ML21118B034

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Safety Analysis Report for the ATR-FFSC Package, Revision 16, May 2021, Part 1
ML21118B034
Person / Time
Site: 07109330
Issue date: 05/01/2021
From:
Orano Federal Services
To: Pierre Saverot
Idaho National Lab, Division of Fuel Management
PSaverot NMSS/DFM/STL 301.415.7505
Shared Package
ML21118B033 List:
References
Download: ML21118B034 (34)
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Text

Safety Analysis Report Advanced Test Reactor Fresh Fuel Shipping Container (ATR FFSC)

Revision 16, May 2021 Docket 71-9330 Prepared by:

Prepared for:

Orano Federal Services LLC Battelle Energy Alliance, LLC (BEA)

This page left intentionally blank.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 i

TABLE OF CONTENTS 1.0 General Information................................................................................................. 1-1 1.1 Introduction..................................................................................................... 1-1 1.2 Package Description........................................................................................ 1-3 1.2.1 Packaging........................................................................................ 1-3 1.2.2 Contents.......................................................................................... 1-8 1.2.3 Special Requirements for Plutonium............................................ 1-23 1.2.4 Operational Features..................................................................... 1-23 1.3 Appendix....................................................................................................... 1-24 1.3.1 Glossary of Terms......................................................................... 1-24 1.3.2 Packaging General Arrangement Drawings.................................. 1-25 2.0 Structural Evaluation................................................................................................ 2-1 2.1 Structural Design............................................................................................. 2-1 2.1.1 Discussion....................................................................................... 2-1 2.1.2 Design Criteria................................................................................ 2-2 2.1.3 Weights and Centers of Gravity...................................................... 2-3 2.1.4 Identification of Codes and Standards for Package Design............ 2-5 2.2 Materials........................................................................................................ 2-10 2.2.1 Mechanical Properties and Specifications.................................... 2-10 2.2.2 Chemical, Galvanic, or Other Reactions....................................... 2-11 2.2.3 Effects of Radiation on Materials................................................. 2-11 2.3 Fabrication and Examination......................................................................... 2-12 2.3.1 Fabrication.................................................................................... 2-12 2.3.2 Examination.................................................................................. 2-12 2.4 General Requirements for All Packages....................................................... 2-12 2.4.1 Minimum Package Size................................................................ 2-12 2.4.2 Tamper-Indicating Feature............................................................ 2-12 2.4.3 Positive Closure............................................................................ 2-13 2.4.4 Valves........................................................................................... 2-13 2.4.5 External Temperatures.................................................................. 2-13 2.5 Lifting and Tiedown Standards for All Packages......................................... 2-13 2.5.1 Lifting Devices.............................................................................. 2-13 2.5.2 Tiedown Devices.......................................................................... 2-16 2.5.3 Closure Handle.............................................................................. 2-19 2.6 Normal Conditions of Transport................................................................... 2-23 2.6.1 Heat............................................................................................... 2-23 2.6.2 Cold............................................................................................... 2-23 2.6.3 Reduced External Pressure........................................................... 2-24 2.6.4 Increased External Pressure.......................................................... 2-24 2.6.5 Vibration....................................................................................... 2-24 2.6.6 Water Spray.................................................................................. 2-25 2.6.7 Free Drop...................................................................................... 2-25 2.6.8 Corner Drop.................................................................................. 2-26 2.6.9 Compression................................................................................. 2-26

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 ii 2.6.10 Penetration.................................................................................... 2-27 2.7 Hypothetical Accident Conditions................................................................ 2-27 2.7.1 Free Drop...................................................................................... 2-29 2.7.2 Crush............................................................................................. 2-33 2.7.3 Puncture........................................................................................ 2-33 2.7.4 Thermal......................................................................................... 2-34 2.7.5 Immersion - Fissile Material........................................................ 2-36 2.7.6 Immersion - All Packages............................................................ 2-36 2.7.7 Deep Water Immersion Test......................................................... 2-36 2.7.8 Summary of Damage.................................................................... 2-35 2.8 Accident Conditions for Air Transport of Plutonium................................... 2-44 2.9 Accident Conditions for Fissile Material Packages for Air Transport.......... 2-44 2.10 Special Form.................................................................................................. 2-44 2.11 Fuel Rods....................................................................................................... 2-44 2.12 Appendices.................................................................................................... 2-45 2.12.1 Certification Tests on CTU-1.................................................. 2.12.1-1 2.12.2 Certification Tests on CTU-2.................................................. 2.12.2-1 2.12.3 Structural Evaluation for MIT and MURR Fuel..................... 2.12.3-1 3.0 Thermal Evaluation................................................................................................. 3-1 3.1 Description of Thermal Design...................................................................... 3-1 3.1.1 Design Features............................................................................... 3-2 3.1.2 Contents Decay Heat..................................................................... 3-3 3.1.3 Summary Tables of Temperatures.................................................. 3-4 3.1.4 Summary Tables of Maximum Pressures....................................... 3-4 3.2 Material Properties and Component Specifications....................................... 3-6 3.2.1 Material Properties.......................................................................... 3-6 3.2.2 Technical Specifications of Components........................................ 3-8 3.3 Thermal Evaluation for Normal Conditions of Transport........................... 3-15 3.3.1 Heat and Cold............................................................................... 3-15 3.3.2 Maximum Normal Operating Pressure......................................... 3-16 3.4 Thermal Evaluation for Hypothetical Accident Conditions........................ 3-20 3.4.1 Initial Conditions.......................................................................... 3-20 3.4.2 Fire Test Conditions...................................................................... 3-21 3.4.3 Maximum Temperatures and Pressure.......................................... 3-21 3.4.4 Maximum Thermal Stresses......................................................... 3-23 3.4.5 Accident Conditions for Air Transport of Fissile Material........... 3-24 3.5 Appendices................................................................................................... 3-30 3.5.1 Computer Analysis Results........................................................... 3-31 3.5.2 Analytical Thermal Model............................................................ 3-31 3.5.3 Thermal Decomposition/Combustion of Package Organics......... 3-47 3.6 Thermal Evaluation for MIT and MURR Fuel Elements............................ 3-59 3.6.1 Description of Thermal Design..................................................... 3-59 3.6.2 Design Features............................................................................. 3-59 3.6.3 Contents Decay Heat................................................................... 3-60

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 iii 3.6.4 Summary Tables of Temperatures................................................ 3-60 3.6.5 Summary Tables of Maximum Pressures..................................... 3-61 3.6.6 Material Properties and Component Specifications...................... 3-63 3.6.7 Thermal Evaluation for Normal Conditions of Transport............ 3-67 3.6.8 Thermal Evaluation for Hypothetical Accident Conditions......... 3-73 3.6.9 Appendices.................................................................................... 3-82 4.0 Containment............................................................................................................ 4-1 4.1 Description of the Containment System........................................................ 4-1 4.1.1 Type A Fissile Packages................................................................. 4-1 4.1.2 Type B Packages............................................................................. 4-2 4.2 Containment under Normal Conditions of Transport.................................... 4-2 4.3 Containment under Hypothetical Accident Conditions................................. 4-2 4.4 Leakage Rate Tests for Type B Packages...................................................... 4-2 5.0 Shielding Evaluation............................................................................................... 5-1 6.0 Criticality Evaluation............................................................................................... 6-1 6.1 Description of Criticality Design.................................................................... 6-1 6.1.1 Design Features Important for Criticality....................................... 6-1 6.1.2 Summary Table of Criticality Evaluation....................................... 6-2 6.1.3 Criticality Safety Index................................................................... 6-6 6.2 Fissile Material Contents................................................................................. 6-8 6.2.1 Fuel Element................................................................................... 6-8 6.2.2 Loose Fuel Plates............................................................................ 6-9 6.3 General Considerations................................................................................. 6-17 6.3.1 Model Configuration..................................................................... 6-17 6.3.2 Material Properties........................................................................ 6-20 6.3.3 Computer Codes and Cross-Section Libraries.............................. 6-21 6.3.4 Demonstration of Maximum Reactivity....................................... 6-21 6.4 Single Package Evaluation............................................................................ 6-29 6.4.1 Single Package Configuration....................................................... 6-29 6.4.2 Single Package Results................................................................. 6-33 6.5 Evaluation of Package Arrays under Normal Conditions of Transport........ 6-38 6.5.1 NCT Array Configuration............................................................. 6-38 6.5.2 NCT Array Results....................................................................... 6-42 6.6 Package Arrays under Hypothetical Accident Conditions............................ 6-57 6.6.1 HAC Array Configuration............................................................ 6-57 6.6.2 HAC Array Results....................................................................... 6-59 6.7 Fissile Material Packages for Air Transport................................................. 6-67 6.8 Benchmark Evaluations................................................................................. 6-77 6.8.1 Applicability of Benchmark Experiments.................................... 6-77 6.8.2 Bias Determination....................................................................... 6-78 6.8.3 Bias Determination for Air Transport Analysis............................ 6-82

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 iv 6.9 Appendix A: Sample Input Files................................................................... 6-91 6.10 Appendix B: Criticality Analysis for MIT and MURR Fuel....................... 6-104 6.10.1 Description of Criticality Design................................................ 6-104 6.10.2 Fissile Material Contents............................................................ 6-105 6.10.3 General Considerations............................................................... 6-114 6.10.4 Single Package Evaluation.......................................................... 6-122 6.10.5 Evaluation of Package Arrays under Normal Conditions of Transport... 6-129 6.10.6 Package Arrays under Hypothetical Accident Conditions......... 6-137 6.10.7 Fissile Material Packages for Air Transport............................... 6-145 6.10.8 Benchmark Evaluations.............................................................. 6-145 6.10.9 Sample Input Files...................................................................... 6-147 6.11 Appendix C: Criticality Analysis for Small Quantity Payloads.................. 6-163 6.11.1 Description of Criticality Design................................................ 6-163 6.11.2 Fissile Material Contents............................................................ 6-164 6.11.3 General Considerations............................................................... 6-170 6.11.4 Single Package Evaluation.......................................................... 6-176 6.11.5 Evaluation of Package Arrays under Normal Conditions of Transport..................................................................................... 6-179 6.11.6 Package Arrays under Hypothetical Accident Conditions......... 6-185 6.11.7 Fissile Material Packages for Air Transport............................... 6-189 6.11.8 Benchmark Evaluations.............................................................. 6-189 6.11.9 Sample Input Files...................................................................... 6-199 6.12 Appendix D: Criticality Analysis for the U-Mo Demonstration Element. 6-201 6.13 Appendix E: Criticality Analysis for the Cobra Fuel Element.................. 6-202 6.13.1 Description of Criticality Design................................................ 6-202 6.13.2 Fissile Material Contents............................................................ 6-203 6.13.3 General Considerations............................................................... 6-211 6.13.4 Most Reactive Fuel Element Model........................................... 6-218 6.13.5 Single Package Evaluation.......................................................... 6-224 6.13.6 Evaluation of Package Arrays under Normal Conditions of Transport................................................................................. 6-227 6.13.7 Package Arrays under Hypothetical Accident Conditions......... 6-232 6.13.8 Fissile Material Packages for Air Transport............................... 6-238 6.13.9 Benchmark Evaluations.............................................................. 6-238 6.13.10 Sample Input File........................................................................ 6-239 6.14 Appendix F: Criticality Analysis for ATR, MURR, MIT, and NBSR LEU Fuel Elements and/or DDEs................................................................................. 6-245 6.14.1 Description of Criticality Design................................................ 6-245 6.14.2 Fissile Material Contents............................................................ 6-247 6.14.3 General Considerations............................................................... 6-251 6.14.4 Package Criticality Calculations................................................. 6-267 6.14.5 Fissile Material Packages for Air Transport............................... 6-290 6.14.6 Benchmark Evaluations.............................................................. 6-292 6.14.7 References................................................................................... 6-297 6.14.8 Sample Input File........................................................................ 6-298

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 v

7.0 Package Operations.................................................................................................. 7-1 7.1 Package Loading............................................................................................. 7-1 7.1.1 Preparation for Loading.................................................................. 7-1 7.1.2 Loading of Contents - ATR Fuel or ATR U-Mo Demonstration Element Fuel Assembly.................................................................. 7-2 7.1.3 Loading of Contents - Loose ATR Fuel Plates............................... 7-3 7.1.4 Loading of Contents - MIT, MURR, or RINSC Fuel Assembly.... 7-4 7.1.5 Loading of Contents - Small Quantity Payloads (except RINSC).. 7-5 7.1.6 Preparation for Transport................................................................ 7-7 7.2 Package Unloading.......................................................................................... 7-9 7.2.1 Receipt of Package from Conveyance............................................ 7-9 7.2.2 Removal of Contents....................................................................... 7-9 7.3 Preparation of Empty Package for Transport................................................ 7-10 7.4 Other Operations........................................................................................... 7-10 8.0 Acceptance Tests and Maintenance Program........................................................... 8-1 8.1 Acceptance Tests............................................................................................. 8-1 8.1.1 Visual Inspections and Measurements............................................ 8-1 8.1.2 Weld Examinations......................................................................... 8-2 8.1.3 Structural and Pressure Tests.......................................................... 8-2 8.1.4 Leakage Tests.................................................................................. 8-2 8.1.5 Component and Material Tests....................................................... 8-2 8.1.6 Shielding Tests................................................................................ 8-2 8.1.7 Thermal Tests.................................................................................. 8-2 8.1.8 Miscellaneous Tests........................................................................ 8-3 8.2 Maintenance Program...................................................................................... 8-3 8.2.1 Structural and Pressure Tests.......................................................... 8-3 8.2.2 Leakage Rate Tests......................................................................... 8-3 8.2.3 Component and Material Tests....................................................... 8-3 8.2.4 Thermal Tests.................................................................................. 8-4 8.2.5 Miscellaneous Tests........................................................................ 8-4 9.0 Quality Assurance................................................................................................... 9-1 9.1 Organization................................................................................................... 9-1 9.1.1 ATR FFSC Project Organization.................................................... 9-1 9.2 Quality Assurance Program........................................................................... 9-3 9.2.1 General............................................................................................ 9-3 9.2.2 ATR FFSC-Specific Program......................................................... 9-4 9.2.3 QA Levels....................................................................................... 9-4 9.3 Package Design Control............................................................................... 9-11 9.4 Procurement Document Control.................................................................. 9-12 9.5 Instructions, Procedures, and Drawings...................................................... 9-13 9.5.1 Preparation and Use...................................................................... 9-14 9.5.2 Operating Procedure Changes....................................................... 9-14 9.5.3 Drawings....................................................................................... 9-14

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 vi 9.6 Document Control........................................................................................ 9-14 9.7 Control Of Purchased Material, Equipment and Services........................... 9-16 9.8 Identification And Control Of Material, Parts and Components................. 9-18 9.9 Control Of Special Processes....................................................................... 9-19 9.10 Internal Inspection....................................................................................... 9-20 9.10.1 Inspections During Fabrication..................................................... 9-21 9.10.2 Inspections During Initial Acceptance and During Service Life.. 9-22 9.11 Test Control................................................................................................. 9-22 9.11.1 Acceptance and Periodic Tests..................................................... 9-23 9.11.2 Packaging Nonconformance......................................................... 9-23 9.12 Control Of Measuring and Test Equipment................................................. 9-23 9.13 Handling, Storage, And Shipping Control................................................... 9-24 9.14 Inspection, Test, And Operating Status....................................................... 9-25 9.15 Nonconforming Materials, Parts, or Components....................................... 9-26 9.16 Corrective Action......................................................................................... 9-28 9.17 Quality Assurance Records.......................................................................... 9-28 9.17.1 General.......................................................................................... 9-29 9.17.2 Generating Records....................................................................... 9-30 9.17.3 Receipt, Retrieval, and Disposition of Records............................ 9-30 9.18 Audits

....................................................................................................... 9-32

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-1 1.0 GENERAL INFORMATION This chapter of the Safety Analysis Report (SAR) presents a general introduction and description of the Advanced Test Reactor (ATR) Fresh Fuel Shipping Container (FFSC).1 This application seeks validation of the ATR FFSC as a Type AF fissile materials shipping container in accordance with Title 10, Part 71 of the Code of Federal Regulations (10CFR71).

The major components comprising the package are discussed in Section 1.2.1, Packaging, and illustrated in Figure 1.2-1 through Figure 1.2-11 and Figure 1.2-16. Detailed drawings of the package design are presented in Appendix 1.3.2, Packaging General Arrangement Drawings. A glossary of terms is presented in Appendix 1.3.1, Glossary of Terms.

1.1 Introduction The ATR FFSC is designated a Type AF-96 packaging per the definition of 10 CFR §71.42, and has been designed to transport a single, unirradiated research reactor fuel element or the associated loose plates. The loose plates may either be flat or rolled to the geometry required for assembly into a fuel element. All fuel elements are of the plate-type. All fueled plates consist of a central fuel matrix meat, clad on both sides and all edges with aluminum alloy cladding. The fuel elements contain various types of fuel matrix containing varying amounts of U-235 ranging between low enrichment ( 20% U-235) and high enrichment ( 94% U-235). Some fuel matrices include burnable poison. Fuel elements containing up to 2 kg of U-235 may be transported by air.

Since the package transports a Type A quantity of radioactive material (see Section 4.1.1, Type A Fissile Packages) and radiation is negligible, the only safety function performed by the package is criticality control. This function is achieved, in the case of a transport accident, by confining the fuel element within the package and by maintaining separation of fuel in multiple packages.

The fuel itself is robust and inherently resists unfavorable geometry reconfiguration while contained within the package. For ease of handling and property protection purposes, each fuel element or loose plate group is contained within a lightweight aluminum housing referred to as the fuel handling enclosure (FHE).

For ATR fuel elements, the criticality control function is demonstrated via full-scale testing of a prototypic package followed by a criticality analysis using a model which bounds the test results, ensuring that the calculated keff + 2 is below the upper subcritical limit (USL) in the most limiting case. Two full-scale prototype models were used to perform a number of performance tests including normal conditions of transport (NCT) free drop and hypothetical accident 1 In the remainder of this Safety Analysis Report, Advanced Test Reactor Fresh Fuel Shipping Container will be abbreviated as ATR FFSC. In addition, the term packaging will refer to the assembly of components necessary to ensure compliance with the regulatory requirements, but does not include the payload. The term package includes both the packaging components and the fresh fuel payload.

2 Title 10, Code of Federal Regulations, Part 71 (10 CFR 71), Packaging and Transportation of Radioactive Material, 1-1-21 Edition.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-2 condition (HAC) free drop and puncture tests. Other fuel elements and loose plates are modeled in various ways as described in Chapter 6, Criticality Evaluation.

The characteristics and criticality safety index (CSI) of each payload are summarized in Table 1.1-1. Additional fuel information is given in Section 1.2.2, Contents. The ATR FFSC packaging is described in Section 1.2.1, Packaging.

Table 1.1 Fuel Types in the ATR FFSC Package Fuel Element U-235 Mass, max, grams Enrichment, max, %

Core (Meat)

Alloy FHE Type CSI ATR HEU 1,200 94 UAlx ATR FHE 4.0 MIT HEU 515 94 UAlx MIT FHE 4.0 MURR HEU 785 94 UAlx MURR FHE 4.0 ATR loose plates 600 94 UAlx LFPB 4.0 RINSC 283 20 U3Si2 RINSC FHE 25.0 Small Quantity Payload 400 94 Various SQFHE 25.0 Cobra HEU 410.3 94 UAlx Cobra FHE 4.0 Cobra LEU 435.8 20 U3Si2 Cobra FHE 4.0 ATR LEU 1,681 20 U-10Mo ATR LEU FHE 6.25 MURR LEU element and DDE 1,660 20 U-10Mo MURR LEU FHE 4.0 MIT LEU element and DDE 1,070 20 U-10Mo MIT FHE 4.0 NBSR DDE 460 20 U-10Mo Blocked with disposable material

4.0 Notes

1. The Small Quantity Payload category includes low fissile quantity fuel elements (such as RINSC fuel elements), research and development plate-type fuels (such as experimental or demonstration elements or foils) and loose fuel element plates. The RINSC fuel element is transported in its own FHE, and Small Quantity payloads are transported in the Small Quantity FHE.
2. MIT and MURR loose plates use UAlx. Cobra loose plates use either UAlx for HEU or U3Si2 for LEU. Other Small Quantity payloads may use a monolithic alloy such as U-Mo or a dispersion alloy such as UAlx or U3Si2.
3. Design Demonstration Element (DDE). Note, All DDEs are one-time shipments.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-3 1.2 Package Description This section presents a basic description of the ATR FFSC. General arrangement drawings are presented in Appendix 1.3.2, Packaging General Arrangement Drawings.

1.2.1 Packaging 1.2.1.1 Packaging Description The ATR FFSC is designed as Type AF packaging for transportation of unirradiated research reactor fuel elements and associated loose plates as described in Section 1.2.2, Contents. The packaging is rectangular in shape and is designed to be handled singly with slings, or by fork truck when racked. Package components are shown in Figure 1.2-1. Transport of the package is by highway truck or by air. The maximum gross weight of the package in any loaded configuration is 290 lbs.

The ATR FFSC is a two part packaging consisting of the body and the closure. The body is a single weldment that features square tubing as an outer shell and round tubing for the payload cavity. Three 1-inch thick ribs maintain spacing between the inner and outer shells. The components of the packaging are shown in Figures 1.2-2, 1.2-3, 1.2-4, and 1.2-5 and are described in more detail in the sections which follow. With the exception of several minor components, all steel used in the ATR FFSC is ASTM Type 304 stainless steel. Components are joined using full-thickness fillet welds (i.e., fillet welds whose leg size is nominally equal to the lesser thickness of the parts joined) and full and partial penetration groove welds.

1.2.1.1.1 ATR FFSC Body The ATR FFSC body is a stainless steel weldment 73 inches long and 8 inches square weighing (empty) approximately 230 lbs. It consists of two nested shells; the outer shell a square stainless steel tube with a 3/16 inch wall thickness and the inner shell a 6 inch diameter, 0.120 inch wall, stainless steel round tube. There are three 1 inch thick stiffening plates secured to the round tube by fillet welds at equally spaced intervals. The tube is wrapped with thermal insulation and the insulation is overlaid with 28 gauge stainless steel sheet. The stainless steel sheet maintains the insulation around the inner shell. This insulated weldment is then slid into the outer square tube shell and secured at both ends by groove welds. Thermal insulation is built into the bottom end of the package as shown in Figure 1.2-3, and the closure provides thermal insulation at the closure end of the package as shown in Figure 1.2-4.

1.2.1.1.2 ATR FFSC Closure The closure is a small component designed to be easily handled by one person. It weighs approximately 10 lbs and is equipped with a handle to facilitate use with gloved hands. The closure engages with the body using a bayonet style design. There are four lugs, uniformly spaced on the closure, that engage with four slots in the mating body feature. The closure is secured by retracting two spring loaded pins, rotating the closure through approximately 45º, and

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-4 releasing the spring loaded pins such that the pins engage with mating holes in the body. When the pins are properly engaged with the mating holes the closure is locked.

A small post on the closure is drilled to receive a tamper indicating device (TID) wire. An identical post is located on the body and is also drilled for the TID wire. For ease in operation, there are two TID posts on the body. There are only two possible angular orientations for the closure installation and the duplicate TID post on the body enables TID installation in both positions.

A cover is placed over the closure handle during transport to render the handle inoperable for inadvertent lifting or tiedown. Figure 1.2-5 illustrates the placement of the handle cover. The profile of the cover depicted in Appendix 1.3.2, Packaging General Arrangement Drawings, is optional and may be modified to fit other handle profiles to ensure lifting and tiedown features are disabled as required by 10 CFR §71.45. As an option, the closure handle may be removed for transport rather than installing the handle cover.

1.2.1.1.3 ATR HEU Fuel Handling Enclosure The ATR HEU Fuel Handling Enclosure (FHE) is a hinged thin gauge aluminum weldment used with the ATR HEU fuel element, as illustrated in Figure 1.2-1. The ATR HEU FHE is a cover used to protect the fuel from handling damage during ATR FFSC loading and unloading operations. It is a thin walled aluminum fabrication featuring a hinged lid and neoprene rub strips to minimize fretting of the fuel element side plates where they are in contact with the container. The ATR HEU FHE is not used with the ATR LEU element, which instead uses the ATR LEU FHE. See Section 1.2.1.1.10, ATR LEU FHE.

During transport the ATR HEU FHE is not relied upon to add strength to the package, or satisfy any safety requirement. For purposes of determining worst case reactivity, the ATR HEU FHE is assumed to be not present.

1.2.1.1.4 MIT Fuel Handling Enclosure The MIT FHE is comprised of two identical machined segments which surround the MIT HEU or LEU fuel element or DDE, and is secured by two end spacers and locked together using ball lock pins (see Figure 1.2-6). The primary purpose of end spacers is to secure the two sections of the FHE prior to loading the FHE into the package. The location of the hole in the end plate of the spacer also facilitates easy removal of the FHE from the package. The MIT FHE is a cover used to protect the fuel from handling damage during ATR FFSC loading and unloading operations. It is an aluminum fabrication featuring machined segments and neoprene rub strips to minimize fretting of the fuel element side plates where they are in contact with the container.

During transport the MIT FHE, including the end spacers, is not relied upon to add strength to the package; however the enclosure does maintain the fuel element within a defined dimensional envelope.

1.2.1.1.5 MURR HEU Fuel Handling Enclosure The MURR HEU FHE is very similar to the MIT FHE and is comprised of two identical machined segments which surround the MURR HEU fuel element secured by two end spacers and locked together using ball lock pins (see Figure 1.2-7). The primary purpose of end spacers

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-5 is to secure the two sections of the FHE prior to loading the FHE into the package. The location of the hole in the end plate of the spacer also facilitates easy removal of the FHE from the package. The MURR HEU FHE is a cover used to protect the fuel from handling damage during ATR FFSC loading and unloading operations. It is an aluminum fabrication featuring machined segments and neoprene rub strips to minimize fretting of the fuel element side plates where they are in contact with the container. The MURR HEU FHE is not used with the MURR LEU element or MURR DDE, which instead use the MURR LEU FHE. See Section 1.2.1.1.11, MURR LEU FHE.

During transport the MURR HEU FHE, including the end spacers, is not relied upon to add strength to the package; however the enclosure does maintain the fuel element within a defined dimensional envelope.

1.2.1.1.6 RINSC Fuel Handling Enclosure The RINSC fuel, although classified as a small quantity payload, has its own dedicated FHE.

The RINSC FHE is very similar to the MURR and MIT FHEs and is comprised of two identical machined segments which surround the RINSC fuel element and are secured by two end spacers and locked together using ball lock pins (see Figure 1.2-8). The primary purpose of end spacers is to secure the two sections of the FHE prior to loading the FHE into the package. The location of the hole in the end plate of the spacer also facilitates easy removal of the FHE from the package. The RINSC FHE is a cover used to protect the fuel from handling damage during ATR FFSC loading and unloading operations. It is an aluminum fabrication featuring machined segments and neoprene rub strips to minimize fretting of the fuel element side plates where they are in contact with the container.

During transport the RINSC FHE does not add strength to the package nor satisfy any safety requirement. For purposes of determining worst case reactivity, the RINSC FHE is assumed to be not present.

1.2.1.1.7 ATR FFSC Loose Fuel Plate Basket The Loose Plate Fuel Basket (LFPB) is comprised of four identical machined segments joined by threaded fasteners (reference Figure 1.2-16). The fasteners joining the segments in the lengthwise direction are permanently installed. The basket is opened/closed using the 8 hand tightened fasteners. For criticality control purposes during transport the loose fuel plate basket maintains the fuel plates within a defined dimensional envelope.

Additional aluminum plates may be used as dunnage to fill gaps between the fuel plates and the basket payload cavity. The dunnage is used for property protection purposes only.

1.2.1.1.8 Small Quantity Payload FHE The small quantity payload FHE (SQFHE) is very similar to the RINSC, MURR, and MIT FHEs. The SQFHE is comprised of two identical machined segments which surround the small quantity payloads and are secured by two end spacers and locked together using ball lock pins (see Figure 1.2-9). The primary purpose of end spacers is to secure the two sections of the FHE prior to loading the FHE into the package. The location of the hole in the end plate of the spacer also facilitates easy removal of the FHE from the package. The SQFHE is a cover used to

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-6 protect the fuel from handling damage during ATR FFSC loading and unloading operations. It is an aluminum fabrication featuring machined components.

During transport the SQFHE does not add strength to the package nor satisfy any safety requirement. For purposes of determining worst case reactivity, the SQFHE is assumed to be not present.

Dunnage is used to fill gaps between the small quantity payloads and SQFHE. Dunnage may be made from aluminum plates, shapes, and sheets, and may include miscellaneous steel or aluminum fasteners. Dunnage may also be made from cellulosic material such as cardboard.

The maximum gap between the fuel plate face and the basket payload cavity is 1/4 inches. The SQFHE does not come with neoprene rub strips like the RINSC FHE, however 1/8 inch thick neoprene rub strips may be used in the SQFHE to minimize fretting of the small quantity payloads where there may be contact with the SQFHE or optional aluminum dunnage. Neoprene rub strips may be used between the SQFHE and the small quantity payloads and/or between the dunnage and the small quantity payloads. The 1/8 inch neoprene rub strips shall not be stacked in more than two layers between the small quantity payload and any interior face of the SQFHE.

1.2.1.1.9 Cobra FHE The Cobra FHE is used for both the HEU and LEU versions of the fuel element, and has a design very similar to the RINSC, MURR, MIT, and Small Quantity Payload FHEs. The Cobra FHE is comprised of two identical machined segments which are secured by two end spacers and locked together using ball lock pins (see Figure 1.2-10). The primary purpose of end spacers is to secure the two sections of the FHE prior to loading the FHE into the package. The location of the hole in the end plate of the spacer also facilitates easy removal of the FHE from the package.

The Cobra FHE serves to protect the fuel from handling damage during ATR FFSC loading and unloading operations. It is an aluminum fabrication featuring machined components and neoprene rub strips to minimize fretting of the fuel element where it is in contact with the container.

During transport the Cobra FHE does not add strength to the package. For purposes of determining worst case reactivity, the Cobra FHE is assumed to be not present.

1.2.1.1.10 ATR LEU FHE The ATR LEU FHE is made in two halves of solid, low density balsa wood. It surrounds and supports the ATR LEU fuel element, and is illustrated in Figure 1.2-17. Other than operational clearances, it occupies most of the space within the package cavity not occupied by the fuel element. The two halves are held together using four straps. Straps may be plastic ties or straps, metal ties or straps, or duct tape. At each end, the FHE is closed by thin aluminum plates having holes to facilitate removal of the FHE from the package. The ATR LEU FHE is used with the ATR LEU fuel element only.

During transport the ATR LEU FHE is not relied upon to add strength to the package, or satisfy any safety requirement. For purposes of determining worst case reactivity, the ATR LEU FHE is assumed to be not present.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-7 1.2.1.1.11 MURR LEU FHE The MURR LEU FHE is very similar to the standard MURR HEU FHE described in Section 1.2.1.1.5, MURR HEU Fuel Handling Enclosure, and is illustrated in Figure 1.2-18. The only difference is the removal of some unnecessary aluminum material in order to reduce its weight.

It is used with the MURR LEU fuel element or MURR DDE. It may also be used with the standard MURR HEU fuel element.

During transport the MURR LEU FHE, including the end spacers, is not relied upon to add strength to the package; however the enclosure does maintain the fuel element within a defined dimensional envelope.

1.2.1.2 Gross Weight The maximum shipped weight of the ATR FFSC (gross weight) with the specified payload is 290 lbs for all payload configurations. Further discussion of the gross weight is presented in Section 2.1.3, Weights and Centers of Gravity.

1.2.1.3 Neutron Moderator/Absorption There are no moderator or neutron absorption materials in this package.

1.2.1.4 Heat Dissipation The uranium payload produces a negligible thermal heat load. Therefore, no special devices or features are needed or utilized in the ATR FFSC to dissipate heat. A more detailed discussion of the package thermal characteristics is provided in Chapter 3.0, Thermal.

1.2.1.5 Protrusions The closure handle protrudes 1 3/8-inches from the face of the closure. The handle is secured to the closure by means of four 10-24 UNC screws. The screws will fail prior to presenting any significant loading to either the closure engagement lugs or the locking pins.

On one face of the package body, two index lugs are secured to the package to facilitate stacking of the packages. The opposite face of the package has pockets into which the index lugs nest as illustrated in Figure 1.2-11. Each index lug is secured to the package by means of a 3/8-16 socket flat head cap screw. Under any load condition, the screw will fail prior to degrading the safety function of the package.

1.2.1.6 Lifting and Tiedown Devices The ATR FFSC may be lifted from beneath utilizing a standard forklift truck when the package is secured to a fork pocket equipped pallet, or in a package rack. Swivel lift eyes may be installed in the package to enable package handling with overhead lifting equipment. The swivel eyes are installed after removing the 3/8-16 socket flat head cap screws and index lugs.

The threaded holes into which the swivel lift eyes are installed for the lifting the package are fitted with a 3/8-16 UNC screw and an index lug (see Figure 1.2-11) during transport. When the

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-8 packages are stacked and the index lugs are nested in the mating pockets of the stacked packages, the index lugs can serve to carry shear loads between stacked packages.

1.2.1.7 Pressure Relief System There are no pressure relief systems included in the ATR FFSC design. There are no out-gassing materials in any location of the package that are not directly vented to atmosphere. The package insulation, located in the enclosed volumes of the package, is a ceramic fiber. The insulation does not off-gas under normal or hypothetical accident conditions. The closure is not equipped with either seals or gaskets so that potential out-gassing of the FHE neoprene material and fuel element plastic bag material will readily vent without significant pressure build-up in the payload cavity.

1.2.1.8 Shielding Due to the nature of the uranium payload, no biological shielding is necessary or specifically provided by the ATR FFSC.

1.2.2 Contents The ATR FFSC is loaded with contents consisting of unirradiated fuel elements, DDEs, small quantity payloads, and ATR loose fuel element plates as listed in Table 1.1-1. The total mass of polyethylene (including the mass of any plastic material such as adhesive tape) in the packaging shall not exceed:

For all HEU fuel elements, ATR Loose Plates, RINSC, Small Quantity, and Cobra LEU:

maximum 100g.

For ATR LEU, MURR LEU, MURR DDE, MIT LEU, MIT DDE, and NBSR DDE:

maximum 200g.

The total mass of neoprene plus any cellulosic material such as paper or cardboard in the packaging shall not exceed 4 kg. The neoprene thickness and arrangement shall be as directed by the drawings in Appendix 1.3.2, Packaging General Arrangement Drawings, or as dictated throughout this Chapter.

The composition of the uranium in any payload is: U-235 (enrichment is specified below for each payload type), up to 1.2 wt.% U-234, up to 0.7 wt.% U-236, and the balance U-238.

1.2.2.1 ATR HEU Fuel Element Each ATR HEU fuel element contains up to 1,200 g U-235, enriched up to 94% U-235. The fuel element (ATR Mark VII) fissile material is uranium aluminide (UAlx). The fuel element weighs not more than 25 lbs, is bagged, and is enclosed in the ATR FHE weighing 15 lbs.

There are four different ATR Mark VII fuel element types designated 7F, 7NB, 7NBH, and YA.

The construction of these fuel elements are identical, varying only in the content of the fuel matrix. In the 7F fuel element, all 19 fuel plates are loaded with enriched uranium in an aluminum matrix with the eight outer plates (1 through 4 and 16 through 19) containing boron as a burnable poison. The fuel element with the greatest reactivity is the 7NB which contains no

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-9 burnable poison. The 7NBH fuel element is similar to the 7NB fuel element except that it contains one or two borated plates. The YA fuel element is identical to the 7F fuel element except that plate 19 of the YA fuel element is an aluminum alloy plate containing neither uranium fuel nor boron burnable poison. The total U-235 and B-10 content of the YA fuel element is reduced accordingly. A second YA fuel element design (YA-M) has the side plate width reduced by 15 mils.

The ATR fuel elements contain 19 curved fuel plates. A section view of an ATR fuel element is given in Figure 1.2-12. The fuel plates are rolled to shape and swaged into the two fuel element side plates. Fuel plate 1 has the smallest radius, while fuel plate 19 has the largest radius. The fissile material (uranium aluminide) is nominally 0.02-in thick for all 19 plates. Fuel element side plates are fabricated of ASTM B 209, aluminum alloy 6061-T6 or 6061-T651 and are approximately 0.19-in thick. The maximum channel thickness between fuel plates is 0.087 inches.

1.2.2.2 MIT HEU Fuel Element Each MIT HEU element contains up to 515 g U-235, enriched up to 94 wt.%. Like the ATR fuel element, the MIT fuel element fissile material is uranium aluminide (UAlx). The fuel element weighs not more than 10 lbs, is bagged, and is enclosed in the MIT FHE weighing 25 lbs.

Each MIT HEU fuel element contains 15 flat fuel plates, as shown in Figure 1.2-13. The fuel plates are fabricated and swaged into the two fuel element side plates. The fuel meat is a mixture of uranium metal and aluminum, while the cladding and structural materials are an aluminum alloy. The fissile material (uranium aluminide) is nominally 0.03-in thick and the cladding is nominally 0.025-in thick. Fuel element side plates are fabricated of ASTM B 209, aluminum alloy 6061-T6 and are approximately 0.19-in thick. The maximum channel thickness between fuel plates is 0.090 inches, excluding the thermal grooves. If the 0.012 inch thermal groove is considered, the maximum channel thickness between fuel plates is 0.114 inches.

1.2.2.3 MURR HEU Fuel Element Each MURR HEU element contains up to 785 g U-235, enriched up to 94 wt.%. Like the ATR fuel element, the MURR fuel element fissile material is uranium aluminide (UAlx). The fuel element weighs not more than 15 lbs, is bagged, and is enclosed in the MURR FHE weighing 30 lbs.

Each MURR HEU fuel element contains 24 curved fuel plates. Fuel plate 1 has the smallest radius, while fuel plate 24 has the largest radius, as shown in Figure 1.2-14. The fuel meat is a mixture of uranium metal and aluminum, while the cladding and structural materials are an aluminum alloy. The fuel plates are rolled to shape and swaged into the two fuel element side plates. The fissile material (uranium aluminide) is nominally 0.02-in thick for all 24 plates. Fuel element side plates are fabricated of ASTM B 209, aluminum alloy 6061-T6 or 6061-T651 and are approximately 0.15-in thick. The maximum channel thickness between fuel plates is 0.090 inches.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-10 1.2.2.4 Small Quantity Payload The small quantity payload consists of a class of research and development plate-type fuels with U-235 as the fissile isotope (i.e., no U-233 or plutonium), with a bounding U-235 loading 400 g, and U-235 enrichment 94%. Fuel types that fall into the small quantity payload category include low fissile quantity fuel elements (such as RINSC fuel elements), research and development plate-type fuels (such as experimental or demonstration elements or foils) and loose fuel element plates.

The acceptable limits for any small quantity payload are the bounding quantity of 400 g fissile mass and 94% enrichment. In addition, the payload shall not include beryllium, carbon, deuterium, or materials with a hydrogen density greater than that of water, except that the payload may contain up to 100 g of polyethylene and up to 4,000 g of neoprene plus cellulosic material such as paper or cardboard. The payload shall be in the form of unirradiated foils, fuel plates or fuel elements and miscellaneous non-fueled associated components. The maximum weight of any small quantity payload, including the SQFHE, is 50 lbs. As stated above, the RINSC fuel element, shown in Figure 1.2-15, is shipped in the dedicated RINSC FHE.

With the exception of RINSC fuel, which utilizes the RINSC FHE, all small quantity payload items fall within the maximum dimensional bounds of the SQFHE, or approximately 55-in x 3.4-in x 3.4-in. The minimum dimensions for a small quantity payload item are approximately 10-in x 1.5-in x 0.008-in.

1.2.2.5 ATR Loose Fuel Plates The maximum weight of the ATR loose plate payload (Figure 1.2-16) is 50 lbs. This weight is made up of the maximum basket contents weight of 20 lbs and the loose fuel plate basket weight of 30 lbs.

The loose plate payload is limited to 600 grams U-235. The plates are limited to those used in ATR fuel elements. The plates may either be flat or rolled to the geometry required for assembly into the fuel element. For handling convenience, the loose plate basket will be loaded with either flat or rolled plates. Additionally, the plates may be banded or wire tied in a bundle.

1.2.2.6 Cobra Fuel Element The Cobra fuel element (shown in Figure 1.2-19) is used at the BR2 reactor in Belgium. This category includes HEU (enriched to 94% U-235 as UAlx dispersed in aluminum powder), and LEU (enriched to 20% U-235 as U3Si2 dispersed in aluminum powder). The bounding loading is 410.3g U-235 in HEU and 435.8g U-235 in LEU. The bounding fuel element weight is 20 lb, is bagged, and is enclosed in a Cobra FHE weighing 28 lb.

1.2.2.7 ATR LEU Fuel Element The ATR LEU fuel element contains up to 1,681 g U-235, enriched up to 20% U-235. The fissile material is U-10Mo. The fuel element weighs no more than 44 lb, is bagged, and is enclosed in the ATR LEU FHE weighing not more than 6 lb.

The ATR LEU fuel element contains 19 curved fuel plates, and has the same structural design as the ATR HEU fuel element, shown in Figure 1.2-12. The fuel meat is U-10Mo alloy, and the

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-11 cladding and structural materials are aluminum alloy. The nominal fissile material thickness varies between 0.008 inches and 0.016 inches, and the plates are nominally 0.050 inches thick, except the innermost plate is 0.080 inches thick and the outermost plate is 0.100 inches thick.

There is a nominally 0.001-inch thick layer of zirconium between the fuel meat and the cladding. The fuel plates are rolled to shape and swaged into the two fuel element side plates.

1.2.2.8 MIT LEU Fuel Element and DDE The MIT LEU fuel element contains up to 1,070 g U-235, enriched up to 20% U-235. The fissile material is U-10Mo. The fuel element weighs no more than 19 lb, is bagged, and is enclosed in the standard MIT FHE, the same FHE as used for HEU fuel elements.

The MIT LEU fuel element contains 19 flat fuel plates, and except for the quantity of plates and their separation distance, is constructed the same way as the MIT HEU fuel element depicted in Figure 1.2-13. The fuel meat is U-10Mo alloy, and the cladding and structural materials are aluminum alloy. The nominal fissile material thickness varies between 0.013 inches and 0.025 inches, and the plates are nominally 0.049 inches thick overall. There is a nominally 0.001-inch thick layer of zirconium between the fuel meat and the cladding. The fuel plates are swaged into the two fuel element side plates.

The MIT LEU DDE has the same number of plates, plate design, swaging design, and maximum mass of fissile material. It is approximately 2 inches shorter than the MIT LEU fuel element, and the end structures are slightly different to allow in-situ measurements during test irradiation.

1.2.2.9 MURR LEU Fuel Element and DDE The MURR LEU fuel element contains up to 1,660 g U-235, enriched up to 20% U-235. The fissile material is U-10Mo. The fuel element weighs no more than 29 lb, is bagged, and is enclosed in the MURR LEU FHE, weighing not more than 21 lb.

The MURR LEU fuel element contains 23 curved fuel plates, and except for the quantity of plates and their separation distance, is constructed the same way as the MURR HEU fuel element depicted in Figure 1.2-14. The fuel meat is U-10Mo alloy, and the cladding and structural materials are aluminum alloy. The nominal fissile material thickness varies between 0.009 inches and 0.020 inches, and the plates are nominally 0.044 inches thick, except the outermost plate is 0.049 inches thick. There is a nominally 0.001-inch thick layer of zirconium between the fuel meat and the cladding. The fuel plates are rolled to shape and swaged into the two fuel element side plates.

The MURR LEU DDE has the same number of plates, plate design, swaging design, and maximum mass of fissile material. It is approximately 0.25 inches longer than the MURR LEU fuel element, and the end structures are slightly different to allow in-situ measurements during test irradiation.

1.2.2.10 NBSR DDE The NBSR DDE contains up to 460 g U-235, enriched up to 20% U-235. The fissile material is U-10Mo. The DDE weighs no more than 23 lb, is bagged, and is blocked and protected within the ATR FFSC package using up to 4 kg of cellulosic material such as cardboard. The plastic bag and any tape used in packing will not exceed 200 g.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-12 The NBSR DDE contains 17 curved fuel plates and inner and outer curved non-fueled plates in each of two sections for a total of 34 separate fuel plates. Like the full size NBSR fuel element, the fuel plates are arranged in two, nominally 13-inch long sections with an empty space nominally 2 inches long between them. The fuel meat is U-10Mo alloy, and the cladding and structural materials are aluminum alloy. The nominal fissile material thickness is 0.0085 inches, and the plates are nominally 0.050 inches thick. There is a nominally 0.001-inch thick layer of zirconium between the fuel meat and the cladding. The fuel plates are rolled to shape and swaged into the two fuel element side plates. The nominal length of the NBSR DDE is 29.1 inches.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-13 Figure 1.2 Overview of the ATR FFSC (Outer Body Shell Shown Transparent)

Figure 1.2 Top End Body Sectional View (HEU)

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-14 Figure 1.2 Bottom End Body Sectional View Figure 1.2 Closure Sectional View

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-15 Figure 1.2 Closure Handle Cover Figure 1.2 MIT HEU or LEU Fuel Handling Enclosure End Spacer (2)

MIT Fuel Element Enclosure (2)

Locking Pin (2)

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-16 Figure 1.2 MURR HEU Fuel Handling Enclosure Figure 1.2 RINSC Fuel Handling Enclosure End Spacer (2)

MURR Fuel Element Enclosure (2)

Locking Pin (2)

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-17 Figure 1.2 Small Quantity Fuel Handling Enclosure Figure 1.2 Cobra Element Fuel Handling Enclosure

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-18 Figure 1.2 Index Lug and Mating Pocket of Stacked Packages Figure 1.2 ATR Fuel Element - Section View

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-19 Figure 1.2 MIT Fuel Element - Section View Figure 1.2 MURR Fuel Element - Section View End Box Fuel Plates (15)

Side Plates End Box Fuel Plates (24)

Side Plates

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-20 Figure 1.2 RINSC Fuel Element - Section View Figure 1.2 Loose Fuel Plate Basket - Exploded View

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-21 Figure 1.2 ATR LEU Fuel Handling Enclosure Figure 1.2 MURR LEU Fuel Handling Enclosure

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-22 Figure 1.2 Cobra Fuel Element

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-23 1.2.3 Special Requirements for Plutonium Because the ATR FFSC does not contain any plutonium, this section does not apply.

1.2.4 Operational Features There are no operationally complex features in the ATR FFSC. All operational features are readily apparent from an inspection of the drawings provided in Appendix 1.3.2, Packaging General Arrangement Drawings. Operation procedures and instructions for loading, unloading, and preparing an empty ATR FFSC for transport are provided in Chapter 7.0, Operating Procedures.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-24 1.3 Appendix 1.3.1 Glossary of Terms ANSI -

American National Standards Institute.

ASME B&PV Code -

American Society of Mechanical Engineers Boiler and Pressure Vessel Code.

ASTM -

American Society for Testing and Materials.

ATR FFSC -

Advanced Test Reactor Fresh Fuel Shipping Container AWS -

American Welding Society.

DDE -

Design Demonstration Element.

HAC -

Hypothetical Accident Conditions.

NCT -

Normal Conditions of Transport.

Closure -

The ATR FFSC package component used to close the package.

Body -

The ATR FFSC package component which houses the payload.

Fuel element Fuel element and fuel assembly are used interchangeably throughout this document as described in Section 1.2.2, Contents.

Index lug -

A thick washer like component secured to the package body at the lift point locations. The index lug provides shear transfer capability between stacked packages.

Pocket -

A recessed feature on the package body that accepts the index lug when packages are stacked.

Fuel Handling Enclosure (FHE) - Aluminum fabrications used to protect the fuel elements from handling damage. The enclosures are faced with neoprene at locations where the fuel element contacts the FHE to minimize fretting of the fuel element at the contact points. The ATR LEU FHE is used only with the ATR LEU fuel element, and is made primarily of solid balsa wood with aluminum ends, and no neoprene.

Loose fuel plate basket (LFPB) - A machined aluminum container in which the unassembled fuel element plates are secured during transport in the ATR FFSC. The loose plate basket is a geometry based criticality control component.

Small Quantity Payload FHE (SQFHE) - see Fuel Handling Enclosure (FHE).

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-25 1.3.2 Packaging General Arrangement Drawings The packaging general arrangement drawings consist of:

60501-10, ATR Fresh Fuel Shipping Container SAR Drawing, 5 sheets 60501-20, Loose Plate Basket Assembly ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet 60501-30, Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet 60501-40, MIT Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet 60501-50, MURR Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet.

60501-60, RINSC Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet.

60501-70, Small Quantity Payload Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet.

60501-90, Cobra Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet.

60501-110, ATR LEU Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet.

60501-111, MURR LEU Fuel Handling Enclosure, ATR Fresh Fuel Shipping Container SAR Drawing, 1 sheet.

Docket No. 71-9330 ATR FFSC Safety Analysis Report Rev. 16, May 2021 1-26 This page left intentionally blank.

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