Text

Redacted, Industrial Nuclear Company - Application for Certificate of Compliance for the Inc IR1 00ST, Revision 0, Docket No. 71-9385
	ML24295A160
Person / Time
Site:	07109385
Issue date:	06/30/2024
From:	Maret Rose; Industrial Nuclear Co
To:	; Division of Fuel Management
Shared Package
ML24295A159	List: ML24295A160;
References
	CAC 001029, EPID L-2024-NEW-0008
	Download: ML24295A160 (1)
	v • d • e

SAFETY ANALYSIS REPORT IR-100ST Exposure Device

Docket No. 71-9385 Revision 0 June 2024

Industrial Nuclear Company, Inc.

14320 Wicks Blvd.

San Leandro, California 94577 (510) 352-6767 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

TABLE OF CONTENTS 1.0 General Information............................................................................................................. 1 1.1 Introduction...................................................................................................................... 1 1.2 Package Description......................................................................................................... 1 1.2.1 Packaging................................................................................................................... 1 1.2.2 Contents of Packaging............................................................................................... 2 1.2.3 Special Requirements for Plutonium......................................................................... 2 1.2.4 Operational Features.................................................................................................. 2 1.3 Appendices....................................................................................................................... 6 1.3.1 General Arrangement Drawings................................................................................ 6 2.0 Structural Evaluation......................................................................................................... 19 2.1 Description of Structural Design.................................................................................... 19 2.1.1 Discussion................................................................................................................ 19 2.1.2 Design Criteria......................................................................................................... 20 2.1.3 Weights and Center of Gravity................................................................................ 21 2.1.4 Identification of Codes and Standards for Package Design..................................... 21 2.2 Materials......................................................................................................................... 21 2.2.1 Material Properties and Specifications.................................................................... 21 2.2.2 Chemical, Galvanic, or Other Reactions.................................................................. 22 2.2.3 Effects of Radiation on Materials............................................................................ 23 2.3 Fabrication and Examination......................................................................................... 23 2.3.1 Fabrication............................................................................................................... 23 2.3.2 Examination............................................................................................................. 23 2.4 General Requirements for All Packages........................................................................ 23 2.4.1 Minimum Package Size........................................................................................... 23 2.4.2 Tamper Indicating Device........................................................................................ 23 2.4.3 Positive Closure....................................................................................................... 24 2.4.4 Valves...................................................................................................................... 24 2.4.5 Package Design........................................................................................................ 24 2.4.6 External Temperatures............................................................................................. 24 2.4.7 Venting..................................................................................................................... 24 2.5 Lifting and Tie-down Devices for All Packages............................................................ 24

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2.5.1 Lifting Devices......................................................................................................... 24 2.5.2 Tie-Down Devices................................................................................................... 24 2.6 Normal Conditions of Transport.................................................................................... 25 2.6.1 Heat.......................................................................................................................... 25 2.6.2 Cold.......................................................................................................................... 26 2.6.3 Reduced External Pressure...................................................................................... 26 2.6.4 Increased External Pressure..................................................................................... 26 2.6.5 Vibration.................................................................................................................. 26 2.6.6 Water Spray............................................................................................................. 26 2.6.7 Free Drop................................................................................................................. 26 2.6.8 Corner Drop............................................................................................................. 26 2.6.9 Compression............................................................................................................ 26 2.6.10 Penetration............................................................................................................... 27 2.7 Hypothetical Accident Conditions................................................................................. 27 2.7.1 Free Drop................................................................................................................. 27 2.7.2 Crush........................................................................................................................ 2 8 2.7.3 Puncture................................................................................................................... 28 2.7.4 Thermal.................................................................................................................... 31 2.7.5 Immersion - Fissile Material................................................................................... 31 2.7.6 Immersion - All Packages....................................................................................... 31 2.7.7 Deep Water Immersion Test (for Type B Packages Containing More than 105 A2) 32 2.7.8 Summary of Damage............................................................................................... 32 2.8 Accident Conditions for Air Transport of Plutonium.................................................... 32 2.9 Accident Conditions for Fissile Material Packages for Air Transport........................... 32 2.10 Special Form............................................................................................................... 32 2.11 Fuel Rods.................................................................................................................... 3 2 2.12 Appendix.................................................................................................................... 33 2.12.1 Certification Tests.................................................................................................... 34 3.0 Thermal Evaluation............................................................................................................ 59 3.1 Description of Thermal Design...................................................................................... 59 3.1.1 Design Features........................................................................................................ 59 3.1.2 Contents Decay Heat.............................................................................................. 59 3.1.3 Summary Tables of Temperatures........................................................................... 59

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3.1.4 Summary Tables of Maximum Pressures................................................................ 59 3.2 Material Properties and Component Specifications....................................................... 59 3.2.1 Material Properties................................................................................................... 59 3.2.2 Component Specifications....................................................................................... 60 3.3 Thermal Evaluation under Normal Conditions of Transport......................................... 60 3.3.1 Heat and Cold.......................................................................................................... 60 3.3.2 Maximum Normal Operating Pressure.................................................................... 61 3.4 Thermal Evaluation under Hypothetical Accident Conditions...................................... 61 3.4.1 Initial Conditions..................................................................................................... 61 3.4.2 Fire Test Conditions................................................................................................. 61 3.4.3 Maximum Temperatures and Pressures................................................................... 62 3.4.4 Maximum Thermal Stresses.................................................................................... 62 3.4.5 Accident Conditions for Fissile Material Packages for Air Transport.................... 62 4.0 Containment....................................................................................................................... 63 5.0 Shielding Evaluation.......................................................................................................... 64 5.1 Description of Shielding Design.................................................................................... 64 5.1.1 Design Features........................................................................................................ 64 5.1.2 Summary Table of Maximum Radiation Levels...................................................... 64 5.2 Source Specification....................................................................................................... 64 5.2.1 Gamma Source......................................................................................................... 64 5.2.2 Neutron Source........................................................................................................ 65 5.3 Shielding Model............................................................................................................. 65 5.4 Shielding Evaluation...................................................................................................... 65 5.4.1 Methods.................................................................................................................... 65 5.4.2 Input and Output Data.............................................................................................. 65 5.4.3 Flux-to-Dose-Rate Conversions............................................................................... 65 5.4.4 External Radiation Levels........................................................................................ 65 6.0 Criticality Evaluation......................................................................................................... 66 7.0 package Operations............................................................................................................ 67 7.1 Package Loading............................................................................................................ 67 7.1.1 Preparation of the IR-100ST for Loading................................................................ 67 7.1.2 Loading the Special Form Payload into the IR-100ST............................................ 67 7.1.3 Preparation for Transport......................................................................................... 67

iii INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

7.2 Package Unloading......................................................................................................... 68 7.2.1 Receipt of Package from Carrier.............................................................................. 68 7.2.2 Removal of Contents from the IR-100ST Package.................................................. 69 7.3 Preparation of an Empty Package for Transport............................................................ 69 8.0 Acceptance Tests and Maintenance Program.................................................................... 70 8.1 Acceptance Tests............................................................................................................ 70 8.1.1 Visual Inspections and Measurements..................................................................... 70 8.1.2 Weld Examinations.................................................................................................. 70 8.1.3 Structural and Pressure Tests................................................................................... 70 8.1.4 Leakage Tests........................................................................................................... 70 8.1.5 Component and Material Tests................................................................................ 70 8.1.6 Shielding Tests......................................................................................................... 70 8.1.7 Thermal Tests........................................................................................................... 70 8.1.8 Miscellaneous Tests................................................................................................. 70 8.2 Maintenance Program.................................................................................................... 71 8.2.1 Structural and Pressure Tests................................................................................... 71 8.2.2 Leakage Tests........................................................................................................... 71 8.2.3 Component and Material Tests................................................................................ 71 8.2.4 Thermal Tests........................................................................................................... 71 8.2.5 Miscellaneous Tests - Shielding.............................................................................. 71

iv INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

1.0 GENERAL INFORMATION This chapter of the IR-100ST Exposure Device Safety Analysis Report presents a general introduction and description of the IR-100ST Exposure Device. A detailed description of the major packaging and payload components is presented in the following sections. Detailed drawings are presented in Appendix 1.3.1, General Arrangement Drawings.

1.1 Introduction The IR-100ST Exposure Device (hereto referred to as the IR-100ST) is a transportation system designed to transport a single, special form iridium-192 (Ir-192) or selinium-75 (Se-75) source capsule. The design is optimized to provide maximum safety during both operations and transport conditions. The packaging consists of a rectangular stainless steel body housing that encloses a depleted uranium (DU) gamma shield, a molded urethane thermoplastic sensor sensor/handle jacket, an outlet port assembly, a lock assembly, and interior polyurethane foam.

Authorization is sought for shipment of a single, special form Ir-192 or Se-75 source capsule (per package) as a Type B(U)-96, special form material package per the definitions delineated in 10 CFR §71.41. The transport index (TI) for the package, determined in accordance with the definition of 10 CFR §71.4, is determined for each shipment. The TI is based on the radiation dose rate at 1 meter from the package surface (method for the transport index is defined in Chapter 7.0, Package Operations).

1.2 Package Description

1.2.1 Packaging The IR-100ST, is a Type B(U)-96 package designed for transportation of a Ir-192 or Se-75 special form capsule. The IR-100ST Exposure Device is a modified INC IR-100 Exposure Device, which is a licensed transportation package2. The modifications consist of removal of the stainless steel handle with a modified stainless steel end sheet, and attaching a molded urethane elastomer sensor/handle jacket assembly around the modified IR-100 stainless steel body. The primary function of the sensor/handle jacket is to provide remote tracking of the IR-100ST package during transport via a Persistence Monitoring (PM) Tag. The maximum gross weight of the package is 58 pounds and its primary components of construction are identified in Figure 1.2-1. The payload is a special form capsule containing Ir-192 or Se-75, and is described in Section 1.2.2, Contents of Packaging. Primary shielding is provided by DU. The DU gamma shield, which is composed of 0.23% U-235, 99.77% U-238, and illustrated in Figure 1.2-2, is a solid form casting. The shield contains 0.0042 Ci (0.00016 TBq) of DU. Detailed drawings of the IR-100ST are provided in Appendix 1.3.1, General Arrangement Drawings.

The sensor/handle jacket assembly consists of a urethane elastomer is molded around the 1/16-inch stainless steel wire rope for the handle structure. Four, 6.6-volt lithium ion (LiFePO4) cathode power cells with a graphite anode supply power to the tracking system and are fully

1 Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material, 1-1-24 Edition.

2 Safety Analysis Report, IR-100 Exposure Device, Industrial Nuclear Company, Inc., NRC Docket No. 71-9157.

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contained in the sealed PM Tag enclosure, which is located in the bottom of the sensor jacket.

The lithium ion power cells comply with the requirements of Par t III, Section 38.3 of the United Nations transportation standard3 for transportation of dangerous goods. Four steel sealed bearing wheel assemblies are secured in the bottom for longitudinal mov ement of the package on a circular surface, e.g., pipe, to facilitate movement during use as a radiography exposure device.

The battery cell chemistry of the power cells is safer than the typical lithium cobalt oxide composition utilized in lithium batteries. The lithium power cell materials also include copper and aluminum within an outer steel casing. The lithium power cells are also designed with the following protective design features:

Vent seals that activate at an internal high pressure of 261 to 348 psi (1.8 to 2.4 MPa)

A current interrupt device (CID) that activates on excessive pressure due an overcharge condition A shutdown separator that activates when cells reach a temperature of 266 °F (130 °C),

which could melt the cells poly-separators The lithium power cells also satisfied the following European Council for Automotive R&D (EURCAR) requirements4:

EURCAR Hazard Level 2 for overcharge EURCAR Hazard Level 3 for over-discharge, external short, and crush EURCAR Hazard Level 4 for nail penetration

1.2.2 Contents of Packaging The IR-100ST is designed to transport a maximum of 120 Ci (4.44 TBq) of Ir-192 or Se-75 within a single, special form capsule. The capsule is attached to a pigtail assembly that, along with the lock assembly and lockball, secures the capsule within the center of the DU gamma shield.

1.2.3 Special Requirements for Plutonium This section does not apply since plutonium is not transported in the IR-100ST.

1.2.4 Operational Features There are no operationally complex features of the IR-100ST. The contents (described in the following section) are confined within the housing, lock assembly, and DU gamma shield.

Overall views of the package are shown in Figure 1.2-2 and Figure 1.2-2. Integral to the housing and the DU gamma shield is the lock assembly that prevents unauthorized removal or unshielded exposure of the contents. The lock assembly, which permits access to the contents, conform to the requirements of 10 CFR §34.235. Attached to the housing body is a molded urethane sensor

3 United Nations Recommendations on the Transport of Dangerous Goods - Manual Tests and Criteria, Sixth Revised Edition, 2015.

4 LithiumWerks 18650 Lithium Ion Power Cell Data Sheet, June 2022, SF000007 Rev. 2, www.lithiumwerks.com.

5 Title 10, Code of Federal Regulations, Part 34 (10 CFR 34), Licenses for Radiography and Radiation Safety Requirements for Radiographic Operations, 1-1-24 Edition.

2 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

surround that can facilitate both tie-down and handling, and pr ovides remote tracking of the package during transportation.

Sequential steps of operation are provided in Chapter 7.0, Package Operations.

HANDLE

POLYURETHANE FOAM OUTLET PORT ASSEMBLY

LOCK ASSEMBLY

DUST COVER DUST COVER SENSOR SURROUND

4 X WHEEL ASSEMBLIES RADIOACTIVE SOURCE DU SHEILD CAPSULE / PIGTAIL ASSEMBLY

PM TAG WITH LITHIUM POWER CELLS

Figure 1.2 Sectional View of the IR-100ST Packaging

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Figure 1.2 Overall View of the IR-100ST Packaging, Lock Assembly End

4 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

Figure 1.2 Overall View of the IR-100ST Packaging, with Keyed Lock Assembly Outlet Port Assembly End

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1.3 Appendices

1.3.1 General Arrangement Drawings

6

INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

2.0 STRUCTURAL EVALUATION This chapter presents the structural design criteria, weights, mechanical properties of material, and structural evaluations which demonstrate that the IR-100ST Exposure Device meet all applicable structural criteria for transportation as defined in 10 CFR 716.

2.1 Description of Structural Design The primary evaluation of the IR-100ST is performed with various full-scale tests. The results of the tests are provided in the following sections.

The IR-100ST consists of four major fabricated components: 1) a stainless steel housing and lock assembly that enclose and secure the contents, 2) a DU gamma shielding that provides shielding,

3) polyurethane foam that provides protection of the DU from moisture, and 4) a urethane sensor surround that provides remote tracking of the package via the P M Tag.

2.1.1 Discussion The IR-100ST is designed to transport a maximum of 120 Ci (4.44 TBq) of Ir-192 or Se-75 in a special form capsule. Since the payload is designated as special form, the IR-100ST is defined as a confinement system. As shown in the section view in Figure 2-1, the primary components of the package are a stainless steel housing (outer shell, including a lock and outlet port assemblies), a DU gamma shield, internal polyurethane foam, and a urethane sensor surround for remote tracking.

The housing is constructed of Type 304 austenitic stainless steel. The polyurethane foam, which surrounds the DU gamma shield, is closed cell. The DU gamma shield is a casting of solid form and optimally designed to provide optimum shielding of the special form capsule. The DU gamma shield is secured in the housing by welded brackets that capture the shield between the bracket and the stainless steel housing.

The stainless steel housing is a rectangular shell structure with 12-gauge (0.105-inch) thick walls, and overall dimensions of approximately 10.65 inch high x 61/4 inch wide x 101/2 inch long. The housing is constructed entirely of Type 304 stainless steel sheet that completely encloses the foam and the DU gamma shield. As shown in Figure 2-1, the urethane sensor surround incorporates a handle for lifting and tie-down is provided to facilitate operations.

Welded to opposite ends of the housing are the lock and outlet port assemblies. These assemblies provide the operationally capability during use and secure the location of the Ir-192 or Se-75 special form capsule during transportation. The lock assembly, which allows access to the contents, conforms to the requirements of 10 CFR §34.237.

Polyurethane foam within the stainless steel housing fills the void between the DU gamma shield and the stainless steel housing. The foam provides moisture protection of the DU during normal operations.

6 Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material, 1-1-24 Edition.

7 Title 10, Code of Federal Regulations, Part 34 (10 CFR 34), Licenses for Radiography and Radiation Safety Requirements for Radiographic Operations, 1-1-24 Edition.

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The urethane elastomer molded sensor/handle jacket is secured around the IR-100ST stainless steel body by eight (8) 3/16 inch stainless steel pop rivets. The primary function of the sensor/handle jacket is to provide remote tracking of the package during transport via the PM Tag.

HANDLE

r/).75 VENT

POLYURETHANE FOAM OUTLET PORT ASSEMBLY

LOCK ASSEMBLY DUST COVER

DUST COVER SENSOR SURROUND

RADIOACTIVE SOURCE DU SHEILD CAPSULE/PIGTAIL ASSEMBLY

PM TAG WITH LITHIUM POWER CELLS

Figure 2 Sectional View of the IR-100ST P ackaging

2.1.2 Design Criteria

2.1.2.1 Basic Design Criteria The IR-100ST is primarily demonstrated to satisfy the requirements of 10 CFR 71 via full-scale tests. For evaluation of tie-down devices, the design criterion is that the structural tie-down members do not exceed the materials yield strength when subjected to the requirements of 10 CFR §71.45(b).

2.1.2.2 Miscellaneous Structural Failure Modes

2.1.2.2.1 Brittle Fracture The structural materials of the IR-100ST packaging include stainless steel and DU. Each material is not susceptible to brittle fracture at temperatures as low as -20 ºF (-29 ºC) as described below.

The housing and lock and outlet port assemblies of the IR-100ST are fabricated from austenitic stainless steel sheet and bar respectively. This material does not undergo a ductile-to-brittle transition in the temperature range of interest [i.e., down to -40 ºF (-40 ºC)], and thus does not require evaluation for brittle fracture.

The DU gamma shield material, which is enclosed by the stainless steel housing, was preconditioned in dry ice (-109.2 °F [-78.5 °C]) for the cold certification testing described in

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Appendix 2.12.1, Certification Tests. In addition, the DU shield mat erial has been previously dropped and puncture tested at temperatures less than -20 ºF (-49 ºF to -23 ºF) in the INC OP-100 package (NRC Docket No. 71-9185) and the INC THSC package (NRC Docket No. 71-9360). As documented in the certification test reports8,9,10, the IR-100ST, the IR-100, and the THSC packages passed all of the tests, which included cumulative damage effects, with no loss of shielding or confinement capability. Based on the low temperature testing of the IR-100ST, OP-100, and THSC packages, the brittle fracture of the DU gamma shield component is not of concern.

2.1.2.2.2 Fatigue Because the IR-100ST is an essentially a rigid body, no structural failures of the confinement boundary due to fatigue will occur.

2.1.2.2.3 Buckling The IR-100ST provides only a confinement boundary. For normal condition and hypothetical accident conditions, the confinement boundary (i.e., the DU gamma shield) will not buckled due to free or puncture drops. This conclusion has been demonstrated via full-scale test of the IR-100ST.

2.1.3 Weights and Center of Gravity The maximum gross weight of the IR-100ST is 58 pounds, and the maximum DU gamma shield weight is 38 pounds. The center of gravity for the package is at approximately 51/2 inches from the bottom package and 73/4 inches from the lock assembly end of the package.

2.1.4 Identification of Codes and Standards for Package Design Since the IR-100ST contains a small quantity of Ir-192 or Se-75 radioactive material, and does not contain a pressure boundary, the package is designed to industrial metal fabrication standards.

2.2 Materials

2.2.1 Material Properties and Specifications Mechanical properties for the stainless steel materials utilize d for the structural components of the IR-100ST are provided in this section. Temperature-dependent material properties for structural components are obtained from Section II, Part D, of the ASME Boiler and Pressure Vessel (B&PV) Code11. Since the evaluation of the IR-100ST is primarily via test, only the material properties that are utilized in the analysis portion of the evaluation are provided. Table 2.2-1 presents the properties of the structural materials utilized in the package. Table 2.2-2 presents the typical properties for the non-structural material s utilized in the package.

8 Orano Federal Services, LLC, Document TR-3026798, Certification Test Report for the Industrial Nuclear Company IR-100ST Package, Revision 0, February 2024.

9 Packaging Technology, Inc., PacTec Document TR-002, Certification Test Report for the OP-100 Package, Revision 1, March 1998.

10 Orano Federal Services, LLC, Document TR-3021828, Certification Test Report for the INC Ten Hole Source Changer, Revision 1, September 2018.

11 American Society of Engineers (ASME) Boiler and Pressure Vesse l Code,Section II, Materials, Part A - Ferrous Material Specifications, and Materials, Part D - Properties, 2017 Edition.

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Table 2.2 Type 304 Stainless Steel Material Properties

Yield Ultimate Design Coefficient of Strength Strength Stress Elastic Thermal (Sy), psi (Su), psi Intensity Modulus, Expansion, x 10-6, Material Temperature, (Sm), psi x 106, psi in/in/ºF Specification ºF

-40 30,000 75,000 20,000 28.8 8.1

-20 30,000 75,000 20,000 28.7 8.2 Type 304 70 30,000 75,000 20,000 28.3 8.5 Stainless Steel 100 30,000 75,000 20,000 28.1 8.6 200 25,000 71,000 20,000 27.5 8.9 300 22,400 66,200 20,000 27.0 9.2 Notes:

ASME B&PV Code,Section II, Part D, Table Y-1 ASME B&PV Code,Section II, Part D, Table U-2 ASME B&PV Code,Section II, Part D, Table 2A ASME B&PV Code,Section II, Part D, Table TM-1, Material Group G ASME B&PV Code,Section II, Part D, Table TE-1, Material Group 3, Mean When necessary, values are linearly interpolated or extrapolated and given in bold text.

The weight density and Poisson's ratio for stainless steel are 0.285 lbm/in3 and 0.29, respectively Table 2.2 Non-Structural Typical Material Properties

Coefficient of Temperature, Flexural Linear Thermal Tensile Strength Modulus, x103, Expansion, x 10-6, Material ºF (St), psi psi in/in/ºF Bayflex 110-50 -22 N/A 115.0 Urethane 73 3,500 52.0 < 42 (Unreinforced) 149 N/A 38.0

ULTEM 1000 Resin 100 16,000 479.0 - 500.0 29.0 - 31.0 Notes:

Bayflex 110-50 Technical Data Sheet, Edition 2012-05-31 SABIC Polyetherimide Resin Technical Data Sheet Applicable from -40 °F to 300 °F

2.2.2 Chemical, Galvanic, or Other Reactions The housing that contains the DU gamma shield casting is fabricated from Type 304 stainless steel.

The stainless steel housing does not have significant reactions with the interfacing components, air, or water. The DU casting, which is coated with enamel paint, is further encased by polyurethane foam. Copper shims are placed between the interface between the DU and stainless steel to prevent a eutectic reaction. The "S" tube (made of either titanium or Zircaloy) does not react or form a galvanic corrosion cell between the DU gamma shield material and the stainless steel.

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The urethane elastomer molded sensor/handle jacket that surrounds the stainless steel housing does not chemically react with the metallic stainless steel housing of the package.

2.2.3 Effects of Radiation on Materials The gamma radiation associated with the Ir-192 or Se-75 radioactive material will have no effect on the austenitic stainless steel and depleted uranium (DU) comprising the structural materials of the IR-100ST. As discussed in Section 2.1.1, Discussion, the interior polyurethane foam only provides moisture protection of the DU gamma shield. The effect of the radiation on the polyurethane foam to provide this protection is negligible. Since the urethane elastomer molded sensor/handle jacket is secured to the stainless steel housing, there is no effect due to radiation on the package to provide its radiation safety functions from the urethane sensor surround.

2.3 Fabrication and Examination

2.3.1 Fabrication The IR-100ST is fabricated utilizing conventional metal forming and joining techniques. Materials are procured in accordance with the standards delineated on the drawings in Appendix 1.3.1, General Arrangement Drawings. All welding procedures and welding personnel are qualified in accordance with Section IX of the ASME Boiler and Pressure Vessel (B&PV) Code12.

2.3.2 Examination The primary safety function of the IR-100ST is to provide gamma shielding of the special form radioactive material. To verify this function, each DU gamma shield is examined by performing a shielding test, as delineated in Section 8.1.6, Shielding Tests, prior to being used in the fabrication of an IR-100ST packaging. In addition, all welds are visually inspected in accordance with the notes identified in Appendix 1.3.1, General Arrangement Drawings.

2.4 General Requirements for All Packages The IR-100ST is evaluated, with respect to the general standards for all packaging specified in 10 CFR §71.436. Results of the evaluations are discussed in the following sections.

2.4.1 Minimum Package Size The smallest overall dimension of the IR-100ST Exposure Device package is 61/4 inches in width.

This dimension is greater than the minimum dimension of 4 inches specified in 10 CFR

§71.43(a). Therefore, the requirements of 10 CFR §71.43(a) are satisfied by the IR-100ST.

2.4.2 Tamper Indicating Device A tamper indicating seal (wire/lead security seal) is attached to the dust cap of the lock assembly (refer to Figure 2-1), which provides visual evidence that the closure was not tampered during transport. Thus, the requirements of 10 CFR §71.43(b) are satisfied.

12 American Society of Mechanical Engineers (ASME), Boiler and Pressure Vessel Code,Section IX, Qualification Standard for Welding and Brazing Procedures, Welders, Brazers, and Welding and Brazing Operators, 2017 Edition.

23 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

2.4.3 Positive Closure The IR-100ST cannot be opened inadvertently. Positive closure of the IR-100ST is provided by the lock assembly that secures the source pigtail assembly in i ts proper shielded position. The lock assembly, which permits access to the contents, conforms to the requirements of 10 CFR

§34.235. Thus, the requirements of 10 CFR §71.43(c) are satisfied.

2.4.4 Valves Because the IR-100ST is a confinement system and designed to transport only a special form radioactive material capsule, there are no valves or other pressure retaining devices on the package. Therefore, the requirements of 10 CFR §71.43(e) are satisfied.

2.4.5 Package Design As shown in Sections 2.6, Normal Conditions of Transport, 3.3, Thermal Evaluation under Normal Conditions of Transport, and 5.4, Shielding Evaluation, the IR-100ST design satisfies the requirements of 10 CFR §71.71. Thus, the requirements of 10 CFR §71.43(f) are satisfied.

2.4.6 External Temperatures The decay heat load of the Ir-192 or Se-75 special form capsule (0.84 W) is negligible. For the PM Tag, the average power per hour while transmitting is 0.122 W, which is negligible. When the power cells are fully discharged, the maximum power during charging in transit is 2.50 W (8.54 Btu/hr), which occurs for a maximum of 4 to 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months . Note that power cell charging only occurs when the IR-100ST is transported within an enclosed secure transport box in an enclosed transport vehicle. Therefore, this shipping configuration will significantly affect the package surface temperature for the conditions specified in 10 CFR §71.43(g). As noted in Section 3.3, Thermal Evaluation for Normal Conditions of Transport, the maximum surface temperature does not exceed 185 ºF (85 ºC) in still air and shade during transport. Thus, the requirements of 10 CFR §71.43(g) are satisfied by the IR-100ST.

2.4.7 Venting With an Ir-192 or Se-75 special form source capsule encapsulating the radioactive material, the package does not incorporate any feature that would permit cont inuous venting during transport.

Thus, the requirements of 10 CFR §71.43(h) are satisfied by the IR-100ST.

2.5 Lifting and Tie-down Devices for All Packages

2.5.1 Lifting Devices The IR-100ST is manually lifted by the handle (refer to Figure 2-1). A force of 182.2 lbf, which greater than 3 times the 58-lbm package gross weight, was applied to the handle without any measurable distortion or damage. Therefore, the handle can support three times the package weight and the requirements of 10 CFR §71.45(a) are met.

2.5.2 Tie-Down Devices Due to the handle design of the sensor surround, an IR-100ST package may be secured utilizing the handle as a tie-down device. There are no other features on the IR-100ST package that may be

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utilized as a tie-down device. Therefore, an IR-100ST package was tested utilizing nylon straps to secure the package to a horizontal surface. Loads of 583.4 lb f, 294.6 lbf, and 120.0 lbf were then applied for a minimum 5 minutes in the longitudinal, lateral, and vertical directions, respectively, as shown in Figure 2.5-1. No damage of the sensor surround or the package was observed.

Therefore, the 10-5-2 requirements of 10 CFR §71.45(b) are met by the 58-lb m IR-100ST package.

Figure 2.5-1 IR-100ST Tie-down Loading Configuration

2.6 Normal Conditions of Transport

2.6.1 Heat During fabrication, the IR-100ST stainless steel housing and DU gamma shield were exposed to a maximum temperature of 250 ºF (121 °C) over 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months during the foam cure. No loss in operational capability or damage occurred. As noted in Chapter 3.0, Thermal Evaluation, the average steady state temperature of any component in an ambient environment of 100 ºF (38 °C) and full insolation is 131 ºF (55 °C). with a maximum surface temperature of 155 ºF (68 °C) on the upper horizontal stainless steel surface.

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2.6.2 Cold The IR-100ST stainless steel body was exposed to a dry ice environment (-109 ºF [-40 ºC]) for an extended period of time in an ice chest without detrimental effects. However, the bottom of the urethane sensor surround was found to be cracked due to exposure to the extreme cold of the dry ice, as noted in Appendix 2.12.1, Certification Tests.

2.6.3 Reduced External Pressure The IR-100ST is a confinement boundary for a special form payload and does not contain a pressure boundary. Therefore, the effect of reduced external pressure is not applicable.

2.6.4 Increased External Pressure The IR-100ST is a confinement boundary for special form payload and does not contain a pressure boundary. Therefore, the effect of increased external pressure is not applicable.

2.6.5 Vibration The effects of vibrations induced by normal conditions of trans port on the IR-100 ST package is bounded by operational experience of the similar IR-100 package that has been subjected to both normal conditions of transport as well as rugged field use over an extended period of time (1982 to present). The packages have not experienced any damage or effects due to the vibrations induced by normal conditions of transport.

2.6.6 Water Spray The materials of construction utilized for the IR-100ST are such that the water spray test identified in 10 CFR §71.71(c)(6) will have a negligible effect on the package.

2.6.7 Free Drop Since the gross weight of the IR-100ST is less than 11,000 pounds, a four-foot free drop is required per 10 CFR §71.71.(c)(7). As discussed in Appendix 2.12.1, Certification Tests, a NCT, four foot drop, aligned the center-of-gravity (CG) over the lock assembly lower edge, was performed on a IR-100ST certification test unit (CTU) as an initial condition for subsequent hypothetical accident condition (HAC) tests. A radiation survey following certification testing demonstrated the ability of the IR-100ST packaging to maintain its shielding integrity.

Therefore, the requirements of 10 CFR §71.71(c)(7) are satisfied.

2.6.8 Corner Drop This test does not apply, since the materials of construction do not include wood or fiberboard, as delineated in 10 CFR §71.71(c)(8).

2.6.9 Compression A 299.7-pound force, which is greater than five times the gross package weight, was applied to the IR-100ST handle while sitting in its normal upright position. No observable deformation or damage was detected. Therefore, the requirements of 10 CFR §71.71(c)(9) are satisfied.

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2.6.10 Penetration For the similar IR-100 package13, a 11/4 inch diameter, 13 pound, hemispherical-end steel bar was dropped from a height of one meter (40 inches) onto the package in an effort to pierce the housing, and possibly damaged the lock assembly. These orientations were an effort to shift the source out of the "safe" area of the DU gamma shielding. Three drop tests were performed. The first two drop tests were onto the outlet end (safety plug end) and a side. Both drops resulted in a 3/16-inch spherical dent in the impacted surface. The third drop test was onto the side of the lock body in an effort to shear or bend the lock assembly away from the housing. The result of the third drop was a 3-degree angular shift of the IR-100 lock assembly relative to the housing. The penetration tests demonstrated that the weld integrity was not compromised, there was no damage to the source pigtail assembly, and there was no loss in operational capability of the basic IR-100 package.

Since the IR-100ST stainless steel body is identical to the IR-100 package, and the lock assembly of the IR-100ST is significantly more robust than the IR-100 lock assembly, the previous penetration tests for the IR-100 package bounds the response of the IR-100ST package to the penetration test. Therefore, the requirements of 10 CFR §71.71(c)(10) are satisfied.

2.7 Hypothetical Accident Conditions When subjected to the hypothetical accident conditions as speci fied in 10 CFR §71.73, the IR-100ST meets the performance requirements specified in Subpart E of 10 CFR 71. This conclusion is demonstrated in the following subsections, where each accident condition is addressed, and the package is shown to meet the applicable design criteria. The method of demonstration is primarily by test. The loads specified in 10 CFR §71.73 are applied sequentially, per Regulatory Guide 7.8.

Test results are summarized in Section 2.7.7, Summary of Damage, with details provided in Appendix 2.12.1, Certification Tests.

2.7.1 Free Drop Subpart F of 10 CFR 71 requires that a 30-foot (9-meter) free drop to be considered for the IR-100ST. The free drop is to occur onto a flat, essentially unyielding, horizontal surface, and the package is to strike the surface in an orientation for which the maximum damage is expected.

The free drop is addressed by test, in which several orientations are used. The free drop proceeds both the puncture and fire tests.

2.7.1.1 Technical Basis for the Free Drop Tests To properly select a worst-case package orientation for the 30 foot (9 meter) free drop event, items that could potentially compromise shielding integrity and/or the special form source of the IR-100ST must be clearly identified. For the IR-100ST design, the foremost item to be addressed is the shielding integrity.

The shielding integrity may be compromised by two methods: 1) significant movement of the special form source from its stored shielded position in the DU gamma shield, and/or 2) as a result of thermal degradation of the DU gamma shield itself by excessive oxidation in a subsequent fire

13 Safety Analysis Report, IR-100 Exposure Device, Industrial Nuclear Company, Inc., NRC Docket No. 71-9157.

27 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

event. Importantly, these methods require significant damage to the stainless steel housing and/or lock assembly that secures the special form source.

For the above reasons, testing must include orientations that affect the lock assembly, which secures the special form source, and/or damage the welded stainless steel body, which may result in an excessive opening into the housing cavity for a subsequent fire event. Therefore, an orientation that places the center-of-gravity (CG) over each of these items was included in the test sequence.

2.7.1.2 Test Sequence for the Selected Tests Based on the above discussions, the IR-100ST was tested for two specific, HAC 30 foot (9 meter) free drop conditions: 1) CG-over-the lock assembly, and 2) an impact on the lower edge of the lock assembly. Although only a single worst-case 30 foot (9 meter) drop is required by 10 CFR

§71.73(c)(1), multiple tests were performed to ensure that the most vulnerable package features were subjected to worst-case loads and deformations. The specific conditions selected for the IR-100ST Certification Test Units (CTUs) are summarized in Table 2.7-1.

2.7.1.3 Summary of Results from the Free Drop Tests Successful HAC free drop testing of the CTUs indicates that the various IR-100ST design features are adequately designed to withstand the HAC 30 foot (9 meter) free drop event. The most important result of the testing program was the demonstrated ability of the IR-100ST to maintain its shielding integrity. Significant results of the free drop testing are as follows:

No evidence of distortion or damage of the lock assembly occurred that resulted in significantly displacement of the special form source from its shielded position.

There was no evidence of rupturing of the welded stainless steel housing that could have resulted in thermal degradation of the DU gamma shield by excessive oxidation in a subsequent fire event.

Further details of the free drop test results are provided in Appendix 2.12.1, Certification Tests.

2.7.2 Crush The crush test specified in 10 CFR §71.73(c)(2) is only required when the specimen has mass not greater than 1,100 lbs. (500 kg), an overall density not greater than 62.4 lbm/ft3 (1,000 kg/m3),

and radioactive contents greater than 1,000 A2, not as special form. The IR-100ST density is greater than 62.4 lbm/ft3 (1,000 kg/m3) and the payload is special form. Therefore, the dynamic crush test of 10 CFR §71.73(c)(2) is not applicable to the IR-100ST.

2.7.3 Puncture Subpart F of 10 CFR 71 requires performing a puncture test in accordance with the requirements of 10 CFR §71.71(c)(3). The puncture test involves a 40 inch (1 meter) drop onto the upper end of a solid, vertical, cylindrical, mild steel bar mounting on an essentially unyielding, horizontal surface. The bar must be 6 inches (15 cm) in diameter, with the top surface horizontal and its edge rounded to a radius of not more than 1/4 inch (6 mm). The minimum length of the bar is required to be 8 inches (20 cm). The ability of the IR-100ST to adequately withstand this specified drop condition is demonstrated via testing of two full-scale, IR-100ST CTUs.

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2.7.3.1 Technical Basis for the Puncture Drop Tests To properly select a worst-case package orientation for the puncture drop event, items that could potentially compromise shielding integrity and/or the special form source of the IR-100ST must be clearly identified. For the IR-100ST design, the foremost item to be addressed is the shielding integrity.

The shielding integrity may be compromised by two methods: 1) movement of the special form source relative to the DU gamma shield, and/or 2) as a result of thermal degradation of the DU gamma shield itself by excessive oxidation in a subsequent fire event. Importantly, these methods require significant damage to the stainless steel housing and/or lock assembly that secures the special form source.

For the above reasons, testing must include orientations that affect the lock assembly, which secures the special form source, and/or damage the welded stainless steel body, which may result in an excessive opening into the housing cavity for a subsequent fire event. Therefore, orientations that places the CG over the lock assembly were included in the test sequence. These orientations were also utilized for the HAC 30 foot (9 meter) free drops and hence, would expect to produce the worst-case cumulative damage to the package.

2.7.3.2 Test Sequence for the Selected Tests Based on the above general discussions, the CTUs were specifically tested for three HAC puncture drop conditions as part of the certification test program. Although only a single worst case puncture drop is required by 10 CFR §71.73(c)(3), multipl e tests were performed to ensure that the most vulnerable package features were subjected to worst case loads and deformations.

The specific conditions selected for the IR-100ST Certification Test Units (CTUs) are summarized in Table 2.7-1.

2.7.3.3 Summary of Results from the Puncture Drop Tests Successful HAC puncture drop testing of the CTUs indicates that the various IR-100ST design features are adequately designed to withstand the HAC puncture drop event. The most important result of the testing program was the demonstrated ability of the IR-100ST to maintain its shielding integrity. Significant results of the puncture drop testing are as follows:

No evidence of distortion or damage of the lock assembly occurred that resulted in significantly displacement of the special form source from its shielded position.

There was no evidence of rupturing of the stainless steel housing that could have resulted in thermal degradation of the DU gamma shield by excessive oxidation in a subsequent fire event.

Further details of the free drop test results are provided in Appendix 2.12.1, Certification Tests.

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2.7.4 Thermal Subpart F of 10 CFR 71 requires performing a thermal test in accordance with the requirements of 10 CFR §71.73(c)(4). To demonstrate the performance capabilities of the IR-100ST package when subjected to the HAC thermal test specified in 10 CFR §71.71(c)(4), a similar full-scale IR-100 CTU was previously exposed to a minimum of 1,475 ºF (800 ºC) for 30 minutes in a vented electric oven. Prior to the thermal test, the IR-100 CTU was subjected to a number of 30-foot free drop and puncture tests, as discussed in Section 2.71, Free Drop, and Section 2.7.3, Puncture of the IR-100 Safety Analysis Report14.

The individual cells within the LiFePO4 power cells would be expected to exceed the threshold temperature necessary to exceed thermal runaway during the HAC thermal test. For power cells, the average temperature rise for the onset of thermal runaway was experimentally determined to be a maximum of 495 °F (257 °C)15. Assuming that all four lithium power cells are fully charged and have undergone thermal runaway, the maximum internal cell temperatures could be over 1,832 °F (1,000 °C), which is significantly lower than the melting temperature of housing stainless steel (2,800 °F [1,538 °C]).

For possible ignition of vented gases from thermal runaway of the batteries, the PM Tag with the lithium power cells is located below the stainless steel housing and are aligned parallel with the longitudinal axis of the IR-100ST. Conservatively assuming the urethane sensor surround does not melt or burn from the 30-minute 1,475 ºF (800 ºC) fire, any flames that may be generated from the batteries would not directly impinge on the stainless steel housing, and would be directed away from the package housing. This potential thermal event would minimize any temperature increase to the stainless steel housing during the HAC thermal test. Therefore, the PM Tag does not affect the safety function of the IR-100ST package to retain and shield the special form radioactive capsule within the stainless steel housing.

Previous successful HAC thermal testing of the IR-100 CTU demonstrates that the various IR-100ST design safety features are adequately designed to withstand the HAC thermal test event. The most significant result of the both the IR-100 and IR-100ST testing programs demonstrated the ability of the IR-100ST packaging to maintain its shielding integrity, as demonstrated by an actual post-test radiation surveys.

2.7.5 Immersion - Fissile Material The IR-100ST does not carry fissile material, and therefore, this section does not apply.

2.7.6 Immersion - All Packages The IR-100ST is a confinement boundary for special form payload and does not have a pressure boundary. Therefore, the effect of pressure is not applicable.

14 Safety Analysis Report, IR-100 Exposure Device, Industrial Nuclear Company, Inc., NRC Docket No. 71-9157.

15 Fire Hazard Analysis for Various Lithium Batteries, DOT/FAA/TC-16/17, March 2017.

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2.7.7 Deep Water Immersion Test (for Type B Packages Containing More than 105 A2)

The IR-100ST contains a maximum of 120 Ci (4.44 TBq) of Ir-192 or Se-75, which have A2 values of 16 Ci (0.6 TBq) and 81 Ci (3.0 TBq), respectively. Since the IR-100ST does not contain more than 105 A2 quantities of radioactive material, this section does not apply.

2.7.8 Summary of Damage As discussed in the previous sections, the cumulative damaging effects of free drop, puncture drop, and thermal tests were satisfactorily withstood by the IR-100ST certification testing.

Subsequent radiation post-test survey and destructive examinations of the CTUs confirmed that shielding integrity was maintained throughout the test series. Therefore, the requirements of 10 CFR §71.73 have been adequately satisfied.

2.8 Accident Conditions for Air Transport of Plutonium This section does not apply since plutonium is not shipped in the IR-100ST.

2.9 Accident Conditions for Fissile Material Packages for Air Transport This section does not apply since fissile material is not shipped in the IR-100ST.

2.10 Special Form The contents of the IR-100ST are a special form Ir-192 or Se-75 source capsule. All source capsules are limited to a maximum of 120 curries. The special form certifications for the Ir-192 or Se-75 capsules are as follows:

Manufacture Model Number Certification Number

Industrial Nuclear Co., Inc. A* USA/0297/S 791* USA/0393/S

Source Production &

Equipment Co., Inc. VSe Source Capsule* USA/0785/S-96

Note: Source capsule is limited to a maximum of 120 Ci of Ir-192 or Se-75 material, as applicable.

2.11 Fuel Rods This section does not apply since fuel rods are not shipped in the IR-100ST.

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2.12 Appendix 2.12.1 Certification Tests

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2.12.1 Certification Tests Presented herein are the results of normal conditions of transport (NCT) and hypothetical accident condition (HAC) test that address free drop and punctu re drop test performance requirements of 10 CFR 7116.

2.12.1.1 Introduction The IR-100ST, when subjected to the sequence of HAC tests specified in 10 CFR §71.73, subsequent to the NCT tests specified in 10 CFR §71.71, is shown to meet the performance requirements specified in Subpart E of 10 CFR 71. As indicated in the introduction to Chapter 2.0, Structural Evaluation, the primary proof of performance for the HAC tests is via the use of full-scale testing. In particular, free drop and puncture drop tests of IR-100ST CTUs confirms that the packaging will retain its shielding integrity following a worst-case HAC sequence.

2.12.1.2 Summary As seen in the figures presented in Section 2.12.1.7, Test Results, successful testing of the CTUs indicates that the various IR-100ST packaging design features are adequately designed to withstand the HAC tests specified in 10 CFR §71.73. The most important result of the testing program was the demonstrated ability of the IR-100ST packaging to maintain its shielding integrity.

Significant results of the free drop tests are as follows:

No evidence of distortion or damage of the lock assembly occurred that would have significantly displaced the special form source from its desired shielded position.

There was no evidence of rupturing of the stainless steel housing that could have resulted in thermal degradation of the DU gamma shield by excessive oxidation in a subsequent fire event.

Significant results of the puncture drop testing are as follows:

No evidence of distortion or damage of the lock assembly occurred that would have significantly displaced the special form source from its desired shielded position.

There was no evidence of rupturing of the stainless steel housing that could have resulted in thermal degradation of the DU gamma shield by excessive oxidation in a subsequent fire event.

2.12.1.3 Test Facilities The drop testing was performed utilizing a horizontal concrete slab, which is approximately 9-12 inches thick 10 feet 15 feet. A 2 inch 48 inch 48 inch steel plate was placed on top of the concrete slab, grouted, and secured to the concrete pad by four (4) 5/8-inch anchor bolts.

Considering only the concrete directly underneath the steel plate as being effective with a minimum thickness of 4 inches, the estimated mass of the drop pad (concrete and steel plate) is 2,100 lbm, which is more than 36 times the mass of the IR-100ST CTU. Based on these characteristics, the drop pad conservatively satisfies the requirement of 10 CFR §71.73 for an essentially unyielding, horizontal surface.

16 Title 10, Code of Federal Regulations, Part 71 (10 CFR 71) Packaging and Transportation of Radioactive Material, 1-1-24 Edition.

34 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

The puncture bar for the puncture tests is a 6 inch (15 cm) diameter 12-inch long solid bar that is orthogonally socket welded through a 3/4-inch 12 inch 12 inch steel plate. The top circumferential edge of the bar has a maximum 1/4-inch (6-mm) radius. The free length of the bar is 12 inches, thus ensuring an adequate length to potentially cause maximum damage to the CTU as required by 10 CFR §71.73(c)(3). Following the thirty foot free drop tests, the 3/4-inch plate of the puncture bar assembly was then welded to the 2-inch thick plate on the drop pad to ensure that the puncture bar is restrained for the puncture drop tests.

2.12.1.4 Certification Test Unit Description The IR-100ST consists of a Zircaloy or titanium source tube surrounded by an epoxy-coated, depleted uranium (DU) gamma shield. The DU gamma shield assembly is encased within a welded, Type 304 stainless steel housing. Stainless steel support brackets, welded to the inner housing surface, capture the DU gamma shield between the support bracket and the inner surface of the stainless steel housing. Copper shim stock is installed between the DU-stainless steel interfaces to preclude a reaction between the two dissimilar metals. The void space between the DU gamma shield assembly and the inner stainless steel housing is filled with approximately 2 pounds of rigid polyurethane foam that prevents moisture from contacting the DU material. The maximum gross weight of the IR-100ST Exposure Device is 58 pounds.

Prior to free drop, puncture, and thermal testing, two IR-100ST CTUs were loaded with a dummy source capsule assembly to simulate the special form capsule. The actual weight of the CTU-1 and CTU-2 were 55.9 and 56.4 pounds, respectively. Aside from the dummy source capsule assembly, the CTUs were identical to the IR-100ST packaging design depicted in Appendix 1.3.1, General Arrangement Drawings.

2.12.1.5 Technical Basis for Tests For the confinement system to fail, the IR-100ST would need to move or separate the radioactive source from the central location within the DU gamma shield assembly. This potential failure mode may only occur if either or both of the following conditions occur:

1. The lock assembly of the IR-100ST is broken free of the stainless steel housing or damaged such that the source is significantly moved from its stored position.

2. The DU gamma shield assembly translates away from the lock assembly/pigtail assembly and the source is significantly moved from its desired stored position.

For either of these potential conditions to be initiated, the IR-100ST would need to sustain significant damage due to the normal and hypothetical accident condition free drops and then sustain further damage due to the 40-inch (1-meter) drop onto a 6-inch diameter vertical steel bar. Therefore, the primary objective of the 4-ft (1.2-meter) normal condition and 30-ft (9-meter) hypothetical accident condition (HAC) free drops is to damage the lock assembly or cause significant movement of the special form source within the DU gamma shield of the IR-100ST packaging. A secondary objective of the 30-ft (9-meter) HAC free drops is to fail the welded stainless steel body such that a potential air pathway into the interior would form. Such a pathway could potentially result in a self-sustaining oxidation reaction of the DU and hence, result in a loss of shielding.

The following sections provide the technical basis for the chosen test orientations and sequences for the IR-100ST CTUs as presented in Appendix 2.12.3.6, Test Sequence for Selected Free Drop and Puncture Drop Tests.

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2.12.1.5.1 Temperature Both hot and cold temperatures were utilized at the time of IR-100ST NCT and HAC free drop certification testing. For the NCT and HAC hot condition, CTU-1 was pre-heated to a temperature greater than 133 °F. For the HAC cold condition, CTU-2 was pre-cooled in dry ice to a temperature below -20 °F. The results of the certification tests demonstrated that extreme temperatures had no effect on the shielding integrity of the DU gamma shield in the IR-100ST.

In addition, the austenitic stainless steel and DU materials are not susceptible to brittle fracture, as delineated in Section 2.1.2.2.1, Brittle Fracture.

2.12.1.5.2 Free Drop Tests The IR-100ST is qualified primarily by full-scale testing, with acceptance criterion being the ability to demonstrate shield integrity. Per 10 CFR §71.73(c)(1), the package is required to strike an essentially unyielding surface in a position for which maximum damage is expected.

Therefore, for determining the drop orientations that satisfy the regulatory maximum damage requirement, attention is focused predominately on the issue of shield integrity.

To maximize the damage to the Model IR-100ST and potentially separating the radioactive source, two orientations have been selected for the free drop testing:

1. CG-Over-Lock Assembly (Hot & Cold): This orientation targets the lock assembly that secures the special form source in the DU shield for both the NCT hot and the HAC -20 °F HAC cold conditions. Should this impact be sufficiently severe, the lock assembly may be dislodged and/or broken and allow the special form source to separate from the package body.

2. CG-Over-Lock Assembly Lower Edge (Hot & Cold): This orientation again targets the lock assembly by attacking the lower edge to potentially pry the assembly off of the body for both the NCT hot and the HAC -20 °F HAC cold conditions. Should this impact be sufficiently severe, the lock assembly may be dislodged and/or broken and allow the special form source to separate from the package body.

2.12.1.5.3 Puncture Drop Tests 10 CFR §71.73(c)(3) requires a free drop of the specimen through a distance of 40 inches (1 meter) onto a puncture bar in a position for which maximum damage is expected. As in Section 2.12.1.5.2, Free Drop Tests, the maximum damage criterion is evaluated primarily in terms of loss of shielding integrity. Loss of shielding integrity could occur directly by dislodging the lock assembly body and/or broken and allow the special form source to separate from the IR-100ST body.

All puncture orientations are oblique per the orientations identified above in Section 2.12.1.5.2, Free Drop Tests. Should a condition surface during the certification testing that results in unanticipated damage, then a new evaluation and assessment to determine most-damaging orientation(s) for the puncture drop test will be performed.

2.12.1.6 Test Sequence for Selected Free and Puncture Drop Tests The following sections establish the selected free drop and puncture drop test sequence for the IR-100ST CTUs based on the discussions provided in Section 2.12.1.5, Technical Basis for

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Tests. The tests sequences are summarized in Table 2.12.1-1 and ill ustrated in Figure 2.12.1-1 and Figure 2.12.1-2.

2.12.1.6.1 Certification Test Unit No. 1 (CTU-1)

All drop tests for CTU-1 are performed in the NCT hot condition.

Free Drop No. 1 is a NCT free drop from a height of 4 feet, with CG over the lock assembly. The 4 foot drop height is based on the requirements of 10 CFR

§71.71(c)(7) for a package weight not exceeding 11,000 pounds. The purpose of this test was to cause maximum damage to the most vulnerable feature (lock assembly) on the packaging.

Free Drop No. 2 is a HAC free drop from a height of 30 feet, with CG over the lock assembly, which is the same impact pint as the NCT Free Drop No. 1. In this way, 4' DR NCT and HAC free drop damage is cumulative. The 30 30' foot drop height is based on the requirements of 10 CFR

§71.73(c)(1). The purpose of this test is to cause l ~

maximum damage to the most vulnerable feature (lock assembly) on the packaging. ~

Free Drop No. 3 is a HAC free drop from a height of 30 feet, impacting the lower edge of the sensor surround at a shallower angle than Free Drop No. 2. The 30 foot drop height is based on the requirements of 10 CFR

§71.73(c)(1). The purpose of this test was intended to cause maximum damage to the sensor surround in an attempt to shear the lock assembly off the housing, and displace the special form capsule from its desired t shielded position. 30' l.....,....,..,.

~

Puncture Drop No. 6 impacts directly onto the damage created by Free Drop Tests 1 and 2, directly on the lower edge of the lock assembly. The puncture drop height is based on the requirements of 10 CFR §71.73(c)(3). The purpose of Puncture Drop No. 6 is to cause maximum damage to the most vulnerable feature (lock assembly) on the packaging.

o*

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2.12.1.7 Test Results The following sections report the results of free drop, puncture drop, and thermal tests following the sequence provided in Section 2.12.1.6, Test Sequence for Selected Free Drop, Puncture Drop, and Thermal Tests. Results are summarized in Table 2.12.1-1 (refer also to Figure 2.12.1-1 and Figure 2.12.1-2).

Figure 2.12.1-3 through Figure 2.12.1-25 sequentially photo-document the certification testing process for the IR-100ST CTUs.

2.12.1.7.1 Certification Test Unit No. 1 (CTU-1)

2.12.1.7.1.1 CTU-1 Free Drop Test No. 1 Free Drop No. 1 is a NCT free drop from a height of 4 feet, with CG over the lock assembly. As shown in Figure 2.12.1-3, the CTU was oriented 74º with respect to the horizontal impact surface (longitudinal angle 74º, rotational angle 0º). The following list summarizes the test parameters:

verified longitudinal angle as 74° +/-1° verified rotational angle as 0º 1º verified drop height as 4 feet (1.2 meter), +3/-0 inches (actual drop height 4 feet) measured surface temperature of SST sheet metal as greater than 133 °F conducted test at 1:28 p.m. on Tuesday, 12/12/2023 The package impacted the drop pad and resulted in the cap for the keyed lock separating from assembly. No further damage was observed. The impact and subsequent damage are shown in Figures 2.12.1-3 and 2.12.1-4.

2.12.1.7.1.2 CTU-1 Free Drop Test No. 2 Free Drop No. 2 is a HAC free drop from a height of 30 feet, with CG-over-the lock assembly. As shown in Figure 2.12.1-5, the CTU was oriented 71º with respect to the horizontal impact surface (longitudinal angle 71º, rotational angle 0º). The following list summarizes the test parameters:

verified longitudinal angle as 71º 1º verified rotational angle as 0º 1º verified drop height as 30 feet (9 meter), +2/-0 inches measured surface temperature of package sheet metal as greater than 133 °F conducted test at 1:53 p.m. on Tuesday, 12/12/2023 Upon observation, it was noted that the test unit did not impact the lock assembly as intended.

Therefore, a second HAC 30-foot free drop was immediately initiated for this orientation. The following list summarizes the test parameters for the subsequent HAC free drop:

verified longitudinal angle as 72° +/-1° verified drop height as 30 feet (9 meter), +2/-0 inches measured surface temperature of package sheet metal as great than 133 °F at time of test conducted 2nd test on Tuesday, 12/12/2023

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The second test impacted the lock assembly as intended. Other than minor deformation on the lock assembly, no further damage was noted. With the lock cap missing, the keyed lock did unlock from the impact, but was push inward and reset without any difficulty. Figures 2.12.1-5 thru 2.12.1-10 photo-documents Test Number 2 for both free drop tests of CTU-1.

2.12.1.7.1.3 CTU-1 Free Drop Test No.3 Free Drop No. 2 is a HAC free drop from a height of 30 feet, with CG-over-lower edge. As shown in Figure 2.12.1-11, the CTU was oriented 53.5º with respect to the horizontal impact surface (longitudinal angle 53.5º, rotational angle 0º). The following list summarizes the test parameters:

verified longitudinal angle as 53.5° +/-1° verified drop height as 30 feet (9 meter), +2/-0 inches measured surface temperature of package sheet metal as greater than 133 °F at time of test conducted test at 2:05 p.m. on Tuesday, 12/12/2023 During the free fall, the package rotated such that the test unit impacted the upper edge of the stainless steel package. The stainless steel upper edge slightly deformed that also deformed the stainless steel sheet metal nameplate. The deformation resulted in no cracks that would have exposed the interior cavity containing the DU gamma shielding. Figures 2.12.1-11 thru 2.12.1-13 photo-document the damage to the CTU.

2.12.1.7.1.4 CTU-1 Puncture Drop Test No. 6 Puncture Drop No. 6 impacted directly onto the damage created by Free Drop Test 2, with the CG-over the upper edge. As shown in Figure 2.12.1-9, the CTU was oriented 66.8º with respect to the horizontal impact surface (longitudinal angle 66.8º, rotational angle 0º). The following list summarizes the test parameters:

verified longitudinal angle as 66.8° +/-1° verified drop height as 40 inch (1 meter), +2/-0 inches temperature of package was at ambient temperature at time of test conducted test at 3:53 p.m. on Tuesday, 12/12/2023 The puncture impact was on the damage to the lock assembly from the Tests 1 and 2. The puncture bar initially struck the lock assembly, impacted the sensor surround that remained, and finally struck the stainless steel upper edge of the inner package. However, no cracks in the stainless steel housing case were observed from the deformation. Figures 2.12.1-14 and 2.12.1-15.

2.12.1.7.2 Certification Test Unit No. 2 (CTU-2)

All drop tests for CTU-2 were performed in the same orientations as CTU-1, except the tests are performed in the NCT cold condition.

2.12.1.7.2.1 CTU-2 Free Drop Test No. 4 Prior to performing the free drop, the bottom of the sensor surround was found to be cracked, as shown in Figure 2.12.1-16. The cracks in the sensor surround were due to the pre-conditioning of the CTU in an ice chest with dry ice to obtain the required regulatory minimum -20 °F cold

39 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

temperature for testing. Since the sensor surround is not required for the shielding safety function of the package, the HAC free drop was performed with the damaged surround as intended.

Free Drop No. 4 a HAC free drop from a height of 30 feet, with the CG-over-the lock assembly.

As shown in Figure 2.12.1-17, the CTU was oriented 71.6º with respect to the horizontal impact surface (longitudinal angle 71.6º, rotational angle 0º). The following list summarizes the test parameters:

verified longitudinal angle as 71.6° +/-1° verified drop height as 30 feet (9 meter), +2/-0 inches measured surface temperature of package sheet metal as less than -21 °F at time of test conducted test at 12:47 p.m. on Tuesday, 12/12/2023 The package impacted the drop pad and resulted in the keyed lock and the source cap being impacted. A small piece of the sensor surround broke off from the package from the impact.

The impact event and damage are shown in Figures 2.12.1-18 thru 2.12.1-20 respectively.

2.12.1.7.2.2 CTU-2 Free Drop Test No. 5 Free Drop No. 4 is a HAC free drop from a height of 30 feet, impacting the lower edge of the sensor surround. As shown in Figure 2.12.1-21, the CTU was oriented 43º with respect to the horizontal impact surface (longitudinal angle 43º, rotational angle 0º). The following list summarizes the test parameters:

verified longitudinal angle as 43° +/-1° verified drop height as 30 feet (9 meter), +2/-0 inches measured surface temperature of SST sheet metal as less than -20 °F at time of test conducted test at 1:00 p.m. on Tuesday, 12/12/2023 The impact broke the lower section of the sensor surround, which separated the lithium power cell pack and circuit board from the test unit. The 3/16-inch stainless steel Pop rivets on the lower half of the sensor surround failed from the impact. No damage to the inner stainless steel package was observed. Figures 2.12.1-22 and 2.12.1-23 photo-document Test Number 5 for the free drop test of CTU-2.

2.12.1.7.2.3 CTU-2 Puncture Drop Test No. 7 Puncture Drop No. 7 was intended to impact directly onto the damage created by Free Drop Test 3, directly impacting the lower edge of the lock assembly at a shallower angle than Free Drop No. 2. This orientation was an attempt to increase the existing opening in the housing and thus, provide more exposure of the DU gamma shield for the subsequent thermal test. As shown in Figure 2.12.1-24, the CTU was oriented 51º with respect to the horizontal impact surface (longitudinal angle 51º, rotational angle 0º). The following list summarizes the test parameters:

verified longitudinal angle as 51º 1º verified rotational angle as 0º 1º verified drop height as 40 inches, +2/-0 inches temperature of package was at ambient temperature at time of tests conducted test at 4:05 p.m. on Tuesday, 12/12/23

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With the lower section of the sensor surround eliminated, the puncture bar struck the lock assembly that resulted in minor deformation. No other damage to the test unit was observed.

The impact damage are shown in Figure 2.12.1-25.

2.12.1.7.2.4 Weight of Test Units Prior to and following the free and puncture drop testing, each IR-100ST CTU assembly was weighed. The recorded pre-test and post test weights are listed in the following table. The 0.4 lbm weight loss for CTU-1 was due to the removal of a damaged nameplate on the top of the package to perform the post-test radiation survey. The 3 lbm weight loss for CTU-2 was due primarily to the loss of the lower half of the sensor surround containing the PM Tag with the lithium power cells and electronics.

Test Unit Number Pre-Test Weight (lb m) Post-Test Weight (lb m)

CTU-1 55.9 55.5 CTU-2 56.4 53.4

2.12.1.7.2.5 Post-Test Radiation Survey Post-test radiation survey of the IR-100ST CTUs were performed on Thursday, 1/25/2024. The post-test radiation survey was performed utilizing an Ir-192 sp ecial form source.

The strength of the source on the day of the survey was 58 Ci (2.15 TBq). To account for the maximum allowable payload of 120 Ci (4.44 TBq) of Ir-192, the measured values were adjusted upward by the ratio of 120/58 or 2.069. The results of the post-test radiation survey are follows:

Test Unit Maximum Dose Rate [Top/Bottom/Side/End] (mrem/hr)

Number Surface 1-meter 2-meter CTU-1 120 29 110 93 2.1 2.1 2.1 2.1 0.0 0.0 0.0 0.0 CTU-2 171 41 114 106 2.1 2.1 2.1 2.1 0.0 0.0 0.0 0.0

As indicated above, the radiation dose levels were below the requirements of 10 CFR §71.47(a) for NCT and 10 CFR §71.51(a)(2) for HAC for a non-exclusive use shipment.

41

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,ii

-~-

,t e

Figure 2.12.1 Schematic Summary of CTU-1 Testing

e

I

-~-

I e ti'-

Figure 2.12.1 Schematic Summary of CTU-2 Testing

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Figure 2.12.1 CTU-1 Free Drop Test No. 1: Immediately Prior to Impact

Figure 2.12.1 CTU-1 Free Drop Test No. 1: Close-up View of Lock Assembly

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Figure 2.12.1 CTU-1 Free Drop Test No. 2: Test Unit Immediately Prior to 1st Free Drop

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Figure 2.12.1 CTU-1 Free Drop Test No.2: Test Unit Prior to Impact 1st Free Drop

Figure 2.12.1 CTU-1 Free Drop Test No.2: Close-up Following 1st HAC Free Drop

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Figure 2.12.1 CTU-1 Free Drop Test No.2: Test Unit Immediately Prior to 2nd Free Drop

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Figure 2.12.1 CTU-1 Free Drop Test No. 2: 2nd Free Drop Test Unit Prior to Impact

Figure 2.12.1 CTU-1 Free Drop Test No. 2: Close-up View Following 2nd Free Drop

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Figure 2.12.1 CTU-1 Puncture Drop Test No. 3: Immediately Prior to Free Drop

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Figure 2.12.1 CTU-1 Free Drop Test No. 3: Close-up View of Impact Area on Upper Edge Due to Rotation During Free Fall

Figure 2.12.1 CTU-1 Free Drop Test No. 3: Close-up Side View of Package

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Figure 2.12.1 CTU-1 Puncture Drop Test No. 6: Test Unit Prior to HAC Puncture Drop

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Figure 2.12.1 CTU-1 Puncture Drop Test No. 6: Close-up View of Upper Edge Damage

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Figure 2.12.1 CTU-2 Free Drop Test No. 4: View of Cracked Sensor Prior to Free Drop

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Figure 2.12.1 CTU-2 Free Drop Test No. 4; Test Unit Prior to Free Drop

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Figure 2.12.1 CTU-2 Free Drop Test No. 4: Test Unit Immediately Prior to Impact

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Figure 2.12.1 CTU-2 Free Drop Test No. 4: Close-up View of Impact Area

Figure 2.12.1 CTU-2 Free Drop Test No. 4; Close-up View of Damage

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Figure 2.12.1 CTU-2 Free Drop Test No. 5; Test Unit Prior to Free Drop

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Figure 2.12.1 CTU-2 Free Drop Test No. 5: Test Unit Immediately After Impact

Figure 2.12.1 CTU-2 Free Drop Test No. 5: Close-up View of Bottom Damage

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Figure 2.12.1 CTU-2 Puncture Drop Test No. 7: Prior to Drop on Lower Edge

Figure 2.12.1 CTU-2 Puncture Drop Test No. 7: Close-up View of Lower Edge

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3.0 THERMAL EVALUATION This chapter establishes the compliance of the IR-100ST transporting a payload of up to 120 Ci (4.44 TBq) of Ir-192 or Se-75 in special form with the thermal requirements of 10 CFR 7117.

3.1 Description of Thermal Design

3.1.1 Design Features The IR-100ST does not contain any specific thermal design features. The thermal performance of the package is demonstrated by test. Therefore, this section does not apply.

3.1.2 Contents Decay Heat The IR-100ST may contain up to 120 Ci (4.44 TBq) of Ir-192 or Se-75 in special form. The radiolytic decay heat of Ir-192 is 7.03 x 10-3 W/Ci18. The radiolytic decay heat of Se-75 is 2.41 x 10-3 W/Ci14. Since the radiolytic decay heat of Ir-192 is greater than the radiolytic decay heat of Se-75, the heatload of Ir-192 payload bounds the Se-75 payload. Therefore, the maximum decay heat load for the IR-100ST package is 0.84 W (2.87 Btu/hr), which is negligible.

3.1.3 Summary Tables of Temperatures Further details of the NCT results are presented in Section 3.3, Thermal Evaluation under Normal Conditions of Transport. Similarly, further discussion of the HAC thermal evaluation is provided in Section 3.4, Thermal Evaluation under Hypothetical Accident Conditions. The maximum surface temperature of the IR-100ST package under NCT conditions in full sunlight is 155 °F (88 °C) on the top horizontal stainless steel surface on the housing.

3.1.4 Summary Tables of Maximum Pressures The containment of the IR-100ST is provided by the special form payload. Gas can freely move from the internal cavity to the environment during all phases of operation. Therefore, there are no internal pressures to be determined since the IR-100ST does not contain any pressure boundaries.

3.2 Material Properties and Component Specifications

3.2.1 Material Properties The IR-100ST is constructed of a 12-gauge (0.105 inch) thick stainless steel outer skin surrounding polyurethane foam and a depleted uranium (DU) gamma shield. Since the structural integrity of the package is established by testing, the only pertinent temperature limits on the components is established by their melting temperatures for the fire based Hypothetical Accident Condition (HAC).

The melting temperatures for DU and stainless steel are 2,071 °F (1,133 °C) and 2,800 °F (1,538 °C),

respectively.

The payload was qualified per Qualification of Special Form Radioactive Material, in 10 CFR §71.75(b)(4).

17 Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material, 1-1-24 Edition.

18 ORIGEN-S Decay Data Library and Half-Life Uncertainties, O. W. Hermann, P. R. Daniel, and J. C. Ryman, Oak Ridge National Laboratory, ORNL/TM-13624, September 1998.

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3.2.2 Component Specifications The IR-100ST does not contain any component or material that is important to the thermal performance of the package. The two primary structural materials are austenitic stainless steel and the DU gamma shield. As noted in Section 2.1.2.2.1, Brittle Fracture, both materials have been tested to temperatures below -20 °F with no loss of structural or shielding capability.

3.3 Thermal Evaluation under Normal Conditions of Transport This section presents the thermal evaluation of the IR-100ST under the normal conditions of transport (NCT) per 10 CFR §71.71.

3.3.1 Heat and Cold Since the total decay heat load is less than 1 W (3 Btu/hr) and the maximum heat from charging the lithium power cells in the PM Tag is 2.5 W (8.54 Btu/hr), a detailed thermal analysis of the IR-100ST internals is unnecessary. The peak internal temperatures will very closely match those on the surface of the package. To determine the NCT maximum package temperatures with and without insolation, an ANSYS steady-state model of the IR-100ST was developed. Per 10 CFR §71.71(c)(1), the worst-case high temperature conditions for the package consist of an ambient temperature of 100 °F (38 °C) and maximum insolation. Under those conditions, the worst-case surface temperature for the IR-100ST would be 155 °F (68 °C) on the top, horizontal metallic surface, with the average surface temperature being 131 °F (55 °C). A view of the ANSYS model results with insolation averaged over a 12-hour period is shown in Figure 3.3-1.

A:s-..,-staten. 1112hrAwer9 Sudlce Temp erlturu Typ e: Temp erature Un it"F Time: 1 J 12/13/202 3 lll 04AM mu..

149 141 m

~ "'

114 11 8 11 2

0.000 4.500 9,000(in)

2.250 6.750

Figure 3.3 IR-100ST NCT Steady-State Temperatures with Full Insolation Given the negligible decay heat, the maximum temperature for all surfaces of the IR-100ST in shade with an ambient temperature of 100 °F (38 °C) will be 100 °F (38 °C). This temperature is below the maximum acceptable surface temperature of 122 °F (50 °C) for non-exclusive use shipments as stipulated in 10 CFR §71.43(g). Similarly, the package temperature will be equal to ambient under the low temperature conditions of -20 °F (-29 °C) and -40 °F (-40 °C).

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3.3.2 Maximum Normal Operating Pressure This section does not apply since the IR-100ST does not contain any pressure boundaries.

Therefore, there is no maximum normal operating pressure (MNOP) for the IR-100ST.

3.4 Thermal Evaluation under Hypothetical Accident Conditions Since the certification testing of the IR-100ST resulted in no breach of the welded stainless steel housing, the thermal performance of the IR-100ST under Hypothetical Accident Conditions (HAC) was determined via previous testing of the similar IR-100 Exposure Device19 in accordance with 10 CFR §71.73. Additional details are provided in the following sections.

3.4.1 Initial Conditions A previously free and puncture dropped IR-100 certification test unit (CTU) package was placed into an oven and exposed to a forced convective environment that resulted in the average surface temperature of the package to at least 1,475 °F (800 °C).

3.4.2 Fire Test Conditions Following the introduction of air and indication of the package surface was at a minimum of 1,475

ºF (800 °C), the package was maintained in the oven for 30 minutes. During the 30-minute test, the surface temperature varied between 1,481 °F and 1,530 °F (805 °C and 832 °C). During heat-up, burning of the polyurethane foam was observed. Following the 30-minute test, the package was removed from the oven and allowed to cool in air.

A post-test examination of the package indicated that the polyurethane foam was completely consumed by the fire, adding its combustion energy to that of the forced convection from the oven. The depleted uranium shielding, and the outer skin of the package were not compromised or appreciably oxidized. Additionally, the peak temperatures recorded in the test were well below the melting temperatures of both stainless steel (2,800 °F) and uranium (2,071°F). An interior view of the disassembled IR-100 CTU showing the interior is presented in Figure 3.4-1.

19 Safety Analysis Report, IR-100 Exposure Device, Industrial Nuclear Company, Inc., NRC Docket No. 71-9157.

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Figure 3.4 Interior View of Disassembled IR-100 CTU Following HAC Thermal Test A post-test radiation survey conducted after the fire test indicated little, if any, degradation in shielding capability. The special form qualification of the payload certifies that it could withstand the fire test without degradation.

Based on the thermal test of the similar IR-100, the IR-100ST package satisfies the HAC thermal requirements set forth in 10 CFR §71.73(c).

3.4.3 Maximum Temperatures and Pressures Based the thermal tests performed on the IR-100, none of the IR-100ST components exceed its temperature limit as described in Section 3.2.1, Material Properties. Specifically, the maximum recorded package temperatures fall more than 500 °F (260 °C) below the melting point of steel and uranium. Additionally, the special form payload does not exceed the temperatures for the special form certification tests.

The containment of the IR-100ST is provided by the special form payload. Gas can freely move from the internal cavity to the environment during all phases of operation, so determination of internal pressures is not required.

3.4.4 Maximum Thermal Stresses The effects of HAC thermal stresses were addressed by the fire test of the IR-100 package. No damage due to thermal stresses was found during post-test examination of the tested CTU.

3.4.5 Accident Conditions for Fissile Material Packages for Air Transport This section does not apply since the IR-100ST does not contain fissile material.

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4.0 CONTAINMENT The IR-100ST is designed as a means of confinement for a special form Ir-192 or Se-75 source capsule. Containment of radioactive material is provided by the special form construction of the payload. The source capsules and their respective special form certification are as follows:

Manufacture Model Number Certification Number

Industrial Nuclear Co., Inc. A* USA/0297/S-96 791* USA/0393/S-96

Source Production &

Equipment Co., Inc. VSe Source Capsule* USA/0785/S-96

Note: Source capsule is limited to a maximum of 120 Ci (4.44 TBq) of Ir-192 or Se-75 material, as applicable.

Since the IR-100ST does not provide containment, subsequent sections of this chapter are not applicable.

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5.0 SHIELDING EVALUATION This section demonstrates the shielding capability of the IR-100ST design for the authorized special form contents. The shielding evaluation is demonstrated via prototypic testing in lieu of an analytical evaluation.

5.1 Description of Shielding Design

5.1.1 Design Features The IR-100ST is a welded structure that contains a depleted uranium (DU) gamma shield, which surrounds a titanium S-tube. A stainless steel special form capsule, which contains 120 Ci (4.44 TBq) of either Ir-192 or Se-75 isotope, is inserted into the S-tube via a pigtail assembly. The radioactive source is positioned at the center of the DU gamma shield to provide the maximum attenuation of the gamma radiation.

5.1.2 Summary Table of Maximum Radiation Levels Table 5-1 provides the maximum measured external radiation levels for the IR-100ST CTUs with the maximum bounding payload content (120 Ci [4.44 TBq] Ir-192) for a non-exclusive use shipment.

Table 5 Maximum Measured External Radiation Levels (Non-Exclusive Use)

Normal Conditions of Transport Hypothetical Accident Conditions Package 10 CFR §71.47(a) 10 CFR §71.51(a)(2)

Measurement Measured* Limit Measured Limit Location mrem/hr (mSv/hr) mrem/hr (mSv/hr) mrem/hr (mSv/hr) mrem/hr (mSv/hr)

Side Surface 114 (1.14) 200 (2) N/A N/A Top Surface 171 (1.71) 200 (2) N/A N/A Bottom Surface 41 (0.41) 200 (2) N/A N/A End Surface 106 (1.06) 200 (2) N/A N/A 40 inches (1 Meter) from Surface* 2.1 (0.02) 10 (0.1) 2.1 (0.02) 1000 (10)

Note: Normal condition measured values are for a test unit that was post-test surveyed from the hypothetical accident conditions tests per 10 CFR §71.73.

5.2 Source Specification

5.2.1 Gamma Source The radioactive content of the IR-100ST is limited to 120 Ci (4.44 TBq) of either Ir-192 or Se-75 isotopes. As shown in Table 5-2, Ir-192 results in a higher unit dose than Se-75 per curie of activity. In addition, the photon energies of Ir-192 (0.380 Me V average) are higher than Se-75 (0.280 MeV average). Therefore, the Ir-192 payload will bound the Se-75 payload for the 120 Ci (4.44 TBq) content limit. Since actual Ir-192 special form capsules are utilized to determine the acceptance of the DU gamma shield, the tabulation of gamma decay source strengths for the special form capsules is not required for the IR-100ST.

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Table 5 Specific Gamma Ray Constants for Iridium and Selenium Isotopes20

Specific Gamma Ray Constant Radionuclide (R-m2/hr-Ci)

Iridium-192 0.460 Selenium-75 0.203

5.2.2 Neutron Source This section does not apply since the IR-100ST does not contain fissile material.

5.3 Shielding Model The shielding capability of the IR-100ST design is demonstrated by physical tests of prototypic packages. Therefore, no analytical shielding model of the package is performed.

5.4 Shielding Evaluation

5.4.1 Methods The method utilized to demonstrate the shielding performance of the IR-100ST is via prototypic testing utilizing a special form capsule containing radioactive Ir-192 material.

5.4.2 Input and Output Data This section does not apply since the shielding performance of the IR-100ST is not performed analytically.

5.4.3 Flux-to-Dose-Rate Conversions This section does not apply since the shielding performance of the IR-100ST is not performed analytically.

5.4.4 External Radiation Levels Following the specified tests of a prototypic package with a 120 Ci (4.44 TBq) of Ir-192 payload per 2.6, Normal Conditions of Transport, and 2.7, Hypothetical Accident Conditions, the maximum radiation level measured on the surface and at 1-meter of the IR-100ST is 171 mrem/hr (1.71 mSv/hr) and 2.1 mrem/hr (0.02 mSv/hr), respectively. As noted in Table 5-1, these levels are significantly below the regulatory limits of 10 CFR §71.47(a) and 10 CFR

§71.51(a)(2).

20 Exposure Rate Constants and Lead Shielding Values for Over 1, 100 Radionuclides, David S. Smith and Michael G. Stabin, Department of Radiology and Radiological Sciences, Vanderbilt University, Nashville, TN, Health Physics Society Journal, March 2012 issue.

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6.0 CRITICALITY EVALUATION

The IR-100ST does not transport fissile material; therefore, this section does not apply.

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7.0 PACKAGE OPERATIONS

7.1 Package Loading This section delineates the procedures for loading a payload into the IR-100ST. Hereafter, reference to specific IR-100ST components may be found in Appendix 1.3.1, General Arrangement Drawings.

7.1.1 Preparation of the IR-100ST for Loading

1) Visually inspect the IR-100ST for damage and/or missing parts.

2) Remove the rear cap and the outlet cap.

3) Inspect the lock assembly for damage. Replace any damage that is noted.

4) Prior to loading an active Ir-192 or Se-75 source into the package, insert a dummy source pigtail, and functionally test the automatic locking device to ensure that all components are operating properly.

5) Pull (retract) the dummy pigtail. The Side Lever will pop-out and lock the dummy source in the stored position. Rotate the key to the locked position and remove the key.

6) Insert the key into the lock, retract (pull) the dummy source pigtail and rotate the key to the unlocked position. Manually depressed the Side Lever to the op erate position and remove the dummy source pigtail.

7.1.2 Loading the Special Form Payload into the IR-100ST

1) Place the special form Ir-192 or Se-75 source pigtail assembly into a source changer.

2) Connect the drive cable housing and the guide tube to the package.

3) Crank the drive cable out through the guide tube and connect it to the Ir-192 source pigtail assembly. Connect the guide tube to the source changer.

4) Unlock the source changer and retract the Ir-192 or Se-75 sourc e pigtail assembly into the package.

5) Survey the package to ensure that the source is in the stored position. Rotate key to locked position and remove the key.

6) Disconnect drive cable, and install the outlet cap and rear cap.

7) Install the Ir-192 or Se-75 source identification plate on the top of the IR-100ST.

7.1.3 Preparation for Transport

1) Install a tamper-indicating seal (security wire/lead seal) in the lock assembly, and verify that the lock keys are attached to the package.

2) Load the IR-100ST onto the transport and secure utilizing tie-down nylon straps, as shown in Figure 7.1-1. (In lieu of using tie-down straps, the IR-100ST may be optionally placed within a secure transport box that is secured within an enclosed transport vehicle.)

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3) Monitor external radiation per the guidelines of 49 CFR §173.44121.

4) Determine the shielding transport index for the loaded IR-100ST per the guidelines of 49 CFR §173.403.

5) Complete all necessary shipping papers in accordance with Subpart C of 49 CFR 17222.

6) IR-100ST markings shall be in accordance with 10 CFR §71.85(c) and Subpart D of 49 CFR 172. Package labeling shall be in accordance with Subpart E of 49 CFR 172. Packaging placarding shall be in accordance with Subpart F of 49 CFR 172.

T IE-DOWN (TYP)

Figure 7.1-1-IR-100ST Tie-Down Configuration Utilizing the Sensor Surround Handle

7.2 Package Unloading This section delineates the procedures for unloading a payload into the IR-100ST. Hereafter, reference to specific IR-100ST components may be found in Appendix 1.3.1, General Arrangement Drawings.

7.2.1 Receipt of Package from Carrier

1) Remove the tie-down devices that secure the IR-100ST to the transport vehicle. If optionally transported within a shielded lock box, remove the IR-100ST from the lock box.

21 Title 49, Code of Federal Regulations, Part 173 (49 CFR 173), Shippers-General Requirements for Shipments and Packagings, 10-1-23 Edition.

22 Title 49, Code of Federal Regulations, Part 172 (49 CFR 172), Hazardous Materials Tables and Hazardous Communications Regulations, 10-1-23 Edition.

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2) Monitor the external radiation to ensure that the IR-100ST was not damaged during shipment.

7.2.2 Removal of Contents from the IR-100ST Package

1) Remove the tamper indicating wire seal from the lock assembly.

2) Remove the outlet cap and rear cap.

3) Connect the drive cable housing and the guide tube to the package. Connect the guide tube to a source changer.

4) Unlock the IR-100ST and extend the Ir-192 or Se-75 source pigtail assembly into the source changer.

5) Secure the Ir-192 or Se-75 source pigtail assembly in the source changer, lock, and remove the key.

6) Disconnect the drive cable from the source changer and retract it.

7) Disconnect the guide tube from the source changer and the IR-100ST. Install the Safety Plug and dust seal on the IR-100ST.

8) Complete all required shipping papers in accordance with Subpart C of 49 CFR 172.

9) IR-100ST marking shall be in accordance with 10 CFR §71.85(c) a nd Subpart D of 49 CFR 172. Package labeling shall be in accordance with Subpart E of 49 CFR 172. Packaging placarding shall be in accordance with Subpart F of 49 CFR 172.

7.3 Preparation of an Empty Package for Transport Previously used and empty IR-100STs shall be prepared and transported per the requirements of 49 CFR §173.428.

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8.0 ACCEPTANCE TESTS AND MAINTENANCE PROGRAM

8.1 Acceptance Tests Per the requirements of 10 CFR §71.85(c), this section discusses the inspections and tests to be performed prior to first use of the IR-100ST.

8.1.1 Visual Inspections and Measurements All IR-100ST materials of construction shall be examined in accordance with the requirements delineated on the drawings in Appendix 1.3.1, General Arrangement Drawings, per the requirements of 10 CFR §71.85(a).

8.1.2 Weld Examinations All IR-100ST welds shall be examined in accordance with the requirements delineated on the drawings in Appendix 1.3.1, General Arrangement Drawings, per the requirements of 10 CFR

§71.85(a).

8.1.3 Structural and Pressure Tests The IR-100ST does not contain any lifting/tie-down devices or pressure boundaries that require testing.

8.1.4 Leakage Tests The IR-100ST does not contain any seals or containment boundaries that require leakage testing.

Therefore, this section does not apply.

8.1.5 Component and Material Tests The IR-100ST does not contain any additional components or materials that require acceptance testing.

8.1.6 Shielding Tests A radiation profile is performed on each depleted uranium (DU) shield prior to being used in the fabrication of an IR-100ST. These measured survey results are ratioed to determine the expected radiation levels for the maximum authorize source strength of 120 Ci (4.44 TBq) for either Ir-192 or Se-75 isotopes. Any radiation profile of a DU gamma shield that results in a dose rate that exceeds the requirements of 49 CFR §173.441 with the maxim um authorized payload shall not be utilized in the manufacture of an IR-100ST.

8.1.7 Thermal Tests The IR-100ST does not contain any thermal features or systems that require testing. Therefore, this section does not apply.

8.1.8 Miscellaneous Tests There are no additional acceptance tests required for the IR-100ST.

70 INC IR-100ST Exposure Device Safety Analysis Report Docket No. 71-9385 Revision 0, 6/2024

8.2 Maintenance Program This section describes the maintenance program used to ensure continued performance of the IR-100ST.

8.2.1 Structural and Pressure Tests The IR-100ST does not contain any lifting/tie-down devices or pressure boundaries that require load testing.

8.2.2 Leakage Tests The IR-100ST does not contain any seals or containment boundaries that require testing.

8.2.3 Component and Material Tests 8.2.3.1 Fasteners All threaded components shall be inspected quarterly for deformed or stripped threads.

Damaged components shall be repaired or replaced prior to further use.

8.2.3.2 Lock Assembly and Outlet Port Prior to each use, inspect the lock assembly and outlet port fo r damage or restrained motion.

Any motion or operational impairing shall be corrected prior to further use.

8.2.4 Thermal Tests No thermal tests are necessary to ensure continued performance of the IR-100ST.

8.2.5 Miscellaneous Tests - Shielding Prior to each shipment, a radiation survey is performed to ensu re that the radiation dose levels do not exceed the requirements of 49 CFR §173.441. This survey co nfirms that the DU gamma shield has maintained its shielding function.

71

v t e Letter Sequence Request
CAC:001029, (Approved, Closed) EPID:L-2024-NEW-0008, (Open)
Initiation Request, Request, Request, Request, Request, Request, Request Acceptance Supplement, Supplement Administration Meeting, Meeting, Meeting, Meeting, Meeting Questions 6 January 2022 29 March 2023 20 September 2024 Answers Results Approval, Approval, Approval Other: ML22005A350, ML22006A352, ML22006A353, ML22006A356, ML22020A185, ML22020A186, ML22111A086, ML22278A001, ML23003A046, ML23067A436, ML23137A441, ML23220A156, ML23333A105, ML24204A233, ML24275A093, ML24275A094
MONTHYEAR2023-05-22 [Table View]

ML24295A160

Text

6.0 CRITICALITY EVALUATION

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