ML21279A215
| ML21279A215 | |
| Person / Time | |
|---|---|
| Site: | 07103099 |
| Issue date: | 02/29/2020 |
| From: | Govt of Canada, Nuclear Safety Commission |
| To: | Office of Nuclear Material Safety and Safeguards |
| Nishka Devaser NMSS/DFM/STL 301-415-5196 | |
| Shared Package | |
| ML21279A212 | List: |
| References | |
| QTP-001 SAR 2020-1, Rev 0 | |
| Download: ML21279A215 (336) | |
Text
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 0-1 Safety Analysis Report 2020-1 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 0-2 Table of Contents The Table of Contents is located at the beginning of each Section of the SAR Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 0-3 Section 0 - Review Statement 0.1 Review Statement The ISO-RAD-TC Safety Analysis Report (SAR) ISO-RAD Canada Inc (ISO-RAD) respectfully requests this Safety Analysis Report (SAR) application be reviewed under the RD-364/NUREG-1886 The Joint Canada-United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages (NUREG-1886). The SAR has been prepared according to the RD-364/NUREG-1886 document.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-1 Safety Analysis Report Section 1 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-2 Section 1 Contents Section 1 Contents ................................................................................................................................ 1-2 Section 1 GENERAL INFORMATION ............................................................................................... 1-3 1.1 Introduction ............................................................................................................................ 1-3 1.2 Transport Package Description .............................................................................................. 1-3 1.2.1 Packaging ........................................................................................................................ 1-4 1.2.1.1 Outer Drum Assembly ............................................................................................. 1-4 1.2.1.2 Bulk and PIC Inner Container (BPIC) - Square ...................................................... 1-6 1.2.1.4 Multi Port Inner Container (MPIC) - Square........................................................... 1-9 1.2.1.5 Multi Port Inner Container (MPIC) - Round ......................................................... 1-10 1.2.1.6 BPIC 2835A .......................................................................................................... 1-11 1.2.2 Contents ........................................................................................................................ 1-12 1.2.3. Special Requirements for Plutonium ............................................................................ 1-14
1.3 Appendix
............................................................................................................................. 1-14 1.3.1 References ..................................................................................................................... 1-14 1.3.2 Drawings of the ISORAD-TC1 transport package ....................................................... 1-16 1.3.2.1 R180831-100 ISORAD-TC1 Outer Drum Assembly - Empty ............................. 1-16 1.3.2.2 R180831-101 ISORAD-TC1 Outer Drum with BPIC - Square Assembly .......... 1-18 1.3.2.3 R180831-200 ISORAD-TC1 BPIC - Square Assembly ....................................... 1-19 1.3.2.4 R180831-201 ISORAD-TC1 BPIC - Square Assembly (No Lid & Plug) ........... 1-20 1.3.2.5 R180831-101-4 ISORAD-TC1 Outer Drum with BPIC - Round Assembly ....... 1-22 1.3.2.6 R180831-400 ISORAD-TC1 BPIC - Round Assembly ....................................... 1-23 1.3.2.7 R180831-401 ISORAD-TC1 BPIC - Round Assembly (No Lid & Plug) ............ 1-24 1.3.2.8 R180831-102 ISORAD-TC1 Outer Drum with MPIC - Square Assembly ......... 1-26 1.3.2.9 R180831-300 ISORAD-TC1 MPIC - Square Assembly ...................................... 1-27 1.3.2.10 R180831-301 ISORAD-TC1 MPIC - Square Assembly (No Lid & Caps) ......... 1-28 1.3.2.11 R180831-102-6 ISORAD-TC1 Outer Drum with MPIC - Round Assembly ...... 1-30 1.3.2.12 R180831-600 ISORAD-TC1 MPIC - Round Assembly ...................................... 1-31 1.3.2.13 R180831-601 ISORAD-TC1 MPIC - Round Assembly (No Lid & Caps) .......... 1-32 1.3.2.14 R180831-101-5 ISORAD-TC1 Outer Drum with BPIC 2835A ........................... 1-34 1.3.2.15 R180831-500 ISORAD-TC1 BPIC 2835A ........................................................... 1-35 1.3.2.16 R180831-501 ISORAD-TC1 BPIC 2835A (No Lid & Plug) ............................... 1-36 1.3.2.17 R180831-XXX Sketch ISORAD-TC1 Prepared for Transport............................. 1-38 List of Figures Figure 1.2a - Isometric View of ISORAD-TC1 Packages .................................................................. 1-4 Figure 2 - Outer Drum External Parts .................................................................................................. 1-4 Figure 3 - Outer Drum Internal Parts ................................................................................................... 1-5 Figure 4 - BPIC Inner Container External Parts .................................................................................. 1-6 Figure 5 - BPIC Inner Container Internal Parts ................................................................................... 1-6 Figure 6 - MPIC Inner Container External Parts ................................................................................. 1-8 Figure 7 - MPIC Inner Container Internal Parts .................................................................................. 1-9 List of Tables Table 1.2a: Isotopes Permitted in the Model ISORAD-TC1 BPIC ..................................................... 1-13 Table 1.2b: Isotopes Permitted in the Model ISORAD-TC1 MPIC .................................................... 1-14 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-3 Section 1 GENERAL INFORMATION 1.1 Introduction ISO-RAD Canada Inc (ISO-RAD) respectfully requests this Safety Analysis Report (SAR) application be reviewed under the RD-364/NUREG-1886 The Joint Canada-United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages (NUREG-1886). The SAR has been prepared according to the RD-364/NUREG-1886 document.
The ISORAD-TC1 Transport Container (ISORAD-TC1) is designed as a transport package (package) and storage container for Type B quantities of special form radioactive materials primarily, but not limited to, Iridium-192 (Ir-192), Selenium-75 (Se-75), Ytterbium-169 (Yb-169), Cesium-137 (Cs-137),
and other radioactive isotopes listed in Tables 1.2a and 1.2b. The package conforms to the General Package Requirements, Type A Package Requirements, and Type B(U) Package Requirements criteria for packaging in accordance with IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6), IAEA Regulations for the Safe Transport of Radioactive Material (TS-R-1), 10 CFR 71, and 49 CFR 173 which were in effect at the time of sign-off of this report. The package was designed and tested to verify compliance with the Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions of Transport (HACT) requirements and tests.
The ISORAD-TC1 package has been designed to withstand the Type A Package (NCT) and Type B Package (HACT) specifications and testing requirements of the International Atomic Energy Agency (IAEA), the Canadian Nuclear Safety Commission (CNSC), the United States Nuclear Regulatory Commission (USNRC), the United States Department of Transportation (USDOT), United Nations (UN) Economic Commission for Europe Committee for Inland Transport, International Air Transport Association (IATA), and other regulatory bodies governing transport packages. The regulations and guides used in the design development of the ISORAD-TC1 are IAEA SSR-6:2018, IAEA TSR-1, SOR/2015-145 CNSC Packaging and Transport of Nuclear Substances Regulations (PTNS Regulations), RD-364/NUREG-1886 The Joint Canada-United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages (NUREG-1886), the 2019 European Agreement concerning the International Carriage of Dangerous Goods by Road (ADR), IATA Dangerous Goods Regulations.
1.2 Transport Package Description
[IAEA SSR-6 220 & 809, IAEA TS-R-1 220 & 807, and 10 CFR 71.33]
The ISORAD-TC1 package, shown in Figure 1.2a, is constructed in accordance with descriptive drawings R180831-100, R180831-200, R180831-300, R180831-400, R180831-500, and R180831-600 in Section 1.3. Its general dimensions are approximately 589 mm (23.1875 inches) high, 403.24 mm (15.88 inches) diameter, and has a maximum weight of 136.08 kg (300 pounds) while configured to transport the BULK and PIC - Square Inner Container assembly (BPIC).
The Outer Drum of the ISORAD-TC1 package is designed to operate as a Type B(U) outer container with multiple inner package configurations, that provide inner shielding and security of approved special form capsules of radioactive material.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-7 1.2.1.2.2 BPIC Top Plate The top plate encloses the top portion of the Outer Casing. The BPIC Top Plate is constructed of austenitic stainless steel that is approximately 170 mm (6.693in) wide, 170 mm (6.693in) deep and 24.35mm (0.959in) thick. The Top Plate contains eight threaded and tapped M8 x 1.25 holes to secure the BPIC Lid, four threaded and tapped M6 x 1 holes to secure the BPIC Plug Assembly, and one hole approximately 61.75 mm (2.431in) diameter to allow the insertion and removal of the BPIC Plug Assembly.
1.2.1.2.3 BPIC Lid The BPIC Lid encloses the top portion of the BPIC Assembly. The BPIC Lid is constructed of austenitic stainless steel that is approximately 170 mm (6.693in) wide, 170 mm (6.693in) deep and mm ( in) thick. The BPIC Lid contains eight recessed holes to accommodate eight M8 x 1.25 bolts to secure the BPIC Lid to the BPIC Assembly and one hole threaded and tapped M8 x 1.25 for the removable eyebolt.
1.2.1.2.4 BPIC Shield The BPIC Shield is either depleted uranium (DU) or tungsten and has a minimum shield thickness of approximately 62.0 mm (2.44in) in every direction. The shield is approximately mm +/- mm
( in +/- in) high and mm +/- mm ( in +/- in) diameter. The DU version has a titanium inner cylinder or tungsten shield has an inner titanium or stainless steel cylinder that has an interior dimension of approximately mm +/- mm ( in +/- in) diameter and mm +/-
mm ( in +/- in) tall. The DU shield rests on a brass bottom support, a brass cylinder that slips over the DU shield and has a brass top shield support which provides a eutectic barrier which prevents stainless steel and DU interaction. The brass creates stability and a eutectic barrier with prevents stainless steel and DU interaction. The tungsten version eliminates most brass pieces.
1.2.1.3 Bulk and PIC Inner Container (BPIC) - Round The BPIC Inner Container is approximately 278 mm (10.945in) high, 170 mm (6.693in) diameter. It has a maximum weight of 87.54 kg (193.0 pounds). It has an Outer Casing, Top Plate and Bottom Plate made of austenitic stainless steel which encases a brass clad depleted uranium (DU) shield or tungsten shield. The DU version has a titanium inner cylinder or tungsten shield has an inner titanium or stainless steel cylinder that has an interior dimension of approximately mm ( in) diameter and 73mm (2.87in) tall. The DU shield rests on a brass bottom support, a brass cylinder that slips over the DU shield and has a brass top shield support which provides a eutectic barrier which prevents stainless steel and DU interaction. The round version eliminates the brass corner spacers. The brass creates stability and a eutectic barrier with prevents stainless steel and DU interaction. The tungsten version eliminates most brass pieces.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-10 1.2.1.4.4 MPIC Shield Assembly The MPIC Shield Assembly creates the source cavities/channels for the source assemblies to be stored or transported. The MPIC Shield Assembly comprises two major parts the MPIC Base Insert and the MPIC Inner Insert and are constructed of either depleted uranium or tungsten. The tube configurations are either one circle of ten tubes or a circle within a circle of five tubes each. The MPIC Shield Assembly is approximately mm ( in) in diameter and mm ( in) tall. The approximate weight is 50.8 kg (111.99 pounds).
1.2.1.5 Multi Port Inner Container (MPIC) - Round The MPIC is approximately 278 mm (10.945in) high and 170 mm (6.693in) diameter. It has a maximum weight of 69.66 kg (153.57 pounds). The Outer Casing, Top Plate, and Bottom Plate are constructed of austenitic stainless steel. It has a two part inner cylindrical shield of depleted uranium (DU) clad in brass, or tungsten, which creates ten shielded cavities for Radioactive Material (RAM) in special form capsules installed in source holders. The cavities are lined with titanium tubes for the DU version and titanium or stainless steel for the tungsten version. Each source holder is secured with an individual locking mechanism and further secured with a threaded source cap. The DU shield rests on a brass base, brass sleeve, and is topped with a brass plate. This creates a eutectic barrier between the DU and stainless steel. The tungsten shield version eliminates some of the brass pieces.
1.2.1.5.1 MPIC Lid The MPIC Lid encloses the top portion of the MPIC Assembly. The MPIC Lid is constructed of austenitic stainless steel that is approximately 170 mm (6.693in) in diameter and mm ( in) thick. The MPIC Lid contains one recessed hole to accommodate the padlock and the M16 x 1.25 Hex Nut to secure the MPIC Lid to the MPIC Assembly and one through hole to slip over the Lock Screw Post.
1.2.1.5.2 MPIC Top Plate The MPIC Top Plate encloses the top portion of the Outer Casing. The MPIC Top Plate is constructed of austenitic stainless steel that is approximately 163.5 mm (6.437in) in diameter and mm ( )
thick. The Top Plate contains ten threaded and tapped M14 x 1.5 holes to mount and secure the Locking Mechanisms the MPIC Top Plate, one threaded and tapped M12 x 1.25 holes to secure the MPIC Lock Screw Post, which is used to secure the MPIC Lid into place.
1.2.1.5.3 MPIC Source Locking Mechanism The MPIC Source Locking Mechanism (SLM) is used to secure the sources in each of the ten source tube channels within the MPIC. The MPIC SLM is Lock Housing is constructed of austenitic stainless steel or titanium with the remaining parts being constructed of austenitic stainless steel. The MPIC SLM is approximately 26.5 mm (1.043 in) long and 24.78 mm (0.976 in) high.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-11 1.2.1.5.4 MPIC Shield Assembly The MPIC Shield Assembly creates the source cavities/channels for the source assemblies to be stored or transported. The MPIC Shield Assembly comprises two major parts the MPIC Base Insert and the MPIC Inner Insert and are constructed of either depleted uranium or tungsten. The tube configurations are either one circle of ten tubes or a circle within a circle of five tubes each. The MPIC Shield Assembly is approximately mm ( in) in diameter and mm ( in) tall. The approximate weight is 50.8 kg (111.99 pounds).
1.2.1.6 BPIC 2835A The BPIC 2835A Inner Container is approximately 248 mm (9.764in) high and 170 mm (6.693in) diameter. It has a maximum weight of 80 kg (176.37 pounds). It has an Outer Casing, Top Plate, Bottom Plate, and Source Cavity made of austenitic stainless steel which uses brass shims where necessary to protect the depleted uranium (DU) shield. The BPIC 2835A is the inner container from the Croft and Associates 2835A transport package. ISO-RAD has tested the package in its ISORAD-TC1 Outer Drum and has met the requirements. The BPIC 2835A has an inner austenitic stainless steel cylinder that has an interior dimension of approximately mm ( in) diameter and mm ( in) tall. The DU shield rests on brass shims as necessary. The design limits available oxygen preventing stainless steel and DU eutectic interaction. The O-ring seals are removed as the BPIC 2835A Inner Container will not be authorized for normal form radioactive material. The removal of the O-rings will enable the container to be open to the air as the other BPICs.
1.2.1.6.1 BPIC 2835A Plug Assembly The BPIC Plug Assembly encloses the top cavity in a DU plug shield. The BPIC Plug Assembly includes a DU Plug Shield that is encased in austenitic stainless steel and is inserted into the upper portion of the interior cavity of the lined DU shield. Brass shims are inserted as necessary in between the DU Plug Shield and the Plug Lid. The BPIC Plug Assembly is secured in place by the BPIC 2835A Lid with six stainless steel M8 x 1.25 bolts.
1.2.1.6.2 BPIC 2835A Top Plate The top plate encloses the top portion of the Outer Casing. The BPIC 2835A Top Plate is constructed of austenitic stainless steel that is approximately 170 mm (6.693in) in diameter and 20.9 mm (0.823in) thick. The Top Plate contains six threaded and tapped M8 x 1.25 holes to secure the BPIC Lid and one hole approximately 60.0 mm (2.36in) to allow the insertion and removal of the BPIC 2835A Plug Assembly.
1.2.1.6.3 BPIC 2835A Lid The BPIC 2835A Lid encloses the top portion of the BPIC Assembly. The BPIC Lid is constructed of austenitic stainless steel that is approximately 170 mm (6.693in) in diameter and mm ( in) thick. The BPIC Lid contains six recessed holes to accommodate six M8 x 1.25 bolts to secure the Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-12 BPIC Lid to the BPIC Assembly, one threaded and tapped hole for the former test point, and one hole threaded and tapped M8 x 1.25 for the removable eyebolt.
1.2.1.6.4 BPIC 2835A Shield The BPIC Shield is either depleted uranium (DU) or tungsten and has a minimum shield thickness of approximately 63.0 mm (2.48in) in every direction. The shield is approximately mm +/- mm
( in +/- in) high and mm +/- mm ( in +/- in) diameter. The DU version has a titanium inner cylinder or tungsten shield has an inner titanium or stainless steel cylinder that has an interior dimension of approximately mm +/- mm ( in +/- in) diameter and mm +/-
mm in +/- in) tall. The DU shield rests on a brass bottom support, a brass cylinder that slips over the DU shield and has a brass top shield support which provides a eutectic barrier which prevents stainless steel and DU interaction. The brass creates stability and a eutectic barrier with prevents stainless steel and DU interaction. The tungsten version eliminates most brass pieces.
1.2.2 Contents
[IAEA-SSR6 Section IV & 809(a), IAES TS-R-1 Section IV & 807(a), and 10 CFR 71.33(b)]
Typically the source capsules are inserted into a capsule insert/holder (300 series stainless steel, 400 series stainless steel, titanium, or tungsten), then placed into the transport packages inner isotope cavity, and secured in-place by the bolted down Shield Plug Assembly which is covered by the bolted down Lid Assembly. During shipments, an optional tungsten insert may be used to provide additional shielding.
The ISORAD-TC1 package is designed as a general purpose transport package to transport special form capsules containing the isotopes listed in Tables 1.2a and 1.2b. The package is designed for radioactive material that emits alpha, beta or gamma radiation. The specified contents do not include materials that emit a significant number of neutrons. The contents are solid or liquid form in approved special form capsules.
The contents may also include inorganic non-radioactive materials associated with the radioactive materials, such as contents holders or fixtures and packing materials. No organic/hydrogenous materials are allowed in the inner cavity.
Fissile materials and irradiated fissile materials containing fission products are not permitted for transport in the ISORAD-TC1. No Pyrophoric radioactive materials are permitted in the ISORAD-TC1 package. The ISORAD-TC1 transports less than 3,000 A1, the package is designated as Category II as defined in NUREG-6407 (Page 5 Table 4) [1.9].
The maximum activity of the radioactive contents is limited principally by the radiation heat load and has excess shielding. The contents heat limit is 60.88 Watts for radioactive materials in special form.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-13 Table 1.2a: Isotopes Permitted in the Model ISORAD-TC1 BPIC Authorized Total Authorized Total Isotope Curies Isotope Curies Forms Watts Forms Watts Actinium-225 Special1 67.75 2.35 Phosphorous-32 Special1 150.66 0.81 Actinium-227 Special1 25 0.0118 Phosphorous-33 Special1 10017 53.6 Actinium-228 Special1 17 0.136 Selenium-75 Special1 10017 53.6 Barium-131 Special1 5076 15.54 Sodium-22 Special1 2700 14.5 Barium-133 Special1 125 0.669 Sodium-24 Special1 5.4 0.029 Cadmium-109 Special1 2700 14.45 Strontium-89 Special1 329.4 1.14 Carbon-14 Special1 1080 0.32 Strontium-90 Special1 50 0.163 Cobalt-57 Special1 300 1.61 Thorium-227 Special1 272.7 9.79 Cobalt-60 Special1 0.5 0.0077 Thorium-228 Special1 150.66 4.84 Cesium-131 Special1 26900 4.44 Tungsten-187 Special1 604.8 2.75 Cesium-134 Special1 190.35 1.95 Tungsten-188 Special1 17.36 0.0104 Cesium-137 Special1 3888 3.927 Yttrium-90 Special1 46.71 0.0259 Copper-67 Special1 18657 30.04 Ytterbium-169 Special1 10017 25.143 Curium-248 Special1 5 0.18 Ytterbium-175 Special1 26900 26.9 Europium-152 Special1 81 0.624 Zinc-65 Special1 500 7.7 Other Isotopes Type A Indium-111 Special1 11475 29.84 Special1 30 Quantity Iridium-192 Special1 10125 60.88 Iridium-194 Special1 55.35 0.296 Iron-55 Special1 2700 0.45 Lutetium-177 Special1 10017 53.6 1 Special Form is defined in IAEA SSR-6, IAEA TS-R-1, 10 CFR 71, and 49 CFR 173.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-14 Table 1.2b: Isotopes Permitted in the Model ISORAD-TC1 MPIC Authorized Curies Total Total Isotope Sources Forms Each Curies Watts Special1 Iridium-192 10 150 1500 9.16 Solid Special1 Selenium-75 10 150 1500 8.025 Solid Special1 Ytterbium-169 10 150 1500 3.765 Solid 1 Special Form is defined in IAEA SSR-6, IAEA TS-R-1, 10 CFR 71, and 49 CFR 173.
1.2.3. Special Requirements for Plutonium The ISORAD-TC1 package is not designed for plutonium.
1.2.4 Operational Features The primary containment system for the ISORAD-TC1 package is the radioactive source capsule(s) meeting special form criteria referred to in Section 4.1 of this Safety Analysis Report.
The source capsule(s) are certified as special form radioactive material under IAEA SSR-6 (2018),
IAEA TS-R-1, CNCS SOR (2015), USNRC 10 CFR Part 71, or USDOT 49 CFR Part 173.
1.3 Appendix
1.3.1 References
[1.1] IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material, Revision 1, International Atomic Energy Agency, Vienna, 2018.
[1.2] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 2005 Edition, International Atomic Energy Agency, Vienna, 2005.
[1.3] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 1996 (Revised),
International Atomic Energy Agency, Vienna, 2000.
[1.4] Packaging and Transport of Nuclear Substances Regulations (CNSC PTNS SOR/2015-145),
Canadian Nuclear Safety Commission, Ottawa, 2015.
[1.5] Title 10, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
[1.6] Title 49, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-15
[1.7] ADR European Agreement Concerning the International Carriage of Dangerous Goods by Road. United Nations, New York and Geneva, 2018.
[1.8] CNSC RD-364/ USNRC NUREG-1886 Joint Canada - United States Guide for Approval of Type B(U) and Fissile Material Transport Packages, USNRC and CNSC, Washington &
Ottawa, 2009.
[1.9] NUREG/CR-6407, INEL-95/0551, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, February 1996
[1.10] Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels, Revision 1, March 1978.
[1.11] Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, Revision 1, U.S. Nuclear Regulatory Commission, Office of Standards Development, March 1989.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 1-38 1.3.2.17 R180831-XXX Sketch ISORAD-TC1 Prepared for Transport Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-1 Safety Analysis Report Section 2 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-2 Table of Contents Section 2 STRUCTURAL EVALUATION ...................................................................................... 2-5 2.1 Description of Structural Design ................................................................................................ 2-5 2.1.1 Discussion ............................................................................................................................ 2-5 2.1.1.1 Outer Drum ................................................................................................................... 2-5 2.1.1.2 Cork Assembly.............................................................................................................. 2-6 2.1.1.3 Inner Container ............................................................................................................. 2-6 2.1.1.4 Containment System ..................................................................................................... 2-8 2.1.2 Design Criteria ..................................................................................................................... 2-9 2.1.2.1 Regulations and Codes .................................................................................................. 2-9 2.1.2.2 Allowable Stress ......................................................................................................... 2-10 2.1.2.3 Buckling ...................................................................................................................... 2-10 2.1.2.4 Fatigue......................................................................................................................... 2-10 2.1.2.5 Brittle Fracture ............................................................................................................ 2-10 2.1.3. Weights and Centers of Gravity ........................................................................................ 2-10 2.1.4 Identification of Codes and Standards for Package Design ............................................... 2-11 2.1.4.1 Package Design ........................................................................................................... 2-11 2.1.4.2 Fabrication & Assembly ............................................................................................. 2-11 2.1.4.3 Maintenance & Use..................................................................................................... 2-12 2.2 Materials ................................................................................................................................... 2-12 2.2.1 Material Properties and Specifications .............................................................................. 2-12 2.2.1.1 Structural Materials ..................................................................................................... 2-15 2.2.1.2 Shielding Material ....................................................................................................... 2-15 2.2.1.3 Cork Assembly............................................................................................................ 2-15 2.2.2 Chemical, Galvanic or Other Reactions............................................................................. 2-15 2.2.3 Effects of Radiation on Materials ...................................................................................... 2-15 2.3 Fabrication and Examination .................................................................................................... 2-16 2.3.1 Fabrication ......................................................................................................................... 2-16 2.3.2 Examination ....................................................................................................................... 2-16 2.3.2.1 Fabrication Tests and Examinations ........................................................................... 2-17 2.3.2.2 Acceptance Tests ........................................................................................................ 2-17 2.4 General Standards for All Packages ..................................................................................... 2-18 2.4.1 Minimum Package Size ................................................................................................ 2-18 2.4.2 Tamper-Indicating Feature............................................................................................ 2-18 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-3 2.4.3 Positive Closure ............................................................................................................ 2-18 2.5 Lifting and Tie-down Standards for All Packages ............................................................... 2-18 2.5.1 Lifting Devices.............................................................................................................. 2-18 2.5.2 Tie-Down Devices ............................................................................................................. 2-19 2.6 Normal Conditions of Transport (NCT) ................................................................................... 2-20 2.6.1 Heat .................................................................................................................................... 2-20 2.6.1.1 Summary of Pressures and Temperatures ............................................................. 2-24 2.6.1.2 Differential Thermal Expansion ............................................................................ 2-26 2.6.1.3 Stress Calculations ................................................................................................. 2-27 2.6.1.4 Comparison with Allowable Stresses .................................................................... 2-29 2.6.2 Cold ............................................................................................................................... 2-29 2.6.3 Reduced External Pressure ........................................................................................... 2-29 2.6.4 Increased External Pressure .......................................................................................... 2-34 2.6.5 Vibration ....................................................................................................................... 2-37 2.6.6 Water Spray .................................................................................................................. 2-37 2.6.7 Free Drop ...................................................................................................................... 2-37 2.6.8 Corner Drop .................................................................................................................. 2-39 2.6.9 Compression ................................................................................................................. 2-39 2.6.10 Penetration................................................................................................................. 2-39 2.7 Hypothetical Accident Conditions of Transport .................................................................. 2-41 2.7.1 Free Drop I .................................................................................................................... 2-42 2.7.1.1 Oblique Drop ......................................................................................................... 2-42 2.7.1.2 Top Down Drop ..................................................................................................... 2-43 2.7.1.5 Summary of Results............................................................................................... 2-44 2.7.2 Crush ............................................................................................................................. 2-44 2.7.3 Puncture ........................................................................................................................ 2-44 2.7.3.1 Top Puncture Drop ................................................................................................ 2-45 2.7.3.2 Side Puncture Drop ................................................................................................ 2-45 2.7.3.3 Bottom Puncture Drop ........................................................................................... 2-45 2.7.4 Thermal ......................................................................................................................... 2-46 2.7.4.1 Summary of Maximum Pressures.......................................................................... 2-46 2.7.4.2 Differential Thermal Expansion ............................................................................ 2-47 2.7.4.3 Stress Calculations ................................................................................................. 2-48 2.7.4.4 Comparison with Allowable Stresses .................................................................... 2-48 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-4 2.7.5 Immersion - Fissile Material ......................................................................................... 2-48 2.7.6 Immersion - All Packages ............................................................................................. 2-48 2.7.7 Deep Water Immersion Test ......................................................................................... 2-50 2.7.8 Summary of Damage .................................................................................................... 2-50 2.8 Accident Conditions for Air Transport of Plutonium .......................................................... 2-51 2.9 Accident Conditions for Fissile Material Packages for Air Transport ................................. 2-51 2.10 Special Form ..................................................................................................................... 2-51 2.11 Fuel Rods .......................................................................................................................... 2-51 2.12 Appendix .......................................................................................................................... 2-52 2.12.1 QTP-001 Test Plan, Revision 1. ............................................................................ 2-52 2.12.2 QTP-001 Test Report Revision 0 .......................................................................... 2-53 2.12.3 SERCO Test Report............................................................................................... 2-54 2.12.4 Special Form Certificates for use with Model ISORAD-TC1 .............................. 2-55 2.12.5 References ............................................................................................................. 2-56 List of Tables Table 2.2.1.A: Mechanical Properties of Principal Transport Package Materials .............................. 2-12 Table 2.2.1.B: Materials of Construction............................................................................................ 2-13 Table 2.6.A: Summary of NCT Initial Conditions ............................................................................. 2-20 Table 2.6.1.A: Radionuclide Decay Energy ....................................................................................... 2-20 Table 2.6.1.B: ISORAD-TC1 Authorized Contents BPIC Inner Container ....................................... 2-21 Table 2.6.1.C: Summary Temperatures Normal Conditions of Transport ......................................... 2-24 Table 2.6.1.D: NCT Maximum Temperature and Pressure in the Special Form Capsule .................. 2-24 Table 2.6.1.E: HACT Maximum Temperature and Pressure in the Special Form Capsule ............... 2-25 Table 2.6.1.F: Summary of Thermal Expansion of BPIC Inner Container ........................................ 2-27 Table 2.7.A: Summary of HACT Initial Conditions........................................................................... 2-41 Table 2.7.4.1.b: Summary Table of Maximum Pressures................................................................... 2-46 Table 2.7.8.a: Summary of Damages During Performance of QTP-001 ............................................ 2-50 List of Figures Figure 2.1.3: Centers of Gravity......................................................................................................... 2-10 Figure 2.6.7.a: NCT Drop Test Orientation 1 and NCT Drop Test Orientation 2 ............................. 2-38 Figure 2.6.10.a: NCT Penetration Test Orientation 1 and NCT Drop Test Orientation 2 ................. 2-40 Figure 2.6.10.b: NCT Penetration Test Orientation 3 and Impact Area ............................................ 2-41 Figure 2.7.1.a: HACT Free Drop 1 Test Orientation 1 ...................................................................... 2-43 Figure 2.7.1.b: HACT Free Drop 1 Test Orientation 2 ...................................................................... 2-43 Figure 2.7.3.a: HACT Puncture Test Orientation 1, 2 & 3 ................................................................ 2-46 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-5 Section 2 STRUCTURAL EVALUATION The Structural Evaluation section identifies and describes the principal structural engineering design of the ISORAD-TC1 packaging, components, and systems important to safety and compliance with the performance requirements of IAEA SR-6, IAEA TS-R-1, and 10 CFR Part 71. The design was influenced in part by the organizations American Welding Society (AWS) and American Society of Mechanical Engineers (ASME).
2.1 Description of Structural Design
[IAEA TS-R-1 220 & 806-808, IAEA SSR-6 220 & 808-810, and 10 CFR 71.33(a)]
2.1.1 Discussion The principal structural components of the ISORAD-TC1 package are the Outer Drum Assembly, the five (5) configurations of Inner Containers forming part of the security containment system, and the special form radioactive material (primary containment vessel). The radioactive contents are carried within the Inner Containers in special form capsules of various sizes and potential configurations. The special form capsules are placed inside capsule inserts within the inner container isotope cavity (see Section 1.2.1).
The Outer Drum is designed to absorb impacts, provide protection during handling/transportation operations, and provide thermal insulation for the inner container during NCT and HACT conditions.
The inner cork packing is designed to absorb the impact loads preventing damage to the inner container under NCT and HACT conditions. The inner container configurations are designed to provide the shielding and security for the special form capsules. The containment system consists of the inner container (provides shielding and secures the special form capsules within the inner container), capsule insert {spacer/holder} (centers the special form capsules, distributes the dose, and may provide additional shielding within the inner container isotope cavity), and special form capsules (are the primary containment vessel for the radioactive isotope). A description of the structural design of each of these components is provided in the following sections.
2.1.1.1 Outer Drum The Outer Drum comprises of a keg, Attachment Ring Assembly, Lid Assembly, Cork Assembly, and Cork Lid assembly as shown in drawing R180831-100 (Section 1.3.2). The keg is a commercially available standard 1/2 barrel 15.5 gallon keg and is constructed from two rolled and formed austenitic stainless steel halves welded in the center to form a cylinder. A base and top with rims are welded to the rolled cylinder to form the keg body. The keg is modified with Attachment Ring Supports and Attachment Ring Assembly, which are permanently welded into place to form the attachment point for the Lid Assembly. The Cork Assembly is assembled into the modified keg. An optional stainless steel Cavity Sleeve can be placed within the Cork Assembly cavity to protect the cork during normal handling operations.
The Outer Drum closure is facilitated by eight M14 x 2 threaded bolt holes for securing the Lid Assembly to the Attachment Ring and two Lock Studs which are permanently welded into position.
The Lid Assembly is a circular plate with ten holes machined for the eight closure bolts and two holes for the Lock Studs. The Lid Assembly is attached to the body with eight M14 x 2 austenitic stainless Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-6 steel hex head bolts with lock and flat washers, two padlocks for security, and tamper-evident seal (lead wire seal, special zip tie, or other tamper-evident device) between the two Lock Studs. The Outer Drum has two handle holes that can be used to lift the package.
2.1.1.2 Cork Assembly The Cork Assembly fits inside the Outer Drum cavity and surrounds the Inner Container. It is designed to reduce impact loads on the Inner Container and provides thermal insulation. The cork surrounds the side walls and the lid of the Inner Container. The cork varies in width from mm ( inches) to mm ( inches) on the side walls due to the variation in configuration of the Inner Container and is a minimum of mm ( inches) thick above the lid. The cork may be coated in a fire-retardant paint.
The cork components are shown in detail in drawing R180831-110 and R180831-111 (Section 1.3.2).
2.1.1.3 Inner Container The ISORAD-TC1 consists of five configurations of Inner Containers.
Configuration 1: The Bulk & PIC Inner Container (BPIC) - Square inner container forms part of the containment system and consists of a body, a removable Shield Plug Assembly bolted together with four M6 x 1 closure screws, and a removable lid assembly bolted together with eight M8 x 1.25 closure screws as shown in drawing R180831-200 (Section 1.3.2).
The body assembly is formed from a stainless steel square casing filled with depleted uranium or tungsten shield which is surrounded by brass. The depleted uranium or tungsten forms the shielding for the walls and base of the inner cavity. The shielding cavity is filled with a titanium cavity sleeve that is press fit into the Top Plate and LocTite, JB Weld, or brazed together. The austenitic stainless steel casing consists of three austenitic stainless steel pieces, the Outer Casing, Top Plate, and Bottom Plate.
Each piece is machined from solid or tubular austenitic stainless steel. The Top Plate and Bottom Plate are welded to the Outer Casing with a circumferential full penetration groove weld which is both visually and liquid penetrant tested according to AWS D1.6/D1.6M or equivalent international standard.
The Shield Plug Assembly is formed from the titanium Plug Housing and Plug Housing Lid filled with the depleted uranium or tungsten Plug Shield and the brass Plug Housing Spacer. The Plug Housing and Plug Housing Lid are welded together with a circumferential fillet weld which is both visually and liquid penetrant tested according to AWS D1.9/D1.9M or equivalent international standard.
Configuration 2: The Bulk & PIC Inner Container (BPIC) - Round inner container forms part of the containment system and consists of a body, a removable Shield Plug Assembly bolted together with four M6 x 1 closure screws, and a removable lid assembly bolted together with eight M8 x 1.25 closure screws as shown in drawing R180831-400 (Section 1.3.2).
The body assembly is formed from a stainless steel round casing filled with depleted uranium or tungsten shield which is surrounded by brass. The depleted uranium forms the shielding for the walls and base of the inner cavity. The shielding cavity is filled with a titanium cavity sleeve that is press fit into the Top Plate and LocTite, JB Weld, or brazed together. The stainless steel casing consists of three austenitic stainless steel pieces, the Outer Casing, Top Plate, and Bottom Plate. Each piece is Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-7 machined from solid austenitic stainless steel. The Top Plate and Bottom Plate are welded to the Outer Casing with a circumferential full penetration groove weld which is both visually and liquid penetrant tested according to AWS D1.6/D1.6M or equivalent international standard.
The Shield Plug Assembly is formed from the titanium Plug Housing and Plug Housing Lid filled with the depleted uranium or tungsten Plug Shield and the brass Plug Housing Spacer. The Plug Housing and Plug Housing Lid is welded together with a circumferential full penetration fillet weld which is both visually and liquid penetrant tested according to AWS D1.9/D1.9M or equivalent international standard.
Configuration 3: The Multi Port Inner Container (MPIC) - Square inner container forms part of the containment system and consists of a body and a removable lid assembly with one closure M16 hex nut and padlock as shown in drawing R180831-300 (Section 1.3.2).
The body assembly is formed from a stainless steel square casing filled with a two part depleted uranium shield which is surrounded by brass. The depleted uranium or tungsten forms the shielding for the ten inner cavities and can be configured as five and five holes in two rows or ten holes in one row.
Each shielding cavity is filled with a titanium source tube that is press fit into the Top Plate and source lock and brazed together. The stainless steel casing consists of three austenitic stainless steel pieces, the Outer Casing, Top Plate, and Bottom Plate. Each piece is machined from solid austenitic stainless steel.
The Top Plate and Bottom Plate are welded to the Outer Casing with a circumferential full penetration groove weld which is both visually and liquid penetrant tested according to AWS D1.6/D.6M or equivalent international standard.
Configuration 4: The Multi Port Inner Container (MPIC) - Round inner container forms part of the containment system and consists of a body and a removable lid assembly with one closure M16 hex nut and padlock as shown in drawing R180831-600 (Section 1.3.2).
The body assembly is formed from a stainless steel round casing filled with a two part depleted uranium or tungsten shield which is surrounded by brass. The depleted uranium forms the shielding for the ten inner cavities and can be configured as five and five holes in two rows or ten holes in one row.
Each shielding cavity is filled with a titanium source tube that is press fit into the Top Plate and Source Lock Assembly and brazed. The stainless steel casing consists of three austenitic stainless steel pieces, the Outer Casing, Top Plate, and Bottom Plate. Each piece is machined from solid austenitic stainless steel. The Top Plate and Bottom Plate are welded to the Outer Casing with a circumferential full penetration groove weld which is both visually and liquid penetrant tested according to AWS D1.6/D.6M or equivalent international standard.
Configuration 5: The Bulk & PIC Inner Container (BPIC) 2835A is a round inner container that forms part of the containment system and consists of a body, a removable Shield Plug Assembly, and a removable lid assembly bolted together with six M8 x 1.25 closure screws as shown in drawing R180831-500 (Section 1.3.2).
The body assembly is formed from a stainless steel round casing filled with depleted uranium tungsten shield which is surrounded by brass. The depleted uranium or tungsten forms the shielding for the walls and base of the inner cavity. The shielding cavity is filled with a stainless steel cavity sleeve that is welded to the Top Plate. The stainless steel casing consists of three austenitic stainless steel pieces, the Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-8 Outer Casing, Top Plate, and Bottom Plate. Each piece is machined from solid austenitic stainless steel.
The Top Plate and Bottom Plate are welded to the Outer Casing with a circumferential full penetration groove weld which is both visually and liquid penetrant tested according to AWS D1.6/D1.6M or equivalent international standard.
The Shield Plug Assembly is formed from the stainless steel Plug Housing and Plug Housing Lid filled with the depleted uranium Plug Shield and the brass shims. The Plug Housing and Plug Housing Lid is welded together with a circumferential full penetration fillet weld which is both visually and liquid penetrant tested according to AWS D1.6/D.6M or equivalent international standard.
2.1.1.4 Containment System The ISORAD-TC1 is designed to transport solid or liquid radioactive materials only in special form capsules. The package will not transport normal form or non-special form radioactive material or gasses. The package may transport liquids providing the liquid is encapsulated in an approved special form capsule. The containment system for the BPIC Series consists of three major components inner container (provides shielding and secures the special form capsules within the inner container),
capsule insert {spacer/holder} (centers the special form capsules, distributes the dose, and may provide additional shielding within the inner container isotope cavity), and special form capsules (provides the primary containment of the radioactive isotope and is considered the containment vessel for the package). The containment system for the MPIC Series consists of the three major components inner container (provides shielding and secures the special form capsules incorporated into a source holder assembly (i.e. radiography source or similar source holder assembly) within the inner container),
source locking mechanism (locks the source holder assembly into the inner container), and special form capsules (provides the primary containment of the radioactive isotope and is considered the containment vessel for the package).
The inner container as described above in 2.1.1.3 and is the outer most layer of the containment system.
The inner container provides the physical security, shielding and the isotope cavity.
The capsule inserts are designed to accommodate the various special form capsules that will be transported in the three configurations of BPIC. The capsule inserts can be constructed of 300 series stainless steel, 400 series stainless steel, titanium, or tungsten. The hole configurations are variable in number and diameter to maximize the isotope contents flexibility. The sizes of special forms vary from country to country and are not standardized. The maximum weight of isotope contents and capsule insert holder is 1.37 kg (3 pounds).
The special form capsules must possess a valid special form certificate issued under conformance to IAEA TS-R-1 or IAEA SSR-6 standard requirements. In addition, a valid and current leak test analysis is required before transport of the special form in the ISORAD-TC1. Meeting special form criteria provides the primary containment of the radioactive isotope being transported.
Any one of the three inserts specified in Section 1.3.2 shall be used to provide further shielding and confinement for the special form capsule contents.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-9 2.1.2 Design Criteria The ISORAD-TC1 package has been designed to withstand the Type A Package (NCT) and Type B Package (HACT) specifications and testing requirements of the IAEA, CNSC, USNRC, USDOT, UN, IATA, and other regulatory bodies governing transport packages. The ISORAD-TC1 package is not a containment vessel, nor transports greater than 3000A1 quantities of radioactive isotopes, and therefore the ASME Boiler and Pressure Vessel Code (BPVC)Section II Part D or Section III (BPVC) was not used as an integral part of the design process, but was referenced during the design process. The containment vessel are the special form capsules. The following regulations and codes were used to design, develop, test and prepare the ISORAD-TC1 Type B application.
2.1.2.1 Regulations and Codes 2.1.2.1.1 Regulatory References A. IAEA SSR-6:2018 [2.1]
B. IAEA TS-R-1:2005 (Revised) [2.2]
B. IAEA TS-R-1:1996 (Revised) [2.3]
C. CNSC PTNS SOR/2015-145 [2.4]
D. USNRC 10 CFR Part 71 [2.5]
E. USDOT 49 CFR Part 173 [2.6]
E. UN ADR:2019 F. IATA Dangerous Goods Regulations 2.1.2.1.2 Regulatory Guides A. CNSC RD-364 / USNRC NURERG 1886 [2.7]
D. USNRC NUREG 6407 [2.10]
E. USNRC Regulatory Guide 7.6 [2.11]
F. USNRC Regulatory Guide 7.8 [2.12]
G. USNRC Regulatory Guide 7.11 [2.13]
2.1.2.1.3 Welding Codes and Guides A. AWS A2.4 Standard Symbols for Welding, Brazing, and Nondestructive Examination
[2.14]
B. AWS A3.0 Standard Terms and Definitions [2.15]
C. AWS D1.6/D1.6M Structural Welding Code: Stainless Steel [2.16]
D. AWS D1.9/D1.6M Structural Welding Code: Titanium [2.17]
2.1.2.1.4 Design Codes and Standards A. 27th Edition Machinerys Handbook [2.18]
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-10 2.1.2.2 Allowable Stress The ISORAD-TC1 is not subject to this section because it is not a pressure vessel. See Sections 2.6.1.1
- 2.6.1.4 for calculations.
2.1.2.3 Buckling The ISORAD-TC1 is not subject to this section because it is not a pressure vessel.
2.1.2.4 Fatigue The hypothetical maximum number of cycles that the ISORAD-TC1 will undergo is approximately 52 cycles/year for 20 years = 1040 cycles. The number of cycles was multiplied by 10 to give 10400 cycles, to give a safety margin. In reality, the average ISORAD-TC1 will undergo 18 cycles/year for 20 years = 360 because of logistical limitations. Providing the same safety margin of 10 as in the hypothetical maximum, the ISORAD-TC1 maximum safety margin cycles is 3,600.
2.1.2.5 Brittle Fracture All the structural components (Outer Drum and Inner Container outer casing components) of the package are fabricated from 300 series stainless steel which is an austenitic stainless steel that is ductile at low temperatures. According to Regulatory Guide 7.11 [2.13], austenitic stainless steel is not susceptible to brittle facture at temperatures encountered in transport. The most destructive hypothetical accident condition 9 meter Free Drop test was conducted on Prototype #1 at -40°C to determine if brittle fracture has any effect on the package, with compliance demonstrated if the containment vessel (special form capsule) is undamaged and leak tight on completion of testing.
2.1.3. Weights and Centers of Gravity The ISORAD-TC1 package maximum weight is approximately 136.1 kg (300 pounds). The heaviest Inner Container (BPIC - Square) weighs approximately 102.1 kg (225.0 pounds), including about 67.8 kg (149.5 pounds) +/- 2.0 kg(4.41 pounds) of DU shielding. The center of gravity of the ISORAD-TC1 transport package is approximately 285 mm (11.22 inches) above the bottom of the drum (See Figure 2.1.3).
Figure 1.1.3: Centers of Gravity Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-11 2.1.4 Identification of Codes and Standards for Package Design The package has been designed to transport special form material in quantities of less than 1000 TBq or 3000A1, and is classified as a Category II package, as defined in NUREG-6407 [2.10 Page 5 Table 4].
The standards to which the package has been designed, fabricated, tested and maintained have been selected based on the guidance provided in Regulatory Guide 7.6 [2.11] and NUREG/CR-3854 [2.9].
The design criteria used to assess the containment boundary have been taken from Regulatory Guide 7.6 [2.11] and the load combinations have been taken from Regulatory Guide 7.8 [2.12] as discussed in section 2.1.2.1. The Inner Container contains austenitic stainless steel welds and inspection according to the AWS D1.6/D1.6M welding code. The Plug Assembly contains titanium welds and inspection according to AWS D1.9.D1.9M welding code.
Section 2.1.1 identifies the major components of the ISORAD-TC1 package and identifies if they provide containment or fulfill the other safety functions such as gamma shielding or support. The drawings in section 1.3 identify whether the items are Important to Safety (ITS), the identification was carried out using the guidance of NUREG/CR-6407 [2.10]. The Assembly drawings from SAR section 1.3 contain the QA Category classification for the various components, the codes or standards used to purchase materials, fabricate the package, inspect and examine the package.
The depleted uranium and the cork insulation and impact limiters are specialist materials for which an ASTM standard does not exist. The depleted uranium is fabricated and tested in accordance with standard industry practices. Chemical composition checks through Material Test Reports (MTRs) like other metal materials used in the package. In addition, Lot numbers are permanently engraved on the shields to track and trace the shields. Each shield is surveyed to ensure regulatory dose level compliance in each completed Inner Container.
The cork is fabricated in accordance with the vendors standard practices and quality inspected to the requirements of drawing R180831-110 or R180831-111.
2.1.4.1 Package Design See Section 2.1.2 relating to design criteria of the package. The regulations, regulatory guides, and codes listed were all referenced and incorporated during the design, development, testing, and safety analysis review of the ISORAD-TC1 package. However, the design was based on the General Package, Type A, and Type B(U) container requirements of IAEA SSR-6, IAEA TS-R-1, CNSC SOR/2015-14, UN ADR, 49 CFR 173, and 10 CFR 71 regulations in effect at the time of the package design.
2.1.4.2 Fabrication & Assembly All component fabrication (including assembly) is controlled under the ISO-RAD Canada Inc. Quality Assurance Program (QAP) previously submitted to the CNSC. All welding under this plan adheres to American Welding Society (AWS), American Society of Mechanical Engineers (ASME), or equivalent international standards appropriate to the materials and designs fabricated. All safety critical hardware meets ASME-B18 standards. All external fabrication and especially items deemed critical to safety are QA inspected by qualified ISO-RAD Canada personnel. The process includes verification to Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-12 manufacturing/fabrication instructions, assembly instructions, and drawings and all records are retained on file for the life of the transport package.
2.1.4.3 Maintenance & Use Maintenance and use of this transport container are described in Sections 7 and 8 of the SAR.
2.2 Materials [SSR-6 614, TS-R-1 613, and 10 CFR 71 .43(d)]
2.2.1 Material Properties and Specifications Table 2.2.1.A lists the relevant mechanical properties (at ambient temperature) of the principal materials used in the ISORAD-TC1 transport package. The sources referred to in the last column are listed after the table.
Table 2.2.1.A: Mechanical Properties of Principal Transport Package Materials 300 Series Brass Titanium Uranium Tungsten Cork Stainless Steel C360/C260/ Grade 2 (Depleted) #4 (Weathered)/ C230 #2 (Painted White)#1 Modulus of 190 GPa 100 GPa 116 GPa 205 GPa 400 GPa 0.02 GPa Elasticity (27 Mpsi) (15 Mpsi) (16.8 Mpsi) (30 Mpsi) (58 Mpsi) (0.0029 Mpsi)
Poissons 0.29 0.33 0.361 0.21 0.28 0.0 Ratio Density 8000 kg/m3 8500 kg/m3 4621 kg/m3 19,000 kg/m3 19,270 kg/m3 154.23 kg/m3 (0.29 lb/in3 (0.32 lb/in3) (0.163 lb/in3) (0.69 lb/in3) (0.70 lb/in3) (0.00544 lb/in3)
(Ultimate) 517 MPa 478 MPa 344 MPa 365 MPa 980 MPa 0.85 MPa Tensile (75 kpsi) (69 kpsi) (49.9 kpsi) (53 kpsi) (142 kpsi) (0.12 kpsi)
Strength Yield 207 MPa 240 MPa 275 MPa 172 MPa 750 MPa N/A Strength (30 kpsi) (35 kpsi) (39.9 kpsi) (25 kpsi) (109 kpsi)
Coefficient 17.3 µm/m°C 20.2 µm/m°C 8.90 µm/m°C 13.3 µm/m°C 4.5 µm/m°C 40 µm/m°C of Thermal (9.9 µin/in°F) (11.2 µin/in°F) (4.94 µin/in°F) (7.4 µin/in°F) (2.5 µin/in°F) (22.2 µin/in°F)
Expansion Thermal 16 W/m-°K 100 W/m-°K 11.4 W/m-°K 27.5 W/m-°K 163.3 W/m-°K 0.04 W/m-°K Conductivity (9.2 Btu/h-ft-°F) (63 Btu/h-ft-°F) (6.6 Btu/h-ft-°F) (16 Btu/h-ft-°F) (1130 Btu/h-ft-°F) (0.023 Btu/h-ft-°F)
Emissivity 0.85/0.992 0.22 0.31 0.15 0.23 (1500°C) 0.95 Specific 500 j/kg °K 390 j/kg °K 522.3 J/kg °K 120 j/kg °K 0.13 j/kg °K 2000 j/kg °K Heat (0.12 Btu/lb-F°) (0.09 Btu/lb-F°) (0.125 Btu/lb-F°) (0.03 Btu/lb-F°) (0.03 Btu/lb-F°) (0.48 Btu/lb-F°)
Melting 1,427 887/954/1000 1,668 1,130 3,410 400 Point °C Resource references:
- 1. American Society for Metals. Metals Handbook, Volume 1, Tenth Edition. Ohio: Materials Park, 1990.
- 2. Lowenstein, Paul. Industrial Uses of Depleted Uranium. American Society for Metals. Metals Handbook, Volume 3, Ninth Edition.
- 3. American Society for Metals. Metals Handbook, Volume 2, Tenth Edition. Ohio: Materials Park, 1990.
- 4. http://rhenium.com/assets/tungsten_properties.pdf Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-13 Table 2.2.1.B: Materials of Construction Outer Drum Material ASTM or Other Specification Form Keg AISI 304 A240/A240M Plate/Sheet Lid Assembly AISI 304 A240/A240M & A276 Plate/Sheet & Bar Attachment Ring AISI 304 A240/A240M Plate/Sheet Attachment Ring Support AISI 304 A240/A240M Plate/Sheet M14 x 2 x 35 Hex Bolt AISI 316 Minimum A4 70 M14 Lock Washer AISI 316 Minimum A4 70 M14 Washer AISI 316 Minimum A4 70 Cork Assembly Cork ISO 7322 Sheet Cavity Sleeve AISI 304 A240/A240M, A276, A312/A312M Plate/Sheet, Bar, or Tube/Pipe Cork Lid Assembly Lid Cork Cork ISO 7322 Sheet Mounting Plate AISI 304 or Aluminum 6061 A240/A240M & A276 or B209/B221 Plate/Sheet or Bar or Plate/Sheet or Bar BPIC - Square & Round Material ASTM or Other Specification Form Shield Depleted Uranium/Tungsten SAE AMS 7730 /B777 Casting or Ingots / Sintered Outer Casing AISI 304 A240/A240M or A312/A312M Plate/Sheet or Round or Square Pipe Top Plate AISI 304 A240/A240M or A276 Plate/Sheet or Bar Bottom Plate AISI 304 A240/A240M or A276 Plate/Sheet or Bar Lid AISI 304 A240/A240M or A276 Plate/Sheet or Bar M8 x 1.25 Skt Head Screw AISI 316 Minimum A4 70 M6 x 1 Socket Head Screw AISI 316 Minimum A4 70 Top Support Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Bottom Support Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Spacers (Square Only) Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Sleeve (Round Only) Brass C360/C260/C230 B16/B16M or B36/B36M or B135/B135M Bar or Plate & Bar or Tube Shield Cavity Grade 2 Titanium (min) B348/B348M or B338/B338M or B861 Bar or Tube/Pipe Shield Plug Assembly Material ASTM or Other Specification Form Shield Plug Depleted Uranium/Tungsten SAE AMS 7730 / B777 Casting or Ingots / Sintered Plug Housing Grade 2 Titanium (min) B348/B348M or B338/B338M or B861 Bar or Tube/Pipe Plug Lid Grade 2 Titanium (min) B348/B348M or B265 Bar or Plate/Sheet Plug Spacer Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-14 Table 2.2.1.B Materials of Construction (Continued)
MPIC - Square & Round Material ASTM or Other Specification Form Shield Base Depleted Uranium/Tungsten SAE AMS 7730 / B777 Casting or Ingots / Sintered Shield Top Depleted Uranium/Tungsten SAE AMS 7730 / B777 Casting or Ingots / Sintered Outer Casing AISI 304 A240/A240M or A312/A312M Plate/Sheet or Round or Square Pipe Top Plate AISI 304 A240/A240M or A276 Plate/Sheet or Bar Bottom Plate AISI 304 A240/A240M or A276 Plate/Sheet or Bar Lid AISI 304 A240/A240M or A276 Plate/Sheet or Bar Lid Support AISI 304 A240/A240M or A312/A312M Plate/Sheet or Round or Square Pipe Source Lock Assembly AISI 304 A240/A240M or A276 Plate/Sheet or Bar Source Caps AISI 304 A240/A240M or A276 Plate/Sheet or Bar M16 Low Profile Hex Nut AISI 316 Top Support Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Bottom Support Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Spacers (Square Only) Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Shield Sleeve Brass C360/C260/C230 B16/B16M or B36/B36M or B135/B135M Bar or Plate & Bar or Tube Tubes Grade 2 Titanium (min) B348/B348M or B338/B338M or B861 Bar or Tube/Pipe BPIC 2835A Material ASTM or Other Specification Form Shield Depleted Uranium SAE AMS 7730 Casting or Ingots Outer Casing AISI 304 A240/A240M or A312/A312M Plate/Sheet or Round or Square Pipe Top Plate AISI 304 A240/A240M or A276 Plate/Sheet or Bar Bottom Plate AISI 304 A240/A240M or A276 Plate/Sheet or Bar Shield Cavity AISI 304 A276 or A312/A312M Bar or Round or Square Pipe Lid AISI 304 A240/A240M or A276 Plate/Sheet or Bar M8 x 1.25 Skt Head Screw AISI 316 Minimum A4 70 All Shims Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Shield Sleeve Brass C360/C260/C230 B16/B16M or B36/B36M or B135/B135M Bar or Plate & Bar or Tube Shield Plug Assembly Material ASTM or Other Specification Form Shield Plug Depleted Uranium SAE AMS 7730 Casting or Ingots Plug Housing AISI 304 A276 or A312/A312M Bar or Round or Square Pipe Plug Lid AISI 304 A276 or A240/A240M Bar or Plate/Sheet Plug Shims Brass C360/C260/C230 B16/B16M or B36/B36M Bar or Plate & Bar Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-15 2.2.1.1 Structural Materials The Inner Container structural materials are fabricated from austenitic stainless steel and titanium. The structural members in the main body (Top Plate, Outer Casing, and Bottom Plate) are fabricated from AISI 304/304L stainless steel in either plate, bar, tube, or pipe form and from Grade 2 titanium in either plate, bar, tube, or pipe for the inner cavity. The closure bolts are minimum AISI 316 SAE Grade A4 (70,000 psi) [2.18, page 1508]. All the insulating and shock absorbing material is fabricated from resin bonded cork.
The structural evaluation of the Outer Drum and Inner Container was assessed under NCT using a temperature range of -40°C to 250°C. In order to carry out the stress analysis a Poisson ratio of 0.3 and a density of 8030 kg/m3 were used for the austenitic stainless steel 304/304L components. A Poisson ratio of 0.3.
2.2.1.2 Shielding Material The shielding is formed from 99% pure depleted uranium with a maximum of 0.45% U-235 by weight.
The mechanical properties of the depleted uranium used in the structural evaluation are presented in Table 2.2.1.A.
2.2.1.3 Cork Assembly The Cork Assembly is manufactured from resin bonded cork in sheet or rolls. The cork may be formed from one piece or from several pieces glued with a contact adhesive. ISO-RAD is using data researched and by information gathered and tested by Croft and Associates and reported in Serco Report SERCO/TAS/002762/01 (See Appendix 2.12.3). In addition, ISO-RAD tested the cork under NCT and HACT conditions and completely protected the Inner Container as the inner container received no damage during testing.
2.2.2 Chemical, Galvanic or Other Reactions
[IAEA SSR-6 614 & 644, IAEA TS-R-1 613 & 642, and 10 CFR 71.43(d)]
The package has been evaluated to determine all the material interactions of chemically or galvanic dissimilar materials. The materials used in the construction of the ISORAD-TC1 package are depleted uranium, tungsten, 300 series stainless steel, cork, brass, aluminum, and titanium. To prevent the possible formation of a eutectic alloy from steel and depleted uranium during the HCAT thermal scenario, defined by 10 CFR 71.73(c)(4), the brass and/or titanium is used as a separator for all steel-uranium interfaces. With this construction, there will be no significant chemical or galvanic reaction between package components during NCT or HACT.
There is no potential for chemical, galvanic or other reactions between the components of the package which are stainless steel and cork in dry conditions, and stainless steel with brass and titanium as eutectic barriers and encapsulated depleted uranium which is sealed and therefore dry. Eutectic formation shall not affect the package performance as the operating temperatures are lower than the eutectic formation temperature.
2.2.3 Effects of Radiation on Materials Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-16
[IAEA SSR-6 614, IAEA TS-R-1 613, and 10 CFR 71.43(d)]
The contents of the package emit one or all of alpha, beta, gamma. Some of the alpha, beta, and gamma sources may emit small amounts of neutron radiation. The use of depleted uranium, tungsten, stainless steel, cork, brass, aluminum, and titanium in Type B(U) transport packaging is well documented and are standard materials of construction. The materials have been in use for decades without degradation of package performance over time. The ISORAD-TC1 will have no material degradation due to irradiation from the package contents. Austenitic stainless steel, brass, depleted uranium, titanium, aluminum, and cork were chosen for the construction of the package because they are durable materials that can withstand the damaging effects from the radiation.
2.3 Fabrication and Examination [IAEA SSR-6 228, 306-309, 640, 809(b), IAEA TS-R-1 232, 306-309, 638, & 807(b), and 10 CFR 71.33(a)(5)]
2.3.1 Fabrication Package components are procured, manufactured and inspected for use under ISO-RAD Canada Incs QA Program submitted to the CNSC. All new fabrication of packaging will be evaluated and documented for compliance to the specifications contained in the drawings provided in Section 1.3 prior to initial use as part of an ISORAD-TC1 transport package. All work performed in the fabrication of the ISORAD-TC1 package is required to be carried out under ISO-RAD Canadas QAP.
The other safety items are fabricated in accordance with the manufacturing instructions created and implemented under the ISO-RAD Canada QAP. All components are machined, cut, shaped, or other process to approved drawings and assemblies are assembled with approved drawings and assembly/
manufacturing instructions. These components are the Outer Drum, Outer Lid Assembly, Attachment Ring Assembly, Inner Container, and Plug Assembly. All welding procedures and personnel shall be qualified in accordance with AWS D1.6/D1.6M and AWS D1.9/D1.9M or equivalent international standard. Welding consumables their supply, certification, control during storage and use, shall comply with the appropriate requirements of AWS D1.6/D1.6M and AWS D1.9/D1.9M or equivalent international standard. The Outer Drum shall be fabricated in accordance with drawing R180831-100.
All welding procedures and personnel shall be qualified in accordance with AWS D1.6/D1.6M and AWS D1.9/D1.9M, or equivalent international standard.
It shall be fabricated using standard industry practices. The cork shall be tested to demonstrate it meets the required specification in drawing R180831-110 and R180831-111.
Any consumables used during manufacture such as bolts, screws, hex nuts, and padlocks shall be procured from commercial suppliers that are approved to a level commensurate with the safety functions of the consumable purchased.
2.3.2 Examination All examinations shall be carried out under the scope of ISO-RAD Canadas QAP. Examinations shall be carried out on materials, components, and finished assemblies throughout the manufacturing process. These tests will assure that the manufactured article meets the critical characteristics to allow Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-17 the safe transport of radioactive material. All tests shall be carried out to approved procedures, with calibrated equipment. Section 8 describes the acceptance testing and routine maintenance requirements for the Inner Containers and other package components used in the ISORAD-TC1 package. The records of the tests will be maintained with the manufacturing records for the life of each package.
The examinations required during manufacture are described below:
Material Tests All metal materials used to fabricate parts of the Outer Drum, Inner Containers, and Inserts shall have Material Test Reports (MTR) with heat numbers, ASTM standard met, and other critical information about the metal being purchased.
Any consumables used during manufacture such as bolts, screws, and hex nuts shall be marked with grade markings by the manufacturer. The minimum grade is A4 (70). Optional exterior padlocks, as a minimum, have a stainless steel shackle and preferable be constructed entirely of stainless steel.
2.3.2.1 Fabrication Tests and Examinations All welds are subjected to non-destructive visual (VT) and liquid penetrant (PT) examination in accordance with AWS D1.6/D1.6M and AWS D1.9/D1.9M or equivalent international standard. The applicable acceptance criteria for the visual examinations are given in drawings R180831-100, R180831-201, R180831-301, R180831-401, R180831-501, and R180831-601. The acceptance standards for the liquid penetrant examination of the welds is in accordance with AWS D1.6/D1.6M and D1.9/D1.9M or equivalent international standard.
All components and assemblies are required to be visually inspected and the dimensions measured using calibrated equipment (calipers, check pins, etc.) to assure compliance with the dimensions shown on the general arrangement drawings. The weight of the finished Inner Container and fully assembled package are required to be measured to ensure the weight requirements are met.
2.3.2.2 Acceptance Tests The completed Inner Containers are open to the atmosphere and do not required pressure testing as the packages is not sealed with O-Rings or other seals. The Outer Drum is also open to the atmosphere and will not require pressure testing.
On completion of manufacture the Inner Container will be surveyed in the Outer Drum forming the completed ISORAD-TC package ensuring proper shielding of radioactive contents and external dose rates.
After completion of all acceptance tests and verification of the manufacturing processes, a Certificate of Conformance for each ISO-RAD-TC1 will be issued for each serial numbered package.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-18 2.4 General Standards for All Packages [IAEA SSR-6, IAEA TS-R-1, and 10 CFR 71.43]
2.4.1 Minimum Package Size [IAEA SSR-6 636, IAEA TS-R-1 634, and 10 CFR 71.43(a)]
The ISORAD-TC1 package exceeds the minimum size requirements since it is 403.24 mm (15.876 inch) in diameter and 589 mm (23.1875 inches) high.
2.4.2 Tamper-Indicating Feature [IAEA SSR-6 637, IAEA TS-R-1 635, and 10 CFR 71.43(b)]
The ISORAD-TC1 package outer lid is secured with eight M14 x 2 stainless steel bolts and two lock studs with optional padlocks. A tamper evident lead wire seal, unique zip tie, or other tamper evident seal or device is used to demonstrate if the package has been tampered with. The tamper seal is placed through the holes and in between the Lock Studs. This seal wire is not readily breakable, therefore if it is broken during transport, it serves as evidence of possible unauthorized access to the contents.
2.4.3 Positive Closure [IAEA SSR-6 641, IAEA TS-R-1 639, and 10 CFR 71.43(c)]
The ISORAD-TC1 does not contain valves or pressure release mechanisms. The radioactive material is sealed inside special form capsules and placed inside the ISORAD-TC1 Inner Container, then Inner Container shielded plug and lid are secured as described in paragraph 2.1.1.4. These features maintain positive closure of the transport package and security of the radioactive material during transport.
2.5 Lifting and Tie-down Standards for All Packages 2.5.1 Lifting Devices
[IAEA SSR-6 503(a) & 608, IAEA TS-R-1 502(b), 606, 607, & 608, and 10 CFR 71.45(a)]
The ISORAD-TC1 package is designed to be lifted using straps or chains through the handles cut through the wall of the drum near the top. The ISORAD-TC1 package is approximately 403.24 mm (15.876 inch) in diameter and 589 mm (23.1875 inches) high with the handle openings through the keg wall are approximately 38.1 mm (1.5 inches) high x 120.65 mm (4.75 inches) wide with at least 38.1 mm (1.5 inches) of stainless steel separating the top of the handle opening to the top of the keg. The maximum stress on the handle openings is:
The following force vectors can be determined by using the given load combination, which consists of three separate forces (where Vector 1 is the force that is horizontal along the direction of the vehicle (hv), Vector 2 is the vertical force (vf), and Vector 3 is the horizontal transverse force (ht)).
Vector 1= 10(300 pounds) = 3000hv + 0vf + 0ht Vector 2 = 2(300 pounds) = 0hv + 600vf + 0ht Vector 3 = 5(300 pounds) = 0hv + 0vf + 1500ht The formula to determine the force magnitude is:
= ( 2 ) + ( 2 ) + ( 2 )
= 30002 + 6002 + 15002 R =9000000 + 360000 + 2250000 = 3407.35 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-19 The assumption is the rounded load of 3407.4 will act fully in each direction in the interest of conservatism and simplicity. The calculations assume the full load will act in each of the vector directions. The angular contribution in the x, y, and z directions will not be considered.
The worst case is assumed with the greatest pulling force being in the y plane or horizontal direction. If 3668 pounds were acting on the x, y, and z planes, then the worst case for the handles of the drum would be the force pulling along the horizontal or y plane.
The area used in the following calculation is an estimated number for extreme conservatism. Each handle opening through the keg wall is approximately 38.1 mm (1.5 inches) high x 120.65 mm (4.75 inches) wide of 7.125 in2 with at least 38.1 mm (1.5 inches) stainless steel separating the top of the handle opening to the top of the keg. ISO-RAD has assumed to be extremely conservative the area used by both handles is approximately 0.50 in2, which gives the following:
3407.4 3407.4
=
= 7.125
478.3
= .5
= 6814.8 600 600
=
= 7.125
84.21
= .5
= 1200 Therefore, the actual stress generated in each handle is 478.3 psi. The stress generated using the conservative number in handle is 6814.8 psi. With an additional safety factor of 3 applied, the maximum stress in the lifting handle is 20,444.4 psi. This is substantially below the 39,000 psi yield strength of the stainless steel. Therefore, the lifting handle can support more than three times the weight of the transport package as required by IAEA SSR-1 636, IAEA TS-R-1 634, and 10 CFR 71.45(a).
The package handles will not fail and if the handle openings failed it would not impede the ability of the package to meet the other safety requirements.
2.5.2 Tie-Down Devices
[IAEA SSR-6 607& 638, IAEA TS-R-1 606 & 636, and 10 CFR 71.45(b)]
The ISORAD-TC1 package has no specifically designed tie-down devices. The normal method of securing the package during transport is expected to be using dunnage, cargo nets or an equivalent system that envelope the package without being attached to it: such a system cannot stress the structure of the package. The package may be secured in either the horizontal or vertical position.
The stresses and safety factors demonstrated in Section 2.5.1 demonstrate the forces applied to the package will not exceed the yield strength with a safety factor of three.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-20 2.6 Normal Conditions of Transport (NCT)
[IAEA SSR-6 648, 659(a), & 719-724, IAEA TS-R-1 646, 657(a), & 719-724, and 10 CFR 71.43(f)
& 71.71]
Table 2.6.A: Summary of NCT Initial Conditions Load NCT Initial Conditions (According to Regulatory Guide 7.8)
ID Ambient Internal Fabrication NCT Insolation Decay heat Temperature Pressure Stress 38°C -40°C Max Zero Max Zero Max Zero 1 Hot 38°C Ambient X X X X Cold -40°C 2 X X X X Ambient Reduced External 3 X X X X X Pressure 25 kPa Increased External 4 X X X X X Pressure 140 kPa 5 X X X X X Vibration 6 X X X X X 7 X X X X X Free Drop (Side) 8 X X X X X 9 X X X X X Free Drop (Top) 10 X X X X X 2.6.1 Heat [IAEA SSR-6 615, 617, 618, 639, 653, 664, & 666, IAEA TS-R-1 615, 617, 618, 637, 651, 662, & 664, 10 CFR 71.71(c)(1), and USNRC Regulatory Guide 7.8]
According to IAEA and 10 CFR, the package must be evaluated in an ambient temperature of 38oC, in still air and insolation. Under these conditions the maximum temperature and pressure generated have been calculated and discussed in Section 2.6.1.1. These temperatures and pressures have then been used to determine the differential thermal expansion in Section 2.6.1.2 and therefore the stresses present in the containment system. The calculated stresses are then used to determine if the containment system meets the structural design criteria.
The largest heat source for the ISORAD-TC1 transport package with the BPIC Series installed is 10,125 Curies (375 TBq) of Iridium-192, generating approximately 6.013 milliwatts per Curie or a total of 60.88 watts (see Table 2.6.1.A). The thermal test was conducted assuming 65 Watts (10,810 curies) of Ir-192 which is approximately 6.75 percent higher than the operational limit 60.88 watts (10,125 curies). The thermal evaluation for the heat test is described in Section 3.
Table 2.6.1.A: Radionuclide Decay Energy Radionuclide Package Activity Ci (TBq) MeV/Decay Watts/Package Iridium-192 10,125 (375 TBq) 1.46 60.88 Se-75 10,000 0.86 51 Yb-169 10,000 0.91 54 Cs-137 500 1.18 4 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-21 Table 2.6.1.B: ISORAD-TC1 Authorized Contents BPIC Inner Container 3000 x A1 Highway Special Special Authorized Gamma Energies Half Life Watt Total No Quantities Route Isotope Form Form Forms (keV) (days) Per Ci Watts Exceed Control Type A1 (Ci) Type B (Ci)
These Limits 1000 TBq Alpha Actinium-225 Special 10 21.6 67.75 0.0346 2.35 64000 27000 5.93 MeV Beta 21.277 Actinium-227 Special 24.3 25 0.000472 0.0118 72900 27000 44.77 Years Beta 6.13 Actinium-228 Special 16.2 17 0.008 0.136 48600 27000 2.12 MeV Hours Beta Barium-131 Special 11.5 54 5076 0.00306 15.54 162000 27000 353.8 Barium-133 Special 80-356 3848.7 81 125 0.00535 0.669 243000 27000 Cadmium-109 Special 88 462.6 810 2700 0.00535 14.45 2430000 27000 Beta 5700 Carbon-14 Special 1080 1080 0.000293 0.32 3240000 27000 156 Years Cobalt-57 Special 122 271 270 300 0.00535 1.61 810000 27000 5.27 Cobalt-60 Special 1.3 MeV 10.8 0.5 0.0154 0.0077 32400 27000 Years X-Ray Cesium-131 Special 9.7 810 26900 0.000165 4.44 2430000 27000 29.5-33.5 Cesium-134 Special 764 753.73 18.9 190.35 0.0102 1.95 56700 27000 Cesium-137 Special 662 10,983 54 3888 0.00101 3.927 162000 27000 Copper-67 Special 184.6 2.58 270 18657 0.00161 30.04 810000 27000 Alpha 340,000 Curium-248 Special 0.54 5 0.0359 0.18 1620 27000 5.08 MeV Years Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-22 Table 2.6.1.B: ISORAD-TC1 Authorized Contents BPIC Inner Container (Continued) 3000 x A1 Highway Special Special Authorized Gamma Energies Half Life Watt Total No Quantities Route Isotope Form Form Forms (keV) (days) Per Ci Watts Exceed Control Type A1 (Ci) Type B (Ci)
These Limits 1000 TBq Europium-152 Special 344-1408 4941 27 81 0.0077 0.624 81000 27000 Indium-111 Special 245.4 2.8 81 11475 0.0026 29.84 243000 27000 Iridium-192 Special 206-612 73.8 27 1015 0.006013 60.88 81000 27000 Iridium-194 Special 147 0.803 8.1 55.35 0.00535 0.296 24300 27000 X-Ray Iron-55 Special 985.5 1080 2700 0.000165 0.45 3240000 27000 5.9-6.5 Lutetium-177 Special 113-208 6.73 810 10017 0.00535 53.6 2430000 27000 Beta Phosphorous-32 Special 14.3 13.5 150.66 0.00535 0.81 40500 27000 1710.4 Beta Phosphorous-33 Special 25.4 1080 10017 0.00535 53.6 3240000 27000 250 Selenium-75 Special 66-401 119.8 81 10017 0.00535 53.6 243000 27000 Sodium-22 Special 511 950.57 13.5 2700 0.00535 14.5 40500 27000 Beta Sodium-24 Special 0.625 5.4 5.4 0.00535 0.029 16200 27000 472 Beta Strontium-89 Special 50.5 16.2 329.4 0.00346 1.14 48600 27000 1.46 MeV Beta 28.8 Strontium-90 Special 8.1 50 0.00346 0.163 24300 27000 546 Years Alpha Thorium-227 Special 18.7 270 272.7 0.0359 9.79 810000 27000 6.7-7 MeV Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-23 Table 2.6.1.B: ISORAD-TC1 Authorized Contents BPIC Inner Container (Continued) 3000 x A1 Highway Special Special Authorized Gamma Energies Half Life Watt Total No Quantities Route Isotope Form Form Forms (keV) (days) Per Ci Watts Exceed Control Type A1 (Ci) Type B (Ci)
These Limits 1000 TBq Alpha Thorium-228 Special 698.25 13.5 150.66 0.0321 4.84 40500 27000 84.4 Gamma Tungsten-187 Special Beta 0.998 54 604.8 0.00454 2.75 162000 27000 Tungsten-188 Special 155 69.78 10.8 17.36 0.000598 0.0104 32400 27000 Beta Yttrium-90 Special 2.67 8.1 46.71 0.000554 0.0259 24300 27000 2.28 MeV Ytterbium-169 Special 8-308 32 108 10017 0.00251 25.143 324000 27000 Beta Ytterbium-175 Special 4.185 810 26900 0.001 26.9 2430000 27000 470 Zinc-65 Special 1115 243.86 54 500 0.0154 7.7 162000 27000 Alpha Other Isotopes Special Beta Any Type A Type A Any 30 Various 27000 Gamma Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-24 2.6.1.1 Summary of Pressures and Temperatures
[IAEA SSR-6 616 & 663, IAEA TS-R-1 615 & 661, 10 CFR]
Table 2.6.1.C: Summary Temperatures Normal Conditions of Transport Temperature Condition ISORAD-TC1 External Surface Min-Max Average Insolation (65 Watts) 51.26-104.21 (124.27 - 219.58) 78.76 (173.77)
(38 in full sun)
Decay Heating (65 Watts) 40.48 - 48.82 (104.86 - 119.88) 44.63 (112.35)
(38 in shade)
The calculated maximum temperatures in the Inner Container and Outer Drum with maximum heat load of 65W under NCT are shown in Section 3, Table 3-1.B through Table 3.1D. The maximum temperature for the Inner Container containment system is 244.46oC with insolation. Some of the stress calculations were carried out assuming a temperature of 250°C. The temperatures are divergent by 5.54oC, which would not cause the results of the test to be any different from those presented here and would represent a worst case.
Table 2.6.1.D: NCT Maximum Temperature and Pressure in the Special Form Capsule Special Form Capsule Precalculated NCT Conversion for Formula Starting Temp (T1) 20°C (68°F) 293.15°K Max Temp (T2) 250.0°C (472.03°F) 523.15°K 1
Starting Pressure (P ) 100 kPa (14.504 psi) 1.0 bar abs = 750.061683 Torr Max Pressure (P2) 178.452 kPa (25.883 psi) 1.785 bar abs = 1338.5 Torr Special Form Capsule NCT Thermal Simulation Conversion for Formula Starting Temp (T1) 20°C (68°F) 293.15°K Max Temp (T2) 244.46°C (472.03°F) 517.61°K Starting Pressure (P1) 100 kPa (14.504 psi) 1.0 bar abs = 750.061683 Torr 2
Max Pressure (P ) 176.57 kPa (25.61 psi) 1.765 bar abs = 1323.2 Torr 2 = 1 x 2 ÷ 1 P2 = (750.061683 Torr x 517.61°K) / 293.15°K P2 = (388,239.42774) / 293.15 P2 = 1324.37124 Torr = 1.7656831 bar abs = 176.56831 kPa = 25.6090682 psi The thermal simulation value has been used in the structural evaluation. A calculation of the actual maximum pressure expected under NCT is provided in (Section 2.6.3 & 2.6.4). The precalculated maximum pressure is 1.785 bar abs (178.452 kPa) or 25.883 psi and with the simulation analysis the maximum pressure is 1.765 bar abs (176.57 kPa) or 25.61 psi under NCT.
As all components are open to the ambient atmosphere, no pressure will build up in any part of the Inner Container or Outer Drum under NTC that would adversely affect package performance or integrity. The maximum pressure build-up under Normal Conditions of transport within the special form capsule is 176.5 kPa (25.61 psi) as demonstrated above. As demonstrated in Section 3.5.3.2 for the HACT thermal assessment, the Inner Container and the special form capsule can withstand this pressure differential with no adverse effect on the containment.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-25 Table 2.6.1.E: HACT Maximum Temperature and Pressure in the Special Form Capsule Special Form Capsule Precalculated HACT Conversion for Formula 1
Starting Temp (T ) 20°C (68°F) 293.15°K Max Temp (T2) 800.0°C (1472.03°F) 1073°K 1
Starting Pressure (P ) 102.5 kPa (14.87 psi) 1.02 bar abs = 768.813227 Torr Max Pressure (P2) 375.367 kPa (54.45 psi) 3.754 bar abs = 2815.4841Torr Special Form Capsule HACT Thermal Simulation Conversion for Formula Starting Temp (T1) 20°C (68°F) 293.15°K Max Temp (T2) 244.46°C (472.03°F) 517.61°K 1
Starting Pressure (P ) 102.5 kPa (14.87 psi) 1.02 bar abs = 768.813227 Torr Max Pressure (P2) 181.075 kPa (26.263 psi) 1.8 bar abs = 1365.11227 Torr Precalculated Pressure 1 2 2 =
1 Where: P1: Atmospheric Pressure at the time of sealing (102.5 kPa) (A conservative assumption)
P2: Pressure resulting from the Maximum Hypothetical Accident Temperature T1: Temperature at the time of sealing (20°C or 293°K) (a conservative assumption)
T2: Maximum Hypothetical Accident Temperature (800°C or 1,073°K)
P2 = P1(T2)/T1 P2 = 102.5(1073)/293 P2 = 109982.5/293 P2 = 375.36689 kPa or 54.4423646 psi rounded to 54.45 psi for Precalculated HACT Calculated Pressure HACT Thermal Simulation 2
1 2
= 1 Where: P1: Atmospheric Pressure at the time of sealing (102.5 kPa) (A conservative assumption)
P2: Pressure resulting from the Maximum Hypothetical Accident Temperature T1: Temperature at the time of sealing (20°C or 293°K) (a conservative assumption)
T2: Maximum Hypothetical Accident Temperature (244.46°C or 517.61°K)
P2 = P1(T2)/T1 P2 = 102.5(517.61)/293 P2 = 53055.05/293 P2 = 181.07517 kPa or 26.262733 psi rounded to 26.263 psi for Thermal Analysis HACT The regulations stipulate that upper pressure experienced by the containment vessel (special form capsule) cannot exceed 700 kPa gauge pressure. The precalculation of the worst case maximum pressure expected under HACT assuming the Special Form Capsule reached 800°C is 3.754 bar abs (375.367 kPa) or 54.45 psi. The maximum pressure is over 46% lower than the limit. The simulation Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-26 thermal analysis showed a maximum pressure of 1.82 bar abs (181.075 kPa) or 26.263 psi under HACT. The thermal simulation maximum pressure is over 74% lower than the limit.
2.6.1.2 Differential Thermal Expansion The finite element analysis model investigated the deformations caused within the containment system as a result of the differing expansion rates of the depleted uranium shielding and the stainless steel outer casing. The results of the analysis included the effect of differential thermal expansion in both the radial and longitudinal directions.
The ISORAD-TC1 Outer Drum is designed to have a mm clearance between the cork and the Cavity Sleeve, mm clearance between the Cavity Sleeve and the Inner Container, and another 6 mm clearance between the cork and the Outer Drum lid. As the cork is free standing within the Outer Drum this allows movement of the top cork of up to mm and hence expansion of the Inner Container of mm. There was no significant expansion of the Inner Container therefore it will not impact the structural integrity of the package.
The model has assumed no gap is present between the depleted uranium, brass, and the stainless steel and determined the stresses within the containment vessel boundary caused as a result of the differing thermal expansion rates. The results of the stress calculations are discussed in section 2.6.1.3.
Any thermal expansion encountered during NCT transport will be insignificant with respect to the manufacturing tolerances of the package.
For example:
Expansion of the Outer Casing circumference of the BPIC - Round Inner Container is approximated by:
Expansion Outer diameter derivative formula [2.18, page 405]
E = DT Where: D = Diameter of the Outer Casing = in
= Coefficient of Thermal expansion 9.9in/in T = Cold temperature differential (from -40 to 100 = 140)
T = Hot temperature differential (from 100°F to 472.028°F = 372.028)
E = (6.693in)(9.9 in/in)(140)
E = (3.14159)(6.693)(0.0000099)(140)
E = (21.02666187)(0.001386)= 0.02915 (cold)
E = (6.693in)(9.9 in/in)(372.028)
E = (3.14159)(6.693)(0.0000099)(372.028)
E = (21.02666187)(0.0036830772)= 0.0775 (hot)
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-27 Table 2.6.1.F: Summary of Thermal Expansion of BPIC Inner Container BPIC - Round Material Cold Temp Hot Temp Coefficient Cold Dia Hot Dia
°C (F°) °C (F°) µmm (µin) mm (in) mm (in) 300 Stainless 60 (140) 188.91 (372.03) 17.3 (9.9) 0.75 (0.0292) 1.97 (0.078)
Steel 400 Stainless 60 (140) 188.91 (372.03) 10.8 (5.7) 0.074(0.0029) 0.196(0.0077)
Steel Brass 60 (140) 188.91 (372.03) 20.2 (11.2) 0.80 (0.0313) 2.11(0.0831)
Titanium 60 (140) 188.91 (372.03) 8.6 (4.95) 0.14 (0.0055) 0.37(0.0145)
Depleted 60 (140) 188.91 (372.03) 13.9 (7.4) 0.52 (0.0205) 1.38(0.0545)
Uranium The full analysis is contained in section 3.5.4. Further, the Inner Container Lid and closure bolts /
hardware will expand at approximately the same rate which will maintain the security of the package. All other package components have similar tolerances. Any expansion in this temperature range will be well within these tolerances.
2.6.1.3 Stress Calculations As shown in Section 2.6.1.2, thermal differentials will have no adverse effect of the interfaces between the package components. Mechanical loads at the maximum package weight (300 lbs.) are well distributed across the package bottom and are small compared to the yield strength of the stainless steel (39,000 psi - See Table 2.2). Assuming the mechanical stress from the Inner Container is localized around the sides of the square casing portion of the Inner Container gives:
Inner Container (outside dimensions) = inches Length x inches Width Area of Inner Container bottom = 44.7951 in2 Mass of Inner Shield = 68.1 kg (150 pounds) rounded up for conservatism Stress on Inner Container bottom = 150 pounds/44.7951 in2 = 3.349 psi Stress Area of Bolt [2.18, page 1502]
Where: A = Stressed Area dn = Nominal bolt diameter n = 1/p = 1/p = number of threads per inch or cm p = pitch in inch or cm (length per thread)
A= /4 (dn-0.9743/n)2 A = (3.14159)/4(6-(0.9743/1.25)2 A = (0.7853975)(6-0.77944)2 A = (0.7853975)(27.2542467136) = 21.405422 mm2 A = 0.03317846 in2 Stresses developed within the Inner Container cavity will be negligible as the cavity is not sealed.
Assuming the Inner Container was a perfect seal and no escaping gasses then the force on the Inner Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-28 Container cavity with just the Plug Assembly bolted in and the Outer Container Lid is not on the package is estimated by:
F = (D2/4)P Where: D = Max Diameter of the Plug Assembly lid = 4.0 in P = Pressure induced by the thermal gradient = 54.45 psi (from Table 6.6.1.F Worst Case)
Worst Case 800°C Inner Container Actual Case 244.46°C Inner Container F = (4.0 in)2/4)54.45 psi F = (4.0 in)2/4) 26.263 psi F = ((3.14159)(16)/4)(54.45) F = ((3.14159)(16)/4)(26.263)
F = (12.567)(54.45) = 684.7315 (lbf) F = (12.567)(26.263) = 330.04712 (lbf)
Where: Fi = Force in each bolt A = Stress area of the bolt = 0.03317846 in2 Worst Case 800°C Inner Container Actual Case 244.46°C Inner Container S = Force/Area = Fi/A S = Force/Area = Fi/A S = 171.06828/0.03317846 S = 82.51178/0.03317846 S = 5,156.0135 psi on each bolt S = 2,486.915 psi on each bolt The cover is held by four (4) M6 x 1 stainless steel bolts. This imparts a force of 171.06828 lbf in each bolt in the worst case scenario and imparts a force of 82.512 lbf in each bolt as a result of the thermal analysis. However, if all the stress is assumed to be taken by only one bolt, then the stress in that bolt equals:
Where: Fo = Force in each bolt A = Stress area of the bolt Worst Case 800°C Inner Container Actual Case 244.46°C Inner Container S = Force/Area = Fo/A S = Force/Area = Fo/A S = 684.7315/0.03317846 S = 330.0471/0.03317846 S = 20,637.869 psi (all stress one bolt) S = 9,947.65 psi (all stress one bolt)
The above calculations represent the worst case scenario of the Inner Container Plug Assembly becoming sealed and the Inner Container reaching 800°C. The results are still below the tensile strength value for the M6 x 1 Stainless Steel Grade A4 bolt (70,000 psi nominal). Once again even though the Inner Container is not gasketed and will not hold pressure, if it did, a single bolt would afford a safety margin of 70,000/20,637.869 = 3.39 times safety margin. With all four bolts installed 70,000/5156.0135 = 13.58 times safety margin. All other Inner Containers would have similar or lower stresses. The actual conditions based on thermal simulation result in a much higher safety margin. The M6 x 1 Stainless Steel Grade A4 bolt (70,000 psi nominal) has a high tensile strength.
The results based on the thermal simulation, assuming the Inner Container is gasketed and will hold pressure, a single bolt would afford a safety margin of 70,000/9,947.65 = 7.037 times safety margin.
With all four bolts installed 70,000/2,486.92 = 28.15 times safety margin.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-29 2.6.1.4 Comparison with Allowable Stresses All stresses calculated in Section 2.6.1 are well below strengths for the materials of construction.
Further, the ISORAD-TC1 package was fully tested and passed under NCT. It is therefore concluded that the ISORAD-TC1 package will satisfy the performance requirements specified by the regulations.
2.6.2 Cold [IAEA SSR-6 616, 639, & 666, IAEA TS-R-1 615, 637, & 664, 10 CFR 71.71(c)(2)]
The ISORAD-TC1 package does not contain carbon steel components that are susceptible to brittle fracture at low temperature. The transport package, however, successfully met Type B(U)-96 Transport Tests requirements at temperatures below -40°C (-40°F), the minimum specified in the regulations. Thus, it is concluded that the ISORAD-TC1 package will withstand the normal transport cold condition.
2.6.3 Reduced External Pressure
[IAEA SSR-6 645 & 621, IAEA TS-R-1 643 & 619, 10 CFR 71.71(c)(3), and 49 CFR 173.412(f)]
The ISORAD-TC1 package does not include gaskets between the Inner Container Top Plate and the Inner Container Lid and therefore is open to the atmosphere. The package must be able to withstand the reduced external pressure of IAEA SSR-6, IAEA TS-R-1, 10 CFR 71 and 49 CFR 173. The ISORAD-TC1 package is designed to only transport special form capsules.
The following analysis demonstrates that under NCT reduced pressure condition, the package will meet or exceed the applicable performance requirements specified in the regulations:
- 1. there will be no loss or dispersal of contents;
- 2. there will be no structural changes that reduce the effectiveness of components required for shielding, heat transfer, or containment;
- 3. any loss of shielding integrity would not result in more than a 20% increase in the radiation level at any external surface of the package; and
- 4. there will be no changes that would affect the ability of the package to withstand the hypothetical accident conditions tests.
The regulations listed above specify four reduced pressure conditions:
"The containment system shall retain its radioactive contents under a reduction of ambient pressure to 60 kPa (8.7 psia)."
- 2. IAEA TS-R-1 (1996) Revised Paragraph 619 state: "Packages containing radioactive material transported by air shall have a containment system able to withstand without leakage a reduction in ambient pressure to 5 kPa.
- 3. IAEA SSR-6 (2018) Paragraph 621 and TS-R-1 (2005) Paragraph 619 state: Packages containing radioactive material to be transported by air shall be capable of withstanding, without leakage, an internal pressure which produces a pressure differential of not less than maximum normal operating pressure plus 95 kPa."
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-30
- 4. 10 CFR 71.71 (c)(3) requires the package to withstand a reduced pressure of 25 kPa.
The analysis will be conducted using the most conservative conditions and have assumed the following:
- 1. The external pressure is only 5 kPa.
- 2. The package is at HACT temperature of 800°C For this analysis, we have considered the package in four elements: Inner Container, Inner Container Contents Cavity, and Containment (Special Form Capsule).
Inner Container The Inner Container (BPIC) is open to the atmosphere so no pressure differential exists within the outer package areas. The stainless steel casing containing the depleted uranium or tungsten shield is sealed by means of the welding the Top Plate and Bottom Plate to the Outer Casing. The space between the shield and the housing is filled brass and a small amount of air.
Inner Container Contents Cavity The Inner Container Contents Cavity (the location where the Special Form capsules are located),
although enclosed, is not sealed; no gaskets or O-rings or other sealing devices are used to seal the Contents Cavity. Therefore, the Contents Cavity is open to the atmosphere and no pressure differential within the Contents Cavity.
A further assumption is that the air behaves as an ideal gas, the pressure within this special form capsule would be calculated from:
1 2 2 =
1 Where: P1: Atmospheric Pressure at the time of sealing (102.5 kPa)
P2: Pressure resulting from the Maximum Hypothetical Accident Temperature T1: Temperature at the time of sealing (20°C or 293°K) (a conservative assumption)
T2: Maximum Hypothetical Accident Temperature (800°C or 1,073°K)
P2 = P1(T2)/T1 P2 = 102.5(1073)/293 P2 = 109982.5/293 P2 = 375.366894198 kPa or 54.4423652 psi The formula result is a maximum pressure from HACT temperature of 800°C is 375.67 kPa or 54.45 psi.
The increase in pressure within the Inner Container between the Inside wall and the Shield, which is welded closed (sealed) would increase stress. The increase in pressure within the Contents Cavity in between the Contents Cavity and the Shield would also result in increased stress. The assumption is Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-33 stress in the Inner Container casing is less than 6% of the yield strength of the austenitic stainless steel at the test temperature.
Containment (Special Form Capsule)
The special form capsule is constructed of austenitic stainless steel, titanium, or vanadium. The calculations will use the austenitic stainless steel because it has a lower yield strength than titanium or vanadium.
A further assumption is that the air behaves as an ideal gas, the pressure within this special form capsule would be calculated from:
1 2 2 =
1 Where: P1: Atmospheric Pressure at the time of sealing (102.5 kPa)
P2: Pressure resulting from the Maximum Hypothetical Accident Temperature T1: Temperature at the time of sealing (20°C or 293°K) (a conservative assumption)
T2: Maximum Hypothetical Accident Temperature (800°C or 1,073°K)
P2 = P1(T2)/T1 P2 = 102.5(1073)/293 P2 = 109982.5/293 P2 = 375.366894198 kPa or 54.4423652 psi The formula result is a maximum pressure from HACT temperature of 800°C is 375.37 kPa or 54.45 psi.
The ISORAD-TC1 is designed and intended to be used with a wide range of Special Form Capsules.
Perhaps the most vulnerable special form capsule would be a long length thin walled special form capsule fabricated from stainless steel. The assumed worst case Special Form Capsule would and have an outside diameter of 13.75 mm and a wall thickness of 0.25 mm. (Titanium and Vanadium have higher yield strengths than austenitic stainless steel, and therefore a capsule with similar dimensions would be less vulnerable.) The tensile stress induced in a thin-walled pressure vessel from an internal pressure can be described by Barlow's formula (Machinery's Handbook, 27th Edition, page 327):
=
2 Where: P: Pressure Differential 375.67 kPa (54.45 psi) di: Inside Diameter 13.25 mm (0.52165354 in) t: Wall Thickness 0.25 mm (0.00984252 in)
= Pdi / 2t
= 375.67(13.25) / 2(0.25)
= 4,977.6275 / 0.5 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-34
= 9,955.255 kPa or 9.955 MPa or 1443.88766 psi (Contents Cavity and the Special Form)
The same hypothetical worst case Special Form capsule (13.75 mm outside diameter and 0.25 mm wall thickness), the pressure differential stress is calculated to be 9.955 MPa (1443.888 psi). At a temperature of 870°C (1600°F) the yield strength of type austenitic stainless steel is 69 MPa (10,000 psi).
The longitudinal stress for Special Form Capsule is calculated from:
2
=
( + 2)2 2 Where: l: Longitudinal Stress Pl: Longitudinal Pressure: (375.67 kPa or 54.45 psi) d: Inside diameter of the Cylinder (13.25 mm or 0.52165354 in) t: Thickness of the Cylinder (0.25 mm or 0.00984252 in) l = P(d2) / (d + 2t)2 - d2 l = 375.67(13.25)2 / (13.25 + 2(0.25))2 - (13.25)2 l = 375.67(175.5625) / (13.25 + 0.5)2 - (175.5625) l = 65,953.564375 / (13.75)2 - 175.5625 l = 65,953.564375 / 189.0625 - 175.5625 l = 65,953.564375 / 13.4495 l = 4,885.630162228231 kPa or 4.886 MPa or 708.600746 psi For the same hypothetical worst case Special Form capsule (13.75 mm outside diameter and 0.25 mm wall thickness), the longitudinal stress is calculated to be 4.886 MPa (708.60 psi). At a temperature of 870°C (1600°F) the yield strength of type austenitic stainless steel is 69 MPa (10,000 psi).
Therefore, under the conditions of the reduced pressure at the HACT thermal temperature, The maximum differential pressure stress generated in the worst cast Special Form Capsule is less than 15% of the yield strength of the austenitic stainless steel at the test temperature. The maximum longitudinal pressure stress in the Special Form Capsule is less than 7% of the yield strength of the austenitic stainless steel at the test temperature. The reduced external pressure evaluation demonstrates that the package satisfies the applicable performance requirements specified in the regulations.
2.6.4 Increased External Pressure [IAEA SSR-6 616, IAEA TS-R-1 615, and 10 CFR 71.71(c)(4)]
The ISORAD-TC1 package does not include gaskets between the Inner Container body and the Inner Container Lid and therefore is open to the atmosphere. The package must be able to withstand the increased external pressure of IAEA and 49 CFR do not provide a kPa (psi) value and 10 CFR of 140 kPa (20 psi).
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-35 The Inner Container and the Contents Cavity will withstand this pressure without loss of structural integrity. The ISORAD-TC1 transport package components are open to the atmosphere and contain no items which could create a differential pressure relative to atmospheric conditions. As demonstrated in Section 2.6.1, the package containment can withstand a pressure differential 54.45 psi, therefore the increased external pressure requirement of 20 psi in 10 CFR 71 will not adversely affect the package containment.
The following analysis demonstrates that under NCT reduced pressure condition, the package will meet or exceed the applicable performance requirements specified in the regulations:
- 1. there will be no loss or dispersal of contents;
- 2. there will be no structural changes that reduce the effectiveness of components required for shielding, heat transfer, or containment;
- 3. any loss of shielding integrity would not result in more than a 20% increase in the radiation level at any external surface of the package; and
- 4. there will be no changes that would affect the ability of the package to withstand the hypothetical accident conditions tests.
The regulations listed above specify two increased pressure conditions:
- 1. IAEA SSR-1 Paragraph 617 and IAEA TS-R-1 Paragraph 615 state: "The design of the package shall take into account ambient temperatures and pressures that are likely to be encountered in routine conditions of transport."
- 2. 10 CFR 71.71 (c)(4) states, increased external pressure. An external pressure of 140 kPa (20 lbf/in2) absolute.
The analysis will be conducted using the most conservative conditions and have assumed the following:
- 1. The external pressure is 140 kPa.
- 2. The internal pressure in the sealed areas is 0 kPa For this analysis, we have considered the package in four elements: Inner Container, Inner Container Contents Cavity, and Containment (Special Form Capsule).
Inner Container The Inner Container (BPIC) is open to the atmosphere so no pressure differential exists within the outer package areas. The stainless steel casing containing the depleted uranium or tungsten shield is sealed by means of the welding the Top Plate and Bottom Plate to the Outer Casing. The space between the shield and the housing is filled brass and a small amount of air. The assumption is the internal air sealed within the container is at 0 kPa.
The assumption is the Inner Container behaves as a cylindrical tube. The collapsing pressure of a tube can be described (Machinery's Handbook, 27 Edition, pages 292 - 298) by:
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-37 Containment (Special Form Capsule)
The special form capsule is constructed of austenitic stainless steel, titanium, or vanadium. The calculations will use the austenitic stainless steel because it has a lower yield strength than titanium or vanadium.
Using the same worst case Special Form Capsule from the reduced external pressure, the following calculation was made:
8 3
= 3.46 x 10 x Where: P: Collapsing Pressure of Capsule d: Outside Diameter 13.75 mm (0.54133858 inch)
T: Wall Thickness = 0.25 mm (0.00984252 inch)
P = 3.46 x 108 x (T/d)3 P = 346000000 x (0.009843/0.54133858)3 P = 346000000 x (0.018180271)3 P = 346000000 x 0.00000600898414 P = 2,079.10851244 kPa or 2.079 MPa or 301.549195 psi The calculation demonstrates the collapsing pressure of the austenitic stainless steel Special Form Capsule is 2.079 MPa (301.55 psi). The NCT increased pressure condition (140 kPa) represents less than 7% of the collapsing pressure.
The increased external pressure evaluation demonstrates that the package satisfies the applicable performance requirements specified in the regulations.
2.6.5 Vibration [IAEA SSR-6 613, IAEA TS-R-1 612, and 10 CFR 71.71(c)(5)]
The ISORAD-TC1 package incorporates the use of bolts, lock washers, and optional padlocks to secure the Outer Drum Lid Assembly. The cork material is a vibration absorber and should perform better than other similar transport containers. It is therefore concluded that the ISORAD-TC1 will withstand vibration normally incident to transport.
2.6.6 Water Spray [IAEA SSR-6 719, 720, & 721, IAEA TS-R-1 719, 720, & 721, and 10 CFR 71.71(c)(6)]
The ISORAD-TC1 packages are constructed of water-resistant materials throughout. Therefore, the water spray test would not reduce the shielding effectiveness or structural integrity of the package.
The test was not conducted.
2.6.7 Free Drop [IAEA SSR-6 722(a), IAEA TS-R-1 722(a), and 10 CFR 71.71(c)(7)]
ISORAD-TC1 Prototype #1 was subjected to two 1.2 meter (4 foot) free drops in accordance with Test Plan QTP-001 (Section 2.12.1). The orientation of the first 1.2 meter (4 foot) free drop (see Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-38 Figure 2.6.7.a) was selected because of its potential to cause significant deformation of the Outer Drum and with the intent of breaking the lid bolts with high stress loads potentially causing the Inner Container to be ejected from the Outer Drum. Prototype #1 was dropped at about a 45 angle onto top side edge of the Outer Drum. The orientation of the second 1.2 meter (4 foot) free drop (see Figure 2.6.7.b) was selected because of its potential to cause significant deformation of the Outer Drum and with the intent of breaking the lid bolts with high stress loads potentially causing the Inner Container to be ejected from the Outer Drum. Prototype #1 was dropped at about straight down onto top side of the Outer Drum.
ISORAD-TC1 Prototype #2 was subjected to two 1.2 meter (4 foot) free drops in accordance with Test Plan QTP-001 (Section 2.12.1). The orientation of the first 1.2 meter (4 foot) free drop (see Figure 2.6.7.a) was selected because of its potential to cause significant deformation of the Outer Drum and with the intent of breaking the lid bolts with high stress loads potentially causing the Inner Container to be ejected from the Outer Drum. Prototype #2 was dropped at about a 45 angle onto top side edge of the Outer Drum. The orientation of the second 1.2 meter (4 foot) free drop (see Figure 2.6.7.b) was selected because of its potential to cause significant deformation of the Outer Drum and with the intent of breaking the lid bolts with high stress loads potentially causing the Inner Container to be ejected from the Outer Drum. Prototype #1 was dropped at about straight down onto top side of the Outer Drum.
Photographs of the drop orientation are provided in Test Plan QTP-001Test Report (QTP-001)
(Section 2.12.2).
Figure 2.6.7.a: NCT Drop Test Orientation 1 and NCT Drop Test Orientation 2 There was minimal damage to the Outer Drum of Prototype #1 from the two 1.2 meter Free Drops.
Specifically, the top rim of the Outer Drum had a small indentation from the impact point of Orientation 1. No damage was indicated from the Orientation 2 impact. The damage to Prototype #2 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-39 was also minimal from the two 1.2 meter Free Drops. The damage from the Orientation 1 Free Drop was like Prototype #1. The Orientation 2 Free Drop resulted in no damage to the Outer Drum.
When Prototype #1 was disassembled to load the radioactive material into the package, no interior damage was noted to any internal structures. When Prototype #2 was disassembled to load the radioactive material into the package, the only damage noted was minimal damage to the first spacer for the Inner Container 2835A configuration. The first spacer is butt up against the Inner Container lid. The lid has a raised point that impaled the first spacer causing a 3mm diameter break in the first spacer. The damage would not impede the safety of the package nor render the package to not be fit for transport.
2.6.8 Corner Drop [IAEA SSR-6 722(b), IAEA TS-R-1 722(b), and 10 CFR 71.71(c)(8)]
This test is not applicable, as the transport package does not transport fissile material, nor is the exterior of the transport package made from either fiberboard or wood.
2.6.9 Compression [IAEA SSR-6 723, IAEA TS-R-1 723, and 10 CFR 71.71(c)(9)]
The QTP-001Test Plan (Section 2.12.2) documents that the ISORAD-TC1 package maintained its structural integrity and shielding effectiveness under the NCT compression test. As a worst case, Prototype #3 was used and was prepared as Prototype #1 and Prototype #2 with the exceptions of no Cork Assembly, no Outer Drum Lid, or any other internal support. Prototype #3 was subjected to an evenly distributed compressive load of 1097.5 kg (2419.57 pounds) based on Prototype #1s maximum weight for a period of 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />, which exceeds five times the maximum transport package weight of 136.1 kg (300 pounds). This weight is also greater than 13 kPa (2 lb/in2) multiplied by the vertically projected surface area of the transport package. Following the test, no damage to the specimen was observed.
2.6.10 Penetration [IAEA SSR-6 724, IAEA TS-R-1 724, and 10 CFR 71.71(c)(10)]
Prototype #1 was subjected to three penetration tests, in accordance with Test Plan QTP-001 Revision 1 (Section 4). Prototype #1 was impacted by a penetration bar in Orientation 1 on the bottom center of the Outer Drum with the intention of puncturing the Outer Drum. The inspection following the test indicated that the bar hit as intended on the package and dented the Outer Drum to a depth of mm ( inches). Prototype #1 was impacted by a penetration bar in Orientation 2 on the side center of the Outer Drum with the intention of puncturing the Outer Drum. The inspection following the test indicated that the bar hit as intended on the specimen, and slightly dented the Outer Drum to a depth of mm ( inches). Prototype #1 was impacted by a penetration bar in Orientation 3 on the top off-center of the Outer Drum Lid in between the lid splines with the intention of puncturing the Outer Drum. The inspection following the test indicated that the bar hit as intended on the package and did not dent the Outer Drum Lid. As described in QTP-001 Test Report (Section 4.5.5), the result was no loss of structural integrity or reduction of shielding efficiency as a result of the three impacts.
Prototype #2 was subjected to three penetration tests, in accordance with Test Plan QTP-001 (Section 4). Prototype #2 was impacted by a penetration bar in Orientation 1 on the bottom center of the Outer Drum with the intention of puncturing the Outer Drum. The inspection following the test Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-40 indicated that the bar hit as intended on the package and dented the Outer Drum to a depth of mm ( inches). Prototype #2 was impacted by a penetration bar in Orientation 2 on the side center of the Outer Drum with the intention of puncturing the Outer Drum. The inspection following the test indicated that the bar hit as intended on the specimen, and slightly dented the Outer Drum to a depth of mm ( inches). Prototype #2 was impacted by a penetration bar in Orientation 3 on the top off-center of the Outer Drum Lid in between the lid splines with the intention of puncturing the Outer Drum. The inspection following the test indicated that the bar hit as intended on the package and did not dent the Outer Drum Lid. As described in QTP-001 Test Report (Section 4.5.5), the result was no loss of structural integrity or reduction of shielding efficiency as a result of the three impacts.
Figure 3.6.10.a: NCT Penetration Test Orientation 1 and NCT Drop Test Orientation 2 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-41 Figure 4.6.10.b: NCT Penetration Test Orientation 3 and Impact Area 2.7 Hypothetical Accident Conditions of Transport [IAEA SSR-6 726, IAEA TS-R-1 726, and 10 CFR 71.73]
The same two prototypes were used to conduct the HACT tests as for the NCT tests. Each prototype consisted of a different Inner Container (Prototype #1 BPIC-Square and Prototype #2 BPIC-2835A).
However, the same tests were performed on each of the specimens. The QTP-001 Test Plan was followed during the execution of the HACT tests.
Table 2.7.A: Summary of HACT Initial Conditions Load HACT Initial Conditions (According to Regulatory Guide 7.8)
ID Ambient Internal Fabrication HACT Insolation Decay heat Temperature Pressure Stress 38°C -40°C Max Zero Max Zero Max Zero 1 9 m Free Drop X X X X X 2 (Oblique) X X X X X 3 9 m Free Drop X X X X X 4 (Top) X X X X X 5 Free Drop X X X X X 6 Puncture (Top) X X X X X 7 Free Drop X X X X X 8 Puncture (Bottom) X X X X X 9 Free Drop X X X X X 10 Puncture (Side) X X X X X Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-42 2.7.1 Free Drop I [IAEA SSR-6 727(a), IAEA TS-R-1 727(a), and 10 CFR 71.73(c)(1)
The design of the ISORAD-TC1 allows only one path for the radioactive contents to come out of the package. The drop orientations are designed to dislodge the inner container from the inside of the Outer Drum. The Inner containers are constructed of a stainless steel outer casing that is welded closed top and bottom with a bolted Shield Plug and an Inner Container Lid that is also bolted down.
The two bolted components secure the special form radioactive contents within the Inner Container.
The Inner Container would have to be ejected from the Outer Drum and both bolted enclosures would need to fail for the special form content to become unshielded. In addition, the Outer Drum is permanently enclosed on the bottom and sides with a removeable bolted down Lid Assembly. The Outer Drum is primarily for protection of the Inner Container and to provide thermal and shock protection during normal transport. Justification for all prototype drop orientations are included in Test Plan QTP-001 (Section 4).
The drop orientations apply the maximum force on the package in an attempt to break the Outer Drum Lid Assembly and possibly dislodging the Inner Container from within the Outer Drum. If the Inner Container becomes dislodged, it is subject to direct damage from further impacts and the puncture pin damage. A dislodged Inner Container alone would not significantly increase radiation levels and certainly not to 1000 mR/hr at one meter. As stated before, both the bolted down Inner Container Lid and the bolted down shield plug would have to fail for the special form capsules to become unshielded.
2.7.1.1 Oblique Drop Prototype #1 was dropped top down and at an oblique approximately 45° angle (as shown in Figure 2.7.1.2a). The first 9 meter Free Drop of Prototype #1 was conducted with the test specimens at or below -40°C (-40°F). Prototype #1 was selected because it weighed the most and would impart the most damage and was most likely to suffer brittle fracture from extreme cold. Orientation 1 was selected based on an assessment because the greatest force would be applied to one area and would potentially impart the most damage to the specimen. The orientation would impart the maximum force onto the top rim and welding points of the Attachment Ring Assembly. This is referenced as the 9 meter Free Drop Orientation 1 in the test plan and report.
The intent of the test was to apply the maximum moment to the Outer Drum Lid, Attachment Ring, enclosure bolts, and weld attachment points. In addition, to determine if (1) impact could cause buckling and/or brittle failure of the stainless steel Outer Drum structure and components, and (2) detachment of the Outer Drum Lid from the Attachment Ring Assembly due to thread failure of the tapped holes in the Attachment Ring and brittle failure of the stainless steel components, which are welded to the Outer Drum. Sketches of the drop orientations are provided in Section 4 of the QTP-001 Test Plan and Photographs of the drop orientations are in the QTP-001 Test Report (Section 5.1).
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-43 Figure 5.7.1.a: HACT Free Drop 1 Test Orientation 1 2.7.1.2 Top Down Drop Vertical Top Down: The intent of the test was to apply the maximum tensile load to the Outer Drum Lid Assembly and Inner Container Lid bolts. The justification for the test is to determine if (1) the Outer Drum Lid Assembly could separate from the Outer Drum, (2) the impact could cause buckling and/or brittle failure of the stainless steel Outer Drum structure, and (3) detachment of the Inner Container from the Outer Drum could occur due to thread failure of the tapped holes in the Outer Drum Attachment Ring Assembly or the Attachment Ring Supports. (Test Specimen XXXX).
Figure 6.7.1.b: HACT Free Drop 1 Test Orientation 2 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-44 2.7.1.5 Summary of Results
[IAEA SSR-6 726, IAEA TS-R-1 726, and 10 CFR 71.73(a) and (b)]
Two prototypes were subjected to the 9 meter (30 foot) free drop in accordance with QTP-001 Test Plan Revision 1 (Section 4). All tests were conducted at ambient temperature except the Orientation 1 for Prototype 1 at or below -40°C (-40°F). Two different orientations were used, as described in Section 4. Photographs of the drop orientations are provided in QTP-001 Test Report Section 5.1.4.
The test results are summarized in the following sections.
2.7.1.5.1 Oblique Drop Results Test Specimen Prototype #1 impacted as intended. The impact deflected the top rim of the Outer Drum one side by approximately mm ( in). One of the padlocks detached from the Lock Stud and the Lid Assembly was slightly bowed out. No damage was noted on any of the eight M14 x 2 bolts. The rest of the Outer Drum remained intact.
Test Specimen Prototype #2 impacted as intended. The impact deflected the top rim of the Outer Drum one side by approximately mm ( in). The Lid Assembly was slightly bowed out.
No damage was noted on any of the eight M14 x 2 bolts and the padlocks remained intact. The rest of the Outer Drum remained intact.
2.7.1.5.2 End Drop Results Prototype #1 impacted as intended. The impact deflected the top rim of the Outer Drum on the entire rim by approximately mm ( in). The Lid Assembly was bowed outward by approximately mm ( in). No damage was noted on any of the eight M14 x 2 bolts. The rest of the Outer Drum remained intact.
Test Specimen Prototype #2 impacted as intended. The impact deflected the top rim of the Outer Drum on the entire rim by approximately mm ( in). The Lid Assembly was bowed outward by approximately mm ( in). No damage was noted on any of the eight M14 x 2 bolts. The rest of the Outer Drum remained intact.
2.7.2 Crush [IAEA SSR-6 727(c), IAEA TS-R-1 727(c), and 10 CFR 71.73(c)(2)]
Not applicable. This package is not used for the Type B transport of normal form radioactive material.
2.7.3 Puncture [IAEA SSR-6 727(b), IAEA 727(b), and 10 CFR 71.73(c)(3)]
The puncture bar is a 152.4 mm (6.0 inch) diameter x 218.44 mm (8.6 inch) long, mild steel solid bar attached to a 1219.2 mm ( 48.0 inch) x 1219.2 mm (48.0 inch) x 50.8 mm (2.0 inch) thick mild steel plate (Drop Test Target). The bar is attached to the base with a 1/4 inch circumferential fillet weld. The puncture is rigidly attached to the Drop Test Target by four 7/16-14 x 1 long stainless steel bolts (See QTP-001 Test Report Appendix 7.2).
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-45 Justification for all prototype puncture orientations are included in Test Plan QTP-001 (Section 4).
Following the 9 meter (30 foot) free drop, the prototypes were subjected to the puncture test, in accordance with Test Plan QTP-001 (Section 4).
2.7.3.1 Top Puncture Drop Prototype #1: As noted above, in the 9 meter (30 foot) Free Drop of Prototype #1, the top rim of the Outer Drum was damaged, and the Lid Assembly was slight bowed outward. The damage from the Orientation 1 (Top Side Down) Puncture Drop was minimal causing no indentation on the Lid Assembly. The impact to the Lid Assembly pushed the bowed portion of the Lid Assembly back in to almost normal. No other damage was noted from the Puncture Drop.
Prototype #2: As noted above, in the 9 meter (30 foot) Free Drop of Prototype #1, the top rim of the Outer Drum was damaged, and the Lid Assembly was slight bowed outward. The damage from the Orientation 1 (Top Side Down) Puncture Drop was minimal causing no indentation on the Lid Assembly. The impact to the Lid Assembly pushed the bowed portion of the Lid Assembly back in to almost normal. No other damage was noted from the Puncture Drop.
2.7.3.2 Side Puncture Drop Prototype # 1: The damage from the Orientation 2 (Side) Puncture Drop was minimal causing a slight indentation approximately mm ( in) diameter and mm ( inch) deep at the area of impact. No other damage was noted from the Puncture Drop.
Prototype # 2: The damage from the Orientation 2 (Side) Puncture Drop was minimal causing a slight indentation approximately mm ( in) diameter and mm ( inch) deep at the area of impact. No other damage was noted from the Puncture Drop.
2.7.3.3 Bottom Puncture Drop Prototype # 1: The damage from the Orientation 3 (Bottom) Puncture Drop was minor causing an indentation approximately mm ( in) diameter and mm ( inch) deep at the area of impact. No other damage was noted from the Puncture Drop.
Prototype # 2: The damage from the Orientation 3 (Bottom) Puncture Drop was minor causing an indentation approximately mm ( in) diameter and mm ( inch) deep at the area of impact. No other damage was noted from the Puncture Drop.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-46 Figure 7.7.3.a: HACT Puncture Test Orientation 1, 2 & 3 2.7.4 Thermal The regulations require that the package can withstand a 30 minute fire with an average flame temperature of 800°C. A hypothetical accident thermal test was not performed on the ISORAD-TC1. The requirement was demonstrated by conducting a simulated thermal analysis using a model of the ISORAD-TC1. The detailed results of the simulation evaluation are presented in Sections 3.3 through 3.5 and Appendices 3.6.
The analysis assumed the Outer Drum was slightly deformed, and the Cork Assembly essentially undamaged, as described in Section 2.7.1. The thermal analysis demonstrates no loss of structural integrity of the package, no loss of shielding effectiveness from the heat, and not escape of radioactive contents under HACT thermal conditions. Based on the Hawk Systems Thermal Simulation Analysis and the ISO-RAD analysis, ISO-RAD concludes the ISORAD-TC1 will maintain its structural integrity and shielding effectiveness under the HACT thermal conditions.
2.7.4.1 Summary of Maximum Pressures Table 2.7.4.1.b: Summary Table of Maximum Pressures Package Fire Conditions Internal Configuration 800 (1,472) Pressure Internal Pressure Limit Developed ISORAD-TC1 BPIC 800 (1,472) 800 (1,472)
Inner Container - Weld Seal 0.3754 MPa (54.45 psi) 4.675 MPa (677.97 psi)
Contents Cavity N/A N/A Open to Air Special Form Capsule 0.3754 MPa (54.45 psi) 4.886 MPa (708.60 psi)
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-48 2.7.4.3 Stress Calculations As was shown in Section 2.7.4.2, thermal differentials will have no detrimental effect on the interfaces between the stainless steel housing of the Outer Casing and the depleted uranium shielding in the Inner Container. Stresses may develop within the cavity, but should be minimal because the Inner container is not gasketed. As shown in Section 3.4.3.2, the maximum pressure which could be generated in the Inner Container if it had a gasket is 54.3 psi. As also shown in Section 3.4.3.2, this pressure is insufficient to cause a lid failure of the closure bolts. Therefore, the containment will remain intact during the thermal test conditions.
2.7.4.4 Comparison with Allowable Stresses All stresses calculated in Section 2.7.4 are well below strengths for the materials of construction.
Further, the ISORAD-TC1 package was demonstrated to pass both the NCT and HACT conditions of transport. It is therefore concluded that the ISORAD-TC1 package will satisfy the performance requirements specified by the IAEA and USNRC regulations.
2.7.5 Immersion - Fissile Material
[IAEA SSR-6 731-733, IAEA TS-R-1 731-733, and 10 CFR 71.73(c)(5)]
Not applicable. The ISORAD-TC1 package does not transport fissile material.
2.7.6 Immersion - All Packages
[IAEA SSR-6 701 & 729, IAEA TS-R-1 701 & 729, and 10 CFR 71.73(c)(6)]
This analysis demonstrates that under the hypothetical accident condition of immersion, the package will satisfy the applicable performance requirements specified in the regulations:
- 1. there will be no loss or dispersal of contents;
- 2. there will be no structural changes that reduce the effectiveness of components required for shielding,
- 3. heat transfer, or containment;
- 4. any loss of shielding integrity would not result in more than a 20% increase in the radiation level at any external surface of the package; and
- 5. there will be no changes that would affect the ability of the package to withstand the hypothetical accident conditions tests.
Paragraph 729 of TS-R-1 (incorporated in Subsection 1(4) of the PTNS Regulations by reference to Paragraph 716 of TS-R-1), and 10 CFR 71.73(c)(6) require an undamaged package to be subjected to water pressure equivalent to immersion under a head of water of at least 15 m (50 ft). This is equivalent to an external water pressure of 150 kPa (21.7 psi).
The ISORAD-TC1 Outer Drum is externally constructed of metallic components and internally with austenitic stainless steel and cork. The Inner Container is constructed entirely of metallic components austenitic stainless steel, brass, and titanium. Prolonged exposure to water will not reduce the shielding efficiency or structural integrity of the package.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-50 l = (365 x 0.2721224) / (0.5216535448 + 0.01968504)2 - 0.2721224 l = (99.324476) / (0.5413385848)2 - 0.2721224 l = 99.324476 / 0.2930474633932667910 - 0.2721224 l = 99.324476 / 0.02092506339326679104 = 4,746.675034302 kPa or 688.447009 psi From this relationship, the minimum collapsing pressure of the source capsule is 301.55 psi, which exceeds the required external pressure increases of 21.7 psig (Normal Water Immersion) and 290 psig (Enhanced Water Immersion) for the respective regulatory references.
Resource references:
- 1. Young, Warren C. Roarks Formulas for Stress & Strain, Sixth Edition. McGraw-Hill: New York, 1989, p. 634.
- 2. Hibbeler, R.C. Mechanics of Materials. 2nd Edition, 1991.
2.7.7 Deep Water Immersion Test (for Type B Packages Containing More than 105A2)
[IAEA SSR-6 659, 660, & 730, IAEA TS-R-1 657, 658, & 730, and 10 CFR 71.61]
Not applicable. This package does not transport normal form radioactive material in quantities exceeding 105A2. The worst case Special Form Capsule would pass the 290 psig requirement with 301.55 psig.
2.7.8 Summary of Damage [IAEA SSR-6 701, 702, 716, & 726, IAEA TS-R-1 701, 702, 716, &
726, and 10 CFR 71.73 (a) & (b)]
Table 2.7.8.a summarizes the results of the Normal Conditions of Transport and Hypothetical Accident testing performed on the ISORAD-TC1, in the sequence that the tests were completed.
Table 2.7.8.a: Summary of Damages During Performance of QTP-001 Specimen Test Performed Test Results Prototype #3 Compression test Pass See QTP-001 Test Report Prototype #1 1 meter (40 inch) penetration bar Pass See QTP-001 Test Report Prototype #2 1 meter (40 inch) penetration bar Pass See QTP-001 Test Report Prototype #1 1.2 meter (4 foot) drop Pass See QTP-001 Test Report Prototype #2 1.2 meter (4 foot) drop Pass See QTP-001 Test Report Prototype #1 9 meter (30 foot) drop Pass See QTP-001 Test Report Prototype #2 9 meter (30 foot) drop Pass See QTP-001 Test Report Prototype #1 1 meter (40 inch) Puncture bar Pass See QTP-001 Test Report Prototype #2 1 meter (40 inch) Puncture bar Pass See QTP-001 Test Report Prototype #1 Post Drop Inspection Pass See QTP-001 Test Report Prototype #2 Post Drop Inspection Pass See QTP-001 Test Report The tests were performed using three test prototypes. In the course of testing, one prototype (Prototype # 3) was used for the NCT compression test, and the two prototypes (Prototype # 1 and Prototype #2) were conservatively subjected to all the NCT Tests, two each 9 meter (30 foot) drop tests, and three each puncture tests without loss of structural integrity or shielding effectiveness.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-51 Based on these results, it is concluded that the ISORAD-TC1 transport package maintains structural integrity and shielding effectiveness during Normal Conditions and Hypothetical Accident Conditions of Transport.
2.8 Accident Conditions for Air Transport of Plutonium [IAEA SSR-6 683, IAEA TS-R-1 680, and 10 CFR 71.55]
Not applicable. The ISORAD-TC1 package is not used for transportation of plutonium.
2.9 Accident Conditions for Fissile Material Packages for Air Transport [IAEA SSR-6 683, IAEA TS-R-1 680, and 10 CFR 71.55]
Not Applicable. The ISORAD-TC1 package is not used for transportation of Type B quantities of fissile material.
2.10 Special Form [IAEA SSR-6 602-604, IAEA TS-R-1 602-604, and 10 CFR 71.75]
The ISORAD-TC1 transport package is designed for use with only special form source capsules with a maximum of 13.75 mm (0.5413 in) outside diameter and 0.25 mm (0.009843) minimum wall thickness. The materials of construction are austenitic stainless steel, 400 series stainless steel, titanium, or vanadium. The source capsule must be qualified as Special Form radioactive material.
Representative Certificates for each special form, or an equivalent with justification, are included in Appendix 2.12.4.
2.11 Fuel Rods Not applicable. The ISORAD-TC1 package is not designed or used for transportation of fuel rods.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-52 2.12 Appendix 2.12.1 QTP-001 Test Plan, Revision 1.
Revision 0 29 February 2020
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 1 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN REVISION AUTHOR CHECKED APPROVED DATE 0 Kevin J. Schehr Rafael Bustillo Kevin J. Schehr 09/26/2019 1 Kevin J. Schehr Rafael Bustillo 12/18/2019 REVISION PAGE NOS. DETAILS DATE 0 ALL Original Document 09/26/2019 1 ALL Added updated information including 12/18/2019 notations NOTE: THOSE ATTACHMENTS INDICATED IN THE TABLE OF CONTENTS AS FORMS MAY NOT BE INCLUDED IN PROCEDURE BUT ARE, RATHER, INCLUDED BY REFERENCE REVISIONS TO THE TEXT OF THIS DOCUMENT ARE INDICATED IN THE MARGIN AS BELOW:
N WHERE N INDICATES THE LATEST REVISION NUMBER When Printed, this document becomes UNCONTROLLED
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 2 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Table of Contents Section 1.0 Introduction .......................................................................................................................................................................................................... 4 Section 2.0 Package Description .......................................................................................................................................................................................... 5 2.1 Outer Container............................................................................................................................................................................................................... 5 2.2 Bulk and PIC Inner Container ...................................................................................................................................................................................... 5 2.3 Multi Port Inner Container ............................................................................................................................................................................................ 5 2.4 2835 Bulk and PIC Inner Container ............................................................................................................................................................................ 6 Section 3.0 Regulatory Requirements ................................................................................................................................................................................. 7 Table 3.1 General Requirements for All Packages ....................................................................................................................................................... 7 Table 3.2 Requirements for Type A Packages............................................................................................................................................................. 10 Table 3.3 Requirements for Type B(U) Packages ....................................................................................................................................................... 13 Section 4.0 Package Orientations....................................................................................................................................................................................... 17 Section 5.0 Evaluation and Assessment of Package Conformance.......................................................................................................................... 28 Table 5.1 Compliance, Evaluation, and Assessment Regulations......................................................................................................................... 28 Section 6.0 Construction and Condition of Test Specimens ...................................................................................................................................... 30 Table 6.1 Preparation of a Specimen for Testing ....................................................................................................................................................... 30 Section 7.0 Test Equipment and Material List ................................................................................................................................................................. 31 Table 7.1 Test Equipment Normal Conditions Tests ................................................................................................................................................. 31 Table 7.2 Test Equipment Hypothetical Accident Conditions Tests ..................................................................................................................... 32 Section 8.0 Test Procedures ................................................................................................................................................................................................ 33 Table 8.1 Type A Packaging /Normal Conditions Tests............................................................................................................................................ 33 Table 8.2 Hypothetical Accident Conditions Tests .................................................................................................................................................... 37 Table 8.3 Normal Conditions Test Procedures ........................................................................................................................................................... 41 Table 8.4 Hypothetical Accident Conditions Test Procedures ............................................................................................................................... 44
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 3 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 9.0 ISORAD-TC1 Test Schedule ........................................................................................................................................................................... 47 Table 9.1 Daily Test Schedule .......................................................................................................................................................................................... 47
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 4 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 1.0 Introduction The test plan was developed by ISO-RAD Canada Ltd (ISO-RAD Canada). ISO-RAD Canada is the designer and owner of the design of the ISORAD-TC1 Type B(U) Multipurpose Transport Package (ISORAD-TC1).
The ISORAD-TC1 has been designed to withstand the Type A and Type B Package specifications and testing requirements of International Atomic Energy Agency (IAEA), the Canadian Nuclear Safety Commission (CNSC), the United States Nuclear Regulatory Commission (USNRC), the United States Department of Transportation (USDOT), United Nations (UN) Economic Commission for Europe Committee for Inland Transport, International Air Transport Association (IATA), and other regulatory bodies governing transport packages. The regulations and guides used in the design development of the ISORAD-TC1 are IAEA SSR-6:2018, IAEA TSR-1, SOR/2015-145 CNSC Packaging and Transport of Nuclear Substances Regulations (PTNS Regulations), RD-364/NUREG-1886 The Joint Canada-United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages (NUREG-1886), the 2019 European Agreement concerning the International Carriage of Dangerous Goods by Road (ADR), IATA Dangerous Goods Regulations.
The ISO-RAD test plan will strive to collect the required data according to the CNSC RD-364 and USNRC NUREG-1886 guide. The various regulations in most cases overlap and the tables developed within this test plan will ensure a requirement is not overlooked. The test plan will demonstrate a difference and notations will be made describing any additional test or specification that must be conducted to demonstrate compliance.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 5 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 2.0 Package Description 2.1 Outer Container The Outer Container is a modified 15.5 Gallon Beer Keg constructed from AISI 304 Stainless Steel. The approximate height is 589 mm [23.1875 in] and has an outside diameter of 403.24 mm [15.88 in]. The outer container modification consists of adding an Attachment Ring and Attachment Ring Supports constructed of AISI 304 Stainless Steel. The Attachment Ring and Attachment Ring Supports are Tungsten Inert Gas (TIG) or Laser welded onto the Outer Container using American Welding Society (AWS) standards. The Outer Container Lid is constructed of AISI 304 Stainless Steel and is removable to allow insertion and extraction of the inner container assembly. The Outer Container Lid is secured in place by eight M14 x 2 300 Series Stainless Steel Stud Bolts and M14 x 2 300 Series Stainless Steel Hex Nuts.
The interior of the Outer Container is first lined with approximately 104 mm [4.1 in] of cork to provide thermal insulation and impact damage limiting for the inner container. The Inner Container Cavity is optionally lined with an austenitic stainless steel sleeve with an approximate wall thickness of 0.8 mm
[0.03125 in]. The stainless steel sleeve provides protection for the Cork. The Cork Lid is constructed of Cork securely mounted onto an aluminum or stainless steel Plate with four Number 14 x 1 300 Series Stainless Steel Screws.
2.2 Bulk and PIC Inner Container The Bulk and PIC Inner Container (BPIC) (Square or Round version) is designed to carry special form capsules in multiple configurations to maximize the capacity of the Isotope Cavity within the BPIC. The Bulk configuration insert is designed to hold Bulk Special Form capsules to hold various radioactive isotopes in bulk quantities. The PIC configuration insert is designed to hold Primary Inner Capsules (PIC)s which are individual capsules used by manufacturers in the construction of outer sources assembly holders for various radioactive isotopes.
The BPIC is constructed from a Depleted Uranium (DU) shield surrounded by a Top Shield Support, a Bottom Shield Support, and Brass Sleeve constructed from C360, C260 or C240 Brass. The Brass is encased in AISI 304 Stainless Steel Bottom Plate, Top Plate and Outer Casing. The Brass pieces provide a eutectic barrier. The inner cavity of the DU Shield is lined with a Grade 2 Titanium Cavity Sleeve. The BPIC Lid is secured in place by eight M8 x 1.25 300 Series Stainless Steel Socket Head Cap Screws and the BPIC Plug is secured in place by four M6 x 1 300 Series Stainless Steel Socket Head Cap Screws.
2.3 Multi Port Inner Container The Multi Port Inner Container (MPIC) (Square or Round version) is designed to carry ten (10) special form Industrial Radiography sources. The MPIC is constructed from a two-part Depleted Uranium (DU) shield surrounded by a Top Shield Support, a Bottom Shield Support, and Brass Sleeve constructed from C360, C260 or C240 Brass. The Brass is encased in AISI 304 Stainless Steel Bottom Plate, Top Plate with locking mechanism, Outer Casing, Lid Support, and Lid Assembly. The Brass pieces provide a eutectic barrier. The ten (10) Source Tubes are made of Grade 2 Titanium. The MPIC Lid is secured by one M16 300 Series Stainless Steel Low Profile Hex Nut.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 6 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN 2.4 2835 Bulk and PIC Inner Container The 2835A Bulk and PIC Inner Container (2835A) is designed to carry special form capsules in multiple configurations to maximize the capacity of the Isotope Cavity within the BPIC. The Bulk configuration insert is designed to hold Bulk Special Form capsules to hold various radioactive isotopes in bulk quantities. The PIC configuration insert is designed to hold Primary Inner Capsules (PIC)s which are individual capsules used by manufacturers in the construction of outer sources assembly holders for various radioactive isotopes.
The 2835A is constructed from a Depleted Uranium (DU) shield surrounded by a Top Shield Shim, a Bottom Shield Shim, Cavity Shims, DU Plug Shims, and Brass Sleeve constructed from C360, C260 or C240 Brass. The Brass is encased in AISI 304 Stainless Steel Bottom Plate, Top Plate, DU Shield Cavity Sleeve, Lid, and Outer Casing. The Brass pieces provide a eutectic barrier. The 2835A Lid is secured by six M8 x 125 300 Series Stainless Steel Socket Head Cap Screws.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 7 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 3.0 Regulatory Requirements The purpose of the ISORAD-TC1 test plan is to ensure compliance with the regulations and ISO-RADs Quality Assurance Program /Management System (QAPMS). The test plan is required by the QAPMS and the derivative work instructions governing design control, document control, and regulatory compliance in the QAPMS and the Quality Management Manual (QMM).
The following Sections 3.1, 3.2, and 3.3 detail the regulatory requirements for General Package Requirements, Type A Package Requirements, and Type B(U) Package Requirements according to the listed regulations. The CNSC PTNS Regulations follow the IAEA regulations with the USNRC, USDOT, and IATA incorporating slight changes to the IAEA regulations.
Table 3.1 General Requirements for All Packages Regulation General Requirements Notations IATA 10.6.0.1 Package and packaging performance specifications, in terms of retention of integrity of containment and shielding, depend upon the quantity and nature of the radioactive material transported.
Performance specifications applied are graded to take into account conditions of transport characterized by the following severity levels:
(a) Conditions likely to be encountered in routine transport (incident free);
(b) Normal conditions of transport (minor mishaps); and (c) Accident conditions of transport.
The performance specifications include design requirements and tests.
IAEA SSR-6 607 The package shall be so designed in relation to its mass, volume and shape that it can be easily and IAEA TSR-1 606 safely transported. In addition, the package shall be so designed that it can be properly secured in or ADR 6.4.2.1 on the conveyance during transport.
IATA 10.6.0.2 The Package must be so designed in relation to its weight, volume, and shape that it can be easily and safely handled and transported. In addition, the package must be so designed that it can be properly secured in the aircraft during transport.
IAEA SSR-6 608 The design shall be such that any lifting attachments on the package will not fail when used in the IAEA TSR-1 607 Intended manner and that if failure of the attachments should occur, the ability of the package to ADR 6.4.2.2 meet other requirements of these Regulations would not be impaired. The design shall take account IATA 10.6.0.3 of appropriate safety factors to cover snatch lifting.
10 CFR 71.45(a) Any lifting attachment that is a structural part of a package must be designed with a minimum safety factor of three against yielding when used to lift the package in the intended manner, and it must be designed so that failure of any lifting device under excessive load would not impair the ability of the package to meet other requirements of this subpart. Any other structural part of the package that could be used to lift the package must be capable of being rendered inoperable for lifting the
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 8 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation General Requirements Notations Continued package during transport, or must be designed with strength equivalent to that required for lifting 10 CFR 71.45(a) attachments.
IAEA SSR-6 609 Attachments and any other features on the outer surface of the package that could be used to lift it IAEA TSR-1 608 shall be designed either to support its mass in accordance with the requirements of para. 608 or shall ADR 6.4.2.3 be removable or otherwise rendered incapable of being used during transport.
IATA 10.6.0.4 IAEA SSR-6 610 As far as practicable, the packaging shall be so designed that the external surfaces are free from IAEA TSR-1 609 protruding features and can be easily decontaminated.
ADR 6.4.2.4 IATA 10.6.0.5 IAEA SSR-6 611 As far as practicable, the outer layer of the package shall be so designed as to prevent the collection IAEA TSR-1 610 and the retention of water.
ADR 6.4.2.5 IATA 10.6.0.6 IAEA SSR-6 612 Any features added to the package at the time of transport that are not part of the package shall not No parts are designed to be IAEA TSR-1 611 reduce its safety. added to the package.
ADR 6.4.2.6 IATA 10.6.0.7 IAEA SSR-6 613 The package shall be capable of withstanding the effects of any acceleration, vibration or vibration IAEA TSR-1 612 resonance that may arise under routine conditions of transport without any deterioration in the ADR 6.4.2.7 effectiveness of the closing devices on the various receptacles or in the integrity of the package as a IATA 10.6.0.8 whole. In particular, nuts, bolts and other securing devices shall be so designed as to prevent them from becoming loose or being released unintentionally, even after repeated use.
IAEA SSR-6 613A 613A. The design of the package shall take into account ageing mechanisms.
IAEA SSR-6 614 The materials of the packaging and any components or structures shall be physically and chemically The design has brass or IAEA TSR-1 613 compatible with each other and with the radioactive contents. Account shall be taken of their titanium providing a eutectic ADR 6.4.2.8 behaviour under irradiation. barrier between all stainless IATA 10.6.0.9 steel and depleted uranium.
IAEA SSR-6 615 All valves through which the radioactive contents could escape shall be protected against unauthorized No valves are included in the IAEA TSR-1 614 operation. ISORAD-TC1 design.
ADR 6.4.2.9 IATA 10.6.0.10 IAEA SSR-6 616 The design of the package shall take into account ambient temperatures and pressures that are likely IAEA TSR-1 615 to be encountered in routine conditions of transport.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 9 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation General Requirements Notations Continued ADR 6.4.2.10 IATA 10.6.0.10 IAEA SSR-6 617 A package shall be so designed that it provides sufficient shielding to ensure that, under routine The design ensures a IAEA TSR-1 616 conditions of transport and with the maximum radioactive contents that the package is designed to significant safety margin in ADR 6.4.2.11 contain, the dose rate at any point on the external surface of the package would not exceed the values respect to the dose rate.
IATA 10.6.0.11 specified in paras 516, 527 and 528, as applicable, with account taken of paras 566(b) and 573.
IAEA SSR-6 618 For radioactive material having other dangerous properties, the package design shall take into account The radioactive isotopes for IAEA TSR-1 617 those properties (see paras 110 and 507). this package do not have ADR 6.4.2.12 other dangerous properties.
IATA 10.6.0.12 IAEA SSR-6 619 For packages to be transported by air, the temperature of the accessible surfaces shall not exceed IAEA TSR-1 618 50°C at an ambient temperature of 38°C with no account taken for insolation.
IATA 10.6.1.1 IAEA SSR-6 620 Packages to be transported by air shall be so designed that if they were exposed to ambient The primary containment is IAEA TSR-1 619 temperatures ranging from 40°C to +55°C, the integrity of containment would not be impaired. the special form capsules and IATA 10.6.1.2 exceed this requirement.
IAEA SSR-6 621 Packages containing radioactive material to be transported by air shall be capable of withstanding, The primary containment is IAEA TSR-1 620 without loss or dispersal of radioactive contents from the containment system, an internal pressure the special form capsules and IATA 10.6.1.3 that produces a pressure differential of not less than maximum normal operating pressure plus 95 exceed this requirement.
kPa.
10 CFR 71.43(a) The smallest overall dimension of a package may not be less than 10 cm (4 in).
10 CFR 71.43(b) The outside of a package must incorporate a feature, such as a seal, that is not readily breakable and A tamper seal is placed that, while intact, would be evidence that the package has not been opened by unauthorized between the lock studs.
persons.
10 CFR 71.43(c) Each package must include a containment system securely closed by a positive fastening device that Special Form Sources only cannot be opened unintentionally or by a pressure that may arise within the package.
10 CFR 71.43(d) A package must be made of materials and construction that assure that there will be no significant chemical, galvanic, or other reaction among the packaging components, among package contents, or between the packaging components and the package contents, including possible reaction resulting from in leakage of water, to the maximum credible extent. Account must be taken of the behavior of materials under irradiation.
10 CFR 71.43(e) A package valve or other device, the failure of which would allow radioactive contents to escape, The package design does not must be protected against unauthorized operation and, except for a pressure relief device, must be include these features.
provided with an enclosure to retain any leakage.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 10 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation General Requirements Notations 10 CFR 71.43(f) A package must be designed, constructed, and prepared for shipment so that under the tests specified in § 71.71 ("Normal conditions of transport") there would be no loss or dispersal of radioactive contents, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging.
10 CFR 71.43(g) A package must be designed, constructed, and prepared for transport so that in still air at 38°C (100°F) and in the shade, no accessible surface of a package would have a temperature exceeding 50°C (122°F) in a nonexclusive use shipment, or 85°C (185°F) in an exclusive use shipment.
10 CFR 71.43(h) A package may not incorporate a feature intended to allow continuous venting during transport.
Table 3.2 Requirements for Type A Packages Regulation Requirements for Type A Packages Notations IAEA SSR-6 635 Type A packages shall be designed to meet the requirements specified in paras 607-618 and, in IAEA TSR-1 633 addition, the requirements of paras 619-621 if carried by air, and of paras 636-651.
ADR 6.4.7.1 IAEA SSR-6 636 The smallest overall external dimension of the package shall not be less than 10 cm.
IAEA TSR-1 634 ADR 6.4.7.2 IATA 10.6.2.4.1.1 10 CFR 71.43(a) The smallest overall dimension of a package may not be less than 10 cm (4 in).
IAEA SSR-6 637 The outside of the package shall incorporate a feature such as a seal that is not readily breakable IAEA TSR-1 635 and which, while intact, will be evidence that the package has not been opened.
ADR 6.4.7.3 IATA 10.6.2.4.1.2 IAEA SSR-6 638 Any tie-down attachments on the package shall be so designed that, under normal and accident IAEA TSR-1 636 conditions of transport, the forces in those attachments shall not impair the ability of the package ADR 6.4.7.4 to meet the requirements of these Regulations.
IATA 10.6.2.4.1.3 10 CFR 71.45(b) (b) Tie-down devices:
(1) If there is a system of tie-down devices that is a structural part of the package, the system must be capable of withstanding, without generating stress in any material of the package in excess of its yield strength, a static force applied to the center of gravity of the package having a vertical component of 2 times the weight of the package with its contents, a horizontal component along the direction in which the vehicle travels of 10 times the weight of the package with its contents,
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(2) Any other structural part of the package that could be used to tie down the package must be capable of being rendered inoperable for tying down the package during transport, or must be designed with strength equivalent to that required for tie-down devices.
(3) Each tie-down device that is a structural part of a package must be designed so that failure of the device under excessive load would not impair the ability of the package to meet other requirements of this part.
IAEA SSR-6 639 The design of the package shall take into account temperatures ranging from 40°C to +70°C for IAEA TSR-1 637 the components of the packaging. Attention shall be given to freezing temperatures for liquids and ADR 6.4.7.5 to the potential degradation of packaging materials within the given temperature range.
IATA 10.6.2.4.1.4 IAEA SSR-6 640 The design and manufacturing techniques shall be in accordance with national or international IAEA TSR-1 638 standards, or other requirements, acceptable to the competent authority.
ADR 6.4.7.6 IATA 10.6.2.4.1.5 IAEA SSR-6 641 The design shall include a containment system securely closed by a positive fastening device that cannot The special form capsule(s)
IAEA TSR-1 639 be opened unintentionally or by a pressure that may arise within the package. are an integral part of the ADR 6.4.7.7 containment system.
IATA 10.6.2.4.2.1 10 CFR 71.43(c)
IAEA SSR-6 642 Special form radioactive material may be considered as a component of the containment system. The special form capsule(s)
IAEA TSR-1 640 are an integral part of the ADR 6.4.7.8 containment system.
IATA 10.6.2.4.2.2 IAEA SSR-6 643 If the containment system forms a separate unit of the package, the containment system shall be IAEA TSR-1 641 capable of being securely closed by a positive fastening device that is independent of any other ADR 6.4.7.9 part of the packaging.
IATA 10.6.2.4.2.3 IAEA SSR-6 644 The design of any component of the containment system shall take into account, where applicable, IAEA TSR-1 642 the radiolytic decomposition of liquids and other vulnerable materials and the generation of gas by ADR 6.4.7.10 chemical reaction and radiolysis.
IATA 10.6.2.4.2.4
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 12 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Requirements for Type A Packages Notations IAEA SSR-6 645 The containment system shall retain its radioactive contents under a reduction of ambient pressure IAEA TSR-1 643 to 60 kPa.
ADR 6.4.7.11 IATA 10.6.2.4.2.5 IAEA SSR-6 646 All valves, other than pressure relief valves, shall be provided with an enclosure to retain any leakage IAEA TSR-1 644 from the valve.
ADR 6.4.7.12 IATA 10.6.2.4.2.6 IAEA SSR-6 647 A radiation shield that encloses a component of the package specified as a part of the containment IAEA TSR-1 645 system shall be so designed as to prevent the unintentional release of that component from the ADR 6.4.7.13 shield. Where the radiation shield and such component within it form a separate unit, the radiation IATA 10.6.2.4.2.7 shield shall be capable of being securely closed by a positive fastening device that is independent of any other packaging structure.
IAEA SSR-6 648 A package shall be so designed that if it were subjected to the tests specified in paras 719-724, it IAEA TSR-1 646 would prevent:
ADR 6.4.7.14 (a) Loss or dispersal of the radioactive contents; IATA 10.6.2.4.1.6 (b) More than a 20% increase in the maximum dose rate at any external surface of the package.
IAEA SSR-6 649 The design of a package intended for liquid radioactive material shall make provision for ullage to The ISORAD-TC1 is not IAEA TSR-1 647 accommodate variations in the temperature of the contents, dynamic effects and filling dynamics. designed to transport liquids ADR 6.4.7.15 not in special form.
IATA 10.6.2.4.3.3 IAEA SSR-6 650 A Type A package designed to contain liquid radioactive material shall, in addition: The ISORAD-TC1 is not IAEA TSR-1 648 (a) Be adequate to meet the conditions specified in para. 648(a) if the package is subjected to the designed to transport liquids ADR 6.4.7.16 tests specified in para. 725; and not in special form.
IATA 10.6.2.4.3.1 (b) Either:
IATA 10.6.2.4.3.2 (i) Be provided with sufficient absorbent material to absorb twice the volume of the liquid contents. Such absorbent material must be suitably positioned so as to contact the liquid in the event of leakage; or (ii) Be provided with a containment system composed of primary inner and secondary outer containment components designed to enclose the liquid contents completely and to ensure their retention within the secondary outer containment components, even if the primary inner components leak.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 13 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Requirements for Type A Packages Notations IAEA SSR-6 651 Type A package designed for gases shall prevent loss or dispersal of the radioactive contents if the The ISORAD-TC1 is not IAEA TSR-1 649 package were subjected to the tests specified in para. 725, except for a Type A package designed for designed to transport gases.
ADR 6.4.7.17 tritium gas or for noble gases.
Continued IATA 10.6.2.4.4.1 Table 3.3 Requirements for Type B(U) Packages Regulation Requirements for Type B(U) Packages Notations IAEA SSR-6 652 Type B(U) packages shall be designed to meet the requirements specified in paras 607-618, the IAEA TSR-1 650 requirements specified in paras 619-621 if carried by air, and in paras 636-649, except as specified in ADR 6.4.8.1 para. 648(a), and, in addition, the requirements specified in paras 653-666.
IATA 10.6.2.5.2 IAEA SSR-6 653 package shall be so designed that, under the ambient conditions specified in paras 656 and 657, heat IAEA TSR-1 651 generated within the package by the radioactive contents shall not, under normal conditions of ADR 6.4.8.2 transport, as demonstrated by the tests in paras 719-724, adversely affect the package in such a way IATA 10.6.2.5.3 that it would fail to meet the applicable requirements for containment and shielding if left unattended for a period of one week. Particular attention shall be paid to the effects of heat that may cause one or more of the following:
(a) Alteration of the arrangement, the geometrical form or the physical state of the radioactive contents or, if the radioactive material is enclosed in a can or receptacle (for example, clad fuel elements), cause the can, receptacle or radioactive material to deform or melt; (b) Lessening of the efficiency of the packaging through differential thermal expansion, or cracking or melting of the radiation shielding material; (c) Acceleration of corrosion when combined with moisture.
IAEA SSR-6 654 package shall be so designed that, under the ambient condition specified in para. 656 and in the IAEA TSR-1 652 absence of insolation, the temperature of the accessible surfaces of a package shall not exceed ADR 6.4.8.3 50°C, unless the package is transported under exclusive use.
IATA 10.6.2.5.5 10 CFR 71.43(h) A package must be designed, constructed, and prepared for transport so that in still air at 38°C (100°F) and in the shade, no accessible surface of a package would have a temperature exceeding 50°C (122°F) in a nonexclusive use shipment, or 85°C (185°F) in an exclusive use shipment.
IAEA SSR-6 655 Except as required in para. 619 for a package transported by air, the maximum temperature of any IAEA TSR-1 653 surface readily accessible during transport of a package under exclusive use shall not exceed 85°C in ADR 6.4.8.4 the absence of insolation under the ambient condition specified in para. 656. Account may be taken
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IAEA SSR-6 656 The ambient temperature shall be assumed to be 38°C.
IAEA TSR-1 654 ADR 6.4.8.5 IATA 10.6.2.5.4 IAEA SSR-6 657 The solar insolation conditions shall be assumed to be as specified in Table 12.
IAEA TSR-1 655 TABLE 12. INSOLATION DATA ADR 6.4.8.6 Case Form and location of surface Insolation IATA 10.6.2.5.4 for 12 h per day (W/m2) 1 Flat surfaces transported horizontally downward facing 0 2 Flat surfaces transported horizontally upward facing 800 3 Surfaces transported vertically 200a 4 Other downward facing (not horizontal) surfaces 200a 5 All other surfaces 400a a a Alternatively, a sine function may be used, with an absorption coefficient adopted and the effects of possible reflection from neighbouring objects neglected.
IAEA SSR-6 658 A package that includes thermal protection for the purpose of satisfying the requirements of the IAEA TSR-1 656 thermal test specified in para. 728 shall be so designed that such protection will remain effective if ADR 6.4.8.7 the package is subjected to the tests specified in paras 719-724 and 727(a) and 727(b) or 727(b) and IATA 10.6.2.5.7 727(c), as appropriate. Any such protection on the exterior of the package shall not be rendered ineffective by ripping, cutting, skidding, abrading or rough handling.
IAEA SSR-6 659 A package shall be so designed that if it were subjected to:
IAEA TSR-1 657 (a) The tests specified in paras 719-724, it would restrict the loss of radioactive contents to not more ADR 6.4.8.8 than 106A2 per hour.
IATA 10.6.2.5.8 (b) The tests specified in paras 726, 727(b), 728 and 729 and either the test in:
Para. 727(c), when the package has a mass not greater than 500 kg, an overall density not greater than 1000 kg/m3 based on the external dimensions, and radioactive contents greater than 1000A2 not as special form radioactive material; or Para. 727(a), for all other packages.
(i) It would retain sufficient shielding to ensure that the dose rate 1 m from the surface of the package would not exceed 10 mSv/h with the maximum radioactive contents that the package is designed to contain.
(ii) It would restrict the accumulated loss of radioactive contents in a period of one week to not more than 10A2 for krypton-85 and not more than A2 for all other radionuclides.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 15 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Requirements for Type B(U) Packages Notations Continued Where mixtures of different radionuclides are present, the provisions of paras 405-407 shall apply, IAEA SSR-6 659 except that for krypton-85 an effective A2(i) value equal to 10A2 may be used. For case (a), the IAEA TSR-1 657 assessment shall take into account the external non-fixed contamination limits of para. 508.
ADR 6.4.8.8 IATA 10.6.2.5.8 10 CFR 71.51(a) A Type B package, in addition to satisfying the requirements of §§ 71.41 through 71.47, must be designed, constructed, and prepared for shipment so that under the tests specified in:
(1) Section 71.71 ("Normal conditions of transport"), there would be no loss or dispersal of radioactive contents--as demonstrated to a sensitivity of 10-6 A2 per hour, no significant increase in external surface radiation levels, and no substantial reduction in the effectiveness of the packaging; and (2) Section 71.73 ("Hypothetical accident conditions"), there would be no escape of krypton-85 exceeding 10 A2 in 1 week, no escape of other radioactive material exceeding a total amount A2 in 1 week, and no external radiation dose rate exceeding 10 mSv/h (1 rem/h) at 1 m (40 in) from the external surface of the package.
IAEA SSR-6 660 A package for radioactive contents with activity greater than 105A2 shall be so designed that if it IAEA TSR-1 658 were subjected to the enhanced water immersion test specified in para. 730, there would be no ADR 6.4.8.9 rupture of the containment system.
IATA 10.6.2.5.10 10 CFR 71.51(d) For packages which contain radioactive contents with activity greater than 105 A2, the requirements of § 71.61 must be met.
IAEA SSR-6 661 Compliance with the permitted activity release limits shall depend neither upon filters nor upon a IAEA TSR-1 659 mechanical cooling system.
ADR 6.4.8.10 IATA 10.6.2.5.11 10 CFR 71.51(c) Compliance with the permitted activity release limits of paragraph (a) of this section may not depend on filters or on a mechanical cooling system.
IAEA SSR-6 662 A package shall not include a pressure relief system from the containment system that would allow IAEA TSR-1 660 the release of radioactive material to the environment under the conditions of the tests specified in ADR 6.4.2.11 paras 719-724 and 726-729.
IATA 10.6.2.5.12 IAEA SSR-6 663 A package shall be so designed that if it were at the maximum normal operating pressure and it were IAEA TSR-1 661 subjected to the tests specified in paras 719-724 and 726-729, the levels of strains in the containment ADR 6.4.8.12 system would not attain values that would adversely affect the package in such a way that it would IATA 10.6.2.5.13 fail to meet the applicable requirements.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 16 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Requirements for Type B(U) Packages Notations IAEA SSR-6 664 A package shall not have a maximum normal operating pressure in excess of a gauge pressure of 700 IAEA TSR-1 662 kPa.
ADR 6.4.8.13 IATA 10.6.2.5.14 IAEA SSR-6 665 A package containing low dispersible radioactive material shall be so designed that any features IAEA TSR-1 663 added to the low dispersible radioactive material that are not part of it, or any internal components ADR 6.4.8.14 of the packaging, shall not adversely affect the performance of the low dispersible radioactive IATA 10.6.2.5.15 material.
IAEA SSR-6 666 A package shall be designed for an ambient temperature range of 40°C to +38°C.
IAEA TSR-1 664 ADR 6.4.8.15 IATA 10.6.2.5.16
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Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 28 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 5.0 Evaluation and Assessment of Package Conformance Table 5.1 Compliance, Evaluation, and Assessment Regulations Regulation Parameters Notations IAEA SSR-6 701 Demonstration of compliance with the performance standards required in Section VI shall be IAEA TSR-1 701 accomplished by any of the following methods listed below or by a combination thereof:
ADR 6.4.12.1 (a) Performance of tests with specimens representing special form radioactive material, or low IATA 10.6.3.1.1 dispersible radioactive material, or with prototypes or samples of the packaging, where the contents of the specimen or the packaging for the tests shall simulate as closely as practicable the expected range of radioactive contents and the specimen or packaging to be tested shall be prepared as presented for transport.
(b) Reference to previous satisfactory demonstrations of a sufficiently similar nature.
(c) Performance of tests with models of appropriate scale, incorporating those features that are significant with respect to the item under investigation when engineering experience has shown the results of such tests to be suitable for design purposes. When a scale model is used, the need for adjusting certain test parameters, such as penetrator diameter or compressive load, shall be taken into account.
(d) Calculation, or reasoned argument, when the calculation procedures and parameters are generally agreed to be reliable or conservative.
10 CFR 71.41 Demonstration of compliance (a) The effects on a package of the tests specified in § 71.71 ("Normal conditions of transport"), and the tests specified in § 71.73 ("Hypothetical accident conditions"), and § 71.61 ("Special requirements for Type B packages containing more than 105 A2"), must be evaluated by subjecting a specimen or scale model to a specific test, or by another method of demonstration acceptable to the Commission, as appropriate for the particular feature being considered.
(b) Taking into account the type of vehicle, the method of securing or attaching the package, and the controls to be exercised by the shipper, the Commission may permit the shipment to be evaluated together with the transporting vehicle.
(c) Environmental and test conditions different from those specified in §§ 71.71 and 71.73 may be approved by the Commission if the controls proposed to be exercised by the shipper are demonstrated to be adequate to provide equivalent safety of the shipment.
(d) Packages for which compliance with the other provisions of these regulations is impracticable shall not be transported except under special package authorization. Provided the applicant demonstrates that compliance with the other provisions of the regulations is impracticable and that the requisite standards of safety established by these regulations have been demonstrated through
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 29 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN means alternative to the other provisions, a special package authorization may be approved for one-time Regulation Parameters Notations Continued shipments. The applicant shall demonstrate that the overall level of safety in transport for these 10 CFR 71.41 shipments is at least equivalent to that which would be provided if all the applicable requirements had been met.
IAEA SSR-6 702 After the specimen, prototype or sample has been subjected to the tests, appropriate methods of IAEA TSR-1 702 assessment shall be used to ensure that the requirements of this section have been fulfilled in ADR 6.4.12.2 compliance with the performance and acceptance standards prescribed in Section VI.
IATA 10.6.3.1.2 IAEA SSR-6 716 Testing the integrity of the containment system and shielding and assessing criticality safety IAEA TSR-1 716 After each test or group of tests or sequence of the applicable tests, as appropriate, specified in paras ADR 6.4.13 718-737:
IATA 10.6.3.2 (a) Faults and damage shall be identified and recorded.
(b) It shall be determined whether the integrity of the containment system and shielding has been retained to the extent required in Section VI for the package under test.
(c) For packages containing fissile material, it shall be determined whether the assumptions and conditions used in the assessments required by paras 673-686 for one or more packages are valid.
10 CFR 71.71(a) Evaluation. Evaluation of each package design under normal conditions of transport must include a determination of the effect on that design of the conditions and tests specified in this section.
Separate specimens may be used for the free drop test, the compression test, and the penetration test, if each specimen is subjected to the water spray test before being subjected to any of the other tests.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 30 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 6.0 Construction and Condition of Test Specimens Table 6.1 Preparation of a Specimen for Testing Regulation Parameters Notations IAEA SSR-6 713 All specimens shall be inspected before testing in order to identify and record faults or damage, including the IAEA TSR-1 713 following:
ADR 6.4.12.3 (a) Divergence from the design; IATA 10.6.3.1.3 (b) Defects in manufacture; (c) Corrosion or other deterioration; (d) Distortion of features.
IAEA SSR-6 714 The containment system of the package shall be clearly specified.
IAEA TSR-1 714 ADR 6.4.12.3 IATA 10.6.3.1.4 IAEA SSR-6 715 The external features of the specimen shall be clearly identified so that reference may be made simply and IAEA TSR-1 715 clearly to any part of such a specimen.
ADR 6.4.12.3 IATA 10.6.3.1.4 Verify the manufacturing documents and records of the dimensions prior to test commencement. Yes _X__ No____
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 31 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 7.0 Test Equipment and Material List Table 7.1 Test Equipment Normal Conditions Tests Regulation/Other Test Equipment Normal Condition Tests Notations Safety General Safety Equipment
- 1) Safety Glasses
- 2) Gloves
- 3) Survey Meter
- 4) Rate Alarm
- 5) Personal Dosimetry Video/ 1) Video Recording Recording 2) Camera Equipment 3) Video / Camera Stand Stacking Test 1) 2,200 pounds of weight
- 2) Scale
- 3) Timing device Penetration 1) Calibrated penetration bar
- 2) Tube to guide penetration bar
- 3) Tape measure IAEA SSR-6 717 The target for the drop test specified in paras 705, 722, 725(a), 727 and 735 shall be a flat, IAEA TSR-1 horizontal surface of such a character that any increase in its resistance to displacement or ADR 6.4.14 deformation upon impact by the specimen would not significantly increase damage to the IATA 10.6.3.3 specimen.
49 CFR 173.465(c)(5) The target for the free drop test must be a flat, horizontal surface of such mass and rigidity that any increase in its resistance to displacement or deformation upon impact by the specimen would not significantly increase the damage to the specimen.
4 Drop Test 1) Calibrated Drop Test Target
- 2) Tape Measure
- 3) Drop Release Mechanism
- 4) Thermometer
- 5) Hoist/Winch
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 32 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Table 7.2 Test Equipment Hypothetical Accident Conditions Tests Regulation/Other Test Equipment Normal Condition Tests Notations Safety General Safety Equipment
- 1) Safety Glasses
- 2) Gloves
- 3) Survey Meter
- 4) Rate Alarm
- 5) Personal Dosimetry Video/ 1) Video Recording Recording 2) Camera Equipment 3) Video / Camera Stand IAEA SSR-6 717 The target for the drop test specified in paras 705, 722, 725(a), 727 and 735 shall be a flat, IAEA TSR-1 horizontal surface of such a character that any increase in its resistance to displacement or ADR 6.4.14 deformation upon impact by the specimen would not significantly increase damage to the IATA 10.6.3.3 specimen.
49 CFR 173.465(c)(5) The target for the free drop test must be a flat, horizontal surface of such mass and rigidity that any increase in its resistance to displacement or deformation upon impact by the specimen would not significantly increase the damage to the specimen.
9 meter Drop Test 1) Calibrated Drop Test Target
- 2) Tape Measure
- 3) Drop Release Mechanism
- 4) Thermometer
- 5) Hoist/Winch
- 6) Manlift 1 meter Drop Test 1) Calibrated Drop Test Target
- 2) Calibrated Puncture Pin
- 3) Tape Measure
- 4) Drop Release Mechanism
- 5) Thermometer
- 6) Hoist/Winch
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 33 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 8.0 Test Procedures Table 8.1 Type A Packaging /Normal Conditions Tests Regulation Regulation Requirements Notations 10 CFR 71.71(b) Initial conditions. With respect to the initial conditions for the tests in this section, the demonstration of compliance with the requirements of this part must be based on the ambient temperature preceding and following the tests remaining constant at that value between -29°C (-20°F) and +38°C
(+100°F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be considered to be the maximum normal operating pressure, unless a lower internal pressure consistent with the ambient temperature considered to precede and follow the tests is more unfavorable.
10 CFR 71.71(c) Conditions and tests.
(1) Heat. An ambient temperature of 38°C (100°F) in still air, and insolation according to the following table:
Form and location of surface Insolation for 12 h per day (W/m2)
Flat surfaces transported horizontally None Other Surfaces 800 Flat Surfaces not transported horizontally 200 Curved Surfaces 400 (2) Cold. An ambient temperature of -40°C (-40°F) in still air and shade.
(3) Reduced external pressure. An external pressure of 25 kPa (3.5 lbf/in2) absolute.
(4) Increased external pressure. An external pressure of 140 kPa (20 lbf/in2) absolute.
(5) Vibration. Vibration normally incident to transport.
(6) Water spray. A water spray that simulates exposure to rainfall of approximately 5 cm/h (2 in/h) for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
(7) Free drop. Between 1.5 and 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the conclusion of the water spray test, a free drop through the distance specified below onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.
(8) Corner drop. A free drop onto each corner of the package in succession, or in the case of a cylindrical package onto each quarter of each rim, from a height of 0.3 m (1 ft) onto a flat, essentially unyielding, horizontal surface. This test applies only to fiberboard, wood, or fissile material rectangular packages not exceeding 50 kg (110 lbs) and fiberboard, wood, or fissile material cylindrical packages not exceeding 100 kg (220 lbs).
(9) Compression. For packages weighing up to 5000 kg (11,000 lbs), the package must be subjected, for a period of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, to a compressive load applied uniformly to the top and bottom of the package in the position in which the package would normally be transported. The compressive load
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 34 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Regulation Requirements Notations Continued must be the greater of the following:
10 CFR 71.71(c) (i) The equivalent of 5 times the weight of the package; or (ii) The equivalent of 13 kPa (2 lbf/in2) multiplied by the vertically projected area of the package.
(10) Penetration. Impact of the hemispherical end of a vertical steel cylinder of 3.2 cm (1.25 in) diameter and 6 kg (13 lbs) mass, dropped from a height of 1 m (40 in) onto the exposed surface of the package that is expected to be most vulnerable to puncture. The long axis of the cylinder must be perpendicular to the package surface.
IAEA SSR-6 719 The tests are the water spray test, the free drop test, the stacking test and the penetration test.
IAEA TSR-1 719 Specimens of the package shall be subjected to the free drop test, the stacking test and the ADR 6.4.15.1 penetration test, preceded in each case by the water spray test. One specimen may be used for all the IATA 10.6.3.4 tests, provided that the requirements of para. 720 are fulfilled IAEA SSR-6 720 The time interval between the conclusion of the water spray test and the succeeding test shall be such IAEA TSR-1 720 that the water has soaked into the maximum extent, without appreciable drying of the exterior of the ADR 6.4.15.2 specimen. In the absence of any evidence to the contrary, this interval shall be taken to be 2 h if the IATA 10.6.3.4.1 water spray is applied from four directions simultaneously. No time interval shall elapse, however, if the water spray is applied from each of the four directions consecutively.
IAEA SSR-6 721 Water spray test: The specimen shall be subjected to a water spray test that simulates exposure to rainfall Not Performed IAEA TSR-1 721 of approximately 5 cm per hour for at least 1 h.
ADR 6.4.15.3 IATA 10.6.3.4.2 49 CFR 173.465(b) Water spray test. The water spray test must precede each test or test sequence prescribed in Not Performed this section. The water spray test must simulate exposure to rainfall of approximately 5 cm (2 inches) per hour for at least one hour. The time interval between the end of the water spray test and the beginning of the next test must be such that the water has soaked in to the maximum extent without appreciable drying of the exterior of the specimen. In the absence of evidence to the contrary, this interval may be assumed to be two hours if the water spray is applied from four different directions simultaneously. However, no time interval may elapse if the water spray is applied from each of the four directions consecutively.
IAEA SSR-6 722 Free drop test: The specimen shall drop onto the target so as to suffer maximum damage in respect of IAES TSR-1 722 the safety features to be tested:
ADR 6.4.15.4 (a) The height of the drop, measured from the lowest point of the specimen to the upper surface of the IATA 10.6.3.4.3 target, shall be not less than the distance specified in Table 14 for the applicable mass. The target shall be as defined in para. 717.
(b) For rectangular fibreboard or wood packages not exceeding a mass of 50 kg, a separate specimen shall be subjected to a free drop onto each corner from a height of 0.3 m.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 35 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Regulation Requirements Notations Continued (c) For cylindrical fibreboard packages not exceeding a mass of 100 kg, a separate specimen shall be The ISORAD-TC1 has a IAEA SSR-6 722 subjected to a free drop onto each of the quarters of each rim from a height of 0.3 m. mass of less than 5000 kg.
IAES TSR-1 722 The test will be conducted ADR 6.4.15.4 TABLE 14. FREE DROP DISTANCE FOR TESTING PACKAGES TO NORMAL CONDITIONS OF TRANSPORT at 1.2 meters.
IATA 10.6.3.4.3 Package mass (kg) Free drop distance (m) package mass < 5 000 1.2 1.2 5 000 package mass < 10000 0.9 0.9 10 000 package mass < 15000 0.6 0.6 15 000 package mass 0.3 0.3 49 CFR 173.465(c) Free drop test. The specimen must drop onto the target so as to suffer maximum damage to the safety features being tested, and:
(1) The height of the drop measured from the lowest point of the specimen to the upper surface of the target may not be less than the distance specified in table 10, for the applicable package mass. The target must be as specified in § 173.465(c)(5). Table 10 is as follows: Table 10 - Free Drop Distance for Testing Packages to Normal Conditions of Transport (2) For packages containing fissile material, the free drop test specified in paragraph (c)(1) of this section must be preceded by a free drop from a height of 0.3 m (1 foot) on each corner, or in the case of cylindrical packages, onto each of the quarters of each rim.
(3) For fiberboard or wood rectangular packages with a mass of 50 kg (110 pounds) or less, a separate specimen must be subjected to a free drop onto each corner from a height of 0.3 m (1 foot).
(4) For cylindrical fiberboard packages with a mass of 100 kg (220 pounds) or less, a separate specimen must be subjected to a free drop onto each of the quarters of each rim from a height of 0.3 m (1 foot).
(5) The target for the free drop test must be a flat, horizontal surface of such mass and rigidity that any increase in its resistance to displacement or deformation upon impact by the specimen would not significantly increase the damage to the specimen.
Package mass Test Required Free drop distance The ISORAD-TC1 has a Kilograms (pounds) Meters (Feet) mass of less than 5000 kg.
The test will be conducted
<Mass 5000 (11,000) Test Required 1.2 (4) at 1.2 meters.
5,000 (11,000) Mass to 10,000 (22,000) N/A 0.9 (3) 10,000 (22,000) Mass to 15,000 (33,000) N/A 0.6 (2)
>15,000 (33,000) Mass N/A 0.3 (1)
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 36 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Regulation Requirements Notations IAEA SSR-6 723 Stacking test: Unless the shape of the packaging effectively prevents stacking, the specimen shall be The vertical projected area IAEA TSR-1 723 subjected, for a period of 24 h, to a compressive load equal to the greater of the following: of the package exceeds ADR 6.4.15.5 the 5 times maximum IATA 10.6.3.4.4 (a) The equivalent of 5 times the maximum weight of the package; weight. The vertical (b) The equivalent of 13 kPa multiplied by the vertically projected area of the package. projected area is 701.499 The load shall be applied uniformly to two opposite sides of the specimen, one of which shall be the base on which the package would typically rest.
49 CFR 173.465(d) Stacking test. The vertical projected area of the package does not (1) The specimen must be subjected for a period of at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to a compressive load exceed the 5 times equivalent to the greater of the following:
maximum weight. The (i) A total weight equal to five times the maximum weight of the package; or vertical projected area is (ii) The equivalent of 13 kilopascals (1.9 psi) multiplied by the vertically projected area of 701.499 the package.
(2) The compressive load must be applied uniformly to two opposite sides of the specimen, one of which must be the base on which the package would normally rest.
IAEA SSR-6 724 Penetration test: The specimen shall be placed on a rigid, flat, horizontal surface that will not move IAEA TSR-1 724 significantly while the test is being carried out:
ADR 6.4.15.6 IATA 10.6.3.4.5 (a) A bar, 3.2 cm in diameter with a hemispherical end and a mass of 6 kg, shall be dropped and directed to fall with its longitudinal axis vertical onto the centre of the weakest part of the specimen so that if it penetrates sufficiently far it will hit the containment system. The bar shall not be significantly deformed by the test performance.
(b) The height of the drop of the bar, measured from its lower end to the intended point of impact on the upper surface of the specimen, shall be 1 m.
49 CFR 173.465(e) Penetration test. For the penetration test, the specimen must be placed on a rigid, flat, horizontal surface that will not move significantly while the test is being performed.
(1) A bar of 3.2 cm (1.25 inches) in diameter with a hemispherical end and a mass of 6 kg (13.2 pounds) must be dropped and directed to fall with its longitudinal axis vertical, onto the center of the weakest part of the specimen, so that, if it penetrates far enough, it will hit the containment system. The bar may not be significantly deformed by the test; and (2) The height of the drop of the bar measured from its lower end to the intended point of impact on the upper surface of the specimen must be 1 m (3.3 feet) or greater.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 37 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Regulation Requirements Notations IAEA SSR-6 725 A specimen or separate specimens shall be subjected to each of the following tests unless it can be IAEA TSR-1 725 demonstrated that one test is more severe for the specimen in question than the other, in which case ADR 6.4.16 one specimen shall be subjected to the more severe test.
IATA 10.6.3.5 (a) Free drop test: The specimen shall drop onto the target so as to suffer the maximum damage in respect of containment. The height of the drop measured from the lowest part of the specimen to the upper surface of the target shall be 9 m. The target shall be as defined in para. 717.
(b) Penetration test: The specimen shall be subjected to the test specified in para. 724 except that the height of drop shall be increased to 1.7 m from the 1 m specified in para. 724(b).
49 CFR 173.466 Additional tests for Type A packagings designed for liquids and gases. Not designed for liquids and gasses.
(a) In addition to the tests prescribed in § 173.465, Type A packagings designed for liquids and gases must be capable of withstanding the following tests in this section. The tests are successful if the requirements of § 173.412(k) are met.
(1) Free drop test. The packaging specimen must drop onto the target so as to suffer the maximum damage to its containment. The height of the drop measured from the lowest part of the packaging specimen to the upper surface of the target must be 9 m (30 feet) or greater. The target must be as specified in § 173.465(c)(5).
(2) Penetration test. The specimen must be subjected to the test specified in § 173.465(e) except that the height of the drop must be 1.7 m (5.5 feet).
Table 8.2 Hypothetical Accident Conditions Tests Regulation Regulation Requirements Special Notes IAEA SSR-6 726 The specimen shall be subjected to the cumulative effects of the tests specified in paras 727 and This test is not being IAEA TSR-1 726 728, in that order. Following these tests, either this specimen or a separate specimen shall be performed.
ADR 6.4.17.1 subjected to the effect(s) of the water immersion test(s), as specified in para. 729 and, if IATA 10.6.3.6.0 applicable, para. 730.
10 CFR 71.73(a) Test procedures. Evaluation for hypothetical accident conditions is to be based on sequential This test is not being application of the tests specified in this section, in the order indicated, to determine their performed.
cumulative effect on a package or array of packages. An undamaged specimen may be used for the water immersion tests specified in paragraph (c)(6) of this section.
10 CFR 71.73(b) Test conditions. With respect to the initial conditions for the tests, except for the water immersion tests, to demonstrate compliance with the requirements of this part during testing, the ambient air temperature before and after the tests must remain constant at that value between -29°C (-20°F) and +38°C (+100°F) which is most unfavorable for the feature under consideration. The initial internal pressure within the containment system must be the maximum normal operating pressure,
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 38 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Regulation Requirements Special Notes Continued unless a lower internal pressure, consistent with the ambient temperature assumed to precede and 10 CFR 71.73(b) follow the tests, is more unfavorable.
IAEA SSR-6 727 Mechanical test: The mechanical test consists of three different drop tests. Each specimen shall be IAEA TSR-1 727 subjected to the applicable drops, as specified in para. 659 or para. 685. The order in which the ADR 6.4.17.2 specimen is subjected to the drops shall be such that, on completion of the mechanical test, the IATA 10.6.3.6.1 specimen shall have suffered such damage as will lead to maximum damage in the thermal test that follows:
IAEA SSR-6 727(a) For drop I, the specimen shall drop onto the target so as to suffer maximum damage, and the IAEA TSR-1 727(a) height of the drop, measured from the lowest point of the specimen to the upper surface of the ADR 6.4.17.2(a) target, shall be 9 m. The target shall be as defined in para. 717.
IATA 10.6.3.6.1.1 10 CFR 71.73(c)(1) Free Drop. A free drop of the specimen through a distance of 9 m (30 ft) onto a flat, essentially unyielding, horizontal surface, striking the surface in a position for which maximum damage is expected.
IAEA SSR-6 727(b) For drop II, the specimen shall drop onto a bar rigidly mounted perpendicularly on the target so as IAEA TSR-1 727(b) to suffer maximum damage. The height of the drop, measured from the intended point of impact ADR 6.4.17.2(b) of the specimen to the upper surface of the bar, shall be 1 m. The bar shall be of solid mild steel of IATA 10.6.3.6.1.2 circular cross-section, 15.0 +/- 0.5 cm in diameter and 20 cm long, unless a longer bar would cause greater damage, in which case a bar of sufficient length to cause maximum damage shall be used.
The upper end of the bar shall be flat and horizontal with its edge rounded off to a radius of not more than 6 mm. The target on which the bar is mounted shall be as described in para. 717.
10 CFR 71.73(c)(3) Puncture. A free drop of the specimen through a distance of 1 m (40 in) in a position for which maximum damage is expected, onto the upper end of a solid, vertical, cylindrical, mild steel bar mounted on an essentially unyielding, horizontal surface. The bar must be 15 cm (6 in) in diameter, with the top horizontal and its edge rounded to a radius of not more than 6 mm (0.25 in), and of a length as to cause maximum damage to the package, but not less than 20 cm (8 in) long. The long axis of the bar must be vertical.
IAEA SSR-6 727(c) For drop III, the specimen shall be subjected to a dynamic crush test by positioning the specimen Not required for ISORAD-TC1.
IAEA TSR-1 727(c) on the target so as to suffer maximum damage by the drop of a 500 kg mass from 9 m onto the This package will not transport ADR 6.4.17.2(c) specimen. The mass shall consist of a solid mild steel plate 1 m x 1 m and shall fall in a horizontal 1000 x A2 curies nor transport IATA 10.6.3.6.1.3 attitude. The lower face of the steel plate shall have its edges and corners rounded off to a radius non-special form contents.
of not more than 6 mm. The height of the drop shall be measured from the underside of the plate to the highest point of the specimen. The target on which the specimen rests shall be as defined in para. 717.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 39 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Regulation Requirements Special Notes 10 CFR 71.73(c)(2) Crush. Subjection of the specimen to a dynamic crush test by positioning the specimen on a flat, Not required for ISORAD-TC1.
essentially unyielding horizontal surface so as to suffer maximum damage by the drop of a 500-kg This package density will (1100-lb) mass from 9 m (30 ft) onto the specimen. The mass must consist of a solid mild steel exceed 1000 kg/m3, will not plate 1 m (40 in) by 1 m (40 in) and must fall in a horizontal attitude. The crush test is required transport 1000 x A2 curies, nor only when the specimen has a mass not greater than 500 kg (1100 lb), an overall density not transport non-special form greater than 1000 kg/m3 (62.4 lb/ft3) based on external dimension, and radioactive contents contents.
greater than 1000 A2 not as special form radioactive material. For packages containing fissile material, the radioactive contents greater than 1000 A2 criterion does not apply.
IAEA SSR-6 728 Thermal test: The specimen shall be in thermal equilibrium under conditions of an ambient The ISORAD-TC1 Thermal IAEA TSR-1 728 temperature of 38°C, subject to the solar insolation conditions specified in Table 12 and subject to the tests to be accomplished with ADR 6.4.17.3 design maximum rate of internal heat generation within the package from the radioactive contents. simulation analysis.
IATA 10.6.3.6.2 Alternatively, any of these parameters are allowed to have different values prior to, and during, the test, provided due account is taken of them in the subsequent assessment of package response. The thermal test shall then consist of (a) followed by (b).
(a) Exposure of a specimen for a period of 30 min to a thermal environment that provides a heat flux at least equivalent to that of a hydrocarbon fuel-air fire in sufficiently quiescent ambient conditions to give a minimum average flame emissivity coefficient of 0.9 and an average temperature of at least 800°C, fully engulfing the specimen, with a surface absorptivity coefficient of 0.8 or that value that the package may be demonstrated to possess if exposed to the fire specified.
(b) Exposure of the specimen to an ambient temperature of 38°C, subject to the solar insolation conditions specified in Table 12 and subject to the design maximum rate of internal heat generation within the package by the radioactive contents for a sufficient period to ensure that temperatures in the specimen are decreasing in all parts of the specimen and/or are approaching initial steady state conditions. Alternatively, any of these parameters are allowed to have different values following cessation of heating, provided due account is taken of them in the subsequent assessment of package response. During and following the test, the specimen shall not be artificially cooled and any combustion of materials of the specimen shall be permitted to proceed naturally.
10 CFR 71.73(c)(4) Thermal. Exposure of the specimen fully engulfed, except for a simple support system, in a The ISORAD-TC1 Thermal hydrocarbon fuel/air fire of sufficient extent, and in sufficiently quiescent ambient conditions, to tests to be accomplished with provide an average emissivity coefficient of at least 0.9, with an average flame temperature of at simulation analysis.
least 800°C (1475°F) for a period of 30 minutes, or any other thermal test that provides the equivalent total heat input to the package and which provides a time averaged environmental temperature of 800°C. The fuel source must extend horizontally at least 1 m (40 in), but may not
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 40 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Regulation Regulation Requirements Special Notes Continued extend more than 3 m (10 ft), beyond any external surface of the specimen, and the specimen must 10 CFR 71.73(c)(4) be positioned 1 m (40 in) above the surface of the fuel source. For purposes of calculation, the surface absorptivity coefficient must be either that value which the package may be expected to possess if exposed to the fire specified or 0.8, whichever is greater; and the convective coefficient must be that value which may be demonstrated to exist if the package were exposed to the fire specified. Artificial cooling may not be applied after cessation of external heat input, and any combustion of materials of construction, must be allowed to proceed until it terminates naturally.
IAEA SSR-6 729 Water immersion test: The specimen shall be immersed under a head of water of at least 15 m for Test not performed IAEA TSR-1 729 a period of not less than 8 h in the attitude that will lead to maximum damage. For demonstration demonstration through ADR 6.4.17.4 purposes, an external gauge pressure of at least 150 kPa shall be considered to meet these calculation.
IATA 10.6.3.6.3.1 conditions.
Enhanced water immersion test for Type B(U) and Type B(M) packages containing more than 105A2 and Type C packages.
10 CFR 71.73(c)(5) Immersion--fissile material. For fissile material subject to § 71.55, in those cases where water Test not required. The inleakage has not been assumed for criticality analysis, immersion under a head of water of at least package is non-fissile.
0.9 m (3 ft) in the attitude for which maximum leakage is expected.
IAEA SSR-6 730 Enhanced water immersion test: The specimen shall be immersed under a head of water of at least Test not required. The IAEA TSR-1 730 200 m for a period of not less than 1 h. For demonstration purposes, an external gauge pressure of package is non-fissile.
ADR 6.4.18 at least 2 MPa shall be considered to meet these conditions.
IATA 10.6.3.6.3.2 10 CFR 71.73(c)(6) Immersion--all packages. A separate, undamaged specimen must be subjected to water pressure Test not performed equivalent to immersion under a head of water of at least 15 m (50 ft). For test purposes, an demonstration through external pressure of water of 150 kPa (21.7 lbf/in2) gauge is considered to meet these conditions. calculation.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 41 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Table 8.3 Normal Conditions Test Procedures Step Test Procedures 8.3.1 Specimen Temperature Measurement The penetration, drop, and puncture tests are to be carried out while the carbon steel portions of the package are at or below -40°C. The package does not contain carbon steel . Tests will be conducted at ambient temperature.
8.3.2 Test Specimens Preparation and Inspection Refer to the Specimen Preparation Form Endure the form is complete and signed before beginning of tests.
8.3.2.1 Inspect the test units to ensure compliance with the requirements of prototype regulatory drawings.
8.3.2.2 Weigh each inner container and record the weight.
8.3.2.3 Weigh each outer container and record weight.
8.3.2.4 Survey each complete package to obtain a radiation profile. Mark and record the highest readings.
8.3.2.5 Fully prepare the package for transport conditions. Record the process.
8.3.2.6 Mark each outer package clearly with permanent marker to ensure proper identification.
8.3.2.7 Weigh each transport package in the complete shipping configuration and record the weight.
8.3.2.8 Ensure all parties sign preparation document.
8.3.3 See Test schedule in Section 9.0 8.3.4 Normal Conditions Stacking/Compression Test The first test is the stacking or compression test, according to IAEA SSR-6 723, 10 CFR 71.71(c)(9), and 49 CFR 173.465(d), in which the package is placed under a weight load of at least 703.1 kg (1,550 pounds). The weight load is the greater of five times the maximum package weight or equivalent of 13 kilopascals (1.9 psi) or pounds of force (lbf) multiplied by the vertically projected area of the package. Refer to Section 7.0 for safety equipment and tools.
5 x 300.00 lbf = 1500.00 lbs 403.24 mm(15.88in) x 589.0 mm(23.25in) X 1.9 psi = 701.499 lbf 8.3.4.1 Stacking/Compression Test Setup (Use form QTP-001 F-01)
Prepare the test specimen for the test 8.3.4.1.1 Review the requirements 8.3.4.1.2 Place the specimen on a flat concrete surface in normal shipping position.
8.3.4.1.3 Gradually place a minimum of 703.1 kg(1550 pounds) uniformly distributed on the top of the test
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 42 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Step Test Procedures Continued specimen.
8.3.4 8.3.4.1.4 Test the specimen in accordance with the regulatory requirements.
8.3.4.2 Stacking/Compression Test Assessment (Use form QTP-001 F-04)
Upon completion of the test, Regulatory and Quality Assurance will jointly perform the following:
8.3.4.2.1 Review the test execution to ensure the test was performed in accordance with IAEA SSR-6 723, 10 CFR 71.71, 49 CFR 173.465, and other regulations.
8.3.4.2.2 Thoroughly assess the damage to the test specimen to determine if damage occurred and can the specimen endure further testing.
8.3.4.2.3 Thoroughly evaluate the test specimen to determine if test results alter the planned drop orientations to achieve maximum damage.
8.3.5 Penetration Test The Penetration Test follows the Stacking/Compression Test from Step 8.3.4. A calibrated penetration bar is dropped from a height of at least 1 meter (40 inches) to impact the predetermined impact points on the package. The penetration bar is dropped through free air.
8.3.5.1 Penetration Test Setup (Use form QTP-001 F-02) 8.3.5.1.1 Place the specimen on the calibrated Drop Target and position the orientation according to Section 9.0 Test Schedule.
8.3.5.1.2 Ensure the target impact area on the package is marked and raise the penetration bar at least 1 meter (40 inches) above the marked impact area. Measure with Tape Measure and record and photograph the measurement.
8.3.5.1.3 Test the specimen in accordance with the regulatory requirements.
8.3.5.2 Penetration Test Assessment (Use form QTP-001 F-05)
Upon completion of the test, Regulatory and Quality Assurance will jointly perform the following:
8.3.5.2.1 Review the test execution to ensure the test was performed in accordance with IAEA SSR-6 723, 10 CFR 71.71, 49 CFR 173.465, and other regulations.
8.3.5.2.2 Thoroughly assess the damage to the test specimen to determine if damage occurred and can the specimen endure further testing.
8.3.5.2.3 Thoroughly evaluate the test specimen to determine if test results alter the planned 1.2 meter (4 foot) drop orientation(s) to achieve maximum damage.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 43 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Step Test Procedures 8.3.6 1.2 Meter (4 Foot) Free Drop Test The final Normal Transport Conditions test is the 1.2 meter (4 foot) free drop as described in IAEA SSR-6, 10 CFR 71.71(c)(7), 49 CFR 173.465, and other regulations. The drop compounds any damage caused in the first two tests. Upon completion of this step, the first intermediate test inspection will be performed.
Refer to Section 7.0 for information about safety equipment and required tools.
8.3.6.1 1.2 Meter (4 Foot) Free Drop Test Setup (Use form QTP-001 F-02)
The test requires the specimen to be released from a height of at least 1.2 meters (48 inches) onto the calibrated Drop Test Target.
8.3.6.1.1 Use the calibrated Drop Test Target 8.3.6.1.2 Place the specimen on the Drop Test Target surface and orient the package according to Section 9.0 Test Schedule.
8.3.6.1.3 Lift the specimen to a height of at least 1.2 meters (48 inches) above the calibrated Drop Test Target as measured from the lowest hanging part of the test specimen.
8.3.6.1.4 Test the specimen in accordance with the regulatory requirements.
8.3.6.2 1.2 Meter (4Foot) Free Drop Test Assessment (Use Form QTP-001 F-05)
Upon completion of the test, Regulatory and Quality Assurance will jointly perform the following:
8.3.6.2.1 Review the test execution to ensure the test was performed in accordance with IAEA SSR-6 723, 10 CFR 71.71, 49 CFR 173.465, and other regulations.
8.3.6.2.2 Thoroughly assess the damage to the test specimen to determine if damage occurred and can the specimen endure further testing.
8.3.6.2.3 Thoroughly evaluate the test specimen to determine if test results alter the planned 1.2 meter (4 foot) drop orientation(s) to achieve maximum damage.
8.3.7 Perform a detailed inspection and assessment of the tested specimen(s). The inspection is to be recorded on form QTP-001 F-07 the Post Test Detailed Inspection and Damage Assessment.
Ensure all damage is assessed, photographed, and documents signed.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 44 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Table 8.4 Hypothetical Accident Conditions Test Procedures Step Test Procedures 8.4.1 Specimen Temperature Measurement The heaviest test specimen for the 9 meter (30 Foot) drop is to be tested at or below -40°C. Temperature measurements will be made by positioning the thermocouple touching the inner container. The Thermometer will be disconnected from the thermocouple immediately before hoisting the test specimen.
8.4.2 Test Specimens Preparation and Inspection Refer to the Specimen Preparation Form Endure the form is complete and signed before beginning of tests.
8.3.2.1 Inspect the test units to ensure compliance with the requirements of prototype regulatory drawings.
8.3.2.2 Weigh each inner container and record the weight.
8.3.2.3 Weigh each outer container and record weight.
8.3.2.4 Survey each complete package to obtain a radiation profile. Mark and record the highest readings.
8.3.2.5 Place thermocouple in the package.
8.3.2.6 Fully prepare the package for transport conditions. Record the process.
8.3.2.7 Mark each outer package clearly with permanent marker to ensure proper identification.
8.3.2.8 Weigh each transport package in the complete shipping configuration and record the weight.
8.3.2.8 Ensure all parties sign preparation document.
8.4.3 See Test schedule in Section 9.0 8.4.4 9 Meter (30 Foot) Free Drop Test The first Hypothetical Accident Conditions test is the 9 meter (30 foot) free drop as described in IAEA SSR-6 727(a) and 10 CFR 71.73(c)(l).
8.4.4.1 9 Meter (30 Foot) Free Drop Test Setup (Use form QTP-001 F-03)
In this test, the test specimen is released from a height of 9 meters (30 feet) and lands on the calibrated Drop Test Target.
This heaviest test specimen will be frozen sufficiently to be at -40°C or below at the time of impact. Follow the instructions Section 9.0 for measuring and recording the test specimen temperature before and after the drop.
To set up a package (Use form QTP-001 F-03) for the 9 meter (30 foot) free drop test:
8.4.4.1.1 Use the calibrated Drop Test Target.
8.4.4.1.2 Measure and record the test specimen temperature to ensure that the package is at the specified temperature.
8.4.4.1.3 Place the test specimen on the drop surface and position it according to the appropriate orientation as in Section 9.0.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 45 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Step Test Procedures Continued 8.4.4.1.1 Raise the package so that the test specimen impact target is at least 9 meters (30 feet) above the Drop Test Target.
8.4.4 8.4.4.1.5 Test the specimen in accordance with the regulatory requirements.
8.4.4.2 9 Meter (30 Foot) Free Drop Test Assessment (Use form QTP-001 F-06)
Upon completion of the test, Regulatory and Quality Assurance will jointly perform the following tasks:
8.4.4.2.1 Review the test execution to ensure that the test was performed in accordance with IAEA SSR-6, 10 CFR 71.73, 49 CFR 173.465 and in accordance with the impact orientation and other conditions specified in this plan.
8.4.4.2.2 Thoroughly assess the damage to the test specimen to determine if damage occurred and can the specimen endure further testing.
8.4.4.2.3 Thoroughly evaluate the test specimen to determine if test results alter the planned Puncture Test drop orientation(s) to achieve maximum damage.
8.4.5 Puncture Test The 9 meter (30 foot) free drop is followed by the puncture test, per IAEA SSR-6 727(b) and 10 CFR 71.73(c)(3), in which the package is dropped from a height of at least 1 meter (40 inches) onto the calibrated Puncture Pin.
The Puncture Pin is to be bolted to the calibrated Drop Test Target used in the free drop tests. The Puncture Pin meets the minimum height 20 cm (8 inches) required in IAEA SSR-6 and 10 CFR 71.73(6)(3). The specimen has no projections or overhanging members longer than 8 inches, which could act as impact absorbers, thus allowing the Puncture Pin to cause the maximum damage to the specimen.
Refer to Section 7.0 for information about safety equipment and required tools. Refer to Section 9.0 for test sequence.
8.4.5.1 Puncture Test Setup (Use form QTP-001 F-03)
To set up a test specimen for the puncture test:
8.4.5.1.1 Measure and record the test specimen temperature to ensure that the package is at the specified temperature.
8.4.5.1.2 Position the test specimen according to Section 9.0.
8.4.5.1.3 Check the alignment of the center-of-gravity with the impact point on the test specimen and the Step Test Procedures
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 46 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Continued Puncture Pin.
8.4.5 8.4.5.1.4 Lift the test specimen to at least 1 meter (40 inches) between the lowest part of the test specimen and the top of the Puncture Pin.
8.4.5.1.5 Test the specimen in accordance with the regulatory requirements.
8.4.5.2 Puncture Test Assessment (Use form QTP-001 F-06)
Upon completion of the test, Regulatory and Quality Assurance will jointly perform the following:
8.4.5.2.1 Review the test execution to ensure the test was performed in accordance with IAEA SSR-6 727(b),
10 CFR 71.71, 49 CFR 173.465, and other regulations.
8.3.5.2.2 Thoroughly assess the damage to the test specimen to determine if damage occurred and can the specimen endure further testing.
8.4.6 Perform a detailed inspection and assessment of the tested specimen(s). The inspection is to be recorded on form QTP-001 F-07 the Post Test Detailed Inspection and Damage Assessment.
Ensure all damage is assessed, photographed, and documents signed.
8.4.7 Thermal Tests are to be conducted via simulation.
8.4.8 Final Test Inspection Perform the following inspections after completion of all the required testing:
8.4.8.1 Measure and record-any damage to the test specimen(s).
8.4.8.2 Measure and record a radiation profile of the test specimen in accordance with requirements.
8.4.8.3 Document and assess the external radiation levels on the surface and at 1 meter from the surface of the test specimen(s).
8.4.8.4 Determine whether it is necessary to dismantle the test specimen(s) for inspection of hidden component damage or failure.
8.4.8.5 If proceeding with the inspection, record and photograph the process of removing any component from the test specimen(s).
8.4.8.6 Measure and record any damage or failure found in the process of dismantling the test specimen(s).
Regulatory and Quality Assurance will make a final assessment of the test specimen(s), and jointly determine whether the specimen meets the requirements of IAEA SSR-6 and 10 CFR 71.
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 47 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Section 9.0 ISORAD-TC1 Test Schedule Table 9.1 Daily Test Schedule Day 1 /Time Normal Condition Test Verify Measurement Perform Result Notes Water Spray N/A N/A Pass The materials of N/A construction are not affected by water spray.
Day 1 - 8:00 Stacking Test. (One Package only)
AM to Day 2 - 8:00 (1) A total weight equal to five times the maximum 1) ______________
AM weight of the package; or (2) The equivalent of 13 kilopascals (1.9 psi) 2) ______________
multiplied by the vertically projected area of the package.
Place ______ pounds of weight must be stacked for the test. Take Photo Start Date and Time ____________________ ___________
End Date and Time ______________________
Day 2 /Time Normal Condition Test Verify Measurement Perform Result Notes 8:00AM - Free Drop from 1.2 meters. Turn on Video at the Video on 10:00 AM beginning of test and record until tests are ___________
complete.
- 1) ISORAD-TC1 Prototype S/N _____________ 1) _______
A) Drop at 45° Angle A) ________
B) Drop on Top B) ________
- 2) ISORAD-TC1 Prototype S/N _____________ 2) _______
A) Drop at 45° Angle A) ________
B) Drop on Top B) ________
Day 2 /Time Normal Condition Test Verify Measurement Perform Result Notes 10:00 AM - Safety Survey Package 10:30 AM Package 1 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 48 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Day 2 /Time Normal Condition Test Verify Measurement Perform Result Notes Continued Package 2 S/N _______________ Top _________
10:00 AM - Bottom_________
10:30 AM Side 1 _________
Side 2 _________
Assess Package Damage Package 1 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Package 2 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
10:30 AM - Penetration test the specimen must be 1 m (3.3) Video on 11:30 AM or greater. _____________
- 1) ISORAD-TC1 Prototype S/N _____________ 1) _______
A) Drop on Bottom A) ________
B) Drop on Top B) ________
C) Drop on Side C) ________
- 2) ISORAD-TC1 Prototype S/N _____________ 2) _______
A) Drop on Bottom A) ________
B) Drop on Top B) ________
C) Drop on Side C) ________
11:30 AM - Safety Survey Package 12:30 AM Package 1 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Package 2 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 49 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Day 2 /Time Normal Condition Test Verify Measurement Perform Result Notes Continued Assess Package Damage 11:30 AM - Package 1 S/N _______________ Top _________
12:30 AM Bottom_________
Side 1 _________
Side 2 _________
Package 2 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
12:30 PM - Unload the container and insert simulated isotope 1) _________
12:45 PM weight load.
12:45 PM- Freeze heaviest package containing Bulk & PIC Photo Start________
Day 3 12:45 inner package to at least -40°C before Hypothetical Photo Finish_______
PM Accident Conditions Tests.
ISORAD-TC1 Prototype S/N _____________ Temp __________
Day 3 /Time Accident Condition Test Verify Measurement Perform Result Notes 1:00 PM - Free Drop from 9 meters. Turn on Video at the Video on 3:00 PM beginning of test and record until tests are ___________
complete.
- 1) ISORAD-TC1 Prototype S/N _____________ 1) _______
A) Drop at 45° Angle A) ________
B) Drop on Top B) ________
- 2) ISORAD-TC1 Prototype S/N _____________ 2) _______
A) Drop at 45° Angle A) ________
B) Drop on Top B) ________
3:00 PM - Safety Survey Package 3:30 PM Package 1 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Package 2 S/N _______________ Top _________
Bottom_________
Side 1 _________
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 50 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Side 2 _________
Day 3 /Time Accident Condition Test Verify Measurement Perform Result Notes Continued Assess Package Damage 3:00 PM - Package 1 S/N _______________ Top _________
3:30 PM Bottom_________
Side 1 _________
Side 2 _________
Package 2 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
3:30 PM - Puncture Free Drop from 1 meter. Turn on Video at Video on 5:00 PM start of test and record until tests are complete. ___________
- 1) ISORAD-TC1 Prototype S/N _____________ 1) _______
A) Drop on Top A) ________
B) Drop on Side B) ________
C) Drop on Bottom C) ________
- 2) ISORAD-TC1 Prototype S/N _____________ 2) _______
A) Drop on Top A) ________
B) Drop on Side B) ________
C) Drop on Bottom C) ________
Day 3 /Time Accident Condition Test Verify Measurement Perform Result Notes 5:00 PM - Safety Survey Package 5:30 PM Package 1 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Package 2 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Assess Package Damage Package 1 S/N _______________ Top _________
Bottom_________
Side 1 _________
Side 2 _________
Procedure Number: QUALITY ASSURANCE PROGRAM / MANAGEMENT SYSTEM Page: 51 of 51 QTP-001 ISO-RAD CANADA INC Revision Number: 1 Date: 18 December 2019 ISORAD-TC1 TEST PLAN Day 3 /Time Accident Condition Test Verify Measurement Perform Result Notes Continued Package 2 S/N _______________ Top _________
5:00 PM - Bottom_________
5:30 PM Side 1 _________
Side 2 _________
Not Free Drop III Performed Not Thermal tests are conducted through computer Performed simulation.
Not Water Immersion Tests not performed Performed Test Witnesses Printed Name Signature Date Regulatory and RSO Quality Assurance President Other Witness Other Witness Other Witness
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-53 2.12.2 QTP-001 Test Report Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 1 of 70 ISORAD-TC1 Transport Package QTP-001 Test Report ISO-RAD Canada Inc Kevin J. Schehr, DBA Managing Director January 29, 2020 Revision 0 Managing Director Engineer Quality Manager Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 2 of 70 Table of Contents 1.0 Introduction ............................................................................................................................................. 5 2.0 Package Description ................................................................................................................................ 6 2.1 Outer Drum ......................................................................................................................................... 6 2.2 Bulk and PIC Inner Container .............................................................................................................. 6 2.3 Bulk and PIC Inner Container (BPIC) 2835A ........................................................................................ 7 3.0 General Testing Information ................................................................................................................... 7 3.1 Weather conditions ............................................................................................................................. 7 3.2 Prototypes used for Normal Conditions of Transport Testing ............................................................ 8 3.2.1 Prototype #1 ISORAD-TC1 BPIC - Square ................................................................................ 8 3.2.2 Prototype #2 ISORAD-TC1 BPIC 2835A .................................................................................. 8 3.2.3 Prototype #3 Empty Outer Drum ................................................................................................ 8 3.3 Pretest Radiation Profile Survey ......................................................................................................... 8 3.3.1 Prototype #1 BPIC - Square ........................................................................................................ 8 3.3.2 Prototype #2 BPIC 2835A........................................................................................................... 9 3.3.3 General Safety and ALARA........................................................................................................ 9 4.0 Normal Conditions of Transport Testing ................................................................................................ 9 4.1 Test Equipment Normal Conditions Tests ........................................................................................... 9 4.1.1 General Safety Equipment........................................................................................................... 9 4.1.2 Video and Recording Equipment ................................................................................................ 9 4.2 Water Spray Test - Not performed ................................................................................................... 10 4.3 Compression Test - Stacking Test ..................................................................................................... 10 4.3.1 Equipment Used ........................................................................................................................ 10 4.3.2 Test Weight ............................................................................................................................... 10 4.3.3 Description of test performed .................................................................................................... 11 4.3.4 Equipment Used ........................................................................................................................ 12 4.3.5 Pass Criteria............................................................................................................................... 12 4.3.6 Describe Physical Damage ........................................................................................................ 12 4.3.7 Radiation Levels........................................................................................................................ 12 4.3.8 Test Result ................................................................................................................................. 12 4.4 Free Drop Test (1.2 Meter)................................................................................................................ 12 4.4.1 Test Height - Free Drop Test .................................................................................................... 12 4.4.2 Description of Test Performed .................................................................................................. 13 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 3 of 70 4.4.3 Equipment Used ........................................................................................................................ 16 4.4.4 Pass Criteria............................................................................................................................... 16 4.4.5 Describe Physical Damage ........................................................................................................ 16 4.4.6 Radiation Levels........................................................................................................................ 17 4.4.7 Test Result ................................................................................................................................. 17 4.5 Penetration Test - Normal Conditions of Transport ......................................................................... 17 4.5.1 Drop Height - Penetration Test ................................................................................................. 17 4.5.2 Description of Test Performed .................................................................................................. 17 4.5.3 Equipment Used ........................................................................................................................ 18 4.5.4 Pass Criteria............................................................................................................................... 19 4.5.5 Describe Physical Damage ........................................................................................................ 19 4.5.6 Radiation Levels........................................................................................................................ 20 4.5.7 Test Result ................................................................................................................................. 21 4.6 Post NCT Testing Damage Assessment ............................................................................................. 21 5.0 Hypothetical Accident Conditions Tests ............................................................................................... 24 5.1 Free Drop I - 9m (30ft)...................................................................................................................... 25 5.1.1 Description of Test Performed .................................................................................................. 25 5.1.2 Equipment Used ........................................................................................................................ 26 5.1.3 Pass Criteria............................................................................................................................... 26 5.1.4 Describe Physical Damage ........................................................................................................ 27 5.1.5 Radiation Levels........................................................................................................................ 35 5.1.6 Test Result ................................................................................................................................. 35 5.2 Free Drop II (1 Meter Puncture Test) ................................................................................................ 35 5.2.1 Description of Test Performed .................................................................................................. 35 5.2.2 Equipment Used ........................................................................................................................ 37 5.2.3 Pass Criteria............................................................................................................................... 37 5.2.4 Describe Physical Damage ........................................................................................................ 38 5.2.5 Radiation Levels........................................................................................................................ 38 5.2.6 Test Result ................................................................................................................................. 38 6.0 Detailed Overall Damage Assessment .................................................................................................. 38 6.1 Detailed Measurements .................................................................................................................... 39 6.2 Damage Assessment Conclusion ....................................................................................................... 50 7.0 Appendix ............................................................................................................................................... 52 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 4 of 70 7.1 Radiation Profile Results ................................................................................................................... 52 7.1.1 Pretest Baseline Radiation Profile ............................................................................................. 53 7.1.2 Post NCT Radiation Profile....................................................................................................... 58 7.1.3 Post HACT Radiation Profile .................................................................................................... 63 7.2 Test Equipment Drawings: ................................................................................................................ 68 7.3
References:
........................................................................................................................................ 70 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 5 of 70 Test Report for the Model ISORAD-TC1 Type B (U) Package 1.0 Introduction ISO-RAD Canada Inc (ISO-RAD) has designed a new Type B(U) package, the model ISORAD-TC1 Transport Container (ISORAD-TC1). The ISORAD-TC1 is a general purpose container used to transport non-fissile special form radioactive material. To demonstrate compliance with the performance standards required IAEA SSR-6, IAEA TS-R-1, and 10 CFR Part 71, a series of required tests were conducted on the ISORAD-TC1 Type B(U) package from on January 3, 2020 through January 6, 2020.
The ISORAD-TC1 package is designed as a transport package and storage container for Type B quantities of special form radioactive materials primarily, but not limited to, Iridium-192 (Ir-192),
Selenium-75 (Se-75), Ytterbium-169 (Yb-169), Cesium-137 (Cs-137), and other radioactive isotopes listed in Tables 1.2a and 1.2b of the Safety Analysis Report (SAR). The package conforms to the General Package Requirements, Type A Package Requirements, and Type B(U) Package Requirements criteria for packaging in accordance with IAEA Regulations for the Safe Transport of Radioactive Material (SSR-6), IAEA Regulations for the Safe Transport of Radioactive Material (TS-R-1), 10 CFR 71, and 49 CFR 173 which were in effect at the time of sign-off of this report.
The package was designed and tested to verify compliance with the Normal Conditions of Transport (NCT) and Hypothetical Accident Conditions of Transport (HACT) requirements and tests.
The ISORAD-TC1 package has been designed to withstand the Type A Package (NCT) and Type B Package (HACT) specifications and testing requirements of the International Atomic Energy Agency (IAEA), the Canadian Nuclear Safety Commission (CNSC), the United States Nuclear Regulatory Commission (USNRC), the United States Department of Transportation (USDOT), United Nations (UN) Economic Commission for Europe Committee for Inland Transport, International Air Transport Association (IATA), and other regulatory bodies governing transport packages. The regulations and guides used in the design development of the ISORAD-TC1 are IAEA SSR-6:2018, IAEA TSR-1, SOR/2015-145 CNSC Packaging and Transport of Nuclear Substances Regulations (PTNS Regulations), RD-364/NUREG-1886 The Joint Canada-United States Guide for Approval of Type B(U) and Fissile Material Transportation Packages (NUREG-1886), the 2019 European Agreement concerning the International Carriage of Dangerous Goods by Road (ADR), IATA Dangerous Goods Regulations.
Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 6 of 70 2.0 Package Description 2.1 Outer Drum The Outer Drum comprises of a keg, Attachment Ring Assembly, Lid Assembly, Cork Assembly, and Cork Lid assembly as shown in drawing R180831-100 (SAR Section 1.3.2). The keg is a commercially available standard 1/2 barrel 15.5 gallon keg and is constructed from two rolled and formed austenitic stainless steel halves welded in the center to form a cylinder. A base and top with rims are welded to the rolled cylinder to form the keg body. The keg is modified with Attachment Ring Supports and Attachment Ring Assembly, which are permanently welded into place to form the attachment point for the Lid Assembly. The Cork Assembly is assembled into the modified keg. An optional stainless steel Cavity Sleeve can be placed within the Cork Assembly cavity to protect the cork during normal handling operations.
The Outer Drum closure is facilitated by eight M14 x 2 threaded bolt holes for securing the Lid Assembly to the Attachment Ring and two Lock Studs which are permanently welded into position.
The Lid Assembly is a circular plate with ten holes machined for the eight closure bolts and two holes for the Lock Studs. The Lid Assembly is attached to the body with eight M14 x 2 austenitic stainless steel hex head bolts with lock and flat washers, two padlocks for security, and tamper-evident seal (lead wire seal, special zip tie, or other tamper-evident device) between the two Lock Studs. The Outer Drum has two handle holes that can be used to lift the package.
2.2 Bulk and PIC Inner Container The Bulk & PIC Inner Container (BPIC) - Square inner container forms part of the containment system and consists of a body, a removable Shield Plug Assembly bolted together with four M6 x 1 closure screws, and a removable lid assembly bolted together with eight M8 x 1.25 closure screws as shown in drawing R180831-200 (SAR Section 1.3.2).
The body assembly is formed from austenitic stainless steel square casing filled with depleted uranium or tungsten shield which is surrounded by brass. The depleted uranium or tungsten forms the shielding for the walls and base of the inner cavity. The shielding cavity is filled with a titanium cavity sleeve that is press fit into the Top Plate and LocTite, JB Weld, or brazed together. The austenitic stainless steel casing consists of three austenitic stainless steel pieces, the Outer Casing, Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 7 of 70 Top Plate, and Bottom Plate. Each piece is machined from solid or tubular austenitic stainless steel.
The Top Plate and Bottom Plate are welded to the Outer Casing with a circumferential full penetration groove weld which is both visually and liquid penetrant tested according to AWS D1.6/D1.6M or equivalent international standard.
The Shield Plug Assembly is formed from the titanium Plug Housing and Plug Housing Lid filled with the depleted uranium or tungsten Plug Shield and the brass Plug Housing Spacer. The Plug Housing and Plug Housing Lid are welded together with a circumferential fillet weld which is both visually and liquid penetrant tested according to AWS D1.9/D1.9M or equivalent international standard.
2.3 Bulk and PIC Inner Container (BPIC) 2835A The Bulk & PIC Inner Container (BPIC) 2835A is a round inner container that forms part of the containment system and consists of a body, a removable Shield Plug Assembly, and a removable lid assembly bolted together with six M8 x 1.25 closure screws as shown in drawing R180831-500 (SAR Section 1.3.2).
The body assembly is formed from a stainless steel round casing filled with depleted uranium tungsten shield which is surrounded by brass. The depleted uranium or tungsten forms the shielding for the walls and base of the inner cavity. The shielding cavity is filled with a stainless steel cavity sleeve that is welded to the Top Plate. The stainless steel casing consists of three austenitic stainless steel pieces, the Outer Casing, Top Plate, and Bottom Plate. Each piece is machined from solid austenitic stainless steel. The Top Plate and Bottom Plate are welded to the Outer Casing with a circumferential full penetration groove weld which is both visually and liquid penetrant tested according to AWS D1.6/D.6M or equivalent international standard.
3.0 General Testing Information 3.1 Weather conditions Testing was conducted 3-6 January 2020 in Houston Texas. The weather was relatively constant with ambient temperature averaging 25.5°C and a relative humidity averaging 47.6 % humidity.
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 8 of 70 3.2 Prototypes used for Normal Conditions of Transport Testing The NCT tests were conducted using three prototype ISORAD-TC1s that would subject the design to the worst case conditions. The goal was to inflict the maximum damage possible to the prototypes for each test condition.
3.2.1 Prototype #1 ISORAD-TC1 BPIC - Square Prototype #1 was the ISORAD-TC1 configured with the Bulk & PIC Inner Container (BPIC) -
Square configuration. This is the heaviest of all configurations of the ISORAD-TC1 and will have the greatest vulnerability to damage from normal and accident conditions.
3.2.2 Prototype #2 ISORAD-TC1 BPIC 2835A Prototype #2 was the ISORAD-TC1 configured with the BPIC 2385A inner container. The configuration was tested because this configuration uses only six bolts to secure the Inner Container Lid, whereas the other BPIC configurations use eight bolts to secure the lid. Additionally, the BPIC 2835A shield plug is held in place by gravity and the Inner Container Lid and is not independently secured except by an additional bolted down inner container lid. All other configurations of the BPIC have the shield plug additionally secured by additional bolts.
3.2.3 Prototype #3 Empty Outer Drum Prototype #3. A third prototype was used for the stacking test only. This prototype was configured with no Inner Container or Cork Assembly. Using Prototype #3 for this test is a worst case because no additional support was provided by the Cork Assembly or the Inner Container would be the most vulnerable to damage from the stacking test.
Both prototypes #1 and #2 underwent all other NCT and HACT tests. If these versions of the package can withstand normal and accident testing, all other configurations will meet the regulatory criteria as well.
3.3 Pretest Radiation Profile Survey 3.3.1 Prototype #1 BPIC - Square Prototype # 1 was loaded on January 4, 2020 with 2317.707 curies of Ir-192 in special form capsules. A radiation profile survey was conducted, and the results were recorded on the Shielding Efficiency Test Survey Form dated 04 JAN 2020. A calibrated ND-2000 Survey Meter S/N 96219 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 9 of 70 was used to conduct the survey. The survey is a baseline for the NCT posttest radiation profile survey. The complete results are in Appendix 7.1 of this test report.
3.3.2 Prototype #2 BPIC 2835A Prototype # 2 was loaded on January 3, 2020 with 2339.571 curies of Ir-192 in special form capsules. A radiation profile survey was conducted, and the results were recorded on the Shielding Efficiency Test Survey Form dated 03 JAN 2020. A calibrated ND-2000 Survey Meter S/N 96219 was used to conduct the survey. The survey is a baseline for the NCT posttest radiation profile survey. The complete results are in Appendix 7.1 of this test report.
3.3.3 General Safety and ALARA For general safety and ALARA, the radioactive material contents were removed before NCT testing was conducted. After all NCT tests were completed, each prototype was again loaded with 2300 +
Curies of Ir-192 and all radiation profile surveys were again taken to compare with the base line reading to determine if the package met the radiation containment requirement for NCT and other pass/fail criteria. After the HACT tests were completed, each prototype was again loaded with 2200
+ Curies of Ir-192 and all radiation profiles surveys were taken again. The baseline for the second survey was the readings after the NCT tests. The results are noted in the appropriate appendix section of the report for each test conducted.
4.0 Normal Conditions of Transport Testing 4.1 Test Equipment Normal Conditions Tests 4.1.1 General Safety Equipment 4.1.1.1 Safety Glasses 4.1.1.2 Gloves 4.1.1.3 Survey Meter 4.1.1.4 Personal Dosimetry 4.1.2 Video and Recording Equipment 4.1.2.1 Video Recording 4.1.2.2 Camera 4.1.2.3 Video / Camera Stand Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 10 of 70 4.2 Water Spray Test - Not performed The ISORAD-TC1 packages are constructed of water-resistant materials throughout. Therefore, the water spray test would not reduce the shielding effectiveness or structural integrity of the package.
The test was not conducted.
4.3 Compression Test - Stacking Test Test Date: 3-5 January 2020 Test Location: Houston, TX 4.3.1 Equipment Used 4.3.1.1 Approximately 1097.5 kg (2,419.573 pounds) of weight 4.3.1.2 Scale 4.3.1.3 Timing device 4.3.2 Test Weight IAEA SSR-6 & TS-R-1 Paragraph 723 requires a total weight equal to five times the maximum mass of the package or the equivalent of 13 kPa multiplied by the vertically projected area of the package. The maximum package weight is 136.08 kg (300.0 pounds) including contents. The test requires a weight of at least 680.39 kg (1500.0 pounds) of weight equally distributed to both ends.
The equivalent of 13 kPa (2.0 psi) or pounds of force (lbf) multiplied by the vertically projected area of the package. The package is 403.24 mm (15.876 inch) in diameter by 589.0 mm (23.1875 inch) in height which results in (15.87 x 23.1875 x 2.0 = 735.97 lbf). The 735.97 lbf is less than the 1,500 pounds. The test weight used was 1097.5 kg (2419.573 pounds), which is greater than 680.389 kg (1500.00 pounds) required to satisfy the test conditions.
Two shipping containers certified by the shipping authority were used as the force to apply the weight to the ISORAD-TC1 for the compression test. One crate weighed 579.0 kg (1276.47lbs) and the other crate weighed 518.5 kg (1143.09lbs) for a combined weight of 1097.5kg (2419.773 lbs).
This weight exceeds the requirement of at least 680.39 kg (1500lb) by over 60%. (Figures 01 & 02).
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 12 of 70 4.3.4 Equipment Used 4.3.4.1 Two certified shipping containers for a combined 1097.5kg (2419.573 pounds) of weight.
4.3.4.2 Timing device 4.3.5 Pass Criteria The criteria set forth in IAEA SSR-6 648, IAEA TS-R-1 646, and 49 CFR 173.412(j) the package must prevent:
- a. Loss or dispersal of the radioactive contents
- b. Significant increase in the radiation levels recorded or calculated at the external surfaces for the condition before the test. (IAEA Significant increase is defined as 20%)
4.3.6 Describe Physical Damage The description of the physical damage of the containment system resulting from the test. No physical damage was observed as a result of this testing.
4.3.7 Radiation Levels The radiation results are located in Appendix 7.1 of this test report.
4.3.8 Test Result No damage occurred to Prototype #3 as a result of the test. Therefore, no loss or dispersal of radioactive contents occurred. Therefore, no increase in the radiation levels as compared to pre-test levels would occur. The ISORAD -TC1 passed the Stacking Test.
4.4 Free Drop Test (1.2 Meter)
Test Date: 4 January 2020 Test Location: Houston, TX 4.4.1 Test Height - Free Drop Test IAEA Paragraph 722 and Part 173.465 (c), the required drop height of a package that weighs
<1100lb (5000 kg) is not less than 4ft (1.2m). This package is NOT designed to transport radioactive material in liquid form, so the more stringent height required by IAEA paragraph 725 and Part 173.466 does not apply. The ISORAD-TC1 has a mass of 136.08 kg (300 pounds) which is less than 5000 kg. The test was conducted at least 1.2 meters (48 inches).
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 15 of 70 of its potential to cause significant deformation of the Outer Drum and with the intent of breaking the lid bolts with high stress loads potentially causing the Inner Container to be ejected from the Outer Drum.
Figure 11: Prototype #2 45° Angle Drop Figure 12: Prototype #2 Topside Down Drop Figure 13: Prototype #2 45° Angle Drop Figure 14: Prototype #2 Topside Down Drop Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 17 of 70 When Prototype #1 was disassembled to load the radioactive material into the package for the posttest radiation survey profile, no interior damage was noted to any internal structures. The Inner Container, Cork Assembly, and Cork Lid were damage free.
Prototype #2: The damage to Prototype #2 was also minimal from the two 1.2 meter Free Drops. The damage from the Orientation 1 Free Drop was like Prototype #1. The Orientation 2 Free Drop resulted in no damage to the Outer Drum.
When Prototype #2 was disassembled to load the radioactive material into the package, the only damage noted was minimal damage to the first spacer for the BPIC 2835A configuration. The first spacer is butt up against the Inner Container lid. The lid has a raised point that impaled the first spacer causing a 3mm diameter break in the first spacer. The damage would not impede the safety of the package nor render the package not fit for transport. The Inner Container, Cork Assembly, and Cork Lid were damage free. There was no damage to the shielding, and the package remained intact.
4.4.6 Radiation Levels The radiation surveys were conducted after the Normal Conditions Tests are completed. See Appendix 7.1 of this test report.
4.4.7 Test Result Damage as noted above occurred to Prototype #1 or #2 as a result of the test. Therefore, no loss or dispersal of radioactive contents occurred. Therefore, no increase in the radiation levels as compared to pre-test levels would occur. The ISORAD-TC1 package passed the 1.2 Meter Free Drop Test.
4.5 Penetration Test - Normal Conditions of Transport Test Date: 4 January 2020 Test Location: Houston, TX 4.5.1 Drop Height - Penetration Test IAEA SSR-6 724 and CFR 173.465 (e) (2) requires the height of the drop of the bar must be 1 m (39.3 in) or greater. ISO-RAD performed the penetration drop from a height of 41 inches (1.04m).
4.5.2 Description of Test Performed Prototype # 1 and Prototype # 2 were each placed on a flat, horizontal surface of solid steel. The package was tested in three orientations each to expose the most vulnerable part of the package to the Penetration Bar (See Figures 17 - 19 & Figures 20 - 22). From a measured height of at least 1 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 19 of 70 4.5.3.4 Tape measure 4.5.4 Pass Criteria The criteria set forth in IAEA SSR-6 648, IAEA TS-R-1 646, and 49 CFR 173.412(j) the package must prevent:
- a. Loss or dispersal of the radioactive contents
- b. Significant increase in the radiation levels recorded or calculated at the external surfaces for the condition before the test. (IAEA Significant increase is defined as 20%)
4.5.5 Describe Physical Damage The Penetration Bar did not pierce the skin of the Outer Drum nor sufficiently dented the skin to impact the Inner Container on either Prototype # 1 or Prototype # 2.
Prototype #1 was tested in three orientations. Orientation 1 was the bottom of the Outer Drum. The Penetration Bar caused an indent of approximately 12.7 mm (0.50 inch) deep. Orientation 2 was the side of the Outer Drum approximately in the middle. The Penetration Bar caused a slight indent of approximately 1mm (0.0394 inch) deep. Orientation 3 was on the top of the Outer Drum Lid Assembly. The Penetration Bar did not indent the Lid Assembly only slightly marring the paint. See Figures 21 - 23 for actual impact damage. There was no visible damage to the Inner Container system or Cork Assembly.
Prototype #2 was tested in three orientations. Orientation 1 was the bottom of the Outer Drum. The Penetration Bar caused an indent of approximately 11.0 mm (0.433 inch) deep. Orientation 2 was the side of the Outer Drum approximately in the middle. The penetration Bar caused a slight indent of approximately 1mm (0.0394 inch) deep. Orientation 3 was on the top of the Outer Drum Lid Assembly. The Penetration Bar did not indent the Lid Assembly only slightly marring the paint. See Figures 24 - 26 for actual impact damage. There was no visible damage to the Inner Container system or Cork Assembly.
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 21 of 70 Table 01: Prototype # 1 Before and After NCT Radiation Profile Comparison Reading Before Test After Test Difference Before Test After Test 1 Difference Pass/
Location Surface Surface 1m m Fail Side 1 57.67 mR/hr 57.67 mR/hr 0.00 mR/hr 1.922 mR/hr 1.922 mR/hr 0.00 mR/hr Pass Side 2 57.67 mR/hr 57.67 mR/hr 0.00 mR/hr 2.402 mR/hr 2.402 mR/hr 0.00 mR/hr Pass Side 3 57.67 mR/hr 57.67 mR/hr 0.00 mR/hr 1.922 mR/hr 1.922 mR/hr 0.00 mR/hr Pass Side 4 62.48 mR/hr 62.48 mR/hr 0.00 mR/hr 1.922 mR/hr 1.922 mR/hr 0.00 mR/hr Pass Top 28.84 mR/hr 28.84 mR/hr 0.00 mR/hr 0.961 mR/hr 0.961 mR/hr 0.00 mR/hr Pass Bottom 38.45 mR/hr 38.45 mR/hr 0.00 mR/hr 1.922 mR/hr 1.922 mR/hr 0.00 mR/hr Pass Table 02: Prototype # 2 Before and After NCT Radiation Profile Comparison Reading Before Test After Test Difference Before Test After Test 1 Difference Pass/
Location Surface Surface 1m m Fail Side 1 66.65 mR/hr 66.65 mR/hr 0.00 mR/hr 1.428 mR/hr 1.428 mR/hr 0.00 mR/hr Pass Side 2 66.65 mR/hr 66.65 mR/hr 0.00 mR/hr 1.904 mR/hr 1.904 mR/hr 0.00 mR/hr Pass Side 3 66.65 mR/hr 66.65 mR/hr 0.00 mR/hr 1.904 mR/hr 1.904 mR/hr 0.00 mR/hr Pass Side 4 66.65 mR/hr 66.65 mR/hr 0.00 mR/hr 1.904 mR/hr 1.904 mR/hr 0.00 mR/hr Pass Top 16.19 mR/hr 15.24 mR/hr -0.95 mR/hr 1.428 mR/hr 1.428 mR/hr 0.00 mR/hr Pass Bottom 66.65 mR/hr 66.65 mR/hr 0.00 mR/hr 1.904 mR/hr 1.904 mR/hr 0.00 mR/hr Pass 4.5.7 Test Result The damage caused by the Penetration Bar to Prototype #1 or #2 was insignificant making only small indentations to the skin of the Outer Drum as a result of the test. No loss or dispersal of radioactive contents occurred. No increase in the radiation levels as compared to pre-test levels occurred. The ISORAD-TC1 package passed the 1 Meter Penetration Test.
4.6 Post NCT Testing Damage Assessment Figure 29: Prototype #1 Post NCT Drop 1 Figure 30: Prototype #1 Post NCT Drop 2 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 22 of 70 Figure 31: Prototype #1 Post NCT Figure 32: Prototype #1 Post NCT Figure 33: Prototype #2 Post NCT Drop 1 Figure 34: Prototype #2 Post NCT Drop 2 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 25 of 70 IAEA SSR-6 and TS-R-1 728 was conducted with computer simulation and calculations. Following these tests, calculations were made to determine the pressure tests from the immersion tests as specified in IAEA SSR-6 and TS-R-1 729 and, if applicable, paragraph 730 were conducted and are presented in the SAR Section 2.
Figure 43: Ambient Temperature at 9 meter Free Drop Test Site 5.1 Free Drop I - 9m (30ft)
Test Date: 6 January 2020 Test Location: Houston, TX 5.1.1 Description of Test Performed For Free Drop I, the package was dropped onto the Drop Test Target so as to suffer maximum damage, and the height of the drop, measured from the lowest point of the specimen to the upper surface of the target, was at least 9 m (30ft). The target was as defined in IAEA paragraph 717.
The design of the ISORAD-TC1 allows only one path for the radioactive contents to come out of the package. The drop orientations are designed to dislodge the inner container from the inside of the Outer Drum. The Inner containers are constructed of a stainless steel outer casing that is welded closed top and bottom with a bolted Shield Plug and an Inner Container Lid that is also bolted down.
The two bolted components secure the special form radioactive contents within the Inner Container.
The Inner Container would have to be ejected from the Outer Drum and both bolted enclosures would need to fail for the special form content to become unshielded. In addition, the Outer Drum is permanently enclosed on the bottom and sides with a removeable bolted down Lid Assembly. The Outer Drum is primarily for protection of the Inner Container and to provide thermal and shock Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 26 of 70 protection during normal transport. Justification for all prototype drop orientations are included in Test Plan QTP-001 (Section 4) and the ISORAD-TC1 SAR Section 2.
The drop orientations apply the maximum force on the package in an attempt to break the Outer Drum Lid Assembly and possibly dislodging the Inner Container from within the Outer Drum. If the Inner Container becomes dislodged, it is subject to direct damage from further impacts and the puncture pin damage. A dislodged Inner Container alone would not significantly increase radiation levels and certainly not to 1000 mR/hr at one meter. As stated before, both the bolted down Inner Container Lid and the bolted down shield plug would have to fail for the special form capsules to become unshielded.
It was determined that the top of the package was the most vulnerable. The Lid Assembly, which is attached to the Attachment Ring Assembly, is the most likely area to be damaged by a 9m (30ft) drop. If the force of impact were to cause the Inner Container to shift and impact the Outer Drum, the lid would be the only area susceptible to damage from the force.
Per paragraph 717, the target for the drop was to be a flat, horizontal surface of such a character that any increase in its resistance to displacement or deformation upon impact by the specimen did not significantly increase damage to the specimen. The target was a solid mild steel measuring 48in L x 48in W x 2in H, with an approximate weight of 592.12 KG (1305.4 pounds). (See Figure 138).
5.1.2 Equipment Used 5.1.2.1 Metal Tower, 14.021 meters (46 feet) high 5.1.2.2 Tape Measure 5.1.2.3 Drop Release Mechanism 5.1.2.4 Drop Test Target 5.1.3 Pass Criteria The requirements set forth in IAEA SSR-6 659 (a) & (b), TS-R-1 657(a) & (b), and 10 CFR 71.51(2).
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 29 of 70 Figure 52: Prototype #1 Free Drop I Figure 53: Prototype #1 Free Drop I Figure 54: Prototype #1 Free Drop I Figure 55: Prototype #1 Free Drop I Figure 56: Prototype #1 Temp Before Free Drop I Figure 57: Prototype #1 Temp After Free Drop I Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 30 of 70 Figure 58: Prototype #1 Free Drop I Figure 59: Prototype #1 Free Drop I 5.1.4.2 Prototype # 1 Free Drop I Orientation 2 Prototype #1 impacted the Drop Test Target as intended. The impact deformed the top rim of the Outer Drum on the entire rim by approximately 9.16 mm (0.36 in). The Lid Assembly was bowed outward by approximately 3.88 mm (0.153 in). No damage was noted on any of the eight M14 x 2 bolts. The rest of the Outer Drum remained intact.
Figures 60 - 62: Prototype #1 Free Drop I Orientation 2 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 31 of 70 Figure 63: Prototype #1 Free Drop I Figure 64: Prototype #1 Free Drop I Figure 65: Prototype #1 Free Drop I Figure 66: Prototype #1 Free Drop I Figure 67: Prototype #1 Free Drop I Figure 68: Prototype #1 Free Drop I Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 34 of 70 Figure 79: Prototype #2 Free Drop I Figure 80: Prototype #2 Free Drop I Figure 81: Prototype #2 Free Drop I Figure 82: Prototype #2 Free Drop I Figure 83: Prototype #2 Free Drop I Figure 84: Prototype #2 Free Drop I Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 35 of 70 Figure 85: Prototype #2 Free Drop I Figure 86: Prototype #2 Free Drop I 5.1.5 Radiation Levels A safety survey was conducted after test. See Appendix 7.1 for Radiation Profile Results 5.1.6 Test Result The damaged inflicted on Prototype #1 or #2 as a result of the two each 9 Meter Free Drop I Tests did not result in damage to the Inner Container. Therefore, no loss or dispersal of radioactive contents occurred. There was a slight increase in the external radiation levels as compared to pre-test levels, but not to the extent of reading 10 mSv/hr at 1 meter. The ISORAD-TC1 package passed the 9 Meter Free Drop I Test.
5.2 Free Drop II (1 Meter Puncture Test)
Test Date: 6 January 2020 Test Location: Houston, TX 5.2.1 Description of Test Performed Each ISORAD-TC1 package was dropped onto the Puncture Bar, which is rigidly mounted perpendicularly on the target using attachment bolts. The drops were once each on its bottom, side and top to ensure suffering maximum damage in all axises. The height of the drops, measured from the intended point of impact of the specimen to the upper surface of the bar, were at least 1 m (40 in). The bar was constructed of solid mild steel of circular cross-section, 15.0 +/- 0.5 cm (6.00 in) in diameter and 20 cm (8.25 inches) long, sufficient to cause maximum damage shall be used (See Figures 88 & 137) . The upper end of the bar was flat and horizontal with its edge rounded off to a radius of not more than 6 mm. The target was mounted on the unyielding target used in the Free Drop I test.
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 36 of 70 Figure 87: Puncture Bar Test Setup Figure 88: Puncture Bar Test Setup Figures 89 - 91: Prototype #1 Free Drop II Puncture Tests Figures 92 - 94: Prototype #1 Free Drop II Puncture Tests Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 37 of 70 Figures 94 - 96: Prototype #2 Free Drop II Puncture Tests Figures 97 - 99: Prototype #2 Free Drop II Puncture Tests 5.2.2 Equipment Used 5.2.2.1 Mild steel bar 3.2cm (6.00 inches) in diameter and (8.5 inches) long.
5.2.2.2 Forklift 5.2.2.3 Tape Measure 5.2.3 Pass Criteria The requirements set forth in IAEA SSR-6 659 (a) & (b), TS-R-1 657(a) & (b), and 10 CFR 71.51(2).
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 38 of 70 (i) It would retain sufficient shielding to ensure that the dose rate 1 m from the surface of the package would not exceed 10 mSv/h with the maximum radioactive contents that the package is designed to contain.
5.2.4 Describe Physical Damage The Prototype #1: There was minimal damage to the Outer Drum of Prototype #1 from the three 1 meter Free Drop II tests. Orientation 1 top of the Outer Drum had no denting or deformation. The Orientation 2 side of the Outer Drum was impacted and a slight indentation 2.71 mm (0.107 inch) in the shape of the Puncture Pin. The Orientation 3 bottom of the Outer Drum was impacted by the Puncture Pin and the bottom deformed inward by 41.92 mm (1.65 inch), but did not tear or break through the skin of the drum.
Prototype #2: There was minimal damage to the Outer Drum of Prototype #2 from the three 1 meter Free Drop II tests. The Orientation 1 top of the Outer Drum had no denting or deformation. The Orientation 2 side of the Outer Drum was impacted and a slight indentation 0.18 mm (0.007 inch) in the shape of the Puncture Pin. The Orientation 3 bottom of the Outer Drum was impacted by the Puncture Pin and the bottom deformed inward by 37.91 mm (1.49 inch), but did not tear or break through the skin of the drum.
5.2.5 Radiation Levels The Radiation Profile results are located in Appendix 7.1 of this test report.
5.2.6 Test Result The damaged inflicted on Prototype #1 or #2 as a result of the two each 1 Meter Free Drop II Tests did not result in damage to the Inner Container. Therefore, no loss or dispersal of radioactive contents occurred. There was a slight increase in the external radiation levels as compared to pre-test levels, but not to the extent of reading 10 mSv/hr at 1 meter. The ISORAD-TC1 package passed the 1 Meter Free Drop II Test.
6.0 Detailed Overall Damage Assessment The overall condition of Prototype #1 and Prototype #2 are good. The cumulative damage after the NCT and the HACT is what would be expected after only NCT testing.
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Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 39 of 70 6.1 Detailed Measurements The damage assessment covers both Prototype #1 and Prototype #2. The assessment compares the baseline measurement from the undamaged Prototype #3 to the Prototype # 1 and Prototype #2 measurements. A carpenters square was used to provide a firm flat surface which to make the measurements. The carpenters square is 2.52 mm thick (See Figure 100). The measurements recorded in the photographs will have 2.52 mm subtracted from the measurement to derive the actual measurement. The caliper used to perform the measurements was calibrated on 16 DEC 2019 and the calibration is valid until 16 JUN 2020 (See Figure 101). Table 03 contains a summary of the damage measurements.
Table 03: Summary of Damage Dimensions Prototype #1 Prototype #2 Prototype Measurement Measurement Difference Measurement Measurement Difference Location Before Test After Test Before Test After Test Top 1 61.71 mm 52.55 -9.16 mm 61.71 mm 56.37 mm -5.34 mm Top 2 61.71 mm 31.48 mm -30.23 mm 61.71 mm 47.54 -14.17 mm Side 6.67 mm 9.38 mm 2.71 mm 6.67 mm 6.49 mm -0.18 mm Bottom 6.78 mm 48.7 mm 41.92 mm 6.78 mm 44.69 mm 37.91 mm Circum. 1244.6 mm 1270.0 mm 25.4 mm 1244.6 mm 1282.7 mm 38.1 mm Height 589.0 mm 558.85 mm -30.16 mm 589.0 mm 571.54 mm -17.46 mm Figure 100: Thickness of Square Figure 101: Calibration Caliper Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 40 of 70 The baseline measurement for the bottom of the Outer Drum is 6.78 mm from the top of the rim to the center of the bottom of the Outer Drum as shown in Figure 102. Prototype #1 measures 48.7 mm deformation towards the interior of the Outer Drum as a result of the Puncture Bar Test (See Figure 103). Prototype # 2 mesures 44.69 mm deformation towards the interior of the Outer Drum as a result of the Puncture Bar Test (See Figure 104).
Figure 102: Baseline Bottom Measurement Figure 103: Prototype #1 After Bottom Puncture Test Figure 104: Prototype #2 After Bottom Puncture Test Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 41 of 70 The baseline measurement for the side of the Outer Drum is 6.67 mm as measured between the chines on the side (See Figure 105). Prototype #1 measures 9.38 mm deformation towards the interior of the Outer Drum as a result of the Puncture Bar Test (See Figure 106). The damage would not inhibit its effectiveness to perform as a Type B package. Prototype # 2 measures 5.94 mm deformation towards the interior of the Outer Drum as a result of the Puncture Bar Test (See Figure 107). The damage would not inhibit its effectiveness to perform as a Type B package.
Figure 105: Baseline Side Measurement Figure 106: Prototype #1 After Side Puncture Test Figure 107: Prototype #2 After Side Puncture Test Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 42 of 70 The baseline measurement for the top portion circumference of the Outer Drum is 1244.6 mm (49.0 inches) as measured between the chines on the side (See Figure 108). Prototype #1 measures 1270.0 mm (50.0 inches) deformation is outwards from the Outer Drum as a result of the 9 meter Free Drop Test (See Figure 109). The damage would not inhibit its effectiveness to perform as a Type B package.Prototype # 2 measures 1282.7 mm (50.5 inches) deformation is outwards from the Outer Drum as a result of the 9 meter Free Drop Test (See Figure 110). The damage would not inhibit its effectiveness to perform as a Type B package.
Figure 108: Baseline Circumfrence Measurement Figure 109: Prototype #1 Free Drop 1 Test Figure 110: Prototype #2 Free Drop 1 Test Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 45 of 70 Figure 116: Prototype #1 Lid After Free Drop 1 Test Figure 117: Prototype #2 Lid After Free Drop 1 Test Figure 118: Prototype #1 Lid After Free Drop 1 Test Figure 119: Prototype #2 Lid After Free Drop 1 Test The baseline Outer Drum Lid Assembly measurement is 6.31 mm with no correction because this is a direct measurement without the carpenters square (See Figure 120). Prototype # 1 lid deformed 3.88 mm outwards creating a slight bow to the lid laying flat (See Figure 121). The damage would not inhibit its effectiveness to perform as a Type B package. Prototype # 2 lid deformed 4.68 mm outwards creating a slight bow to the lid laying flat (See Figure 122). The damage would not inhibit its effectiveness to perform as a Type B package.
Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 46 of 70 Figure 120: Baseline Thickness of the Outer Drum Lid Measurement Figure 121: Prototype #1 Lid After Free Drop 1 Test Figure 122: Prototype #2 Lid After Free Drop 1 Test Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 47 of 70 The baseline Outer Drum height measurement is 586 mm (23.1875 inch) with no correction because this is a direct measurement with a tape measure (See Figure 123). The Prototype # 1 Outer Drum deformed 13.018 mm downwards forcing the side to bulge outwards (See Figure 124). The maximum deformation of Prototype # 1 Outer Drum was 30.163 mm downwards folding inwards from the 9 Meter Free Drop I (See Figure 125). The damage would not inhibit its effectiveness to perform as a Type B package. The Prototype # 2 Outer Drum deformed 11.113 mm downwards forcing the side to bulge outwards (See Figure 126). The maximum deformation of Prototype # 2 Outer Drum was 17.463 mm downwards folding inwards from the 9 Meter Free Drop I (See Figure 127). The damage would not inhibit its effectiveness to perform as a Type B package.
Figure 123: Baseline Height of Outer Drum Measurement Figures 124 & 125: Prototype #1 Height of Outer Drum After HACT Test Measurement Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 48 of 70 Figures 126 & 127: Prototype #2 Height of Outer Drum After HACT Test Measurement The BPIC -Sqaure Inner Container which was installed inside of Prototype #1 did not suffer any damage as a result of the NCT or HACT conditions tests (See Figures 128 - 131). The BPIC -
Squared was fully functional and maintained its shielding integrity. The BPIC remained frozen to -
57.5°C more than three hours after the HACT 9 meter Free Drop (See Figure 129).
Figure 128: Prototype #1 After Free Drop 1 Test Figure 129: Prototype #2 After Free Drop 1 Test Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 50 of 70 Figure 134: BPIC 2835A Figure 135: BPIC 2835A 6.2 Damage Assessment Conclusion The damage suffered by Prototype #1 did not degrade the integrity of the package or the radioactive contents. The Prototype #1 configuration weighed 136.08 kg (300.00 pounds) with the BPIC -
Square Inner Container is the heaviest configuration of the ISORAD-TC1. The Inner Container was undamaged and did not lose its shielding ability. The Outer Drum was damaged, but not to the extent that the Inner Container would or could become dislodged from within the Outer Drum. After the cumulative mechanical damage inflicted from the NCT and HACT testing, the ISORAD-TC1 package would still be able to be transported.
The damage suffered by Prototype #2 did not degrade the integrity of the package or the radioactive contents. The Prototype #2 configuration weighed 114.02 kg (251.37 pounds) with the BPIC 2835A Inner Container. The Inner Container was undamaged and did not lose its shielding ability. The Outer Drum was slightly less damaged than the Prototype #1, but not to the extent that the Inner Container would or could become dislodged from within the Outer Drum. After the cumulative mechanical damage inflicted from the NCT and HACT testing, the ISORAD-TC1 package would still be able to be transported.
Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 51 of 70 Table 04: Prototype # 1 Before and After HACT Radiation Profile Comparison Reading Before Test After Test Difference Before Test After Test Difference Pass/
Location Surface Surface 1m 1m Fail Side 1 57.67 mR/hr 58.75 mR/hr 1.08 mR/hr 1.922 mR/hr 2.445 mR/hr 0.52 mR/hr Pass Side 2 57.67 mR/hr 58.75 mR/hr 1.08 mR/hr 2.402 mR/hr 2.445 mR/hr 0.43 mR/hr Pass Side 3 57.67 mR/hr 63.65 mR/hr 5.98 mR/hr 1.922 mR/hr 1.958 mR/hr 0.04 mR/hr Pass Side 4 62.48 mR/hr 63.65 mR/hr 1.17 mR/hr 1.922 mR/hr 1.958 mR/hr 0.04 mR/hr Pass Top 28.84 mR/hr 29.38 mR/hr 0.54 mR/hr 0.961 mR/hr 0.979 mR/hr 0.02 mR/hr Pass Bottom 38.45 mR/hr 44.07 mR/hr 5.62 mR/hr 1.922 mR/hr 2.445 mR/hr 0.52 mR/hr Pass Table 05: Prototype # 2 Before and After HACT Radiation Profile Comparison Reading Before Test After Test Difference Before Test After Test Difference Pass/
Location Surface Surface 1m 1m Fail Side 1 66.65 mR/hr 68.54 mR/hr 1.89 mR/hr 1.428 mR/hr 1.958 mR/hr 0.53 mR/hr Pass Side 2 66.65 mR/hr 68.54 mR/hr 1.89 mR/hr 1.904 mR/hr 1.958 mR/hr 0.05 mR/hr Pass Side 3 66.65 mR/hr 78.34 mR/hr 11.69 mR/hr 1.904 mR/hr 2.938 mR/hr 1.03 mR/hr Pass Side 4 66.65 mR/hr 68.54 mR/hr 1.89 mR/hr 1.904 mR/hr 2.449 mR/hr 0.55 mR/hr Pass Top 16.19 mR/hr 19.58 mR/hr 3.39 mR/hr 1.428 mR/hr 2.938 mR/hr 1.51 mR/hr Pass Bottom 66.65 mR/hr 73.45 mR/hr 6.80 mR/hr 1.904 mR/hr 3.917 mR/hr 2.01 mR/hr Pass ISO-RAD concludes that the ISORAD-TC1 Transport Container meets or exceeds the requirements set for in IAEA SSR-6, IAEA TS-R-1, CNSC SOR/2015-145 , 10 CFR Part 71, 49 CFR Part 173, UN ADR, and IATA.
Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 52 of 70 7.0 Appendix 7.1 Radiation Profile Results Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 53 of 70 7.1.1 Pretest Baseline Radiation Profile ISORAD-TC1 Transport Package Shielding Efficiency Test Report ISO-RAD Canada Inc Kevin J. Schehr, DBA Managing Director January 15, 2020 Revision 0 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 54 of 70 Shielding Efficiency and Pre NCT Radiation Profile Test Report By: Rafael A. Bustillo Date: 15 JAN 2020
Subject:
Shielding Efficiency Test of ISORAD-TC1 Transport Package Prototype #1 BPIC - Square Inner Container and Prototype #2 BPIC 2835A Inner Container The radiation survey was performed by Rafael A. Bustillo using an NDS ND-2000 Survey Meter S/N 96219, which was calibrated on 03 JAN 2020 with Certificate of Calibration number NDTT-20000301. These profiles are the shielding efficiency of the shield and the pre-NCT survey to establish the baseline radiation levels.
Prototype #2 was loaded with 2339.571 curies of Ir-192 on 03 JAN 2020 and a radiation profile survey was performed. The entire outer surface of the ISORAD-TC1 was surveyed using the survey meter and the highest exposure rate observed on each surface was recorded on the Shielding Efficiency Test Survey Form labeled with Prototype # 2 BPIC 2835A.
Prototype #1 was loaded with 2317.707 curies of Ir-192 on 04 JAN 2020 and a radiation profile survey was performed. The entire outer surface of the ISORAD-Tc1 was surveyed using the survey meter and the highest exposure rate observed on each surface was recorded on the Shielding Efficiency Test Survey Form labeled with Prototype # 1 BPIC - Square.
To determine if a correction factor was to be applied, ISO-RAD first used the IAEA Safety Guide TS-G-1.1 (ST-2) and ANSI N43.9-2015 Standard Table 1 [B1]. According to Table 1[B1], the correction factor is 1.0 based on the package dimensions and the probe distance. Secondly, ISO-RAD used the Surface Correction Factor (SFC) formula to determine if the distance would match the IAEA and ANSI guidance below:
2
=
1 Where d1 = Center of Source to Drum Surface 201.00036 mm (7.9134 in) d2 =Center of Source to Center of Source Detector 214.2236 mm (8.5276 in)
SFC = d2/d1 SFC = 8.5276/7.9134 SFC = 1.0776 ISO-RAD applied a correction factor of 1.1 for conservatism. See the attached survey sheets with the survey meter readings, extrapolated and distance corrected data.
Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 56 of 70 Shielding Efficiency Test (Pre NCT)
Survey Form Prototype S/N Prototype # 1 Sealed Source S/N Multiple (See List)
Isotope Ir-192 Curie Strength 2317.707__________
Container ISORAD-TC1 BPIC - Square Survey Meter S/N 96219__________
Date of Survey 04 JAN 2020 Surveyor Rafael A. Bustillo_________
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Sv/hr Surface Reading mR/hr Side 1 0.12 12 0.004 0.4 Side 2 0.12 12 0.005 0.5 Side 3 0.12 12 0.004 0.4 Side 4 0.13 13 0.004 0.4 Top 0.06 6 0.002 0.2 Bottom 0.08 8 0.004 0.4 Method to determine extrapolation factor: Max curie strength for container = M; Actual curies strength of test source = A; and Extrapolation factor = E. M = 10125 /A= 2317.707 =E 4.3685418.
E was rounded to 4.369. Extrapolated to 10125 Curies of Ir-192.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.12 X 4.369 = 0.524 _12_ X 4.369 = 52.43 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Side 2 0.12 X 4.369 = 0.524 _12_ X 4.369 = 52.43 0.005 X 4.369 = 0.0219 0.5 X 4.369 = 2.190 Side 3 0.12 X 4.369 = 0.524 _12_ X 4.369 = 52.43 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Side 4 0.13 X 4.369 = 0.568 _13_ X 4.369 = 56.80 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Top 0.06 X 4.369 = 0.262 _6__ X 4.369 = 26.22 0.002 X 4.369 = 0.0087 0.2 X 4.369 = 0.874 Bottom 0.08 X 4.369 = 0.350 _8__ X 4.369 = 34.96 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Distance Correction Factor (From ANSI N43.9 Table 1[B.1]: Distance from center of probe to package surface 1.56 cm x Half Linear Dimension of Package 69.957 = Correction Factor 1.0, but used 1.1 for conservatism.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.524 X 1.1 = 0.576 52.43 X 1.1 = 57.67 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Side 2 0.524 X 1.1 = 0.576 52.43 X 1.1 = 57.67 0.0219 X 1.1 = 0.0241 2.190 X 1.1 = 2.409 Side 3 0.524 X 1.1 = 0.576 52.43 X 1.1 = 57.67 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Side 4 0.568 X 1.1 = 0.625 56.80 X 1.1 = 62.48 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Top 0.262 X 1.1 = 0.288 26.22 X 1.1 = 28.84 0.0087 X 1.1 = 0.0096 0.874 X 1.1 = 0.962 Bottom 0.350 X 1.1 = 0.385 34.96 X 1.1 = 38.46 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 57 of 70 Shielding Efficiency Test (Pre NCT)
Survey Form Prototype S/N Prototype # 2 Sealed Source S/N Multiple (See List)
Isotope Ir-192 Curie Strength 2339.571__________
Container ISORAD-TC1 BPIC 2835A Survey Meter S/N 96219__________
Date of Survey 03 JAN 2020 Surveyor Rafael A. Bustillo_________
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Sv/hr Surface Reading mR/hr Side 1 0.14 14 0.003 0.3 Side 2 0.14 14 0.004 0.4 Side 3 0.14 14 0.004 0.4 Side 4 0.14 14 0.003 0.3 Top 0.034 3.4 0.004 0.4 Bottom 0.14 14 0.004 0.4 Method to determine extrapolation factor: Max curie strength for container = M; Actual curies strength of test source = A; and Extrapolation factor = E. M = 10125 /A= 2339.571 =E 4.3277164.
E was rounded to 4.328. Extrapolated to 10125 Curies of Ir-192.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.14 X 4.328 = 0.606 _14_ X 4.328 = 60.59 0.003 X 4.328 = 0.0123 0.3 X 4.328 = 1.299 Side 2 0.14 X 4.328 = 0.606 _14_ X 4.328 = 60.59 0.004 X 4.328 = 0.0173 0.4 X 4.328 = 1.731 Side 3 0.14 X 4.328 = 0.606 _14_ X 4.328 = 60.59 0.004 X 4.328 = 0.0173 0.4 X 4.328 = 1.731 Side 4 0.14 X 4.328 = 0.606 _14_ X 4.328 = 60.59 0.004 X 4.328 = 0.0173 0.4 X 4.328 = 1.731 Top 0.03 X 4.328 = 0.147 _3.4 X 4.328 = 14.72 0.003 X 4.328 = 0.0123 0.3 X 4.328 = 1.299 Bottom 0.14 X 4.328 = 0.606 _14_ X 4.328 = 60.59 0.004 X 4.328 = 0.0173 0.4 X 4.328 = 1.731 Distance Correction Factor (From ANSI N43.9 Table 1[B.1]: Distance from center of probe to package surface 1.56 cm x Half Linear Dimension of Package 69.957 = Correction Factor 1.0, but used 1.1 for conservatism.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.61 X 1.1 = 0.667 60.59 X 1.1 = 66.65 0.0123 X 1.1 = 0.0143 1.299 X 1.1 = 1.429 Side 2 0.61 X 1.1 = 0.667 60.59 X 1.1 = 66.65 0.0173 X 1.1 = 0.0191 1.731 X 1.1 = 1.904 Side 3 0.61 X 1.1 = 0.667 60.59 X 1.1 = 66.65 0.0173 X 1.1 = 0.0191 1.731 X 1.1 = 1.904 Side 4 0.61 X 1.1 = 0.667 60.59 X 1.1 = 66.65 0.0173 X 1.1 = 0.0191 1.731 X 1.1 = 1.904 Top 0.15 X 1.1 = 0.162 14.72 X 1.1 = 16.20 0.0123 X 1.1 = 0.0143 1.299 X 1.1 = 1.429 Bottom 0.61 X 1.1 = 0.667 60.59 X 1.1 = 66.65 0.0173 X 1.1 = 0.0191 1.731 X 1.1 = 1.904 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 58 of 70 7.1.2 Post NCT Radiation Profile ISORAD-TC1 Transport Package Shielding Efficiency Test Report ISO-RAD Canada Inc Kevin J. Schehr, DBA Managing Director January 16, 2020 Revision 0 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 59 of 70 Shielding Efficiency and Pre NCT Radiation Profile Test Report By: Rafael A. Bustillo Date: 16 JAN 2020
Subject:
Shielding Efficiency Test Post-NCT of ISORAD-TC1 Transport Package Prototype #1 BPIC - Square Inner Container and Prototype #2 BPIC 2835A Inner Container The radiation survey was performed by Rafael A. Bustillo using an NDS ND-2000 Survey Meter S/N 96219, which was calibrated on 03 JAN 2020 with Certificate of Calibration number NDTT-20000301. These profiles are the post-NCT survey to establish the passing the 20% increase pass/fail criteria for NCT testing. In addition, this is the baseline radiation levels for the post-HACT profile.
Prototype #2 was loaded with 2317.707 curies of Ir-192 on 04 JAN 2020 and a radiation profile survey was performed. The entire outer surface of the ISORAD-TC1 was surveyed using the survey meter and the highest exposure rate observed on each surface was recorded on the Shielding Efficiency Test Survey Form labeled with Prototype # 2 BPIC 2835A.
Prototype #1 was loaded with 2317.707 curies of Ir-192 on 04 JAN 2020 and a radiation profile survey was performed. The entire outer surface of the ISORAD-Tc1 was surveyed using the survey meter and the highest exposure rate observed on each surface was recorded on the Shielding Efficiency Test Survey Form labeled with Prototype # 1 BPIC - Square.
To determine if a correction factor was to be applied, ISO-RAD first used the IAEA Safety Guide TS-G-1.1 (ST-2) and ANSI N43.9-2015 Standard Table 1 [B1]. According to Table 1[B1], the correction factor is 1.0 based on the package dimensions and the probe distance. Secondly, ISO-RAD used the Surface Correction Factor (SFC) formula to determine if the distance would match the IAEA and ANSI guidance below:
2
=
1 Where d1 = Center of Source to Drum Surface 201.00036 mm (7.9134 in) d2 =Center of Source to Center of Source Detector 214.2236 mm (8.5276 in)
SFC = d2/d1 SFC = 8.5276/7.9134 SFC = 1.0776 ISO-RAD applied a correction factor of 1.1 for conservatism. See the attached survey sheets with the survey meter readings, extrapolated and distance corrected data.
Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 61 of 70 Shielding Efficiency Test (Post-NCT)
Survey Form Prototype S/N Prototype # 1 Sealed Source S/N Multiple (See List)
Isotope Ir-192 Curie Strength 2317.707__________
Container ISORAD-TC1 BPIC - Square Survey Meter S/N 96219__________
Date of Survey 04 JAN 2020 Surveyor Rafael A. Bustillo_________
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Sv/hr Surface Reading mR/hr Side 1 0.12 12 0.004 0.4 Side 2 0.12 12 0.005 0.5 Side 3 0.12 12 0.004 0.4 Side 4 0.13 13 0.004 0.4 Top 0.06 6 0.002 0.2 Bottom 0.08 8 0.004 0.4 Method to determine extrapolation factor: Max curie strength for container = M; Actual curies strength of test source = A; and Extrapolation factor = E. M = 10125 /A= 2317.707 =E 4.3685418.
E was rounded to 4.369. Extrapolated to 10125 Curies of Ir-192.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.12 X 4.369 = 0.524 _12_ X 4.369 = 52.43 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Side 2 0.12 X 4.369 = 0.524 _12_ X 4.369 = 52.43 0.005 X 4.369 = 0.0219 0.5 X 4.369 = 2.190 Side 3 0.12 X 4.369 = 0.524 _12_ X 4.369 = 52.43 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Side 4 0.13 X 4.369 = 0.568 _13_ X 4.369 = 56.80 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Top 0.06 X 4.369 = 0.262 _6__ X 4.369 = 26.22 0.002 X 4.369 = 0.0087 0.2 X 4.369 = 0.874 Bottom 0.08 X 4.369 = 0.350 _8__ X 4.369 = 34.96 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.750 Distance Correction Factor (From ANSI N43.9 Table 1[B.1]: Distance from center of probe to package surface 1.56 cm x Half Linear Dimension of Package 69.957 = Correction Factor 1.0, but used 1.1 for conservatism.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.524 X 1.1 = 0.576 52.43 X 1.1 = 57.67 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Side 2 0.524 X 1.1 = 0.576 52.43 X 1.1 = 57.67 0.0219 X 1.1 = 0.0241 2.190 X 1.1 = 2.409 Side 3 0.524 X 1.1 = 0.576 52.43 X 1.1 = 57.67 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Side 4 0.568 X 1.1 = 0.625 56.80 X 1.1 = 62.48 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Top 0.262 X 1.1 = 0.288 26.22 X 1.1 = 28.84 0.0087 X 1.1 = 0.0096 0.874 X 1.1 = 0.962 Bottom 0.350 X 1.1 = 0.385 34.96 X 1.1 = 38.46 0.0175 X 1.1 = 0.0193 1.750 X 1.1 = 1.925 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 62 of 70 Shielding Efficiency Test (Post NCT)
Survey Form Prototype S/N Prototype # 2 Sealed Source S/N Multiple (See List)
Isotope Ir-192 Curie Strength 2317.707__________
Container ISORAD-TC1 BPIC 2835A Survey Meter S/N 96219__________
Date of Survey 04 JAN 2020 Surveyor Rafael A. Bustillo_________
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Sv/hr Surface Reading mR/hr Side 1 0.14 14 0.003 0.3 Side 2 0.14 14 0.004 0.4 Side 3 0.14 14 0.004 0.4 Side 4 0.14 14 0.003 0.3 Top 0.032 3.2 0.004 0.4 Bottom 0.14 14 0.004 0.4 Method to determine extrapolation factor: Max curie strength for container = M; Actual curies strength of test source = A; and Extrapolation factor = E. M = 10125 /A= 2317.707 =E 4.3685418.
E was rounded to 4.369. Extrapolated to 10125 Curies of Ir-192.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.14 X 4.369 = 0.612 _14_ X 4.369 = 61.17 0.003 X 4.369 = 0.0131 0.3 X 4.369 = 1.311 Side 2 0.14 X 4.369 = 0.612 _14_ X 4.369 = 61.17 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.748 Side 3 0.14 X 4.369 = 0.612 _14_ X 4.369 = 61.17 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.748 Side 4 0.14 X 4.369 = 0.612 _14_ X 4.369 = 61.17 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.748 Top 0.03 X 4.369 = 0.140 _3.2 X 4.369 = 13.98 0.003 X 4.369 = 0.0131 0.3 X 4.369 = 1.311 Bottom 0.14 X 4.369 = 0.612 _14_ X 4.369 = 61.17 0.004 X 4.369 = 0.0175 0.4 X 4.369 = 1.748 Distance Correction Factor (From ANSI N43.9 Table 1[B.1]: Distance from center of probe to package surface 1.56 cm x Half Linear Dimension of Package 69.957 = Correction Factor 1.0, but used 1.1 for conservatism.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.61 X 1.1 = 0.673 61.17 X 1.1 = 67.29 0.0131 X 1.1 = 0.0143 1.311 X 1.1 = 1.442 Side 2 0.61 X 1.1 = 0.673 61.17 X 1.1 = 67.29 0.0175 X 1.1 = 0.0191 1.748 X 1.1 = 1.923 Side 3 0.61 X 1.1 = 0.673 61.17 X 1.1 = 67.29 0.0175 X 1.1 = 0.0191 1.748 X 1.1 = 1.923 Side 4 0.61 X 1.1 = 0.673 61.17 X 1.1 = 67.29 0.0175 X 1.1 = 0.0191 1.748 X 1.1 = 1.923 Top 0.14 X 1.1 = 0.152 13.98 X 1.1 = 15.18 0.0131 X 1.1 = 0.0143 1.311 X 1.1 = 1.442 Bottom 0.61 X 1.1 = 0.673 61.17 X 1.1 = 67.29 0.0175 X 1.1 = 0.0191 1.748 X 1.1 = 1.923 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 63 of 70 7.1.3 Post HACT Radiation Profile ISORAD-TC1 Transport Package Shielding Efficiency Test Report ISO-RAD Canada Inc Kevin J. Schehr, DBA Managing Director January 16, 2020 Revision 0 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 64 of 70 Shielding Efficiency and Pre NCT Radiation Profile Test Report By: Rafael A. Bustillo Date: 17 JAN 2020
Subject:
Shielding Efficiency Test Post-HACT of ISORAD-TC1 Transport Package Prototype #1 BPIC - Square Inner Container and Prototype #2 BPIC 2835A Inner Container The radiation survey was performed by Rafael A. Bustillo using an NDS ND-2000 Survey Meter S/N 96219, which was calibrated on 03 JAN 2020 with Certificate of Calibration number NDTT-20000301. These profiles are the Post-HACT survey to establish the passing the no dispersal of radioactive contents and less than 1000 mR/hr at 1 meter.
Prototype #1 was loaded with 2274.59 curies of Ir-192 on 06 JAN 2020 and a radiation profile survey was performed. The entire outer surface of the ISORAD-Tc1 was surveyed using the survey meter and the highest exposure rate observed on each surface was recorded on the Shielding Efficiency Test Survey Form labeled with Prototype # 1 BPIC - Square.
Prototype #2 was loaded with 2274.59 curies of Ir-192 on 06 JAN 2020 and a radiation profile survey was performed. The entire outer surface of the ISORAD-TC1 was surveyed using the survey meter and the highest exposure rate observed on each surface was recorded on the Shielding Efficiency Test Survey Form labeled with Prototype # 2 BPIC 2835A.
To determine if a correction factor was to be applied, ISO-RAD first used the IAEA Safety Guide TS-G-1.1 (ST-2) and ANSI N43.9-2015 Standard Table 1 [B1]. According to Table 1[B1], the correction factor is 1.0 based on the package dimensions and the probe distance. Secondly, ISO-RAD used the Surface Correction Factor (SFC) formula to determine if the distance would match the IAEA and ANSI guidance below:
2
=
1 Where d1 = Center of Source to Drum Surface 201.00036 mm (7.9134 in) d2 =Center of Source to Center of Source Detector 214.2236 mm (8.5276 in)
SFC = d2/d1 SFC = 8.5276/7.9134 SFC = 1.0776 ISO-RAD applied a correction factor of 1.1 for conservatism. See the attached survey sheets with the survey meter readings, extrapolated and distance corrected data.
Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 66 of 70 Shielding Efficiency Test (Post-HACT)
Survey Form Prototype S/N Prototype # 1 Sealed Source S/N Multiple (See List)
Isotope Ir-192 Curie Strength 2274.59___________
Container ISORAD-TC1 BPIC - Square Survey Meter S/N 96219__________
Date of Survey 06 JAN 2020 Surveyor Rafael A. Bustillo_________
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Sv/hr Surface Reading mR/hr Side 1 0.12 12 0.005 0.5 Side 2 0.12 12 0.005 0.5 Side 3 0.12 13 0.004 0.4 Side 4 0.13 13 0.004 0.4 Top 0.02 2 0.002 0.2 Bottom 0.09 9 0.005 0.5 Method to determine extrapolation factor: Max curie strength for container = M; Actual curies strength of test source = A; and Extrapolation factor = E. M = 10125 /A= 2274.59 =E 4.4513516.
E was rounded to 4.451. Extrapolated to 10125 Curies of Ir-192.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.12 X 4.451 = 0.534 _12_ X 4.451 = 53.41 0.005 X 4.451 = 0.0222 0.5 X 4.451 = 2.223 Side 2 0.12 X 4.451 = 0.534 _12_ X 4.451 = 53.41 0.005 X 4.451 = 0.0222 0.5 X 4.451 = 2.223 Side 3 0.13 X 4.451 = 0.579 _13_ X 4.451 = 57.86 0.004 X 4.451 = 0.0178 0.4 X 4.451 = 1.780 Side 4 0.13 X 4.451 = 0.579 _13_ X 4.451 = 57.86 0.004 X 4.451 = 0.0178 0.4 X 4.451 = 1.780 Top 0.06 X 4.451 = 0.267 _6__ X 4.451 = 26.71 0.002 X 4.451 = 0.0089 0.2 X 4.451 = 0.890 Bottom 0.09 X 4.451 = 0.401 _9 _ X 4.451 = 40.06 0.005 X 4.451 = 0.0223 0.5 X 4.451 = 2.223 Distance Correction Factor (From ANSI N43.9 Table 1[B.1]: Distance from center of probe to package surface 1.56 cm x Half Linear Dimension of Package 69.957 = Correction Factor 1.0, but used 1.1 for conservatism.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.534 X 1.1 = 0.588 53.41 X 1.1 = 58.75 0.0222 X 1.1 = 0.0245 2.223 X 1.1 = 2.445 Side 2 0.534 X 1.1 = 0.588 53.41 X 1.1 = 58.75 0.0222 X 1.1 = 0.0245 2.223 X 1.1 = 2.445 Side 3 0.579 X 1.1 = 0.636 57.86 X 1.1 = 63.65 0.0178 X 1.1 = 0.0196 1.780 X 1.1 = 1.958 Side 4 0.579 X 1.1 = 0.637 57.86 X 1.1 = 63.65 0.0178 X 1.1 = 0.0196 1.780 X 1.1 = 1.958 Top 0.267 X 1.1 = 0.294 26.71 X 1.1 = 29.38 0.0089 X 1.1 = 0.0098 0.890 X 1.1 = 0.979 Bottom 0.401 X 1.1 = 0.539 40.06 X 1.1 = 44.07 0.0222 X 1.1 = 0.0245 2.223 X 1.1 = 2.445 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 67 of 70 Shielding Efficiency Test (Post-HACT)
Survey Form Prototype S/N Prototype # 2 Sealed Source S/N Multiple (See List)
Isotope Ir-192 Curie Strength 2274.59___________
Container ISORAD-TC1 BPIC 2835A Survey Meter S/N 96219__________
Date of Survey 06 JAN 2020 Surveyor Rafael A. Bustillo_________
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Sv/hr Surface Reading mR/hr Side 1 0.14 14 0.004 0.4 Side 2 0.14 14 0.004 0.4 Side 3 0.16 16 0.006 0.6 Side 4 0.14 14 0.005 0.5 Top 4 4 0.006 0.6 Bottom 0.15 15 0.008 0.8 Method to determine extrapolation factor: Max curie strength for container = M; Actual curies strength of test source = A; and Extrapolation factor = E. M = 10125 /A= 2274.59 =E 4.4513516.
E was rounded to 4.451. Extrapolated to 10125 Curies of Ir-192.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.14 X 4.451 = 0.623 _14_ X 4.451 = 62.31 0.004 X 4.451 = 0.0178 0.4 X 4.451 = 1.780 Side 2 0.14 X 4.451 = 0.623 _14_ X 4.451 = 62.31 0.004 X 4.451 = 0.0178 0.4 X 4.451 = 1.780 Side 3 0.16 X 4.451 = 0.712 _16_ X 4.451 = 71.22 0.006 X 4.451 = 0.0267 0.6 X 4.451 = 2.671 Side 4 0.14 X 4.451 = 0.623 _14_ X 4.451 = 62.31 0.005 X 4.451 = 0.0223 0.5 X 4.451 = 2.226 Top 0.04 X 4.451 = 0.178 _4 X 4.451 = 17.80 0.006 X 4.451 = 0.0267 0.6 X 4.451 = 2.671 Bottom 0.15 X 4.451 = 0.668 _15_ X 4.451 = 66.77 0.008 X 4.451 = 0.0356 0.8 X 4.451 = 3.561 Distance Correction Factor (From ANSI N43.9 Table 1[B.1]: Distance from center of probe to package surface 1.56 cm x Half Linear Dimension of Package 69.957 = Correction Factor 1.0, but used 1.1 for conservatism.
Side of Highest Surface Highest Surface 1 m from Highest 1 m from Highest Container Reading mSv/hr Reading mR/hr Surface Reading Surface Reading Side 1 0.62 X 1.1 = 0.685 62.31 X 1.1 = 68.54 0.0178 X 1.1 = 0.0196 1.780 X 1.1 = 1.958 Side 2 0.62 X 1.1 = 0.685 62.31 X 1.1 = 68.54 0.0178 X 1.1 = 0.0196 1.780 X 1.1 = 1.958 Side 3 0.71 X 1.1 = 0.783 71.22 X 1.1 = 78.34 0.0267 X 1.1 = 0.0294 2.671 X 1.1 = 2.938 Side 4 0.62 X 1.1 = 0.685 62.31 X 1.1 = 68.54 0.0223 X 1.1 = 0.0245 2.226 X 1.1 = 2.449 Top 0.18 X 1.1 = 0.196 17.80 X 1.1 = 19.58 0.0267 X 1.1 = 0.0294 2.671 X 1.1 = 2.938 Bottom 0.67 X 1.1 = 0.735 66.77 X 1.1 = 73.45 0.0356 X 1.1 = 0.0392 3.561 X 1.1 = 3.917 Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 68 of 70 7.2 Test Equipment Drawings:
Figure 136: Penetration Bar Figure 137: Puncture Pin Revision 0 January 29, 2020
Model: ISORAD-TC1 QTP-001 Test Report Project: QTP-001 QTP-001- Page 70 of 70 7.3
References:
[7.2.1] IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material, Revision 1, International Atomic Energy Agency, Vienna, 2018.
[7.2.2] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 2005 Edition, International Atomic Energy Agency, Vienna, 2005.
[7.2.3] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 1996 (Revised),
International Atomic Energy Agency, Vienna, 2000.
[7.2.4] Packaging and Transport of Nuclear Substances Regulations (CNSC PTNS SOR/2015-145), Canadian Nuclear Safety Commission, Ottawa, 2015.
[7.2.5] Title 10, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
[7.2.6] Title 49, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
[7.2.7] ADR European Agreement Concerning the International Carriage of Dangerous Goods by Road. United Nations, New York and Geneva, 2018.
Revision 0 January 29, 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-55 2.12.4 Special Form Certificates for use with Model ISORAD-TC1 Deleted at the request of CNSC Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-56 2.12.5 References
[2.1] IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material, Revision 1, International Atomic Energy Agency, Vienna, 2018.
[2.2] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 2005 Edition, International Atomic Energy Agency, Vienna, 2005.
[2.3] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 1996 (Revised),
International Atomic Energy Agency, Vienna, 2000.
[2.4] Packaging and Transport of Nuclear Substances Regulations (CNSC PTNS SOR/2015-145),
Canadian Nuclear Safety Commission, Ottawa, 2015.
[2.5] Title 10, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
[2.6] Title 49, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
[2.7] Joint Canada - United States Guide for Approval of Type B(U) and Fissile Material Transport Packages (CNSC RD-364 / USNRC NURERG 1886), USNRC and CNSC, Washington & Ottawa, 2009.
[2.8] NUREG CR/3019, UCRL-53044 Recommended Welding Criteria for Use in the Fabrication of Shipping Containers for Radioactive Materials, March 1984.
[2.9] NUREG/CR-3854, Fabrication Criteria for Shipping Containers, U.S. Nuclear Regulatory Commission, Washington D.C., April 1984.
[2.10] NUREG/CR-6407, INEL-95/0551, Classification of Transportation Packaging and Dry Spent Fuel Storage System Components According to Importance to Safety, February 1996.
[2.11] Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels, Revision 1, March 1978.
[2.12] Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, Revision 1, U.S. Nuclear Regulatory Commission, Office of Standards Development, March 1989.
[2.13] Regulatory Guide 7.11, Fracture Toughness Criteria of Base Material for Ferritic Steel Shipping Cask Containment Vessels with a Maximum Wall Thickness of 4 Inch (0.1 m), U.S.
Nuclear Regulatory Commission, Office of Standards Development, June 1991.
[2.14] AWS A2.4 Standard Symbols for Welding, Brazing, and Nondestructive Examination Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 2-57
[2.15] AWS A3.0 Standard Terms and Definitions
[2.16] AWS D1.6/D1.6M Structural Welding Code: Stainless Steel
[2.17] AWS D1.9/D1.6M Structural Welding Code: Titanium
[2.18] Oberg, E., Jones, F., Horton, H., and Ryffel, H., 27th Edition Machinerys Handbook, A Reference Book for the Mechanical Engineer, Designer, Manufacturing Engineer, Draftsman, Toolmaker, and Machinist, 2004 Industrial Press Inc. New York
[2.19] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section II, Part D, Materials, 2001 Edition with Addenda through July 1, 2003.
[2.20] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Division 1, Code Cases: Nuclear Components, Case N-284-1, Metal Containment Shell Buckling Design Methods, Class MC, 2001 Edition with Addenda through July 1, 2003
[2.21] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NB, Class 1 Components, 2001 Edition with Addenda through July 1, 2003.
[2.22] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Division 1, Subsection NF, Supports, 2001 Edition with Addenda through July 1, 2003.
[2.23] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Division 3, Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste, 1977 Edition.
[2.24] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section III, Division 3, Containment Systems for Storage and Transport Packagings of Spent Nuclear Fuel and High Level Radioactive Material and Waste, 2001 Edition with Addenda through July 1, 2003.
[2.25] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section V, Non Destructive Examination, 2001 Edition, with Addenda through July 1, 2001.
[2.26] American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Ferrous Material Specification, 2007 Edition with Addenda through July 1, 2009.
[2.27] American National Standards Institute, for Radioactive Material, Leakage Tests on Packages for Shipment, ANSI N14.5-1997.
[2.28] Vibration, Measurement and Analysis, JD Smith, Butterworth-Heinemann Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-1 Safety Analysis Report Section 3 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-2 Table of Contents Section 3 - THERMAL EVALUATION .............................................................................................. 3-3 3.1 Description of Thermal Design ................................................................................................... 3-3 3.1.1 Design Features .................................................................................................................... 3-3 3.1.2 Decay Heat of Contents ....................................................................................................... 3-4 3.1.3 Summary Tables of Temperatures ....................................................................................... 3-4 3.1.4 Summary Tables of Maximum Pressures ............................................................................ 3-6 3.2 Material Properties and Component Specifications.................................................................... 3-7 3.2.1 Material Properties ............................................................................................................... 3-7 3.2.2 Component Specifications ................................................................................................... 3-8 3.3 General Considerations ............................................................................................................... 3-9 3.3.1 Evaluation by Analysis ........................................................................................................ 3-9 3.3.2 Evaluation by Test ............................................................................................................... 3-9 3.4 Thermal Evaluation Under Normal Conditions of Transport ................................................... 3-10 3.4.1 Heat and Cold .................................................................................................................... 3-10 3.4.2 Temperatures Resulting in Maximum Thermal Stresses ................................................... 3-12 3.4.3 Maximum Normal Operating Pressure .............................................................................. 3-12 3.5 Thermal Evaluation Under Hypothetical Accident Conditions ................................................ 3-12 3.5.1 Initial Conditions ............................................................................................................... 3-12 3.5.2 Fire Test Conditions ........................................................................................................... 3-12 3.5.3 Maximum Temperatures and Pressure............................................................................... 3-12 3.5.4 Temperatures Resulting in Maximum Thermal Stresses ................................................... 3-16 3.5.5 Fuel/Cladding Temperatures for Spent Nuclear Fuel ........................................................ 3-28 3.5.6 Accident Conditions for Fissile Material Packages for Air Transport .............................. 3-28 3.6 Appendix ................................................................................................................................... 3-29 3.6.1 References .......................................................................................................................... 3-29 3.6.2 Hawk Ridge Systems Technical Report SVC-01129 December 2019 .............................. 3-31 List of Tables Table 3.1.A: Selected Summary Table of Inner Container Contents.................................................... 3-4 Table 3.1.B: Summary Table of BPIC Maximum Temperatures of Major Components ...................... 3-4 Table 3.1.C: Summary Table of BPIC Maximum Temperatures of Major Components ...................... 3-5 Table 3.1.D: Summary Table of BPIC Special Form Capsule and DU Shield Temperatures ............. 3-5 Table 3.1.E: Summary Table of Maximum Internal Pressures ............................................................. 3-6 Table 3.2.A: Thermal Properties of Principal Transport Package Materials ..................................... 3-7 Table 3.2.B: Thermal Properties of Special Form Capsules (Containment Vessel) ............................ 3-7 Table 3.4.A: Variable Surface Temperatures by Adjusting the Decay Heat Load ............................. 3-11 Table 3.5.A: NCT Thermal Expansion Diameter ............................................................................... 3-26 Table 3.5.B: HACT Thermal Expansion Diameter ............................................................................. 3-26 Table 3.5.C: NCT Thermal Expansion Length.................................................................................... 3-27 Table 3.5.C: HACT Thermal Expansion Length ................................................................................. 3-27 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-3 Section 3 - THERMAL EVALUATION
[IAEA SSR-6 644, 653-657, 662-664, 728, & 736, IAEA TS-R-1 642, 651-655, 660-662, 728 &
736, and 10 CFR 71.33(b)(5), 71.33(b)(7), 71.43(d), 71.43(g), 71.55(f)(1)(iv), 71.71(c)(1),
71.71(c)(2), & 71.73(c)(4)]
3.1 Description of Thermal Design This section identifies the key thermal design features for the ISORAD-TC1 package. The maximum temperatures at both NCT and HAC conditions have been calculated for these features by a Finite Element Analysis (FEA) and a thermal model of the package. The FEA and the thermal model of the package were validated against thermal analysis of other similar transport containers.
The test procedure and results of the simulation tests and the FEA are presented and discussed within this section.
The ISORAD-TC1 package is a passive thermal device having no mechanical cooling system or relief valves to aid in cooling the package. All cooling of the transport package is through free convection and radiation. The maximum actual heat source is 375 TBq (10,125 Ci) of Iridium-192 (Ir-192) generating 60.88 Watts of decay heat, the thermal analysis was conducted with a heat source of 400.37 TBq (10,810 Ci) of Iridium-192 (Ir-192) generating 65.0 Watts of decay heat. The corresponding decay heat generation rate is approximately 60.88 Watts (actual) versus 65.0 Watts (simulated), which is approximately 6.75% lower heat load providing an additional safety margin for the NCT requirements (See Section 2.6.1, Heat). The design of the ISORAD-TC1 package is that of a non-fissile Type B(U) package and will not be tested under IAEA SSR-6 736, IAEA TS-R-1 736, or 10 CFR 71.55(f)(1)(iv).
The maximum operational temperatures determined for the maximum contents heating have been listed and shown to be lower than the maximum design temperatures of the package.
3.1.1 Design Features The ISORAD-TC1 package is as described in Section 1 and was tested to 65 Watts and will have an operation limit of 60.88 Watts or 375 TBq (10,125 curies) of Ir-192. The Outer Drum and the Cork Assembly provide the Inner Container and the special form capsules with protection from impact and fire. Under HACT fire conditions the Outer Drum skin is designed to heat up very quickly with the Cork Assembly providing insulation to the Inner Container and special form capsules, as a result of its low thermal conductivity and ablation properties. Since heating of the Cork Assembly during the HACT fire may cause gas evolution within the Outer Drum cavity, the Outer Drum is not sealed with gaskets and is open to the atmosphere and pressure build up should not occur. Features uniquely relevant to thermal performance are detailed below.
3.1.1.1 Outer Drum The protective Outer Drum increases distance from the Inner Container surface to the package surface. The Outer Drum also includes cork as insulation and shock absorbing material. The Outer Drum design elevates the main body of the drum off the ground and limits the ground contact to Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-4 only the outside bottom rim. The top Attachment Ring configuration incorporates an air gap of approximately mm ( inch) to aid in heat dissipation.
3.1.1.2 Inner Container The Inner Container uses depleted uranium surrounded by brass or tungsten shielding fully enclosed in a welded austenitic stainless steel square or round structure. The construction with the brass filling the air voids prevents oxidation by severely limiting oxygen from reaching the depleted uranium shield. The tungsten shield versions will not have brass as the threat of possible depleted uranium oxidation will not exist.
The Inner Container provides the shielding and security closure for the special form capsules (containment vessel(s)). The special form capsules provide the primary containment for the radioactive contents.
3.1.2 Decay Heat of Contents [IAEA SSR-6 653, IAEA TS-R-1 651, and 10 CFR 71.33(b)(7)]
The maximum content activity for the ISORAD-TC1 package is 10,125 Ci of Ir-192. The corresponding decay heat generation rate for the content activity is approximately 60.88 Watts (See Table 2.6.1.A). All other contents generate significantly less heat (See Table 2.6.1.A) for complete list of contents.
Table 3.1.A: Selected Summary Table of Inner Container Contents BPIC BPIC 2835A MPIC Isotope mW/Ci Ci TBq W Ci TBq W Ci TBq W Ir-192 6.013 10125 375 60.88 10125 375 60.88 1512 56 9.092 Se-75 5.35 10017 371 53.6 10017 371 53.6 1512 56 8.09 Yb-169 2.51 10017 371 25.15 10017 371 25.15 1512 56 8.09 Lu-177 5.35 10017 371 53.6 10017 371 53.6 N/A N/A N/A See Table 2.6.1.A for other isotopes 3.1.3 Summary Tables of Temperatures The maximum temperatures reached under NCT and HACT conditions have been determined using an FEA thermal model detailed in the Hawk Ridge Technical Report SVC-01129 appended in Section 3.5.2. Table 3.1.B summarizes the results of this report and presents the maximum temperatures reached in the Inner Container cavity and the Iridium Capsule with internal heat loads of 65 Watts under NCT and HACT thermal conditions.
Table 3.1.B: Summary Table of BPIC Maximum Temperatures of Major Components Temperature Condition Outer Inner 1 Iridium/ Insulation 65 Watts Heat Load Drum Container Special Form (Cork)
NCT Decay Heating (38 in shade) 48.82°C 237.51°C 237.9°C 231.16°C NCT Isolation (38 in full sun) 104.1°C 245.55°C 245.97°C 235.53°C HACT Maximum During Fire Test 801.3°C 244.07°C 244.56°C 809.94°C HACT Maximum 30 minutes 210.7°C 247.04°C 247.5°C 364.76°C Post-Fire Test Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-5 1
Note that the special form capsules will not reach 800°C, but theoretically could reach a maximum temperature of 800°C during the HACT thermal test when transporting the maximum activity of 10,125 Ci of Ir-192 if the Inner Container is ejected from the Outer Drum during an accident.
All components, except for the Special Form Capsules, are open to the atmosphere. As such, no pressure will build up in these parts of the package under HACT conditions. As noted in Section 2.7.4, any associated pressure build-ups will be insufficient to detrimentally effect the package containment of the Special Form Capsules.
The maximum temperatures within the Inner Container and special forms are generated at 65 Watts; therefore, the temperatures reached at critical locations with this heat load were calculated in the Hawk Ridge Test Report under NCT and are summarized here in Table 3.1.B and 3.1.C. The maximum temperatures calculated are all within the acceptable temperature limits for the package components.
Table 3.1.C: Summary Table of BPIC Maximum Temperatures of Major Components Temperature NCT Decay NCT Isolation HACT Maximum HACT Maximum Condition Heating (38 in During Fire 30 minutes 65 Watts Heat Load (38 in shade) full sun) Test Post-Fire Test Outer Drum 48.82°C 104.10°C 801.30°C 210.70°C Insulation (Cork) 231.16°C 235.53°C 809.94°C 364.76°C Cork Lid Plate 231.65°C 239.63°C 238.29°C 243.40°C DU Shield 236.51°C 244.45°C 243.07°C 246.10°C Inner Container Shell 231.71°C 239.63°C 238.33°C 242.05°C 1
Iridium/Special Form 237.90°C 245.97°C 244.56°C 247.50°C DU Plug Shield 235.54°C 243.56°C 242.18°C 245.26°C Brass Supports 231.72°C 239.67°C 238.34°C 241.80°C Brass Plug 229.89°C 237.97°C 236.50°C 240.10°C Titanium Plug 237.51°C 245.55°C 244.07°C 247.04°C The temperature of the Inner Container contents under NCT and HACT conditions with contents emitting 65 Watts has been determined from the data in the Hawk Ridge Technical Report as from 1.39°C to 1.42°C above that of the Depleted Uranium shield (See Table 3.1.D): this is based on the worst case assumption that all the heat from the contents, which is emitted as radiation, is absorbed within the special form capsules. The maximum resulting temperatures of the special form capsules calculated are presented in Table 3.1.D: these temperatures are within the acceptable temperature limits for the all the components of the Inner Container.
Table 3.1.D: Summary Table of BPIC Special Form Capsule and DU Shield Temperatures Temperature NCT Decay NCT Isolation HACT Maximum HACT Maximum Condition Heating (38 in During Fire 30 minutes 65 Watts Heat Load (38 in shade) full sun) Test Post-Fire Test 1
Iridium/Special Form 237.90°C 245.97°C 244.56°C 247.50°C DU Shield 236.51°C 244.45°C 243.07°C 246.10°C Difference 1.39°C 1.52°C 1.49°C 1.4°C DU Shield 236.51°C 244.45°C 243.07°C 246.10°C Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-6 Inner Container Shell 231.71°C 239.63°C 238.33°C 242.05°C Difference 4.8°C 4.82°C 4.74°C 4.05°C 1
Iridium/Special Form 237.90°C 245.97°C 244.56°C 247.50°C Inner Container Shell 231.71°C 239.63°C 238.33°C 242.05°C Difference 6.19°C 6.34°C 6.23°C 5.45°C 3.1.4 Summary Tables of Maximum Pressures All package components are open to the atmosphere. As such, no pressure will build up in these components under either NCT or HACT conditions. Normal operating conditions will generate negligible pressure differential within the package. The package can withstand elevated atmospheric pressure because all components except the Special Form Capsules are open to the atmosphere.
Any pressure generated within the Special Form Capsule is significantly below that which would be generated during the HACT thermal test, which is shown in Sections 2.6.3 to result in no loss of structural integrity or containment.
Table 3.1.E: Summary Table of Maximum Internal Pressures Package Fire Conditions Internal Configuration 800 (1,472) Pressure Internal Pressure Limit Developed ISORAD-TC1 BPIC 800 (1,472) 800 (1,472)
Inner Container - Weld Seal 0.3754 MPa (54.45 psi) 4.675 MPa (677.97 psi)
Contents Cavity N/A N/A Open to Air Special Form Capsule 0.3754 MPa (54.45 psi) 4.886 MPa (708.60 psi)
ISORAD-TC1 BPIC 2835A 800 (1,472) 800 (1,472)
Inner Container - Weld Seal 0.3754 MPa (54.45 psi) 4.675 MPa (677.97 psi)
Contents Cavity N/A N/A Open to Air Special Form Capsule 0.3754 MPa (54.45 psi) 4.886 MPa (708.60 psi)
ISORAD-TC1 MPIC 800 (1,472) 800 (1,472)
Inner Container - Weld Seal 0.3754 MPa (54.45 psi) 4.675 MPa (677.97 psi)
Contents Cavity N/A N/A Open to Air Special Form Capsule 0.3754 MPa (54.45 psi) 4.886 MPa (708.60 psi) 1 - Pressure assuming transport of the greatest decay heat generating source term of 10,125 Ci Ir-192.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-7 3.2 Material Properties and Component Specifications 3.2.1 Material Properties Table 3.2.A lists the relevant thermal properties of the important materials in the transport package.
The sources referred to in the last column are listed below the table.
Table 3.2.A: Thermal Properties of Principal Transport Package Materials Material Density Melting/ Thermal Thermal Emissivity Modulus Source kg/m3 Combustion Expansion Conductivity 1
Of (lb/in3) Temperature µm/m°C W/m-°K Elasticity
(µin/in°F) (Btu/h-ft-°F) GPa (Mpsi)
Depleted 19,000 1,130 13.3 27.5 0.15 205 #2, #1, Uranium (0.69) (2,066) (7.4) (16) (30) 6-11 Tungsten 19,270 3,370 4.5 163.3 0.23 400 #1, (0.70) (6,098) (2.5) (1130) (1500°C) (58) 6-51 Brass C360 8500 887°C(1628.6°F) 20.2 100 0.22 100 Brass C260 (0.32) 954°C (1749.2°F) (11.2) (63) (15)
Brass C230 1000°C(1832°F)
Titanium 4621 1,690°C 8.90 11.4 0.31 116 GPa #4, Grade 2 (0.163) (3,074°F) (4.94) (6.6) (16.8) 6-50 Stainless 8000 1,427 17.3 16 0.852 190 #1, Steel - 304 (0.29) (2,600) (9.9) (9.2) 0.9923 (27) 6-11 Cork 154.23 400 40 0.044 0.95 0.02 (0.00544) (752) (22.2) (0.023) (0.0029)
Table 3.2.B: Thermal Properties of Special Form Capsules (Containment Vessel)
Material Density Melting/ Thermal Thermal Emissivity Modulus Source kg/m3 Combustion Expansion1 Conductivity Of (lb/in3) Temperature µm/m°C W/m-°K Elasticity
(µin/in°F) (Btu/h-ft-°F) GPa (Mpsi)
Titanium 4621 1,690°C 8.90 11.4 0.31 116 #4, Grade 2 (0.163) (3,074°F) (4.94) (6.6) (16.8) 6-50 Stainless 8000 1,427 17.3 16 0.36 190 #1, Steel 300 (0.29) (2,600) (9.9) (9.2) (27) 6-11 Series Stainless 7800 1,427 10.3 305 0.36 215 #5, Steel 400 (0.28) (2,600) (5.7) (17) (31) 20 Series Stainless 0.28 1,420°C 11in/in 17.9 0.36 190 #5, Steel (2,588°F) (6.1) (12.4) (27) 10 4PH Vanadium 5953.5 1735°C 7.92 30.7 0.42 160 #7 (0.221) 3155°F (4.4) (17.7) (23.2)
MP35N 0.30 1,315°C 14in/in 11.2 0.23 215 #6, (2,399°F) (6.4) (31) 13 1Note: The thermal expansions of the materials in this table are temperature dependent 2
Note: Weathered Stainless Steel has a higher emissivity of 0.85.
3 Note: Painted Stainless Steel has a higher emissivity Flat White 0.992 and Flat Black 0.92.
4 Note: Cork Thermal Conductivity varies significantly as it absorbs heat (See Table 6) Hawk Ridge Report 5
Note: 400 Series Stainless Steel Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-8 The Cork Assembly will experience temperatures exceeding 800°C during a HACT fire. The best data available on cork under these conditions and under the same use in transport packages is from Croft and Associates Ltd. (Croft). No measurements of cork properties at HACT high temperatures are available. However, the HACT thermal test has been performed on the Safkeg-LS 3979A package as detailed in Crofts report CTR 2009/21, this package uses approximately the same cork specification as the ISORAD-TC1 package design. The ISO-RAD simulation conducted by Hawk Ridge Systems uses the Safkeg-LS and Safkeg-HS thermal report data. The Safkeg data was used to validate the ISO-RAD model and, to demonstrate the acceptability of the thermal properties assumed for the cork. Croft found that, in order to obtain agreement with the measured temperatures, the thermal conductivity of the cork needed to be increased by 50%. Croft validated the thermal properties against an actual furnace test and are effective properties that include any effects of charring and shrinkage of the cork.
The NCT thermal test performed on the ISORAD-TC1 package has been simulated using the Croft cork thermal characteristics applied to the model. As cork is a natural material, a degree of variation in conductivity may well be possible because granular size and shape irregularities exist.
3.2.2 Component Specifications All components are specified and described on the descriptive drawings included in Appendix 1.3.
The components that are important to thermal performance are the Outer Drum, the Cork Assembly, the Inner Container, and special form capsules (containment vessel). The outer shell of Outer Drum and the outer shell of the Inner Container vessel are manufactured from austenitic stainless steel with the special form capsules having a range of materials as described in Table 3.2B and all having been tested to at least 800°C during special form testing.
The allowable service temperatures for all the components cover the maximum and minimum temperatures anticipated during NCT and HACT conditions of transport apart from cork which may combust at temperatures ranging from 400°C to 450°C. The minimum allowable service temperature for all components is less than or equal to -40°C. The maximum service temperature for each component is determined from the temperatures calculated from the thermal model.
The upper NCT temperature reached by the austenitic stainless steel in the Outer Drum is 104.21°C for continuous operations and 801.32°C for short term under HACT conditions. The upper temperature reached by the austenitic stainless steel in the Inner Container 239.63°C for continuous operations and 243.4°C for short term operations under HACT conditions.
The allowable temperature limit for steel when relied upon for structural support is 427°C as specified in USNRC Regulatory Guide 7.6 [3.8 Page 7.6-3]. During the HACT test the temperature of the Outer Drum shell skin exceeds this temperature for a short period of time. During a fire the austenitic stainless steel is providing fire shielding to the cork from the direct exposure of the flames, its main function is not providing structural support therefore the maximum allowable temperature it can reach is 1427°C, which is the melting point of austenitic stainless steel.
The depleted uranium shielding reaches a maximum temperature of 243.07°C during HACT conditions and 246.1°C 30 minutes after the fire conditions. The depleted uranium does not provide Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-9 any structural function therefore it is limited by its melting point of 1130°C. Tungsten shielding would reach a similar temperature, but has a melting point of 3370°C.
The cork is essentially unaffected by temperatures up to 300°C and hardens during temperatures from 320°C until it reaches combustion temperature. The Cork Assembly average temperature during NCT insolation operations is 137.0°C with maximum temperature NCT where small portions of the cork may reach 235.53°C along a thin layer of the cork adjacent to the Inner Container. Under HAC conditions the Cork Assembly reaches a maximum temperature of 809.94°C. Cork chars and self-extinguishes under high temperatures and leaves a low-density carbonaceous layer which provides insulation equivalent to still CO2. The cork is very effective in protecting the Inner Container and the radioactive contents in a HACT event.
The upper temperature reached by the Iridium/special form capsule(s) is 245.97°C for continuous operation (NCT Insolation conditions), and 244.56°C for short term operation (HACT conditions) and 247.50°C 30 minutes after. These temperatures are within the allowable range of the special form materials listed in Table 3.2B.
3.3 General Considerations 3.3.1 Evaluation by Analysis Evaluations by analysis are described in the section they apply to in this Safety Analysis Report or when applicable in the documents contained in Appendix 2.12 and Appendix 3.6.2. The analysis was conducted by Hawk Ridge Systems using Solidworks Thermal Simulation software. The assumptions used were conservative and the wattage used was 65 Watts, which is 6.75% higher than the planned ISORAD-TC1 package wattage limit of 60.88 Watts.
The NCT thermal conditions for the 38°C in the shade test was conservatively conducted in a steady state analysis. For the 38°C with solar insulation test, a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> transient test with twelve hours cycles was used alternating solar insolation and no solar insolation.
The HACT thermal conditions used a transient test analysis for the 30 minutes in an 800°C fire and for a period of time following the removal of the 800°C fire.
The ISORAD-TC1 model used was a simplified version of the actual model. The simplified model removed bolts, other hardware, chamfers, and most spacing from in between parts. The heat source used was 65 Watts in a single austenitic stainless steel special form capsule containing Iridium metal. The decay heat generated was deposited completely as heat within Iridium contents (65 W).
ISO-RAD contends this is a conservative assumption, as in actual use, not all of the decay power would be completely deposited as heat in the source capsule; much of the gamma decay power would be distributed beyond the capsule to be deposited in the uranium or tungsten shield.
3.3.2 Evaluation by Test Evaluations by direct testing are documented in the Test Plans contained in Appendix 2.12 or are described in the section they apply to in this Safety Analysis Report.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-10 3.4 Thermal Evaluation Under Normal Conditions of Transport The ISORAD-TC1 package has been evaluated for compliance with IAEA SSR-6, IAEA TS-R-1and 10 CFR 71 by thermally modeling the package. The thermal model has been validated by comparison against other thermal evaluations on similar Type B transport containers.
3.4.1 Heat and Cold 3.4.1.1 Insolation and Decay Heat Thermal Model The analytical model is described in detail by the Hawk Ridge Test Report SVC-01129 (Section 3.6.2). The thermal evaluation in the report is modeled on the ISORAD-TC1 containing the largest thermal heat source of 65 Watts (10,810 Ci of Ir-192) and is therefore representative of the worst case temperature condition. The maximum surface temperature caused by the effects of solar input and content decay is 104.21°C (219.58°F) approximately located at the center external surface of the Outer Drum Lid. Please note the simplified thermal model presented the Outer Drum Lid Assembly as completely flat and the actual Lid Assembly contains ridges and may slightly lower the external temperature reading. This temperature is well below the maximum service temperature for the materials used in the ISORAD-TC1 transport package components important to safety. This temperature constitutes the most onerous NCT thermal condition. Based on the package materials of construction for components important to safety, this temperature will not be sufficient to adversely affect the containment or shielding integrity as it is well below the maximum service temperature for the materials. As such, the package complies with the requirements of this section.
3.4.1.2 Still Air (shaded) Decay Heating
[IAEA SSR-6 654 & 656, IAEA TS-R-1 652 & 654, and 10 CFR 71.43(g)]
The Hawk Ridge report (see Section 3.6.2) determines the maximum surface temperature of the ISORAD-TC1 package in the shade (i.e., no insolation effects), assuming an ambient temperature of 38°C (100°F), per IAEA SSR-6 654 & 656, IAEA TS-R-1 652 & 654, and 10 CFR 71.43(g).
The thermal evaluation in the Hawk Ridge report is modeled on the ISORAD-TC1 containing the largest thermal heat source of 65 Watts (10,810 Ci of Ir-192) and is therefore representative of the worst case temperature condition. The highest temperature on the ISORAD-TC1 package under these conditions is 48.82°C (119.88), located on the side of the Outer Drum slightly below the center of the drum. The highest temperature on the ISORAD-TC1 in a readily accessible location on the package is 48.82 (119.88). Based on the analysis in the report, it is demonstrated that the surface temperature for any readily accessible surface of the ISORAD-TC1 package in still air at 38 in the shade will not exceed 50. The thermal analysis is also 65 Watts versus 60.88 Watts in actual contents. The actual contents would lower the external temperature to 48.138°C (118.65°F) as presented in Table 3.4.A. below.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-11 aT = 38.026 + (0.1016(W)) mT = 38.026 + (0.1661(W))
Table 3.4.A: Variable Surface Temperatures by Adjusting the Decay Heat Load Decay Heat Load (W) Average Surface T Max Surface T 0.0 38°C 38°C 10.0 (MPIC) 39.04°C (102.27°F) 39.69°C (103.44°F) 30.0 (BPIC) 41.074°C (105.94) 43.009°C (109.42) 45.0 (BPIC) 42.60°C (108.68) 45.501°C (113.90) 60.88 (BPIC) 44.21°C (111.58) 48.138°C (118.65°F) 65.0 (BPIC) 44.63°C (112.34) 48.823°C (119.88) 72.1 (BPIC) 45.35°C (113.63) 50.002°C (122.00) 3.4.1.3 Cold Effected Materials The ISORAD-TC1 does not contain carbon steel components. The components of the ISORAD-TC1 package are not affected by the low Normal Transport temperature (-40). Although austenitic series stainless steel is not known to be susceptible to brittle fracture at -40°C, during testing, shock induced stresses could cause the stainless steel to fail in brittle fracture. As such, to demonstrate that brittle fracture will not occur, the first 9 meter (30 foot) Free Drop will be conducted with the package at or below -40°C at the time of drop. Damage during Normal and Hypothetical Accident Conditions of Transport are described in Section 2.7 of the SAR.
All materials exhibit some contraction due to lower temperatures. However, in this limited temperature range, the ISORAD-TC1 package was not adversely affected as the Prototype #1 passed the Normal and Hypothetical Accident drop, and puncture testing and maintained its structural integrity and shielding.
3.4.1.4 ISORAD-TC1 Series Type B(U) Source Capsule Thermal Analysis This analysis demonstrates that the pressure inside the hypothetical ISORAD-TC1 special form capsule, when subjected to the HACT thermal test, does not exceed the pressure which corresponds to the minimum yield strength at the thermal test temperature.
The source capsules to be used in the ISORAD-TC1 are all special form capsules tested and approved. The thermal test from the HACT requires heating the package externally to a temperature of 800°C for a period of 30 minutes. From the Hawk Ridge Technical Report (see 3.6.2), the special form capsules will reach a maximum temperature of 247.50°C (477.5°F) 30 minutes post fire test when loaded to the maximum decay heat load 65 Watts (10,810 Ci) of Ir-192 during the thermal test.
Hawk Ridge Technical Report SVC-01129 demonstrated that the special form capsule will withstand the stresses induced by the thermal test when loaded to the maximum source term and will retain their integrity under the Thermal test conditions. Therefore, it is concluded that the container and contents meet the requirements of this section.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-12 3.4.2 Temperatures Resulting in Maximum Thermal Stresses The temperature and pressure variations described in Sections 3.4.1 and 3.4.3 will not adversely affect the transport package during normal transport since the melting temperatures of all safety critical components are well above these temperatures and the package will experience no pressures sufficient to cause package failure. It is therefore concluded that the ISORAD-TC1 package will maintain its structural integrity and shielding effectiveness under the normal transport thermal stress conditions.
3.4.3 Maximum Normal Operating Pressure All package components except for the special form capsules are open to the atmosphere. As such, pressure will not build up in these parts of packages during NCT conditions. This condition is not time dependent once steady state is achieved. However, the following assumes the Inner Container and Contents Cavity were gasketed for conservatism. If the Content Cavity is sealed under the lowest temperatures (-40°C) and then allowed to heat up to the highest temperature (245.97°C from Section 3.1.3.) a small pressure differential will be created.
The result is a pressure differential of 18.26 psi. All other package components, except for the special form capsules, will exhibit a pressure differential of 0 psi as they are open to the atmosphere with no means for creating a pressure differential. Hawk Ridge Technical Report SVC-01129 (see 3.6.2) demonstrated the source will maintain its containment integrity under HACT Thermal test conditions which created a pressure differential of 68.49 psi (See Section 3.5.2.3). The increased pressure differential of 18.26 psi will have no adverse impact on the source containment under NCT operating pressure conditions. No other contributing gas sources are present.
3.5 Thermal Evaluation Under Hypothetical Accident Conditions 3.5.1 Initial Conditions The ISORAD-TC1 package is assumed to start at the NCT solar insolation temperature values with an external value of 104.21°C (219.58°F) on the Outer Drum and an internal value of 245.97°C (474.75°F) in the Special Form Capsule. The Inner Container is not expected to reach the thermal test temperature of 800°C (1,472°F). As demonstrated in the Hawk Ridge Technical Report SVC-01129 (see 3.6.2), the maximum temperature for the special form capsule during the HACT thermal test condition when loaded to the maximum heat load is 244.56°C (472.21°F).
3.5.2 Fire Test Conditions Not applicable. Actual NCT and HACT Fire Tests were not conducted. The tests were conducted via computer simulation.
3.5.3 Maximum Temperatures and Pressure See Sections 3.1.3 and 3.1.4.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-13 3.5.3.1 Temperatures The cork in the ISORAD-TC1 package will be partially destroyed when subjected to the HACT thermal test conditions. The other package materials, however, are suitable for use at 800 (1,472) (see Table 3.2.A). The depleted uranium, which is susceptible to oxidation, is surrounded by brass and enclosed within austenitic stainless steel and would not be exposed to oxygen. The transport package will undergo no loss of structural integrity or shielding. The pressures generated have been demonstrated in Section 3.5.3.2 to be less than the yield strength of the material and will not adversely affect the package integrity.
Based on the minimal damage caused by the mechanical destructive testing, no breach of the weldment occurred, therefore no air path to the shield was created which could cause uranium oxidation and resultant loss of effective shielding around the sources.
It is concluded that the ISORAD-TC1 package therefore meets the requirements of the fire test based on an analysis of its construction and the condition of the test units after the preceding destructive testing.
3.5.3.2 Pressure All package components except for the special form capsules are open to the atmosphere. As such, pressure will not build up in other package components during the Hypothetical Accident conditions.
If the Special Form Capsule is sealed under the lowest temperatures (-40°C) and then allowed to heat up until the HACT thermal test temperature of 247.5°C (477.5°C) a small pressure differential will be created.
Using the Ideal Gas Law and equating for two standard scenarios the result is:
P1/T1 = P2/T2 Substituted: P2 = (P1/T1) x T2 Where: P1 = Ambient pressure at sealing = 14.87 psi T1 = -40°C = 233°K P2 = Pressure at temperature (245.97)
T2 = 247.5 = 520.65°K P2 = (P1/T1) x T2 P2 = (14.87/233) x 520.65 P2 = 0.0638197 x 520.65 P2 = 33.227726 Which is a pressure differential of 18.36 psi. No other contributing gas sources are present. See Section 2.7.4.3.
If the Special Form Capsule hypothetically reaches 800°C, the maximum internal pressure can be found by assuming that the internal temperature will reach 800°C (1,470°F) under the HACT Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-14 thermal test conditions. Using the ideal gas law and requiring the air to occupy a constant volume, the internal air pressure differential could reach 68.49 psi shown as follows:
Using the Ideal Gas Law and equating for two standard scenarios the result is:
P1/T1 = P2/T2 Substituted: P2 = (P1/T1) x T2 Where: P1 = Ambient pressure at sealing = 14.87 psi T1 = -40 (233.0°K)
P2 = Pressure at temperature (800)
T2 = 800 (1073.15°K)
P2 = (P1/T1) x T2 P2 = (14.87/233) x 1073.15 P2 = 0.0638197 x 1073.15 P2 = 68.488111 or P1/T1 = P2/T2 Substituted: P2 = (P1/T1) x T2 Where: P1 = Ambient pressure after NCT Solar Insolation = 33.227726 psi T1 = 245.97 (520.65°K)
P2 = Pressure at temperature (800)
T2 = 800 (1073.15°K)
P2 = (P1/T1) x T2 P2 = (33.227726/520.65) x 1073.15 P2 = 0.0638196 x 1073.15 P2 = 68.488003 Therefore, the internal pressure within the Special Form Capsule is 68.49 psi. The stress area of the M6 x 1 bolt is 0.03318 in2. Multiplying this by 4 (number of Plug Assembly bolts) gives a total stress area of the bolts as 0.1327136 in2. Therefore, the maximum stress on each bolt is 217.05 lbs/0.13272 in2 = 1,635.4766 psi. At a temperature of 870 (1,600), the yield strength of type 316 stainless steel is 10,000 psi (
Reference:
Department of Defense Aerospace Structural Metals Handbook, Metals and Ceramics Information Center, Battelle, 1991 Edition). Thus, the maximum stress in the bolts is less than 17% of their yield strength Which is a pressure differential of 35.26 psi. The maximum stress would occur in the four M6 x 1 bolts securing the Contents Cavity lid.
The maximum stress is calculated by:
= F/A Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-16 Therefore, the force on all 4 bolts is (54.2614 psi)(4 in2) = 217.05 lbs. The stress area of the M6 x 1 bolt is 0.03318 in2. Multiplying this by 4 (number of Plug Assembly bolts) gives a total stress area of the bolts as 0.1327136 in2. Therefore, the maximum stress on each bolt is 217.05 lbs/0.13272 in2
= 1,635.4766 psi. At a temperature of 870 (1,600), the yield strength of type 316 stainless steel is 10,000 psi (
Reference:
Department of Defense Aerospace Structural Metals Handbook, Metals and Ceramics Information Center, Battelle, 1991 Edition). Thus, the maximum stress in the bolts is less than 17% of their yield strength at 870.
This analysis demonstrates that the maximum internal gas pressure at 800 (1,472) would be 54 psi. Under these conditions, the maximum stress in the capsule would be less than the yield strength of the material.
3.5.4 Temperatures Resulting in Maximum Thermal Stresses The temperature and pressure variations described in Sections 3.4.1 and 3.4.3 will not adversely affect the transport package during normal transport since the melting temperatures of all safety critical components are well above these temperatures and the package will experience no pressures sufficient to cause package failure. The thermal stresses introduced under the hypothetical accident condition criteria were evaluated in Section 3.5.3 and again demonstrated to be insufficient to cause package failure. It is therefore concluded that the ISORAD-TC1 package will maintain its structural integrity and shielding effectiveness under the hypothetical accident condition transport thermal stress conditions.
There will be no thermal stress as a result of differential thermal expansion of the materials of the Inner Container. The construction of the Inner Container positions the uranium shield by surrounding by Brass Spacers resting on austenitic stainless steel parts. Brass has a larger linear coefficient of thermal expansion (11.2 µin/in°F) than Uranium (7.4 µin/in°F). Therefore, as the temperature increases from 20°C to 800°C, the clearance between these parts will increase.
Similarly, austenitic stainless steel has a larger linear coefficient of thermal expansion (9.9
µin/in°F) is slightly less than Brass (11.2 µin/in°F). Therefore, as the temperature increases from 20°C to 800°C, the clearance between these parts will decrease, but sufficient tolerance exists to not create stress. Consequently, there will be no thermal stress as a result of differential thermal expansion of the materials of the Inner Container.
The construction of the Inner Container Shield Plug Assembly positions the uranium shield by surrounding with Titanium with a Brass Plug Spacer on top of the Depleted Uranium Plug Shield is a Titanium Plug Lid. Brass has a larger linear coefficient of thermal expansion (11.2 µin/in°F) than Uranium (7.4 µin/in°F). Therefore, as the temperature increases from 20°C to 800°C, the clearance between these parts will increase. The Titanium has a smaller linear coefficient of thermal expansion (4.94 µin/in°F) than Brass (11.2 µin/in°F) or Uranium (7.4 µin/in°F). Therefore, as the temperature increases from 20°C to 800°C, the clearance between these parts will decrease, but sufficient tolerance exists to not create stress. Consequently, there will be no thermal stress as a result of differential thermal expansion of the materials of the Inner Container Shield Plug Assembly.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-28 At a temperature of 800°C, the surface of the uranium shield would be expected to oxidize.
However, the air-filled space between the depleted uranium shield and the housing has a volume of approximately cm3 ( liter). For analysis, we assume this volume is filled with air at atmospheric pressure. This is a conservative assumption, as during the seal welding process, the temperature of the austenitic stainless steel, and therefore the contained air, would be elevated, thereby reducing the initial pressure. However, for conservativism, no account is taken of this reduced pressure. Assuming air occupies 22.4 L/mol at atmospheric pressure, and further assuming that oxygen represents only 20% of the volume of air, this volume would contain 1.19 x 1021 molecules of O2 or 2.38 x 1021 atoms of oxygen.
The metal uranium forms several oxides:
- Triuranium octoxide (U3O8, the most stable uranium oxide, yellowcake typically contains 70 to 90 percent Triuranium octoxide)
- Uranyl peroxide (UO4)
ISO-RAD conservatively assumed that UO2 would be the only oxide formed in the Inner Container, UO2 is the most "uranium-rich" oxide. ISO-RAD will also conservatively assume, all oxygen atoms are consumed in the formation of uranium oxides, then the 2.38 x 1021 atoms of oxygen would consume 1.19 x 1021 atoms of uranium, which would represent approximately 0.469 grams.
Assuming the 0.469 grams of uranium were removed uniformly over only the outside surface of the depleted uranium shield (1052 cm2), then the reduction of shield thickness would be less than 0.5
µm. For Iridium-192, this would represent an increase in the exposure rates of less than 0.010%, far below the HACT requirement of less than 10 mSv/hr (1000 mR/hr) permitted by the regulations.
Outer Drum The Outer Drum is enclosed, but is not sealed. Therefore, there will be no pressure build up in this part of the container during the HACT thermal test.
3.5.5 Fuel/Cladding Temperatures for Spent Nuclear Fuel Not applicable. This package is not used for transport of spent nuclear fuel.
3.5.6 Accident Conditions for Fissile Material Packages for Air Transport Not applicable. This package is not used for transport of Type B quantities of fissile material.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-29 3.6 Appendix 3.6.1 References
[3.1] IAEA SSR-6, Regulations for the Safe Transport of Radioactive Material, Revision 1, International Atomic Energy Agency, Vienna, 2018.
[3.2] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 2005 Edition, International Atomic Energy Agency, Vienna, 2005.
[3.3] IAEA TS-R-1, Regulations for the Safe Transport of Radioactive Material, 1996 (Revised),
International Atomic Energy Agency, Vienna, 2000.
[3.4] Packaging and Transport of Nuclear Substances Regulations (CNSC PTNS SOR/2015-145),
Canadian Nuclear Safety Commission, Ottawa, 2015.
[3.5] Title 10, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
[3.6] Title 49, Code of Federal Regulations, Part 71, Office of the Federal Register, Washington D.C.
[3.7] Joint Canada - United States Guide for Approval of Type B(U) and Fissile Material Transport Packages (CNSC RD-364 / USNRC NURERG 1886), USNRC and CNSC, Washington & Ottawa, 2009.
[3.8] Regulatory Guide 7.6, Design Criteria for the Structural Analysis of Shipping Cask Containment Vessels, Revision 1, March 1978.
[3.9] Regulatory Guide 7.8, Load Combinations for the Structural Analysis of Shipping Casks for Radioactive Material, Revision 1, U.S. Nuclear Regulatory Commission, Office of Standards Development, March 1989.
[3.10] Holman, J. P., Heat Transfer, 10th Edition, McGraw-Hill, New York 2010.
[3.11] Summary of the Physical Properties and Composition of Resin Bonded Cork, CTR 2001/11, Issue D, 2002.
[3.12] The Emissivity of Various Materials Commonly Encountered in Industry, Land pyrometers Technical Note 101.
[3.13] Advisory Material for the IAEA Regulations for the Safe Transport of Radioactive Material, 2005 Edition, IAEA Safety Guide No. TS-G-1.1 (Rev. 1), 2008.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-30 3.6.1.1 References Table 3.2.B
- 1. Eugene A. Avallone and Theodore Baumeister III, Mark's Standard Handbook for Mechanical Engineers, Tenth Edition, New York: McGraw-Hill, 1996. Pages 6-7, 6-11, 6-50, 6-51,
- 2. Lowenstein, Paul. Industrial Uses of Depleted Uranium. American Society for Metals. Metals Handbook, Volume 3, Ninth Edition.
- 3. Smith, L. P., The Language of Rubber: An Introduction to the Specification and Testing of Elastomers, Oxford: Butterworth-Heinemann, 1993.
- 4. Marks Standard Handbook for Mechanical Engineers, Tenth Edition, E.A. Avallone, T.
Baumeister. Page 6-50
- 5. Stainless Steel: Tables of Technical Properties, Materials and Applications Series, Volume 5.
Euro Inox: The European Stainless Steel Development Association. Luxembourg 2007.
- 6. SPS Technologies Data Sheet. Page 6
- 7. Vanadium http://www-ferp.ucsd.edu/LIB/PROPS/PANOS/v.html
- 8. ASM Specialty Handbook Stainless Steels, Ed. J.R. Davis, 1994. Page 10 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 3-31 3.6.2 Hawk Ridge Systems Technical Report SVC-01129 December 2019 Revision 0 29 February 2020
Analysis Services Report Analysis Consulting Service SVC-01129 Vision Hawk Ridge Systems Analysis Consulting Service provides insight into engineering problems by applying the SOLIDWORKS Simulation suite of software.
The following is a summary of the results obtained from the application of SOLIDWORKS Simulation suite of software to the customers engineering problem.
Project Description Thermal flow analysis of an iridium material container following the U.S.NRC thermal testing requirements.
Prepared For Kevin J. Schehr, DBA ISO-RAD Canada Inc.
Herein referred to as the customer.
Prepared By Paolo Olmos Analysis Services Engineer Hawk Ridge Systems Herein referred to as the service provider.
Reviewed By Glenn Whyte Product Manager Hawk Ridge Systems Prepared On Date: 12-30-2019 2300 Contra Costa Blvd, Suite 630 Pleasant Hill, CA 94523 l www.hawkridgesys.com US:1.877.266.4469 l CANADA: 1.866.587.6803 Pleasant Hill
- Mountain View
- Fairfield
- Woodland Hills
- Ventura
- Ontario
- Costa Mesa
- Redmond
- Portland Reno
- Las Vegas
- Winnipeg
- Toronto
- Richmond
- Edmonton
- Calgary
- Minneapolis
- Brillion
- Cincinnati Page 1 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Table of Contents Vision....................................................................................................................................................................... 1 Project Description .................................................................................................................................................. 1 Prepared For ........................................................................................................................................................... 1 Prepared By ............................................................................................................................................................ 1 Reviewed By ........................................................................................................................................................... 1 Prepared On............................................................................................................................................................ 1 Table of Contents.................................................................................................................................................... 2 1 Report Details .................................................................................................................................................. 3 2 Results Summary ............................................................................................................................................ 4 3 Results Discussion .......................................................................................................................................... 6 4 Analysis Details ............................................................................................................................................... 7 4.1 Analysis Description ................................................................................................................................. 7 4.2 List of Key Assumptions......................................................................................................................... 13 5 Results ........................................................................................................................................................... 14 5.1 Condition 1: Steady State Test .............................................................................................................. 14 5.2 Condition 2: Solar Test........................................................................................................................... 15 5.3 Condition 3 Fire Test .............................................................................................................................. 20 Appendix A: Terms ............................................................................................................................................... 25 A.1 Review of Engineering Principles .......................................................................................................... 25 A.2 Internal and Confidential Material .......................................................................................................... 25 A.3 References and Publicity ....................................................................................................................... 25 Appendix B: Definitions......................................................................................................................................... 26 Page 2 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 1 Report Details Background Information:
ISOFLEX Radioactive LLC has an iridium material container that they are trying to evaluate under various thermal testing conditions to prevent leaks. They are looking to evaluate their design under the U.S.NRC thermal test requirements for conditions 1-3.
Figure 1: Test requirements Goal(s):
- 1. Condition 1: Perform a steady-state natural convection heat transfer analysis with an internal heat load from the radioactive material.
- 2. Condition 2: Perform a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> transient analysis. 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with insolation and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> without insolation.
- 3. Condition 3: Perform a transient analysis simulating the container engulfed in fire for 30 minutes then turned off, tracking temperatures through 60 minutes.
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Analysis Services Report Analysis Consulting Service SVC-01129 2 Results Summary Below is a summary of the maximum temperature results for the thermal flow analyses performed based on the requirements for the U.S.NRC thermal test conditions 1-3. See Section 4 for detailed results which include the minimum and average temperatures, temperature cut plots, as well as temperature time history plots.
Table 1 below shows the maximum temperatures reached for the condition 1 steady state test of each component and surface of interest included in the analysis model.
Melt Max Component Material Temperature Temperature Drum Outside Surface 300 Series Stainless Steel 1427 °C 48.82 °C Uranium Shield Uranium (Depleted) 1130 °C 236.51 °C Uranium Plug Uranium (Depleted) 1130 °C 235.54 °C Brass Plug Brass C360 887 °C 229.89 °C Insulation Cork 400 °C 231.16 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 231.65 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 231.71 °C Brass Support Brass C360 887 °C 231.72 °C Titanium Plug Titanium Grade 2 1668 °C 237.51 °C Iridium Iridium 2450 °C 237.90 °C Table 1: Condition 1-Steady state test maximum temperatures The results from the condition 1 steady state test show that the outside surface of the drum reaches a maximum temperature of 48.82 °C, therefore satisfying the requirement for condition 1, no accessible surface of a package would have a temperature exceeding 50 °C in a nonexclusive use shipment. The other components are also well below the melting and combustion temperature specifications of their assigned materials.
Table 2 below shows the maximum temperatures reached for the condition 2 solar test of each component and surface of interest included in the analysis model. The maximum temperatures provided are at the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> time periods.
Melt Max Temperature Max Temperature Component Material Temperature (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />)
Drum Outside Surface 300 Series Stainless Steel 1427 °C 104.21 °C 49.53 °C Uranium Shield Uranium (Depleted) 1130 °C 244.45 °C 242.83 °C Uranium Plug Uranium (Depleted) 1130 °C 243.56 °C 241.90 °C Brass Plug Brass C360 887 °C 237.97 °C 236.19 °C Insulation Cork 400 °C 235.53 °C 233.83 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 239.63 °C 237.95 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 239.63 °C 237.95 °C Brass Support Brass C360 887 °C 239.67 °C 238.00 °C Titanium Plug Titanium Grade 2 1668 °C 245.55 °C 243.92 °C Iridium Iridium 2450 °C 245.97 °C 244.35 °C Table 2: Condition 2-Solar test maximum temperatures Page 4 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 The results from the condition 2 solar test show that the temperatures reached under ambient conditions (still air temperature at 38 °C and solar insolation conditions) will meet the containment and shielding requirements when comparing against the melting and combustion temperatures for each component.
Table 3 below shows the maximum temperatures reached for the condition 3 fire test of each component and surface of interest included in the analysis model. The maximum temperatures provided are at the 30 minute and 60 minute time periods.
Melt Max Temperature Max Temperature Component Material Temperature (30 minutes) (60 minutes)
Drum Outside Surface 300 Series Stainless Steel 1427 °C 801.32 °C 210.70 °C Uranium Shield Uranium (Depleted) 1130 °C 243.07 °C 246.10 °C Uranium Plug Uranium (Depleted) 1130 °C 242.18 °C 245.26 °C Brass Plug Brass C360 887 °C 236.50 °C 240.10 °C Insulation Cork 400 °C 809.94 °C 364.76 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 238.29 °C 243.40 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 238.33 °C 242.05 °C Brass Support Brass C360 887 °C 238.34 °C 241.80 °C Titanium Plug Titanium Grade 2 1668 °C 244.07 °C 247.04 °C Iridium Iridium 2450 °C 244.56 °C 247.50 °C Table 3: Condition 3-Fire test maximum temperatures The results from the condition 3 fire test show that the outside surface of the drum reaches a maximum temperature of 801.32 °C, which is well below the melting temperature of the assigned material. This does not fully predict weather or not the escape of radioactive material will occur since a thermal stress analysis was not performed. The actual test requires calculating the most severe thermal stress conditions that result during the fire test and subsequent cooldown. Another result to point out is that the insulation cork material has a combustion temperature of 400 °C but the maximum temperature reached in testing is more than double at 809.94 °C. The combustion temperature however is not through the entire thickness of the insulation. The temperature profile can be seen in Figure 2. Note that the red indicates temperatures equal to or above the cork combustion temperature of .400 °C.
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Analysis Services Report Analysis Consulting Service SVC-01129 Figure 2: Condition 3-Fire test temperature cut plot through the front plane at 30 minutes, the maximum temperatures shown have been adjusted to the 400 °C combustion temperature of cork 3 Results Discussion In initial runs of the steady state test, it was discovered that the primary factor influencing the outer temperature of the shell is the rate at which energy is lost from the outer surfaces. With a 63W applied heat load, in the steady-state condition, that amount of energy needs to be dissipated from the shell of the container surface, independent of the internal construction of the vessel. The main factors influencing this are the convective rate off the surface of the container, which cannot be altered in this test, and the radiation emissivity of the surface. Based on these factors we implemented an emissivity of 0.91, and used a final emissivity of 0.992, which both resulted in passing results.
In initial runs of the solar test, we used a fixed value for the thermal conductivity of the cork material - at 0.04 W/m-°K. This treated the cork as a very effective heat insulator, and makes it difficult for the internal heat generated to get out, resulting in high temperatures over 400 °C in the iridium core, and in the interior layers of the cork insulation. After reviewing this assumption, it was decided to move to a temperature-dependent conductivity value for the cork. This technique had been used in previous test protocols shared by the customer, and while this choice is less conservative, it more closely reflects the real-life behavior of that material, resulting in more realistic temperature predictions inside the container.
It was also found that for all studies, the ceramic thermal blanket was having a negative effect on the internal temperatures. While it was providing some shielding to resist the external temperatures during the solar and fire tests, it was also keeping heat in and contributing slightly to making the internal temperatures too high. It was therefore decided to eliminate the ceramic thermal blanket from the analysis model.
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Analysis Services Report Analysis Consulting Service SVC-01129 4 Analysis Details 4.1 Analysis Description This section defines the full extent of the analysis or analyses that was performed.
Software and Versions Used:
SOLIDWORKS and SOLIDWORKS Flow Simulation 2019 SP5.0.
3D Model:
HRS created a simplified version of the container assembly from 180831-101 Alternate Drum Assembly Bulk and PIC.SLDASM, which was provided by the customer.
Figure 3: Model provided by the customer with the global orientation showing isometric view and a section view through the front plane Simplifications included, but were not limited to:
o Removed threads and bolts from the assembly.
o Filled gaps between parts and combined components in contact and of the same material.
o Removed small features such as chamfers from components.
Figure 4: Simplified model created by HRS with the global orientation showing isometric view and a section view through the front plane Page 7 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Figure 5: Components in the thermal model Study Definition(s):
Study 1 (Condition 1): External Steady-state natural convection and radiation heat transfer analysis with an internal heat load from the radioactive material.
Study 2 (Condition 2): 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> transient analysis, 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with the heat flux on and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> with the heat flux off.
Study 3 (Condition 3): A transient analysis simulating the container engulfed in fire for 30 minutes then turned off, tracking temperatures through 60 minutes.
Fluid(s):
SOLIDWORKS Flow Simulation default air at 101325 Pa was used.
Air Properties Specific heat ratio: 1.399 Molecular mass: 0.02896 kg/mol Figure 6: Thermal conductivity, dynamic viscosity, and specific heat of air Page 8 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Material(s)
Material assignments for each component in the thermal model are shown in Table 4 below:
Component Material Drum 300 Series Stainless Steel Uranium Shield Uranium (Depleted)
Uranium Plug Uranium (Depleted)
Brass Plug Brass C360 Insulation Cork Stainless Steel Sleeve 300 Series Stainless Steel Bulk Steel Capsule 300 Series Stainless Steel Brass Support Brass C360 Titanium Plug Titanium Grade 2 Iridium Iridium Table 4: Thermal properties Material properties are shown in Table 5 below:
Material Density Thermal Conductivity Specific Heat Melting Point Emissivity 300 Series Stainless Steel 8000 kg/m^3 16 W/m-°K 500 J/kg °K 1427 °C 0.992*
Brass C360 8500 kg/m^3 100 W/m-°K 390 J/kg °K 887 °C ** 0.22 Titanium Grade 2 4621 kg/m^3 11.4 W/m-°K 522.3 J/kg °K 1668 °C 0.31 Uranium (Depleted) 19000 kg/m^3 27.5 W/m-°K 120 J/kg °K 1130 °C 0.15 Cork 154.224 kg/m^3 See Table 6*** 2000 J/kg °K 400 °C**** 0.95 Iridium 22650 kg/m^3 147 W/ m-°K 1300 J/kg °K***** 2450 °C 0.87*****
Table 5: Thermal properties
- Flat White Paint with an emissivity of 0.992 is applied on the exterior of the drum.
- Used brass C360 as worse case, but C260 and C230 have melting temperatures of 954 °C and 1000 °C respectively.
- From Croft 3977 Type B Thermal Analysis ML12277A403.pdf, p.23.
- Used 400 °C as worse case, but the combustion range for cork is 400-450°C.
- From report ML15014A130.pdf provided by the customer for iridium.
Temperature Thermal Conductivity
-0.15 °C 0.055 W/ m-°K 32.85 °C 0.06 W/ m-°K 87.85 °C 0.071 W/ m-°K 124.85 °C 0.0777 W/ m-°K 146.85 °C 0.081 W/ m-°K Table 6: Thermal conductivity for cork Page 9 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Boundary Conditions:
Study 1: Steady State Test o Quarter symmetry was used.
o Temperature of the air was 38.0 °C.
o Gravity equal to 9.81 m/s^2 was applied in the negative Y direction.
o The component highlighted in blue in Figure 7 has a total heat generation rate of 65W (16.25W applied to the quarter symmetric model).
o Heat transfer due to radiation was considered.
Emissivity on outer surfaces of the drum was 0.992.
Figure 7: Component generating 65W Study 2-Solar Test o Quarter symmetry was used.
o Boundary conditions and initial conditions from Study 1 were transferred.
o Time dependency was enabled for a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> total time with a 0.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time step, with results saved every 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
o A heat flux of 800 W/m^2 was applied to the top lid external surface following the time curve in Figure 8.
o A heat flux of 400 W/m^2 was applied to the side external drum surfaces following the time curve in Figure 8.
o Heat transfer due to radiation was considered.
Emissivity on outer surfaces of the drum was 0.992.
Figure 8: Heat flux time curve over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Page 10 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Study 3-Fire Test o Quarter symmetry was used.
o Boundary conditions from Study 1 were transferred.
o Initial conditions came from 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> results from study 2 o Temperature of the air was at a maximum of 800 °C and reached a minimum of 38.0 °C, following the time curve in Figure 9.
o Time dependency was enabled for a 60 minute total time with a 0.2 minute time step, with results saved every 2 minutes.
o The heat transfer coefficient on the external surface of the lid and drum has a maximum of 56.1 W/m^2*K at 800 °C, following the time curve in Figure 9.
o Heat transfer due to radiation was considered.
Emissivity on outer surfaces of the drum was 0.992.
Figure 9: Heat transfer coefficient and temperature time curve over 60 minutes Idealization Solar radiation is represented by a heat flux applied to the external surfaces of the model.
The drum being engulfed in flames is represented by the air temperature and heat transfer coefficient on the external surfaces of the model.
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Analysis Services Report Analysis Consulting Service SVC-01129 Mesh and Numerical Accuracy:
Higher numerical accuracy was used. Best practices were used to determine the size of the computational domain and mesh size.
Higher physical accuracy was used. Some simplifications to the model were made and the results are based on the assumptions used.
Figure 10: Computational domain from the top view and mesh through the front plane Page 12 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 4.2 List of Key Assumptions The following is a best effort to define the key assumptions that are associated with the proposed analysis or analyses.
Components excluded from the analysis did not significantly affect the results.
The surfaces are perfectly smooth.
The components are free from manufacturing defects.
Components did not deform, expand or contract due to thermal or flow effects.
Materials are isotropic.
Forced convection was not considered (i.e. cooling from the wind).
Heat input from fire is approximated as a heat transfer coefficient relative to a defined temperature.
Solar radiation is approximated as a heat flux (does not account for cloud cover, intensity or position of the sun).
The ground was not included in the analyses.
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Analysis Services Report Analysis Consulting Service SVC-01129 5 Results 5.1 Condition 1: Steady State Test Melt Min Average Max Component Material Temperature Temperature Temperature Temperature Outer Drum Surface 300 Series Stainless Steel 1427 °C 40.48 °C 44.63 °C 48.82 °C Uranium Shield Uranium (Depleted) 1130 °C 230.00 °C 231.84 °C 236.51 °C Uranium Plug Uranium (Depleted) 1130 °C 229.87 °C 231.16 °C 235.54 °C Brass Plug Brass C360 887 °C 229.74 °C 229.83 °C 229.89 °C Insulation Cork 400 °C 44.02 °C 116.52 °C 231.16 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 217.93 °C 227.45 °C 231.65 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 224.12 °C 229.81 °C 231.71 °C Brass Support Brass C360 887 °C 229.53 °C 230.81 °C 231.72 °C Titanium Plug Titanium Grade 2 1668 °C 228.88 °C 230.59 °C 237.51 °C Iridium Iridium 2450 °C 236.35 °C 237.37 °C 237.90 °C Table 7: Condition 1-Steady state test temperatures Figure 11: Condition 1-Steady state temperature plot through the front plane Page 14 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 5.2 Condition 2: Solar Test Melt Min Average Max Component Material Temperature Temperature Temperature Temperature Outer Drum Surface 300 Series Stainless Steel 1427 °C 51.26 °C 78.76 °C 104.21 °C Uranium Shield Uranium (Depleted) 1130 °C 238.10 °C 239.81 °C 244.45 °C Uranium Plug Uranium (Depleted) 1130 °C 238.00 °C 239.21 °C 243.56 °C Brass Plug Brass C360 887 °C 237.84 °C 237.92 °C 237.97 °C Insulation Cork 400 °C 54.18 °C 137.00 °C 235.53 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 228.12 °C 236.00 °C 239.63 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 233.61 °C 237.89 °C 239.63 °C Brass Support Brass C360 887 °C 237.66 °C 238.80 °C 239.67 °C Titanium Plug Titanium Grade 2 1668 °C 237.07 °C 238.69 °C 245.55 °C Iridium Iridium 2450 °C 244.54 °C 245.48 °C 245.97 °C Table 8: Condition 2-Solar test temperatures at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Figure 12: Condition 2-Solar test temperature plot at 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> through the front plane Page 15 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Melt Min Average Max Component Material Temperature Temperature Temperature Temperature Outer Drum Surface 300 Series Stainless Steel 1427 °C 40.63 °C 44.80 °C 49.53 °C Uranium Shield Uranium (Depleted) 1130 °C 236.31 °C 238.13 °C 242.83 °C Uranium Plug Uranium (Depleted) 1130 °C 236.21 °C 237.47 °C 241.90 °C Brass Plug Brass C360 887 °C 236.05 °C 236.13 °C 236.19 °C Insulation Cork 400 °C 43.90 °C 118.31 °C 233.83 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 224.69 °C 233.86 °C 237.95 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 231.13 °C 236.08 °C 237.95 °C Brass Support Brass C360 887 °C 235.82 °C 237.09 °C 238.00 °C Titanium Plug Titanium Grade 2 1668 °C 235.19 °C 236.90 °C 243.92 °C Iridium Iridium 2450 °C 242.90 °C 243.86 °C 244.35 °C Table 9: Condition 2-Solar test temperatures at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Figure 13: Condition 2-Solar test temperature plot at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> through the front plane Page 16 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Condition 2-Solar Test Minimum Solid Temperature Time History Plot 245.00 240.00 235.00 230.00 225.00 220.00 215.00 210.00 205.00 200.00 195.00 190.00 185.00 180.00 175.00 170.00 165.00 160.00 Outer Drum Surface 155.00 Uranium Shield Temperature, °C 150.00 Uranium Plug 145.00 Brass Plug 140.00 Insulation 135.00 130.00 Stainless Steel Sleeve 125.00 Bulk Steel Capsule 120.00 Brass Sleeve Support 115.00 Titanium Plug 110.00 105.00 Iridium 100.00 95.00 90.00 85.00 80.00 75.00 70.00 65.00 60.00 55.00 50.00 45.00 40.00 35.00 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Physical Time, Hours Figure 14: Condition 2-Solar test minimum solid temperature time history plot Page 17 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Condition 2-Solar Test Average Solid Temperature Time History Plot 245.00 240.00 235.00 230.00 225.00 220.00 215.00 210.00 205.00 200.00 195.00 190.00 185.00 180.00 175.00 170.00 165.00 Outer Drum Surface 160.00 Uranium Shield 155.00 Temperature, °C 150.00 Uranium Plug 145.00 Brass Plug 140.00 Insulation 135.00 Stainless Steel Sleeve 130.00 Bulk Steel Capsule 125.00 Brass Sleeve Support 120.00 115.00 Titanium Plug 110.00 Iridium 105.00 100.00 95.00 90.00 85.00 80.00 75.00 70.00 65.00 60.00 55.00 50.00 45.00 40.00 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Physical Time, Hours Figure 15: Condition 2-Solar test average solid temperature time history plot Page 18 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Condition 2-Solar Test Maximum Solid Temperature Time History Plot 250.00 245.00 240.00 235.00 230.00 225.00 220.00 215.00 210.00 205.00 200.00 195.00 190.00 185.00 180.00 175.00 170.00 Outer Drum Surface 165.00 Uranium Shield Temperature, °C 160.00 Uranium Plug 155.00 150.00 Brass Plug 145.00 Insulation 140.00 Stainless Steel Sleeve 135.00 Bulk Steel Capsule 130.00 125.00 Brass Sleeve Support 120.00 Titanium Plug 115.00 Iridium 110.00 105.00 100.00 95.00 90.00 85.00 80.00 75.00 70.00 65.00 60.00 55.00 50.00 45.00 0 1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 Physical Time, Hours Figure 16: Condition 2-Solar test maximum solid temperature time history plot Page 19 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 5.3 Condition 3 Fire Test Melt Min Average Max Component Material Temperature Temperature Temperature Temperature Outer Drum Surface 300 Series Stainless Steel 1427 °C 794.31 °C 797.40 °C 801.32 °C Uranium Shield Uranium (Depleted) 1130 °C 236.68 °C 238.45 °C 243.07 °C Uranium Plug Uranium (Depleted) 1130 °C 236.53 °C 237.75 °C 242.18 °C Brass Plug Brass C360 887 °C 236.37 °C 236.45 °C 236.50 °C Insulation Cork 400 °C 167.54 °C 392.41 °C 809.94 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 230.51 °C 235.26 °C 238.29 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 233.57 °C 236.59 °C 238.33 °C Brass Support Brass C360 887 °C 236.42 °C 237.49 °C 238.34 °C Titanium Plug Titanium Grade 2 1668 °C 235.63 °C 237.23 °C 244.07 °C Iridium Iridium 2450 °C 243.07 °C 244.05 °C 244.56 °C Table 10: Condition 3-fire test temperatures at 30 minutes Figure 17: Condition 3-Fire test temperature plot at 30 minutes through the front plane Page 20 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Melt Min Average Max Component Material Temperature Temperature Temperature Temperature Outer Drum Surface 300 Series Stainless Steel 1427 °C 129.43 °C 155.14 °C 210.70 °C Uranium Shield Uranium (Depleted) 1130 °C 240.32 °C 241.89 °C 246.10 °C Uranium Plug Uranium (Depleted) 1130 °C 240.13 °C 241.19 °C 245.26 °C Brass Plug Brass C360 887 °C 240.00 °C 240.05 °C 240.10 °C Insulation Cork 400 °C 152.23 °C 276.08 °C 364.76 °C Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 237.31 °C 241.00 °C 243.40 °C Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 239.51 °C 240.61 °C 242.05 °C Brass Support Brass C360 887 °C 240.27 °C 241.15 °C 241.80 °C Titanium Plug Titanium Grade 2 1668 °C 239.63 °C 240.83 °C 247.04 °C Iridium Iridium 2450 °C 246.10 °C 247.02 °C 247.50 °C Table 11: Condition 3-fire test temperatures at 60 minutes Figure 18: Condition 3-Fire test temperature plot at 60 minutes through the front plane Page 21 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Condition 3-Fire Test Minimum Solid Temperature Time History Plot 800.00 780.00 760.00 740.00 720.00 700.00 680.00 660.00 640.00 620.00 600.00 580.00 560.00 540.00 520.00 500.00 Outer Drum Surface 480.00 Uranium Shield Temperature, °C 460.00 Uranium Plug 440.00 Brass Plug 420.00 Insulation 400.00 Stainless Steel Sleeve 380.00 360.00 Bulk Steel Capsule 340.00 Brass Sleeve Support 320.00 Titanium Plug 300.00 Iridium 280.00 260.00 240.00 220.00 200.00 180.00 160.00 140.00 120.00 100.00 80.00 60.00 40.00 0.00 5.00 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 50.00 55.00 60.00 Time, Minutes Figure 19: Condition 3-Fire test minimum solid temperature time history plot Page 22 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Condition 3-Fire Test Average Solid Temperature Time History Plot 820.00 800.00 780.00 760.00 740.00 720.00 700.00 680.00 660.00 640.00 620.00 600.00 580.00 560.00 540.00 Outer Drum Surface 520.00 Uranium Shield Temperature, °C 500.00 Uranium Plug 480.00 Brass Plug 460.00 Insulation 440.00 Stainless Steel Sleeve 420.00 Bulk Steel Capsule 400.00 Brass Sleeve Support 380.00 Titanium Plug 360.00 Iridium 340.00 320.00 300.00 280.00 260.00 240.00 220.00 200.00 180.00 160.00 140.00 120.00 0.00 5.00 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 50.00 55.00 60.00 Time, Minutes Figure 20: Condition 3-Fire test average solid temperature time history plot Page 23 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Condition 3-Fire Test Maximum Solid Temperature Time History Plot 820.00 800.00 780.00 760.00 740.00 720.00 700.00 680.00 660.00 640.00 620.00 600.00 580.00 Outer Drum Surface 560.00 Uranium Shield 540.00 Temperature, °C Uranium Plug 520.00 Brass Plug 500.00 Insulation 480.00 Stainless Steel Sleeve 460.00 Bulk Steel Capsule 440.00 Brass Sleeve Support 420.00 Titanium Plug 400.00 Iridium 380.00 360.00 340.00 320.00 300.00 280.00 260.00 240.00 220.00 200.00 180.00 0.00 5.00 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 50.00 55.00 60.00 Time, Minutes Figure 21: Condition 3-Fire test maximum solid temperature time history plot Page 24 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Appendix A: Terms A.1 Review of Engineering Principles Hawk Ridge Systems Analysis Consulting Services are performed by degreed mechanical engineers with extensive knowledge of SOLIDWORKS Simulation software. The engineer will take utmost care to follow sound engineering practices, make appropriate assumptions, and communicate any assumptions and limitations of the solution method clearly to the customer. However, the customer retains sole responsibility for the final engineering review of the validity of the results, and any decisions about subsequent design choices based on these results.
A.2 Internal and Confidential Material The customer will identify all confidential material as such when presented to Hawk Ridge Systems. Hawk Ridge Systems will, by default, treat all material provided by the customer as internal information, not to be distributed outside of Hawk Ridge Systems without the customers consent. Hawk Ridge Systems will abide by the terms of any confidentiality/privileged information agreements currently in place between the customer and Hawk Ridge Systems.
A.3 References and Publicity The customer consents to the use by the service provider of the customers name, together with descriptions of any work performed, in the service providers promotional material, and the promotional material of Hawk Ridge Systems partners. Additionally, the customer agrees that upon request, it will inform clients and prospective clients of Hawk Ridge Systems that Hawk Ridge Systems has performed its obligations under the statement of work.
Page 25 of 26 - V1
Analysis Services Report Analysis Consulting Service SVC-01129 Appendix B: Definitions Analysis: Application of CAE (Computer Aided Engineering) tools such as CFD (Computational Fluid Dynamics) software and/or FEA (Finite Element Analysis) software to solve engineering problems.
Numerical Accuracy:
Standard: Best practices will be followed with regard to numerical parameters, i.e. mesh size, time step, number of included frequencies, etc.
Higher: Two variance tests will be run on each of the numerical parameters, to check that the results are trending towards convergence in the area of interest.
Highest: Variance tests will be run on each of the numerical parameters until results convergence is achieved in the area of interest. A validation example can be provided by the customer to ensure the setup of the problem is behaving as expected.
Physical Accuracy:
Standard: Model will be simplified as necessary to balance result accuracy with solve time. Best practices will be followed for removal and/or simplification of bodies/parts/assemblies. Some assumptions will be made, but results between analyses will be comparable. The accuracy of results is based on the assumptions made.
Higher: Model will be kept similar to the actual physical product, but some simplifications will be made to non-essential areas i.e. thread removal from fasteners. Minimal assumptions will be made. The accuracy of results is based on the assumptions made.
Highest: Model will be kept as close to the actual physical product as possible. Minimal assumptions will be made. The accuracy of results is based on the assumptions made. Any deviation from the actual physical model and its potential effects on the results will be researched and documented.
Page 26 of 26 - V1
Analysis Services Report Addendum Analysis Consulting Services SVC-01129 Notes The results were compiled using SOLDIWORKS 2020 SP2.0.
Solar Test Results The revised Table 2 below shows the maximum temperatures reached for the condition 2 solar test of each component and surface of interest included in the analysis model. The maximum temperatures provided are at the 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the maximum from all time steps, and at 180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br />. The temperature difference provided is between the maximum at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and the maximum found across all time steps for all components inside the drum and insulation.
Max Max Max Melt Temperature Max Temperature Component Material Temperature Temperature Temperature Temperature Difference (Across All Time Steps)
(12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) (24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) (180 hours0.00208 days <br />0.05 hours <br />2.97619e-4 weeks <br />6.849e-5 months <br />) 105.50°C at Drum Outside Surface 300 Series Stainless Steel 1427 °C 104.10°C 49.55°C 104.45°C 174.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 251.41°C at Uranium Shield Uranium (Depleted) 1130 °C 244.47°C 242.84°C 8.57°C 250.93°C 157.80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> 250.50°C at Uranium Plug Uranium (Depleted) 1130 °C 243.57°C 241.92°C 8.58°C 250.03°C 157.70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> 244.87°C at Brass Plug Brass C360 887 °C 237.99°C 236.21°C 8.66°C 244.43°C 157.60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> 242.33°C at Insulation Cork 400 °C 235.55°C 233.84°C 241.88°C 157.70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> 246.54°C at Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 239.64°C 237.97°C 8.57°C 246.09°C 157.70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> 246.54°C at Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 239.64°C 237.97°C 8.57°C 246.09°C 157.70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> 246.59°C at Brass Sleeve Support Brass C360 887 °C 239.69°C 238.02°C 8.57°C 246.13°C 157.70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br /> 252.50°C at Titanium Plug Titanium Grade 2 1668 °C 245.56°C 243.94°C 8.56°C 252.02°C 157.80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> 252.92°C at Iridium Iridium 245.98°C 244.37°C 8.55°C 252.44°C 158.00 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> TABLE 2: CONDITION 2- SOLAR TEST MAXIMUM TEMPERATURES
The plots provided below are for the components inside the Drum.
Uranium Shield Maximum Temperature 254 252 250 248 Temperature (Solid) [°C]
246 244 242 240 238 236 234 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Uranium Plug Maximum Temperature 252 250 248 246 Temperature (Solid) [°C]
244 242 240 238 236 234 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Brass Plug Maximum Temperature 246 244 242 240 Temperature (Solid) [°C]
238 236 234 232 230 228 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Insulation Maximum Temperature 244 242 240 238 Temperature (Solid) [°C]
236 234 232 230 228 226 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Stainless Steel Sleeve Maximum Temperature 248 246 244 242 Temperature (Solid) [°C]
240 238 236 234 232 230 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Bulk Steel Capsule Maximum Temperature 248 246 244 242 Temperature (Solid) [°C]
240 238 236 234 232 230 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Brass Sleeve Support Maximum Temperature 248 246 244 242 Temperature (Solid) [°C]
240 238 236 234 232 230 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Titanium Plug Maximum Temperature 254 252 250 248 Temperature (Solid) [°C]
246 244 242 240 238 236 0 20 40 60 80 100 120 140 160 180 200 Physical time [h]
Iridium Maximum Temperature 254.00 252.00 250.00 248.00 Temperature (Solid) [°C]
246.00 244.00 242.00 240.00 238.00 236.00 0.00 20.00 40.00 60.00 80.00 100.00 120.00 140.00 160.00 180.00 200.00 Time [h]
Fire Test Results The revised Table 3 below shows the maximum temperatures reached for the condition 3 fire test of each component and surface of interest included in the analysis model. The maximum temperatures provided are at the 30 minute, 60 minute, the maximum from all time steps, and at 360 minutes. The temperature difference provided is between the maximum at 60 minutes and the maximum found across all time steps for all components inside the drum and insulation.
Max Max Max Melt Temperature Max Temperature Component Material Temperature Temperature Temperature Temperature Difference (Across All Time Steps)
(30 minutes) (60 minutes) (360 minutes) 801.79°C at Drum Outside Surface 300 Series Stainless Steel 1427 °C 801.73°C 211.68°C 55.37°C 26.00 minutes 266.27°C at Uranium Shield Uranium (Depleted) 1130 °C 245.19°C 248.61°C 17.66°C 265.71°C 279.00 minutes 265.38°C at Uranium Plug Uranium (Depleted) 1130 °C 244.32°C 247.80°C 17.58°C 264.81°C 278.40 minutes 259.66°C at Brass Plug Brass C360 887 °C 238.78°C 242.75°C 16.91°C 259.01°C 272.60 minutes 813.49°C at Insulation Cork 400 °C 810.95°C 375.71°C 259.68°C 22.00 minutes 261.37°C at Stainless Steel Sleeve 300 Series Stainless Steel 1427 °C 240.45°C 247.07°C 14.3°C 260.77°C 275.40 minutes 261.41°C at Bulk Steel Capsule 300 Series Stainless Steel 1427 °C 240.49°C 244.62°C 16.79°C 260.81°C 275.60 minutes 261.42°C at Brass Sleeve Support Brass C360 887 °C 240.50°C 244.34°C 17.08°C 260.82°C 275.60 minutes 267.29°C at Titanium Plug Titanium Grade 2 1668 °C 246.18°C 249.55°C 17.44°C 266.73°C 279.40 minutes 267.77°C at Iridium Iridium 246.66°C 250.00°C 17.77°C 267.22°C 280.00 minutes TABLE 3: CONDITION 3- FIRE TEST MAXIMUM TEMPERATURES
The plots provided below are for the components inside the Drum.
Uranium Shield Maximum Temperature 270.00 265.00 260.00 Temperature (Solid) [°C]
255.00 250.00 245.00 240.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Uranium Plug Maximum Temperature 270.00 265.00 260.00 Temperature (Solid) [°C]
255.00 250.00 245.00 240.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Brass Plug Maximum Temperature 265.00 260.00 255.00 Temperature (Solid) [°C]
250.00 245.00 240.00 235.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Insulation Maximum Temperature 900.00 800.00 700.00 600.00 Temperature (Solid) [°C]
500.00 400.00 300.00 200.00 100.00 0.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Stainless Steel Sleeve Maximum Temperature 265.00 260.00 255.00 Temperature (Solid) [°C]
250.00 245.00 240.00 235.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Bulk Steel Capsule Maximum Temperature 265.00 260.00 255.00 Temperature (Solid) [°C]
250.00 245.00 240.00 235.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Brass Sleeve Support Maximum Temperature 265.00 260.00 255.00 Temperature (Solid) [°C]
250.00 245.00 240.00 235.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Titanium Plug Maximum Temperature 270.00 265.00 260.00 Temperature (Solid) [°C]
255.00 250.00 245.00 240.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Physical time [min]
Iridium Maximum Temperature 270.00 265.00 260.00 Temperature (Solid) [°C]
255.00 250.00 245.00 240.00 0.00 50.00 100.00 150.00 200.00 250.00 300.00 350.00 400.00 Time [min]
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 4-1 Safety Analysis Report Section 4 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 4-2 Table of Contents Section 4 - CONTAINMENT ...........................................................................................................4-3 4.1.1 Special Requirements for Damaged Spent Nuclear Fuel ......................................................4-3 4.2 Containment Under Normal Conditions of Transport .......................................................4-3 4.3 Containment Under Hypothetical Accident Condition .......................................................4-4 4.4 Leakage Rate Tests for Type B Packages .............................................................................4-5 4.5 Appendix ..................................................................................................................................4-5 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 4-3 Section 4 - CONTAINMENT
[IAEA SSR-6 648, 658, 660, & 661, IAEA TS-R-1 646, 656, 658, & 659, 10 CFR 71.43(f), and 10 CFR 71.51]
4.1 Description of the Containment System The ISORAD-TC1 package containment system is the Special Form Capsule(s) that contains the radioactive contents. This source capsule shall be qualified as Special Form radioactive material under IAEA SSR-6, IAEA TS-R-1, 49 CFR 173, or other equivalent regulatory body.
The containment for the radioactive contents is a seal welded 300 Series stainless steel, 400 Series stainless steel, titanium, or vanadium certificated as a Special Form Capsule. The Special Form Capsule may have diameters ranging up to 13.75 mm (0.541 in) and have wall thicknesses as small as 0.25 mm (0.009843 in). The welded radioactive source capsule, certified as special form radioactive material, must possess a valid special form certificate of Competent Authority issued under conformance to IAEA SSR-6, IAEA TS-R-1 or 49 CFR Part 173 standard requirements.
In addition, a valid and current leak test analysis is required before transport of the special form in the ISORAD-TC1. Meeting special form criteria provides the primary containment of the radioactive isotope being transported.
There are no penetrations of the containment. There are no drain or fill ports, valves, seals, test ports, pressure relief devices, lids, cover plates or other closure devices. The source capsules are seal welded using a weld process that has been demonstrated to be acceptable for certification of the radioactive source capsule as Special Form radioactive material under an IAEA Certificate of Competent Authority.
There is no mechanical closure as the containment is seal welded.
4.1.1 Special Requirements for Damaged Spent Nuclear Fuel Not applicable. This package is not used for the transport of spent nuclear fuel.
4.2 Containment Under Normal Conditions of Transport
[IAEA SSR-6 658(a), IAEA TS-R-1 656(a), 10 CFR 71.51(a)(1), 10 CFR 71.71, 10 CFR 71.75, and 49 CFR 173.469]
The seal welded and certified Special Form Capsules used in conjunction with the ISORAD-TC1 package have satisfied the requirements for the special form radioactive material as prescribed in IAEA SSR-6, IAEA TS-R-1, 10 CFR 71.75, and 49 CFR 173.469. There will be no release of radioactive material under the NCT.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 4-4 Pressurization of the source capsules and contents cavity under the conditions of the HACT thermal test resulted in stresses below the yield strength of the Inner Container Contents Lid bolts as provided in Section 2.6.1.3.
The NCT criteria listed in IAEA SSR-6 719-724, IAEA TS-S-1 719-724, and 10 CFR 71.71 will result in no loss of transport package containment as prescribed in IAEA SSR-6 658(a), IAEA TS-R-1 656(a), and 10 CFR 71.51(a)(1). This conclusion is based on calculations and data provided in Sections 2 and 3 of the SAR. The certified Special Form Capsules have demonstrated the leak-tightness specified in this section (<1 µPa m3 s-1 as prescribed by ISO 9918 and IAEA TS-R-1). This testing provides sufficient demonstration that the containment criteria of 10CFR71.51 for normal conditions of transport have been satisfied. This demonstrates that there will be no release of radioactive material content under the NCT.
There will be no generation of combustible gases within the special form capsule and the package will not transport fissile material.
4.3 Containment Under Hypothetical Accident Condition
[IAEA SSR-6 658(b), IAEA TS-R-1 656(b), and 10 CFR 71.51(a)(2)]
The HACT requirements outlined in IAEA SSR-6 726-729, IAEA TS-R-1 726-729, and 10 CFR 71.73 will result in no loss of transport package containment. This conclusion is based on information presented in Section 2.7 and Section 3.4.
The pressurization of the Special Form Capsules and package under HACT accident conditions was determined to have no detrimental effect on the capsules ability to maintain containment. In addition, the Inner Container cover bolts provide an additional measure of security in ensuring pressurization of the package under the accident conditions. The containment (Special Form Capsule) will withstand the pressure variations of transport. Sections 2.7 and 3.5 of the SAR demonstrate the transport package meets the containment requirements of IAEA SSR-6 658(b), IAEA TS-R-1 656(b), and 10 CFR 71.51(a)(2).
The Special Form Capsules used as containment in this package must be certified as special form radioactive material as delineated in IAEA SSR-6, IAEA TS-R-1, and 10 CFR 71.
The performance of the most vulnerable special form source capsule under the HACT thermal condition is described in Section 2.6.4 and 2.7.6. Under the conditions of the HACT, the maximum differential pressure stress generated in the worst case Special Form Capsule is less than 15% of the yield strength of the austenitic stainless steel at the test temperature. The maximum longitudinal pressure stress in the Special Form Capsule is less than 8% of the yield strength of the austenitic stainless steel at the test temperature.
No thermal stress will result from differential thermal expansion of the materials of the Special Form Capsule as the capsule is fabricated from a single material. In each special form capsule design, there is sufficient clearance between the contents and the capsule to preclude any thermal stress as a result of differential thermal expansion of the radioactive contents against the material of the Special Form Capsule.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 4-5 4.4 Leakage Rate Tests for Type B Packages
[IAEA SSR-6 658, IAEA TS-R-1 656, 10 CFR 71.51, and ISO9978:1992(E)]
The primary containment for the radioactive material contents in the ISORAD-TC1 packages are the radioactive Special Form Capsules. Only certified Special Form Capsules are authorized for Type B transport in the ISORAD-TC1. The certification must be under IAEA SSR-6, TS-R-1, 10 CFR Part 71, or 49 CFR Part 173. The Special Form Capsule must be certified leakage free after manufacture, and again once every six months thereafter prior to transport, the source capsules are leak tested in accordance with ISO9978:1992(E) (or more recent editions) to ensure that containment of the source does not allow release of more than 0.005 Ci of radioactive material. These fabrication and periodic tests ensure that contamination release from the package does not exceed the regulatory limits.
The ISORAD-TC1 package does not have leakage test requirements as it does not contain seals and is open to the atmosphere.
Reference ISO9978:1992(E) - Radiation Protection - Sealed Radioactive Sources -
Leakage Test Methods.
4.5 Appendix No Additional Documents.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-1 Safety Analysis Report Section 5 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-2 Table of Contents Section 5 - SHIELDING EVALUATION ............................................................................................... 5-3 5.1 Description of Shielding Design ............................................................................................... 5-3 5.1.1. Design Features .................................................................................................................. 5-3 5.1.2 Summary Table of Maximum Radiation Levels ................................................................ 5-3 5.2 Source Specification ...................................................................................................................... 5-5 5.2.1 Gamma Source ........................................................................................................................ 5-5 5.2.2 Neutron Source .................................................................................................................. 5-5 5.3 Shielding Model ........................................................................................................................ 5-5 5.3.1 Configuration of Source and Shielding.............................................................................. 5-5 5.3.2 Material Properties ............................................................................................................. 5-5 5.4 Shielding Evaluation ................................................................................................................. 5-6 5.4.1 Methods.............................................................................................................................. 5-6 5.4.2 Input and Output Data ........................................................................................................ 5-6 5.4.3 Flux-to-Dose-Rate Conversion .......................................................................................... 5-7 5.4.4 External Radiation Levels .................................................................................................. 5-7 5.5 Appendix ................................................................................................................................... 5-8 5.5.1 Profile Measurements of the ISORAD-TC1 Containing Ir-192. ....................................... 5-8 5.5.2 Calculated Measurements of the ISORAD-TC1 containing Se-75 and Yb-169 ............. 5-13 List of Tables Table 5.1.A: ISORAD-TC1 BPIC Prototype #1 Summary Table of External Radiation Levels ............ 5-4 Table 5.1.B: ISORAD-TC1 BPIC 2835A Prototype #2 Summary Table of External Radiation Levels 5-4 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-3 Section 5 - SHIELDING EVALUATION
[IAEA SSR-6 523, 526-528, 572, 647, 648(b), & 658(b)(ii)(i), IAEA TS-R-1 526, 530-532, 572, 645, 646(b), & 656(b)(ii)(i), 10 CFR 71.47, and 10 CFR 71.51(a)(1)]
The shielding design was confirmed to be adequate for the overall package design by actual direct measurements using radiation profiles from the prototype shield, and by actual measurements of resulting radiation levels after the numerous tests performed for NCT and HACT conditions.
Extrapolation of the curies used to perform the survey versus the ISORAD-TC1 package curies limit and the distance correction factor for the survey meter probe. Theoretical calculations were used to evaluate the ISORAD-TC1 shielding for the Se-75 and the Yb-169 and the results are located in Appendix 5.5.2.
5.1 Description of Shielding Design 5.1.1. Design Features The principal shielding in the ISORAD-TC1 package is the depleted uranium or tungsten shield assembly and optional tungsten insert(s) used when necessary to obtain allowable dose rates as well as fix source capsule locations within the shield.
The BPIC Shield is either depleted uranium (DU) or tungsten and has a minimum shield thickness of approximately 62.0 mm (2.44in) in every direction. The shield is approximately +/- mm
( in +/- in) high and mm +/- mm ( in +/- in) diameter. The DU version has a titanium inner cylinder or tungsten shield has an inner titanium or stainless steel cylinder that has an interior dimension of approximately mm +/- mm ( in +/- in) diameter and mm +/-
mm ( in +/- in) tall. The DU shield rests on a brass bottom support, a brass cylinder that slips over the DU shield and has a brass top shield support which provides a eutectic barrier which prevents stainless steel and DU interaction. The brass creates stability and a eutectic barrier with prevents stainless steel and DU interaction. The tungsten version eliminates most brass pieces.
The BPIC Plug Assembly encloses the top cavity in a DU or tungsten plug shield. The BPIC Plug Assembly includes a DU or Tungsten Plug Shield that is encased in titanium and is inserted into the upper portion of the interior cavity of the DU or tungsten shield. In between the DU or tungsten Plug Shield and the Plug Lid is a brass spacer. The BPIC Plug Assembly is secured to the Top Plate of the BPIC with four stainless steel M6 x 1 bolts.
The MPIC Shield Assembly creates the source cavities/channels for the source assemblies to be stored or transported. The MPIC Shield Assembly comprises two major parts the MPIC Base Insert and the MPIC Inner Insert and are constructed of either depleted uranium or tungsten. The tube configurations are either one circle of ten tubes or a circle within a circle of five tubes each. The MPIC Shield Assembly is approximately mm +/- mm ( in +/- in) in diameter and mm +/-
mm ( in +/- in) tall. The approximate weight is 50.8 kg (111.99 pounds).
5.1.2 Summary Table of Maximum Radiation Levels Table 5.1.A and 5.1.B include radiation profile data obtained from the ISORAD-TC1 packages that were tested to the Normal and Hypothetical Accident Conditions of Transport under QTP-001 Test Plan (see Appendix 2.12.2 QTP-001 Test Report).
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-4 Table 5.1.A: ISORAD-TC1 BPIC Prototype #1 Summary Table of External Radiation Levels Extrapolated to Capacity of 375 TBq (10125 Curies) of Ir-192 (Non-Exclusive Use)4.5,6 S/N: Prototype # 1 Package Surface 1 Meter from Package mSv/h (mrem/h) mSv/h (mrem/h)
Normal Conditions Top Side Bottom Top Side Bottom of Transport 0.2884 0.6248 0.3845 0.0096 0.0241 0.0193 Gamma (28.84) (62.48) (38.45) (0.962) (2.41) (1.93)
Neutron 0 (0) 0 (0) 0 (0) 0 (0) 0 (0) 0 (0) 0.2884 0.6248 0.3845 0.0096 0.0241 0.0193 Total (28.84) (62.48) (38.45) (0.962) (2.41) (1.93)
TS-R-1 530 & 531, SSR-6 526 & 527, or 2 (200) 2 (200) 2 (200) 0.1 (10)* 0.1 (10)* 0.1 (10)*
10 CFR 71.47 1 Meter from Package Hypothetical Accident Conditions mSv/h (mrem/h) 0.0098 0.0245 0.0245 Gamma (0.979) (2.445) (2.445) 0 0 0 Neutron (0) (0) (0) 0.0098 0.0245 0.0245 Total (0.979) (2.445) (2.445)
TS-R-1 657(b)(ii) SSR-6 559(b), & 10 CFR 71.51 10(10,000) 10(10,000) 10(10,000)
- Transport Index may not exceed 10 Table 5.1.B: ISORAD-TC1 BPIC 2835A Prototype #2 Summary Table of External Radiation Levels Extrapolated to Capacity of 375 TBq (10125 Curies) of Ir-192 (Non-Exclusive Use)4.5,6 S/N: Prototype # 2 Package Surface 1 Meter from Package mSv/h (mrem/h) mSv/h (mrem/h)
Normal Conditions Top Side Bottom Top Side Bottom of Transport 0.162 0.6665 0.6665 0.0143 0.0190 0.0190 Gamma (16.20) (66.65) (66.65) (1.43) (1.904) (1.904)
Neutron 0 (0) 0 (0) 0 (0) 0 (0) 0 (0) 0 (0) 0.162 0.6665 0.6665 0.0143 0.0190 0.0190 Total (16.20) (66.65) (66.65) (1.43) (1.904) (1.904)
TS-R-1 530 & 531, SSR-6 526 & 527, or 2 (200) 2 (200) 2 (200) 0.1 (10)* 0.1 (10)* 0.1 (10)*
10 CFR 71.47 1 Meter from Package Hypothetical Accident Conditions mSv/h (mrem/h) 0.0294 0.0294 0.0392 Gamma (2.938) (2.938) (3.917) 0 0 0 Neutron (0) (0) (0) 0.0294 0.0294 0.0392 Total (2.938) (2.938) (3.917)
TS-R-1 657(b)(ii) SSR-6 559(b), & 10 CFR 71.51 10(10,000) 10(10,000) 10(10,000)
- Transport Index may not exceed 10 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-5 5.2 Source Specification 5.2.1 Gamma Source The Special Form Sources allowed for transport in the ISORAD-TC1 transport package are specified in Sections 2.10 and 2.12.4.
5.2.2 Neutron Source Not applicable. The Model ISORAD-TC1 transport packages are not used for the transportation of neutron emitting sources.
5.3 Shielding Model 5.3.1 Configuration of Source and Shielding
[IAEA SSR-6 572, IAEA TS-R-1 772, and 10 CFR 71.74(b)]
The shielding evaluations were performed with Iridium-192 Special Form Capsules in a prototype model of the ISORAD-TC1 with the BPIC - Square and the BPIC 2835A inner containers. The Inner Container was installed into the Outer Drum as normally prepared for transport. Shielding surveys were performed before and after NCT testing and after HACT tests. The only effect of the tests on the packaging and its contents under NCT and HACT were denting and crushing of the Outer Drum. The BPIC Inner Containers position inside of the Outer Drum will not be affected under NCT and HACT conditions. Section 5.1.2 contains detailed information regarding the dose rate at the top, bottom and side of the ISORAD-TC1 BPIC - Square and the BPIC 2835A when loaded with 375 TBq (10,125 Ci) of Iridium-192. The dose rates were significantly lower than regulatory requirements.
The minimum amount of depleted uranium or tungsten shielding in the BPIC - Square or Round Inner Container in any direction from the closest side of the center Contents Cavity is 62 mm (2.44 in) or (DU 22.18 or Tungsten 18.77) half value layers. The minimum amount of depleted uranium or tungsten shielding in the BPIC 2835A Inner Container in any direction from the closest side of the center Contents Cavity is 63 mm (2.5 in) or (DU 22.73) half value layers. For the MPIC - Square or Round Inner Container, the minimum amount of depleted uranium or tungsten shielding in any direction from the end / bottom of the Contents Channel is 47.375 mm (1.87 in) or (DU 17.00 or Tungsten 14.38).
The locations with the maximum radiation levels were in the center of the side of the outer container, and the maximum radiation levels on the top and bottom were also located in the center.
5.3.2 Material Properties Not applicable. A shielding model was not used as the primary justification for these packages.
Shielding justification was based on direct measurement using Ir-192. Additional radionuclides were justified based on calculations derived from the Ir-192 direct measurements.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-6 5.4 Shielding Evaluation 5.4.1 Methods Shielding justification for Ir-192 was based on direct measurement. See QTP-001 Test Report (see Section 2.12.2) for results of Ir-192 radiation surveys of the ISORAD-TC1 packages after testing and Section 5.5.3 for Ir-192 profile results without any added shielding within the Inner Container Content Cavity.
Two prototype ISORAD-TC1 packages were used for all NCT and HACT tests. The prototypes were surveyed using Ir-192, radiation profiles were taken on both prototypes. The prototypes were each profiled three times: before testing, after the NCT 1.2 meter (4 foot) Free Drop and Penetration tests, and after the HACT 9 Meter Free Drop and Puncture tests. In Test Report QTP-001, the Ir-192 data was extrapolated to 10,125 curies for comparison of relative dose rate changes before and after testing when profiles were performed using sources with less activity. These results are shown in Tables 5.1A and 5.1.B. All Ir-192 radiation profile data are within regulatory acceptance limits, as shown in Tables 5.1.A and 5.1.B.
Radiation shielding information for the package configuration containing Yb-169 or Se-75 were obtained by calculation using derived calculations from the Ir-192 direct measurements.
5.4.2 Input and Output Data Radiation measurements included in this Section were adjusted to the maximum activity capacity for the package (e.g., activity correction factor) and the surface measurements were also adjusted to correct for off-set of the survey meter probe from the true surface of the package.
Activity correction factors (CFA) were obtained by using the following relationship:
CFA = Maximum Package Activity (Ac)
Actual Profile Activity (Ap)
For Example: if Ap = 2317.707Ci and Ac = 10125Ci then CFA = 10125Ci = 4.369 2317.707Ci Therefore, all original surface and 1 meter profile measurements would be multiplied by a factor of 4.369 for a package profiled using 2,317.707 Ci and a package capacity of 10,125 Ci. Radiation measurements at the surface of the container were also adjusted to compensate for the off-set of the survey meter probe from the true surface of the package.
Surface correction factors (SCF) were obtained by using the following relationship:
To determine if a correction factor was to be applied, ISO-RAD first used the IAEA Safety Guide TS-G-1.1 (ST-2) and ANSI N43.9-2015 Standard Table 1 [B1]. According to Table 1[B1], the correction factor is 1.0 based on the package dimensions and the probe distance. Secondly, ISO-RAD used the Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-7 Surface Correction Factor (SFC) formula to determine if the distance would match the IAEA and ANSI guidance below:
Where d1 = Center of Source to Drum Side Surface mm ( in) d2 =Center of Source to Center of Source Detector mm ( in)
SFC = d2/d1 SFC =
SFC = 1.0776 ISO-RAD applied a correction factor of 1.1 for conservatism. See the attached survey sheets with the data.
Therefore, in the example shown, all original surface profile measurements located along the side of the package would also be multiplied by a factor of 1.1 to account for surface correction of the detector to the drum. Different SCFs would be calculated for the any dimension of the container where the minimum distance from the center of the activity to the center of the radiation probe is different.
5.4.3 Flux-to-Dose-Rate Conversion Not applicable. Flux rates were not used to convert to dose rates in any shielding evaluations.
5.4.4 External Radiation Levels
[IAEA SSR-6 524, 524A, 526-528, 572, 647, 648(b), & 659(b), IAEA TS-R-1 526, 530-532, 572, 645, 646(b), & 656(b)(ii)(i), and 10 CFR 71.47(a)]
Radiation surveys for all ISORAD-TC1 configurations showed maximum surface and 1 meter radiation levels from the transport packages within regulatory limits. Radiation surveys of ISORAD-TC1 transport packages after undergoing normal and accident condition transport testing were also well within the regulatory limits.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-8 5.5 Appendix 5.5.1 Profile Measurements of the ISORAD-TC1 Containing Ir-192.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 5-13 5.5.2 Calculated Measurements of the ISORAD-TC1 containing Se-75 and Yb-169 Revision 0 29 February 2020
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Model: ISORAD-TC1 SAR 2020-1 Rev Project: QTP-001 Page 6-1 Safety Analysis Report ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev Project: QTP-001 Page 6-2 Section 6 Table of Contents Section 6 - CRITICALITY EVALUATION ........................................................................................ 6-3 6.1 Criticality .................................................................................................................................... 6-3 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev Project: QTP-001 Page 6-3 Section 6 - CRITICALITY EVALUATION 6.1 Criticality All parts of this section are not applicable. The ISORAD-TC1 packages are not used for shipment of Type B quantities of fissile material.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-1 Safety Analysis Report Section 7 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-2 Table of Contents Section 7 PACKAGE OPERATIONS .................................................................................................. 7-3 7.1 Package Loading .................................................................................................................... 7-3 7.1.1 Preparation for Loading .................................................................................................. 7-3 7.1.1.1 Authorized Package Contents .................................................................................. 7-3 7.1.1.2 ISORAD-TC1 Loading Instructions ........................................................................ 7-3 7.1.2 Loading of Contents ........................................................................................................ 7-4 7.1.2.1 For the Bulk & PIC Inner Container (BPIC) ........................................................... 7-4 7.1.2.2 For the Multi Port Inner Container (MPIC)............................................................. 7-4 7.1.2.3 For the Bulk & PIC Inner Container 2835A (BPIC 2835A) ................................... 7-5 7.1.2.4 Loading the Inner Container into the Outer Drum .................................................. 7-5 7.1.3 Preparation for Transport ................................................................................................ 7-6 7.2 Package Unloading................................................................................................................. 7-6 7.2.1 Receipt of Package from Carrier..................................................................................... 7-6 7.2.2 Removal of Contents....................................................................................................... 7-7 7.2.2.1 General for All Configurations ................................................................................ 7-7 7.2.2.2 For the BPIC and BPIC 2835A ............................................................................... 7-8 7.2.2.2 For the MPIC ........................................................................................................... 7-8 7.3 Preparation of an Empty Package for Transport .................................................................... 7-8 7.4 Other Operations .................................................................................................................... 7-9 7.5 Appendix ................................................................................................................................ 7-9 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-3 Section 7 PACKAGE OPERATIONS
[IAEA SSR-6 IAEA 809(d), TS-R-1 807(d), 10 CFR 71.31(c), 10 CFR 71.35(c), 10 CFR 71.43(g),
10 CFR 71.47(b-d), 10 CFR 71.87, and 10 CFR 71.89.
Operation of the ISORAD-TC1 transport packages must be in accordance with the operating instructions supplied with the transport package, per the above referenced regulations.
7.1 Package Loading 7.1.1 Preparation for Loading 7.1.1.1 Authorized Package Contents The ISORAD-TC1 transport packages are designed for use with only special form source capsules as approved under IAEA SSR-6, IAEA TS-R-1 or U.S. Department of Transportation special form certification. The approved isotopes and maximum package activity limits are shown in Section 2 Table 2.6.1.A and Table 2.6.1.B. Details of encapsulation as well as chemical and physical form of the radioactive material will comply with specifications approved under the Canadian Nuclear Safety Commission (CNSC) Certificate of Compliance, U.S. Department of Transportation revalidation Certificate of Competent Authority, and the valid Special Form Certificates issued under IAEA SSR-6, IAEA TS-S-R-1, U.S. Department of Transportation, or other Competent Authority special form certifications.
7.1.1.2 ISORAD-TC1 Loading Instructions Ensure that written instructions exist for preparing the ISORAD-TC1 for loading. These written instructions must include the following steps at a minimum:
Step 1. Verify that the contents are special form and appear on the Certificate of Competent Authority issued by the regulating authority, and that the Activity contained is less than or equal to the authorized activity.
Step 2. Ensure that the package complies with the conditions of approval in the Certificate of Competent Authority. Verify that the package meets the design identified in the package approval. The package must have a nameplate that has the Type B certificate number permanently on the plate.
Step 3. Visually inspect bolts for cracks, and replace any unsuitable bolts with those supplied by the manufacturer.
Step 4. Verify that the package is in unimpaired physical condition. Surface dents and small holes are allowable in the outer housing, the drum, providing that the inner container is secured in the center of the drum. Surface dents that do not impair the proper transport and operation of the inner container are also allowed.
Step 5. Verify that the package is properly labeled.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-4 Step 6. There are no special controls or precautions for handling the package.
7.1.2 Loading of Contents Ensure that written instructions exist for loading the ISORAD-TC1. These written instructions must include the following steps at a minimum:
7.1.2.1 For the Bulk & PIC Inner Container (BPIC)
Step 1. Load the special form capsules into the center cavity working from a properly shielded area, shielded vault, or hotcell.
Step 2. Insert Shield Plug Assembly into container over the center cavity, and bolt it in place with four M6 x 1.0 stainless steel bolts.
Step 3. Place the BPIC Lid over the bolted down Shield Plug Assembly. When properly aligned, it should be flat and flush with the top of the BPIC.
Step 4. Bolt the container lid in place using eight M8 x 1.25 bolts and install the M8 x 1.25 eyebolt (assists in lifting) in the center hole.
Step 5. The BPIC is ready to be lifted and installed into the Outer Drum cavity.
Step 6. Using a manual or mechanical lifting device, place the BPIC in the cavity in the outer package.
Step 7. Remove the eyebolt.
7.1.2.2 For the Multi Port Inner Container (MPIC)
Step 1. The MPIC can load up to 10 sealed sources assemblies, one each, into the tubes working from a properly shielded area, shielded vault, or hotcell.
Step 2. Secure each source assembly using the locking mechanism.
Step 3. Install a Source Cap fitting over each source channel end.
Step 4. Install the MPIC Lid over the Locking Stud and either Step 4A. Thread the special M16 eyebolt tight to the stud to enable lifting of the MPIC; Step 4B. Thread the M16 Hex Nut tight securing the Lid to the MPIC body and use the special lifting attachment installed through the padlock hole in the Locking Stud.
Step 5. Place the MPIC into the Outer Drum cavity.
Step 6. Using a manual or mechanical lifting device, place the BPIC in the cavity in the outer package.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-5 Step 7. Remove the eyebolt or lifting attachment.
7.1.2.3 For the Bulk & PIC Inner Container 2835A (BPIC 2835A)
Step 1. Load the special form capsules into the center cavity working from a properly shielded area, shielded vault, or hotcell.
Step 2. Insert Shield Plug Assembly into container over the center cavity.
Step 3. Place the BPIC 2835A Lid over the Shield Plug Assembly. When properly aligned, it should be flat and flush with the top of the BPIC 2835A.
Step 4. Bolt the container lid in place using six M8 x 1.25 bolts and install the M8 x 1.25 eyebolt (assists in lifting) in the center hole.
Step 5. The BPIC 2835A is ready to be lifted and installed into the Outer Drum cavity.
Step 6. Using a manual or mechanical lifting device, place the BPIC in the cavity in the outer package.
Step 7. Remove the eyebolt.
CAUTION: Failure to remove the eyebolt may cause damage to the Cork Lid Assembly.
Step 8. Install BPIC 2835A cork space(s) on top of the BPIC 2835A in the Outer Drum cavity.
7.1.2.4 Loading the Inner Container into the Outer Drum Step 1. Ensure the lifting eyebolt or lifting attachment is removed.
Step 2. Place the Cork Lid Assembly into the upper cavity of the Outer Drum.
Step 3. No additional equipment may be transported in the space between the Cork Lid Assembly and the Outer Drum Lid.
Step 4. Install the Outer Drum Lid. Bolt the Outer Drum Lid to the drum by using the eight M14 x 2 stainless steel bolts, M14 lock washers, and M14 flat washers.
Step 5. Install two padlocks through the two Lock Studs.
Step 6. Apply tamper-evident seal.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-6 7.1.3 Preparation for Transport Ensure that written instructions exist for preparing the ISORAD-TC1 for transport. These written instructions must include the following steps at a minimum:
Step 1. The level of removable radioactive contamination must be as low as reasonably achievable (ALARA). The contamination may be determined by measuring the activity on wipes taken from representative areas using an absorbent material and moderate pressure. Packages may be shipped on a nonexclusive use basis only if outer surface contamination levels are less than the values given below. It is it is the shipper's responsibility to ensure that the following conditions are met. IAEA SSR-6 508, TS-R-1 508, and 49 CFR 173.443 require that the non-fixed (removable) contamination on the external surfaces of the outer package being shipped on a nonexclusive use basis not to exceed 10-4
µCi/cm2 (0.0001 µCi/cm2) averaged over a 300 cm2 (46.5 in2) area of any part of the surface assuming a wipe efficiency of 0.10 (or the actual wipe efficiency may be used).
Step 2. Measure the maximum surface radiation level of the outer package, the drum. This Radiation level must be as low as reasonably achievable, and not exceed 2 mSv/hr (200 mR/hr). Measure the maximum radiation level at one meter (39.4 in) from the surface. This radiation level must not exceed 0.1 mSv/hr (10 mR/hr).
Step 3. Ensure that the padlocks and tamper-evident seal is properly applied.
Step 4. There are no pressure relief valves utilized on the ISORAD-TC1.
Step 5. The Outer Drum handles are the only structural parts of the package designed to meet lift and tie down requirements.
Step 6. The outer package will be properly marked, labeled and described on a shipping paper in accordance with IAEA SSR-6 531-536, 538-539, and 542; IAEA TS-R-1 535-540, 542-543, and 546; and 49 CFR 172.310 and 49 CFR 172.403.
Step 7. The ISORAD-TC1 is designed to meet nonexclusive use requirements when loaded with the maximum radioactive. Written instructions to the carrier are required for packages that require exclusive use shipment.
Step 8. There are no special instructions needed to safely open the ISORAD-TC1 package.
7.2 Package Unloading 7.2.1 Receipt of Package from Carrier The consignee must establish written procedures for receiving and safely opening the ISORAD-TC1package in accordance with applicable regulations. These procedures must include, at a minimum, these requirements.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-7 Step 1. Schedule to either receive the package when the carrier offers it for delivery, or to take possession of the package expeditiously at the carrier's terminal.
Step 2. Monitor the external surface radiation level as soon as practical after receipt. The package must be monitored (surveyed) within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the package is received during normal working hours, or not later than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after start of business the next working day if it is received after working hours.
NOTE: The ISORAD-TC1 package is only authorized to transport special form radioactive material, so it is only required to monitor the external surfaces of the package for removable contamination if there is evidence of degradation of package integrity, such as package that is crushed or otherwise damaged.
Step 3. Check the external package for excessive dents or damage if they might reasonably be expected to impair the inner package, including the presence of the tamper seal and lid bolt indicating that the package is, and has been, securely closed.
Step 4. Make any required notifications if the above requirements have not been met prior to proceeding to unload the package.
Step 5. The MPIC may be unloaded one source at a time into an industrial radiography device or source changer using an exchange tube. For the BPIC containers and other circumstances, a shielded vault or hotcell may be the only way to provide adequate shielding during the unloading process.
7.2.2 Removal of Contents CAUTION: The BPIC inner containers may be hot to the touch, thick thermally insulated gloves are recommended for use in handling the BPIC containers.
7.2.2.1 General for All Configurations Step 1. Remove the tamper-evident seal and padlocks from the Outer Drum Lid.
Step 2. Unbolt the eight M14 x 2 bolts from the Outer Drum Lid.
Step 3. Lift the Cork Lid Assembly from the Outer Drum to expose the BPIC or MPIC by using the finger holes in the top of the lid.
Step 4. Use the appropriate eyebolt(s) or lifting attachment to assist in lifting the BPIC or MPIC from the Outer Drum.
Step 5. Remove the BPIC or MPIC to a suitable location for transferring its radioactive contents.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-8 7.2.2.2 For the BPIC and BPIC 2835A Follow the remaining steps to remove the contents:
Step 1. Loosen and remove the (BPIC) eight M8 x 1.25 bolts or (BPIC 2835A) six M8 x 1.25 bolts from the Lid of the inner container.
Step 2. Remove the BPIC Lid, exposing the bolted down Shielded Plug Assembly.
Step 3. Remove the four M6 x 1.0 bolts from the Shielded Plug Assembly.
Step 4. Lift the Shielded Plug Assembly, use appropriate measures, the plug weighs approximately 4.5 kg (9.9 pounds).
Step 5. Transfer special forms and store in shielded vault or hotcell.
7.2.2.2 For the MPIC Follow these remaining steps to remove the contents:
Step 1. Remove padlock from the Locking Stud.
Step 2. Loosen and remove the M16 Nut the MPIC Lid.
Step 3. Remove the MPIC Lid, exposing the Source Caps and Locking Mechanism.
Step 4. Loosen the Source Caps and transfer source assemblies.
7.3 Preparation of an Empty Package for Transport Step 1. Verify the ISORAD-TC1 package does not contain radioactive sources by surveying the package for higher radiation levels. This test should be performed by trained and authorized personnel who are monitored and possess a calibrated and properly operating survey instrument.
Step 2. The empty packaging may contain Depleted Uranium and may be shipped as either a labeled radioactive material package or as an excepted package, article manufactured from depleted Uranium as required by applicable U.S. Department of Transportation regulations. For the Tungsten shield versions use the appropriate empty packaging marking.
Step 3. Perform a contamination survey of the internal surfaces of the package (inner container cavity and underside of the lid). If the non-fixed surface contamination exceeds local requirements for empty package shipment, decontaminate as necessary.
Step 4. Assemble the package in the same manner as a loaded package, (with the exception that there will not be any radioactive material loaded in the package. Secure the BPIC or MPIC in the Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 7-9 Outer Drum, install the Cork Lid Assembly, install the Outer Drum Lid, and tighten the eight M14 x 2 bolts.
Step 5. Install the padlocks and apply tamper-evident seal.
Step 6. Perform a radiation survey to confirm that the package is empty and meets the requirements for shipment of empty packages.
Step 7. Apply the appropriate marking and labeling for the empty package.
7.4 Other Operations Not Applicable; there are no special operational controls needed.
7.5 Appendix None.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 8-1 Safety Analysis Report Section 8 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 8-2 Section 8 Table of Contents Section 8 Table of Contents .................................................................................................................. 8-2 Section 8 - ACCEPTANCE TESTS AND MAINTENANCE PROGRAM ........................................ 8-3 8.1 Acceptance Test ..................................................................................................................... 8-3 8.1.1 Visual Inspections and Measurements ............................................................................ 8-3 8.1.2 Weld Examinations ......................................................................................................... 8-3 8.1.3 Structural and Pressure Tests .......................................................................................... 8-3 8.1.4 Leakage Tests.................................................................................................................. 8-4 8.1.5 Component and Material Tests ....................................................................................... 8-4 8.1.6 Shielding Tests ................................................................................................................ 8-4 8.1.7 Thermal Tests.................................................................................................................. 8-4 8.1.8 Miscellaneous Tests ........................................................................................................ 8-4 8.2 Maintenance Program ............................................................................................................ 8-5 8.2.1 Structural and Pressure Tests .......................................................................................... 8-5 8.2.2 Leakage Tests.................................................................................................................. 8-5 8.2.3 Component and Material Tests ....................................................................................... 8-5 8.2.4 Thermal Tests.................................................................................................................. 8-5 8.2.5 Miscellaneous Tests ........................................................................................................ 8-5 8.3 Appendix ................................................................................................................................ 8-5 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 8-3 Section 8 - ACCEPTANCE TESTS AND MAINTENANCE PROGRAM
[IAEA SSR-6 501 & 809(d), IAEA TS-R-1 501 & 807(d), and 10 CFR 71 Subpart G]
8.1 Acceptance Test
[IAEA SSR-6 501, IAEA TS-R-1 501, and 10 CFR 71.85]
8.1.1 Visual Inspections and Measurements 8.1.1.1 Visual Inspection Visual inspections and measurements will be performed in accordance with ISO-RAD Canadas, Quality Assurance Program. Visually inspect each transport package component to be shipped to assure the following:
Step 1. The transport package was assembled properly to the applicable drawing referenced on the Type B transport certificate.
Step 2. Evaluate the transport package for shielding integrity to ensure the transport dose rate requirements are met when the container is loaded to capacity.
Step 3. All fasteners as required by the applicable drawings referenced on the Type B transport certificate are properly installed and secured.
Step 4. The relevant labels are attached, contain the required information, and are marked in accordance with IAEA SSR-6 530-536A & 538-540, IAEA TS-R-1 534-539 &
542-544, or 10 CFR 20.1904, 10 CFR 40.13(c)(6)(i), 10 CFR 34, and 10 CFR 71 or equivalent Agreement State regulations.
8.1.2 Weld Examinations Weld examinations will be performed in accordance with the applicable drawings requirements, welding code (AWS D1.6/D1.6M and AWS D1.9/D1.9M or equivalent standards), and in accordance with ISO-RAD Canadas, approved Quality Assurance Program.
8.1.3 Structural and Pressure Tests Prior to first use as part of a Model ISORAD-TC1 Transport Package, container structural conformance will be evaluated in accordance with the applicable drawing requirements and in accordance with ISO-RAD Canadas, Quality Assurance Program. The containment system is not designed to require increased or decreased operating pressures to maintain containment during transport, therefore pressure tests of package components prior to first use is not required.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 8-4 8.1.4 Leakage Tests The source capsules (primary containment) are wipe tested for leakage of radioactive contamination upon initial manufacture. The removable contamination must be less than 185 Bq (0.005 µCi). The source capsules will also be subjected to leak tests under ISO9978:
1992(E) (or more recent editions). The source capsules are not used if they fail any of these tests.
8.1.5 Component and Material Tests Component and material compliance are achieved in accordance with the requirements in ISO-RAD Canadas approved Quality Assurance Program.
8.1.6 Shielding Tests The radiation levels at the surface of the transport package and at 1 meter from the surface are measured upon manufacture. This survey is performed in a low background area using Ir-192 where the activity is determined as content activity. The survey involves a slow scan survey of the entire surface area as well as one meter from the surface of the package. This survey is used to identify any significant void volumes or shield porosity which could prevent the finished device from complying with the dose limits in IAEA SSR-6 526-528, IAEA TS-R-1 530-532, or 10 CFR 71.47.
The radiation profile survey is made with the radiation detector housing in contact with the surface of the package and then also at one meter from the surface of the container. These radiation levels, when extrapolated to the rated capacity of the transport package, must not exceed 200 mR/hr at the surface, nor 10 mR/hr at 1 meter from the surface of the transport package. Failure of this test prevents use of the package as a Type B(U) package.
Rejected packages which do not comply with the construction requirements on the applicable drawings referenced on the Type B certificate, or that do not comply with the radiation profile requirements will not be distributed as approved Type B(U) packages.
8.1.7 Thermal Tests As demonstrated in Sections 2 and 3, the source content will have no adverse effect on the package surface temperature and therefore no additional testing is necessary to evaluate thermal properties of the packaging.
8.1.8 Miscellaneous Tests Not applicable.
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Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 8-5 8.2 Maintenance Program
[IAEA SSR-6 306(b) 809(d), IAEA TS-R-1 310(b) & 807(d), and 10 CFR 71.31(c), 10 CFR 71.37(b), 10 CFR 71.87(b), 10 CFR 71.87(g), & 10 CFR 71.93(b)]
8.2.1 Structural and Pressure Tests Not applicable. Material certification is obtained for Safety Class A components used in the transport package prior to their initial use. Based on the construction of the design, no additional structural testing during the life of the package is necessary if the container shows no signs of defect when prepared for shipment in accordance with the requirements of Section 7 of the SAR. The ISORAD-TC1 packaging is not designed to require increased or decreased operating pressures to maintain containment during transport, therefore pressure tests of package components prior to individual shipment is not required.
8.2.2 Leakage Tests As described in Section 8.1.4, Leakage Tests, the radioactive source assembly is leak-tested at manufacture. In addition, the sources are leak tested in accordance with that Section at least once every six months thereafter if being transported to ensure that removable contamination is less than 185 Bq (0.005 Ci).
8.2.3 Component and Material Tests The transport package is inspected for tightness of fasteners, proper seal wires, general condition and fitness for use prior to each use (see Section 7.1.1).
Prior to each use, a radiation survey of the transport package is made to assure that the radiation levels do not exceed 200 mR/hr at the surface, nor 10 mR/hr at 1 meter from the surface.
8.2.4 Thermal Tests Not applicable. The source content of the Model ISORAD-TC1 package has no adverse effect on the package surface temperature and therefore no additional testing is necessary to evaluate thermal properties of the packaging prior to shipment.
8.2.5 Miscellaneous Tests Inspections and tests designed for secondary users of this transport package under the general license provisions of 10 CFR 71.17(b) are provided in Section 7.
8.3 Appendix None.
Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 9-1 Safety Analysis Report Section 9 ISO-RAD Canada, Inc Ottawa, ON Canada Model: ISORAD-TC1 Type B(U)-96 Transport Package February 29, 2020 Revision 0 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 9-2 Table of Contents Chapter 9 - QUALITY ASSURANCE ................................................................................................ 9-3 9.1 United States Quality Assurance Program Requirements ...................................................... 9-3 9.2 Canadian Quality Assurance Program Requirements ............................................................ 9-3 9.3 Appendix ................................................................................................................................ 9-3 9.3.1 ISO-RAD Canada Inc. Quality Assurance Program/Management System .................... 9-3 Revision 0 29 February 2020
Model: ISORAD-TC1 SAR 2020-1 Rev 0 Project: QTP-001 Page 9-3 Chapter 9 - QUALITY ASSURANCE 9.1 United States Quality Assurance Program Requirements Not applicable.
9.2 Canadian Quality Assurance Program Requirements ISO-RAD Canada Inc maintains a Quality Assurance Program/Management System (QAP) that meets the requirements of IAEA SSR-6 306, IAEA TS-R-1 310, PTNS SOR/2015-145 Paragraph 24 of the regulations, and 10 CFR 71 Subpart H. IAEA Safety Series 113 and IAEA Safety Guide TS-G-1.4 were followed in the establishment of the program. All elements applicable to designers of transport packages have been included in the QAP per IAEA Safety Series 113 Table I.
9.3 Appendix 9.3.1 ISO-RAD Canada Inc. Quality Assurance Program/Management System Revision 0 29 February 2020