ML19211A696
| ML19211A696 | |
| Person / Time | |
|---|---|
| Site: | 07109027 |
| Issue date: | 11/29/1979 |
| From: | Munro J TECH/OPS, INC. (FORMERLY TECHNICAL OPERATIONS, INC.) |
| To: | Macdonald C NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS) |
| References | |
| 14892, NUDOCS 7912200461 | |
| Download: ML19211A696 (70) | |
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Tech / Ops Radiation Products Division 40 North Avenue
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' G-ss M' " 3!b~ ON 29 November 1979 Mr. Charles E. MacDonald, Chief Transportation Branch Division of Fuel Cycle and Material Safety U.S. Nuclear Regulatory Comnission Washington, DC 20555
Dear Mr. MacDonald:
We request renewal of USNRC Certificate of Compliance No. 9027 issued for Technical Operations Models 741 and 741E Type B Packages.
In accordance with your letter of 23 July 1979, eight copies of a consolidated application for this package are enclosed. In accordance with 10CFRlTO.31, Item ll.E, we are also enclosing.a check for $150 for the renewal fee.
We are simultaneously applying to the U.S. Tepartment of Transportation for an International Atomic Energy Agency Certificate of Competent Authority issued under the 1973 Revised Edition of IAEA Safety Series No. 6 for Type B(U) packaging.
We trust that this application satisfies your requirements for renewal of this certificate.
Singerer, ql'd 1629 233 J
Jahn J.[L o III chnical Director JJM/fb Fr.cl.
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TechiOps Rad,ation Products Divison 40 North Avenue Burling'on. Massachusetts 01803 Te!ephone (617) 272 2000 29 November 1979 Mr. Charles E. MacDonald, Chief Transportation Eranch Division of Fuel Cycle and Material Safety U.S. Nuclear Regulatory Co=rission Washin6 ton, DC 20555
Dear Pr. MacDonald:
We request renewal of USNRC Certificate of Compliance No. 9027 issued for Technical OIerations Models 741 and 7 lE Type B Packages.
In accordance h
with your letter of 23 July 1979, eight copies of a consolidcted application for this Inckage are enclosed.
In accordance with 10CFRlTO.31, Item ll.E, we are also enc 1csing a check for $150 for the renewal fee.
We are sinultaneously applying to the U.S. reparttent of Transportation for an International Atomic Energy Agency Certificate of Competent Authority issued under the 1973 Revised Edition of IAEA Safety Series No. 6 for Type B(U) packaging.
We trust that this application satisfies your requirements for renewal of this certificate.
Sincere ",
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1 John J.hlanro & "
1629 234 III Technical Director s_./
JJN/fo Encl.
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R.R. Rawl, USDOT L/s /
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Tech / Ops r; o-eo dha Radiation Products Division 40' North Avenue Burhngton, Massachusetts 01803 Te ephone (617) 272-2000 29 November 1979 Mt. Richard R. Ravl Health Fnysicist Office of Hazardous IMterials Regulation Materials Transportation Bureau United States Dapartment of Transportation Research and Special Prograns Administration Washington, DC 20590 D2ar Fr. Rawl:
We request issuance of an International Atomic Energy Agency Certificate of Cor;etent Authority for Type B(U) packaging under the 1973 Revised Edition of IASA Safety Series No. 6 for Technical Operations Models 741 and Th1E Type B Tuckages, USA /9027/B. We are requesting this certificate due to the serious difficulties we have encountered in transporting this packag'2 internationally with its approval based on the 1967 Edition of IAEA Safety Series No. 6.
We are enclosing two complete copies of the package description for the Fodels 741 and ThlE. We are simultaneously applying to the U.S. Nuclear Rerulatory Cor=ission for renewal of USNRC Certificate of Compliance No. 9027 issued for this package. Eight copies of this package description have been forwarded to USNRC.
We trust that thi s request contains the '.nformation you require for your review. Your prompt action vould be g"catly appree's ted.
Sincereh
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o III te/nnical' Director 1629 235 JJM/fb Encl.
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C.E. Pacronald, USi!RC
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Rad:ation Products Divison 40 North Avenue Burlington, Massachusetts 01803 Telephone (617J 272 2000 PACKAGE DESCRIPTION TECHIIICAL OIZ4ATIOIIS l
I*ODEL V+1 USA /9027/B 1629 236
1.
General Information 1.1 Introduction The Tech / Ops Models 741 and ThlE are designed for use as gat =a ray projec-tors and shipping containers for Type B quantities of radioactive material in special form. The Model 741 differs from the Model ThlE only by the addition of an electric circuit, which provides compatibility with Tech / Ops Model 657 Autmatic Exposure Device. Throughout this evaluation, the Models 741 and ThlE are considered interchangeable, except where specifically designated.
The Model 741 conforts to the criteria for Type B packaging in accordance with 10CFR71 and satisfies the criteria for Type B(U) packaging in accor-dance with IAEA Safety Series No. 6,1973 The sources to be used in conjunction with the Model 741 are Tech / Ops sealed source asseiblies Models Nos. A424-9 and A424-18. The source assemblies vill contain a maxitum of 240 curies of iridium-192 and 33 curies of cobalt-60 respec-tively, as special form.
1.2 Package Description 1.2.1 Thekaging The Model Th1 is 11.25 inches (236=m) high,191 inches f h86mm) long, and 14 inches (365mm) wide in overall dimension. The gross weight of the package is 300 rounds (136kg).
The radioactive source assembly is stored in a zircalloy or titanium "S" tube in the geometric center of the package.
The weight of the uranium chield is 200 pounds (91kg). The shield is provided with a paint finish.
The shield is enclosed in a shell fabricated of inch (6.35mm) thick hot rolled steel. The shield is fixed in position within the shell by the retaining bar assemblies. The void space betvaen the shield and the shell is filled with a cestable rigid polyurethane fcam.
Steel-uranium interfaces are seIarated with 0.010 inch (0.25 hem) thick copper separators.
Attached to the sides of the container are 0.625 inch (15 9:n) thick hot rolled steel side frames used for lifting the package.
Mounted at each end of the "S" tube are positioning devicen. The source assembly is locked in Iosition by neans of the control cable connector and additionally secured by zeans of a shipping plug. A protective shippin6 plate ([ inch thick steel) is mounted over the control cable assembly.
Tamperproof seals are provided during shipt.ent of these sources.
Assembly joints which are not 2eak-tight provide passageways for the escape of any gas generated from decc1 position of the potting foam in the event the projector is involved in a fire accident. The outer packaging is designed to avoid the collection and retention of water. The rackage is painted and finished to provide for easy decontamination. The radicac".ive r.aterial in sealed inside a source capsule, which is the containment vessel of the package.
REVISION O 1629 237 1-1 NOV 2 C "o" j~ ~
W
. The Model 741 has been previously approved for use as a Type B package under USNRC Certificate of Compliance No. 9027, Rev. 1 (enclosed in Section 1 3).
1.2.2 Operational Features The source assembly is secured in the proper position by the control cable connector and lock assembly. This assembly requires a key for operation, and thus provides positive closure. A inch (6.35mm) thick steel shipping plate is used to protect the assembly during shipment.
Additionally, the source assembly is secured by means of a shipping plug inserted in the oppo-site end of the "S" tube. This plug is seal vired and provided with a tamperproof seal.
1.2 3 Contents of Packaging The Model 741 is designed fc2-a capacity of up to 33 curies of cobalt-60 as Tech / Ops Source Assembly Ah24-18 and up to 240 curies of iridium-192 asTech/OpsSoureAssemblyAh24-9 The assemblies are in special form as prescribed in 10CFR71 and IAEA Safety Series No. 6,1973 1629 238 1-2 REVISICII O NOV. 2 91970
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13 APPENDIX
- USNRC Certificate of Compliance No. 9027, Rev.1
- Inscriptive Assembly Drawing, Model 741 1629 239 i
i 1-3 REVISIC:: O NOV. 2 91979
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F or rn N H C 618 U.S. NUCL L AR RE GUL ATORY COYMISSIOct CERTIFICATE OF COMPLI ANCE 0CR O,
for Riviioactive Materials Packages 1 (a) Certificat e Number 1.(b) Revision No.
1.[c) Pack age identif.c.ation No.
1.(d ) Fases No 1.(el Total No. Pa9 9027 1
USA /9D27/Bf i 1
1
- 2. PRE AMBLE 2.(a)
This certifica. it issued to satisfy Sections 173 393a.173.394,173.395, and 173396 of the Department of Transportation Hazard Materials Reg tations (49 CFR 170-189 and 14 CF R 103) and Sections 146-19-10a and 14G-19-100 of the Department of Transportation Dangerou: Cargoes Regulations (46 CF R 146-149), as amer 6d.
2.(b)
The packaging and contents described in item 5 below, meets the safet y stand.wt's wt forth in Sut:part C of Title 10. Code of F Metal Re gulations. Part 71. " Packaging of R Aficactive Materials for Transpor, ard Transportai on of Radioactive f.'aterial Under Certain Conditions.'
2.f c)
Diis cer1ificate does not reheve the consignor from compf.ance with any seguise.*ent of the segu!ations of the U.S. Department of Transportation or cther applicable regulatory agencies. including the government od any country through or into which the packa;e will be transported.
- 3. This certificare is issuN on the basis of a safety analysis report of the pacb age design or a;4.' cation-3.ta)
Prepared by (*.'ame and address):
3.(b)
Title and identiGcation of report or a;rlication:
Technical Operations, Inc.
Technical Operations, Inc. application fiorthwest Industrial Park dated August 15, 1974, as supplemented.
Burlington, Massachusetts 01803 3 <c)
D oc k ei No. 71-9027 4.
CONDITIONS This c ertificate is conditional upon the fu! filling of the requirements of Subpart D of 10 CJ R 71, as applicable, and the conditions specifie-in i:em 5 below.
- 5. Description af f aciapng and Authorized Contents. Modet Numner. P>ssile Class. Other Corm.ons, ard Ref erences-(a )
Packaging (1 ) Models flos.
741 and 741E (2) Description A steel encased, uranium shielded Gamma Ray Projector.
Primary components consist of an outer steel shell, internal bracing, polyurethane potting material, depleted uranium shield, and a zircalloy "S" tube.
The contents are securely positioned in the zircalloy "S" tube by a source cable locking ~ device and shipping pl ug.
Tamper-proof seals are provided on the packaging and a 1/4 inch thick steel shipping plate isTolted over the source locking mechanism for additional protection during transport.
The total weight of the package is approximately 300 pounds.
?00R ORG E un m M
REVISION O AOL'. 2 9 1979
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Page 2 - Certificate No. 9027
' Revision No.1 - Docket No. 71-9027 5.
(a) Packaging (continued)
(3) Drawings The packaging is constructed in accordance with the following Technical Operations, Inc. Drawings Nos.:
B66001-1 thru 8 D74101, Rev. A B66001-12, 20 B74101-1, 4 A66001-4, 5, 6, 11 A74101-2, 5, 7, 8 B65502 C74102 B65502-1 D74102-1 B65503 A74102-2, 3 B655E01 C74103 B65501 -6 074103-1 65502 Bill of Mat'ls.
C74103-2 CSK 1923 74104 (b) Contents (1 ) Type and form of material Cobalt-60 or Iridium-192 as sealed sources which meet the requirements of special form ad defined in 571.4(o) of 10 CFR Part 71.
(2) Maximum quantity of material per package 33 Curies of Cobalt-60; or 240 Curies of Iridium-192 6.
The source assemblies authorized for use in this packaging are limited to Models Nos. A424-9 or A424-18 as shown in Technical Operations, Inc. Drawing No. C42400, Rev. F, Sheet 2 of 3.
7.
The name plates 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 os Paragraph 71.12(b) of 10 CFR Part 71.
9.
Expiration date:
January 31, 1980.
jf}g }4j HEVISION O NOV. 3 91979
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Q Page 3 - Certificate No. 9027 - Revision No.1 Doc %t No. 71-9027 REFERENCES Technical Operations, Inc. application dated August 15, 1974.
Supplement dated:
December 19, 1974.
FOR THE U.S. NUCLEAR REGULATORY COMMISSION
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w Charles E. MacDonald, Chief Transportation Branch Division of Fuel Cycle and Material Safety a, 2 S 197 7 Date:
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Structural Evaluation 2.1 Structural Design 2.1.1 Discussion Structurally the Model 741 consists of five components: a source capsule, shield assembly, outer shell, side frames and lock assembly. The source capsule is the primary containment vessel.
It meets the requirements for special form radioactive material as outlined in 10CFRTl (See Section 2.8).
The shield is 200 pounds (91kg) of depleted uranium. The shield assembly fulfills two functions:
It provides shielding for the radioactive material and, together with the positioning mechanisms, insures proper positioning of the source. The shield assembly is supported with retaining bars which are forced together by means of hex nuts threaded on adjusting screws. The adjusting screws and retaining bars are secured with jam nuts. The entire shield assembly is potted in a castable rigid polyurethane fos: and encased in a inch (6.35mm) thick hot rolled steel shell.
Steel-uranium interfaces are separated with copper. Attached to a the shell are side frames made of 0.625 inch (15 9mm) thick steel which are bolted together with 7/16-20 UNF hex head bolts. These are designed as lifting devices and impact limiters.
The key operated lock assembly and control cable connector secure the source in the shielded position.
4 inch (6.35mm) thick steel shipping plate is installed to protect the lock from damage.
Positive proof of source position is evidenced by use of a seal wired shipping plug.
2.1.2 Design Criteria The Model 741 is designed to comply with the requirements of 10CFR71 and IAEA Safety Series No. 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 10CFRTl or IAEA Safety Series No. 6, lot 3 2.2 Weights and Centers of Gravity The Model 741 projector weighs 300 pounds (136kg). The shield assembly contains 200 pounds (91kg) of depleted uranium. The center of gravity is located approximately at the geometric center of the package.
2.3 Mechanical Troperties of Materials The Model 741 Gamma Ray Projector shell is made of hot rolled steel. This 2
material has a yield strength of 40,000 pounds per square inch (276ME/m ),
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Reference:
Machinery's Handbook, 20th Edition,1976, p. 452) 2.4 General Standards for All Packages 2.h.1 Chemical and Calvanic Reactions The materials used in the construction of the Model 741 Gamma Ray Projector 1629 250 REVISION O 2-1 NOV. 2 91979 4
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are uranium metal, steel, beryllium copper, bronze, copper and circalloy or titanium. There vill be no significant chemical or galvanic action between any of these components.
The possibility of the forcation of the eutectic alloy of iron uranium at temperatures below the melting temperatures of the individual metals was considered. The iron uranium eutectic alloy temperature is approximately 1337 F (725 c). However, vacuum conditions and extre=e cleanliness of the surfaces are necessary.to produce the alloy at this lov temperature.
Due to the conditions under which the shields are mounted, sufficient contact for this effect does not exist.
In support of this conclusion, the following test results are presented.
A thermal test of a sample of bare depleted uranium retal was perfomed by Nuclear Metals, Inc.
The test indicated that the uranium sample oxidized such that the radial dimension was reduced by 1/32 inch. A subsequent test was performed in which a sample of bare, depleted uranium netal was placed on a steel plate and subjected to the the=al test conditions. The test showed no alloying or nelting 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 likelihood of the formation of an iron-uranium eutectic alloy is recote, copper separators are used at steel-uranium interfaces.
2.h.2 Positive Closure The Model 741 source cannot be exposed without opening a key-operated lock.
Access to the lock requires the removal of the shipping plate.
Additionally, there is a shipping plug which is seal wired and provided with a tamperproof scal.
2.4.3 Liftinc Devices The Model Thl is designed to be lifted by the side frames. Each is secured by four,'/16-20 UNF SAE-Jh29 crade 5 hex head bolts. These bolts are installed with 7/16 lock washers. The yield strength of these 7/16-20 UNF bolts is 10,900 pounds (h8.6kn). As there is a thread engagement of 3/4 inch between the Lolt and the tapped rod, the yield strength is less than the stripping stren6th (approximately 21 A
tcrque of 30 foot-pounds (,000 pounds) and thus is the limiting factor.
41U-m) is applied to these bolts. This corresponds to a tension of approxinately h030 pounds (18.2kn).
The total tensile load-ing on each bolt is h 10 pounds (19.2kN), due to the bolt tcrq2e and three 3
times the weight of the package. The total torsional loading is 6100 pounds (27.lkN). Foth loads are less than the yield strength of the bolts.
The veld joining the side frame to the side frate insert on che Fodel 741 is a 3/16 inch fillet veld.
The Merican Welding Society "Coie for Arc and Gas Welding in Building Construction" permits the stress on a fillet veld 1629 251 2-2 EVISION O t.W. 2 S T3
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2 to be 13,600 psi (89 6Icz/m ).
As the shear stress on the thrcat of the fillet veld is the limiting factor, the allowable stress on a 3/16 inch fillet veld (throat dimension, 0.133 inch, 3 3Scm) is calculated to be 1,800 pounds per linear inch (320 i/r:m) As the perineter of the side frame insert is 46 inches (1.2m), the allevable load is 82,800 pounds (369kN).
Hence, the allowable load on the side frame insert veld is greater than the yield strength of the bolt.
2.4.2 Tiedown D2 vices The tiedown devices on the Model Thl are the side frames. As indicated in 2.4.3 above, these frames can safely support the package.
25 Standards for Type B and large Quantity Packages 251 Load Resistance Considering the package as a simple beam supported on both ends with a uniform load of 5 times the package weight evenly distributed along its ler6th, the maximum stress can be computed from:
F1 S
=
M where:
S: maximum stress F: total 3 cad (1500 pounds) 1:
Length of beam (19 1 inches) 3 Z:
section modulus of beam (52.6 in )
(
Reference:
I/aehinery's Handbook, 20th ed.,1976, p. 442)
Thus, the maximum stress generated in the. beam is 68 pounds per square inch 2
(470kir/m ), which is far below the yield strength of the material, 40,000 psi 2
(276?cz/m ).
252 External Fressure The Model Thl is open to the atmosphere; thus, there vill be no differential pressure actirc on it.
Tne collapsing pressure of the source capsules can be found:
86,670 t_
- 1386 P
=
D where P:
collapsing pressure in pounds per square inch t: vall thickness in inches (0.020 inch)
D:
outside diameter in inches (0.25 inch)
(
Reference:
Machinery's Handbook, 20th ed.,1976, p. hh8) 1629 252 2-3 REVISION O NCE 2 91979 m--,,y--..,,-.
...------,.w.
7.
. y.
,,, u _,,.___.,, _.
.c--
- ~
The collapsing pressure of the capsules is calculated to be 5550 pounds per square inch (38.8'C;/m ).
Therefore, the capsule can withstand an 2
external pressure of 25psig.
2.6
!!ortal Conditions of Transport _
2.6.1 Heat _
From The thermal evaluation is perforred in Chapter 3 of this application.
it can be concluded that the Model 741 can withstand the this evaluation, normal heat transport conditions.
2.6.2 Cold The retals used in the tanufacture of the Model 741 can all withstand temperatures of -40 F (-h0 C). The lover operating limit of the polyure-0 thane foa= is -100 F (-73 C). Thus, it is concluded that the Model 741 vill withstand the normal transport cold conditions.
2.6.3 Pressure The Model 741 is open to the atmosphere; thus, there vill be no differential In Section 3 5 4, the source capsules are deconstrated pressure acting on it.
to be able to withstand an external pressure reduction of 0 5 atmospheres e
(50.7kt;/c ),
2.6.h Vibr ation The Fodel 741 has been in use five years.
During that tir.e there has never been a vibrational failure reported. Thus, we contend the Model Thl vill not undergo a vibrational failure in transport.
2.6.5 Water Spray Test The water spray test was not actually performed on the Model Thl.
We contend that the materials used in construction of the Model Thl are all highly water resistant and that exposure to vater will not reduce the shielding or affect the structural integrity of the package.
2.6.6 Pree Drop The drop analysis perforced in Hypothetical Accident conditions (See Section 2.7 1) is sufficient to satisfy the requirenents outlined for tha normal transport free drop condition in 10CFRTl and IAEA Safety Series !!o. 6,1973 the free drop On this basis, we conclude that the Fodel Thl can withstand without impairnent of the shielding cr package integrity.
2.6.7 Corner Drop 1629 253 iot applicable REVISION O 2h Nov.2 9 % %
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,2.6.8 Penetration e
A penetration test of the Vodel 741 was not actually perforned. However, the similar Model 63h was subjected to the penetration test with no resultant loss of shielding or package integrity (a copy of the test report is enclosed in Section 2.10).
The following analysis demonstrates that the maximum damage exhibit;ed by the Model Thl due to the penetration test is less than that of the Podel 634.
The taximum stress observed in a flat rectangular plate supported on all edges due to concentrated central loading is:
0.62F in L I
+
0 577 S
=
2 2r t
o4 where F:
total load t:
thickness of plate (inches)
L:
length of longest side (inches) r:
0 325t (inches) o
(
Reference:
Machinery's Handbook, 20th ed.,1976, p. 4h4)
The appropriate dinensions for the Model 741 and Model 63h are:
Podel Thl Podel 68h t
0.25 inch (6.35mm) 0.1875 in. (4.76=m)
L 19 1 in;h (h85mm) 17.in. (432m=)
r 0.0812 in. (2.06mm) 0.0609 in. (1 55mm) g The calculated stress for the Fodel Thl is 53 0F; for the Model 68h it is 97 3F.
In both cases the load F (40 inch drop of a 13 pound hemispherical billet) and the material of construction (hot rolled steel) are the sane.
The maxirum stress, and thus the maximum danage, to the flat plate occurs in the Podel 63h. The shipping plate which protects the lock techanism is the same in the two rodels. As the Model 68h successful 3y withstood the penetration condition, we conclude that the Model Thl can undergo the penetration test with no loss of structural integrity or shieldin6 (A
copy of the test report for the Fodel 63h is enclosed in 02cticn 2.10).
2.6.9 Compression The gross weight of the Model Th1 is 300 rounds (136kg). The maximum 2
cross sectional area of the package is 268 square inches (0.173s ).
- Thus, 2 pounds per square inch times the cross rectional area (600 pounds, 2'(3kg) is less than five tines the package weight (1500 pounds, 682kg).
For this analysis, the load vill be taken to be 1500 pounds.
The taximum stress Eenerated in a flat rectangular steel plate with all edges fixed and a load distributed uniformly over the surface of the plate 1629 254 ntu o
liOV. 2 g ig7g
];
u~
<- c;. ::. : -
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m.
.... -;7 ; y c,. - - - ~.
, - - ~
I
can be computed from:
0 5F S
=
0.623 / Af 5 2 [ l_
+
t (v/
,v 4
vhere: S: maximum stress F: total load (1500 pounds) t: thickness of plate (0.25 inches) vidth of plate (14 iriches) v:1: length of plate (19 1 incht:c)
(
Reference:
Machinery's Handbook, 20th ed., 1976, p. 444, Eq. 13)
From this relationship, the r:aximum stress generated in the plate is 2780 pounds per square inch (19 2MN/m ).
This figure is greatly below the yield 2
2 strength of the material, 40,000 pouncis per square inch (276fG/m ).
- Thus, it can be concluded that compression vill not adversely affect the package.
27 Hypothetical Accident Conditions 271 Free Drop The Model Th1 was not actually submitted to the 30 foot drop test. However, the Model 655 was submitted to the drop test (the test report appears in Section 2.10).
The Model 741 has approximately the same weight and is constructed from the same caterials as the Model 655:
Model 655 Fodel Thl Length 19-3/4 inches (502mm) 19-1/8 inchas (486mm)
Width 11 inches (279mm) lh inches (356mm)
Height 10-1/4 inches (260:=n) ll-1/h inches (286mm)
Weight of Shield 280 lbs.
200 lbs.
Gross Weight of Container 385 lbs.
30,lbs.
Side Frame Material 3/h inch thick (19.lmm) 5/8 inch thick (15 9mm) ductile iron hot rolled steel Shell Material d inch thick (19 3r=)
[ inch thick (15 9mm) steel hot rolled steel Eased on the satisfactory performance of the Model 655, we conclude that the Model Th1 vill undergo no loss of shielding or structural integr5ty as a result of the 30 foot drop test.
2.7.2 Puncture The Model 741 was not submitted to the puncture test of 10CFRT1. However, the similar Model 655 (see Section 2 7 1) was rubmitted to the puncture test. There was no resultant damage to the container nor reduction in shielding.
(A copy of the test report appears in Section 2.10).
The 1629 255 nrvmon o 2-6 Nov. 2 9 Ing
shipping plate used in the Model 7h1 is the same as that used in the Model 676. The Model 676 puncture test report (included in Section 2.10) shows that the shippin6 plate withstood the puncture test. On this basis, we conclude that the Model 7h1 can successfully withstand the penetration condition of 10CFR71.
273 Thermal The thermal analysis is presented in Section 3 5 There it is shown that the melting points of the caterials, except the potting compound, used in the construction of Model 741 are all greater than lh75 F (800 C).
Also indicated is the previous acceptability of this design (HRC Certificate of Cam.pliance No. 9027, Rev.1) using this evaluation.
Thus, it is concluded that the Model 7h1 satisfactorily reets the require-tents for the hypothetical accident-thermal evaluation as set forth in 10CFRT1.
2.7.h Water Irrersion Not applicable 275 Summary of Ihmage The tests designed to induce techanical stress (drop, puncture) caused minor defomation, but no reduction in the safety features of the package.
The themal test resulted in no reduction of the safety of the package.
It can be concluned that the hypothetical accident conditions have no adverse effect on the shielding effectiveness and structural integrity of the package.
2.8 Special Form The Model Thl Garma Ray Trojector is designed for use with Tech / Ops Source Assemblies Ah24-9 and Ah2h-18. These source assenblies have been certified as special form radioactive material (IAEA Certificate of Competent Authority T s. USA /^'65/S and USA /OlSh/S.) We contend that these certificates are sufficient evidence that the requirements for special form radioactive caterials, as established in IAEA Safety Series No. 6,1973, are satisfied.
29 Fuel Rods Not applicable.
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b
2.10 APPENDIX
- Nuclear Metals, Inc., Test Report:
- Test Report: Penetration Test, Model 684
- Test Report: Drop and Puncture Tests, Model 655
- Teut Report: Puncture Test, Model 676
- Dascriptive Assembly Drawings, Source Assemblies
- IAEA Certificates of Competent Authority Hos. USA /0165/S, USA /0154/S 1629 257
~
2-8 REVISIOII O NOV 2 9197;
- yz- --. -- _ - ; s. 7,.
- - - - - - - ~
/
[L (gg x ucte u sie u os,i~c.
??
2229 M AIN ST RE ET CON C O R C. M A5 5 ACw e5E T T 5 0 742 T L L t poeGNE 687 3 E 9 -SeiO 28 January 1974 Technical Operations, Inc.
Radiation Products Division South Avenue Burlington, Massachusetts 01803 Attention:
Mr. J. Lima Gentlemen:
In response to a reouest by Joe Lima of Tech Ops, a simulated fire test was performed on samples of bare depleted uranium in contact with mild steel, the object being to determine what, if any, alloying or melting would occur under these conditions.
TEST DATA:
{
A 3/4-inch diameter x 5/8-inch long bare depleted uranium specimen was set on a 1-inch diameter x 1/8-inch thick 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:
!!o reaction was evidenced between the two metals.
Both separated readily and showed no alloying or melting charact eristics.
Oxidation of th'e uranium was about the same degree as that reported to Joe Lima on an earlier experiment.
The test was performed by l'MI on 25 January 1974.
1629 258 Very truly you C.
~
/
b
(/ohn G. Powers J
Project Engineer 2-9 REVISION 6
' 3 t.
2 9 M7; c.
m
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,__-.g_,
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TEST REPORT RADIATION PRODUCTS DIVISION BY:
John J. Munro III DATE:
5 September 1979 SUPJECT:
Model 684 Fenetration Test On 5 Septer.ber 1979, a penetration test was performed on a Technical Operations Model 684 Shipping Container in accordance with 10CFR71 Appendix A.8 and IAEA Safety Series No. 6,1973, paragraphs 714a and 71hb.
The hemispherical end of a vertical steel cylinder 1.25 inch in diameter weighing lb pounds was dropped frc= the height of 40 inches onto the geometric center of the bottom surface of the Model 684. There was no deformation and no damage which would affect the shielding or structural integrity of the package.
A second test was conducted using the same cylinder.
It was dropped from the height of h0 inches onto the shipping plate. There was no deformation and no damage which would affect the shielding or structural integrity of the package.
Documentry photocraphs are enclosed.
Perfomed by Witnessed by A
tw Y
(Johr J.' Munro III Arg'elo Kiklis 1629 259 2-to REVISION O NOV 2 9 !g7f O
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4 TEST REPORT DESCRIPT10N: Model 655 - 30' Drop DATE March 18,1970 The first drop test produced no damage. The second drop test broke the corner of the side plate off. Two tie-rod bolts were sheared off one side and one sheared off on the other side.
D. U. Shield remained in place. The sout ce Tube was straight and the front nut turned freely.
The panetare test (41" drop on to a 6 dia, steel billet) left a 1/16" deep x 1/8" wide x 1" long gouge on the botton.
CONCLUS10N:
BY Richard Evans WITNESSED BY Fred IIauser
(
2 - 14 REVISIO:I C
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a-1 FROM LE FT, MODELS G70, 655 AND 672 G AMMA R AY PROJECTORS AFTER CONCLUSION OF 30 FOOT DROP TEST AND PUNCTURE TEST N
300RORGNAL 1629 264 i
1 2 - 15 REVISION O
.ya;. o a 3b.:.
u
t
(,
TEST REPORT DESCRIPTION:
DATE 9 December lo7h Puncture Test of Model b76 Containe" Connector A Model 676 Carta Ray Projector with Shipping Flate installed was dropped fro: a height of 40 inches onto a six inch disteter, eight inch high steel Billet as shown in Figure 1 a. The Container impeeted on the Shipping Plate as shown in Figure 1 b.
CONCLUSION:
tio damage to the container, shipping plate or control cable connector resulted. There was no reduction of shielding effectiveness nor loss of Fadioactive Material.
i pa--e LA BY Don Brasceur WI TN ESS ED BY [T hn J. !'un o 111 t
v v
1629 265 2 - 16 nsvistou o j
2 s :su
~
(
s3
- FUNCTURE it,ST COIiTROL CABLE CONIECTOR ASSEMBLY j
Model 676 Gamma Fny Projector (veight 545 poundo) dropped from a height of 40 inches onto a steel cylinder (61n die x 8 in high)
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l 1629 267 Model 670 at Conclusion of Puncture Test e _ 18 REVISION O NC;. 2 9 072 8
e g
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.MODEL CAPACITY CAPSULE STYLE DlM A DlM 8 (CURIES)
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9 'MG
/.4 24 22 6001I LG 0001 N.A.
Il %
A424-4 55 6001i, 60000 N.A 20 A4 24-5 6
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7L A424-7 165 600lR ; 6 0002 N.A.
ITW A4 24 -8 I]O 6001I I$0060 N.A.
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1629 268 M ATE RI ALS TECHNIC AL OPER ATIONS I N C.
7 RADIATION PRODUCTS DIVISION f,
B U R LIN G TON, MA 01803 FINISH
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A M O60 R EY.
ATE DESCRIPTION IN NER CA?SJLE :
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~
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OUTER CAPSULE OllTE R PLU G
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A60012 -I AGOOIR-2 AGoat2 -5 3
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DWG TITLE
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RADIATION PRODUCTS DIVISION
/,
B U R LIN G TO N, MA 01803 N
"*""EOBALT 60 SOURCE REFERENCE DRAWN BY g((S,5,ggjf,wyyctsant p
(SINGLE ENCAPSULATl0N)
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PAGE 2-21
-- _ _. ~. -
A
/2-:5-7?
ffE DM27 2 rJr P00RDlGNAL MODEL CAPACITY CAPSULE DlM A DIM B (CURIES)
STYLE A424-(
120
[@ R 4 373 f
c A424-9 (20 6gpfcgR l 225 T
A814of I20 Sk%
- l.875 79(6 A68509 12 0 C66510
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- 1629 271 MATERIALS TECHNIC AL OPERATIONS IN C.
RADIATION PRODUCTS DIVISION
_7 IIUISH BURLINGTON, M A 01003 DWG TITLE 192 oa>*" sv
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IRlDiUM SOURE REFERENCE h#6 ~f7/8
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7 RA'- r #' *M
! t RADIATION PRODUCTS DIVISION f
B U R LI N G T O N. MA 01803 FINISH OWG TITLE 192
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.a N-Dl_P ARTMENT OF 1 RANSPCHl ATION 3
RESEARCH AND SPECI AL FROGRAMS A DMINt' I R ATION j
w A s m N C.T O N.
D C.
20!90 3
%,,,e
~~
I AEA CERTIFICATE OF CO.'4?ETENT AUTHORITY hEFER 70-Special Form Radioactive Material Encapsulation Certificate Number USA /0165/S (Eevision 0)
This cert i fi es that the encapsulated sources, as described, when loaded with the authorized radioactive contents, have been demonst rat ed t o racet the regulatory requirerents for special form radioactive material as I
2 prescribed in IAEA and USA Regulations for the transport of radioactive caterials.
1 I.
Source Description and Radioactive Contents - The sources described by this c e r t.i f i c a t e consi st of the following Technical Operations, Inc.,
models which are welded capsules constructed of either 304 or 304L stainless steel to the list ed capsule designs (see Appendix A) and which contain not more than the listed cuantit ies of Cobalt-60 in met allic ferm:
i Model Capsul _c Style Activity (Curi es)_
j A424-2 60011, 60001 22 A424-3 60011, 60001 22 4424-4 60011, 60000 55 4424-5 60011, 60001 6
A424-7 60012, 60002 165 A424-8 60011, 60000 110 A4 2 4 -10 60011, 60004 6
A424-11 60011, 60004 55 A 4~2 4 -12 60011, 60004 110 A424-13 60012, 60002 330 A424-14 60011, 60004 110 A424-15 60011, 60004 11 A424-16 60011, 60000 55 A424-17 60011, 60000 55 A424-18 60011, 60000 33 A42?-19 60001, 60004 0.11 A453-1 60011, 60000 110 A453-2 60012, 60002 165 A453-5 60012, 60002 550 A453-6 60013, 60003 1100 A453-7 60011, 60000 110 A453-8 60011, 60000 55 A453-9 60011, 60000 55 A453--10 60011, 60000 55 1629 273 L
2 - 2 'l REVISIC O f
hS. 2 s 13y
~
~
e Certificate Number USA /0165/S, Kevision 0 Page 2
~
.(.
II.
This cer tificat e, unless renewed, expires on September 30, 1982.
This certificate is issued in accordance with paragraph 803 of the IAEA Fegulations and in response to the July 26, 1979, petition by Technical Operations, Inc., Burlington, Massachusetts, and in consideration of the associated information therein.
Certified by'.
k!
_[
dN'l 6
(Date)
R.
R. Eawl Designated U.S.
Competent Authority for the International Transportation of Kadioactive Materials Office of Hazardous Materials Regulation Materials Transportation Bureau U.S. Department of Transportation 1" Safety Series No. 6, Fegul at ions for the Safe Tr;> n s po r t of Eadioactive Materials, 1973 Revised Edition" published by the International
(
Atomic Energy Agency (IAEA), Vienna, Aust ri a.
Title 49, code of Federal Eegulations, Part 170-178, USA.
1629 274 9
(~
FIVISION O t On 2 2 EO 2 - 25 e
e a
e gg.
g e
'g
s -
g
/.n'.e,*.
p DEPARTMENT OF TRANSPORTATION
[N h.
RESLARCH AND SPECI AL PROGRAMS A DMINISTR ATION
- j q
j W AsW N GT ON. D C.
20590
'5.2,, '
1AEA CERTIFICATE OF COMPETENT AUTHORITY N (
S e c i a_1_Fo rm Ra d i_o_a c t i ve Ma t e r i a l_En c ap s ul a t i on P
,, m,, o s
Certificate Number USA /0154/S l
Thic certifies that the encapsulated sources, as described, when loaded with the authorized radioactive contents, have been demon-I strated to meet the regulatory requirementy for spegial form radioactive material as prescribed in IAEA and USA' regulations fcr the transport of radioactive materials.
1.
So_u r c e De s c rip _t i on - The sources described by this certificate are identi.~ied as the Technical Operations, Inc., Models which are described and constructed as follows:
i Model No.
Capsule S t yl e Approximate Size Mn_ inches, diameter x lencth)
A424-1 B60001 or B60004
.25 x.97 A424-6 B60001 or B60004
.25 x.97 A421.-9 B60001 or B60004
.25 x.97 A424-20 B60001 or B60004
.25 x.97 A38101 B60006 Pellet, Wafer or Large Wafer
.25 x.90 A6E309 C68310 Pellet or Wafer
.25 x.78 A81401 B60001 or B60004
.25 x.97 B69701 B60001 or B60004
.25 x.97 All capsules are constructed of either 304 or 304L stainless steel l
and conform with the following design drawings:
f f
Capsule Stvie Drawing Number B60001 B60001 - 1 Rev. H and - 2 Rev. F B60004 B60001 - 1 Rev. H and B60004 - 1 Rev. D B60006 Pellet B60006 - 1 Rev. H and B60001 - 2 Rev. F B60006 Wafer B60006 - 1 Rev. H and B60004 - 1 Rev. D E60006 Large Water B60006 - 2 and B60001 - 2 Rev. F C68310 Pellet C68310 Rev. E and B68310-3 i
C68'1D Wafer C68310 Rev. B II.
Radioactive Cont ents - The authorized radioactive cont ents of these sources consist of not more than the following amount s of Iridium-192 as solid metal:
1629 275 REVISION O
!.'0V. 2 9 Gn 2-zG mm-
~
s r.
Certificate '; umber USA /0154/S Page 2 j.
~'
Model No.
Contents (Curies)
- - _1 A424-1 120 A424-6 120 A424-9 120 A424-20 240
~~
A58101 240 A68309 120 A81401 120 B69701 120 III.
This certificate, unless renewed, expires December 31, 1981.
This certi ficat e is issued in accordance with paragraph 803 of the IAEA RegulationsI, and in response to t he ';ovember 3,1978, pet ition by Technical Operations, Inc., Burlington, Massachusetts, and in consideration of the associated information therein.
}
Certified by:
f i
Y 1-R. R, Rawl, liealth Thysicist- -_- _.
(Date)
U.
S. Depa r t ment of T ransport at ion Office of I!azardous Materials Regulation k'ashington, D.
C.
20590.
i I
t i
I" Safety Series No. 6, Regulations for the Safe Transport of Radioactive Mat erial s, 1973 Revised Edition", published by the International Atomic Energy Agency (IAEA), Vienna, Austria.
1 i
Title 49, Coae of Federal Regulations, Part 170-178, USA.
e 1629 276 REVISION O NOV. 2* 9197(
2-27 D
i t
t f
3 Themal Evaluation F
f 31 Discussion The Model 741 Ga=a Ray Projector is a completely passive themal device and has no mechanical cooling systems or relief valves. All cooling of the
[
1.v,hage is through free convection and radiation. The maximum heat source is the 2h0C1 tridium-192 source. The corresponding decay heat is 2 5 vatts (see Section 3.4.1).
32 summary of Thermal Properties of Materials The melting points of the metals used in the construction of the Fodel 741 are:
D2pleted Uranium :/etal 20700F (11330C)
Carbon Steel 2453 F (1345 C)
U Copper 1940 F (1060 C)
Bronze 1840 F (1005 C)
(
Reference:
!3achinery's Handbook, 20th ed.,1976)
Titanium 3300 F (1820 C)
Beryllium Copper 1600 F ( 8'(0 C)
Zircalloy 3350 F (1845 C)
(
Reference:
Metals Handbook, 1961)
The rigid polyurethane foam has a miniraus operating range of -10dUF to 200 F 0
(-72 C to 93 C).
It will decompose at the fire test temperature cf 14'(5 F0 (300C). reccmposition vill result in gaseous byproducts which vill burn in air.
33 Technical Specifications of Components I;ot applicable 3.4 Iormal Conditions of Transport _
3.h.1 Themal Model The maximum heat source in the Fodel 741 results from the decay of 2h0Ci of iridium-192.
Iridium-192 decays by electron capture and beta crission. The decay energy for both procecses is approxirately 1.h5MeV.
(
Reference:
Radiological Health Handbook, p.403). Thus:
10
- l~'
l.h5MeV x 3 7 x 10 disint x 1.6 x 10 J x 2h0C1 = 2.06 vatts s-Ci MeV The decay heat source is conservatively taken as 2 5 vatts.
REVISION O
[
3-1 A0V. 2 e im
Li-
!I To qualify as a Type B(U) package the requirements of IAEA Safety Series No. 6, 1973, paragraphs 231 and 232 tust be caticfied.
The calculational todel used to demonstrate compliance with these regulations is described in detail in Section 3.6, along with the results of the analysis. Essen-tially, it is assumed that one-fourth of the entire decay heat load is deposited uniformly in each of six sides. The smallest of the sides is assured to reach the maximum surface te Ierature. Heat transfer from the i
side is restricted only to convective heat transfer from the upper face of the plate.
To meet the additional requirecents of Iaragraph 240 of the IATA regulations, a separate analysis was Ierforred. To do this a heat balance vac set up i
over the surface of the package, using the insolation data in Table III of the IAEA regulations.
The decay heat source was considered negligible.
The outer shell was assuned to be insulated from the interior of the package. Heat transfer from the package was taken to occur by radiation, and over specific curface areas by free convection.
A detailed description of the model is given in the analysis in Section 3.6.
3.h.2 Maximu-Temperatures An examination of the telting points of the materials used in construction of the Fodel 741 chow that the maxinun tomperatures encountered under nomal conditic ns of transport engender no loss of structural integrity or loss of shielding of the package. The specific Type B(U) analyses (Sectb n 3 6) show the package temperature to be below 40 C (lok F) in the shade and 0
0 below 65 C (lh9 F) when insolated.
3.4.3 Minimum Temperatures O
U The nininum normal operating temperature of the Fodel 7h1 is -hC C ( LC F).
This temperature vill have no adverse effect on the Jackage.
3.h.h Maximum Internal Fressure Normal oIerating conditions generate negligible internal pressures.
Any pressure generated is significantly below that of the hypothetical accident
[
pressure, which is shown to result in no loss of shielding or containment.
3.h.5 Maximum Thermal Stresces The raxinum temperatures that occur during normal transport are low enough to insure that thermal gradients will cause no significant therral stresses.
3.4.6 zaluation of Package Terforrance for Horr.al Conditions of Transport o
The therral conditions of normal transport are obviously insignificant from a functional Ioint of view for the Model 7h1. Also, the applicable conditions of IARA Regulations for Type B(U) packages have been shown to be satisfied by the Model 741.
t 3-2 REVISION o h;V.2 9 1H3
35 Evpothetical Acc! dent Thernal Evaluation 351 Thermal Model The Model 741, including the source assembly, is assumed to reach the fire j
test temperature of 800 C (1475 F).
At this temperature the polyurethan potting ec= pound will have decomposed and the resulting gases vill have escaped the package through the assembly joints which are not leak-tight.
352 Package Conditions and Environment The Model 741 is considered to have undergone no significant damage during the free drop and puncture tests; thus, the package in this analysis is assured to be free from functional da. age.
353 rackace Terperatures As indicated in 3 5 1, the rackage reaches a n.axittro of 800 C (lh75 F) 0 throu6 out. An examination of the melting Icints of the natorials used h
in the construction of the Model Th1 (except the pottirc compound, as noted) indicates that there vill be no damage to the package as a result of this tenparature. The possibility of the formation of the iron-uranium eutectic alloy was addressed in Section 2.h.1, where it was concluded that the formation of the alloy was unlikely.
3 5.h Maxiru-Internal Wessures The Model 741 packaging is open to the otrosphere, insuring that there vill be no pressure buildup within the package.
In Section 3 6 there is an analysis of the source capsules under the fire test conditions.
It is shown that the mxitum internal gas pressure at this temperature is 5h.7 pai 2
(0 3TTid;/m ),
The critical location for failure of the capsule is the veld. An internal 2
pressure of 5h.7 psi (0 37?cI/m ) vill generate a naxirum stress of 287 psi 2
(193'CJ/m ) in the veld. At a temperature of 870 C (1600 F) the yield strength of Type 304 or 30hL stainless steel is 10,000 psi (69. cal /m ),
2 r
Thus, at 800 C (14'(5 F), the taximun stress in the capsule would only be 3%
cf the yield strength at that point.
355 Maxirum Ther'.nal Stresses There are no significant thermal strusses generated during the thermal test.
3 5.6 maluation of rackage rerforrance The Model Thl vill undergo no loss of structural integrity or shieldin6 when subjected to the conditions of the hypothetical themal accident. The pressures and terperatures generated have been demonstrated to be within acceptable limits.
i 1629 279 3 _ ~,
NOV. 2 91979
3.6 AFFE!OIX
- Model Thl Thernal Armlysis:
IAEA Cafety Series no. 6, 1973, paragraphs 231, 232
- Fodel Thl Thermal Analysis:
IAEA Safety Series No. 6, 1973, paragraph 240
- Thermal Analysis, 0.25 inch 0.D Capsules 1629 280 t
REV.75:0:: 0 f.:v. 2 s r,7s
?
3 L
i i
Model 7kl - Thermal Analysis IO t
Type E(U), Paragraphs 231, 232, IAEA Safety Series No. 6,1973 i
i This analysis is perfcrmed to demonstrate that the Model Thl Gamma Ray Pro-f jector meets the specific Type B(U) thermal requirements of paragraphs 231 and 232 of IAEA Safety Series No. 6,1973, i.e.,
that the naximum surface 0
temperature does not exceed 50 C in the shade assuming 38oC ambient temper-ature.
To acsure conservatism, it is assumed that:
(1) the entire decay heat (2 5 vatts) is deposited in the exterior faces of the Model Thl, (2) the interior of the Model Thl is perfectly insulated, providing heat trar.sfer from the vall only to the atmosphere.
The rectangular sha pe of the con-tainer means that each face eclipses a different amount 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 3; of the total source (0.63 vatts) uniformly distributed over the face.
Considering the smallest face as undergoing c::e-dimensional convective heat transfer:
/
,t T
T q
Interior (Insulated)
UAir V
/
vall T
q T
vnere:
=
a hA T: temperature at the vall outer surface y
q: (decay) heat source (0.63 vatts) i 2
A: surface area of the strallest face (0.10 m )
h: free convective heat transfer coefficient for air 5 watts /reter - C (Feference:
Heat Trar:sfer. J. P. Ho2ran, 4 th Edition, p. 13)
Thus, the maximum temperature at the vall Tw is 40cc (10h0F) under normal con-ditions of transport. This satisfies the requirements of the aforementioned regulations.
1629 281
+
i.
3-5 REVISIO:IO NOV. 2 91979
Model 741 - Thermal Analysis Type B(U), Faragraph 2h0, IAEA Safety Series No. 6. 1973 This analysis is performed to demonstrate that the Model 741 Gamma Bay Pro-jector meets the specific Type B(U) thermal requirements of paragraph 240, IAEA Safety Series Uc. 6, 1973 This paragraph recuires that the maximum surface temperature of a Type B(U) package not exceed 82 C (180 F) under 0
0 normal conditions of transport, given insolation as outlined in Table III of the regulations and an anbient temperature of 380C_(LOO F).
O The calculational model used censists of taking a steady state heat balance over the surface of the package. To facilitate calculations, certain sim-plifying assumptions are made.
These are outlined belev:
Insolation 800 cal /cm -12hr (775 W/m2) for the top surface, 200 cal /cm -12hr (19h W/m2) 2 for the sides and side frames, none for the base as outlined in Table III of I AEA Safety Series No. 6.
The package is finished with russett enamel.
The solar abscrptivity of this enatel is 0.81 (
Reference:
Thermal Radiation Properties Survey, G. G. Cubareff et. al., 2nd ed., 1960, p. 260).
A conservative figure of 0 90 was used as the package absorptivity.
Decay Heat Load The decay heat load (maximum 2 5 watts) is assumed negligible.
Package Orientation The package rests on the side frames, i.e., in the normal transport orientation.
Jiest Transfer Mechanisms The Model 741 is assumed to undergo free convection and to radiate to the envi-ro nme nt.
The inside fae:s are considered to be insulated, so there is no conduc-tion into the package.
- Further, there are no temperature gradients present.the sides are taken to be thin enough so Radiation: The package is assmed to radiate from the outer shell only, i.e.,
a cub 9 10.1 ine:res (257rm) x 12.6 inches (32mm) x 13 inches (33> m).
This assumption provides for conservatism by not considering any radiative heat loss through the side frames.
Convection, top:
The upper surface of the ooter shell is taken to undergo free convection.
To provide conservatism, the apper surfaces of the side fra~es are considered not to undergo consection.
The heat transfer coefficient of a horizoltal flat plate is I
given by:
h 1629 282 nrvam o 3-G 2 9 U2d
I 1 0.25 h.
1 32 a__T.
=
lL;
(
Reference:
Heat Transfer, J. P. Holman, hth ed., 1976, p. 253) where L is the average of the lengths of the sides, 0 325m.
i 1 7hfAT)
- h
=
t Convection, sides: The vertical components of the outer shell are consi-dered to exhibit free convective heat transfer. For conser-vatism, the side frares are taken to insulated.
Effectively, the vertical convec';ive Feet transfer area is that of a ver-tical plate 0.257m high x 2(0 32 1 + 0 33Dm) long, convecting on one side only.
The heat transfer coefficient for a vertical flat plata is:
- 1. 42 6 T h
=
s
]
(
Reference:
Heat Transfer, J. P. Holma n, hth ed., 1976, p. 253) where L is the height cf the plate,
- 0. 25 7m.
1 99 (6T)U*D h
=
3 Taking a heat balance over the packaEe surface:
heat in = heat out f ra d'.
+ con. top + cony. sides) 91n " 9 rad
- 9et
- Ecs gi, = 0 90 (775 W x A
+ 19h W x A
)
t 2
2 s
e m
m
= 0 93 (775 x 0.11 + 19h x 0.h8) = 161 vatts l
h h) 9 rad "60'^r IT
- T V
a
-8 q
h (0.8) (5.669 x 10 W
) (0 55m2)
=
7
- (311 k) 2-h m
k q
=hA (oT) where oT = T
-T ct g t y
a 1.Th x (oT) *
(0.106m2) (oT)
=
= Qs ( T) 1629 283 ocs 199 (AT)o.25 (0.48m2) (AT)
=
[
i REVISION O 3 NOV. 2 91979 i
"lieratica yields a vall temperature T of 65 C (149 F).
Thus, the Model 741 y
satisfies the requirements of paragraph 2h0, IAEA Safety Series 'Jo. 6,1973 1629 284 h
4 6
P REVISION O I'E' 2 9 157; 3-6
0.25 Inch 0.D. Source Capsules - Themal Analysis Eypothetical Fire Conditions This analysis is intended to demonstrate that Tech / Ops source capsules which are of 0.25 inch (6.35mm) diameter, seal velded to a minimum penetration of 0.020 inch (0 53mm), made of Type 304 or 304L stainless steel, and licensed as special form containers under IAEA Safety Series No. 6,1973, also neet the re iuirements of raragraph 238, IAEA Safety Series No. 6, 19'T3, i.e.,
containment under specified thermal test
- conditions.
I The actual containment vessel for the radioactive naterial is the velded source capsule.
These capsules are all 0.25 inches (6.35mm) in dimeter and Jess than 1 inch (25.L m) in lereth.
t The internal volume of the source capsules contains only iridium-192 or ecbalt-60 metal (as a solid) and air.
It is assumed at the time of Icading that the entrapped air in the capsule is at standard temperature and 0
pressure (20 C, 0.101 f.'eganewtons per squsre reter). We contend that this is a conse_wative assuxption because, during the velding process the in-ternal air is heated, causing sore of the air nass to escape before the capsule is scaled. Vnen the valded capsule returns to anhient temperature, the internal pressure is somewhat reduced.
As described in Tech / Ops standard source encapsulation Ivocedure, the mininun weld penetration is 0.020 inch (0 51mm).
Under conditions of internal pressure, the critical location for failure is this veld.
Since the capsule has an outside dianeter of 0.25 inch (6.3gms), this veld has a cross-sectional area of 0.014 square inches (9 3'hm ).
Under conditions of paragraph 238 of IAEA Safety Series No. 6 it is assumed that the capsule could reach a tenperature of l G'5dF(bOOOC).
~
f Using tic ideal gas law and requiring the air to occupy a constant volume:
P PT
=
2 12 T1 2
initial pressure (0.3 033G/m )
p
=
y initial tenrerature (293 k)
T e
final temperature (1093 k)
=
The internal cas pressure could reach 0 377IG/m2 It is assm ed that the capsule can be treated as a thin-va31ed, cylindrical pressure vessel.
1629 285.
3-9 REVISION O NOL'. 2 9 w79
The r:aximum lon6itudinal tensile stress can be calculated by writing a longitudinal force balance through the weld:
=0 stress x area ~ E##3U"#* * "#
s "P
2 2
2 S v % - M_)
-NE t
=0 y
h where S1=
longitudinal stress outer diereter ( 6. 35cm)
D
=
o Di=
inner diereter ( 5 35cm) pressure (0 377!s/rr.2),
P
=
2 Tnu s, the longitudinal stress is 0 922'd/m The hoop stress can be found in a similar fashion.
Taking a longitudinal cress-section and suming forces:
=0 hoop stress x area - pressure x areap s
Lt - pD L = 0 2Sh i
where Sn' I " U 3
~
L = length of cylinder t = thickness of 921d (0.Shn m)
~
Tnu s, the hoop stress is 1 93pt;/m2 At a temr erature of 1600 F (870 C) the yield strength of type 30h stainles's steel is 10,000 psi ( 69.OM!i/ni )
Thus, the pressure induced 2
O stresses are less than 3% of the yield strength at 80C C.
1629 286
-2 REVISION O
- -10 NOV. 2 G 973
4.
Containment 4.1 Containment Eoundary 4.1.1 Containment Vessel The containment system for the Model T l Gamma Ray Projector is either h
Tech / Ops Model Ah24-9 or Model Ah24-18 Source Assembly. These source certified (IAEA Certificates of Competent Authority assemblies are currently /015h/S) as special form containment for radioactive Nos. USA /0165/S and USA materials.
The actual containment vessel is the velded source capsule, either styles 60004, 60011 or 60000.
Ti-capsules are made of Type 304 or 30hL stainless steel. They are seal velded with a minimum veld renetration of 0.020 inch
( 0 51==). The capsules are rounded cylinders 0.25 inches (6 35nn) in diameter and less than 1 inch (25.htm) in length. Appropriate descriptive drawings are enclosed in Section 2.10.
h.l.2 Contairment Penetrations There are no renetrations of contaimrent. The source ca psule is seal velded to provide conformity to sIecial form requirement'.
4.1 3 Seals and Welds The containment vessel is tungsten inert gas velded. This is done in accordance with Tech / Ops standard source encapsulation procedure (see section 7.4).
The minimum veld penetration is 0.020 inches (0 51==).
This has proved acceptable for licensing this vessel as special form.
h.1.4 Closure Not Applicable 4.2 Reauirements for Normal Conditions of Transport 4.2.1 Release of Radioactive Material The source assetblies used all meet the requirements of special form radio-active material as delineated in IAEA Safety Series No. 6,1973 and 10CTR71.
Thus, there vill be no releese of radioactive materials under conditions of normal trans5crt.
4.2.2 Fressurination of Containment Vessel The source assemblies used all meet the requirements of s;ecial form radioactive material.
Fressure buildup due to the conditions of the hypo-thetical thermal ace
- dent has been shown to create stresses well below the structural limits of the capsule (see Section 3 5). Thus, the con-tainment vessel vill withstand the pressure variations of normal transport.
h-1 1Z70 707 IUL/
LUI REVISION 6 NOV. 2 91979
h.2 3 coolant contamination Not Applicable 4.2.4 Coolant Ioss Not Applicable 4.3 containment.:equirements for the Itcpothetical Accident condition 4.3 1 rission cas Products Not applicable 4.3 2 Release or contents The hypothetical accident conditions as outlined in 10CFRT1, Appendbc B, l.,
2., and 3. have been shown (sections 2 71, 2 7 2 and 3 5 respectively) to result in no loss or package containment.
1629 288 k
l t
t r
REVISION O
{
h-2 N0l'. 2 G 1979
5 Shielding Evaluation 51 Discussion and Results The Model 741 is shielded with 200 pounds (91kg) of depleted uranium.
Tt e uranium metal is cast around the zircalloy or titanium "S" tube which holds The storage position for the source 1e at the inflection in the l
the source.
"S" tube.
j t
Radiation profiles of the Model 741 containing 33Ci of cobalt-60 and 240 Ci of iridium-192 (see Section 5 5) were made. The results are presented in Table 5 1.
From this data, and from previous acceptability (NRC Certificate of Compliance No. 9029, Rev. 2) it is concluded that the Model 741 complies with the regulatory standards in loCFR71 and IAEA Safety Series No. 6,1973 TABLE 5.1
SUMMARY
OF MAXIMJM DOSE RATES (mRlhr)
Contact At 1 Meter Side Top Bottom Side Top Ecttom Gamma 140 90 135 2.0 2.0 2.0 Neutron Not Applicable Not Applicable Total 140 90 135 2.0 2.0 2.0 Hypothetical accident conditions vill result in essentially no chan6e in the above readings.
52 Source Specification 5 2.1 Gamma Source f
The garna sources used are encapsulated cobalt-60 in quantities of up to 37 I
curies, or iridium-192 up to 240 curies.
r L
5 2.2
_ Neutron Source t
Not Applicable 53 Model S;ecifications Not Applicable 5.4 Shielding Evaluation The shielding evaluation was Ierformed on the Model 741 containinc 33 curies of cobalt-60 and 240 curies of iridium-192. The radiation profile is included in Section 5 5 Extrapolation of this data to the r:aximum capacity of the u
r 1
1629 289 L
5-1 BEVISION O NOV. 2 91977
package (Section 51) clearly indicates that the I:odel 741 confoms to regulatory radiation limits. As the hypothetical a*
ident evaluation (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 vill be within acceptable limits.
1629 290 I
5-2 REVISICII O NUV. 2 91,79
55 APPENDIX
- Model Thl: Badiation Profile 1629 291 5-3 REVISION 0 NOV. 2 91979
RADIATION PROFII2 Model 741 Serial Number 186 60 Containing 33 0 Curies of Cobalt (AN/PDR-27(J))
Iocation At Contact At 1 Meter Top 90 1.0 Bottom 135 2.0 Front 140 2.0 Rear 120 2.0 Left 115 2.0 Right 110 15 Model 741 Serial Number 150 Containing 2hD Curies of 192 Iridium (AN/PDR - 27(J))
Top 1.6 Iess than 1 Bottom 1.4 Less than 1 Front 2.8 Less than 1 Rear 23 Less than 1 Inft 1.6 Less than 1 Right 19 Less than 1 NOTES: All intensities are expressed in units of millircentgens per hour.
Intensities given are the maximum intensities on the measured side.
1629 292 5-4 REVISION O M V 2 9 1979
6.
Criticality Evaluation Not Applicable 1629 293 i
I e
6-1 REVISION O h
NOV. 2 91979
7 Operating Procedures 9
71 Procedures for Loading the Package Section 7.h describes the procedu-e for fabricating the special fom source encapsulation.
Section 7.4 also contains the pre,cedure for loading this source assembly into the package und preparing tne package for transport.
7.2 Procedures for Unloading the Package Section 7.h contains the procedure for unloading the cource assembly from the package.
73 Preparation of an Empty Package for Transport Section 7.4 describes the procedure for preparing an empty package for transport.
1629 294 i
t i
i t
7-1
{
REVISION O 6
fiOV 2 01979
7.h APPENDIX
- Encapsulation of Sealed Sources
- Technical Operations Model 741:
Procedures of Loading - Unloading the Package 1629 295 I
i REVISION O 7-2 f.?V 2 D G7s
" cr.r ?? _. ~
uI oo bat RADIATION SAFfI'Y PJJIUAL
~
Part II In Plant Operations Section 2 ENCAPSULATION OF. SEALED SOURCES A.
Personnel Requirements On}y an individual qualified as a Senior Radiological Technician shall perform the operations associated with the encapsulation of 192 Iridium. There must be a second qualified Radiological Technician available in the building when these operations are being performed.
B.
General Reauirements The 192 Iridium loading cell shall be used for the encapsulation of solid metallic 192 s such Iridium and the packaging of sealed souregO 169Ytterbiu=.
Solid metallic Cobalt as 170 Thulium, 137 Cesium and not exceeding one curie may be handled in this cell also.
The maximum amount of 192Iridium to be handled in this cell at aY s The maximum amount of 13 C one time shall not exceed 1000 curies.
to be handled in this cell at any one time shall not exceed 100 curies.
t This cell is designed to be operated at less than etmospheric pressure.
The exhaust blower provided shall not be turned off except when the cell is in a decontaminated condition.
Sources shall not be stor6d 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-wall" tools such as the velding fixture or transfer pigs are removed, the openings are to be closed with the plugs provided. These tools shall be decontaminated whenever they are removed from the hot cell.
C.
Preparatory Procedure l.
Check welding fixture, capsule drawer and manipulator fingers from cell and survey for contamination. ' If contamination in excess of 0.001,4 Ci of removable contamination is found, these items must be decontaminated.
2.
If the welding fixture or the electrodes have been changed, perform the encapsulation procedure omitting the insertion of any activity. Ihamine this du a capsule by sectioning thru veld.
Weld lenetration must be not less than 0.020 inch.
REVISION O 1629 296 11.2.1 2 9rre 7-3
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 measured by the manometer 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.
FYior to use, assemble and visually ins;ect the two capsule components to determine if weld zone exhibits any misalignment and/or separation.
Defective capsules shall be rejected.
2.
Decrease capsule components in the Ultrasonic Bath, using isopropyl alcohol as degreasing agent, for a period of 10 tinutes.
Ery the capsule components at 100 C for a minimum 0
of twenty minutes.
3 Insert capsule components into hot cell with the posting bar.
4.
Place capsule ' r we]d positioning device.
5 Nove drawer of source transfer container into hot cell.
6.
Place proper amount of activity in capsule.
Disposable-funnel must be used with pellets and a brass rivet with wafers to prevent contamination of weld zone.
7 Eemove unused radioactive material from the hot cell by with-drawing the drawer of the source transfer container from the cell.
8.
Remove runnel or rivet.
9 Assemble capsule components.
- 10. Weld adhuring to the following conditione:
lf}g }g7 a.
Electrode spacing.021" to.024" centered on joint +.002"; use jig for this purpose.
b.
Preflow argon, flush 10 seconds.
c.
Start 15 amps.
d.
Weld 15 amps.
e.
Slope 15 amps.
,l f.
Fest flow 15 seconds II-2-2 REVISION O 7_g r ev. 2 s h63
11.
Visually increet the veld.
An acceptable veld must be ecntinuous without cratering, cracks or evidence of blow out.
If the ve3d is defective, the capsule must be cleaned and revelded to acceptable conditions or disposed of as radioactive vaste.
12.
Check the capsule in height gau6e 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 must show no more than.005/4Ci of contamination.
If the patch shows more than.005 juCi the capsule must be cleaned and reviped.
If the revipe patc$ still shows more than 0.005 pCi of contamina-tion, steps 8 through 11 must be reIcated.
15 Vacuum bubble test the capsule.
Place t he velded capsule in a glass vial contrining isopropyl alcohol.
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 alcohol.
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 more than 1500 pounds.
- 17. ; Apply the tensile test to assesbly between the capsule and connector by applying proof load of T5 lbs.
extension under the load shall not exceed 0.1 inch.
If the extension exceeds 0.1 inch, the source must be disposed of as radioactive vaste.
18.
Position the source in the exit port of hot cell. Withdraw all personnel to the control area.
Use remote control to insert source in the ion chamber and position the source for maximum response.
Record the meter reading. Compute the activity in curies and fill out a temrorcry source tag.
19 Using remote control, eject the source from cell into source changer through the tube gauze wipe test fixture. Monitor before reentering the hot cell arma to 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 p Ci.
If contar_ination is in excess of this level, the source is leaking and shall be rejected.
20.
Complete a Source Loading Log (Figure II.2.1) for the operation.
1629 298 11.2 3 REVISION O 7-5 60U.0 9 :n>
Technical Operations Model Thl Procedures for Loading - Unloading the Tackage Wear personnel monitoring devices during all source changing procedures.
Mon-itor all operations with a calibrated, operable survey meter.
Note: All the precautions used when making radiographic exposures must be followed.
1.
Survey the projector to ensure that the source is in the proper po-sition.
2.
Locate the projector and source changer 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 914m), and the guide tube bead radius should not be less than 20 inches (0 508m).
3 Set the source changer for operation.
h.
Attach one end of a gu.de tube fitting to the fitting above the empty i
chanber in the source changer and the other end to the projector.
5 Attach the control cable to the projector:
a.
Unlock the projector with the key provided and turn the connector selector ring from the LOCK position to the COSNECT position. When the ring is in the CONVECT position, the storage cover vill disen-gage from the projector.
b.
Slide the control cable collar back and open the jaws of the svi-vel connector, exposing the rale portion of the connector.
Engage the male and female portions of the swivel connector by depressing the spring loaded locking pin toward the projector with the thu-b-nail.
Release the locking pin and test that the connection has been
- nade, c.
Close the jaws of the control cable connector over the swivel type connector, d.
Slide the control cable collar ov' - the conr.ector jaws.
Hold the control cable collar flush aEainst the projector connector and rotate the selector ring from the COT ECT position to the OPERATE position.
6.
Crank the source into source changer.
a.
Survey this operation with a survey reter to be sure the socree REVISION O ?-6 1629 299 nov. 2 s 13n
has been transferred from projector to changer.
b.
With a survey meter verify radiation level does not exceed 200 mR/hr at the surface of the changer.
7 Disconnect the control cable from the source assenbly.
Disconnect the guide tube from the source changer.
Secure the source in the source changer.
0-IF THE M CTOR TS TO REMAIN EIG W '
a.
Fully retract the control cable.
Disengage the control cable from the projector and lock the projector.
b.
Attach the identification plate of the source to the source changer.
Affix a green "etpty" tag to projector.
c.
d.
Perform a vipe test of the projector to assure that the radiation observed is less than 0.001 microcuries per 100 square centimeters.
e.
Survey the projector to assure that the radiation levels do not exceed 200mR/hr at the surface nor 101R/hr at three feet fro = the surface.
f.
Mark the projector:
Radicactive - "LSA".
Affix the proper ship-ping labels to the Iwekege.
g.
Complete the proper shipping papers as specified in Tech / Ops Radi-ation Safety Manual II.6.3E(4), (5 ), (6).
9 IF THE PRCJECTOR -IS TO EE RELOADED: connect the source changer end of the guide tuce to the fitting above the new source in the source changer.
10.
Crank source to full retraction within the projector.
a.
Survey this operation with a survey reter to be sure the source has been transferred into the projector.
b.
With e survey teter verify radiation level does not exceed 2CDar/hr at the surface of the projector.
11.
Disconnect the control cable and lock the projector.
12.
Disconnect the source guide tube from the projector and source changer.
13 Arfix the identification plate of the new source to the projector and attach the identification plate of the old source to the source chn :_ge r.
14.
Freiere for shipment:
a.
Again survey projector to insure that the radietion level does not REVISION O 1629 300 Sov 2 s is2, 7-7
exceed 200mr/hr at the surface of the projec. tor.
b.
Survey the radiation level at a distance of three feet from the surface of the projector.
This radiation level should not exceed 10mr/hr. The highest radiation level measured at three feet from the container is used to determine the Transport Index in accor-dance with 49CFRlT3 389(h).
c.
Affix the proper shipping labels.
1629 301 REVI5 ICE O SOV. 2 s 197s 7-8
8.
Acceptance Tests and Maintenance Program 8.1 Acceptance Tests 8.1.1 Visual Acceptance The package is visually examined to assure that the appropriate fastenere are seal wired properly and that the package is properly marked.
The seal veld of the radioactive source capsule is visually inspected for proper closure.
8.1.2 Structural and Pressure Tests The svage coupling between the source capsule and cable is subjected to a static tensile test with a load of seventy-five pounds.
Failure of this test will prevent the source assembly from being used.
8.1 3 Leak Tests The radioactive source capsule (the primary containment) is vipe tested for leakage of radioactive contamination. The source capsule is subjected to a vacuum bubble leak test. The capsule is then subjected to a second vipe test for leakage of radioactive contamination. These tests are described in Section 7.4.
Failure of any of these tests vill prevent use of this source assembly.
8.1.4 Conionent Tests The lock assembly of the package is tested to assure that security of the source vill be maintained. Failure of this test will prevent use of the package until the lock assembly is corrected and retested.
8.15 Tests for Shielding Integrity The radiation levels at the surface of the packaEe and at three feet from the surface are measured using a small detector survey instrument (e.g.,
AN/PER-27). These radiation levels, when extrapolated to the rated capacity of the package, must not exceed 200 milliroentgens per hour at the surface nor ten milliroentgens per hour three feet from the surface of the package.
Failure of this test will prevent use of the package.
8.1.6 Thermal Acceptance Tests 1629 302 Not Applicable 8.2 Maintenance Frogram 8-1 REVISION O NOV. 2 91979
I V
8 2.1 Structural and Pressure Tests 7
not Applicable i
8.2.2 Leak Tests As described in Section 8.13, the radioactive source assembly is leak tested f
at manufacture. Additionally, the source assembly is vipe tested for leakage
[
of radioactive contamination every six months.
8.2 3 Subsystem Maintenance The lock assembly is tested as described in Section 31.4, prior to each use of the package. Additionally, the package is inspected for tightness of fasteners, proper seal wires and Eeneral condition prior to each use.
8.2.4 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 not exceed 200 milliroent ens per hour at the surface E
nor ten milliroentgens per hour at three feet from the surface.
8.2.6 Thermal not applicable
- 8. 2. '(
Miscellaneous Inspections and tests designed for secondary users of this package under the general license provisions of 10CFR71.12(b) are included in Section 7.4.
1629 303 8-Z REVISIO;; 7 NOV. 2 91979 Id83[d