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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 19 records
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Technical Information Needs and Regulatory Considerations for Front-End Transportation Activities of HALEU Fuel Feed Material

This report fulfills task order number 31310022F0033 Task 1 Assessment of technical information needs and considerations for front-end transportation activities for HALEU fuel feed material. This assessment primarily focuses on the transportation outlook and packaging requirements for transporting high-assay low-enriched uranium (HALEU) fuel feed material. The regulatory framework is described in section 3.0. In section 4.0 the specific issues associated with reviewing HALEU feed material packages are discussed. The rest of the report discusses the need for new packages and Department of Energy progress on HALEU production. Through this evaluation it is shown that the 5% enrichment barrier between low-enriched uranium and HALEU is a significant challenge to transportation packaging. However, there are licensed designs and concepts that can be used to transport HALEU feedstock. No significant issues were identified through the HALEU range up to 20% enriched.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

A new HALEU isotopic certified reference material

Candidate certified reference material (CRM) U196 was prepared by gravimetrically mixing and dissolving NNSA NRMP CRM 112-A (natural uranium metal) and CRM 116-A (highly enriched 93.2% 235 U metal) to achieve a high-assay low-enriched uranium (HALEU) isotopic abundance reference material. CRM U196 will be certified for the 233 U/ 238 U, 234 U/ 238 U, 235 U/ 238 U, and 236 U/ 238 U ratios. Gravimetrically determined uranium isotope amount ratios and values derived from the isotope data were calculated from the buoyancy corrected mass data and from certified values for the metal CRMs used to create the solution. The 235 U isotope amount fraction (·100) and expanded uncertainty is 19.5846 ± 0.0028. For attribute verification, total evaporation and modified total evaporation isotope ratio measurements were performed. In conclusion, the reference material is to be distributed as a unit containing 5 mg uranium as dry uranyl nitrate.

High-assay low-enriched uranium↗
DOE OSTI2021

Analysis of the 30B UF 6 Container for Use with Increased Enrichment

Commercial light-water reactor (LWR) operators and fuel vendors are pursuing advancements in fuel and reactor design that include increasing 235 U enrichment from low enrichment (<5 wt.% 235 U) to low-enriched uranium+ (LEU+) (10 wt.% > 235 U > 5 wt.%) and high-assay low-enriched uranium (HALEU) (20 wt.% > 235 U > 10 wt.%). These advancements will necessitate the ability to transport LEU+ and HALEU fuel materials. Assessment of Existing Transportation Packages for Use with HALEU(ORNL/TM-2020/1725) assesses the potential to use currently licensed transportation packages for the transportation of HALEU by evaluating a representative package for each uranium fuel form category (for example, oxide and/or metal in fuel assemblies, fuel pins, powder, pellets, and uranium hexafluoride [UF 6 ].) This report expands on the UF 6 analysis.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Addressing HALEU Demand

Presentation to Industry leaders regarding HALEU Demand.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

HALEU Production in FCF

This is one more version of a PowerPoint presentation that has been given for many years.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Chlorination of HALEU regulus casting dross

A scoping study was performed for chlorinating dross formed during uranium casting operations. Here, the purpose is to minimize the losses of uranium to dross wastes. The dross is primarily a mixture of uranium metal and uranium oxide with a minor fraction of crucible and crucible coating materials. The reaction chemistries were performed in a carrier salt of LiCl-KCl eutectic at 500 °C. The addition of FeCl 2 chlorinated the uranium metal to UCl 3 , by reducing the FeCl 2 to iron metal. After the uranium metal was chlorinated, zirconium metal was added to the salt. The residual FeCl 2 chlorinated the zirconium metal to ZrCl 4 , by reducing the FeCl 2 to iron metal. In turn, the ZrCl 4 chlorinated the uranium oxide to UCl 3 , by converting the ZrCl 4 to zirconium oxide. The effectiveness of the chlorination reactions was qualitatively verified by cyclic voltammograms that indicated the presence or absence of FeCl 2 and UCl 3 in the salt.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Critical Experiments in Support of Current Reactors and Advanced Reactor Deployment [Slides]

Limited, specific needs exist to support current reactor fleet and fuel cycle. Primarily, MOX experiments representative of low burnup LWR fuel are needed. LEU+ power distribution measurements may also be useful. Advanced reactors present many challenges and opportunities. New fuel forms and associated processes lack validation. Initial core criticality calculations can be validated with appropriate, highly similar experiments. Transportation of LEU+/HALEU material will be essential, and will rely on new experiments to support validation

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Technology Development of a High-Capacity High-Assay Low Enriched Uranium Transportation Concept

This paper discusses the technology development (TD) efforts that led to the development of a High-Capacity High-Assay Low Enriched Uranium Transportation (HALEU) transportation concept. In 2018, the Department of Energy (DOE) Office of Nuclear Technology and Research Development tasked Idaho National Laboratory (INL) to investigate strategies to transport large quantities of HALEU. To complete this task, INL collaborated with Pacific Northwest National Laboratory and Oak Ridge National Laboratory. The project was completed in 2020, and one of the project outcomes was a transportation concept that consisting of five individual Type B packages transported on a single legal-weight truck (LWT). The total payload capacity of this concept is 1,881 kg (4,149 lb) of HALEU in the form of uranium dioxide (UO2) powder. The concept utilizes an existing Type B packaging design carrying a novel fuel basket design with an incorporated flux trap. The basket can be loaded with 18 individual fuel canisters. The research collaboration investigated the U.S. certification potential of this concept. This part of the project included evaluations of criticality safety, radiological safety, thermal safety, structural integrity, and confinement under hypothetical accident scenarios of transport. The results of these evaluations demonstrated a promising potential for U.S. certification of this concept. Eventually, the described efforts led to the pursuance and issuance of a U.S. patent, thus, protecting the associated intellectual property (IP). Current short-term goals include making this IP available to private industry partners for licensing, directly supporting DOE’s objectives of accelerating commercialization of national laboratory-generated IP. If additional funding becomes available, long-term research goals could include exploring the feasibility of transporting other uranium chemical forms (e.g., UF4) with this concept, or refining operational procedures to load or unload the packagings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Candidate Core Designs for the Transformational Challenge Reactor

Early cycle activities under the Transformational Challenge Reactor (TCR) program focused on analyzing and maturing four reactor core design concepts: two fast-spectrum systems and two thermal-spectrum systems. A rapid, iterative approach has been implemented through which designs can be modified and analyzed and subcomponents can be manufactured in parallel over time frames of weeks rather than months or years. To meet key program initiatives (e.g., timeline, material use), several constraints—including fissile material availability (less than 250 kg of HALEU), component availabilities, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small (less than one cubic meter in volume) cores with near-term viability. The fast-spectrum designs did not meet the fissile material constraint, so the thermal-spectrum systems became the primary design focus. Since significant progress has been made on advanced moderator materials (YH x ) under the TCR program, gas-cooled thermal-spectrum systems using less than 250 kg of HALEU that occupy less than 1 m 3 are now feasible. The designs for two of these systems have been evolved and matured. In both thermal-spectrum design concepts, bidirectional coolant flow is used. Coolant flows down through YH x moderator elements and is reversed in a bottom manifold and core support structure, and then flows up though or around the fuel elements. The main difference between the two thermal-spectrum design concepts is the fuel elements—one uses traditional UO 2 ceramic fuel, and the other uses UN-bearing TRISO fuel particles embedded inside a SiC matrix. Finally, core neutronics and thermal performance for these systems are assessed and summarized herein.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

High Assay Low Enriched Uranium - A material in demand and the supply efforts underway

The presentation briefly describes the HALEU demand, sources of supply, & work underway @ INL to meet some of the demand. It highlights HALEU recovery from EBR-II, process enhancements including recasting via multi tier drip cast crucible method to produce reduced dose and reduced physical size uranium ingots referred to as regulus. Includes additional polishing, to improve applicability to meeting HALEU needs and briefly touches on innovative recycling concepts under development (i.e. Zircex). HALEU enables reactors of reduced size, provides for increased fuel efficiency and longer core life, all of which contribute to less waste generation. Supplying HALEU from recycled material has the potential for further reductions in waste generation through via reductions in SNF disposition.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

A Review of Candidates for a Validation Data Set for High-Assay Low-Enrichment Uranium Fuels

Many advanced reactor concept designs rely on high-assay low-enriched uranium (HALEU) fuel, enriched up to approximately 19.75% 235 U by weight. Efforts are underway by the US government to increase HALEU production in the United States to meet anticipated needs. However, very few data exist for validation of computational models that include HALEU, beyond a few fresh fuel benchmark specifications in the International Reactor Physics Experiment Evaluation Project. Nevertheless, there are other data with potential value available for developing into quality benchmarks for use in data- and software-validation efforts. This paper reviews the available evaluated HALEU fuel benchmarks and some of the potentially relevant benchmarks for fresh highly enriched uranium. It then introduces experimental data for HALEU fuel irradiated at Idaho National Laboratory, from relatively recent irradiation programs at the Advanced Test Reactor. Such data should be evaluated and, if valuable, collected into detailed benchmark specifications to meet the needs of HALEU-based reactor designers.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

FOR-868 Microreactor Applications Research Validation and Evaluation (MARVEL) Project

The Microreactor Applications Research Validation and Evaluation (MARVEL) reactor will offer experimental capabilities that are not currently available at DOE’s national laboratories. The test bed will perform R&D on the fundamental features, operations, and behaviors of microreactor technologies and help industry partners quickly test, develop, and demonstrate their technologies. The liquid- metal cooled microreactor will produce energy using small amounts of high- assay, low-enriched uranium (HALEU) from available research materials. Its design is primarily based on existing technology and will be built using off-the- shelf components to the extent practical. The reactor will be built inside the Transient Reactor Test (TREAT) Facility Micro-Reactor Experiment Cell (T- REXC), a separate project governed by the requirements of FOR-684, “Transient Reactor Test (TREAT) Facility Micro-Reactor Experiment Cell (T REXC),” and outside the scope of this document.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗