Text

Rev 3 to SAR Tech/Ops Rpd,Inc Model Ai 500 Su Type B(U) Package USA/9006/B(U)
	ML20203P568
Person / Time
Site:	07109006
Issue date:	03/31/1986
From:	; TECH/OPS, INC. (FORMERLY TECHNICAL OPERATIONS, INC.)
To:	;
Shared Package
ML20203P565	List: ML20203P564; ML20203P568;
References
	NUDOCS 8605070467
	Download: ML20203P568 (115)
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Insertion Instructions focyRevision 3 to Safety Analysis Report for 4

Tech / Ops RPD, Inc. Model AI 500 SU' Type B(U) Package USA /9996/B(D) t Remove Page Insert Page 1-3 Rev 2 1-3 Rev 3 1-4 Rev 0 1-4 Rev 3 1-5 Rev 0 1-5 Rev 3 1-6 Rev 0 1-6 Rev 3 1

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SAFETY ANALYSIS REPORT l

TECH /0PS RPD, INC.

l MODEL At 500 su TYPE B(U) PACKAGE 1

USA /9006/B(U) j i

Revision 3 t

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ID1IDdus11DD Tech / Ops, Inc. Model AI 500 SU is designed for use as a radiographic exposure device and a transport package for Type B quantities of radioactive material in special form.

The Model AI 500 SU conforms to the criteria for Type B(U) packaging in accordance with 10 CFR 71 and IAEA Safety Series No. 6, 1973 Revised Edition (as amended).

1.2 Rach3gg DREGriptiDD 1.2.1 Raghaging The Model AI 500 SU is 279 millimeters (11.00 inches) long, 124 millimeters (4.88 inches) wide, and 141 millimeters (5.56 inches) high.

The packaging incorporates a handle which extends 38 millimeters (1.5 inches) from the top surface and is 190 millimeters (7.5 inches) long.

The total mass of the package is 27 kilograms (65 pounds).

The radioactive material is sealed in a source capsule which conforms to the requirements for special form radioactive material.

This source capsule is installed onto a source holder assembly.

The source holder assemblies used in conjunction with the Model AI 500 SU are listed in the appendix.

The source holder assembly is housed in one of two titanium source tubes.

Each source tube has an outside diameter of 11.1 millimeters (0.438 inch) and an inside diameter of 9.5 millimeters (0.375 inch).

One end of the source tube is enclosed by means of a plug which is soldered to the source tube.

The opposite end of the source tube is inserted in the lockbox which is welded to the front plate.

The source tubes are surrounded by uranium metal as shield-ing material.

The uranium shielding is cast in place around the source tubes.

The mass of the uranium shield is 18 kilograms (39 pounds).

The uranium shield is encased in a Type 304 stainless steel housing.

The housing is 3.4 millimeters (0.135 inch) thick.

Mounted on the f ront plate is the lockbox assemblies.

The locking assembly is used to secure the radioactive source and source holder assembly in the shielded position during transport.

l 1-1 Revision 0 1 March 1985 O

Surrounding the lockbox assembly is the protective cover..

This cover is fabricated from Type 304 stainless steel and' O-is attached to the package shell by means of a hinge and a

seal bolt.

The head of the seal bolt is drilled to provide a means for i

attaching tamperproof seal wire during transport.

On certain

packages, additional shitiding plates are installed to reduce the surface radiation level.

These plates are 6.4 mm (0.25 inch) thick and welded to the out-t side of the package as described on drawing AI 500 SU 93, The outer packaging is designed to avoid the collection and retention of water.

The package has a

smooth, unpainted 4

stainless steel finish to provide for easy decontamination.

The radioactive material is sealed inside a stainless steel source capsule.

This capsule acts as the containment vessel for the radioactive material.

1.2.2 Dpgrat19Dal EgDLuxgs The source holder assembly is secured in the proper shielded storage position by means of the locking assembly.

The source assembly is held in the shielded storage position by means of a

lock slide.

The lock slide creates an interference with the stop ball, preventing its movement out O

of the shielded. position.

The lock slide is held in position by a latch pin and by a key operated lock.

The 4

lock slide cannot be moved from the lock position until both

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the key operated lock is unlocked and the latch pin is depressed by the insertion of a source guide tube into the lock body.

Additionally, a cap is installed over the source l

holder assembly and attached to the lockbox.

This cap is seal wired to prevent inadvertent loosening.

The cap also 4

provides an additional means to prevent movement of the lock slide.

An outer protective cover is installed over the lock box and is locked in the closed positin to prevent unauthorized access.

1.2.3 CDDhgDhs of kbg RaGkB99 The Madel AI 500 SU is designed for the transport of iridium-192 in quantities up to 120 curies in Tech / Ops source assemblies listed in the appendix.

The source cap-sule used with each of these source assemblies satisfies the criteria for special form radioactive material in accor-dance with 10 CFR 71 and IAEA Safety Series No.

6, 1973 Revised Edition (as amended).

1-2 Revision 2 8 August 1985 s

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1.3 APPENDIX i

i Drawing Al500SU90 Sheets 1 through 7 i

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i Drawing Al500SU91 Sheets I and 2 1

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2.0 Etructural EsaluatiDD 2.1 Structural Danign 2.1.1 Discuss 19D The Model AI 500 SU is comprised of five structural compo-nents:
a source capsule, source holder assembly, shield assembly, outer housing assembly and locking assembly.
The source capsule is the primary containment vessel.
It satis-fies the criteria.for special form radioactive material.
The shield assembly provides shielding for the radioactive material and, together with the source holder assembly and locking assembly, assures proper positioning of the radio-active source.
The outer housing is f abricated f rom 3.4 millimeter (0.135 inch) thick Type 304 stainless. steel.
The housing provides the structural integrity of the package.
The package outer cover, also fabricated from 3.4 millimeter (0.135 inch) Type 304 stainless steel provides protection to the lockbox assembly.
The lockbox assembly secures the source holder assembly in the shielded position at the bottom of the source tube and assures positive closure.
2.1.2 Design Exitaria The Model AI 500 SU is designed to comply with the require-ments for Type B(U) packaging as prescribed by 10 CFR 71 and IAEA Safety Series No. 6, 1973 Revised Edition (as amended).
l All design criteria are evaluated -by a straightforward application of the appropriate section of 10 CFR 71 or IAEA Safety Series No. 6, 2.2 Enights and CsDhaza 91 Gxaxity l
The total mass of the Model AI 500 SU is 30 kilograms (65 pounds).
The shield assembly consists of 18 kilograms (39 pounds) of depleted uranium.
The center of gravity was located experimentally.
It is located 90 millimeters (3.56 inches) f rom the rear, 70 millimeters (2.75 inches) above the bottom surf ace, and 62 millimeters (2.44 inches) from the right surface.
2-1 Revision 0 1 March 1985 l
- -..,.-.-.-,-.~.--..:
2.3 BEGhaDiCal RE992EkiRS 92 BRh2ElB1B The outer housing of the Model AI 500 SU is fabricated from Type 304 stainless steel.
This material has a yield strength of 207 MPa (30,000 psi).
Drawings of the source capsules used in conjunction with the Model AI 500 SU are enclosed in Section 2.10.
These source capsules are f abricated f rom either Type 304 or Type 304L stainless steel.
2.4 GBDRED1 SkDDdDEdB L9L b11 ERGhDSBS 2.4.1 CbRaical DDB GalyaDiG EBRGhiDDR The materials used in the construction of the Model AI 500 SU are uranium metal, stainless steel, brass, titanium, and epoxy.
There will be no significant chemical or galvanic action between any of these components.
The possibility of the formation of the eutectic alloy iron uranium at temperatures below the melting temperatures of the individual metals has been considered.
The iron uranium eutectic alloy temperature is approximately 725 C (1337 F).
However, vacuum conditions and extreme cleanliness of the surfaces are necessary to produce this alloy at this low O
temperature.
Due to the conditions in which the shield is mounted in the Model AI 500 SU, sufficient contact for this
'i effect would not exist.
In support of this conclusion, the following test results are presented.
On 28 November 1973, a thermal test of a sample of bare depleted uranium metal was performed.by Nuclear Metals, Inc., Concord, MA.
The sample was placed in a ceramic crucible and inserted in a furnace preheated to 800 C (1475 F) and remained there for thirty minutes.
The sample was then removed and allowed to cool.
The test indicated that the uranium sample oxidized such that the radial dimension was reduced by 0.18 millimeters (0.007 inch).
On 25 January 1974, a subsequent test was performed by Nuclear Metals, Inc.
In this test, a sample of bare depleted uranium metal was placed on a eteel plate and subjected to the thermal test conditions.
The test revealed no melting or alloying characteristics in the sample and the degree of oxidation was the same as experienced in the earlier test.
2-2 Revision 0 1 March 1985 l
Notwithstanding these test results, there are not iron-uranium interfaces in the Model AI 500 SU.
2.4.2 292.1%iy2 C19DK12 The source assembly in the Model Ai 500 SU cannot be moved to an unshielded position without movement of a
holddown cap.
The holddown cap is held in the engaged position by a
seal wire.
Access to the holddown cap requires opening of the outer cover.
The outer cover is bolted to the package and the bolt is padlocked in the engaged position.
Additionally, the source assembly is held in the shielded storage position by means of a lock slide which is secured by an independent key operated lock.
Therefore, positive closure is maintained during transport.
2.4.3 L1111D9 Dayls22 The Model Ai 500 SU is designed to be lifted by its handle.
The handle is attached to the package by means of handle bails at each end of the handle.
A static tensile test of this handle arrangement was made.
A report of this test is included in section 2.10.
This test demonstrates that the lifting device can support five times the weight of the package without exceeding the yield strength of the material.
2.4.4 TiedswD Dsyisms
(
The Model AI 500 SU can be tied down by means of the handle.
An analysis of this tiedown arrangement under the load conditions of 10 CFR 71.45 (b) is presented in Section 2.10.
This analysis demonstrates that the maximum stress generated in the handle would be less than the yield strength of the material.
Therefore, we conclude that the Model AI 500 SU satisfies the tiedown condition of 10 CFR 71.45 (b) (1).
Additionally, if the handle were to fail under excessive
load, the ability of the package to maintain its structural integrity and shielding efficiency would not be impaired.
Therefore, the package tiedown design satisfies the criteria of 10 CFR 71.45 (b) (3).
2.5 S%DDDDIBS 191 Zy29 B RDGhD9tB 2.5.1 L9aC EssistaDan Considering the package as a simple beam supported on both 2-3 Revision 2 8 August 1985 O
ends with a uniform load of five times the package weight O
evenly distributed along its length, the maximum stress can j
be computed from F1 Cmax =
8z where 7 max = maximum stress F
assumed load (1448N or 325 pounds) kngth of beam (279mm or 11.00 inches) 4 3
3 z
Sa; tion Modulus (76,051mm or 4.67 in )
(Ref: Machinery's Handbook, 22nd Edition, p. 294 & 261)
The load is assumed to be 1,448 newtons (325 pounds).
The container is assumed to be a rectangular solid 141 milli-meters (5.56 inches) high, 124 millimeters (4.88 inches) wide, a wall thickness of 3.4 millimeters (0.135 inch) and a length of 279 millimeters (11.0 inches).
Consequently, the 3
3 section modulus of the beam is 76,051 mm (4.67 in
).
Therefore, the maximum stress generated in the beam under these conditions would be 0.135 MPa (95 lb/in*) which is far below the yield strength of the material.
f 1
2.5.2 ExtexDal Exgssure
[L The Model AI 500 SU is open to the atmosphere.
Thus there will be no differential pressure acting on it.
The collap-sing pressure of the source capsule is calculated assuming that the capsule is a thin wall tube with a wall thickness equal to the minimum depth of weld penetration which is 0.5 millimeter (0.020 inch).
The collapsing pressure is calcu-lated from:
P = 597.6 t/d - 9.556 where P: Collapsing Pressure in MPa t: Wall Thickness (0.5mm or 0.02 inch) d: Outside Diameter (6.4 mm or 0.250 inch)
(Ref: Machinery's Handbook, 22nd Edition, p. 330)
From this relationship, the collapsing pressure of the source capsule is calculated to be 37.1 MPa
(." 5 4 8 ps i).
Therefore, the scurce capsule could withstand an external pressure of 0.17 MPa (25 psi).
2.6 59xmal CDDslikiDDR DL 2xnDspDxt 2-4 Revision 0 1 March 1985 O
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2.6.1 Beat O'
The thermal evaluation of the Model AI 500 SU is presented in Section 3.
From this evaluation, it is concluded that the Model AI 500 SU will maintain its structural integrity and shielding effectiveness under the normal transport heat condition.
2.6.2 Cold The metals used in the manuf acture of the Model AI 500 SU can all withstand a temperature of -40*C (-40*F).
The outer package housing and the primary containment are all fabri-cated f rom Type 304 stainless steel.
As stated in Draft Regulatory Guide, Division 7, Task MS 144-4, austenitic stainless steels are not susceptible to brittle fracture at temperatures encountered in transport.
The epoxy used in the Model AI 500 SU have an operating temperature range of -43*C to 104 C.
From this data, it is concluded that the Model AI 500 SU will maintain its struc-tural integrity and shield!.ng effectiveness under the normal t.ransport cold condition.
2.6.3 Endy.ggd Ernssuza An external pressure test of the Model AI 500 SU was conduc-ted by Automation Industries, Inc.
The device was placed in O'
a pressure chamber and the pressure was reduced to 0.25 kg/cm and maintained its shielding efficiency and struc-4 tural integrity.
A report of this test is presented in a letter to Mr. R. Raul f rom Mr. M. Santoro dated 4 December 1981 and included in Section 2.10.
A demonstration of the ability of the source capsules to withstand an external pressure of 0.5 atmosphere is presented in Section 3.5.4.
On the basis of these data, it is concluded that the Model AI 500 SU will maintain its structural integrity and shield-ing effectiveness under the normal transport pressure condi-tion.
2.6.4 YlbIA119D i
The Model AI 500 SU has been in use for more than ten years.
In this period, there has been no evidence of vibration-induced failure.
4 2-5 Revision 0 1 March 1985 O
t I
,--_n
On the basis of this history, it is concluded that the Model AI 500 SU will maintain its structural integrity and fsv) shielding effectiveness under the normal transport vibration condition.
2.6.5 WaigI Spray The water spray test was not acually performed on the Model AI 500 SU.
The materials used in the construction of the Model AI 500 SU are highly water resistant.
Therefore, it is concluded that the Model AI 500 SU will maintain.its structural integrity and shielding effectiveness under the normal transport water spray condition.
2.6.6 EIgg Disp A prototype Model AI 500 SU was subjected to the hypotheti-cal accident free fall condition.
This is described in Section 2.7.1.
On the basis of this test, it is concluded that the Model AI 500 SU will maintain its structural inte-grity and shielding effectiveness under the normal transport free drop condition.
2.6.7 CoragI Disp Not applicable 2.6.8 EgnetratiDD A prototype Model AI 500 SU was subjected to a penetration test by Automation Industries. The package was impacted by the penetration bar in three different attitudes.
As a result of these impacts, there was no loss of structural integrity nor reduction of shielding efficiency.
A report of this test is presented in a letter to Mr. B.
Rawl f rom Mr. M. Santoro dated December 1981 and included in Section 4
2.10.
On the basis of this test it is concluded that the Model AI 500 SU will maintain its structural integrity and shielding effectiveness under the normal transport penetration condi-tion.
2.6.9 CpapIsssign A prototype Model AI 500 SU was subjected to the compression condition.
The total mass of the. package is 30 kilograms 2-6 Revision 0 1 March 1985 v
(65 pounds).
The maximum cross sectional area of the
- O package is 0.04 m ' ( 61. 2 in' - ).
Thus, five times the weight of the package (1448 newtons or 325 pounds) is greater than 13.8 KPa (two pounds per square inch) times the maximum cross sectional area (545 newtons or 123 pounds).
I The package was subjected to a compressive load of 1515 newtons (340 pounds).
The load was applied for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
At the conclusion of this test, there was-no loss of structural integrity nor reduction of shielding efficiency.
A report of this test is presented in a letter to Mr. R. Rawl from Mr. M Santoro dated 4 December 1981 and on the basis of this test, it is concluded that the Model AI 500 SU will maintain its structural integrity and shielding effectiveness under the normal transport compression condition.
' 2.7 Hypothetical Accident Conditions 2.7.1 Free Drop The Model AI 500 SU was subjected to the conditions of the free drop by Automation Industries, Inc.
The O
tareet used in this free aron te t ooa i tea of o11d concrete apron with a thickness of 0.2 meters (8
inches).
A steel plate with a
thickness of 16 millimeters (0.63 inches) was placed in intimate contact with the concrete apron.
During the test, the package fell from a height of 10.3 meters (34 feet) onto the target.
As a result of this
test, there was no loss of structural integrity nor loss of chielding efficiency.
A_ report of this test by Mr. M.-Santoro dated 24 July 1973 is presented in Section 2.10.
On the basis of l
this test, it is concluded that the Model AI 500.SU will maintain its structural integrity and shielding t
effectiveness under the hypothetical free drop accident condition.
Additionally, a Model AI 500 SU package with a modified i
locking arrangement was twice subjected to the conditions of the Free Drop Test by Tech / Ops RPD,_Inc.
A report of this test is included in Section 2.10.
On the basis of this test, it is concluded that the Model I
AI 500 SU package with the modified locking arrangement will maintain its structural integrity and shielding 2-7 Revision 3
~
_ _31 yarch 1986
efficiency under the hypothetical free drop accident condition.
2.7.2 Puncture At the conclusion of the free drop test, the prototype Model AI 500 SU was twice subjected to the puncture condition by Automation Industries.
The target for-the puncture test was a steel billet 76 mm (3 inches) in diameter and 203.mm (8 inches) high mounted on the target used in the free drop test.
During this test, the package dropped from the height of one meter (40 inches) onto the billet.
As a
result of this
test, there was no loss of structural integrity nor reduction in shielding efficiency.
A report of this test by Mr.
M.
Santoro dated 24 July 1973 is presented in Section 2.10.
on the basis of these tests, it is concluded that the Model AI 500 SU will maintain its structural integrity and shielding effectiveness under the hypothetical puncture accident condition.
l Additionall', a Model AI 500 SU package with a modified y
locking arrangement was three times subjected to the O
ccaaitic er the ru=cture re t av reca<or
=>o, c-1 A report of this test is included in Section 2.10.
On the basis of this test, it is concluded that the Model.
AI 500 SU package with the modified locking arrangement will maintain its structural integrity and shielding efficiency under the hypothetical puncture accident condition.
l 2.7.3 Thermal l
At the conclusion of the free drop and puncture tests, i
the prototype Model AI 500 SU was subjected to the thermal condition by Automation Industries.
The package was placed into a kerosene and fuel oil fire.
The package remained there for 47 minutes.
At - the I
conclusion of the test, the fire ' was extinguised and the package was allowed to cool only by natural convection and radiation.
No artificial cooling methods were used.
As a
result of this
test, there was no loss of structural integrity nor reduction in shielding efficiency.
A report of this test by Mr.
M.
Santoro dated 24 July 1973 is presented in Section 2.10.
On 2-8 Revision 3 31 March 1986
>)
Q the basis of this test, it is concluded that the Model V
N AI 500 SU will maintain its structural integrity and shielding effectiveness under the hypothetical thermal accident condition.
2.7.4 Water Inumersion Not Applicable.
2.7.5 Sununary of Damage The tests designed to induce mechanical stress (free
drop, puncture) caused minor deformation but no reduction in structural integrity nor impairment of any safety features.
The thermal test had no adverse affect on the package.
As a result of these tests, there was no loss of structural integrity nor release of any contents.
Prior to the conduct of these tests and subsequent to the conduct of these
tests, measurements of the radiation intensity in the vincinity of the package were made.
The results of these measurements demonstrate that there was no reduction in shielding O
erricieocv as a resu1e or taese tests-2.8 Special Form The Model AI 500 SU is designed to transport Tech / ops source capsules.
These source capsules have been certified as a special form radioactive material under IAEA Certificate of Competent Authority Number USA /0154/S, USA /0279/S, and USA /0335/S.
These certificates are presented in Sect. ion 2.10.
2.9 Fuel Rods Not applicable.
O 2-9 Revision 3 31 March 1986
2.10 Appendix IAEA Certificate of Competent Authroity USA /0154/S lAEA Certificate of Competent Authority USA /0279/S IAEA Certificate of Competent Authority USA /0333/S Drawing 60050 Drawings SK2332-7, -8,
-9, -10 Test Report: Model Al 500 SU Tiedowa Arrangement Letter to Mr. R.
Rawl, U.S.
Department of Transportation i
from Mr.
M. P.
Santoro, Automation Industries, Dated 4 December 1981, describing the results of the Model 500 SU Penetration Reduced Pressure and Compression Tests.
Automation Industries Test Report by Mr.
M.
P.
Santoro, dated 24 July 1973, describing the results of the Model 500 SU Free Drop, Puncture and Thermal Tests.
Test Report: Model Al 500 SU Free Drop Test Test Report: Model Al 500 SU Puncture Test O
.w:
w:
I O
2-10 Revision 3 31 March 1986-
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+
US Decotment
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V Research and speclo w & M lAEA CERTIFICATE OF COMPETENT AUTHORITY Administration Special Form Radioactive Material Encopsulation Certificate Number USA /0154/5 (Revision 4)
This certifies 4 hot the encapsulated sources, os described, when loaded with the authorized radioactive contents, have been demonstrated to meet the regulatory requirements for special form radiooctive materials os prescribed in IAEA l/ and USA 2/ regulations for the transport of radioactive materials.
1.
Source Description - The source capsules described by this certificate are identified as the Technico! Operations, Inc., Models which are described and j.
concructed as follows:
Copsule Model Approximate Size (in inches, diameter x length) 60001.
.25 x.97 60004-
.25 x.97 60006 Pellet, Wofer or Large
.25 x.90 Wofer 68310 Pellet or Wofer
.25 x.78 O
~~
60017 i
.25 x.97 60018
.25 x.97 60020
.25 x.97 60021
.25 x.97 All capsules are constructed of either 304 or 304L stainless steel and conform with the following design drawings:
Copsute Model Drawing Number 60001 6000) - 1, Rev. M and 60001-5
.60004 60004 - 1, Rev. E and 60004 - 2 50006 Pellet 60001 - 5 and 60006 - 3 60006 Wofer 60006 - 1, Rev. H and 60004 - 1, Rev. E 60006 Large Wafer 60006 - 1, Rev. H and 60001 - 5, Rev. F 68310 Pellet 68310 - 9 and 68310 - 10, Rev. A 68310 Wofer 68310 - 1, Rev. F and 68310 - 2, Rev. G 60017 60017 - 1, Rev. A and 60001 - 2 60018 60017 - 1, Rev. A and 60004 - 1, Rev. E 60020 60020 - 3 and 60001 - 5 60021 60020 - I and 60004 - 1, Rev. E il.
Radioactive Contents - The authorized radioactive contents consist of metallic fridium-192 with not more than 240 Curies in models 60001, 60004, 60006 Pellet, Wofer and Large Wofer or 120 Curies in models 60017, 60018, 60020, 60021, 68310 Pellet and Wofer.
Revision 0 2-11 1 Mar 1985 1
4 i
e
-.n.-,-.
2 N
Certificate Number USA /0154/5, Revision 4 111.
This certificate, unless renewed, expires December 31,1989.
This certificate is issued in occordance with parograph 803 of the IAEA Regulations I/ and in response to the November I,1984, petition by Technical Operations, Inc.,
Bu,rlington, Mossochusetts, and in consideration of the associated information therein.
Certified by:
/0B0nn Ahoaht.9 /9W Richard R. Rowl (Date)
~
Chief, Rodioactive Materials Bronch Office of Hazardous Materiots Regulation Materiots Transportation Bureau y "Sofety Series No. 6, Regulations for the Safe Transport of Radioactive Materials, 1973 Revised Edition", published by the International Atomic Energy Agency (IAEA),
Vienna, Austrio.
2/ Title 49, Code of Federal Regulations, Ports 170-178, US A.
Revision 4 - incorporated new drawing nos; extended expiration date.
O#
Revision 0 2-12 1 Mar 1985 4
of Transportaton
w
0 f
IAEA CERTIFICATE OF COMPETENT AUTHORITY Special Form Radioactive Material Encuesulation Certificate Number USA /0279/S Revision 0 This certifies that the encapsulated sources, as described, when loaded with the authorized radioactive contents'have been demonstrated to meet the regulatory requirements for special form radioactive material as prescribed in IAEA I/ and USA 2/ regulations for the transport of radioactive materials.
I. Source Descriotion - The sources described by this certificate are identified as Automation Industries Models 500-W8 and 500-W10 which are tungsten-inert-gas welded 316 stainless steel encapsulations which measure 0.25 inches (G.4 mm) in diameter by 0.75 inches (19 mm)in length (AI drawing 500-W8 & W10).
II. Radioactive Contents - The authorized radioactive contents of these sources consist of not more than 300 curies of Iridium-192 metal wafers.
III. This certificate, unless renewed, expires April 30,1988.
This certificate is issued in accordance with paragraph 803 of the IAEA' Regulations 1/, and in response to the October 19, 1983 petition by Automation O
Industries,Inc. Phoenixville, PA and in consideration of the associated information therein.
Certified by:
Y?fGM l0hb Richard R. Rail (bate) U
~
Chief, Radioactive Branch Office of Hazardous Materials Regulation Materials Transportation Bureau 1/ " Safety Series No. 6, Regulations for the Safe Transport of Radioactive
?.laterials,1973 Revised Edition", published by the International Atomic Energy Agency (IAEA) Vienna, Austria.
2/ Title 49, Code of Federal Regulations, Part 170-178, USA.
Revision 0 2-13 1 Mar 1985 O
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W Ub Devo tment
'..- " s e.. '. v.
of 1ro wportoton
' '"* l ' (
e
Research ond IAEA CERTIFICATE OF COMPETENT AUTHORITY 5pecw Progroms Administration 5 ccial Form Radioactive Materials Encapsulation Certificate Number USA /0335/5 Revision 0 This certifies that the encapsulated source, as described, when loaded with the authorized radioactive contents, has been demonstrated to meet the regulatory requirements for special form radioactive material as prescribed in IAEA 1/ and USA 2/ regulations ior the transport of radioactive materials.
1.
Source Description - The source described by this certificate is identified as Tech Ops Model 875 source capsule assembly which is a single, welded encapsulation constructed of type 304 or 304L stainless steel and which measures approximately 24 mm (0.95 inch) in length by 6.4 mm (0.25 inch) in diameter. Contents may be f urther contained in stainless steel or titanium inner secondary encapsulations along with springs and spacers.
11.
Radioactive Contents - The authorized radioactive contents of this source consist of not more than:
Radionuclide Maximum Activity O
Ir-192 240 Ci Co-60 220 Ci Yb-169 200 Ci Cs-137 TO Ci Tm-170 30 Ci 111.
This certificate, unless renewed, expires Jul'y 15,1989.
This certif.icate is issued in accordance with paragraph 803 of the IAEA Regulations 1/, and in response to the May 22, 1984 petition by Tech Ops, Burlington, MA and in consideration of the associated information therein.
Certified by:
5 mf lan U /W4 Richard R. Rawl (gate)(
Chief, Radioactive Materials Branch Office of Hazardous Materials Regulation Materials Transportation Bureau 1/ " Safety Series No. 6, Regulations for the Safe Transport of Radioactive Materials, 1973 Revised Edition", published by the International Atomic Energy Agency (IAEA),
Vienna, Austria.
2/ Title 49, Code of Federal Regulations, Part 170-178, USA.
Revision 0 2-14 1 Mar 1985
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e TEST REPORT RADIATION PRODUCTS DIVISION By:
George W. Parsons'.I, f.
John J. Munro III Date:
21 March 1985-
Subject:
Model AI 500 SU Package Tiedown Test A Model AI 500-SU Type 8(U) Package, Serial Number-572. was subjected to the tiedown load standard of 10 CFR 71.45(b).
The package was tied down by means of cables attached'to each side of each ball connecting the handle to the-package.
^
A force of ten times the weight of the package (650 pounds) in the hori:<,ntal direction in which the vehicle travels and a force of five times the weight of the package (325 pounds ) in the transverse hori: ental direction. produce a resultant hori: ental force of 738 pounds.
It was concluded that the most severe application of this force was along-the direction of the long ax,is of the package. The actual hori: ental force applied in this direction during the test was 750 pounds. The actual vertical force applied to the package during this test was 325 pounds or five times the weight of the package.
The package was subjected to these forces for one hour.
At the conclusion of this test, there was no failure of any component of the
(
package or o f ! P, tiedowr. arrangement.
There was no evidence of any yield of any component.
Thers 7re, it Is concluded that the package can withstand the
.own load condittons without generating any stress in excess of the yield ngth of the material.
Revision 0 2-20 1 Mar 1985 C(
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Revision 0 2-21 1 Mar 1985
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TEST. REPORT.
RADIATION PRODUCTS DIVISION f
Sy George W. Parsons,h.h John J. Munro III Date:
21 March 1985
Subject:
Model AI'500'SU Package Lifting Test 4
A Model AI 500 SU Type 8(U) Package, Serial Number 572. was subjected to the
,i.
lif ting load standard of 10 CFR 71.45(a).
The package was secured to a-platform, with a mass in excess of ten times the mass of the package, by means i
of cables attached to each side of each bail that connects the handle to the package.
I 1
A force in e.< cess of three times the weight of the package (250 pounds) was l
applied to the package handle in the vertical direction.
This force was applied for one hour.
I At the conclusion of this test, there was no failure of any component of the i
package or of the handle. There was no evidence of any yield of any component.
Therefore it is concluded that the package can withstand the lifting load l
conditions without generating any stress in excess of the yield strength of the material.
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AUTOMATION INDUSTRIES, INC.
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SPERRY PRODUCTS DIVISION r's P.O. Box 245 (V
)
PHOENIXVILLE. PA.19460
'l (215) 933 8961 December 4, 1981 Mr. R. R. Rawl Office of Hazardous Materials Regulation Materials Transportation Bureau U.S. Department of Transportation 400 Seventh Street, S.W.
Washington, D.C.
20590 Re: Application For B(U) Certification Of Two (2) Of Automation's Existing Type "B" Packages Having Assigned Identification Numbers As Follows:
USA /9006/B Model 500-SU, IR-192 Source Changer USA /9007/B Model 520 Iriditron Exposure Device
Dear Mr. Rawl:
O<
The above referenced packages were originally certified as Type "B" packages by your office in 1973. Automation Industries has recently subjected both of these packages to the basic and specific additional test requirements for Type B(U) packages. All additional tests were performed in accordance to, and in sequential order as outlined in the IAEA Safety Series No. 6, 1973 Revised Edition (as auended 1979).
I.
SPECIAL NOTES AND CONDITIONS RELATING TO ADDITIONAL TESTS:
A.
There have been no design changes to either package such as size, mass, structural configuration, or shielding medium since the original Type "B" approvals were issued in 1973.
B.
Each package was subjected to the additional tests separately; and during each test, the test package contained the same en-capsulated source of Iridium-192.
C.
The contained source of Iridium-192 had a calibrated strength of 17.0 curies on September 1, 1981, and is identified by Automation's serial no. IR-14361.
D.
One test specimen of each package design was used throughout all of the additional test requirements.
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Revision 0 2-23 1 Mar 1985 O
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Mr. R. R. Rawl Page December 4,1981 E.
Each test package was surveyed for external surface radia-tion levels and radiation levels measured at one (1) meter from any external surface; both prior to, and after each test. To further insure structural and containment integ-rity, an intimate wipe test of the contained IR-192 sealed source and container "S" or "J" tube was performed both prior to and after each test.
Af ter the WATER IMMERSION TEST, Sect. VII, Par. 721, the residual water after each 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months hydrostatic test was col-lected and the volume determined. Two (2) 10ml samples of each batch were assayed by well-counting for a 10 minute period. This procedure was to further assure the integrity of containment.
After completion of all tests covering additional requirements for Type B(U) packaging, the contained source of Iridium-192, serial no. IR-14361 was further leak tested to meet the require-ments of Appendix "A" of ANSI Standard N542, Sections:
A2.1.1. Wipe (Stear) Test A2.1.3. Immersion with Boiling Test.
V NOTE: All of Automation's sealed sources of Iridium-192 have an internal void volume of less than 0.10 ml.
F.
Considering the materials of construction, compact design con-figuration, and relatively light mass of both packages, some of the additional test (such as water spray or internal heat generation) were omitted by reasoned consideration or where con-servative and reliable engineering judgement clearly obviates the need for testing. These particular tests will be referenced in this report with the notation-- " Test Not Required".
Reference to the cumulative ef fects of the mechanical tests and thermal test performed in 1973 for qualifying as Type "B" Pack-ages will be included in this report with the notation-- "Com-pliance Demonstrated, 1973".
l G.
A set of photographs will be included as part of this test report to illustrate the various equipment and apparatus used in per-forming the applicable tests.
l H.
All of the additional tests for B(U) packaging were performed during the period September 1, 1981 through September 11, 1981.
Revision 0 2-24 1 Mar 1985 l O
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1 Mr. R. R. Rawl Page December 4, 1981 II.
EQUIPMENT AND INSTRUMENTS EMPLOYED A.
TEST EQUIPMENT 1.
Vacuum Pump Welch Duo Seal Model 1400, S/N 112839 2.
Hydrostatic Pump Neptune Model HP-1, 750 PSIG Max.
3.
Pressure Vessel Fabricated from 8" std. vt.,
sch. 40, seamless steel pipe, fitted with 300 lb.
RF slip-on blind flange as removable top, and two pressure taps.
Internal Dimensions: 7.981" I.D. X 14" high 4.
Dead weight for compression test.
Cylindrical, lead ahielded, Cobalt-60 transport container. Container S/N SC-204 Gross weight " Empty": 340 Lbs.
5.
Cylindrical Steel Bar for Penetration Test Bar diameter: 3.2 cm (1.25")
Bar weight: 6.0 kg (13.25 Lb.)
Bar length: 97.2 cm (38.25")
Striking end: Hemisherical B.
TEST INSTRUMENTS 1.
Survey Meter Radiac Set, Model 68-27R S/N I-130, Calibrated 6-5-81 & 9-7-81 2.
Well Counter For leak test analysis Eberline Mini Scaler Model MS-2, S/N 151 with bicron 2"X2" crystal
& 1" diam. X 1-1/2" deep cavity Bicron Model 2MW2-PQ 3.
Rate Meter For calibrating source Victorcen Model 555 S/N 279.
Revision 0 s/
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Mr. R. R. Rawl Page December 4, 1981 4.
Pressure Gauge Marshaltown #G27477 2-1/2" Dial 0-400 PSIG Calibration: ANSI Grade B, 2%
5.
Vacuum Gauge Marshaltown #G14489 2-1/2" Dial, 0-30" Hg. Vac.
Calibration: ANSI Grade F
'l%
6.
Various laboratory supplies such as hot plate, beakers, burettes, syringes, radiac wash, dis-tilled water, rinse bottles, etc...for leak test-ing and evaluating radioactive concentrations.
C.
TEST SPECIMENS 1.
Model 500-SU Source Changer Type "B" Certificate No. USA /9006/B Serial No. 670 2.
Model 520 Iriditron, Exposure Device Type "B" Certificate No. USA /9007/B gj Serial No. 831 3.
Scaled Source of Iridium-192 Encapsulation: Special Form Calibrated Strength: 17.0 Curies (9-1-81)
Automation's S/N IR-14361 III. RADIATION SURVEYS OF PACKAGES A 17 curie source of Iridium-192 was inserted into the Model 500-SU source changer and also into the Model 520 Iriditron. Both packages were surveyed for surface radiation levels and for rad-iation levels measured at one (1) meter from any external surface of the package. These surveys were performed prior.to subjecting the packages to the basic and specific additional tests for B(U) packaging. Results of these initial surveys also established a baseline for comparison of radiation levels and integrity of con-tainment prior to and after each sequential test. Results of these initial " baseline" surveys are tabulated below:
A.
Model 500-SU source changer, S/N 670 Results of " baseline" radiation surveys Area Surveyed mrem /h at Surface mrcm/h at 1 Meter L. Side 2.5 0.1 R. Side 11.0 0.15 Front 1.5 0.1 O
Rear 10.5 0.15 b
Top 2.3 0.1 Bottom 9.5 0.1 Revision 0 2-26 i Mar 1985
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Mr. R. R. Rawl Page M December 4,1981 B.
Model 520 Iriditron, S/N 831 Results of " baseline Radiation Surveys Area Surveyed mrem /h at Surface mrem /h at 1 Meter L. Side 26.0 0.1 R. Side 28.0 0.15 Front (lock end) 14.0 0.1 Rear 6.0 0.1 Top 26.0 0.1 Bottom 8.0 0.1 IV.. ADDITIONAL PERFORMANCE TESTS Each package was tested separately while containing the 17 curie sealed source of Iridium-192. Each package was subjected sequen-tially to the following prescribed tests required by IAEA Safety Series #6.
A.
Reduced Pressure - Sec. II, Par. 221 Specimens placed in pressure vessel and chamber air pressure reduced to 0.25 kg/cm2 (3.56 #/in2 Ab.). Specimens were held p) at this reduced pressure for a period of 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months .
(
1.
Evaluation---Visual examination after reduced pressure tests to containers revealed no apparent damage to con-ponents.
2.
Radiation surveys on all exterior surfaces and at I meter distance showed no significant variation from " baseline" surveys.
3.
Post contamination wipe (smear) test results on both con-tainer surfaces and source capsule were less than 0.0001 micro curies /sq. cm.
B.
Water Spray Test---Sec. VII, Pars. 710 & 711
" Test Not Required" C.
Free Drop Test---Sec. VII, Par. 712
" Compliance Demonstrated, 1973" D.
Compression Test---Sec. VII, Par. 713(a)
Siw e the vertical projected area of the test packages are racher small, a load in excess of five (5) time the actual veight of either package was used in performing this test.
)
Revision 0 2-27 1 Mar 1985
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(O Mr. R. R. Rawl Page December 4, 1981 The Model 500-SU source changer Geighs 58 pounds.
The Model 520 Iriditron weighs 40 pounds.
We employed a cylindrical, lead shielded transport container weighing 340 pounds for performing the compression test.
Each package was placed in its normal upright position with the base supported by a flat concrete floor. The 340 pound load was positioned on the top surface of each package, and the compressive load applied for an uninterrupted period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
1.
' valuation-Visual examination of each container af ter the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months load period revealed no apparent damage to containers or components.
2.
Radiation surveys on all exterior surfaces and at I meter distance showed no significant variation from " baseline" surveys.
3.
Post contamination wipe (smear) test results on both con-tainer surfaces and source capsule were less than 0.0001 micro curies /sq. cm.
D E.
Penetration Test---Sec. VII, Par. 714 Each package was subjected to the following penetration test.
Each package was placed on flat, horizontal concreto floor.
Each package received three (3) impacts from the p netrating bar. The height of free fall of the bar was one meter, mea-sured from the lower hemispherical end of the bar to the upper surface of the test specimen.
The penetrating bar consisted of cylindrical steel rod having a 3.2 cm. (1.25") diameter, and a hemispherical striking end.
The overall length of the bar was 38-1/4", and the total weight was 13-3/8 pounds.
The Model 500-SU source changer was subjected to a total of three (3) impacts; one each to the following exterior surfaces:
(a) Center of top surface (b) Center of right side (c) Center of rear surface The Model 520 Iriditron was subjected to a total of three (3) impacts; one each to the following exterior surfaces:
(a) Center of left side (b) Center of rear end plate (c) Rear of locking mechanism Revision 0 2-28 1 Mar 1985
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a Mr. R. R. Rawl Page December 4, 1981 1.
Evaluation--Visual examination revealed that the penetration test results in little or no damage to either package. A small indentation, barely discern-ible was noted at each point of impact. There was no deformation of container surfaces, and all velds and locking devices remained intact.
2.
The penetrating bar was not deformed as a result of the 6 drops.
3.
Radiation surveys on all exterior surfaces of the pack-ages, and at 1 meter distance showed no significant var-iation from " baseline" surveys.
4.
Post contamination wipe (smear) test results on both container surfaces and source capsule were less than 0.0001 micro curies /sq. em.
F.
Free Drops (I &.II)---Sec. VII, Par. 719
" Compliance demonstrated, 1973" v)
G.
Thermal Test---Sec. VII, Par. 720
" Compliance demonstrated, 1973" H.
Water Immersion Test---Sec. VII, Par. 721 Both packages were tested separately while containing the 17 curie source of Iridium-192. A pressure vessel was used in conjunction with a hydrostatic pump to exert an external water pressure of 1.5 kg/cm2 (Sauge). The actual test pres-sure used for thes.e tests was 30 PSIC maintained for a period
-oi 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months ..
1.
Evaluation--Visual examination revealed that the water immersion test resulf k,j no physical damage to either t
package.
2.
Radiation surveys on all exterior surfaces of the pack-age.s and at 1 meter distance showed no significant var-iation from " baseline" surveys.
3.
Post contaminatign wipe (smear) test results on both container surfaces and source capsule were less than 0.0001 micro curies /sq. em.
Revision 0 i2-29 i Mar 1985 0'J
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Mr. R. R. Rawl Page JfEEd December 4, 1981 4.
The total volume of pressure vessel equals 700 cu.
in. = 11.47 liters.
Residual water after Model 500-SU source changer was removed from pressure vessel = 7.78 liters.
Residual water after Model 520 Iriditron was removed from pressure vessel = 7.99 liters.
For purpose of calculations lets use 8 liters of residual water for both packages.
A 10 m1 sample (1/800 of total residual water) from each test package was assayed by well counting over a 10 minute period. Results of these assayes showed no activity levels above background.
V.
INTEGRITY OF CONTAINMENT AND SHIELDING A.
Our well counting instrument is sufficiently sensitive to detect the preser.cc of gamma emission down to 0.005 nei over a 10 minute counting period.
B.
After each performance test both packages were evaluated
(N for extent of non-fixed removable surface contamination.
y,)
This was determined by hand wiping a 300 sq. cm. area of the container surface using a dry cotton vad. Results of well counting demonstrated that in no instance did concen-trations of non-fixed contamination exceed 0.0001 micro curies /sq. em.
C.
After each performance test the contained sealed source of Iridium-192 was leak tested by using the wipe (smear) test method described in Appendix "A" of ANSI-N542, Section A2.1.1.
All exterior surfaces of the capsule were wiped thoroughly with cotton pipe stem cleaners moistened with a 5% solution of radiac wash. The amount of radioactivity removed from the source before and af ter each performance test was used as an indicator in evaluating the integrity of the package in restricting the loss of radioactiveicontents. In no case were results of capsule wipe tests in excess of 0.J001 micro curies.
D.
To further assure the packages ability to restrict loss of radioactive content, an additional leak test was performed on the contained sealed source of Iridium-192 after the packages had been subjected to the required cumulative tests.
This test consisted of " Immersion with Boiling" as specified in Appendix "A" of ANSI-N542, Section A2.1.3.
b V
Revision 0 2-30 1 Mar 1985
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%8 Mr. R. R. Rawl Page 1 December 4, 1981 Immersion with boiling was accomplished by using a stand-ard radiographic exposure device which was equipp,ed with a source guide tube fitted with a perforated " screen-type" end cap which terminated into a 1000 ml beaker containing a 500 mi solution of~5% radiac wash. This arrangement per-mitted remote handling and positioning of the source during the test. The Iridium-192 source was expo' sed and immersed into the 500 ml of solvent. The solvent.was then boiled for a period of 10 minutes. The sealed source was then re-moved and allowed to cool. The' boiling solvent was retained.
The test source was then rinsed in a 150 ml fresh solution of solvent. The rinse solvent was then added to the origi-nal boiling solvent- -This operation was repeated twice, for a total of three tests, using the original solvent for the boiling. Two 10 m1 samples of the original solvent were each assayed by well counting for a 10 minute period.
Results of these assays showed no concentrations above 0.002 nei/ml of solvent. Since the total residual solvent after boiling was equal to 600 m1, then the total activity in the solvent is (0.002 nci/ml X 600 ml) equals 1.2 nci total act-ivity.
VI.
SUMMARY
AND COMMENTS A.
Test results demonstrate that the two (2) packages have successfully met the basic and specific additional test requirements for Type B(U) packaging relating to:
(1)
Structural integrity.
(2)
Retention of shielding and shielding integrity.
(3)
Restricting loss of radioactive contents to less than: A X 10-6 per hour.
2 and X 10-3 per week A2 where A2 = 20 ci. for Iridium-192.
s B.
A set of 3 photographs are included as part of this test re-port which illustrate the various equipment and test appara-tus employed.
C.
Automation Industries, Inc. requests that both of the above referenced containers be certified as type B(U) packages, and that appropriate Certificates of Competent Authority be established.
Sincerely,
$4Y&
A Michael P. Santoro, j
v l
' Nuclear Products 12-4-81 i
l MPS: deb
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Revision 0 Enc: (8) Photographs 2-31 1 Mar 1985
$ AUTOMATION INDUSTRIES, INC.
fm SPERRY DMSiON ik)j P.o toX 245 M*
TEST RESULTS-FVR TYPE "B" PACKAGING IRIDIUM SOURCE CHANCER-SPENT URANIUM TYPE AND MODEL 520 IRIDITRON EXPOSURE DEVICE In an effort to reduce transportation costs and also as an effort towards product improvement. Automation Industries, Inc has de-signed and constructed a new Iridium Source Changer for servicing our Iridium-192 customers.
Our new Source Changer will use Spent Uranium as the shielding medium. By utilizing Spent Uranium, our new changer is approxi-mately one quarter the volume and one-third the weight of our presert lead shielded Changers. The gross shipping weight of the new Source Changer will be sixty (60) pounds. Bis reduction in size and weight renders a very compact, rugged, and structurally sound design, capable of withstanding severe abuse encountered during shipment and in field use.
The container design consists of a rectangular box, approximately 5" wide x 7" high x 11" long, fitted with a hinged cover to permit I
access to the internal compartment. Se entire unit is fabricated from type 18-8 Stainless Steel sheet, #10 cage, (0.140" thick),
with all corners and seams continuously welded. We Spent Uranium shield is completely encased in an all-welded leak tight compart-(^N ment.
The shield is fixed in the compartment by seal welding the U) two (2) exit tubes through the partitioning sheet, and then potting t
with a high temperature solid epoxy. We unit is designed to meet the requirements of D.O.T.-55 Specification.
In essence, this new changer is a miniature version of our present lead shielded changer. All threaded connections, threaded seal caps, transfer tubes, method of packaging, securing and sealing are the same for both units. The procedures to be followed for effect-ing a source change in the field are also the same for both units.
The same operating instructions will apply for both source Changers.
As designed, the Spent Uranium Changer is a D.O.T.-55 shipping con-tainer; however, at this time, Automation Industries, Inc. would also like to qualify it as a Type "B" shipping container. In order to substantiate this qualification, the Spent Uranium Source Changer was subjected to the following test requirements of the International Atomic Energy Agency. Tests were performed sequentially in the order listed below:
A.
NECHANICAL TEST Free Drop #1 The package was dropped from a height of thirty-four (34) feet onto a flat, horizontal, 5/8 inch thick steel plate.)
~
July 24,1973 M
e
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Michael P. Santoro
{Rev.#1 Q
Product Manager, Nuclear j
Dated 5-8-80 Rev. #2 Dated 1-27-83 See added Appendix "A" Page 8 %
See added Appendix "B", P. =s 9 thru 13.
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$ AUTOMATION INDUSTRIES, INC.
/7 SPERRY DMSiON
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rue.no Free Drop #2 The package was dropped from a height of forty (40) inches onto the upper end of a steel circular bar which was perpendicular to the concrete pad.
The target surface of the circular bar was flat with its outer edges rounded off to a radius of six (6) em.
Devia tion s :
(a) For the Free Drop fl, the 5/8" steel plate was not wet floated onto the concrete pad.
However, the eight (8) inch thick concrete apron was flat and horizontal, as was the steel plate, and the contact interface was intimate.
(b) For the Free Drop #2, the circular steel bar was three (3) inch diameter in lieu of the fif teen (15) cm. diameter called for.
OBSERVATIONS :
Subsequent to Drops #1 and #2, visual in-spection and radiation surveys indicate, that the container and/or containers, as presently designed, would have sustained more severe test conditions, and still g' 'S maintained integrity. Results of the t
drop tests showed negligible effects.
%d B.
THERMAL TEST The container and/or containers were suspended by wire rope from an "A" frame, centered over a 66 inch by 66 inch fuel pan, having five (5) inch sides.
The container and/or containers were positioned approximately twelve (12) inches above the surface of the fuel. The fuel consisted of a 50/50 percent mixture of Kerosene and #2 Fuel 011. Total time of exposure to flace was fourty-seven (47) minutes.
This exceeded the required test period by seventeen (17) minutes, since we were unable to extinguish the flace by use of three (3) conventional CO2 extin-guishers. Two (2) local Fire Companies arrived with suitable foam generating equipment to blanket the flame. The container and/or containers were allowed to cool naturally for a period of three (3) hours.
3eviations:
Total time of exposure to flame was fourty seven (47) minutes, due to difficulty encountered in attempting to extinguish the fire at the thirty (30) minute mark.
July 24,1973 v
Michael P. Santoro Product Manager, Nuclear f
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nj Revision 0 2-33 i Mar 1985 i
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$ AUTOMATION INDUSTRIES, INC.
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j SPERRY DIVISION v
gh, mm mem For ecological reasons and also potential fire haz-ards, the same fire was used to expose two (2) dif-ferent containers sinultaneously. (The Spent Uranium Source Changer and Our Model 520 Iridieron Exposure Device.)
An attempt to monitor the flame temperature, using a Weed Model 8000 Temperature Indicator with platinum resistance probe (0 to 1600 *F) failed, due to a malfunction in the instrument or a short in the probe element.
OBSERVATIONS:
Af ter the three (3) hour cooling period, both contain-ers were inspected visually for structural damage, and also monitored for any radiation hazards. There were no apparent high surface radiation levels. Both con-tainers exhibited bulged or " pregnant" attitudes, due to extremely high internal pressures resulting from the decomposition of the trapped epoxy potting resins at the elevated temperatures. The internal gas pres-sures had to be exceedingly high in order to permanently set a convex bow on all surfaces of the containment shells (#10 Gage, 0.140 inch thick, type 304 Stainless Steel Plate).
The Spent Uranium Source Changer showed no structural
/
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(,)
failure, nor any loss of shielding integrity, as a result of the fire test.
The Model 520 Iridieron did spring about 507. of the weld seam on the bottom portion of the rear end plate, (Lock Box End).
This occurred at approximately the fourty (40) minute elapsed time mark, while the firemen were preparing to blanket the fire with foam. The weld failure was clearly evidenced by a cuffled explosion, followed b a rapid release of expanding gases or epoxy vapors. y Close up inspection showed that the lower segment of the rear end plate had pulled away from the shell tube, forming an angle of approximately 15*off the perpen-dicular. Accordingly, this rotation of the rear end plate caused the lockbox to be cocked upward approxi-mately 15* off the horizontal.
One end of each hold down tube on either side of the ledel 520 base, had separated from the end plates.
These separations are not relevant to containmen*. of shielding integrity, however, they do attest to the tremendous forces that were built up and released, in order to cause these weld failures.
MN AItc '
July 24, 1973 Michael P. Santoro Product Manager, thclear fN
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Revision 0 1
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$ AUTOMATION INDUSTRIES, INC.
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SPERRY DIVISION
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POBOXSG PA 19440 It was also noted that the leather handles, and alumi-num Radiation Warning Tags and Labels, had completely disintegrated. Stainless Steel Nameplates and chemi-cally etched engravings remained legible. On the Model 520, the aluminum Source Identification Plate was 60%
{
melted away, and the remaining portion not legible.
C.
COVTAINMENT AND SHIELDING INTEGRITY During each test, each container was loaded with a sealed source of Iridiumr192 of following strengths:
1.
SPENT URANIUM SOURCE CHANCER Drop Tests 1 and 2, 34 Curies of Iridium-192 (See Chart "A")
30 Minute Fire Test, 30.7 Curies of Iridium-192 (See Chart "C")
2.
MODEL 520 IRIDITRON (EXPOSURE DEVICE)
Drop Tests 1 and 2, 24 Curies of Iridiumm192 (See Chart "B")
30 Minute Fire Test, 21.5 Curies of Iridium-192
(N (See Chart "D")
)
\\~ /
Pbdiation Surveys were performed on bo,th containers, prior to and after being subjected to each test.
Dose rates were measured and recorded for surface levels; at six (6) inches from external surface, and at one (1) meter from the external surface. See Charts "A" to "D" for radiation survey results.
(a) The Spent Uranium Source Changer did not exhibit in any significant change in dose rates af ter each and cumulative tests.
(b) The Model 520 Iridicron did not exhibit any significant change in dose rates af ter Drop Tests 1 and 2.
However, af ter the Thermal Test, some of the surface and six (6) inch dose rates increased, while others directly opposite in location decreased, (see Chart "D").
The cumulative average of dose rates did not shift more than 15%. This could be considered negligible. This change in radia-tion dose levels can be attributed partly to the complete loss of epoxy resin, when the rear end plate seal weld ruptured and reliev-ed the internal pressure. It was calculated 3 /'
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July 24,1973 Michael P. Santoro Product NWnager, Nuclear m
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that the epoxy offered approximately 1% of the shielding value. The major cause of shift in dose levels was caused when the lockbox cocked upwards. This motion of the lockbox pulled the Iridium-192 source capsule out of center position in the shield; approximately 3/16" off center.
The source pigtail is positioned and fixed into the shield by means of the lock prongs and limiting orifice when the unit is in locked position. Accordingly, when the locking mechanism rotated upward, the source pigtail was displaced an equal a mount. This explains why some dose read-ings increased, while others decreased.
Since the Model 520 Shield is overdesigned, with resulting good safety factor; and a 1/2" safe dwell position in the center of the "S" tube, the source capsule can be translated at least 1/4" to eigher side of the theoretical center before appreciable changes in radiation dose levels are noticed.
This feature was designed into the unit.
D.
PHOTOGRAPHS eN The following photograph numbered one (1) through
(
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twenty-three (23), illa t. rate the set up, progress,
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and effects of tests on the containers:
1------------------ 30 Foo t Free Drop
2------------------ Spen t Uranium Change r -
Prior To Test
3------------------ Fbde l 520 Iriditron -
Prior To Test
4, #5-------------- Fbdel 520 Iridieron -
Af ter 34 Foot Free Drop f6------------------ Spen t Uranium Change r -
After 34 Foot Free 3 rop
7------------------Hodel 520 And Sou rce Change r -
Af ter All Mechanical Tests
8,
9--------------Fbdel 520 Iridieron -
Before And Af ter Drop Onto Circular Pin
10, ill------------Source Change r -
Before And After Drop Onto Circular Pin
12, fl3------------Close Up Of Fuel Pan -
Prior To Thermal Test
14, #15, #16------- Progress Of The rmal Test
17, #18------------Source Changer And Model 520 -
Af ter Thermal Test Note Source Pigtail Connectors Protruding From Lockboxes.
Note Seam Weld Rupture On tbdel
$20 Iridieron.
/#f July 24, 1973 Fuchael P. Santoro Product Manager, Nuclear
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fl9-----------------Side View Of Model 520 -
Af ter Thermal Test x
Note Cocked Position Of Lockbox
' #20, #21------------ Radiation Surveys
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22, #23------------Radia tion Surveys Ed INTEGRITY OF SOURCE CAPSULE (ENCAPSULATION)
Af ter completion of all tests, both source capsules were leak tested to determine whether there was any leakage of radioactive contents. Both leak test re-sults were negative with respect to leakage.
Source Serial Number IR-6964, contained in Nbdel 520 Iridie ron.
s Leak Tested on June 6, 1973 Af ter Fire Test Removable Contamination: Less than 0.001 micro curies Source Serial Number IR-7056, contained in Spent Uranium Source Changer.
Leak Tested on June 6, 1973 Af ter Fire Test Removable Contamination: Less than 0.001 micro curies Visual inspection of the two (2) Source Pigtails when inspected under magnification while viewing through our Hot-Cell viewing window, indicated f-g that there was no anchanical damage imparted to v) either capsule assembly as result of tests, F.
RECOMMENDATIONS Due to the severity of the Theruni Test, (011 Fire),
we would suggest that anyone performing this test, do so in a very isolated area, removed from any flaar mable equipment or buildings. It is also recommended that trained professional fire-fighting personnel and equipment, be cn hand to terminate the test, and for obvious safety reasons. The heat intensity of this test is so overwheledng, that it is impossible to approach the flame with conventional hand-held fire extinguishers, when attempting to extinguish the flame after the thirty (30) minute exposure.
G.
CONCLUSION It is our opinion that both of these containers satis-g factorily met the Type "B" test requirements of The International Atomic Energy Agency, and that they be certified as such by aJsignment of individual Certifi-cation Marks as issued by the U.S. Department Of Trans-porta tion. (D.O.T.).
We' desire that this certification 9d' dIr r July 24, 1973 Michael P. Santoro Product Manager, Nuclear G
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or permit be acceptable for both domestic and export shipments.
Since the mdel 520 Iriditron is an Exposure Device, used in Industrial Radiography, and the Spent Uranium Source Changer is a shipping Container for transport-ing new replacement sources to our customers, and the returning of decayed sources to our Phoenixville faci-lity for ultimate disposal, it is 11o2gd that this type "B" certification will not require that our
~
domestic customers register with the D.O.T. to enable authorized receipt, use, or transshipment of these containers. (They are D.O.T. 55 Spec. containers, and we do not feel that our domestic customers should be burdened with additional registration, simply be-cause we tested and proved that the containers will withstand the more rigorous tests of Type "B" packaging.)
H.
SPECIAL FORM MATERIAL (a) The Spent Uranium Source Changer and the Model 520 Iriditron, will be used only as shielded containers for the isotope Iridium-192 in solid metallic form. ne wafers of Iridium-192 are encapsulated into stainless steel capsules using a 1770*F silver braze,
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(Eutectic Welding Alloys Conpany, #1807),
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for the sealing process. We sealed sources are decontaminated and leak tested prior to insertion into the shielding units.
(b) To date we have distributed over 7400 Iridium-192 Sealed Source Capsules of this design to licensed recipients.
(c) The isotope Iridium-192 in solid metallic form, is an noble metal and meets all the require-ments of celting point, sublimation, percussion friability, low solubility or dissolution, and chemical stability tests, as outlined in the International Atomic Energy Agency regulations.
Since we encapsulate no wafers of metallic Iridium-192 which have any dimension less than 0.5 nsa, the radioactive material in itself is Special Form.
I.
REOUEST FOR APPROVALS (a) We request that the Spent Uranium Source Changer be certified as a Type "B" package for shipping Special Form, Sealed Sources of Iridium-192, up to, but not exceeding, 300 Curies.
Specific Activity Range: 300 To 400 Curies / gram YdM July 24, 1973 Michael P. Santoro Product Manager, Nuclear n
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k' d Po. sot 243 PMoENIKVs,4 PA 194e0 msima e (b) We request that the Nodel $20 Iriditron be certifiwd as a Type "B" package for shipping Special Form, Sealed Sources of Iridium-192, up to, but not ex-ceeding,120 Cudes.
Specific Activity Range: 300 To 400 Curies / gram.
(c) We request that our Iridium-192 Isotope be certified as Special Form Shipment and/or Special Form Material, e
de July 24, 1973 Michael P. Santoro Product Manager, Nuclear REVISION #1.
APPENDIX "A" (ADDED 5-8-80)
ADDITIONAL DROP TEST PERFORMED 10 HODEL $20 FHIELD TO INSL1tE THAT FORCE OF IMPACT OCCURRED TO MOST VULNERABLE PART OF THE DEVICE.
(a) Date of Test: April 2, 1975 (b) Description of Test 30 ft. free drop onto unyielding target consisting of 6 inch thick flat steel plate.
(c) Orientation of Device: A guy wire was employed to maintain proper orientation
[esj seal plus connector.
to assure that maximum damage occurred to the locking mechanism and protruding Y!
(d) Test Specimen: The 520 shield was the same unit which had previously been subjected to sequential drop, puncture, and fire exposure when qualifying the unit for Type "B" Packaging.
RESULTS & OBSERVATIO*iS (a) After lepact: The locking mechanism remained locked-The protruding seal plug connector was slightly deformed--And the guy wire still maintained the 520 shield in a vertical attitude.
(b) Af ter removing guy wire, deformed seal plug connector was removed--
There was no evidence of any damage to the source pigtail connector - And the source pigtail assembly was still retained in the proper locked, stored position.
(c)
Conclusion:
Results of this test clearly demonstrates the inherent stability of the Model 520 sbield and that the device would withstand tests of this severity.
M 4 des- ' May 8, 1980.
Michael F. Santoro General Manager, Nuclear Products, mfI G
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APPENDIX "B" (ADDED !=27-83)
I.
SCOPE: TO DOCLHENT RESULTS OF MECHANICAL TESTS PERFORMED ON W E COMBINATION PACKAGE CONSISTING OF AUTOMATION INDUSTRIES MODEL 500-SU IRIDILH-192 SOL 1tCE CHANGER (CONTAINMENT VESSEL) WHEN ENCASED WITHIN A STEEL DRUM OVERPACK.
A.
PACKACE DESCRIPTION The complete package' assembly consists of Automation Industries Model 500-SU 1ridium-192 Source Changer (Containment Vessel) completely encased within a #18 gage steel drum which measures 15-1/8" outside diameter and a total height of 13-3/4".
The drum is fitted with a top opening lid which is secured with a clamp ring and seal bolt. Tuo (2) close fitting solded hair-packs are fitted within the drum to receive and main-tain the containment vessel in a fixed position during transport.
(See Automation's Dwg. No. D-5005U-OP for coaplete specifications.)
5.
CROSS VEICHT OF COMPLETE PACKACE (CONTAISHENT VESSEL PLUS DRLM OVERPACK): 80 Pounds.
C.
MODEL NLHBER OF COMPLETE PACKACE: Model 500SU-OP.
D.
TYPE AND FORM OF RADI0 ACTIVE CONTENT: Meta 111e wafers of Iridium-192 as sealed sources which meet the requirements of special-form encapsulation; and depleted uranium shield casting of the 500-SU.
E.
MAXIMLH QUANTITY OF RADI0 ACTIVE CONTENT PER PACKACE: 300 Curies of Iridium-192, plus 39 pounds of depleted uranium (shield cast-ing of the 500-SU containment vessel).
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11. PERFORMANCE TESTING - CENERAL FOR ALL PACKACINC - SI'BPART C A.
71.31(c)(I) LIFTING ' DEVICES: The complete package assembly was supported by the two (2) lifting handles and a static load of 260 pounds was distributed over the top surface of the package. Load-ing was maintained for a period of 5 minutes. At completion of test, visual inspection revealed no evidence of afsteri f ailure or deformation to any of the supporting components that would indi-cate stresses in excess of the yield strength. (See attached photograph #1).
B.
71.31(c)(4) LIFTINC DEVICES. EXCESSIVE LOAD: (Test performed by the testing laboratories of W.B. Coleman Company, Philadelphia, PA).
The package was fixtured onto a tensile testing machine and an up-ward vertical force was applied to one of the lifting handles and gradually increased until failure. The handle grip loop failed at 2040 pounds. Failure consisted of the grip loop separating and pull-ing out of the handle mounting bracket. Visual inspection revealed that the excessive load to the handle did not generate excessive stress to the drum housing nor impair the conteinment or shielding properties of the package. (See attached photograph #2).
I' W
Jan. 27, 1983 M.P. Santoro, Mgr. Engr.,
Nuclear Products.
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71.31(d)(1) TIE-DOWN DEVICES: A static force applied through the center of gravity of the package having a vertical component equal to two (2) times the weight of the package (160 pounds), a hort-montal camponent equal to ten (10) times the weight of the package (800 pounds), and a horizontal component in the transverse direc-tion equal to five (5) times the weight of the package (400 pounds).
Since we cannot control the orientation of the package in relation to the direction of vehicle travel, we have selected the centerline passing through the two (2) handles as the direction of the 800 pound horizontal vector. The resultant of these ecaponents is equal to a static force of 910 pounds rotated through 27 degrees of f the centerline passing through the handles, and having an angle of de-c11 nation equal to 10 degrees. (This test was performed by the test-ing laboratories of W.B. Coleman Company, Philadelphia, PA).
The package was fixtured onto a tensile testing machine with steel banding straps securely tightened to the hold-down loops to main-tain proper orientation of the package and also support the pack-age during the test. A one (1) inch diameter rae was attached to the crosshead to transfer and direct the 910 pound force in line with the center of gravity of the package. This load was maintained for five (5) minutes.
At full loading (910 pounds) there tas no f ailure in the tie-down system---The tie-down straps remained intact as did the handle loops.
Further inspection revealed that the test load did not generate ex-cessive stress to the drum body nor to the tie-down system. (See attached photograph f3).
D.
71.31(d)(3) TIE-DOWN DEVICES, EXCESSIVE LOADt (Test performed by the testing laboratories of W.B. Coleman Company, Philadelphia, PA).
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(
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loading would be applied to the tie-down loop at approximately 45
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degrees from the horizontal when the package would be in its normal upright shipping attitude. Loading was gradually increased until failure. The tie-down handle loop failed at 1385 pounds. Failure consisted of the handle loop separating and pulling out of the han-die mounting bracket. Visual inspection revealed that excessive loading to the tie-down device to point of failure did not generate excessive stress to the drum housing nor impair the containment or shielding properties of the package. (See attached photograph #4).
E.
71.32(a) LOAD RESISTANCE: (Test performed by the testing labora-tories of W.B. Coleman Company. Philadelphia, PA). The package was tested as a simple beam supported at its ends along both its verti-cal axis and horizontal axis. A static load of 400 pounds was applied normal to and uniformly distributed along the length of its upper surface. For both tests the load was maintained for a duration of five (5) minutes.
Visual examination at completion of tests showed no evidence that any surface of the package was stressed to its yield point. (See attached photograph #5).
M t/ M u Ian. 27, 1983 H.P. Santoro, Mgr. Engr.,
Nuclear Products.
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!!I. PERFORMANCE TESTINC - FOR NORMI, CONDITIONS OF TRANSPORT " APPENDIX "A" A.
APPENDIX "A-6". FREE DROP The package was subjected to three (3) f ree drops through a distance of four (4) feet onto a flat hori-sontal surface (blacktop roadway). The points of impact were to the top clamping ring, the bottom rim and to the side wall of the drum.
These tests imparted very little damaRe to the drum other than re-moval of paint and slight surface scratches to the metal at points of impact. (See attached photograph #6).
5.
APPENDIX "A-7". CORNER DROP: The package was subjected to one (1) foot corner drops onto each quarter of the top clamping ring, and onto each quarter of bottom ria (total of eight drops). The sur-face of impact was a flat horizontal blacktop roadway.
Again, damage was very minimal consisting of paint removal and slight surface scratches to the metal at points of impact. (See attached photograph #7).
C.
APPENDIX "A-8". PENETRATION: The package was supported by a flat horizontal blacktop roadway. The package was subjected to two (2) impacts from the penetrating bar. One g1) drop onto the center of the top lid, and the second drop onto the center of the drum's bot-tom with the drum resting in the inverted position.
The penetrating bar consisted of a cylindrical steel rod having a 1-1/4" diameter, and a hemispherical striking end. The overall length of the bar was 38-1/4", and the total weight was 13-3/8 pounds. The height of free fall of the bar was 40 inches measured from the lower hemispherical end of the bar to the upper surface of the test package.
Visual examination revealed that the penetration drops resulted in very little damage to the two (2) anrfaces of impact. Damage to both surfaces at point of impact consisted of a hemispherical de-pression approximately 3/4" in diameter and 1/4" deep. There was no piercing or tearing of either test surface. (See attached photo-graph #8).
D.
APPENDIX "A-9". CCMPRESSION: A static load of 460 pounds was equally distributed over the top surface of the drum while the drum was in its normal upright transport position and supported by a flat hort-tontal concrete ficor slab. The compressive load was maintained continuously for a period of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .
Visual examinations after testing revealed no evidence of any damage to the package resulting from this loading, nor any indications of deformation to any of supporting components which would indicate stresses in excess of the yield strength. (See attached photograph
9).
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(f44 Jan._27, 1983 i
M.P. Santoro, Mgr. Engr..
Nuclear Products.
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PERFORMANCE TESTING - HYPOTHETICAL ACCIDENT CONDITIONS " APPENDIX-B" A.
APPENDIX "B-1", FREE DROP THROUCH 30 FEET A sealed source of IR-192, serial number 15202, having a calibrated strength of 8.5 curies on January 12, 1983 was loaded into and secured within a Model 500-SU Source Changer, serial number SU-644 This will be designated as the containment vessel.
The containment vessel was then inserted into the drum overpack f and properly positioned and secured within the cavity of the molded hair-packs. The drum was then closed with the top removable lid and secured and locked by the clamp ring and seal bolt assembly.
This complete assembly will be designated as the Model 500sU-OP overpack assembly, and will have a gross shipping weight of 80 pounds.
Prior to drop, the package was surveyed for external surface radiation levels and radiation levels measured at three (3) feet from all external surfaces of the package. Results of these surveys indicated the highest levels recorded on any ex-terior surface of the drum was 1.8 MR/HR and 0.1 MR/HR when mea-sured at a distance of three (3) feet from any exterior surface.
The package was hoisted to the rooftop of a 34 foot high build-ing, and allowed to drop onto a flat horizontal essentially un-yielding blacktop roadway.
Impact occurred at the lower side surface of the drum between the area of the lower chine bead and the bottom rim. Essentially it approximated a bottom corner drop.
(N The damage to the package was minimal considering the amount of (V) energy absorbed. The structural soundness of the package was unchanged and the integrity of closures and containment remained intact. The only physical damage consisted of flattening of the bottom rim of the drum at surface of impact and a slight perfora-tion of the side wall also within the area of impact. The perfora-tion resembled a right angle allt having a width of 3/32" and un-equal legs of 1-1/2" X 1".
The perforation was actually the in-print of the rear lower corner of the containment vessel shich compressed the hair pack at the instant of impact with sufficient force to locally shear the side wall of the drum by the combined actions of shear, elongation, and forging.
Radiation surveys on all exterior surfaces of the package, and at a distance of 3 feet from all exterior surfaces demonstrated that there was no increase in surface radiation levels or transport index resulting from the free drop test. The highest surface radiation level remained at 1.8 MR/HR as did the transport index at 0.1 MR/HR.
tpon removal of the seal bolt, clamping ring, top lid, and upper hair-pack mold from the package, we noted that containment vessel was still properly positioned by the orientation bracket and pro-perly nested within the lower hair-pack cavity. This demonstrated that the hair-packs have an excellent capacity for energy absorb-tion, combined with good elasticity and resiliency. Examination of the containment vessel revealed no damage to any of exterior surfaces, seal wires or closures. (See attached photograph #10).
Y Jan. 27, 1963 M.P. Santoro, f.ngr. Mgr.,
Nuclear Products.
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APPENDIX "B-2", PUNCTURE: The package was subjected to four (4) drops through a distance of 40 inches measured from the lowest surface of the package to the top of a cylindrical mild steel bar which was mounted vertically on a flat horizontal unyielding sur-face. The bar was 6 inches in diameter, 8 inches in length, with a J!st horizontal top surf ace whose edge was rounded to a 1/4" radius.
The four (4) drops were so directed so that impacts would occur to the following surfaces of the package Center of top lid, cente.. of bottom, mid-point of side with drum falling hortrontal-ly, ad onto the seal bolt and locking lugs of the clamp ring assemaly.
Visual examinations after testing indicated that damage to the package resulting from these drops were very minimal; consist-ing of very sifght crescent shaped indentations to the flat sur-faces of drum lid and bottom. (See attached photograph #11).
V.
SUMMARY
AND COMMENTS:
A.
Test results demonstrate that the combination containment vesse' and drum overpack assembly Automation Industries, Inc. Ndel 500SU-OP has successfully met the test requirements for both normal and accident conditions of transport relating to (1) Structural integrity (2) Retention of shielding and shielding integrity (3) Restricting loss of ary radioactive content A
f I
B.
The above radioactive material packages will continue to be de-(j signed and fabricated in accordance to Automation Industries' Quality Assurance Program, NRC Docket No. 71-0264, approved October 9, 1979.
C.
All tests covered by Revision #2, Appendix "B" of this report were conducted daring the period of January 4 through January 21, 1983.
D.
Automation Industries, Inc. requests that the combination con-tainment vessel and drum overpack assembly, Model 5005U-OP be certified as a Type "B(U)" package for transporting sealed sources of special-form Iridium-192 up to a maximum of 300 cur'ies per package, and that the appropriate certificates of competent authorities be established.
M Jan. 27, 1983 4*
Michael P. Santoro. Mgr. Engr.,
Nuclear Products n/
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TEST REP 0RT TO:
John J. Murwo III FROM:
Cathleen M. Roughan (nif-DATE:
6 March 1986
SUBJECT:
MODEL AI 500 SU DROP TEST On 27 February 1986, a Model AI 500 SU outfitted with the modified lock box was subjected to a free drop test in accordance with the requirements of 10 CFR 71.73 (c)(1) and IAEA Safety Series No. 6, Paragraph 719 (a). The test was performed at Valley Tree Service in Groveland, PA.
The Model AI 500 SU was dropped twice from a height of 9.1 meters (30 feet) onto a target. The target consisted of a concrete cube, each side measuring 1.2 meters (48 inches) upon which had been wet-floated a steel plate 0.9 meters (36 inches) wide, 0.9 meters (36 inches) long and 25 mm (one inch) thick. This target conforms to the guidance for an essentially unytelding surface as prescribed in Paragraph 701 of IAEA Safety Series No. 37.
)
In the first drop, the front plate of the package impacted the target.
'd This resulted in a slight deformation of the package, approximately 6m
(.25 inch).
In the second drop, the package impacted the target on the cover plate over the lock box mechanism. The cover plate was shifted to the right side approximately 12 mm (0.5 inch); the padlock was bent but remained in place, locking the bar and the cover plate in the secured position.
As a result of these tests, there was slight deformation in the package but no impairment of the safety features of the package. There was no struc-tural damage to the lock box assembly nor any release of the package contents.
A shielding efficiency test performed subsequent to the completion of the Model AI 500 SU test program demonstrated that the free drop tests did not reduce the shielding efficiency of the packace.
Revision 3 2-45 31 Mar 1986 O
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Free Drop Target:
Consists of a concrete cube, each side 3
measuring 1.2 m (48 inches) upon which had been wet floated a steel plate 0.9 m (36 inches) wide, 0.9.m (36 inches) long and 25 mm (1 inch) thick.
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FREE DROP TEST 2-47 j
i Revision 3 31 March 1986
TectVOps Q
TEST REP 0RT T0:
John J. Munro FROM:
Cathleen M. Roughan L&
DATE:
March 21, 1986
SUBJECT:
MODEL AI 500 SU PUNCTURE TEST On February 27, 1986, a Model AI 500 SU source changer outfitted with a mod-ified lock bcx was subjected to the puncture test in accordance with the requirements of 10 CFR 71.73(C)(2) and IAEA Safety Series No. 6, Paragraph 719(b). This test was performed at Valley Tree Service in Groveland, Mass.
Immediately following the free drop tests, the Model AI 500 SU was dropped three times from a height of one meter onto a target. The target consisted of a right circular cylindrical steel billet,152 mm (6 inches) in diameter and 203 mm (8 inches) high, mounted onto the target used in the free drop tests.
During the first drop, the package impacted the cover plate over the lock n
box. The padlock was removed and the lock box was mechanically tested.
)
The lock box did not have any observable damage and worked properly.
In the second drop, the cover plate was secured in the open position so the lock box would impact the target without the protection of the cover plate. The package impacted the target on the rear face. As a result of this, the lock box was slightly loosened but still functioned properly and was still securely attached to the package. The head of one of the screws securing the top plate of the lock box had broken off in the impact.
In the third drop, the cover plate was again secured in the open position in order to impact the lock box without the protection of the cover plate.
The package impacted the target on the rear face and there was no struc-tural damage to the lock box. The lock box was tested for proper opera-tion and functioned satisfactorily.
As a result of these tests, therewas no impairment of any design or safety features of the package. There was no structural damage to the locking assembly. There was no release of the package content:;.
A shielding efficiency test performed subsequent to the completion of the Model AI 500 SU test program demonstrated that these puncture tests did not reduce the shielding efficiency of the package.
Revision 3 2-48 31 Mar 1986
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l Punture Test Cover Plate Closed evision 3 2-52 31 Mar 1986 i
3.0 1bexmal ExaluatiDD
/~'N 3.1 DisssssipD O
The Model AI 500 SU is a ccupletely passive thermal device and has no mechanical cooling system nor relief valves.
All cooling of the package is through f ree convection and radiation.
The heat source is 120 curies of iridium-192.
The corresponding decay heat generation rate is 1.03 watts.
3.2 Summary 92 Zbgxmal R199g111g5 DL BakgIlalR The melting temperatures of the metals used in the con-struction of the Model AI 500 SU are:
Brass 930*C (17 06
F)
Uranium 1133 C (2070*F)
Steel 1345 C (2453*F)
Tungsten 3370'C (60988F)
Titanium 1820'C (3308*F)
The epoxy used in this device has an operating temperature range of -43 C to 104 C (-45 F to 220 F).
3.3 TggbDigal EpsgitigatigDs of CompgDgnts Not applicable
/s 3.4 UDImal CDDditiDDS 92 TIaDEP9It U
3.4.1 Thermal Bsdel The heat source in the Model AI 500 SU is a maximum of 120 curies of iridium-192.
Iridium-192 decays with a total energy liberation of 1.45 MeV per disintegration or 8.58 milliwatts per curie.
Assuming all the decay energy is transformed into heat, the heat generation rate for the 120 curies of iridium-192 would be 1.03 watts.
To demonstrate compliance with the requirements of 10 CFR 71.43 (g) and paragraph 230 of IAEA Saf ety Series No. 6, a separate analysis is presented in Section 3.6.
The thermal model employed is described in that analysis.
To demonstrate compliance with the requirements of paragraph 240 of IAEA Safety Series No. 6 for Type E(U) packaging, a separate analysis is presented in Section 3.6.
The thermal model employed is described in that analysis.
3-1 Revision 0 17 November 1983 O
3.4.2 Bazimum Zgmpggatuzgs p
The maximum temperatures encountered under normal
\\
conditions of transport will have no adverse effect on the structural integrity or shielding efficiency of the package.
As presented in Section 3.6, the maximum tempera-ture in the shade would not exceed 43*C (119 F) and the maximum temperature when insolated would not exceed 64 C (146"F).
3.4.3 BiDisum Tamp 2 Intuits The minimum normal operating temperature of the Model AI 500 SU is -40*C (-40*F).
This temperature will have no adverse effect on the structural integrity or shielding efficiency of the package.
3.4.4 bazimam laternal Exssanzes Normal operating conditions will generate negligible internal pressures.
Any pressure generated is signifi-cantly below that which would be generated during the hypothetical thermal accident condition, which is shown to result in no reduction in structural integrity or shielding efficiency.
3.4.5 Enzimum Zb2xmal 21122222 The maximum temperatures which will occur during normal transport are sufficiently low to assure that thermal gradients will cause no significant thermal stresses.
3.4.6 Eyaluation gi Raskagg Egxigsmange ynggx yggmal gonfigiong DL ExaDs99xt The normal transport thermal condition will have no adverse effect on the structural integrity or shielding efficiency of the package.
The applicable conditions of IAEA Safety Series No. 6 for Type B(U) packages are shown to be satisfied by the Model AI 500 SU.
3.5.0 Bypothetical Tb21 mal bscident EyaluatinD 3.5.1 Tb2Imal Esdel A prototype Model AI 500 SU package was subjected to the hypothetical thermal accident condition by Automation Industries Inc.
3-2 Revision 0 1 March 1985 O
i 3.5.2 Rackass CDDBikicos and EDyix9Danch
[
The prototype Model AI 500 SU package which was subjected to the hypothetical thermal accident condition had previously been subjected to f ree drop test and puncture test.
The package had suffered minor structural deformation during these-mechanical tests, but suffered no reduction in structural integrity or shielding efficiency.
3.5.3 Eackagg Ism &sInhDIns During the hypothetical thermal accident test, the package was placed into a kerosene and fuel oil fire.
The package remained there for 47 minutes.
The possibility of the formation of an iron-uranium eutectic allow was addressed in Section 2.4.1 where it was concluded that the formation of the alloy was not a likely possibility.
There was no indication of any melting or alloy formation as a result of this thermal test.
3.5.4 Bazimum 1DhexDal Exassures The maximum internal pressure generated during the thermal test is described in the Automation Industries Test Report by Mr.
M.
Santoro, dated 24 July 1973 and presented in Section 2.10.
O In Section 3.6.3, an analysis of the source capsule, which serves as the primary containment, under the thermal test condition is presented.
This analysis demonstrates that the maximum internal gas pressure at 800*C would be 373 kPa (54 psi).
The critical location for failure is the source capsule weld.
The analysis shows that an internal pressure of 373 kPa (54. psi) would generate a maximum stress of 1.96 MPa (284 psi).
At 87 0'C (16 0 0 *F), the yield strength of Type 304 stainless steel is 69 MPa (10,000 psi).
Therefore, if the source capsule were to reach a temperature of 800*C, the maximum stress in the capsule would be only 3% of the yield strength of the material.
3.5.5 Enzimum Zbermal Shxgsses There were no significant thermal stresses generated during the thermal test.
A description is presented in the Automation Industries Test Report by Mr. M. Santoro dated July 1973 and presented in Section 2.10.
3-3 Revision 0 1 March 1985 O
3.5.6 EsalvatiDD 92 Rackags ERIL9xmaDGk s
As a result of the hypothetical thermal accident. test, i
there was no significant deformation or damage.
There was no impairment of any design or safety features.
There was no damage to the locking assembly or package closure.
There was no release of the package contents.
A shielding efficiency test was performed at the conclusion of the thermal test.
This shielding efficiency test demonstrated that there was no reduction in shielding efficiency as a result of the thermal test condition.
A report of the thermal test by Automation Industries' Mr.
M. Santoro dates 24 July 1973 is presented in Section 2.10.
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3-4 Revision 0 1 March 1985 O
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3.6.0 App 2Ddl2 j
O 3.6.1 Model AI 500 SU Type B(U) Thermal Analysis:
10 CFR 71.43 (g) and paragraph 230 of IAEA Safety Series No. 6 3.6.2 Model' AI 500 SU Type B(U) Thermal Analysis:
Paragraph 240 of IAEA Safety Series No. 6 3.6.3 Iridium-192 Source Capsule Thermal Analysis s
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4 3-5 Revision 0 1 March 1985
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3.6.1 Model AI 500 SU Type B(U) Thermal Analysis q
10 CFR71.43 (g) and paragraph 230 of g
IAEA Safety Series No. 6 This analysis demonstrates that the maximum surface temperature of the Model AI 500 SU will not exceed 50* C with the package in the shade and an ambient temperature of 38*C.
To assure conservatism, the following_ assumptions are used:
(a) The entire decay heat (1.03 watts) is deposited in the exterior surfaces of the package.
(b)
The interior of the package is perf ectly insu-lated and heat transfer occurs only from the exterior surface to the environment.
j (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.
(d)
The only heat transfer mechanism is free convection.
Using these assumptions, the maximum wall temperature is q=hA (T
-T) where q = Heat deposited per unit time in the face of interest (0.26 watts) h = Free convection heat transfer coefficient for air
= 1.32 (OT U waWE T d
A = Area of the face of interest (0.0175m' )
T,=
Maximum temperature of the surface of the package T=
Ambient Temperature (38*C) a From this relationship, the maximum temperature of the su rf ace is 4 2.6*C.
This satisfies the requirement of 10 CFR 71.43 (g) and paragraph 230 of IAEA Safety Series No.
6.
3-6 Revision 0 1 March 1985 O
u
3.6.2 Model AI 500 SU Type B(U) Thermal Analysis paragraph 240 of IAEA Safety Series No. 6
\\
This analysis demonstrates that the maximum surface temperature of the Model AI 500 SU will not exceed 82*C when the package is in an ambient temperature of 38 C and is insolated in accordance with 10 CFR 71.71(c) (1) and Table III of IAEA Safety Series No. 6.
The calculational model consists of taking a steady state heat balance over the surface of the package.
In order to assure conservatism, the following assumptions are used.
(a)
The package is insolated at the rate of 775 W/m' L
(800 cal /cm'-12 hr) on the top surface,194 W/m (200 cal /cm
-12 hr) on the side surfaces and no insolation on the bottom surface.
(b) The decay heat load is added to the solar heat load (c)
The package has an unpainted stainless steel surface.
The solar absorptivity is assumed to be 0.9.
The solar emissivity is assumed to 0.8.
(d)
The package is assumed to undergo free convection from the sides and top, and undergo radiation f rom the sides, top and bottom.
The inside faces pd are considered perfectly insulated so there is no conduction into the package. The faces are consi-dered to be sufficiently thin that no temperature gradients exist in the faces.
(e)
The package is approximated as a rectangular parallele piped 279mm (11 in) long,141 mm (5.56 in) wide, and 124 mm (4.88 in) high transported on the bottom.
The surface area of the top and L
bottom are 0.0393 m.
The total surf ace area of the sides is 0.104 m%
The maximum surface temperature is computed from a steady state heat balance relationship:
in " Nout 3-7 Revision 0 1 March 1985 4
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i The heat load applied to the package is 1
O qin "
9s d
where w:
solar absorptivity (0.9) q, :
solar heat load (50.63 watts) qd:
decay heat load G.O wans)
The heat dissipation is expressed as S
out e
r where gj convective heat transfer q :
radiative heat transfer The convective heat transfer is (hA)
+ (hA)
~
a' q
=
sides w
where h:
convective heat transfer coefficient A:
area of the surface of interest T :
Temperature of the surface w
T, :
Ambient Temperature (3 8*C)
The radiative heat transfer is q
=g A (T
-T
)
r where a:
StefanBoltzmann Constant (5.669 x 10-8W/m* *K)
E. :
Emissivity (0.8)
Interation of this relationship yields a maximum wall temperature of 63.3 *C which satisfies the requirements of paragraph 240 of IAEA Safety Series No. 6.
3-8 Revision 0 1 March 1985 O
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3.6.3 Model AI 500 SU Type B(U) Source ' Capsule Thermal Analysis Paragraph 238 of IAEA Safety Series No. 61973 This analysis demonstrates that the pressure inside the source capsule used in. conjunction with the Model AI 500 SU, when subjected to the hypothetical thermal accident condition, does not exceed the pressure which corresponds to _the minimum yield strength at the thermal test temperature.
The source capsule is fabricated from stainless steel, either Type 304 or 304L.
The outside diameter of the capsule is 6.35 mm (0.250 inch).
The source capsule is seal welded.
The minimum weld penetration is 0.5 mm (0.02 inch).
Under conditions of. internal pressure, the critical.
location for failure is this weld.
The internal volume of the source capsule contains only iridium metal (as a solid) and air.
It is assumed at the time of loading the entrapped air is at standard temperature and pressure (20* C and 10 0 kPa).
This is a conservative assumption because, during the welding process, the internal air is heated, causing some of the air mass to escape before I
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 temper-ature of 800*C (1475 F).
Using the ideal gas law and requiring the air to occupy a constant volume, the internal gas pressure could reach 373 kPa (54 psi).
The capsule is assumed to be a thin walled cylindrical pressure vessel with the wall thickness equal to the depth of weld penetration.
The maximum longitudinal stress is calculated from:
cr Ai - pap z
where 53:
Longitudinal Stress g:
Stress Area
~
p:
Pressure Pressure Area g:
From this relationship, the maximum longitudinal stress is g
calculated to be 900 kPa (129 psi).
3-9 Revision 0 1 March 1985 O
~
The hoop stress is calculated from s
1a [{'= Pde where cg : Hoop Stress L: Length of the cylinder t : Thicknessof the cylinder (0.5mm ore.02 inch)
From this relationship, the hoop stress is calculated to be 1.96 MPa (284 psi).
At a temperature of 870*C (1600 F), the yield strength of Type 304 stainless steel is 69 MPa (10,000 psi).
Therefore, under the conditions of paragraph 238 of IAEA Saf ety Series No. 6, the stress generated is less than 3%
of the yield strength of the material.
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3-10 Revision 0 1 March 1985 O
4.0 C9DhalDESDL 4.1.0 C9DhalDmgDk D99Dbary 4.1.1 CDDhalDmgDL Hg&Sel The containment system for the Model AI 500 SU is the radioactive source capsule as described in Section 1.2.3 of this application.
This source capsule is certified as special form radioactive material in IAEA Certificate of Competent Authority Number USA /0279/S or USA /0154/S.
4.1.2 CDD%alDmgDL EgDgkratigDB There are no penetrations of t*se containment.
4.1.3 Seals and 89162 The containment is seal welded by a tungsten inert gas welding process which is described in Tech / Ops Standard Source Encapsulation Procedure presented in Section 7.4.
The minurum weld penetration is 0.5 mm (0.02 inch).
4.1.4 C19sure Not applicable.
4.2.0 EggDixsagots igx ynxmal CDDBigigDs 91 2xannygxt 4.2.1 Egleang at Badioachixg Bataxial The source capsules used in conjunction with the Model AI 500 SU have satisfied the requirements for special form radioactive material as prescribed in 10 CFR 71.77 and IAEA Safety Series No.
6.
There will be no release of radioactive material under the normal conditions of transport.
4.2.2 ErgssyrlzakigD 92 gbg ggDgalDmgDg yggggl Pressurization of the source capsules under the conditions of the hypothetical thermal accident was demonstrated to generate stresses well below the yield strangth of the capsule material as described in Section 3.6.3.
Therefore, the containment will withstand the pressure variations of normal transport.
4-1 Revision 0 1 March 1985 O
4.2.3 C991ADL CDDhamlDB%iDD Not applicable.
4.2.4 C991aD1 L9Es Not applicable.
4.3 C9DhalDERDL B2GMLE252DLs 191 hbt BX99hbtki9al b991dRDL CDDDitiDD 4.3.1 EinsigD Gas Ex9dM9hs Not applicable.
4.3.2 Eglgang 92 CDDggDLs The hypothetical accident conditions of 10 CFR 71.73 will result in no loss of package containment.
This conclusion is based on information presented in Sections 2.7.1, 2.7. 2,
2.7.3, 2.7.4, and 3.5.
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4-2 Revision 0 1 March 1985 l
5.0 Shielding ExaluatlDD 5.1 Dis 9yssipD and Ensults The principle shieldjng of the Model AI 500 SU is the uranium shield assembly.
The mass of the uranium shield.is 18 kilograms (39 pounds).
A shielding efficiency test of a Model AI 500 SU was made.
The package contained 106 curies of iridium-192.
A report of this test is presented in Section 5.5.
Extrapolation of these data to a capacity of 120 curies of iridium-192 is presented in Table 5.1.
Since the Model AI 500 SU contains no neutron source, the gamma dose rates are the total does rates which are presented.
As shown in Table 5.1, the maximum dose rates are below the regulatory requirements.
Table 5,1 Summary 92 BazimDD D92g Bates mBLbx At Surface At One Meter from Surface TDP Side B9119m T9P Sid2 E91195 164 172 189 1.1 1.1 1.1 5.2 29ux99 Spn91LiDatinD 5.2.1 gamma 29yx9g The gamma source is iridium-192 in a sealed capsule as special form radioactive material in quantities up to 120 curies.
5.2.2 Egug19D 29sx9g Not applicable.
5.3 896gl SpgDifi9atiDD Not applicable.
5.4 Sbigidlog EyalyagiDD A shielding ef ficiency test of a Mode 1 AI 500 SU containing 106 curies of iridium-192 was performed.
The results of 5-1 Revision 0 1 March 1985 O
m 4
l this
test, which are presented in Section 5.5, O,
demonstrate that the dose rates surrounding this package are within the regulatory limits.
An additional shielding efficiency test was made prior to and subsequent to the free fall and puncture test made on a Model AI 500 Su package equipped with a modified locking arrangement.
The results of these tests are presented in Section 5.5.
These results demonstrate that the Model AI 500 SU with the modified locking arrangement maintained its shielding efficiency under the hypothetical free fall and puncture accident conditions.
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.J 5-2 Revision 3 31 March 1986
4 t-55 Appendix l
Test Report: Shielding Efficiency Test Model Al 500 Su Serial Number 688 Test Report: Shielding Efficiency Test of Model Al 500 Su after Free Fall and Puncture Tests r
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i 5-3 Revision 3' 31 March 1986
~. -.
4 TEST REPORT RADIATION PRODUCTS DIVISION BY:
John J. Munro III DATE:
15 February 1985
SUBJECT:
Model AI 500 SU Shielding Efficiency Test On 19 October 1984 Model AI 500 SU serial number 688 was subjected to a shielding efficiency test.
The package contained source assembly Model A424-9 serial number 6319 with an activity of 106 curies of iridium-192.
The package tested was prepared as for transport including installation of the outer package cover.
The test consisted of measuring the radiation intensity in the vicinity of the package using a small. remote probe survey instru-ment.
The instrument used in this test was an AN/PDR-27 (J).
The radiation detector has an active diameter of 6.4 mm and an active length of 11.1 mm.
The detector is installed in a housing with a diameter of 15.9 mm.
The instrument was most recently calibrated on 5 August 1983.
The first phase of the test consisted of measuring the radiation intensity at the surf ace of the package.
A slow scan survey was O
made over the entire surface of the package with thc detector housing in contact with the surface.
The maximum radiation inten-sity indicated on each surface was recorded. This measured value was corrected for the surface intensity by multiplying the measured result by the square of the ratio of source-to-detector axis dis-tance to source-to-surf ace distance.
This result is presented in the following table.
The second phase of the test consisted of measuring the radiation intensity at a distance of one meter from the surface of the package.
A slow scan survey was made adjacent to the entire sur-face of the package with the detector axis located one meter from the package surface.
The maximum radiation intensity indicated at one meter f rom each surface was recorded and is' presented in the s
following table.
i This test demonstrated that the radiation intensities associated.
with the Model AI 500 SU, when extrapolated to a capacity of 120 curies of iridium-192 are within the regulatpry limits as specified in 10 CFR 71.47,10 CFR 34.21 and IAEA Safety Series No. 6.
5-4 Revision 0 1 March 1985 O
4
RADIATION PROFILE MODEL AI 500 SU SERIAL 688 19 October 1984 Containing Source Assembly Model A424-9 Serial No. 6319 l
Activity of 106 Curies of Iridium-192 Maximum Dose Rates (mR/hr)
At Surface At One Meter From Surface To 145 1.0 Front Side 123 1.0 Right Side 152 1.0 Rear Side 44 1.0 O,
Left Side 89 1.0 Bottom 167 1.0 Measurements made with AN/PDR-27(J) survey instrument serial number 7930 calibrated 1 August 1984 5-5 Revision 0 i
1 March 1985 r.
r
--e-.
, _ -, +..,,, -, - -,,,
RADIATION PRODUCTS DIVISION
,m I
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i TEST REP 0RT BY:
Cathleen M. Roughan (nif-DATE:
21 March 1986
SUBJECT:
MODEL AI 500 SU SHIELDING EFFICIENCY TEST On 26 February 1986, Model AI 500 SU serial number 659 was subjected to a shielding efficiency test. The package contained source assembly Model A424-9 serial number 8804 with an activity of 107.5 curies of Iridium-192.
The package tested was prepared as for transport including installation of the outer package cover.
The test consisted of measuring the radiation intensity in the vicinity of the package using a small remote probe survey instrument. The instrument used in this test was an AN/PDR-27(J). The radiation detector has an active diameter of 6.4 mm and an active length of 11.1 mm.
The detector is installed in a housing with a diameter of 15.9 mm.
The in'strument was most recently calibrated on 3 February 1986.
The first phase of the test consisted of measuring the radiation intensity (O'")
at the surface of the package. A slow scan survey was made over the entire surface of the package with the detector housing in contact with the sur-face. The maximum radiation intensity indicated on each surface was re-corded. This measured value was corrected for the intensity by multiplying the measured result by the square of the ratio of source-to-detector axis distance to source-to-surface distance. This result is presented in the following table.
The second phase of the test consisted of measuring the radiation intensity at a distance of one meter from the surface of the package. A slow scan survey was made adjacent to the entire surface of the package with the detector axis located one meter from the package surface. The maximum radiation intensity indicated at one meter from each surface was recorded and is presented in the following table.
The same shielding efficiency test was performed on 27 February 1986 at the conclusion of the drop and puncture tests.
The Model AI 500 SU serial number 659 contained source assembly Model A424-9 serial number 8804 with an activity of 106.5 curies of Iridium-192.
These tests demonstrate that the radiation intensities associated with the Model AI 500 SU, when extrapolated to a capacity of 120 curies of Iridium-192, are within the regulatory limits as specified in 10 CFR 71.47, 10 CFR 34.21 and IAEA Safety Series No. 6.
/
(
l 5-6 Revision 3 31 March 1986
()
RADIATION PROFILE MODEL AI 500 SU SERIAL 659-26 FEBRUARY 1986 BEFORE DROP AND PUNCTURE TESTS Containing Source Assembly Model A424-9 Serial No. 8804 Activity of 107.5 Ceries of Iridium-192 Maximum Dose Rates (mR/hr) i l
At Or:e Meter i
At Surface From Surface 1
Top 57 0.5 Front Side 84 0.7 1
Right Side 124 0.7 Rear Side 22 0.2 Left Side 87 0.7 Bottom 158 0.7 Measurements made with AN/PDR-27(J) survey instrument serial number I-130 calibrated 3 February 1986 i
i l
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i 5-7 Revision 3 31 March 1986
.c-.. -, _
.-,-r
-y ww d
we*
r--m-g-u+-
-r-
-e---m-,v v
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c - vmw
&w'ee-*1
,,w-wv g-mt----w=ww m--*~w=+v+-ee+-+-ee-++,-ww rn-we--
,,w.v e w +e
RADIATION PROFILE 4
MODEL AI 500 SU SERIAL 659 I
27 FEBRUARY 1986 i
AFTER DROP AND PUNCTURE TESTS Containing Source Assembly Model A424-9 Serial no. 8804 Activity of 106.5 curies of Iridium-192 Maximum Dose Rates (mR/hr) 1 At One Meter At Surface From Surface l
Top 34 0.2 Front Side 99 0.3 Right Side 176 0.7 i
Rear Side 18 0.5 Left Side 103 0.5 Bottom 116 0.7 Measurements made with AN/PDR-27(J) survey instrument serial number I-130 calibrated 3 February 1986 I
e
.. 58 Revision 3 31 March 1986 l
~_,b
..-,.,.,-..-,..m,.____-,-,.--,-,__,---,_..-,...~.....---
- -,. - -. - - - ~,. -, - -,... ~,.
6.0 Czi.ticallt2 E2DlDDLlDD Not applicable.
O 6' 1 Revision 0 1 March 1985 O
7.0 DygratiDS 219G2d9125 v
7.1 R19ceduxs 19x L9ading kbg Eackage The procedure for fabricating the special form source capsule is presented in Section 7.4.1.
The procedure for loading the source assemblies into the package is also included in Section 7.4.1.
7.2 219seduza 191 DDlgading kbg Enchage The procedure for unloading the package is presented in Section 7.4.2.
7.3 Exg2DrablDD DL a RaChagg 19x TxaDS991%
The procedure for preparation of a package for transport is included in the Model AI 500 SU Instruction Manual presented in Section 7.4.2.
O C.
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7-1 Revision 0 1 March 1985 O
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7.4 b99BDbli 7.4.1 Procedure for-Encapsulation of Sealed Sources l
7.4.2 Model AI 500 SU Source Changer Operation Manual i
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7-2 Revision 1 j
19 April 1985 t
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_ _ _ _,... -.. - ~. -., ~.. _... _ _
T.4.1 Procedure for Encapsulation of Sealed sources RADIATION SAFETY MANUAL Part B - In Plant Operations Section 2 ENCAPSULATION OF SEALED SOURCES A.
Personnel Recuirements Only an' individual qualified as a Radiological Technician shall perform the operations associated with the encapsulation of sealed sources.
A second Radiological Technician must be available in the building when these operations are being performed.
B.
General Recuirements l
1. In the Burlington, MA facility, a loading cell shall be used for the encapsulation of sealed sources and repackaging of sealed sources.
The maximum amount and form of radioactive material j
which may be handled in the loading cell is specified below:
j Radioisotope Form Maximum Activity Iridium-192 Solid Metallic 2000 curies Cobalt-60 Solid Metallic or sealed sources 1 curie Cesium-137 Sealed Sources only 100 curies Ytterbium-169 Sealed Sources only 100 curies i
Tantalum-182 Sealed Metallic or 100 curies Colid Carbide Limits for any other radioisotopes or forms shall be specified by the Radiation Safety Committee.
2. In the Pho'enixville, PA facility, the general purpose hot cell shall be used for the encapsulation of sealed sources and i
repackaging of sealed sources.
The maximum amount and form of radioactive material which may be handled in the cell is specified below:
Radioisotope Form Maximum Activitv_
Iridium-192 Solid Metallic 20,000 curies Cobalt-60 Sqlid Metallic 1 curie Cobalt-60 Sealed Sources 300 curies i
Cesium-137 Sealed Sources 200 curies Ytterbium-169 Sealed Sources 100 curies Tantalum-182 Solid Metallic or 100 curies 4
Solid Carbide l
/'
Limits for other radioisotopes or forms shall be specified by the Radiation Safety Committee.
7-3 Revision ]
19 April 1985
.- -.... N.
- - - - - l- - L*:5.-. :- X-- - --
IC--TT:7===
3. The loading and general purpose hot cellc ar'e-designed to be operated at less than atmospheric pressure.
The exhauct-blower rhould not be turned of f during the operation er at any time that radioactive material is in the cell.
4. Unencapsulated radioactive material shall not be stored in these cells when the cell is unattended.
Material may only be stored inside these cells in welded capsules or screw top capsules.
When radioactive material is stored in these cells, a radioactive material tag identifying the types, quantities, locations and storage cates of all such material shall be l
attached to the manipulator or to the cell body adjacent to the window.
5. When any "through the wall" tool-is removed, the opening shall be closed with the plug provided.
All tools shall be decontaminated whenever they are removed from the cell.
6. Each individual performing this operation must wear a film badge and pocket dosimeter at waist level _and a second film badge and pocket dosimeter in the. vicinity of the head.
All operations must be monitored with a calibrated and operational radiation survey meter.
C.
Precaratory Procedure
1. Record the names and initial pocket dosimeter readings for the personnel performing the loading operation on the Loading Log Sheet.
()
2. Check the cell lights for preper operation.
Check the cell 1
manipulators both visually and operationally.
Assure that all cell ports are plugged.
3. Assure that the exhaust system is operational.
Record the manometer reading on the Loading Log Sheet.
If the manometer reading is less than 0.5 inch or greater -than 2.0 inches of water, the filter must be changed.
I
4. Assure that the air scmpling system is operational and that sample filters are in place.
5. Perform the preoperational contamination survey as indicated on the Loading Log Sheet.
Record the results on the Loading Log Sheet.
6. Perform the encapsulation procedure omitting the insertion of any activity.
Examine this dummy capsule weld.
If this weld is acceptable, preparation of active capsules may proceed.
If the weld is not acceptable, the condition responsible for this unacceptable weld must be corrected prior to proceeding.
This step must also must performed each time the welding electrode is changed.
4 Revisim 0 7-4 i March 198 5
_,.m
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--..,,,,,,-.,-,.w,,
D.
Encapsulation Procedure int Burlinoton. MA'Pacility
1. Prior to use, assemble and visually inspect the two capsule components to assure the weld zone does not exhibit any misalignment and/or separation.
Defective capsules shall be 7-rejected.
t, ]s
2. Degrease capsule components in the Ultrasonic Bath, using isoprop31 alcohc1 as degreasing agent, for a period of 30 minutes.
Dry the capsule components at 100
C for a minimum of 20 minutes.
3. Insert capsule components irto hot cell with the posting bar.
4. Place capsule bottom in weld positioning device.
Withdrew the posting bar.
5. Move the drawer bar of the source transfer container into the loading cell.
Open the screw top capsule.
6. Withdraw the proper amount of activity f rom the screw top cap-sule and place it in the capsule bottom.
A brass rivet must be used with wafers to prevent contamination of weld zone.
7. Assure that all unused radioactive material is removed f rom the loading cell by installing it in the screw top capsule and withdrawing the drawer bar of the source transfer container f rom the cell.
8. Remove the rivet (if applicable).
9. Assemble the capsule components.
'J
10. Weld adhering to the written welding procedure for the capsule being welded.
11. Visually inspect the weld.
An acceptable weld must be continuous without cratering, cracks or evidence of blow out.
If the weld is defective, the capsule must be cleaned and rewelded to acceptable conditions or disposed of as radioactive
~
l waste.
12. Check the capsule in the height gauge to be sure that the weld is at the center of the capsule.
13. Wipe the exterior of the capsule with a flannel patch wetted with EDTA solution or equivalent.
14. Count the patch with the scaler counting system.
The patch must show no more than.005 microcurie of contamination.
If the patch shows more than.005 microcurie, the capsule must be cleaned and rewiped.
If the rewipe patch still shows more than 0.005 of contamination, steps 10 through 14 must be repeated.
Revision 0 7-5 1 March 1985 a
i
.e.,
,w
15. Vacuum bubble test the capsule.
Place the welded' capsule in a glass _ vial containing isopropyl alcohol.
Apply a vacuum of 15 in Hg(Gauge).
Any visual detection of bubbles will indicate a leaking source.
If the scarce is determined to be leaking, place the source in a dry vacuum vial and boil of f the residual alcohol.
Reweld the capsule; repeat steps 10 through 15.
r-wg O
16. Transfer the welded source capsule to the sealed source section of the loading cell.
17. For wire mounted source capsules, transfer the capsule to the swaging fixture.
Insert the wire and connector assembly and swage.
Hydraulic pressure should not be less than 1250 nor more than 1500 pounds.
For source holder mounted source capsu'les, transfer the capsule to the appropriate source holder loading fixture.
Insert the source capsule into the source holder.
Screw the source holder together and install the roll pin.
Check to assure that the pin does not protrude on either side.
18. Apply the tensile test to assembly between the capsule and connector by applying proof load of 100 lbs.
Extension under the load shall not exceed 0.05 inch.
If the extension exceeds 0.05 inch, the source must be disposed of as radioactive waste.
19. Assure that the cell tunnel door is closed.
Position the source in the exit port of the loading cell.
Use the remote control to insert the source into the ion chamber and position the source for maximum response.
Record the meter reading.
Compute the activity in curies and fill out a temporary source tag.
()
20. Again using remote control, eject the source from cell into source changer through the tube gauze wipe test fixture.
Monitor the radiation level as the source changer shielded door is opened.
Remove the tube gauze and count with scaler counting system.
This assay must show no more than 0.005 microcurie.
If contamination is in excess of this level, the source is leaking and shall be rejected.
21. At the end of the day's operations, perform the post-operational contamination survey as indicated on the Loading Log Sheet.
Record the results on the Leading Log Sheet.
22. Record the final pocket dosimeter readings for the personnel performing the loading operation on the Loading Log Sheet.
23. Record the daily Air Sample results on the Loading Log Sheet.
Revision 0 7H5 1 March 1985
F.
Encapsulation Eroccdnig fnr Phoenixville, EA Facility
1. Prior to use,- assemble and visually inspect the two capsule components to at sure the weld zone does not exhibit any Defec'ive capsules shall be misalignment and/or separation.
t rejected.
2. Degrease capsule components in the Ultrasonic Bath, using isopropyl alcohol as degreasing agent, for a period of 30 minutes.
Dry the capsule components at 100* C for a minimum of 20 minutes.
3. Insert capsule components into hot cell with the posting bar.
4. Place capsule bottom in weld positioning device.
Withdraw the posting bar.
5. Remove a storage rod from the storage pit in the cell floor.
Remove the threaded end cap from the storage rod and remove the screw top capsule.
Open the screw top capsule.
6. Withdraw the proper amount of activity from the screw top capsule and place it in the welded capsule bottom.
A brass rivet must be used with wafers to prevent contamination of the weld zone.
7. Assure that all unused radioactive material is removed from the hot cell by installing it-into the screw top capsule.
Close the screw top capsule and place it into the threaded end cap of the storage rod.
Reinstall the end cap onto the storage rod.
Replace the storage rod into the storage pit in the cell floor.
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8. Remove the rivet (if applicable).
9. Assemble the capsule components.
10. Weld adhering to the written welding procedure for the capsule being welded.
4
11. Visually inspect the weld.
An acceptable weld must be continuous without cratering, cracks or evidence of blow out.
If the weld is defective, the capsule must be cleaned and rewelded to acceptable conditions or disposed of as radioactive waste.
i
12. Check the capsule in the height gauge to be sure that the weld is at the center of the capsule.
13. Wipe the exterior of the capsule with a flannel patch wetted with EDTA solution or equivalent.
' Revision 0 7-7 1 March 1985 tks
14. Count the patch with the scaler counting system.
The patch must show no more than.005 microcurie of contamination.
If the patch shows more than.005 microcurie, the cepsule must be c3aaned and rewiped.
If the rewipe patch still shows more than 0.005 of contamination, steps 10 through 14 must be repeated.
O
(
j
15. Vacuum bubble test the capsule.
Place the welded capsule in a glass vial containing isopropyl alcohol.
Apply a vacuum of 15 in Hg(Gauge).
Any visual detection of bubbles will indicate a leaking source.
If the source is determined to be leaking, place the source in a dry vacuum vial and boil of f the residual alcohol.
Reweld the capsule; repeat steps 10 through 15.
16. Transfer the welded source capsule to the sealed source section of the hot cell.
17. For wire mounted source capsules, transfer the capsule to the swaging fixture.
Insert the wire and connector assembly and swage.
Hydraulic pressure should not be less than 1250 nor more than 1500 pounds.
For source holder mounted source capsules, -transfer the capsule to the appropriate source holder loading fixture.
Insert the source capsule into the source holder.
Screw the source holder together and install the roll pin.
Check to assure that the pin does not protrude on either side.
18. Apply the tensile test to assembly between the capsule and connector by applying proof load of 100 lbs.
Extension under the load shall not exceed 0.05 inch.
If the extension exceeds 0.05 inch, the source must be disposed of as radioactive waste.
I
19. Assure that the cell tunnel door is closed.
Position the source s/
in the exit port of the loading cell.
Use the remote control to insert the source into the ion chamber and position the source for maximum response.
Record the meter reading.
Compute the activity in curies and fill out a temporary source tag.
20. Again using remote control, eject the source from cell into source changer through the tube gauze wipe test fixture.
Monitor the radiation level as the source changer shielded door is opened.
Remove the tube gauze and count with scaler counting system.
This assay must show no more than 0.005 microcurie.
If contamination is in excess of this level, the source is leaking and shall be rejected.
21. At the end of the day's operations, perform the post-operational contamination survey as indicated on the Loading Log Sheet.
Record the results on the Loading Log Sheet.
22. Record the final pocket dosimeter readings for the personnel performing the loading operation on the Loading Log Sheet.
23. Record the daily Air Sample results on the Loading Log Sheet.
Revision 0 7-8 i March 1985
i Tech / Ops, Inc.
O Rad;ation Products DMnon 40 North Avenue 7.4.2 MODEL AI 500 SU l
surungion. uanachusetts oisas (qt)
Twephone (su) 272-2000 SOURCE CHANGER OPERATION MANUAL NOTICE This device is used as a radiographic source changer and Type i
B(U) transport package for Tech / Ops, Inc. radioactive sources I
listed in this manual.
.Tha user should hacDag thoroughly familiar with.the instruction manual before attempting operation 2f the ecuioment.
In order to use this equipment to perform source changes within the United States, the user must be specifically licensed to do so.
Application for a license should be filed on Form NRC-313R with the appropriate U.S.
Nuclear Regulatory Commission Regional Office listed in Appendix D of 10 CFR 20 or with the appropriate agreement state office.
Prior to initial use of the source changer as a transport package, the user in the United States must register his name, license number and package identification number with:
Director
\\#
Office of Nuclear Material Safety and Safeguards U.S. Nuclear Regulatory Commission Washington, DC 20555 The user must have in his possession a copy of USNRC certificate of Compliance Number 9006 issued for this package.
Prior to t.he first export shipment of this source changer f rom the United States, the user must also register his identity with:
Office of Hazardous Materials Regulation Materials Transportation Bureau U. S. Department of Transportation Washington, DC 20590 The user must have in his possession a copy of International Atomic Energy Agency Certificate of Competent Authority Number USA /9006/B(U)T issued for this source changer.
Users of this equipment outside the United States must comply with the regulatory, licensing and transportation rules and regulations as they apply in their respective countries.
IU 7-9 Revision 1
..19 April 1985
e i
I General b)
The AI Model 500 SU source changer is used primarily for the transfer of encapsulated radioactive sources into radiographic exposure devices.
The source changer is designed to contain the radioactive sources during transport and to permit the field exchange of sources.
The source changer is five inches (127 mm) wide, six inches (152 mm) high and eleven inches (279 mm) long.
The total weight of the source changer is 60 pounds (27 kg).
The source changer contains 39 pounds (18 kg) of depleted uranium as shielding.
The source changer is approved as a Type B(U) transport package under USNRC Certificate of Compliance Number 9006 and IAEA Certificate of Competent Authority Number USA /9006/B(U)T.
The capacity of the source changer is 120 curies of iridium-192 as one of the source assemblies listed in Table 1.
Shioment Data An envelope accompanies each shipment and contains a source certificate with decay data and leak test certification, a source identification plate for attachment to the user's radiographic exposure device, return shipping labels, tamper proof seals and an instruction manual.
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Padiation Safety Considerations Pursuant to USNRC and agreement state regulations, all personnel present during radiographic and source changing operations are required to wear a direct reading pocket dosimeter and either a film badge or a thermoluminescent dosimeter (TLD).
The pocket dcsimeter must be recharged at the start of each shift.
The operator should f requently check the pocket dosimeter reading throughout the shift.
Dosimeter readings must be recorded at the end of each shift.
Records of the initial and final readings of the pocket dosimeter must be kept for inspection by the USNRC.
In the event that a person's pocket dosimeter is found to be off scale, that person must stop all work with radiation immediately.
His film badge (or TLD) must be sent in immediately for processing, and he must not reenter a restricted area until it has been determined that he received less than the maximum allowed occupational exposure as defined in 10 CPR Part 20.101.
Personnel performing source changing operations should also have a calibrated and operable radiation survey meter capable of measuring from 2mR/hr to at least 1000 mE/hr to determine radiation levels when performing these operations, m[V i
Areas in which source changing is performed must be identified.
If a permanent radiographic installation is used, it must have the appropriate personnel access control devices as defined in 10 Revision 0 7-10 1 Mar 1985
e.
CFR 20.203.
Otherwise, certain areas must be established as follows:
Access to the Restricted Area must be controlled.
A Restricted Area is defined in 10 CFR 20.105 as the area where an individual could receive an exposure in excess of two milliroentgens in any one hour, or 100 milliroentgens in seven consecutive days or 500 milliroentgens in one year.
The Restricted Area should also be posted with signs reading
" Caution (or Danger) - Radiation Area."
Signs reading
" Caution (or Danger) - High Radiation Area" should be posted around the perimeter where an individual could receive an exposure in excess of 100 milliroentgens in any one hour.
The radiographer or radiographer's assistant must guard against unauthorized entrance into these areas at all times.
No personnel should be allowed into the restricted area without a direct reading pocket dosimeter and either a film badge or TLD.
Receiot nf Radioactive Material The consignee of a package of radioactive material must make arrang.ments to receive the package when it is delivered.
If the package is to be picked up at the carrier's terminal, 10 CFR Part 20.205 requires that this be done expeditiously upon notification
-(s) of its arrival, wJ Upon receipt, survey the source changer with a survey meter as soon as possible, preferably at the time of pickup and no more than three hours later if it was received during working hours, or no more than 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months later if it was received af ter normal working hours.
Radiation levels should not exceed 200 milliroentgens per hour at the surface of the source changer nor 10 milliroentgens per hour at a distance of three feet f rom the surface.
Actual radiation levels should be recorded on the receiving report. If the radiation levels exceed these limits, the container should be secured in a Restricted Area, and the appropriate personnel notified.
All components should be inspected for physical damage.
The radioisotope, activity, model number and serial number of the source and the package model number and serial number should be recorded.
Ooeration 1.
Locate the source changer and the radiographic exposure device in a restricted area.
Arrange them so that they can be connected by the source guide tube extension which accompanies the source changer.
tC/
2.
Locate the control unit to the radiographic exposure device as far away as possible from the exposure device and preferably behind radiation shielding.
Revision 0 7-11 1 Mar 1985
.=_..
1 3.
Unlock the source.. changer and open the cover.
Remove the 4
envelope containing the documentation and remove the source
.[
}
guide tube extension.
4.
Connect the drive cable to the source assembly in the exposure device and connect the control unit to the exposure l
devie in accordance with the manufacturer's instructions and the company's operating and emergency procedures.
4 5.
Connect the guide tube extension to the radiographic exposure device and to the fitting of the source changer i
over the empty tube.
Raise the lock slide to the OPEN position.
4 6.
Assure no unauthorize' personnel are in the restricted area.
At the expocure devic
controls, crank the source assembly a
rapidly from the txposure device to the source changer.
observe t.he survey meter during this operation.
The radiation intensity should greatly increase as the source is first exposed, decrease slightly as the soruce moves through the guide tube and reduce to background when the source enters the source changer.
7.
Approach the exposure device with a survey meter; survey the i
exposure device on all sides, survey the guide tube and survey the source changer on all sides to assure the source 4
has been properly transferred and stored.
The radiation level should be less than 200 mR/hr and at the surface of l
the source changer and less than le mR/hr at one meter fror the surface of the source changer.
x Whilemaintainingpositiveforwardforcel 8.
Move the lock slide to the LOCK position (down).
Engage the key operated lock.
on the drive cable and source
assembly, disconnect the source guide tube extension from the source changer and expose the drive cable to source assembly ' connection.
Disconnect the drive cable from the. source assembly.
Install the holddown cap over the source assembly.
)
9.
Remove the holddown cap from the other source assembly.
[
Connect the drive cable to the source assembly.
Connect the guide tube extension to the source changer fitting.
Unlock i
the key operated lock and raise the lock slide to the OPEN position.
l 10.
Assure no unauthorized personnel are in the restricted area.
At the exposure device controls, crank the source assembly rapidly from the source changer to the exposure device.
1 l
7-12 Revision 2 l
8 August 1985 s
O l
f l
Observe the survey meter during this operation.
The f.
radiation intensity should greatly increase as the source is first exposed, increase slightly as the source moves through the guide tube and decrease to background i.s the source assembly enters the exposure device.
11.
Approach the exposure device with the survey
meter, survey the exposure device on all sides, survey the length of the guide
tube, and survey the source changer on all sides to assure the source has been properly transferred to its storage position in the exposure device.
Radiation levels should be less than 200 mR/hr at the surface and less than le mR/hr at one meter from the surface of the exposure device and the source changer.
12.
Lock the exposure device.
Move the lock slide to the LOCK position.
Remove the guide tube extension from the source changer and exposure device and install it in the compartment of the source changer.
13.
Affix the identification plate of the new source to the exposure device.
Place the identification plate for the old source in the source changer.
14.
Assure that the proper holddown cap is installed over the source in the source changer (see Table 1).
Seal wire the cap in place.
Assure that the key operated locks are
()
engaged.
15.
Close the source changer cover and seal wire and lock the cover in place.
ShipatD% 91 2B619bG%i29 591999 1.
Assure that the source assembly is secured in the proper storage position and the source changer is locked.
Assure that the soruce changer cover is sealwired with a
tamperproof seal.
2.
If the source changer is to be packaged ina crate or other outer packaging, the outer packaging must be strong enough to withstand the normal conditions of transport.
These requirements are outlined in 10 CFR 71.71.
The source changer should be put in the outer package with sufficient blocking to prevent shifting during transportation.
7-13 Revision 2 8 August 1985 O
3.
Survey the package with a survey meter at the surface and at a
distance of one meter from the surface to determine the O
proper radioactive shipping labels to be applied to the package as required by 49 CFR Part 172.403.
The radiation-exposure limits for each shipping label are given in figure 1.
If radiation levels above 200 mP/hr at the surface or le n3/hr at one meter from the surface are
measured, the i
package must not be shipped.
4.
Properly complete two shipping lables indicating the radioisotope, activity and the Transport Index.
The Transport Index is used only on Yellow II and Yellow III labels and is defined as the maximum radiation level in milliroentgens per hour measured at a distance of one meter from the surface of the package.
Put these two labels on opposite sides of the package after making sure any previous labels have been removed.
The package should be marked with the proper shipping name (Radioactive
Material, Special s'
/'.
Form, n.o.s.) and the identification number (UN 2974).
If g
the source changer is packaged inside an outer container, mark the outside package "INSIDE PACKAGE COMPLIES WITH O
i i
7-14 Fevision 2 8 August 1985
O I
1
_ _ _. _ _ ~,,
O i
Maxinum Radiation Level at Surface at One Meter Radioactive White I e.4 O
0.5 mR/hr None RAW 0 ACTIVE I j 7
Radioactive Yellow II 4.4 O
50 mR/hr 1.0 -' 'Sr j
RADIDACTIVE 11/
l Q '~
7 Radioactive Yellos III l
4.4 O
200 mR/hr 10 mR/hr RA010ACilVE Illj2' L_ J
'I i
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l l
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i Revision 0 7-15 i Har 1985 i
i i
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O PRESCRIBED SPECIFICATIONS-TYPE B (U) USA /9996/B (D). "
5.
Complete the appropriate shipping papers - Examples are shown in Figure 2 and 3.-
These shipping papers must include:
a.
Proper Shipping Name (Radioactive Material, Special Form, n.o.s.) and Identification Number (UN2974).
b.
Name of Radionuclide (Iridium-192) c.
Activity of the Source (in Curies) l d.
Category of Label Applied (i.e. Radioactive Yellow II) e.
Transport Index f.
Package Identification Number (i.e.
USA /9006/B(U) Type B) g.
Shipper's Certification "This is to certify that the above named materials are properly classified, described, packaged, marked and labeled and are in proper condition for transport according to the applicable regulations of the Department of Transportation."
l Notes:
1.
For air shipments, the following shipper's j
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 according to j
applicable national governme. ital regulations."
2.
For air shipments to, f rom or through the United States, a " CARGO AIRCRAFT ONLY" label and the shipping papers must state:
"THIS SHIPMENT IS WITHIN THE LIMITATIONS PRESCRIBED FOR CARGO ONLY AIRCRAFT.*
l 6.
Due to the depleted uranium used as shielding in the source l
changer, a notice must also be enclosed in or on the package included with the packing list, or otherwise forwarded with the package.
This notice must include the name of the consignor or consignee and the following statement:
j O
"This package conforms to the conditions and limitations specified in 49 CPR 173.424 for excepted radioactive material, articles manuf actured f rom depleted uranium, UN i
2909."
}
l Revision 0 7-16 I Har 1985 l
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STRAIGHT BILL CF LADING. SHORT FCRM Shipper's ORIGINAL - NOT NEGOTIABLE Carrier's No.
CARRIER:
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RADIATION PROD. DIV. OF
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Consigned to Radiation Products Division D'5tination 40 North Avenue Burlington, Massachusetts 01803 USA Car initials and Number Route f
p[,84, I Kind of Pecaege, oescription of Articles, Specisi Marks and Enceptione g
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Metal Container ( AI Model 500 SU )
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a Radioactive Material, Soecial Form N.O.S. UN2974 l
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RELEASED VALUE NOT TO EXCEED 40s PER POUND This is to certify that the above named materials are properly classified, described, packaged, marked and labeled and are in proper condition for transportation according to the applicable regulations of the Department of Transportation.
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Tech / Ops, Inc.
IhIEP'I* Per Agent Per Po,menent po.t-office sodress of shipper 40 North Avenue Burhngton. Massachusetts Ot803 U.S. A.
Revision 0 l
7-17 i Mar 1985
SHIPPER'S DECLAR ATION FOR DANGEROUS GOODS 3
j Air Waybill No.
Page i of I Pages Shipper's Reference Number 1
.topoonso Consignee Tech / Ops, Inc.
Radiation Products Division 40 North Avenue Burlington, Massachusetts 01803 USA Two completed and signed cooses of mis Declaration must be handedto the operator WARN lNG TRANSPORT DETAILS Failure to comply in all respects with the applicable This shipmentis wimen me Dangerous Goods Regulations may be in breach of the Airport of Departure:
applicable law, subject to legal penalties. This Declaration tmetations presenbed for:
must not, in any circumstances, be completed and/or l'**#8*'
signed by a consolidator, a forwarder or an IATA cargo N
CARGO agent.
$NLY Shepment type me r, ann.e.no6eewi Airport of Destination:
BOSTON l
l RADIOACTIVE l NATURE AND QUANTITY OF DANGEROUS GOODS Dangerous Goocs toensscaten j
j
sues. i.
&*nt'ry and
! Pacson9 f Autrio -
Class ;
UN o,
type of pacsong inst o,
Proper snco ng Name
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. son No
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......................'...j....
7 l UN2974 i l Iridium 192 Metal Solid
II i Special Radioactise Material Special Form N.O.S.
l l Yellow iFonn l
l n curies j j Certificate
USA /0154/S T. I. : Type B(U) j Package l
l 0.4 j Certificate l
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j j USA / 9006B(U) l l1 Type B(U) Packag-l l
{ Dimensions; l
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5 in x i
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.b_______
Addtional Handling Information I hereby declare that the contents of this consignment are fully and accurately Name/ Title of Signatory desenbed above by proper shipping namc and are classified, packed, marked Radiologica1 Technician and labelled. and are in all respects in the proper condition for transport by air Place and Date s
I accoreng to the apphcable Intemational and National Govemment Regula-g Burlington, M A tions Signature (see wanng above) 7-18 Revision 0 1 Mar 1985
7.
For shipment of an empty source changer, assure that there is no source in the container.
If the radiation level is below 9.5 mR/hr at the surface, and there is no measurable radiation level at one meter from the container, no label is required.
Mark the outside of the package with the proper shipping name (Radioactive material, articles manufactured f rom depleted uranium UN 2999).
Mark the outside of the package:
l
" Exempt from specification packaging, shipping paper and certification, marking and labeling and exempt from the requirements of Part 175 per 49 CFR 173.421-1 and 49 CPR 173.424."
Additionally, a notice must be enclosed in or on the package included with the packing list or otherwise forwarded with the package.
This notice must include the name of the consignor or consignee and the statement:
"This package conforms to the conditions and limitations specified in 49 CFR 173.424 for excepted radioactive materials, articles manufactured from depleted uranium, UN 2909."
8.
Return the container to Tech / Ops, Inc. according to proper g
procedures for transporting radioactive material as
.)
established in Title 49 Code of Federal Regulations part 172-178.
Note:
The U. S. Department of Transportation, in 49 CPR 173.22 (c) requires each shipper of Type B quantities of radioactive material to provide prior notification to the consignee of the dates of shipment and expected arrival.
l O
l
~
Revision 0 7 - 19 1 Mar 1985
e.
Table 1 Source Assemblies Used In Conjunction With Automation Industries Model 500 SU Source Changer Automation Industries Source Assembly Designs 41701, Al 200-520-008, N1 39990, A10
[ Cap N]
200-520-009, N2 130013, All 200-520-010, N3
[ Cap N]
200-520-008, N4 l
41706, B1 200-520-011, N5 41706, B2
[ Cap B & N]
41706, B3 36910, T1
[ Cap T]
36910, T2
[ Cap T]
O~s 41708, C1 200-660-809, T3 (Cap N]
41708, C2
[ Cap C & N]
36910, T5
[ Cap T]
41708, C3 200-660-009, T6
[ Cap N]
41039, H1
[ Cap H & N]
39998, G1
[ Cap N]
41045, J1 (Cap J & NJ 39998, G2
[ Cap N]
39998, G3
[ Cap G]
39998, GS
[ Cap N]
39998, G6
[ Cap G]
Tech / Ops Source Assembly Designs A424-1
[ Cap T]
A424-9
[ Cap N]
814 (Cap NJ 848 (Cap G]
866
[ Cap P]
[ Cap X]:
Holddown cap identification for use with Model 500 SU Source Changer l
O Revision 0 7-20 1 Mar 1985
i 8.0 Acceptance ZRBh2 ads Enlnkanancs ExDgzam 8.1 ACSRptADc2 Z2BkB.
8.1.1 Visual InspssilpD The package is visually examined to assure that the appro-priate fasteners are properly seal wired.
The package is inspected to assure that the proper marking and labeling is present.
The seal weld of the radioactive source capsule is visually inspected for proper closure.
8.1'. 2 Struchusal and ExnanDxs Z2BLs The swage coupling between the source capsule and cable is subjected to a
static tensile test with a load of one hundred pounds.
8.1.3 Leak Zasts The radioactive source capsule, which serves as the primary containment, is wipe tested for leakage of radioactive contamination.
The source capsule is subjected to a vacuum bubble leak test.
These tests are described in Section 7.4.1.
Failure of either of these tests will prevent use of this source assembly.
8.1.4 Component Zmsts The lock assembly (source holddown assembly) of the package is tested to assure that the security of the source will be maintained. A simulated (dummy) source assembly is installed in the source changer and the appropriate source holddown cap is installed.
Verification is made that the source holddown cap contacts the source assembly prior to the threads being fully bottomed.
A check is made that the holddown cap is securely attached to the device.
Additionally, the locking arrangement and lock slides are tested to assure that the lock slide secures the source assembly and the lock secures the lock slide.
Failure of any of these tests will prevent use of the package until the-cause of the failure is corrected and' retested.
8.1.5 Tests isI Shleiding Imingxlty With the package containing a source assembly, the radia-tion levels at the surface of the package and at one meter from the surface of the package are measured using a small detector survey instrument.
These radiation levels, when 8-1 Revision 2 8 August 1985 O
{
l extrapolated to the rated capacity of the package, must not exceed 200 milliroentgens per hour at the surface of the
,(
package.
8.1.6 Zbssmal bcssakangs Zsska Not applicable.
8.2 BalDkRDDDs2 2199Ebs 8.2.1 Skzusku1A1 DDb EERan9E9 2RBkn Not applicable.
8.2.2 Leah Issks As described in Section 8.1. 3,
the radioactive source assembly is leak tested at manufacture.
Additionally, the source assembly is wipe tested for leakage of radioactive contamination every six months.
8.2.3 Buksysham BalDLRDDDs2 The source holddown 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, proper seal wires, and general condition prior to each use.
8.2.4 yalyss, Euphyrs Dissa, and gashghs 1
Not applicable.
8.2.5 ShiglBiDg 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 surface nor ten milliroent-gens per hour at three feet from the surface.
8.2.6 TbgImal Not applicable.
8.2.7 BissellAD29Ds Inspections and tests designed for secondary users of this package under the general license provisions of 10 CPR71.12 (b) are included in Section 7.4 8-2 Revision 1 19 April 1985
._

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