ML19344A497

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SAR Technical Operations Model 920 Type B(U) Package
ML19344A497
Person / Time
Site: 07109143
Issue date: 08/01/1980
From:
TECH/OPS, INC. (FORMERLY TECHNICAL OPERATIONS, INC.)
To:
Shared Package
ML19344A496 List:
References
16969, NUDOCS 8008200493
Download: ML19344A497 (50)
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Text

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e-The outer packaging is designed to avoid the collection and re-tension of water. The package has a smooth finish to facilitate decontamination. The radioactive material is sealed inside a stainless steel source capsule. The capsule acts as the containment vessel for the radio-active material. 1.2.2 Operational Features The source assembly is secured in the proper storage position by means of the locking assembly. A lock slide engages an undercut in the source assembly preventing movement in either the forward or rearward direction. The lock slide is held in position by the selector ring. The selector ring is secured in the " LOCK" position by means of a key operated lock mounted on the rear plate of the projector. In the event that the lock failed, the source assembly would be contained within the package. The source position indicator slide would prevent movement in the forward direction. Interfer-ence between the source assembly and source tube would prevent move-ment in the recrward direction. 1.2.3 Contents of Package The Model 920 is designed for the transport of iridium-192 in quantities of up to 240 curies as Tech / Ops source assembly 90003. This source assembly contains either Tech / Ops Model 90004 or 90005 source capsule which satisfies the criteria for special form radio-active material in accordance with 10CFR71 and LAEA Safety Series No. 6, 1972 Edition. Revision 0 1 Aug 1980 1-2 ~ G

e - i 1 i j 1.3 APPENDIX i Descriptive Assembly Drawings, Model 920 1 Dwg 92090 Sheet 1 of 3 Dwg 92090 Sheet 2 of 3 Dwg 92090 Sheet 3 of 3 i l Dwg 90090 Sheet 3 of 4 Dwg 90090 Sheet 4 of 4 1 Dwg 90091 Sheet 1 of 1 4 I i I r j Revision O 1 Aug 1980 i 1-3 4 ~ee, w-- ,-,..--enm-,- w.,---.,---w-e- -,rw, ,--,,,,n,,,,, ,,,_en,,e,.,, m.v,,,,,,,_v-,v,,--~.,--r,wcm.n n m,-w -. y g gm -r,-m.-,,,,--,-, -,.wa-we,~---

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2. Structural Evaluation 2.1 Structural Design l 2.1.1 Discussion Structurally, the Model 920 consists of four components: source capsule, shield assembly, outer housing and locking assembly. The source capsule is the primary containment vessel. It satisfies the criteria for special form radioactive material. The shield assembly fulfills two functions. It provides shielding for the radioactive material and, together with the locking assembly, assures proper positioning of the source. The outer housing provides the structural strength of the package. The locking assembly secures the source assembly in the shielded position in the package and assures positive closure. 2.1.2 Design Criteria The Model 920 is designed to comply with the requirements of ICCFR71 and IAEA Safety Series No. 6,1973 Revised Edition. 2.2 Weights and Centers of Gravity The Model 920 weighs 47 pounds (21kg). The shield assembly contains 31 pounds (14kg) of depleted uranium. The center of gravity was located experimentally. It is located approximately 4.5 inches (114mm) from the front surface, 3.0 inches (76 mm) from the bottom surface and 2.7 inches (68cm) from the right surface. 2.3 Mechanical Properties of Materials ^ The housing of the Model 920 is fabricated from Type 304 stainless steel. This material has a yield strength of 35,000 pounds per square inch 2 (241MN/m ), Drawings of the Models 90004 and 90005 source capsules are enclosed in Section 2.10. These source capsules are fabricated from Type 304 or Type 304L stainless steel. Each capsule is sealed by tungsten inert gas welding. 2.4 Ceneral Standards for All Packages 2.4.1 Che=ical and Galvanic Reactions The materials used in the construction of the Model 920 are uranium =etal, stainless steel, tungsten, bronze, carbon steel, and aluminum. There will be no chemical or galvanic action between any of these components. There is no iron-uranium interf ace in this package. Therefore, there is no possibility of the formation of an iron-uranium eutectic alloy at elevated te=peatures. Revision 0 1 Aug 1980 2-1

2.4.2 Positive Closure The source assembly in the Model 920 cannot be exposed without opening a key operated lock. Thus, positive closure is provided. 2.4.3 Lifting Devices The Model 920 is designed to be lif ted by its handle. The handle is fabricated from aluminum and attached to the end plates by means of four No. 10-32 UNC flat head machine screws. The weakest area of the handle is the screw attachment. The stress area of each screw is 0.0175 inz (11,32,2). The yield strength of these screws 2 is greater than 40,000 pounds per square inch (276MN/m ). Therefore each screw can support 700 pounds (3.1kN) or more than fourteen times the weight of the package without exceeding the yield strength of the material. 2.4. 4 Tiedown Devices The handle is also used as a tiedown device. As de=onstrated in Section 2.4.3, each of the four screws attaching the handle can support more than fourteen times the weight of the package with-out exceeding the yield strength of the material. 2.5 Standards for Type B and Large Ouantity Packages 1.5.1 Load Resistance onsidering the package as a simple beam supported on both ends with a uniform load of five times the package weight evenly distributed along its length, the maxi =um stress generated can be computed from: a - !1. 8Z where a: Maximum Stress Generated F: Total Load (235 pounds; 1047 newtons) 1: Length of Bean (13 inches; 330mm) a 3 Z: Section Modulus (2.32 in ; 38,018mm )

Reference:

Machinery's Handbook, 21st Edition, p.404) The package is assumed to be a rectangular shell 5.04 inches (128=m) wide and 5.75 inches (146mm) high with a wall thickness of 0.06 inch 3 (1.5mm). Consequently, the section modulus is 2.32 in. From this relationship, the maxi =um stress generated in the beam is 2 165 pounds per square inch (1.14 MN/m ) which is f ar below the yield strength of the material. Revision 0 1 Aug 1980 2-2 e m, e-.

2.5.2 External Pressure The Model 920 is og.en to the atmosphere. Therefore, there will be no differential pressure to act upon the package. The collapsing pressure of the source capsule is calculated assuming that the capsules are thin wall tubing with the wall thickness equal to the minimum depth of weld penetration (0.020 inch; 0.5=m). The collapsing pressure is calculated from: P = 86,670 t - 1386 d where P: collapsing pressure in pounds per square inch c: Wall thickness (0.020 inch) d: Outside diameter (0.205 inch) (Re ference : Machinery's Handbook, 21st Edition, p. 440) From this relationship, the collapsing pressure is computed to be 2 7070 pounds per square inch (49MN/m ). Therefore, the source capsule can withstand an external pressure of 25psig without adverse affect. 2.6 Normal Conditions of Transport 2.6.1 Heat The thermal evaluation of the Model 920 is presented in Chapter 3. From this evaluation, it can be concluded that the Model 920 can withstand the normal heat transport condition. 2.6.2 Cold The metals used in the manuf acture of the Model 920 can all withstand a temperature of -40 C. The lower operating limit of the polyurethane 8 foan is -100 F'(73 C). Thus, it is concluded that the Model 920 will 8 8 withstand the normal transport cold condition. 2.6.3 Pressure The Model 920 is open to the atmosphere. Therefore, there will be no differential pressure to act upon it. In Section 3.6.4, it is demonstrated that the source capsule is able to withstand an external pressure reduction of 0.5 atmosphere 2 (50.7kN/m ), 2.6.4 Vibration A vibration test of the Model 900 was conducted. The package was Revision 0 1 Aug 1980 2-3 c

vibrated for seventy minutes with a maximum acceleration of 9.8m/s2 at each of the folicwing frequencies: 5, 8, 12, 20, 32 and 80Hz. At the conclusion of this test, the source assembly remained secured in the proper storage position. A copy of this test report is included in Section 2.10. The Model 920 is identical to the Model 900 with the exception of the shield assembly. Because of the satisfactory performance of the Model 900 during the vibration test and the similarity of the Model 920 to the Model 900, it is concluded that the Model 920 will withstand the normal transport vibration condition. 2.6.5 Water Spray Test The water spray test was not actually performed on the Model 920. The materials used in the construction of the package are all highly water resistant. Exposure to water will not affect the structural integrity nor reduce the shielding effectiveness of the package. 2.6.6 Free Drop The drop analysis presented in Section 2.7.1 demonstrates that the Model 920 will withstand the normal transport free drop condition without loss of shielding effectiveness nor loss of structural integrity. 2.6.7 Corner Drop 2.6.8 Penetration A penetration test of the Model 900 was performed. There was no reduction of shielding effectiveness nor loss of structural integrity as a result of this test. A copy of this test report is included in Section 2.10. Because of the satisf actory performance of the Model 900 during the penetration test and the similarity of the Model 920 to the Model 900, it is concluded that the Model 920 will withstand the penetra-tion test condition. 2.6.9 Compression 2 The maximum cross sectional area of the Model 920 is 100 in. The weight of the package is 47 pounds. Therefore, five times the weight of the package is greater than two pounds per square inch times the maximum cross sectional area. The load used is 232 pounds. Revision 0 1 Aug 1980 i 2-4 c

The Model 900 was subjected to the conditions of the conpression test. There was no reduction of shielding effectiveness nor loss of structural integrity as a result of this test. A copy of this test report is included in Section 2.10. Because of the satisfactory performance of the Model 900 during the compression test and the similarity of the Model 920 to the Model 900, it is concluded that the Model 920 will withstand the compression test condition. 2.7 Hypothetical Accident Conditions . 2.7.1 Free Drop The Model 900 was subjected to a drop test through a distance of 30 feet (9.1m) onto a steel plate. There was no loss of shielding effectiveness nor loss of atructural integrity as a result of this test. A copy of the test report is included in Section 2.10. Be-cause of the similarity of the Model 920 to the Model 900, it is concluded that the Model 920 will withstand the free drop test con-dition. 2.7.2 Puncture The Model 900 was subjected to a free drop from the height of one meter onto a steel billet which was six inches in diameter and eight inches long. As a result of this test, there was no loss of shielding effectiveness nor loss of package integrity. A copy of this test report is included in Section 2.10. Because of the similarity of the Model 920 to the Model 900, it is concluded that the Model 920 will withstand the puncture test condition. 2.7.3 Thermal The thermal analysis is presented in Section 3.5. It is shown that the melting temperatures of the materials used in the construction of the Model 920 except the golyurethane foam, and the aluminum handle, 0 are all in excess of 1475 F (800 C) To demonstrate that the radioactive source assemblies will remain in a shielded position following the hypothetical thermal accident, the following analysis is presented. At the conclusion of the thermal test, it is assumed that the polyurethane foam has completely escaped from the package. The shield assembly is prohibited from movement by the front housing, rear plate and lock assembly which are attached by the spacer rods. Thus, it is concluded that the Model 920 satisfactorily meets the requirements for the hypothetical thermal accident condition of 10CFR71. Revision 0 1 Aug 1980 2-5 e s -,- =

m. .,m. o I d i l 4 i 2.10 APPENDIX Descriptive Assembly Drawings - Source Capsule Test Report: Vibration Resistance Test Test Report: Penetration Test Test Report: Compression Test i Test Report: Free Fall Test Test Report: Puncture Test b t j i 1 1 l Revision 0 1 Aug 1980 2-7 --c. -....,,,. - -,, - -.,, = -..,. _,..,., ,,--..-,,n..- ..e..,- .a_,v,

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^ TEST REPORT RADIATION PRODUCTS DIVISION BY: George Parsons DATE: 30 January 1980

SUBJECT:

Model 900 Vibration Resistance Test A vibration resistance test of the Model 900 radiographic exposure device-Type B package was conducted by Associated Testing Laboratory, Burlington, MA, on 9 January 1980 in accordance with International Standard ISO 3999. The device was fastened to the platform of a vibrating machine and was continuously scanned through a frequency range of 5Ha to 80Ha with a 2 maximum acceleration of 9.8m/sec to search for a resonant frequency. This scan was performed on each of three rectilinear axis. Over this range, there was no resonant frequency found. The device was then vibrated along its longitudinal axis (the axis of the source tube) for 70 minutes at each of the following frequencies: 5, 8, 12, 20, 32, 80Hz. As a result of this test, there was no loss of structural integrity nor loss of shielding efficiency. There was no loosening of fasteners. The device functioned normally at the conclusion of the test. GP/fb IEVEICN O ADE. 1 134u 2 -13

-a TEST REPORT RADIATION PRODUCTS DIVISION BY: David Marzilli DATE: 14 January 1980

SUBJECT:

Model 900 Penetration Test On 14 January 1980 a penetration test was performed on a Model 900 Gamma Ray Projector. This projector already had been submitted to the shielding efficiencyy vibration resistance and shock resistance tests outlined in International Standard ISO 3999. A 13 pound steel bar, with a hemispherical end 1% inches in diameter, was dropped more than 40 inches onto the lock mechanism of the Model 900. A guide tube, 1 5/16 inches square along its inside dimensions, was used to insure positioning over the lock. The test was done twice. The lock mechanism functioned before and after the tests, and no other damage was noted. Thus, the Model 900 will satisfy the penetration test require =ents of 10CFR71. Witnessed: AM)i e ~ "" Aggelo Kiki'Is 4EVISIO O

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TEST REPORT RADIATION PRODUCTS DIVISION BY: David Marzi11i DATE: 14 January 1980

SUBJECT:

Model 900, 30 Ft. Free Fall Test On 11 January 1980, a Model 900 Gamma Ray Projector was submitted to a free fall of 30 feet, as outlined in 10CFR71 and IAEA Safety Series No. 6, 1973. The drop was done twice, onto a concrete driveway covered with a 7/16 inch steel plate. On the first drop the machine impacted on its front face, bounced and impacted the rear. The second drop caused the machine to impact on its rear lower left corner (as the machine is faced from the rear). The first drop caused the indicator cover plate screws (4) and the lower four front end plate screws to shear. The front housing remained secured by the screws mounting it to the spacer rod and the rear plate. The second drop caused deformation of the rear end plate and shell. As a result of these tests, there was no loss of shielding effectivem.'es nor f\\ loss of security of the source assembly. Thus, it is concluded that che Model 900 satisfies the requirements of the free fall test as described in 10CFR71, IAEA Safety Series No. 6, 1973 and ISO 3999. Witnessed: / ohn JVMunro III ,ffAA / Aptelo C. Kiklis REWSION O 56-I !?n)

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TEST REPORT RADIATION PRODUCTS DIVISION BY: David Marzilli DATE: 14 January 1980 ?UBJECT: Model 900 Puncture Test On 11 January 1980 a Model 900 Ga=ma Ray Projector was twice submitted to a free fall of 40 inches ento a six inch diameter steel billet. The projector had already been twice submitted to a 30 foot free fall (See Test Report dated 14 January 1980). The =ost vulnerable portion of the Model 900 was deemed to be the lock mechanism. The package was dropped 40 inches onto its lock, once in shear and once in cocpression. There was no structural or functional damage. Thus, we conclude that the Medel 900 meets the require =ents of the puncture test as described in 10CFR71 and IAEA Safety Series No. 6, 1973. Witnessed: A #.- ./hIs b d Mgelo KiAlis V Francis E./Ro ~ f

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3.0 Thermal Evaluation 3.1 Discussion The Model 920 is a completely passive thermal device and has no mechanical cooling system nor relief valves. All cooling of the package is through free convection and radiation. The heat source is 240 curies of iridium-192. The corresponding decay heat is 2.06 watts. 3.2 Summary of Thermal Properties of Materials The melting points of the materials used in the construction of the Model 920 are: Steel 2453'F (1345 C) 0 Uranium 2070'F (ll33*C) 0 0 Tungsten 6098 F (3370 C) 8 0 Bronze 1840 F (1005 C) Aluminum 1220*F ( 600 C) 0 8 The polyurethane foam has a mini =um operating range of -100 F 0 (-73 C) to 200 F (93'C). It will deco =pcse at the fire test temperature (800'C). Decomposition will cesult in gaseous by-products which will burn in air. 3.3 Technical Specification of Ccmponents Not Applicable 3.4 Normal Conditions of Transport 3.4.1 Thermal Model The heat source in the Model 920 is a maximum of 240 curies of iridium-192. Iridium-192 decays with a total energy liberation of 1.45 MeV per dis-intergration or 8.58 milliwatts per curie. Assuming that all of the decay energy is transformed into heat, the heat generation rate for the 240 curies of iridium-192 would be 2.06 watts. To demonstrate compliance with the requirements of paragraphs 231 and 232 of IAEA Safety Series No. 6,1973 Edition for Type B(U) packaging, an analysis is presented in Section 3.6.1. The thermal model e= ployed is described in that section. To demonstrate compliance with the requirements of paragraph 240 of IAEA Safety Series No. 6,1973 Edition for Type B(U) packaging, an analysis is presented in Section 3.6.2. The thermal model employed is described in-that section. Revision 0 1 Aug 1980 3-1

1 3.4.2 Maximum Temperatures The maximum temperatures encountered under normal conditions of transport will have no adverse effect on the structural integrity or shielding. As presented in Section 3.6, the maximum temperature 8 in the shade would be less than 42 C and the maxi =um temperature 0 when insolated would be less than 62 C. 3.4.3 Mini =um Temperatures 0 The minimum normal operating temperature of the Model 920 is -40 C (-40 F). This temperature will have no adverse affect on the 0 package. 3.4.4 Maximum Internal Pressures Normal operating conditions generate negligible internal pressures. Any pressure generated is significantly below that generated during the hypothetical thermal accident, which is shown to result in no loss of shielding nor containment. 3.4.5 Maximum Thermal Stresses The maxi =um temperatures that occur during normal transport are low enough to insure that thermal gradients will cause no significant thermal stresses. 3.4.6 Evaluation of Package Performance for Normal Conditions of Transport The thermal conditions of normal transport are insignificant from a functional viewpoint for the Model 920. The applicable conditions of IAEA Safety Series No. 6,1973 Edition for Type B(U) packages have been shown to be satisfied by the Model 920. 3.5 Hypothetical Accident Thermal Evaluation 3.5.1 Thermal Model The Model 920, including the source assembly, is assumed to reach 0 the thermal test temperature of 800 C. At this temperature the polyurethane foam will have decomposed and the resulting gases will have escaped the package through vent holes and non-leak tight assembly joints. 3.5.2 Package Conditions and Environment The Model 920 underwent no significant damage during the free drop and puncture tests. The package used in this analysis is considered undamaged. Revision 0 1 Aug 1980 i 3-2 i

3.5.3 Package Temperatures As indicated in Section 3.5.1, the entire package is assumed to 8 reach a temperature of 800 C. Examination of the melting tempera-tures of the materials used in the construction of the Model 920 indicates that there will be no da= age to the package as a result of this temperature. The possibility of the for=ation of an iron-uranium eutectic alloy was addresaed in Section 2.4.1 where it was concluded that the formation of the alloy was not a likely eventuality. 3.5.4 Maximum Internal Pressures The Model 920 packs.ging is open to the atmosphere. Therefore, there will be no pressure buildup within the package. In Section 3.6, an analysis of the source capsules under the thermal test condition 8 demonstrates that the maximum internal gas pressure at 800 C is 54 psi 2 (373kN/m ). In Section 3.6.3, an analysis is presented which demonstrates that the maximum stress generated in the source capsule (containment) under the thermal test conditions could only be 3% of the yield strength of the material at the test te=perature. ( 3.5.5 Maxi =um Thermal Stresses There are no significant thermal stresses generated during the ther=al test. 3.5.6 Evaluation of Package Performance The Model 920 will undergo no loss of structural integrity or shielding when subjected to the thermal accident condition. The pressures and temperatures have been demonstrated to be within acceptable limits. Revision 0 1 Aug 1980 3-3 E'

1 o 3.6 APPENDIX 3.6.1 Model 920 Type B(U) Thermal Analysis: Paragraphs 231 and 232 of IAEA Safety Series No. 6,1973 Edition 3.6.2 Model 920 Type B(U) Thermal Analysis: Paragraph 240 of IAEA Safety Series No. 6, 1973 Edition 3.6.3 Model 920 Type B(U) Source Capeule Thermal Analysis: Paragraph 238 of IAEA Safety Series No. 6,1973 Edition i i Revision 0 1 Aug 1980 i~ 3-4 3 -- - - - -,. ~. .-n. c, ~-

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. i I 3.6.1 Model 920 Type B(U) Thermal Analysis Paragraphs 231 and 232 of IAEA Safety Series No. 6, 1973 Edition This analysis demonstrates that the maximum surface temperature 8 8 of the Model 920 will not exceed 50 C (122 F) with the package in 8 0 the shade and at an ambient temperature of 38 C (100 F). To assure conservatism, the following assu=ptions are used: a. The entire decay heat (2.06 watts) is deposited in the exterior faces of the Model 920. b. The interior of the Model 920 is perfectly insulated and heat transfer occurs only from the exterior wall to the atmosphere. c. Because each face of the package eclipses a different solid angle, it is assumed that twenty five percent of the total heat is deposited in the smallest face (front), d. The only heat transfer mechanism is free convection. Using these assumptions, the maximum wall temperature is found from: q = hA(T -T) w a and h = 1.42 "T - T," y 1 where q: Heat deposited per unit time in the face of interest (0.52 watts) h: Free convective heat transfer coefficient 2 o for air in watts /n c 2 A: Area of the face of interest (0.026m ) T,: Maximum temperature of the wall of the package 8 '~ T,: Ambient temperature (38 c) Height of the face of interest (0.20m) 1 : 0 From this relationship, the =aximum temperature of the wall is 44.0 C (111 F). This satisfies the requirement of paragraphs 231 and 232 of 8 IAEA Safety Series No. 6, 1973 Edition. Revision 0 1 Aug 1980 3-5 7 -m,_

e 3.6.2 Model 900 Type B(U) Thermal Analysis Paragraph 240 of IAEA Safety Series No. 6, 1973 Edition This analysis demonstrates that the maximum surface temperatures 0 8 of the Model 920 will not exceed 82 C (180 F) when the package is 0 0 in an ambient temperature of 38 C (100 F) and insolated in accordance with paragraph 240 of IAEA Safety Series No. 6, 1973 Edition. The calculational model consists of taking a steady state heat balance over the surface of the package. The following assumptions are used. 2 The package is insolated at the rate of 775 /m a. 2 2 (800 cal /cm -12h) on the top surface, 194 w/m 2 (200 cal /cm -12h) on the sides and no insolation on the bottom. b. The decay heat load is added to the insolation heat load. c. The solar absorptivity is assu=ed to be 0.9. The solar emissivity is assumed to be 0.8. d. The package is assumed to undergo free convection from the top and sides and undergo radiation from the top, sides and bottom. The inside faces are considered \\ insulated so there is no conduction into the package. The f aces are considered to be sufficiently thin that no temperature gradients exist. The package is approximated as a rectangular solid of e. 0.33m length, 0.20m high and 0.14m wide. The maximum surface temperature is established from a steady state heat balance relationship. 3 +9d"9 +9 4i e r 1 where a : Absorptivity (0.9) q: Solar Heat Load (72.12 watts) g q: Decay Heat Load (206 watts) d q: Convective Heat Transfer q: Radiative Heat Transfer Revision 0 1 Aug 1980 3-6 E n 7,,- y

1 The convective heat transfer is: c _(hA) top + (hA) sides, (T -T) q = w a where h: Convective Heat Transfer Coefficient A: Area of the Surfaca of Interest T,: Temperature of the Wall T,: Ambient Temperature The heat transfer due to radiation is: g = acA (T * - T,") _ o g' ') where c : Stephan Boltzman Constant (5.67 x 10 s yfm2 c: Emissivity (0.8) Iteration of this relationship demonstrates that the wall temperature 0 is 62.0 C which satisfies the requirement of paragraph 240 of IAEA Safety Series No. 6,1973 Edition. .~ i i Revisicu O 1 Aug 1980 3-7 2 g-- e---r.,

1 3.6.3 Model 920 Type B(U) Source Capsule Thermal Analysis, Paragraph 238 of IAEA Safety Series No. 6, 1973 Edition This analysis demonstrates that the pressure inside the Model 90004 or Model 90005 source capsule, when subjected to the ther=al test, does not exceed the pressure which corresponds to the minimum yield strength of the material at the thermal test temperature. The source capsules are fabricated from stainless steel, Type 304 or 304L. The outside diameter of each capsule is 0.205 inch (5.2mm). The source capsule is seal welded. The minimum weld penetration is 0.020 inch (0.5mm). Under the conditions of internal pressure, the critical locr. tion for failure is this weld. The internal volume of the source capsule contains only iridium metal as a solid and air. It is assumed that the air is at standard tempera-0 2 ture and pressure (20 C; 100kN/m ) at the time of loading. This is a conservative assumption because, during the velding process, the internal air is heated, causing some of the air mass to escape before the capsule is sealed. When the welded capsule returns to ambient temperature, the internal pressure would be somewhat reduced. Under the conditions of paragraph 238 of IAEA Safety Series No. 6, it is assumed that the capsule could reach a temperature of 800'C ( (14 75' F). Using the ideal gas law and requiring the air to occupy a 2 constant volume, the internal gas pressure could reach 373kN/m (54 psi). The capsule is assumed to be a thin walled cylindrical pressure vessel. The maxi =um longitudinal stress is calculated from: oat 7 = PAp where a: Longitudinal Stress t 2 2 g) A: Stress Area = n(r -r 2 P: Pressure (373kN/m ) A: Pressure Area = ur.2 P 1 From this relationship, the maxi =um longitudinal stress is calculated 2 to be 686kN/m (99p,t), i l Favision 0 1 Aug 1980 3-8

1 The hoop stress can be found by: 2a it = Pld; h or c = Prg h t where c : hoop stress h 1: length of the cylinder t : thickness of the cylinder 2 From this relationship, the hoop stress is calculated to be 1.54MN/m (223 psi). At a temperature of 870'C (1600 F), the yield strength of Type 304 0 2 stainless steel is 69EN/m (10,000 psi). Thus, under the conditions of paragraph 238 of IAEA Safety Series No. 6, 1973, the stress generated is less than 3% of the yield strength of the material. ( Revision 0 1 Aug 1980 3-9 E' w

4. Containment 4.1 Containment Boundary 4.1.1 Containment Vessel The containment system for the Model 920 is the radioactive source capsule as described in Section 2.10. The source capsule is f abricated from either Type 304 or Type 304L stainless steel. The capsule is cylindrical in shape with a diameter of 0.205 inch (5.2mm) and a length of 0.650 inch (16.5mm). 4.1.2 Containment Penetrations There are no penetrations of the containment. 4.1.3 seals and Welds The containment is seal welded by a tungsten inert gas welding process which is described in Tech / Ops Standard Source Encapsulation Procedure (Section 7.4). The minimum weld penetration is 0.020 inch (0.51mm). 4.1.4 Closure ( Not Applicable 4.2 Requirements for Normal Conditions of Transport 4.2.1 Release of Radioactive Material This source capsule has satisfied the requirements for special form radioactive material as delineated in LAEA Safety Series No. 6,1973 Edition and 10CFR71. Therefore, there will be no release of radioactive material under the normal conditions of transport. 4.2.2 Pressurization of the Containment Vessel Pressurization of the source capsule under the conditions of the hypothetical thermal accident was demonstrated to generate stresses well below the structural limits of the capsule (Sections 3.5.4, 3.6.2). Thus, the containment will withstand the pressure variations of normal transport. 4.2.3 Coolant Contamination Not Applicable Revision O 1 Aug 1980 4-1 t c-

1 4.2.4 Coolant Loss Not Applicable 4.3 Containment Requirements for the Hypothetical Accil nt Conditions 4.3.1 Fission Gas Products 4.3.2 Release of Contents The hypothetical accident conditions of 10CFR71, Appendix B will result in no loss of package containment as described in Sections 2.7.1, 2.7.2 and 3.5. ( Revision 0 1 Aug 1980 4-2 i

I 5. Shielding Evaluation 5.1 Discussion and Results The Model 920 contains 31 pounds (14 kilograms) of depleted uranium shielding. The uranium shielding is cast around the tungsten source tube. The shielding evaluation is based upon actual radiation profile measurements of a Model 900 container (Certificate of Compliance No. 9141) and experimentally generated transmission data for iridiumist gamma rays in uranium. The maximum dose rates for a 182 Model 920 containing 240 curies of iridium are presented in Table 5.1. As the Model 920 contains no neutron source, the gamma dose rates are the total dose rates which are presented. As shown la Table 5.1, the =aximum dose rates associated with the Model 920 are below the regulatory requirements. Table 5.1 Summary of Maximum Dose Rates (mR/hr) At Surface At One Meter I Side Top Bottom Side Top Bottom 185 185 150 1.1 1.1 0.9 5.2 Source Specification 5.2.1 Gamma Source is2 The gamma source is iridium in a sealed capsule as special form in quantities up to 240 curies. 5.2.2 Neutron Source, Not Applicable 5.3 Model Specification Not Applicable Revision 0 1 Aug 1980 5-1

i 5.4 Shielding Evaluation The shielding evaluation of the Model 920 is based upon actual radiation survey results of a Model 900 container and experi-mentally developed transmission data for iridium ts2 gamma rays in uranium. Section 5.5.1 presents the results of a radiation profile of 182 Model 900, Serial Number P1 containing 96 curies of iridium Section 5.5.2 presents the results of this survey extrapolated to a capacity of 240 curies of iridium 1s2, From these extrapolated data, the additional shielding requirements for the Model 920 were calculated using experimentally developed transmission data. Drawing 92090, Sheet 3 (page 1-6) presents the Model 920 shield assembly and compares it to the Model 900 shield assembly. Section 5.5.3 presenir, the results of the calculation of radiation intensities for the Model 920. I, Revision 0 1 Aug 1980 5-2 I

1' l i i 5.5 APPENDIX 5.5.1 Radiation Profile - Model 900 Serial No. P1 containing 1 96 curies of iridium 1:2 r 5.5.2 dadiation Profile - Model 900 Serial No. P1 extrapolated 3 to a capacity of 240 curies of iridium tsz 5.5.3 Calculated Radiation Profile of Model 920 containing 240 curies of iridium 1s2 l 4 t d 4 4 s Revision 0 1 Aug 1980 5-3 I 1

50 50 60 70 70 70 60 60 50 80 80 90 f00 fl0 90 70 70 70 70 80 90 l00 11 0 90 70 80 5 50 60 60 70 80 70 60 50 50 TOP 70 60 $0 60 60 70 90 90 70 60 60 85 85 50 60 60 80 90 80 70 do' 60 go ao go 10 60 70 90 12 0 l00 90 70,70 70 60 to 80 100 120 100 80 70 70 ao 60 70 90 11 0 150 11 0 90 80 70 60 60 70 90 110 f3D 10 0 70 70 l 70 h 50 90 l 50 40 %W 90 60 70 00 11 0 130 foo ao 70 70 50 60 70 90 14 0 (20 90 70 70 so 60 60 80 90 100 90 70 70 60 70 70 60 to 80 12 0 100 60 80 60 0 70j 0 80) 40 I 60 I FRONT LEFT REAR RIGNT 50 90 50 50 50 50 50 g 50 50 50 60 60 60 60 50 50 50 5.5.1 ~x Radiation Profile yo 50 60 GO 60 60 50 50 50 F, Model 900 S.N. P1 50 50 50 50, yo 50 m yo go -o Containing 96Ci of Iridium-192 8o BOTTOM

125 12 5 150 11 5 175 175 15 0 150 125 200 200 225 250 175 N5 175 IT5 115 175 200 225 250 275 225 175 200 15 12 5 150 150 17 5 200 175 s50 125 125 g TOP ~ ~ b 'II 215 215 125 150 150 ')00 225 200 115 150 15 0 soo soo 200 125 12 15 0 175 225 225 175 15 0 150 10 0 125 150 175 22 5 500 250 225 175 175 175 150 150 200 250 }do 250 Joo 175 175 foo u 150 150 175 225 525 500 250 175975 3 its hf25 225 l 150 175 225 275 325 275 225 2c0 175 h 22r 100 J250 150 175 225 175 325 2W 200 175 175 125 15 0 177 225 350 foo 225 ITF 175 t 150 150 150 200 225 250 225 115 175 150 175 125 150 150 200 NO 250 200 200 15 0 00 2001 11 5 175] 150 I 100 } FR0rlT LEFT REAR RIGHT 12 5 125 125 125 125 125 12 5 125 12 5 5.5.2 y, $ d. 125 115 150 150 150 150 125 12 5 125 Extrapolation of oo en Radiation Profile Data g o' 125 12 5 15 0 150 150 15 0 125 125i 12 5 M del 900 S.N. PL 0 12 5 125 125 125 12 5 12 5 125 i[5 125 l To A Capacity of 240Ci of Iridium-192 BOTTOM

O 12 5 12 5 125 90 75 90 125 ISO 12 5 150 17F 18 5 150 12 0 120 150 175 175 150 lif 18 5 IJO 120 12 0 150 175 l75 125 175 12 5 90 90 90 125 12 5 05 TOP l50 179 17 5 125 15 0 05 11 0 100 110 145 150 150 175 100 175 125 150 125 800 100 120 Hs 150 150 Of 150 15 0 00 110 11 0 18 5 17 5 175 115 150 150 165 1% 150 130 16 5 17 5 175 goo 150 150 150 120 14 0 16 0 185 17 5 17 5 l 115 125 hits 11 5 l 150 11 5 les 145 W MS 185 175 175 17 5 i 10 0 B75 9 150 175 185 14 5 14 0 15 0 16 5 11 5 17 5 11 5 150 150 120 15 0 160 165 lif 17 5 150 150 150 165 120 110 120 I45 115 150 175 125 150 125 llo 13 0 I)0 IM 115 150 175 175/ IIT5 8751 I 100 / \\ 158 ,/ FRONT LEFT REAR RIGNT - :o l 125 12 5 12 5 125 925 125 125 125 US 5.5.3 c-"{. 12 5 12 7 150 150 e50 150 05 125 as calculated Radiation Profile H del 920 ho 12 5 125 Ifo 150 150 150 12 5 05 containing 240 curiea or i25 i25 ny us las u5 a5 as of Iridium-192 BOTTOM i

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I 6. Criticality Evaluation Not Applicable ( / . = = Revision 0 1 Aug 1980 6-1 3

7. Operating Procedures 7.1 Procedures for Loading the Package The procedure for fabricating the special form source capsule is presented in Section 7.4. The procedure for loading the source assembly into the package is presented in Section 7.4. 7.2 Procedures for Unioading the Package The procedure for unloading the package is presented in Section 7.4. 7.3 Preparation of an Empty Package for Transport The procedure for preparation of an empty package for transport is presented in Section 7.4. ( l Revision 0 1 Aug 1980 7-1 8 r4

~ RADIATION SAFETY MANUAL Part II In Plant Operations I Section 2 ENCAPSULATION OF. SEALED SOURCES A. Personnel Requirements Only an individual qualified as a Senior Radiological Technician shall perform the operations asscciated with the encapsulation of 192 Iridium. There must be a second qualified Radiological Technician available in the building when these operations are being perfor:::ed. B. General Requirements The 192 Iridium loading cell shall be used for the encapsulation of s such solid metallic 192IridiumandthepackagingofsealedsouregO 169 Ytterbium. Solid metallic Cobalt as1 Thulium,137esiumand C not exceeding one curie may be handled in this cell also. The maximum amount of 192 Iridium to be handled in this cell at an' one ti:ne shall not exceed 1000 curies. The =aximum amount of 13 s 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 atmospheric pressure. The exhaust blewer provided shall not be turned off except when the cell is in a decontaminated condition. Sources shall not be stored in this cell overnight or when cell is unattended. Unencapsulated material shall be returned to the transfer containers and encapsulated sources transferred to approved source containers. When any of the "through-the-vall" tools such as the velding fixture or transfer pigs are re=oved, the openings are to be closed with the plugs provided. These tools shall be decontaminated whenever they are removed from the hot cell. ~ C. Preceratory Procedure 1. Check welding fixture, capsule draver and manipulator fingers from cell and survey for contamination. If contamination in excess of 0.001 jrCi of removable contamination is found, these ite=s must be decontaminated. 2. If the velding fixture or the electrodes have been changed, perform the encapsulation procedure omitting the insertion of any activity. Examine this dur::::y capsule by sectioning thru veld.' Weld penetration =ust be not less than 0.020 inch. ,-2 REVISION O II.2.1 Aug. 1 my 2

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 prepara*;ory procedure repeated. 3 Check pressure differential across first absolute filter, as measured by the manometer on the left side of the het cell. Thisisabout}inchofwaterforanewfilter. When this pressure differential rises to about 2 inches of water, the filter must be changed. D. Encapsulation Procedure 1. Prior to use, asse=ble and visually inspect the two capsule components to determine if veld zone exhibits any misalignment and/orseparation. Defective capsules shall be rejected. 2. Degrease capsule components in the Ultrasonic Bath, using isopropyl alcohol as degreasing agent, for a period of 10 0 minutes. Dry the capsule components at 100 C for a mini =um of twenty minutes. 3 Indert capsule components into hot cell with the posting bar. 4. Place capsule in veld positioning device. i 5 Move drawer of source transfer container into hot cell. ( 6. Place proper amount of activity in capsule. Disposeble funnel must be used with pellets and a brass rivet with vafers to prevent contamination of veld zone. 7 Remove unused radioactive material from the hot cell by with-drawing the drawer of the source transfer container from the cell. 8. Remove funnel or rivet. 9 Assemble capsule components.

10. Weld adhering to the following conditions:

a. Electrode spacing.021" to.02k" centered on joint +.002"; use jig for this purpose. b. Preflow argon, flush 10 seconds. c. Star, 15 amps. d. Weld 15 amps, e. Slope 15 amps. f. Post flov 15 seconds EEVI3 ION O 7-3 AUG' 1 :95~' II.2.2 .-.., -.. ~. -, ~... - w

11. Visually inspect the vold. An acceptable vald =ust be continuous without cratoring, cracks or evidenen of blev cut.

If the veld is defective, the capsule =ust be cleaned and revelded to acceptable conditions or disposed of as radioactive vaste. 12. Check the capsule in heiSht gauge to be sure that the veld is at the center of the capsule. 13 Wipe exterior of capsule with flannel patch vetted with EDPA solution or equivalent.

14. Count the patch with the scaler counting system. Patch =ust show no more than.005p ci of contamination. If the patch shows more than.005 pCi the capsule =ust be cleaned and reviped.

If the revipe patcb still shows more than 0.005 pCi 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 containirq isopropyl alechol. Apply a vacuum of 15 in Hg(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 vacuum vial and boil off the residual alechol. Reveld the capsule.

16. Transfer the capsule to the svaging fixture. Insert the vire and connector assembly and svage. Hydraulic pressure should not be less than 1250 nor more than 1500 pounds.
17.. Apply the tensile test to asse=b1;y between the capsule and

. connector by applying proof load of 75 lbs. Extensien under (' the load shall not exceed 0.1 inch. If the extension exceeds 0.1 inch, the scurce =ust be disposed of as radioactive vaste.

18. Position the source in the exit port of hot cell. Withdraw all perscnnel to the control area. Use re=ote centrol to insert source in the ion chamber and position the scurce for caximum response. Record the meter reading. Compute the activity in curies and fill cut a temporary source tag.

19 Using re=ote centrol, eject the source frem een into source changer through the tube gauze wipe test fixture. Monitor before reentering the hot cell areato be sure that the source is in the source changer. Remove the tube gauze and count with scaler counting system. This assay must show no more than 0.005 pC1. If contamination is in excess of this level, the source is leaking and shall be rejected. 20. Ccaplete a Source Leading Leg (Figure II.2.1) for the operatien. II.2 3 PIVISION O 7 AUG. 11 53 ,n

i Tech / Ops Model 900 and Model 920 Garma Ray Projector Operation Manual Technical Data Size: 13 in. long, 7.7 in. high 5.3 in. wide (330mm long,195mm high,135mm wide) Model 900 Model 920 Weight: 44 lbs (20kg) 47 lbs (21kg) Shielding (Uranium) 28 lbs (13kg) 31 lbs (14kg) Capacity (ir-192) 100Ci 200Ci Transport Status Type B USA /9141/B USA / /B General The Models 900 and 920 are designed for use as radiographic exposure devices, storage containers and transport packages for Tech / Ops Model 90003 source assembly. The US Nuclear Regulatory Commission allows the use of these devices only by ( persons who are specifically authorized under the terms of their license. Application for a license to use this device should be made to: Radioisotope Licensing Branch Division of Fuel Cycle and Material Safety US Nuclear Regulatory Commission Washington, DC 20555 Frior to the first shipment of these devices, the user, in addition, should register with: Transportation Certification Branch Division of Fuel Cycle and Material Safety US Nuclear Regulatory Com mission Washington, DC 20555 Revision 0 1 Aug 1980 7-5

1 Tcch / Ops .n f . a Radiation Products Omson 40 North Avenue Burkngton. Massachusetts 01803 l Telephone (617) 272 2000 Receipt 1. Upon receipt of the projector system, survey the projector on all sides to ensure that radiation levels do not exceed the following: At Surface 200mR/hr At 6 Inches from Surface 50mR/hr At 3 Feet from Surface 10mR/hr I 2. Check the projector, control unit and guide tubes for obvious damage. 3. Check packing list and Bill of Lading to ensure that all are intact and are representative of the shipment. 4. Place the projector in a restricted area until ready to use. Operation - NOTE - t Personnel using this projector =ust have a calibrated and operable survey meter with a range of at least 0 to 1000mR/hr. In addition, personnel =onitoring devices =ust be worn during these operations. They are direct reading pocket dosimeter and either a film badge or a thermoluminescent dosimeter. Radiographic operations must be conducted in a restricted area and the area must be posted as required in 10CFR20. 1. Survey the projector on all sides and ensure that the radiation levels do not exceed 50 milliroentgens per hour at six inches from the surface. 2. Locate the projector and controls to afford the cperator as much shielding as possible. . REVISION 6 AUS. 1 Inu ~.. - -. 7 ..-_9 _s ymy-w g,-_,-.-pg,. y 9 -,y v qq .-9 4 ,~g yyy gws-,, w y--

1 Tcch / Ops ,e f Aacate P,oducts Omson 40 Nym Averme Burbrgton. Massachusetts 01803 Tewphone (617) 272 2000 3. At the radiographic focal point, position and secure the source stop of the guide tube assembly. 4. Connect together as many guide tube assemblies as necessary (maximum of three). Lay out the guide tube assembly as straight as possible (bend radius must be greater than 20 inches). 5. Position the projector at the end of the guide tube assembly. 6. From the projector, lay out the control cables as straight as possible (bend radius must be greater than three feet). 7. Unlock the projector with the key and rotate the selector ring v$pD' from the LOCK position to the CONNECT position. The storage cover s' will disengage from the prcjector. s 8. agage the male connector of the driving cable to the female connector of the source assembly. 9. Slide the control cable connector forward into the locking 1 assembly of the projector. Rotate the selector ring from the 'A CONNECT position to the LOCK position. Depress the lock and remove the key.

10. Attach the guide tube assembly to the exit port of the projector.

fl. Thoroughly check all cable connections, bend radii and the position g of the source stop. V M2.* INSURE THAT NO ONE IS WITHIN THE BOUNDARY OF THE RESTRICTED AREA. l' s f 13. With a survey meter, approach the projector. Unlock the projector with the key and turn the selector ring from the LOCK position to the OPERATE position. 14. Fully raise the source position indicator knob. The knob will stay in the raised position. - NOTE - If cranking beco=es difficult at any time, return the source immediately to the stored position. REVISION O 7-7 AUi 1 taca

1 e . i TGeh /Opa ,2-Radation Products Omsco 40 North Avenue Burtington. Massa:nusetts 01803 Teephone (617) 272-2tOO

15. At the control unit, rapidly rotate the hand crank in the EXPOSE direction. Continue to rotate the hand crank until the source assembly reaches the source stop. The odometer should read the approximate distance of source travel.
16. At the end of the exposure time, rapidly rotate the hand crank in the RETRACT direction. Continue to rotate the hand crank until the source assembly reaches the storage position in the projector, which serves as a mechanical stop for the source asse=bly. The source position indicator knob should drop to the closed position and the odometer should read approximately 000.
17. Approach the projector with a survey meter. Survey the projector on all sides, survey the guide tube and survey the source stop to assure that the source assembly is in its proper stored position.

The radiation level should not exceed 50 milliroentgens per hour at six inches from the projector. 18. Rotate the selector ring to the LOCK position. Lock the projector.

19. Disconnect the guide tube assembly from the projector.

20. Unlock the projector. Rotate the selector ring from the LOCK position to the CONNECT position. Slide th'c control cable connector back and disconnect the drive cable frcs the source assembly. 21. Install the storage cover into the locking assembly and rotate the selector ring to the LOCK-position. Depress the lock and remove the key. Daily Inspection Daily inspection of radiography equipment is reco= mended to assure that the equipment is in proper working condition. Daily inspection should be performed prior to the start of each shift. 1. Inspect the entire length of each guide tube section and insure that each section is free from dents. Inspect the end fittings to insure that they are tightly attached to the guide tube section. Inspect the threads to insure that they are not galled or damaged. Do not use damaged guide tubes. REVISIO"? O JT na, IGM c -~

= 2. Inspect the entire length of control cable housings and insure that each section is free from dents. Inspect the end fittings to insure that they are tightly attached to the control cable housings. Check the control cable connector for damage. Check the male source connector for damage. Do not use a damaged control unit. 3. During the first radiographic operation, note the operation of the ~ selector ring and lock assembly. If operation is difficult, do not operate the equipment. 4. During the first radiographic operation, note the operation of the control crank. If operation is difficult, retract the source to the stored position, and survey the equipment in accordance with the operating instructions. Periodic Inspection and Maintenance Periodic inspection and maintenance of radiography equipment is recommended to assure that the equipment remains in proper working condition. Periodic inspection and maintenance should be performed at intervals not to exceed three months. Projector 1. Remove the source from the projector and install it in a source changer following the projector and source changer operating instructions. 2. Remove the rear plate from the projector. Disassemble the lock asse=bly. 3. Clean the components of the lock assembly. Examine the com-ponents for damage, excessive wear, galling and burrs. Lightly lubricate the lock slide and locking pin with grease MIL-G-23827 or equivalent. 4. Reassemble the locking assembly and test for proper operation. 5. Remove the source position indicator rod. Remove the front end plate and the indicator cover plate. Remove the indicator slide, locking pin, and slide tube. 6. Clean these components. Examine the components for damage, excessive wear, galling and burrs. Lightly lubricate the in-dicator slide and locking pin with grease, MIL-G-23827 or equivalent. 7. Clean the source tube. REVISICS o Kuk 1 tges c m m

8. Reassemble the slide tube, locking pin, indicator slide, indicator cover plate and front end plate. Reinstall the source position indicator rod. Reinstall the rear end plate. Test the projector for proper operation. Control Unit 1. Crank the drive cable in the expose direction until the ~ stop spring reaches the crank gear. Disassemble the control housing from the crank assembly. Remove the stop spring. Remove the drive cable from the control unit. 2. Clean the drive cable. Examine the drive cable for damage or excessive wear. Test the male connector with Tech / Ops no-go gage. Check the male connector for proper connection to the driving cable. 3. Remove the control housings from the control unit. Examine carefully for internal damage by flexing the housings by hand. Internal damage to the reinforcing braid or flexible metallic tube will be evidenced by a crunch feeling when the cable housing is flexed. Cut, flattened or burnt cable housings should be replaced. Superficial cuts or burns may be sealed and reinforced with tape. Clean housings by syringing a few ounces of solvent into bore, and blow out with low pressure air (not more than 20 psi). Do not allow solvent to remain in housings. Do not soak in solvent. Check end fittings for secure attachment. 4. Disassemble the crank unit. Wash parts in solvent. Check inside of housing for evidence of galling and wear. A deeply scored (more than 020 deep) line where the cable contacts the inner wall of the housing indicates the need for replacement. Check clearance between the hubs of the wheel and the bushings. More than.005 clearance indicates need for replacement. Examine teeth of wheel for damage. A bent tooth may be filed off. Two or more adjacent bent teeth will require replacement of the wheel. Lubricate the gear with grease and MIL-G-23837 or equivalent. 5. Reinstall the control housings to the crank unit. Lubricate the driving cable with grease MIL-G-23827 or equivalent. Reinstall the driving cable, installing the stop spring. Source Guide Tubes 1. Check for cuts, burns or crushed tubes. Check fittings for secure attachment. Examine and test screw threads for function. EEVISIC O lDE.

  • Y 3 -10

-.4

l l Clean bore of tube with water or solvent and drain out promptly. Do not soak in solvent. Check for free passage of source by holding tube vertical and dropping du=my source assembly through tube. The dummy source assembly should fall through freely. Final Inspection 1. Check the system for proper reassembly. Check fittings for tightness. 2. Reinstall the source into the projector following the projector and source changer operating instructions. Check for proper operation of the control unit, source position indicator and locking assembly.

3. Survey the projector on all sides to assure that the radiation levels do not exceed the following:

At surface: 200 mR/hr At 6 inches: 50 mR/hr At 3 feet 10mR/hr

4. Assure that the projector is properly labeled.

Leak Testing The source assembly in the Models 900 and 920 must be leak tested at intervals not to exceed six months. This can be accomplished using Tech / Ops Model 518 Leak Test Kit.

1. Place the projector in a restricted area.
2. Moisten the wipe test patch with EDTA solution.
3. Wipe the exit port of the projector and the female connector assembly.
4. Place the wipe test patch in the plastic envelope.
5. Set the survey meter on its most sensitive range and place the meter in a low background area. Move the patch to the meter, not the meter to the patch.
6. If the meter indication is less than 0.2mR/hr above background, place the plastic envelope into the mailing box and mail to Technical Opera-tions. BE SURE TO COMPLETE AND RETURN THE IDENTIFICATION SHEET.
7. If the meter indication is more than 0.2mR/hr, DO NOT MAIL. Contact Technical Operations for special instructions.

Revision 0 1 Aug 1980 7 - 11 4 -

1 The wipe test swab will be subjected to a precise radioassay when received by Tech / Ops and a leak test certificate will be mailed promptly. This certificate must be kept with your records. It is subject to NRC inspection. Transportation 1. Assure that the source assembly is in the proper storage position in the projector following the operating instruc-tions. Be sure that the storage cover is installed. 2. Safety lock wire the source position indicator knob and crimp the lead seal. 3. Survey the projector on all sides at the surface and at one meter and determine the proper shipping label to be applied in accordance with Table I. TABLE I Su rface 3 Feet RADIOACTIVE-WHITE.I.. / s ,' b, N s g /( )) 0.SmR/hr None N RA010 ACTIVE s..: / \\ / % i\\ / RADIOACTIVE-YELLOW II '^N N

o. 'N O

N 50mR/ h r 1.0mR/hr sQsA010ACllVEW > / N ~::.? l / s- -/ / N rN/ RADIOACTIVE-YELLON III /s \\ N N 200mR/hr 10mR/hr ( s A0l0 ACTIVE tj,2 s 'n -. / s =5 ),' 1 s v REWSiC.! O M '.

pg 4. Fill out information requested on label indicating: a. Contents (Isotope) b. No. of Curies c. Transport Index The Transport Index is determined by observing the maximum reading at 1 meter from the source container. This reading beco=es the Transport Index., 5. Remove all old shipping labels. - NOTE - Do not remove metal container identification label. 6. Affix new shipping labels to two opposite sides. 7. Properly complete the shipping papers indicating: Proper shipping name (i.e. Radioactive Material, Special Form, n.o.s.) Name of Radionuclide (i.e. Iridium) Physical or chemical form (or Special Form) Activity of source (expressed in curies or millicuries) Category of label applied (i.e., Radioactive Yellow III) Transport Index USNRC Identification Number For export shipments, IAEA Identification Number Shipper's Certification: "This is to certify that the above named materials are properly classified, j described, packaged, marked and labeled and are in proper condition for transport according to the applicable regulations of the Department of i Transportation." l* t 7 - e "5 REVI5ICM e l AU E.

3sC T

i ~ ^* Notes: 1. For air shipments, the following shipper's certification may be used: "I hereby certify that the contents of this consignment are fully and accurately described above by proper shipping name and are classified, packed, marked and labeled and are in proper condition for carriage by air accord-ing to applicable national governmental regulations." 2. For air shipments, the package must be labeled with a "CARCO AIRCRAFT ONLY" label and the shipping papers must state: "THIS SHIPMENT IS IJITHIN THE LIMITATIONS PRESCRIBED FOR CARCO-ONLY AIRCRAr*T." preparation of an Empty Package for Transport 1. To prepare an empty package for transport, follow the in-structions of the procedure above beginning with Step 2 with the following exceptions: a. The package must be marked " Radioactive Material - LSA-n o.s. b. The proper shipping name is Radioactive Material - LSA-n.o.s. c. Radionuclide is depleted uranium. 9 7 -14 g3 vg--g9 , A!".. I !??? t + --,e --ng <a-a ~ m r- -nw

i o

  • 8.

Acceptance Tests and Maintenance Program 8.1 Acceptance Tests 8.1.1 Visual Inspection The package is visually examined to assure that the appropriate fasteners are properly secured and the package is properly marked. The seal weld of the radioactive source capsule is visually inspected for proper closure. 8.1.2 Structural and Pressure Tests The source assembly is subjected to a static tensile test with a load of seventy five pounds. Failure of this test will prevent the source sssembly from being used. 8.1.3 Leak Tests the radioactive source capsule (the primary containment) is wipe tested for leakage of radioactive contamination. The source capsule is also subjected to a vacuum bubble leak teet. These tests are described in Section 7.4. Failure of any of these tests will prevent use of this source capsule. 8.1.4 Component Tests The lock assembly of the package is tested to assure that the security of the source asse=bly will be maintained. Failure of this test will prevent use of the package until the lock assembly is corrected and retested. 8.1.5 Tests for Shielding Integrity The radiation levels at the surf ace of the package and at three feet from the surf ace are measured using a small detector survey instrument (i.e. AN/PDR-27). These radiation levels, when extrapolated to the rated capacity of the package, must not exceed 200 milliroentgens per hour at the surf ace nor 10 milliroentgens per hour at thr e feet from the surface of the package. Failure of this test will prevent use of the package. 8.1.6 Thermal Acceptance Test Not Applicable Revision 0 1 Aug 1980 8-1 ~ -. -

1 .s o 8.2 Maintenance Program 8.2.1 Structural and Pressure Tests Not Applicable 8.2.2 Leak Tests As described in Section 8.1.3, the radioactive source assembly is leak tested at manufaccure. Additionally, the source assembly is wipe tested for leakage of radioactive contamination every six months. 8.2.3 Subsystem Maintenance The lock assembly is tested as described in Section 8.1.4 prior to each use of the package. Additionally, the package is inspected for tightness of fasteners and general condition prior to each use. 8.2.4 Valves, Rurture 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 not exceed 200 milliroentgens per hour at the surf ace nor ten milliroentgens per hour at three feet from the surface. 8.2.6 Thermal Not Applicable 8.2.7 Miscellaneous Inspections and tests designed for secondarj users of this package under the general license provisions of 10CFR71.12(b) are included in Section 7.4. Revision 0 1 Aug 1980 8-2 s .w 4,, --m-- -y ,q -,,,}}

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