Results for “Highly Enriched Uranium” — Page 21

NASA NTRSDate not supplied

Affordable Development Strategy for NEP Nuclear Systems

One nuclear electric propulsion (NEP) reactor systems under consideration is a hydride moderated thermal spectrum reactor fueled by high assay low enriched uranium (HALEU). While such a reactor is expected to yield the lightest HALEU reactor design, its development challenges grow exponentially with increasing mission demands, most notably power output, specific weight of the overall system (which may require operation at temperatures exceeding 1200 K), service lifetime, and human-rated reliability. Two of the greatest cost drivers are full-powered nuclear demonstrations and extensive material development campaigns, so it is important to consider options that can minimize the need for or complexity of such tasks. This paper discusses a structured framework being developed for assessing how NEP design choices, such as materials selection, neutronic features, and heat-removal technologies, can translate into project risk and how project performance goals can be traded with development cost.Reactors operating at high temperatures often require cutting-edge heat transfer technologies and creep-resistant materials. Use of new materials in high temperature reactors brings additional complication beyond those common to any new space materials development campaign. For example, such materials may not possess necessary neutronic cross-sectional or neutronic irradiation data. Similarly, use of new materials may significantly influence core neutronics; in some extreme cases, neutronic reactivity feed-back of certain new materials can vary during their service life as radiation damage impacts the scattering cross-section. In an affordable development approach, high fidelity modeling and simulation tools are used to identify and characterize potential ‘knees-in-the-curves’ in the relationship that exists between the mission characteristics and the project risk. Of particular significance is use of modern uncertainty management and variance reduction methods to perform gap analyses that feed into phenomena identification and ranking tables (PIRT) commonly used to communicate nuclear readiness levels. Model-based measurements techniques are used to design sub-scale experiments as a substitute to minimize orcompletely eliminate the need for nuclear demonstrations.This paper will describe the approach and present preliminary results. It will lay the groundwork for developing a set of metrics that can be broadly characterized as system nuclear readiness levels and advancement degree of difficulty for nuclear systems. Equally importantly, a goal of this paper is to initiate a dialogue among stakeholders on what is the sufficient level of maturity that is required for launching a demonstration unit.

Dasari V Rao↗

DOE OSTI2026

Nuclear Security Risks for HALEU Fuels

There is growing interest in high-assay low-enriched uranium (HALEU) for use in advanced nuclear reactors as a high-energy fuel source. The primary objectives of this report are to identify the security risks that directly result from HALEU and to identify the gaps and challenges it presents from a theft and sabotage perspective. This study focuses on HALEU security risks for the front end of the fuel cycle and includes a review of the supply chain, fuel fabrication, and transport for terrestrial reactors.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2024

Superoxide Radicals in Uranyl Peroxide Solids: Lasting Signatures Identified by Electron Paramagnetic Resonance Spectroscopy

Abstract U(VI) peroxide phases (studtite and meta‐studtite) are found throughout the nuclear fuel cycle and exist as corrosion products in high radiation fields. Peroxides are part of a family of reactive oxygen species (ROS) that include hydroperoxyl and superoxide species and are produced during alpha radiolysis of water. While U(VI) peroxides have been thoroughly investigated, the incorporation and stability of ROS species within studtite have not been validated. In the current study, electron paramagnetic resonance (EPR) spectroscopy was used to identify the presence of free radicals within a series of U(VI) peroxide samples containing depleted, highly enriched, and natural uranium. Density functional theory calculations indicated that the predicted EPR signals matched well with a superoxide (O 2 − ⋅) species incorporated into the studtite structure, confirming the presence of ROS in the material. Further analysis of samples that were synthesized between 1945 and 2023 indicated that there is a correlation between the radical signal and the product of specific activity multiplied by age of the sample.

Scherrer, Sarah K.↗

DOE OSTI2024

Superoxide Radicals in Uranyl Peroxide Solids: Lasting Signatures Identified by Electron Paramagnetic Resonance Spectroscopy

Abstract U(VI) peroxide phases (studtite and meta‐studtite) are found throughout the nuclear fuel cycle and exist as corrosion products in high radiation fields. Peroxides are part of a family of reactive oxygen species (ROS) that include hydroperoxyl and superoxide species and are produced during alpha radiolysis of water. While U(VI) peroxides have been thoroughly investigated, the incorporation and stability of ROS species within studtite have not been validated. In the current study, electron paramagnetic resonance (EPR) spectroscopy was used to identify the presence of free radicals within a series of U(VI) peroxide samples containing depleted, highly enriched, and natural uranium. Density functional theory calculations indicated that the predicted EPR signals matched well with a superoxide (O 2 − ⋅) species incorporated into the studtite structure, confirming the presence of ROS in the material. Further analysis of samples that were synthesized between 1945 and 2023 indicated that there is a correlation between the radical signal and the product of specific activity multiplied by age of the sample.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗

DOE OSTI2020

Strategies for thorium fuel cycle transition in the SD-TMSR

Liquid-fueled Molten Salt Reactor (MSR) systems represent advances in safety, economics, and sustainability. The MSR has been designed to operate with a Th/ 233 U fuel cycle with 233 U used as startup fissile material. Since 233 U does not exist in nature, we must examine other available fissile materials to start up these reactor concepts. This work investigates the fuel cycle and neutronics performance of the Single-fluid Double-zone Thorium-based Molten Salt Reactor (SD-TMSR) with different fissile material loadings at startup: High Assay Low Enriched Uranium (HALEU) (19.79%), Pu mixed with HALEU (19.79%), reactor-grade Pu (a mixture of Pu isotopes chemically extracted from Pressurized Water Reactor (PWR) spent nuclear fuel (SNF) with 33 GW d/tHM burnup), transuranic elements (TRU) from Light Water Reactor (LWR) SNF, and 233 U. The MSR burnup routine provided by SERPENT-2 is used to simulate the online reprocessing and refueling in the SD-TMSR. The effective multiplication factor, fuel salt composition evolution, and net production of 233 U are studied in the present work. Additionally, the neutron spectrum shift during the reactor operation is calculated. The results show that the continuous flow of reactorgrade Pu helps transition to the thorium fuel cycle within a relatively short time (≈ 4.5 years) compared to 26 years for 233 U startup fuel. Finally, using TRU as the initial fuel materials offers the possibility of operating the SD-TMSR for an extended period of time (≈ 40 years) without any external feed of 233 U.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

Transformational Challenge Reactor preconceptual core design studies

In the nuclear industry, a manufacturing-informed design approach has the potential to yield the most benefit from advanced manufacturing. By leveraging advanced materials, data science, and rapid testing and deployment, manufacturing-informed design can drive down costs and development times, ultimately improving future commercial viability. This approach is being demonstrated in the US Department of Energy Office of Nuclear Energy (DOE-NE) Transformational Challenge Reactor (TCR) program. Preconceptual design activities for TCR have been focused on analyzing and maturing four reactor core design concepts: two fast-spectrum and two thermal-spectrum systems. Furthermore, the designs were iteratively modified and analyzed, and subcomponents were manufactured in parallel over weeks instead of months or years. To meet key program initiatives (e.g., timeline and material use), several constraints—including fissile material availability, component availability, materials compatibility, and additive manufacturing capabilities—were factored into the design effort, yielding small cores less than one cubic meter in volume with near-term viability. Additionally, the TCR program has made significant progress on development of advanced moderator materials such as yttrium hydride, advancing the feasibility of gas-cooled thermal spectrum systems using less than 250 kg of high-assay low enriched uranium (HALEU) and occupying less than 1 m3. Each of the two resulting thermal designs uses a different fuel form: traditional UO2 ceramic fuel and tristructural isotropic (advanced TRISO) fuel particles embedded inside a SiC matrix. Core neutronics and thermal performance for these systems were assessed and summarized. Evaluation of the performance metrics for these two moderated designs has yielded the downselected TCR design: a TRISO-fueled and yttrium hydride moderated gas-cooled reactor.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2025

Transient fuel performance analysis for the preliminary fuel concept of general atomics fast modular reactor

This study investigates the transient fuel performance of General Atomics Fast Modular Reactor (GA-FMR) during accident scenarios, focusing on the behavior of its innovative fuel system that combines high-assay low enriched uranium dioxide (HALEUO2) fuel with SiGA® ceramic matrix composite silicon carbide cladding. The preliminary fuel design’s response was analyzed during reactivity-initiated accidents (RIA) and loss of coolant accidents (LOCA) using BISON fuel performance analysis code, which included both the diffusion enhanced and BISON-FASTGRASS coupled UO 2 models. The RIA analysis demonstrated that effective reactivity control reduced fuel temperature, though with transient fission gas release resulting in additional tensile stress state on the cladding. LOCA simulations revealed differing predictions between the two models: the BISON UO 2 model showed more transient fission gas release but minimal pellet expansion, while the BISON-FASTGRASS UO 2 model predicted less pronounced fission gas release but more fuel swelling and thermal expansion, potentially leading to pellet-cladding mechanical interaction. Here, these findings highlight critical areas for fuel design optimization and identify knowledge gaps requiring further experimental and computational investigation to advance GA-FMR fuel development.

Lee, Soon K. [Argonne National Laboratory (ANL), A↗

DOE OSTI2021

Isotopic Characterization of HALEU from EBR-II Driver Fuel Processing

Provide the isotopic characteristics of the High Assay Low Enriched Uranium (HALEU) product recovered during the used nuclear fuel treatment operations occurring at Idaho National Laboratory. Revision one of this document (TEV-3537 Rev 1) provides isotopic characterization for material that has been recast using the drip cast crucible gravity casting technique used to transform traditional EBR-II uranium ingots into regulus.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

An Assessment of Applying Pyroprocessing Technology to Advanced Pebble-Type Fuels

With an expected increased demand for high-assay low enriched uranium (HALEU) to supply advanced reactors, the possibility of recovering actinides from various used nuclear fuels (UNF) is being reassessed. Pebble-type fuels such as tristructural isotropic (TRISO) fuel are intended to be directly disposed after use, but the HALEU remaining in each pebble is a resource with increasing value. An economically feasible recovery strategy would maximize the amount of HALEU recovered for reuse in new fuel without increasing the waste volume relative to the direct disposal of used TRISO fuel. We are evaluating the technical feasibility of recycling TRISO or pebble fuels to recover actinides for reuse as new fuel for advanced reactors by using pyrochemical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2021

Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages

Microreactors are very small nuclear reactors with a power output of about 20 megawatts electric or less that are designed to be factory-built, modular in nature, and highly portable. These compact reactors will be small enough to be transported by truck or even air and could help solve energy challenges in a number of areas, ranging from remote commercial or residential locations to military bases. Pacific Northwest National Laboratory is tasked to develop and evaluate transportation licensing options for microreactors. The work is funded by the National Reactor Innovation Center a National Department of Energy program led by Idaho National Laboratory for the Office of Nuclear Energy Research and Development which support demonstration of microreactor technology. Key transportation steps include the (1) initial movement of high-assay low enriched uranium fresh fuel, (2) transportation of an intact, but never-operated microreactor, and (3) transportation of an intact, previously-operated microreactor. The deliverables on the project consists of a documentation of applicable regulations and regulatory authority for transportation. The objective of this report is to propose a risk-informed regulatory framework for the licensing of the transportation of microreactors, including the transportation of irradiated nuclear fuel that is assumed to be an integral component of the microreactor transportation package. The framework lays out a viable regulatory pathway, including decision points for regulatory options and the supporting technical evaluations for those options in phases from near to long term. This report includes discussion of the (1) general microreactor design concepts including representative microreactor source terms, (2) options for regulatory approval of microreactor transportation based on current regulation and historical precedence, (3) regulatory basis for including risk information in microreactor transportation licensing activities, and (4) description of a risk-informed regulatory framework.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2021

Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages

Microreactors are very small nuclear reactors with a power output of about 20 megawatts electric or less that are designed to be factory-built, modular in nature, and highly portable. These compact reactors will be small enough to be transported by truck or even air and could help solve energy challenges in a number of areas, ranging from remote commercial or residential locations to military bases. Pacific Northwest National Laboratory is tasked to develop and evaluate transportation licensing options for microreactors. The work is funded by the National Reactor Innovation Center a National Department of Energy program led by Idaho National Laboratory for the Office of Nuclear Energy Research and Development which support demonstration of microreactor technology. Key transportation steps include the (1) initial movement of high-assay low enriched uranium fresh fuel, (2) transportation of an intact, but never-operated microreactor, and (3) transportation of an intact, previously-operated microreactor. The deliverables on the project consists of a documentation of applicable regulations and regulatory authority for transportation. The objective of this report is to propose a risk-informed regulatory framework for the licensing of the transportation of microreactors, including the transportation of irradiated nuclear fuel that is assumed to be an integral component of the microreactor transportation package. The framework lays out a viable regulatory pathway, including decision points for regulatory options and the supporting technical evaluations for those options in phases from near to long term. This report includes discussion of the (1) general microreactor design concepts including representative microreactor source terms, (2) options for regulatory approval of microreactor transportation based on current regulation and historical precedence, (3) regulatory basis for including risk information in microreactor transportation licensing activities, and (4) description of a risk-informed regulatory framework.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2023

An Assessment of Applying Pyroprocessing Technology to Advanced Pebble-Type Fuels

With an expected increased demand for high-assay low enriched uranium (HALEU) to supply advanced reactors, the possibility of recovering actinides from various used nuclear fuels (UNF) is being re-assessed. Pebble-type fuels such as tristructural isotropic (TRISO) fuel are intended to be directly disposed after use, but the HALEU remaining in each pebble is a resource with increasing value. An economically feasible recovery strategy would maximize the amount of HALEU recovered for reuse in new fuel without increasing the waste volume relative to the direct disposal of used TRISO fuel. We are evaluating the technical feasibility of recycling TRISO or pebble fuels to recover actinides for reuse as new fuel for advanced reactors by using pyrochemical methods.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2023

Baseline Hypothetical Facility for the Production of 131 I, 99 Mo, and 133 Xe from HALEU Fission Targets

The report describes the development of a hypothetical facility to produce the pharmaceutical radioisotope 131 I at an amount of 60 curies per week via the fission of a High Assay Low Enriched Uranium (HALEU) target, along with the chemical and physical processes and equipment needed to separate the 131 I and co-produced 99 Mo and 133 Xe. The hypothetical design was carried out using a 10 MWt research reactor. The irradiation calculation determined that three HALEU targets with aluminum cladding can be used to produce 60 Ci/week of 131 I, 750 Ci/week of 99 Mo and 560 Ci/week of 133 Xe. The process selected for the baseline design uses caustic dissolution of the target material with ion exchange processes to separate and purify the iodine and molybdenum. The xenon is processed using a cryogenic carbon bed separation process. Target processing occurs in seven shielded hot cells with a total footprint of 16m 2 and waste management occurs in an eight hot cell with a 2.5 m 2 footprint. Waste generated from the processing of the three targets per week would generate less than 4 shielded drums of waste annually. These hot cells would need varying levels of shielding due to the amount of fission products being handled in the unit. Hot cell facilities also require support services including QA/QC, health physics, administrative staff, operator changing room, radiological buffer areas, and waste storage. The overall facility would require a footprint of 1050 m 2 with 20.25 m 2 of shielded hot cells.

62 RADIOLOGY AND NUCLEAR MEDICINE↗

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Application of Analytical Hierarchy Process for Narrowing Down Nep Candidate Reactor Designs

Nuclear electric propulsion (NEP)-powered vehicles have been contemplated for human Mars missions. The nuclear power system contemplates using a high temperature light-weight nuclear reactor for the production of electrical power in the range of 2-5 MWe with a 3-10 year service life. A myriad of technology options exist for achieving these mission objectives. The analytic hierarchy process (AHP), a multi-attribute decision method, is being used to narrow down the candidate designs. The AHP is a structured decision process that fuses model-supplied quantitative data with subjective assessments to facilitate decisions that involve multiple competing criteria. Proven end-to-end nuclear design and systems analysis tools will be used to provide quantitative performance data such as end-to-end system reliability, system robustness to recover from off-normal conditions, system specific weight (α in kg/kWe), and the ability to meet the service life-time and power level requirements. Three high assay low enriched uranium (HALEU) reactor concepts, namely, gas-, heat pipe- and pumped liquid-cooled nuclear cores coupled to a He/Xe gas or supercritical CO2- Brayton power conversion system are being modeled using the AHP to understand the trade-offs associated with these design combinations. In addition to the reactor and power conversion options, there are a myriad of additional components and subsystems – e.g., radiators, heat exchangers, and recuperators – that also figure into the evaluation process. The choice of the right overall system is a multidimensional problem that has to include not only quantitative data, but also the so called “external factors”, examples of which include the component and subsystem Technology Readiness Levels (TRLs), the associated Advancement Degrees of Difficulty (AD2), the cost and schedule required to achieve a technical maturity consistent with mission infusion, alignment with the priorities of NASA’s Space Nuclear Propulsion program, and alignment with other ongoing government and commercial investments in micro reactors. Focused expert elicitations form the basis for qualitative data set. As a final step, decision-makers individually express their opinions regarding the relative importance of the criteria and preferences among the alternatives through pairwise comparisons. The paper will describe the AHP approach, progress to-date applying it to the NEP human Mars mission problem, and preliminary results. Use of AHP provides sufficient flexibility for incorporating industry input at different stages as the technologies evolve through additional research and development. It is expected that the decision process will be ongoing and expanded to examine additional options and technology choices, culminating in a defensible set of candidate reactor concepts that will form the basis for developing a multi-year NEP technology maturation strategy. .

Dasari V Rao↗

NASA NTRSDate not supplied

Fuel and Moderator Development Strategy for NEP Systems

Nuclear electric propulsion (NEP)-powered vehicles have been contemplated for human Mars missions. The nuclear power system contemplates using a high temperature light-weight nuclear reactor for the production of electrical power in the range of 2-5 MWe with a 3-10 year service life. There are several reactor and moderator options that exist for an NEP reactor with these operational needs. One option under consideration is a hydride moderated reactor using a fuel comprised of high assay low enriched uranium (HALEU) compounds. The NEP fuel-moderator mixtures share development challenges with NASA’s Nuclear Thermal Propulsion (NTP) and Fission Surface Power (FSP) projects. The operational requirements also align well with development and investments in commercial microreactorsolutions. This paper will describe a structured approach for developing a multi-year strategy for maturing NEP fuel and moderator materials. Touchpoints where NASA could benefit from and be of benefit to the ongoing commercial investments are also described and discussed.

Dasari V Rao↗

DOE OSTI2023

Carbide-based fuel assembly for thermal propulsion applications

Carbide-based fuel assembly includes outer structural member of ceramic matrix composite material (e.g., SiC—SiC composite), insulation layer of porous refractory ceramic material (e.g., zirconium carbide with open-cell foam structure or fibrous zirconium carbide), and interior structural member of refractory ceramic-graphite composite material (e.g., zirconium carbide-graphite or niobium carbide-graphite). Spacer structures between various layers provide a defined and controlled spacing relationship. A fuel element bundle positioned between support meshes includes a plurality of distributively arranged fuel elements or a solid, unitary fuel element with coolant channels, each having a fuel composition including high assay, low enriched uranium (HALEU). Fuel assemblies are distributively arranged in a moderator block and the upper end of the outer structural member is attached to a metallic inlet tube for hydrogen propellant and the lower end of the outer structural member is interfaced with a support plate, forming a nuclear thermal propulsion reactor.

Barringer, Eric A.↗

DOE OSTI2024

A Comprehensive Assessment of the Viability of Small Modular Reactors in Africa: A Nuclear Security and Nonproliferation Perspective

In response to escalating global energy demands driven by industrialization and the pressing need for decarbonization, this paper explores the potential of Small Modular Reactors (SMRs) as a sustainable energy solution in Africa. Focusing on nuclear security and non-proliferation concerns, the study assesses Africa's energy landscape, emphasizing the need for diverse and reliable power sources. While highlighting the scalability and cost-effectiveness of SMRs, the analysis acknowledges potential challenges associated with their introduction, particularly concerning nuclear security and non-proliferation. Given the recommendation to use High Assay Low Enriched Uranium (HALEU) in some SMRs, it is important to critically consider the security implications of transporting nuclear materials, the proximity of the public to the plant, and the time taken to respond to planned assaults. The possibility of using HALEU makes the nuclear material more prone to adversaries such as sabotage, theft, and terrorism. Utilizing PESTLE analysis, this research seeks to outline the detailed political, economic, social, technological, environmental, and legal readiness of Africa to embrace the first-of-a-kind technology (SMR) while fulfilling its mandate to the Non-Proliferation Treaty (NPT). Examining regulatory frameworks, international cooperation, and safety protocols, the study underscores the importance of regional collaboration to prevent the misuse of nuclear technology for military and malicious intent. Drawing insights from successful case studies, the paper concludes by synthesizing key findings and proposing recommendations for policymakers and stakeholders. These recommendations encompass regulatory enhancement, capacity building, technology transfer, and diplomatic efforts to strengthen nuclear security, non-proliferation, and safeguards in Africa.

Prah, Christina↗

DOE OSTI2023

Development of Instrumented Advanced Test Reactor Irradiation Capsule Experiment for In-situ Thermal Conductivity Measurements of High-Density Fuels

Idaho National Laboratory (INL) is developing a first-of-a-kind leadout instrumented capsule experiment design to enable in-situ measurement capabilities in the Advanced Test Reactor (ATR) core. The Ceramic Advanced Thermal Evolution Research (CRATER) experiment supports the aLEU program objective to accelerate fuel performance irradiation testing for identifying alternative high-assay, low enriched uranium (HALEU) fuel systems. CRATER is a fueled, instrumented capsule experiment to measure in-situ temperature and thermal conductivity of ceramic fuels. Two ceramic fuel types will be used, uranium mono-nitride (UN) and uranium mono-carbide (UC), with a third metallic fuel used for comparison (UMo). The three fuel specimens will use a stainless-steel cladding. Programmatic objectives include linear heat generation rates (LHGR) of 210 ± 25 Watts per cm. and an inner clad temperature of 300-450 °C. The evolution of fuel thermal conductivity during irradiation has never been successfully measured in-situ for these systems and this experiment is designed to use advances in measurement sciences to characterize how thermal transport properties evolve while in reactor. Neutronic simulations of the experiment and its surrounding reactor environment were conducted using the Monte Carlo N-Particle Transport code (MCNP) and result in optimized fuel enrichment to meet target linear heat generation rates (LHGRs) influencing fuel temperatures, and fuel burnup requirements. Fabrication research and development (R&D) efforts are underway to produce annular right cylinder UC and UN pellets using carbothermic reduction and nitridation (or hydride-dehydride-nitride) synthesis methods, followed double-action die cold isostatic pressing.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗