the following Technical Ope. ations, Inc. drawings:

l D65000-Rev. C B65000-il C65000-2 A65000-15 065000-3 C65000-16 CS5000-5 860010-5 C65000-6 A65001 C65000-7 A65001-1 C65000-8 Bill of Mat'Is.

065000 B65000-15-Rev. A C65000-17-Rev. B V

EVISION O

,_h AUG.

3 1973

Page 2 - Certificate No. 9032 - Revision No. 1 - Docket No. 71-9032 1

5.

(b)

Contents (1)

Type and.orm of material Iridium-192 as sealed sources which meet the requirements of special form as defined in 571.4(o) of 10 CFR Part 71.

(2) Maximum quantity of material per package 240 Curies 6.

The source shall be secured in the shielded position of the packaging by the source assembly.

The source assembly must be 0

fabricated of materials capable of resisting a 1475 F fire environ-ment for one-half hour and maintaining their positioning function.

The cable of the source assembly must engage the source holdown assembly.

The flexible cable of the source assembly must be of sufficient length and diameter to provide positive positioning of the source at the crimp of the "U" tube.

7.

The nameplates shall be fabricated of materials capable of resisting the fire test of 10 CFR Part' 71 and maintaining their legibility.

8.

The package authorized by this certificate is hereby approved for use under the general license provisions of Paragraph 71.12(b) of 10 CFR Part 71.

9.

Expiration date:

January 31, 1980.

REFERENCES Technical Operations, Inc. application dated November 1,1974, requesting approval to deliver special form Iridium-192 sources in Model 650 source changer to a carrier for transport.

Supplements dated November 11, December 27, 1974, and February 27, 1976.

FOR THE U.S. NUCLEAR REGULATORY COMMISSli

{ Nfsv

^Y Charles E. NacDonald, Chief Transportation Branch Office of Nuclear Material Safety and Safeguards D a t e__ _ _# # " _ _'. _' _ D5_____..____

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1-5 EVISION O

' i' A;3.

3 1973

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Structure 2 Evaluation 2.1 Structural Design 2.1.1 Discussion Structurally the Model 650 consists of four components:

A source capsule, shield assembly, outer casing, and source holddown assembly. The ocurce capsule is the primary contain=ert vessel. It =eets the require =ents of special form contain=ent as outlined in 1CCFR71 (see Section 2.8).

The shield assembly fulfills two functions:

It provides shielding for the radioactive caterial and, together with the source holddown assembly, insures proper positioning of the source. The outer casing is a double thickness of steel, 0.047 inch (1.1 h =) outer, 0.1345 inch (3.k2==) inner, in the side directions and single thickness (0.1345 inch, 3.k2n=) in the top and bottom directions. Attached to the top cover ic a lid of 0.1345 inch steel. The casing provides the structural strength of the package.

The casing joints are butt joints. The top lid protects the source hold-down assembly. The holddevn assembly secures the source assembly in the shielded position at the cri=p in the "U" tube, and assures positive closure.

2.1.2 Design Criteria The Model 650 is designed to co= ply with the require =ents of ICCFR71 and IAEA Safety Series IIo. 6,1973 The device is simple in design, such that there are no design criteria which cannot be evaluated by straight-forward application of the appropriate section of 10CFR71 or IAEA Regula-tions.

2.2

'Jeights and Centers of Gravity The Model 650 Source Changer weighs 70 pounds (31.8kg). The shield asse=bly contains 35 pounds (15 9kg) of depleted uranium. The center of gravity was located experimentally h.25 inches (108==) frc= the bottom of the package along the cylindrical axis.

23 Mechanical Procerties of Materials The Model 650 Source changer casing ic cold rolled steel. This =aterial has a yield strength of 40,0C0 pounds per square inch (276 m /=2,)

(

Reference:

Machinery's Handbook, 20th edition, 1976, page h52.)

The radioactive sources to be used with the Model 650 consist of a

?/pe 30h or a Tvre 3ChL stainless steel capsule to which is svaged a "Tel flex" steel drive cable. The capsules are sealed by tungsten inert gas velding and the svage joints are tensile tested on a production basis to 75 pounds.

2.h General Standards for All Packages

[n REVISICN O 2-1

~

AUG.

3 ;973

r B

2.4.1 Chemical and Galvanic Reactions The =aterials used in the construction of the Model 650 Source Changer 9re uranium rets1, steel, titaniu=, and copper. There vill be no significant chemical or galvanic action betvcen any of these ec=ponents.

The possibility of the phenomena of the formation of the eutectic alloy of iron uranium at te=peratures belev the melting te=peratures of the individual metals was ccasidered. The iren uraniu= eutectic alloy te=-

perature is approximately 1337 F (725 C).

However, vacuum conditions anu extre=e cleanliness of the surfaces are necessary to produce the alloy at this loi tenperature. Due to the conditions under which the shields are mounted, sufficient contact for this effect does nct exist.

In support of this conclusion, the folleving test results are presented.

A thernal test of a sample of bare depleted uraniu =etal was perfomed by Nuclear Metals, Inc. The test indicated that the uranium sa=ple oxidized such that the radial dimension was reduced by 1/32 inch. A subsequent test was perfomed in which a sa=ple of bare, depleted uraniu=

cetal was placed on a steel plate and subjected to the themal test conditions. The test shcved no alloying or =elting characteristics in the sample, and the degree of oxidation was the same as evidenced in the first test. A copy of the test report appears in Section 2.10.

Although the liklihood of the formation of an iron-uraniu= eute-tic alley is re=ote, ccpper separators are used at iron-uraniu= interfaces.

2.h.2 Positive Closure The source assembly cannot be exposed in the Model 650 without first re-coving the source holddown asse=bly. The source holddown assembly is bolted and sealvired. Access to the scurce holddown asse=bly requires r uoval of the lid. The lid is bolted and sealvired, and is provided with a tamperproof seal. The top and bottc= covers are lockwired to insure that the retaining bolts are not accidentally loosened.

2.h.3 Liftir4 Device s The Model 650 is designed to be lifted by the botto= or top covers or top lid. Each is secured by four 5/16 hex bolts (the botta= and top cover bolts are safety vired). Each bolt has a cross-sectional area of 0.0500 in2 (32 3==2).

The yield strength of the =aterial is 40,000 pounds per square inch. Thus, each bolt can support approxicately 2,000 pounds (909kg), or approximately 30 tices the package weight.

2.h.h Tiedevn Ihvices The tiedevn devices on the Model 650 are the top and bottc= covers. As indicated in (2.h.3) abcve, the bolts attaching these devices can each support approximately 30 times the package weight.

(,

Q.J 2-2 REVISION O AUG.

3 G73

251 Standards for Type B and Large Quantity Packages Considering the package as a simple bea= supported on both ends with a unifer= load of 5 times the package weight evenly distributed alors its length, the maximum stress can be cceputed frc=:

F1 S

=

E where:

S: =aximum stress F: total load (350 pounds) 1: length of been (13 25 inches)

Z:

section of =odulus of beam (2.05 in3)

(

Reference:

Machinery's Handbeck, 20th ed.,1976, p442)

The load is taken to be 350 pounds (159kg). The container is assumed to be a cylindrical tube 13 25 inches (337==) long with a vall thickness of 0.0h7 in.

(1. N ) and inside diameter of 7.4h inches (189==).k The section modulus is cc=puted to be 2.05 cubic inches (3 36 x 10==3).

Thus, the maxirc.n stress generated in the bea= is 283 pounds per square in. (195bCI/=2), which is far belev the yield strength of the material.

252 External Pressure The Mcdel 650 is open to the at=csphere; thus, there vill be no differen-tial pressure acting on it.

The collapsing pmssure of the source capsules can be fcund 86,670 t - 1386 P

=

D vhere P: collapsing pressure in pounds per square inch t: vall thickness in inches (0.020 inch)

D: Outside dia=eter in inches (0.25 inch)

(

Reference:

b'achinery's Handbock, 20th ed., p.hh8)

'Ihe collapsing pressure of the capsules is calculated to be 5,550 pcunds per square inch (38.3 32T/=2). Therefore, the capsule can withstand an external pressure of 25 psig.

2-3 REVISICN O 1,U1 3 1973

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2.6

?!ormal Conditions of Transoort 2.6.1 Heat The ther=al evaluation is perfor=ed in Chapter 3 cf this application.

Frc= this evaluation, it can be concluded that the Model 650 can with-stand the normal heat transport conditions.

2.6.2 Cold The metals used in the manufacture of the Model 650 can all withstand temperatures of -40 F (-40 C). The lcwer operating limit of the poly-urethane fea= is -100 F (-73 C). Thus it is concluded that the Model 650 vill withstand the nomal transpor( cold conditions.

2.6.3 Pressure The Model 650 is open to the atmosphere; thus, there vill be no differen-tial pressure acting on it.

In Section 3 5 4, the source capsules are demonstrated to be able to withstand an external pressure reduction of 0 5 at=ospheres (50 7kN/=2),

2.6.h Vibration The Model 650 has been in use ten years.

During that time there have been no vibrational failures reported.

Thus, we contend the Model 650 vill not undergo a vibrational failure in transpcrt.

2.6.5 Water Spray Test Tne water spray test was not actually perfor=ed on the Model 650. We

ontend that the materials used in construction of the Mot'el 650 are all highly water resistant and that exposure to vater vill not educe the shielding or affect the structural integrity of the package.

2.6.6 Free Drop The drop analysis perfor=ed in Hypothetical Accident Conditions (see Section 2 7.1) is sufficient to satisfy the requirements outlined for the nor=al transport frea drop condition in 1CCFR71 and IAEA Safety Series No. 6, 1973 On this basis, we conclude that the Model 650 can withstand the free drop without impaiment of the shielding cr package integrity.

2.6.7 Corner Drop Not Applicable 2h REVISION O AUG.

3 197S a p t-

/

I

2.6.8 Penetration A penetration test of the Model 650 was perfor=ed. There was no loss of shielding or structural integrity folleving th2 penetration test. A copy of the test report appears in Section 2.10.

2.6.9 compression The groas veight of the Model 650 is 70 pounds (3g.8kg).

The maxi =u=

cross-sectional area is 133 square inches (0.086= ).

Thus, five times the weight (350 lbs.) is greater than 2 pounds per square inch ttnes the cross-sectional area (266 lbs. ).

For this analysis, the 1 cad vill be taken to be 350 pounds.

The maximum stress generated in a flat rectangular steel plate with all tges fixed and a unircr=ly distributed load over the surface of the plate can be ec=puted frc=:

S=

\\v}f]

2 rl G.623 (1

+

..v where: S: =aximum stress F: total load = 350 pounds t: thicknass of plate = 0.1345 inches v: vidth of plate = 8 3 inches 1:

length of plate = 10 inches

(

Reference:

Machinery's Handbcok, 20th Edition, p.44h, Eq.13)

Pro = this relationship, the maximum stress generated in the plate is 3hh6 pounds per square inch (23.8MN/=2). This figure is greatly belev the yield strength cf the material, h0,000 pcunds per square inch. Thus, it can be concluded that ec=pression vill not adversely affect the packace.

2.7 Hyrcthetical Ace'l_ent Conditions 271 Free Drop The Mcdel 650 was drop tested through a distance of 30 feet onto a steel plate. The container was found to experience no loss of shieldin6 cr integrity as a result of this test.

The test report is enclosed in Section 2.10.

2-5 REVISION O AUG.

3 1973 NOO O

Eased on this test result, we conclude that the Mcdel 650 can vithstand the Free Drop Condition.

2 7.2 Puncture The Model 650 was subjected to the puncture test of 10CFR71. A copy of the test report is enclosed in 2.10.

Based on this test result, we conclude that the Model 650 can withstand the puncture conditions of the hypothetical accident.

273 Thermal The thermal analysis is presented in Section 3 5 There it is shcvn that the =elting point of the caterials, except the potting cc= pound, used in the construction of the Model 650 are all greater than lkT5 F (800 C).

Also indicated is the previcus acceptability of this design (NRC Certificate of Cc=pliance No. 9032, Rev.1) using this evaluation.

To de=cnstrate that the radioactive scurce assembly will retain in a shielded position folleving the hypothetical thermal accident, the folleving analysis is presented. At the conclusion of the thermal test, it is assumed that the polyurethane foa= has ccepletely escaped frc= the container. The shield assembly is restricted frc= rotational =ovement by the titaniu= "U" tube which protrudes thrcugh the top of the bottc=

section of the container. Hence, the only freedc= of movement available to the shield is vertically tcvard the scurce holddown asse=bly.

Thus, it is concluded that the Model 650 satisfactorily =cets the recuire-

=t nts f or the hypothetical accident-ther=al evaluation as set fcrth in 10cPh(1.

2.7.h

'4ater In=ersion Not Applicable.

2.75 Su _ary of Datage The tests designed to induce mechanical stress (drop, puncture) caused mincr deformation, but no reduction in the safety features of the package.

The thermal test resulted in no reduction of the safety of the package.

It can be concluded that the hypothetical accident conditions have no adversa effect on the shielding effectiveness and structural integrity of the package.

2.S Scecial Form The Medel 650 Scurce Changer is designed for use with Tech / Ops scurce asse=blies Ah2b-1, Ah2h-9, A31401, E69702. These scurce asse=blies have

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2-6 hEVISION O AUG.

3 1973

been previcusly certified as special form radicactive materials (IAEA Certificate of Cc=petent Authority No. USA /015h/S, see Section 2.10).

We contend that these certificates are sufficient evidence that the requirements for special form radioactive =aterials, as established in IAEA Safety Series No. 6, Rev.1973, are satisfied.

29 Pael Pods Not Applicable

.'s EVISION O S_7

% 5.

3 1979

2.10 APPE TDDC

- Nuclear Metals, Inc., Test Report: Iron Uranium Alloying Test Report: Penetraticn Test, Model 650

- Test Report: Puncture and Drop Tests, Model 650 Descriptive Assembly Dravings, Source Assemblies

- IAEA Certificate of Cc=petent Authcrity :0. USA /015h/S 2-8 EEVISION O

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3 1979 t

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N U C L E A I! M E T A L S, I N C.

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...,...,...n 28 January 1974 Technical Operations, Inc.

Radia ion Products Division South Avenue Burlington, Massachusetts 01803 Attention:

fir. J. Lima Gentlemen:

In response to a request by Joe Lima of Tech Ops, a simulated fire test was performed on samples of bare denleted uranium in contact with mild steel, the object being to deteimine what, if any, alloying or melting would occur under these conditions.

TEST DATA:

/9 A 3/4-inch diameter x 5/8-inch lonq hare donleted uranium sperimen was set on a 1-itjch diameter x 1/3-inch thici; mild steel plate, placed in a thin wall ceramic crucible.

A mild steel cover plate was used on top of the crucible to act as a partial air seal.

The crucible was loaded in a preheated 1450 F resistance heated furnace, held for 35 minutes, then removed and allowed to air cool under a ventilated hood.

RESULTS:

l'o reaci inn was evidenced between the two metals.

Both separated readily and showed no alloying or meltina characteristics.

Oxidation of the uranium was about the same degree as that reported to Joe Lima on an earlier experiment.

The test was performed by fiMI on 25 January 1974.

Very truly your.5, C

kw 6johnG. Powers Project Engineer C\\

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c. s ' E/ISION O

?003 DRBINL

TEST REPORT RADIATION PRODUCTS DIVISION BY:

David Marcilli DATE:

26 July 1979 SUPJECT:

650 Penetration Test On 26 July 1979, a penetration test in accordance with 10CFR71, Appendix A, Paragraph 8 and IAEA Safety Series No. 6,1973, Paragraphs 714(a) and 714(b),

was perferned on a Tech / Cps Model 650 Source Changer, Serial Number 10h.

The = cst vulnerable location was deemed to be the vide side of the top cover, at a position where the holddown asse=bly vcula exrerience maximum shear if penetration cccurred.

1.

The hemispherical end of a vertical steel cylinder 1 inches in diameter, weighing 14 pounds, was dropped frc= a height of h0 inches ento the location specified.

The resultant damage was an indentation apprcximately 1 inch india =eterandlessthaniinchdeep. There was no damage which would affect the shielding or structural integrity cf the projector.

2.

The test in step (1) was repeated at the apprcximate midpoint of the cylindrical steel shell. Another small indentation (apprcximately 1 inch diameter by less than inch deep) was

=ade.

The resultant annnge in no way affected the shielding or structural integrity of the package. Documentary photcgraphs are enclesed.

Perfered by:

Witnessed by:

3J co-Divid Marcilli A. Kiklis DM/'d Encis.

2-10 REVISION O AUG.

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SECOND DROP

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W TEST REPORT DESCRIPTION:

DATE 9-4-G9 Shipping Container Model No. 650 Drop Test 1.

a)

The shipping container was dropped from a height of 30 ft. with the cover and a dummy source instalW The first drop landed in the corner of the hose plate, and bent it over.

b)

The second drop landed on the cover and distorted it.

There was no dam-age in either dre;' to the body of the e an eir.e:. The dummy source and z

cover were both removeable.

2.

The contain.er was dropped frem a height of 40 inches onto a sjeel bar of a)

G inch radius and 10 inches in length, it was dropped so as to land on the body of the container. The only apparent damage was a slight dent in the body. The dummy source and cover were both removeable and there was no apparent damage to the uranium shield.

CONCLUSION:

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B U R LIN G TO N, MA 01803 FIN!SH OWG TITLE 192 IRIDIUM SOURCE REFERENCE fyg,y,,, (==;>ny-y K

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DEPARTMENT OF TRANSPORTATION gg - 1 RESEARCH AND SPECI AL PROGRAMS A D MINf STR A TION 4'

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W A sHING TON. O C.

20590

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s' IAEA CERTIFICATE OF COMPETENT AUTHORITY Soecial Form Radioactive Material Encapsulation

,,, y o ;

Certificate Number USA /0154/S This certifies that the encapsulated sources, as described, when Icaded with the authori;:ed radioactive contents, have been demon-strated to meet the regul'atory requirements forspegial form radicactive mat erial as prescribed in IAEA and USA regulations for the transport of radioactive materials.

I.

Source Description - The sour.es described by this certificate i

identified as the Technical Operations, Inc., Models which are are described and constructed as follows:

Model No.

Caosule Stvle Approximate Size (in inches, diameter x length)

A424-1 B60001 or B60004

.25 x.97 A424-6 360001 or B60004

.25 x.97 A424-9 B60001 or B60004

.25 x.97 m i A410-20 E60001 or B60004

.25 x.97 ASa101 360006 Pellet, Wafer or Large Wafer

.23 x.90 A68309 C68310 Pellet or Wafer

.25 x.78 AF1401 360001 or B60004

.25 x.97 B69701 B60001 or B60004

.25 x.97 All capsules are constructed of either 304 or 304L stainless steel and conform with the following design drawings:

Cap sul e Style Drawing ';unber 360001 B60001 - 1 Rev. E and - 2 Rev. F l

E60004 B60001 - 2 Rev. E and B60004 - 1 Eev. D P60006 Pellet 960006 - 1 Rev H and 360001 - 2 nev. r 360006 Wafer B60006 - 1 Rev. E and B60004 - 1 Rev. D B60006 Large Wafer B60006 - 2 and B60001 - 2 Rev. F C68310 Pellet C68310 Rev. E and B68310-3 C68310 Wafer C68310 Rev. B II.

r,a d i o a c t i ve Contents - The authorined radicactive contents of these sources consist o Enot uore than the folic.ing acounts of Iridiu:-192 as solid metal:

EVIECN O r

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Cont ent s (Curies)

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A424-1 120 A424-6 120 A424-9 120 A424-20 240 A38101 240 A68309 120 A81401 120 B69701 120 III.

This certificate, unless renewed, expires December 31, 1981.

This certificate is issued in accordance with paragraph 803 of the IAEA RegulationsI, and in response to the November 3, 1978, petition by Technical Operations, Inc., Burlington, Massachusetts, and in consideration of the associated information therein.

Certified by:

/./h_ -

A fWE.

R.

R. Rawl, Health Thysicist (Date)

U.

S. Department of Transportation Office of Hazardous Materials Regulation

' ashington, D. C.

20390.

I" Safety Series No.

6, Regulations for the Safe Transport of T.adioactive Materials, 1973 Revised Edition", published by the International Atomic Energy Agency ( LAEA), Vi enna, Austria.

Title 49, code of Federal Regulations, Part 170-178, USA.

s REVISION O

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AUl 3 1373 t am O

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Ther=al Evaluation 31 Discussion The Model 650 Source Changer is a completely passive thermal device end has no mechanical cooling syste=s or relief valves. All cooling of uhe package is thrcugh free convection and radiation. The only heat sot ree is the (maxi =um) 2kOCi iridium-192 source. A conservative figure f ar the corresponding decay heat is 2 5 watts (see Section 3 4.1).

32 Su -ary of Thermal ~>roperties of Materials The celting points of the metals used in the construction of the Mcdel 650 are:

Depleted Uranium Metal 2070 F (1133 C) 0 Carbon Steel 2k53 F (1345 C) 0 Titanium 5308 F (1820 C)

Copper 19h0 F (1080 C)

(

Reference:

Machinery's Handbook, 20th ed. )

C The ridid colyurethane foa ha s a mini =u operating range of -10C F to 200c? (- 3 to 93'C).

It will decompose at the fire test tenperature of 1475cp ( 300C).

Decomposition vill result in gaseous byproducts which will burn in air.

33 Technical Specifications of Lc=conents fict Applicable 3.h

cr.al Ccnditions of Transmrt 3.h.1 Therral Model The heat source in the Model 650 is a maxi =um of 2LCCi of iridium-192.

The decay heat generated by the scurce can be calculated kncving that iridium-192 decays by electron capture and beta emission. The decay energy for both prccesses is approxf =ately 1.45 MeV (Refere:ce: Radio-logical Health Handbock, p. 403):

disint) x (1.6 x 10-13 ) x 2h0C1 = 2.C6 Watts lO J

1.h5 "eV x (3 7 x 1C s -Ci MeV The iecay heat source is ccnservatively taken as 2 5 watts.

3-1 REVISION O auG.

3 1979 n:

U, ' O J L

e 6

t To qualii'y as a Type B(U) package the additional requirements of IAEA Safety Series No. 6,1973, paragraphs 231 and 232 =ust be satisfied. The calculational =odel used to de=enstrate co=pliance with these regulations is described in detail in Section 3.6, along with the results of the analysis.

E?sentially, it is assumed that one-fourth of the entira decay heat lead is deposited unifomly in each of six sides. The smallest of the sides is assumed to reach the =axi=u= surface temperature. Heat trnasfer frc=

the side is restricted only to convective heat transfer frc= the upper face of the plate.

To =eec the additional requirements of paragraph 240 of the IAEA regula-tions, a separate analysis was performed. To do this, a heat balance was set up over the surface of the paedge, using the insolation data in Table III of the LEA Regulations. The decay heat source was considered negligible. The cuter shell was assumed to be insulated frc= the interior of the package. Heat transfer frc= the package was taken to occur by radiation, and over specific surface areas by free convection, A detailed description of the model is given in the analysis, in Section 3.6.

3.h.2 yaxi=um Te=peratures An examination of the =eltir4 points of the matcrials used in construction of the Model 650 show that the =axi=u= te=peratures encountered under normal conditicns of transport engender no loss of structural integrity or loss of shielding of the package. The specific Type B(U) analyses (Section 3.6) show the package te=perature to be below ho C (lc43F) in o

0 the shade and belev Tc C (1700F) when insolated.

3.k.3 Mini =u Te=peratures The mini =u= nc=al operating te=perature of the Model 650 is -4C C ( 40cF).

C This te=perature vill have no adverse effect on the package.

3 k.h "aximum Internal Pre oures Nomal >perating conditions generate negligible internal pressin es.

Any pre s sur-eenerated is significantly belcv that of the hypothetical accident pressur e, which is shavn to result in r less of shielding or contain=ent.

3.4.5 Maxi =ur The=al Stresses

~'he =axi=u= temp 2ratures that occur curir4 normal tranarcrt are icv enough to insure that thermal g:1dients vi'l cause no significant thermal stresses.

3.h.6 Evaluaticn of Mckage Performance fcr ';c=al Ccnditiens of Tran;;crt The the mal ecnditions of nc mal transport are obvicusly 'nsignificant frc= a functicnal point cf 712v for the Mcdel 650. A1so, the applic'ble

,_o REVI3 ION O W

~

x s,

F U

U U 't AU3.

3 1979 e

d

~.

conditions of IA?J Regulations for Type 3(U) packages have bean shewn to be satisfied by the Model 650.

35 Hypothetical Accident Themal Evaluation 351 Themal Ncdel The Model 650, including the scurce asse=bly, is assumed to reach the fire test te=rerature of 3000C (1475 F).

At this temperature the polyurethane potting cc= pound will have deccx:;osed and the resulting gases vill have escaped the package through the vent holes and the asse=bly joints which are not leak-tight.

352 package cenditions and Environ =ent The I'odel 650 underwent nt significant danage during the free drop and puncture tests. The package used in this analysis is considered undenged.

353 Package Te=reratures As indicated in 3 5 1, the package reaches a taximu= of 800 C (lh75 ?)

throughcut. An examination of the =elting points of the materials used in the construction of the Model 650 (except the potting cc= pound, as noted) indicates that there vill be no da=a6e to the package as a result of this te=rerature. The possibility of the for=ation of the iren-uraniu=

eutectic alley was addressed in Section 2.4.1, where it was concluded that the for=ation of t' 711cy was not a likely eventuality.

3 5.h Maximum Internal oressures The Model 650 packaging is open to the at=osphere, incuring that there vill be no pressure buildup vithin the package.

In section 3 6 there is an analysis of the scurce capsules under the fire test conditions. It is shcvn that the =aximum internal gas 54.6 pounds per square inch (377kN/=2). pressure at this temperature is The critical location for failure is the veld. An internal pressure cf Sh.6 pcunds per square inch (377k:I/=2) vill generate a =axt=u= stress of 291 pcunds per square inch (2.CIC;/=2) in the veld. At a te=rerature of 16CCe? (37C C) the yield strength of Type 3c4 stainless steel is lo,coopsi O

(6?:C:/ =" ).

0 Thus, at 800 C, the taxtnu= stress in the package vould be only 3% of the yield strmgth at that point.

355 t'aximum ~ne mal St n sses There are no significant the=al stresses generated during the theral test.

EVISION O

.u.

A U 3.

3 l979

' J.I dbJ e

356 Evaluation of Package arformance The 'edel 650 vill undergo no loss of structural integrity or shielding when subjected to the conditions of the hypothetical thermal ace?. dent.

The pressures and tenperatures generated have been demonstrated to be within acceptable limits.

3h REVISICu u A U S.

3 1973 I

i.

! :)U

'e

~.

1 3.6 AFPriDIX

- Model 650 Thernal Analysis:

IAEA Safety Series !!o. 6, 1973, paragraphs 231, 252

- Model 650 Ther=al Ana4 sis: IAEA Safety Series !?o. 6, 1973, parrigraph 240

- Iridit= Cap ules - Therral Analysis 3-5 REVISION O

~1 3 1973 l'_ ;,(

L' n /

e 6

2 Model 650 - Ther=al Analysis Type B(U), Paraersphs 231, 232, IAEA Safety Series :Jo. 6,1973 This analysis is perferred to de=onstrate that the Model 650 Source Changer =eets the specific Type B(U) ther=al requirements of paragraphs 231 and 232 of IAEA Safety Series I;o. 6,1973, i.e., that the maxi =um surface te=perature does not exceed 50 C in the shade, assuming 38 c a=bient air te=perature.

To assure conservatis=, it is assumed that: (1) the entire decay heat (2 5 uatts) is deposited in the exterior faces of the Model 650, (2) the interior of the Model 650 is perfectly insulated, providing heat transfer from the vall enly to the atmosphere. This is equivalent to assu=ing that each vall has a heat source. The rectargular shape of t he container means that each face eclipses a different a=ount of the solid angle through which the radiation (and thus decay heat) is distributed. To (conservatively simplify, it is assumed that each of the six exterior faces receives,, of the total scurce (0.63 vatts) uniformly distributed over the face.

Considering the smallest face as undergoing one-dimensional convective heat transfer:

(

~/ >

/

q Interior (Insulated)

T T

a a

v

.\\\\

/.

Air gall

+ T where:

Tv

_a

=

a hA temperature at the vall outer surface T

=

y ambient air temperature (38 c)

T

=

3 (decay) heat source (0.63 vatts) q

=

A

- surface area of the smallest face (0.^53 square =eters) free convective heat transfer coefficient for air h

=

5 watts /=eter2-C, (

Reference:

Heat T n nsfer, J.P.

3 fj i, [',

Holman, k5th Editien, p.13)

' REVISION O t-6 rol J 13e3 O

6

2 0

h nomi condi-Thus the =axi=u temperature at the vall m 4

s tions of transfort. This sati8fies +whe,equire=ents of the aforementioned regulations,

(

c,

/ E'ESION O al 3 !s79 u

-7 O

O

1 Model 650 - Themal Analysis Type B(U), Paragraph 240, IAEA Safety Series !!o. 6,1973 This analysis is perfomed to demonstrate that the Model 650 Source Changer meets the specific Type 3(U) themal require =ents of paragraph 240, IAEA Safety Series ? o. 6,1973 This paragraph requires that the maximu= surface te=perature of a Type E(U) package not exceed 820C under nomal conditions of transport, given insolation as outlined in Table III of the regulations and an ambient temperature of 38 c (10coF).

~

The calculational model consists of taking a steady state heat balance over the surface of the package. To facilitate calculations, certain simplifying assu=ptions are made. These are outlined below:

Insolation:

800 cal /c=2-12hr (775 W/=2) for the top surface, h00 cal /c=2-12hr (338 W/=2) for the cylindrical outer shell, 200 cal /c=2-12hr (19h W/=2) for the sides of the top lid. It is assumed that the insolated surface consists of: the top face at the lid, the cuter shell (a cylinder 8 3 inches, 211==

high and 7 53 inches,191== in diameter), and the rectangular surfaces of the top lid. This is a conservative area, as the overhangin6 edges of the top cover and top lid provide shade over sc=e of the side area.

The package is finished with either :ussett or green enamel. The solar abscrptivitics of these enamels are 0.81 and 0 76, respectively (

Reference:

Thermal Radiation Properties Survey, G. G. Gubareff et. al., 2nd ed.,

1900, p.200).

A conservative figure of 0 90 was chosen as the package absorptivity.

IMeay Heat Load The decay heat load (2 5 vatts) is assumed negligible.

Package Orientation The package rests en the bottc= cover, i.e.,

in the no mal transport orientation.

Heat Transfer Mechanis=s The Model 650 is assumed to undergo free convection and to radiate to the environ =ent.

The inside faces are considered to be insulated, so there is no ecnduction into the package. Further, the sides are taken to be thin enough so there are no te=perature gradients present.

-1 3-8 EVISION O AUG.

3 1973 p ' ' fj

(

O O

1 Radiaticn:

The package is assuced to radiate frc= the top surface of the top lid, bottc= surface of the bottc= lid, frc= the rectangular sides of the lid, and frc= the cylindrical cuter shell. This surface area is smaller than the actual surface area of the package, providing for a conservative radiative heat 1 css estinate.

The solar emissivity is ccnservatively chosen to be 0.8 (

Reference:

Thermal Radiation Procerties Survey, G.G. Gubareff, et.al., 2nd ed 1900, pp 217, 220).

Convective Top:

The top surface of the package is assu=ed to undergo free convection as a horizontal flat plate. The heat transfer coefficient is given by:

1.32 [a T }

h.

=

( L /

(

Reference:

Heat Transfer, J. P. Holman, h h ed t

197o, p.253 )

Where L is the average of the length of the sides, 2h2:= : Thuc:

h

= 19 @ T) t Convection Sides:

The Model 650 is constructed in such a way that there are overharging edges which eculd interfere with convection in the vertical direction. On the top lid this overhang is assumed to ec=pletely prevent convection over the sides. On the cuter shell, it is assumed that free convection occurs only where this overhang is less than 3==.

This is true for 1/3 of the cylinder:

Top viev

,m q

1 i

)

(

1

's

)

\\

~

a I'

^ 2 c' -g 1

=

',a REVISION O Ats.

3 t373 m

e O

1 Thus, for convection along the sides, the 650 is considered to be a right cylinder of 0.oh23:2side surface area.

The convection coefficient is taken to be:

1.42 [a T )

h

=

s (L /

(

Reference:

Heat Transfer, J.P. Holman, 4th ed., 1976, p 253) where L is the height of the cylinder, 211=m.

Thus:

2.10 (AT)

• h

=

3 Taking a heat balance over the surface of the ptekage:

91n " 9 rad

• 9et cs 9 n

• 91 V 1

9 rad *8 6A (T Ta )

r v

= (5.669 x 10-8) (o.8) (0.289) [T (311 k) 3 v

h JL aT vhere o T 9et T - Ta

=

=

y

~

= [19 AT)

]

(0.0532)6 T

/

h As AT 9c3

=

3

= (2.-10 (AT)

] (o.cho3) o T Iteration yields a vall te=perature T of 76.h C.

Thus, it is concluded y

that the Model 750 is in cc=pliance with paragraph 2ho of IA::A Safety Series No. 6, 1973

.?

t

~'" 3 - 10 REVISION Q AUG.

3 1979

1 Iridium Source Capsules - Themal Analysis gypothetical Fire Conditions This analysis is intended to demonstrate that Tech / Ops source capsules which are of 0.25 inch (6.35==) dia=eter, seal velded to a mintau=

penetration of 0.020 inch (0 51==), =ade of Type 30h or 304L stainless steel, and licensed as special for containers under IAEA Safety Series No. 6,1973, also =eet the requirements of paragraph 238, IAEA Safety Series No. 6,1973, i.e.,

contain=ent under specified ther=al test conditions.

The actual contain=ent vessel for the radioactive caterial is the velded source capsule.

These capsules are all 0.25 inches (6.35==) in dia=eter and less than 1 inch (25.L==) in length.

The internal volume of the source capsules contains only iridiu= =etal (as a solid) and air.

It is assumed at the time of loading that the entrapped air in the capsule is at standard te=perature and pressure (2coC,1.03 kilograms per square centi =eter). We contend th a conservative assumption because, during the velding proces,at this is s, the in-ternal air is heated, causing sc=e of the air mass to escape before the capsule is sealed. When the velded capsule returns to ambient te=perature, the internal pressure vould be sc=ewhat reduced.

As described in Tech / Ops standard source encapsulation procedure, the

=inicum veld penetration is 0.020 inch (0 51==).

Under conditions of internal pressure, the critical location for failure is this veld.

Since the capsule has an outside diameter of C.25 inch (6 35==)E)this veld has a cross-sectional area of 0.014 square inches (0.093c=

Under conditions of paragraph 238 of IAEA Safety Series, No. 6, it is assu=ed that the capsule could reach a temperature of 1475o7 (800 C).

0 Using the ideal 6as law and requiring the air to occupy a constant vclure:

PTI2 P

=

p Tl 1

initial pressure (1.03?kg/c=2) p

=

initial te=perature (293 k)

T1

=

T2 fin 1 te=perature (1093 k)

=

2 The internal gas pressure could reach 3 8h kg/cs.

It is assured that the capsule can be treated as s thin-valled, cylif.drical pressure vessel.

'.) 3 - 11 REVISION O AUG.

3 1975 e

1 The maximu= longitudinal tensile stress can be calculated by writir6 a longitudinal force balance through the veld:

stress x area - pressure x area

=0 s

p (Do - Di2) 2 2

-P D4

=0 S,-

4 7

where S1=

1 ngitudihal stress Do=

outer diameter (0.635c=}

Di=

inner dianeter (0 535c=)

pressure (3 8hkg/c=2)

P

=

2 Thus, the longitudinal stress is 128 newtons /cm,

The hoop stress can be found in a st=ilar fashion.

Taking t longitudinal cross-section and st==ing forces:

=0 hoop stress x areas - pressurc x areap 2S L

_DL=0 i

h where Sh = hoop stress L = length of cylinder t = thickness of wald (0 51m =)

Thus,- the hcop stress is 201 newtons /c="o At a te=cerature of 1600 ? (870 C) the y$ eld strength is of type 3Ch 0

i stainles's steel is 10,000 psi (6900 n/c= ).

Thus, the pressure induced stresses are less than 3% of the yield strength at 800 C.

~

U94

-a_

REVISICN O

, _m

sgic s w a, w

e O

t 4.

Contairment 4.1 Contai=ent Foundary 4.1.1 Contain=ent Vessel The contain=ent syste= for the Model 650 source Changer is the scurce asse=bly. The actual contain=ent for the radioactive material is the velded source capsule, either style E60001 or E60004. Both source styJes are currently certified (IAEA Certificate of Cc=petent Authority :To.

USA /015h/s) as special for= contairrent for radioactive materials.

The capsules are made of Type 304, or 3chL stainless steel. They are seal velded with a mini =u= veld penetration of 0.020 inch (0 51==).

The capsules are rcunded cylinders 0.25 inches (6.35:=) in diameter x 0 97 inches in length (2h.6==).

Appropriate drawings are enclosed in section 2.10.

h.1.2 Containment penetrations There are no penetrations of contai=ent. The entire source asse=bly is seal velded to provide confor=ity to special for= requirements.

h.l.3 Seals and Welds The contain=ent vessel is tungsten inert gaa welded as described in Tech / Ops standard source encapsulation precedures (see section T.h).

The =ini=u=

veld penetration is 0.020 inch (0 51==).

This has proved acceptable for certifying this vessel as special fer=.

h l.4 Closure

!!ct Applicable 4.2 Require =ents for ?Ic nal Conditionc of Transport h.2.1 Release of Eedioactive Material The source assemblies used all meet the recuire=ents of special fcr=

radicactive =aterial as delineated in IAEA safety Series ;o. 6.,1973 and 10CFRT1. Thus, there vill be no release of radioactive =aterials under conditions of nor=al transport.

h.2.2 Prcssuricatica of Contairrent Vessel The scurce asse=blies used all =eet the require =ents of special for.1 radicactive =aterial. Pressure buildup due to the conditions cf tha hypothetical 'hermal accident has been shcvn to create stresses well "i

REVISION O E 3.

3 !a79 L

.i/J e

t below the structural li=its of the capsule (see Section 3 5 ).

Thus, the conteirrent vessel vill withstand the pres ure variatiens of nc: al transport.

h.2 3 gelant Contamination Ilot Applicable 4.2.4 Coolant Loss Not Applicable k.3 Contain=ent Recuirerents for the Hycothetical Accident Condition 431 Fissicn Gas h oducts Not Applicable k.3 2 Releases of Contents Tr i hypothetical accident conditiens as outlined in 10CFR71, Appendix E, l., 2, and 3. have been sho m (Sections 2.7.1, 2 7 2 and 3 5 respectively) to result in no loss of package contain=ent.

'-2 REVISION O AUG.

3 1319 L

tll0

t 5

Shielding Evaluation 51 Discussion and Pesults The Model 65L ic shielded with 35 pounds of depleted u-aniu=.

The uranium metal is cast around the titaniu= "U" tube which holds the source.

A radiation profile of Model 650 S.II. 257 containing 220 5Ci of iridiu=-192 (see Section 5 5) was made. An extrapolation for a 60Ci source yielded the results which are presented in Table 51 From this data and from previous acceptability (NRC Certificate of Ccepliance No. 9032, Rev.1 enclosed in Secticn 13), it is concluded that the Model 650 co= plies with the regulatory standards in 1CCFR71 and IAEA Safety S'eries No. 6, 1973 TABLE 5 1 SLEMARY OF MAXIMUM DCSE RATES

(=Rjhr)

Contact At 1 Meter Side Top Ecttom Side Top Bottcm Gn-n 173 76 43 1.0 1.0 1.0 Neutron Not Applicable at Appliceble Total 173 76 43 1.0 1.0 1.o Hypothetical accident conditions will result in essentially no change in the above readings.

52 Source Specification 5 2.1 Gn -a Source The g,-n source used is encapsulated iridium-192 in the quantity of up to 2EO curies.

5 2.2 Neutron Source Not Applicable 53 Model specificaticn Not Applicable 5-1 REVISION O AU;.

3 1979

/

s

!j

\\

()

=

1 5.k Shielding Evaluation The Model 650 shieldire evaluation was perfor=ed on Model 650 Serial

h:tber 257 containing 220 5 curies of iridiu=-192. Tne radiation profile is included in Section 5 5 attrapolation of this data to the capacity of 240 curies, (Section 51) clearly indicates that the Model 650 confcI=s to the regulatcry radiation li=its. As the hypothetical accident tests (Section 2 7) revealed no change in the shielding arrangement, it is concluded that shielding after the hypothetical accident is essentially unchanged. Therefore, the radiation profile indicates the package will be within acceptable limits.

i 5-2 REVISION O AUG.

3 1979

1 55 APPE:TDIX

- Radiation Profile Model t350 Serial Nu: ber 257 c.:

REVISION O AUG.

S 1979 t

o

=

Top 1

Front i

T t

Left Right k

)

Rear I

og Ecttc=

RADIATION FROFILE Model 650 serial ::u=ter 257 Containing 220 5C1.192Iridium Location At Centact at 1 Meter Top TO

< l.0 Ecttc=

43

< l.0 Front 160

< l.0 Rear 130

< l.o Ieft 130

< l.o Right 110

< l. 0 Notes:

1.

All intensities are expressed in units of Millircentgens per hour.

2.

Intensities exprassed are the caximu=_ intensities on the ceasured side.

3 Measurements were made with an A/:I P:R 27-J survey =eter.

- u o, 5-h REVISION O AUG.

3 1973 p

+

ge 4

--.g

--m

-m e

wwhme-,

,.mmeg

s 1

6.

criticality Evaluation Ilot Applicable 6-1 REVISION O AUG.

3 1979

's 0 i t

/

3

7 Operating Precedures 71 Procedures for Loadira the Package Section 7.h describes the procedure for fabricating the special form source encapsulation. Section 7.4 contains the procedure for loading this source assembly into the package and preparing the package for transport.

72 Procedures for Unicading the Package Section 7.4 contains the procedure for unlcading the source asse=bly frc=

the package.

73 Preparation of an Empty Package for Transport Section 7.k has the procedure for preparing an empty package for transpcrt.

U-t

., l REVISION O

'~

AUG.

3 1979 o

't 7.4 APPE'TDIX

- Source Encapsulation Procedure

- Source Charging Precedure

- (E=pty) Packa6e Transport Procedure 7-2 REVISION O AUG.

3 1979

,)

t

L RADIATION SAFETY MANUAL Part II In Plant Operations Section 2 ENCAPSULATION OF. SEAIZD SOURCES A.

Personnel Require =ents Only an individual qualified as a Senior Radiological Technician shall perfor= the operations associated with the encapsulation of 192 Iridium. There cust be a second qualified Radio 1cgical Technician available in the building when these operations are being perfor=ed.

3.

General Requirements The 192Iridiu= loadi g cell shall be used for the encapsulation of solid retallic 192Iridiu= and the packaging of sealed sources such DO 169 as 17 h uliu=, 137Cesiu= and Ytterbiu=.

Solid metallic Cobalt not exceeding one curie =ay be handled in this cell also.

The maxiru= a= cunt of 192 Iridium to be handled in this cell at an,y one time shall not exceed 1000 curies. The maximu= accunt of 133Cs to be handled in this cell at any one time shall not exceed 100 curies.

This cell is designed to be operated at less than et=ospheric pressu e.

The exhaust blcwer provided shall not be turned off except when the cell is in a decontaminated conditicn.

Sources shall not be stored in this cell overnight or when cell is unattended.

Unencapsulated caterial shall be returnad to the transfer centainers and encapsulated sources transferred to approved source ecntainers.

7nen any of the "through-the-vall" tools such as the velding fixture or transfer pigs are removed, the openings are to be closed with the plugs provided. These tools shall te decenta=inated whenever they are recoved fro = the hot cell.

C.

Precaratory Procedure l.

Check velding fixture, capsule drawer and =anipulator fingers frc= cell and survey for centa=ination.' If contarination in excess of 0.001 ACi cf removable contamination is found, these ite=s rust be decentatinated.

2.

If the welding fixture er the electrodes have teen changed, perfer= the encapsulaticn precedure crittin6 the insertica of any activity.

Examine this du==y capsule by sectioning thru veld.

Weld penetraticn rust be not less than 0.C20 inch.

BEE 3ICN O

.-.e._

ii,;9 y,,

~

n a c.

= l313

______"O I

If veld is sound and penetration is at least 0.020 inch, the preparation of active capsules may proceed.

If not, the condition responsible for an unacceptable veld must be corrected and the preparatory procedure repeated.

3 Check pressure differential across first absolute filter, as

=easured by the manoteter on the left side of the hot cell.

This is about } inch of water for a new filter. When this pressure differential rises to about 2 inches of water, the filter must be changed.

D.

Encapsulation Procedure 1.

Prior to use, assemble and visually inspect the two capsule compenents to determine if veld none exhibits any =isalignrent and/or separation. Defective capsules shall be rejected.

2.

Degrease capsule cc=ponents in the Ultrasonic Bath, using isopropyl alcohol as degreasing agent, for a period of 10 0

minutes.

Dry the capsule co=penents at 100 C for a =inicu=

of twenty =inutes.

~

3 Insert capsule ec=penents into hot cell with the posting bar.

k.

Place capsule in veld positiening devi:e.

5 Move drawer of source transfer container into hot cell.

c.

Place proper a= cunt of activity in capsule.

Disposable funnel =ust be used with pellets and a brass rivet with vafers to prevent contamination of weld none.

7 Recove unused radicactive material frc= the hot call by with-drawing the drawer of the source transfer container frc= the cell.

8.

Be cve funnel cr rivet.

9 Assemble capsule components.

10. Weld adherin6 to the following conditions:

Electrode spacing.C21" to.C2h" a.

centered en joint +.002"; use jig for this purpose.

b.

Freficv argon, flush 10 seconds.

c.

Start 15 amps.

d.

Weld 15 amps.

r I J e.

Slope 15 amps.

f.

Post ficv 15 seconds

.2.2 RTl!SICN O

'a AU 3.

3 157S e

5

i 1

11.

Visually inspect the veld.

An acceptable weld cust be continuous without cratering, cracks or evidence of ble. out.

If the veld is defective, the capsule cust be cleaned and revelded to acceptable conditions or disposed of as radioactive vaste.

12.

Check the capsul 3 in height gauge to be sure that the veld is at the center of the capsule.

13 Wipe exterior of capsule with flannel patch vetted with EDTA solution or equivalent.

14. Count the patch with the scaler counting system. Patch cust show no core than.005/4Ci of contamination. If the patch shows more than.005juci, the capsule =ust be cleaned and reviped.

If the revipe patch still shows = ore than 0.005jaCi of contamina-tion, steps 8 through 11 must be repeated.

15 Vacuum bubble test the capsule.

Place t he velded capsule in a glass vial containing isopropyl alcohol. Apply a vacuum of 15 in Eg(Gauge). Any visual detection of bubbles vill indicate a leaking source.

If the source is determined to be leaking, place the source in a dry vacuu= vial and boil off the residual alechol.

Reveld the capsule.

16. Transfer the capsule to the svaging fixture.

Insert the wire and connector assembly and svage.

Hydraulic pressure should not be less than 1250 nor core than 1500 pounds.

17 Apply the tensile test to assembly between the capsule and connector by applying proof load of 75 lbs. Extension under the load shall not exceed 0.1 inch.

If the extension exceeds 0.1 inch, the source cust be disposed of as radioactive vaste.

18. Position the source in the exit port of hot cell. Withdraw all personnel to the centrol area.

Use remote control to in. sert source in the ion chamber and position the source for taximum response. Record the ceter reading. Compute the activity in curies and fill out a temporary source tag.

19 Using remote control, eject the source from cell into source changer through the tube gauce wipe test fixture. Monitor before reentering the hot cell arma to be sure that the source is in the source changer. Re=ove the tube gauce and count with scaler counting system. This assay =ust show no more than 0.005,aci.

If contamination is in excess of this level, the source is leaking and shall be rejected.

20.

Cc=plete a Scurce Loading Log (Figure II.2.1) for the operation.

,D tOb r

II.2 3 7-5 REVI5'ON O Agi 3 ;579

Technical Operations Model 650 Standard Source Charging Procedure Wear personnel =onitoring devices durirg all source chargir4 operations.

Monitor all operations with a calibrated, operable survey meter.

Note: All the precautions used when making radiographic exposures cust be followed.

1.

Upon receipt of the source changer, survey the source changer to ensure that the source is in the proper storage position.

2.

Locate the source charger and projector in a restricted area.

Locate the devices so as to avoid sharp bends in the guide tube or control housing.

The control cable housing bend radius should not be less than 36 inches (0 91h=), and the guide tube bend radius should not be less than 20 inches (0 508=).

3 Set the projector as for an exposure.

4.

Remove the cover frc= the source changer by breaking the seal wire and re=oving the bolts.

5 Re=ove the source holddown cap by breaking the seal wire and unbolting.

CAITTION: When the source holddown cap is removed, the source connector is exposed. Care cust be taken to ensure the source is not dislodged when handling the changer.

6.

Connect one end of a guide tube extension to the projector and the other end to the fitting above the empty cha=ber in the source changer.

7 Close and latch the source guides.

8.

Crank source into the source charger.

a.

Survey this operation with a gnmn survey =eter to be sure scurce has been transfered frc=

projector to changer, b.

With a survey =eter verify radiation level dces not exceed 2CC=r/hr at the surface of changer.

i i., REVISIT: O 7-6 AUG.

3 ;g7; s

'l 9

Open the source guides. Disconnect the drive cable frc= the source assembly by =oving the lock pin devn and sliding the drive cable connecter out through the keyway.

10.

Disconnect the guide tube frc= the source charger. Connect the guide tube to the fitting above the chamber containing the new source.

11.

Couple the drive cable to the source by depressing the lock pin, sliding the drive cable connector into the keyway, and releasing the lock pin. Test for proper en8agement.

12.

Close and latch the source guides.

13 Crank source to full retraction within projector.

a.

Survey this operation with a gnr-m survey =eter to be sure scurce has been transferred into the projector.

b.

With a survey =eter verify radiation level does not exceed 200 =r/hr at the surface of the projector.

IL.

Lock the projector.

15 Disconnect the source guide tube frc= the source changer.

16. Affix the identification plate of the new source to the projector and attach the identification plate of the old scurce to the source holddown cap.

17 Bolt the source holddcvn cap in place and seal wire.

18.

Prepare source charger for shipping:

a.

Attach ID plate of old source to source changer.

b.

Eolt changer cover in place and seal.

c.

Again survey the source changer to insure that the radiation level does not exceed 200 =r/hr at the surface of the changer, d.

Survey the radiation level at a distance of 3 feet frc= the surface of the scurce changer. This I

radiation level shculd not exceed 10 =r hr.

The highest radiaticn level measured at three feet frc= the container is used to deter =ine the Transport Index in acccriance with h9CFRlT3 389(h).

l Affi< the prcper shipping labels and return to e.

Technical Operations, Inc.

REVISION C zug 3 gg 7-7 m

1 Technical Operations Model 650 Procedure for Shipment of Source Q. anger (Empty)

Wear personnel conitcring devices during all source changing operaticns.

Monitor all operations vi+.h a calibrated, operable survey meter.

Note: All the precautions used when taking radicgraphic exposures must be follcwed.

1.

Upon receipt of the source changer, survey the source changer to ensure that the source is in the proper storage position.

2.

Locate the source changer and projector in a restricted area.

Lccate the devices eo as to avoid sharp bends in the guide tube cr control housing.

The control cable housing bend radius shculd not be less than 36 inches (0 914=), and the guide tube bend radius should not be less than 20 inches (0 508m).

3 Set the projector as for an exposure.

4.

Re=ove the cover frem the source changer by breaking the seal wire and re=cving the bolts.

5 Remove the souree holddown cap by breaking the seal vire and unbolting.

CAIRION: When the scurce holddown cap is re=cved, the source connector is exposed. Care cust be taken to ensure the source is not dislodged when handling the changer.

6.

Connect one end of a guide tube extension to the projector and the other end to the fitting above the new scurce in the scurce changer.

7 Close and latch the source guides.

8.

Couple the drive cable to the scurce by depressing the lock pin, sliding the drive cable connector inte the keyway, and releasing the icek pin. Test for proper engagement.

9 C1cce and latch the scurce guides.

10.

Crank source to full retraction within projector.

a.

Survey this operation with a gn-a survey reter to be sure source has been transferred into the prcjec cr.

3-t iU/

i-REVISICN O 7-8 AUG.

3 1973

1 b.

With a survey meter verify radiatica level does not exceed 200 tr/hr at the surface of the projec-tor.

11.

Disconnect the source guide tube frcm the scurce charger.

12.

Affix the identification plate of the new scurce to the projector and attach the identification plate cf the old source to the source holddcvn cap.

13 Bolt the source holddcun cap in place and seal wire.

14.

Affix a green "e=pty" tag to source changer.

15 Ferfcrm a vipe test of the source changer to assure that the radiation observed is less than 0.001 microcuries per 100 square centimeters.

16. Survey the source changer to assure that the radiation levels do not exceed 200tR/hr at the surface ncr 1CtR/hr at three feet frcm the surface.

17

' ark the source changer: Radiunctive "LSA".

Affix the proper shipping labels to the package.

13.

Ccaplete the proper shippire papers as specified in Tech / Ops Radiation Safety Manual II.6.3E(h),(5),(6).

(

l:0

,u HEVISION O 7-9 '

AUG.

3 1979 _

e

1 8.

Acceptance Tests and Maintenance Program 8.1 A :ectance Tests 8.1.1 Visual Inspection The package is visually exa=ined to assure that the appropriate fasteners are seal wired properly and that the package is properly =arked.

The seal veld of the radicactive scurce capsule is visually inspected for proper closure.

3.1.2 Structural and Pressure Tests The svage coupling between the scurce capsule and cable is subjected to a static tensile test with a lead cf seventy five pcunds. Failure of this test will prevent the source asse=bly frc= being used.

8.1 3 Leak Tests The radicactive scurce capsule (the primary containment) is wipe tested for leakage of radicactive contamination. The source capsule is subjected to a vacuu= bubble leak test. The capsule is then subjected to a second wipe test for leakage of radicactive contamination. These tests are described in Section 7.h.

Failure of any of these tests will prevent use of this source assembly.

S.I.h Cc=ponent Tests All ec=penents of the package are inspected on a 100% basis for confer =ance with engineering drawings and specifications. The uraniu= shield is tested for shielding integrity by radiation measurement pricr to inecrporation into the package.

3.1 5 Tests fer shielding Integrity The radiation levels at the surface of the package and at three feet frc=

the surface are =easured using a small detector survey instrument (i.e.

AN/PER-27). These radiation levels, when extrapolated to the rated capacity cf the package, must not exceed 200 millircentgens per hour et the surface nor ten =illircentgens per hcur at three feet frc= the surface of the package. Failure of this test vill prevent use cf the package.

3.1.0 Thermal Acceptance Tests L

I Not Applicable g_1 EASION O AUG.

3 1979 e

3.2 Maintenance Pro gam 8.2.1 Structural and Pressure Tests Not Applicable 8.2.2 Leak Tests As described in Section 8.13, the radioactive source asse=bly is leak tested at manufacture. Additionally, the scurce asse=bly is vipe tested for leakage of radioactive conta=ination every six tonths.

8.2 3 subsyetem Maintenance The package is inspected for tightness cf fasteners, proper seal vires ana general ecndition pricr to each use.

8.2.k Valves, Rupture Discs and Gaskets Not Applicable 8.2 5 Shielding Prior to each use, a radiation survey of the package is made to assure that the radiation levels do nct exceed 200 millircentgens per hcur at the sur-face nor ten millircentgens per hcur at three feet frce the surface.

8.2.6 Thermal Not Applicable 8.2 7 Miscellanecus Inspections and tests designed for secondary users of this package under the general license provisions of 10CFR71.12(b) are included in Section 7.h.

t m

3-2 REVISICh V AUG.

3 1979

.s S

O